Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3: PBNC 2022, 1 - 4 November, Beijing & Chengdu, China [1 ed.] 9789811988981, 9789811988998, 9811988986

Table of contents :
Acknowledgement
Contents
Contributors
On the Issues for Legislation of Spent Fuel in China
1 Introduction to the Current Administration and Legislation of Spent Fuel in China
2 The Deficiencies of Current Legislation for the Administration of Spent Fuel
2.1 The Regulation on the Administration of Spent Fuel Has not Yet Been Established
2.2 The Ownership of Spent Fuel and Its Reprocessing Products Are Still Undefined
2.3 It Lacks a Value Preservation and Appreciation Mechanism for Spent Fuel Fund
2.4 The Role of Public Influenced Is not of Enough Significance in the Whole Process of Spent Fuel Management
3 Suggestions to Improve the Legal System of Spent Fuel in China
3.1 Draft and Enact the Regulation on the Administration of Spent Fuel
3.2 Define the Ownership of Spent Fuel and Its Reprocessing Products Explicitly
3.3 Establish a Value Preservation and Appreciation Mechanism for Spent Fuel Fund
3.4 Give Full Attention to the Influenced Public in the Whole Management Process of Spent Fuel
4 Conclusions
References
Identification and Investigation on Thermal Hydraulic Phenomena Related to Core Make-Up Tank
1 Introduction
2 CMT Design and Function Overview
3 Phenomenon Identification Based on CMT Overall Effect Test
4 Analysis of Thermal Hydraulic Phenomena in CMT
5 Summary and Suggestions on Thermal Hydraylic Phenomena Related to CMT
References
Alara Methodology for Reactor Building Shielding Design in HPR1000 PWR
1 Introduction
2 Review of In-Service Reactor Building Shielding Design
3 ALARA Methodology of Reactor Building Shielding Design
3.1 Overall Review
3.2 Specific Analysis
3.3 Evaluation Review
4 Application of ALARA Methodology for Reactor Building Shielding Design
4.1 Overall Review
4.2 Specific Analysis
4.3 Evaluation Review
5 Conclusion
References
The Study of Shielding Benchmark of Heating Exchanger for Operating NPP
1 Introduction
2 Development of Shielding Benchmark
3 Methodology of Shielding Benchmark Establishment
3.1 Introduction of Simulation Software
3.2 Establishment Process of Shielding Benchmark
4 Establishment of Heating Exchanger Shielding Benchmark
4.1 Measurement of Source Term and Effective Dose Rate
4.2 Modeling of Shielding Benchmark
4.3 Shielding Benchmark Establishment and Validation
5 Conclusion
References
An Anisotropic Porous Model for Heat Exchanger Modeling of Fluoride-Salt-Cooled High-Temperature Advanced Reactor -- FuSTAR
1 Introduction
2 Methodologies
2.1 Mathematical and Physical Model
2.2 Description of Calculation Case
3 Results and Discussion
3.1 Comparison of 2D Shear Flow Case
3.2 Comparison of Anisotropic Flow in PCHE
4 Conclusions
References
Study on the Prediction of Critical Heat Flux in Uniformly Heated Round Tube by Multilayer Perceptron
1 Introduction
2 MLP Model
3 Training Dataset
3.1 Experiment Data
3.2 Data Processing
4 Training and Inference
4.1 Uniform MLP
4.2 Un-Uniform MLP
5 Conclusions
References
Neutronics and Thermal-Hydraulics Coupling Analysis of Integral Inherently Safe Fluoride-Salt-Cooled High-Temperature Advanced Reactor - Fustar
1 Introduction
2 Layout oF FuSTAR
3 Coupling Scheme and Mathematics
3.1 The Coupling Scheme
3.2 Mathematical Model and Pressure-Drop Correlation
4 CFD Modeling and Coupling Analysis
4.1 CFD Modeling
4.2 Coupling Analysis
5 Conclusions
References
Application of Light Curtain Test Technology in Dynamic Deformation Test of Rotating Machinery Rotor
1 Introduction
2 Overview of Light Curtain Test Device
3 Light Curtain Sensor Error Confirmation
4 Analysis of Test Results
4.1 Analysis of Radial Deformation Test Results
4.2 Axial Deformation Data Analysis
4.3 Measurement Uncertainty
4.4 Comparison Between Test Results and Numerical Calculation Results
5 Conclusions
Reference
Influence of Social Responsibility of Nuclear Power Companies on Public Acceptance
1 Introduction
2 Materials and Methods
2.1 Scales
2.2 Subjects
2.3 Reliability and Validity Analysis
3 Results and Discussion
3.1 Common Method Deviation Test
3.2 Correlation Analysis
3.3 Mediating Effect Test
3.4 Results
4 Additional Experiment
4.1 Experiment Design
4.2 Experiment Results
5 Discussion
6 Conclusion
Appendix
References
Local Correlation and NpNn Linearity of Nuclear Electric Moments
1 Introduction
2 Local Correlation of Electric Moments
3 Linear Evolution of Electric Moment in NpNn Logarithmic Coordinates
4 Conclusions
References
Research on Financial Risk Analysis of Nuclear Power Project Based on Monte Carlo Simulation
1 Introduction
2 Introduction to Monte Carlo Simulation
3 Financial Risk Analysis Model of Nuclear Power Project Based on Monte Carlo Simulation
4 Case Analysis
5 Conclusions
References
Research on Low Carbon and Energy Saving Technology Path of Nuclear Power HVAC System
1 Introduction
2 Cooling Heat Recovery
3 Hot and Cold Air Distribution
4 Cold Storage
5 Conclusion
References
Multi-objective Optimization Design of Plate Heat Exchanger for Spent Fuel Pool Cooling System
1 Introduction
2 The Establishment of Evaluation Model for Heat Exchanger
2.1 Establishment of Mathematical Model of Heat Exchanger
2.2 Investment Cost Calculation Model
2.3 Establishment of Evaluation Model Program
2.4 Validation of Evaluation Model Procedures
3 Optimization Algorithm Program
4 Optimization Example and Result Analysis
4.1 Selection of Optimization Variables
4.2 Optimization Results and Analysis
4.3 Sensitivity Analysis
4.4 Selection and Analysis of Optimization Scheme
5 Conclusion
References
Fuel Economy Analysis of ATF Assembly with SiC Cladding and UO2(BeO) Pellets in CPR1000
1 Introduction
2 Model and Calculation Method
3 Design Criteria and Method
3.1 Design Criteria
3.2 Method
4 Fuel Management Design
4.1 Design of Benchmark Loading Pattern with M5 AFA3G Fuel
4.2 Design of the ATF Core
4.3 Fuel Management Calculation Results
5 Fuel Management Economy Analysis
5.1 Batch Discharging Burn-Up Analysis
5.2 Analysis of Uranium Consumption
5.3 Annual Consumption Assemblies Assessment
6 Summary
7 Conclusion
References
Thermodynamic Analysis and Optimization of Nuclear Closed Air Brayton Cycle
1 Introduction
2 Methodology
3 Results and Discussion
4 Conclusion
References
Affecting Factors and Aspects for Improving Customized Nuclear Public Acceptance Strategies
1 Characteristics of Public Perception in the International Nuclear Power
1.1 Categories of Nuclear Power Markets Worldwide
1.2 Characteristic of Public Perception in Developed Countries
1.3 Characteristic of Public Perception in Nuclear Emerging Countries
2 Good Practices and Strategies for Improving Public Acceptance
2.1 Focus on the Analysis of Energy Transformation to Enhance the Public's Rational Understanding of Nuclear Energy
2.2 Focus on Improvement of Nuclear Legislation and Consolidation on the Responsibilities of All Stakeholders
3 Enhancement of Technology and Management to Ensure Steady and Sustainable Development of Nuclear Energy
4 Promotion of Education and Diversification of Communication Channels
5 Conclusion
References
Initial Response Study of Nuclear Power Plant Operations After Local COVID-19 Outbreak
1 Unit Safety Control
1.1 Operations Scheduling Adjustment
1.2 Operations Plan Adjustment
1.3 Operations Risk Control
1.4 Unit Defects Assessment
1.5 Production Material Guarantee
2 Personnel Safety Guarantee
2.1 Arrival of Key Personnel
2.2 The Arrival of the Personnel in the Secondary Key Positions
2.3 Alternate Shift Arrangements
2.4 All Staff Was Screened for Covid-19
3 NPP Site Partition Control
3.1 NPP Site Partition Control
3.2 Operators Stay in the Dormitory for Shift Personnel
3.3 Closed Management for Shift Personnel at Work
4 Refined and Graded Management of Epidemic Prevention and Control
5 Conclusion
References
Discussion on Nuclear Arms Control Based on the International Situation in 2022
1 Introduction
2 Analysis of International Nuclear Arms Control
2.1 Introduction of International Situation of Nuclear Arms Control
2.2 Analysis of International Situation of Nuclear Arms Control
3 Some Current Research in Nuclear Arms Control
4 Conclusions
References
SNSTC’s Capabilities and Practices on Performance Testing for Nuclear Security System and Equipment
1 Introduction
2 Testing Methods and Industry Standards
3 Testing of RPMs
4 Coordinated Research Projects of IAEA
5 Summary
Practical Analysis of Radon Protection Methods in an Underground Project
1 Harm Caused by Radon
2 Effect Analysis of Three Common Radon Protection Measures
2.1 Ventilation
2.2 Shield
2.3 Adsorption
3 Conclusions
References
Numerical Calculation of Flow Heat Transfer in a Single Square Tube Under a Blockage Accident
1 Introduction
2 Numerical Methods
2.1 Basic Parameters
2.2 Geometrical Model
2.3 Blockage Simulation Method
2.4 Method Mesh Sensitivity Analysis and Fully Develop Turbulence Test
3 Results and Discussion
3.1 Blockage Material
3.2 Blockage Thickness
3.3 Blockage Area Proportion
3.4 Blockage Shape
4 Conclusion
References
Research on the Siting Problem of Economy-Oriented Small Heating Nuclear Reactor Based on System Dynamics
1 Introduction
2 Literature Review
3 Methodology
3.1 Model Method and Assumptions
3.2 General Form of EOSSD Model
4 Empirical Analysis
4.1 The Problem of Location Selection When the Rate Is Linear
4.2 The Problem of Location Selection When the Rate Is Non-linear
5 Conclusions
References
Technical and Economic Analysis of Nuclear Heating Based on HPR1000
1 Introduction
2 Potential Market Analysis
3 Technical Feasibility
3.1 Site Layout Requirements
3.2 Heating Scheme
3.3 Feasibility Analysis of Key Technologies
3.4 Environmental Safety Impacts
4 Economic Analysis
4.1 Economic Analysis of the Heating Scheme
4.2 Economic Analysis of Steam Supply Scheme
4.3 Brief Summary
5 Conclusions
References
The Application of Renal Dynamic Imaging in Measuring Renal Function of En-Bloc Pediatric Kidneys Transplanted into Recipients
1 Introduction
2 Materials and Methods
2.1 Patients
2.2 99mTc-DTPA-based Renal Dynamic Imaging (GATE’s Method)
2.3 Estimated Glomerular Filtration Rate (eGFR)
2.4 Kidney Transplant Biopsy
3 Statistical Analysis
4 Results
4.1 Subject Characteristics
4.2 Correlation Analysis
4.3 Comparison of Main Indexes and Renal Function Early and Late After Transplantation
4.4 Renal Function Assessed by Serum Creatinine and gGFR
4.5 Impact of the Occurrence of Complications and Kidney Biopsy on Renal Function
5 Discussion
6 Conclusions
References
Study of Cost Index System for Nuclear Power Plant
1 Introduction
2 Cost Index System for General Design
2.1 Guidelines of Establishing the List of Cost Accounts
2.2 Guidelines of Calculating the Unit Cost of Accounts
2.3 Guidelines of Establishing Major Equipments and Materials Index
3 Conclusion and Future Work
References
Application of Intelligent Innovation in Accident Procedure of Nuclear Power Plant
1 Introduction
2 Accident Procedure of Nuclear Power Plant
3 Intelligent Requirements for Accident Procedures
4 System Structure
4.1 System Hardware Structure
4.2 System Software
5 System Running
6 Feedback
7 Conclusion
References
Experimental Study on Thermal-Hydraulic Characteristics of Air Heat Exchanger for Sodium Cooled Fast Reactor
1 Introduction
2 Theoretical Calculation Method
3 Experimental Device
4 Experimental Research and Result Analysis
4.1 Experimental Content and Experimental Conditions
4.2 Data Processing
4.3 Experimental Results and Analysis
5 Conclusion
References
Effect of Coating Temperature on Microstructure and Properties of the SiC Layer in TRISO-Coated Particles
1 Introduction
2 Experimental Procedure
2.1 Preparation of TRISO-Coated Particles
2.2 Characterization Methods
3 Results and Discussion
3.1 Microscopic Structures and Densities
3.2 Mechanical Properties
4 Conclusions
References
An Improved Super-Resolution Model for Bubble Feature Extraction Process
1 Introduction
2 Related Work
3 Architecture Specification
3.1 High-Order Degradation Model
3.2 Gradient Map Loss
3.3 Generator Architecture
4 Experiment
4.1 Datasets and Evaluation Metrics
4.2 Training Details
4.3 Qualitative Results
4.4 Preprocessing Process of 2D Bubble Images
4.5 Contour Extraction Results
5 Conclusion
References
Analysis of Current Status and Development Trend of Nuclear District Heating
1 Introduction
2 Development Status and Trend of Central Heating in China
3 Nuclear Heating Faces Major Opportunities
3.1 China Supports Nuclear Heating
3.2 Rapid Development of Urban Central Heating Area
3.3 China Has Good Nuclear Heating Practice
4 Overview of Nuclear Heating
4.1 Single Nuclear Heating Mode
4.2 Cogeneration Heating Mode
5 Advantages and Industrialization Promotion of Nuclear Heating
5.1 Good Environmental Benefits
5.2 Nuclear Heating is Economically Competitive
5.3 Promotion and Trend Analysis of Nuclear Heating Industrialization
6 Problems in the Development of Nuclear Heating
7 Conclusion
Research on High-Speed Cold Cutting Technology for Combined Spider
1 Introduction
2 Research Subjects
2.1 Combined Spider
2.2 Choice of Precision Cold Cutting Technology
3 Research Results
3.1 Test Conditions
3.2 Water Jet Scheme and Part Design
3.3 Water Jet Surface Quality Tests
3.4 Brazing Performance Test of Water Jet Surface
3.5 Water Jet Process Parameters
3.6 Water Jet Accuracy Control
3.7 Results of Small Batch Cutting Studies
4 Conclusions
References
Research and Development Practice of Equipment Condition Monitoring and Intelligent Diagnosis Platform Based on Big Data and AI at TNPS
1 Introduction
2 Typical Algorithms
3 Overall Architecture
4 Functional Structure
5 Typical Functions
6 Conclusion
References
Study on the Design Standard of Water-Intake Breakwaters in Coastal Nuclear Power Plants
1 Introduction
2 Provisions by Safety Guides and Codes
3 Determination of the Design Standard
4 Engineering Example
5 Conclusions
References
A Study of Stability and Coagulation Behavior of Broth Based the Internal Gelation Process for Uranium Dioxide Microspheres
1 Introduction
2 Experiment
3 Results and Discussion
3.1 The Effect of the Preparation and Storage Temperature on the Stability of the Broth
3.2 Analysis of Causes of Broth Solidification
3.3 The Effect of Urea/U on the Stability of the Broth
3.4 Analysis of Post-treatment of Scrapped Broth
4 Conclusions
References
The Preliminary Thermo-hydraulics Calculation of Unfolding Process of the Lotus Reactor
1 Introduction
2 Modelling and Methodology
2.1 Tools
2.2 Overset Mesh
2.3 Subcooled Boiling
2.4 Modeling
3 Results and Discussion
3.1 Steady-State Simulation
3.2 Transient Simulation
4 Conclusion
References
Key Technology of a New Potential Economical Desalination Method for Small Nuclear Power
1 Introduction
2 Methodology
2.1 Expert Knowledge and Comparative Analysis
2.2 Hydrate Equilibrium Prediction
3 Result and Discussion
4 Conclusions
References
Research on On-Line Measurement Technology of Fuel Rod Deformation Based on LVDT
1 Introduction
2 Methods and Principles
3 Results and Discussion
4 Conclusion
References
Numerical Study on Flow Distribution Characteristics of Parallel Square Flow Channels Under Blockage
1 Introduction
2 Numerical Methods
2.1 Geometrical Model
2.2 Mesh Sensitivity Analysis and Fully Develop Turbulence Test
3 Results and Discussion
3.1 Inlet Header Sizes
3.2 Different Blockage Location
3.3 Different Blockage Ratio at the Inlet
3.4 Maximum Wall Temperature
4 Conclusion
References
Computer Package for Calculating the Fast-Neutron Fluence in Reactor Pit Based on the Reactor Operational History and the Monte-Carlo Method
1 Introduction
2 Calculation Mothod and Models
2.1 Calculation Process
2.2 Input Data Preprocessing
3 Verification
4 Conclusion
References
Research on Classification and Coding of Nuclear Power Plant Status Reports Based on Machine Learning
1 Introduction
2 Classification and Coding System for Nuclear Power Plant Status Reports
3 Experiments and Analysis of Results
3.1 Experimental Preparation
3.2 Classification Model Evaluation Indicators
3.3 The Effect of Training Set Sample Size on Model Classification Performance
3.4 The Effect of Learning Objects on Model Classification Performance
3.5 The Effect of Word Segmentation Mode on Model Classification Performance
4 Conclusions
References
Atomic Scale Simulation on Liquid Metal Embrittlement Induced by Segregation of Lead Element in Lead-Cooled Fast Reactor
1 Introduction
2 Simulation Methods
3 Results and Discussion
3.1 Results for PB Segregation
3.2 Results for Uniaxial Tensile Loading Test
3.3 Effects of Pb Segregation on the Mechanical Properties of Ferritic Steels
4 Conclusion
References
Application of Construction Phase Process Cost Management in Nuclear Power Projects
1 Introduction
2 Overview of Construction Process Cost Control Methods
3 On-Budget Cost Management
4 Off-Budget Expense Management
5 Conclusion
References
Characterization of Residual Uranium Resources in the Decommissioned Final Mining Area of a Uranium Deposit in Xinjiang
1 Introduction
1.1 A Subsection Sample
2 Research Background
2.1 Hydrogeological Setting of the Deposit
2.2 Production History and Problems
3 Characterization and Analysis of Uranium Resources in the Target Mining Area Zxcvbnm
3.1 Selection of Target Area and Drilling
3.2 Analysis of the Distribution of Residual Uranium Resources
3.3 The Validity Analysis of Logging Curves
4 Secondary Development Potential Analysis
5 Conclusion
References
Study on Influence of Process Parameters on Image Quality in X-ray Digital Radiography Inspection of Fuel Rod Weld
1 Introduction
2 Process Experiment
2.1 Equipment of Experiment
2.2 Process Experiment
2.3 Indicators of Evaluation
3 Results and Discussion
3.1 The Multi-index Comprehensive Balance Analysis
3.2 Range Analysis
3.3 The Variance Analysis
3.4 Analysis of the Influence of Factors on Image Quality
4 Verification Experiment
5 Conclusions
References
Construction of CDS/TIO2/HGS Composite Materials and Photocatalytic Reduction of Hexavalent Uranium in Wastewater
1 Introduction
2 Experimental
2.1 Materials
2.2 Catalysts Preparation
2.3 Characterization
2.4 Photocatalysis Activity
3 Results and Discussion
3.1 Characterization of Materials
3.2 Photocatalysis Test
3.3 Mechanism of Photocatalytic Reduction
4 Conclusions
References
Experimental Study on Shell Side Flow Distribution of Fast Reactor Steam Generator
1 Introduction
2 Experimental Set-Up
3 LDV Measuring Device and Measuring Point Layout
3.1 LDV Measuring Device
3.2 Layout of Measuring Points
4 Test Results and Analysis
4.1 Results of Cross Flow Velocity in Tube Bundle Area
4.2 Results of Axial Flow Velocity in Tube Bundle Area
4.3 Results of Three-Dimensional Velocity Distribution at the Outlet of Porous Ring Plate
5 Conclusions
References
Vibration Analysis of Heating-Used Steam Extraction Pipelines in Nuclear Power Plant
1 Introduction
1.1 A Subsection Sample
2 Two-Phase Flow Induced Vibration in the Extraction Pipeline
3 Extraction Piping Vibration Monitoring
3.1 Layout of Monitoring Points
3.2 Monitoring Result
4 Analysis
4.1 Model Analysis
4.2 Exciting Sources
4.3 Drainage
5 Conclusions and Future Perspectives
References
Experimental Study on Aerosol Removal Effected by Spray Characterization
1 Introduction
2 Experimental Apparatus
2.1 Overview of the Facility
2.2 Experimental Parameters
3 Experimental Results
3.1 Spray Droplet Size Distribution
3.2 Aerosol Global Removal Efficiency
4 Aerosol Removal Model
4.1 Equation of Droplets Motion
4.2 Improved Aerosol Removal Model
4.3 Removal Efficiency of Mechanical Mechanisms
5 Conclusions
References
Study on Public Acceptance of Restart of Inland Nuclear Power Plants Based on TAM/TPB Integration Model
1 Introduction
2 Theoretical Framework and Research Hypotheses
2.1 Perceived Usefulness
2.2 Perceived Risk
2.3 Subjective Norms, Behavioral Attitudes, and Perceived Behavior Control
2.4 Trust
3 Research Methodology
3.1 Sample and Data Collection
3.2 Questionnaire Design and Variable Measurement
3.3 Reliability and Validity Tests
4 Data Analysis and Results
4.1 Measurement Model Test
4.2 Structural Model Test
5 Discussion
6 Conclusion and Policy Implications
References
Low Temperature Performance of LBE Oxygen Sensors with Different Reference Electrodes
1 Introduction
2 Experiment
3 Results and Discussions
3.1 Calculate of Theoretical E
3.2 Performance Test
4 Conclusion
References
The Development of the Combined Fission Matrix Theory in Burnup Calculation
1 Introduction
2 Method
2.1 Model
2.2 Fission Matrix Theory
2.3 Fission Matrix Combination Method
2.4 Correct Ratio
2.5 Fission Matrix Database
3 Result
4 Conclusion
References
Study of the Electric Device of RRI019/020VN in Qinshan Nuclear Power Plant Cannot Close Completely in Long Signal Valve Closing
1 RRI019/020VN Operation
2 Introduction of Valve Electric Device Control Mode
2.1 Limit Valve Closing Control
2.2 Torque Valve Closing Control
3 Troubleshooting Analysis of RRI019/020VN
4 Solution Analysis
5 Conclusion
Reference
Numerical Study of Influence of 15N Enrichment on Burnup Performances for a Heat Pipe Cooled Traveling Wave Reactor
1 Introduction
2 Core Design and Computational Tool
3 Results
3.1 Reactivity Analysis
3.2 Power Distribution and Burnup
3.3 Consumption and Breeding of Nuclear Fuel
4 Conclusions
References
Parallelization of Thermal Hydraulic Real-Time Simulation Program Based on Two-Phase Drift Flux Model
1 Introduction
2 Introduction and Performance Analysis THEATRe Program
2.1 Physical Model
2.2 Calculation Procedure
2.3 Hotspot Module of Serial THEATRe Program
3 Parallel Scheme
3.1 MPI Parallel Scheme
3.2 OpenMP Parallel Scheme
4 Effectiveness and Performance of Parallel THEATRe Code
4.1 Effectiveness of Parallel THEATRe Code
4.2 Performance of Parallel THEATRe Code
5 Conclusions
References
Simulation of Photon-Counting Detector Based on OpenModelica
1 Introduction
2 Method and Modeling
2.1 Introduction of OpenModelica
2.2 X-ray Tube
2.3 Semiconductor Detector
2.4 Preamplifier
2.5 Pulse Shaper
2.6 Pile-Up Rejection Circuit
2.7 Overall Structure
3 Results
3.1 Collection Efficiency
3.2 PREamplifier’s and Pulse SHAper’s Response
3.3 Test of Pile-Up Rejection Circuit
4 Conclusion
References
Research on Boiling and Coupled Heat Transfer of OTSG with Drift Flow Single Fluid Model
1 Introduction
2 Improved Drift Flow Single Fluid Model
2.1 Governing Equation
2.2 Slip Velocity Model
2.3 Interphase Mass Transfer Model
2.4 Flow Pattern Identification and Heat Transfer Model
3 Model Validation
3.1 Simulation Description
3.2 Result Analysis
4 Conclusions
References
Study on the Crystallization Rate of Uranium Peroxide Based on Ultrasound and Microchannel Technology
1 Introduction
2 Materials and Methods
2.1 Experiment Reagent
2.2 Experimental Facility
2.3 Determination Method of Nucleation Rate and Growth-Rate
2.4 Analytical Method
3 Results and Discussion
3.1 Effect of Ultrasonic Power on Nucleation Rate and Growth Rate of Uranium Peroxide Crystal in Tubular Process
3.2 Effect of Ultrasonic Power on Nucleation Rate and Growth Rate of Uranium Peroxide Crystal in Microchannel Process
3.3 Comparison of Crystallization Process Under Different Mixing Modes and Ultrasonic Power Conditions
4 Conclusion
References
Construction of Nuclear Emergency Awareness System Based on “Cloud-Edge-Terminal”
1 Introduction
2 Problems Existing in the Construction of Intelligent Nuclear Emergency
3 Edge Computing Technology
4 Intelligent Emergency Awareness Architecture of “CLoud-Edge-Terminal”collaboration
5 Conclusions
References
Sodium Spray Fire Analysis with Combustion Space Multi-node Model for Sodium-Cooled Fast Reactor
1 Introduction
2 Model
2.1 Droplet Combustion Model
2.2 Motion Model
2.3 Combustion Space Multi-nodes Model
3 Model Validation
3.1 FAUNA Experiments
3.2 SNL T3 Experiments
4 Conclusions
References
Analysis of Fission Product Diffusion Behavior of Fully Ceramic Micro-encapsulated Fuel
1 Introduction
2 Sensitivity Analysis of TRISO Particle Parameters
2.1 UO2 Kernel Diameter
2.2 Buffer Layer Thickness
2.3 IPyC Layer Thickness
2.4 SiC Layer Thickness
2.5 OPyC Layer Thickness
3 Sensitivity Analysis of SiC Matrix Parameters
3.1 Diffusion Coefficient of Fission Products in SiC Matrix
3.2 Fuel Pellet Diameter
3.3 Thickness of Fuel-Free Zone
4 Summary
References
Optimization of Sintering Parameters of Lead Oxide Particles for Solid Oxygen Control in Lead Cooled Fast Reactor
1 Introduction
2 Experimental
2.1 Material Preparation
2.2 Characterization Methods
3 Resulis and Discussion
3.1 Relative Density
3.2 Hardness
3.3 Flexural Strength
3.4 SEM Micrographs
3.5 Optimum Sintering Parameter
4 Conclusions
References
Research on Fission Products Selection in the Primary Coolant of PWR During Normal Operation
1 Introduction
2 Fission Products Selection Methodology
3 Radionuclides Selected
3.1 Generation Mechanisms
3.2 Selection Based on Radionuclides Half-Life and Activity Levels
3.3 Selection Based on RGP (Relevant Good Practice)
3.4 Insights from OPEX
3.5 Recognition of Technical Users ‘Needs’
4 List of Selected Fission Products
5 Comparison
6 Conclusions
References
Development of a Depletion Code with a Control-Rod-Adjusting Program Based on OpenMC
1 Introduction
2 Development of Automatic Control-Rod-Adjusting Depletion Calculation Program
2.1 OpenMC Depletion Calculation Module
2.2 Development of Critical Control Rod Position Search
2.3 Development of Depletion Element Update and Physical Parameter Storage
2.4 Coupling of Critical Control Rod Position Search and Depletion
3 Verification and Applications
3.1 Verification Model
3.2 Critical Control Rod Position Update
3.3 Nuclear Data Storage
4 Conclusions
References
Study of Fine Particle Deposition of Porous Medium in Heat Pipe of Heat Pipe Reactor Under Gravity
1 Introduction
2 Research Subject
2.1 Geometric Models
2.2 Structure and Medium Parameters
2.3 Concentration Parameters
2.4 Grid-Independent Validation
3 Calculation Formula
3.1 Equation for the Flow of a Working Mass
3.2 Equation for Particle Movement
3.3 The Equation for Particle Deposition Volume Calculation
4 Calculation Results and Analysis
4.1 Calculation of Deposition Volume with Time and Concentration for Vertical Placement
4.2 Calculation of Deposition Volume with Time and Particle Concentration When Placed Horizontally
4.3 Comparative Calculation of Deposition in Different Cases of Heat Pipe Porous Medium
5 Conclusion
References
Recommendations to Improve the Public Acceptance of Nuclear Environmental Protection
1 Introduction
2 People’s Acceptance of Nuclear Environmental Protection
3 Recommendations
3.1 Enhance the Publicity of Nuclear
3.2 Strengthen the Education on Nuclear Environmental Protection
3.3 Promote Programs of Nuclear Environmental Protection
4 Conclusions
References
Coupled Neutronics and Thermal-Fluid Calculation for Prismatic High-Temperature Gas-Cooled Reactor Core at Steady State
1 Introduction
2 Method for Steady-State Neutronics and Thermal-Hydraulics Coupling of the Core
3 Moose-Based Development of Steady-State Neutronics and Thermal-Hydraulics Coupling Program
3.1 Development of the Core Neutron Diffusion Module
3.2 Development of the Core Thermal-Fluid Module
3.3 Multiapp Development
3.4 Data Exchange Between Modules
3.5 Two-Dimensional Cross-Section Interpolation Method
4 Calculation of MHTGR-350MW Benchmark
4.1 Three-Dimensional Coupling Model
4.2 Calculation Conditions
4.3 Calculation Results
5 Conclusion
References
Study on the Preparation of Metallographic Samples and the Characterization of Volume Content and Size Distribution of Inclusions in Uranium Metal
1 Introduction
2 Experiment
2.1 Experimental Program
2.2 Sample Preparation
3 Results and Analysis
3.1 Comparison of Different Metallographic Sample Preparation Methods for Inclusions Showing
3.2 Identification of Inclusions
3.3 Quantitative Metallographic Study on the Volume Fraction and Size Distribution of Inclusions
4 Conclusions
References
Development Background and Research Progress of UN-U3Si2 Composite Fuel
1 Introduction
2 Properties of UN, U3Si2 and UN-U3Si2 Composite Pellet
3 Research Progress of UN-U3Si2 Composite Fuel
4 Conclusions and Prospects
References
Study on Fabrication and Characterization of SiCf/SiC Composite Cladding
1 Introduction
2 Experimental Process
2.1 Preparation of Interface Layer
2.2 CVI Densification
2.3 Performance Characterization
3 Results and Analysis
3.1 Interface Layer
3.2 Micro-morphology and Structure
3.3 Density
3.4 Tensile Strength and Elongation
4 Conclusions
References
Life-Cycle Cost Study for a Near-Surface Disposal Repository of Low-Level Waste in China
1 Introduction
1.1 Background
1.2 Research Status
1.3 Research Aims and Significance
2 Life-Cycle Cost Components and Calculation Model for LLRW Disposal Repository in China
2.1 Construction Cost
2.2 Operating Costs
2.3 Closure Costs
3 Life-Cycle Cost of a LLRW Disposal Repository in China
3.1 Introduction
3.2 Calculated Boundaries
3.3 Cost Entries
4 Conclusion
References
Numerical Study on External Environmental Radiation of Nuclear Ramjet Reactor
1 Introduction
1.1 A Subsection Sample
2 System Description
3 Numerical Method
3.1 Mathematical Model
3.2 Modeling Method
4 Results and Discussion
4.1 Source Term Analysis
4.2 Radiation Analysis
5 Conclusion
References
Optimization on Design of Logic Degradation for Safety I&C System of Nuclear Power Plant
1 Introduction
2 The Safety I&C System of Nuclear Power Plant
3 Logic Degradation Design
4 Problems in Logic Degradation Design
4.1 Applicable Status/Condition Is Not Considered
4.2 The Influence of Unit Conditions on Measurement Is Not Considered
4.3 Degradation Logic Is Not Separated from the Shutdown Protection Logic
5 Improvement of the Logic Degradation
5.1 Coolant Pump Power
5.2 DNBR
5.3 Liquid Level of Pressurizer (Narrow Range)
5.4 Separation of Degradation Logic and Shutdown Protection Logic
6 Summary
References
Study on Gamma-Ray Spectra Feature Recognition and Isotope Composition Analysis of Plutonium Based on Convolutional Neural Networks
1 Introduction
2 Machine Learning and Convolutional Neural Network
3 Data Set Creation
4 Result and Discussion
5 Conclusions
References
Study on the Coupled Heat Transfer Characteristics of Liquid Lead-Bismuth Eutectic and Supercritical CO2
1 Introduction
2 Research Object
2.1 Analysis Model
2.2 Calculate Parameters and Boundary Conditions
2.3 Meshing and Independent Verification
3 Computational Model
3.1 Governing Equation
3.2 Turbulence Model
3.3 Heat Transfer Equation
4 Results and Analysis
4.1 Temperature Distribution
4.2 Heat Exchange
4.3 Heat Transfer Coefficient
4.4 Pressure Drop
5 Conclusion
References
Adsorption Behaviors of Hydrogen on Equal Atomic Ratio TiZrV Film Applied in AB-BNCT by Density Functional Theory Study
1 Introduction
2 Model
3 Hydrogen Adsorption
4 Conclusion
References
Research on Interface Technology of Coupling Thermal-Hydraulics and Other Codes
1 Introduction
2 Coupling Platform
3 Coupling Schemes
3.1 Coupling Methods
3.2 Spatial Mesh Correspondence
3.3 Time Step Coupling
3.4 Coupled with Other Software
4 Tests
4.1 Test Data
4.2 Test Results
5 Conclusion
References
Development and Verification of Transport-Activation Coupling Capability in CosRMC
1 Introduction
2 Internal Coupling Transport-Activation Calculation Function Development
3 Computational Model and Comparison Code
4 Calculation Results and Analysis
4.1 Constant Energy Spectrum
4.2 Constant/Variable Energy Spectrum
4.3 Continuous Energy/Multi-group Cross Section
5 Conclusions
References
Research on Shielding Deep Penetration Calculation Based on MC Variance Reduction Techniques
1 Introduction
2 Geometric Model of Particle Transport
3 Application of Variance Reduction Techniques
4 Calculation Results and Analysis
4.1 Single Variance Reduction Technique
4.2 Combined Variance Reduction Techniques
5 Conclusions
References
Prospects for Next Generation Nuclear Power System
1 Introduction
2 Basic Design Considerations
2.1 Economic
2.2 Safety
2.3 Environment-Friendly
2.4 Resource Utilization
3 Features of Next Generation Nuclear Power System
4 Conceptual Design of Next Nuclear Power System
4.1 Intrinsically Safe Multipurpose Lead Cold Traveling Wave Reactor
4.2 Technical and Research Difficulties
5 Conclusions
References
Safety Analysis of Internal Flooding Under Main Feedwater Pipeline Rupture Accident Based on Small Modular Reactor ACP 100
1 Introduction
2 Method
2.1 Internal Flooding Safety Evaluation
2.2 Application of CNIFA for Internal Flooding Safety Evaluation
2.3 Internal Flooding Analysis of Main Feedwater Supply Pipeline Rupture Accident
3 Result and Discussion
3.1 Simulation Results of Main Feedwater Supply Pipeline Accidents
3.2 Design Optimization and Improvement
3.3 Optimization and Improvement Results
4 Conclusions
References
Research on Nuclear Seawater Desalination Technology Based on Small Modular Reactor ACP100
1 Introduction
2 Method
2.1 Seawater Desalination Process
2.2 Radioactivity of Nuclear Desalination
2.3 Cost Calculation of Nuclear Desalination
3 Result and Discussion
3.1 Analysis of Radioactivity for MED Nuclear Desalination
3.2 Cost of Different Nuclear Desalination Methods
4 Conclusions
References
Research Status and Prospect of Comprehensive Utilization Technology of Nuclear Energy
1 Introduction
2 Comprehensive Utilization of Nuclear Energy
2.1 Nuclear Heating
2.2 Nuclear Steam Supply
2.3 Nuclear Desalination
2.4 Nuclear Hydrogen Production
2.5 Nuclear Energy-Storage-Renewable Energy Coupling
3 Scene Construction
4 Suggestions on the Development of Comprehensive Utilization of Nuclear Energy
4.1 Strengthen the Top-Level Design and Coordinate the Promotion of Comprehensive Nuclear Energy Utilization Planning
4.2 Improve Scientific and Technological Research and Development Capabilities and Make Breakthroughs in Key Technology Application Innovations
4.3 Promote the Reform of the System and Mechanism, and Improve the Construction of the Standard System for Comprehensive Utilization of Nuclear Energy
4.4 Strengthen Industry Supervision and Continuously Improve the Safety Performance of Nuclear Energy
References
Hydrogen Adsorption Mechanism of Non-equal Atomic Ratio TiZrV NEG Films Surface
1 Introduction
2 Model Building
3 Results and Disscussion
3.1 Adsorption Energy
3.2 Density of States
4 Conclusion
References
Study of Fast-Start Heat Transfer Characteristics of Potassium Heat Pipe
1 Introduction
2 Research Subject
2.1 Geometric Model
2.2 Calculation Parameters
2.3 Grid Irrelevance Verification
3 Calculation Formula
4 Results and Discussion
4.1 Heat Pipe Routine Start-Up Calculation
4.2 Heat Pipe Condensing Section External Heating Start-Up Calculation
4.3 Heat Pipe Internal Heating Start-Up Calculation
5 Conclusion
References
Prediction of CHF in Narrow Rectangular Channel Based on GA-BP Neural Network
1 Introduction
2 CHF Database
3 BP Neural Network
4 GA-BP Neural Network
5 Conclusions
References
Study on 131I Source Term in Reactor Coolant Under Shutdown Transient Condition
1 Introduction
2 Cause Analysis of 131I Peak Release
3 Analysis of 131I Peak Factor and Release Rate
3.1 Analysis of 131I Peak Factor
3.2 131I Release Analysis
4 Conclusions
References
Study on Aggregation Forming of Cathode Liquid Cerium Metal in Molten Salt Electrorefining Process
1 Introduction
2 Experimental Section
2.1 Experimental Materials and Equipment
2.2 Experimental Methods and Procedure
3 Analysis of Factors Affecting Cathodic Metal Cerium Forming
4 Results and Discussion
4.1 Passivation Film Confirmation and Electrode Structure Optimization
4.2 Study on Parameter Optimization of Four—Electrode System
5 Conclusions
References
Thermal Engineering Calculation of Control Drum Misoperation of Heat Pipe Reactor
1 Introduction
2 Research Object
2.1 Geometric Model
2.2 Calculation Parameters
3 Computational Model
3.1 Calculation Parameters
3.2 One Dimensional Heat Conduction Model
3.3 Reactivity Introduction and Feedback Model
3.4 Model Validation
4 Calculation Results and Analysis
4.1 Hottest Fuel Assembly Neutron Flux
4.2 Hottest Fuel Assembly Hot Spot Temperature
4.3 Normalized Power of Hottest Fuel Assembly
5 Conclusion
References
CFD Analysis on the Multiphase Flow of the Wickless Sodium Heat Pipe with Non-condensable Gas
1 Introduction
2 Experimental Apparatus
3 CFD Model and Solution Description
4 Results and Discussion
5 Conclusions
References
Influence of Impedance Characteristics of Perforated Plate on Pump-Induced Pulsation in Reactor
1 Introduction
2 Theory of Acoutisc Damping
2.1 Fluid Damping
2.2 Boundary Impedance Characteristics
3 Computational Model
3.1 Finite Element Model
3.2 Model Parameter
4 Results and Discussing
4.1 Modal Analysis
4.2 Sound Pressure Response Without Considering Impedance Boundaries
4.3 Sound Pressure Response Considering Impedance Boundaries
5 Conclusion
References
Development of Design and Verification Platform for Heat Pipe Cooled Nuclear Reactor
1 Introduction
2 Reactor Structure
2.1 Core Body System
2.2 Heat Pipe Thermoelectric Power Generation System
2.3 Residual Heat Removal System
3 Architecture of HPR Design Verification Platform
3.1 Introduction to Simulation Platform
3.2 System Models
3.3 Model Interface
4 Result of Coupling Calculation
5 Conclusion
References
Development of Curriculum for International Students in the Field of Nuclear Energy Engineering at Harbin Engineering University
1 Introduction
2 Analysis of International Students in the Field of Nuclear Energy Engineering
2.1 Background of International Students
2.2 Thinking and Countermeasures
3 Design of Curriculum for Master Program of Nuclear Energy Engineering
3.1 Common Course Modules
3.2 Professional Basic Course Modules
3.3 Professional Course Modules
3.4 Management Course Modules
3.5 Practice Modules
3.6 Curriculum
4 Conclusions
References
Technological Upgrading Analysis on Diversity Transformation of Reactor Protection System in a Fast Reactor Project
1 Introduction
2 Reactivity Control in Water Reactor
3 Design of Reactor Protection System of Other Fast Reactor
4 Design of Reactor Protection System of One Fast Reactor in China
5 Design and Optimization of Fast Reactor Protection System
6 Possible Upgrading Direction
7 Conclusions
References
Study on Top Architecture Design of Nuclear Power System Based on an Incremental Iteration Generation Method
1 Introduction
2 Method Introduction
3 Functional Architecture Generation for Nuclear Power System
4 Logical Architecture Generation for a Nuclear Power System
4.1 Functional and Logical Component Definitions
4.2 Logical Structure Inference of Nuclear Power System Based on Functional-Logical Mapping
4.3 Definition of Faults in the Logical Structure of Nuclear Power Systems
4.4 Internal Association of Logic Structure of Nuclear Power System
5 Case Study: Normal Residual Heat Removal System of Nuclear Heating Reactor
5.1 Functional Decomposition of Heating Reactor Under Normal Operating Conditions
5.2 Definition of Abnormal Conditions
5.3 Internal Associations of the System Logical Architecture
6 Summary
References
Study on Optimization of Economic Evaluation Methods for Generation III Reactors in China
1 Introduction
2 Overview of the 2011 Energy Standard
2.1 The Financial Evaluation Method
2.2 Evaluation Parameters
3 Economic Evaluation Method Improvement Analysis
3.1 Economic Evaluation Parameters
3.2 Nuclear-Related Expenses
3.3 Power Market Reform Factors
3.4 Operation and Maintenance Costs Division
4 Case Calculation
4.1 Evaluation Parameters
4.2 Evaluation Results
5 Conclusion
References
Preliminary Design and Thermal Analysis of Intermediate Heat Exchanger Based on Irradiation Test Loop with Alternative Coolants
1 Introduction
2 Structure Design and Key Parameters of the Heat Exchanger
2.1 Description of the Heat Exchanger Structure
2.2 Thermal Design Input
2.3 Key Design Parameters of Heat Exchanger
3 Method and Thermal Calculation
3.1 Convective Heat Transfer Coefficient of Tube Side
3.2 Convective Heat Transfer Coefficient of Shell-Side LBE
3.3 Heat Exchange Accounting
4 Thermal Analysis Under Alternative Coolants Conditions
4.1 Influence of Thermal Parameters Under Liquid Metal Conditions
4.2 Influence of Thermal Parameters Under Helium or Molten Salt Condition
5 Conclusions
References
Development of High-Fidelity Neutron Transport Solver of Alpha
1 Introduction
2 Theory
2.1 Methodology of MOC-EX
2.2 Methodology of 3D CMFD
3 Implementation
3.1 Parallel Planar MOC Algorithm
3.2 Parallel Planar MOC Algorithm
4 Numerical Results
4.1 The 3D C5G7 MOX Benchmark Problem
4.2 The 3D KAIST Benchmark Problem
5 Conclusions
References
Study on Welding Process Between the Hexagonal Tube and the Grid Frame Component of MOX Fuel Assembly in CEFR
1 Introduction
2 Materials and Equipments
2.1 Experimental Materials
2.2 Experimental Equipment
3 Quality Requirements and Experimental Methods
3.1 Welding Quality Requirements
3.2 Experimental Method
4 The Results and Welding Procedure Qualification
4.1 Influence of Welding Current
4.2 Influence of Electrode Pressure
4.3 Influence of Welding Time
4.4 Welding Procedure Qualification
5 Conclusions
References
Unattended Monitoring Apparatus Data Analysis Technicals Used for Nuclear Safeguards
1 Introduction
2 Basic Concepts of Unattended Monitoring of Nuclear Safeguards
3 Research on Unattended Data Analysis Techniques for Nuclear Material Management
3.1 Data Acquisition and Exchange Systems
3.2 Data Information Management System
4 Conclusions
References
Economic Analysis of Nuclear Cogeneration System Based on High-Temperature Gas-Cooled Reactor
1 Introduction
2 Thermal Economic Model of HTGR Cogeneration System
3 Results and Discussion
3.1 Thermal Economic Analysis of Thermo-power Cogeneration System
3.2 Thermal Economic Analysis of Hydrogen-Power Cogeneration System
4 Conclusions
References
Study on Social Stability Risk Assessment of Nuclear Facilities
1 Introduction
2 SSRA Procedure
3 SSRA Contents
4 SSRA Scope and Methods
5 SSRA Risk Index Database
6 SSRA Evaluation Criteria and Weight Calculation
6.1 SSRA Evaluation Criteria
6.2 SSRA Weight Calculation
7 Case Analysis
8 Conclusion
References
Investigation on the Fabrication of High-Density Micro Pits on Zirconium Alloy by Micro Imprinting
1 Introduction
2 Experimental
2.1 Sample Preparation
2.2 Fabrication of Die
2.3 Imprinting Process
3 Results and Discussion
3.1 Traditional Imprinting Forming
3.2 “Step by Step” Imprinting Process
4 Conclusions
References
Application and Analysis of Airborne Electromagnetic Method in Engineering Exploration of High-Level Radioactive Waste Repository
1 Introduction
2 Introduction of Airborne Electromagnetic Method in Exploration of Qinghai Province
3 Analysis of the Ability of Airborne Electromagnetic Method to Solve Geological Problems
4 Application Prospect of Airborne Electromagnetic Method in Exploration of High-Level Radioactive Waste Disposal Repository
5 Conclusion
References
Optimization of Power Supply Scheme and Economic Benefit Analysis of a Project in a Newly-Built Large Nuclear Chemical Plant
1 Introduction
2 Load Analysis
2.1 Project Overview
2.2 Load Analysis
2.3 Safety Level Loads
3 Power Supply Scheme and Economic Benefit Analysis
3.1 Power Supply Scheme
3.2 Innovation and Comparison
3.3 Economic Benefits
4 Conclusion
References
Air-Cooler Structure Optimization of HTGR Passive Cavity Cooling System
1 Introduction
2 PCCS Air Cooler
3 Numerical Simulation
3.1 Grid
3.2 Numerical Methods
3.3 Data Reduction
3.4 Grid Independence Analysis
4 Flow Uniformity of PCCS Air Cooler
4.1 Flow Uniformity Comparison
4.2 Static Pressure Recovery
4.3 Flow Field Comparison
4.4 Optimization Analysis of Central-Type Arrangement
5 Conclusions
References
Research on the Innovative Development Trend and Competition Pattern of Domestic Nuclear Heating from the Perspective of Patents
1 Introduction
2 Research Methods
2.1 The Patent Search Process
2.2 The Patent Search Strategy
2.3 The Selection of Patent Information Features
2.4 The Determination of Competitive Situation Analysis Dimension
3 Conclusions
4 Proposal
References
Study on Beam Trajectory Deflection and Spatial Dose Distribution in Magnetic Resonance-Guided Proton Therapy
1 Introduction
2 Materials and Methods
2.1 Simulation Model
2.2 Magnetic Field Distribution Data
3 Results
3.1 Beam Trajectory Deflection with Fringe Magnetic Field
3.2 Study on the Spatial Dose Distribution of Secondary Particles Guided by Magnetic Fields
3.3 Study on Spatial Dose Distribution Before and After Beam Trajectory Correction
4 Conclusions
References
Analysis on the Global Competition of Digital Reactor from the Perspective of Patents
1 Introduction
2 Research Method
3 Research Contents
3.1 Global Patent Competition Landscape
3.2 Patent Competition Pattern in China
3.3 Patents in Key Technical Fields
4 Conclusions
References
A Monte Carlo Code Developed for Radiation Shielding Calculations Based on RMC
1 Introduction
2 Features Suited for Shielding Calculation
2.1 Gamma Transport and Coupled Transport
2.2 Flexible Fixed Source Definition and Calculation
2.3 Variance Reduction Function
2.4 Others
3 Validation and Application
3.1 PWR Model
3.2 Results
4 Summary
References
Performance Analysis and Optimization of Heat Pipe-Based Radiator for Space Fission Power System Thermal Management
1 Introduction
2 Design of Heat Pipe Radiator
3 Analysis of Heat Dissipation Process of Heat Pipe-Fin Unit
4 The Influence of Design Parameters on the Weight of Radiator
4.1 The Influence of lhpc on Radiator Weight
4.2 The Influence of lf on Radiator Weight
4.3 The Influence of Tf11 on Radiator Weight
5 Simulation Analysis of Heat Pipe Radiator
5.1 Analysis of Flow Characteristics of Alkali Metals
5.2 Analysis of Heat Transfer Characteristics of Alkali Metals
6 Conclusion
References
Study on Flow and Heat Transfer of Liquid Gallium
1 Introduction
2 Research Objectives
2.1 Geometric Model
2.2 Calculation Parameters
3 Heat Transfer Model
3.1 Liquid Gallium Physical Properties Calculation Model
3.2 Turbulence Model
3.3 Heat Transfer Model
4 Results and Discussion
4.1 Effect of Inlet Velocity
4.2 Effect of Heat Flux
4.3 Effect of Tube Diameter
4.4 Effect of Fluids
5 Conclusions
References
Assessing the Effect of Some ATF Materials and Uncertainties on Their Properties Under Normal Operation Conditions by Means of the Transuranus Code
1 Introduction
2 Material Properties Implemented in Transuranus
3 Case Description
4 Results and Discussion
4.1 Effect of ATF Materials
4.2 Uncertainty Analysis
5 Summary and Conclusions
References
Author Index

Citation preview

Springer Proceedings in Physics 285

Chengmin Liu Editor

Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3 PBNC 2022, 1–4 November, Beijing & Chengdu, China

Springer Proceedings in Physics

285

Indexed by Scopus The series Springer Proceedings in Physics, founded in 1984, is devoted to timely reports of state-of-the-art developments in physics and related sciences. Typically based on material presented at conferences, workshops and similar scientific meetings, volumes published in this series will constitute a comprehensive up to date source of reference on a field or subfield of relevance in contemporary physics. Proposals must include the following: – Name, place and date of the scientific meeting – A link to the committees (local organization, international advisors etc.) – Scientific description of the meeting – List of invited/plenary speakers – An estimate of the planned proceedings book parameters (number of pages/articles, requested number of bulk copies, submission deadline). Please contact: For Americas and Europe: Dr. Zachary Evenson; [email protected] For Asia, Australia and New Zealand: Dr. Loyola DSilva; [email protected]

Chengmin Liu Editor

Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3 PBNC 2022, 1–4 November, Beijing & Chengdu, China

Editor Chengmin Liu Nuclear Power Institute of China Chengdu, China

ISSN 0930-8989 ISSN 1867-4941 (electronic) Springer Proceedings in Physics ISBN 978-981-19-8898-1 ISBN 978-981-19-8899-8 (eBook) https://doi.org/10.1007/978-981-19-8899-8 © The Editor(s) (if applicable) and The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 This work is subject to copyright. All rights are solely and exclusively licensed by the Publisher, whether the whole or part of the material is concerned, specifically the rights of translation, reprinting, reuse of illustrations, recitation, broadcasting, reproduction on microfilms or in any other physical way, and transmission or information storage and retrieval, electronic adaptation, computer software, or by similar or dissimilar methodology now known or hereafter developed. The use of general descriptive names, registered names, trademarks, service marks, etc. in this publication does not imply, even in the absence of a specific statement, that such names are exempt from the relevant protective laws and regulations and therefore free for general use. The publisher, the authors, and the editors are safe to assume that the advice and information in this book are believed to be true and accurate at the date of publication. Neither the publisher nor the authors or the editors give a warranty, expressed or implied, with respect to the material contained herein or for any errors or omissions that may have been made. The publisher remains neutral with regard to jurisdictional claims in published maps and institutional affiliations. This Springer imprint is published by the registered company Springer Nature Singapore Pte Ltd. The registered company address is: 152 Beach Road, #21-01/04 Gateway East, Singapore 189721, Singapore

Acknowledgement

Thanks to all members of the PBNC 2022 Conference Committees.

Contents

On the Issues for Legislation of Spent Fuel in China . . . . . . . . . . . . . . . . . . . . . . . Jiu Liu Identification and Investigation on Thermal Hydraulic Phenomena Related to Core Make-Up Tank . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Jilin Tang, Shuguang Wang, Yusheng Liu, Sichao Tan, and Dongyang Li Alara Methodology for Reactor Building Shielding Design in HPR1000 PWR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Guanghao Zeng, Yonghai Zhou, Qianqian Huang, Shouhai Yang, and Weifeng Lv The Study of Shielding Benchmark of Heating Exchanger for Operating NPP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zeng Guanghao, Yang Shouhai, Huang Xinming, Jiang Zhenyu, and Gong Quan

1

11

23

38

An Anisotropic Porous Model for Heat Exchanger Modeling of Fluoride-Salt-Cooled High-Temperature Advanced Reactor -- FuSTAR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Xinyu Li, Dalin Zhang, Xinze Li, Xingguang Zhou, Xinan Wang, Tongan Yang, Wenxi Tian, and Suizheng Qiu

47

Study on the Prediction of Critical Heat Flux in Uniformly Heated Round Tube by Multilayer Perceptron . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Xiang Zou, Lei Lei, Ma GuoQiang, Jiao Feng, and Chen Shijun

59

Neutronics and Thermal-Hydraulics Coupling Analysis of Integral Inherently Safe Fluoride-Salt-Cooled High-Temperature Advanced Reactor - Fustar . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Xingguang Zhou, Dalin Zhang, Xinyu Li, Xin Min, Wenqiang Wu, Lei Zhou, Wenxi Tian, and Suizheng Qiu

73

Application of Light Curtain Test Technology in Dynamic Deformation Test of Rotating Machinery Rotor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Han Xuan and Yi Jianhua

88

Influence of Social Responsibility of Nuclear Power Companies on Public Acceptance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zhong Xu and Jianhong Ma

96

viii

Contents

Local Correlation and NpNn Linearity of Nuclear Electric Moments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Y. Xiao, D. Liu, Z. Z. Qin, and Y. Lei

111

Research on Financial Risk Analysis of Nuclear Power Project Based on Monte Carlo Simulation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Chunhua Lu, Xiaoyuan Lin, and Ruomin Zhang

121

Research on Low Carbon and Energy Saving Technology Path of Nuclear Power HVAC System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yang Li, Guo Chuang Chen, Zhang Xin, and Wang Chong

127

Multi-objective Optimization Design of Plate Heat Exchanger for Spent Fuel Pool Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yuquan Shang, Weiguang Zhao, and Changqi Yan

134

Fuel Economy Analysis of ATF Assembly with SiC Cladding and UO2 (BeO) Pellets in CPR1000 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Tingting Zou, Wei Gao, and Xin Wang

147

Thermodynamic Analysis and Optimization of Nuclear Closed Air Brayton Cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Huawei Fang, Jingwei Yi, Xiaoyu Zhang, Tiebo Liang, Yiran Qian, Xin Tang, and Weixiong Chen

160

Affecting Factors and Aspects for Improving Customized Nuclear Public Acceptance Strategies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Haomiao Lin

166

Initial Response Study of Nuclear Power Plant Operations After Local COVID-19 Outbreak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Li Lianhai, Ren Xiaojiang, Zhang Xianggui, Wu Wenqi, and Yuan Xia

176

Discussion on Nuclear Arms Control Based on the International Situation in 2022 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yanfeng Lyu and Xuesheng Lyu

188

SNSTC’s Capabilities and Practices on Performance Testing for Nuclear Security System and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Wang Shuo, Yang Changjie, Lu Hong, He Jialin, and Chen Chen

193

Practical Analysis of Radon Protection Methods in an Underground Project . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Jie Tian, Guobo Zhong, Fei Wu, and Xiangwei Wang

198

Contents

ix

Numerical Calculation of Flow Heat Transfer in a Single Square Tube Under a Blockage Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Daheng Li, Shuwen Yu, Ahmed A. Ghani, and Changhong Peng

203

Research on the Siting Problem of Economy-Oriented Small Heating Nuclear Reactor Based on System Dynamics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Haitao Luo, Changle An, Lu Zhao, and Ruomin Zhang

214

Technical and Economic Analysis of Nuclear Heating Based on HPR1000 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Qian Yu, Jiang Hu, Lijuan Chen, Xin Shang, and Mei Rong

225

The Application of Renal Dynamic Imaging in Measuring Renal Function of En-Bloc Pediatric Kidneys Transplanted into Recipients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Ruolin Wu, Daijuan Huang, Zhendi Wang, Kun Li, Fan Hu, Cheng Wan, Yajing Zhang, Xiaoli Lan, Zairong Gao, and Xiaotian Xia

233

Study of Cost Index System for Nuclear Power Plant . . . . . . . . . . . . . . . . . . . . . . Li Wenan, Hu Jiang, Shi Yang, Shang Xin, and Rong Mei

246

Application of Intelligent Innovation in Accident Procedure of Nuclear Power Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Jie Zhang and Yang Lu

253

Experimental Study on Thermal-Hydraulic Characteristics of Air Heat Exchanger for Sodium Cooled Fast Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Lina Zhu, Zhenjia Chen, and Yuanwu Ye

263

Effect of Coating Temperature on Microstructure and Properties of the SiC Layer in TRISO-Coated Particles . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Qiang Zeng, Zongbei He, Xiaoqiang Pan, Kai Wang, Chong Yu, Jiancai Peng, and Liang Zhang An Improved Super-Resolution Model for Bubble Feature Extraction Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Heng Zhang, Lingpeng Zhong, Qin Hang, Xue Lyu, Bo Liu, Jinchao Liu, and Guoyin Wang Analysis of Current Status and Development Trend of Nuclear District Heating . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yanrui Li, Chao Chen, Jian Hu, Jiqiang Su, and Hongjun Liu

275

286

301

x

Contents

Research on High-Speed Cold Cutting Technology for Combined Spider . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Guo-peng Qin, Li-ying Zhang, Yan-ru Qin, and Lan-yan Tan Research and Development Practice of Equipment Condition Monitoring and Intelligent Diagnosis Platform Based on Big Data and AI at TNPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . He Fang and Qiang Yang Study on the Design Standard of Water-Intake Breakwaters in Coastal Nuclear Power Plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Bilu Xiong, Zaohua Zheng, and Zhi Lin A Study of Stability and Coagulation Behavior of Broth Based the Internal Gelation Process for Uranium Dioxide Microspheres . . . . . . . . . . . . Te Wang, Zhangxian Lu, Xinhao Li, Feixiang Zeng, Jia Li, and Yali Shang

308

317

325

332

The Preliminary Thermo-hydraulics Calculation of Unfolding Process of the Lotus Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yantao Luo, Tian Zhang, and Xiang Wang

341

Key Technology of a New Potential Economical Desalination Method for Small Nuclear Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Siguang Li, Yiran Gan, Yi Li, and Tiebo Liang

356

Research on On-Line Measurement Technology of Fuel Rod Deformation Based on LVDT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Lingjie Xu, Junping Si, Hui Zhang, and Wenlong Zhang

362

Numerical Study on Flow Distribution Characteristics of Parallel Square Flow Channels Under Blockage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Daheng Li, Xinyan Xu, Shuwen Yu, Jun Xiao, and Changhong Peng

369

Computer Package for Calculating the Fast-Neutron Fluence in Reactor Pit Based on the Reactor Operational History and the Monte-Carlo Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yongjun Gao, Fang Liu, Feifei Tang, and Feng Xue Research on Classification and Coding of Nuclear Power Plant Status Reports Based on Machine Learning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Ning Dong, Lian Zhang, Hankun Cai, Pan Hu, Jianjun Sun, and Xiaodong Wang

383

391

Contents

Atomic Scale Simulation on Liquid Metal Embrittlement Induced by Segregation of Lead Element in Lead-Cooled Fast Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zhu Bida

xi

406

Application of Construction Phase Process Cost Management in Nuclear Power Projects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Jinxu Shi and Lili Zhang

414

Characterization of Residual Uranium Resources in the Decommissioned Final Mining Area of a Uranium Deposit in Xinjiang . . . . . . . . . . . . . . . . . . . . . . Xu Ying, Chen Bihua, Chen Hong, Zhao Linxin, and Jia Hao

422

Study on Influence of Process Parameters on Image Quality in X-ray Digital Radiography Inspection of Fuel Rod Weld . . . . . . . . . . . . . . . . . . . . . . . . . Yu Wenxin and Cao Hui

434

Construction of CDS/TIO2 /HGS Composite Materials and Photocatalytic Reduction of Hexavalent Uranium in Wastewater . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yan Song, Shusen Chen, Yantao Su, Ziming Li, Fengju Wang, Yangfei Gou, Haizhen Wang, and Hua Chang

443

Experimental Study on Shell Side Flow Distribution of Fast Reactor Steam Generator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Xing Shuai, Wu Zhiguang, Zhou Lijun, Zhang Ao, and Wang Bo

453

Vibration Analysis of Heating-Used Steam Extraction Pipelines in Nuclear Power Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Wu Yanhui, Zhuang Yaping, and Ji Tengfei

464

Experimental Study on Aerosol Removal Effected by Spray Characterization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Peizheng Hu, Zhen Xu, Junlong Wang, Lili Tong, and Xuewu Cao

473

Study on Public Acceptance of Restart of Inland Nuclear Power Plants Based on TAM/TPB Integration Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Wangping Li, Dongyan Su, and Xiaoxing Chen

488

Low Temperature Performance of LBE Oxygen Sensors with Different Reference Electrodes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Mu-Ran Qi, Yun-Peng Tang, Hui-Ping Zhu, Xiao-Bo Li, Rui-Xian Liang, Yun-Gan Zhao, Yi-Feng Wang, and Feng-Lei Niu

