Boiling Water Reactors 0128213612, 9780128213612

Table of contents :
Boiling Water Reactors
Copyright
Contributors
About the Authors
Preface of JSME series in Thermal and Nuclear Power Generation
Preface
Editing working group for volume 4: Boiling water reactors
Abbreviations
History of BWR development
Nuclear energy development in Japan
Primary energy supply
Electric power generation
Nuclear power generation
Nuclear power generation legislation
References
Establishment and realization of BWR technologies
Established stage
Introduction
Development of BWR by the Argonne National Laboratory in the United States
Realizing stage [3,6-22]
The early stage of GEs BWR development
The further stage of GEs BWR development
ABWR development with international cooperation
Next BWR development
References
Improvement and standardization program in Japan
Technology importation
First improvement and standardization program
Countermeasure for SCC [9,11,12,17,18]
Improved primary containment vessel (improved PCV)
Second improvement and standardization program [12,16]
Improvement of fuel-core design and CRD
Third improvement and standardization program
Development program of ABWR [9,13,20-27]
References
Improvement of system and construction
Reduction of the construction period of BWRs [1-3]
Large block and module technologies
Management by computer and information technologies
Improvement of ABWR system and construction [4-8]
Large block-module construction method
All-weather construction method
References
Construction experience and operation performance
Introduction period [1-4]
Improvement and standardization programs of Japan [2-7]
Recent status
References
Features of BWR plant
Introduction
Reference
Reactor
Overview
Reactor system
Reactor pressure vessel (RPV)
Reactor internals
Core and fuel
Control rods and control rod drive
Reactivity control system
Control rod drive system
Standby liquid control system
Core monitoring system
References
Reactor coolant system and connectedsystems
Overview
Nuclear boiler system
Main steam system
Feedwater system
Reactor recirculation system
Reactor water cleanup system
Residual heat removal system
Leak detection system
References
Engineered safety features
Overview
Containment system
Overview
Primary containment vessel (PCV)
Primary containment isolation system (PCIS)
Primary containment vessel gas control systems
Overview
Atmospheric control system (AC)
Flammability control system (FCS)
Containment heat removal system
Secondary containment
Standby gas treatment system (SGTS)
Emergency core cooling system
Overview of ECCS
Reactor core isolation cooling (RCIC) system
High-pressure core flooder system
Low-Pressure Flooder (LPFL)
References
Instrumentation and controls
Introduction
Overall architecture (example of ABWR)
Major control systems and auxiliary control systems
Major control systems
Auxiliary control systems
Safety systems
Process computer system
Human-machine interface
Electric power
Overview
Function
Configuration/main equipment (example of ABWR)
Grid connection
Transformers
Auxiliary medium-voltage distribution buses
Emergency diesel generators
DC power supply system
AC instrumentation power supply system
Auxiliary system
Overview
Auxiliary system
Fuel pool cooling and cleanup system [1]
Reactor building cooling water system [2]
Reactor building service water system [2]
Turbine building cooling water system [2]
Turbine building service water system [2]
Makeup water condensate system [2]
Instrument air system [2]
High-pressure nitrogen gas supply system [2]
Sampling system [2]
Heating ventilating and air conditioning system [2]
References
Steam and power conversion systems
Overview
Steam and power conversion systems
Turbine generator [1]
Main steam system, auxiliary steam system, and turbine bypass system [1]
Extraction steam system [1]
Turbine gland steam system [1]
Feedwater heater drain and vent system [1]
Condenser [1]
Circulating water system [1]
Condensate and feedwater system [2]
Off-gas system [2]
References
Nuclear reactor dynamics and thermal hydraulics of reactor core and fuel assembly
Reactor internals and coolant flow paths in a reactor pressure vessel
Unique basic characteristics of the BWR core
Application of negative void reactivity
BWR core configuration and basic design concept
Reactor core support structure and other reactor internals
Coolant flow paths and the BWR operating map
Coolant flow paths
Operating map
References
Advances of reactor core and fuel assembly
High burnup fuel design
Introduction
Reliability improvement (1970s)
Operational improvement
Economical improvement-Step I fuel and core
Economical improvement-Step II fuel and core
Economical improvement-Step III fuel and core
Summary
References
MOX fuel design
Thermal-hydraulic design
Thermal-hydraulic design basis of the reactor core
Nuclear thermal-hydraulic stability
Flow-induced vibration
References
Introduction
Basic information about Pu
Characteristics of Pu should be considered for utilization
MOX fuel assembly design
MOX core design
Summary
References
Countermeasures and cause of fuel rodfailure
Overview of fuel failures in BWRs
Countermeasures and cause of fuel rod failure
References
Proving test on the thermal-hydraulicperformance of BWR fuel assembly
Introduction
Proving test on thermal-hydraulic performance of a BWR fuel assembly
Void fraction measurement test for BWR fuel assembly [11-13]
Development of thermal-hydraulic correlations based on the full-scale BWR fuel assemblies data
References
Advances in reactor core and fuel assemblyanalysis
Nuclear analysis in BWRs
2D lattice calculation
3D core calculation analysis
Validation with measurements
References
General reference for nuclear analysis
Thermal-hydraulic system analysis code
Thermal-hydraulic subchannel analysis code
References
Advances in containment vessel design
Thermal hydraulics of severe accidents
Introduction
Initiation of fuel melt
Progression of core melt
Water-Zircaloy reaction accelerating fuel melt
Melting relocation inside the RPV
Melting jet structure and behaviors (from the RPV bottom to the PCV floor)
FP aerosol behaviors [3,4]
Accident management for BWR
Summary of AM
Defense in depth
International event scale (INES)
Selection of BWR AM measures
Typical BWR core damage sequence
In-vessel phenomena (from core melt to RPV bottom leak)
Ex-vessel phenomena after RPV failure
AMs for existing BWR
AMs for the recently operated and planned plants (also with PWR)
References
Advances in safety analysis codeand safety systems
Various BWR analysis codes
Importance of nuclear analysis codes
Best estimate code and evaluation model code
Verification and validation (VandV) of simulation [3,4]
BWR analysis code (EM code) [5]
LOCA analysis code (BE code)
SA progression analysis code
Computational fluid dynamic (CFD) analysis code
Large-scale test facility for code verification and obtaining correlations
BWR safety systems for severe accident
Passive safety concept
Reinforcement for passive safety
Lineup of passive safety systems
References
Fukushima Daiichi nuclear power plant accident and analysis evaluation
Outline of accident
Event progress and analysis evaluation at Unit 1
Event progress and analysis evaluation at Unit 2
Event progress and analysis evaluation at Unit 3
Hydrogen explosion at Unit 4
Avoiding severe accidents at Fukushima Daini NPS
Overview of emergency response at Fukushima Daini NPS
Fukushima Daini Unit 1 response and station behavior
Response status at the time of tsunami arrival
Reactor cooling water injection and PCV cooling
RHR restoration and reactor cold shutdown
Continuous ERC planning activities
Lessons learned from Fukushima Daiichi accident
Causes of severe accidents and countermeasures
Measures for severe accidents installed in the United States and European NPPs
Filtered containment venting system
Special emergency heat removal system
Tsunami protection
New nuclear regulatory requirements in Japan
New nuclear regulatory requirements
Tsunami protection examples
Tornado protection examples
Example of compliance with new regulatory standards for PWRs that can be used as a reference for BWRs
BWR NPS to be reviewed for new requirements or restarting
Activities toward decommissioning Fukushima Daiichi
Current status of reactors at Units 1 through 4
Finding contaminated water leak path for leak shutdown from PCV
Isolation of groundwater flow from contaminated water
Contaminated water management
Preparation for fuel-debris removal
Important lessons learned from Fukushima Daiichi NPS accident
References
BWR innovations
Trans-uranic (TRU) burner reactor and reduced-moderation water reactor
TRU burner reactor
Introduction
RBWR concept
RBWR specifications
RBWR core characteristics
Progressive introduction of RBWR [10]
References
Reduced-moderation light water reactor
Introduction [1-4]
Research and development of the cost-reduced low-moderation spectrum BWR
References
Design innovation of BWR and high-pressureBWR
Introduction
Objective of LSBWR design
Natural circulation core concept
Conceptual design of long cycle core of LSBWR
Examination of plant operating pressure and plant thermal efficiency
Safety system and PCV concept
Module fabrication and construction
Ship hull structure for reactor building
General arrangement of LSBWR and LLBWRs building design
Construction methodology and evaluation
Summary of design innovation of LSBWR, LLBWR, and high-pressure BWR
References
Power uprate in BWR
Current status and trend of reactor power uprates
Benefits and safety of constant rated reactor thermal power operation
Possibilities and issues on constant rated reactor thermal power operation
Current status of reactor power uprate with equipment modification
Reactor thermal power and electric power
Reactor power uprate with constant rated reactor thermal power operation
Relationship between reactor thermal power and electric power outputs
Issues and safety in constant rated reactor thermal power operation
Experiences in BWR operation with constant rated reactor thermal power operation
Power uprate with equipment modification
Uprate by measurement uncertainty recapture
High accuracy leading edge flowmeter (LEFM) for nuclear reactor feedwater measurement in MU
Inevitable issues on the accuracy of the PF in high accuracy ultrasonic flowmeters and new-concept flowmeter pos ...
Recent implementation and issues of uprates in the United States
References
Post-BT standard for BWR power plant
Introduction
Standard for the assessment of fuel integrity under anticipated operational occurrences
The method for predicting the change of rod temperature during post-BT operation
The criteria of fuel integrity after BT [9,10,13,20]
References
Core catcher
Overview of core melt stabilization and cooling
Core catcher of EU-ABWR
Concept of core catcher
Performance evaluation test
Core catcher for the existing BWR
Concept of core catcher
Performance evaluation test
References
Steam injector
Introduction
Principle and application of SI
SI analysis model
Visualized fundamental tests
Test apparatus and measurements
Test results
Application of steam jet-type SI to PCIS
High-pressure tests and analysis
Application of water jet type SI to RLP
Confirmation of analysis method
Scale-up examination of SI for application to RLP
High-pressure tests using scale models
Simplified feed water system by SI
Scaled model tests of simplified feedwater system
Analysis for improving SI-FWH
Transient test result of the first stage
Advantages of SI introduction to ABWR in volume and mass reduction
Steam injector (SI) pump-up water system to refill pool for passive containment cooling isolation condenser (PCC/I ...
Concept of SIPOWER
Evaluation of PCC/IC pool water level transient by SIPOWER
Full-scale mock-up test to confirm feasibility of SIPOWER
Air-purge analysis in PCC/IC pool for SIPOWER
Summary of SIPOWER
References
Built in upper internal control rod drives(CRDs) for ABWR-III
Introduction of merits and technical tasks for internal CRD
Plant concepts of ABWR-III
Power devices for the internal CRD
Magnet coupling power connector
Magnet coupling signal connector
Internal CRDs mechanism
Latch mechanism for scram operation and lift a control rod
Development of heatproof motor
Ceramics coil radiation durability test
Neutron flux at the internal CRD
Evaluation of ABWR-III conditions
Durability test of ball bearing
Two-phase flow and structural integrity
LOCA and pressure transient analysis
Aseismic analysis results
Summary
References
Index

Citation preview

Boiling Water Reactors

JSME Series in Thermal and Nuclear Power Generation

Boiling Water Reactors Volume 4 Edited by

Koji Nishida Institute of Nuclear Safety System, Mihama, Fukui, Japan

Shinichi Morooka Former Waseda University, Shinjuku, Tokyo, Japan

Michitsugu Mori Graduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan

Yasuo Koizumi The University of Electro-Communications, Chofu, Tokyo, Japan

Elsevier Radarweg 29, PO Box 211, 1000 AE Amsterdam, Netherlands The Boulevard, Langford Lane, Kidlington, Oxford OX5 1GB, United Kingdom 50 Hampshire Street, 5th Floor, Cambridge, MA 02139, United States Copyright © 2023 Elsevier Inc. All rights reserved. No part of this publication may be reproduced or transmitted in any form or by any means, electronic or mechanical, including photocopying, recording, or any information storage and retrieval system, without permission in writing from the publisher. Details on how to seek permission, further information about the Publisher’s permissions policies and our arrangements with organizations such as the Copyright Clearance Center and the Copyright Licensing Agency, can be found at our website: www.elsevier.com/permissions. This book and the individual contributions contained in it are protected under copyright by the Publisher (other than as may be noted herein). Notices Knowledge and best practice in this field are constantly changing. As new research and experience broaden our understanding, changes in research methods, professional practices, or medical treatment may become necessary. Practitioners and researchers must always rely on their own experience and knowledge in evaluating and using any information, methods, compounds, or experiments described herein. In using such information or methods they should be mindful of their own safety and the safety of others, including parties for whom they have a professional responsibility. To the fullest extent of the law, neither the Publisher nor the authors, contributors, or editors, assume any liability for any injury and/or damage to persons or property as a matter of products liability, negligence or otherwise, or from any use or operation of any methods, products, instructions, or ideas contained in the material herein. ISBN: 978-0-12-821361-2 For information on all Elsevier publications visit our website at https://www.elsevier.com/books-and-journals

Publisher: Charlotte Cockle Editorial Project Manager: Sara Valentino Production Project Manager: Prasanna Kalyanaraman Cover Designer: Miles Hitchen Typeset by STRAIVE, India

Contributors

Tetsushi Hino Hitachi, Ltd., Hitachi, Ibaraki, Japan Haruhiko Ikeda Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki, Japan Chikako Iwaki Toshiba Energy Systems & Solutions, Corp., Yokohama, Kanagawa, Japan Yukihiro Katayama Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki, Japan Hiroki Kawai Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki, Japan Yasuo Koizumi The University of Electro-Communications, Chofu, Tokyo, Japan Masayoshi Matsuura Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki, Japan Hiroshi Miyano Former Toshiba Corporation, Kawasaki; Senior Adviser of Japan Society Maintenology, Taito, Tokyo, Japan Michitsugu Mori Graduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan Shinichi Morooka Former Waseda University, Shinjuku, Tokyo, Japan Tadashi Narabayashi Tokyo Institute of Technology, Meguro, Tokyo, Japan Koji Nishida Institute of Nuclear Safety System, Mihama, Fukui, Japan Akira Nishimura Tokyo Institute of Technology, Meguro, Tokyo, Japan Hisatoshi Shirahama Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki, Japan Hirotsugu Suzuki Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki, Japan Kazuhiko Wakasugi AESJ Fellow, and Former Technical Counselor of NSC of Japan, Zushi, Kanagawa, Japan

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Contributors

Yukihisa Yabushita CSA of Japan Co., Ltd., Minato, Tokyo, Japan Seiichi Yokobori Tokyo City University, Setagaya, Tokyo, Japan Yuichiro Yoshimoto Former Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki, Japan Kenichi Yoshioka Toshiba Energy Systems & Solutions, Corp., Kawasaki, Kanagawa, Japan

About the Authors

Tetsushi Hino started his research career in the study of boiling water reactors (BWRs) at Hitachi, Ltd., in 1999, where he was engaged in BWR core design and analysis methods. Since 2011, he has been engaged in the research and development of the light-water-cooled fast reactor (RBWR). After working for Hitachi Europe Ltd. in a joint study with UK research organizations from 2016 to 2018, his current work includes the development of the core concept and design of innovative light water reactors in Hitachi, Ltd. He is a member of the Atomic Energy Society of Japan (AESJ). Haruhiko Ikeda joined Hitachi, Ltd., in 1999 and was assigned as a turbine island system engineer for Japanese BWR plants. In 2001, he was transferred to the project management division, where he was assigned as a project engineer for overseas projects. He returned to becoming a turbine island system engineer in the system engineering division in 2003 and was engaged in commissioning tests of overseas newly built nuclear power plants, power uprate feasibility studies, equipment replacements of Japanese BWRs, and the Generic Design Assessment (GDA) process for the newly built UK advanced boiling water reactor (ABWR). Chikako Iwaki started her research and development career at Toshiba Corporation in 1989 and earned a doctor of engineering degree from the University of Tsukuba in 2003. She moved to Toshiba Energy Systems & Solutions Corporation in 2019 and continues to work there. Her main research areas are nuclear thermal hydraulics and two-phase flow. She has been engaged in research and development related to improving the safety and efficiency of nuclear power plant systems, targeting various components and systems such as the nuclear core, steam separator and dryer, steam injector, and passive cooling systems including in-vessel retention and core catcher. Yukihiro Katayama started his engineering career in electrical system design of BWRs at Hitachi, Ltd., in 2002, and was engaged in the design of electrical systems for Japanese ABWRs. He was the subject matter expert for the Electrical Systems of the GDA project in the UK from 2013 to 2017. Since 2020, he has been assigned as the section manager of the Electrical & Instrumentation Design Section of Hitachi-GE Nuclear Energy, Ltd. Hiroki Kawai started his engineering career in the mechanical process system design of BWRs at Hitachi, Ltd., in 2004. He has over 10 years’ experience in process system engineering and over 5 years’ experience in fire safety engineering for ABWRs. He has contributed to several newly built projects and preventive maintenance projects for nuclear power plants in Japan and the United Kingdom. From 2012 to 2017,