508

xii

Contents

The Development of the Combined Fission Matrix Theory in Burnup Calculation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Ruishuang Gao, Xiaojing Liu, and Donghao He Study of the Electric Device of RRI019/020VN in Qinshan Nuclear Power Plant Cannot Close Completely in Long Signal Valve Closing . . . . . . . . Deng Peiqiang, Zha Weihua, Liu Dongbing, Li Ping, Jin Yidan, Yang Kaiyu, Qian Wei, and Wu Hao

519

530

Numerical Study of Influence of 15 N Enrichment on Burnup Performances for a Heat Pipe Cooled Traveling Wave Reactor . . . . . . . . . . . . . . Kunfeng Ma and Po Hu

538

Parallelization of Thermal Hydraulic Real-Time Simulation Program Based on Two-Phase Drift Flux Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Tingwei Ren and Lei Li

549

Simulation of Photon-Counting Detector Based on OpenModelica . . . . . . . . . . . Shishuai Wang and Xiang Wang Research on Boiling and Coupled Heat Transfer of OTSG with Drift Flow Single Fluid Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zhenguo Zhang, Xiaochang Li, Sichao Liu, Sichao Tan, Hai Zhao, and Ruifeng Tian

562

578

Study on the Crystallization Rate of Uranium Peroxide Based on Ultrasound and Microchannel Technology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Wei Gong, Yongdong Xu, Wenxing Cui, and Yingxi Zhu

590

Construction of Nuclear Emergency Awareness System Based on “Cloud-Edge-Terminal” . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Weidong Yang, Jie Zhou, and Shixian Liu

600

Sodium Spray Fire Analysis with Combustion Space Multi-node Model for Sodium-Cooled Fast Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zhenyu Zou, Lili Tong, and Xuewu Cao

607

Analysis of Fission Product Diffusion Behavior of Fully Ceramic Micro-encapsulated Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Liying Zhang, Defeng Yang, Xiaoxia Wang, Aijun Mi, and Liangzhi Cao

624

Optimization of Sintering Parameters of Lead Oxide Particles for Solid Oxygen Control in Lead Cooled Fast Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Xu Tu, Huiping Zhu, Yunpeng Tang, Muran Qi, Xudong Liu, and Tinxu Yan

634

Contents

xiii

Research on Fission Products Selection in the Primary Coolant of PWR During Normal Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Mingliang Dai, Pengtao Fu, and Zhenyu Jiang

646

Development of a Depletion Code with a Control-Rod-Adjusting Program Based on OpenMC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zhiqiang Wu, Pengyu Chen, Jinsen Xie, Zedong Zhou, and Tao Yu

656

Study of Fine Particle Deposition of Porous Medium in Heat Pipe of Heat Pipe Reactor Under Gravity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Dong Wei, Tao Zhou, Shang Mao, Wenbin Liu, Chunhui Xue, Huaichang Lu, and Yitong Zhang Recommendations to Improve the Public Acceptance of Nuclear Environmental Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Jiahang Zhang Coupled Neutronics and Thermal-Fluid Calculation for Prismatic High-Temperature Gas-Cooled Reactor Core at Steady State . . . . . . . . . . . . . . . . Juanjuan Guo, Kun Yan, Pu Tong, Fang He, Shanfang Huang, and Kan Wang

664

682

688

Study on the Preparation of Metallographic Samples and the Characterization of Volume Content and Size Distribution of Inclusions in Uranium Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yanhua Du, Hong Guo, Chunxia Chi, Yueqing Qian, Jinming He, Wei Liu, and Hongbo Wang

706

Development Background and Research Progress of UN-U3 Si2 Composite Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Qiming Sun, Ye Yang, Mingyang Li, Pengbo Ji, and Wentao Liu

715

Study on Fabrication and Characterization of SiCf /SiC Composite Cladding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Qiming Sun, Yongheng Lu, Yueqing Qian, Yu Li, and Ying Meng

724

Life-Cycle Cost Study for a Near-Surface Disposal Repository of Low-Level Waste in China . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Hong-Ying Yu, Dong Zhao, Xin Shang, Fang Wang, and Li He

733

Numerical Study on External Environmental Radiation of Nuclear Ramjet Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yongjiang Wen, Chenglong Wang, Wenxi Tian, Guanghui Su, and Suizheng Qiu

742

xiv

Contents

Optimization on Design of Logic Degradation for Safety I&C System of Nuclear Power Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yan-nong Wu, Yi-min Zhang, Wei Zhu, Qin-mai Hou, and Shi-xian Liu Study on Gamma-Ray Spectra Feature Recognition and Isotope Composition Analysis of Plutonium Based on Convolutional Neural Networks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Haocheng Zhao, Lei Bai, and Lixia He

759

768

Study on the Coupled Heat Transfer Characteristics of Liquid Lead-Bismuth Eutectic and Supercritical CO2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . Tao Zhou, Yao Yao, Yi Jiang, and Dong Wei

780

Adsorption Behaviors of Hydrogen on Equal Atomic Ratio TiZrV Film Applied in AB-BNCT by Density Functional Theory Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yaocheng Hu, Jie Wang, Yaqiong Su, Qiuyu Sun, Qingyu Si, Yupeng Xie, Pengyu Huang, Xiaoqing Liang, and Sheng Wang

792

Research on Interface Technology of Coupling Thermal-Hydraulics and Other Codes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Shuyong Zhou, Zesong Huang, Junming Tang, Bing Hao, Zhiguo Qin, Wei Zhen, and Wei Gao Development and Verification of Transport-Activation Coupling Capability in CosRMC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Shengzhe Wang, Shichang Liu, Guoping Quan, Chenglin Zhu, and Yixue Chen Research on Shielding Deep Penetration Calculation Based on MC Variance Reduction Techniques . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Aikou Sun, Zhenping Chen, Leiming Li, Chengwei Liu, and Tao Yu Prospects for Next Generation Nuclear Power System . . . . . . . . . . . . . . . . . . . . . Yaoxiang Zhang, Hongxing Yu, Bangyang Xia, and Wenjie Li Safety Analysis of Internal Flooding Under Main Feedwater Pipeline Rupture Accident Based on Small Modular Reactor ACP 100 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Liang Ding, Xiaoyun Huang, Chenhui Wang, and Ji Xing Research on Nuclear Seawater Desalination Technology Based on Small Modular Reactor ACP100 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Liang Ding, Ran Zhang, Chenhui Wang, and Ji Xing

800

810

821

834

843

853

Contents

Research Status and Prospect of Comprehensive Utilization Technology of Nuclear Energy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Ji Xing, Liang Ding, Chenhui Wang, Feixue Liu, and Ziwei Zhao Hydrogen Adsorption Mechanism of Non-equal Atomic Ratio TiZrV NEG Films Surface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Qiuyu Sun, Jie Wang, Yaqiong Su, Yaocheng Hu, Qingyu Si, Yupeng Xie, Pengyu Huang, Xiaoqing Liang, and Sheng Wang Study of Fast-Start Heat Transfer Characteristics of Potassium Heat Pipe . . . . . Wenbin Liu, Tao Zhou, Dong Wei, Shang Mao, Chunhui Xue, Huaichang Lu, and Yao Yao

xv

867

881

890

Prediction of CHF in Narrow Rectangular Channel Based on GA-BP Neural Network . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Enpei Wang and Lei Li

904

Study on 131 I Source Term in Reactor Coolant Under Shutdown Transient Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Jing Fu-ting, L. V. Huan-wen, X. I. A. Ming-Ming, Y. E. Zhang-Han, and Gao Xi-long

913

Study on Aggregation Forming of Cathode Liquid Cerium Metal in Molten Salt Electrorefining Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zhengang Tian, Yiqiang Zhong, Zejie Wang, Meixun Qin, Xiaoqi Li, Shuwen Meng, Yiwu Chen, Jinsheng Liu, Yingxi Zhu, Kaifa Du, Huayi Yin, and Dihua Wang

920

Thermal Engineering Calculation of Control Drum Misoperation of Heat Pipe Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Peng liu, Tao Zhou, Huaichang Lu, and Tianyu Gao

932

CFD Analysis on the Multiphase Flow of the Wickless Sodium Heat Pipe with Non-condensable Gas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zilin Su, Yuchuan Guo, Zeguang Li, and Kan Wang

945

Influence of Impedance Characteristics of Perforated Plate on Pump-Induced Pulsation in Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Sun Yu, Cai Fengchun, Qian Sheng, Huangfu Yuzhao, and Zhang Ke

957

Development of Design and Verification Platform for Heat Pipe Cooled Nuclear Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Chao Tan, Fuchang Shan, Suizheng Qiu, Chenglong Wang, and Zeqin Zhang

967

xvi

Contents

Development of Curriculum for International Students in the Field of Nuclear Energy Engineering at Harbin Engineering University . . . . . . . . . . . . Gao Puzhen, Zhao Donglei, and Cui Yuan

980

Technological Upgrading Analysis on Diversity Transformation of Reactor Protection System in a Fast Reactor Project . . . . . . . . . . . . . . . . . . . . . Liu XingQing, Zhang HaiBin, Tan Pin, Li Bin, and Chen DeLin

986

Study on Top Architecture Design of Nuclear Power System Based on an Incremental Iteration Generation Method . . . . . . . . . . . . . . . . . . . . . . . . . . . Pan Xinxin, Song Chunjing, and Ming Yao

993

Study on Optimization of Economic Evaluation Methods for Generation III Reactors in China . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1007 Juanjuan Guo, Di Shen, and Fang He Preliminary Design and Thermal Analysis of Intermediate Heat Exchanger Based on Irradiation Test Loop with Alternative Coolants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1021 Yixiong Sun, Liqing Qiu, Liang Zhang, Wenhua Yang, and Sheng Sun Development of High-Fidelity Neutron Transport Solver of Alpha . . . . . . . . . . . 1036 Hang Zou, Qian Zhang, Peitao Song, Liang Liang, and Qiang Zhao Study on Welding Process Between the Hexagonal Tube and the Grid Frame Component of MOX Fuel Assembly in CEFR . . . . . . . . . . . . . . . . . . . . . . 1051 Yonglong Yu, Liang Guo, Wande Zhang, Tongyu Zhu, and Shunxiao Zhang Unattended Monitoring Apparatus Data Analysis Technicals Used for Nuclear Safeguards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1065 Xiao Fan and Lixia He Economic Analysis of Nuclear Cogeneration System Based on High-Temperature Gas-Cooled Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1071 Yan Luo, Gen Li, and Tao Lu Study on Social Stability Risk Assessment of Nuclear Facilities . . . . . . . . . . . . . 1082 Rongxu Zhu, Feng Zhao, Xiaofeng Zhang, and Meng Zhang Investigation on the Fabrication of High-Density Micro Pits on Zirconium Alloy by Micro Imprinting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1090 Tong Niu and Yuanxin Luo

Contents

xvii

Application and Analysis of Airborne Electromagnetic Method in Engineering Exploration of High-Level Radioactive Waste Repository . . . . . 1102 Pei-jian Wang, Wei Zhang, Jiangkun Li, Guomin Hu, and Qisen Zheng Optimization of Power Supply Scheme and Economic Benefit Analysis of a Project in a Newly-Built Large Nuclear Chemical Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1107 Shuo Gao, Zhi Huang, and Shizhong Tian Air-Cooler Structure Optimization of HTGR Passive Cavity Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1112 Haiqi Qin, Xiaowei Li, Li Zhang, and Xinxin Wu Research on the Innovative Development Trend and Competition Pattern of Domestic Nuclear Heating from the Perspective of Patents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1124 Anna Gao, Yading Zhang, Ran Su, and Ningyuan Wang Study on Beam Trajectory Deflection and Spatial Dose Distribution in Magnetic Resonance-Guided Proton Therapy . . . . . . . . . . . . . . . . . . . . . . . . . . . 1139 Xiao-Qing Ren, Ming Wang, Xin-Chen Wang, Meng Li, Guo-Dong Li, and Lei Zhang Analysis on the Global Competition of Digital Reactor from the Perspective of Patents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1151 Dongbin Li, Zaojing Chen, Weiwei Hu, Chongyu Su, and Chenglin Sun A Monte Carlo Code Developed for Radiation Shielding Calculations Based on RMC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1167 Dongchuan Ying, Huanwen Lyu, Songqian Tang, Xiaotong Shang, and Feng Xiao Performance Analysis and Optimization of Heat Pipe-Based Radiator for Space Fission Power System Thermal Management . . . . . . . . . . . . . . . . . . . . 1174 Zengen Li, Haochun Zhang, Dong Zhang, Qi Wang, and Yan Xia Study on Flow and Heat Transfer of Liquid Gallium . . . . . . . . . . . . . . . . . . . . . . . 1191 Shang Mao, Tao Zhou, Peng Xu, and Lanyu Zhou Assessing the Effect of Some ATF Materials and Uncertainties on Their Properties Under Normal Operation Conditions by Means of the Transuranus Code . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1200 P. Van Uffelen, A. Schubert, and Z. Soti Author Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1215

Contributors

Changle An Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai, China Zhang Ao China Nuclear Power Operation Technology Corporation, Wuhan of Hunan Prov., China Lei Bai China Institute of Atomic Energy, Beijing, China Zhu Bida Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Chen Bihua Beijing Research Institute of Chemical Engineering Metallurgy, CNNC, Beijing, China Li Bin NingDe, FuJian, China Wang Bo Dongfang Electric Co. I.Td, Chengdu, China Hankun Cai Research Institute of Nuclear Power Operation, Wuhan, China Liangzhi Cao Xi’an Jiaotong University, Xi’an, China Xuewu Cao School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai, China Hua Chang CNNC Key Laboratory on Uranium Extraction from Seawater, Beijing Research Institute of Chemical Engineering and Metallurgy, Beijing, China Yang Changjie State Nuclear Security Technology Center (SNSTC), Beijing, China Chao Chen China Institute of Nuclear Industry Strategy, Beijing, China Chen Chen State Nuclear Security Technology Center (SNSTC), Beijing, China Guo Chuang Chen Hualong International Nuclear Power Technology Co. Ltd., Shen Zhen, Guang Dong, China Lijuan Chen China Nuclear Power Engineering Co., Ltd., Beijing, China Pengyu Chen School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China; Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China Shusen Chen CNNC Key Laboratory on Uranium Extraction from Seawater, Beijing Research Institute of Chemical Engineering and Metallurgy, Beijing, China Weixiong Chen State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China

xx

Contributors

Xiaoxing Chen School of Economics, Management and Law, University of South China, Hengyang, China Yiwu Chen China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Yixue Chen School of Nuclear Science and Engineering, North China Electric Power University, Beijing, China Zaojing Chen China Institute of Nuclear Industry Strategy, Beijing, China Zhenjia Chen China Institute of Atomic Energy, Beijing, China Zhenping Chen School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China; Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China Chunxia Chi China North Nuclear Fuel Co., Ltd., Baotou, Inner Mongolia, China Wang Chong Hualong International Nuclear Power Technology Co. Ltd., Shen Zhen, Guang Dong, China Song Chunjing Shanghai Nuclear Engineering Research and Design Institute Co. Ltd, Shanghai, China Wenxing Cui The Fourth Filial Company of 404 Company Limited, CNNC Lanzhou, Gansu, China Mingliang Dai China Nuclear Power Technology Research Institute, Shenzhen, China Chen DeLin NingDe, FuJian, China Liang Ding Harbin Engineering University, Harbin, China; China Nuclear Power Engineering Co., LTD, Beijing, China Ning Dong Research Institute of Nuclear Power Operation, Wuhan, China Liu Dongbing CNNC Nuclear Power Operation Management Co., Ltd, Jiaxing, Zhejiang, China Zhao Donglei College of Nuclear Science and Technology, Harbin Engineering University, Harbin, Heilongjiang, China Kaifa Du School of Resource and Environmental Science, Wuhan University, Wuhan, China Yanhua Du CNNC Key Laboratory on New Materials Research and Application Development, Baotou, Inner Mongolia, China; China North Nuclear Fuel Co., Ltd., Baotou, Inner Mongolia, China Xiao Fan Beijing, China He Fang Jiangsu Nuclear Power Corporation, Lianyungang, Jiangsu, China

Contributors

xxi

Huawei Fang Key Laboratory of Nuclear Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, Sichuan, China Jiao Feng Nuclear and Radiation Safety Center, MEE, Beijing, China Cai Fengchun Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan, China Pengtao Fu China Nuclear Power Technology Research Institute, Shenzhen, China Jing Fu-ting Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Yiran Gan Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan, China Anna Gao China Institute of Nuclear Industry Strategy Beijing, Beijing, China Ruishuang Gao Shanghai Jiao Tong University, Shanghai, China Shuo Gao China Nuclear Power Engineering Co., Ltd, Beijing, China Tianyu Gao Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, China Wei Gao China Nuclear Power (Beijing) Simulation Technology Co., Ltd., Shenzhen, Guangdong, China; China Nuclear Power Technology Research Institute Co., Ltd., Shenzhen, Guangdong, China Yongjun Gao Suzhou Nuclear Power Research Institute of CGNPC, Suzhou, Jiangsu, China Zairong Gao Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Ahmed A. Ghani University of Science and Technology of China, Hefei, China Wei Gong The Fourth Filial Company of 404 Company Limited, CNNC Jiayuguan, Gansu, China Yangfei Gou CNNC Key Laboratory on Uranium Extraction from Seawater, Beijing Research Institute of Chemical Engineering and Metallurgy, Beijing, China Zeng Guanghao Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China Hong Guo CNNC Key Laboratory on New Materials Research and Application Development, Baotou, Inner Mongolia, China; China North Nuclear Fuel Co., Ltd., Baotou, Inner Mongolia, China

xxii

Contributors

Juanjuan Guo China Institute of Nuclear Industry Strategy, Beijing, China Liang Guo The 404 Company Limited, Lanzhou, Gansu, China Yuchuan Guo Department of Engineering Physics, Tsinghua University, Beijing, China Ma GuoQiang Nuclear and Radiation Safety Center, MEE, Beijing, China Zhang HaiBin NingDe, FuJian, China Qin Hang Chongqing University of Posts and Telecommunications, Chongqing, China Bing Hao China Nuclear Power (Beijing) Simulation Technology Co., Ltd., Shenzhen, Guangdong, China Jia Hao Beijing Research Institute of Chemical Engineering Metallurgy, CNNC, Beijing, China Wu Hao CNNC Nuclear Power Operation Management Co., Ltd, Jiaxing, Zhejiang, China Donghao He Shanghai Jiao Tong University, Shanghai, China Fang He China Institute of Nuclear Industry Strategy, Beijing, China Jinming He CNNC Key Laboratory on New Materials Research and Application Development, Baotou, Inner Mongolia, China; China North Nuclear Fuel Co., Ltd., Baotou, Inner Mongolia, China Li He Economic Evaluation Division, China Nuclear Power Engineering Co., Ltd., Beijing Institute of Industrial Engineering, Beijing, China Lixia He China Institute of Atomic Energy, Beijing, China Zongbei He Nuclear Power Institute of China, Chengdu, SiChuan, China Chen Hong Beijing Research Institute of Chemical Engineering Metallurgy, CNNC, Beijing, China Lu Hong State Nuclear Security Technology Center (SNSTC), Beijing, China Qin-mai Hou NSC, Beijing, China L. V. Huan-wen Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Daijuan Huang Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Pengyu Huang School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Qianqian Huang Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China

Contributors

xxiii

Shanfang Huang Tsinghua University, Beijing, China Xiaoyun Huang China Nuclear Power Engineering Co., LTD, Beijing, China Zesong Huang China Nuclear Power (Beijing) Simulation Technology Co., Ltd., Shenzhen, Guangdong, China Zhi Huang China Nuclear Power Engineering Co., Ltd, Beijing, China Fan Hu Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Guomin Hu Airborne Survey and Remote Sensing Center of Nuclear Industry, Shijiazhuang, China Jian Hu China Institute of Nuclear Industry Strategy, Beijing, China Jiang Hu China Nuclear Power Engineering Co., Ltd., Beijing, China Pan Hu Research Institute of Nuclear Power Operation, Wuhan, China Peizheng Hu School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai, China Po Hu School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai, China Weiwei Hu China Institute of Nuclear Industry Strategy, Beijing, China Yaocheng Hu Shaanxi Engineering Research Center of Advanced Nuclear Energy, Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, School of Nuclear Science and Technology, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Cao Hui CNNC Jianzhong Nuclear Fuel Co., Ltd, Yibin, Sichuan, China Pengbo Ji China North Nuclear Fuel Co., Ltd., Baotou, China He Jialin State Nuclear Security Technology Center (SNSTC), Beijing, China Hu Jiang China Nuclear Power Engineering Co., Ltd, Beijing, People’s Republic of China Yi Jiang Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Beijing, China; School of Nuclear Science and Engineering, North China Electric Power University, Beijing, China Zhenyu Jiang China Nuclear Power Engineering Co., Ltd., Shenzhen, China Yi Jianhua Institute of Physical and Chemical Engineering of Nuclear Industry, Tianjin, China Yang Kaiyu CNNC Nuclear Power Operation Management Co., Ltd, Jiaxing, Zhejiang, China

xxiv

Contributors

Zhang Ke Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan, China Xiaoli Lan Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Lei Lei Nuclear and Radiation Safety Center, MEE, Beijing, China Y. Lei School of National Defense Science and Technology, Southwest University of Science and Technology, Mianyang, China Liang Liang Nuclear Energy Creative Power Co., Ltd., Xi’an, Shaanxi, China Rui-Xian Liang China Institute of Atomic Energy, Beijing, China Tiebo Liang Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan, China; Key Laboratory of Nuclear Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, Sichuan, China Xiaoqing Liang School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Li Lianhai Jiangsu Nuclear Power Corporation, Lianyungang, China Daheng Li School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, China Dongbin Li China Institute of Nuclear Industry Strategy, Beijing, China Dongyang Li Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin, Heilongjiang, China Gen Li School of Mechanical and Electrical Engineering, Beijing University of Chemical Technology, Beijing, China Guo-Dong Li School of Nuclear Technology and Automation Engineering, Chengdu University of Technology, Chengdu, Sichuan, China Jia Li Nuclear Power Institute of China, Chengdu, China Jiangkun Li Airborne Survey and Remote Sensing Center of Nuclear Industry, Shijiazhuang, China Kun Li Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Lei Li Key Discipline Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang, China; Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin, Heilongjiang, People’s Republic of China Leiming Li School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China;

Contributors

xxv

Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China Meng Li School of Nuclear Technology and Automation Engineering, Chengdu University of Technology, Chengdu, Sichuan, China Mingyang Li CNNC Key Laboratory of New Materials Research and Application Development, Baotou, China; China North Nuclear Fuel Co., Ltd., Baotou, China Siguang Li Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan, China Wangping Li School of Economics, Management and Law, University of South China, Hengyang, China Wenjie Li Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Xiao-Bo Li China Institute of Atomic Energy, Beijing, China Xiaochang Li Harbin Engineering University Harbin, Heilongjiang, China Xiaoqi Li China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Xiaowei Li Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, China Xinhao Li Nuclear Power Institute of China, Chengdu, China Xinyu Li Xi’an Jiaotong University, Xi’an, Shaanxi, China Xinze Li Xi’an Jiaotong University, Xi’an, Shaanxi, China Yang Li Hualong International Nuclear Power Technology Co. Ltd., Shen Zhen, Guang Dong, China Yanrui Li China Institute of Nuclear Industry Strategy, Beijing, China Yi Li Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan, China Yu Li CNNC Key Laboratory of Fabrication Technology of Fuel Assembly, Baotou, China; China North Nuclear Fuel Co., Ltd., Baotou, China Zeguang Li Department of Engineering Physics, Tsinghua University, Beijing, China Zengen Li School of Energy Science and Engineering, Harbin Institute of Technology, Harbin, China Ziming Li CNNC Key Laboratory on Uranium Extraction from Seawater, Beijing Research Institute of Chemical Engineering and Metallurgy, Beijing, China

xxvi

Contributors

Zhou Lijun China Institute of Atomic Energy, Beijing, China Haomiao Lin China National Nuclear Corporation Overseas Ltd., Beijing, China Xiaoyuan Lin Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai, China Zhi Lin China Nuclear Power Design Company (Shenzhen), Shenzhen, Guangdong, China Zhao Linxin Beijing Research Institute of Chemical Engineering Metallurgy, CNNC, Beijing, China Bo Liu Chongqing University of Posts and Telecommunications, Chongqing, China Chengwei Liu School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China; Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China D. Liu School of Science, Southwest University of Science and Technology, Mianyang, China Fang Liu Suzhou Nuclear Power Research Institute of CGNPC, Suzhou, Jiangsu, China Feixue Liu China Nuclear Power Engineering Co., LTD, Beijing, China Hongjun Liu China Institute of Nuclear Industry Strategy, Beijing, China Jinchao Liu China Nuclear Power (Bei Jing) Simulation Technology Co., Ltd, Beijing, China Jinsheng Liu China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Jiu Liu Harbin Engineering University, Harbin, Heilongjiang, China Shi-xian Liu NSC, Beijing, China Shichang Liu School of Nuclear Science and Engineering, North China Electric Power University, Beijing, China Shixian Liu Nuclear and Radiation Safety Center, Beijing, China Sichao Liu Harbin Engineering University Harbin, Heilongjiang, China Wei Liu CNNC Key Laboratory on New Materials Research and Application Development, Baotou, Inner Mongolia, China; China North Nuclear Fuel Co., Ltd., Baotou, Inner Mongolia, China Wenbin Liu Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, Jiangsu, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, Jiangsu, China;

Contributors

xxvii

National Engineering Research Center of Power Generation Control and Safety, Nanjing, Jiangsu, China Wentao Liu CNNC Key Laboratory of Fabrication Technology of Fuel Assembly, Baotou, China; China North Nuclear Fuel Co., Ltd., Baotou, China Xiaojing Liu Shanghai Jiao Tong University, Shanghai, China Xudong Liu Key Laboratory of Passive Nuclear Power Safety and Technology, North China Electric Power University, Beijing, China Yusheng Liu Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin, Heilongjiang, China; Nuclear and Radiation Safety Center, MEE, Beijing, China Peng liu Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, China Chunhua Lu Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai, China Huaichang Lu Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, Jiangsu, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, Jiangsu, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, Jiangsu, China Tao Lu School of Mechanical and Electrical Engineering, Beijing University of Chemical Technology, Beijing, China Yang Lu Suzhou Nuclear Power Research Institute Co., Ltd, Shenzhen, Guangdong, China Yongheng Lu CNNC Key Laboratory of Fabrication Technology of Fuel Assembly, Baotou, China; China North Nuclear Fuel Co., Ltd., Baotou, China Zhangxian Lu Nuclear Power Institute of China, Chengdu, China Haitao Luo Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai, China Yan Luo School of Mechanical and Electrical Engineering, Beijing University of Chemical Technology, Beijing, China Yantao Luo College of Nuclear Science and Technology, Harbin Engineering University, Harbin, China

xxviii

Contributors

Yuanxin Luo College of Mechanical and Vehicle Engineering, Chongqing University, Chongqing, China Weifeng Lv Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China Huanwen Lyu National Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, China Xue Lyu Chongqing University of Posts and Telecommunications, Chongqing, China Xuesheng Lyu China Institute of Atomic Energy, Beijing, China Yanfeng Lyu China Institute of Atomic Energy, Beijing, China Jianhong Ma Department of Psychology and Behavioral Sciences, Zhejiang University, Hangzhou, China Kunfeng Ma School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai, China Shang Mao Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, Jiangsu, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, Jiangsu, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, Jiangsu, China Rong Mei China Nuclear Power Engineering Co., Ltd, Beijing, People’s Republic of China Shuwen Meng China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Ying Meng China North Nuclear Fuel Co., Ltd., Baotou, China Aijun Mi China Nuclear Power Engineering Co, Beijing, China Xin Min Xi’an Jiaotong University, Xi’an, Shaanxi, China X. I. A. Ming-Ming Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Feng-Lei Niu China Institute of Atomic Energy, Beijing, China Tong Niu College of Mechanical and Vehicle Engineering, Chongqing University, Chongqing, China Xiaoqiang Pan Nuclear Power Institute of China, Chengdu, SiChuan, China Deng Peiqiang CNNC Nuclear Power Operation Management Co., Ltd, Jiaxing, Zhejiang, China Changhong Peng School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, China

Contributors

xxix

Jiancai Peng Nuclear Power Institute of China, Chengdu, SiChuan, China Tan Pin NingDe, FuJian, China Li Ping CNNC Nuclear Power Operation Management Co., Ltd, Jiaxing, Zhejiang, China Gao Puzhen College of Nuclear Science and Technology, Harbin Engineering University, Harbin, Heilongjiang, China Mu-Ran Qi China Institute of Atomic Energy, Beijing, China Muran Qi Key Laboratory of Passive Nuclear Power Safety and Technology, North China Electric Power University, Beijing, China Yiran Qian State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Yueqing Qian China North Nuclear Fuel Co., Ltd., Baotou, Inner Mongolia, China Guo-peng Qin CNNC Jianzhong Nuclear Fuel Co., Ltd, Yibin, Sichuang, China Haiqi Qin Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, China Meixun Qin China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Yan-ru Qin CNNC Jianzhong Nuclear Fuel Co., Ltd, Yibin, Sichuang, China Z. Z. Qin School of Science, Southwest University of Science and Technology, Mianyang, China Zhiguo Qin China Nuclear Power (Beijing) Simulation Technology Co., Ltd., Shenzhen, Guangdong, China Liqing Qiu Nuclear Power Institute of China, Chengdu, Sichuan, China Suizheng Qiu School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi, China Gong Quan Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China Guoping Quan National Energy Key Laboratory of Nuclear Power Software, State Power Investment Corporation Research Institute, Beijing, China Tingwei Ren Key Discipline Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang, China Xiao-Qing Ren School of Nuclear Technology and Automation Engineering, Chengdu University of Technology, Chengdu, Sichuan, China Mei Rong China Nuclear Power Engineering Co., Ltd., Beijing, China

xxx

Contributors

A. Schubert European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, Karlsruhe, Germany Fuchang Shan CNNC Key Laboratory on Nuclear Industry Simulation, Wuhan, Hubei, China Xiaotong Shang Department of Engineering Physics, Tsinghua University, Beijing, China Xin Shang Economic Evaluation Division, China Nuclear Power Engineering Co., Ltd., Beijing Institute of Industrial Engineering, Beijing, China Yali Shang Nuclear Power Institute of China, Chengdu, China Yuquan Shang Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin, China Di Shen China Institute of Nuclear Industry Strategy, Beijing, China Qian Sheng Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan, China Jinxu Shi The Fourth Research and Design Engineering Corporation of CNNC, Shijiazhuang, Hebei, China Chen Shijun Suzhou Nuclear Power Research Institute Co., Ltd., Shenzhen, China Yang Shouhai Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China Xing Shuai China Institute of Atomic Energy, Beijing, China Wang Shuo State Nuclear Security Technology Center (SNSTC), Beijing, China Junping Si Nuclear Power Institute of China, Chengdu, Sichuan, China Qingyu Si School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Peitao Song Radiation Dosimetry Laboratory, Department of Health Physics, China Institute for Radiation Protection, Taiyuan, Shanxi, China Yan Song CNNC Key Laboratory on Uranium Extraction from Seawater, Beijing Research Institute of Chemical Engineering and Metallurgy, Beijing, China Z. Soti European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, Karlsruhe, Germany Chongyu Su China Institute of Nuclear Industry Strategy, Beijing, China Dongyan Su School of Economics, Management and Law, University of South China, Hengyang, China Guanghui Su Xi’an Jiaotong University, Xi’an, Shaanxi, China Jiqiang Su China Institute of Nuclear Industry Strategy, Beijing, China

Contributors

xxxi

Ran Su China Institute of Nuclear Industry Strategy Beijing, Beijing, China Yantao Su CNNC Key Laboratory on Uranium Extraction from Seawater, Beijing Research Institute of Chemical Engineering and Metallurgy, Beijing, China Yaqiong Su School of Chemistry, Xi’an Key Laboratory of Sustainable Energy Materials Chemistry, State Key Laboratory of Electrical Insulation and Power Equipment, Xi’an Jiaotong University, Xi’an, Shaanxi, China Zilin Su Department of Engineering Physics, Tsinghua University, Beijing, China Aikou Sun School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China; Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China Chenglin Sun China Institute of Nuclear Industry Strategy, Beijing, China Jianjun Sun Research Institute of Nuclear Power Operation, Wuhan, China Qiming Sun CNNC Key Laboratory of Fabrication Technology of Fuel Assembly, Baotou, China; China North Nuclear Fuel Co., Ltd., Baotou, China; Shaanxi Engineering Research Center of Advanced Nuclear Energy, Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, School of Nuclear Science and Technology, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Qiuyu Sun School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China; Shaanxi Engineering Research Center of Advanced Nuclear Energy, Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, School of Nuclear Science and Technology, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Sheng Sun Nuclear Power Institute of China, Chengdu, Sichuan, China Yixiong Sun Nuclear Power Institute of China, Chengdu, Sichuan, China Chao Tan CNNC Key Laboratory on Nuclear Industry Simulation, Wuhan, Hubei, China Lan-yan Tan CNNC Jianzhong Nuclear Fuel Co., Ltd, Yibin, Sichuang, China Sichao Tan Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin, Heilongjiang, China Feifei Tang Suzhou Nuclear Power Research Institute of CGNPC, Suzhou, Jiangsu, China Jilin Tang Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin, Heilongjiang, China; Nuclear Power Institute of China, Chengdu, Sichuan, China

xxxii

Contributors

Junming Tang China Nuclear Power (Beijing) Simulation Technology Co., Ltd., Shenzhen, Guangdong, China Songqian Tang National Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, China Xin Tang State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Yun-Peng Tang China Institute of Atomic Energy, Beijing, China Yunpeng Tang Key Laboratory of Passive Nuclear Power Safety and Technology, North China Electric Power University, Beijing, China Ji Tengfei Shanghai Nuclear Engineering Research and Design Institute, Shangai, China Jie Tian Naval University of Engineering, Wuhan, Hubei, China; Unit of PLA, Sanya, Hainan, China Ruifeng Tian Harbin Engineering University Harbin, Heilongjiang, China Shizhong Tian China Nuclear Power Engineering Co., Ltd, Beijing, China Wenxi Tian Xi’an Jiaotong University, Xi’an, Shaanxi, China Zhengang Tian China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Lili Tong School of Mechanical Engineering, Shanghai Jiao Tong University, Minhang, Shanghai, China Pu Tong China Institute of Nuclear Industry Strategy, Beijing, China Xu Tu Key Laboratory of Passive Nuclear Power Safety and Technology, North China Electric Power University, Beijing, China P. Van Uffelen European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, Karlsruhe, Germany Cheng Wan Department of Nephrology, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Chenglong Wang School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi, China Chenhui Wang China Nuclear Power Engineering Co., LTD, Beijing, China Dihua Wang School of Resource and Environmental Science, Wuhan University, Wuhan, China Enpei Wang Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin, Heilongjiang, People’s Republic of China

Contributors

xxxiii

Fang Wang Economic Evaluation Division, China Nuclear Power Engineering Co., Ltd., Beijing Institute of Industrial Engineering, Beijing, China Fengju Wang CNNC Key Laboratory on Uranium Extraction from Seawater, Beijing Research Institute of Chemical Engineering and Metallurgy, Beijing, China Guoyin Wang Chongqing University of Posts and Telecommunications, Chongqing, China Haizhen Wang CNNC Key Laboratory on Uranium Extraction from Seawater, Beijing Research Institute of Chemical Engineering and Metallurgy, Beijing, China Hongbo Wang China North Nuclear Fuel Co., Ltd., Baotou, Inner Mongolia, China Jie Wang Shaanxi Engineering Research Center of Advanced Nuclear Energy, Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, School of Nuclear Science and Technology, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Junlong Wang Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Kai Wang Nuclear Power Institute of China, Chengdu, SiChuan, China Kan Wang Tsinghua University, Beijing, China Ming Wang School of Nuclear Technology and Automation Engineering, Chengdu University of Technology, Chengdu, Sichuan, China Ningyuan Wang China Institute of Nuclear Industry Strategy Beijing, Beijing, China Pei-jian Wang Airborne Survey and Remote Sensing Center of Nuclear Industry, Shijiazhuang, China Qi Wang School of Energy Science and Engineering, Harbin Institute of Technology, Harbin, China Sheng Wang School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Shengzhe Wang School of Nuclear Science and Engineering, North China Electric Power University, Beijing, China Shishuai Wang College of Nuclear Science and Technology, Harbin Engineering University, Harbin, China Shuguang Wang Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin, Heilongjiang, China Te Wang Nuclear Power Institute of China, Chengdu, China Xiang Wang College of Nuclear Science and Technology, Harbin Engineering University, Harbin, China Xiangwei Wang Unit of PLA, Sanya, Hainan, China

xxxiv

Contributors

Xiaodong Wang Research Institute of Nuclear Power Operation, Wuhan, China Xiaoxia Wang China Nuclear Power Engineering Co, Beijing, China Xin Wang China Nuclear Power Technology Research Institute Co., Ltd., Shenzhen, Guangdong, China Xin-Chen Wang School of Nuclear Technology and Automation Engineering, Chengdu University of Technology, Chengdu, Sichuan, China Xinan Wang Xi’an Jiaotong University, Xi’an, Shaanxi, China Yi-Feng Wang China Institute of Atomic Energy, Beijing, China Zejie Wang China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Zhendi Wang Department of Urology, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Dong Wei Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, Jiangsu, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, Jiangsu, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, Jiangsu, China Qian Wei CNNC Nuclear Power Operation Management Co., Ltd, Jiaxing, Zhejiang, China Zha Weihua CNNC Nuclear Power Operation Management Co., Ltd, Jiaxing, Zhejiang, China Yongjiang Wen Xi’an Jiaotong University, Xi’an, Shaanxi, China Li Wenan China Nuclear Power Engineering Co., Ltd, Beijing, People’s Republic of China Wu Wenqi Jiangsu Nuclear Power Corporation, Lianyungang, China Yu Wenxin CNNC Jianzhong Nuclear Fuel Co., Ltd, Yibin, Sichuan, China Fei Wu Naval University of Engineering, Wuhan, Hubei, China Ruolin Wu Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Wenqiang Wu Xi’an Jiaotong University, Xi’an, Shaanxi, China Xinxin Wu Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, China Yan-nong Wu NSC, Beijing, China

Contributors

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Zhiqiang Wu School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China; Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China Gao Xi-long Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Bangyang Xia Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Xiaotian Xia Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Yan Xia Beijing Institute of Spacecraft System Engineering, China Academy of Space Technology, Beijing, China Yuan Xia Jiangsu Nuclear Power Corporation, Lianyungang, China Zhang Xianggui Jiangsu Nuclear Power Corporation, Lianyungang, China Feng Xiao National Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, China Jun Xiao Nuclear and Radiation Safety Center, Ministry of Ecology and Environment (MEE) of the People’s Republic of China, Beijing, China Y. Xiao School of Science, Southwest University of Science and Technology, Mianyang, China Ren Xiaojiang Jiangsu Nuclear Power Corporation, Lianyungang, China Jinsen Xie School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China; Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China Yupeng Xie School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China Shang Xin China Nuclear Power Engineering Co., Ltd, Beijing, People’s Republic of China Zhang Xin Hualong International Nuclear Power Technology Co. Ltd., Shen Zhen, Guang Dong, China Ji Xing China Nuclear Power Engineering Co., LTD, Beijing, China Liu XingQing NingDe, FuJian, China Huang Xinming Daya Bay Nuclear Power Operations and Management Co., Ltd., Shenzhen, Guangdong, China

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Contributors

Pan Xinxin Shanghai Nuclear Engineering Research and Design Institute Co. Ltd, Shanghai, China Bilu Xiong China Nuclear Power Design Company (Shenzhen), Shenzhen, Guangdong, China Lingjie Xu Nuclear Power Institute of China, Chengdu, Sichuan, China Peng Xu Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Beijing, China; School of Nuclear Science and Engineering, North China Electric Power University, Beijing, China Xinyan Xu School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, China Yongdong Xu The Fourth Filial Company of 404 Company Limited, CNNC Lanzhou, Gansu, China Zhen Xu School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai, China Zhong Xu CNNP Nuclear Power Operation Management Co. Ltd., Haiyan, China; Department of Psychology and Behavioral Sciences, Zhejiang University, Hangzhou, China Han Xuan Institute of Physical and Chemical Engineering of Nuclear Industry, Tianjin, China Chunhui Xue Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, Jiangsu, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, Jiangsu, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, Jiangsu, China Feng Xue Suzhou Nuclear Power Research Institute of CGNPC, Suzhou, Jiangsu, China Changqi Yan Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin, China Kun Yan China Institute of Nuclear Industry Strategy, Beijing, China Tinxu Yan Key Laboratory of Passive Nuclear Power Safety and Technology, North China Electric Power University, Beijing, China Defeng Yang China Nuclear Power Engineering Co, Beijing, China Qiang Yang Jiangsu Nuclear Power Corporation, Lianyungang, Jiangsu, China Shi Yang China Nuclear Power Engineering Co., Ltd, Beijing, People’s Republic of China

Contributors

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Shouhai Yang Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China Tongan Yang Xi’an Jiaotong University, Xi’an, Shaanxi, China Weidong Yang Nuclear and Radiation Safety Center, Beijing, China Wenhua Yang Nuclear Power Institute of China, Chengdu, Sichuan, China Ye Yang China North Nuclear Fuel Co., Ltd., Baotou, China Wu Yanhui Shandong Nuclear Power Company, Yantai, Shandong, China Ming Yao Shanghai Nuclear Engineering Research and Design Institute Co. Ltd, Shanghai, China Yao Yao Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, Jiangsu, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Beijing, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, Jiangsu, China Zhuang Yaping Shandong Nuclear Power Company, Yantai, Shandong, China Yuanwu Ye China Institute of Atomic Energy, Beijing, China Jingwei Yi Key Laboratory of Nuclear Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, Sichuan, China Jin Yidan CNNC Nuclear Power Operation Management Co., Ltd, Jiaxing, Zhejiang, China Huayi Yin School of Resource and Environmental Science, Wuhan University, Wuhan, China Dongchuan Ying National Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, China Xu Ying Beijing Research Institute of Chemical Engineering Metallurgy, CNNC, Beijing, China Chong Yu Nuclear Power Institute of China, Chengdu, SiChuan, China Hong-Ying Yu Economic Evaluation Division, China Nuclear Power Engineering Co., Ltd., Beijing Institute of Industrial Engineering, Beijing, China Hongxing Yu Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Qian Yu China Nuclear Power Engineering Co., Ltd., Beijing, China Shuwen Yu School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, China

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Contributors

Sun Yu Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan, China Tao Yu School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China; Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China Yonglong Yu The 404 Company Limited, Lanzhou, Gansu, China Cui Yuan College of Nuclear Science and Technology, Harbin Engineering University, Harbin, Heilongjiang, China Huangfu Yuzhao Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan, China Feixiang Zeng Nuclear Power Institute of China, Chengdu, China Guanghao Zeng Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China Qiang Zeng Nuclear Power Institute of China, Chengdu, SiChuan, China Dalin Zhang Xi’an Jiaotong University, Xi’an, Shaanxi, China Dong Zhang School of Energy Science and Engineering, Harbin Institute of Technology, Harbin, China Haochun Zhang School of Energy Science and Engineering, Harbin Institute of Technology, Harbin, China Heng Zhang Chongqing University of Posts and Telecommunications, Chongqing, China Hui Zhang Nuclear Power Institute of China, Chengdu, Sichuan, China Jiahang Zhang CNNC Environmental Protection Co., Ltd., Beijing, China Jie Zhang Suzhou Nuclear Power Research Institute Co., Ltd, Shenzhen, Guangdong, China Lei Zhang School of Nuclear Technology and Automation Engineering, Chengdu University of Technology, Chengdu, Sichuan, China Li Zhang Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, China Li-ying Zhang CNNC Jianzhong Nuclear Fuel Co., Ltd, Yibin, Sichuang, China Lian Zhang Research Institute of Nuclear Power Operation, Wuhan, China Liang Zhang Nuclear Power Institute of China, Chengdu, Sichuan, China

Contributors

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Lili Zhang The Fourth Research and Design Engineering Corporation of CNNC, Shijiazhuang, Hebei, China Liying Zhang Xi’an Jiaotong University, Xi’an, China; China Nuclear Power Engineering Co, Beijing, China Meng Zhang Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu, China Qian Zhang Laboratory for Advanced Nuclear Energy Theory and Applications, Department of Physics, Zhejiang Institute of Modern Physics, Zhejiang University, Hangzhou, Zhejiang, China Ran Zhang China Nuclear Power Engineering Co., LTD, Beijing, China Ruomin Zhang Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai, China Shunxiao Zhang The 404 Company Limited, Lanzhou, Gansu, China Tian Zhang College of Nuclear Science and Technology, Harbin Engineering University, Harbin, China Wande Zhang The 404 Company Limited, Lanzhou, Gansu, China Wei Zhang Airborne Survey and Remote Sensing Center of Nuclear Industry, Shijiazhuang, China Wenlong Zhang Nuclear Power Institute of China, Chengdu, Sichuan, China Xiaofeng Zhang Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu, China Xiaoyu Zhang Key Laboratory of Nuclear Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, Sichuan, China Yading Zhang China Institute of Nuclear Industry Strategy Beijing, Beijing, China Yajing Zhang Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China Yaoxiang Zhang Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, China Yi-min Zhang NERO, Beijing, China Yitong Zhang Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, Jiangsu, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, Jiangsu, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, Jiangsu, China Zeqin Zhang School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi, China Zhenguo Zhang Harbin Engineering University Harbin, Heilongjiang, China

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Contributors

Y. E. Zhang-Han Fuqing Nuclear Power Co., Ltd, Fuqing, China Dong Zhao Economic Evaluation Division, China Nuclear Power Engineering Co., Ltd., Beijing Institute of Industrial Engineering, Beijing, China Feng Zhao Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu, China Hai Zhao Harbin Engineering University Harbin, Heilongjiang, China Haocheng Zhao China Institute of Atomic Energy, Beijing, China Lu Zhao Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai, China Qiang Zhao Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, College of Nuclear Science and Technology, Harbin Engineering University, Harbin, Heilongjiang, China Weiguang Zhao Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin, China Yun-Gan Zhao North China Electric Powder University, Beijing, China Ziwei Zhao China Nuclear Power Engineering Co., LTD, Beijing, China Wei Zhen China Nuclear Power (Beijing) Simulation Technology Co., Ltd., Shenzhen, Guangdong, China Qisen Zheng Airborne Survey and Remote Sensing Center of Nuclear Industry, Shijiazhuang, China Zaohua Zheng China Nuclear Power Design Company (Shenzhen), Shenzhen, Guangdong, China Jiang Zhenyu Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China Wu Zhiguang China Institute of Atomic Energy, Beijing, China Guobo Zhong Unit of PLA, Sanya, Hainan, China Lingpeng Zhong Chongqing Chongqing, China

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Yiqiang Zhong China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Jie Zhou Nuclear and Radiation Safety Center, Beijing, China Lanyu Zhou China Nuclear Power Engineering Co., Ltd., Beijing, China Lei Zhou Xi’an Jiaotong University, Xi’an, Shaanxi, China Shuyong Zhou China Nuclear Power (Beijing) Simulation Technology Co., Ltd., Shenzhen, Guangdong, China

Contributors

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Tao Zhou Department of Nuclear Science and Technology, School of Energy and Environment, Southeast University, Nanjing, Jiangsu, China; Institute of Nuclear Thermal-Hydraulic Safety and Standardization, Nanjing, Jiangsu, China; National Engineering Research Center of Power Generation Control and Safety, Nanjing, Jiangsu, China Xingguang Zhou Xi’an Jiaotong University, Xi’an, Shaanxi, China Yonghai Zhou Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China Zedong Zhou School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, China; Virtual Simulation Experiment Teaching Center on Nuclear Energy and Technology, University of South China, Hengyang, Hunan, China Chenglin Zhu National Energy Key Laboratory of Nuclear Power Software, State Power Investment Corporation Research Institute, Beijing, China Hui-Ping Zhu China Institute of Atomic Energy, Beijing, China Huiping Zhu Key Laboratory of Passive Nuclear Power Safety and Technology, North China Electric Power University, Beijing, China Lina Zhu China Institute of Atomic Energy, Beijing, China Rongxu Zhu Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu, China Tongyu Zhu The 404 Company Limited, Lanzhou, Gansu, China Wei Zhu NSC, Beijing, China Yingxi Zhu The Fourth Filial Company of 404 Company Limited, CNNC Lanzhou, Gansu, China; China National Nuclear Industry Corporation 404, Jiayuguan City, Gansu Province, China Hang Zou Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, College of Nuclear Science and Technology, Harbin Engineering University, Harbin, Heilongjiang, China Tingting Zou China Nuclear Power Technology Research Institute Co., Ltd., Shenzhen, Guangdong, China Xiang Zou Nuclear and Radiation Safety Center, MEE, Beijing, China Zhenyu Zou Shanghai Jiao Tong University, Minhang, Shanghai, China

On the Issues for Legislation of Spent Fuel in China Jiu Liu(B) Harbin Engineering University, Harbin, Heilongjiang, China [email protected]

Abstract. Background China has already been the country with the fastest development of civil nuclear power industry in the world. With the rapid development of civil nuclear power industry, more and more spent fuel has been produced and accumulated year by year. According to the closed fuel cycle scheme determined by China at the beginning of the development of civil nuclear power industry in the 1980s, the spent fuel reprocessing policy and laws are being implemented to maximize the utilization of resources and minimize the environmental pollution. However, there are still some problems in the current legal system pertaining to spent fuel, and the current legal system is difficult to meet the practical needs of spent fuel management in China. Methods To highlight the problems in the current legal system pertaining to spent fuel, a legislation study is used to analyze the current content of related legislation and regulations. Comparative methodology is also adopted in this paper to analyze the legislative and administrative experience of other countries with advanced nuclear power industry and summarize the problems in the current legal system pertaining to spent fuel management in China. Results The Regulation on the administration of spent fuel has not yet been promulgated; the ownership of the spent fuel has not been defined; the management and application of spent fuel fund needs to be innovated; and the public is not as significant as they should be in the whole process of spent fuel management. Conclusions Some suggestions are proposed in this paper: for example, the related legislation and regulations should be improved, especially the Regulations on the administration of spent fuel had better be established soon, which should define the ownership of the spent fuel clearly. And the public should be attached importance to during the whole process of spent fuel management. Keywords: Spent fuel · Civil nuclear power industry · Reprocessing policy · Ownership · Spent fuel fund

© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 1–10, 2023. https://doi.org/10.1007/978-981-19-8899-8_1

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1 Introduction to the Current Administration and Legislation of Spent Fuel in China According to Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radiation Waste Management issued by International Atomic Energy Agency (IAEA), “spent fuel” means nuclear fuel that has been irradiated in and permanently removed from a reactor core [1]. Thus, the spent fuel is not nuclear waste. It is a nuclear fuel that can be reused after scientific and reasonable reprocessing, which contains many unused proliferating materials including Uranium 238 or Thorium 232, and unburned and newly generated fissile materials such as Plutonium 239, Uranium235 or Uranium233, and some transuranium elements as Neptunium, Americium and Curium produced by nuclear fuel during irradiation, as well as fission elements Strontium90, Cesium137 and Technetium99 [2]. The nuclear power plant must shut down and replace fuel every 12 or 18 months, due to the strong radioactivity and high heat release of the newly unloaded spent fuel, the safety requirements and costs of transportation and off reactor storage are very high [3]. According to international practice, if there is no hurry for post-treatment, it is generally necessary to cool the reactor pool for more than 5 years and then transport it to the off reactor storage facility or post-treatment plant for storage or treatment. With the further development of China’s civil nuclear industry, the generation and accumulation of spent fuel in China are becoming more, which brings new requirements and challenges to China’s spent fuel closed cycle strategy. The so-called “closed cycle” refers to the reprocessing of spent fuel by proton bombardment to transmute elements with a half-life of hundreds of thousands of years into elements with a half-life of hundreds of years. This concept is also related to “open cycle”, which means the spent fuel is no longer used. According to the disposal method of high-level radioactive waste, the spent fuel shall be stored for a long time after cooling and special packaging, or buried 500–1000 m deep to shield the difference between radioactivity and decay heat. At the beginning of the development of civil nuclear industry in the 1980s, China established the closed fuel cycle scheme and implemented the spent fuel reprocessing policy, which has been applied since then. Although it is undeniable that the spent fuel has strong radioactivity, long half-life and high chemical toxicity, if the spent fuel is deeply disposed according to the “one-time pass” method, the half-life of some elements can reach hundreds of thousands of years. But, the closed fuel cycle scheme can recover useful uranium, plutonium and other materials, it can not only improve the utilization rate of resources, but also reduce the radioactive hazard and chemical toxicity of waste. Now many nuclear power plants in China will be fully integrated in the near future, and are actively and steadily promoting the outward transportation. Therefore, China is promoting the development of spent fuel transportation, storage and reprocessing. At present, China’s legal system on spent fuel management can be divided into three levels: laws related to spent fuel management formulated by the National People’s Congress, administrative regulations about nuclear and radiation protection management by the State Council, and departmental rules pertaining to spent fuel management. In addition, it also includes Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radiation Waste Management, which has been acceded to and transformed into China’s domestic legislation in order to strengthen the safety management

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of spent fuel and radioactive waste, broaden foreign cooperation and promote the healthy development of China’s nuclear industry [4]. Currently the increasing accumulation of spent fuel not only puts forward new requirements for spent fuel management, transportation, intermediate storage and reprocessing, but also for relevant legal systems so as to better adapt to new problems.

2 The Deficiencies of Current Legislation for the Administration of Spent Fuel 2.1 The Regulation on the Administration of Spent Fuel Has not Yet Been Established Although there are many laws and regulations pertaining to the whole process management of spent fuel in China’s current legal system, there is no legislation or regulation specialized for the whole process of spent fuel management. Therefore, for the whole management process of spent fuel, the current legislation is not sufficient to support the standardized management of spent fuel, and can not comprehensively solve the comprehensive issues. At present, there are merely several legislation and regulations specialized in some aspects of spent fuel management as follows: (1) Measures for the Project Management of Spent Fuel Treatment and Disposal Fund of Nuclear Power Plants aims to standardize the project management of spent fuel treatment and disposal fund of nuclear power plants and the rational and effective use of such fund. (2) Interim Provisions on the Administration of Road Transportation of Spent Fuel in Nuclear Reactors aims to strengthen the administration of road transportation of spent fuel in nuclear reactors and ensure the safety of spent fuel transportation. This regulation focus on the transportation subject and management subject involved in road transportation of spent fuel. (3) Safety Requirements for Spent Fuel Reprocessing Facilities (for Trial Implementation) not only makes restrictive stipulation on the spent fuel reprocessing facilities, but also on the site selection, design, construction and commissioning of spent fuel reprocessing facilities. (4) Nuclear Safety Supervision Requirements for Spent Fuel Dry Storage System in Nuclear Power Plants (for Trial Implementation) also makes special requirements for spent fuel dry storage system in nuclear power plants [5]. (5) Interim Measures for the Administration of Collection, Use and Management of Spent Fuel Treatment and Disposal Fund of Nuclear Power Plants aims to standardize the collection, use and management of spent fuel treatment and disposal fund of nuclear power plants. Although the above regulations can provide for some processes of spent fuel management, the provisions are not detailed and there are no corresponding special provisions on the generation of spent fuel, environmental protection and etc. Thus, they alone or together can not cover all aspects of spent fuel management in the whole process. As for Nuclear Safety Law, Regulations on Safety Management of Radioactive Goods Transportation and etc., although they involve provisions on management of spent fuel, they are only macro provisions, and can not provide comprehensive and effective legal provisions for the whole management process. Therefore, there is a lack of legislation and regulations covering the whole process of spent fuel management in China.

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2.2 The Ownership of Spent Fuel and Its Reprocessing Products Are Still Undefined Substantially speaking, spent fuel is property in civil law. It truly has the characteristics of property in law, but spent fuel also has its own particularity lying in its radioactivity and reusable value. Thus, the ownership of spent fuel in China has not been defined now. Legally speaking, the ownership of spent fuel and its reprocessing products includes possession, usage, getting profits and disposal. Possession is the actual control of property. The possession of spent fuel shall be protected by law and shall not be illegally infringed upon. The purpose of the usage of spent fuel is giving full play to the value of the property. The users can not only further extract useful resources from spent fuel by themselves, but also entrust others to extract useful resources from spent fuel. Getting profits is the economic benefit obtained through possession and usage of spent fuel. Therefore, the owner of spent fuel can obtain profits from the resources extracted from spent fuel. Disposition refers to the final disposition of the property, spent fuel, in both fact and law. It is sure that the spent fuel can be disposed of as waste, if it can no longer reused. However, according to China’s existing legislation and regulations and the national nuclear fuel closed cycle policy, the ownership of spent fuel and its reprocessing products has not been defined yet, although Nuclear Safety Law and other legislation and regulations clarify the safe and secure responsibility of spent fuel. The Atomic Energy Law (Draft) issued in 2018 also avoided defining the ownership of spent fuel, therefore, until now, China has not clearly define the ownership of spent fuel and its reprocessing products in any law. The ownership of spent fuel is not an isolated issue, which is related to the source of financing for the construction and operation of spent fuel reprocessing plants, the nature of reprocessing funds (government funds or enterprise reserves) and involves various complex issues such as the interests of various groups, technology and law, which need to be considered together and made an overall decision. From the legislative level, defining the ownership of spent fuel and its reprocessing products is a prerequisite for strengthening China’s spent fuel safety management. 2.3 It Lacks a Value Preservation and Appreciation Mechanism for Spent Fuel Fund As the investment of nuclear power enterprises continually, the spent fuel fund has already accumulated more than 10 billion CNY now. If the spent fuel fund is used as the static fund in the account, it is actually a waste, which may lead to the failure to meet the demand in future withdrawal, and affect the specific implementation of spent fuel safety management. For the spent fuel fund of nuclear power plants, it is necessary to ensure that there are sufficient fund to carry out spent fuel treatment and disposal efficiently and reasonably in the future, thus, it is important to find appropriate investment channels for the fund. Considering that with the continuous and rapid development of China’s spent fuel reprocessing industry, the reprocess and disposal of spent fuel may face huge expenses in the future, if the spent fuel fund lacks such a value preservation and appreciation mechanism, it will undoubtedly bring great pressure to the spent fuel

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reprocessing industry. In addition, it is very easy to cause high financial risks under the condition of large accumulation of spent fuel fund, which is not conducive to the effective utilization of the spent fuel fund. 2.4 The Role of Public Influenced Is not of Enough Significance in the Whole Process of Spent Fuel Management The public are not given full right-to-know and right-to-participate. In the absence of real and accurate information disclosed to the public in the candidate places in time, public would be misled by all kinds of medias, resulting in panic. Although there are relevant provisions about information disclosure and participation in current laws and regulations, the realization of right-to-know and public participation has not been effectively implemented in practice. In the whole management process of spent fuel, most of the relevant information related to spent fuel matters is released to the public through the official website and news newspapers of the operators of nuclear facilities. However, most people are not particularly familiar with these information platforms. Therefore, although relevant departments and factories have fulfilled the obligation of information disclosure, they have not really given the people involved the practical right-to-know and right- to-participate. In addition, the publicity time of such information on local government websites is always relatively short, so the public are difficult to fully and effectively understand the relevant information related to their own interests in such a short time, and the public whose right-to-know has not been fully realized can not realize their right to participate in a few days, which also hinders the public’s in-depth understanding of civil nuclear projects and facilities and information related to spent fuel. The science popularization of spent fuel is also insufficient. Even if the information is open and transparent, the lack of spent fuel related knowledge will influence the public’s right-to-know, which makes the public’s right-to-participate very limited. In addition, due to the characteristics of the network age, false information can be rapidly spread through various self-Media, causing misunderstanding and panic among the people without knowledge about nuclear energy and spent fuel. In practice, the public’s fear and worry about spent fuel facilities largely stems from the lack of relevant popular science education and knowledge. Therefore, Chap. 5 of Nuclear safety Law stipulates the disclosure, publicity and education of nuclear safety information. However, it usually only provides relevant information to the public, but do not pay any attention to the systematic popular science publicity and education of spent fuel related contents for the public. Therefore, the spent fuel public science publicity is not sufficient, and there is a lack of shaping the public’s nuclear safety concept and awareness.

3 Suggestions to Improve the Legal System of Spent Fuel in China 3.1 Draft and Enact the Regulation on the Administration of Spent Fuel Whether a country treats spent fuel as radioactive waste determines its need for legislation and regulations of the administration of spent fuel. For countries that treat spent fuel as waste, it is not necessary to specially establish legislation and regulation pertaining to

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spent fuel, because the laws for radioactive waste can be applied to spent fuel. However, since China has determined the closed fuel cycle scheme and implemented the spent fuel reprocessing policy, it is certainly not enough for China to merely regulate the whole process of spent fuel management by the radioactive waste disposal legislation and regulations. Thus, the Regulation on the Administration of Spent Fuel should be drafted and enacted in time. China’s future Regulation on the Administration of Spent Fuel should take “standardizing and strengthening the whole management process of spent fuel, effectively utilizing resources, protecting environment and promoting the safe, effective and sustainable development of nuclear energy” as its legislative purpose. The scope of the Regulation in the future should involve the storage of spent fuel in and off reactor; spent fuel transportation; the whole process of spent fuel reprocessing and uranium plutonium recycling and the treatment and disposal of spent fuel. As for structure, the future Regulation shall be divided into nine parts: general provisions, management requirements, spent fuel storage, spent fuel transportation, spent fuel reprocessing and recycling, special spent fuel management, international cooperation, legal liability and supplementary provisions. The general provisions should include legislative purpose, scope of application and responsibilities of government departments. Management requirements shall first determine the ownership of spent fuel, and specify the accessibility of operators engaged in spent fuel management, nuclear safety, security and emergency requirements, spent fuel fund, etc. International cooperation part shall stipulate the principles and regulations that China should follow in carrying out international cooperation related to spent fuel. The Regulation shall cover the whole management process of spent fuel “from life to death”, including production, storage, transportation, post-treatment and other links. The establishment of spent fuel management regulation will help to improve China’s nuclear legal system, further coordinate the relationship between government administration and the interests of operators and realize the management of spent fuel industry by law. This is not only the expectation of the spent fuel reprocessing industry, but also a necessary measure for the construction of China’s nuclear legal system. The detailed structure and content of the future Regulation are outlook in the chart below (Table 1). 3.2 Define the Ownership of Spent Fuel and Its Reprocessing Products Explicitly Defining the ownership of spent fuel and its reprocessing products in national legislation is a prerequisite for strengthening China’s spent fuel safety management. Clarifying the ownership of spent fuel is conducive to the effective management of spent fuel and promoting the reuse of spent fuel as a resource. According to the provisions of China’s Civil Code in property right, the ownership of spent fuel produced by nuclear power plants should belong to the owners of nuclear power plants when there is no transfer of property rights. The owner of the nuclear power plant has the right to dispose of the spent fuel according to related legislation and regulations. Once the transfer of spent fuel happen, its ownership should be defined based on whether the spent fuel would be reused: if the nuclear power plant no longer uses the spent fuel and its reprocessing products, the spent fuel would be disposed of as waste and its ownership should not be registered; if the spent fuel would be used for national

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Table 1. The outlook of future regulation on the administration of spent fuel Regulation on the administration of spent fuel (outlook) Chapter I General Provisions (Legislative purpose) (Scope) (Spent fuel management definition) (Management policy) (Responsibilities of responsible departments) (Technology) (Personnel) (Standardization) Chapter II Management requirements (Ownership of spent fuel) (Planning and Implementation) (Project application and approval) (Admittance) (Nuclear material license) (Nuclear safety permit) (Nuclear safety requirements and responsibilities) (Nuclear security requirements) (Nuclear emergency requirements) (Spent fuel fund) (Information report system and management) (Information disclosure) Chapter III Spent fuel storage (Subject) (Offsite storage facility siting) (Administratio) (Cost) Chapter IV Spent fuel transportation (Transportation management) (Transportation responsibility) (Transportation equipment and facilities) (Transport plan) Chapter V Spent fuel reprocessing and recycling (Post processing responsibility subject) (Spent fuel reprocessing) (Waste management) Chapter VI Special spent fuel management (Spent fuel management of commercial heavy water reactor and high temperature reactor) (Research reactor spent fuel management) (Cost management) Chapter VII International co-operation (peaceful uses of nuclear energy) (Compliance with relevant conventions) (International reprocessing services) Chapter VIII legal responsibility Chapter IX Supplementary articles (Interpretation of terms) (Entry-into-force time)

defense, it should belong to the state; and if the nuclear power plant would reapply the spent fuel and its reprocessing products, the ownership should not be transferred, during the whole period, spent fuel reprocessing can be regarded as the producing activities entrusted by the nuclear power plant. However, this is only a conclusion made from China’s Civil Code nowadays. Till now there is still no legislation and regulations in China about the definition for ownership of spent fuel. It is agreed that under China’s current spent fuel fund system, it is the nuclear power plants who have the responsibility to pay for the spent fuel treatment and disposal fund. Thus, it is unfair for the nuclear power plants if the spent fuel and its reprocessing products belong to State by law. So, the ownership of spent fuel and its reprocessing products to the spent fuel producer, that is, the nuclear power plants, is more conducive to rationalize the market-oriented operation of spent fuel reprocessing and more appropriate for market-oriented economy under Civil Code. In this way, it is the nuclear power plants

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which own the spent fuel and its reprocessing products and it is the state government that is responsible for formulating the spent fuel management mechanism and establishing the plan for spent fuel reprocess, so as to facilitate enterprises to manage spent fuel in a safe and secure way. And the responsibility of storage and transportation of spent fuel and the management of recovered uranium and plutonium products belong to the nuclear power plants, which can reduce the burden on the government. 3.3 Establish a Value Preservation and Appreciation Mechanism for Spent Fuel Fund In terms of the spent fuel, on the one hand, it must be ensured that there are sufficient funds for spent fuel treatment and disposal in the future so that risks from the management system can be prevented. On the other hand, appropriate investment channels for the spent fuel fund should be found or established. Therefore, it is necessary to understand the operation mode of similar types of fund in China and how other countries maintain and increase the value of spent fuel fund. However, the investments for several main types of government fund in China, such as railway construction fund, civil aviation development fund and waste electrical and electronic products treatment fund, all have been taken as a subsidy. In the other words, almost all these fund are not used for investment, so to some extent, they are merely static fund which cannot produce any profits themselves. And it would be difficult to keep the value of such types of fund in this way. Thus, it should be necessity to establish a value preservation and appreciation mechanism for spent fuel fund. As a country with developed spent fuel reprocessing industry, Sweden gives great example for ways of maintaining and increasing the value of spent fuel funds. In Sweden, the spent fuel reprocessing fund had been initially deposited in the Swedish Central Bank. Later, the nature of the fund was transferred from the internal reserve of the enterprise to the national fund, and the financial guarantee was provided by National Debt Institution. And the fund committee invests a large amount of fund through National Debt Office to ensure a fixed return at the market interest rate for a long time. At the same time, the committee also use the mechanism of interest marketization to increase the rate of return through reinvestment in different interest periods. Therefore, China should establish a special department and mechanism by government for fund management and financial planning in order to make full use of the market and improve the utilization efficiency of spent fuel treatment and disposal fund. Establishing such department and mechanism can also avoid the problem of non-disclosure in the fund disposal. 3.4 Give Full Attention to the Influenced Public in the Whole Management Process of Spent Fuel According to Principle 10 of Rio Declaration of 1992, environmental issues are best handled with the participation of all concerned citizens [6]. And in order to ensure the public are able to know and understand what is happening in the environment around them and ensure the public is able to participate in an informed manner, Convention on Access to Information, Public Participation in Decision-making and Access to Justice

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in Environmental Matters (Aarhus Convention) of 1998 requires a system that enables the public to request and receive environmental information from public authorities and requires a system through which public authorities collect environmental information and actively disseminate it to the public without request [7]. Moreover, the International Atomic Energy Agency recommends that “Information disclosure and public involvement principle requires that bodies involved in the development, use and regulation of nuclear energy make available all relevant information concerning how nuclear energy is being used, particularly concerning incidents and abnormal occurrences that could have an impact on public health, safety and the environment” [8]. In view of the particularity of spent fuel itself and the public’s sensitivity to environmental protection issues, the future Regulation on spent fuel management should pay attention to the information disclosure, social supervision and public participation and promote the establishment of an effective spent fuel management information feedback mechanism so as to further enhance the public’s understanding and trust in spent fuel management and promote the healthy development of spent fuel reprocessing industry. The earlier the public knows and participates, the more it can break the pattern of information asymmetry and improve the public’s acceptance of spent fuel reprocessing facilities. In this regard, it is necessary to implement the whole process information disclosure system. The information disclosure of governments and operators should follow the principle of “mandatory disclosure first, supplemented by voluntary disclosure”, timely publish the related information on government websites, operators’ websites and local newspapers, and centrally publish various announcements such as site selection environmental assessment, stability assessment. Meanwhile, it is necessary to strengthen the popular science publicity for spent fuel. In the professional field, nuclear power is a safe and clean energy, and spent fuel is not nuclear waste. However, most people are not professional and lack a clear understanding of this. Good science publicity is the most powerful weapon to overcome such nuclear fear. Strengthening such science publicity and education is the most important link to popularize the nuclear safety culture and cultivate the correct nuclear safety concept of the public. China’s 12th Five Year Plan for nuclear safety and radioactive pollution prevention and 2020 Vision also points out that “it is necessary to increase the popular science knowledge of nuclear and radiation safety in basic education, establish a long-term nuclear safety education and publicity mechanism, fulfill the public’s demand for nuclear safety related information, and enhance the public’s understanding and confidence in the safety of the use of nuclear energy and nuclear technology”. Establish a long-term publicity mechanism, the sooner the better. We should always patiently answer the influenced public’s questions and doubts. In terms of content, it shall be objective and comprehensive, including the basic knowledge of spent fuel, industry attributes, environmental protection, safety and etc. It is suggested to pay attention to the role of the media, make full use of all kinds of media to carry out science publicity and education for the public, enhance the neutrality of media publicity and education, and improve the sense of authority and trust of the media, so as to eliminate the doubts of the public through the media.

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To sum up, while providing the public with information related to spent fuel, allowing and advocating their participation in relevant decisions, we should pay more attention to the full and acceptable science publicity and education for the public, so that the public can truly accept relevant projects, better understand relevant projects and fully realize their right to participate when realizing their right-to-know.

4 Conclusions Nowadays the generation and accumulation of spent fuel in China are increasing. Spent fuel has strong radioactivity, long half-life and high chemical toxicity. If the spent fuel is directly disposed in deep geology, its radioactivity will take more than 100 thousand years to decay to the level of natural uranium ore. In order to reduce waste emission, protect the environment to the greatest extent and achieve sustainable development, China has determined the closed cycle scheme for nuclear fuel. At present, China does not have specialized laws and regulations for the whole management process of spent fuel, but there are a large number of existing laws, administrative regulations and rules of relevant departments that can be applied to the whole process of spent fuel management. By combing and interpreting the laws, regulations and departmental rules related to spent fuel management, this paper summarizes the legal problems existing in China’s spent fuel management system, such as the absence of the Regulation on the Administration of Spent Fuel has not been issued, and the role of the influenced public in the whole management process of spent fuel has not been given enough attention. By further studying the existing legal system of spent fuel management in China, some ways forward are mentioned, such as drafting and enacting the Regulation on the Administration of Spent Fuel, and giving full attention to the influenced public in the whole management process of spent fuel. Acknowledgment. This research is funded by Fundamental Research Funds for the Central Universities under grant number 3072022WK1312.

References 1. International Atomic Energy Agency (IAEA): Storage of Spent Nuclear Fuel, No. SSG-15 of Safety Standard, p. 2 2. Lu, W.: Status and analysis of spent fuel after treatment in the world. Chem. Ind. Jiangxi 10–12 (2018) 3. Hu, Y., Li, H., Deng, C., Xu, T., Mo, Z.: Study on rapid estimation method of water temperature in spent fuel storage tank. Nucl. Power Eng. 35(03), 102–104 (2014) 4. http://www.npc.gov.cn/wxzl/gongbao/2006-05/24/content_5350144.htm 5. Yuan, C., Liu, Y., Mo, H.: Research on the application of dry storage technology of PWR spent fuel. Nucl. Sci. Eng. 37(03), 470–476 (2017) 6. Principle 10 of Rio Declaration 7. Article 4 and 5 of Aarhus Convention 8. Stoiber, C., Baer, A., Pelzer, N., Tonhauser, W.: Handbook on Nuclear Law, p. 10. IAEA (2003)

Identification and Investigation on Thermal Hydraulic Phenomena Related to Core Make-Up Tank Jilin Tang1,2(B) , Shuguang Wang1 , Yusheng Liu1,3 , Sichao Tan1 , and Dongyang Li1 1 Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin

Engineering University, Harbin, Heilongjiang, China [email protected] 2 Nuclear Power Institute of China, Chengdu, Sichuan, China 3 Nuclear and Radiation Safety Center, MEE, Beijing, China

Abstract. Aiming at the core makeup tank system (CMT) used in advanced passive PWR technology, this paper analyzes and summarizes the design functions and operation characteristics of CMT. Based on the investigation and analysis of thermal hydraulic researches related to CMT at home and abroad, and combined with integral effect test or separate effect test, the thermal hydraulic phenomena of CMT and its branches are studied. At the end, the thermal hydraulic phenomena related to CMT are summarized, and some suggestions on its experimental research are put forward. During CMT test, different scaling criteria should be selected for different phenomena concerned, the interaction between natural circulation characteristic time and phenomenon characteristic time of concern should also be considered. Keywords: Passive safety system · Core make-up tank · Phenomena identification

1 Introduction The AP1000, a kind of passive safety pressurized water reactors, is an innovative nuclear power technology designed and developed by Westinghouse. The safety system of AP1000 adopts the design concept of passive safety, it drives the operation of the safety system based on gravity, heat transfer, inertia and other passive methods, which not only simplifies the system design, but also improves the reliability of the safety system and the economy of the nuclear power plant. The Core Make-up Tank (CMT) is one of the most distinctive passive components, which is mainly used for core cooling and coolant replenishment during high-pressure operation [1].

© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 11–22, 2023. https://doi.org/10.1007/978-981-19-8899-8_2

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2 CMT Design and Function Overview The AP1000 is equipped with two Core Make-up Tanks, which are filled with lowtemperature concentrated boron water, exposed in the containment, without heating or thermal insulation devices, and the water temperature is consistent with the ambient temperature of the containment. The safety injection flow diagram is shown in Fig. 1 and the structure is shown in Fig. 2. CMT is designed as a steel forged welding container with hemispherical upper and lower heads, with stainless steel surfacing on the inner wall. The upper and lower heads are provided with one inlet and outlet connecting pipe with safety end, one man hole is provided in the cylinder, 17 measuring instrument connecting pipes are provided in the cylinder and head, and 8 support columns are welded in the lower head to support the shell. In addition, in order to avoid rapid condensation inside the CMT, a steam sparger with circumferential opening above CMT [2].

Fig. 1. Schematic of the AP1000 passive safety injection system schematic

The operation of CMT is very important for the Passive Core Cooling System (PXS) to complete its emergency core cooling function, because the injection of CMT directly replaces the high-pressure injection pump used in the traditional PWR, which is the only coolant source in the high-pressure safety injection stage. The working principle of CMT is mainly to use the height difference and density difference between CMT and the core to form a driving head under the action of gravity to drive the boron containing cold water in CMT to inject into the Reactor Pressure Vessel (RPV) to realize core water make-up and cooling. According to the accident type and severity, there are two main modes of CMT operation, natural circulation mode and steam substitution mode. The natural circulation

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Fig. 2. Structural diagram of the CMT

mode refers to that the hot water in the core flows into the CMT through the Pressure Balance Line (PBL), and the boron containing cold water stored in the CMT is injected into RPV through the Direct Vessel Injection Line (DVI). The CMT is gradually filled with hot water and the cold water is replaced, the driving force of natural circulation is gradually weakened, and the whole cycle process takes water as the working medium without involving steam. Steam substitution mode refers to that the steam in the reactor core flows into CMT through PBL. During this process, steam may condense on the CMT wall and liquid surface, and a small amount of condensed water compensates for the water level. It involves steam water phase change, CMT drainage, continuous drop of liquid level and other phenomena. In Loss of Coolant Accident (LOCA), the two operation modes of CMT and their operation time are mainly determined by the size of the break. When the size of the break is small, the natural circulation mode can be maintained for a long time. When the break size is large, the natural circulation process will be very short. With the drainage of CMT, the steam in the system will continue to flow into CMT and enter the steam alternative operation mode.

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3 Phenomenon Identification Based on CMT Overall Effect Test The working process of CMT is closely related to the state of reactor coolant system. At the same time, there are complex processes such as steam condensation flow and thermal stratification. Therefore, scholars at home and abroad have carried out a lot of research on the thermal hydraulic phenomena related to CMT. Based on the existing research results, this paper identifies the key thermal hydraulic phenomena, analyzes the mechanism of the phenomenon, and combs the thermal hydraulic phenomena according to the different levels of phenomena. The two operation modes of CMT are natural circulation processes with the density difference between cold CMT and hot PBL as the driving force. The natural circulation mode is a single-phase natural circulation process, and the driving force is the density difference between the cold water in CMT and the hot water in Reactor Coolant System (RCS). The steam substitution mode is a two-phase natural circulation process, and the driving force is the density difference between the cold water in CMT and the saturated steam in RCS. Under the two circulation modes, the circulation flow is determined by the matching result of the driving force generated by the density difference between the cold and hot ends and the loop resistance. For different operation modes of CMT, domestic and foreign scientific research institutions have carried out a large number of experimental research based on the Integral Effect Test Facility (IETE). The European Union has carried out a number of CMT experimental studies using PACTEL [3]. PACTEL is an IETF designed with Loviisa VVER-440 as the reference prototype. A series of Small Break Loss of Coolant Accident (SBLOCA) in the heat pipe section was carried out on the PACTEL facilities, focusing on simulating the gravity driven core cooling process in the CMT natural circulation mode. During the test, it was observed that the rapid condensation of steam at the top of CMT would interrupt the flow of emergency core cooling for many times. In the project, the EU further studied the effects of factors such as the size and location of the breach, the size and location of CMT, the removal of steam distributor, the initial water temperature of CMT and PBL, the connection location of PBL and the flow resistance of safety injection line on the thermal hydraulic behavior of passive safety injection system during LOCA. In addition, the phenomenon of heat transfer in the fluid wall and its delamination were also studied. The test shows that the steam distributor at the top of CMT plays an important role in limiting rapid condensation. Japan has carried out the experimental research on the CMT of Westinghouse passive nuclear power plant by using ROAS-AP600 test facility [4]. ROAS-AP600 is an IETF with 1 /48 volume ratio, full height and full pressure designed and modified for AP600 based on the Large-scale Test Facility (LSTF) simulated by Japan Atomic Energy Research Institute (JAERI). ROAS-AP600-CMT test confirmed that during the natural circulation flow between the core and CMT, significant thermal stratification occurred in the CMT, because the hot water from the cold pipe section gathered in the upper area of the CMT. However, axial heat conduction and diffusion in the CMT was weak, the movement of thermal stratification was basically one-dimensional downward. The test also shows that the effect of the size and location of the break on the natural circulation rate of CMT is not significant except for the PBL break. During LOCA, after the Automatic Depressurization System (ADS) is put into operation, the hot water layer

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above CMT flashes and the drainage rate of CMT increases to a certain extent under the pressure relief process of RCS. Aiming at the independently developed AC600 design, Nuclear Power Institute of China carried out the full pressure CMT experiment [5]. In the experiment, 36 thermocouples were set in the CMT simulator to measure the temperature of steel wall and fluid, the water level was measured simultaneously by electric contact water level gauge and differential pressure water level gauge, the discharge flow was measured by turbine flowmeter, and the pressure in CMT was also measured. The test results show that at the initial stage of discharge after CMT is put into operation, due to the strong condensation effect of CMT cold wall and cold water surface, the steam entering the CMT simulation body from the pressurizer will lead to pressure pulsation in the CMT simulation body, resulting in the characteristics of short-term low-level pulsation platform of mass flow rate in the discharge pipe. When the break size is small, the CMT simulator is in the process of gravity discharge after entering the stable discharge period. Under the steam substitution mode, CMT injects water into RPV, and steam flows into CMT from pressurizer, resulting in a large temperature difference between the inner and outer walls of CMT. South Korea has carried out experimental research on CMT of modular small reactor using SMART-ITL test facility [6]. SMART reactor is an integrated reactor designed and developed in Korea. The main components of its main coolant system are contained in pressure vessels, such as pressurizer, reactor core, steam generator and reactor coolant pump. The CMT test carried out by SMART-ITL shows that there are three stages in the safety injection process of smart CMT, namely circulation stage, oscillation stage and stable injection stage. The PBL temperature and its interface position, CMT water level and other factors have an important impact on each stage. The two parameters of PBL temperature and CMT water level determine the operation stage of CMT and its branches, that is, the recirculation stage mainly depends on the CMT water level, and the oscillation stage is mainly dominated by PBL evaporation. When the fluid temperature in the upper area of PBL and CMT is equal, it enters the stable injection stage. Lee [7] tested and analyzed the ability of direct contact condensation in CMT of Passive High Pressure Injection System (PHPIS) in CARR passive reactor (CP1300). The steam generator is used to provide steam in the test, and the steam is directly injected into CMT cold water. The discharge flow and water level of CMT were monitored, and the axial water temperature of CMT was measured by thermal resistance. The results show that the greater the undercooling of water, the later the start-up time of CMT water injection, and both steam distributor and natural circulation of hot water can accelerate the start-up of gravity driven water injection. Lee divided direct contact condensation into three modes, sonic injection, subsonic injection and steam cavity, and proposed a condensation model suitable for CMT [8]. During the research and development of Guohe-1, China used ACME test facility to study the transient response and thermal hydraulic behavior of CMT under SBLOCA [9]. The research results show that the thermal stratification mode of fluid in CMT is similar in different experimental processes, and the hot fluid layer can separate hot steam and cold fluid. ADS depressurization will lead to CMT flash, which will increase the CMT-RCS differential pressure, and significantly reduce the thermal stratification area.

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However, after the system pressure is stable, the thermal stratification will recover and continue to the long-term cooling stage. The flash process can cool the CMT wall and cause the reverse heat transfer on the CMT wall. In the process of CMT circulation and discharge, the heat storage on the CMT wall will be released gradually after ADS is started. The CMT tests of the above different overall test facilities are summarized. The results show that under LOCA accident conditions, the transient response of CMT can be divided into two significantly different stages, natural circulation stage and steam substitution stage. In the natural circulation stage, the more obvious phenomena or parameters that can be identified include the natural circulation rate of the system, boron migration, thermal stratification and break positions. In the steam substitution stage, the more obvious phenomena or parameters that can be identified include flashing, steam condensation, wall heat storage and release, etc.

4 Analysis of Thermal Hydraulic Phenomena in CMT Under different operation modes, the natural circulation process between core and CMT will be affected by phenomena or parameters in CMT as described in Sect. 2.1, such as thermal stratification caused by temperature difference in CMT, migration and mixing of boron, condensation between steam and CMT free fluid surface, condensation between steam and CMT wall, wall heat storage, etc. Based on the working mode and operating parameter conditions of CMT, the main local phenomena in CMT can be further analyzed and discussed. (1) Migration and mixing of boron Under accident conditions, the boron contained cold water in CMT will migrate and mix with the flow of safety injection, which mainly occurs in the early stage of natural circulation mode. According to the research of Zhao, boric acid has good follow feature in water, and its concentration change is mainly affected by the flow rate and temperature of safety injection fluid [10]. In terms of influencing factors, the safety injection flow rate is mainly determined by the natural circulation phenomenon between the core and CMT, that is, the matching between the driving force of density difference and the resistance of CMT branch. The temperature distribution in CMT is mainly determined by the phenomenon of thermal stratification. Therefore, the migration and mixing of boric acid are mainly affected by the thermal hydraulic phenomenon in CMT, and the influence of this phenomenon itself on the flow and temperature distribution in CMT can be ignored. Therefore, this phenomenon can be decoupled from other thermal and hydraulic phenomena in CMT and studied separately. (2) CMT level The mixing liquid level in CMT plays an important role, because this parameter is often related to the setting value of safety injection system [2, 6], which has a significant impact on the process of LOCA accident. However, in terms of CMT thermal hydraulic research,

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this parameter is affected by multiple physical processes such as steam condensation, cold and hot fluid mixing and CMT core natural circulation in CMT at the same time, which is the result of the coupling and joint action of these processes. Therefore, the research on CMT level should rely on the accurate reproduction and coupling operation of phenomena such as steam condensation, cold and hot fluid mixing and CMT core natural circulation, and it is of little significance to carry out experimental research alone. (3) Flash The flashing phenomenon in CMT mainly occurs in the working stage of automatic pressure relief system. Its mechanism is that flashing will lead to instantaneous evaporation of hot water and produce a large amount of steam. The direct result is that the hot fluid in CMT will be reduced due to vaporization, and the thermal stratification phenomenon will be affected; The indirect result is that the steam formed by CMT hot fluid layer will change the density difference between CMT and core. Under the action of gravity, the density difference will affect the cyclic driving force of CMT branch. The characteristic time of the direct effect of flash phenomenon is consistent with the time scale of the depressurization process, and the characteristic time of the indirect effect is equivalent to the time scale of the natural circulation [11]. Flash phenomenon is a complex heat and mass transfer process, and there is no ideal correlation so far. The existing literature research shows that, the main parameters affecting flash are pressure drop, pressure reduction rate and initial temperature [12]. Therefore, for the experimental verification of prototype design, isobaric mode should be adopted whether overall effect experimental research or single effect experimental research. Otherwise, the flash test phenomenon obtained will be quite different from the prototype design. (4) Steam condensation Under the steam substitution mode, it can be divided into CMT hydrothermal layer condensation and wall condensation according to different condensation locations [13]. For the design without steam sparger on the top of CMT, the condensation phenomenon is mainly free fluid surface condensation, and the direct contact condensation of steam and water is dominant [14]. Since the heat transfer coefficient of direct contact condensation is very dependent on the flow state, it is necessary to correctly identify the flow state before making a correct prediction of the relevant heat transfer rate. The research on this phenomenon should be carried out separately. The rate and duration of direct contact condensation are affected by the thermal stratification of CMT liquid. Therefore, the study of direct contact condensation in CMT usually needs to consider the thermal stratification. At the same time, studies have shown that, in this design scheme, the steam condensation process will have a great impact on the driving head of CMT drainage, resulting in fluctuations in the drainage process, and even stagnation of safety injection process in serious cases. In order to avoid the adverse effect of rapid condensation, a steam distributor is configed on the top of CMT to change the steam injection direction, so as to avoid steam heating the liquid on the upper part of CMT.

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For the design of steam sparger on the top of CMT, after the steam enters CMT, its flow direction is changed to spray directly to the wall, and the steam tends to contact the wall preferentially, resulting in the condensation on the wall is much greater than that on the hydrothermal layer. At the same time, due to the higher thermal conductivity of the metal wall, the condensation amount of steam on the wall is more dominant than that in the hydrothermal layer. Therefore, it can be considered that the condensation in CMT is mainly the wall condensation process in this kind of design [15]. (5) Wall heat storage The wall condensation process is closely related to the wall conditions [15, 16]. Due to the large CMT tank and the large design wall thickness due to the pressure bearing requirements, the thick steel wall of CMT will act as a good medium for condensing steam under accident conditions. The CMT tank itself forms a cold source with large capacity. Therefore, the heat conduction phenomenon and overall heat storage release inside the CMT wall also have a great impact on the condensation process in CMT, which also needs to be considered in modeling. In the working stage of ADS, with the pressure drop of the RCS, flash occurs in the CMT, resulting in the decrease of the internal hot fluid temperature. At this time, the phenomenon of wall reverse heat transfer may occur. The phenomenon of wall reverse heat transfer should be considered from the point of view of heat transfer. Different from the cold source release of the tank steel wall during wall condensation, the temperature difference between the hot fluid and the wall is usually small during reverse heat transfer, and the heat storage release rate is slow. Therefore, this phenomenon has a very limited impact on the water of CMT. Considering the time characteristics of physical process, the characteristic time of condensation process is mainly related to the action scale of wall condensation heat transfer, and the characteristic time of wall heat storage and release is mainly related to the action scale of heat conduction process in CMT wall. (6) Thermal stratification According to the test results described in Sect. 2.1 and the research of Li, there are obvious thermal stratification in CMT under the two operation modes of circulation mode and steam substitution [17]. In the operation mode of nature circulation, the movement and increase of hydrothermal layer are mainly determined by the flow of single-phase natural circulation, and its characteristic time is consistent with that of single-phase natural circulation. In the steam substitution stage, the movement and increase of the hydrothermal layer are mainly affected by the liquid flow after steam condensation. Therefore, the moving process of thermal stratification is mainly affected by CMT core natural circulation and wall condensation. For the stability of thermal stratification, according to the research of Yu Pei and Wang Shengfei [18, 19], in the stratification phenomenon of cold and hot fluid in CMT, the stability of stratification interface is mainly related to the temperature difference of cold and hot fluid, the initial height difference of cold and hot fluid area and the external disturbance velocity. The external disturbance velocity can be considered as the

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moving speed of thermal stratification location in CMT, which is mainly determined by the gravity drainage process of CMT.

5 Summary and Suggestions on Thermal Hydraylic Phenomena Related to CMT CMT and its branches are taken as the research object, and the branch is decomposed systematically, as shown in Fig. 3. In combination with Fig. 3, the above phenomena are summarized and analyzed, as shown in Table 1. In Table 1, the levels of various phenomena and the characteristic time scales of their functions are also divided, and the thermal hydraulic phenomena of CMT are summarized.

Fig. 3. Break down diagram of the CMT and its related loop

In combination with Table 1, research suggestions on thermal hydraulic phenomena related to CMT are as follows: (1) The natural circulation of CMT and its branches is a system level phenomenon, and the similarity criteria of this phenomenon should be followed first when conducting experimental research. (2) The physical characteristic time of flash, boron migration and mixing phenomenon has little to do with the natural circulation process, so it can be tested and studied separately. (3) The phenomenon of direct contact condensation is mainly related to thermal stratification, its physical characteristics are related to thermal parameters, and has little to do with natural cycle process, so it can be tested and studied separately.

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(4) The phenomena of wall condensation, wall heat storage, and CMT mixed liquid level are all related to the natural circulation process. In the case of studying these Table 1. Analysis of thermal hydraulic phenomena in CMT Thermal hydraulic phenomena

Operating characteristics

Phenomenal hierarchy

Phenomenon characteristics/time scale

Mechanism description

Single phase natural circulation

Circulation mode

System

Cycle speed - loop length

Occur in the early stage of the break and is driven by the density difference between the cold and hot ends

Two phase natural circulation

Drainage mode

System

Cycle speed - loop length

The generation of two phases is related to the system pressure, and the driving force is gravity

Migration and mixing of boron

Initial stage of circulation mode

Local

Determined by natural circulation flow rate

Do not affect other CMT thermal hydraulic phenomena

Flash

Steam substitution Local mode

Heat and mass transfer process

Related to ADS pressure relief and break

Direct contact condensation

Steam substitution Local mode

Heat and mass transfer process

Without steam sparger, surface condensation is dominant, and the condensation rate is affected by the temperature of the heated liquid layer

Wall condensation

Steam substitution Local mode

Characteristic time With steam of condensation heat sparger, wall transfer condensation is dominant

Wall heat storage

Circulation mode

Heat conduction characteristic time

Local

Wall cold storage release is conducive to maintaining the density difference (continued)

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Table 1. (continued) Thermal hydraulic phenomena

Thermal stratification

CMT level

Operating characteristics

Phenomenal hierarchy

Phenomenon characteristics/time scale

Mechanism description

Steam substitution Local mode

Heat conduction characteristic time

With the release of wall cold storage, the steam condensation rate decreases, the cold storage affects the condensation amount and time

Circulation mode

Local

Single phase natural circulation characteristic time

Mainly determined by natural circulation speed

Steam substitution Local mode

Single phase natural circulation and condensation heat transfer characteristic time

The increase of the hydrothermal layer is determined by the amount of condensation

Steam substitution Local mode

Single phase natural circulation and condensation heat transfer characteristic time

Determined by drainage rate and condensation

phenomena above or analyzing the relevant test processes, the interaction between the characteristic times of natural circulation process and of these phenomena of concern in the design prototype should be considered.

References 1. Schulz, T.L.: Westinghouse AP1000 advanced passive plant. Nucl. Eng. Des. 236(14) (2006) 2. Lin, C.: Passive Safety Advanced Nuclear Power Plant AP1000. Atomic Energy Press (2008) 3. Tuunanen, J., Vihavainen, J., ‘Auria F.D., et al.: Assessment of passive safety injection systems of ALWRs (1999) 4. Yonomoto, T., Kondo, M., Kukita, Y., et al.: Core makeup tank behavior observed during the Rosa-AP600 experiments. Nucl. Technol. 119(2) (1997) 5. Ji, F.: Study on Core makeup water experiment of AC600 makeup water tank. Nucl. Power Eng. (1999) 6. Bae, H., Ryu, S.U., Jeon, B.-G., et al.: Core makeup tank injection characteristics during different test scenarios using SMART-ITL facility. Ann. Nucl. Energy 126 (2019)

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7. Lee, S.I., No, H.C., Bang, Y.S., et al.: Assessment of RELAP5/MOD3.1 for Gravity-Driven Injection Experiment in the Core Makeup Tank of the CARR Passive Reactor (CP-1300). Office Sci. Tech. Inf. Tech. Rep. (1996) 8. Sang, I.L., No, H.C.: Improvement of direct contact condensation model of RELAP5/MOD3.1 for passive high-pressure injection system. Ann. Nucl. Energy 25(9), 677–688 (1998) 9. Li, Y., Ye, Z., Zhong, J., et al.: Core makeup tank behavior investigation during ACME integral effect tests. Nucl. Eng. Des. 364 (2020) 10. Zhao, T. : Study of Boron Concentration Distribution in the Downcomer Using Laser Diagnosic Technology. Harbin Engineering University (2017) 11. Zuber, N., Wilson, G.E., Ishii, M., et al.: An integrated structure and scaling methodology for severe accident technical issue resolution: development of methodology. Nucl. Eng. Des. 186(1) (1998) 12. Yingli, G., Wei, D., Junjie, Y., et al.: Influence of the initial condition on pool water instantaneous flash evaporation. J. Eng. Thermophys. 08, 1335–1338 (2008) 13. Li, K.: Investigation of Direct Contact Condensation of Saturater Steam on Free Subcooled Liquid. Chongqing University (2003) 14. Zhu, M., Chang, H., Wang, H., et al.: Evaluation of distortion of wall stored energy in core make-up tank test facility. Prog. Nucl. Energy 100 (2017) 15. Peng, Y.: Effect of Initial Condition on Temperature Distribution and Steam Condensation Behavior During CMT Drainage. Chongqing University (2003) 16. Chengcheng, D., Huajian, C., Benke, Q., et al.: CMT scaling analysis and distortion evaluation in passive integral test facility. At. Energy Sci. Technol. 47(11), 2026–2032 (2013) 17. Kuining, L., Yudong, L., Mingwei, T.: Experimental research of direct contact condensation of steam on subcooled liquid. J. Eng. Thermophys. 06, 977–979 (2007) 18. Yu Pei, L., Haijin, Y.C.: Experimental research and analysis on location of stratified interface on density lock. Nucl. Power Eng. 31(03), 83–87 (2010) 19. Wang, F., Yan, C., Gu, H., et al.: Development of steady-state heat tramsfer model in density lock. At. Energy Sci. Technol. 44(02):183–187 (2010)

Alara Methodology for Reactor Building Shielding Design in HPR1000 PWR Guanghao Zeng(B) , Yonghai Zhou, Qianqian Huang, Shouhai Yang, and Weifeng Lv Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China [email protected]

Abstract. The reactor building is the area with highest radiation risk in nuclear power plant (NPP) and a systematic shielding design optimization procedure is absent in shielding design process. In order to find out more shielding design improvement issues in reactor building, the as low as reasonably achievable (ALARA) methodology is established in combination with ALARA principle and optioneering procedure. By applying ALARA methodology for fuel transfer canal shielding design, the shielding design around fuel transfer canal is optimized and the optimal shielding design option can reduce the construction cost, speed up the design schedule and satisfy the nuclear safety. Keywords: Alara · Reactor building · Shielding design · Fuel transfer canal · Optioneering

1 Introduction The main radioactive sources of a nuclear power plant during normal operation are the reactor core, the reactor coolant, the nuclear auxiliary systems and the spent fuel assemblies stored in fuel building. As the reactor building contains the reactor core and the reactor coolant, the radioactive material not only includes neutron and gamma sources created by reactor core and reactor coolant during power operation, but also includes N16 generated by the activation of oxygen in the reactor coolant, the fission products leaked from fuel element cladding and the corrosion products that are activated when passing through reactor core. Hence, the reactor building is the area with the highest radiation risk in nuclear power plant. Meanwhile, systematic shielding design optimization as well as a methodology as shielding design guidance is absent during the shielding design of inservice reactor building. Hence, the establishment of ALARA methodology to optimize reactor building shielding design is necessary.

2 Review of In-Service Reactor Building Shielding Design Daya Bay NPP and in service Chinese pressurized water reactor (CPR) [1–3] required that the reactor building shielding design should ensure that the area dose rate does not exceed the dose rate limit during normal operation in order to meet the requirements of © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 23–37, 2023. https://doi.org/10.1007/978-981-19-8899-8_3

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Regulations for the Safe Transport of Radioactive Material [4]. Moreover, the concept of ALARA is implemented to policy, design and operation in nuclear power plant. HPR1000 pressurized water reactor (PWR) inherits and develops the design concept of in service CPR [5], and apply ALARA concept to improve shielding design, such as radiation protection improvement of airborne radioactivity, addition of radioactive gas protective equipment like gas suit and gas mask. Meanwhile, barriers and warning signs around high radiation area are set up and working permission is mandatory to enter high radiation area. In terms of shielding design, the design experience of the in-service CPR is applied to HPR1000 PWR the treatment of corrosion products is optimized to reduce source term, such as application of corrosion-resistant alloys and improvement of water chemistry. For the European PWR in Taishan nuclear power station, ALARA is applied to the area dose rate assessment around reactor core including the reactor pit, the compartment outside the primary shielding wall, and on the operating platform. In China, balance of cost and benefit in shielding design is comprehensively analyzed in AP1000 advanced passive plant by considering ALARA such as material optimization [6] to reduce effective dose rate in reactor building. Meanwhile, ALARA is also applied in layout, pollution control, material selection and field operation [7, 8]. In UK [9– 11], ALARA is applied to optimize the shielding design of equipment and system in reactor building, including decentralization of in-containment refueling water storage tank (IRWST) pool injection valves and optimization of the secondary side waste heat discharge system of the steam generator. In Sellafield nuclear plant [12], ALARA have been applied to continuously optimize shielding design in 30-year-cycle safety assessment, such as the addition of neutron detection equipment to detect corrosion deposits in product finishing facilities. European Pressure Reactor (EPR) NPP in France [13] and German have applied ALARA [14] to shielding design by referring national laws and regulations. Analysis of relative good practice and optioneering of optimal option is also applied [15, 16]. UK ABWR NPP [17] emphasizes the need to ensure ALARA throughout NPP design and operation and the concept is applied in risk assessment, identification of risk reduction measures, judgment of whether measures are feasible and identification of other measures [18–20] to ensure the design is ALARA and meet the regulatory requirements [21, 22]. By reviewing the reactor building shielding design of international in-service NPP, an ALARA methodology as guidance for reactor building shielding design is absent during design process, which is necessary to establish.

3 ALARA Methodology of Reactor Building Shielding Design ALARA methodology [23–26] has been often applied to optimize design program and assess design risk to find out a balance between risk reduction and cost in different fields. Based on the international standard ISO17776 and relevant good project experience [22, 27] in the field of nuclear industry, the general ALARA methodology process is as follows:

Alara Methodology for Reactor Building Shielding

1. 2. 3. 4. 5. 6. 7.

25

Hazard identification; Selection of an appropriate risk matrix; Risk assessment (Probability times consequence); Ranking the risks, for systematic consideration; Identification of potential risk reduction measures; All risk identified as intolerable must be reduced until tolerable; All risks are reduced to a level ALARA.

In the field of the reactor building shielding design, the ALARA methodology process is developed based on general ALARA methodology process which is shown in Fig. 1. The process includes the following steps: 1. Overall review: this step includes design review, analysis of relevant good practice and risk as well as option generation; 2. Specific analysis: this step includes optioneering, confirmation and implementation of the optimal option; 3. Evaluation review: the review of optimal option ensures ALARA by considering the latest improvements.

3.1 Overall Review A. Design Review The main radioactive components with potential high radioactivity in reactor building include the following parts: 1. 2. 3. 4. 5. 6. 7.

Containment; Primary shielding wall; Secondary shielding wall; Operating floor; Reactor pit; Reactor cavity; Fuel transfer canal.

Over the years, the reactor building shielding design of the above parts is continuously improved along with the development of nuclear technology which relate to the high radioactive area by being reviewed against the relevant good practice and operating experience. B. Analysis of Relevant Good Practice and Risk In general, the review and analysis shall be undertaken primarily against codes and standards. In addition, throughout the project, input from authority shall be taken into consideration as part of the process of identifying potential improvements. Additionally, improvement items from previous NPP should be considered.

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Fig. 1. Process of ALARA methodology for reactor building shielding design

C. Option Generation When potential design improvements are identified, these may be grouped and then processed through the following step which is generation of feasible shielding design options. According to international standard ISO17776: 2000 [27], the general hierarchy of risk-reducing measures which help generate options should be carried out as follows: 1. 2. 3. 4. 5.

Prevention (elimination of the hazard); Detection; Control (reduce probability, reduce exposure); Mitigation of the consequences; Emergency respond (curative measures).

Alara Methodology for Reactor Building Shielding

27

Particular attention should always first be given to risk-reducing measures that have the effect of eliminating or reducing the probability of a hazardous event occurring. Based on the ISO17776 standard while referring to the ALARA methodology used in UK ABWR PCSR chapter twenty [11] during General Design Review as well as the UK HSE Design Risk Reduction Guideline [27], the shielding design of HPR1000 PWR developed Eliminate, Reduce, Isolate, Control and Personal Protective Equipment (ERIC/PPE) methodology to identify the potential improvements and generate feasible reactor building shielding design options. The ERIC/PPE methodology is described as follows: 1. Elimination: (a) Removing the radiation source around the area of activities; (b) Replacing the manual tasks with fully automatic tasks; or (c) Replacing the local operations by remote operations. 2. Reduction: Reduce the source term or the frequency and duration of the high dose activities, for example by: (a) Optimizing the material selection, surface finishes, reactor water chemistry regime and decontamination to reduce the source term; (b) Optimizing the process design, system design and equipment design to be more convenient and robust in order to reduce the frequency and duration of the operation and maintenance. 3. Isolation: Isolate the radiation sources or contamination from workers, for example by: (a) Performing radiation zoning to isolate the high radiation areas from low radiation areas; (b) Adopting appropriate shielding or containment to isolate the radioactive sources from workers in order to achieve the shielding optimization target by applying permanent or temporary shielding measures. 4. Control: Implement administrative control to avoid unauthorized entrance or spread of contamination. 5. Personal Protective Equipment (PPE). Within the radiation protection topic, the use of PPE is essential for providing a last level of defense against radioactive contamination. This report has focused on the provision

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of protection against direct radiation dose, and therefore PPE will not be assessed further within this ALARP analysis because it is not considered appropriate for new build power stations to require the use of shielding PPE items. 3.2 Specific Analysis For the feasible options generated by the above ERIC/PPE methodology, the optimal option is carried out by the following steps: optioneering, option implementation and option confirmation. A. Optioneering A comparison and selection of feasible options is carried out. The steps of optioneering are as follows: 1. Selection of the optioneering criteria; 2. Evaluation of the weighting factor of optioneering criteria; 3. Quantitative evaluation method to evaluate the score of each optioneering criteria and the total score for each option. The following criteria are specified to be used for the optioneering: (a) Environment: waste generation, increase in active effluent, increase in non-active effluent and detergents; (b) Safety: contamination spread, response time to avert safety/security incident, familiarization of plant, conventional safety; (c) Discipline-specific factors: compliance with codes and standards; compliance with operation experience; (d) Alignment with design principles and hierarchy: hierarchy of hazard controls and protection; (e) Costs: Waste, cleaning, training, facilities or PPE; (f) Ease of operation: Fleet management, Procedures, Availability of staff; (g) Ease of implementation: Space available, Impact of different approach; Weightings address the relative importance of the various assessment criteria and are applied after the initial scoring. The definition of criteria weighting score is shown in Table 1. Table 1. Definition of criteria weighting score Criteria weighting score 1

2

3

4

No significant impact

Important during operation

Important during design phase

Important in full life

Alara Methodology for Reactor Building Shielding

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Each option obtains a raw score according to its performance in each criterion. The definition of option score is shown in Table 2: Table 2. Definition of option score Option score 1

2

3

4

5

Option not acceptable

Significant difference between the option and the best option

Option acceptable

Minor difference between the option and the best option

Best option

The option with the highest score is the optimal option. Each option in each criterion has a raw score (RS for short) and a weighting score (WS for short). The relation between WS and RS is calculated by the formula WS = RS × W, where W is the weighting factor. B. Confirmation and Implementation of Optimal Option Based on the result of optioneering, the optimal option is confirmed and is implemented according to the management procedure specification. 3.3 Evaluation Review Since new improvements may be generated during the design process, it is necessary to review the optimal option by considering the latest improvements by applying the procedure described in Sects. 3.1 and 3.2 to establish a new optimal option.

4 Application of ALARA Methodology for Reactor Building Shielding Design The reactor building is the area with highest radiation risk in NPP not only because the existence of main radioactive sources in reactor building, but also workers have to perform activities with potential dose uptake inside reactor building during shutdown, which cause high risk in collective dose and individual dose. The spent fuel assemblies in reactor core are transferred underwater through the fuel transfer canal and transported to fuel building during shutdown. The spent fuel transportation is one of the activities with highest radiation risk in reactor building during shutdown as strong gamma and neutron flux leak from spent fuel assemblies. Based on the fact that the spent fuel assemblies are high radioactive source, the shielding design of the nearby area around fuel transfer canal needs to be optimized to achieve ALARA.

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4.1 Overall Review A. Design Review The shielding design of the fuel transfer canal in Daya Bay NPP and in-service CPR1000 PWR, refers to the NPP in France and is shown in Fig. 2. Figure 2 shows that the fuel transfer canal is composed of steel pipeline and is filled with water during the fuel assembly transportation. As the gamma ray and neutron are emitted from the fuel assemblies which has high radiation exposure risk to personnel, the fuel transfer canal itself and the concrete structure around fuel transfer canal build up the shielding structure. The measure data shows that during fuel assemblies transportation, the effective dose rate in the accessible area around the fuel transfer canal exceeds the area upper dose rate limit. To prevent from high personnel radiation exposure, the area around fuel transfer canal is restricted during fuel assembly transportation to ensure personnel radiation safety.

Fig. 2. Elevation view of fuel transfer canal in Daya Bay NPP and CPR1000

The fuel transfer canal of in service EPR NPP is shown in Fig. 3. The concrete is the main shielding material and lead bricks are not considered around fuel transfer canal. To avoid direct irradiation of fuel transfer canal to the personnel in nearby area through seismic joints, a special cylinder shielding made by steel surrounds the fuel transfer canal as a specific shielding design improvement. As shown in Fig. 4, in order to optimize the previous shielding design of in service nuclear power plant, the fuel transfer canal in HPR1000 PWR is optimized to meet the requirements that the area dose rate of the accessible area around the fuel transfer canal does not exceed the area dose rate limit and the passage of personnel is not restricted as well. By comparing the structure in Daya Bay NPP and CPR1000, the concrete structure is optimized and lead bricks are also applied to enhance the shielding effect. Moreover, fence gates are applied at the entrance of the labyrinth to prevent personnel from entering the labyrinth by accident. A. Analysis of Relevant Good Practice and Risk

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Fig. 3. 3D dimension view of fuel transfer canal of in service EPR NPP

Fig. 4. Elevation view of fuel transfer canal in HPR1000 PWR

Radiation risk item is found at the entrance of the fuel transfer canal: the dose rate at the entrance of the fuel transfer canal exceeds the upper dose rate limit of green zone and personnel are at risk of exceeding the exposure dose around annulus when fuel

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assemblies are transporting through the fuel transfer canal as the shielding effort of annulus structure around the fuel transfer canal is not enough. B. Generation of Options Based on the operation experience of in service NPP, the ERIC/PPE methodology is used to identify feasible shielding design options for fuel transfer canal. The radioactive sources in fuel transfer canal cannot be eliminated as the spent fuel assemblies must be transported from the reactor building to the fuel building through the fuel transfer canal during refueling. Therefore, no new options can be generated by applying the measures of radiation source elimination or radiation intensity reduction of radiation source. By following the isolation measures, the following options are generated: 1. Enhancement of the shielding wall of the labyrinth around fuel transfer canal: the enhancement of the shielding wall will increase the wall thickness of the labyrinth which would too narrow for staff to go through, this will cause the emergency safety problem in escape. Therefore, this option is not feasible. 2. Application of heavy concrete: The heavy concrete has better shielding effect and will not compress the passage space inside labyrinth, therefore it is a feasible solution. 3. Application of steel shielding door to replace the fence gate at the entrance of the fuel transfer canal: it can reduce the effective dose rate to the dose rate limit of green zone at the entrance of the fuel transfer canal, therefore it is a feasible solution. By following the control measures, the following option is generated: the position of the fence gates at the entrance of the fuel transfer canal at the annular space and annulus are moved to the entrance of the stairs, which can prevent personnel from entering the fuel transfer canal by accident. Meanwhile personnel control measures in reactor building during fuel assemblies transfer are also implemented to further prevent the staff from entering the labyrinth of fuel transfer canal by accident. Hence, this plan is feasible. Meanwhile, since the area around fuel transfer canal are annular space and annulus in reactor building which are areas without restriction and personnel do not need to wear additional radiation protection equipment in order not to affect the activity convenience. Therefore, “personal protection equipment” cannot form a new option. Based on the above analysis, the shielding design options of the fuel transfer canal are summarized as follows: Option 1: Application of heavy concrete to build up the shielding structure around fuel transfer canal; Option 2: Application of steel shielding door at the entrance of the labyrinth of fuel transfer canal; Option 3: Position adjustment of the fence gate and application of personnel control measures during fuel assembly transportation. 4.2 Specific Analysis A. Optioneering

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33

As feasible options generated from the above process, the optioneering of the fuel transfer canal shielding design option in reactor building is carried out. The criteria to be considered as well as the weighting and score of options in each criterion are shown in Table 3. B. Confirmation and Implementation of Optimal Options As is shown in Table 3, option 3 has the highest raw score and weighting score among three feasible options, which means that option 3 has the best performance by considering all related criteria. Hence, the position adjustment of the fence gate and application of personnel control measures during fuel assembly transportation is the optimal option to improvement the shielding design of fuel transfer canal. 4.3 Evaluation Review During the optimal option establishment process of fuel transfer canal shielding design, no new improvement issues are shown up. Hence the optimal option does not need to redevelop fuel transfer canal shielding design by using ERIC/PPE methodology in combination with newly identified improvements to ensure ALARA. Therefore, the position adjustment of the fence gate and application of personnel control measures during fuel assembly transportation is the optimal option which ensures that the shielding design is ALARA.

5 Conclusion Based on the fact that a systematic shielding design optimization procedure is absent in shielding design process of reactor building, the ALARA methodology for reactor building shielding design in HPR1000 PWR is established. By applying the ALARA methodology, the reactor building shielding design could be optimized to ALARA by considering safety, environment, cost and schedule and related criteria.

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Criteria

W

Performance

Safety

4

Environment

Schedule

Cost

4

3

4

RS

WS

Option 1: The heavy concrete with sufficient thickness can reduce the radiation exposure risk and meet area dose rate limit

5

20

Option 2: A shielding door with sufficient thickness can reduce the radiation exposure risk and meet area dose rate limit

5

20

Option 3: Position adjustment of the fence gate and application of personnel control measures can prevent personnel from entering the fuel transfer canal and personnel can be not affected by high radiation exposure

5

20

Option 1: The application of heavy concrete does not generate radioactive waste and has no impact on the environment

5

20

Option 2: The application of steel shielding door does not produce radioactive waste and have no impact on the environment

5

20

Option 3: Position adjustment of the fence gate and application of personnel control measures does not affect the working environment as it does not produce radioactive waste

5

20

Option 1: additional design and deployment of heavy concrete components is required to complete the shielding structure which affects the schedule

2

6

Option 2: The installation procedure of shielding door is complicated enough that it affects the schedule

3

9

Option 3: Position adjustment of the fence gate has little influence on schedule as it needs extra work to change the gate position. Meanwhile, the application of personnel control measures has no impact on schedule as it is a measure during operation

4

12

Option 1: Additional manufacture of heavy concrete material and working time is required, which means additional material cost and labor cost

2

8

(continued)

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Table 3. (continued) Criteria

Ease of operation

Ease of implementation

Total Score

W

3

2

Performance

RS

WS

Option 2: Addition of steel shielding door increases the cost compared with the fence gate

2

8

Option 3: Position adjustment of the fence gate and application of personnel control measures may increase labor cost in management

4

16

Option 1: Application of heavy concrete has no effect on the NPP operation

5

15

Option 2: The addition of shielding door at the entrance of the labyrinth has no effect on the operation of the nuclear power plant

5

15

Option 3: The control measure takes place during fuel assemblies transportation and few staff is in reactor building. Hence, it does not affect the normal operation of the NPP

5

15

Option 1: Additional manufacture of heavy concrete material needs additional working time and specific method to implemented

3

6

Option 2: The application of steel shielding door is more difficult to implement compared to the fence gate as the installation craft is more complicated

2

4

Option 3: The application of new personnel control measures increases the personnel training cost which cause some implementation difficulty

3

6

Option 1

22

75

Option 2

22

76

Option 3

26

89

Acknowledgement. The work presented in this paper is supported by the flexible shielding material project (No. K-A2020.110) and is managed by CGN (China General Nuclear Power Corporation).

References 1. Chen, D., He, Y., Yang, M., et al.: Practice and experience of radiation protection and optimization (ALARA) management system in Daya Bay NPP during the past 10 years. Radiat. Prot. (Taiyuan) 24(3–4), 131–143 (2004)

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2. Yang, M., Chen, D.: The preliminary success of ALARA implementation in Daya Bay NPP. Radiat. Prot. (Taiyuan) 20(3), 129–138 (2000) 3. Lun, Z., Gao, X., Li, B., et al.: Experience on ALARA management at DAYA Bay NPP. Radiat. Prot. (Taiyuan) 37(4), 287–292 (2017) 4. Series, I.S.S.: Regulations for the Safe Transport of Radioactive Material. TS, (1996) 5. Lyu, W., Ran, W., Liu, J., et al.: Development and validation of rapid 3D radiation field evaluation technique for nuclear power plants. Sci. Technol. Nucl. Installations 2020 (2020) 6. Li, H., Zanf, Y., Hu, Y.: Analysis on radiation protection design of AP1000 unit of Sanmen nuclear power plant. Radiat. Prot. 41(1), 44–49 (2021) 7. Su, R., Tang, Y.: Layout design of reactor coolant system of AP1000 NPP. China Nucl. Power 7(1), 4–8 (2014) 8. Cui, Y., Wu, M., Zhao, W., et al.: Test of dissolution of simulated corrosion-oxides in primary loop of the AP1000 reactor. J. Radioanal. Nucl. Chem. 326(2), 1151–1158 (2020) 9. Doehnert, B.: Design of the Ap1000 Power Reactor. Westinghouse Electric, Belgium (2006) 10. Ali, S., Qureshi, K., Gulfam, S., et al.: Safety assessment of advanced PWR. In: 2008 International Conference on Power Generation Systems and Renewable Energy Technologies (PGSRET), pp 1–6. IEEE, 2018 11. Hitachi-GE Nuclear Energy, Ltd.: UK ABWR Generic Design Assessment - Generic PCSR Chapter 20: Radiation Protection, GA91-9101-0101-20000 (XE-GD-0652), Revision C, December (2017) 12. Bounds, A.: 30 years of safety assessment at Sellafield. Saf. Reliab. 31(1), 27–34 (2011) 13. Sas, A.N.P., Electricité de France, S.A.: Public report on the generic design assessment of new nuclear reactor designs (2009) 14. Xiaoyu, M., Kan, W.: EPR: doses and radiation protection: a clear step forward in comparison with former generation of PWRs in France and Germany. China: N. p., 2005. Web 15. Miniere, D., Beneteau, Y., Le Guen, B.: E-mail: [email protected]. Optimisation of radiation protection for the new European pressurized water reactor (EPR). Argentina: N. p., 2008. Web 16. Stricker, L., Dollo, R.: The ALARA policy of Electricite de France. Radioprotection 30(1), 47–60 (1995) 17. Suzuki, N., Ishizawa, N., Yamazaki, K., Sato, Y., Ohsumi, K., Akamine, K., Hosokawa, K.: An ABWR Water Chemistry Control Design Concept for Low Radiation Exposure and the Operating Experience at the first ABWR. France: N. p., 2002. Web 18. Fujii, M., Morooka, S., Heki, H.: Application of Probabilistic Safety Analysis in Design and Maintenance of the ABWR. Advances in Light Water Reactor Technologies, pp. 1–30. Springer, New York, NY (2011) 19. Ishiwatari, Y., Hisamochi, K., Hirokawa, N., et al.: Risk-informed design for UK ABWR project. In: Topical Issues in Nuclear Installation Safety: Safety Demonstration of Advanced Water Cooled Nuclear Power Plants. V. 1. Proceedings of an International Conference, 2018 20. Ang, M.L., Mohamud, N., Chitose, H., et al.: A Demonstration of Practical Elimination of Early or Large Fission Product Release for the UK ABWR Generic Design 21. Ionising Radiations Regulations 2017. UK: www.leguslation.gov.uk (2017) 22. ONR UK.: Safety Assessment Principles for Nuclear Facilities (2014) 23. Valeur, J.R., Ruggiero, L.: ALARP methodology used for BAT environmental impact reduction. In: SPE International Conference and Exhibition on Health, Safety, Security, Environment, and Social Responsibility. OnePetro (2016) 24. Valeur, J.R.: Integrated HSE ALARP assessment-reducing HSE risks and impacts in the same Processes. In: SPE Middle East Health, Safety, Environment & Sustainable Development Conference and Exhibition. OnePetro, 2014 25. Valeur, J.R., Petersen, J.: Use of the ALARP principle for evaluating environmental risks and impacts of produced-water discharged to sea. Oil and Gas Facil. 2(06), 92–100 (2013)

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26. Nesticò, A., He, S., De Mare, G., et al.: The ALARP principle in the cost-benefit analysis for the acceptability of investment risk. Sustainability 10(12), 4668 (2018) 27. ISO BS. 17776: Petroleum and natural gas industries-offshore production installationguidelines on tools and techniques for hazard identification and risk assessment, 1–45 (2000) 28. Salmon, D.: HSE Information sheet: Guidance on Risk Assessment for Offshore Installations. Offshore Information Sheet No.3/2006, Health & Safety Executive (HSE), London, UK (2006). http://www.hse.gov.uk/offshore/sheet32006.pdf 29. HSE, UK.: Guidance on Risk Assessment for Offshore Installation (2013)

The Study of Shielding Benchmark of Heating Exchanger for Operating NPP Zeng Guanghao1(B) , Yang Shouhai1 , Huang Xinming2 , Jiang Zhenyu1 , and Gong Quan1 1 Shenzhen CGN Engineering Design Co., Ltd., Shenzhen, Guangdong, China

[email protected] 2 Daya Bay Nuclear Power Operations and Management Co., Ltd., Shenzhen, Guangdong,

China

Abstract. The shielding benchmark of Chinese NPP lacks unified requirements and standards. The uneven quality database greatly influences the application of shielding benchmark data bank. In order to further improve Chinese standard shielding benchmark, the shielding benchmark of heating exchanger for operating NPP by applying the SuperMC code based on the Monte Carlo (MC) method is established by applying the measured data of source strength and environmental effective dose rate. The comparison between the calculation result and the measured data of environmental effective dose rate caused by the heating exchanger is within acceptable error, which can provide data verification support for other software and the verification results can be included in the benchmark data bank. Keywords: Shielding design · Shielding benchmark · Heating exchanger · Monte Carlo · Supermc

1 Introduction After years of research, a certain amount of experimental data and operation data of nuclear power plant (NPP) as well as shielding benchmark have been accumulated in established benchmark data bank in China. However, the current data bank is still not sufficiently representative and brings difficulties to the development and the autonomous process of safety analysis software. Furthermore, the uneven quality of the data bank makes itself not recognized internationally, which affects the application of the data bank. Hence, more adequate and in-depth work needs to be carried out in conjunction with operating nuclear power plants to establish more shielding benchmark to enrich the shielding benchmark data bank. By collecting the effective dose rate measured data of in-service NPP and modeling heating exchanger with SuperMC code which is a domestic self-developed Monte Carlo simulation software, the heating exchanger shielding benchmark for operating NPP is established by correcting with measured data, which accomplish the production of autonomous domestic shielding benchmark.

© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 38–46, 2023. https://doi.org/10.1007/978-981-19-8899-8_4

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2 Development of Shielding Benchmark The project Shielding Integral Benchmark Archive and Database (SINBAD), which started in 1996 [1], aims to establish a set of radiation shielding and dosimetry data relative to experiments relevant in reactor shielding, fusion blanket neutronics and accelerator shielding [2]. As for reactor shielding, SINBAD has 37 experiments and the benchmark has been applied for computer code validations or nuclear data. The International Thermonuclear Experimental Reactor (ITER) [4] benchmark model is applied mainly to test the CAD/Monte Carlo codes developed by different researchers. To verify and demonstrate the theoretical calculation results by different software based on Monte Carlo method such as MCNP and SuperMC, the ITER benchmark model is used for software testing and verification [5]. In China, researchers have been working on establishment and enrichment of shielding benchmark data bank. Domestic companies in the field of nuclear power and scientific research institutions have accumulated a certain amount of operation data of NPP and benchmark examples. Although some test database has been established, a large amount of operating measure data in domestic operating NPP is still missing for a complete data bank. Hence, it is necessary to accumulate more operating data of in-service NPP and improve the database quality of shielding benchmark.

3 Methodology of Shielding Benchmark Establishment 3.1 Introduction of Simulation Software Monte Carlo (MC) method can simulate accurately shielding benchmark problem with complicated and large-scale geometry as well as large energy range and complex energy spectrum structure of photons in NPP [6]. SuperMC code is a computer aided design (CAD)-based Monte Carlo program for integrated simulation of nuclear system [7]. It is a multi-functional calculation program for nuclear design and safety evaluation, which has already become a typical nuclear analytical code used for theoretical nuclear simulations of reactor design, shielding assessment, reactor criticality studies, etc. [8]. 3.2 Establishment Process of Shielding Benchmark To establish a shielding benchmark with high accuracy, the key factor is to carry out the theoretical analysis by theoretical shielding benchmark model correction through iterative calculation with measured data in order to establish an accurate shielding benchmark [9]. Based on the key factor, the establishment process of shielding benchmark which is presented in Fig. 1 is carried out by following the main steps: Step 1: Effective dose rate and source term measurement of the equipment which is the objective of shielding benchmark. It should be noted that the measurement environment should also be recorded such as the position of the measurement point and the operating condition of the radioactive equipment etc.

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Step 2: SuperMC code is applied to accomplish high accuracy modeling by referring the equipment technical parameters. The calculation detector point of effective dose rate is identical to the measurement point. Step 3: By analyzing the error between calculation results and measurement data, the shielding benchmark model is corrected through iterative calculation to approach the measurement model.

Fig. 1. Establishment process of shielding benchmark

4 Establishment of Heating Exchanger Shielding Benchmark The heating exchanger is a typical radioactive equipment in NPP. It contains radioactive primary coolant and corrosion product attached to the inner surface of the equipment. Hence, it is typical to take the heating exchanger as shielding benchmark study objective.

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4.1 Measurement of Source Term and Effective Dose Rate As shown in Fig. 2, the heating exchanger is a convection U-shaped equipment with coolant tube inside the shell. The down-flow with radionuclides from the primary coolant system passes through the tube side of the heating exchanger, and the purified upper charge flow passes through the shell side, so as to achieve the cooling effort inside heating exchanger. As shown in Fig. 3, a series of measurement points are located outside the surface of heating exchanger. The distance between each measurement point is 20cm which could ensure a reasonable proportional attenuation of effective dose rate.