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About the Authors

he worked for the UK ABWR GDA as process system and fire safety engineer. His current work involves the management of project costs in Hitachi-GE Nuclear Energy. Yasuo Koizumi is currently a research promoter and an invited researcher at the University of Electro-Communications. Prior to this, he was an invited researcher at the Japan Atomic Energy Agency (JAEA) for 5 years. He received his PhD from the University of Tokyo in 1977. He started his research career at the Japan Atomic Energy Research Institute in 1977 as a research engineer for nuclear reactor safety. He worked at the Idaho National Engineering Laboratory from 1981 to 1983. He moved to the Department of Mechanical Engineering of Kogakuin University in 1989. Then, he moved to the Department of Functional Machinery and Mechanics of Shinshu University in 2008. He retired as professor in 2014 and has been with JAEA since then. His research is focused on the areas of pool and flow boiling, critical heat flux, condensation heat transfer, and two-phase flow. He is also interested in heat transfer and fluid flow at the microscale. Since his field of research is closely related to energy systems, he has great interest in thermal and nuclear power stations and energy supply in society. Masayoshi Matsuura started his research and development career in BWRs at Hitachi, Ltd., in 1987. He was engaged in the design of nuclear power plant systems and management of the next-generation reactor development projects. His current work involves the management of Hitachi-GE’s R&D and the development of new reactor concepts. He is certified as a professional engineer of Japan (P.E. Jp) for the nuclear and radiation field and is a member of AESJ. Hiroshi Miyano started his career in Toshiba in 1971. At Toshiba Corporation, he was engaged in research on flow-induced vibration (FIV) and the development of BWR systems at the research institute, working on primary system design and maintenance tasks as a leader in the System Design Department, and became the Nuclear Chief Engineering Officer and was then promoted to Director of the Group Company. During his career at Toshiba, he devoted himself to societies of scientists and engineers such as the Japan Society of Mechanical Engineers (JSME) and AESJ, where he was the chair of the Codes and Standards Committee. In 2014, he was the chair of the Fukushima Decommissioning Committee. Then, he committed to JSME and AESJ as “Fellow.” He enrolled in Advanced Management Program AMP-159 of Harvard Business School during 2000 and earned his PhD from Hokkaido University in 2012. From 2011 to 2019, he was professor in the System Design Division at Hosei University and was a member of The Engineering Academy of Japan Inc. in 2020. He is senior adviser in the Japan Society of Maintenology and the chair of the Fukushima Decommissioning Committee of AESJ. Michitsugu Mori is currently an invited guest professor of the Graduate School of Engineering, Hokkaido University. His research career started at NSRR (Nuclear Safety Research Reactor) of JAEA on the quenching cooling process of a fuel rod during reactivity-initiated accidents and then succeeded in the Department of Nuclear Engineering, Graduate School of Engineering, Tohoku University, Japan, and he was awarded a doctor of engineering degree in 1981. He researched the Brayton-cycle of a modular gas-cooled reactor (MGR) and its components at the Department of Nuclear

About the Authors

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Science & Engineering of the Massachusetts Institute of Technology (MIT), the United States, from 1987 to 1989. He joined the R&D Center of Tokyo Electric Power Company (TEPCO), where he researched the thermal-hydraulics and advanced measurement technologies of a light water reactor (LWR) including next-generation reactors. He became a full professor at Hokkaido University in 2012 and performed experiments and simulations on nuclear system safety, e.g., reactor core injection system, plant transient and debris behaviors, and next-generation reactors. He devoted himself to academic societies as the president of the Japanese Society for Multiphase Flow (JSMF) and the vice-president of the Heat Transfer Society of Japan (HTSJ) and Atomic Energy Society of Japan (AESJ), and the member of Board Directors of Japan Society of Mechanical Engineers (JSME). He currently is a fellow of JSME and AESJ, and honorary member of JSME, HTSJ, and JSMF. Shinichi Morooka is professor emeritus of Waseda University. He received a doctor of engineering from Waseda University in 1980. His research is focused on thermal hydraulics of nuclear power plants. He worked at Toshiba Corporation in the thermal hydraulics R&D Center of nuclear power plants for approximately 30 years. He has a great deal of experience in developing components for actual nuclear power plants. He returned to Waseda University as a professor in 2010. Currently, he works on the optimization of the heat transfer performance for LWR components using computed fluid dynamics code and experimental technologies, with nuclear fuels, separator systems, steam generators as target components. He can construct flow mechanisms, develop original simulation code based on flow mechanisms, and predict the heat transfer performance of fuel assemblies. Tadashi Narabayashi graduated the master course of Nuclear Engineering, Tokyo Institute of Technology, and entered Toshiba Corporation in 1978. He got a doctor degree in 1991 from Tokyo Institute of Technology. He was the chief specialist of Reactor Components and Two-Phase Flow in Toshiba Corp. He was a professor at Hokkaido University from 2005 to March 2018. He is the Professor Emeritus of Hokkaido University. He returned to the Tokyo Institute of Technology in April 2018. He has been involved in investigating the causes of accidents and developing countermeasures for other nuclear power plants in Japan, as an advisory meeting member of the Nuclear Regulation Authority (NRA) for the Fukushima Daiichi Accident Investigation Team. He was also a member of the Nuclear Program Advisory Panel (NPAP) for Khalifa University in UAE from 2012 to 2015. He was the recipient of the Outstanding Professor of the Year Award given by the ISOE (IAEA/OECD-NEA) in 2018. Koji Nishida received his doctor of engineering degree from Kobe University in 1987 for his study on convective film boiling heat transfer. He joined Hitachi Research Laboratory where he started researching the thermal hydraulics of BWRs. He was engaged in developing high burnup fuel bundles and high-performance next-generation BWRs including the small modular reactor (SMR). After the Fukushima Daiichi Nuclear Power Station accident in 2011, he was engaged in analyzing the accident progression. He moved to the Institute of Nuclear Safety System in 2017. Currently, he is conducting research on severe accidents and safety systems for pressured water reactors.

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About the Authors

Akira Nishimura graduated from the physics faculty of Tokyo University and joined Hitachi, Ltd., in 1971. His major background was core nuclear design for BWRs and maintenance technologies of nuclear plants. His main contribution was the development of axial uranium enrichment distribution technology for BWR cores, which is the current basic technology followed by Step I, II, and III fuel cores. In 2000, he joined Global Nuclear Fuel-Japan (GNF-J) as General Manager of Engineering and established the BWR fuel and core engineering team. In 2003, he became mixed oxide (MOX) fuel leader and led the rationalization of the Japanese MOX program for transportation of MOX fuel and inspection; he also created and an audit system among all Japanese BWR utilities. Since 2011, he has been a member of Tokyo Institute of Technology as a professor and a coordinator of the Japanese University Network for Global Nuclear Human Resource Development with 18 Japanese universities and 2 overseas universities to provide fundamental nuclear education to young generations inside and outside Japan. He has been a member of the Standardization Committee of the Japan Nuclear Society, the International Committee of Japan Nuclear Contribution, and the Japan Society of Newer Metals. Hisatoshi Shirahama started his engineering career in the mechanical process system design of BWRs at Hitachi, Ltd., in 2000. He has been certified as a professional engineer of Japan (P.E. Jp) for the nuclear and radiation field since 2009. He was in charge of project management for the GDA project in the UK from 2013 to 2015. Since 2020, he has been assigned as the section manager of the Plant and Engineering Design Section of Hitachi-GE Nuclear Energy, Ltd. Hirotsugu Suzuki started his engineering career in instrument and control (I&C) system design of BWRs at Hitachi, Ltd., in 1997, and was engaged in the design of the I&C systems for Japanese ABWRs. He was the subject matter expert for I&C systems for the GDA project in UK from 2015 to 2017. Then, he moved to Hitachi Nuclear Energy Europe Ltd. as the I&C engineering manager. Currently, he has returned to Hitachi-GE Nuclear Energy, Ltd., and has been assigned as the unit manager of the instrumentation design unit. Kazuhiko Wakasugi started his career at the Central Research Laboratories of Toshiba Corp. in 1962 after graduating from the Department of Mechanical Engineering, Nagoya Institute of Technology. His research is focused on irradiation performance and mechanical design of LWR fuel. Then, he moved to Japan Nuclear Fuel Corp. in 1971, which was founded in 1967, where he was engaged in BWR fuel manufacturing technology, including 2 years’ experience in overseas dispatch at GE Wilmington to introduce BWR fuel manufacturing technology. Then, he moved to the Secretariat of the Nuclear Safety Commission as Technical Counselor in 2000, where he was engaged in audits on the MITI for fuel cycle-related regulation and surveys of overseas nuclear regulatory systems. He is a fellow of the AESJ and currently dedicates his time to participating in the activities of dialog with university students as a member of the senior network of the AESJ in order to transmit nuclear technology to younger generations.

About the Authors

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Yukihisa Yabushita is a system analyst who has been engaged mainly in nuclear power plant accident analysis using thermal-hydraulic analysis codes for more than 40 years. He received his bachelor’s degree in mathematics from the Faculty of Science and Engineering, Waseda University, in 1972, after which he joined CRC Co., Ltd. He started carrying out numerical simulations of thermal-hydraulic experiments for Japan Atomic Energy Research Institute (JAERI). He jointly established Nippon Energy Inc., which was a professional company for accident analysis, in 1981. He earned his doctor of engineering degree from Tokyo Institute of Technology in 1996. His main interest is in performing numerical simulations of the complicated thermal-hydraulic behavior of nuclear power plants that have had accidents. In 1999, he established his own engineering consulting company, CSA of Japan Co., Ltd., to continue his life work. Seiichi Yokobori graduated from the mechanical engineering department at the University of Tokyo in 1975 and entered higher grades of master’s and doctor’s courses. After graduating and receiving a doctorate in 1980, he joined the Toshiba Nuclear Engineering Laboratory. His main interest was to investigate BWR safety thermal hydraulics experimentally. During his stay in Toshiba for 27 years, he presented the BWR rewetting correlation, developed the SAFER03 safety model, confirmed the PCCS heat removal performance, and applied the BWR drywell cooler for accident management. Most of these experimental results were awarded by AESJ. After retiring from Toshiba in 2007, he moved to the Musashi Institute of Technology to establish the Nuclear Safety Department, which has produced many nuclear engineers for technical succession. Currently, he is a professor emeritus of Tokyo City University and also a fellow of JSME, AESJ, and JSFM, and an honorary member of HTSJ. Yuichiro Yoshimoto graduated in mechanical engineering from Waseda University in 1971; he received his MS in control engineering from Tokyo Institute of Technology in 1973 and his PhD in quantum engineering and system science from the University of Tokyo in 1995. He has many years of experience in nuclear plant (BWR) design and construction at Hitachi, Ltd. During his stint at Hitachi, he has been involved in many research and development activities such as engineering for reactor systems, core thermal hydraulics and stability, high burnup fuel, and ABWR and ESBWR development. Kenichi Yoshioka started his research and engineering career at Toshiba Corporation in 1993 and earned a doctor of engineering degree from Osaka University in 2015 after receiving bachelor’s and master’s degrees from Osaka University. Then, he moved to Toshiba Energy Systems & Solutions Corporation, and to his present position in 2019. His main research interests are reactor physics and radiation detection. At present, he is working on the decommissioning and dismantling of a research reactor, the development of LWR fuels using critical experiments, the criticality safety of debris retrieval in Fukushima Daiichi, and the development of burnup analysis code with a continuous Monte Carlo technique. He has a Japanese national qualification of the chief engineer of reactors.

Preface of JSME series in Thermal and Nuclear Power Generation

Electric power supply is a fundamental and principal infrastructure for modern society. Modern society is based on the power produced from heat. This series of books consists of eight volumes describing thermal and nuclear power generation, taking Japan as the example, and referring to other countries. Volume 1 discusses how power supply is attained historically, focusing on thermal and nuclear power generation along with the minimum-required scientific and technological fundamentals to understand this series of books. Then, the present status of thermal and nuclear power generation techniques is presented in detail in Volumes 2–8. The rehabilitation and reconstruction of Japan after World War II was initiated through the utilization of a large amount of coal for boilers of the thermal power plants. Meanwhile, environmental pollution caused by coal combustion became a serious issue, following which oil was introduced for use in the boilers. Due to two worldwide oil crises and because of carbon dioxide issues, natural gas has also begun to be used in boilers. Current thermal power generation in Japan is based on coal and gas utilization. As a result of enough power supply, Japan has become one of the leading countries in the world, economically and technologically. The thermal power technology, which began with the introduction of technology from overseas, has transformed Japan into one of the most advanced countries in the world through the research and development efforts of Japanese industry, government, and academia during this process. Global warming related to excess carbon dioxide emissions has become a worldwide issue in recent years. Reducing carbon dioxide emission in thermal power generation is important to help cope with this issue. One direction is to change the fuel in the boiler from coal to gas that emits less carbon dioxide. Another important direction is to endeavor to enhance the thermal efficiency of coal thermal power plants as well as oil and gas. Many developing countries in the world will need more thermal power plants in the future. Although oil and/or gas thermal power plants may be introduced in these countries, it is assumed that coal thermal power plants will still be used due to economic reasons. Considering these factors, the publication of this series of books, which presents and explains the development history and the present status of the most advanced thermal power plants in Japan and other developed countries, is timely for engineers and researchers in developed countries to pursue further advancements and for engineers and researchers in developing countries to learn and acquire this knowledge. Nuclear power generation technology in Japan started, after being introduced from overseas, approximately 60 years ago. Then, it reached the mature stage of nuclear power technology through untiring research and development efforts. However, nuclear power plants at the Fukushima Daiichi Nuclear Power Station were heavily damaged by

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Preface of JSME series in Thermal and Nuclear Power Generation

huge tsunamis caused by the Great East Japan Earthquake in 2011, which also resulted in the contamination of large areas around the power station. Measures to improve the nuclear power reactors and to make it more robust are currently underway by analyzing the factors that caused this serious accident. Technical vulnerability can be solved by technology. Nuclear power generation technology is one of the definite promising technologies that should be used in the future. Nuclear power generation is still expected to be one of the main ways to supply electricity in the framework of the basic energy plan of Japan in addition to thermal power generation. This implies that the construction of new power reactors will be required to replace the nuclear reactors that will reach their useful lifetime. Looking overseas, many developing countries are introducing nuclear power generation technology as a safe and economically excellent way to obtain electricity. Transfer of the nuclear power generation technology developed and matured in Japan to those countries is naturally the obligation of Japan. In this situation, the need for human resource development in the field of nuclear power generation technology in the developing countries, as well as in Japan, is beyond dispute. Thus, there is an urgent need to summarize the nuclear power generation technology acquired by Japan and make it available to the developing countries as well as Japan. The Power and Energy Systems Division (PESD) of the Japan Society of Mechanical Engineers is celebrating its 30th anniversary from its establishment in 1990. This division is entrusted with handling power supply technology in mechanical engineering. Responding to the urgent demands for thermal and nuclear technologies mentioned in the previous paragraphs cannot be done by others but only by the PESD that is composed of leading engineers and researchers in this field from Japan. In view of these circumstances, summarizing Japan’s and other countries’ power generation technology and disseminating it not only in Japan but also overseas seems significantly important. So, it was decided to execute this book series, publishing it as one of the 30th anniversary events of the PESD. The authors of this book series are those who have been actually engaged in the most advanced research and development for thermal power and nuclear power generation in Japan and Canada. Their experience and knowledge are reflected in their writing. It is not an introduction of what others have done, but living knowledge based on their own experiences and thoughts. We hope that this series of books becomes learning material that is not yet in existence in this field. We hope that readers acquire a way of thinking as well as whole and detailed knowledge by having this book series in hand. This series is the joint effort of many individuals generously sharing and writing from their expertise. Their efforts are deeply appreciated. We are very thankful for the unbiased and heartfelt comments from the many reviewers that have helped to improve this series. Special thanks go to Maria Convey and Sara Valentino of the editorial staffs at Elsevier. Yasuo Koizumi, Editor-in-Chief The University of Electro-Communications, Chofu, Tokyo, Japan Mamoru Ozawa, Editor-in-Chief Kansai University, Takatsuki, Osaka, Japan

Preface

The need to develop human resources in the field of nuclear power in Japan and overseas is undisputed. It is the mission of Japanese nuclear researchers and engineers to summarize the nuclear technologies that Japanese reactor manufacturers and utility operators have acquired and improved so far, and to provide reliable books that detail these technologies to the world. In addition, many researchers and engineers from the early development stage of nuclear power in Japan have reached retirement age, and there is the risk that the knowledge accumulated on these technologies will disappear. In recent years, lectures in English have become commonplace at universities in Japan, in which many international students are enrolled. So it is expected that more lectures will be given internationally in the future. Basic information can be collected by anyone, but to cultivate human resources, it is essential to put together an international book containing reliable information about the actual design concepts that were used by the researchers and engineers who designed and built the nuclear power plants (NPPs). The target readers of the present book are engineers, researchers, and students worldwide in the technical field of nuclear power. This volume begins with a description of the history of the boiling water reactor (BWR) development followed by brief explanations of BWR structure and its systems and components including its safety systems. Overviews of progressive developments are outlined in thermal-hydraulic and nuclear performance improvements for fuel rods, fuel bundles, and core design. Improvement and standardization of BWRs in Japan are described followed by a description of the advanced boiling water reactor (ABWR) development activities by Japanese public and private sectors. Subsequently, the thermal-hydraulic analysis codes that are effective in improving the design and safety of NPPs are described. Then, the serious accident at the Fukushima Daiichi Nuclear Power Station (NPS) Units 1–4 caused by the tsunami following the Great East Japan Earthquake in March 2011 is described, mainly focusing on accident progression, the results of the post accident analysis and investigation, and the improved countermeasures for serious BWR accidents based on lessons learned from this accident. Finally, the book presents future prospects for technological developments of innovative BWRs. The descriptions in the book explain the design concepts in as much detail as possible, and in addition to design specifications that are provided in many cases, actual information such as photographs and drawings of specific plants are shown as examples. Chapter 1 describes the development history of BWRs, the situation of research and development over time, and the design features of the first unit constructed in the United States. In addition to describing the technical issues of introducing domestic