Fig. 2. Heating exchanger in NPP

Meanwhile, three heating exchangers with the same function and size in different units is chosen as measure objective. The measurement position and the dosimeter is identical for each measurement. The source term analysis data of primary coolant inside heating exchanger is shown in Fig. 4. The energy is mainly concentrated in the low energy group. The effective dose rate of three heating exchanger is measured and 12 effective dose rate measurement data of heating exchanger is shown in Fig. 5. Meanwhile, the effective dose rate contribution ratio of each radionuclide in the heating exchanger is shown in Fig. 6. It can be seen from Fig. 6 that Co-60 and Ag-110 m are the main effective dose rate contributors. In the following step of shielding benchmark establishment, the radionuclide contribution ratio is respected so that the effective dose rate distribution and the radionuclide contribution ratio of the shielding benchmark model will be closer to the actual situation.

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Fig. 3. Design drawing of heating exchanger and position of measure point

Fig. 4. Fluid Source term in different heating exchanger

4.2 Modeling of Shielding Benchmark After the completion of effective dose rate measurement, the modeling of shielding benchmark is the next step by following the methodology in Sect. 3.2. The technical parameters of heating exchanger are shown in Fig. 7 while the position of measurement point is also presented which determines the calculation detector position. By applying SuperMC code, the heating exchanger model is established by referring the heating exchanger technical parameter and the measurement point position. The model of heating exchanger shielding benchmark is shown in Fig. 7.

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Fig. 5. Effective dose rate measurement data of heating exchanger

Fig. 6. Corrosion product source term of heating exchanger

4.3 Shielding Benchmark Establishment and Validation Based on the source term measurement data of the heating exchanger, by applying the established heating exchanger model, the effective dose rate of the three different heating exchanger is calculated and the values are shown in Fig. 8. The Fig. 8 shows that the error between calculation data and measurement data is within the range of 76% ~ 88% which is not accepted and the heating exchanger benchmark model needs to be corrected in order to approach the measurement data. In order to optimize the calculation values of effective dose rate and minimized the error between calculation data and measurement data, the iterative calculation process is applied by correcting the source term of corrosion product due to the complexity of corrosion product immigration and the difficulty of corrosion product source term measurement.

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Fig. 7. Model of heating exchanger shielding benchmark

Fig. 8. Calculation values and measure data of effective dose rate of heating exchanger

By applying the effective dose rate measurement data and referring the operating experience of corrosion product source term, the shielding benchmark correction factor is defined and is applied to correct the total source strength of corrosion product in order that the calculation values of effective dose rate in iterative calculation process approach the measurement data. As shown in Fig. 9, the results present a good agreement between measurement data and calculated effective dose rate of new heating exchanger shielding benchmark model. Most of the error between these two results is below 10%, which is acceptable. Hence, the established heating exchanger shielding benchmark could be applied to verify other calculation software.

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Fig. 9. Effective dose rate measurement data of heating exchanger and iteration calculation

5 Conclusion By applying the shielding benchmark establishment methodology with SuperMC code, the shielding benchmark model of heating exchanger is built up and the calculation result of effective dose rate has good coincidence with measurement data within acceptable error. Moreover, the heating exchanger shielding benchmark can be applied to the verification of other calculation software or nuclear data. Since the corrosion product source term is difficult to measure and operating experience from NPP is applied to evaluate the effective dose rate contributed by corrosion product, the heating exchanger shielding benchmark could be optimized with more accurate corrosion product source term as input with the help of real-time sampling measurement. Acknowledgement. The work presented in this paper is supported by the flexible shielding material project (No. K-A2020.110) and is managed by CGN (China General Nuclear Power Corporation).

References 1. Kodeli, I.A., Sartori, E.: SINBAD-Radiation shielding benchmark experiments. Ann. Nucl. Energy 159, 108254 (2021) 2. Feng, X.Y., Zhang, P., Lee, H., et al.: Validation of MCS code for shielding calculation using SINBAD. Nucl. Eng. Technol. (2022) 3. Chen, C., Yang, Q., Wu, B., et al.: Validation of shielding analysis capability of SuperMC with SINBAD. In: EPJ Web of Conference. EDP Science, vol.153, p. 02009 (2017) 4. Wu, Y., Song, J., Zheng, H., et al.: CAD-based Monte-Carlo program for integrated simulation of nuclear system SuperMC. Ann. Nucl. Energy 82, 161–168 (2015) 5. Wilson, P.P.H., Feder, R., Fischer, U, Loughlin, M., Petrizzi, L., Wu, Y., et al.: State-of-the-art 3-D model transport methods for fusion energy systems. Fusion Eng. Des. 83, 824–833 (2008) 6. Song, J., Sun, G., Chen, Z., et al.: Benchmarking of CAD-based SuperMC with ITER benchmark model. Fusion Eng. Des. 89(11), 2499–2503 (2014)

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7. Baidoo, I.K., Li, B., Zou, J., et al.: Analysis of gas-cooled fast reactor integral shielding experiment: “Wuerenlingen iron benchmark experiment (PROTEUS)” using SuperMC. Ann. Nucl. Energy 152, 108016 (2021) 8. Yilmaz, L.: Applying Monte Carlo Simulation in New Tech. In: Public Sector Crisis Management. IntechOpen (2020) 9. Cheng, E.T., Forrest, R.A., Pashchenko, A.B.: Report on the second international activation calculation benchmark comparison study. International Atomic Energy Agency (1994)

An Anisotropic Porous Model for Heat Exchanger Modeling of Fluoride-Salt-Cooled High-Temperature Advanced Reactor -FuSTAR Xinyu Li, Dalin Zhang(B) , Xinze Li, Xingguang Zhou, Xinan Wang, Tongan Yang, Wenxi Tian, and Suizheng Qiu Xi’an Jiaotong University, Xi’an, Shaanxi, China [email protected]

Abstract. The birthplace of the Chinese nuclear industry, the vast western region offered enough strategic space maneuvers and deep defense, and it is urgent to develop safe and efficient multi-level energy supply systems. In consequence, led by Xi’an Jiaotong University, a new generation of no water cooling reactor technology was put forward: Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR) system, which is equipped with the Printed Circuit Heat Exchanger (PCHE). To predict global temperature, flow field, and stress estimation of PCHE, a porous media model needs to be established for rapid design and analysis. Given the apparent anisotropic geometry in PCHE, the conventional porous media method is no longer appropriate. Therefore, in this paper, the anisotropic tensor form of permeability was derived based on the Brinkmann equation, and an empirical anisotropic porous media model was constructed based on the existing resistance coefficient correlations. In addition, a three-dimensional CFD was used to verify the results. The results showed that the anisotropy porous model can simulate the apparent velocity and flow pressure drop characteristics of the whole field well, and the anisotropy is realized by assigning permeability tensor in each zone, which lays a foundation for the subsequent heat transfer coupling design. The methodology presented in this paper may also be applied to the modeling and design of porous media in other complex heat exchangers. Keywords: Anisotropic porous · PCHE · FuSTAR · Resistance correlations · CFD

1 Introduction The Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR) system was proposed by Xi’an Jiaotong University, with high temperature and low-pressure operation, no water cooling, inherent safety, the characteristics of compact structure. FuSTAR can work in dry areas to achieve efficient power, to produce above 700 °C hightemperature process heat, and also be used for the high-temperature hydrogen production, desalination of salt water, mineral exploitation, etc., especially fit the demands of multi-level energy in remote areas in western China [1]. © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 47–58, 2023. https://doi.org/10.1007/978-981-19-8899-8_5

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In FuSTAR, the Printed Circuit Heat Exchanger (PCHE) is widely used to achieve the miniaturization of the reactor [2]. Considering the maturity of the project and the ease of processing, the channel structure is mainly used, and the flow and heat transfer process can be summarized as a one-dimensional (1D) process. Therefore, accurate and efficient prediction of PCHE in a one-dimensional flow and heat transfer process can obtain the relevant physical phenomena in the stage of design and guide early engineering. At present, the design and calculation in the processes of one-dimensional to threedimensional (3D) flow and heat transfer of a single plate of PCHE have been realized for FuSTAR, where the 1D partial differential equations are used for the flow process, and 3D partial differential equation is used for the solid heat transfer process. This can be used to analyze the 3D field of temperature, mass flow rate, and pressure of a single plate, and the results can be applied to analyze the Von Mises stress. Figure 1 shows the diagram of calculation on flow - heat transfer - mechanics in 1D − 3D, which is executed on Intel i7-9700 for 2 min.

Fig. 1. Temperature, stress, and deformation fields calculated in 1D + 3D

The 1D − 3D method can accurately get the location of local high thermal stress, especially for cross-flow channels, and its computable scale is much larger than full 3D modeling. However, the 1D − 3D method can only calculate the medium-scale schemes even though it greatly simplifies the computation. In our test, an Intel 6258R processor will take 5 h to calculate the 30 channels for a single board. To make matters worse, in FuSTAR, a single board of PCHE has hundreds or even thousands of channels, so such long computing times make the design and optimization almost impossible. In addition, CFD or empirical methods based on single-channel cannot reflect the heterogeneous effects of cross-flow between the hot and cold channels, while the maximum thermal stress may occur there [3]. Therefore, there is still no sufficiently efficient method to know the 3D stress field in PCHE at present, and it can only be estimated roughly.

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In this paper, an anisotropic porous model is proposed to simulate the threedimensional overall flow characteristic of PCHE in FuSTAR. In addition, the 1D in 3D space flow model was also used for comparison with the anisotropic porous model. The porous model takes into account the anisotropy of the flow channel to obtain the distribution of pressure and mass flow rate, which lays the foundation for the development of an equilibrium or non-equilibrium porous heat transfer model.

2 Methodologies 2.1 Mathematical and Physical Model Anisotropic Porous Media Model The coolant in FuSTAR is FLiBe, and the working medium in the power circulation system is CO2 , where the flow and heat transfer in PCHE with Ma < 0.3. Therefore, the N-S equation is used to couple the Brinkman equations and standard k-ε turbulence model without the source terms of mass and volume force [4, 5], see (1):     u  d ⊗ ud = α∇ · (−pI + K) − α μκ −1 + cF κ −1 ρ|ud | · ud ∇· ρ ⎧α  T  

⎪ ⎨ (μ + μ ) ∇ ud + ∇ ud − 2 ∇ · ud I − 2 ρkI, α = 1 T α α 3 α 3 K= ⎪ ⎩ 0, α ∈ (0, 1) (1) ⎡ ⎤ ⎧κ > 0 ⎪ κxx 0 0 ⎨ xx Vf α= , κ = ⎣ 0 κyy 0 ⎦, κyy > 0 ⎪ Vs + Vf 0 0 κzz ⎩ κzz > 0 where ρ is the density, ud is the Darcy velocity tensor, α is the porosity, p is the pressure, I is the unit tensor of second-order, K is the viscosity tensor, μ is the dynamic viscosity, κ is the permeability tensor, cF is the Forchheimer parameter, μT is the turbulent dynamic viscosity, k is the turbulent kinetic energy, ε is the dissipation rate of turbulent kinetic energy, Vf is the volume of fluid and Vs is the volume of solid. There are two unknown parameters in Eq. (1), one is the permeability κ, and the other is the Forchheimer parameter cF . Therefore, a 1D flow model of the channels is introduced to close the equation. Equation (1) is integrated along the flow direction:     u  d ∇· ρ ⊗ ud = α∇ · (−pI + K) − α μκ −1 + cF κ −1 ρ|ud | · ud α    ud  ud ⊗ = ∇ · (−pI + K) − μκ −1 + cF κ −1 ρ|ud | · ud →∇· ρ α α  out  out  u ud  d ⊗ dx = ∇· ρ ∇ · (−pI + K)dx α α in in  2   2    out  ρud ,x ρud ,x μ cF + √ ρ|ud | ud dx → − − 2 κ α α2 κ in out

in

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 cF + √ ρ|ud | ud dx κ κ

out  μ

 = pin − pout + Kout − Kin −

in

(2)

The physical properties, Darcy velocity, and Forchheimer parameters change very little in pure flow processes, so the assumptions of integral mean and weak compressible are introduced:  2  ⎧ 2  ρud ,x ρud ,x ⎪ ⎪ ⎪ − ≈0 ⎪ 2 ⎪ α α2 ⎪ ⎨ out

in

Kout − Kin ≈ 0, ⎪ ⎪     ⎪ ⎪ out  μ ⎪ cF cF μ ⎪ ⎩ + √ ρ|ud | ud dx ≈ + √ ρ|ud | ud x κ κ κ κ in    μ cF  + √ ρ ud ,x  ud ,x x → pin − pout = κ κ

(3)

Considering the Darcy-Weisbach formula: pin − pout

u u x ρ  dα,x  dα,x =f De 2

where f is the Darcy friction coefficient, De is the hydraulic diameter. Combining (3) and (4): u u  μ cF  f ρ  dα,x  dα,x  ud ,x + √ ρ ud ,x ud ,x = κ De 2 κ √ ⎧ κ f μ ⎪  ⎪ ⎨ cF = 2D α 2 − √  κρ ud ,x  e →   ⎪ ⎪ ⎩ f = f (Re, . . .), Re = ρDe  ud ,x  = ρDe |ux | μ α μ

(4)

(5)

where ux is the real velocity of free flow in the channels. Equation (5) can be substituted into Eq. (1) to calculate cF . Finally, the permeability κ needs to be determined. Since permeability is a geometric parameter, it should not change with the state of flow, so it is considered a constant and can be defined by Darcy’s flow law:       A cm2 3 −1 2 = κ cm p(MPa) qV cm · s μ(MPa · s)L(cm)  2      3 2  A 10−2 m   → qV 10−2 m · s−1 = κ 10−2 m p(Pa) μ(Pa · s)L 10−2 m       A m2 3 −1 2 = κ m p(Pa) → qV m · s μ(Pa · s)L(m) A → qV = κp (6) μL

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where qV is the volume flow rate, A is the global flow cross-sectional area of fluid and solid, and L is the flow distance. Rewrite Eq. (6) in the same form as the above equations: ud ,x A pin − pout → Af = κA μL α μx ud ,x Af → pin − pout = μx κα A

qV = κp

(7)

where Af is the cross-section area of the fluid. Moreover, in the structure of channels, Darcy friction coefficient and Darcy’s law at low Reynolds number are introduced. Taking the circular pipe as an example: u u ⎧ x ρ  dα,x  dα,x ⎪ ⎪ ⎪ ⎨ pin − pout = f D 2 x ud ,x e (8) → pin − pout = 32 2 μ 64 64 ⎪ De α ⎪ f = =   ⎪ ⎩ ρDe  ud ,x  Re μ

α

The expression of permeability κ can be obtained by combining Eqs. (7) and (8): ⎧ ud ,x Af ⎪ ⎪ μx ⎨ pin − pout = κα A x ud ,x ⎪ ⎪ ⎩ pin − pout = 32 2 μ (9) De α → 32

Af De2 ud ,x Af x ud ,x = μx → κ = μ 2 De α κα A A 32

For periodic flow channels, there is the following equation: α=

Af A

(10)

Substituting Eq. (9) into Eq. (1) to make it completely closed. In addition, the flow process in multi-channel of PCHE can be described by defining permeability in three directions. Finally, all the main equations are summarized as follows:    u ud  μ cF d ⊗ = ∇ · (−pI + K) − + √ ρ|ud | ud ∇· ρ α α κ κ √ ⎧ ρDe  ud ,x  f κ μ ⎪ ⎪   , f = f (Re, ...), Re = c = − √   ⎪ F ⎪ 2De α 2 μ α ⎪ κρ ud ,x  (11) ⎨ ⎡ ⎤ κxx 0 0 ⎪ Af De2 ⎪ ⎪ ⎣ 0 κyy 0 ⎦ , κ = κ = ⎪ ⎪ ⎩ A 32 0 0 κzz Along the flow direction, κij = κ, otherwise, κ → 0. Since zero cannot be directly divisible, a value much smaller than κ along the flow direction is adopted. In addition, for flow channels with different geometric cross-sections, an equivalent, flow-independent permeability can be obtained by changing the form of the Eq. (9).

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Equation (11) can be solved in several methods. In this paper, the first-order finite element was adopted, and streamline diffusion and crosswind diffusion are applied [6]. The partial Jacobi-Newton method is used to solve the problem. All the above processes are completed in software COMSOL. 1D in 3D Space Flow Model for Comparison The 1D momentum equation is established in 3D space. Consider the weakly compressible form: ∇p · e + f

ρ|u|u + ρg · e = 0 2De

(12)

where e is the vector in the direction of flow, and the Darcy resistance coefficient f can be calculated by Churchill relations [7]: ⎧   1   12 ⎪ ⎪ 8 12 37530 16 ⎪ −1.5 ⎪ f = 8 + + f , f = (f ) ⎪ A B A ⎨ Re Re     0.9   16 ⎪ ⎪ ⎪ 7 e ⎪ ⎪ f = −2.457 ln + 0.27 ⎩ B Re De

(13)

The model is used as the comparison in the 1D calculation. 2.2 Description of Calculation Case In this paper, only the flow characteristics were researched, so three cases of isothermal flow conditions were used for comparison. All the walls of porous media adopt a slip surface. The first case is a 2D shear flow with several channels, see Fig. 2. The 2D fields of velocity and pressure are calculated with the standard k-ε turbulence model and the anisotropic porous model respectively. The pressure drop and the distribution of mass flow rate are compared. The second case is an anisotropic flow in a plate of PCHE considering the real applications of FuSTAR. In one condition, FLiBe was used as the working medium, and FLiNaK was used as the working medium in the other. The main parameters are shown in Fig. 3 and geometries are shown in Fig. 4. Although the solid region is modeled, the calculations on heat transfer are turned off, and only the anisotropic flows of FLiNaK were calculated and compared. The results of laminar coupled 1D flow were used as the benchmark to compare the results of the porous media model. Above and below the geometry are periodic boundaries. The slip wall is used on the side of porous media, and the no-slip wall is used on the side of the free-flow region. The results of pressure drop and distribution of mass flow rate are compared.

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0.015 m/s H2 = 10mm

H1 = 20mm D = 4mm

D = 4mm

L = 40mm

Fig. 2. Sketch of geometry and parameters in the 2D case

Fig. 3. The main parameters of the two isothermal flow conditions

3 Results and Discussion 3.1 Comparison of 2D Shear Flow Case The pressure drop with different Reynolds numbers and inlet velocities are shown in Fig. 5. The pressure drop with the low Reynolds number is of a small order of magnitude and therefore presents a large relative deviation. But in general, the absolute deviation in pressure drop is very low, which proves that the porous media model is correct to calculate the pressure drop.

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Inlet fully developed Vin = 0.01-0.5 m/s

Outlet pressure = 0Pa

Fig. 4. The geometries of the 1D + 3D model and porous model Re 1600

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Fig. 5. Pressure drop with different Reynolds numbers and inlet velocities in the 2D case

Figure 6 shows the actual velocity field when inlet velocity Vin is 0.015 m/s. The porous media model can reflect the overall distribution of velocity at a macro level. Figure 7 shows the distribution of velocity at the middle cross-section of the channels. The shape of the distribution calculated in porous media is close to that of the NS method. With sliding walls, the velocity distribution calculated by the porous media model is still steeper than that calculated by the NS method, and the relative deviation is about 10%. In the future, it is necessary to introduce corrections or to consider the non-slip wall conditions.

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Fig. 6. Actual velocity field with inlet velocity Vin = 0.015 m/s. Normalized velocity 0.1778

0.1866

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Vin=0.015,porous Vin=0.015,free Vin=0.01,porous Vin=0.01,free Vin=0.006,porous Vin=0.006,free Vin=0.003,porous Vin=0.003,free Vin=0.0025,porous Vin=0.0025,free Vin=0.002,porous Vin=0.002,free Vin=0.0015,porous Vin=0.0015,free Vin=0.001,porous Vin=0.001,free Number of flow channels

Fig. 7. Distribution of velocity at the middle cross-section of the channels

3.2 Comparison of Anisotropic Flow in PCHE The distribution of the velocity field at an inlet velocity of 0.01 m/s is shown in Fig. 8. Overall, anisotropic permeability can simulate the flow state of a right-angle pipe very clearly. The Darcy velocity can be divided by the porosity to obtain the actual velocity. The remarkable advantage of porous media is that it can calculate at a similar scale no matter how many channels of PCHE. However, the calculation time of the 1D + 3D model increases significantly with the increase in the number of channels, making it very difficult to calculate thousands of them. Figure 9 shows the trend of pressure drop with inlet velocity or Reynolds number. With the increase of the Reynolds number, the predicted pressure drop of porous media is higher than that of the 1D pipeline flow model, and the deviation degree is larger. When Re exceeds 1000, the pressure drop deviation from 1% went up to 13%. This shows that the anisotropic model of the Brinkman equation coupled with the N-S equation, Darcy flow

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Fig. 8. Distribution of the velocity field at an inlet velocity of 0.01 m/s

theory, and Darcy Weissbach formula in this paper can only predict flow characteristics in the range of several hundred Reynolds numbers. Therefore, other corrections need to be introduced for high Reynolds number flows over 1000. This also indicates that at a high Reynolds number, the deviation caused by ignoring the viscosity force at the inlet and outlet of the control body will be amplified in Eq. (3). In the future, it can be considered that the viscous force is no longer solved in this model. Re 35000

0

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Fig. 9. Pressure drop with different Reynolds numbers and inlet velocities in the anisotropic flow case

Figure 11 shows the field of velocity across the central section of the PCHE, and the axis of the section is shown in Fig. 10, which also reflects the distribution of mass flow rate. The results show that with the increase of velocity and Reynolds number, the fields of the velocity in porous media and those of the 1D pipeline flow model become more uneven, which is mainly reflected at both ends of the plate, reflecting an obvious historical effect and is greatly influenced by upstream and downstream. In porous media, the velocity at both ends decreases as the flow distance increases, However, in 1D pipeline

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flow, the velocity of the edge pipe does not change much on the flow path. In addition, the uneven distribution of velocity may also cause an increase in the deviation of the pressure drop, so it should be considered to correct the distribution in the future.

Fig. 10. The axis of the section to calculate the field of velocity

2.5 Vin=0.5,porous Vin=0.5,1D+3D

Velocity (m/s)

2.0

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1.0

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Vin=0.1,porous Vin=0.1,1D+3D Vin=0.2,porous Vin=0.2,1D+3D

0.5

0.0 0.000

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0.010

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y (m) Fig. 11. Field of velocity across the central section of the PCHE

Fortunately, for the molten salt in FuSTAR, the Re is usually less than 100 to reduce pressure drop. Therefore, the model proposed in this paper can be directly used in the design of PCHE in FuSTAR.

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4 Conclusions To efficiently calculate the distribution of mass flow rate and flow heat transfer characteristics of large-scale channels in PCHE, an anisotropic porous media model is proposed in this paper. The model has coupled with the Brinkman equation, weakly compressible N-S equation, Darcy’s flow law, and Darcy Weissbach formula, which completely describe and close the model. A 2D multi-channel shear flow case and a 1D + 3D single-plate flow case in PCHE are used to verify the validity of the model. The results show that the porous model can replace the refined model well in the range of 10,000 Re in 2D flow without considering the bending structure, and similar distribution of mass flow rate and pressure drop can be obtained. However, the distribution of mass flow rate is steeper than the results of free flow. In the 1D + 3D flow case, the deviation of pressure drop will increase to more than 10% after Re exceeds 1000, which may be due to the neglect of the viscosity force, and the field of velocity also appears malformation. The porous media model in this paper can directly serve the design of heat exchangers in FuSTAR, but some modifications should be made in the future to offset the deviation of pressure drop and the distribution of mass flow rate to improve the universality. Acknowledgement. This research is supported by the National Key Research and Development Program of China (Grant No. 2020YFB1902000).

References 1. Jiang, D., et al.: Fluoride-salt-cooled high-temperature reactors: review of historical milestones, research status, challenges, and outlook. Renew. Sustain. Energy Rev. 161, 112345 (2022).https://doi.org/10.1016/j.rser.2022.112345 2. Liu, G., Huang, Y., Wang, J., Liu, R.: A review on the thermal-hydraulic performance and optimization of printed circuit heat exchangers for supercritical CO2 in advanced nuclear power systems. Renew. Sustain. Energy Rev. 133, 110290 (2020). https://doi.org/10.1016/j. rser.2020.110290 3. Ma, T., Li, L., Xu, X.-Y., Chen, Y.-T., Wang, Q.-W.: Study on local thermal–hydraulic performance and optimization of zigzag-type printed circuit heat exchanger at high temperature. Energy Convers. Manage. 104, 55–66 (2015). https://doi.org/10.1016/j.enconman.2015.03.016 4. Nield, D.A., Bejan, A.: Convection in Porous Media. New York, NY: Springer New York (2013). https://doi.org/10.1007/978-1-4614-5541-7 5. Panton, R.L.: Incompressible flow. J. Appl. Mech. 52(2), 500–501 (1985). https://doi.org/10. 1063/1.881530 6. Hughes, T.J.R., Mallet, M.: A new finite element formulation for computational fluid dynamics: III. The generalized streamline operator for multidimensional advective-diffusive systems. Comput. Methods Appl. Mech. Eng. 58(3), 305–328 (1986). https://doi.org/10.1016/00457825(86)90152-0 7. Churchill, S.W.: Friction-Factor Equation Spans all Fluid-Flow Regimes (1977) Accessed: Mar. 29, 2022. [Online]. Available: http://www.researchgate.net/publication/279898130_Fri ction-Factor_Equation_Spans_all_Fluid-Flow_Regimes

Study on the Prediction of Critical Heat Flux in Uniformly Heated Round Tube by Multilayer Perceptron Xiang Zou1 , Lei Lei1 , Ma GuoQiang1 , Jiao Feng1 , and Chen Shijun2(B) 1 Nuclear and Radiation Safety Center, MEE, Beijing, China

[email protected] 2 Suzhou Nuclear Power Research Institute Co., Ltd., Shenzhen, China

[email protected]

Abstract. The paper trains and tests over 200 multilayer perceptron (MLP) models with different input parameter conditions and structural characteristics, based on 7537 CHF experimental data, to predict critical heat flux (CHF) in uniformly heated round tube, and compares the best MLP with tradition CHF predict methods. The results show that the best MLP accurately predicts CHF in both numerical value and trend, and is better than traditional prediction methods in training data regions; the prediction accuracy of inlet condition is the highest and local condition is the lowest; MLPs which are deeper and have more neurons have a better prediction, but also prone to untrained; MLPs with decreasing structure can reduce the parameter scale while maintaining the prediction accuracy. Finally, the MLP used in the paper can accurately predict CHF in a large range of parameters, and the trend of different factors on prediction accuracy can provide a reference for the relevant study. Keywords: Multilayer perceptron (MLP) · Deep learning · Artificial neural network (ANN) · Critical heat flux (CHF)

1 Introduction Critical heat current density (CHF) is an important phenomenon in the reactor safety analysis. When the heat flux on the heating surface reaches CHF, there will be a significant degradation of heat transfer. If the heat flux continues to increase, the temperature will rise sharply, leading to the burning of fuel rods. CHF involves very complex phenomena, and it is difficult to propose analytical solutions. To predict the critical heat current density, a lot of studies and experiments have been carried out worldwide since the middle of the twentieth century, and more than a thousand empirical relations have been proposed, but these empirical relations only have good prediction accuracy in a small range. An artificial neural network (ANN) has strong predictive power for non-linear complex regression problems and can obtain good prediction results without focusing on the physical mechanism. Mazzola [1] trained a three-layer fully connected artificial neural network [2] using 1808 CHF experimental © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 59–72, 2023. https://doi.org/10.1007/978-981-19-8899-8_6

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data; Lang and Jianqiang [2] trained a three-layer fully connected BP network [3] with 6941 data points; Salman Zaferanlouei (2010) and others trained 5-layer ANFIS network using 513 CHF experimental data and CHF Look-up Table [4]. These studies successfully used the simple structure ANN to predict the CHF in a certain range of parameters and demonstrated that the neural network has the advantages of high accuracy, a wide range of parameter applicability, and ease to update. Usually, a complex network with more neurons and deeper hidden layers has better prediction ability but is also hard to train. This paper collects 10,172 experimental data of CHF on uniformly heated circular tubes in a larger range of parameters, and trains and tests over 200 fully connected multi-layer perceptron (MLP) with different input parameter conditions (inlet, outlet, and local), structural features (uniform, increasing, decreasing, and mixed), hidden layers and hidden layer neurons. The effect of these factors on the ability of MLP to predict CHF was systematically investigated, and MLP with good prediction accuracy over a large range. Finally, the MLP used in the paper can accurately predict CHF in a large range of parameters, and the trend of different factors on prediction accuracy can provide a reference for relevant studies.

2 MLP Model MLP is a type of artificial neural network, typically structured with an input layer, multiple hidden layers, and an output layer. The number of neurons in the input layer and the output layer is determined by the input parameters and the output parameters, and the number of hidden layers and the number of neurons in each layer are set according to the specific problems. The MLP of uniform, increasing, decreasing and mixed structures are defined as follows: (1) Uniform MLP, with an equal number of neurons in all hidden layers (Fig. 1); (2) Increasing MLP, the head of the uniform hidden layer adds multiple hidden layers, and the number of its neurons is doubled layer by layer (Fig. 2); (3) Decreasing MLP, multiple hidden layers are added at the tail of the uniform hidden layer, and the number of its neurons is halved layer by layer (Fig. 3); (4) Mixed MLP, a combination of (2) and (3), adding increasing and decreasing hidden layers to the head and tail of the uniform hidden layer (Fig. 4). Each layer is fully connected, and the parameter relationship is as follows:   yl = a Wl · yl−1 + bl

(1)

In the formula, l is the index of layers, yl is the output of the layer, W is the weight matrix, N l is the number of neurons in the layer, b is the bias vector, a is the nonlinear activation function. In this paper, the loss function L uses MSE, the activation function use tanh with zero mean, and the optimized loss function method use the Adam stochastic gradient

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Fig. 1. Structure of uniform MLP

Fig. 2. Structure of increasing MLP

descent method. LMSE =

1  ||ypredict,i − ygt,i ||2 2n

(2)

ex − e−x ex + e−x

(3)

i∈[0,n]

atanh =

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Fig. 3. Structure of decreasing MLP

Fig. 4. Structure of mixed MLP

3 Training Dataset 3.1 Experiment Data In previous studies, three types of input parameter conditions are usually used to predict CHF: (1) Inlet condition: CHF = f (P, D, L, G, Tsub ); (2) Outlet condition:CHF = f (P, D, G, x); (3) Local condition:CHF = f (P, D, L, G, x)。 P is pressure, D is diameter, L is heated length, G is mass flow, Tsub is subcooling temperature, x is vapor ratio. This paper collects 10,172 [5–7] uniformly heated round tube CHF experiment data suitable for the three types of input parameters (see Table 1 for the range of experiment

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parameters), which have been widely cited in regulatory and research institutions around the world. 3.2 Data Processing The data in this paper comes from different experiments. Due to experimental deviation, recording error, printing error, and other reasons, some wrong data need to be eliminated. The main screening criteria are: (1) The thermal balance equation is not satisfied; (2) The trend is significantly different from other test data obtained under similar conditions; (3) Measured under unstable flow conditions. After the screening, 7537 valid CHF experimental points were used as the CHF dataset of this paper. Since the experimental data range spans multiple orders of magnitude, to accelerate the loss function convergence of MLP, avoid gradient explosion and correspond to the activation function with zero mean, the data were normalized to [ −1, 1] using the following formula: xnorm = 2 ·

x − xmin −1 xmax − xmin

(6)

4 Training and Inference According to the inlet conditions, outlet conditions, and local conditions, the MLP with different structures, layers, and neurons were constructed, respectively. 75% of the CHF data set was selected as the training set to train these MLPs, and the remaining 25% was used as the test set to evaluate the prediction accuracy of the MLP. 4.1 Uniform MLP For uniform MLP with an equal number of neurons in the hidden layer, 210 models with a different number of layers and neurons are trained on the training set and predict CHF on the test set. Using the ratio of how many errors between the model prediction value and the experimental value are less than 20% as the model accuracy. The results are shown in Figs. 5, 6 and 7. Comparison of the accuracy between different MLPs shows: (1) As the number of layers and neurons in the hidden layer increases, the prediction ability of the model is enhanced; but the MLP prediction ability drops down when it is too large, which shows that the model is too complex and requires more data or better model structure to fully train. After the hidden layers are over 20, the model prediction accuracy is basically below 50%, this is due to the disappearance of the gradient in the backpropagation, so it is impossible to effectively train the model.

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Fig. 5. Prediction accuracy of uniform MLP of inlet condition

Fig. 6. Prediction accuracy of uniform MLP of outlet condition

(2) For the inlet condition MLP, when the number of neurons exceeds 256, the accuracy almost no longer increases, and both shallow and deep MLP can obtain good prediction accuracy. (3) For the outlet condition MLP, the influence trend is similar to the inlet condition MLP, and the best prediction ability is comparable to the inlet condition; When the number of neurons is small, the prediction ability is inferior to the inlet condition MLP, which shows that the impact on CHF of Tsub is more direct than x, so more neurons are needed to learn the relationship between x and CHF.

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Fig. 7. Prediction accuracy of uniform MLP of local condition

(4) For the local condition MLP, the accuracy of the model prediction is significantly worse than the inlet condition and outlet condition MLP due to the least input parameters. Meanwhile, it shows that the geometric parameters (length and diameter) of the tube have an obvious influence on the CHF, which also has been confirmed in previous studies. When the number of neurons is large enough, the accuracy of the deep model is significantly better than the shallow model, which indicates that the predicted CHF by the local condition is more challenging, and the prediction ability of the deep model declines significantly earlier. The uniform MLP structure parameters and prediction accuracy with the best prediction ability are shown in Table 2, and the error distribution is shown in Figs. 8, 9, 10. It shows that for the three input parameter condition types, MLP can predict CHF well within the parameters range of the CHF dataset. The traditional prediction method is shown in Table 3. By comparison, we can find that the accuracy of the traditional prediction is much lower than that of MLPs because the data has exceeded the applicable scope of the traditional formula. 4.2 Un-Uniform MLP As discussed in Sect. 3.1, MLPs have better prediction ability with more neurons and a deeper neural network, but the training is also more difficult and even untrainable. Using high accuracy and stable training with smaller size uniform MLP as the baseline model (see Table 4), build increasing, decreasing, and mixed MLPs, when the minimum hidden layer neurons equal the uniform layer neurons, the un-uniform model degenerates to uniform MLP. The same CHF training set and test set were used to train and test un-uniform MLPs, using the ratio of how many errors between the model prediction value and the

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Source

K. M. Becker

H. C. Kim

B. Thompson

D/mm

3.94–24.95

6–12

1.0–37.5

H/m

0.4–3.75

0.3–1.77

0.0254–3.6576

P/MPa

0.22–10.1

0.106–0.951

0.1–20.7

Tsub/K

30–240

11–154

0–337.8

G/ kg m−2 s−1

120–5450

20–277

9.9–18580.4

x

0–1.0

0.323–1.251

− 0.459–1.577

Number of data

5270

513

4389

Table 2. Structure of uniform MLP with best prediction accuracy Condition

Inlet

Outlet

Local

Layers

8

8

8

neurons

1024

1024

1024

Error < 10%

95.01%

89.33%

69.53%

Error < 20%

98.67%

96.71%

84.55%

Error < 50%

99.58%

99.36%

96.71%

RMS

7.93%

11.36%

29.79%

Avg. error

0.08%

− 0.72%

− 0.62%

Table 3. Prediction accuracy of tradition method Method

Bowring correlation (%)

Look-UP table 2006 [8] (%)

Error < 10%

27.76

29.35

Error < 20%

42.73

55.20

Error < 50%

92.99

88.48

RMS Avg. error

31.38

62.96

− 16.52

− 3.65

experimental value are less than 20% as the model accuracy, which is shown in Figs. 11, 12, 13. Comparison of the accuracy between different MLPs shows: (1) Decreasing MLP can improve the prediction ability of models, and reach a level comparable to the best uniform MLP (Table 3) because the smooth transition

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Fig. 8. Error distribution of uniform MLP of inlet condition with best prediction accuracy

between the uniform hidden layer and output layer is conducive to the transmission of feature information; And compared to the best uniform MLP, the number of weight matrix parameters for decreasing MLP decreased significantly, take the inlet condition MLP as an example, the number of best uniform MLP weight matrix parameters is about 7300000, the number of weight matrix parameters for the best decreasing MLP (minimum hidden layer neurons is 16) is about 370,000, comparable prediction accuracy was obtained with only a 5% parameter size. (2) The predictive power of the increasing MLP is significantly worse than the baseline MLP. It can be seen that the smooth transition between the input layer and the

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Fig. 9. Error distribution of uniform mlp of outlet condition with best prediction accuracy

hidden layer prevents the large-size hidden layer from extracting the features of the input parameters. (3) Because the mixed MLP is a combination of increasing MLP and decreasing MLP, its prediction ability is between both, and as the increasing hidden layer of the model head hinders the feature extraction, so the prediction ability is also inferior to the baseline model. (4) When the minimum number of hidden layer neurons is too small, the uniform MLP also cannot effectively train the model due to the deep number of layers.

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Fig. 10. Error distribution of uniform MLP of local condition with best prediction accuracy Table 4. Baseline model Condition

Layers

Neurons

Error < 20% (%)

Inlet

6

256

96.13

Outlet

6

256

92.62

Local

8

128

81.42

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Fig.11. Prediction accuracy of un-uniform MLP of inlet condition

Fig. 12. Prediction accuracy of un-uniform MLP of outlet condition

5 Conclusions This paper trained and tested 264 fully connected MLPs with different input parameter conditions, structural features, hidden layers, and neurons in each layer. By comparing the prediction accuracy of different MLPs and analyzing the parameter influence trend, the following conclusions can be obtained: (1) The MLP has a very good prediction accuracy within the parameter range covered by the training data; (2) The prediction accuracy of the inlet condition is the highest, and the outlet condition is the second, which indicates that the inlet condition has a more direct impact on CHF, and inlet condition parameters are better for MLP to predict CHF. The local

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Fig. 13. Prediction accuracy of un-uniform MLP of local condition

condition has the least information, it is hard to predict CHF only with them when the heated length effect is significant, so the prediction accuracy is the lowest; (3) The prediction ability of MLPs with deeper hidden layer and more neurons have a higher upper limit, but the training is also more difficult and even untrainable; (4) The MLPs with shallow hidden layer and a small number of neurons is easy to train, but it has insufficient prediction ability for complex problems; (5) Decreased MLP can improve the prediction ability of the MLP. The best prediction accuracy is comparable to that of the uniform MLP, but it can greatly reduce the parameter scale of the model, improve the training speed and save the computing resource; Finally, this paper obtains the MLP with good prediction accuracy, which can make an accurate prediction of CHF in a large range, and analyzing the best MLP structure parameters and the influence trend can provide an important reference for subsequent related studies.

References 1. Mazzola, A.: Integrating artificial neural networks and empirical correlations for the prediction of water subcooled critical heat flux. Rev. Gen. Therm. 36(11), 799–806 (1997) 2. Lang, H., Jianqiang, S.: Study on tube critical heat flux data treatment with artificial neural networks. At. Energy Sci. Technol. 39(1), 69–72 (2005) 3. Zaferanlouei, S., Rostamifard, D., Setayeshi, S.: Prediction of critical heat flux using ANFIS. Ann. Nucl. Energy 37(6), 813–821 (2010) 4. Groeneveld, D.C., et al.: The 2006 CHF look-up table. Nucl. Eng. Des. 237(15–17), 1909–1922 (2007) 5. Thompson, B., Macbeth, R.V.: Boiling Water Heat Transfer Burnout in Uniformly Heated Round Tubes: A Compilation of World Data With Accurate Correlations: AEEW-R-356[R]. UK: UKAEA (1967)

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6. Kim, H.C., Baek, W.-P., Chang, S.H.: Critical heat flux of water in vertical round tubes at low pressure and low flow conditions. Nucl. Eng. Des. 199(1–2), 49–73 (2000) 7. Becker, K., Hernborg, G., Bode, M., Eriksson, O.: Burnout Data for Flow of Boiling Water in Vertical Round Ducts, Annuli and Rod Clusters: AE-177. Aktiebolaget Atomenergi, Sweden (1965) 8. Tanase, A., Cheng, S.C., Groeneveld, D.C., Shan, J.Q.: Diameter effect on critical heat flux. Nucl. Eng. Des. 239(2), 289–294 (2009) 9. Goodfellow, I., Bengio, Y., Courville, A.: Deep Learning. MIT Press, US (2016)

Neutronics and Thermal-Hydraulics Coupling Analysis of Integral Inherently Safe Fluoride-Salt-Cooled High-Temperature Advanced Reactor - Fustar Xingguang Zhou, Dalin Zhang(B) , Xinyu Li, Xin Min, Wenqiang Wu, Lei Zhou, Wenxi Tian, and Suizheng Qiu Xi’an Jiaotong University, Xi’an, Shaanxi, China [email protected]

Abstract. The integral inherently safe fluoride salt cooled high-temperature advanced reactor – FuSTAR is designed by Xi’an Jiaotong University with the support of the National Key Research and Development Program of China. To capture the thermal-hydraulics feedback and neutronics response of the FuSTAR under the steady-state and near-critical conditions. A three-dimensional core neutronics and thermal-hydraulics coupling simulation and analysis for FuSTAR based on STAR-CCM+ and MCNP are carried out. High fidelity modeling is used to carry out core neutronics calculation, and the thermal non-equilibrium porous media is used to carry out the core thermal-hydraulics calculation. The coupling computation and data transmission are completed automatically through the user code in STAR-CCM+. The coupling analysis shows that the overall temperature rise of coolant in FuSTAR is 50K, and the maximum fuel temperature is 1041.2K, which is lower than the thermal limitation. The feedback coefficient of coolant is negative, yet the effect is not significant. The feedback coefficient of fuel is negative because of the Doppler effect, which improves the inherent safety of the reactor. This analysis can provide the calculation reference for FuSTAR design and other deterministic method codes coupling analysis. Keyword: Coupling analysis · Porous media · Thermal-hydraulics · Neutronics

1 Introduction The coupling analysis of multi-physics of nuclear reactors [1–3] is an important technology for comprehensive analysis of modern nuclear systems, especially the advanced GenIV nuclear reactors. In traditional reactor analysis, neutronics and thermal-hydraulics are calculated independently of each other. Therefore, the traditional analysis technology can not accurately capture the dynamic feedback characteristics of reactor thermalhydraulics and neutronics coupling. Hence, the results are conservative. And this will make the reactor engineering designers reserve a high safety margin for the design of the reactor, and then, may increase the core construction cost [4–6]. © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 73–87, 2023. https://doi.org/10.1007/978-981-19-8899-8_7

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In 2021, Xi’an Jiaotong University researched and designed an integral inherently safe fluoride salt cooled high-temperature advanced reactor – FuSTAR [7]. The thermal power of FuSTAR is 125MW, and the outlet temperature of the coolant is 973K, which can supply an amount of clean and efficient heat for the western region of China. A coupling analysis for FuSTAR is carried out by STAR-CCM+ and MCNP in this article, in which STAR-CCM+ completes thermal-hydraulics analysis with Computational Fluid Dynamics (CFD) in Finite Volume Method (FVM), and MCNP completes neutronics analysis in Monte-Carlo method.

2 Layout oF FuSTAR FuSTAR is a type of small modular fluoride salt cooled high-temperature reactor (SmFHR). Hence, it is convenient for modular assembly and transportation. FuSTAR has the characteristics of small volume, light weight, deep burnup, long life, and inherent safety. The core layout and dimensions of 1/6 FuSTAR as shown in Fig. 1. FuSTAR is a thermal-spectrum reactor, that uses graphite to moderate and reflect neutron for better economy. The height of the core is 340 cm, which contains the 20 cm height upper reflector, 20 cm height bottom reflector, and the 300 cm active region. The radius of FuSTAR is 147.21 cm, such a small radius is easily for transportation by an engineering truck. FuSTAR also uses two enrichment fuels, inner assemblies use the 15% enrichment fuel and the outer assemblies use the 17.5% enrichment fuel. This design can deepen the burnup and effectively flatten the radial power distribution of the core.

3 Coupling Scheme and Mathematics 3.1 The Coupling Scheme The coupling scheme of FuSTAR is shown in Fig. 2. At the beginning of coupling analysis, MCNP calculates the neutron transportation, that can obtain the fission power density and the neutron flux. The fission power density of each section in each assembly should be reconstructed by the total thermal power of FuSTAR by Eq. (1). PN ,i =

Pi 125 · K 6  Pk

(1)

k=1

where PN,i is the reconstructed fission power, Pi is the fission power calculated by MCNP. Thus, we can get the actual fission power when FuSTAR is in near-critical condition, which will be transferred to STAR-CCM+ as an energy source term to complete the thermal-hydraulics calculation. In thermal-hydraulics, the thermal non-equilibrium porous media model is used to simplify the model of core and reduce the calculation workload. The thermal nonequilibrium model can capture the different temperatures between porous media and

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(a). The layout and dimensions of FuSTAR in axial

(b). The layout and dimensions of FuSTAR in radial Fig. 1. The layout and dimensions of 1/6 FuSTAR

fluid, which means getting more precise results based on less calculation workload. This method is widely used by GeN-Foam [8–10], a well-known open-source coupling solver. When the thermal-hydraulics calculation is done, the convergence criterion will be judged. The convergence criterion is the fluctuation of k-eigenvalue (k eff ) less than 1e−3. If met the criterion, the coupling analysis will exit and output the results. If not, the data transmission will transfer the feedback of thermal-hydraulics to MCNP and start the new cycle of the coupling analysis until satisfying the convergence criterion.

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MCNP

Monte-Carlo Neutronics Transportation Neutron Flux Fission Power Reconstruction Data Transmisson

Feedback of ThermalHydraulics

Thermal Nonequilibrium Porous Media

Next Turn of Coupling

Feedback of Fuel, Coolant, cladding and Reflector Convergence Criteria NO ˂keff 3). When CSR perception is not included, there is no multicollinearity among variables,

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the regression coefficient is still significant to different degrees, and R2 is basically unchanged. As shown in the Table 7. Table 5. Correlation analysis results 1 1 CSR-legal

2

3

4

5

6

7

8

1

2 0.644** 1 CSR-economic & other 3 Trust

0.734** 0.732** 1

4 Knowledge

0.552** 0.246** 0.445** 1

5 Unknown risk perception

− − − − 1 0.426** 0.331** 0.385** 0.465**

6 Dread risk perception

− − − − 0.584** 1 ** ** ** ** 0.379 0.210 0.279 0.513

7 Benefit perception

0.618** 0.386** 0.564** 0.538** − − 1 0.372** 0.337**

8 Public acceptance

0.589** 0.316** 0.512** 0.567** − − 0.715** 1 0.312** 0.415**

M

1.526

2.069

1.916

1.988

3.704

3.936

1.724

1.855

SD

0.566

0.680

0.645

0.990

1.047

1.191

0.696

0.950

Note * p < 0.05, ** p < 0.01 Table 6. Regression analysis results 1(standardized) Coefficient

Dependent variable Trust

β t

0.807 19.786***

R2

0.652

F

391.479***

Knowledge

Unknown risk perception

Dread risk perception

0.425

− 0.412

− 0.316

6.781***

− 6.542*** − 4.820***

0.180

0.170

0.100

42.803***

23.234***

45.982***

Benefit perception

Public acceptance

0.541

0.485

9.308***

8.014***

0.293 86.646***

0.235 64.224***

Note * p < 0.05,** p < 0.01,*** p < 0.001,two-tailed.CSR Perception As Independent Variable

3.3 Mediating Effect Test According to the method introduced by Wen [9],with the process 3.3 plug-in of SPSS, use bootstrap method and repeat sampling 5000 times to test the variable mediation

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Variable

Public acceptance as dependent variable

Public acceptance as dependent variable

β

β

t

VIF

t

VIF

CSR perception

0.031

0.399

3.048

Trust

0.107

0.880

3.122

0.129

2.281*

1.581

0.174

2.851**

1.785

0.196

3.265**

1.785

0.133

2.282*

1.715

0.127

2.177*

1.695

1.721

− 0.177

1.769

0.524

Knowledge Unknown risk perception Dread risk perception

− 0.169

Benefit perception

0.496

R2

0.586

F

− 2.872** 8.154***

− 2.996** 8.839***

1.716 1.739

0.585

48.077***

57.898***

Note * p < 0.05,** p < 0.01,*** p < 0.001,two-tailed

effect. The results show that the direct effect is not significant, the mediating effects of trust and unknown risk perception are not significant, while the mediating effects of Knowledge, dread risk perception and benefit perception are significant. See Table 8 for details. Table 8. Bootstrap results of mediation effects Effect type

Effect value

Boot SE

Bootstrap 95%CI Lower limit

Total

0.814

0.102

0.614

Upper limit 1.015

Relative proportion (%) 100.00

Direct

0.053

0.132

− 0.208

0.313

6.47

Total indirect

0.762

0.143

0.486

1.045

93.53

Trust

0.145

0.112

− 0.076

0.372

17.77

Knowledge

0.140

0.057

0.043

0.264

17.19

− 0.090

0.051

− 0.199

0.001

− 11.05

Dread risk perception

0.093

0.042

0.022

0.186

11.43

Benefit perception

0.474

0.102

0.279

0.679

58.18

Unknown risk perception

Note CSR perception is the independent variable

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3.4 Results The Chinese public’s perception of nuclear power CSR includes two parts: legal responsibility, economic&other responsibilities. Among them, the perception of legal responsibility has a positive impact on public acceptance, and these impacts are partially realized through the mediating effect of Knowledge, dread risk perception and benefit perception, as shown in Fig. 1. 0.169*

Knowledge 0.552***

0.618***

CSR-Legal

0.644**

0.154*

Benefit Perception

-0.379*** CSR-Economic&other

0.489***

Public Acceptance

-0.168** Dread Risk Perception

Fig. 1. The influence path of CSR of nuclear power companies on public acceptance

4 Additional Experiment 4.1 Experiment Design The experiment adopts a subject design, selects the relevant contents of the 2018 and 2019 CSR reports of China Nuclear Power Co., LTD., and makes two videos introducing legal responsibilities, economic and other responsibilities. The participants were set to 3 groups, Group A watch economic&other responsibility video, Group B watch legal responsibility video, Group C watch no video. Then, the above-mentioned questionnaire was used to measure the participants’ perception of CSR, public acceptance and other variables. Finally, SPSS software was used to analyze data for intergroup comparison. See the experimental process in Fig. 2. 4.2 Experiment Results The details of the subjects are shown in Table 9. The results of ANOVA are shown in Table 10. The effect size of Knowledge variable was 0.002 (η2 p < 0.01), with no significant difference between groups. The effect size of unknown risk perception was 0.048 (η2 p < 0.06). The effect size of panic risk perception was 0.117 (η2 p < 0.14). The effect sizes of other variables were all greater than 0.14. Then pair comparison showed that group A was significantly different from group C

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Fig. 2. Supplementary experiment flow

in six variables including legal liability perception, economic and other responsibility perception, trust, panic risk perception, benefit perception and public acceptance. Compared with group C, group B showed significant differences in 7 variables including legal liability perception, economic and other responsibility perception, trust, unknown risk perception, panic risk perception, income perception and public acceptance. However, there were significant differences only in unknown risk perception and panic risk perception between group A and group B. Detailed comparison data between groups are shown in Fig. 3.

5 Discussion The public perception of the CSR of nuclear power companies in China is divided into two dimensions: legal responsibility, economic&other responsibilities, which is inconsistent with classical models. Carroll’s social performance model believes CSR includes four aspects: legal responsibility, economic responsibility, ethical responsibility and discretionary responsibility[10]. Transnational research in consumer domain shows that consumers believe that CSR includes three aspects: legal responsibility, ethical responsibility and discretionary responsibility[11]. The discrepancy not only shows that the content and boundary of CSR are constantly changing with social development, but also reflects the particularity of nuclear power companies’ social responsibility. The mean scores of CSR perception and public acceptance were similar to the research results of ZOU Shu-liang’s team [12], both lower than 2.5, indicating that no matter how the dimensions were differentiated, CSR perception and public acceptance of nuclear power companies were still at a low level.

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Table 9. Supplementary demographic information of subjects (N = 195) Attribute

Category

Number

Proportion

Sex

Male

103

52.80

Female

92

47.20

Under 20

4

2.10

21–30

36

18.50

31–40

100

51.30

41–50

43

22.10

Age

Job

51–60

8

4.10

Above 60

4

2.10

Government

60

30.80

Enterprise staff

26

13.30

Professional

56

28.70

Workers

19

9.70

Service/business

11

5.60

Students

11

5.60

Military

2

1.00

NGOs

1

0.50

Unemployed

3

1.50

Retired

6

3.10

Trust is significantly correlated with public acceptance (correlation coefficient 0.512), but the regression is not significant, indicating that trust may also affect public acceptance through other mediating variables. Subsequent studies can further investigate the relationship between trust and other variables. According to the experimental data, the publicity behavior of nuclear power companies can indeed improve the public’s perception of their social responsibility, and also improve the public’s acceptance of nuclear power. Moreover, the emphasis of publicity content does not affect the promotion effect. This result is a major incentive for nuclear power companies to continue to invest resources to fulfill and publicize their social responsibility. Comparing the data of this study with that of professor Zou Shuliang’s team, it is found that residents in Haiyan still have a low perception and acceptance of the CSR of nuclear power companies. This may be because previous popular science activities focused on scientific and technology knowledge, while there was little introduction of the CSR of nuclear power companies. Another possibility is that the objects of public communication are quite different from the sources of the subjects. According to the introduction of the wechat official account of Qinshan Nuclear Power Plant, the visitors it receives are mostly government staff, journalists, teachers and students, and

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DV CSR-legal

Contrast Error

CSR-economic & other

Contrast

Trust

Contrast

Knowledge

Contrast

Error Error Error

Sum of squares

DOF

Mean square

F

Significance

η2 p

273.341

2

136.671

281.245

0.000

0.746

175.869

0.000

0.647

173.117

0.000

0.643

0.171

0.843

0.002

4.810

0.009

0.048

12.744

0.000

0.117

157.741

0.000

0.622

24.333

0.000

0.202

93.302 177.849 97.081 172.571 95.697

192 2 192 2

0.506 86.286

192

0.498 0.063

70.546

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1.350

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288.400

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201.056

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Public acceptance

Contrast Error

88.924

2

0.126

Unknown risk perception

Error

0.486

259.296 38.285

60.481 238.613

2

2 192

19.142 1.502 100.528 0.637 30.241 1.243

Fig. 3. Comparison of public acceptance of nuclear power among groups

many residents around the new nuclear power plant, but it does not communicate much with local residents.

6 Conclusion As the public’s perception of CSR changes with social development and the particularity of the nuclear power industry, the Chinese public’s perception of CSR of nuclear power

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companies can be divided into legal responsibility, economic responsibility and other responsibilities. Among them, the perception of legal liability has a positive effect on the public acceptance of nuclear power through a complete intermediary effect, and the specific effect path is shown in Fig. 1. Although the Chinese public’s perception of the CSR of nuclear power companies is divided into two parts, the two parts have the same influence on the public acceptance of nuclear power in the actual publicity. However, in order to reduce risk perception, emphasis should be placed on promoting nuclear power companies to fulfill economic and other responsibilities.

Appendix The corresponding relationship between items and variables in the revised scale is shown in Table 11.

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Z. Xu and J. Ma Table 11. Corresponding items and variables adjusted

Variable

Item

CSR-legal

NPPs strictly abide by legals and carry out radioactive level monitoring and evaluation NPPs strictly abide by legals and invest sufficient resources to ensure nuclear safety The managers of NPPs try to comply with the legal NPPs seek to comply with all legals regulating hiring and employee benefits NPPs are committed to energy conservation, emission reduction and environmental protection NPPs provide complete and accurate radioactive monitoring data to the public NPPs have a comprehensive code of conduct NPPs are regarded as trustworthy

CSR-economic & other

NPPs have flexible policies enable employees to better coordinate work and personal life NPPs closely monitor employees’ productivity NPPs strive to lower operating costs NPPs supports employees who acquire additional education NPPs have been successful at maximizing profits NPPs top management establish long-term strategies NPPs have programs that encourage the diversity of our workforce(in terms of age, gender, and race) NPPs have internal policies prevent discrimination in employees’ compensation and promotion

Trust

NPPs and I have a sharing relationship I see no reason to doubt Nuclear power staff competence and preparation for the job I can talk freely to this individual about difficulties I am having at work and know that NPPs will want to listen If I shared my problems with NPPs, I know they would respond constructively and caringly Most people trust and respect NPPs I can rely on NPPs not to make my job more difficult by careless work (continued)

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Table 11. (continued) Variable

Item Other work associates of mine who must interact with NPPs consider them to be trustworthy I would have to say that NPPs and I have both made considerable emotional investments in our relationship NPPs staff approach their jobs with professionalism and dedication I would feel a sense of loss if NPPs transferred

Knowledge

I do not feel very knowledgeable about Nuclear Power.* Compared to most other people, I know less about Nuclear Power.* When it comes to Nuclear Power, I really don’t know a lot.* Among my circle of friends, I’m one of the “experts” on Nuclear Power I know quite a lot about Nuclear Power

Unknown risk perception

I think the potential risks of NPPs are Not known to exposed.* I think the potential risks of NPPs are Not known to science.* I think the potential risks of NPPs are Delayed. * I think the potential risks of NPPs are New.* I think the potential risks of NPPs are Not Equitable.*

Dread risk perception

I think the potential risks of NPPs are Global Catastrophic.* I think the potential risks of NPPs are Dread.* I think the potential risks of NPPs are Uncontrollable.* I think the potential risks of NPPs are Consequences fatal.*

Benefit perception

Building NPPs can boost the local economy Building NPPs will increase employment opportunities for local residents In the long run, nuclear power will ensure a stable and reliable energy supply Using nuclear power can mitigate climate change

Public acceptance

I support the construction of NPPs in China I support the construction of NPPs in our province I support the construction of NPPs in our city I support the construction of NPPs in our county

Note* Reverse scoring

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References 1. Tian, Z., Wang, R., Yang, W.: Consumer responses to corporate social responsibility (CSR) in China. J Bus Ethics 101(2), 197–212 (2011) 2. Deng, X., Zhang, T., Xu, Y., Long, X.: A study of the influence of the CSR on consumers’ purchase intention. Chin. J. Manage. 13(07), 1019–1027 (2016) 3. Maignan, I.: Consumers’ perceptions of corporate social responsibilities: a cross-cultural comparison. J. Bus. Ethics 30(1), 57–72 (2001) 4. Ding, Z., Ng, F.: Reliability and validity of the Chinese version of McAllister’s trust scale. Constr. Manag. Econ. 25(11), 1107–1117 (2007) 5. van Rijnsoever, F.J., van Mossel, A.: Public acceptance of energy technologies: the effects of labeling, time, and heterogeneity in a discrete choice experiment. Renew. Sustain. Energy Rev. 45, 817–829 (2015) 6. Slovic, P.: Perception of risk. Science 236(4799), 280–285 (1987) 7. Visschers, V.H.M.: Find the differences and the similarities: relating perceived benefits, perceived costs and protected values to acceptance of five energy technologies. J. Environ. Psychol. 40, 117–130 (2014) 8. Roh, S., Lee, J.W.: Differentiated influences of risk perceptions on nuclear power acceptance according to acceptance targets: evidence from Korea. Nucl. Eng. Technol. 49(5), 1090–1094 (2017) 9. Wen, Z., Ye, B.: Analyses of mediating effects: the development of methods and models. Adv. Psychol. Sci. 22(05), 731–745 (2014) 10. Carroll, A.B.: A three-dimensional conceptual model of corporate performance. Acad. Manag. Rev. 4(4), 497–505 (1979) 11. Maignan, I., Ferrell, O.C.: Nature of corporate responsibilities: perspectives from American, French, and German consumers. J. Bus. Res. 56(1), 55–67 (2003) 12. Zou, S., Teng, F., Yu, X.: Research on definition of social responsibility of nuclear power corporate and analysis of measuring dimension. J. Univ. South China (Soc. Sci. Ed.) 16(02), 1–5 (2015)

Local Correlation and NpNn Linearity of Nuclear Electric Moments Y. Xiao1 , D. Liu1 , Z. Z. Qin1 , and Y. Lei2(B) 1 School of Science, Southwest University of Science and Technology, Mianyang 621900, China 2 School of National Defense Science and Technology, Southwest University of Science and

Technology, Mianyang 621900, China [email protected]

Abstract. In this work, we suggest that, as another important observable strongly related to nuclear structure evolution, the electric moment of the first 2+ state, denoted by Q here, is also governed by a similar local correlation of adjacent nuclei according to the Hartree-Fock method and anharmonic vibrator model. We made a survey of currently available experimental data of the Q values across the whole nuclear landscape, which confirms this local correlation of the Q values. On the other hand, it also has been long recognized that Q can be organized in the NpNn scheme. Thus, we believe such an NpNn plot of Q values should also be constrained by their local correlation, which would present linearity in a logarithmic scale of NpNn. We verified such linearity with available experimental data, and some exceptions are brought out, and discussed. In a word, the local correlation of Q values out of theoretical models leads to a linear evaluation against logarithmically scaled NpNn. We verified our finding with experimental data, and believe it may serve further experimental analysis and theoretical prediction of Q values. Keywords: Electric moment · Local correlation · NpNn scheme · Nuclear structural evolution · Nuclear systematics

1 Introduction The evolution of nuclear structure in the main shell is mainly driven by proton-neutron interaction (p-n interaction) [1–3]. If the coupling of each pair of protons and neutrons has a similar interaction, the interaction of the remaining protons and neutrons in the nucleus should be roughly proportional to the product of NpNn. Here, Np and Nn refer to the number of valence protons and neutrons respectively. Therefore, the NpNn product has been widely used in the study of nuclear structures [4–9]. On the other hand, the regional evolution law of the first 21 + state energy, E2, and its reduced transition probability of downward transition, B(E2), or even and even nuclei also reflects the structural evolution characteristics of atomic nuclei. Therefore, these observable values should show a systematic law under the organization of NpNn product parameters. Inspired by this, Casten proposed a systematic research model of NpNn, corresponding these observable values to the NpNn product of heavy nuclei, observing its smooth evolution law, © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 111–120, 2023. https://doi.org/10.1007/978-981-19-8899-8_10

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and exploring its abnormal evolution factors and their corresponding micro- mechanisms [10–12]. Although the systematic study of NpNn has achieved some success, the specific correlation between the valence nucleon number and the proton-neutron residual interaction has not been quantified, and the relationship between the proton-neutron residual interaction and various nuclear observables is only qualitatively described, so it is difficult to quantitatively and analytically connect the NpNn product with nuclear observables. To solve this problem, predecessors tried to fit the systematic law of NpNn through preset analytical formulas, to serve the experimental prediction. Long before NpNn was systematically proposed, Hamamoto found that the experimental data of B(E2) was roughly proportional to NpNn [9], and further proposed that Q may also meet this proportional relationship. Casten, Sun Xiaojun, etc. tried to connect the nuclear observability with the exponential function of the NpNn product [7, 8] and achieved some quantitative understanding. Recently, we have noticed that some nuclear observables, such as E2 and B(E2), in addition to obeying the NpNn systematic law, also satisfy some local systematic correlations: the physical quantities of adjacent nuclei are connected by some simple algebraic relations [13, 14], and the micro differentiation of these local relations can constrain the NpNn systematic law and obtain its analytical expression [15]. These local correlations have been used to predict unknown nuclear data, such as E2 and B(E2) [14, 16], nuclear mass [17–21], single nucleon separation energy [22], nuclear charge radius [23], and so on. Compared with NpNn systematicness, the theoretical basis of nuclear structure of local correlation may be clearer. This is because of the localization of the nuclear structure model: if two nuclei are far away from each other on the nuclide diagram, their theoretical model parameters should be very different, and even if there are obvious differences in the model framework, so it is impossible to directly give the NpNn systematic law in a large shell through a unified nuclear model. However, the model description of adjacent nuclei is convergent, so it is possible to extract the local systematic relationship of nuclear observability from such a model. For example, the local relationship between E2 and B(E2) is derived from the Hartree-Fock model [24]. Therefore, if the local correlation of atomic nuclei can give an analytical expression of NpNn systematic law, it is equivalent to establishing a certain connection between NpNn systematicness and the atomic nucleus model, which is conducive to the subsequent theoretical development and experimental data study. Hamamoto once pointed out that electric moment Q, similar to E2 and B(E2), is an important characterization of nuclear quadrupole deformation, which should be systematically described by the NpNn product [9]. If the electric moment also has a local system similar to E2 and B(E2), then the analytical expression of the systematic law of electric moment NpNn can be further given. However, compared with E2 and B(E2), the experimental data on the electric moment are relatively scarce, and the discussion of NpNn systematic evolution law and local correlation of electric moment is relatively rare. With the increasing abundance of electric moment experimental data, we should try to carry out this work. Therefore, this paper will first build the local correlation of electric moment through the theoretical model and use the experimental data to verify

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these local correlations. Follow the logic in literature [15]: if the local systematic relationship of the electric moment is similar to E2 and B(E2), it should also be linear in the NpNn plot of logarithmic coordinates. This paper will further use experimental data to verify whether the electric moment also has this logarithmic NpNn linear relationship, analyze the abnormal factors of exceptional data, and provide a new way for the analysis and prediction of experimental data of electric moment.

2 Local Correlation of Electric Moments This section will deduce the possible local relationship of Q value from the theoretical model. For different types of nuclei, different phenomenological models of nuclear structure are used. The Q value of rotating nuclei with axisymmetric deformation is proportional to the intrinsic electric moment [25] Q=−

  2 2  φ0 |Q|φ0 = − i|Q|i 7 7 Z





(1)

i=1

|φ 0  is the Hartree-Fock ground state constructed by the deformed single particle state |i, and Q is the electric operator. Assuming that adjacent nuclei have similar deformed proton single-particle states, Q(Z + 2, N + 2) + Q(Z, N ) − Q(Z + 2, N ) − Q(Z, N + 2) Z+2 Z Z+2 Z   2   i|Q|i = 0 =− 7 

(2)

i=1 i=1 i=1 i=1

The non-zero Q value of vibrational nuclei comes from the configuration mixing of phonon states [26]. To simplify the derivation, the derived model space contains only one phonon state and two phonon states, which are represented by |1 and |2 respectively. The corresponding Hamiltonian matrix is given by the following formula  ω λ H= (3) λ 2ω where ω is the phonon excitation energy; λ is the phonon configuration mixing energy. 21 + state corresponds to low excitation configuration mixing of |1 and |2

+

2 = α|1 + β|2 (4) 1 where, α and β are real numbers, corresponding to the proportion of the two phonon states in the 21 + state. The phonon state mixing of typical vibrational nuclei is not strong, i.e., λ ≪ ω, making βλ/ω, and α1. In this way, the Q value of 21 + state can be calculated by the following formula [27]   8χ λ 2π   + + ˜ 1||b||2 (5) Q = 21 |Q|21 = 5ω 7 

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where Q = χ (b† + b*) is the E2 electromagnetic transition operator; χ is the intensity parameter of the Q operator; b† and b* are phonon generation operators and time inversion operators, respectively. ω and 1||b||2 matrix elements can be regarded as smooth change functions of Z and N. In adjacent nuclei, they should remain approximately unchanged. The configuration mixing energy λ can be attributed to p-n interaction [1]. Therefore, the Q value of the vibrational kernel should be a smooth function of the NpNn product. This is in line with Hamamoto’s assumption of the systematic law of electric moment NpNn [9]. Vibrational nuclei generally have little deformation, fewer valence nucleons, and smaller NpNn. Therefore, Q can be expanded by Taylor around NpNn = 0: Q(Np , Nn ) = Q|Np Nn =0 + Np Nn

dQ + o2 d (Np Nn )

(6)

Then the Q value has a local second-order difference, such as δQ = Q(Np + 2, Nn + 2) + Q(Np , Nn ) − Q(Np + 2, Nn ) − Q(Np , Nn + 2) dQ 4 d (Np Nn )

(7)

Note that dQ/d(NpNn) is undetermined here, and the local correlation strictly similar to formula 2 is not obtained. However, considering that when NpNn = 0, there is no p-n residual interaction in the nuclear system, which corresponds to no mixed phonon excitation at λ = 0. According to Formula 5, Q≡0. Then the average Q value of the four adjacent nuclei should be‾Q = (Np + 1)(Nn + 1)dQ/d(NpNn). For non-magic vibrational nuclei, Np2, Nn2. Therefore, there is‾Q  δQ. In other words, even in vibrational nuclei, the value of the local second-order difference δQ relative to the atomic nuclear moment itself is still very small. Considering the uncertainty of the experimental data, it can be considered that δQ(Z + 1, N + 1) = Q(Z + 2, N + 2) + Q(Z, N ) − Q(Z + 2, N ) − Q(Z, N + 2)  0

(8)

Therefore, whether it is a rotating nucleus or a vibrating nucleus, theoretically, the electric moment of the atomic nucleus should meet Formula 8. The reliability of Formula 8 can be verified by using the experimental data of electric moment [28]. Figure 1 (a) shows the Q values available for all experiments. Figure 1 (b) shows the value of δQ calculated by Formula 8. It can be seen that the values of δQ in all nuclear regions are close to zero within the experimental error range. It has the same reliability in the light nuclear region and the heavy nuclear region. We also note that E2 and B(E2) have similar relative proportional stability [24], indicating that this local correlation may be a general feature of the collective nature of the quadrupole of the nucleus.

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Fig. 1. The relationship between Q, δQ and its relative proportion and mass number. (a) Q value comes from the literature [28]; (b) δQ value is calculated by formula 8. The error comes from the experimental error transfer of four adjacent nuclei. The dotted line represents zero value.

3 Linear Evolution of Electric Moment in NpNn Logarithmic Coordinates Like E2 and B(E2), if Q not only has a regular evolution but also meets the local systematic correlation in Formula 8, then Q should evolve linearly in logarithmic NpNn coordinates. The specific derivation is as follows: firstly, the local relationship of Q is slightly differentiated into sections that will deduce ∂ 2Q 0 ∂Np ∂Nn

(9)

Since the electric moment, Q is a function of the NpNn product, the above partial derivative equation can be transformed into a second-order ordinary differential equation with NpNn as the independent variable d 2Q dQ ∂ 2Q + Np Nn = 0 ∂Np ∂Nn d (Np Nn ) d (Np Nn )2

(10)

The general solution of the above ordinary differential equation is Q = c1 ln(Np Nn ) + c2

(11)

where c1 and c2 are arbitrary constants. This is the linear evolution law of electric moment in NpNn logarithmic coordinates.

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Table 1. The nucleon numbers of the three nuclear regions involved in the NpNn systematic study in this paper. The nuclear region is divided according to the main shell separated by three magic numbers 50, 82, and 126. Z

N

A130

50–66

66–82

A150

50–66

82–104

A190

66–82

104–126

To verify this linear correlation, it is necessary to plot the evolution of electric moment experimental data with ln(NpNn) in different nuclear regions. In the light nuclear region, because the nuclear shell is small, the variation range of the NpNn value is small, and the systematicness of various observable NpNn in atomic nuclei is not obvious. As for the electric moment of atomic nuclear power, although it has a universal local relationship, its linear correlation in NpNn logarithmic coordinates is not obvious. Therefore, this section only analyzes the logarithmic NpNn linear behavior of the electric moment experimental data [28] in the three heavy core regions (A 130, A150, and A190). These three heavy nuclear regions are separated by magic numbers 50, 82, and 126. The specific nucleon number range is shown in Table 1. General characteristics of quadrupole collective nature. Table 2. Experimental measurement of Ba isotope electric moment. Data are from the literature [28]. The data is divided into two columns: “satisfying linearity” and “not satisfying linearity”, which correspond to the experimental data of Ba electric moment satisfying logarithmic NpNn linearity and the experimental data not satisfying linearity in Fig. 2 (a). Satisfying linearity(eb)

Not satisfying linearity(eb)

130 Ba

− 1.0(2)

− 0.1(2)

134 Ba

− 0.32(6)

+ 0.09(6); − 0.20(6); + 0.21(6)

136 Ba

− 0.19(6)

+ 0.07(7)

Figure 2 shows the evolution law of electric moment with ln(NpNn) in the above nuclear region. It can be seen that the Q values of the three core regions described in Fig. 2 (especially the A150 core region) not only have the NpNn systematic law but also show an obvious logarithmic NpNn linear correlation. There are also specific reasons for the deviation of some data from linearity in the other two nuclear regions, as detailed below. In the nuclear region of A130, as shown in Fig. 2 (a), some experimental measurements of Ba isotope Q do not seem to meet the NpNn logarithmic linear correlation. This deviation mainly comes from the uncertainty of experimental measurement. Since the nuclei of the A130 nuclear region are located in the transition region from the

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near-spherical nucleus to the deformed nucleus [29, 30], the same experimental measurement can give many different values of electric moments [31, 32]. Some of the measured values meet the logarithmic NpNn linear relationship, while others deviate. Previous experiments cannot rule out this data divergence a priori. However, these divergent experimental measurement results are listed in detail in Table 2, and they are divided into "satisfying linearity" and "not satisfying linearity", which correspond to the experimental data satisfying linearity and the experimental data not satisfying logarithmic NpNn linearity in Fig. 2 (a) respectively. In Fig. 2 (a), data points that “not satisfying linearity" is further marked with blue dots. After excluding these Q values “not satisfying linearity", the regularity of the experimental data is more obvious. We believe that the "linear" Q experimental measurement is more consistent with the systematic law of NpNn and more reliable. Therefore, the logarithmic NpNn linear law of Q can provide a new criterion for the reliability evaluation of electric moment measurement results to a certain extent. In the A190 nuclear region, except for a few Pt isotopes (Z = 78) and 204 Hg, the overall experimental data show a linear trend. These electric data points of Pt and 204 Hg are also highlighted with blue dots in Fig. 2 (c). The abnormal behavior of Pt isotopes in NpNn systematics around A190 has been noticed [33] and attributed to their nuclear shape complexity [34]. The deformation of these Pt isotopes includes soft triaxial deformation, oblate ellipse, and long ellipse. The intrinsic electric moments of different deformations are quite different, resulting in a local deviation from NpNn systematicness. On the other hand, the Q value of 204 Hg is very small and does not conform to linearity. This is because 204 Hg is very close to the magic number, and its Q value itself should be a small quantity. For vibrational nuclei, when the NpNn product approaches zero, the Q value should also approach zero. Therefore, in Fig. 2 (c), we draw a dotted line connecting the main endpoint and the point (NpNn0, Q = 0), which schematically reflects the evolution trend of the Q value at NpNn → 0. The 204 Hg data point is right on this dotted line. Therefore, we believe that the evolution of the electric moment in the A190 core region may correspond to a piecewise linear function, which is similar to the linear evolution of B(E2) and E2 [15]. Because the Q experimental data of A190 nuclear region at NpNn → 0 is not rich enough, this piecewise linear evolution trend at NpNn → 0 still needs to be verified by experiments.

4 Conclusions To sum up, we obtain Q local systematization based on the axisymmetric rotor model and the non-harmonic oscillator model. If Q also satisfies the systematic law of NpNn, then Q should show a linear correlation feature in the logarithmic coordinate NpNn scheme. We use experimental data to verify the local correlation of Q value and its linear evolution in NpNn logarithmic coordinates. With a few interpretable exceptions, local correlation and logarithmic NpNn linear laws are generally established. The linear law of Q in NpNn logarithmic coordinates solves the problem of uncertainty in the experimental measurement of Ba isotope Q value to some extent. This puts forward new requirements and expectations for the experimental measurement of the electric moment in this nuclear region. In a word, the local systematicness of the electric

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Fig. 2. The Evolution of the electric moment with NpNn logarithm in A130, 150, and 190 nuclear regions. The experimental data are from the literature [28]. The red line in the figure schematically highlights the linear characteristics of the observability of the three nuclei. Some data deviating from linear evolution are highlighted in blue. In particular, the Ba isotope in figure (a) deviates from the uncertainty of the experimental data (see Table 2). Figure (c) also draws a dotted line connecting the endpoint of the main line and the point (NpNn0, Q = 0). See the text for a specific explanation.

moment and its linear law in logarithmic NpNn coordinates provide a clearer vision for the study of nuclear structure evolution, which may be helpful to the verification and prediction of the experimental data of the electric moment. Acknowledgments. We thank the National Natural Science Foundation of China for supporting this work under Grant No. 12105234. The work was also partially supported by the Sichuan Science and Technology Program under Grant No. 2019JDRC0017. We also acknowledge the support of Southwest University of Science and Technology under the Doctoral Program (Grant No. 18ZX7147), College Education Research Program (Grant No. 19GJZX19), and Online Course Program (Grant No. 22ZXKC18).

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29. Qiang, Y.H., Petrache, C.M., Guo, S.: Identification of high-K rotation in 130Ba: testing the consistency of electromagnetic observables. Phys. Rev. C 99, 014307 (2019) 30. Guo, S., Zhou, H.B., Petrache, C.M.: Collective motion in 130Ba. Nucl. Phys. Rev. 37(3), 530–535 (2020) 31. Rothschild, P.J., Baxter, A.M., Burnett, S.M.: Quadrupole moment of the first excited state of 136Ba. Phys. Rev. C 34, 732 (1986) 32. Burnett, S., Baxter, A., Gyapong, G.: Electric quadrupole moments of the first excited states of 130, 134, 138Ba. Nucl. Phys. A 494(1), 102 (1989) 33. Zhao, Y.M., Casten, R.F., Arima, A.: Generalization of the NpNn scheme and the structure of the valence space. Phys. Rev. Lett. 85, 720 (2020) 34. Gyapong, G., Spear, R., Esat, M.: Electric quadrupole moments of the first excited states of 194Pt, 196Pt, and 198Pt. Nucl. Phys. A 458(1), 165 (1986)

Research on Financial Risk Analysis of Nuclear Power Project Based on Monte Carlo Simulation Chunhua Lu, Xiaoyuan Lin(B) , and Ruomin Zhang Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai, China {luchunhua,linxiaoyuan,zhangruomin}@snerdi.com.cn

Abstract. Taking the economic evaluation method of nuclear power plant construction projects as the basic model, the profitability index commonly used in financial analysis as economic evaluation index, the typical probability distribution of engineering construction projects as the probability distribution of risk factors, the Monte Carlo simulation method is used to calculate the economic evaluation index. The probability distribution of economic evaluation index and the of risk factors are closely related through the financial analysis method of nuclear power plant construction projects. Through the cumulative probability distribution of economic evaluation index and the risk tolerance of investors, the project economic evaluation risk indicators are determined, to provide more comprehensive support for nuclear power project investment decisions. Keywords: Nuclear power project · Financial risk analysis · Monte Carlo simulation

1 Introduction With the proposal of the carbon peaking and carbon neutrality goals, our country requires the orderly development of nuclear power under the premise of strict supervision and absolute safety. Therefore, nuclear power is an important option for our country to improve its energy structure and achieve carbon peaking and carbon neutrality goals. Nuclear power projects have high safety requirements, large investment scale, and long operation cycles. In order to make project approval decisions more scientific and rational, to reduce insufficient awareness among decision makers of changes in project environment and improper response leads to large deviations between project economics and actual operation, we must pay attention to and strengthen the financial risk analysis of nuclear power construction projects [1]. The economic evaluation guidelines of the nuclear power plant construction project (NB/T 20048-2011) on the risk analysis method standard proposes that the expert investigation method, the analytic hierarchy process, etc. can be used [2], but there are no specific methods and steps to guide the risk analysis. Generally speaking, most of the data used in the economic evaluation of nuclear power projects come from forecasting

© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 121–126, 2023. https://doi.org/10.1007/978-981-19-8899-8_11

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and estimation, which has a certain degree of uncertainty. In order to analyze the impact of changes in the main data used on the evaluation indicators, and to estimate the risks that the project may take, risk assessment can be carried out. Thus, the early warning of project risk and corresponding countermeasures are put forward to provide reference for investment decision-making.

2 Introduction to Monte Carlo Simulation The basic idea of Monte Carlo simulation is to solve problems in mathematics, physics, engineering technology and production management. First build a probabilistic model or stochastic process with parameters equal to the solution of the problem; then calculate the statistical characteristics of the desired parameters by sampling experiments (simulations) of the model or process; finally, the required approximate values are given, and the accuracy of the solution can be expressed by the standard error of the estimated value or other statistical characteristics. Specifically, it can be expressed as follows: According to actual engineering or management problems, a probability model g(x1 , x2 , . . . , xn ) is established with the target value, and random variables (x1 , x2 , . . . , xn ) are settled by sampling technique, substitute it into the probability model g(x1 , x2 , . . . , xn ) to get a random target value. In the same way, N similar random target values are generated. Combined with the law of large numbers, the mean value, variance and other statistical parameters of these N random target values can be calculated, to obtain an approximate solution to the target value of practical project or management problems [3]. Therefore, the general steps of Monte Carlo simulation modeling are: (1) Data collection and statistical test. Collect the factors related to the target value, namely {x1 , x2 , . . . , xn }, and analyze them, estimate their distribution functions and distribution parameters, and make statistical test. For the determination of the distribution function of some important factors and its parameters. (2) Create target problem. First of all, it is necessary to clarify the essence of the problem to be solved, combine the environment associated with the analysis and calculation to abstract the target problem into a probability model g (x1 , x2 , . . . , xn ). Then draw the calculation flow chart and write a calculation method that the computer can accept. (3) According to the accuracy requirements, determine the number of simulations N. (4) Sampling of random variables. According to the distribution type and parameters of the determined random variables {x1 , x2 , . . . , xn }, a group of variables (x1 , x2 , . . . , xn ) corresponding to the random variables {x1 , x2 , . . . , xn } are randomly selected. (5) The objective function value is calculated according to the probability model established in step (2), that is, a sample value of a random event is obtained. (6) Repeat (4) and (5) N times to obtain N random sample values. (7) Statistical analysis is performed on N sample values to obtain a distribution curve and test its probability distribution and other statistical characteristics.

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3 Financial Risk Analysis Model of Nuclear Power Project Based on Monte Carlo Simulation Taking the financial evaluation method of nuclear power plant construction projects as the basic model, the profitability index commonly used in financial analysis is used as the economic evaluation index. The typical probability distribution of engineering construction projects is selected as the probability distribution of risk factors, and the Monte Carlo simulation method is used to calculate the economic evaluation index. The main steps are: (1) Determine economic evaluation indicators, including project financial internal rate of return, financial net present value, payback period, etc. (2) Determine the risk factors corresponding to the economic evaluation indicators, including one or more of construction investment, load factor, electricity sales price, and nuclear fuel price. (3) Setting probability distributions for the risk factors, including triangular distribution, normal distribution, etc. (4) The random sampling method of Monte Carlo simulation is used to extract the values of a group of risk factors from the probability distribution of the risk factors. (5) Calculate the economic evaluation index result based on the value of the risk factor extracted in step (4). (6) Repeat steps (4) and (5) until the present sampling times are reached. (7) According to all the economic evaluation index results calculated in step (5), draw the probability distribution and cumulative probability distribution diagram of the economic evaluation index results. (8) According to risk tolerance, set economic evaluation indicators (Fig. 1).

4 Case Analysis Taking one nuclear power project as a case to carry out financial risk analysis, first determine the financial internal rate of return of project investment as an economic evaluation index. According to the sensitivity analysis chapter of The economic evaluation guidelines of the nuclear power plant construction project, the sensitive factors include construction investment, load factor, electricity sales price and nuclear fuel price. Through the financial analysis of actual data of multiple nuclear power projects, construction investment, load factor, the electricity sales price has a greater impact on the rate of return, so take the overnight cost per kilowatt, the load factor and the electricity sales price are as risk factors. The commonly used probability distribution functions are normal distribution, α distribution, β distribution, triangular distribution and average distribution, etc. According to the characteristics of construction projects and the characteristics of risk factors, the triangular distribution and average distribution are preliminarily determined as the risk probability distribution, as shown in Table 1.

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Fig. 1. Flow chart of financial risk analysis of nuclear power projects based on Monte Carlo simulation

Taking the financial evaluation method of nuclear power plant construction projects as the basic model, the Monte Carlo simulation is used to randomly sample 1000 times, and the financial internal rate of return of all investments is drawn into a probability distribution and a cumulative probability distribution diagram, where KDE represents

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Table 1. Risk factor table No.

The name of the risk factors

Unit

Probability distribution

Minimum

Possible value

Maximum

1

Overnight cost per kilowatt

yuan/kW

Triangular distribution

10,000

12,000

15,000

2

Load factor

%

Even distribution

70

80

90

3

Feed-in tariff including tax

yuan/MWh

Triangular distribution

380

390

430

the nuclear density function of the internal rate of return, PDF represents the probability density function of the internal rate of return, and CDF represents the cumulative distribution function of the internal rate of return, as shown in Fig. 2.

Fig. 2. Probability distribution and cumulative probability distribution diagram of economic evaluation index results

In Fig. 2, the cumulative probability of 70% corresponds to an internal rate of return of 5%, that is, the probability that the project is greater than 5% is 70%, and investors will bear about 30% of the risk, and the risk appetite is conservative. The internal rate of return corresponding to a cumulative probability of 30% is 8%, that is, the probability of a project greater than 8% is 30%, and investors will bear about 70% of the risk, and the risk appetite is aggressive. According to the above quantitative risk analysis, nuclear power project investors can determine the appropriate internal rate of return according to their own investment risk preferences.

5 Conclusions Carry out risk probability analysis on the financial evaluation of nuclear power projects based on the Monte Carlo simulation method, so that the investors of nuclear power projects can clearly understand the possibility of the emergence of economic evaluation indicators under different risk factors according to the probability distribution of risk

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evaluation indicators, and change the original financial results of a single parameter and a single result. It presents the financial evaluation results of the joint action of various risk factors, so as to provide more comprehensive support for the investment decision of nuclear power projects.

References 1. Economic Evaluation Methods and Parameters of Construction Projects, 3rd edn. 2. The Economic Evaluation Guidelines of the Nuclear Power Plant Construction Project (NB/T 20048-2011) 3. Application of Monte Carlo Simulation in Economic Evaluation of Petrochemical Projects

Research on Low Carbon and Energy Saving Technology Path of Nuclear Power HVAC System Yang Li(B) , Guo Chuang Chen, Zhang Xin, and Wang Chong Hualong International Nuclear Power Technology Co. Ltd., Shen Zhen, Guang Dong, China [email protected]

Abstract. Nuclear power is the most realistic choice to cope with climate change and low-carbon transformation of energy structure, and it is also the inevitable trend for China to achieve the “double carbon goal”. At present, the design task of nuclear power HVAC is not only to meet the functional requirements of users, but also to fulfill the tasks of efficient use of resources, environmental protection, energy saving and emission reduction. This paper discusses several “low carbon” technology paths that can be researched, developed and selected in the design process, including cold storage, pipeline optimization, cooling heat recovery, cold and hot air distribution, etc., analyzes the main problems and applicable conditions, and provides reference suggestions for the sustainable development of HVAC design and operation of nuclear power in China. Keywords: Nuclear power · Low-carbon · Nuclear HVAC · Cold storage · Pipeline optimization · Cooling heat recovery

1 Introduction In the multi-wheel drive energy system of coal, oil, gas, electricity, nuclear, new energy and renewable energy, which has been formed in China. Nuclear power has high energy density, large single power, long-term stable operation, clean and efficient application for bearing the base load of large power grid and necessary peak shaving, and has outstanding advantages compared with other energy sources. It is the most realistic strategic choice for China to achieve the goal of “double carbon” and build a low-carbon energy system [1]. It is estimated that the installed capacity of nuclear power in China will account for more than 10% of the total power generation by 2030 and more than 20% by 2060. For the nuclear power industry, the future nuclear power engineering industries such as nuclear power design, construction and commissioning will face unprecedented opportunities and challenges. Nuclear power plant can be regarded as a large-scale system with precise cooperation of multidisciplinary systems, and each “subsystem” has different operation modes and energy consumption modes. HVAC is also one of them, which consists of about 20 subsystems such as ventilation, air conditioning, smoke control and chilled water. In © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 127–133, 2023. https://doi.org/10.1007/978-981-19-8899-8_12

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the past design work, subject to the application requirements of mature technologies, energy saving and low carbon are not important tasks of design procedure, but only whether the relevant Standard is achieved as a measure. In other industries, especially in the construction industry, the energy consumption of HVAC system is listed as the primary energy consumption in the operation stage, and the goals and measures of “green” and “low carbon” have been generally accepted and implemented [2], and the corresponding technologies tend to be mature and reliable. Based on the above situation and the wide application of HVAC system in nuclear power plants, this paper discusses several energy-saving technology paths that can be studied and selected in the design process, including cooling heat recovery, cold and hot air distribution, cold storage, etc., analyzes the applicable conditions and basic ideas, and provides reference suggestions for the “low carbon” idea in the design and operation stage of nuclear power HVAC.

2 Cooling Heat Recovery The air conditioning system of the main control room is used to maintain the appropriate temperature and humidity in the habitable area of the main control room during the running state and accident condition of the plant, and provide suitable environmental conditions for the correct operation of equipment and personnel [3]. Ensure the habitability of the habitable area of the main control room in the event of radioactivity pollution incident in Yard. This system does not directly participate in the three major security functions of plant, but as a support system, it provides suitable operating environment conditions for the systems and equipment participating in the three major security functions of plant. In the existing nuclear power technology, the main control room of the nuclear island is equipped with the air conditioning system of the main control room. When the system is in normal operation, it is in the mode of fresh air plus primary return air and all air. The supplied air is uniformly treated by the air handling unit and then sent to the area to maintain the appropriate environmental conditions of temperature and humidity. After the air is cooled and dehumidified by the surface cooling section of the air handling unit (connected with the chilled water system serving the area), it needs secondary heating to reach the temperature and humidity of the supply air. At present, it is used by Electric heater, at the end of Duct. The heat dissipation of nuclear power plant equipment is large, and the air conditioning system in the main control room runs continuously all the year round. The above-mentioned methods have the situation that high-grade cold and heat sources cancel each other. Therefore, under the above conditions, the cooling water (minimum 15 °C and maximum 45 °C) corresponding to the refrigerator of the chilled water system can be recovered by adding a hot water Coil heating section in the air handling unit of the air conditioning system of the main control room; A connecting pipeline loop and an electric valve are added to the cooling water from the heating section of the hot water Coil to the refrigerator of the original main control room air conditioning system corresponding to the chilled water system, and part of the cooling water is introduced to heat and raise the air temperature after cooling and dehumidification by the surface cooler, so as to meet the air supply temperature and humidity required by the indoor environment. Its principle is shown in Fig. 1. The method can realize the heat energy recovery and secondary utilization of the cooling water of the refrigerator, not only can reduce the electric heat energy consumption

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Fig. 1. Schematic diagram of cooling water heat recovery in main control room of nuclear island

of the heated air, but also can improve the refrigerant efficiency. The required pipeline modification is also easily coupled with the existing air conditioning and chilled water systems, with low economic cost, basically no influence on other systems and high practicability.

3 Hot and Cold Air Distribution Most of the electrical instruments and equipment in the nuclear island plant are concentrated in the electrical room, control cabinet room, communication computer room command center and other rooms of the whole plant. All kinds of electrical cabinets, control cabinet and other electrical equipment have a large amount of heat dissipation, and the HVAC system mainly uses exhaust air to dissipate heat and has a large amount of ventilation. Take part of the ventilation [3]. At present, the air distribution scheme adopted by the electric instrument computer room served by the system is the way of side air supply and side air return, which is no different from other functional rooms. This paper thinks that we can refer to the design mode of civil data center [4]. Consider establishing targeted air flow organization of sending and returning air, which can reduce the mixed flow of cold and hot air flow validity and improve the utilization rate of cold air and cold energy. On this basis, the air distribution situation can be air supply and return under the overhead floor [5]. That is to say, the overhead floor is used as the static pressure box, and the cold air is sent out from the air supply outlet located in the cold channel in front of the cabinet, sucked by the fan in the Power supply of the cabinet, and then discharged into the hot channel after absorbing heat, and then entered the return (exhaust) air pipe located at the top to reciprocate. The air supply mode has the main advantage of less electric energy loss compared with the traditional mode in operation.

4 Cold Storage In some large-scale air conditioning water systems, in order to achieve the function of continuous refrigeration, the main measures adopted are cold storage technology. It should be considered that the air conditioning system can still maintain normal operation

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when the power system supplied by the system has unexpected power failure. Therefore, it is necessary for the water system to configure cold storage tanks to store the cooling load needed by all systems in stable operation after switching from commercial power to diesel power. There are three kinds of cold storage media commonly used in engineering [6]. Water: The temperature change of water is used to store sensible heat. The cold storage temperature is generally above 4 °C, and the temperature difference of cold storage is 5–11 °C. The accumulated cooling capacity per unit volume of water is small, ranging from 5.8 to 12.77 kWh/m3 , and the volume of cold storage tank is large. The refrigerator can operate under conventional air conditioning conditions or slightly lower temperature. Ice: the latent heat of ice dissolution is used to store cold energy, and the ice making temperature is generally − 4 to – 8 °C; The storage capacity of ice is 40– 50 kWh/m3 , and the storage capacity of ice storage tank is smaller than that of water storage tank. Ice storage can provide lower water supply temperature, which is suitable for air conditioning projects with large temperature difference and low temperature water supply, low temperature air supply, regional cooling and unconditional design of water storage. However, at least one refrigerator is of double working conditions, and its efficiency will decrease during ice making. Eutectic salt: The mixture of inorganic salt and water is called eutectic salt, and the phase transition temperature of eutectic salt is generally 5–7 °C. The storage capacity is about 20.8 kWh/m3 , and the volume of storage tank is between ice storage and water storage. The refrigerator can operate under conventional air conditioning conditions, but the maturity and reliability of eutectic salt materials need to be tested by time. Water cold storage is a cold storage mode with the smallest accumulated cold capacity per unit, which requires the largest volume of cold storage tank. However, water cold storage does not need double-working refrigeration units, and does not need special cold storage forms commonly used in refrigeration and mass interception centers. It is well combined with the original conventional chilled water system. At present, cold water storage tanks are used in chilled water systems of nuclear power plants in recent years. For example, the operating chilled water system of a nuclear power plant in is equipped with four large-capacity main chilled water production columns [3], each column consists of a chiller and a Circulating Water Pump, one of which is standby. Considering the lowload operation in winter, another cold storage tank is set to avoid the low-load operation of large-capacity cooler. A bypass Regulating pipeline is set between the main water supply and return pipes of the system, so as to ensure the constant flow of cold water through the chiller when the flow rate of users changes. Two expansion constant pressure tanks are used for constant pressure. According to the demand of ventilation system, the system determines the number of coolers and corresponding pumps put into operation. The cold storage tank is used to improve the thermal inertia of the system, and avoid the frequent start-up and stop of the cold water cooler caused by the low cooling load of the users served by the system under the working condition of “complete discharge of the reactor in winter”, and the continuous cooling capacity of the chilled water system in Guaranty under the condition of short-term power loss. However, there are many problems in this scheme at present. Due to the large heat dissipation of various process equipment pipelines on the nuclear island, and nuclear power safety considerations, the safety margin is high, the system capacity is set according to the maximum, and the main

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equipment of the system (including water chillers, water storage tanks, water pumps, diaphragm constant pressure tanks, etc.) occupies a large area and space cost, which is not conducive to the optimal design of the nuclear island plant; In normal operation, chillers and corresponding pumps are easy to be in the low efficiency range of equipment, which not only causes a lot of energy waste, but also is an unfavorable factor for the maintenance of the equipment system itself; In the case of losing the Power supply outside the plant, the chilled water supply from the Guaranty of the electric refrigeration chiller caused the system equipment to occupy a large load on the emergency diesel engine; The operation chilled water service area is large, and the system water capacity is high. The system is equipped with water storage cold storage tanks, which further increases the system water capacity and increases the leakage risk and layout difficulty to a certain extent. The diaphragm type constant pressure tank designed by calculation is large in volume, and it is difficult to arrange it in the factory building. In this paper, a scheme of ice-storage and cold-storage chilled water system for nuclear island operation is proposed for discussion. Two water-cooled electric chillers with a load capacity of 2 × 50% and corresponding water pumps are arranged in parallel in the system (the cooling water is taken from the nuclear island chilled water system); Set up two ice storage tanks with a capacity of 50% of the system load and supporting cold release facilities; Two water chillers and two cold storage tanks can be used as standby for each other; Set up one high-level expansion water tank; Set up corresponding pipelines, electric valves, Strainer, etc. to provide chilled water for the normal operation of nuclear island unsafe air conditioning and ventilation system and other systems requiring chilled water cooling source, to release cold from ice storage cold storage device in case of losing the Power supply of the plant, and to provide chilled water for users who need to operate under this working condition within a certain period of time. The schematic diagram of the scheme is shown in Fig. 2.

Fig. 2. Diagram of the scheme

Under the premise of safety redundancy, the scheme should set as few electric refrigeration chillers (capacity 2 × 50%), parallel ice storage tanks (2 × 50%) and supporting cold release facilities as possible to meet the functional requirements. The total occupied area of system equipment is reduced, which is beneficial to the optimization of plant layout; During normal operation, two chillers and two ice storage cold storage tanks can be used as standby for each other; The cold storage tank can also be used as

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a continuous and uninterrupted chilled water cooling source when the external power (LOOP) is lost, and only the cold release water pump loads the emergency diesel engine Power supply, which greatly reduces the corresponding emergency diesel engine load; During normal operation, the cold storage device can release the cold flow according to the end load Regulating, so as to take the total chilled water supply in Regulating and the Guaranty electric refrigeration chiller to operate in the high energy efficiency ratio range, which improves the operation efficiency and economy of the system; Because of the large water capacity of the system, the high-level expansion water tank of the system is selected for constant pressure, which is smaller than the diaphragm constant pressure tank and occupies a small area, and the constant pressure water replenishment effect is stable; Ice storage has small cold loss (2–3%), which is phase change storage. Compared with water storage, ice storage has small volume and saves land. The cold release end can realize closed system, and the energy consumption of water pump is also small, all of which achieve the goal of relative energy saving. Ice storage technology is widely used in civil construction industry in China, and its reliability and experience have been accumulated for a long time. It is also feasible to use it in non-safety system of nuclear island, which is worthy of in-depth discussion.

5 Conclusion In this paper, three preliminary energy-saving schemes are put forward, including heat recovery of cooling water, application of air distribution in cold and hot channels and application of cold storage technology in chilled water system of nuclear island, and their application scenarios and feasibility are given. At the same time, the new scheme will always bring new problems. The recovery and reuse of cooling water heat energy in the main control room area can not only reduce the electric heat energy consumption of heated air, but also improve the refrigerant efficiency. The required pipeline modification is easily coupled with the existing air conditioning and chilled water systems, which has low economic cost, basically has no influence on other systems and high practicability. However, due to the division of areas, the relevant pipelines can only be laid in this area, and the total energy recovered is low. The air distribution method of air supply and return under the overhead floor makes the circulation of air supply and return in the electrical equipment room smoother, absorbs heat more accurately, saves a lot of energy consumption in ventilation system operation, and the selection of main equipment such as Fan in ventilation system configuration can also reduce the risk of “hot spots” of electrical equipment, which is helpful for equipment maintenance. The scheme is also feasible in plant equipment layout. But this scheme needs to increase a lot of interfaces with electrical equipment specialty, and it needs to increase the accuracy of design calculation, so as to save energy and not reduce the reliability of the system at the same time. The scheme of ice storage nuclear island chilled water system applies the principle of phase change cold storage to the load Regulating of unsafe nuclear island chilled water system running continuously all year round, which not only reduces the initial investment cost and operation energy consumption of the system, but also makes the system cope

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with various load states. This scheme is quite different from the existing scheme, and there are still many problems to be further studied. Although some main equipments, such as double-working condition chillers, ice storage tanks and ethylene glycol solution pumps, are widely used in other industries, there is no reference experience in nuclear power plants. Therefore, research and Arguments also need time and accumulation, and need further study. To sum up, under the goal of energy saving and low carbon of nuclear power HVAC system, engineering designers should constantly optimize the design scheme in design procedure, plant Province based on the principles of energy saving, recovery and recycling, and find a balance between reliable technology and “green and low carbon”, so as to provide reference and power for the sustainable development of nuclear power HVAC design and operation in China.

References 1. Zhang, Y.: Analysis of China’s nuclear power development trend under the dual carbon target. Nucl. Sci. Eng. 6, 1347–2135 (2021) 2. Du, Z., Zhou, L.: On the importance of building energy efficiency and its technical application. Urban Archit. 8, 133 (2013) 3. Design Criteria for Heating, Ventilation and Air Conditioning Systems Outside Containment of PWR Nuclear Power Plants in CNPEC NB/T 20095-2012, pp. 23–26. Institute of Nuclear Industry Standardization, Beijing (2012) 4. Zhang, L., Ouyang, S., Hu, Y., et al.: Study on the influence of cold channel closure in data Central Computer Room on air distribution and energy efficiency in Computer Room. Power Inf. Commun. Technol. 5, 12–16 (2018) 5. Shen, G., Zhu, X., Qiao, L., et al.: Energy-saving design and principle of cold and hot passages in Computer Room, data. Build. Technol. 2, 116–118 (2016) 6. Lu, Y.: Handbook of Practical Heating and Air Conditioning Design, pp. 969–971, 211–2175. China Building Industry Press, Beijing (2008) 7. Zheng, X.: Application of green technology in HVAC design under low carbon background. Theor. Res. Urban Constr. (Electron. Ed.) 18, 54 (2019)

Multi-objective Optimization Design of Plate Heat Exchanger for Spent Fuel Pool Cooling System Yuquan Shang(B) , Weiguang Zhao, and Changqi Yan Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, Harbin 150001, China [email protected]

Abstract. In order to improve the economic benefit of the spent fuel pool cooling system and the scientific nature, taking a plate heat exchanger in the spent fuel pool cooling system as the main research object, with the investment costs, cold side of the heat exchanger of the resistance to flow and the flow resistance of hot side as the optimization goal, through a new hybrid genetic algorithm, a multi-objective optimization design of heat exchanger is analyzed. The calculation results show that, within the optimized range, there is an opposite trend between the investment cost of the heat exchanger and the flow resistance of the cold side and the hot side, and the same trend between the cold side and the hot side. When compared with the parent value, the three optimization objectives can get better solutions simultaneously. Therefore, the heat exchanger design of spent fuel pool cooling system has a large optimization space and a variety of optimization angles. The research results can provide theoretical reference for the optimization design of spent fuel pool system and improve the scientific and economic design of the system. Keywords: Spent fuel pool cooling system · Plate heat exchanger · Economy · Optimal design · Multi-objective

1 Introduction The spent fuel pool cooling system is an important auxiliary safety system to ensure the normal operation of nuclear power plants. In normal and accident conditions, it is necessary to ensure the availability of the system and to perform its safety functions [1]. The spent fuel pool cooling system can cool the spent fuel in the spent fuel pool through plate heat exchanger and cooling water of the equipment, so as to realize the function of exporting the waste heat of spent fuel. In order to ensure that the system has the ability to export enough heat, the early design of its heat exchanger was too conservative, and the design was mainly based on the accumulation of engineering experience and the judgment of the designer [2], resulting in high quality of heat exchanger, high investment cost and low operating efficiency. Therefore, it is necessary to optimize the design of

© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 134–146, 2023. https://doi.org/10.1007/978-981-19-8899-8_13

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the plate heat exchanger in order to reduce the weight, reduce the cost and improve the operation efficiency. In this paper, by means of computer-aided, using a new hybrid genetic algorithm, on the premise of meeting the working requirements of the heat exchanger in the spent fuel pool cooling system, taking the investment cost of the heat exchanger, the flow resistance on the cold side and the flow resistance on the hot side as the optimization objectives, the optimal design of the heat exchanger in the spent fuel pool cooling system is studied. The main process is as follows: ➀ firstly, the evaluation model of the plate heat exchanger is established, including the establishment of the mathematical model of the heat exchanger and the determination of the mathematical model for calculating the investment cost of the heat exchanger. The mathematical model of the heat exchanger includes the heat exchange mathematical model on the cold and hot sides and the pressure drop mathematical model; ➁ Then select the appropriate optimization algorithm; ➂ Finally, based on the determined model and optimization algorithm, the appropriate optimization variables and constraints are selected, and their numerical ranges are determined. The optimization variables are found through the optimization algorithm. Within the range of optimization variables and constraints, the optimal design results of the investment cost of the heat exchanger, the flow resistance on the cold side and the flow resistance on the hot side are found.

2 The Establishment of Evaluation Model for Heat Exchanger 2.1 Establishment of Mathematical Model of Heat Exchanger Before optimization design, design parameters and related process parameters of heat exchanger should be determined [3]. The main design parameters of plate heat exchanger include the flow combination of heat exchanger, the number of heat exchange plates, the heat exchange area of single plate, the length of flow channel and the distance between plates. The process parameters mainly include heat load of heat exchanger, mass flow of cold and hot side fluid and inlet and outlet temperature. On the basis of these parameters, the design of plate heat exchanger is mainly divided into two aspects, heat transfer calculation and flow resistance calculation. Heat Transfer Calculation Model The design calculation of a typical plate heat exchanger is to determine the heat transfer area required by the heat exchanger and the combination of process and flow channel on the premise that the mass flow, inlet and outlet temperature and allowable pressure drop of the fluid at the cold and hot sides are known [4]. (1) Calculation of heat transfer area The heat transfer area of heat exchanger A can be determined by the average temperature difference method [5]: A=

Q K · t

(1)

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where, A is the heat transfer area, m2 ; Q is thermal load, W; K is the total heat transfer coefficient, W/(m2 K); t is the average temperature difference of heat transfer, °C. Considering that both sides of the plate heat exchanger in the spent fuel pool cooling system are single-phase fluid and there is no phase transformation, the logarithmic average temperature difference formula is used to calculate the average temperature difference: t = ϕ

tmax − tmin ln

(2)

tmax tmin

where, ϕ is the logarithmic average temperature difference correction coefficient, which is related to the flow direction of interplate fluid [6]. tmax and tmin are respectively the maximum and minimum temperature difference at the ends of hot and cold fluids during countercurrent heat transfer. The total heat transfer coefficient K of plate heat exchanger can be obtained according to the thermal resistance relation equation of the total thermal resistance as the sum of the various thermal resistance in series, mainly including five thermal resistance:  K=

δ 1 1 + R1 + + R2 + h1 λ h2

−1 (3)

where, h11 and h12 are the thermal resistance of fluid on both sides of the plate; R1 and R2 respectively are thermal resistance of dirt layer on both sides; λδ is the thermal conductivity resistance of the plate. Considering the actual operating conditions of heat exchanger in the spent fuel pool system, the calculation formula of convective heat transfer coefficient h is as follows: h= Nu = CRen

Nuλ d

(4)

0.3or0.4

Pr (μ/μw )p

Re =

vd υ

(5) (6)

where, υ is kinematic viscosity, m2 /s; Pr is Prandtl number; μ and μw are respectively the dynamic viscosity of water under the corresponding water temperature and wall temperature; v is the flow rate, m/s; d is equivalent diameter of flow passage; C, n, p are constants. (2) Division of the process Whether the flow design of plate heat exchanger is reasonable has a great influence on the heat transfer resistance performance of heat exchanger and the initial investment of equipment [7]. Plate heat exchanger process refers to the heat exchanger medium to a direction of the flow of a group of channels, has an important role in the flow distribution

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of heat exchanger. The flow is evenly distributed within each channel, so the interplate flow rate is: v=

G nρA1

(7)

where, G is the total flow, kg/s; n is the number of flow channels; ρ is fluid density, kg/m3 ; A1 is the circulation area, m2 . Flow Resistance Calculation Model Considering that both sides of the plate heat exchanger in the spent fuel pool cooling system are unidirectional fluids, the pressure drop of this type of plate heat exchanger can be calculated according to Euler-Reynolds equation [8]: P = bRed ρv2 M

(8)

where, b and d shall be different values depending on the plate type in the plate heat exchanger, and d shall be a negative value; v is the flow velocity of the flow passage, m/s; M is the number of processes. 2.2 Investment Cost Calculation Model The calculation of the investment cost of heat exchanger is related to the heat exchange area, and Hall correlation formula can be used to obtain [9]: C = a1 + a2 Aa3

(9)

where, C is the investment cost; a1 , a2 and a3 are constant coefficients and are related to the type and material of heat exchanger; A is the heat exchange area of the heat exchanger. 2.3 Establishment of Evaluation Model Program According to the above model formula, on the premise of determining the heat exchange plate model, the established evaluation model program flow is as follows: (1) Determine the flow rate of cold and hot side fluid, inlet and outlet temperature and heat load; (2) Assuming the initial heat transfer area, the number of heat transfer plates is selected according to the heat transfer plate model, and the corresponding flow channel combination is determined; (3) Calculate the heat transfer coefficient and heat transfer area according to the number of heat exchanger plates and processes, and compare with the initial heat transfer area. If the error standard is met, step (4) is carried out; otherwise, step (2) is carried out until the error meets the requirements; (4) Calculate the flow resistance on the hot and cold side, and judge whether the resistance meets the requirements. If the requirements are met, step 5 shall be carried out; otherwise, the flow channel combination shall be re-determined and step (2) shall be carried out; (5) Calculate the investment cost of heat exchanger.

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2.4 Validation of Evaluation Model Procedures In order to verify the accuracy of the calculation of the established evaluation model program, the parameters of plate heat exchanger in the spent fuel pool cooling system of a nuclear power plant were selected and compared with the calculated results. The comparison results are shown in Table 1 (the parent type values in the table are the actual parameters of heat exchanger). It can be seen that the relative error between the calculated parameters and the actual parameters is within the allowable range, so the established evaluation model program has certain applicability to the plate heat exchanger in the spent fuel pool cooling system, and can be used for the evaluation of the plate heat exchanger. Table 1. Comparison between actual parameters and calculated results of heat exchanger Design parameters

Parent value

Programmed value

Relative error (%)

Hot side outlet temperature (°C)

37.08

37.08

–

Cold side outlet temperature (°C)

35

35

–

Hot side flow (m3 /h)

450

450

–

Cold side flow (m3 /h)

450

450

–

Number of processes (m3 /h)

1

1

–

Number of heat exchange plates

159

157

1.25

Heat transfer coefficient (W/m2 /K)

7264

7359

1.3

3 Optimization Algorithm Program Genetic algorithm is a meta-heuristic algorithm based on the principles of natural inheritance and natural selection [10]. It gradually approaches the optimal solution in an iterative manner by simulating the process of replication, crossover and variation of biological evolution. Genetic algorithm has been widely used and improved in equipment optimization design. Kuining et al. [3] adopted genetic algorithm to optimize the design of shell and tube heat exchanger. Sarangi et al. [11] used genetic algorithm to optimize the geometry of curved trapezoidal fins in heat exchangers. Jing et al. [12] adopted compound form-genetic algorithm to optimize the design of condenser. In this paper, a new hybrid genetic algorithm is used to optimize the heat exchanger design. The new hybrid genetic algorithm combines simplex algorithm with genetic algorithm, which has higher convergence efficiency and deeper search depth compared with genetic algorithm. At the same time, the multi-objective processing method in NSGAII is integrated, so that the algorithm can be applied to multi-objective optimization design.

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4 Optimization Example and Result Analysis 4.1 Selection of Optimization Variables Based on the above evaluation model and the selected optimization algorithm, the plate heat exchanger in the spent fuel pool cooling system is optimized. Considering that the heat transfer structure of plate heat exchanger is heat transfer plate, the number of heat transfer plate determines the heat transfer area and the number of flow channels. The heat exchange area and the number of channels affect the heat exchange capacity of heat exchanger, and the heat exchange area determines the investment cost of equipment, and the number of channels affects the flow resistance of the fluid in the heat exchanger. So in the new hybrid genetic algorithm to optimize the plate heat exchanger, choose investment cost and the flow resistance of hot and cold side as the optimization goal, on the cold side flow, heat flow rate, outlet temperature of the hot side and cold side inlet temperature as optimization variables, the parent value would be included value range of optimized variables to combination of optimization variables, considering the actual situation at the same time, The inlet temperature of the hot side of the heat exchanger, the flow resistance of the hot and cold side and the investment cost are set certain constraints. The search range of optimization variables is shown in Table 2. Table 2. Optimization variable search range Optimization variables

Parent value

Lower limit of value

Upper limit of value

Hot side outlet temperature (°C)

37.08

35

45

Cold side outlet temperature (°C)

35

30

38

Hot side flow (m3 /h)

450

300

495

Cold side flow (m3 /h)

450

300

495

4.2 Optimization Results and Analysis Through the program calculation, under the condition of no constraint on the flow rate on the hot and cold side of the heat exchanger, the optimization results of the three objectives are shown in Fig. 1. The red dots in the figure are the results of the parent design scheme, and the black dots are the final results of the optimized design scheme. It can be seen that the calculated optimization results have good continuity, and compared with the parent value, there is a large optimization space. Considering that the heat exchanger is in the cooling system of the spent fuel pool, the flow rate of the cold side and the hot side should not be too different, so as to avoid the difference of the flow resistance of the cold side and the hot side being too large. Therefore, certain constraints are set on the difference between the cold and hot side flows. On this basis, the three-objective optimization results of the plate heat exchanger are shown in Fig. 2. It can be seen that after increasing the constraints, the continuity of the

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calculation results becomes stronger. After adding the flow constraint, the projections of the obtained three-objective optimization results in any two of the optimization objective spaces are shown in Fig. 3. Optimize calculation results Parent calculation results











           



 















ide

flo w







ts













Ho







flow resis tanc e/Pa

res is



/ Pa

 



Cold side

ce

Y

                    

tan

Investment cost/CN

       

Fig. 1. Three-objective optimization results (without constraints on the hot and cold side flows)

Optimize calculation results Parent calculation results

































 

 

  

Ho ts ide flo w







 

a /P ce an st si re

  

ow fl





de si







 

d ol







C

res isi tan ce /Pa

Y

                    



Investment cost/CN

        







Fig. 2. Three-objective optimization results (with constraints on the hot and cold side flows)

It can be seen from Fig. 3a that the flow resistance on the cold side changes linearly with the flow resistance on the hot side, and the numerical values are basically the same. This is because during the optimization calculation, the difference between the flows on the cold side and the hot side is constrained, and the numerical values of the flows on both sides are basically the same. When the flow on one side increases, the flow on the other side also increases, resulting in the increase of the flow velocity between the fluid plates on both sides, Thus, the flow resistance on both sides increases; On the contrary, when the flow rate on one side decreases, the flow rate on the other side also decreases,

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resulting in the decrease of the flow velocity between the fluid plates on both sides, resulting in the decrease of the flow resistance on both sides.

Hot side flow resistance/Pa

Optimize calculation results Parent calculation results

Cold side flow resistance/Pa

(a) Spatial projection of flow resistance on hot and cold sides

Investment cost/CNY

Optimize calculation results Parent calculation results

Hot side flow resistance/Pa

(b) Spatial projection of hot-side flow resistance and investment cost

Investment cost/CNY

Optimize calculation results Parent calculation results

Cold side flow resistance/Pa

(c) Spatial projection of cold-side flow resistance and investment cost Fig. 3. Spatial projection of optimization results on any two optimization objectives

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It can be seen from Fig. 3b and c that, when compared with the results of the parent model, the results of the three optimization objectives are all better than the results of the parent model, which indicates that the design scheme of plate heat exchanger in the spent fuel pool cooling system has a large space for optimization. It can be seen from the change of optimization results that the flow resistance on both sides of the heat exchanger has an opposite trend to the investment cost. This is because when the investment cost of heat exchanger is reduced, the heat exchange area will also be reduced, resulting in a reduction in the number of heat exchange plate plates and the number of channels, so that the flow rate between the plates on both sides of the cold and hot increases, resulting in an increase in the resistance on both sides of the cold and hot; On the contrary, when the investment cost of heat exchanger increases, the heat exchange area will also increase, resulting in the increase of the number of plates and the number of flow channels of the heat exchange plate, so that the flow rate between the cold and hot sides decreases, resulting in the increase of the resistance of the cold and hot sides. Therefore, it is necessary to consider the relationship between the three optimization objectives comprehensively when choosing the optimization scheme. 4.3 Sensitivity Analysis Based on the established heat transfer calculation model, flow resistance calculation model and investment cost evaluation model, the analysis results for a certain optimization variable can be given. During the analysis, a set of design parameters can be determined by changing the value of one optimization variable and keeping the other variables consistent with the parent scheme. According to the established model, the corresponding optimization objective results under the combination of design parameters can be calculated. The influences of cold side inlet temperature and cold side flow rate on cold side flow resistance and investment cost of heat exchanger are given below. Influence of Cold Side Inlet Temperature Figures 4 and 5 respectively show the influence of cold side inlet temperature on cold side flow resistance and investment cost. It can be seen that at the same flow rate, with the increase of the inlet temperature of the cold side, the flow resistance of the cold side tends to decline, and the investment cost tends to rise. This is because when other variables remain unchanged, the increase of cold side inlet temperature will increase the number of heat exchanger plates designed, so that the increase of heat transfer area and flow channel number will increase, and the increase of heat transfer area will lead to the increase of investment costs, and the increase of flow channel number will reduce the flow rate of fluid, resulting in the decline of flow resistance. Influence of Cold Side Flow Rate Figures 6 and 7 show the influence of cold-side flow on cold-side flow resistance and investment cost. It can be seen that with the increase of the cold side flow, the cold side flow resistance presents a rising trend, and the investment cost presents a declining trend. This is because when other variables remain unchanged, the increase in the flow rate of the cold side will reduce the number of plates of the designed heat exchanger, reduce the heat transfer area and the number of flow channels, and the reduction of the heat transfer

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Cold side flow resistance/Pa

Cold side flow resistance

Cold side inlet temperature/

Fig. 4. Influence of cold side inlet temperature on flow resistance

Investment cost/CNY

Investment cost

Cold side inlet temperature/

Fig. 5. Influence of cold side inlet temperature on investment cost of heat exchanger

area will lead to the reduction of investment costs, and the reduction of the number of flow channels will increase the flow rate of the fluid, resulting in the increase of flow resistance. 4.4 Selection and Analysis of Optimization Scheme Among the three optimization objectives, the investment cost of heat exchanger and the flow resistance on the hot and cold side have an opposite trend, and the resistance on the hot and cold side has the same trend. Therefore, when looking for the optimization scheme, the three optimization objectives need to be considered comprehensively. In order to better describe the optimization results, this paper selected 6 typical optimization results and analyzed them from different perspectives. The selection results are shown in Table 3.

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Cold side flow resistance/Pa

Cold side flow resistance

Cold side inlet flow/(m3/h)

Fig. 6. Influence of cold side flow on cold side flow resistance

Investment cost/CNY

Investment cost

Cold side inlet flow/(m3/h)

Fig. 7. Influence of cold side flow on investment cost of heat exchanger

As can be seen from the table, with the relative increase of investment cost, cold and hot side resistance will decrease correspondingly in the selected schemes. Among them, the investment cost in plan1 is the least. Compared with the mother model value, the investment cost is reduced by 39.1%, but the cold and hot side resistance is increased by 22.7% and 21.7% respectively. This is because the reduction of investment cost will reduce the heat transfer area, and the number of heat transfer plates of the heat exchanger will also be reduced, resulting in the increase of the number of flow channels and the increase of the flow rate between plates. This causes an increase in flow resistance. Among the selected optimization schemes, when the investment cost increases to a certain value, the schemes superior to the parent value can be obtained for all three optimization objectives, as shown in plan 4 to 6 in the table. Under the condition of constant heat transfer, the inlet temperature and flow rate of the cold side of the heat exchanger are reduced to a certain extent, and the outlet temperature of the hot side is

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increased. At the same time, the number of heat transfer plates is reduced, resulting in the optimization of the investment cost of the plate heat exchanger and the flow resistance of the cold and hot sides to a certain extent. Among them, compared with the parent model value, cold and hot side resistance in plan 6 decreased more, 4.85% and 5.53% respectively, but compared with plan 4 and plan 5, investment cost increased. Therefore, from the perspective of investment economy, the optimization result of plan 1 is better. From the perspective of operation economy, the optimization result of plan 6 is better. It can be seen from the results of the optimization scheme that there is still a large space for optimization of plate heat exchanger in the spent fuel pool cooling system. Table 3. Typical results and parent values Plan

Investment cost

Parent ratio (%)

Hot side resistance

Parent ratio (%)

Cold side resistance

Parent ratio (%)

Parent value

451,940.2

100.00

128,710

100.00

128,850

100.00

Plan 1

275,241.75

60.90

156,638.58

121.70

158,093.76

122.70

Plan 2

276,755.65

61.24

154,810.88

120.28

156,366.18

121.36

Plan 3

291,021.5

64.39

141,282.04

109.77

142,508.63

110.60

Plan 4

308,895.19

68.35

126,274.68

98.11

127,360.78

98.84

Plan 5

314,592.47

69.61

122,004.63

94.79

123,078.43

95.52

Plan 6

315,286.61

69.76

121,589.28

94.47

122,604.64

95.15

5 Conclusion In this paper, the new hybrid genetic algorithm is applied to the multi-objective optimization design of plate heat exchanger in the spent fuel pool cooling system. According to the optimization results, the following conclusions are drawn: (1) When optimizing the investment cost, cold side flow resistance and hot side flow resistance of the plate heat exchanger in the spent fuel pool cooling system, it can be found that the investment cost of the heat exchanger and the flow resistance on the cold and hot side have opposite changes. There is the same change trend between the cold and hot side resistance, so it needs to be considered comprehensively when designing the optimization scheme. (2) In the optimization results listed, the investment cost of heat exchanger can be reduced by 39.1% at most, and in some optimization schemes, the results of the three optimization objectives can be better than the mother type value, which indicates that there is still a large space for optimization of plate heat exchanger in the spent fuel pool cooling system.

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(3) The optimization results are only applicable to the range of optimization variables and constraints considered in this paper, while in practical engineering applications, the engineering itself and other factors affecting the system should be considered when optimizing the plate heat exchanger in the spent fuel pool cooling system. However, the research results of this paper can provide theoretical reference for optimizing the design of spent fuel pool system and improving the scientificity and economy of system design.