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Preface

BWRs, it also presents the cooperation between the public and private sectors that worked together to address the technical issues at that time, such as reducing the failure rate, improving the capacity factor, and reducing the exposure of workers. Chapter 2 explains the features of BWRs. The main equipment and systems of BWR, namely reactor coolant system, emergency core cooling system, instrumentation and controls, electric power, and auxiliary system are described. Chapter 3 explains the advancements in the fuel bundle structure and the development of materials to realize uranium savings and high burnup fuel. The step-by-step history of fuel development is described, in which fuel integrity was always maintained. In addition, the causes of fuel rod failures and its countermeasures are described. The characteristics of the mixed oxide (MOX) fuel, in which plutonium extracted from spent fuels is reused, are also explained. The design and the safety analysis of the BWR core and the fuel require the handling of complex steam-water two-phase flow phenomena. Chapter 3 also describes the development of thermal-hydraulic design and safety analysis methods to ensure reactor safety, the thermal-hydraulic tests using mock-up fuel bundles, and the concept of improving thermal-hydraulic and nuclear performances. In addition, the fuel bundle tests carried out to verify the BWR fuel performance are explained. Furthermore, Chapter 3 describes the events in the primary containment vessel as well as in the reactor pressure vessel during design based and severe accidents. The analysis codes and safety systems for design base and severe accidents are presented. The severe accident that occurred at the Fukushima Daiichi NPS was initiated by the inoperability of the facilities due to the damage caused by the tsunami that followed the earthquake. In Chapter 4, the event progress of the severe accident at Units 1–3 of the Fukushima Daiichi NPP and the cause of the hydrogen explosion at Unit 4 are explained. In addition, the event progress based on post accident research and severe accident analyses are described. The chapter also summarizes the factors that allowed the NPP at the Fukushima Daini NPS to escape from the severe accident, the lessons learned from the severe accident at the Fukushima Daiichi NPS, and the improvements following this accident. BWRs are attractive reactor systems and it is expected that the development of BWRs will continue in the future. This is due to their favorable features such as voids that promote the natural circulation in the core and also enable the moderator to change the density. Chapter 5 concludes the book with descriptions of innovative BWRs including trans-uranic burner, reduced-moderation water, high-pressure reactors, which take advantage of the features described previously. It explains the power uprates of BWRs that have already been implemented mainly in the United States and the research and development of high-pressure and reduced-moderation BWRs. In addition, it describes innovative elemental technologies such as post boiling transition (BT) criteria, the core catcher, the steam injector, and the built-in upper control rod drive system in the upper vessel. The Power and Energy Systems Division of the Japan Society of Mechanical Engineers (JSME) organized the editing working group for this volume in the publishing planning committee of the division. Many valuable and helpful suggestions and

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xxix

comments were received from the working group, for which the editors express their appreciation to the group members. Finally, we offer our sincere regards and gratitude to Mr. Kazuhiro Kidoguchi of the Central Research Institute of the Electric Power Industry, who was the chairman of the publishing planning committee of the Power and Energy Systems Division of JSME and supported us in writing this volume (Volume 4) on Boiling Water Reactors. Koji Nishida Institute of Nuclear Safety System, Mihama, Fukui, Japan Shinichi Morooka Former Waseda University, Shinjuku, Tokyo, Japan Michitsugu Mori Invited Guest Professor of Graduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan Yasuo Koizumi The University of Electro-Communications, Chofu, Tokyo, Japan

Editing working group for volume 4: Boiling water reactors

Kazushige Fujii Naoto Imai Kazuhiro Kidoguchi Yasuo Koizumi Michitsugu Mori Shinichi Morooka Koji Nishida Tetsuaki Takeda Taichi Takii

Toshiba Energy Systems & Solutions, Corp., Kawasaki, Kanagawa, Japan Tokyo Electric Power Company Holdings, Inc., Chiyoda, Tokyo, Japan Central Research Institute of Electric Power Industry, Yokosuka, Kanagawa, Japan The University of Electro-Communications, Chofu, Tokyo, Japan Invited Guest Professor of Graduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan Former Waseda University, Shinjuku, Tokyo, Japan Institute of Nuclear Safety System, Mihama, Fukui, Japan University of Yamanashi, Kofu, Yamanashi, Japan Hitachi-GE Nuclear Energy, Ltd., Hitachi, Ibaraki, Japan

Abbreviations

2D 3D ABB ABWR AC ACRS ACS ADS AEBL AEC AECJ AEG AESJ AET AIST ALPS ALWR AM ANL ANT AO AOO APR AS ASD ASME AST ATLAS ATWS ATR BAF BARS BEST BiMAC BOP BORAX BT BWR CAD

two dimensional three dimensional Asea Brown Boveri advanced boiling water reactor alternating current, atmospheric control system Advisory Committee on Reactor Safeguards reactor/turbine auxiliary control system automatic depressurization system atomic energy basic law Atomic Energy Commission Atomic Energy Commission of Japan Allgemeine Elektricit€ats-Gesellschaftt Atomic Energy Society of Japan Advanced Engineering Team National Institute of Advanced Industrial Science and Technology advanced liquid processing system advanced light water reactor accident management Argonne National Laboratory auxiliary normal transformer air off take system, air operated valve anticipated operational occurrence automatic power regulator (system) turbine auxiliary steam system adjustable speed drive The American Society of Mechanical Engineers auxiliary standby transformer GE’s ATLAS heat transfer facility anticipated transient without scram advanced thermal reactor bottom of active fuel BWR with an advanced recycle system BWR experimental loop for stability and transient test basemat-internal melt arrest and coolability balance of plant boiling reactor experiments boiling transition boiling water reactor computer-aided design

xxxiv

CAE CAMS CB CCFL CCFP CD CDF CF CFD CFDW C&I CIV COBRA-TF CPR CR CRD CRGT CSDM CSP CS CST CUW CV CVS CW CWP DBA DC DCH DF DG DiD DOE DTM DW, D/W DWC DWC EBR EBWR ECCS ECPR EDG, ED/G EDF EHC EM ENDF ENSI EOP

Abbreviations

computer-aided engineering containment atmospheric monitoring system circuit breaker counter current flow limiting, counter current flow limitation conditional containment failure probability condensate demineralizer (system) core damage frequency condensate filter (system) computational fluid dynamics condensate and feedwater (system) control and instrumentation combined intermediate valve coolant boiling in rod arrays, two fluid critical power ratio control rod control rod drive control rod guide tube cold shut down margin condensate storage pit core spray condensate storage tank reactor water cleanup (system) containment vessel containment venting system circulating water (system) circulating water pump design basis accident direct current direct containment heating decontamination factor diesel generator defense in depth Department of Energy digital trip module drywell drywell cooler drywell cooling system experimental breeder reactor experimental boiling water reactor emergency core cooling system experimental critical power ratio emergency diesel generator  Electricit e de France electro-hydraulic control (system) evaluation model evaluated nuclear data file The Swiss Federal Nuclear Safety Inspectorate emergency operation process

Abbreviations

EPR EPRI ERDA EQL ERC ES ESBWR ESF ESTA FA FAC FARN FBR FCI FCS FCV FCVS F/D FDW FDWC FIV FMCRD FP FPC FS FWH GCR GDCS GDS GE GE-H GEN GETAB GETR GEXL GIS GLS GIRAFFE GNF-A GNF-J GSC GSE GSEXH GT HCU HD HECW HEPA Hitachi-GE

xxxv

European pressurized water reactor Electric Power Research Institute, Inc. Energy Research and Development Administration equalizing line emergency response center extraction steam system economic simplified boiling water reactor engineered safety feature eighteen-degree sector test apparatus fuel assembly flow-accelerated corrosion Force d’Action Rapide Nucleaire fast breeder reactor fuel coolant interaction flammability gas control system flow control valve filtered containment venting system filter-demineralizer feedwater (system) feedwater control (system) flow-induced vibration, fluid-induced vibrations fine motion control rod drive fission product fuel pool cooling and cleanup (system) field switch feedwater heater gas-cooled reactor gravity-driven cooling system generator disconnecting switch General Electric Company GE-Hitachi generator General Electric BWR thermal analysis basis GE test reactor GE critical quality boiling length (correlation) gas insulated switchgear generator load switch gravity-driven integral full height test for passive heat removal Global Nuclear Fuel-Americas, LLC Global Nuclear Fuel-Japan Co., Ltd. gland steam condenser gland steam evaporator gland steam exhauster generator transformer hydraulic control unit feedwater heater drain (system) HVAC emergency cooling water (system) high efficiency particulate air Hitachi-GE Nuclear Energy, Ltd.

xxxvi

HMI HNCW HP HPCF HPCI HPCP HPCS HPDP HPDT HPIN HP-T HS HSCR HSK HV HVAC IAE IC ICS ICPR I&C IA IAEA INEL INES Internal CRD IVR JAEA JAPC JENDL JNES JP JPDR JSME LANL LCW LD LDS LEFM LLBWR LO LOCA LOOP LP LPCI LPCS LPDP LPDT LPFL

Abbreviations

human-machine interface HVAC normal cooling water (system) high pressure high-pressure core flooder (system) high-pressure core injection (system) high-pressure condensate pump high-pressure core spray (system) high-pressure drain pump high-pressure drain tank high-pressure nitrogen gas supply (system) high-pressure turbine heating steam (system) heating steam and condensate water return (system) Swiss Federal Nuclear Inspectorate feedwater heater vent (system) heating ventilating and air conditioning (system) Institute of Applied Energy isolation condenser isolation condenser system initial critical power ratio instrumentation and control instrument air (system) International Atomic Energy Agency Idaho National Engineering Laboratory International Nuclear and Radiological Event Scale internal mounted upper entry control rod drive in-vessel retention Japan Atomic Energy Agency Japan Atomic Power Company Japanese Evaluated Nuclear Data Library Japan Nuclear Energy Safety Organization jet pump Japan Power Demonstration Reactor The Japan Society of Mechanical Engineers Los Alamos National Laboratory low conductivity waste treatment (system) load driver leak detection system leading edge flow meter load-following and long operating symbiotic BWR lubricating oil (system) loss of coolant accident, loss-of-coolant accident loss of off-site power low pressure low-pressure core injection low-pressure core spray low-pressure drain pump low-pressure drain tank low-pressure core flooder (system)

Abbreviations

LPRM LP-T LSBWR LWR M/C MCC MCCI MCPR MCR METI MEXT MG MIT MITI MLHGR MO MOX MPV MS MSIV MSR MSV MU, MUR MUWC NaI(Tl) NB NCA NISA NMS NPP NPS NPT NRA NRC NSC NUPEC OECD OG OLMCPR OLU PAR P/C PCCS PCI PCIOMR PCIS PCS PCS PCT

xxxvii

local power range monitor low-pressure turbine long operating simplified BWR light water reactor metal-clad switchgear motor control center molten core and concrete interaction minimum critical power ratio main control room Ministry of Economy, Trade and Industry Ministry of Education, Culture, Sports, Science and Technology motor generator Massachusetts Institute of Technology Ministry of International Trade and Industry maximum linear heat generation rate motor operated valve mixed oxide fuel Monte Carlo code for particle transport calculation on vector processor main steam system main steam isolation valve moisture separator reheater main stop valve measurement uncertainty recapture makeup water condensate (system) thallium-doped sodium iodide scintillator nuclear boiler (system) Toshiba nuclear critical assembly Nuclear and Industrial Safety Agency neutron monitoring system nuclear power plant nuclear power station Treaty on the Nonproliferation of Nuclear Weapons Nuclear Regulatory Authority Nuclear Regulatory Commission Nuclear Safety Commission Nuclear Power Engineering Corporation Organization for Economic Cooperation and Development off-gas (system) operating limit of minimum critical power ratio output logic unit passive autocatalytic recombiner power center passive containment cooling system pellet clad interaction preconditioning interim operating management recommendations primary containment isolation system, passive core injection system plant computer system process control systems peak cladding temperature

xxxviii

PCV PCVB PF PGBR PHOENICS PIV PRA PS PSA Puf PWR QM R/A R/B RBMK RBWR RBWR-AC RBWR-TB RCCV RCIC RCPB RCS RCW RD RELAP RFP RFP-T RHR RIP RMC RMISS RMP RMU RPS RPV RRS RSC RSW RV SA SAM SBO SBWR S/C SCC SEHR SFP SGTS

Abbreviations

primary containment vessel primary containment vessel boundary profile factor plutonium generation boiling water reactor parabolic, hyperbolic, or elliptic numerical integration code series particle image velocimetry probabilistic risk assessment power supply probabilistic safety assessment fissile plutonium pressurized water reactor, pressurized light-water reactor quality management reactor area reactor building Reaktor Bolshoy Moshchnosti Kanalnyy, Reaktory Bolshoi Moshchnosti Kanalynye resource-renewable boiling water reactor RBWR actinide recycler RBWR TRU burner reinforced concrete containment vessel reactor core isolation cooling (system) reactor coolant pressure boundary reactor coolant system reactor building cooling water (system) radioactive drain transfer (system) reactor excursion and leak analysis program reactor feedwater pump reactor feedwater pump turbine residual heat removal (system) reactor internal pump recirculation motor cooling (system) recirculation motor inflatable shaft seal (system) recirculation motor purge (system) remote multiplexing unit reactor protection system reactor pressure vessel reactor recirculation system Reactor Safeguards Committee reactor building service water (system) reactor vessel severe accident sampling system, severe accident management station blackout simplified boiling water reactor suppression chamber stress corrosion cracking special emergency heat removal spent fuel pool standby gas treatment system

Abbreviations

SI SI-FWH SIPOWER SJAE SLC SLU SMR SNL SP, S/P SRNM SRV SSC SSLC T/A TAF T/B TBP TBV TCV TCW TDR TEPCO TGS TIP TLU TMI TMI-2 TOF TOSIA TRAC TRACE TRU TSW UHS UK UPS US US NRC USSR UTP VBWR V&V WEC WW

xxxix

steam injector steam injector feedwater heater steam injector-driven pool water refill steam jet air ejector standby liquid control (system) safety logic unit small modular reactor Sandia National Laboratory suppression pool start-up range neutron monitor safety relief valve systems, structures, and component safety system logic and control turbine area top of active fuel turbine building turbine bypass system turbine bypass valve temperature control valve turbine building cooling water (system) time domain reflectometry water level gauge Tokyo Electric Power Company, Inc. turbine gland steam system traversing in-core neutron probe trip logic unit Three Mile Island Three Mile Island Unit-2 time of flight Toshiba steam injector analysis code transient reactor analysis code TRAC/RELAP5 advanced computational engine transuranium element turbine building service water (system) ultimate heat sink United Kingdom uninterruptible (A.C.) power supply United States US Nuclear Regulatory Commission Union of Soviet Socialist Republics upper tie plate Vallecitos Boiling Water Reactor verification and validation Westinghouse Electric Corporation wet well

History of BWR development

1

Yasuo Koizumia, Michitsugu Morib, Shinichi Morookac, and Koji Nishidad a The University of Electro-Communications, Chofu, Tokyo, Japan, bGraduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan, cFormer Waseda University, Shinjuku, Tokyo, Japan, dInstitute of Nuclear Safety System, Mihama, Fukui, Japan

Chapter outline 1.1 Nuclear energy development in Japan 2 1.1.1 Primary energy supply 2 1.1.2 Electric power generation 4 1.1.3 Nuclear power generation 5 1.1.4 Nuclear power generation legislation 13 References 15

1.2 Establishment and realization of BWR technologies 16 1.2.1 Established stage 16 1.2.2 Realizing stage 19 References 35

1.3 Improvement and standardization program in Japan 36 1.3.1 Technology importation 36 1.3.2 First improvement and standardization program 39 1.3.3 Second improvement and standardization program 42 1.3.4 Third improvement and standardization program 44 References 48

1.4 Improvement of system and construction 50 1.4.1 Reduction of the construction period of BWRs 50 1.4.2 Improvement of ABWR system and construction 53 References 54

1.5 Construction experience and operation performance 55 1.5.1 Introduction period 55 1.5.2 Improvement and standardization programs of Japan 56 1.5.3 Recent status 57 References 58

Boiling Water Reactors. https://doi.org/10.1016/B978-0-12-821361-2.00004-0 Copyright © 2023 Elsevier Inc. All rights reserved.

2

1.1

Boiling Water Reactors

Nuclear energy development in Japan Yasuo Koizumia and Michitsugu Morib a

The University of Electro-Communications, Chofu, Tokyo, Japan, bGraduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan

1.1.1

Primary energy supply

No one doubts that modern society is based on energy. After the Edo Tokugawa Shogunate period ended in 1867 in Japan, the modern Japanese society came into existence. Primary energy supply change in Japan is presented in Fig. 1.1.1 [1]. In the Edo period, the primary energy came mainly from wood. The modernized government of Japan started to use coal. Fortunately, many coal mines were found in Japan in areas such as Kyushu, Joban, and Hokkaido. Commercial electric power generation was initiated in 1887 at the 25-kW Kayaba-cho power generation site of Tokyo Light by coal thermal power generation. Then, hydropower was also, of course, introduced. It was tried to find the resources of oil and gas in Japan. However, the trial was not successful enough and oil was almost imported from abroad and gas was produced from coal

Fig. 1.1.1 Long-term trends in the primary energy supply of Japan. Reproduced from JAEA, ATOMICA, Figure 1 Long-term trends in primary energy supply of Japan, https://atomica.jaea.go.jp/data/pict/01/01030101/01.gif, 2009.

History of BWR development

3

mined in Japan. The largest source of primary energy in Japan came from coal and partly from hydro around 1940, although wood still played some part. Japan’s industrial infrastructure was totally damaged during the World War II from 1939 to 1945. The restoration of Japan was initiated in 1945 after the end of the Word War II. The main resources of primary energy in Japan in those days were coal mined domestically and hydro. Then, as the restoration was progressing, the demand for energy increased explosively. Oil import from overseas restarted as a matter of course to meet the demand for more energy supply for the restoration and started to increase around 1950. The outbreak of the Korean War in 1950 stimulated Japanese economic and industrial restoration, and development. In response to it, Japan entered the high economic miracle period from 1955 to 1973. As shown in Fig. 1.1.1, primary energy usage increased drastically during that period. Oil was easier to handle and became cheaper than coal. Coal supply was unstable in those days because of the labor dispute at coal mines in Japan. The Government of Japan and Japanese industries changed the main energy supply from coal to oil. So oil got to play a principal role in the primary energy supply, although coal was still playing some part. The increase in the primary energy supply was mainly driven by oil. During that period, the industrial structure of Japan changed from the light industry to the heavy chemical industry. The first oil shock triggered by the Fourth Middle East War befell Japan in 1973, and then the second oil shock triggered by the Iranian Revolution occurred in 1979. These caused oil price jumps and made us realize the importance of securing the supply of energy. It aroused the movement of energy efficiency innovation and the diversity of energy importing countries. Although the amount of oil usage decreased as a result of technical improvement for a while, the amount of oil usage did not decrease significantly and returned to the former state of increasing as the economy grew. The initiation of drastic increase in the oil usage from around 1960 was also motivated by environmental cause. As a result of huge coal usage pursuing economic growth, the air was polluted by coal firing. Conversion of energy source from coal to oil in newly introduced thermal power stations was carried out to cope with it. From around 1970, natural gas and nuclear energy were also introduced as primary energy sources to solve the environmental problem and to ensure the diversity of energy sources. The diversity was necessary to avoid relying on just oil and geopolitical concentration from ensuring the energy security. As a result, natural gas and nuclear energy bore roughly 30% of the primary energy supply around 2000. Under such circumstances, the Great East Japan Earthquake happened in March 2011, and the Fukushima Daiichi Nuclear Power Station accident occurred due to the huge tsunami that followed the earthquake. That tsunami attack caused severe damage to four nuclear power plants. The severe damage resulted in severe accidents at Unit-1, Unit-2, and Unit-3. Then all nuclear power plants in Japan were shut down to reinspect safety and to meet newly introduced regulatory guidelines in which the lessons learned from the Fukushima Daiichi Nuclear Power Station accident were reflected. Although some pressurized light-water reactors (PWRs) have been restarted, nuclear energy accounts for only 3.5% of the primary energy supply as of 2019 [2].

4

1.1.2

Boiling Water Reactors

Electric power generation

Primary energy is converted into other forms such as electricity, power, and heat and then consumed in industry, commerce, homes, and transportation. Electricity is easy to transport once power transmission lines are spread, and is also easy to control and to keep the environment clean. Thus, the role of electricity has increased to take the large part of primary energy conversion as shown in Fig. 1.1.2 [3]. Almost half of the primary energy has been used to produce electricity as of 2018 in Japan. In the early stage of the restoration of Japan from the World War II, electricity was mainly generated by hydro as depicted in Fig. 1.1.3 [4]. Then coal was introduced. Oil started to be used around 1960 as mentioned in the preceding section. It is clearly shown in the figure that electricity has been mainly produced by thermal energy since the 1960s although hydropower has still kept some part firmly. Nuclear power generation started in 1966 and then increased its presence. The nuclear power generation in Japan became approximately one-fourth of the total electric power generation in 2010. However, it decreased drastically after the severe accidents at the Fukushima Daiichi Nuclear Power Station in March 2011. The nuclear’s role in electric power generation in 2018 was only 6.2%. On the other hand, electricity from the thermal energy of coal, gas, and oil accounts for 76.9%. Even electricity generated by renewable sources was 16.9%. The state of largely relying on coal, gas, and oil that exhaust carbon dioxide related to global warming is factitious and immediate remedy is expected.

Fig. 1.1.2 Change of electrification ratio; percentage of energy that is used to produce electricity in primary energy. Reprinted from Japan Atomic Energy Relations Organization, Graphical Flip-chart of Nuclear & Energy Related Topics, 1-2-9, Percentage of Electric Power in Primary Energy (Electrification Ratio), https://www.ene100.jp/www/wp-content/uploads/zumen/e1-2-9.pdf, 2021.

History of BWR development

5

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)LVFDO < Momentum > = Advection of the momentum þ fPressure forceg variation in the ¼ > > : ; into or out of the volume control volume ( ) Volumetric wall drag þ fbody forceg þ of the vapor or liquid 8 9 Momentum > > > > > > < = þ fInterfacial dragg Variation due to > > > > > > : ; the phase change (3.3.2.4) Each term of Eq. (3.3.2.4) can be shown with the variables of Fig. 3.3.2.3 as follows: 



Vapor momentum variation in the control volume

 ¼

Advection of the momentum



into or out of the volume fVapor pressure forceg ¼ 



∂ αg ρg V g AΔzΔt ∂t

¼



∂ αg ρg V g V g AΔzΔt ∂z

 ∂ αg Pg AΔzΔt ∂z

fbody force for vaporg ¼ Fgg AΔzΔt where Fgg is the body force for the vapor/gas mixture phase per unit volume (N/m3). 

Volumetric wall drag of the vapor

 ¼ Fwg AΔzΔt

where Fwg is the wall drag for the vapor/gas mixture phase per unit volume (N/m3). fInterface drag for vapor phase g ¼ Fig AΔzΔt where Fig is the interfacial drag for the vapor/gas mixture phase per unit volume (N/m3). 

Vapor momentum variation due to the phase change

 ¼ Γ g V ig AΔzΔt

The relations for momentum conservation of the liquid phase can be shown by the same way as for the vapor phase.

276

Boiling Water Reactors





Liquid momentum variation

¼

in the control volume 

Advection of the momentum



into or out of the volume fLiquid pressure forceg ¼ 

∂ ðα ρ V ÞAΔzΔt ∂t l l l

¼

∂ ðα ρ V V ÞAΔzΔt ∂z l l l l

∂ ðα P ÞAΔzΔt ∂z l l

fbody force for liquidg ¼ Fgl AΔzΔt where Fgl is the body force for the liquid phase per unit volume (N/m3). 

Volumetric wall drag of the liquid

 ¼ Fwl AΔzΔt

where Fwl is the wall drag for the liquid phase per unit volume (N/m3). fInterface drag for liquid phase g ¼ Fil AΔzΔt where Fil is the interfacial drag for the liquid phase per unit volume (N/m3). 

Liquid momentum variation due to the phase change

 ¼ Γ l V il AΔzΔt

The preceding relations for each phase can be substituted into Eq. (3.3.2.4) to get the respective phase momentum conservation equation as follows: For the vapor phase,



 ∂ ∂ ∂ α g ρg V g + αg ρg V g V g ¼ Fwg  Fig  Fgg  α P ∂t ∂z ∂z g g + Γ g V ig

(3.3.2.5)

can be derived. For the liquid phase, ∂ ∂ ∂ ðα ρ V Þ + ðαl ρl V l V l Þ ¼ Fwl  Fil  Fgl  ðαl Pl Þ + Γ l V il ∂t l l l ∂z ∂z

(3.3.2.6)

can be derived. The above two equations are equivalent to Eqs. (1-21) and (1-22) of Ref. [2]. –

Derivation of the energy equations In the control volume as shown in Fig. 3.3.2.4, there are some energy terms that are to be considered to set basic energy conservation: the energy term from the wall surface to the

Nuclear reactor dynamics and thermal hydraulics of reactor core

277

Fig. 3.3.2.4 Schematic representation of the energy conservation (two-fluid model). vapor phase, qwg, and to the liquid phase, qwl, and the energy from the interphase to the vapor phase, qig, and to the liquid phase, qil. The basic principle of conservation of energy can be stated as 

Energy variation in the control volume



 ¼

   Advection of the energy The workdueto + into  or out of the volume  the wall shear stress   The workdue to the The work due The work due + + + to pressureforce  interfacial shear stress  to body force Energy variation Energy added from the + + dueto the phase change outside of the control volume (3.3.2.7)

Each term of Eq. (3.3.2.7) can be shown with the variables of Fig. 3.3.2.4 as follows: The relations for the energy conservation of the vapor/gas mixture phase can be shown as follows: 



Vapor energy variation in the control volume

 ¼

n

o ∂ 1 αg ρg eg + V 2g AΔzΔt ∂t 2

Advection of the vapor energy into or out of the volume



    Pg ∂ 1 2 ¼ + V g V g AΔzΔt α ρ e + ∂z g g g ρg 2

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Boiling Water Reactors



The work due to the wall



shear stress for the vapor 

 ¼ Fwg V g AΔzΔt

The work due to the interfacial



shear stress for the vapor 

The work due to the body force

 ¼ Fig V g AΔzΔt



for the vapor 

The work due to the



vapor pressure force 

 ¼ Fgg V g AΔzΔt

n o ∂ ¼  αg Pg V g AΔzΔt ∂z

Vapor energy variation



due to the phase change



1 ¼ Γ g hig + V 2g AΔzΔt 2

where hig means the interfacial enthalpy (J/kg) and Vig means the interfacial velocity (m/s). 



Energy added to the vapor phase from the outside of the control volume



¼ qwg + qig AΔzΔt

where qwg is the energy transfer rate from the wall to the vapor/gas mixture phase (J/m3 s) and qig is the energy transferred from the interface to the vapor/gas mixture phase (J/m3 s). The relations for momentum conservation of the liquid phase can be shown by the same way as for the vapor phase. Each term of the above equation can be shown with the variables of Fig. 3.3.2.4 as follows: 

Liquid energy variation



in the control volume 

¼

n

o ∂ 1 αl ρl el + V 2l AΔzΔt ∂t 2

Advection of the liquid energy

 ¼

into or out of the volume   

The work due to the wall shear stress for the liquid

    ∂ P 1 αl ρl el + l + V 2l V l AΔzΔt ρl ∂z 2

 ¼ fFwl V l gAΔzΔt

The work due to the interfacial shear stress for the liquid The work due to the body force for the liquid

 ¼ fFil V l gAΔzΔt 

 ¼ Fgl V l AΔzΔt

Nuclear reactor dynamics and thermal hydraulics of reactor core



The work due to the



liquid pressure force 

Liquid energy variation due to the phase change

279

n o ∂ ¼  ðαl Pl V l Þ AΔzΔt ∂z 



1 ¼ Γ l hil + V 2l AΔzΔt 2

where hil means the enthalpy and Vil means the velocity at the interface. 

Energy added to the liquid phase from the outside of the control volume

 ¼ ðqwl + qil ÞAΔzΔt

The preceding relations for each phase can be substituted into Eq. (3.3.2.7) to get the respective phase energy conservation equations as follows: For the vapor phase, n

o n

o ∂ 1 ∂ 1 + αg ρg eg + V 2g αg ρg eg + Pg=ρg + V 2g V g ∂t 2 ∂z 2

 ∂ 1 ¼ Fwg V g  Fig V g  Fgg V g  αg Pg V g + Γ g hig + V 2g + qwg + qig ∂z 2 (3.3.2.8) can be derived. For the liquid phase, n

o n

o ∂ 1 ∂ 1 + α1 ρ1 e1 + V 21 α1 ρ1 e1 + P1=ρ1 + V 21 V 1 ∂t 2 ∂z 2

∂ 1 ¼ Fw1 V 1  Fi1 V 1  Fg1 V 1  ðα1 P1 V 1 Þ + Γ 1 hi1 + V 21 + qw1 + qi1 ∂z 2 (3.3.2.9) can be derived. The above two equations are equivalent to Eqs. (1-23) and (1-24) of Ref. [2]. (b) The closure problem for the two-fluid model The basic equations of mass conservation are Eqs. (3.3.2.2) and (3.3.2.3), the basic equations of momentum conservation are Eqs. (3.3.2.5) and (3.3.2.6), and the basic equations of energy conservation are Eqs. (3.3.2.8) and (3.3.2.9); therefore, each phase equation has 14 variables (average void fraction, αk, density, ρk, phasic velocity, Vk, phasic pressure, Pk, wall shearing stress, Fwk, external force, Fgk, internal energy, ek, wall surface heat flux, qwk, mass generation rate, Г k, interfacial shear stress, Fik, interfacial heat flux, qik, interfacial velocity, Vik, interfacial pressure, Pik, and interfacial enthalpy, hik, on the gasliquid interface for each k-phase (k ¼ g, l)). So, the number of the variables is 28 in this whole equation set. That is, in order to close the system of equations and solve this equation set, we should prepare the additional 22 correlations because six basic equations of the two-fluid model have already been prepared as above. The relationship for the void fractions of the two-phase flow can be prepared as follows:

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Boiling Water Reactors

αg + αl ¼ 1

(3.3.2.10)

The jump conditions for mass conservation, momentum conservation, and energy conservation on the gas-liquid interface is as follows: Γg + Γl ¼ 0

(3.3.2.11)

Fig + Fil ¼ 0

(3.3.2.12)

hgi Γ g + qig + hli Γ l + qil ¼ 0

(3.3.2.13)

The density and the enthalpy in the basic equations are determined for the function of the temperature and the pressure with an equation of state, and the body force on each phase, Fgk can be defined by the external function such as a gravity force. If the variables of one phase Fwk, qwk, Vki, Pik, hik, Г k, Fgk, and qik are defined, the variables of the other phase are easily determined by using the Eqs. (3.3.2.11)–(3.3.2.13). The variables Vki, Pik, and hik on the interface can be defined by the proper assumption of the thermal property, that is, for the variable Vik, the assumption that the phase change is not so big, the following equation is approximately derived below: V ig ¼ V il ð¼ V i Þ

(3.3.2.14)

The interfacial velocity Vi is also related with the phasic velocity Vg and Vl based on each flow pattern condition and thermal condition. That is, for example, for the bubbly flow regime, vapor/gas mixture velocity in the bubble can be supposed to be uniform; so, Vi ¼ Vg

(3.3.2.15)

For the mist flow regime, the opposite condition can be assumed; therefore, Vi ¼ Vl

(3.3.2.16)

By the assumption that the influence of surface tension on the gas-liquid interface is not so large, the following relation can be obtained: Pig ¼ Pil ð¼ Pi Þ

(3.3.2.17)

The pressure on the interface, Pi, is related to the vapor pressure, Pg, because the density of the vapor/gas mixture phase is adequately smaller than the density of the liquid; the contribution of the kinetic energy to the pressure is rather small. In this case, the following relation: Pi ¼ Pg

(3.3.2.18)

is applied. In the case of the two-fluid model, the thermal equilibrium condition between the vapor phase and the liquid phase is not assumed; but, in the neighborhood of the interface the equilibrium condition can be assumed. Therefore, the relation between the vapor temperature, Tgi, and the liquid temperature, Tli, can be supposed to be equal: T ig ¼ T il ð¼ T i Þ

(3.3.2.19)

Nuclear reactor dynamics and thermal hydraulics of reactor core

281

and the interfacial temperature, Ti, equals the saturation temperature, Tsat of the interfacial pressure in the case of a one-component system. By using the proceeding formation, the variables on the vapor-liquid interface, Vki, Pki, and hki can be expressed by using other variables of the two-phase flow parameter. In the case of the two-fluid model, the shear stress on the wall with each phase, Fwg and Fwl, the wall surface heat flux for each phase, qwg and qwl, are needed to close the equations system, but only whole shear stress, Fw and whole heat flux, qw, on the wall surface can be obtained from the experimental data. So, the void fraction is used for partitioning to get each phase value, Fwg, Fwl, qwg, and qwl from the whole values, Fw and qw. As shown above, decreasing the number of unknown parameters can be decreased by introducing the correlations of constitutive equations to basic equations. Finally, we can solve the system of equations of the two-fluid model and set the solution. The interfacial shear stress, Fig, is the characteristic for a two-fluid model and is very important to affect the behavior of the two-phase phenomena. Because this variable is caused by the difference of two-phase velocities; these are calculated by weighting the twophase effect on the friction factor for the single bubble and single droplet. In order to close the equation system described above, the following four valuable constitutive models, – interfacial drag model (see pp. 123–170 of Ref. [2]) – wall drag model (see pp. 170–186 of Ref. [2]) – interfacial heat transfer model (see pp. 197–240 of Ref. [2]) – wall heat transfer model (see pp. 245–357 of Ref. [2]) describe in detail the background of their development.

3.3.2.2

Thermal-hydraulic subchannel analysis code

In this part, the author will introduce the software to be used mainly for the analysis of the thermal-hydraulic behavior in the fuel rod assembly (namely the subchannel, see Fig. 3.3.2.5), which is generally called “thermal-hydraulic subchannel analysis code.” The subject of this subchannel analysis is the prediction of the coolable condition on the fuel surface by determining the mass of the subchannel. In the early stage of twophase flow studies, as stated above, the homogeneous model was used for an ideal two-phase model. It is very simple to use and it could be applied to design a nuclear fuel. However, in order to develop a further more safe and highly efficient fuel design, we need a more accurate model for the thermal-hydraulic analysis. So, we selected a nonhomogeneous model (separated flow model) to simulate more complicated flow phenomena such as the thermal-hydraulic behavior around the fuel rod, the entrained droplet in the vapor phase the de-entrained droplet after a grid spacer, etc. In the separated flow model, the phase is assumed to flow side by side. In the case of the subchannel model, as shown in Fig. 3.3.2.5, we generally use an additional entrained droplet field equation for the two-fluid model consisting of a liquid field (liquid film on the rod surface) and a vapor/gas mixture field, because the behavior of the droplet, such as entrainment and deentrainment on the fuel rod surface, is very important to determine the coolable condition on the fuel rod surface, that is, whether dryout occurs or not. Separate equations are written for each phase and the interaction between phases is also considered with the constitutive correlations. Therefore, as the

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Boiling Water Reactors

Fig. 3.3.2.5 Schematic representation of the flow pattern in the fuel assembly (threefield model).

two-phase model is getting more accurate, we should prepare more constitutive equations to close the new thermal-hydraulic system of equations for the three-fluid model, which includes the vapor field, liquid film field on the fuel rod surface, and entrained droplet field in the vapor phase. The Coolant Boiling in Rod Arrays—Two-Fluids (COBRA-TF) code [5] was selected as the typical thermal-hydraulic subchannel analysis code to study the new approach of a three-fluid model for an advanced subchannel analysis code. The relationship of the three fluid fields in the three-fluid model is shown in Fig. 3.3.2.5. We should mainly note the advanced constitutive equations, which are needed to be fully renewed to be applied to new complicated phenomena. There are a wide range of studies and a long history in the background of this new approach. Although the entrained droplet due to the liquid boiling effect is shown in Fig. 3.3.2.5, this mechanism is not explicitly modeled in the current code [5]. Fortunately, a lot of the valuable information about this code is available for us now. Once we study the new common knowledge of the typical analysis code such as COBRA-TF, the reader of this textbook is advantageously placed to communicate about the latest research with the people of the world. (a) Derivation of the field equation set of the subchannel analysis code The basic equations of the three-fluid model [5] can be driven from mass conservation, momentum conservation, and energy conservation for the separated vapor, liquid, and entrained droplet phase flows. The procedure of deriving the system of equations is basically rather difficult due to strict mathematical derivation; so, it may be burdensome for the reader. Therefore, some references [3, 4] are listed to support the readers’ understanding about the detailed derivations of the basic field equations. In the current subchannel approach, we should consider the lateral cross flow model between the fuel rods without two-dimensional spatial differential terms. Therefore, once

Nuclear reactor dynamics and thermal hydraulics of reactor core

283

the lateral flow leaves a subchannel, it flows into the adjacent subchannel in an instant. Thus it is confirmed that this approach is suitable and effective to reduce the calculation time under the condition of axially-dominated flow in a reactor fuel bundle. – Derivation of the mass equations The basic conservation principle of general mass can be stated as 

 Advection of the mass into or out of the subchannel in the control subchannel     Mass transfer due to the Lateral mass tranfer þ þ phase change rate through the gap   Mass variation due to the þ turbulent mixing and void drift Mass variation





¼

(3.3.2.20)

Each term of Eq. (3.3.2.20) can be shown with the variables of Fig. 3.3.2.6. This figure consists of three fields and each arrowhead indicates the direction of mass transfer between fields. The relations for mass conservation of the vapor/gas mixture phase can be shown as follows: The term of the left-hand side of Eq. (3.3.2.20) for the vapor phase is determined by using physical variables, that is, 

Vapor mass variation in the control subchannel

 ¼



∂ αg ρg AΔzΔt ∂t

where A is the cross-section area (m2) of the control subchannel.