References 1. Zhang Yong, L., Hang, L.M.: Research on the layout principle of spent fuel pool cooling system. Ind. Technol. Forum 18(07), 81–82 (2019) 2. Guijing, L.: Multi-objective Parameter Optimization of Nuclear Power Equipment and Systems. Harbin Engineering University (2015) 3. Kuining, L., Jiguang, Z., Jin, L., Junfeng, G.: Application of genetic algorithm to optimal design of shell and tube heat exchanger. J. Chongqing Univ. 34(08), 97–102 (2011) 4. Songwen, Q.: Heat Exchanger Design Manual. China Petrochemical Press, Beijing (2002) 5. Shiming, Y., Wenshuan, T.: Heat Transfer. Higher Education Press, Beijing (2006) 6. Jingfei, Y.: Research on Optimization Design of Plate Heat Exchanger. Xinjiang University (2013) 7. Shuzhen, H.: Research on the Realization Method and Performance of Multiple Processes in a Single Channel of a Plate Heat Exchanger. Shandong University (2015) 8. Zhang, Y., Yanlong, J., Hexu, W., Chengbin, S.: Experimental research on fluid resistance characteristics of large plate heat exchangers. Power Stn. Syst. Eng. 33(02), 1–4 (2017) 9. Taal, M., Bulatov, I., Klemeš, J., Stehlík, P.: Cost estimation and energy price forecasts for economic evaluation of retrofit projects. Appl. Therm. Eng. 23(14) (2003) 10. Xiong, Y., Dong, W.: A review of the application of genetic algorithms in classical cryptanalysis. J. Univ. Inf. Eng. 22(05), 577–583 (2021) 11. Sarangi, S.K., Mishra, D.P., Ramachandran, H., Anand, N., Masih, V., Brar, L.S.: Analysis and optimization of the curved trapezoidal winglet geometry in a high-efficiency compact heat exchanger. Int. J. Therm. Sci. 164 (2021) 12. Jing, Z., Changqi, Y., Jianjun, W.: Optimal design of condenser weight. Nucl. Power Eng. 32(03), 134–138 (2011)

Fuel Economy Analysis of ATF Assembly with SiC Cladding and UO2 (BeO) Pellets in CPR1000 Tingting Zou(B) , Wei Gao, and Xin Wang China Nuclear Power Technology Research Institute Co., Ltd., Shenzhen, Guangdong, China [email protected]

Abstract. Silicon carbide (SiC) and enhanced UO2 (BeO) are the candidate cladding material and fuel pellet for application with accident-tolerant fuel respectively. As cladding material, SiC is expected to reduce hydrogen and heat produced by oxidation reaction with high temperature steam. A small amount of BeO oxide added to the existing UO2 pellet can enhance the thermal conductivity of fuel pellet. Using the PCM software developed by CGN Research Institute, this paper intends to study the fuel management economy of the existing CPR1000 with the candidate ATF fuel which is the combination of SiC cladding and UO2 (BeO) pellet. Based on the 18-month fuel cycle mode, the core fuel management design of CPR1000 is carried out with the ATF candidate fuel. Compared to the reference fuel management scheme with different batch refueling numbers and M5 AFA3G assembly with U-235 enriched at 4.45%, the assembly discharging burn-up, annual fuel assembly consumption and uranium consumption per power generation are analyzed and compared to evaluate the ATF fuel utilization rate. The study shows that if the initial loading capacity of U-235 is consistent in the reference core and in the ATF core, the ATF core has higher assembly batch discharging burn-up, lower annual fuel consumption and lower uranium consumption per power generation. Compared to reference fuel management scheme, the ATF fuel has better fuel utilization rate. Keywords: SiC · UO2 + BeO · Fuel management · Economic analysis

1 Introduction After the Fukushima nuclear accident in 2011, how to develop nuclear power safely and efficiently has become an important goal of nuclear power safety, and the research on accident tolerant fuel (ATF) is of great significance to improve nuclear power safety. The main purpose of ATF fuel design is to improve the safety and economy of reactors. Since ATF fuel design is mainly used in current and future light water reactors, its design should maintain the current fuel management economy level of reactors. Therefore, for the core loaded with ATF fuel, whether its design can keep the original cycle length, discharging burn-up and the number of fuel assembly required for annual power generation are the key issues in ATF fuel economy analysis. © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 147–159, 2023. https://doi.org/10.1007/978-981-19-8899-8_14

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In the Fukushima accident in Japan, hydrogen generated by the reaction of Zr alloy, a fuel cladding material with high temperature steam, is the main factor leading to the serious accident. Therefore, in order to reduce the effect of hydrogen generated by the reaction of high temperature steam and Zr alloy in the reactor, SiC material is selected as the key fuel cladding material for ATF fuel. High thermal conductivity is a remarkable characteristic of SiC as nuclear-grade material, especially in TRISO fuel pellets [1] and full-ceramic micro package pellets (FCM) [2]. However, the thermal conductivity of SiC decreases sharply [1, 3] (1–2 orders of magnitude) due to the generation of irradiation Defect in irradiation environment, and the degree of thermal conductivity decrease is related to irradiation temperature. For SiC and SiC composite material, the thermal conductivity decreases rapidly from 20–30 W/(m K) to 3–5W/(m K) by neutron irradiation at LWR coolant temperature. The sharp decrease of thermal conductivity and the increase of the gap between the cladding and pellets will eventually lead to the temperature difference between the inside and outside of the cladding reaching about 100K, which will cause radial irradiation swelling of the cladding to different degrees, and then produce strain in the cladding tube, which may eventually lead to cladding cracking [4]. At this time, adding 10% BeO in volume to the fuel pellet can improve the thermal conductivity of the pellet. At high temperature, BeO has good chemical inertness, small neutron capture cross section and good moderation effect. The research shows that adding 10% BeO in volume to UO2 pellet can improve the thermal conductivity of the fuel pellet by about 50% [5–8], and the temperature drops about 100k [9]. In this paper, SiC cladding and enhanced UO2 (UO2 /BeO) pellet combination of ATF fuel is selected as the research target, and the fuel management design of this selected fuel is carried out to explore whether ATF fuel can maintain the fuel management economy of existing reactors.

2 Model and Calculation Method The traditional CPR1000 PWR core is selected as reference core. The core of CPR1000 PWR consists of 157 groups of Fuel assembly, the activity height of the core is 365.76 cm, the fuel assemblies are M5 AFA3G Fuel assembly consists of UO2 ceramic pellets with an enrichment of 4.45%. Choosing this mature core is beneficial to compare and analyze the fuel management economy of ATF fuel. The cladding of ATF fuel is made of SiC material, and the pellet is made of enhanced UO2 (UO2 /BeO), in which the volume fraction of BeO is 10%. Other design parameters are consistent with the M5 AFA3G Fuel assembly. In order to facilitate the comparative analysis with the M5 AFA3G Fuel assembly, two kinds of U-235 enrichment in ATF fuel pellets are selected: one is 4.45%, which is consistent with the M5 AFA3G Fuel assembly; the other is 4.95%, which makes the loading capacity of U-235 equivalent to that of M5 AFA3G Fuel assembly. In this paper, the equilibrium cycle of the traditional 18-month refueling strategy is selected as a reference, and 48–72 assemblies of different batch refueling numbers are selected to design the benchmark loading pattern. Then, according to the benchmark loading pattern, the equilibrium cycle with the same batch refueling number is designed

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by using the ATF fuel, and the ATF core parameters are compared with those of the benchmark loading pattern.

3 Design Criteria and Method 3.1 Design Criteria In order to facilitate the comparison with the benchmark, the benchmark core design and the ATF core design use the same safety limit assumptions. Based on the expected excellent capability of ATF fuel, the core safety limits used in this design analysis are as follows: nuclear enthalpy rise factor (FH ) is less than 2.0; the hot spot factor (FQ ) of condition I is less than 3.0; the temperature coefficient of moderator must be negative or zero at various power levels, so that the reactor has negative feedback characteristics; shutdown margin is more than 2300pcm; the core satisfies 1/4 rotational symmetry. At the same time, in order to further improve the fuel economy of ATF, the maximum burn-up of fuel rods is extended up to 70 GWd/tU. 3.2 Method In this paper, PCM software package developed by CGN Research Institute [10, 11] is used for design, the software package mainly includes: assembly lattice transport calculation code PINE and core diffusion calculation code COCO. The main function of PINE is to carry out two-dimensional transport-burnup calculation for PWR fuel assembly, and provide the equivalent homogenization parameters of fuel assembly (including neutron diffusion coefficient, average macroscopic cross sections of various types of assemblies and discontinuity factors of assemblies surface) needed for the calculation of coarse nodes in the core. COCO uses non-linear semi-analytical nodal method to solve neutron diffusion equation. COCO can simulate the whole burnup process from the beginning to the end of cycle, and can also realize the functions of fuel rod power reconstruction calculation, control rod insertion calculation and refueling calculation at various burn-up. The neutron cross section database for COCO is provided by PINE.

4 Fuel Management Design 4.1 Design of Benchmark Loading Pattern with M5 AFA3G Fuel Based on CPR1000 PWR core, the equilibrium cycle used M5 AFA3G assemblies with 4.45% enrichment and batch refueling numbers of 48, 56, 64, 68 and 72. Following the general principle of low leakage pattern, the core adopts classical checkerboard arrangement, with the third and fourth reloading assemblies arranged in the outermost periphery. When the batch refueling number is large, a certain number of second reloading assemblies will be arranged in the outermost periphery of the core. The schematic diagram of core loading patterns is shown in Fig. 1.

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4.2 Design of the ATF Core The core main parameters of ATF fuel are the same as those of the benchmark. The selected ATF fuel cladding is SiC, and the pellets are enhanced UO2 (adding 10% volume fraction of BeO) ceramic pellets (hereinafter referred to as ATF fuel). Due to the addition of 10% BeO in the pellet, the U content in ATF fuel is reduced by 10% compared to the M5 AFA3G fuel adopted in the benchmark loading pattern. For the convenience of comparison, this paper selects two different enrichment of ATF fuel to design the scheme: (1) ATF fuel with enrichment of 4.45%, which is consistent with the enrichment in the benchmark; (2) ATF fuel with enrichment of 4.95%, which keeps the loading capacity of U-235 in the assembly nearly consistent with that in the benchmark. Similar to the benchmark loading pattern, ATF fuel loading pattern selects batch refueling numbers of 48, 56, 64, 68 and 72 for equilibrium cycle, which follows the general principle of low leakage pattern. The core adopts classic checkerboard arrangement, with the third and fourth reloading assemblies arranged in the outermost periphery. When the batch refueling number is large, a certain number of second reloading assemblies are arranged in the outermost periphery of the core. The schematic diagram of core loading pattern is shown in Fig. 1. 4.3 Fuel Management Calculation Results The core calculation results of the benchmark designed with the M5 AFA3G fuel with an enrichment of 4.45% are shown in Table 1, the calculation results of the ATF core loading pattern with an enrichment of 4.45% are shown in Table 2, and the calculation results of the ATF core loading pattern with an enrichment of 4.95% are shown in Table 3. Tables 1, 2 and 3 give the results of cycle length, number of batch refueling assemblies and discharge burnup of assembly of different core design schemes, and the safety parameters such as nuclear enthalpy rise factor (FH ), hot spot factor (FQ ), moderator temperature coefficient and shutdown margin meet the requirements of the design criteria in Sect. 3.1.

5 Fuel Management Economy Analysis Both of the cladding and pellet materials used in the fuel selection analyzed in this paper are new materials, which are still in the research and development stage, and the cost after industrial application in the future is still unknown. Therefore, according to the above design results of fuel management, from the aspects of utilization of core fuel and the number of assemblies consumed annually under different load factors, the comparisons of fuel batch discharging burn-up and uranium consumption under different fuel managements are given, and the number of assemblies can be saved annually after ATF fuel is given. 5.1 Batch Discharging Burn-Up Analysis Table 4 gives detailed data and comparison of uranium loading and U-235 loading of ATF core and benchmark core. Table 5 shows the comparison between the main calculation

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Fig. 1. Schematic diagram of core loading pattern Table 1. Main calculation results of benchmark loading pattern with different numbers of batch refueling using M5 AFA3G assemblies with 4.45% enrichment Cycle Length of cycle

Equilibrium cycle (EFPD)

502

483

463

420

377

(MWd/tU)

20,209

19,441

18,632

16,932

15,168

72

68

64

56

48

Average discharging burnup of assemblies

44.07

44.89

45.71

47.47

49.61

Max discharging burnup of assemblies

47.89

48.05

48.95

54.41

51.74

Max discharging burnup of rods

50.35

52.21

52.68

59.53

55.92

Critical boron concentration BOL, HZP, ARO (ppm)

2146

2106

1978

1899

1900

Max FH (HFP, ARO) (without uncertainty)

1.51

1.44

1.49

1.57

1.56

MTC (BOL, HZP, ARO) (pcm/°C, without uncertainty)

− 2.22

− 2.78

− 5.40

− 6.28

− 7.43

EOL shutdown margin (pcm)

2957

2982

2965

2941

3169

Number of batch refueling assemblies Discharging burnup (GWd/tU)-EOL

152

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Table 2. Main calculation results of different batch refueling numbers of ATF assembly with 4.45% enrichment Cycle Length of cycle

Equilibrium cycle (EFPD)

468

449

433

393

352

(MWd/tU)

20,869

20,014

19,288

17,508

15,707

72

68

64

56

48

Average discharging burnup of assemblies

44.82

46.23

47.31

49.08

51.38

Max discharging burnup of assemblies

49.77

50.66

50.51

56.15

53.09

Max discharging burnup of rods

52.88

56.67

54.41

61.90

58.24

Critical boron concentration BOL, HZP, ARO (ppm)

1820

1730

1782

1727

1731

Max FH (HEF, ARO) (without uncertainty)

1.49

1.50

1.49

1.56

1.57

MTC (BOL, HZP, ARO) (pcm/°C, without uncertainty)

− 0.04

− 3.32

− 1.66

− 2.30

− 3.18

EOL shutdown margin (pcm)

2901

2918

3071

3050

3267

Number of batch refueling assemblies Discharging burnup (GWd/tU)-EOL

results of equilibrium cycle with ATF fuel with different enrichment and with different batch refueling numbers. Comparing the design results of ATF fuel, it shows that when the number of batch refueling assemblies is the same, the increase of enrichment can significantly improve the batch discharging burnup of assemblies and improve the utilization rate of fuel; Comparing different batch refueling number strategies, when the assembly enrichment is the same, the batch discharging burn-up of assembly can be increased by about 1–5% for every 4 refueling assemblies reduction, and the number of batch refueling assemblies is smaller, the discharging burn-up of assemblies is higher, which means the higher fuel utilization rate. Combine Table 4 with Table 5, it shows that when using ATF fuel, the batch discharging burn-up of the core is improved while keeping the same enrichment or the similar initial loading U-235. The improvement of batch discharging burn-up can improve the fuel utilization rate, that is, the fuel economy can be improved. Under the condition that the enrichment is not improved (4.45% enrichment), the uranium loading of assemblies is reduced by 10%, the discharging burn-up of the ATF core is increased by about 1.5– 3.5% due to the superior neutron capability of SiC material and good moderation effect of BeO. When ATF fuel with 4.95% enrichment is selected, the initial U-235 loading capacity of the core is nearly the same (only 0.11% different from that of the benchmark core (Table 4)). Compared with ATF cores with different batch refueling numbers and

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Table 3. Main calculation results of different batch refueling numbers of ATF assembly with 4.95% enrichment Cycle Length of cycle

Equilibrium cycle (EFPD)

519

498

480

436

389

MWd/tU)

23,151

22,184

21,391

19,445

17,348

72

68

64

56

48

Number of batch refueling assemblies

Discharging burnup (GWd/tU)-EOL Average discharging burnup of assemblies

50.49

51.22

52.50

54.51

55.71

Max discharging burnup of assemblies

54.05

55.37

56.71

62.33

59.75

Max discharging burnup of rods

58.42

61.32

63.10

69.11

65.63

Critical boron concentration BOL, HZP, ARO (ppm)

2047

1908

2036

1988

1997

Max FH (HFP, ARO) (without uncertainty)

1.52

1.52

1.53

1.56

1.60

MTC (BOL, HZP, ARO) (pcm/°C, without uncertainty)

− 0.58

− 1.20

− 0.30

− 0.32

− 2.44

EOL shutdown margin (pcm)

4327

4130

4422

3987

3552

Table 4. Comparison of core uranium loading with M5 AFA3G benchmark loading pattern Core initial loading

Uranium (kg)

Uranium deviation (benchmark-ATF)/benchmark

U-235 (kg)

U-235 deviation (benchmark-ATF)/benchmark

M5 AFA3G-4.45% (benchmark)

72,158.77

/

3211.07

/

ATF-4.95%

64,942.89

10.00%

3214.67

− 0.11%

ATF-4.45%

64,942.89

10.00%

2889.96

10.00%

M5 AFA3G benchmark core, when the batch refueling number is the same, the batch discharging burn-up of assemblies can be increased by about 15%, and the utilization of the fuel has been greatly improved (Table 5). 5.2 Analysis of Uranium Consumption Table 5 shows the calculation results of uranium consumption per power generation of equilibrium cycles using ATF fuels with different enrichment at different batch refueling numbers, compared with the corresponding benchmark loading pattern of M5 AFA3G assemblies. Uranium consumption per power generation means the average consumption of nuclear fuel corresponding to every 1 kWh of electricity produced by nuclear power

154

T. Zou et al. Table 5. Comparison of main calculation results of different batch refueling numbers

72 - number of batch refueling assemblies Assembly type

Cycle length

Batch discharging burnup (GWd/tU)

Uranium consumption (1.0E−3gU/kWh)

Relative variation (ATF/benchmark-1) Batch discharging burnup

Uranium consumption

MWd/tU

EFPD

M5 AFA3G-4.45% (benchmark)

20,209

502

44.07

2.51

0.00

0.00

ATF-4.95%

23,151

519

50.49

2.19

0.146

− 0.127

ATF-4.45%

20,869

468

44.82

2.47

0.017

− 0.017

Batch discharging burnup (GWd/tU)

Uranium consumption (1.0E−3gU/kWh)

Relative variation (ATF/benchmark-1) Batch discharging burnup

Uranium consumption

68 - number of batch refueling assemblies Assembly type

Cycle length MWd/tU

EFPD

M5 AFA3G-4.45% (benchmark)

19,441

483

44.89

2.47

0.00

0.00

ATF-4.95%

22,184

498

51.22

2.16

0.141

− 0.124

ATF-4.45%

20,014

449

46.23

2.40

0.030

− 0.029

Batch discharging burnup (GWd/tU)

Uranium consumption (1.0E−3gU/kWh)

Relative variation (ATF/benchmark-1) Batch discharging burnup

Uranium consumption

64 - number of batch refueling assemblies Assembly type

Cycle length MWd/tU

EFPD

M5 AFA3G-4.45% (benchmark)

18,632

463

45.71

2.42

0.00

0.00

ATF-4.95%

21,391

480

52.5

2.11

0.149

− 0.129

ATF-4.45%

19,288

433

47.31

2.34

0.035

− 0.034

Batch discharging burnup (GWd/tU)

Uranium consumption (1.0E−3gU/kWh)

Relative variation (ATF/benchmark-1) Batch discharging burnup

Uranium consumption

47.47

2.33

0.00

0.00

56 - number of batch refueling assemblies Assembly type

M5 AFA3G-4.45% (benchmark)

Cycle length MWd/tU

EFPD

16,932

420

(continued)

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Table 5. (continued) 72 - number of batch refueling assemblies Assembly type

Cycle length

Batch discharging burnup (GWd/tU)

Uranium consumption (1.0E−3gU/kWh)

Relative variation (ATF/benchmark-1) Batch discharging burnup

Uranium consumption

MWd/tU

EFPD

ATF-4.95%

19,445

436

54.51

2.03

0.148

− 0.129

ATF-4.45%

17,508

393

49.08

2.26

0.034

− 0.033

Batch discharging burnup (GWd/tU)

Uranium consumption (1.0E−3gU/kWh)

Relative variation (ATF/Benchmark-1) Batch discharging burnup

Uranium consumption

48 - number of batch refueling assemblies Assembly type

Cycle length MWd/tU

EFPD

M5 AFA3G-4.45% (benchmark)

15,168

377

49.61

2.23

0.00

0.00

ATF-4.95%

17,348

389

55.71

1.99

0.123

− 0.109

ATF-4.45%

15,707

352

51.38

2.16

0.036

− 0.034

plants in a period of time, and the calculation method is as follows: Ur Ur = PH × EFPD PC × Ket × EFPD 1 1 1 Nb × Ma × = 41.67 × × = PC × EFPD Ket Bub Ket

RUr =

In the formula, RUr is uranium consumption per power generation (gU/kWh), Ur is nuclear fuel consumption, PH is nominal electricity power of nuclear power plant (MWe), PC is nominal core thermal power (MWt), Bub is batch average Burnup (MWd/tU), EFPD is equivalent full power day, Ket is the ratio of nominal electricity power to nominal core thermal power, Nb is number of batch refueling assemblies, and Ma is the mass of metallic uranium (t) of every Fuel assembly. It can be seen from Table 5 that after ATF fuel is used, the uranium consumption per power generation is reduced when the enrichment of pellets is equivalent to that of the benchmark core or the initial loading of U-235 is quite similar. The reduction of uranium consumption for power generation further indicates that the utilization rate of fuel has been improved, that is, the economy of fuel has been improved. Under the assumption that the enrichment is not improved (the 4.45% enrichment) and the uranium loading of assemblies is reduced by 10%, the uranium consumption of ATF core is reduced by about 1.7–3.4% due to the superior neutron capability of SiC material and good moderation effect of BeO, and the maximum reduction is achieved when the batch refueling number is 48 (Table 5). When ATF fuel with 4.95% enrichment is selected, the initial U-235 loading capacity of the core is similar to that of the benchmark core (Table 4). Compared with ATF cores with different batch refueling numbers and M5 AFA3G benchmark cores,

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when the batch reloading numbers are the same, the uranium consumption per power generation can be reduced by about 12.9%, and the utilization rate of the fuel has been greatly improved (Table 5). 5.3 Annual Consumption Assemblies Assessment According to the relevant data in Table 5, the calculation method of the number of assemblies consumed annually is adopted. Table 6 gives the number of assemblies consumed annually by the unit under different load factors with ATF fuels and M5 AFA3G fuels with different batch refueling N y = Nc 365η L strategies. Table 6 also gives the comparison of the number of assemblies consumed annually with the ATF cores and benchmark cores. The calculation method of the number of assemblies consumed annually is as follows: Ny = Nc

365η L

In the formula, Nc is the number of new fuel assemblies required for a cycle, L is the equivalent full power cycle length and η is the load factor. The plant load factor considers 90%, 85% and 80%. According to the results in Tables 4 and 6, when keeping the same enrichment, the uranium loading in the core will be reduced by 10%, but if the initial loading of the fissile nuclide U-235 in the core is nearly similar, it can save about 1.5 assemblies every year after using ATF fuel. When the plant load factor is 90%, it can be save at most 1.61 assemblies and at least about 1.29 assemblies with different batch refueling numbers; When the plant load factor is 85%, the maximum number of assemblies saved is about 1.52, and the minimum is about 1.22 with different batch refueling numbers; When the plant load factor is 80%, the maximum number of assemblies saved is about 1.43, and the minimum number is about 1.15 with different batch refueling numbers. According to the results in Tables 4 and 6, when keeping the same enrichment, both of the uranium loading in the core and the initial loading of fissile nuclide U-235 will be reduced by 10%. In this condition, the annual consumption of assemblies increases slightly after using ATF fuel. When the plant load factor is 90%, the number of additional assembly annual consumption is from 2.97 to 3.50; When the plant load factor is 85%, the number of additional assembly annual consumption is from 2.81 to 3.31; When the plant load factor is 80%, the number of additional assembly annual consumption assemblies is from 2.64 to 3.11.

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Table 6. Annual consumption assemblies with different batch refueling numbers 72 - number of batch refueling assemblies Assembly type

Cycle length

Number of annual consumption assemblies corresponding to different load factors

Variation of number of assemblies consumed annually

EFPD

90%

85%

80%

90%

85%

80%

M5 AFA3G-4.45% (benchmark)

502

47.12

44.50

41.88

0.00

0.00

0.00

ATF-4.95%

519

45.57

43.04

40.51

− 1.54

− 1.46

− 1.37

ATF-4.45%

468

50.54

47.73

44.92

3.42

3.23

3.04

68 - number of batch refueling assemblies Assembly type

Cycle length

Number of annual consumption assemblies corresponding to different load factors

Variation of number of assemblies consumed annually

EFPD

90%

85%

80%

90%

85%

80%

M5 AFA3G-4.45% (benchmark)

483

46.25

43.68

41.11

0.00

0.00

0.00

ATF-4.95%

498

44.86

42.36

39.87

− 1.39

− 1.32

− 1.24

ATF-4.45%

449

49.75

46.99

44.22

3.50

3.31

3.11

64 - number of batch refueling assemblies Assembly type

Cycle length

Number of annual consumption assemblies corresponding to different load factors

Variation of number of assemblies consumed annually

EFPD

90%

85%

80%

90%

85%

80%

M5 AFA3G-4.45% (benchmark)

463

45.41

42.89

40.36

0.00

0.00

0.00

ATF-4.95%

480

43.80

41.37

38.93

− 1.61

− 1.52

− 1.43

ATF-4.45%

433

48.55

45.86

43.16

3.15

2.97

2.80

56 - number of batch refueling assemblies Assembly type

Cycle length

Number of annual consumption assemblies corresponding to different load factors

Variation of number of assemblies consumed annually

EFPD

90%

85%

80%

90%

85%

80%

M5 AFA3G-4.45% (benchmark)

420

43.80

41.37

38.93

0.00

0.00

0.00

ATF-4.95%

436

42.19

39.85

37.50

− 1.61

− 1.52

− 1.43

ATF-4.45%

393

46.81

44.21

41.61

3.01

2.84

2.67

48 - number of batch refueling assemblies

(continued)

158

T. Zou et al. Table 6. (continued)

72 - number of batch refueling assemblies Assembly type

Assembly type

Cycle length

Number of annual consumption assemblies corresponding to different load factors

Variation of number of assemblies consumed annually

EFPD

90%

90%

Cycle length

Number of annual consumption assemblies corresponding to different load factors

85%

80%

85%

80%

Variation of number of assemblies consumed annually

EFPD

90%

85%

80%

90%

85%

80%

M5 AFA3G-4.45% (benchmark)

377

41.82

39.50

37.18

0.00

0.00

0.00

ATF-4.95%

389

40.53

38.28

36.03

− 1.29

− 1.22

− 1.15

ATF-4.45%

352

44.80

42.31

39.82

2.97

2.81

2.64

6 Summary According to the above results, after using ATF fuel, due to the small thermal neutrons absorption cross section of SiC cladding and good moderation effect of BeO, the thermal neutrons utilization of ATF fuel has been improved effectively. Although 10% BeO has been added to the pellet to reduce the uranium loading of the assembly, the batch discharging burnup of core fuel using ATF fuel is still higher than that of benchmark core using traditional M5 AFA3G assembly, and the uranium consumption is also lower than that of benchmark core using traditional M5 AFA3G assembly. So, the utilization of core fuel using ATF fuel is better. For the annual consumption assemblies, when the enrichment of ATF fuel is kept at 4.45%, the uranium loading in ATF core is reduced by 10% compared with that in benchmark core, the annual consumption assemblies increase slightly; When the enrichment of ATF fuel is increased to 4.95%, the uranium loading in ATF core is reduced by 10% compared with that in benchmark core, but the initial loading of U-235 in ATF core is equivalent to the benchmark core, the unit can save a small amount of fuel assembly every year.

7 Conclusion Based on the main parameters of the traditional CPR1000 PWR 18-month refueling mode, the fuel management design and three-dimensional core calculation of ATF fuel are completed with two ATF enrichment: 4.45% and 4.95%, and five batch refueling numbers: 48, 56, 64, 68 and 72. The results show that all the parameters of the fuel management with two enrichment and five batch refueling numbers meet the core design criteria, and can be used for subsequent analysis. Based on the batch discharging burn-up of core fuel, the consumption of uranium resources per unit power generation and the number of assemblies consumed annually

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under different plant load factors, a preliminary economic analysis of fuel management using ATF fuel is carried out to compare with the M5 AFA3G core with an enrichment of 4.45%. After using ATF fuel, because the thermal neutrons absorption cross section of SiC cladding is small and BeO can improve the moderation effect of neutron, the thermal neutrons utilization rate of ATF fuel has been improved. Although 10% BeO has been added to the pellet to reduce the uranium loading of the assembly, the batch discharging burn-up of ATF core is higher and the uranium consumption per power generation is lower, that is, the utilization of ATF fuel is better. For the annual consumption assemblies, since the uranium loading capacity of ATF core is reduced by 10% compared to that of benchmark core, it is necessary to increase the enrichment of ATF fuel to 4.95% to keep the initial loading capacity of U-235 equivalent to that of benchmark core, so that the unit can save a small amount of fuel assembly every year. Both the cladding (SiC) and pellet (UO2 (BeO)) used in the fuel selection analyzed in this paper are new materials, which are still in the research and development stage. On the premise of certain economy, whether the commercial application of the fuel can be realized in the future needs further study of the relevant capability of materials, especially the thermo-mechanical performance under different working conditions. Due to the limited space, this paper only gives the preliminary research results from the aspect of fuel management economy.

References 1. Snead, L.L., Nozawa, T., Katoh, Y., et al.: Handbook of SiC properties for fuel performance modeling. J. Nucl. Mater. 371(1–3), 329 (2007) 2. Chun, J., Lim, S., Chung, B., et al.: Safety evaluation of accident-tolerant FCM fueled with SiC-coated zircalloy cladding for design basis-accidents and beyond DBAs. Nucl. Eng. Des. 289, 287 (2015) 3. Snead, L.L., Zinkle, S.J., White, D.P.: Thermal conductivity degradation of ceramic materials due to low temperature, low dose neutron irradiation. J. Nucl. Mater. 340(2–3), 187 (2005) 4. Katoh, Y., Snead, L.L., Szlufarska, I., et al.: Radiation effects in SiC for nuclear structural applications. Curr. Opin. Solid State Mater. Sci. 16(3), 143 (2012) 5. Ishimoto, S., Hirai, M., Ito, K., et al.: J. Nucl. Sci. Technol. 33(2), 134–140 (1996) 6. Fink, J.K.: J. Nucl. Mater. 279, 1–18 (2000) 7. Fourcade, J., Sarma, K.H., Lee, S.G., et al.: Trans. Am. Nucl. Soc. 92, 179–180 (2005) 8. McCoy, K., Mays, C.: J. Nucl. Mater. 375, 157–167 (2008) 9. Naramore, M.J.: High Thermal Conductivity UO2 -BeO Nuclear Performance Assessments and Overview of Fabrication (2010) 10. Lu, H., Mo, K., Li, W., et al.: Development of self-reliant three-dimensional core nuclear design code COCO. At. Energy Sci. Technol. 47(suppl.), 327–330 (2013) 11. Lu, H., Jun, C., Wang, J., et al.: Verification and validation of self-reliant core nuclear design code COCO. At. Energy Sci. Technol. 51(8), 1459–1463 (2017)

Thermodynamic Analysis and Optimization of Nuclear Closed Air Brayton Cycle Huawei Fang1 , Jingwei Yi1 , Xiaoyu Zhang1 , Tiebo Liang1 , Yiran Qian2 , Xin Tang2 , and Weixiong Chen2(B) 1 Key Laboratory of Nuclear Reactor System Design Technology, Nuclear Power Institute of

China, Chengdu, Sichuan, China 2 State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University,

Xi’an, Shaanxi, China [email protected]

Abstract. In order to improve the flexibility and adaptability of micro nuclear power systems, air is considered as an optional working fluid of nuclear Brayton cycle. It has ability to avoid shortage of working fluid under extreme conditions, and ensure the system operating with tight supply. Compared with the traditional open cycles, the closed cycle has advantages of higher cycle efficiency, smaller equipment size and higher power density. Few of closed air power cycle in nuclear power system has been studied by now. In this paper, cycle parameters were analyzed and designed, and optimal cycle performance was compared between cycle constructions. Four closed air Brayton cycles are proposed and modeled in EBSILON 15.0, including simple regenerative cycle (SRC), intercooling regenerative cycle (IRC), reheat regenerative cycle (RRC) and intercooling-reheat regenerative cycle (IRRC). Thermodynamic analysis of the key parameters is conducted, including pressure ratio, turbine inlet temperature, compressor inlet temperature and pressure. In addition, the genetic algorithm (GA) is applied for multi-parameters optimization for the maximal cycle efficiency. Result shows that with the increase of pressure ratio, the efficiency of SRC increases firstly and then decreases. With the increase of turbine inlet temperature, the pressure ratio corresponding to the maximal cycle efficiency increases gradually. Specific work gradually increases with pressure ratio. In higher turbine inlet temperature, the rising velocity of specific work is faster. Cycle efficiency increases with compressor inlet pressure, and this trend tend to be slower in high pressure region. With the increasing of compressor inlet pressure, optimal pressure ratio decreases and finally approaches to 2.0. The cycle minimal cycle pressure was selected as 6 MPa. Maximal efficiency and corresponding pressure ratio increase with turbine inlet temperature increasing and decrease with compressor inlet temperature increasing. Comparing with SRC, RRC brings 1.16% cycle efficiency promotion, and IRC presents significant specific work increasing. In cycle efficiency distribution contour map of pressure ratio-middle pressure ratio, design point should be selected in high efficiency area. IRRC reaches 39% maximal cycle efficiency and maximal specific power 0.118 MW/(kg/s) of CABC. Keywords: Microreactor · Air Brayton cycle · Intercooling · Reheat · Pressure ratio © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 160–165, 2023. https://doi.org/10.1007/978-981-19-8899-8_15

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1 Introduction Microreactor is designed compact and transportable. Most microreactors can produce 1–20 MW of thermal energy that can be used directly as heat or converted to electric power [1]. Microreactor power conversion systems using closed Brayton cycle have been proposed for recent 10 years, such as Megapower, HOLOS and eVinci [2–4]. In order to improve the flexibility and adaptability of the microreactor power system, air is proposed as working fluid of Brayton cycle. Although the thermodynamic performance is inferior to SCO2, closed air Brayton cycle (CABC) has good environmental adaptability and high maturity of turbine design technology. It has ability to avoid shortage of working fluid under extreme conditions, and ensure the system operating with tight supply. Compared with the traditional open cycles, the closed cycle has advantages of higher cycle efficiency, smaller equipment size and higher power density. Feasibility of closed air cycle applied for solar power generation has been researched for several years [5], but few of CABC nuclear power system has been noticed yet. In this paper, effect of key parameters on performance of basic cycle construction is analyzed. Four different constructions of closed air Brayton cycle (CABC) are modeled and optimized for highest cycle efficiency.

2 Methodology As shown in Fig. 1, four types of systems have been researched in this paper: (a) simple recuperative air cycle (SRC), (b) intercooled recuperative air cycle (IRC), (c) reheat regenerative cycle (RRC), and (d) intercooling-reheat regenerative cycle (IRRC). Basic cycle configuration (SRC) consists of reactor, turbine, compressor, recuperator, cooler and generator. In IRC, an additional cooler is arranged between two stage compressors to reject the heat generated in compression process, which can reduce compression work of the second stage compressor. In RRC, air flow intends reactor again and heated to promote power output of second turbine. IRRC is the most complicated configuration which combines intercooler and reheater with SRC. In cycle design process, reactor operating temperature and environment temperature are assumed constant. A 5 MWt microreactor is selected as heat source. Pressure loss of heat exchangers are considered and other components design parameters maintain unchanged. Thermodynamic model of SRC is established to determine CABC basic design parameters. Effect of key design parameters on cycle performance are analyzed, including pressure ratio (PR), turbine inlet temperature (TIT), compressor inlet temperature (CIT) and pressure (CIP). Cycle efficiency and specific work are calculated to evaluate cycle performance. Cycle efficiency is a ratio of electricity generator output power to reactor heat power. Specific work is a ratio of electricity generator output power to cycle mass flow. Cycle boundary conditions and a part of basic design parameters are listed in Table 1. It can be predicted that intercooling and reheat can intensify cycle performance. Additional components bring more thermal parameters, such as intercooler inlet pressure and reheat pressure. Optimal design parameters are determined by Genetic algorithm (GA) multi-parameter optimization for highest cycle efficiency. Thus, the highest efficiency and corresponding design parameters combination of four cycles configurations in

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different turbine inlet temperatures are calculated and compared. Thermodynamic analysis, parameters optimization and performance comparison are calculated on EBSILON Professional 15.0 platform.

Fig. 1. Four CABC cycle configurations (a) SRC (b) IRC (c) RRC (d) IRRC

Table 1. Cycle design parameters and boundary conditions Parameters

Valve

Parameters

Valve

Turbine efficiency

0.88

Pressure loss of recuperator [MPa]

0.05

Compressor efficiency

0.8

Turbine inlet temperature [°C]

700

Recuperator effectiveness

0.95

Compressor inlet temperature [°C]

35

Generator efficiency

0.986

Reactor power [MW]

5

Pressure loss of main heater [MPa]

0.1

Compressor inlet pressure [MPa]

6

Pressure loss of cooler [MPa]

0.05

3 Results and Discussion Thermodynamic analysis of SRC is conducted at first. Cycle parameters are consistent with Table 1 except variables. Figure 2(a) and (b) shows the effect of pressure ratio (PR) on cycle efficiency and specific work in 500 °C, 700 °C and 900 °C turbine inlet

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temperatures (TIT). Figure 2(a) indicates that with the increasing of pressure ratio, cycle efficiency firstly increases and then decreases. Obviously, an optimal pressure ration exists which is corresponding to maximal cycle efficiency. With the increase of turbine inlet temperature, the pressure ratio corresponding to the maximum cycle efficiency gradually increase. In Fig. 2(b), the specific work gradually increases with pressure ratio. In higher turbine inlet temperature, the rising velocity of specific work is faster. All of this means that cycle with high pressure ratio needs less air working fluid filled in pipes and components, and system size should be smaller. Unlike open air turbine cycle, a CABC should has an accurate design compressor inlet pressure (CIP) above atmospheric pressure. Figure 2(c) shows that compressor inlet pressure has a remarkable influence on maximal cycle efficiency and optimal pressure ratio. Figure 3(a) further shows that cycle with 8MPa CIP has approximately 4.5% promotion of maximal cycle efficiency. With the increasing of CIP, optimal pressure ratio decreases and finally approaches to 2.0. As CIP above 6MPa, the promotion of cycle efficiency declines and nearly proportionally increases with CIP. Thus, performance promotion by CIP increasing has limitation, and cycle lowest pressure is selected as 6 MPa. Reactor operating temperature and environment temperature restrict cycle highest and lowest temperature range, which limits cycle performance. Figure 3(b) and (c) shows effect of turbine inlet temperature and compressor inlet temperature on cycle efficiency maximum and corresponding optimal pressure ratio. Efficiency reaches 35% when TIT equal to 660 °C, and 40% when TIT equal to 770°C. Maximal efficiency and corresponding PR increase with increasing TIT and decrease with increasing CIT. In basic cycle design, TIT and CIT are selected as 700°C and 35 °C respectively.

Fig. 2. (a), (b) Influence of PR on SRC performance. (c) Influence of compressor inlet pressure on SRC efficiency.

In IRC and IRRC, intercooler is added between two stages of compressors. Ratio of the first compressor outlet pressure and inlet pressure is named middle pressure ratio (MPR), which should be considered in thermodynamic analysis. Figure 4(a), (b) shows cycle efficiency distribution contour map of PR-MPR. There has a high efficiency area between PR 2.0–3.5 and MPR 1.0–2.0, which is optimal design point. An optimal group of PR-MPR makes efficiency maximum 37.25%. Similar shape curves of PR-MPR appears in RRC and efficiency maximum is 38.16%. GA algorithm is used to optimize design parameters of four cycles configurations, and results are summarized in Table 2. IRRC reaches 39% maximal cycle efficiency and maximal specific

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Fig. 3. (a)–(c) Influence of CIP, TIT and CIT on SRC highest efficiency and optimal pressure ratio.

Fig. 4. (a), (b) Influence of PR-MPR on IRC and RRC performance. (c) CABC configurations maximal efficiency comparison in different pressure ratio.

power 0.118 MW/(kg/s) of CABC, because intercooling and reheat brings significant performance promotion. Figure 4(c) shows maximal efficiency against pressure ratio of 4 CABC configurations, in which IRRC has a high and flat efficiency curve in high PR range.

4 Conclusion In this paper, we establish thermodynamic system model and analyze key parameters influence on CABC performance. SRC, IRC, RRC and IRRC configurations are optimized and compared. All the results can be reference of CABC design and parameters selecting. The main conclusions of this paper are as follows: (1) With the increase of pressure ratio, the SRC cycle efficiency firstly increases and then decreases. With the increase of turbine inlet temperature, the SRC pressure ratio corresponding to the maximal cycle efficiency gradually increases. Specific work gradually increases with pressure ratio. In higher turbine inlet temperature, the rising velocity of specific work is faster. (2) Cycle efficiency increases with compressor inlet pressure, and this trend tend to be slower in high pressure region. With the increasing of compressor inlet pressure, optimal pressure ratio decreases and finally approaches to 2.0. Maximal efficiency and corresponding pressure ratio increase with turbine inlet temperature increasing and decrease with compressor inlet temperature increasing.

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(3) Comparing with SRC, RRC brings 1.16% cycle efficiency promotion, and IRC presents significant specific work increasing. In cycle efficiency distribution contour map of pressure ratio-middle pressure ratio, design point should be selected in high efficiency area. IRRC reaches 39% maximal cycle efficiency and maximal specific power 0.118 MW/(kg/s) of CABC.

Acknowledgments. This work was financially supported by the Pioneering and Innovative Scientific Program of China National Nuclear Corporation and Operation Fund of Key Laboratory of Nuclear Reactor System Design Technology.

References 1. Office of Nuclear Energy: What is a Nuclear Microreactor? (2021). https://www.energy.gov/ ne/articles/what-nuclear-microreactor 2. McClure, P., Poston, D., Rao, D.V., Reid, R.: Design of Megawatt Power Level Heat Pipe Reactors. Los Alamos National Laboratory (2015) 3. Filippone, C., Jordan, K.A.: The Holos Reactor: A Distributable Power Generator with Transportable Subcritical Power Modules. Holos Generators (2017) 4. Arafat, Y., Wyk, J.V.: eVinciTM micro reactor: our next disruptive technology. Nucl. Plant J. (2019) 5. Rovense, F., Amelio, M., Scornaienchi, N.M., Ferarro, V.: Performance analysis of a solaronly gas micro turbine, with mass flow control. In: 72nd Conference of the Italian Thermal Machines Engineering Association (2017)

Affecting Factors and Aspects for Improving Customized Nuclear Public Acceptance Strategies Haomiao Lin(B) China National Nuclear Corporation Overseas Ltd., Beijing, China [email protected]

Abstract. Nowadays, the issues of climate change is moved from margins to the mainstream. The vital role of nuclear energy in the global fight against climate change has been significantly highlighted by IAEA at COP26 last year. As a low carbon base-load electricity provider, nuclear power can become progressively an important option of countries for achieving its sustainable development goals. Comparing with traditional energy supply, nuclear power projects are characterized by complex technology, long construction period, large capital demand and high sensitivity. Currently, the international nuclear power demand market is mainly concentrated in the emerging countries, most of which have a poor industrial infrastructure, “not in my backyard” phenomenon is prominent, which is seriously interrupting the government’s energy development decision. The current world nuclear power market can be generally divided into two categories, i.e. the market in developed countries (Category I), where the demand for energy has become saturated and the demand for nuclear power mainly comes from the extension of old units and the replenishment of units decommissioning. These countries have developed nuclear power for tackling climate change, and most of them are beneficiaries of nuclear power, with relatively sound nuclear industrial system, high public support and strong public awareness. Such markets attach great importance to the low-carbon energy sources with relatively stable demand for electricity. They pay more attention to the privileges of nuclear energy compared with intermittent renewable energy as well as its shortcomings and solutions. The other major market for nuclear power in the world is emerging economies and developing countries (Category II) that actively seek energy transition. These countries generally attach great importance to the strategic significance of nuclear energy and fully realize the opportunities that nuclear energy brings to national economic development, industrialization and employment. In addition to the countries who developed nuclear science and technology, most of these markets are newcomer countries, whose systemic experience is insufficient. This paper takes number of typical countries as examples to analyze the issues affecting public acceptance in different markets. Through the analysis of the external environment of different markets and the influencing factors of public awareness, it can be seen that although these countries are at different stages of nuclear energy development, factors affecting public acceptance have common characteristics. i.e. nuclear energy or intermittent renewable energy, institutional

© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 166–175, 2023. https://doi.org/10.1007/978-981-19-8899-8_16

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guarantee, nuclear safety and information transparency. Then, this paper puts forward suggestions from the aspects of energy transition cognition, nuclear regulation, technology management and public communication, further to provide reference for improving public acceptance. Keywords: Nuclear · Public · Safety · Regulation · Characteristic

Foreword With the increasingly severe environmental problems and the increasingly tight energy supply, as a kind of efficient and clean energy, the development of nuclear power has become an important energy strategy in various countries. However, the development of nuclear power requires advanced and safe nuclear power technology, favorable financing conditions and mature nuclear power construction and management experience, communication with the public in particular. Currently, the international nuclear power demand market is mainly concentrated in the emerging countries, most among which have poor nuclear industrial infrastructure and lack of talent reserve. Additionally, the “not in my backyard (NIMBY)” phenomenon is prominent, which leads to problems such as wavering decision-making and unclear planning of nuclear power development. At present, nuclear power enterprises are actively developing the international nuclear power market by carrying out cooperation on public communication, nuclear science popularization and technical exchanges with emerging nuclear power countries. From the perspective of the growing public awareness demand in market countries, enhancing public understanding and establishing effective communication mechanisms have become an important starting point for market development and international cooperation, which are of great significance to enhance mutual trust and create a good market environment.

1 Characteristics of Public Perception in the International Nuclear Power At present, a total of 28 countries have expressed interest in nuclear power and are considering, planning or actively working to include it into their energy mix [1]. According to the world Nuclear Association, the main growth of nuclear power will come in countries where the technology is already well established in the foreseeable future. However, in the longer term, the trend to urbanization in less-developed countries will greatly increase the demand for electricity, and especially that supplied by base-load plants such as nuclear. Thus, countries in the Asia, Middle East and Africa have shown the great potentials in pursuing nuclear power for the first time [2] (Fig. 1). 1.1 Categories of Nuclear Power Markets Worldwide Today’s world nuclear power market can be generally divided into two categories, i.e., the market in developed countries (Category I), where the demand for energy and power has become saturated and the demand for nuclear power mainly comes from the extension

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Fig. 1. Number of countries in different regions are considering, planning or starting nuclear power programmes

of old units and the replenishment of units decommissioning. These countries have developed nuclear power for net-negative emissions, and most of them are beneficiaries of nuclear power, with relatively sound nuclear industrial system, high public support and strong public awareness. Such markets attach great importance to the low-carbon development of energy with relatively stable demand for electricity. They pay more attention to the privileges of nuclear energy compared with intermittent renewable energy as well as its shortcomings and solutions. Therefore, to improve public awareness, it is necessary to face up to the advantages and disadvantages of nuclear energy, quantify the problems and propose solutions. The other major market for nuclear power in the world is emerging economies and developing countries (Category II) that actively seek energy transformation. These countries urgently need to develop nuclear power in terms of strategic development, economic energy diversification, and filling the gaps of electricity supply. Such markets generally attach great importance to the strategic significance of nuclear energy and fully realize the opportunities that nuclear energy brings to national economic development, industrialization and employment. In addition to the countries who developed nuclear science and technology, most of these markets are newcomer countries, whose systemic experience is insufficient. The governments are more focused on how to ensure the safety of nuclear power construction and operation, how to address the project funds and how to build human capacity when elaborating a nuclear power project, and the phenomenon of “NIMBY” also hinders government decision-making. Therefore, for such markets, it is necessary to further strengthen public awareness of nuclear safety, technical economy, environment and risk, build a dialogue platform to improve public acceptance. To sum up, although the market categories are different, they all inevitably face challenges from public acceptance. It is not difficult to find out that public awareness level and public concerns vary with different national conditions, systems and development stages of nuclear energy in market countries. This paper takes UK, France, Kazakhstan, Vietnam and other typical countries as examples to analyze the public acceptance of nuclear energy and the main influencing factors in different markets.

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1.2 Characteristic of Public Perception in Developed Countries The UK today became the first major economy in the world to pass laws to end its contribution to global warming by 2050. Like most countries, it has met with different opinions on the development of nuclear power. Public interests in renewable energy was consistently higher than nuclear power in many polls in this early century in the UK. In the face of civil anti-nuclear, NGO pressure and the NIMBY trend, the UK government further captured public concerns through investigations, interviews to understand the main factors affecting public acceptance: one reason is lack of trust in nuclear power with false information and risks disclosed by governments and regulatory body; the other reason is the cost of nuclear power project has risen due to stricter safety standards, meanwhile the cost of renewable energy encounters a sharp drop which has led to doubts about the economic benefits of a nuclear power project. Nuclear power takes the largest share in electricity generation in France, the majority of the country’s electricity comes from nuclear sources and it is the world’s largest net exporter of electricity, and its carbon emissions are also much lower than those of other developed countries. As one of main nuclear countries without major accidents for many years in a row, France has long adhered to the standardization and serialization of nuclear power technology, and has maintained a good record of economy and safe operation. Its open and transparent information sharing system has also won long-term support and understanding from the public. However, in recent years, a sample survey conducted by its relevant departments shows that the public is increasingly concerned about the reprocessing of spent fuel, which has gradually become a potential factor affecting public acceptance. 1.3 Characteristic of Public Perception in Nuclear Emerging Countries For emerging countries with the needs of developing nuclear power, feasibility studies, infrastructure level and investment assessments and human capacity building are crucial for a nuclear power project, they also need to give full consideration to public acceptance as well. Public awareness in such countries is still poor with more focuses on nuclear radiation and safety, environmental impact and other aspects. Kazakhstan, a central Asian country with abundant energy reserves, is seriously considering the possibility of building two large commercial nuclear power units to further optimize its energy structure and realize the transition to a green economy. In this regard, the government and KNPP(the owner company) are intensively carrying out feasibility studies to develop its first NPP while one of the great difficulties is public’s opposition. Thus, the Government declared that it needs to hold public hearings and listen widely to the public before making a further step decision. Vietnam’s nuclear ambitions break a silent wait-and-see situation to nuclear power in southeast Asia. In order to ensure the security of future energy supply and break the dilemma of lacking traditional energy import channels and supply shortage, Vietnam turned its attention to nuclear energy. Vietnam has also developed a series of governmentled nuclear power publicity programs to promote nuclear energy to the public and actively build nuclear power talent programs by sending personnel abroad for training prgoramme [3]. However, the government finally confirmed the cancellation of the project last year,

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citing economic reasons, since all preparation work was ready to go. The government is planning to invest in 6000 MW of coal-fired and gas-fired power stations by 2030 as an alternative and considering to further expand electricity imports from China, Laos and Malaysia. In addition, Malaysia, once a front runner in nuclear energy in Southeast Asia, has abandoned nuclear power due to the risks of nuclear radiation, radioactive waste management and terrorist attacks etc., and has turned to renewable energy. Through the above cases, it is not hard to tell that no matter what level of development the country is in, the decision-makers have to carefully weigh the pros and cons of intermittent renewable energy versus nuclear science and technology, and should be objective and transparent when addressing public concerns, and extensive listening to public opinions in diverse ways is also helpful to enhance public understanding (Table 1).

2 Good Practices and Strategies for Improving Public Acceptance Through the analysis of the external environment of different types of nuclear power market and the influencing factors of public awareness, it can be seen that although these countries are at different stages of nuclear energy development, factors affecting public acceptance have common characteristics. That is to say, the public generally pays more attention to the comparison between nuclear energy and intermittent renewable energy, institutional guarantee, nuclear safety and information transparency. Based on the above cases, this paper puts forward suggestions from the aspects of energy transition cognition, nuclear regulation, technology management and public communication, further to providing reference for improving public acceptance. 2.1 Focus on the Analysis of Energy Transformation to Enhance the Public’s Rational Understanding of Nuclear Energy The nuclear market is being squeezed from intermittent low-carbon renewable energy, whether from the perspective of fighting climate change or the cost of generating electricity. The public and stakeholders will show their strong interests in renewable energy instead of nuclear if without strong government support. As it turns out, the choice to develop intermittent renewable energy is mostly based on the short-term energy transition goal to fill the low-carbon energy gaps quickly. However, short-term-oriented decisionmaking will mislead the country’s energy transformation strategy, and there are practical problems in the energy structure. For example, intermittent renewable energy (such as wind and solar) is highly dependent on external energy supply, and fossil energy is often used as its backup power, additionally its intermittent characteristics are difficult to maintain the balance between supply and demand of power grid, thus increasing the burden of power grid. In energy transformation, the stability of long-term energy supply should be taken into consideration, as well as economic objectives and technical security. The benefits of nuclear energy lie in the fact that it does not need to store electric energy, and that can provide a large amount of base charge to ensure the security of large and interconnected power grids. In addition, nuclear energy has no dependence on

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meteorological conditions and can steadily and continuously supply power to meet the increasing demand for electricity. In light of above cases, a lack of energy knowledge can be seen when the public considering energy transition. Under these circumstances, it is suggested that relevant authorities should further strengthen education and communication with the public on energy issues based on reality, optimize communication Table 1. An overview of public influencing factors in different categories of markets Country category

Country description

Characteristics of external environment

Major public concerns

I

Developed countries with a clear need for nuclear power development, such as France and the UK

- Political stability - High public acceptance - High nuclear cognition - Nuclear facilities are open to the public in a safe and secure way - Abundant nuclear talent - Complete nuclear industrial chain - Independent public communication agency - High corporate social responsibility - High NGOs influence - Strong awareness of environmental protection - Seeking low-carbon clean energy (to combat climate change)

- Security of energy supply - Electricity price stability - Information transparency - Radioactive waste management and disposal - Environmental impact

II1

Developing countries with a clear need for nuclear power development, such as Saudi Arabia, Kazakhstan, Egypt, South Africa, etc.

- Political system is relatively stable - Nuclear cognition is unbalanced - The regulatory system is relatively sound - A certain reserve of nuclear talents - Public communication agencies exist - Seeking transformation of energy structure (National Energy Strategy)

- Security of energy supply - Electricity price advantage - Radioactive waste management and disposal - Environmental problems - Nuclear technology - International supplier selection

(continued)

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Country category

Country description

Characteristics of external environment

Major public concerns

II2

Developing countries that are willing to develop nuclear power, such as the Philippines and Indonesia, Ghana, etc.

- Fragile political system - Low public acceptance - The regulatory system is not sound - Shortage of nuclear talents - Poor infrastructure - Low level of industrialization - Public communication agency (not exist) - Seeking to fill energy sources gaps (electricity shortages)

- Nuclear accident - The radiation - Energy selection - Environmental problems - Project funding - Employment security

strategies and publicity guidance, enable public to make rational and objective analysis and judgment, so as to reach a final consensus. 2.2 Focus on Improvement of Nuclear Legislation and Consolidation on the Responsibilities of All Stakeholders Nuclear safety is an important basis for achieving public acceptance, nuclear regulation and supervision of nuclear energy and radiation protection is an important guarantee of nuclear energy safety [4]. A sound regulatory system will fully implement the whole process of nuclear energy development and effectively enhance public confidence. The legislation of nuclear safety regulations appears like a pyramid hierarchy: Law (top), Government regulations, Department rules, Safety guides and technical documents (bottom) (Fig. 2). In recent years, facing the problems of lack of laws and scattered regulations, China has been vigorously accelerating the formulation of a series of nuclear laws and regulations, such as the Atomic Energy Law, Regulation on the Nuclear Damage Compensation, Regulation on Safety Management of Radioactive Waste, and Regulation on Nuclear Emergency Preparedness and Response, etc. The white paper on Nuclear Safety in China stated the need to establish a mechanism for extensive public participation, and the further disclosure of nuclear safety information, popularization of nuclear science and public participation are also formulated to ensure the public’s right to know, to participate and to supervise. Practice shows that this kind of comprehensive nuclear safety supervision is effective, and clear responsibility from government to stakeholders, local governments and the public where nuclear projects are located also plays a good role layer by layer in enhancing trust [5].

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Fig. 2. The hierarchy of legislation in China

3 Enhancement of Technology and Management to Ensure Steady and Sustainable Development of Nuclear Energy High cost is one of the obstacles to the development of nuclear power, which is undoubtedly an important factor to influence government and the public’s decision. The United States has adopted a series of preferential policies to encourage the development of nuclear power, including increasing taxes on other energy sources and lowering taxes on nuclear power, better yet France relies on the its standardization of technology and remarkable nuclear engineering management. The core value is that the French government only authorizes EDF to be the sole owner, operator and the overall engineering management unit of all domestic nuclear power plants. This mode can ensure that the experience in construction and operation is constantly fed back to the design and manufacturing departments, thus to promote the design work and the equipment quality to be continuously optimized. The effective operation of this mode has not only enabled France to rank among the lowest industrial and household electricity prices in the world, but also increased public trust in French nuclear authorities [6]. In contrast, some countries have different nuclear power technologies with scattered owners and operators, which cause the experience cannot be shared in a closed way with poor scale effect, so that the construction and operating costs are relatively high. In this regard, it is suggested that the mid-and-long term strategy of nuclear power development should be fully considered while selecting nuclear power technology, that will lay a solid foundation to steadily carry out the serialization and standardization. Additionally, benchmarking of advanced engineering management mode is also a privilege to obtain management tools and skills to ensure the safe and efficient development of nuclear power, and to enhance public acceptance.

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4 Promotion of Education and Diversification of Communication Channels Public access to open, transparent, accurate and structured information will greatly contribute to their understanding of nuclear energy. In aspect of publicity, it is suggested to build multi-dimensional public communication channels from three levels: government, social organizations and nuclear-related enterprises. At the government level, the public can become more familiar with nuclear energy and get more engagement through press conferences, public consultation and popular science education to receive reliable information. At the level of social organizations, nuclear societies, associations and other organizations can fully integrate nuclear-related resources from the government, universities, enterprises, think tanks, etc. to put spotlights on the focus issues of public concern, e.g., energy security, potential risks, environmental impact, nuclear radiation and spent fuel management, etc., through holding exhibitions, expert interviews, technical exchanges and dialogues, to achieve full communication with the public and share solutions. At the enterprise level, enterprises should pay attention to the performance of the corporate social responsibility, in addition to information disclosure, site visits, publicity to raise public awareness, a social economic planning and contribution to local industrial activities, economy and employments of a nuclear power project are also a leverage to get public support.

5 Conclusion Climate change is the greatest environmental challenge of our times. The dominant mindset of most countries to tackle with climate change is to change the current electricity system by replacing those fossil-fuel plants with clean generation. As a resilient energy, nuclear power can help to decarbonize hard to abate sectors such as transportation and industry and play a significant role in 2050 carbon neutrality, including by producing lowcarbon hydrogen. Furthermore, its flexibility and reliability can also stabilize electrical grids and ensure the security of electricity supply. For these reasons, most countries are actively saving a seat for nuclear at the table. Nuclear safety and public acceptance are widely recognized as the two major factors for a sustainable and healthy development of a nuclear industry. A high degree of nuclear safety is an important foundation for public acceptance, under which a sound regulatory system and a firmed nuclear energy development strategy are essential to define stakeholders’ responsibilities. In addition, improvement of technology and management level and diversification of communication methods and channels could also be effective means to address public concerns and promote the sustainable development of nuclear energy.

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References 1. IAEA, GOV/INF/2021/32-GC (65)/INF/6. International status and prospects for Nuclear Power 2021 2. WNA, Information Library, Emerging Nuclear Energy Countries 3. Nguyen Viet Phuong, Public Communication Programme for the first NPP (2012) 4. IAEA-TECDOC-1076: Communications on Nuclear, Radiation, Transport and Waste Safety: A Practical Handbook (1999) 5. IAEA No. NG-T-1.4: Stakeholder Involvement Throughout the life Cycle of Nuclear Facilities (2011) 6. EPR-Public Communication: Communication with the Public in a Nuclear or Radiologic Emergency (2012)

Initial Response Study of Nuclear Power Plant Operations After Local COVID-19 Outbreak Li Lianhai, Ren Xiaojiang, Zhang Xianggui, Wu Wenqi(B) , and Yuan Xia Jiangsu Nuclear Power Corporation, Lianyungang, China [email protected]

Abstract. Safety is the lifeblood of the nuclear industry. When the local COVID19 epidemic hit, it was particularly important for the nuclear power plant (NPP) operations department to quickly formulate and implement the initial response measures for epidemic prevention and control to ensure the safe and stable operation of nuclear power units, and also to obtain initiative in follow-up epidemic control and carry out various safety production work. The initiative of the arrangement can mitigate the impact of the epidemic. This paper mainly expounds the six “level 0” response actions and six “level 1” response actions, a total of twelve priority response actions formulated by the operations department. There was zero Covid-19 infection in the core area of the NPP’s sealed control area by means of racing against time, and quickly establishing an epidemic prevention and power protection system by the NPP operations department, and thus ensuring the safe, stable, efficient and environmentally friendly operation of the NPP units. Keywords: Crew safety · Personnel safety · Divisional management and control · Epidemic prevention and power protection · Key-position · Comprehensive prediction

Preface On March 5, 2022, a local case of COVID-19 was confirmed positive in the city where a nuclear power plant is located in China. After the outbreak of COVID-19, the local government took decisive measures and quickly launched a number of administrative measures including “regional control and restrictions” to prevent the spread of the epidemic, but it also had an impact on the daily operation order of the nuclear power plant. When the local COVID-19 hit, personnel from most NPP departments were able to work from home and reduced the negative impact of the epidemic. However, it was necessary to fully understand the special attributes of the NPP operations department, especially the shift personnel who must always carry out work such as monitoring, inspection, alarm processing or response to transient of the unit to guarantee nuclear safety. At the same time, the technical specifications for nuclear power plants approved by the National Nuclear Safety Administration (Chinese Nuclear Regulator) also clearly stipulate the minimum number of the operations department’s shift personnel. The epidemic was an order, and prevention and control were a responsibility. This nuclear power plant was the first nuclear power plant to encounter the local COVID-19 © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 176–187, 2023. https://doi.org/10.1007/978-981-19-8899-8_17

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outbreak, and there was no other domestic power plant’s successful disposal experience that can be used for reference. Starting from the work attributes of the operations department, the NPP initiated a number of epidemic prevention and control measures, and successfully finished the task of “epidemic prevention and power protection”. The initial response of the operations department was sorted out and formulated during the epidemic prevention and control period. Namely measures including six “level 0” and six “level 1” response actions to effectively ensure the safe, reliable, efficient and environmentally friendly operation of the NPP units were implemented during the epidemic according to the logical order of priorities and the daily work division of the management of the operations department (Fig. 1).

Fig. 1. The nuclear power plant’s “epidemic prevention and power protection” system after the local epidemic breakout

1 Unit Safety Control 1.1 Operations Scheduling Adjustment Due to the characteristics of the COVID-19 epidemic, in the early stage of the outbreak, more people worked from home, or worked in closed areas and control areas. How to improve the efficiency of work collaboration in such a situation of extreme dispersion of personnel became an urgent problem.

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(1) After the outbreak of the epidemic, the on-duty shift supervisor immediately started the production on-call standby, and at least one person in charge from the relevant departments in the production field came to the plant site on-call standby, who would efficiently organize and coordinate the work in his department. (2) Confirm the availability of the online office system as soon as possible to ensure that all personnel in key positions in scheduling can use the online office system smoothly. (3) The morning meeting in the production field was held online, so that the personnel of all departments in the production field were able to grasp the continuous status of the unit. After the meeting, the participants shared the crew status by the social platform app timely. When the unit’s parameters deviated or had important defects, the on-duty shift supervisor should immediately notify the person in charge of the operations department, organize and coordinate the maintenance department in time to ensure the nuclear safety. (4) Establish a mechanism for personnel in key positions to hold online meetings regularly. Weekly regular tasks were deployed on online meetings, and decision-making meetings were held online for emergencies to conduct full discussions and control risks. (5) During the epidemic period, the online work ticket process was activated according to the requirements that the operators in the main control room did not meet the field operators, and the isolation manager did not meet the work leader. 1.2 Operations Plan Adjustment The operations department organized and discussed work plans for operations and preventive maintenance, and evaluated the recent periodic tests to minimize the number of operations in the early stage of the epidemic. (1) After evaluation and reference to historical experience data, it was compulsive to minimize the number of high-risk operations in the early stage of the epidemic. (2) According to the number of maintenance personnel on site, rationally reduce unnecessary preventive maintenance work, and reduce the number of work orders during the epidemic by appropriately adjusting equipment maintenance windows, and changing maintenance contents. (3) For non-QSR periodic tests, predict equipment status and performance curves, and cancel unnecessary equipment switching; for QSR periodic tests, make full use of the relevant provisions of the test cycle, and reasonably adjust the test within the allowable grace period of the cycle time, to prevent potential risks due to lack of manpower in operations and maintenance departments. 1.3 Operations Risk Control For risk control of operating units, predict in advance and conduct reasonable management to safeguard the safety of the units.