Fig. 3.3.2.6 Schematic representation of the mass transfer between the fields (three-fluid model).

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Boiling Water Reactors

The first term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 

Advection of the vapor mass into or out of the subchannel

 ¼



∂ αg ρg V g AΔzΔt ∂z

where Vg is the axial vapor/gas mixture phasic velocity (m/s). The second term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 



Lateral vapor mass transfer rate through nk gaps into or out of the subchannel

¼

nk X

αg ρg W g



m¼1

m

AΔzΔt

where Wg is the lateral gap vapor/gas mixture phasic velocity (m/s) and nk is the number of “gaps” connected with the subchannel. The third term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, (

Vapor mass transfer rate due

¼

)

to the phase change 8 Vapor generation rate from liquid to vapor: Γ e0 > > > > > < Vapor condensation rate from vapor to liquid: Γ c0

9 > > > > > = AΔzΔt

> + Vapor generation rate from droplet to vapor: Γ ed0 > > > > > > > > > : ; Vapor condensation rate from vapor to droplet: Γ cd ¼ ðΓ e  Γ c þ Γ ed  Γ cd ÞAΔzΔt

where Г e is the vapor generation rate per unit volume (kg/m3 s) from the liquid phase to the vapor phase, Г c is the vapor condensation rate per unit volume (kg/m3 s) from the vapor phase to the liquid phase, Г ed is the vapor generation rate per unit volume (kg/m3 s) from the entrained droplet phase to the vapor phase, Г cd is the vapor condensation rate per unit volume (kg/m3 s) from the vapor phase to the entrained droplet phase. The fourth term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 

Mass variation due to the turbulent mixing and void drift



 ¼

Vapor mass variation due to the



turbulent mixing and void drift ¼ mtg AΔzΔt

where mtg is the vapor mass variation due to turbulent mixing and void drift.

Nuclear reactor dynamics and thermal hydraulics of reactor core

285

The relations for mass conservation of the liquid phase can be shown by the same way as for the vapor phase. The term of the left-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is,   Liquid mass variation ∂ ¼ ðαl ρl ÞAΔzΔt ∂t in the control subchannel The first term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 

Advection of the liquid mass into or out of the subchannel

 ¼

∂ ðα ρ V ÞAΔzΔt ∂z l l l

where Vl is the axial liquid phasic velocity (m/s). The second term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is,   nk X Lateral liquid mass transfer rate ¼ ðαl ρl W l Þm AΔzΔt through nk gaps into or out of the subchannel m¼1 where Wl is the lateral gap liquid phasic velocity (m/s). The third term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, ( ) Vapor mass transfer rate due

¼

to the phase change 8 Vapor generation rate from liquid to vapor: Γe0 > > > > > < + Vapor condensation rate from vapor to liquid: Γc0

9 > > > > > = AΔzΔt

> > > Droplet entrainment rate from liquid to droplet: SE0 > > > > > > > : ; + Droplet deentrainment rate from droplet to liquid: SD ¼ ðΓ e þ Γ c  SE þ SD ÞAΔzΔt

where SE is the entrained droplet rate per unit volume (kg/m3 s) from the liquid field to the entrained droplet field and SD is the deentrained droplet rate per unit volume (kg/m3 s) from the entrained droplet field to the liquid film field. The fourth term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 

Mass variation due to the turbulent mixing and void drift





Liquid mass variation due to the turbulent mixing and void drift ¼ mtl AΔzΔt



¼

where mtl is the liquid mass variation due to turbulent mixing and void drift.

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Boiling Water Reactors

The relations for the mass conservation of the droplet field can be shown by the same way as for the liquid phase. The term of the left-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 

Droplet mass variation in the control subchannel

 ¼

∂ ðα ρ ÞAΔzΔt ∂t d d

The first term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 

Advection of the droplet mass into or out of the subchannel

 ¼

∂ ðα ρ V ÞAΔzΔt ∂z d d d

where Vd is the axial droplet phasic velocity (m/s). The second term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 

Lateral droplet mass transfer rate through nk gaps into or out of the channel

 ¼

nk X

ðαd ρd W d Þm AΔzΔt

m¼1

where Wd is the lateral gap droplet phasic velocity (m/s). The third term of the right-hand side of Eq. (3.3.2.20) is determined by using physical variables, that is, 

 Mass transfer rate due to the phase change 8 9 Entrained droplet rate from liquid film to droplet: SE0 > > > > > > > > > < Deentrained droplet rate from droplet to liquid film: SD0 > = ¼ AΔzΔt > + Vapor generation rate from droplet to vapor: Γ ed0 > > > > > > > > > : ; + Vapor condensation rate from vapor to droplet: Γ cd ¼ ðSE  SD  Γed þ Γcd ÞAΔzΔt

The preceding relations for each phase can be substituted into Eq. (3.3.2.20) to get the respective field mass conservation equation as follows: For the vapor/gas mixture field, nk



X

∂ ∂ αg ρg + αg ρg V g + αg ρg W g ∂t ∂z m m¼1

¼ Γ e  Γ c + Γ ed  Γ cd + mtg

(3.3.2.21)

Nuclear reactor dynamics and thermal hydraulics of reactor core

287

For the liquid (liquid film) field, nk X ∂ ∂ ðαl ρl Þ + ðαl ρl V l Þ + ðαl ρl W l Þm ∂t ∂z m¼1

¼ Γ e + Γ c  SE + SD + mtl

(3.3.2.22)

For the entrained droplet field, nk X ∂ ∂ ðαd ρd Þ + ðαd ρd V d Þ + ðαd ρd W d Þm ¼ SE  SD  Γ ed + Γ cd ∂t ∂z m¼1

(3.3.2.23)

The above three equations are equivalent with Eqs. (2-26)–(2-28) of Ref. [5]. –

Derivation of the momentum equations In the control subchannel as shown in Fig. 3.3.2.7, there are the forces from the wall surface to the vapor phase, Fwg, and to the liquid phase, Fwl, the forces from the interphase to the vapor phase, Fig, and to the liquid phase, Fil, and the partial pressure forces of the vapor phase, Pg, and of the liquid phase, Pl. The basic conservation principle of momentum can be stated as,

Fig. 3.3.2.7 Schematic representation of the momentum transfer between the fields (threefluid model).

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Boiling Water Reactors

Momentum variation in the control subchannel



 ¼

Advection of the momentum into or out of the subchannel   Lateral momentum + transfer rate through gap



+ fpressure forceg + fbody forceg   Volumetric wall drag of + the vapor or liquid phase + fInterfacial dragg   Momentum variation due to phase change + and entrainment or de  entrainment   Momentum source due to + turbulent mixing and void drift

(3.3.2.24)

The relations for the momentum conservation of the vapor-gas mixture field can be shown as follows: Each term of Eq. (3.3.2.24) can be shown with the variables of Fig. 3.3.2.7 as follows: 





Vapor momentum variation in the control subchannel

 ¼



∂ α ρ V AΔzΔt ∂t g g g

Advection of the vapor momentum into or out of the subchannel Lateral vapor momentum transfer rate through nk gaps



 ¼

¼



∂ αg ρg V g V g AΔzΔt ∂z

nk X



αg ρg V g W g

m¼1

m

AΔzΔt

  ∂Pg + Fgg AΔzΔt fPressure forceg + fBody forceg ¼  αg ∂z 

Volumetric wall drag of the vapor phase

 ¼ Fwg AΔzΔt

where Fwg is the wall drag force per unit volume (N/m3).  fInterfacial dragg ¼ Fig  Fidg AΔzΔt where Fig is the interfacial drag force per unit volume (N/m3) between the interface (liquid film and vapor) and the vapor phase and Fidg is the interfacial drag force per unit volume (N/m3) between the interface (entrained droplet and vapor) and the vapor phase. 

Vapor momentum variation due to phase change and  entrainment or deentrainment ¼ V ig Γ c + V il Γ e  V ig Γ cd + V id Γ ed AΔzΔt



where VigГ c is the momentum exchange rate due to the mass exchange of vapor condensation from the vapor phase to the liquid phase, VilГ e is the momentum exchange

Nuclear reactor dynamics and thermal hydraulics of reactor core

289

rate due to the mass exchange of vapor generation from the liquid phase to the vapor phase, VigГ cd is the momentum exchange rate due to the mass exchange of vapor condensation from the vapor phase to the entrained droplet, and VidГ ed is the momentum exchange rate due to the mass exchange of vapor generation from the entrained droplet. 



Vapor momentum source due to turbulent mixing and void drift

¼ Ftg AΔzΔt

where Ftg is the vapor momentum exchange rate (N/m3) due to the vapor-phase turbulent mixing and void drift. The relations for momentum conservation of the liquid phase can be shown by the same way as for the vapor phase.   Liquid momentum variation ∂ ¼ ðαl ρl V l ÞAΔzΔt ∂t in the control subchannel 



Advection of the liquid momentum into or out of the subchannel Lateral liquid momentum transfer rate through nk gaps

 ¼

 ¼

nk X

∂ ðα ρ V V ÞAΔzΔt ∂z l l l l

ðαl ρl V l W l Þm AΔzΔt

m¼1



∂P fPressure forceg + fBody forceg ¼  αl l + Fgl AΔzΔt ∂z   Volumetric wall drag ¼ Fwl AΔzΔt of the liquid phase fInterfacial dragg ¼ fFil gAΔzΔt 

Liquid momentum variation due to phase change



and  entrainment or deentrainment ¼ V ig Γ c  V il Γ e  V il SE + V id SD AΔzΔt where VilSE is the momentum exchange rate due to the entrainment of the droplet from the liquid film to the entrained droplet field and VidSD is the momentum exchange rate due to the deentrainment of the droplet from the entrained droplet field to the liquid film. 

Liquid momentum source due to turbulent mixing and void drift

 ¼ Ftl AΔzΔt

where Ftl is the liquid momentum exchange rate (N/m3) due to the liquid-phase turbulent mixing and the void drift. The relations for momentum conservation of the entrained droplet field can be shown by the same way as for the liquid phase.   Droplet momentum variation ∂ ¼ ðαd ρd V d ÞAΔzΔt ∂t in the control subchannel

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Boiling Water Reactors





Advection of the droplet momentum into or out of the subchannel Lateral droplet momentum transfer rate through nk gaps

 ¼

 ¼ nk X

∂ ðα ρ V V ÞAΔzΔt ∂z d d d d

ðαd ρd V d W d Þm AΔzΔt

m¼1



∂P fPressure forceg + fBody forceg ¼  αd d + Fgd AΔzΔt ∂z   Volumetric wall drag ¼0 of the droplet phase  fInterfacial dragg ¼ Figd AΔzΔt where Figd is the interfacial drag force per unit volume (N/m3) between the interface (vapor and droplet) and the vapor phase. 

Droplet momentum variation due to phase change and  entrainment or deentrainment ¼ V ig Γ cd  V id Γ ed + V il SE  V id SD AΔzΔt



The preceding relations for each phase can be substituted for Eq. (3.3.2.24) to get the respective phase momentum conservation equation as follows: For the vapor/gas mixture field,  



Xnk

∂Pg ∂ ∂ αg ρg V g + αg ρg V g V g + + F α ρ V W ¼  α g g g g gg g m¼1 ∂t ∂z ∂z m Fwg  Fig  Fidg  V ig Γ c + V il Γ e  V ig Γ cd + V id Γ ed + Ftg (3.3.2.25) can be derived. For the liquid (liquid film) field,

Xnk ∂ ∂ ∂P1 + F ðα1 ρl V 1 Þ + ðα1 ρ1 V 1 V 1 Þ + ð α ρ V W Þ ¼  α 1 1 1 1 g1 1 m m¼1 ∂t ∂z ∂z Fwg  Fi1  V ig Γ c + V il Γ e  V i1 SE + V id SD + Ft1

(3.3.2.26)

can be derived. For the entrained droplet field,   Xnk ∂ ∂ ∂Pd + F ðαd ρd Vd Þ + ðαd ρd Vd Vd Þ + ð α ρ V W Þ ¼  α d d d d gd d m m¼1 ∂z ∂t ∂z  Figd + Vig Γ cd  Vid Γ ed + Vi1 SE  Vid SD

(3.3.2.27)

can be derived. The above three equations are equivalent to Eqs. (2-29)–(2-31) of Ref. [5].

Nuclear reactor dynamics and thermal hydraulics of reactor core

–



291

Derivation of the energy equations In the control subchannel as shown in Fig. 3.3.2.8, there are some energy terms that are to be considered to set basic energy conservation: the energy term from the wall surface to the vapor phase, qwg, and to the liquid phase, qwl, and the energy term from the interphase to the vapor phase, qig, and to the liquid phase, qil. The basic conservation principle of energy can be stated as    Energy variation Advection of energy into ¼ in the control subchannel or out of the subchannel     Lateral energy transfer rate Energy variation + + through the gap due to the phase change     Energy added from the outside energy source due to + + of the control subchannel turbulent mixing and void drift

(3.3.2.28)

Each term of Eq. (3.3.2.28) can be shown with the variables of Fig. 3.3.2.8 as follows: The relations for energy conservation of the vapor/gas mixture field can be shown as follows: 

Vapor energy variation in the control subchannel

 ¼

n

o ∂ 1 αg ρg eg + V 2g AΔzΔt ∂t 2

where eg is the vapor internal energy (J/kg).       Pg ∂ 1 Advection of vapor energy αg ρg eg + ¼ + V 2g V g AΔzΔt ∂z 2 ρg into or out of the subchannel

Fig. 3.3.2.8 Schematic representation of the energy transfer between the fields (threefluid model).

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n

Lateral vapor energy transfer rate through the gap o

 ¼

nk  X m¼1

   Pg 1 αg ρg eg + + V 2g W g AΔzΔt 2 ρg m



Vapor energy variation due to the phase change ¼ +qig + qidg ; qil  qigd AΔzΔt

where qig is the energy transfer rate per unit volume (J/m3 s) due to the phase change of liquid vaporization to the vapor phase, qil is the energy transfer rate per unit volume (J/ m3 s) due to the phase change of vapor condensation to the liquid phase, qidg is the energy transfer rate per unit volume (J/m3 s) due to the phase change of entrained droplet vaporization to the vapor phase, and qigd is the energy transfer per unit volume (J/m3 s) due to the phase change of vapor condensation to the entrained droplet. 

Energy added to the vapor phase from the outside of the control subchannel





¼ qwg AΔzΔt

where qwg is the energy transfer from the outside of the control subchannel to the vapor/gas mixture phase. 



Vapor energy source due to

turbulent mixing and void drift



¼ qtg AΔzΔt

where qtg is the vapor energy source per unit volume (J/m3 s) due to the turbulent mixing and void drift. The relations for energy conservation of the liquid phase can be shown by the same way. Each term of the above equation can be shown with the variables of Fig. 3.3.2.8 as follows:  





Liquid energy variation in the control subchannel

 ¼

Advection of liquid energy into or out of the subchannel Lateral liquid energy transfer rate through nk gaps

n

o ∂ 1 αl ρl el + V 2l AΔzΔt ∂t 2

 ¼

 ¼

    ∂ P 1 αl ρl el + l + V 2l V l AΔzΔt ∂z 2 ρl nk  X m¼1

   P 1 αl ρl el + l + V 2l W l AΔzΔt 2 ρl m

Liquid energy variation due to the phase change and entrainment or deentrainment  ¼ qil  qig  hil SE + hid SD AΔzΔt



where hil is the interfacial enthalpy (J/kg) of the liquid phase, hid is the interfacial enthalpy (J/kg) of the droplet, SE is the entrained droplet rate (kg/m3 s), and SD is the deentrained droplet rate (kg/m3 s).

Nuclear reactor dynamics and thermal hydraulics of reactor core



Energy added to the liquid phase from the outside of the control subchannel

293

 ¼ ðqwl ÞAΔzΔt

where qwl is the energy transfer from the outside of the control subchannel to the liquid phase. 

 Liquid energy source due to ¼ ðqtl ÞAΔzΔt turbulent mixing and void drift

where qtl is the liquid energy source per unit volume (J/m3 s) due to turbulent mixing and void drift. The relations for energy conservation of the entrained droplet field can be shown by the same way as for the liquid. Each term of the above equation can be shown with the variables of Fig. 3.3.2.8 as follows:  







Droplet energy variation in the control subchannel

 ¼

Advection of droplet energy into or out of the subchannel

n

o ∂ 1 αd ρd ed + V 2d AΔzΔt ∂t 2

 ¼

    ∂ P 1 αd ρd ed + d + V 2d V d AΔzΔt ∂z 2 ρd

    nk  X Lateral droplet energy P 1 ¼ αd ρd ed + d + V 2d W d AΔzΔt transfer rate through nk gaps 2 ρd m m¼1 Droplet energy variation due to the phase change and entrainment or deentrainment Energy added to the droplet phase from the outside of the control subchannel



n o ¼ qigd  qidg + hil SE  hid SD AΔzΔt

 ¼0

The preceding relations for each phase can be substituted for Eq. (3.3.2.28) to get the respective phase energy conservation equation as follows: For the vapor/gas mixture phase, n

o n

o ∂ 1 ∂ 1 αg ρg eg + V 2g αg ρg eg + Pg=ρg + V 2g V g + ∂t 2  ∂z 2   Xnk  Pg 1 2 + αg ρg eg + + Vg Wg m¼1 2 ρg m ¼ qig + qidg  qil  qigd + qwg + qtg can be derived. For the liquid phase,

(3.3.2.29)

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Boiling Water Reactors

n

o n

o ∂ 1 ∂ 1 α1 ρ1 e1 + V 21 α1 ρ1 e1 + P1=ρ1 + V 21 V 1 + ∂t 2  ∂z 2   Xnk  P1 1 2 + α1 ρ1 e1 + + V1 W1 m¼1 2 ρ1 m ¼ qi1  qig  hi1 SE + hid SD + qw1 + qt1

(3.3.2.30)

can be derived. For the entrained droplet field, n

o n

o ∂ 1 ∂ 1 αd ρd ed + V 2d αd ρd ed + Pd=ρd + V 2d V d + ∂t 2  ∂z 2   Xnk  Pd 1 2 + αd ρd ed + + Vd Wd m¼1 2 ρd m ¼ qigd  qidg + hi1 SE  hid SD

(3.3.2.31)

Considering material properties, the liquid of the droplet and the liquid of the liquid film are almost the same; so, ρd ¼ ρl, ed ¼ el can be assumed; then, Eqs. (3.3.2.30) and (3.3.2.31) are coupled together to get the liquid field equation as follows:          ∂ 1 1 ∂ 1 ¼ α1 ρ1 e1 + V12 + αd ρ1 e1 + Vd2 α1 ρ1 e1 + P1 =ρ1 + V12 V1 ∂t ∂z 2  2 2  ∂ 1 2 P1 + αd ρ1 e1 + =ρ1 + Vd Vd ∂z 2    Xnk  P1 1 2 W1 α ρ e + + + V 1 1 1 1 m¼1 ρ1 2  m  Xnk  P1 1 2 + αd ρ1 e1 + + V Wd m¼1 ρ1 2 d m ¼ qi1  qig + qigd  qidg + qw1 + qt1 (3.3.2.32) The above two equations, Eqs. (3.3.2.29) and (3.3.2.32) are equivalent to Eqs. (2-35) and (2-36) of Ref. [5]. (b) The closure problem for the three-fluid model for the subchannel analysis code The basic equations of mass conservation for the vapor/gas mixture field, liquid field, and entrained droplet field are Eqs. (3.3.2.21), (3.3.2.22) and (3.3.2.23), respectively; the basic equations of momentum conservation for the vapor/gas mixture field, liquid field, and entrained droplet field are Eqs. (3.3.2.25), (3.3.2.26), and (3.3.2.27), respectively, and the basic equations of energy conservation for the vapor/gas mixture field and the liquid field are Eqs. (3.3.2.29) and (3.3.2.32), respectively. Therefore,

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each phasic equations have 15
3 ¼ 45 variables (average void fraction, αk, density, ρk, axial phasic velocity, Vk, lateral phasic velocity, Wk, partial phasic pressure, Pk, wall shear stress, Fwk, external body force, Fgk, internal energy, ek, wall surface heat flux, qwk, vapor generation rate from liquid to vapor, Г e, vapor condensation rate, Г c, entrained droplet from the liquid film to the droplet field, SE, deentrained droplet rate from the entrained droplet field to the liquid film, SD, vapor generation rate from the entrained droplet to the vapor field, Г ed, vapor condensation rate from the vapor field to the entrained droplet, Г cd, interfacial shear stress, Fik, interfacial heat flux, qik, interfacial velocity, Vik, interfacial pressure, Pik, and interfacial enthalpy, hik regarding the gas-liquid interface for each k-phase (k ¼ g, l, d)). Although the number of the variables is 40 in this whole equation set, using the preceding assumptions (ρd ¼ ρl, ed ¼ el), we reduced one droplet energy equation. That is, in order to close the system of equations and solve this equation set, we should prepare 5 jump conditions, 24 constitutive equations (interfacial drag model, wall drag model, interfacial heat transfer model, wall heat transfer model, and turbulent mixing model), and 3 equations of state for the additional 32 correlations because 8 basic equations of the three-fluid model have already been prepared as above. We will add below a brief explanation for the important correlations. The relationship for the void fractions of two-phase flow can be prepared as follows: αg + αl + αd ¼ 1

(3.3.2.33)

The jump conditions for mass conservation, momentum conservation, and energy conservation on the gas-liquid interface are as follows: Γe + Γc ¼ 0

(3.3.2.34)

Γ ed + Γ cd ¼ 0

(3.3.2.35)

SE + SD ¼ 0

(3.3.2.36)

Fig + Fil ¼ 0

(3.3.2.37)

Fidg + Figd ¼ 0

(3.3.2.38)

The density and the enthalpy in the basic equations are determined for the function of temperature and pressure with an equation of state, and the body force on each phase, Fgk, can be defined by an external function such as a gravity force. If the variables of one phase for Fwk, qwk, Vki, Pik, hik, Г k, Fgk, and qik are defined, the variables of the other phase are easily determined by using Eqs. (3.3.2.34)–(3.3.2.38). The variables Vki, Pik, and hik on the interface can be defined by the proper assumption of the thermal property; that is, for the variable Vki, the assumption that the phase change is not so big that the following relation is approximately established below. V gi ¼ V li ð¼ V i Þ

(3.3.2.39)

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Boiling Water Reactors

The interfacial velocity, Vi, is also related to the phasic velocities Vg and Vl based on each flow pattern condition and thermal condition. That is, for example, for the bubbly flow regime, vapor velocity in the bubble can be supposed to be uniform; so, Vi ¼ Vg

(3.3.2.40)

For the mist flow regime, the opposite condition can be assumed; therefore, Vi ¼ Vl

(3.3.2.41)

By the assumption that the influence of surface tension on the gas-liquid interface is not so large, the following relation can be obtained: Pgi ¼ Pli ð¼ Pi Þ

(3.3.2.42)

The pressure on the interface, Pi, is related to the vapor pressure, Pg, because the density of vapor is adequately smaller than the density of liquid; the contribution of kinetic energy to pressure is rather small. In this case, the following relation Pi ¼ Pg

(3.3.2.43)

is applied. In the case of the three-fluid model, the thermal equilibrium condition between the vapor phase and the liquid phase is not assumed; but, in the neighborhood of the interface, the equilibrium condition can be assumed. Therefore, the relation between the vapor temperature, Tgi, and the liquid temperature, Tli, can be supposed to be equal: T gi ¼ T li ð¼ T i Þ

(3.3.2.44)

and the interfacial temperature, Ti, equals the saturation temperature, Tsat of the interfacial pressure in the case of a one-component system. By using the proceeding formation, the variables on the vapor-liquid interface, Vki, Pki, and hki can be expressed by using other variables of the two-phase flow parameter. In the case of the two-fluid model, the shear stress on the wall with each phase, Fwg and Fwl, the wall surface heat flux for each phase, qwg and qwl, are needed to close the equations system; but, only whole shear stress, Fw, and whole heat flux, qw, on the wall surface can be obtained from the experimental data. So, the void fraction is used for partitioning to get each phase value, Fwg, Fwl, qwg, and qwl from the whole values, Fw and qw. As shown above, decreasing the number of unknown parameters can be decreased by introducing the correlations of constitutive equations to basic equations. Finally, we can set the system of equations of the three-fluid model and get the solution. The interfacial shear stress, Fig, is the characteristic for a two-fluid model and is very important to affect the behavior of the two-phase phenomena. Because of that the variable is caused by the difference of two phasic velocities; these are calculated by weighting the two-phase effect on the friction factor for the single bubble and single droplet. In order to close the equation system, which is described above, next four valuable constitutive models describe in detail the background of their development: – interfacial drag model (see pp. 104–115 of Ref. [5]) – wall drag model (see pp. 83–92 of Ref. [5]) – interfacial heat transfer model (see pp. 116–125 of Ref. [5]) – wall heat transfer model (see pp. 189–224 of Ref. [5])

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In the study of two-phase flow in the fuel assembly, the most important flow regime has been confirmed to be an annular/mist flow regime; so, the detailed flow pattern selection algorithm is prepared for the subchannel analysis code to specify the annular/mist flow regime (see pp. 62–75 of Ref. [5]). – turbulent mixing and void drift for a special model of subchannel analysis (See pp. 189–224 of Ref. [5]) For the subchannel analysis, predicting the behavior of both entrainment and deentrainment of droplets around the fuel rod is very important; so, special constitutive models are prepared as follows: – entrainment and de-entrainment model (see pp. 126–133 of Ref. [5]) – spacer grid model for a special model of subchannel analysis (see pp. 134–138 of Ref. [5]) The background of the development of those constitutive models will be described in detail.

References [1] D. Liles, et al., TRAC-PF1/MOD1: An Advanced Best-Estimate Computer Program for Pressurized Water Reactor Analysisy, NUREG/CR-3858, LA-10157-MS Jul, 1986. [2] TRACE V5.0 Theory Manual, Field Equations, Solution Methods, and Physical Models (V4.160). Division of Risk Assessment and Special Projects, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001, https:// www.nrc.gov/docs/ML1200/ML120060218.pdf (Accessed 16 September 2021). [3] M. Ishii, Thermo-fluid Dynamic Theory of Two-phase Flow, Eyrolles, Paris, Scientific and Medical Publication of France, NY, 1975. [4] M. Ishii, K. Mishima, Liquid transfer and entrainment correlation for droplet-annular flow, in: Proc. 7th Int. Heat Transfer Conf., Munich, paper TF20, Sept. 1982. [5] R.K. Salko, M.N. Avramova, COBRA-TF Subchannel Thermal-Hydraulics Code (CTF) Theory Manual Revision 0, CASL-U-2015-0054-000, 2015.

3.4

Advances in containment vessel design Seiichi Yokobori Tokyo City University, Setagaya, Tokyo, Japan

3.4.1

Thermal hydraulics of severe accidents

3.4.1.1 Introduction As an introductory remark, it would be better to define severe accident (SA) first of all. Postulating both multiple occurrences of accidents and/or abnormal events and multiple inconveniences of safety systems, serious core damage and resultant radioactive release become predominant. These phenomena are defined as severe accidents. When a nuclear fuel rod fails to be fully cooled due to the lack of a heat removal system and beyond DBA as a result, the fuel melts and the nuclear power plant is seriously damaged. SA is also characterized by these heat transfer mismatches around the

298

Boiling Water Reactors

fuel rod. Although the definition of the term “SA” includes a somewhat indecisive range, it can be well said that SA has almost the same meaning as core melt. In this chapter, SA thermal hydraulic behaviors will be studied and the technical measures that mitigate system damage will be summarized as accident management (AM) in the next chapter.

3.4.1.2 Initiation of fuel melt Fission heat generates inside the fuel rod, and the high heat flux is transferred from the pellet center to the cladding surface. Fig. 3.4.1 shows the fuel structure, which is composed of nest-of-boxes. The overall heat transfer resistances are composed of the following four resistances: (I) (II) (III) (IV)

heat conduction of pellets with inner heat generation heat convection inside the gap without gap flow heat conduction inside the cladding tube without inner heat generation forced convection on the tube surface by water (nuclear boiling heat transfer)

The overall heat transfer is completed by these mechanisms. By combining these four heat transfer resistances, the rod temperature distribution in the radial direction can be expressed by the following equation [1] (Eq. 3.4.1). In this equation, To is the pellet center temperature and Tb is the bulk temperature of water. T0  Tb ¼

 000 000 q Rf 2 q Rf 2 1 1 R 1 + + ln c + 4kf 2 R2 h c Rf ke hRc

(3.4.1)

where kf is the thermal conductivity of the pellet, h is the nucleate boiling heat transfer 000 coefficient, q is the heat generation volumetric density, Rf is the radius of the pellet, and ke is the thermal conductivity of the cladding tube, respectively.

㻲㼡㼑㼘㻌㼜㼑㼘㼘㼑㼠

㻳㼍㼜

㻯㼘㼍㼐㼐㼕㼚㼓㻌㼠㼡㼎㼑

T0

㻯㼛㼛㼘㼍㼚㼠 Tb

0 Rf R2

Rc

Fig. 3.4.1 Fuel cladding structure.

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Of these four heat transfer resistances, heat transfer resistances (I)–(III) are strictly determined irrespective of the situation. Only the fourth resistance (IV) depends on the boiling heat transfer condition. The outer diameter of a BWR fuel rod, including cladding tube thickness, is as small as 10 mm. The temperature difference between the pellet center and the tube surface (To  Tb) is as large as 1500 K. Such a steep temperature gradient is one of the largest of all industrial devices. In case of normal operation, the fuel is well cooled and the pellet center temperature To is maintained so as to avoid melting. However, when nucleate boiling is not activated due to some reason, heat transfer coefficient h asymptotes to 0 and thus the fourth heat resistance term of Eq. (3.4.1) increases to infinity. The pellet center temperature To increases accordingly and reaches its melting temperature. Actual pellet heating is considered to be accelerated by other factors such as the water-Zircaloy reaction and eutectic phenomena, which is mentioned in Section 3.4.1.4. Fuel melting is thus initiated and melting progresses rapidly. The molten fuel is called debris or corium, hereafter. The main component of debris is UO2 is quite heavy. Its specific gravity is about 10. Moreover, its viscosity is quite small like water and molten jet finally penetrates along the PCV floor in a short time. Although SA is thought to be accompanied by complicated motions, the simplest way to imagine is the molten liquid jet behavior along the RPV and PCV paths. Along the flow penetration path, many sedimentation, fragmentation, and solidification patterns can be frequently appeared. The Fukushima Daiichi nuclear accident in 2011 is considered to be well related to such a mechanism.

3.4.1.3 Progression of core melt The BWR fuel assembly (F/A) is composed of several metallic substances such as the channel box, tie-plate, cladding tube, and UO2 pellets. The substance of molten material is not single but mixed with many different melting temperature materials. The melting temperatures of each material are summarized in Table 3.4.1. When the ECCS water is not supplied from outside, the RPV water level becomes low and fuel assembly (F/A) melting initiates. All components of the F/A come to be Table 3.4.1 Melting temperature of F/A component.

F/A component

Material

Range of melting point temperature (K) (including inaccuracy)

① ② ③ ④ ⑤ ⑥

Stainless steel Zircaloy Zircaloy UO2 B4C B4C + Fe

1720 2030 2030 3120 2620 1420

Tie plate Channel box Cladding tube Pellet Control rod blade Eutectic crystal

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Cladding Tube Melt

UO2 Pellet Expose

Pellet Axial Center Melt

Fuel Collapse

Debris Sedimentaon

Fig. 3.4.2 An example of molten core collapse progress.

heated by radiation from the pellet, even if the material itself does not generate heat. The stainless steel of the channel box, whose melting temperature is the lowest of all materials, becomes molten first of all. The UO2 pellet still retains its original shape. Following that, Zircaloy melting starts and the pellet finally melts. In addition to these, when two metals with different melting temperatures are mixed, eutectic crystals are produced and the mixture’s melting temperature is generally lower than those of original metals. For example, for a eutectic crystal of B4C and Fe, the melting point temperature decreases by more than 1000 K. Considering these, F/A melting is considered to be accelerated. The actual melting propagation and the resultant molten material fall are considered to depend on the situation. Fig. 3.4.2 shows an example of molten core collapse progress. According to the melting temperature, Zircaloy of the channel box and cladding tube firstly melt. Following that, the center of the long pellet is molten, according to the axial peaking distribution. As the axial peaking generally takes its maximum at the center, the pellet is considered to melt from the center plane. Finally, all materials are melted to debris sedimentation.

3.4.1.4 Water-Zircaloy reaction accelerating fuel melt As for the important phenomena that will accelerate the fuel melt, it has to be emphasized that the steam and Zircaloy reaction can be vigorous at higher temperatures. When the RPV water level decreases without water makeup, the upper portion of the fuel (TAF) is exposed by steam. Although the rod surface is cooled by steam, the steam convection heat transfer rate is quite low and the chemical reaction between steam and Zircaloy of the fuel surface becomes predominant in the following equation: Zr + 2H2 O ! ZrO2 + 2H2 + 609 kJ=mol

(3.4.2)

From this reaction, it is easily understood that hydrogen is generated by Zircaloy oxidation and the rod surface temperature increases due to heat generation. This chemical

Nuclear reactor dynamics and thermal hydraulics of reactor core

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reaction is accelerated at the temperature of 900°C. This heat up accelerates the reaction with the Arrhenius-type expression. W 2 ¼ K 0 exp ðQ=RT Þt

(3.4.3)

where W is the reaction rate, T is the temperature, and t is the time. Zircaloy is not only a component material for the cladding but also a component for the channel box. The volumetric occupation of Zircaloy on the fuel assembly is quite large.