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(1) Identify the potential risks in each work ticket of the operating units, conduct risk assessment and take mitigation measures, implement double approval. Namely an operations engineer would approve the work ticket online, and check all the implementation conditions of work tickets meet the requirements. Then another operations engineer would independently review whether the operation risk had been identified, whether the risk analysis was complete, and whether the measures taken could effectively alleviate the operation risks. (2) For the operation work with medium and high risks, observation and coaching was adopted. Any operation conducted in the main control room was supervised by Deputy Shift Supervisor. Operation involving high risks was observed and coached by any person in charge of the operations department. Operation involving medium risks was observed and coached by any section chief from the operations department with support from operation support engineers and operation technical engineers. The level of management and control measures for medium and high-risk work was in place to ensure the correctness of operation. (3) Actively respond to transients and unexpected equipment defects of the unit. The relevant production departments provided a list of emergency standby personnel in various positions during the epidemic. The process for dismantling insulation, scaffolding and other support work was optimized to ensure that there was enough manpower for emergency maintenance. 1.4 Unit Defects Assessment After the epidemic outbreak, the person in charge of the operations department should quickly organize the evaluation of the status of the operating units, sort out the current defects of the units and constraints of important defects affecting the production, and address them at different levels. (1) As for the progress of the major defects handling, evaluate whether the defects handling would be affected by personnel changes, and whether the delay of the defects handling would cause the downgrade of the unit safety. Take temporary control measures, negotiate with relevant production departments, establish a maintenance rescue team, and ensure that the rescue team was staffed by experienced and skilled personnel to ensure high-quality handling of unit defects. (2) For the important deficiencies that have not yet been carried out, evaluate the impact of deficiencies on the unit, whether it will lead to the degraded operation of the unit, or exceed the time limit of the technical specifications. Conduct special discussions or formulate special plans, and take appropriate actions to mitigate the impact on the unit. If it is impossible to eliminate the defect on a daily basis after professional evaluation, a temporary mitigation measure will be formulated. (3) For the emergency defect of the unit, which may lead to the deviation of the unit’s state, use the existing production standby and emergency duty management process to establish an emergency defect elimination team, stabilize the staffing, and require them to be stationed at the NPP 24 h. The above was done to ensure that personnel were safe from epidemic and were ready to handle defects during emergency.

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1.5 Production Material Guarantee After the epidemic outbreak, the person in charge of the operations department should immediately dispatch the on-duty operations shift to check the inventory of the materials required for the daily operation of the unit, such as the inventory of bulk chemicals, oil, etc., and estimate how long the unit can operate based on the current consumption. The material administrator of the operations department immediately consulted with the commercial department to verify whether the stock of production materials was sufficient according to the actual situation of the unit. For materials that were estimated to be insufficient within half a month, full consideration should be given to the traffic shutdown during the epidemic period. Coordinate in advance to go through the formalities for transportation vehicles’ access to roads so that the procurement of various production materials was ensured [1].

2 Personnel Safety Guarantee 2.1 Arrival of Key Personnel The NPP technical specifications approved by the National Nuclear Safety Administration clearly stipulate the minimum number of shift personnel for the operating unit, and thus the number of shift personnel during normal operation cannot be less than the number required in the technical specifications. After the outbreak of the epidemic, personnel living in the closed and controlled areas could not arrive at their posts in time. Due to the restriction posed by required licenses in certain operation positions, personnel could not be replaced at will. The number of personnel in key positions in the operations department was the biggest challenge for the safety of the nuclear power plant during the epidemic. 2.1.1 Key-Position Personnel in Centralized Residence and Their Health Screening Personnel in key positions in the operations department refer to shift supervisors, deputy shift supervisors, senior operators, etc. After the outbreak of local Covid-19 epidemic, the local government would issue a series of epidemic prevention and control policies [2]. Common measures included traffic control to reduce the flow of people. However, there was a time lag from the formulation to the implementation of the above policies. Therefore, the operations department must bring key-position personnel within the controllable range of the nuclear power plant prior to the implementation of local government control policies. After the outbreak of the epidemic, the person in charge of the operations department immediately organized and verified the number of personnel in key positions in each operations shift, and coordinated the logistics department to prepare sufficient accommodation for personnel in key positions in operations department within the controllable range of the nuclear power plant, such as dormitory for shift personnel in the nuclear power plant.

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The management and control level of each community in the city where the epidemic broke out was usually classified as closed control area, control area and prevention area with alert level from high to low. Personnel in key positions in the operations department living in the prevention area and below should wear N95 masks for personal protection, and drive directly to the nuclear power plant by the shortest route. Parking or contact with others was prohibited along the way. During the period of centralized residence, personnel in key positions strictly avoided the contact with catering personnel, cleaning personnel and security personnel, and nucleic acid test for personnel in key positions was arranged as soon as possible. The person in charge of the operations department designated a person to specifically formulate the detailed information about the arrival of key personnel in the operations department, which would be dynamically updated and released in real time according to the arrival status of the above-mentioned personnel. At the same time, the operations department compiled the “Contingency Plan for Covid-19 Epidemic Emergencies”. The plan should specify the administrative region where the person in charge of the operations department, Shift Supervisors, Deputy Shift Supervisors and Section Chiefs of the operations department live, and stipulate that the personnel in key positions implement the emergency plan and arrive at the workplace during the epidemic outbreak in each administrative region. 2.1.2 Quick Check on the Itinerary of Key Personnel and Their Family Members In order to prevent and control the unexpected Covid-19 cases, refined management for each person in key position was implemented, and each one’s own itinerary was verified including the trajectory of his family members living together, whether the family member’s work place was related to local confirmed cases, health code, itinerary code, etc. [3]. Timely nucleic acid test was organized to assess the possibility of contracting Covid-19 to ensure that the personnel entering the NPP did not have the following conditions: (1) The itinerary code and health code of the employee and his/her family members were “non-green code”. (2) According to the travel trajectories of the confirmed cases announced by the government, identify those who were called the “time and space accompanying people”. (3) Preliminarily judge that a person would become a close or sub-close contact. For those involved in the epidemic who were initially judged to need key monitoring, an independent area would be arranged for home observation. If a person was judged to be a close contact or sub-close contact, immediately evacuate him from the centralized protection area of the NPP by a special vehicle, and disinfect the room and activity area where they stayed.

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2.2 The Arrival of the Personnel in the Secondary Key Positions Personnel in the secondary key positions in the operations department refer to operators, auxiliary system operators, and squad leaders of field operators who have the qualifications required in nuclear power plant technical specifications. The person in charge of the operations department asked each shift supervisors to immediately sort out the information of the personnel in the secondary key positions, and made it clear that all the above-mentioned personnel were not allowed to go outdoors except in an emergency. They should stay at home, and pack their luggage. They were prepared to enter the plant at any time according to the requirements of the operations department of the nuclear power plant. At the same time, organize the shift supervisors to evaluate the number of personnel in each shift. If there were not enough people for five shifts, the operations department will uniformly allocate the number of personnel in each position, and consider switching the shift mode to four shifts. 2.3 Alternate Shift Arrangements The shift personnel of the operations department live in different administrative districts of the city where the nuclear power plant is located. After the outbreak of the local Covid19 epidemic, due to the different restriction levels of epidemic prevention and control in the residential areas, some shift personnel were unable to arrive at their posts in time. In order to ensure the safe operation of the unit, the person in charge of the operations department should immediately start the backup personnel to replace them, and made it clear that the shift personnel in key positions and secondary key positions were prohibited from turning over the shift when the backup personnel did not arrive yet, so as to ensure the continuous parameter monitoring of the unit and guarantee emergency response capabilities. Afterwards, a 2–3 day temporary replacement schedule for personnel should be formulated according to the data of personnel in key and secondary-key positions in the operations department and issued on the day of epidemic control. 2.4 All Staff Was Screened for Covid-19 The person in charge of the operations department organized various shifts and sections to carry out health screening for all staff, including nucleic acid results, health codes and travel trajectories, etc., combined with the government’s announcement of the travel trajectories of confirmed cases, to confirm whether they are “time and space accompanying personnel”, and issued the results immediately. According to the minimum office requirement at the NPP site, it was required that employees with doubtful travel trajectories were prohibited from going to work. If employees moved into the dormitory for shift personnel, they should immediately move to a single room for isolation and observation, and stop working on shift. The daily administrative staff of the operations department take turns to work in office in an orderly manner or work from home according to the minimum office requirement of the nuclear power plant.

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3 NPP Site Partition Control 3.1 NPP Site Partition Control According to the overall epidemic prevention and control policy of the nuclear power plant, special mechanisms for management and control were in place to ensure that priority place and key personnel in the nuclear power plant would not be infected with Covid-19 [4]: (1) A special mechanism for management and control in the core area of the plant, namely the main control room of the operating units and the dormitory for shift personnel. (2) A special mechanism for management and control in the enclosed and controlled area in the plant, i.e., the work field of the operating units and the centralized residence of contractor maintenance personnel. (3) A special mechanism for management and control in off-site area in the plant, namely off-site residence areas of the NPP personnel. (4) A special mechanism for management and control in the construction area. (5) A special mechanism for personnel. The dormitory for shift personnel was under the leadership of Director of the operations department. Shift personnel were directly managed by shift supervisors who reported to Director of the operations department. Operations support personnel and operations technicians lived in the office building. The office building was under the leadership of Section Chiefs in the operations department. 3.2 Operators Stay in the Dormitory for Shift Personnel The person in charge of the operations department organized a flow survey of the operations personnel daily to assess the possibility of the personnel contracting the virus. In principle, as long as they were not the vulnerable personnel to the epidemic, the shift personnel on the day of shift and subsequent shifts would immediately move into the dormitory for shift personnel, and were strictly prohibited from leaving the NPP. In accordance with the principle of “high floor for main control room operators - low floor for field operators”, a boundary was set to make sure that these two types of operators did not come into contact with each other. They took different elevators. They went to work and arrived in their respective positions by different buses, during different time periods and by different routes. At the same time, immediately coordinate the logistics department to clear the non-shift personnel out of the dormitory for shift personnel according to the above principles [5]. Afterwards, special personnel were designated to coordinate the resources of the nuclear power plant, and a logistics security and supply team for the dormitory for shift personnel in the plant area was immediately established to guarantee the supply of clothing, food, lodging, transportation and epidemic prevention materials for residents in the dormitory for shift personnel.

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3.3 Closed Management for Shift Personnel at Work 3.3.1 Implement the Closed Management of the Main Control Room During the Work (1) The main control room of each unit and the office area of the field operators were physically isolated; (2) The main control room personnel and field operators “did not meet” or dine within each shift; (3) The shift on duty did not meet and turn over the shift to the next shift; (4) Start the online process of electronic work tickets. 3.3.2 Field Operators Avoided Close Contact with Maintenance Personnel (1) Field operators for each shift should try to avoid contact with non-operations personnel, and if necessary, keep a social distance of at least 1 m during the contact. (2) All contacts should be truthfully recorded and released in a summary to ensure that subsequent abnormality could be immediately investigated and traced.

4 Refined and Graded Management of Epidemic Prevention and Control During the epidemic, various health screenings, personnel statistics and other statistical reporting work were arranged according to the priorities, and systematic thinking and refined management methods were used to ensure that people and things were within a clear and controllable range. Statistical principles were as follows: summarize the projects that must be completed on the day after the epidemic outbreak, as well as the projects that continued to be carried out on a daily basis. Table 1 showed various statistics items.

5 Conclusion Through the above analysis and research, when the local Covid-19 hit, the NPP operations department should quickly formulate the following response actions, as shown in Table 2. In terms of power plant safety management and control, through adjustment of operations scheduling, operations plan adjustment, operations risk control, defects assessment, and production materials guarantee, the safe, stable, efficient and environmentally friendly operation of the units can be guaranteed. In terms of the management and control of operations personnel, the number of personnel in key positions such as Shift Supervisors and Deputy Shift Supervisors was guaranteed through centralized residence, health monitoring and rapid inspection of travel trajectories for personnel in key positions. Operational shift work was guaranteed by means of adjusting shifts and adding alternative personnel. Confirmed Covid-19 cases

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Table 1. Refined and graded management of epidemic prevention and control No.

Statistics for level 0 Statistics for level 1 action tasks action task

On-site office tasks

Tasks when working from home

1

Statistics table for the arrival of key-position personnel

Statistics table of occupancy in dormitory for shift personnel

Checklist of Tracking table for epidemic prevention non-shift personnel and control in working from home offices

2

Statistics table of personnel in secondary key positions

Statistics table of meal delivery to dormitory for shift personnel

Feedback to the Human Resources Department on the arrival of personnel from the operations department on a daily basis

Statistics table for non-shift personnel working from home who go outdoors

3

Reorganize shift schedules and temporary replacement schedules

Time table for dormitory for shift personnel

Statistics table of employees working in the office building

NPP entry screening table for backup shift personnel who normally stay at home

4

Quick checklist for the itinerary of key-position personnel and their family members

Statistics table of personnel who are potentially vulnerable to Covid-19 after health screening

Inventory statistics table of masks and other anti-epidemic supplies in the operations department

Task tracking table for backup shift personnel who stay at home and work from home

5

Preparation of emergency reserve roster

Table of contact history between field operators and non-operations personnel

Follow-up table of corrective actions to address special inspection issues related to epidemic prevention and control

Statistics table of life trajectory of employees and their families outside NPP centralized residence

6

Daily report of key-position personnel in the field of production and operations

Statistics table of various production materials required for daily production of operating units

Nucleic acid testing table for all staff

were avoided by screening the health of all employees, nucleic acid testing and recording the contact details between field operators and non-operations personnel. In terms of NPP site partition control, the core area and the enclosed area in the plant were established and the close contact between the operations personnel and the non-operations personnel were limited to reduce the risk of infection of the operations personnel in the core area. Furthermore, the face-to-face contact between operations

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personnel in the main control room and field operators were limited. The number of personnel in key positions were guaranteed, and the number of key-position personnel from the operations department met the requirement in the NPP technical specifications. Table 2. Initial response actions of the NPP operations department after local Covid-19 epidemic outbreak No.

Action

Response level

Responsible person

1

Unit safety control

1.1

Operations scheduling adjustment

Level 0

Head of the operations department

1.2

Operations plan adjustment

Level 0

Person in charge of operations department who supervises support section

1.3

Operations risk control

Level 0

Head of the operations department

1.4

Unit defects assessment

Level 0

Person in charge of operations department who supervises support section

1.5

Production materials guarantee

Level 1

Person in charge of operations department who supervises management section

2

Personnel safety guarantee

2.1

Arrival of key-position personnel at NPP site

Level 0

The person in charge of the operations department who supervises shifts

2.2

The arrival of the personnel in the secondary key positions at NPP site

Level 1

The person in charge of the operations department who supervises shifts

2.3

Alternate shift arrangements in the operations department

Level 0

The person in charge of the operations department who supervises shifts

2.4

The health screening and nucleic acid test for all staff

Level 1

Person in charge of operations department who supervises management section

3

Site partition control

3.1

Partition control of the nuclear power plant

Level 1

Head of the operations department

3.2

The dormitory for shift personnel

Level 1

Person in charge of operations department who supervises management section (continued)

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Table 2. (continued) No.

Action

Response level

Responsible person

3.3

Closed management for shift personnel at work

Level 1

The person in charge of the operations department who supervises shifts

Note Six “level 0” response actions, that is, immediate response after the outbreak of the local epidemic; Six “level 1” response actions, that is, response on the same day of the outbreak of the local epidemic

References 1. Notice on Doing a Good Job in Guaranteeing Coal, Power, Oil and Gas Transportation During the Period of Epidemic Prevention and Control, Liannengbao Joint [2022] No. 1 2. The State Council’s Guides on Response to the Covid-19 Epidemic/Joint Prevention and Control Mechanism on Doing a Good Job in the Covid-19 Epidemic/Normalized Prevention and Control, Guofa MingDian [2020] No. 14 3. Requirements for access to Tianwan Nuclear Power Plant during the period of intensive control of the epidemic risk, GE-TR-OHS-000-2022005 4. Tianwan Nuclear Power Plant Covid-19 Epidemic Risk Control Work Plan, GE-TR-OHS-0002022004 5. Temporary Management Instructions for Epidemic Prevention and Control in the Main Control Room of Tianwan Nuclear Power Plant Units 1–6, GE-TR-QAC-110-2022003

Discussion on Nuclear Arms Control Based on the International Situation in 2022 Yanfeng Lyu(B) and Xuesheng Lyu China Institute of Atomic Energy, Xinzhen Sub District, Fangshan District, Beijing, China [email protected]

Abstract. In 2022, major changes took place in the international situation. Although China, the United States, Russia, Britain and France signed the Joint statement of the leaders of the five Nuclear-Weapon States on preventing nuclear war and avoiding arms races, war broke out between Russia and Ukraine, NATO and Russia frequently hinted at each other’s threat of nuclear war, and President Putin personally directed the nuclear deterrence military exercise, which undoubtedly worsened the already low tide of international nuclear arms control. This paper holds that if the situation of nuclear arms control wants to be improved, the strength of major international forces needs to be relatively balanced. Under the condition that the strength of the multipolar world pattern is basically balanced and can restrict each other, high reliability verification means such as zero knowledge protocol can contribute value and strength to the further development and construction of the international nuclear arms control system. Keywords: International situation · Nuclear arms control · Zero knowledge protocol

1 Introduction In 2022, we have witnessed many major events in the world, which have impacted our lives, affected our society and changed our international environment. Although this year, the leaders of China, Russia, the United States, Britain and France made their first joint statement on the issue of nuclear weapons, issued the joint statement on the prevention of nuclear war and the avoidance of arms race. It seems to have found a new way of solving the stalemate in the form of international nuclear non-proliferation and nuclear arms control. However, the form of international security took a sharp turn. The sudden outbreak of war between Russia and Ukraine shocked the world. War and sanctions had a huge impact on the world’s economy and power balance. Compared with the economy, this great change has cast doubt on the peaceful and stable international environment. At the same time, covid-19 virus is still prevalent all over the world, making the already fragile world economy worse. The form of nuclear arms control and nuclear non-proliferation has never been so bad in the past.

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2 Analysis of International Nuclear Arms Control 2.1 Introduction of International Situation of Nuclear Arms Control In 2022, we witnessed the worst international security environment since the Second World War. Although we have made some achievements in the field of nuclear arms control in recent years, for example, the United States and Russia agreed to extend the new strategic arms reduction treaty in February 2021 [1], and the joint statement signed by the five nuclear weapon states this year. These incidents could have served as a way to revitalize the nuclear arms control process, pave the way for further reduction of nuclear weapons and nuclear disarmament [2]. However, due to the war in Ukraine, the risk of nuclear war has been reassessed, and the conflict among the world’s major forces has become more and more intense. Before the war broke out this year, Ukrainian President Zelensky claimed at the Munich Security Conference that if Ukraine could not get security guarantees, it would consider the Budapest memorandum ineffective, and all the package decisions of 1994 were questioned. Ukraine joined the Budapest security memorandum in 1994 and gave up the world’s third largest nuclear capability in exchange for security. Zelensky’s speech hinted that Ukraine might abandon its decades old commitment to a non-nuclear state. Meanwhile, Russian President Vladimir Putin was absent from the Munich Security Conference and did not send any official representatives. During the Munich Security Conference, Russian President Vladimir Putin personally took the lead in commanding the annual all-round nuclear strike exercise of Russia’s strategic deterrent. Although Russia holds such exercises every year [3], the Black Sea fleet, which has never participated in the exercises before, joined them this year for the first time in history. Previously, such exercises were usually held at the end of each year, but the time of this year’s exercise has obviously changed. After the war beginning, with the increasing spillover effect of the Ukrainian battlefield, the confrontation between Russia and NATO gradually extended to the Baltic Sea. At present, the United States has about 200 b-61 tactical nuclear weapons stored in six military bases in Belgium, Germany, Italy, Netherland and Turkey. However, affected by the conflict between Russia and Ukraine, the idea of NATO deploying tactical nuclear weapons to Eastern European countries has been put forward and discussed more than once. In April this year, senior Polish officials also called on NATO to be open to the deployment of us tactical nuclear weapons in Poland. Not only did the United States try to adopt more radical nuclear weapons deployment, but after Lithuania imposed a blockade on Kaliningrad, President Lukashenko of Belarus believed that Lithuania’s blockade was equivalent to a declaration of war. Lukashenko stressed that NATO warplanes are conducting flight training with nuclear weapons on the border of Belarus, which seriously threatens Belarus’ national security. Belarus must have equivalent defense measures. Belarus needs to be fully prepared. Even if it uses the most powerful weapons, it must protect the country. This may mean that Belarus will regain nuclear weapons. On February 27 this year, Belarus deleted the nuclear weapon free clause by amending the constitution, making this possibility never so great. In fact, outside the Russian Ukrainian war, the international form of nuclear nonproliferation is also very dangerous. In September, 2021, Australia suddenly announced that it would tear up its submarine contract with France and instead purchase American

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or British nuclear submarines under the framework of AUKUS. The nuclear submarine reactors of the United States and Britain use highly enriched weapons grade enriched uranium, which can be completely made into nuclear weapons if taken out. In accordance with the provisions of the nuclear non-proliferation treaty, nuclear states prohibit the export of weapons grade uranium to non-nuclear states. If the United States and Britain take the lead in ignoring this provision, the nuclear non-proliferation situation that has been maintained for decades may be broken because of several nuclear submarines. 2.2 Analysis of International Situation of Nuclear Arms Control At present, due to the imbalance of the main forces in the world, some western countries have mastered advanced technology and leading international status. With the help of their public opinion advantages, they improperly exaggerate the confrontational atmosphere, put the strategic competition of big powers above arms control, and believe that the competition of big powers is the new internal core of their global strategy. In the field of nuclear arms control, these countries resorted to opportunism and repeatedly undermined the atmosphere of arms control negotiations and consultations [4]. For example, the number of nuclear weapons in China is one order of magnitude less than Russia or the United States, but China has repeatedly been asked to reduce the number of nuclear weapons. The consequence of doing so is that it directly threatens China’s national defense security under the current seriously unbalanced balance of nuclear power. For another example, the United States has advantages in sea based and air-based medium range missiles, while China has advantages in ground-based medium range missiles. Every time the United States mentions the issue of medium range missiles, it excludes sea based and air-based guided missiles. In fact, sea-based and air-based medium range missiles have even more serious damage to regional and global strategic stability. For these problems, some western countries do not seek pragmatic solutions, but blindly put pressure on other countries to negotiate and undermine the balance of world power. At the same time, the international crisis management logic of some western countries is also very dangerous, which intensifies the risk of arms race and military confrontation. Crisis management and control need to curb the occurrence and escalation of the crisis. The new core of the global strategy of great power competition in some western countries is the root cause of a crisis. If a crisis occurs, all other countries must cooperate with these countries to control the crisis and make the crisis develop in accordance with their national interests. In this international environment, nuclear arms control is difficult to continue, but this does not mean that we give up the research and exploration of nuclear arms control related technologies. We will make advance scientific and technological preparations for a peaceful and stable world in the future.

3 Some Current Research in Nuclear Arms Control When a stable international situation comes, a nuclear arms control agenda focusing on transparency, communication, regulation and reducing the leakage of sensitive information will be necessary. As early as the last century, when the United States and the

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Soviet Union negotiated large-scale reduction of strategic weapons, they started relevant research and testing on nuclear warhead verification technology. It is found that the nuclear warhead can be reliably identified by gamma ray line spectrum measurement and gamma ray penetration measurement, but this will cause the leakage of sensitive information. Finally, according to the protocol between the two countries, simple neutron detectors, thermo luminescence dosimeters and ionization radiometers are used for on-site verification to confirm the existence of nuclear warheads and count the number of nuclear warheads through simple counting methods. However, this method gives up the goal of verifying fake nuclear warheads, and the two countries believe that fake nuclear warheads have no military significance. However, in the process of nuclear disarmament, if the verification of fake nuclear warheads is abandoned, nuclear warheads may be replaced by fake nuclear warheads. This will cause the nuclear warhead to be disassembled on paper and still be hidden in fact. This makes it necessary to study a system that does not leak sensitive information and can test the authenticity of nuclear warheads. In the past decades, experts and professors have proposed many methods to solve this problem, such as the information barrier method [5]. However, those methods have limitations, which will make the hardware measure sensitive information and cause the possibility of information disclosure. In 2014, Professor Alexander proposed the method of using zero knowledge protocol to manufacture bubble neutron detector array covered by X-ray irradiation radiograph according to the corresponding nuclear warhead template, detect the statistical information after the warhead is irradiated by external neutron source, and compare these data with the template data to verify whether the mass abundance of the nuclear warhead matches the template [6]. This method cannot detect sensitive information by hardware itself, eliminating the possibility that the verifier may bypass the information barrier to obtain sensitive information from the hardware level. In 2017, Professor Jie Yan proposed the method of replacing X-ray film coverage with random HDPE plates to interfere with the fission information of nuclear warheads irradiated by external neutron sources [7]. This makes the induced fission signal completely distorted, which enhances the credibility of the verification results and greatly reduces the possibility of sensitive information leakage of the verification results compared with Professor Alexander’s method. But at present, all the verification methods of nuclear warheads based on zero knowledge protocol need to be template based. This method is not so much to detect the true or false of the nuclear warhead as to detect whether the nuclear warhead matches the template. However, the template itself contains sensitive information. This point is not touched at present. The current method is suitable for detecting the authenticity of a batch of nuclear warheads with the same design, including the same mass, shape and abundance. However, if nuclear warheads with different designs are checked according to the current zero knowledge protocol, they will certainly fail. How to solve the problem of relying on templates will be the key to the next nuclear weapon verification research based on zero knowledge protocol.

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4 Conclusions The nuclear arms control situation in 2022 is at an extremely low point in history. The current international environment for nuclear non-proliferation is not optimistic, but we still need to study nuclear arms control related technologies. Under the condition that the strength of the multipolar world pattern is basically balanced and can restrict each other, high reliability verification means such as zero knowledge protocol can contribute value and strength to the further development and construction of the international nuclear arms control system.

References 1. Woolf, A.F.: Promoting nuclear disarmament through bilateral arms control: will new START extension pave the path to disarmament? J. Peace Nucl. Disarm. 4(2), 309–320 (2021) 2. Weber, H.A., Parthemore, C.: Cruise control: the logical next step in nuclear arms control. J. Peace Nucl. Disarm. 2(2), 453–467 (2019) 3. Podvig, P.: Russia’s current nuclear modernization and arms control. J. Peace Nucl. Disarm. 1(2), 256–267 (2018) 4. Neuneck, G.: The deep crisis of nuclear arms control and disarmament: the state of play and the challenges. J. Peace Nucl. Disarm. 2(2), 431–452 (2019) 5. Min, L.: Discussion on nuclear detection techniques used in verification of nuclear warhead. In: Proceedings of the 7th National Conference on Nuclear Electronics and nuclear detection technology Lushan: China Electronics Society, pp. 50–52 (1994) 6. Glaser, A., et al.: A zero-knowledge protocol for nuclear warhead verification. Nature 510, 497–502 (2014) 7. Yan, J.: A random and interactive inspection protocol for nuclear warheads verification. Progress Report on China Nuclear Science and Technology, vol. 5, pp. 286–291 (2017)

SNSTC’s Capabilities and Practices on Performance Testing for Nuclear Security System and Equipment Wang Shuo(B) , Yang Changjie, Lu Hong, He Jialin, and Chen Chen State Nuclear Security Technology Center (SNSTC), Beijing, China [email protected]

Abstract. The State Nuclear Security Technology Center (SNSTC) was established with the approval of the Chinese central government in November 2011. As an affiliate to China Atomic Energy Authority (CAEA), SNSTC’ primary mission is to provide technical support for the government management on nuclear security, nuclear materials control, nuclear export & import control and nonproliferation; and to conduct international exchanges and cooperation. SNSTC is also the operator of China’s Center of Excellence (COE) on Nuclear Security. Since the operation of COE in March 2016, SNSTC has received the ISO-9001:2015 QMS certificate and its laboratories have been certified by China National Accreditation Service for Conformity and Assessment (CNAS) and China Metrology Accreditation (CMA) in 2017. With the comprehensive testing capabilities on function/performance, environmental applicability and electromagnetic compatibility for nuclear security related system and equipment, SNSTC has completed more than 450 tests for about 100 sets of radiation detection and physical protection equipment for nuclear facilities, customs, universities and other relevant stakeholders. As a third-party testing agency, SNSTC was commissioned by General Administration of China Customs to take performance testing and acceptance testing for the radiation portal monitors to be deployed at border ports in 2017–2018. The testing included 46 test items, covered the various aspects such as radiation detection function, radiation detection performance, environmental adaptability, electromagnetic compatibility and long-term reliability, etc. In addition, SNSTC led the technical review and infield acceptance testing for several physical protection system upgrading projects in China, and conducted the physical protection system effectiveness evaluations for many times as requested by nuclear facility operators. Based on the works above mentioned, SNSTC also developed a series of technical documents, such as the Management Measures on Acceptance of Physical Protection Engineering, the Technical Guidance on Acceptance Test of Physical Protection System in Nuclear Facilities, the Technical Specifications for Central Control Room of Physical Protection System in Nuclear Facilities, the Technical Specifications of Digital Radiation Imaging Device used for Vehicle Access Control in Nuclear Facilities and so on. This article will provide a briefing introduction to SNSTC’s capabilities and practices on performance testing for nuclear security related systems and equipment. Some typical cases on performance testing in lab and field conditions for radiation detection equipment and physical protection systems should be demonstrated. The general objects, procedures and requirements of performance testing

© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 193–197, 2023. https://doi.org/10.1007/978-981-19-8899-8_19

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W. Shuo et al. will be introduced. Furthermore, the necessity of and lessons-learned from performance testing, and its importance for nuclear security sustainability on national and facility levels will be discussed. Keywords: Performance testing nuclear security system and equipment

1 Introduction The State Nuclear Security Technology Center (SNSTC) was established with the approval of the Chinese central government in November 2011. As an affiliate to China Atomic Energy Authority (CAEA), SNSTC’ primary mission is to provide technical support for the government management on nuclear security, nuclear materials control, nuclear export and import control and nonproliferation; and to conduct international exchanges and cooperation. In September, SNSTC has been the Collaborating Center on Nuclear Security of International Atomic Energy Agency (IAEA). SNSTC is also the operator of China’s Center of Excellence (COE) on Nuclear Security. Since the operation of COE in March 2016, SNSTC has received the ISO9001:2015 QMS certificate and its laboratories have been certified by China National Accreditation Service for Conformity and Assessment (CNAS) and China Metrology Accreditation (CMA) in 2017. With the comprehensive testing capabilities on function/performance, environmental applicability and electromagnetic compatibility for nuclear security related system and equipment, SNSTC has completed more than 1000 tests for about 100 projects of radiation detection and physical protection equipment for nuclear facilities, customs, universities, institutes and other relevant stakeholders. At present, the Environmental Reliability and Electromagnetic Compatibility Laboratory has obtained more than 500 testing projects approved by CNAS and CMA. It has already possessed the functions, performance, environmental applicability and electromagnetic compatibility testing capabilities of relatively complete nuclear security related products, and has laid a solid foundation for testing of nuclear security system and equipment.

2 Testing Methods and Industry Standards In 2016, SNSTC completed the Testing and Evaluation Program of Radiation Portal Monitors (RPMs) according to the national standard GB/T 31836-2015 “Radiation Protection Instrument for Spectral Analysis-Based RPMs for Detecting and Identifying the Transport of Illegal Radioactive Materials”, and the development of the “RPMs Performance Testing” Automated Testing System. The system can perform key performance testing such as nuclide identification and radiation response when carrying the radioactive source/nuclear material through the RPMs at different speeds and different heights. The Testing and Evaluation Program of the RPMs and the establishment of the Automated Testing System not only standardize the testing method of the PRMs, but also greatly save the testing time and improve the testing efficiency, and lay a solid foundation for the standard testing of RPMs in the future.

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In 2017, three internal control standards for nuclear security, namely, “intrusion detection equipment testing standards—general provisions”, “testing standards for portal radiation detection system based on spectral analysis” and “testing standards for handheld radiation detection equipment”, were completed. In 2018 SNSTC completed “the Portable Trace Explosives Detector Testing”, “X-ray Security Inspection Equipment Testing” and “the Hand-Held Metal Detector Testing” three internal control standards on nuclear security. The above documents provide standards and specifications for the future testing of nuclear security equipment, and provide a technical basis for the standardization of the nuclear security equipment industry.

3 Testing of RPMs As a third-party testing agency, SNSTC was commissioned by General Administration of China Customs to take performance testing and acceptance testing for the radiation portal monitors to be deployed at border ports in 2017–2018. The testing included 46 test items, covered the various aspects such as radiation detection function, radiation detection performance, environmental adaptability, electromagnetic compatibility and long-term reliability, etc. (Figs. 1 and 2).

Fig. 1. Performance testing.

Fig. 2. Electromagnetic testing

According to the Testing and Evaluation Program for new-type RPMs, the testing items are mainly divided into five categories: product configuration testing, radiation

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detection function testing, radiation detection performance testing, environmental adaptability testing and long-term reliability testing. Each major test is divided into several specific testing scenarios. The scenarios are shown in Table 1. Table 1. Testing scenarios Testing items

Scenarios No.

Testing sample No.

Product configuration Gamma-ray detector testing Neutron detector

A1

1#

A2

1#

Radiation detection function testing

Basic function

B1

1#

Detailed function

B2

1#

Radiation detection performance testing

False alarm

C1

1#

Gamma-ray response

C2

1#

Neutron response

C3

1#

Background effect

C4

1#

Nuclide identification – single nuclide identification

C5

1#、2#

Nuclide identification simultaneous identification for multiple nuclide

C6

2#

Nuclide identification – nuclides not in the library

C7

1#

Temperature testing

D1

3#

Environmental adaptability testing

Testing scenarios

Humidity testing

D2

3#

Dust proof testing

D3

5#

Waterproof testing

D4

3#

Vibration testing

D5

4#

Electrostatic discharge testing

D6

3#

Radio frequency immunity testing

D7

3#

Radio frequency emission testing

D8

3#

Conducting disturbance testing

D9

3#

Magnetic field immunity testing

D10

4# (continued)

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Table 1. (continued) Testing items

Long-term reliability testing

Testing scenarios

Scenarios No.

Surge impact testing

Testing sample No.

D11

3#

Electrical fast transient burst D12

3#

Sand blowing testing

5#

D13

Salt spray testing

D14

5#-A1~5#-A4

Passive detection testing

E1

6#~15#

Active detection testing

E2

6#~15#

This testing activity is the first comprehensive and systematic testing of the functions, performance, environmental applicability and electromagnetic compatibility of newtype RPMs in China. It has pioneered the overall performance testing of RPMs and laid a solid foundation for testing similar products in the future. SNSTC has been highly praised by the General Administration of Customs.

4 Coordinated Research Projects of IAEA From August 2017 to May 2018, SNSTC undertook the IAEA Coordinated Research Projects (CRP) “Building Fast Assessment Method of Initial Alarm from Vehicle Radiation Detection Portal Monitor (CRP J02005)”. SNSTC prepared research programs, collected alarm data, constructed theoretical models, analyzed statistical data, etc., and provided core data and model support for IAEA’s development of TRACE software. This work helps to improve the performance of RPMs, reduce false alarm rate, and improve radiation detection efficiency at customs ports. Currently, IAEA has applied this method in 24 countries on 5 continents.

5 Summary In addition, SNSTC led the technical review and in-field acceptance testing for several physical protection system upgrading projects in China, and conducted the physical protection system effectiveness evaluations for many times as requested by nuclear facility operators. Based on the works above mentioned, SNSTC also developed a series of technical documents, such as the Management Measures on Acceptance of Physical Protection Engineering, the Technical Guidance on Acceptance Test of Physical Protection System in Nuclear Facilities, the Technical Specifications for Central Control Room of Physical Protection System in Nuclear Facilities, the Technical Specifications of Digital Radiation Imaging Device used for Vehicle Access Control in Nuclear Facilities and so on. In the future, as the IAEA Collaborating Center on Nuclear Security Technology, SNSTC will carry out more and more nuclear security equipment performance testing projects, provide technical support for IAEA, and work with other organizations to contribute to the worldwide security capacity building.

Practical Analysis of Radon Protection Methods in an Underground Project Jie Tian1,2(B) , Guobo Zhong2 , Fei Wu1 , and Xiangwei Wang2 1 Naval University of Engineering, Wuhan, Hubei, China

[email protected] 2 No. 91515 Unit of PLA, Sanya, Hainan, China

Abstract. Radon and its progeny are gaseous radioactive substances, colorless and odorless, widely present in ground buildings and underground projects. Especially in underground engineering, due to the continuous decay of uranium contained in the soil, radon will be continuously released from the soil into the air in the underground engineering, causing continuous internal exposure damage to the staff. Radon is the most important contributor to the natural background radiation exposure of humans, and the internal exposure dose caused by inhalation of radon and its progeny accounts for more than half of the total natural radiation dose received by the public (Charles in J. Radiol. Prot. 21:83–86, 2001 [1]). Based on the main properties of radon, three common radon reduction measures were tested in an underground project, and the optimal radon reduction method was studied. Radon and its daughters are gaseous radioactive substances widely existed in underground projects, which will seriously harm workers’ healthy. Based on the main properties of radon, three common measures for radon concentration reducing purposes, e.g. ventilation, shielding and adsorption, were tested and validated in an underground project. The statistical analysis results show that ventilation costs lower and its radon-reducing effect behaves better. By ventilating 3 times per hour, the radon concentration can be reduced by 65.6%. Wearing a mask can also effectively prevent radon from entering the body through the respiratory tract. Mask of type Nanhe 95 can filter 89.17–95.82% of radon and its daughters. In contrast, for underground projects with big room, radon was released continuously, and small-scale radon adsorption device would do little on radon concentration reduction. Keywords: Radon · Underground engineering · Measures to reduce radon

1 Harm Caused by Radon Radon is a radioactive gaseous element whose radioactive isotopes are 222 Rn, 219 Rn, and 220 Rn, with half-lives of 3.83 days, 3.96 s, and 55.6 s, respectively; and 222Rn, 219Rn and 222Rn are fission products of 238U, 232Th, and 235U, respectively. Radon progeny is a term generally for short-lived decay products of radon, including 218Po, 214Po, 214Pb, 214Bi, 210Pb and other nuclides. Radon progeny are heavy charged metal solid particles with strong diffusivity and wall attachment effect. They can easily © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 198–202, 2023. https://doi.org/10.1007/978-981-19-8899-8_20

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combine with condensation nuclei in the air to form radioactive aerosols suspended in the air. When inhaled, they will precipitate on the tracheal wall or on lung slices. α-rays caused by the decay of radon progeny is the main source of radiation hazards caused by radon [2]. Inhalation of radon and its progeny will cause internal radiation damage to the human body. The deterministic effect is to cause significant changes in blood cells, while the random effect is to induce cancer at some certain probability.

2 Effect Analysis of Three Common Radon Protection Measures Three methods of ventilation, shielding and adsorption are usually adopted to reduce radon in existing underground projects. By field measurement and inspection, abovementioned three methods are compared and analyzed. 2.1 Ventilation Ventilation is the most common and direct way to reduce radon. Due to the fact that radon concentration in the air of the external environment is low, air from the outer space of the underground project is introduced via the ventilation system into the project, replacing the original inner fractions by discharging the original air in the project through the exhaust system, thus reducing the radon concentration in the air. Three areas were randomly selected in the underground project, and the RTM2200 portable radon measuring instrument was used for continuous measurement in the center of such areas. During test, the ventilation of the selected area was set as “on” and “off” each, and the ventilation was reset after the radon concentration stabilized. The results are as follows (Table 1). Table 1. Comparison of radon concentration when turning on and off the ventilation system Serial number

Workshop

Radon concentration when ventilation is turned off (Bq/m3 )

Radon concentration when ventilation is turned on (Bq/m3 )

Ventilation frequencies (times/h)

1

A

1606.52

500.85

3

2

B

1466.86

541.43

3

3

C

1668.70

565.56

3

According to the measurement results, ventilation has a significant effect on reducing the radon concentration of underground projects, and radon concentration was averagely reduced by 65.6% of in such areas. The ventilation system is an essential facility for underground projects, which is used to control the temperature and humidity inside the project and deliver the staff oxygen which is necessary. While achieving the above functions, it can reduce radon concentration as well. Ventilation was proved to be one of the most economical and suitable for radon concentration reduction in underground projects.

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2.2 Shield Radon reduction by shielding is achieved by using radon shielding coating and other covering materials. It covers the contact surface between the project and the mountain, and shield the radon precipitated from the mountain. Commonly used radon shielding coatings include cement-based composite materials [3] and gypsum composite materials [4], etc. However, this method is costly since the contact area between underground engineering and the mountain is large and the price of the coating layer is correspondingly high. Another method that takes effect is wearing protective masks to isolate the way radon enters the human respiratory tract. To verify the radon protection effect of masks, a simple filtering effect experiment was conducted. Methodology: use an atmospheric sampling pump to extract air samples near the breathing zone of personnel in the underground project, and filter the samples with filter paper (radon adsorption efficiency 1); the flow rate of the sampling pump is 100 L/h, and the sampling time varies from 10 to 60 min. Samples for comparison: use a mask (Nanhe type 95, radioactive dust protective mask) to cover the filter paper, while other conditions remain the same as above. Using the FJ-347 scintillation probe and the FH463B automatic scaler to measure the two filter paper samples respectively, two sample counts were obtained (see Table 2 for details). It can be seen that wearing a mask can effectively prevent radon from entering the human body through the respiratory tract. The two kinds of masks had distinct protective effects on radon and its progeny, and the effects were basically the same. The unobstructed filter paper counts in Table 2 is not proportional to the sampling time, which is related to the sampling time. Meanwhile, the radon concentration fluctuating up and down was observed, we analyzed such situation was resulted from the influence of ventilation, temperature and humidity. Table 2. Comparison of experimental counts of protective efficiency of masks Sampling duration (min)

Measure duration (s)

Filter paper without any cover

Filter paper filtered by Nanhe mask

Filter paper filtered by radioactive dust protective mask

Counts

Counts

Counts

Efficiency (%)

Efficiency (%)

10

600

5732

621

89.17

542

90.54

20

600

12,791

1332

89.59

1246

90.26

30

600

5126

214

95.83

87

98.30

40

600

10,213

1106

89.17

1175

88.50

50

600

15,421

1253

91.87

1378

91.06

60

600

17,988

865

95.19

769

95.72

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2.3 Adsorption Radon reduction by adsorption is based on activated carbon and other materials with strong radon adsorption to adsorb radon inside the project. The commonly used radon adsorption material is activated carbon, which can adsorb and retain radon in the air by utilizing the characteristics of activated carbon with large specific surface area, developed microporous structure, weak chemical properties, and strong adsorption of inert gases. The higher the quality of activated carbon, the stronger its adsorption and retention capacity. However, it will lead to an increase of the volume and cost of the radon reduction device. Increasing the gas adsorption pressure can improve the adsorption coefficient of activated carbon, namely pressure swing adsorption technology, but it has a saturation effect on the increase of activated carbon adsorption coefficient [5]. In order to verify the effect of radon reduction by adsorption, we used JD-80 type adsorption radon reduction device to conduct radon reduction experiments. The device uses activated carbon as the adsorption material, and a radon reduction was completed via a four-stage circle: adsorption, heating, analysis, and cooling. Adsorption: the fan inhales the air in the environment, and under the action of the adsorption carbon bed, the radon in the air is adsorbed, and it needs to run for 120 min; Heating: the electric heating tube heats the adsorption carbon bed, so that the radon is released from the activated carbon It needs to run for 100 min; Analysis: turn on the vacuum pump, and transfer the radon released by the adsorption carbon bed to the retention carbon tank for storage, so that the adsorption carbon bed can restore the radon adsorption capacity; Lastly, it needs to run for 140 min for cooling, during which stage it takes 60 min to cool down the adsorption bed. Methodology: According to the volume size, two areas in the underground project were selected for testing. The test area volume was 610 m3 and 270 m3 , which were marked as area D and area E respectively. RTM2200 portable radon meter was on operation before the device was running, and the device remained running throughout the process. The data is as follows (excerpt) (Table 3). It can be seen from that the two regions did not achieve significant radon reduction effect. On the one hand, the operation efficiency of this type of radon reduction device is low, and the effective radon reduction time in a work process is 120 min, which only accounts for 28.6% of each work cycle; on the other hand, Radon and its daughters are constantly exhaled from the soil and the intermittent adsorption method is difficult to meet the demand for radon reduction.

3 Conclusions Based on experimental researches on radon reduction measures in underground engineering, it was found that ventilation can effectively reduce radon concentration, meanwhile, wearing a mask can effectively prevent radon from entering the human body through the respiratory tract. However, due to the inherent saturation of activated carbon, small activated carbon adsorption devices failed in application of radon concentration reduction of underground engineering with high radon activity concentration, since the activated carbon quickly comes to saturation, and in the long process of antipyretic, desorption, and cooling, the radon in the soil is continuously precipitated.

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Table 3. Counting of radon concentration changes with the adsorption radon reduction device Area D

Area E

Running duration

Radon concentration (Bq/m3 )

Running duration

Radon concentration (Bq/m3 )

Before running

2162

Before running

2215

4h

1991

4h

1815

8h

1980

8h

2355

12 h

1647

12 h

2078

16 h

1972

16 h

2088

20 h

2142

20 h

2307

24 h

2036

24 h

2513

28 h

1819

28 h

2154

32 h

2001

32 h

2212

36 h

1830

36 h

2181

40 h

1822

40 h

2302

44 h

1679

44 h

2175

48 h

1826

48 h

2139

References 1. Charles, M.: Unscear report 2000: sources and effects of ionizing radiation. United Nations scientific committee on the effects of atomic radiation. J. Radiol. Prot. 21(1), 83–86 (2001) 2. Weiwei, W., et al.: Research on radon protection methods and capabilities of underground engineering. Radiat. Prot. 42(1), 48–53 (2022) 3. Ting, W., et al.: Preparation of a gypsum composite radon shielding coating. Nonmet. Miner. 44(2), 31–33 (2021) 4. Ting, W., et al.: Preparation and properties of cement-based composite radon shielding coating. Nonmet. Miner. 44(3), 30–32 (2021) 5. Feng, X., et al.: Research on high pressure adsorption technology of radon gas on activated carbon. At. Energy Sci. Technol. 50(4), 763–768 (2016)

Numerical Calculation of Flow Heat Transfer in a Single Square Tube Under a Blockage Accident Daheng Li, Shuwen Yu(B) , Ahmed A. Ghani, and Changhong Peng University of Science and Technology of China, Hefei, China [email protected]

Abstract. The first wall of the fusion Water Cooled Ceramic Breeder blanket is a multi-channel parallel square channel. Due to the small size of the flow channels, it may be easy to initiate block accidents. In this paper, the commercial computational fluid dynamics software ANSYS FLUENT was used to conduct numerical simulation and research discussion on different blockage aspects such as blockage materials, thickness, shares, and blockage shapes in the middle of a single square channel. By simulating three different blockages shapes in the flow channel, the results show a great influence on the flow field downstream, and the flow resistance is changed, which provides some reference for the following flow distribution problems. Through comparison, it is found that the wall temperature at the blockage’s positions is inversely proportional to the thermal conductivity of the material and directly proportional to the thickness. Keywords: Square flow channel · Channel blockage · Numerical simulation

1 Introduction The continuous consumption of non-renewable fossil energy and a series of atmospheric and environmental pollution problems caused by it makes people have to consider the development of cleaner and more efficient renewable energy. As clean energy with the greatest potential to solve the energy crisis in human society, nuclear fusion has attracted much attention from all countries. CFETR (China Fusion Engineering Test Reactor) is an important scientific project to accelerate the practical application of nuclear fusion. According to the principle of thermonuclear fusion inside the sun, a controllable "artificial sun" is designed and built to solve the human energy crisis. Xu et al. [1] introduced the new types of structural materials of fusion cladding, such as modified RAFMs and noneMA ODSs, and looked ahead to the development of advanced structural materials with better performance. Shi et al. [2] used the Monte Carlo particle transportation method to calculate the neutron radiation damage. The damage of the first wall is compared when beryllium and tungsten are used as plasma-facing materials. Tungsten is more suitable than beryllium as the first wall armor material of the blanket because tungsten has a better protective effect on the structural material of the blanket. Li et al. [3] analyzed the neutronic characteristics of the WCCB (Water Cooled Ceramic Breeder) blanket of © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 203–213, 2023. https://doi.org/10.1007/978-981-19-8899-8_21

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CFETR and found that the results met the design requirements of CFETR. Yang et al. [4] used RELAP5/mod3.4 to investigate in-box LOCA of WCCB for CFETR based on PWR conditions. The results show that the coolant after the break can enter the vacuum chamber due to the existence of a safety valve. This ensures the integrity of the blanket. Wang et al. [5] used RELAP5 to carry out steady-state analysis and accident simulation of HCCB (Helium Cooled Ceramic Breeder) blanket and found that the design of the WCCB blanket met the safety requirements in case of leakage accident. Liu et al. [6] used the Relap5/Scdapsim/Mod3.2 program to simulate 100% blockage at the inlet of a box of standard fuel assemblies of China advanced research reactor. The results show that the fuel assembly with the blocked flow will burn out, but will not expand to adjacent components. Song et al. [7] studied the inlet blockage of the plate fuel assembly and found that the coolant will not boil even if the blockage ratio is 100%. Davari et al. [8] simulated the blockage accident in the flow channel of the Teheran research reactor (TRR). When the blockage is less than 50%, it is in a safe state. When the blockage is more than 70%, nucleate boiling occurs. For the WCCB blanket, due to the small size of the flow channel, it is likely to cause blockage accidents, resulting in radioactive leakage and other serious consequences. It is particularly important to study the blockage accident. There is no sufficient research on the WCCB blanket for fusion reactors. This paper aims to contribute to this field. In this paper, ANSYS FLUENT software has been used to study the fluid velocity and wall temperature changes in a parallel square channel of the first wall of the WCCB blanket when the blockage accident occurs, which provides a reference for safe operation for the CFETR reactor. Also, the flow resistance is increased by adding different blockages in the flow channel, which has guiding significance for the flow distribution.

2 Numerical Methods 2.1 Basic Parameters Figure 1 shows the front view, side view, and cross-section view of the parallel channel. Because the size of the flow channel is only 8 mm × 8 mm, in some accidents, such as impurities entering the flow channel, it is very likely to cause blockage. The heat source on the left and right is simplified as the surface heat source on the outer wall. See Table 1 for specific parameters such as boundary conditions and initial conditions. Tables 2 and 3 show the variation of physical parameters of water and RAFM steel with temperature, respectively. 2.2 Geometrical Model To simplify the numerical calculation, only one flow channel is simulated. As shown in Fig. 1(a), the coolant channel is 8 mm × 8 mm. The wall thickness on the left side is 3mm, the wall thickness on the upper and lower sides is 7 mm, and the wall thickness on the right side is 9 mm. The −Y-direction is the direction of gravitational acceleration, and the − Z direction is the direction of the velocity vector. During the simulation setup, 13.45 MW/m3 volumetric heat generation was added to the solid area. 0.52 MW/m2 and 0.12 MW/m2 surface heat sources were added to the left and right sides respectively.

Numerical Calculation of Flow Heat Transfer Square tube

205

1200

8×8

Unit: mm Inlet

7 3

8

Blockage

Outlet Y

9

X

7

Z

(b)

Y Z

Annular blockage X

Triangular blockage (c)

Rectangular blockage

(a)

Fig. 1. Parallel channels (a) front view, (b) side view and (c) cross-section view of different blockages including annular blockage, triangular blockage and rectangular blockage.

Table 1. Basic parameters. X dimension

20 mm

Y dimension

22 mm

Z dimension

1200 mm

Flow channel size

8 mm × 8 mm

Material of structure (solid area)

RAFM steel

Pressure

15.5 MPa

Inlet velocity

4 m/s

Inlet temperature

473.15 K

Volumetric heat generation

13.45 (MW/m3 )

Heat flux of left outer wall surface

0.52 (MW/m2 )

Heat flux of right outer wall surface

0.12 (MW/m2 )

Turbulence model

Realizable k-ε model

2.3 Blockage Simulation Method Considering that common gaskets are generally made of rubber, copper, or aluminum, this paper chooses the above materials for the blockage. The location of the blockage in the flow channel is shown in Fig. 1(b). z = 1200 is the location of the inlet and z = 0 is the location of the outlet. To simplify the numerical calculation, three different kinds of shapes blockages were simulated, which are rectangular, triangle, and annular respectively. The front view of the section is shown in Fig. 1(c). By changing the blockage material, thickness (in the Z direction), and height (in the Y direction), the effects of material, thickness, and blockage ratio on flow and heat transfer can be obtained respectively.

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Temperature (K)

Density (kg/m3 )

Specific heat (J·kg−1 ·K−1 )

Thermal conductivity (W·m−1 ·K−1 )

Dynamic viscosity (kg·m−1 ·s−1 )

473.15

874.9

4419.5

0.675

1.38E−04

513.15

825.5

4646.2

0.643

1.14E−04

558.15

755.4

5150.8

0.585

9.44E−05

598.15

666.5

6414.5

0.507

7.84E−05

617.15

598.3

8741.4

0.452

6.88E−05

Table 3. Physical properties of RAFM steel. Temperature (K)

Density (kg/m3 )

Specific heat (J·kg−1 ·K−1 )

Thermal conductivity (W·m−1 ·K−1 )

350

7798

511.37

32.9

473.15

7798

511.37

32.9

573.15

7798

543.16

33.4

673.15

7798

582.54

33

773.15

7798

637.15

32.7

873.15

7798

716.61

32.3

973.15

7798

848.73

31.9

2.4 Method Mesh Sensitivity Analysis and Fully Develop Turbulence Test Four groups of meshes were set up, to verify the mesh sensitivity. Figure 2 shows the temperature distribution along the channel centerline from the inlet (z = 1200 mm) to the outlet (z = 0 mm) without blockage. As can be seen from Fig. 2, the calculated results of the four grids are almost the same. The selected in this paper is 4,754,035. Starting from the inlet, select the horizontal center-line of the flow channel every 50 mm (z = 1200 mm to z = 900 mm) and named line-1 to line-7, respectively, as shown in Fig. 3. It shows that the velocity curves coincide with line-6 and line-7, indicating that turbulence has been fully developed.

3 Results and Discussion 3.1 Blockage Material The effects of different blockage materials on temperature distribution were studied and compared by keeping the blockage area share and thickness constant. Rectangular shape, with 50% blockage and 2 mm was taken into account, as shown in Fig. 4, an expanded view around 600 mm (Z-coordinate) is shown in the same figure. The selected materials to

Numerical Calculation of Flow Heat Transfer 510

mesh-1105155 mesh-2521695 mesh-4754035 mesh-6384675

505

Temperature (K)

207

500 495 490 485 480 475 1.2

1.0

0.8

0.6

0.4

0.2

0.0

z-coordinate (m) Fig. 2. Mesh sensitivity analysis. 5.5 line-1 line-2 line-3 line-4 line-5 line-6 line-7

Velocity (m/s)

5.0

4.5

4.0

3.5

3.0

2.5 0.000

0.002

0.004

0.006

0.008

x-coordinate (m) Fig. 3. Fully developed turbulence test.

be compared were rubber, aluminum, copper, and RAFM steel respectively. The thermal conductivity of rubber, aluminum, and copper is 0.1 W·m−1 ·K−1 , 202 W·m−1 ·K−1 , and 387 W·m−1 ·K−1 , respectively. The thermal conductivity of RAFM steel is given in Table 3. As the thermal conductivity of aluminum and copper is much higher than that of the RAFM steel so that the temperature at the blockage point is lower. At the blockage position, the temperature tends to increase when the thermal conductivity decreases.

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Blockage=0 Al Cu Rubber RAFM

Temperature (K)

520 515 510 505 500 495 490 485 480 1.2

1.0

0.8

0.6

0.4

0.2

0.0

z-coordinate (m)

Fig. 4. Effects of different blockage materials.

3.2 Blockage Thickness The effects of different blockage thicknesses on the temperature distribution were compared by keeping the blockage ratio and materials unchanged, as shown in Fig. 5, an expanded view around 600 mm (Z-coordinate) is shown in the same figure. To study the effect of the thickness we take the blockage ratio = 50%, and rubber as material. In Fig. 5, we can observe that, when the blockage thickness increases the temperature increase as well, which means the thickness is directly proportional to the temperature. Figure 6 shows the distribution of axial streamline with the thickness of 2 mm, 8 mm, and 14 mm, respectively. It can be seen that the rectangular blank area is the blocking position, and blocking upstream flow results in the increase of the temperature. However, a small vortex is formed near the blockage, which makes the temperature drop. Also, a formation of a larger vortex can be noticed near the downstream, which makes the temperature drop more rapidly. The side of the blockage also has a backflow area and a local vortex, and, with the increase of the thickness, the radius of the vortex increases. 2mm 4mm 6mm 8mm 10mm 12mm 14mm

Temperature (K)

600 580 560 540 520 500 480 1.2

1.0

0.8

0.6

0.4

0.2

0.0

z-coordinate (m)

Fig. 5. Effects of different blockage thickness

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Fig. 6. Axial velocity streamlines. The blockage is 2 mm, 8 mm, and 14 mm from top to bottom.

3.3 Blockage Area Proportion To study the effects of thickness and blockage ratio on temperature distribution, rubber material was selected. As shown in Fig. 7, Case one: when the thickness is small such as 2 mm, the blockage ratio has a great influence on the local temperature, and the large blockage decreases the temperature sharply. Case two: Figs. 8, 9 and 10, when the thickness is more than 6 mm, the larger the blockage ratio, the higher temperature at the blockage. This is because when the blockage is thin, the bottom inner wall surface at the blockage place can be cooled by the fluid before and after the blockage, but as the thickness increases, the wall at the center of the blockage is difficult to be cooled. 530

0% 20% 40% 50% 60% 80%

Temperature (K)

520

510

500

490

480

1.2

1.0

0.8

0.6

0.4

0.2

0.0

z-coordinate (m)

Fig. 7. Effects of the different blockage ratio at thickness of 2 mm

D. Li et al.

Temperature (K)

560

20% 40% 50% 60% 80%

540

520

500

480

1.2

1.0

0.8

0.6

0.4

0.2

0.0

z-coordinate (m)

Fig. 8. Effects of the different blockage ratio at thickness of 6 mm

20% 40% 50% 60% 80%

Temperature (K)

580 560 540 520 500 480 1.2

1.0

0.8

0.6

0.4

0.2

0.0

z-coordinate (m)

Fig. 9. Effects of the different blockage ratio at thickness of 10 mm

20% 40% 50% 60% 80%

600

Temperature (K)

210

580 560 540 520 500 480 1.2

1.0

0.8

0.6

0.4

0.2

0.0

z-coordinate (m)

Fig. 10. Effects of the different blockage ratio at thickness of 14 mm

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3.4 Blockage Shape The blockage ratio, material, and thickness are kept unchanged, and the effects of the shape of the blockage on the temperature distribution are compared, as shown in Fig. 11, an expanded view around 600 mm (z-coordinate) is shown in the same figure. To study the influence of the blockage shapes, we take the rubber as the material the blockage ratio is 50% with 2 mm. As shown in Fig. 12, for rectangular blockage, the downstream shows only a larger vortex right after the blockage area. For the triangle, cut from the central line, the left and right blockage of the flow channel is asymmetric. Due to the left-right asymmetry, the fluid mix increases, resulting in a large vortex, also a few small eddies can be observed. In the area between the large and small vortices, the flow is poor, so there is a bit of temperature rise. For annular shape, a jet flow is formed downstream of the blockage, rather than a vortex. 530

Rectangle Triangle Annulus

Temperature (K)

520 510 500 490 480 1.2

1.0

0.8

0.6

0.4

0.2

0.0

z-coordinate (m)

Fig. 11. Effects of different blockage shapes

4 Conclusion In this paper, ANSYS FLUENT numerical calculation software was used to simulate the flow blocking of a single square tube facing the plasma side of the first wall. The effects of blockage material, thickness, share, and shape on velocity distribution and temperature distribution are discussed. The conclusions are as follows: (1) The effects of the blockage material are local. The lower the thermal conductivity of the material, the higher the wall temperature at the blockage position. (2) The thicker blockage area, resulting in a larger wall temperature rises. Besides, when you measure near the center of the blockage, it will be observed that the temperature of the wall is getting higher. Also, in the horizontal direction, whenever the width increases, the temperature rises which means the width is directly proportional to the thickness. With the increase of the thickness, the vortex radius on the side of the block also increases.

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Fig. 12. Distribution of axial streamlines. The shape is the annular blockage, triangular blockage, and rectangular blockage.

(3) If the thickness is less than 8 mm, the smaller the blockage ratio, the higher temperature at the blockage. If the thickness is more than 8 mm, the larger the blockage ratio, the higher temperature at the blockage. (4) The shape of the blockage has a great influence on the downstream flow field. For the rectangular blockage, a larger vortex is formed downstream, while for the triangular blockage, in addition to forming a larger vortex, some small vortices will also be formed. For annular blockage, jets are formed downstream instead of vortices.