3.4.1.5 Melting relocation inside the RPV From the information given above, Fig. 3.4.3 summarizes the rough sketch of debris progress inside the RPV. From its birth at the core to the RPV bottom rupture, progress is divided in four phases. The main milestones are summarized in the figure. (a) Blowdown Even if cooling water is terminated, water is still maintained in the RPV for a short while; but, the remaining water is evaporated by rod heat generation. Moreover, much vapor is generated by depressurization from two-phase flow. Steam strongly blows inside the RPV gas space. Hydrogen generation is initiated from uncovered surface steam. (b) Core melt After the blowdown phase is terminated, hydrogen starts to generate from the uncovered surface. All F/A materials are molten and debris is formed. Its mechanism is described in Section 3.4.1.2.

BLOWDOWN

CORE MELT

SUPPORT PLATE BREAK

Fig. 3.4.3 Four phases of molten core progress inside the RPV.

RPV BOTTOM LEAK

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(c) Break of the support plate Debris jet falls down and is accumulated on the support plate. Then, the support plate is broken by the debris heat. Though the molten debris falls into the bottom pool, vigorous explosion is not considered to occur even though some amount of water remains in the RPV bottom. (d) RPV bottom leak Water in the RPV bottom pool evaporates completely and most of the debris melts through the RPV bottom. Concrete debris falling will start by leakage from the tiny hole of the control rod (CR) guide tube.

3.4.1.6 Melting jet structure and behaviors (from the RPV bottom to the PCV floor) To understand the debris scattering motion, it is useful to clarify single jet behavior phenomenologically. Fig. 3.4.4 shows the molten jet falling motion with Reynolds number [2]. The metal was Wood’s metal, whose melting temperature is small as 350 K. The jet starts to issue from the nozzle (D) by gravitational force. The jet falls down and breaks and impinges on the floor intermittently. This jet break was noticed due to instability between jet shear force and stagnant air. After the impingement of the floor, wall jet penetrates radially along the floor. During radial movement, the jet loses its enthalpy and started to solidify. When the jet front progression is stopped by solidification, the following jet cannot penetrate more downward and the melting jet stays inside the pond. These motions can be roughly classified into the following six patterns: (I) (II) (III) (IV) (V) (VI)

Jet issuing from the nozzle. Free jet starting to break by instability. Intermittent impingement. Wall jet starting to penetrate. Jet penetration terminating due to solidification and a circular pond forming on the floor. Continuing jet staying inside the pond.

Fig. 3.4.4 Typical flow pattern of the falling molten jet (Re ¼ 16,000 to 2000) [2].

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The parameters determining these motions are listed in the following: D (jet nozzle diameter), H (height), V (Initial velocity), t (initial jet temperature), and ν (viscosity). The actual motion includes some discrepancy.

3.4.1.7 FP aerosol behaviors [3,4] From the thermal-hydraulics point of view, the effect of SA on nuclear safety can be considered in two aspects. One aspect is maintaining vessel integrity by debris jet motion (Section 3.4.1.6) and the other is to mitigate the radioactive influence by FP aerosol. As far as vessel failure threatening is concerned, both thermal and mechanic behaviors of the molten core are significant concerns. However, SA is not only characterized by the molten jet but also by FP aerosol motion in the air space. Most volatile FP aerosol is released to the RPV. Although volatile fission products (FPs) of molten fuel include many species, representative species are limited. Specifically, the main species of FP regarding SA are CsI and CsOH. It has been confirmed that these two species were typically detected by the PHEBUS-FP Project [3]. In this project, an actual modified fuel rod was forced to melt and FPs were detected. As the Cs fraction of the core is about 80%, the FP behavior is closely connected with the environmental dose. As another approach, NRC reported NUREG-1465 [5], which is revised based on TID-14844 based on BNL’s source term. The report concluded describing important volatile FPs such as iodine, cesium, and tellurium. The jet behaviors are visualized by directly watching the jet motion but the diameter of aerosols was as small as 10 μm and the FP motion is mainly affected by the convected steam flow. FP-accompanied space is roughly classified into two regions. One space is the RPV and PCV dome; the other is the suppression pool through the vent line. When FPs are transported to the RPV, convected gas and steam press the RPV pressure. When the FPs and gas are transported to the suppression pool, the three phase flows with FPs (solid particle), steam (gas), and pool water (liquid) are vigorously mixed. This scrubbing motion plays an important role in nuclear safety (Fig. 3.4.5). This is reviewed in Section 3.4.2.6. To drywell

To RPV To wetwell

Fig. 3.4.5 Schematic representation of FP transport to RPV, drywell, and wetwell.

304

3.4.2

Boiling Water Reactors

Accident management for BWR

3.4.2.1 Summary of AM Under normal operational conditions, water can remove fission heat by nucleate boiling and the peak cladding temperature (PCT) is maintained below 1200°C. Accordingly, the pellet peak temperature is well below its melting temperature. When the heat removal system fails, a control rod is inserted and fission is stopped. Nevertheless, decay heat is still generated from the pellet and the rod peak temperature exceeds its melting temperature. Thus, fuel starts to melt gradually and molten fuel lumps (called debris hereafter) are accumulated. Debris falls from the tiny holes of the RPV vessel bottom. This molten jet continues to fall down through the RPV bottom and finally penetrates to the PCV floor. Even in that situation, technical control measures to maintain the debris inside the PCV wall must be taken. This technical measure is called accident management (AM). The fatal situation without AM was shown in the previous figure (Fig. 3.4.3). Plant damage is strongly dependent on the timely adoption of the AM procedure. As the AM covers a wide space from RPV to PCV, it is classified into two phases. The earlier AM, which is effective in mitigating core damage frequency (CDF), is called Phase I AM, while the AM, which is applied to mitigate the conditional containment failure probability (CCFP), is called the phase II AM. In Japan, based on the persuasion of Nuclear Safety Commission (NSC), in 1992 the MITI (Ministry of Trade and Industry) requested electric power companies to prepare AMs for their own self-improvement. As response to this request, all Japanese utilities presented their AM plans unique to the plant in 1994. Furthermore, the electric companies sent the PSA report and confirmed the effect in 2002. After that, the Fukushima Daiichi Nuclear Accident occurred in 2011. Reflecting on the disaster, the Nuclear Regulation Authority (NRA) decided on a new regulatory standard (Table 3.4.2). In the new standard, the following items are included under obligation. The timing of the plant reoperation depends on the preparatory schedule.

Table 3.4.2 New regulatory standard.

1 2 3 4 5 6 7 8 9

Item

New/reinforce

Filtered vent Preparation of second control room Station in case of emergency Hazard Airplane against terrorism Protect of PCV rupture Redundancy of power source Redundancy of cooling device Dislocation earthquake

Newly Newly Newly Reinforce Reinforce Reinforce Reinforce Reinforce Reinforce

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3.4.2.2 Defense in depth Since the early stage of the nuclear power plant constructed in the 1970s, all nuclear personnel in Japan including the regulatory section, electric power companies, and utilities have paid strong attention to plant safety and promoted the “defense in depth” concept as the top priority of the safety policy. Once a safety system fails, a preparatory back system compensates. The safety research and design goals have been consistently focused on to achieve this concept technically. However, the nuclear safety culture varies with time and “the defense in depth” concept has been reflected on and modified. After the Fukushima Daiichi Nuclear Accident in 2011, this policy contains five structures as listed in Table 3.4.3 [6,7]. From the first to the second structure, prevention action is strongly encouraged. At the first structure, the normal operational state is maintained and early detection of abnormal conditions is emphasized. Through appropriate countermeasures, normal operation can be easily recovered. In this structure, many safety designs are deployed. In these items, safety culture, quality management (QM) system, redundancy design, diversity design, high education, and continuous training are included. As for the second structure, prevention of abnormal progress and measures to stop the accident state are required. Until the second structure, the fuel itself is not damaged. The plant can be recovered to the normal operational condition. The third structure is important. A small scale of accident is premised in this stage. So, mitigation is the important criterion. Even if serious damage is effected to the plant, core damage is strongly avoided. Thus, early detection of plant status and reactor shutdown are the main measures of this stage. Application of ECCS is a typical measure. Both the fourth and fifth structures are premised to the core damage. In the fourth structure, the early detection of core damage and mitigating its effect is the main consideration. Finally, as the fifth structure proceeds, much radioactive dose are released to the PCV outside and its effect on the people outside the PCV should be minimized. Formerly, the “defense in depth” concept was composed of the former three structures (e.g., from the first to the third structure), but it has been extended to five structures reflecting on the Fukushima accident. Considering SA, the fourth and fifth structures were reinforced up to the present. The defense in depth concept is summarized in Table 3.4.3.

Table 3.4.3 BWR Defense in depth [6,7]. Level

Countermeasure

Avoidance

1 2 3 4 5

Maintaining of normal condition Prevention Mitigation (ECCS) SA measure (AM) Outside measure

Abnormality Abnormality (transient) progress, accident Accident progress, core melt FP release, minimize dose Minimize public dose

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3.4.2.3 International event scale (INES) The accident/event magnitude of the nuclear power plant is internationally determined by the IAEA. This table is called INES (International Nuclear and Radiological Event Scale) and summarized in Table 3.4.4. and Fig. 3.4.6. In case of abnormality detection, the electrical power company has to report the accident magnitude to the ministry based on the international event scale. This scale classifies three standards. Standard I: Effect of the fission product on the plant outside Standard II: Effect on inside workers Standard III: Degradation due to damage

Table 3.4.4 Accidents in the INES table. Level

Standard I

7

Major accident

6 5

Serious accident Accident with off-site risk Accident without significant off-site risk Serious incident Incident Anomaly No safety significance

4 3 2 1 0

Accident

Incident

Deviaon

Standard II

Standard III

Chernobyl (1986), Fukushima Daiichi (2011) TMI-2 (1979) JCO (1999)

Mihama (1991) Monju (FBR)

䠓

• Level 7: Major Accident

䠒

• Level 6: Serious Accident

䠑

• Level 5: Accident with Off-Site Risk

䠐

• Level 4: Accident without Off-Site Risk

䠏

• Level 3: Serious Incident

䠎

• Level 2: Incident

䠍

• Level 1: Anomaly

䠌

• Level 0: No safety significance

Fig. 3.4.6 International Nuclear and Radiological Event Scale (INES).

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Moreover, within each standard, levels are classified into seven categories according to their seriousness. Level Level Level Level Level Level Level

1: 2: 3: 4: 5: 6: 7:

Anomaly Incident Serious incident Accident without off-site risk Accident with off-site risk Serious accident Major accident

Standard 1 is classified into Levels 3–7, Standard II is classified into Levels 2–5, and Standard III is classified into Levels 0–3. This level is determined according to the dose released outside. As the SA is characterized by core damage, SA is closely related to the Level between 5 and 7. From this table, within each standard, the Three Mile Island (TMI) Unit-2 accident (1979) is set to be Level 5, while the Chernobyl Unit-4 accident (1986) is at Level 7. Though the TMI accident induced small LOCA, the released dose was not so serious and Level 5 was judged as the appropriate level. The Fukushima Daiichi Nuclear Accident is tentatively set to Level 7. This is because the released dose (3.7–6.3
1017 Bq) is thought to exceed the Level 7 dose (1016 Bq).

3.4.2.4 Selection of BWR AM measures The AM measures should be finally determined considering the following conditions: (1) to utilize existing hardware exhausting the design margin (2) to select effective AMs based on PSA phase I AMs to mitigate CDF phase II AMs to mitigate CCFP (3) to prepare the manual/documents of the AM guideline (4) to prepare recovery documents for important systems such as RHR and DG (5) to enable operators to act easily and surely (6) to maintain the safety inherent in normal operational conditions

The Japanese AM measures should be finally determined considering the conditions unique to Japan. Nevertheless, most of the AM procedures are considered to be universal. The US BWR Group has modified the Mark I plant. Modified procedures are listed in Table 3.4.5. The AMs have been frequently modified.

3.4.2.5 Typical BWR core damage sequence In order to determine the efficiency of the BWR AM, the appropriate BWR accident sequential events should be postulated by the utilities based on the PRA results of the individual plants. Then, the plant’s unique AM measures should be finally determined by the Japanese utilities considering the possible sequential scenarios. Representative BWR core damage sequences are listed in Table 3.4.6. Each sequence is complicated and abbreviations are commonly used. If mitigation of AM is not taken up, all accidents result in both RPV and PCV failure.

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Table 3.4.5 BWR AM modified in US.

1 2 3 4 5 6

Major modifications and upgrades to USBWRs with Mark I containment systems

Date

Added spare diesel generator and portable water pump Added containment vent More batteries in event of station blackout Strengthened torus Control room reconfiguration Back-up safety systems separated

2002 1992 1988 1980 1980 1979

Table 3.4.6 Typical BWR core damage sequence. Sequence

Concrete event progression

After core damage

TQUX

Transient(T) followed by loss of feedwater(Q) ! HP injection failure(U) ! depressurization failure(X) ! core damage Transient(T) ! HP injection failure(U) ! depressurization success ! LP injection failure(V) ! core damage LOCA(A) ! HP/LP injection failure(E) ! core damage Loss of AC power(B) ! RCIC trial ! DC power exhausted ! loss of injection

RPV failure ! PCV failure

TQUV

AE TB (Station blackout) TW

TC (ATWS)

Transient(T) ! injection success ! residual heat removal failure(W) ! PCV break by overpressure ! core damage Transient/LOCA(T) ! scram failure(C) ! PCV break by overpressure ! core damage

RPV failure ! PCV failure RPV failure ! PCV failure RPV failure ! PCV failure RPV failure

RPV failure

Of these scenarios, TQUX is most typical SA scenario. Core damage is faster. In the sequence of TW and TC, PCV breaks faster than RPV failure.

3.4.2.6 In-vessel phenomena (from core melt to RPV bottom leak) Several phenomenological events from core melt initiation to debris penetration to the RPV bottom are summarized in the following. Melt jet behaviors are described for the ex-vessel in the next chapter. (a) Fuel melt The fuel melting mechanism was explained in Section 3.4.1.5. (b) Water-Zr reaction and hydrogen generation This reaction accelerates fuel melting. This was explained in Section 3.4.1.4. (c) Melt acceleration by eutectic crystals Eutectic crystals are also produced and the metal melting temperature goes down compared to that of single metal.

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(d) Possibility of recriticality inside the RPV (FCI) In the BWR plant, the possibility of criticality has two conditions: (1) Geometrical clearance of the fuel assembly is maintained. (2) Water is filled around the fuel rod. Even when the channel box melts, if the fuel assembly still keeps its original structure and water remains at the RPV bottom, it is possible that recritical conditions are satisfied. Nevertheless, supplying boron water by the SLC can avoid recriticality [8]. (e) Possibility of steam explosion in the RPV When the fuel and coolant interaction (FCI) is predominant inside the RPV, vigorous vaporization will result in explosion. In the BWR lower plenum, however, many structures such as the control rod/blade and instrumentation tube are closely bristled and the molten jet does not smoothly fall touching the structure and FCI is hardly thought to occur. Even if small-scale vaporization happens, the mechanistic energy (α-mode failure), which forced the RPV roof open, will not be accumulated. This assumption has been confirmed by a specialists’ meeting (SERG-2) [9]. (f) FP transported to the RPV wall/space Most of the debris falls down by jet motion but the remaining debris is volatilized to gaseous fluid and transported upwardly to the RPV space. The fission product (FP) is a tiny aerosol-like powder as small as 10 μm. It is possible that FPs are delivered with steam flow. Some of them attach to the RPV wall, while most FPs are released to the environment through the tiny leakage path. Most of the volatile FPs (e.g., CsI or CsOH) are transported to the suppression pool (SP) through the vent line. (g) FP transported to the Suppression pool (scrubbing) Volatilized FP aerosol is transported not only to the RPV but also to the suppression pool (SP) through the vent line. Most of the FP is captured by the pool water. Fig. 3.4.7 shows the schematic representation of the scrubbing flow structure. FP contained in the bubble touches the water and is captured while rising. The capture efficiency is estimated by the decontamination factor (DF). The DF for the Mark I type PCV is more than 10,000 and is quite effective. After RPV failure, when the FP aerosol is transported to the SP through the vent line, the overall DF is expected to be more than 100. Therefore, wetwell ventilation is strongly recommended for FP removal.

Liquid surface

Rising bubbles Gravity Vent line

Fig. 3.4.7 BWR Scrubbing [10].

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The FP scrubbing effect is considered quite important to nuclear safety. Many related tests have been conducted internationally. Many projects such as ACE, LACE (EPRI), GE, EPSI (JAERI), and BWR Utility have shown both useful experimental and analytical results. Through small- and medium-size scale tests, the Japanese BWR utilities have conducted scrubbing effect tests with large-scale test facilities (10 m as height and 1.5 m as diameter). It has been demonstrated that DF in case of SP depressurization is almost constant, irrespective of the vapor flow due to bulk boiling [10]. (h) IVR learned from TMI (PWR) [11,12] In 1979, in the PWR plant of Three Mile Island Unit 2, a double pump trip occurred and resulted in the shutdown of the feedwater system. Several errors were continuously followed. Due to misunderstanding of the two-phase water level and wrong operation by the operator, small LOCA occurred. Although the upper half of the core was failed, the bottom-half fuel maintained its original shape. Molten fuels remained inside the RPV. It was assumed that the RPV vessel bottom was somewhat cooled by the water supply [11]. This assumption was named IVR (In Vessel debris Retention) and many IVR studies were developed. Boiling heat transfer characteristics along the downward semispherical surface were investigated. Although this phenomenon was unique to the PWR plant unique, this accident strongly affected the BWR AM.

3.4.2.7 Ex-vessel phenomena after RPV failure After RPV bottom failure, molten jet falls from the RPV bottom and is delivered to the PCV wall. Several important phenomena postulated inside the PCV during SA are closely connected to threatening the PCV wall attack. The avoidance of such PCV failure by mitigating these phenomena is the ex-vessel AM. Major SA phenomena are shown in Fig. 3.4.8. (a) Fuel coolant interaction (FCI) [13,14] The possibility of steam explosion is discussed. It is generally pointed out that steam explosion occurs by the following four mechanisms (Fig. 3.4.9).

DCH 䠢䠟䠥 (Vapor Explosion)

Shell Aack Fig. 3.4.8 Ex-vessel phenomena of SA.