References 1. Xu, Y., Lyu, Y., Zhou, H., Luo, G.: A review on the development of the structural materials of the fusion blanket. Mater. Rev. 32(9A), 2897–2906 (2018) 2. Shi, W., Zeng, Q., Li, W., Chen, H.: Primary analysis of radiation damage on first wall and the outboard blanket on equatorial plane for CFETR. Nucl. Tech. 39(12), 120601–120606 (2016) 3. Li, X., Ma, X., Lu, P., Zheng, Y., Xu, K., Liu, S.: Neutronics analyses of water-cooled ceramic breeder blanket for the updated CFETR model. Nucl. Fusion Plasma Phys. 40(3), 275–282 (2020) 4. Yang, X., Ge, Z., Li, Y., Nie, X., Liu, J.: Investigation into the in-box LOCA of water cooled solid breeder blanket for CFETR based on PWR conditions. J. Univ. Sci. Technol. China 47(6), 479–484 (2017) 5. Wang, J., Song, Y., Lei, M., Liu, S., Guo, C., Xu, K., Salah Uddin, K.: Thermal safety analysis of CFETR helium cooled ceramic breeder blanket module with RELAP5. At. Energy Sci. Technol. 51(10), 1778–1784 (2017)

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6. J.H.-j. Liu, T.-c., Yuan, L.-z.: Flow blockage accident analysis for China advanced research reactor. Nucl. Power Eng. 44(5), 32–35 (2006) 7. Song, L., Guo, Y., Zeng, H.: Numerical analysis on transient flow and temperature field during inlet flow blockage accidents of plate-type fuel assembly. Nucl. Power Eng. 35(3), 6–10 (2014) 8. Davari, A., Mirvakili, S.M., Abedi, E.: Three-dimensional analysis of flow blockage accident in Tehran MTR research reactor core using CFD. Prog. Nucl. Energy 85, 605–612 (2015)

Research on the Siting Problem of Economy-Oriented Small Heating Nuclear Reactor Based on System Dynamics Haitao Luo, Changle An(B) , Lu Zhao, and Ruomin Zhang Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., Shanghai, China [email protected]

Abstract. With the advancement of engineering technology and the discussion of the simplification of the contingency planning area for small nuclear reactors, the traditional disruptive factors in the siting process of nuclear energy projects are gradually solved. However, this comes at the cost of increased investment in the project. Therefore, it must be considered whether the economic cost will become a new “disruptive” factor. Research on the economical oriented siting methods of small nuclear reactors can help to provide more information for project siting decisions earlier through economic analysis. This study uses a system dynamics approach to establish a model framework for economy-oriented project site selection, named EOSSD (Economy-Oriented Siting of Small Nuclear Reactors System Dynamics Model). The model framework seeks the theoretically optimal location of a small nuclear reactor site with the goal of minimizing the total cost of heating. The case study takes the small heating nuclear reactor in Qiqihar, Northeast China, which is in the project conception stage as a research object, and the simulation results are analyzed and the following conclusions are drawn: (1) The price of heat supply adopts a fixed unit price, that is, when there is no difference in the rate of heat source (supply point) to each heat exchange station (demand point), small nuclear reactors tend to be arranged at the “economic center of gravity”; (2) When the unit price of heating varies with the distance between the heat exchange station and the small nuclear reactor, the distance is an important factor affecting the cost of heating, and the coordinates of the small nuclear reactor tend to coincide with a certain point; (3) In China, the impact of the product types of small reactors on the economics of site selection is not obvious. Finally, through the analysis and conclusions, the paper discusses two application suggestions: (1) Based on the microscopic perspective, the model with refined parameters will provide a certain degree of reference for the improvement of nuclear reactor design by providing site opinions in terms of cost and economy. (2) Based on the macroscopic perspective, the model results show that the layout of the small stack should be close to or coincide with the “demand”. Simplifying the contingency planning area of small reactors and bringing them closer to the market point will help their economical improvement, which provides economic support for the “relaxation” of relevant policies. Keywords: Small reactors · Nuclear heating · Siting · Economics

© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 214–224, 2023. https://doi.org/10.1007/978-981-19-8899-8_22

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1 Introduction Although people from all walks of life have reached a broad consensus on the emission reduction potential of nuclear energy and are focusing on promoting the deployment of nuclear energy development projects, just as the 2021 Chinese government work report changed the expression of nuclear power development for the first time in many years, referring to the “active and orderly development of nuclear power”. However, the development of nuclear energy projects still faces many challenges. In 2017, the “Plan for Clean Heating in Winter in the Northern Region (2017–2021)” jointly formulated by ten departments including the National Development and Reform Commission, National Energy Administration, and the Ministry of Environmental Protection of China clearly stated that “research and explore nuclear energy heating, and promote active nuclear power units to supply surrounding areas, heat and safe development of heating demonstration” requirements. In this context, in the face of the strong demand for clean heating in northern China, nuclear-related companies have successively proposed a variety of nuclear energy heating reactor solutions, which have been installed in Heilongjiang, Jilin, Hebei, Shandong, Ningxia, Qinghai and other provinces and regions. The universal selection of plant sites and industrial promotion of nuclear energy heating pilot projects have been carried out. With the changes in social and economic conditions, the selection of small reactor sites is more economical than the site selection of large nuclear power plants that focus on safety including DHR400 and CAP200, are facing market challenges in the process of large-scale commercial use. With the advancement of engineering technology, geological and hydrological problems that may form disruptive factors in the siting process of traditional nuclear energy projects can gradually be solved through engineering measures, but at the cost of increasing project investment, it must be considered whether the economic cost will become a new “disruptive” factor. At present, in the stage of general election of nuclear energy project sites, technical and economic professionals are increasingly required to intervene in advance. However, in the stage of general election, there are sometimes problems such as difficulty in collecting project data and shallow design depth, which are not enough to support technical and economic majors to carry out in-depth work. This has become a prominent contradiction in the current stage of nuclear energy universal suffrage. This study will construct an economical-oriented project system dynamics site selection method framework, with the overall goal of the lowest heating cost, and seek the best location of the site for small reactor heating under different modes. The model will provide a certain degree of support for the site selection of nuclear heating reactors, and become a useful exploration and supplement to the current nuclear heating reactor site selection methods.

2 Literature Review In China, there are a series of regulations and guidelines for the site selection of large nuclear power plants. The essential is to protect public health, protect the environment, and at the same time ensure social security and stability [1]. The main factors include radiation protection, atmospheric dispersion, population distribution, and hydrogeology.

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Internationally, the US Nuclear Regulatory Commission (NRC) stipulates that nuclear power plants must meet the standard requirements in ten aspects such as Geological Seismic, atmospheric conditions, exclusive distribution, population distribution, and ecological and biological area protection. In the case of safety as the most important factor affecting the siting decision of nuclear power plants, foreign scholars such as A. Alons, Ilan Yaar and Ping K wan have focused on the influence of the combination of extreme natural phenomena and seismic geological factors on the comprehensive evaluation of nuclear power plants sites [2–4]. Shaoyu Ji summed up the influencing factors of large-scale nuclear power site selection by examining domestic and foreign documents, there are 25 natural factors, safety factors and technical factors in total. Economic factors only include the payback period, construction investment cost and on-grid electricity price [5]. Regarding the site selection of small reactors, Peng [6], Wang [7] and others believe that the comprehensive utilization of nuclear energy should not be limited to power generation, and small reactors have good prospects in industrial gas heating, urban heating, seawater desalination, etc., and proposed that the biggest advantage of small reactors is that they are close to the community, can make full use of the remaining heat, and reduce the distance loss. Zhiyuan Zuo divided the heating mode of nuclear energy into three types: the heating mode in which power generation is the main and the supplemented by heating; the heating mode in which heating is the main and the supplemented by power generation; single heating mode [8]. The first heating mode usually adopts the method of extracting steam from the steam turbine of a large nuclear power plant. Since the large nuclear power plant must be far enough away from the city, the investment in the long-distance heating pipe network is large, and the long transmission distance also causes substantial heat loss. However, due to the limitation of heat distributing network capacity and the reduction of demand, the installed capacity of the latter two heating modes is much smaller than that of large nuclear power plants. The reactor type is generally designed with lower power and lower parameters, which also greatly improves its safety, operational stability and site adaptability. As a result, small stacks can be built closer to centers of heat load, such as residential areas. It can be seen from the above-mentioned comparison of relevant literature on the siting of large-scale nuclear power and nuclear energy heating reactors at home and abroad that the influencing factors to be considered in the siting of the two are basically the same, but when the site adaptability is higher than that of large nuclear power plants, the location of the heat supply reactor should fully consider the heat load center and user needs, also its economic orientation becomes very obvious. The reasons are: on the one hand, from the perspective of the supply port, the heating pipe network attached to the nuclear energy heating reactor is essentially a “local area network”. The transmission and supply side of heat cannot reconcile the price of electricity from multiple sources and then sell it uniformly, as the power supply company does. Therefore, urging the heat source unit to reduce the price of heat supply is its essential need to improve its own profitability; from the heat side, due to the variety of heat production methods, nuclear energy heating is highly substitutable, consumers are significantly more sensitive to heat prices than electricity, so heat source units need to reduce costs to improve their competitiveness. On the other hand, the current site selection of heating reactors is still carried out with

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reference to the siting criteria of large nuclear power plants, and a set of economicoriented siting methods that meet the needs of the market and owners has not yet been established. With the continuous improvement of the supporting regulations, standards and approval systems for small reactors, the 5 km development restriction area usually required by traditional large nuclear power plants and the off-site emergency response will be greatly relaxed, which will greatly ease the development of small reactors. The thermal reactor can be located in a wider area closer to the load center, which also provides the possibility to reduce heating costs and improve economics. To sum up, it is necessary to explore the economic factors of siting of small reactors (represented by nuclear heating reactors), with emphasis on economics, and to establish a set of siting models suitable for market demands.

3 Methodology This study will make basic assumptions for the model and construct an economy-oriented siting model under the heating mode. The heating mode referred to in this paper consult the division method of Zuo [8], selects a single heating mode and a combined electricheating mode. These two models can be abstracted into single-product and multi-product location problems. Single-product siting problem, that is, a nuclear heating plant site only provides hot steam. 3.1 Model Method and Assumptions In this paper, the system dynamics method is chosen for modeling. The System Dynamics (SD) method was founded in the 1960s by Jay W. Forrester. SD uses causal loop diagrams and stock-and-flow diagrams to describe systems of complex relationships in the model. System dynamics is widely used in management decision and supply chain research [9]. The model is constructed using Vensim, a graphical programming software dedicated to system dynamics. First, simplify the model to make it abstract and feasible. Combined with the actual situation of heating in the northern region, the general assumptions of the model are as follows: (1) The surface is a single uniform plane, and each position point can be converted into a point in the rectangular coordinate system. (2) The scope of cost calculation is from nuclear heating reactor to thermal substation. This assumption is based on the operation mechanism in which the products emitted by the heat source are shunted by the thermal station and then enter the user node. Since the pipeline network from the thermal station to the user node is not constrained by any site conditions, the cost of these household pipeline networks is a common and fixed part, which will not have a biased impact on investment and site selection, and will not be included in the model cost calculation. Consider. (3) The commodity flows at a uniform speed in the pipeline, and the loss of hot steam in the pipeline transportation process is not considered for the time being. This assumption is based on the fact that the steam from the thermal station node is

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transported over a long distance, its pressure and temperature will decrease, but its weight will remain the same. (4) During the model simulation period, the output and consumption of heat and power products are fixed values, the supply and demand are equal, and the market is cleared. 3.2 General Form of EOSSD Model Heat reactor siting methods are modeled with the aid of system dynamics. The modeling idea should be to calculate the initial heat supply cost according to the initial assumed heat source coordinates, use the exact center-of-gravity approach, iterate to find the optimal cost, and finally reach the process of selecting the site range [10]. Assuming that there are n places in the intended area that need one or more products, seek the best location of the heat source center to minimize the transportation cost. Put the proposed site area into the positive coordinate grid, each coordinate unit represents 2 km, the model can be transformed into an objective programming problem. The cost calculation equation (1) and the heating stack coordinate calculation equation (2) and (3) are expressed as follows: Minci =

n 

pi qi di

(1)

i=1

In the formula: i is the serial number of the thermal power station, pi is the transportation cost of transporting the unit product to the ith thermal power station in unit distance, qi is the quantity of products delivered to the ith thermal power station, and di is the straight-line distance from the coordinates of the center of the heating stack to the ith thermal station. ci is the cost of the heating stack to deliver the product to the ith thermal station. Xc = ( Yc = (

n  xi pi qi

i=1 n  i=1

di

)/(

yi pi qi )/( di

n 

i=1 n 

pi qi /di )

(2)

pi qi /di )

(3)

i=1

In the formula: the central coordinate of the heating reactor is (Xc , Yc ), and its initial value is (0, 0), xi is the abscissa of the serial number i thermal station, and yi is the ordinate of the serial number i thermal station. The general model framework is established using Vensim as shown in Fig. 1. Xc and Yc in the figure are the center coordinates of the heating stack.

4 Empirical Analysis Referring to the relevant data of Qiqihar small heating nuclear reactor site selection, four heat supply demand points in this area are randomly selected, and for the single product site selection problem, two cases of linear and nonlinear heating unit price are discussed respectively.

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C Δc

ci

pi

qi

di xi

Yc

yi

Xc

Fig. 1. The general form of the model

4.1 The Problem of Location Selection When the Rate Is Linear The simplest case of single-product site selection is discussed first. It is assumed that the heating unit price is constant, that is, the rate is linear. The central coordinates, heat demand, and unit price of heat supply of the four heat exchange stations are shown in Table 1. The aforementioned model form is instantiated to obtain the model form shown in Fig. 2. Table 1. Overview of heat exchange station Number

Coordinate

Heating demand (GJ)

Heating unit price (yuan/GJ)

14

(2, 1)

24,882.00

8.264

20

(3.3, 3.5)

24,882.00

8.264

28

(3.8, 5)

622.05

9.444

43

(5.5, 5)

29,029.00

9.444

The model is simulated for 100 periods with a step size of 1, and each state variable is assigned an initial value of 0. The result data obtained from the calculation simulation are shown in Table 2 and Fig. 3. The simulation results of the model are basically stable from the 15th time. From the 55th time until the end of the iteration, the calculation result is stable at the point (3.3, 3.5), and the corresponding total cost increase (that is, the cost incurred in each period) is 1.318679 million yuan. Through this simplest site selection situation, it can be proved that the construction of the basic framework of the model is effective.

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Fig. 2. Site selection model of heating reactor with linear rate and single product

Fig. 3. Iterative results of coordinates and costs

4.2 The Problem of Location Selection When the Rate Is Non-linear On the basis of the linear model framework and case data in the previous section, this section will continue to discuss the single-product heating reactor location problem when rates are non-linear. Non-linear rate changes the unit price of heating, making it a function of variables such as distance and demand. Usually, the tax-included heating price of coal-fired heating units adopts the costprofit pricing method, which is in the form of (fuel and power cost + direct labor cost + period cost)/heat supply or heating area × (1 + cost profit rate) × (1 + tax rate). The

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Table 2. Simulation results of the model with linear rate and single product Iteration Coordinate Coordinate Cost X Y added value

Iterations Coordinate Coordinate Cost X Y added value

0

0.00000

0.00000

352.35889 8

3.40370

3.48700

132.66527

1

3.08011

2.49625

156.06509 9

3.36587

3.49288

132.35913

2

3.08011

2.49625

156.06509 10

3.36587

3.49288

132.35913

3

3.37511

3.09981

139.27744 11

3.34222

3.49457

132.17764

4

3.37511

3.09981

139.27744 …

…

…

…

5

3.45535

3.41835

133.46484 55

3.30000

3.50000

131.86789

6

3.45535

3.41835

133.46484 …

…

…

…

7

3.40370

3.48700

132.66527 100

3.30000

3.50000

131.86789

cost calculation method in the minimum spanning tree method of heat source location and heat network wiring considers that the heat source and source network cost are approximately proportional to their heat loads. Combining the above two calculation methods, the calculation formula of pi when the rate is nonlinear is obtained: pi = (1 + π ) · (1 + ρ) ·

αqit + βqik d i qi

(4)

In the formula: i is the serial number of the thermal power station, pi is the transportation cost of transporting the unit product to the ith thermal power station in unit distance, and di is the straight-line distance from the coordinates of the center of the heating reactor to the ith thermal power station. π is the cost profit margin, which is usually not higher than 3% due to the government pricing of urban central heating in China, ρ is the ratio of direct labor cost and fuel cost to the period cost, αqit is the period cost of heat investment, βqik di is the period cost of source network investment, α, β, t, k is the respective cost parameter and empirical data is usually used. Substitute Eq. (4) and the formula for calculating the distance between two points into Eq. (1) to obtain the non-linear cost calculation formula of the rate: Minci = (1 + π)(1 + ρ)

n 

   αqit (x − xi )2 + (y − yi )2 + βqik (x − xi )2 + (y − yi )2

i=1

(5) For the point position determined by any data, the variables other than x and y in formula (5) are all fixed values. It has been verified that the Hessian matrix of formula (5) > 0, so the formula is a convex function, that is, its local minimum value is the global minimum value. Therefore, continue to use the data of the previous example to simulate the model for 100 periods with a step size of 1, assign the initial value of 0 to each state variable, and calculate the minimum cost and coordinates as shown in Table 3.

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Iteration Coordinate Coordinate Cost X Y added value

Iterations Coordinate Coordinate Cost X Y added value

0

0.00000

0.00000

566.95026 8

3.37350

3.38761

168.72858

1

3.11009

2.58573

187.27885 9

3.33709

3.47100

166.88185

2

2.96898

2.35697

196.94510 10

3.35651

3.44798

167.54274

3

3.35406

3.14988

172.28337 11

3.32289

3.48121

166.49783

4

3.19956

2.84828

179.23857 …

…

…

…

5

3.39410

3.39875

168.90784 53

3.30000

3.49999

165.84935

6

3.34584

3.20041

171.28439 …

…

…

…

7

3.36321

3.45400

167.56738 100

3.30000

3.49999

165.84935

It can be seen from the simulation results of the model in Fig. 4 that the numerical values are basically stable since the 11th time. From the 53rd time until the end of the iteration, the calculation result is also stable at the point (3.3, 3.5), and the corresponding total cost increase (that is, the cost incurred in each period) is 1.658494 million yuan. Compared with the case where the rate is linear, when the unit price of heating varies with the heating load and the heating distance, the heating cost borne by the user will increase slightly, but this situation can better reflect the relationship between market supply and demand by the heat source.

Fig. 4. Iterative results of coordinates and costs

5 Conclusions This study proposes a framework of an economy-oriented siting method through a simplified system dynamics model (EOSSD), and discusses the economics of small heating nuclear reactor siting. The framework seeks the theoretically optimal location of small

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nuclear reactor sites with the lowest total heating cost, which will help to provide more information for decision-making in the early stage of Chinese small reactor project siting through economic analysis. And at the same time, the investment in the construction of small reactors within the scope also has certain reference significance to the world. The inspirations that can be obtained from the analysis of calculation methods and cases are as follows: (1) When the price of heat supply adopts a fixed unit price, that is, there is no difference in the rate of heat source (supply point) to each heat exchange station (demand point), small nuclear reactors tend to be arranged at the “economic focus”; (2) When the unit price of heating varies with the distance between the heat exchange station and the small nuclear reactor, the distance is an important factor affecting the cost of heating, and the coordinates of the small nuclear reactor tend to coincide with a certain point; (3) The impact of the type of small reactor on the economics of site selection is not clear in China, since we eliminated differences in different reactor types in our study. In fact, although this model framework is relatively shallow, it has certain application value. At the micro level, the EOSSD model framework can be combined with other engineering cost databases. The refined model will provide a certain degree of reference and guidance for nuclear reactor design improvement by providing site suggestions on cost and economy, so as to promote the economic improvement of the model by means of cost quota design. In the macro aspect, the model results show that the layout of small-scale reactors should be close to or coincide with the “demand”. Simplifying the emergency planning area of small-scale reactors to make them closer to the market point will help improve their economy, which provides economic support for the “loosening” of relevant policies. The EOSSD model framework can be refined in the future, and the in-depth research directions include: (1) further discuss the situation of heat and power combining; (2) increase the time parameter to study the impact of heat load changes in different development periods on the cost of heat source site selection, in order to solve the problem of site selection with the lowest total heating cost during the period; (3) The current nuclear heat supply reactor is usually not built in the central area, and it is necessary to consider the location of the nuclear heat supply reactor at the periphery of the heat exchange station; (4) Current unit transportation cost function of product does not really reflect the actual situation. In the future, the function pi of cost should be optimized according to the investigation of the actual application area.

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References 1. Xiaoqiu, C., Huiguo, Y.: Some issues on radiological consequences analysis in site selection for nuclear power stations. Nucl. Saf. 4, 31–36 (2006) 2. Alonso, A.: Infrastructure and Methodologies for the Justification of Nuclear Power Programmes. Elsevier (2012) 3. Yaar, I., et al.: Possible Sites for Future Nuclear Power Plants in Israel, vol. 298, pp. 90–98 (2016) 4. Wan, P.K., Carson, A.C., Chan, D.W.: Climate change considerations in sustainable development of nuclear power plants in the United States. In: International Conference on Nuclear Engineering, vol. 49347, pp. 403–407 5. Shaoyu, J.: Site Selection and Decision Making Model of Inland Small Modular Reactors Considering Public Participation. North China Electric Power University, Beijing (2019) 6. Peng, J., Peng, F.: Development trend of comprehensive utilization of nuclear energy. China Sci. Technol. Inf. 2, 107–108 (2019) 7. Jianqiang, W., Zhimin, D., Hongjie, X.: Research status and prospect of comprehensive utilization of nuclear energy. J. Chin. Acad. Sci. 34, 460–467 (2019) 8. Zuo, Z.: Discussion on the development and advantages of urban nuclear heating. Res. Urban Constr. Theory 10, 152–153 (2018) 9. Forrester, J.W., Collins, F.: World dynamics. J. Dyn. Syst. Meas. Control 339 (1972) 10. XingDong, M.: Application of System Dynamics in Supply Chain Modeling. Southwest Jiaotong University, ChengDu (2006)

Technical and Economic Analysis of Nuclear Heating Based on HPR1000 Qian Yu(B) , Jiang Hu, Lijuan Chen, Xin Shang, and Mei Rong China Nuclear Power Engineering Co., Ltd., Beijing, China [email protected]

Abstract. HPR1000 is China’s third-generation nuclear power technology with independent intellectual property rights, and is the main technology choice for the safe and efficient development and mass construction of nuclear power in China. This paper investigated the situation of nuclear heating at home and abroad, combined with the demand of heating and steam supply, analyzed the market demand and commercial layout. The technical feasibility, safety reliability and economic suitability of HPR1000 heating and steam supply are studied and demonstrated. Through the analysis of typical cases, the research results show that the proposed scheme of heating system can meet the requirement of heat source reliability. By adding isolation between the secondary circuit and the heating system, the migration of radioactive materials into the heating system can be effectively prevented and the environmental safety impact assessment is acceptable. Through adaptive modification of the control system, automatic regulation of reactor, steam turbine and heating system can be realized to ensure the safe and reliable operation of the unit. In terms of economy, the economic benefits of HPR1000 heating can be competitive with coal, better than gas. The calculated price level of steam supply also has certain market competitiveness. It provides research support for the follow-up HPR1000 heating project. Keywords: HPR1000 · Nuclear heating · Technical and economic

1 Introduction Nuclear energy heating has been included in the national clean heating plan, nuclear energy steam supply and heating have been listed as encouraged industries by the state, and nuclear energy heating will become an important expansion direction for the comprehensive application of nuclear energy in the “14th Five-Year Plan” and in the medium and long term. HPR1000 is a progressive third-generation nuclear power technology based on mature technology, with good maturity, and is the main technical choice for the safe and efficient development and batch construction of nuclear power in China. This paper investigates and understands the situation of nuclear energy heating at home and abroad, combines the domestic heating and steam supply needs, and analyzes the market demand and commercial layout. It provides a reference for the subsequent development and utilization of the heating market, especially the expansion of the application of HPR1000 in heating [1, 2]. © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 225–232, 2023. https://doi.org/10.1007/978-981-19-8899-8_23

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2 Potential Market Analysis Nuclear energy is a safe, reliable, clean and economical energy source. Worldwide, nuclear energy deserves to be more widely used because of its low resource consumption, low environmental impact and strong supply capacity. Exploring the use of nuclear energy to supply steam and district heat to industry and replace traditional energy combustion and heating is a potentially important way to reduce pollutant emissions, improve the atmospheric environment, and save energy and reduce emissions [3]. There is a large demand for heating in northern China, but the current heat source is mainly coal-fired, with the implementation of the national clean heating plan and carbon peak, carbon neutrality action plan, coal-fired heat source will gradually be replaced by low-carbon clean heat source [4]. Due to the difference in energy supply and heating time, the heat price varies greatly from place to place. For example, HPR1000 is built in the northern region, combined with the current long-distance heating technology, the market is easier to implement. With the increasing improvement of people’s living standards, some conditional areas in the south also have heating needs. China’s industrial steam supply market capacity is large, and the users have a high tolerance of the industrial steam supply price. According to their different products, the acceptable steam supply price range is also large. In the economically developed southeast coastal areas, the price of industrial steam is generally 200–280 yuan/ton; in the southwest and northwest areas, the price of industrial steam is generally 100–150 yuan/ton. At present, HPR1000 is mainly built in coastal areas. According to preliminary understanding, there are seven petrochemical industrial bases in coastal areas of China, of which the total amount of steam planned for the Lianyungang petrochemical base in Jiangsu Province exceeds 10,000 ton. It can be seen that the petrochemical industry base and the coastal area industrial base in China may serve as a potential steam supply market for HPR1000.

3 Technical Feasibility The technical feasibility is mainly analyzed from the site layout requirements, the heating scheme feasibility, the feasibility of the key technology and the environmental safety impact. 3.1 Site Layout Requirements Referring to the current level of long-distance heating technology, HPR1000 can heat urban areas within a distance of about 100 km in winter (the heating distance is not limited by heating parameters, but determined by the technical and economic feasibility of the heat network); Medium pressure saturated steam of about 2.5 MPa can be provided for users within a distance of 10 km, and low-pressure saturated steam of about 1.0 MPa can be provided for users within a distance of 40 km. The special requirements of the layout of nuclear power plant sites make the problem that nuclear energy heating usually requires steam or hot water long-distance transportation is no longer a constraint, but the increase in the transmission distance will lead to

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an increase in the cost of heat network transportation, which will have a certain impact on the economy of nuclear energy heating projects, and the specific projects need to be determined in combination with technical and economic feasibility studies. 3.2 Heating Scheme The heating medium of the urban heat network is usually high temperature hot water, and when HPR1000 is used, the heat source can be pumped by steam turbines. The steam turbine pumps steam into the heat grid heater to heat the circulating water of the heat grid, releases the heat to form a hydrophobic water, and the hydrophobic water is heated by the heat grid hydrophobic cooler after the condensate is heated, and finally sent back to the steam turbine condenser. The HPR1000 heating system scheme is similar to the coal-fired cogeneration heating system scheme, and is consistent with the foreign nuclear power heating technology route, and the technology is relatively mature. Due to the radioactive risk of HPR1000 secondary circuit steam in the event of SG heat pipe rupture accident, when HPR1000 is used to supply industrial steam, the secondary circuit steam cannot be directly used for external supply. The steam supply system is proposed to adopt steam conversion technology, using HPR1000 main steam to produce industrial steam through steam conversion equipment, and increase the isolation between the secondary steam and industrial steam. The technology has not yet been verified by the project, but it has been reviewed by domestic industry experts, believing that the technology is reasonable and feasible. 3.3 Feasibility Analysis of Key Technologies The proposed heating system scheme study shows that it can meet the requirements of heat users for the reliability of heat sources; Through the adaptive transformation of the control system, the automatic adjustment of reactors, steam turbines and heating systems can be realized to ensure the safe and reliable operation of the unit; In terms of key heating equipment, the key heating equipment is heat network heater and heat network circulating water pump, which are widely used in conventional urban heating systems and have mature and reliable technology. The core equipment of the steam supply scheme is the steam conversion equipment, which usually includes superheaters, steam generators, hydrophobic tanks, feed water preheaters and feed water deaerators. After researching the mainstream potential equipment suppliers in China, there is no supply performance of large-scale steam conversion equipment, but the relevant equipment program research work has been carried out. 3.4 Environmental Safety Impacts Since the pressure on the circulating water side of the heat grid is greater than the pressure on the auxiliary steam side, it is difficult for radioactive materials to enter the circulating water of the heat network under normal operating conditions, so the radioactive materials in the circulating water of the heating network are negligible, and

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the radiation impact on the heating user is negligible. In terms of steam supply, by adding the isolation between the second circuit and the heating system, it can effectively prevent the migration of radioactive materials to the heating system, and the environmental safety impact assessment is acceptable [5].

4 Economic Analysis 4.1 Economic Analysis of the Heating Scheme According to the proposed design plan description, compared with the Pure Generator Set of HPR1000 in the heating plan, the first station of the supporting heat network and the conventional island steam extraction and transformation facilities have been added. At the same time, due to heating and steam pumping, some freshwater production facilities need to be added, and it is initially estimated that this part of the adjustment will cause the total investment to increase by about 300–500 million. Due to the impact of the site conditions, the total investment of the construction project varies greatly in the cost of earthwork, roads and drainage projects. Referring to the current cost level of the HPR1000 nuclear power unit under construction in China, considering the impact of site differences and the limitation of the depth of the scheme, the total investment of the proposed plan corresponding to the HPR1000 pumped steam heating is about 43 billion to 45 billion yuan (the investment does not include the heat network transmission and distribution part, and the boundary is the plant wall) [6]. There are seven petrochemical industry bases in China’s coastal areas, among which the total amount of steam planned for the petrochemical base in Lianyungang Petrochemical Base in Jiangsu Province exceeds 10,000 tons, and the area has a certain scale of heating demand. Therefore, the economic analysis of this part assumes that the construction site of a nuclear power project is located in the coastal area of Jiangsu Province, Jiangsu Province belongs to the economically developed area, the industrial demand is strong, and its coal-fired benchmark electricity price is at the national median level, which has a certain representativeness [7]. The economic analysis part is mainly based on the conclusions of the financial evaluation, and the national economic evaluation is not considered for the time being. According to the proposed plan, the market competitiveness is judged by measuring the price of the product. • Proposed design input main parameters: Electrical rating: 1200 MWe. The annual heating capacity of the two units: 1092 × 106 GJ/year. Heating and steam extraction leads to a decrease in power generation: 122 MWe. Annual water consumption: 25 ten thousand tons. • Main parameters of financial analysis: Construction investment: 17,000 yuan/kW (the completed price investment is about 41 billion yuan, Excluding the deductible VAT during the construction period).

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Return on capital: 8%. Other parameters: Same as the HPR1000 conventional generator set. • Estimated results: Taking the current coal-fired benchmark electricity price of 391 yuan/MWh (including tax) in Jiangsu Province as the input parameter, under the premise of ensuring the rate of return on project capital of 8%, the calculated unit price of heating is about 44 yuan/GJ, which is close to the heat price of coal-fired units (which is inferior in the market price range) and better than the gas heat price。If the current heat price is reduced by 20% to 35 yuan/GJ, the electricity price needs to rise by 1.65% (397.5 yuan/MWh) or increase the annual utilization hours by 1.78% (7125 h/year), which is very competitive compared with the traditional coal-fired heating supply. For regional project construction with a coal benchmark price higher than 397.5 yuan/MWh, it also has a high market competitiveness. 4.2 Economic Analysis of Steam Supply Scheme According to the proposed design plan, compared with the HPR1000 pure generator set, the steam supply scheme adds steam conversion stations and conventional island steam extraction and transformation facilities. At the same time, due to the large steam supply, it is necessary to increase some freshwater production facilities (seawater desalination, etc.), and it is initially estimated that this part of the adjustment will cause the total investment to increase by about five hundred million to 1 billion yuan [8]. Due to the impact of the site conditions, the total investment of the construction project varies greatly in the cost of earthwork, roads and drainage projects. Referring to the current cost level of the HPR1000 nuclear power unit in China, considering the impact of site differences and the limitation of the depth of the scheme, the total investment of HPR1000 steam supply corresponding to the scheme described is about 43.5 billion to 45 billion yuan (the investment does not include the steam transmission and distribution part outside the plant, and the boundary is the plant wall). It should be noted that, due to the depth of the programme, this level of investment is a preliminary estimate of the proposed programme. The impact of the heating and desalination scheme on the scale and design scheme of the original nuclear island, conventional island and BOP of the nuclear power plant is not considered, and the costs related to the control of the reactor under the heating scheme are not considered. In addition, the investment in heating and steam supply related facilities and supporting water production facilities will also change accordingly with the adjustment of the heating scale. Financial evaluation: • Proposed design input main parameters: Electrical rating: 1200 MWe. The annual steam supply capacity of the two units: 4.8 million tons/year.

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Heating and steam extraction leads to a decrease in power generation: 175 MWe. Annual water consumption: 4.88 million tons. • Main parameters of financial analysis: Construction investment: 17,000 yuan/kW (the completed price investment is about 41 billion yuan, Excluding the deductible VAT during the construction period). Return on capital: 8%. Other parameters: Same as the HPR1000 conventional generator set. • Estimated results: Taking the current coal-fired benchmark electricity price in Jiangsu Province 391 yuan/MWh (including tax), 410.6 yuan/MWh (up 5%), 371.5 yuan/MWh (5% down) as input parameters, under the premise of ensuring the return on project capital of 8%, the calculated unit price of steam supply is about 134 yuan/t, 76 yuan/t, 191 yuan/t (the calculated steam supply price is the factory price, excluding the pipeline network transmission and operation and sales parts). According to relevant surveys, in the economically developed areas of the southeast coast, the terminal price of industrial steam is generally 200–280 yuan/ton. Judging from the calculation results, under the premise of ensuring a reasonable transportation distance, the steam price has a strong market competitiveness in the southeast coastal area. Electricity price goes down by 1% (387 yuan/megawatt hour), and the steam price will increase by 8% to 145 yuan/t, which still has a certain market competitiveness compared with the current market price. For regional project construction with a coal-fired benchmark price higher than 391 yuan/MWh, it has high market competitiveness and good economic benefits. 4.3 Brief Summary According to the design plan, the main output product of the heating project is electricity, and the output of heating products leads to a small proportion of power generation reduction, of which the reduction of steam supply power accounts for 7.30% of the total power, and the reduction of heating power accounts for 5.08% of the total power. Therefore, the change of electricity price is very sensitive to the unit price of heating products, and the results of financial analysis can also be verified that the fluctuation of electricity prices has a very large impact on heating products. If the electricity price rises by 1.65%, the unit price of heating drops by 20%. The unit price of heating products in this paper is calculated under the premise of ensuring the return on project capital of 8%, and the LPR has been declining in the form of the recent decline. If the project return rate can be accepted by 7%, under the same parameter, the estimated unit price of heating is about 25 yuan/GJ, the unit price of steam supply is about 90 yuan/t (including tax), the unit price of heating can be greatly reduced, heating and steam supply will have a very strong market competitiveness. It should be noted that the above economic analysis conclusions are based on the calculation results of specific schemes and specific conditions, and the heat price and

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industrial steam price of the current market are generally linked to the price changes in the coal-fired market, and the calculation does not consider the impact of the transmission and distribution network. Therefore, nuclear energy heating is generally feasible, but the landing of specific projects also needs to be further analyzed in combination with the site conditions and heating environment. It is worth noting that under the new normal of the economy, China’s overall economy has changed from high-speed growth to medium-high-speed growth, the level of interest rates has been declining, and the investment income of various industries has also shown a downward trend, and investors are bound to lower the expected yield in the future. The economic evaluation and calculation period of nuclear power heating projects is only 30 years, while the design life of the third generation nuclear power plant is 60 years. Therefore, if the overall economic environment and the complete operation period of nuclear power plants are considered, the actual economic benefits of nuclear energy heating will be better than the calculated economic benefits, the profitability will be stronger, and it will have better economic feasibility [9, 10].

5 Conclusions From the perspective of technological innovation, nuclear energy heating is the comprehensive development and innovative utilization of nuclear energy. From the perspective of economic development, nuclear energy heating projects can provide employment opportunities for tens of thousands of people during the construction period, which will further promote the development of local transportation, communications, building materials, education and other municipal facilities and welfare services, and is of great significance to accelerating local economic development. From the perspective of environmental protection, nuclear energy heating projects have significant environmental benefits in reducing sulfur dioxide, carbon dioxide and nitrogen oxide emissions, and are an important direction for building a low-carbon energy system. It can be seen that under the set boundary conditions, the construction of nuclear heating project is feasible and the economy is acceptable. However, if the boundary conditions change greatly, coupled with the influence of the supporting pipe network and other factors, the specific project also needs to be further analyzed and calculated combined with the plant site conditions and the heating environment.

References 1. IAEA: Market Potential for Non-electric Applications of Nuclear Energy (2002) 2. International Energy Agency: Nuclear Power in a Clean Energy System. IEA Publications, Vancouver (2019) 3. Leurent, M., da Costa, P., Rämä, M.: Cost-benefit analysis of district heating systems using heat from nuclear plants in seven European countries. Energy 149, 454–472 (2018) 4. Li, Y., Bai, Y., Han, S.: Analysis of current status and development trend of nuclear district heating. Progress Report on China Nuclear Science & Technology, vol. 6 5. VisualAir. 2019 World air quality report: Region & city PM2.5 ranking (2019)

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6. Wang, J., Dai, Z., Xu, H.: Research status and prospect of comprehensive utilization of nuclear energy. Bull. Chin. Acad. Sci. 34(4), 460–467 (2019) 7. Cholewinski, M., Tomków, Ł: Modeling of a nuclear combined heat and power station supplying heat to remote municipal customers: the case of Poland. J. Power Technol. 98(3), 255–266 (2018) 8. Yan, Y., Tian, J., Cai, Q.: Technical and economic analysis of deep pool reactor district heating. Gas Power 13(4), 47–53 (1993) 9. Hao, W., Zhang, Y.: Research and development of NHR200-II. China Nucl. Power 12(5), 518–521 (2019) 10. IAEA: Guidance on Nuclear Energy Cogeneration. IAEA, Vienna (2019)

The Application of Renal Dynamic Imaging in Measuring Renal Function of En-Bloc Pediatric Kidneys Transplanted into Recipients Ruolin Wu1 , Daijuan Huang1 , Zhendi Wang2 , Kun Li1 , Fan Hu1 , Cheng Wan3 , Yajing Zhang1 , Xiaoli Lan1 , Zairong Gao1(B) , and Xiaotian Xia1(B) 1 Department of Nuclear Medicine, Union Hospital, Tongji Medical College, Huazhong

University of Science and Technology, Wuhan, China {gaobonn,xiaotian_xia}@hust.edu.cn 2 Department of Urology, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China 3 Department of Nephrology, Union Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, China

Abstract. Purpose: En-bloc kidney transplantation (EBKT) from pediatric donors involves transplantation of two kidneys from a deceased pediatric donor, en-bloc, into a recipient. Little is known about accurately reflecting measured renal function in EBKT patients. Long-term allograft outcomes are linked to renal function after transplantation. The primary purpose of this article is to assess the association between 99m Tc-DTPA-based renal dynamic imaging and posttransplantation renal allograft function. Methods: This was a retrospective study, analyzed imaging data of EBKT patients who underwent 99m Tc-DTPA-based renal dynamic imaging at our institute from April 2017 to November 2019. All subjects were patients after their first kidney transplant. Electronic patient records were examined for follow-up data. The data were analyzed using χ2 test, t-test, and Mann-Whitney U test. The Pearson correlation test was used to investigate the correlation between two variables. Results: Sixteen patients received en-bloc pediatric kidney transplants of which ten females. The mean age was 30.3 ± 10.4 years and the mean recipient weight was 50.6 ± 10.2. Glomerular filtration rate measured by renal dynamic imaging Gate’s method (gGFR) showed a correlation with the time intervals post transplantation, serum creatinine, blood urea nitrogen, cystatin-C. The gGFR was much higher in the later-period group than in the early-period group (107.64 ± 27.54 ml/min vs 52.88 ± 17.86 ml/min, P < 0.001). Some difference was obtained for estimated glomerular filtration rate (eGFR) over gGFR in the early-period group with statistical significance (67.50 ± 32.23 ml/min vs 52.88 ± 17.86 ml/min, P = 0.044). Furthermore, gGFR increased significantly in patients with normal serum creatinine levels (P = 0.008). However, there was no significant difference in serum creatinine between the time-interval groups. No significant intergroup based on gGFR value (60 ml/min or 90 ml/min) differences in serum creatinine values were observed. Conclusion: Renal dynamic imaging could sensitively reflect renal function changes rather than serum creatinine and estimated glomerular filtration rate (eGFR) in EBKT patients, especially in the 12 months after transplantation. © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2023 C. Liu (Ed.): Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3, SPPHY 285, pp. 233–245, 2023. https://doi.org/10.1007/978-981-19-8899-8_24

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R. Wu et al. Keywords: Renal transplantation · En-bloc kidney transplantation (EBKT) · Pediatric donor · Renal dynamic imaging · Glomerular filtration rate (GFR)

1 Introduction En-bloc pediatric kidney transplantation (EBKT) from pediatric donors involves transplantation of two kidneys from a deceased pediatric donor, en-bloc, into a recipient. An initial report on the technique was published in 1972 [1]. More than 2000 EBKTs were carried out in the United States between 1987 and 1993 alone [2]. There was an increase in the number of EBKT surgery in subsequent years [3]. The present studies have demonstrated that the long-term allograft performance and total patient survival associated with this procedure of kidney transplantation are excellent [3–5]. It has become an accepted and successful remedy for thousands of patients worldwide. Undeniably, a highly meticulous surgical procedure is required for EBKT because the risk of vascular complications is higher than with adult donation after circulatory death [6]. For best outcomes, it must be viewed and treated as a distinct entity from conventional transplantation, focusing on the unique challenges particular to its cohort. The proper assessment of renal function is critical for EBKT patients after surgery/treatment [7]. This EBKT technique does not necessarily mean that both graft outcomes are always synchronous. As a consequence, split renal function must be investigated to precisely evaluate renal damage [8]. Nevertheless, expert consensus recommendations addressing the specific challenges of measuring transplanted kidney function in EBKT have been lacking, as there is a lack of randomized controlled trials and the available evidence is limited to a few observational studies. In current practice, routine clinical assessment of graft function in EBKT is performed by either one or a combination of the following methods. As it is well known, glomerular filtration rate (GFR) is considered the best overall index of kidney function [9] and is defined as the amount of ultrafiltrate produced by the kidneys per unit of time. Most of the available studies evaluated kidney function using serum creatinine (Scr) and/or estimated glomerular filtration rate (eGFR), which are simple and cost-effective. Nevertheless, serum creatinine is a late marker of kidney dysfunction, lacking sufficient sensitivity to detect patients with mild renal damage. And these results might underestimate the importance of the normal contralateral kidney in masking biochemical changes [10]. Also, kidney biopsy is the gold standard of evaluation of transplant parenchyma [11]. Angioregression and loss of peritubular capillaries appear to be associated with the development of interstitial fibrosis and graft dysfunction in human allograft biopsy studies [12–15]. However, it is an intrusive procedure that is prone to sample errors [11, 16]. In addition, several non-invasive imaging methods have been used to evaluate renal allografts, like renal scintigraphy [7], color Doppler ultrasonography, digital subtraction angiography, and magnetic resonance imaging [17]. It is worth mentioning that renal scintigraphy is the most common method that allows clinicians to quantify real changes in the amount of renal function of the treated kidney. In some EBKT cases, we found that renal dynamic imaging was frequently applied to postoperative monitoring [18, 19]. Renal dynamic imaging, also known as Gate’s method, is an effective method for measuring GFR due to the following advantages: achievement of unilateral renal blood flow

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and kidney function; no blood or urine collection is needed, and its result is not affected by the diet of the patient. However, to our knowledge, there are no prior studies evaluating/discussing the clinical diagnostic benefit of renal dynamic imaging and other methods to measure renal function in EBKT patients. Thus, our attempt is to find out which method is most suitable for detection and monitoring of graft function for EBKT patients over various periods of time, especially in the early period post transplantation.

2 Materials and Methods 2.1 Patients This was a retrospective study, analyzed imaging data of 16 EBKT patients who underwent 99m Tc-DTPA-based renal dynamic imaging at our institute from April 2017 to November 2019. All subjects were patients after their first kidney transplant. The panel reactive antibody (PRA) and Lymphocyte cross-matching were negative for the recipients. Clinical data were recorded from the medical record, including age, gender, body weight, height, blood pressure (BP), liver and renal function indicators, and the time interval after transplantation. This study’s main blood biochemical indexes and renal dynamic imaging were measured during the same admission. Alanine aminotransferase (ALT), aspartate aminotransferase (AST), albumin (ALB) and total protein (TP) were the measures of liver function used. Indicators of kidney function included serum creatinine (Scr), blood urea nitrogen (BUN), serum uric acid, cystatin-C, β2 -microglobulin, and 24-h urine protein. The study performed subgroup analyses based on specific characteristics of the patients, such as the indicated time after transplantation, serum creatinine values, kidney biopsy data and the occurrence of complications. This study was performed in compliance with the Helsinki Declaration with the approval of the ethical committee of Union Hospital, Tongji Medical College, Huazhong University of Science and Technology (No.20210646). And the Ethics Committee on Organ Transplantation of Union Hospital reviewed and approved each case of all sixteen en-bloc pediatric kidney transplantation. 2.2

99m Tc-DTPA-based

Renal Dynamic Imaging (GATE’s Method)

Gate’s glomerular filtration rate (gGFR) was measured by 99m Tc-DTPA-based renal dynamic imaging method. The patients were hydrated with 300–500 mL water 20 min before the examination. Dynamic renal scintigraphy was performed using an integrated SPECT/CT system equipped with low-energy high-resolution collimators (Discovery 640, GE Healthcare; Symbia T6, SIEMENS). The scanning probe was placed ventrally above the transplanted kidney in the iliac fossa, and the skin-kidney distance was set as 5 cm. Before radiotracer injection, CT scan was performed with the same SPECT/CT system during free breathing. Both upper extremities were placed on the chest outside the scanning range to reduce artefacts. The CT system was a cone-beam system using a flat-panel x-ray detector. Imaging at 1 location including the renal hilum was performed

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in a localization fast mode. Patients were manually injected with 5 mCi (185 MBq) of 99m Tc-DTPA as a rapid bolus. Coronal images were acquired every 2 s for the first 60 s and then one frame/min for 10–20 min. The syringe radioactive counting rate was measured using the central probe before and after injection. The total injected dose was determined by subtracting the post-count from the pre-count for 10 s each. All data were analyzed at the post-processing workstation with transplant kidney pattern (Xeleris, GE Healthcare). The region of interest (ROI) was manually drawn on the frame of the kidney, and a semi-lunar background was placed outer renal margin. Renal depth was measured on the low-dose CT images. An observer selected a center slice of the CT for each kidney, visually identified the kidney’s most anterior and posterior points on the image, and measured perpendicular distances from the ventral skin to these 2 points. The 2 distances were then averaged to define the renal depth (2-point method) [20, 21]. After the patient’s renal depth was entered into an online computer, the GFR was automatically calculated by commercially available software according to the Gate’s algorithm. 2.3 Estimated Glomerular Filtration Rate (eGFR) Serum creatinine was used in assessment of renal function based on eGFR equation from the Chronic Kidney Disease Epidemiology Collaboration (CKD-EPI)-study [22]. The normal range of serum creatinine (Scr) measured by the enzymatic method was 0.5 ~ 1.2 mg/dL (male, 0.6 ~ 1.2; female, 0.5 ~ 1.0). 2.4 Kidney Transplant Biopsy Renal allograft biopsies were obtained from renal transplant recipients when indicated for allograft dysfunction or proteinuria. EBKT patients had transplant protocol biopsies during the same admission of renal dynamic imaging. According to the Banff classification, two experienced nephropathologists evaluated the biopsies [11, 23]. These were assessed modified Banff’s fibrosis scores and inflammation scores from the biopsy report [24].

3 Statistical Analysis Qualitative variables were presented as percentage and analyzed by χ2 test. Normality of the data was inspected graphically and tested with Shapiro-Wilk test. Quantitative variables with normal distribution were presented as mean ± SD and assessed by t-test. Quantitative variables with skewed distribution were presented as median (IQR) and assessed by Mann-Whitney U test. In addition, correlations were calculated according to Pearson or Spearmen correlation. A P-value < 0.05 was considered to indicate a statistically significant difference. All statistical analyses were performed using SPSS software (version 23.0).

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4 Results 4.1 Subject Characteristics Table 1 shows the gender, age, body weight, height, main parameters, the time interval after transplantation and GFR of the 16 EBKT patients. 16 EBKT patients were included, ten females and six males (mean age, 30.13 ± 10.42 years, range 15–53 years). The average body mass index (BMI) was 19.1 ± 2.6 kg/m2. The patients underwent renal dynamic imaging at different time intervals posttransplantation, ranging from 0 to 41 months, with an average length of 15.25 months. A total of two patients had a history of previous alcohol abuse (maximum alcohol consumption: 250g/day). All but one patient had no history of tobacco use. And they reported complete abstinence from smoking and alcohol after kidney transplantation. 4.2 Correlation Analysis The results of Pearson or Spearmen correlation are shown in Table 2. In 16 EBKT patients, gGFR showed a positive correlation with the time intervals post transplantation (r = 0.725, P < 0.001), and the linear regression equation parameters and R2 were (y = 2.27x + 45.638) and (R2 = 0.525). Furthermore, gGFR negatively correlated with Scr (r = − 0.688, P < 0.05), BUN (r = − 0.862, P < 0.05), cystatin-C (r = − 0.708, P < 0.05). 4.3 Comparison of Main Indexes and Renal Function Early and Late After Transplantation Cohorts of patients were analyzed at the indicated time after transplantation. According to their post-transplantation time, they were divided into two groups: the early-period group (less than or equal to 12 months) and the later-period group (more than 12 months). Table 3 shows main indexes and GFR in two groups. Gender, age, height, weight, BP, TP, ALB, AST, ALT, Scr, uric acid, BUN/Scr ratio and blood uric acid/Scr ratio of the patients were not significantly different between the early-period group and the later-period group. BUN and cystatin-C were much lower in the later-period group than in the early-period group (P < 0.05). gGFR was much higher in the later-period group than in the early-period group (P < 0.001). Moreover, during an observation period of 12 months after transplantation, some difference was obtained for eGFR over gGFR in the early-period group with statistical significance (67.50 ± 32.23 ml/min vs 53.88 ± 17.86 ml/min, P < 0.05), the correlation coefficient was 0.931(P < 0.05). The eGFR and gGFR were not significantly different in the later-period group (111.98 ± 14.92 ml/min vs 107.64 ± 27.54 ml/min, P > 0.05). 4.4 Renal Function Assessed by Serum Creatinine and gGFR It did not seem like there was a significant difference in Scr between the time-interval groups. Meanwhile, subgroup analyses on the gGFR values (i.e., < 60 vs. ≥ 60 mL/min or

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< 90 vs. ≥ 90 mL/min) were performed (Table 4). No significant intergroup differences in serum creatinine values were observed (P > 0.05). However, gGFR increased significantly in patients with normal serum creatinine levels (90.62 ± 31.09 mL/min vs 35.37 ± 15.74 mL/min, P < 0.05). 4.5 Impact of the Occurrence of Complications and Kidney Biopsy on Renal Function In all EBKT patients, kidney function was solely provided by the transplanted kidney. Six of sixteen patients had postoperative complications. Of these, five patients presented with surgical complications, including renal artery stenosis and perirenal hematomas. One patient had cytomegalovirus (CMV) antigenemia for one month after surgery. The average gGFR in patients with postoperative complication was lower than that in those with no postoperative complications (96.50 ± 33.77 mL/min vs 53.18 ± 21.15mL/min, P < 0.05). A total of nine patients underwent kidney biopsies within one week of renal dynamic imaging. One of them with other signs of antibody-mediated rejection. The patient attained their improved status due to timely adjustments to immunosuppressive drug regimens. Apart from this, these no meeting “inflammation/fibrosis” cut-off were considered to be in the “no inflammation/fibrosis” for the majority of patients underwent kidney biopsies. Only a fraction belonged to a group of mild inflammatory changes.

5 Discussion EBKT from pediatric donors was initially conducted reluctantly because of its technical complexity, insufficient nephron mass, donor-recipient mismatch, and danger of hyperfiltration injury, which resulted in a high rate of complications and poor transplant survival [25]. Postoperative long-term follow-up monitoring of these patients, however, found that graft outcomes of EBKT are equivalent to those of deceased-donor kidney transplantation and living-donor kidney transplantation [26, 27]. To reduce the risk of postoperative thrombosis, Dai et al. developed a novel en-bloc kidney transplantation technique using a modified arterial inflow and outflow tract can prevent vascular thrombosis and provide adequate graft function [28]. Rapid compensatory growth in donor kidney volume develops because of mild hyperfiltration in EBKT patients. Consecutive ultrasound examinations revealed a constant rise in the size of the kidney to the near-adult size over the first year following transplantation, which can adapt to adult recipients’ greater metabolic demand [7]. Our local policy is to select recipients of low body mass for EBKT with modest blood pressure lowering (blood pressure controlled < 140/90 mmHg), although a strict cut-off was not applied. Reliable assessment of transplant kidney function in EBKT patients has become critical in implementing this technique. Renal dynamic imaging is simple, fast and less expensive for the determination of GFR. The glomeruli filter 99mTc-DTPA without reabsorption by the renal tubules. The renogram’s accurate information can help clinicians diagnose urinary system obstruction, monitor transplanted kidneys, and the effects of treatment [29]. Therefore, we retrospectively analyzed the renal dynamic imaging

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and clinical data of 16 cases of EBKT at our hospital to investigate their clinical value in estimating renal function. The anterior image acquisition and CT assisted measurement of the depth of transplanted kidneys were innovatively used, which solved the problem that the conventional renal dynamic imaging could be inaccurate in the renal depth. Our quantitative analysis showed good correlations between gGFR and the time intervals after transplantation and some kidney function indicators. We found that the GFR measured by renal dynamic imaging in the later-period group was significantly higher than that in the early-period group, which showed good surgical results matched with high GFR. Moreover, although GFR measurements in EBKT patients within one year are low, it supports that this is correlated with the kidney source of pediatric donors [30]. Clinicians should not misdiagnose that low GFR in transplanted kidneys equals graft failure. Early after donation, increased early graft loss and lower GFR were found in recipients of en-bloc pediatric kidneys compare to living-related donor kidneys, but this difference was not found any more at five years [30, 31]. Superior long-term outcomes in en-bloc recipients are likely due to the increase in size of small pediatric kidneys with the growth of the recipient, resulting in a progressive improvement in GFR. Caution needs to be raised, and we should observe the images accurately and differentiate a typically developing transplanted kidney from one with complications resulting in low GFR. BUN, cystatin-C, and β2 -microglobulin are all excellent identified markers of renal function and showed good variability in the subgroups. However, Scr values were not significantly different between the two groups, which could be related to many factors that influence serum creatinine levels, such as dehydration, dietary protein intake, muscle mass, diuretics [32]. We reclassified based on serum creatinine values and the gGFR values, we can observe that an increase in serum creatinine may not be observed until a substantial decrease in GFR has occurred. It is well known that serum creatinine may not be the most accurate and sensitive parameter for early changes in renal function, as the full effect on serum creatinine may only occur hours or even days after the initiation of renal damage. At the same time, we also found that the estimated glomerular filtration rate based on CKD-EPI methods was higher than the GFR values measured by renal dynamic imaging. Many studies have shown that the CKD-EPI formula based on a large dataset pooled from research and clinical populations is more accurate and more widely applied than other methods of GFR determination in Chinese CKD patients and that it can be applied generally in clinical practice [33]. However, gender, age, and Scr are the main variables in the CKD-EPI formula, so any changes in these variables will affect the GFR obtained using this formula. Notably, this method applies to native kidneys instead of renal transplants in daily clinical practices. We found that the evaluation of renal function within one year of transplantation in EBKT is best assessed reliably using renal dynamic imaging by this research. Moreover, eGFR calculated by CKD-EPI creatinine-based equation could adapt for GFR assessment more than one year after renal transplantation as a simple assessment of renal function in the present study. The finding should be validated in prospective studies. Furthermore, biopsy can also help clinicians assess the risk versus benefit of treatment in EBKT patients. We only found minor inflammatory/fibrosis changes in these biopsies after transplantation. These results might be related to the use of en-bloc pediatric kidney

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transplantation. However, in patients with kidney failure due to IgA nephropathy, IgA deposits can recur in a subsequent kidney transplant [34]. During the course of long-term follow-up, we found that graft survival in some patients with IgA nephropathy is considered good despite this risk of recurrent glomerulonephritis. Moreover, it is essential to closely monitor and timely adjust the blood concentration of immunosuppressive drug for helping prolong graft survival. This study had some limitations. Firstly, our study is the small number of patients and its retrospective nature. Further studies are needed to validate these preliminary findings. Secondly, individual renal graft placed in the iliac fossa may have mild torsion and small degree of overlap. This is still being studied by our group that changing scanning modes and modifying measured parameters could provide more accurate and real-time information. Thirdly, our biopsy data is not comprehensive enough to make better conclusions for guiding quality improvement in postprocedural surveillance because of minor inflammatory/fibrosis changes in biopsies. Further research might help clinicians to decide if a patient requires biopsy and to avoid unnecessary biopsies.

6 Conclusions Our study verified that renal dynamic imaging could sensitively reflect renal function changes in EBKT patients. GFR measured by renal dynamic imaging is more accurate and sensitive than the estimated glomerular filtration rate based on CKD-EPI equation and serum creatinine, especially during an observation period of 12 months after transplantation. Accurate assessment of GFR is essential for interpreting the symptoms, signs, and laboratory abnormalities that may indicate postoperative complications and drug dosing and assessing the prognosis. Renal dynamic imaging should be further popularized to monitor renal function in patients after transplantation.

Female

Male

Female

Female

14

15

16

39

16

27

31

33

25

15

21

30

39

28

22

48

53

29

29

Ages (years)

1.48

1.56

1.60

1.64

1.75

1.55

1.73

1.63

1.73

1.65

1.68

1.55

1.60

1.70

1.57

1.50

Height (m)

40.5

40

50

55

65

46.2

49

55

55

53

71

36

52

62

46.5

34

Weight (kg)

18.49

16.44

19.53

20.45

21.22

19.23

16.37

20.70

18.38

19.47

25.16

14.98

20.31

21.45

18.86

15.11

BMI (kg/m2 )

63.6

60.1

95.5

124.5

108.7

70.1

121.8

77.6

121.1

53

148.3

18.1

60.6

48.9

39.1

73.1

gGFR (ml/min)

100.9

95.5

118.3

126.9

113.4

105.3

134.1

89.9

98.3

78.1

114.0

14.0

66.0

45.1

39.9

96.0

eGFR (ml/min)

19

5

24

19

41

11

13

16

35

12

24

0

12

1

6

6

Time Intervals (Months)

Ureteral stricture

Ureteral stricture

No occurrence

No occurrence

No occurrence

No occurrence

No occurrence

Perirenal hematoma

No occurrence

No occurrence

Ureteral stricture

Perirenal hematoma

Ureteral stricture

No occurrence

CMV infection

No occurrence

Postoperative Complications

BMI: body mass index; BP: blood pressure; eGFR: estimated glomerular filtration rate; gGFR: Gate’s glomerular filtration rate

Male

Male

8

13

Female

7

12

Female

6

Female

Female

5

11

Female

4

Male

Male

3

Male

Female

2

10

Female

1

9

Gender

Patient number

Table 1. Gender, age, main parameters and the time intervals of EBKT patients

The Application of Renal Dynamic Imaging in Measuring Renal 241

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R. Wu et al. Table 2. Correlation analysis in gGFR and other indexes Correlation coefficient

P value

Time Intervals (Months)

0.725

0.001

Scr (mg/dl)

−0.688

0.003

BUN (mg/dl)

−0.862

0.000

Blood uric acid (mg/dl)

−0.165

0.542

Cystatin-C (mg/L)

−0.708

0.003

β2-microglobulin (mg/L)

−0.606

0.064

24-h urine protein (mg/24h)

−0.345

0.298

Table 3. Main indexes and GFR in the early-period group and the later-period group Characteristic

The early-period group

The later-period group

p-value

Number of patients (n)

8

8

Gender (M/F)

1/7

5/3

0.119

Age (years)

33 ± 13

28 ± 7

0.393

Height (m)

1.59 ± 0.06

1.66 ± 0.09

0.095

Weight (kg)

46.2 ± 9.4

55.1 ± 9.5

0.082

BMI (kg/m2 )

18.23 ± 2.42

20.04 ± 2.60

0.279

MAP (mmHg)

95 ± 8

102 ± 8

0.094

TP (g/L)

70.9 ± 7.5

68.2 ± 2.7

0.364

ALB (g/L)

44.4 (3.4)

44.0 ± 3.0

0.798

AST (U/L)

19.5 (11.3)

16.4 ± 3.5

0.279

ALT (U/L)

22.5 (22.3)

14.8 ± 8.4

0.500

Scr (mg/dL)

0.97 (0.86)

0.84 ± 0.19

0.083

Uric acid (mg/dL)

5.63 (1.13)

5.42 ± 0.82

0.382

BUN (mg/dL)

21.06 (16.10)

12.97 ± 3.56

0.028

Cystatin-C (mg/L)

1.43 (0.92)

1.14 ± 0.43

0.021

β2-microglobulin (mg/L)

0.8 ± 0.3

0.4 ± 0.2

0.033

24-h urine protein (mg/24h)

263 (192)

76 (218)

0.376

eGFR

67.50 ± 32.23

111.98 ± 14.92

0.005

gGFR

52.88 ± 17.86*

107.64 ± 27.54

0.000

GFR (ml/min)

* P < 0.05 compared with eGFR during an observation period of 12 months after transplantation

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Table 4. Renal function assessed by Serum creatinine, Blood urea nitrogen and Cystatin-C Scr (mg/dL)

BUN (mg/dL)

Cystatin-C (mg/L)

< 60ml/min

2.14 ± 1.45

29.44(60.56)

2.43 ± 0.98

60ml/min

0.85 ± 0.16

13.78 ± 4.01

1.21 ± 0.34

0.172

0.002

0.003

< 90ml/min

0.97(0.88)

19.05(13.43)

1.37(0.89)

90ml/min

0.80 ± 0.17

11.60 ± 2.93

1.15 ± 0.51

0.093

0.007

0.099

gGFR (60)

p-value gGFR (90)

p-value

References 1. Meakins, J.L., Smith, E.J., Alexander, J.W.: En bloc transplantation of both kidneys from pediatric donors into adult patients. Surgery 71, 72–75 (1972) 2. Dharnidharka, V.R., Stevens, G., Howard, R.J.: En-bloc kidney transplantation in the United states: An analysis of united network of organ sharing (UNOS) data from 1987 to 2003. Am. J. Transplant. 5, 1513–1517 (2005) 3. Maluf, D.G., Carrico, R.J., Rosendale, J.D., et al.: Optimizing recovery, utilization and transplantation outcomes for kidneys from small,