MCCI Aack

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Fig. 3.4.9 Vapor explosion mechanism [13,14].

(1) High-temperature molten lumps falling into the water pool. (2) After falling, stable film boiling is initiated with fragmentation into tiny metallic particles. Accordingly, the surface area of the molten metal is increased. (3) The stable vapor film collapses and triggers the pressure wave into the pool. (4) Due to pressure wave propagation, molten substance segmentation is accelerated and much steam is produced simultaneously and explosion is induced resultantly. (b) Possibility of steam explosion outside the RPV [13] For the above mechanism to proceed smoothly, much water volume has to be prepared in the pool. For this, pool water volume should be deeper and the subcooling must be larger. However, for the actual BWR RPV bottom, many structures are hanging. Both control rod device (CRD) and instrumentation tubes are closely bristled. From these situations, conditions to induce explosion are not so often satisfied. Up to the present, by many studies, it has been concluded that large-scale steam explosion hardly occurs with pressure and subcooling. This assumption is supported by the COTES experiments performed by NUPEC [15]. As for the FCI-related tests, since the effect of steam explosion on the public risk was pointed out by the WASH-1400 report in 1975, many large-scale tests were conducted. For example, FARO (Europe, Ispra), KROTOS (JRC Ispra), ALPHA (JAERI), and COTELS (NUPEC) are typical tests. (c) MCCI (molten core and concrete interaction) MCCI is a typical postulated phenomenon for AM. When the molten core touches the PCV floor made of concrete, the heated concrete is eroded. The floor is possible to be penetrated. Moreover, the reaction between gas and metal becomes predominant and much noncondensable gas such as CO2 is produced. The PCV pressure increases resultantly. Although the PCV risk seems to become higher, debris jet may fall intermittently to the pedestal and a large-scale pond will not be made. A large magnitude of MCCI is hardly considered to occur. As for the AM procedure, prefilling water is a possible measure.

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(d) DCH (direct containment heating) [16] In case of RPV break under high-pressure condition, molten fuel fragmentation is scattered into the PCV. The PCV gas space is heated directly by this tiny powder-like fragmentation and space pressure becomes higher. In addition to this, generated hydrogen gas accelerates the PCV pressure. These are important reasons for the PCV’s early rupture. Once ruptured, much molted fuel sedimented on the floor is lifted up by the gas flow and the entire space is contaminated. To avoid this, the PCV pressure is reduced to 2 MPa. This phenomenon is possible for both BWR and PWR. (e) Shell attack (PCV direct contact) When molten materials fall to the PCV floor and penetrate to the stainless PCV wall, it is possible that debris directly contacts the wall and melts resultantly. Under the dry conditions, the PCV wall surely fails. This situation is especially possible in case of the Mark-I type PCV. The best way to avoid PCV failure is to prevent debris from touching the wall, and thus water is initially filled to the pedestal. From the US research results, it is reported that PCV failure possibility by water filling is quite small.

3.4.2.8 AMs for existing BWR As the existing BWRs were already constructed, innovative AMs need hardly to be prepared. In spite of such difficulties, utilities have prepared the following AM plans. Some of these are included in the measures in Table 3.4.5. (a) Reinforcement of reactor shutdown This is a Phase-I AM that supports the inadequacy of the control rod motion. This is considered as an ATWS measure. (b) Reinforcement of supplying water to the RPV/PCV Reactor core degradation is planned to be prevented by water makeup to RPV. Further, even in case of core damage, the degraded core has to be cooled. For this, the vessel depressurization system is designed to be automatically activated. Then, a low-pressure core cooling system can be available. Moreover, debris is cooled by supplying water to the PCV pedestal. The PCV is also cooled by PCV spray. To achieve this, normal systems such as the MUWC pool are utilized as auxiliary makeups. (c) Reinforcement of the heat removal system from the RPV/PCV By utilizing a normal system such as the PCV cooler, an alternate cooling system is activated. This is classified as Phase II AM. To avoid PCV failure, noncondensable gas and/or steam confined inside the PCV is forced to be vented. (d) Reinforcement of support of safety action To reinforce the instrumentation system in case of SA station blackout, the power system is accommodated with a neighboring plant. The emergency diesel generator is planned to be located at a higher elevation considering recovery.

3.4.2.9 AMs for the recently operated and planned plants (also with PWR) For the recently operating plants, innovative AMs can be prepared. In most of the plants, both debris cooling and containment cooling are prepared; these are listed in Table 3.4.7. As for the debris cooling device, a core catcher is mainly designed,

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Table 3.4.7 AMs for next-generation reactor. Debris cooling

Containment cooling Steel PCV external cooling Active cooling Vertical type PCCS Vertical type PCCS

1

AP1000

IVR

2 3 4

EPR VVR-1000 ESBWR

5 6

SWR1000 EU-ABWR

Debris spreading type core catcher Core catcher Pipe jacket type core catcher (BiMAC) IVR Dish type core catcher

Inclined PCCS Horizontal type PCCS

while a passive containment cooling system (PCCS) is prepared as the containment cooling. (a) AP1000 (Westinghouse, PWR) The containment vessel of the recently operating PWR is designed to be made of steel. In case of accidents, atmospheric water/air can be supplied from outside [19]. Due to the higher heat conductivity of steel, it is expected that the CV (containment vessel) pressure can be kept lower. Four plants of AP1000 were operated in China in 2018. As an ultimate heat sink, seawater is not utilized but atmospheric air is. (b) EPR (AREVA) [20] As the debris cooling device, a spreading compartment is prepared so that molten debris is cooled while spreading. The compartment area is as wide as 170 m2. Fig. 3.4.10 shows the compartment. (c) ESBWR (GE/Hitachi) [21,22] General Electric and Hitachi Nuclear Energy (GEH) extended the SBWR and ABWR design and developed the ESBWR. ESBWR stands for economic BWR. It employs passive safety design features. It is a simplified reactor design, allowing faster construction and lower operating costs. Passive design features such as passive containment cooling reduce the number of active systems and increase safety.

Pressure vessel Spreading compartment

Fig. 3.4.10 Core catcher of EPR [20].

314

Boiling Water Reactors

As the debris cooling device, a core catcher is prepared to remove heat by a parallel pipe jacket passively (BiMAC). The decay heat is removed by vertical type PCCS. (d) SWR1000 (Siemens) [22] This plant has hybrid safety systems. The safety concept for accident control is based on interaction between active systems and passive safety equipment. The passive systems include the hydraulic scram system for rapid control rod insertion and so on. As for containment cooling, condensers for containment heat removal are employed. (e) EU-ABWR (Toshiba, Westinghouse) [23] The European Advanced Boiling Water Reactor (EU-ABWR) was developed by Toshiba and Westinghouse. As the debris cools, the dish-type core catcher is selected and PCV cooling is used by a horizontal-type PCCS, which is located outside the PCV.

References [1] [2] [3] [4] [5] [6] [7] [8] [9]

[10] [11] [12] [13] [14] [15] [16] [17] [18] [19] [20] [21] [22] [23]

M. Akiyama, Nuclear Thermal Engineering (in Japanese), 1978, pp. 62–63. Private information by courtesy of T. Suzuki (2021). Review of FP Behavior of Phebus Project (in Japanese), AESJ (2017). AESJ, Fission Product Behavior under Severe Accident, Research Committee on Fission Product Behavior under Severe Accident, 2021. L. Soffer, et al., Accident Source Terms for Light-Water Nuclear Power Plants, NUREG1465, 1995. INSAG-10, IAEA (1996). INSAG-12, IAEA (1999). Frid, et al., Severe accident recriticality analysis (SARA), Nucl. Eng. Des. 209 (2001) 97–106. S. Basu, et al., NUREG-1524 (1996) or USNRC, A reassessment of the Potential for an Alpha Mode Containment Failure and Review of the Current Understanding of Broader Fuel-Coolant Interaction Issue, NUREG-1524, 1996. M. Akiba, Aerosol Decontamination Effect Test, in: 10th NRA Research report meeting (in Japanese), 1917. J. Wolf, et al., TMI-2 vessel investigation project integration report, NUREG/CR-6197, 1994. T.G. Theofanous, et al., In-Vessel Coolability and Retention of A Core Melt, OECD/ID10460, 1996. AESJ Technical Standard. T. Takashima, Y. Iida, Science of Steam Explosion (in Japanese), Shokabo, 1998. M. Katoh, H. Nagasaka, COTELS fuel coolant interaction tests under ex-vessel conditions, in: JAERI-Conference 2000-015, 2000. M.M. Pilch, et al., NUREG-6075, 1994. T.G. Theofanous, et al., NUREG/CR-6025, 1993. http://www.osti.gov/scitech/biblio/10107249. C.P. Keegan, et al., ICAPP-7489, 2007. M. Fischer, ICAPP-5001, 2005. GE Hitachi Nuclear Energy, ESBWR. Passive Safety Systems and Natural Circulation in Water Cooled Nuclear Power Plants, IAEA-TECDOC-1624 (2009). K. Kamei, et al., Conceptual Design of European Advanced Boiling Water Reactor (EU-ABWR), ICONE18, 2010.

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3.5

315

Advances in safety analysis code and safety systems Seiichi Yokobori Tokyo City University, Setagaya, Tokyo, Japan

3.5.1

Various BWR analysis codes

3.5.1.1 Importance of nuclear analysis codes In order to keep nuclear safety, it has been confirmed that the core should be fully cooled with enough margin. Nuclear power plants are large systems and it is impossible to confirm the normal/transient/accident phenomena by experiment only. Thus, it is required that design adequacy is checked by a safety analysis code. The contribution of the safety analysis code to the nuclear power plant system design is quite important. Particularly, in the nuclear system, many parameters are included. Each parameter is conjugated to the other. So, analysis of mutual feedback of parameters is not ignored. Further system response performance cannot be clarified by tests with varying parameters. The analysis is useful to check parameter sensitivity. The analysis has also the advantage that the unmeasured information can be obtained. Of many nuclear simulations, safety analysis is important above all. In the safety analysis, for example, the long-term water level transient has to be predicted. As the circulating flow is not single phase but two-phase flow, the relation between vapor and gas is incorporated as correlation. In order to get higher accuracy data, the analysis code has to be well qualified. Safety analysis codes are generally classified into two codes. One code is the best estimate (BE) code with high accuracy. The analysis object is not very wide but is somewhat limited. Instead, the phenomenon is precisely predicted. On the contrary, the other is design-related code and is called the evaluation model (EM) code. Longterm system responses of parameters such as pressure/temperature/water level are calculated with time. The results are expected to simulate the overall tendencies of the parameter. Apart from these two codes, by utilizing computational fluid dynamics (CFD), numerical activities are also focused on. This approach is mainly used for monophase flow prediction (Table 3.5.1).

3.5.1.2 Best estimate code and evaluation model code Safety analysis is classified into two codes (e.g., a best estimate code and evaluation model code). Typical output images are different. Using the best estimate code, important physical phenomena can be intensively analyzed. This analysis is also called the separate effect analysis. The computational area is drawn by a fine mesh and initial/boundary conditions are given accurately. As a

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Table 3.5.1 Classification of safety analysis code. Code

Best estimate (BE) code

Evaluation model (EM) code

Phenomena description Representative code

Reappearance with higher accuracy TRAC, RELAP

Overall tendency trace with conservative margin LAMB, SCAT, SAFER

Severe accident code is not described.

typical example of this separate effect test, a saturated CCFL test is introduced. CCFL is an important phenomenon for safety. Suppose upward vapor flow and liquid falls in the narrow flow area; water cannot enter against the high steam flow. The prediction of water level is important for safety. Tests were conducted to obtain the drainage value in every narrow flow area. The ESTA saturated CCFL tests [1] were analyzed to verify the CCFL model of TRAC-BD1. Comparison of the saturated CCFL characteristics is shown in Fig. 3.5.1. Using the node of ESTA, this data was analyzed by the TRAC-BD1. Good agreement between the measured and the predicted CCFL characteristics indicate that TRAC-BD1 can calculate multibundle CCFL phenomena at the tie-plate. The EM code is also called the system response analysis code. This is because a long-term system response is predicted using this code. For example, the peak vessel pressure or temperature can be clarified. The key components are roughly divided by nodes and energy balance is analyzed. Fig. 3.5.2 shows the comparison of heater surface temperatures at the middle position in all the break area spectrum tests [2]. Both SAFER03 analysis and the experiments showed that the peak cladding temperatures (PCTs) occurred in high-power bundles in all break tests. The dryout and quenching behaviors were closely dependent upon a vapor-water two-phase mixture level in the

Fig. 3.5.1 Example of calculated output (TRAC-BD1 code for ESTA) [1].

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Fig. 3.5.2 Example of calculated output (rod surface temperature at PCT). Comparison between test and SAFER03 code [2].

bundle. The two-phase mixture level was predicted by SAFER03, due to the incorporation of the CCFL model at several locations, as shown in Fig. 3.5.2.

3.5.1.3 Verification and validation (V&V) of simulation [3,4] Recently, in most industrial fields, reinforcement of quality management (QM) has been one of the most important issues. The enhancement of quality maintenance of computer-aided-engineering (CAE) has to be fully verified by the confirmation test data. This is the fundamental standard and is called “modeling V&V.” Both modeling and simulation were mainly developed by the ASME V&V [3]. The modeling V&V concerns simulation technology. According to the ASME V&V30, this standard is decisively structured and has four elements and uncertainty analysis. Element 1: Development of model Element 2: Numerical modeling Element 3: Physical modeling Element 4: Prediction judgment of simulation model Final: Actual prediction with uncertainty

However, the recent code QM is required not only for modeling V&V but also for another V&V. This latter V&V is “maintenance V&V” or “product V&V” [4]. This V&V may be somewhat difficult to understand. It also has to be checked how the code is rightly developed according to the customer requirement. Even if the code can predict the phenomena perfectly, if the code is not appropriate for the prediction, it is not suitable for the customer’s requirement. The product V&V takes the process seriously. Prior to the actual analysis, some standard analysis is cross checked with the theoretical results or experimental data.

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3.5.1.4 BWR analysis code (EM code) [5] With the combination of reactor components and small scale analysis, conservative analysis can be done. Fig. 3.5.3 shows the example of BWR RPV noding. The reactor core component is modeled with 7 nodings. On the other hand, lower plenum, guide tube, bypass region, upper plenum, subcooled region, saturated region, and steam dome are modeled by one node. As for water-steam two-phase flow, a homogeneous model is incorporated. Although the old architecture seems to be still used, proper conservatism is included and this has been continued up to the present. (a) SAFER03 [6,7] SAFER03 was developed by implementing the accurate thermohydraulic models with the previous computer code (SAFE and REFLOOD). SAFER03 is a computer code used to calculate long-term reactor vessel inventory and PCT for a BWR LOCA. Toshiba mainly developed this code and had carried out various experimental studies to get the new correlations for the core cooling characteristics and CCFL phenomena. These new correlations were incorporated into SAFER03. As typical SAFER03 noding, the pressure vessel is divided into 8–13 major regions. Other parts are distributed one node: lower plenum, guide tubes, core, core bypass, upper plenum, the initially subcooled region outside the core shroud, and the dome. In order to calculate the water level, the CCFL characteristics of major parts have to be clarified. The CCFL as well as subcooled CCFL breakdown phenomena are considered at all flow limitations between the major regions. Obtained data sets were correlated by the modified Wallis type correlation (e.g., (jf)1/2 vs (jg)1/2). It was confirmed by CCFL tests that correlations predicted the reasonable liquid downflow, as compared with the test data. This modified Wallis-type CCFL correlation is applied to SAFER03 to calculate the allowable liquid downflow. SAFER03 utilizes several correlations and physical models, based on joint studies between the utility and the maker, for development. Using this facility, the following correlations were obtained.

Steam Dome Upper Plenum

Saturated Region

Feedwater

Subcooled Region Bypass Guide Tube

Lower Plenum Fig. 3.5.3 Example of EM code noding (BWR RPV) [5].

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UTP CCFL and CCFL breakdown Core inlet orifice CCFL Bypass top CCFL Multichannel effect Bypass bottom CCFL Single-phase steam cooling heat transfer Mist flow heat transfer Spray heat transfer Based on these findings and other aids, SAFER code, which can better estimate the thermal-hydraulics of the BWR transient by a simple one-dimensional method, has been developed by BWR vendors to replace the existing SAFE-REFLOOD-CHASTE code system. As these important models had not been incorporated into the previous code, the main model modification of SAFER03 is summarized in Table 3.5.2 [7]. The new code predicts the peak-cladding-temperature for the standard jet-pump BWR LOCA transients to be as high as 500–600°C, which is well below the current regulatory limitation. LOCA/ECCS constraints will be reduced for future BWR management, ECCS equipment surveillance, etc., by the use of new code. (1) (2) (3) (4) (5) (6) (7) (8)

Table 3.5.2 Newly incorporated SAFER model [7]. Model item 1 2 3 4 5 CCFL

Heat transfer

Others

1 2 3 4 5 1 2 3 4 5 6 7 8 1 2

SAFER03 model RPV noding divison Core axial noding division Heat slub Heat up calculation rod number Conservation law of heat transfer calculation Upper tie plate(UTP) CCFL Bypass top Side entry orifice Bypass bottom CCFL CCFL breakdown Spray heat transfer Steam cooling heat transfer Mist flow heat transfer Transient boiling heat transfer Film boiling heat transfer Spacer effect Radiation heat transfer Wetting of channel box and fuel rod Multi bundle effect Void rising velocity

ABWR 9 Maximum 13 RPV 7 node 4 Mass, energy, momentum Wallis type correlation Wallis type correlation Wallis type correlation Wallis type correlation Joint study [7] Dittus Boerlter

Joint study [7] Andersen Wilson correlation or drift-flux

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3.5.1.5 LOCA analysis code (BE code) (a) TRAC [8,9] TRAC is a computer code for transient analysis of light water reactors. TRAC was originally developed for the PWR LOCA analysis at the Los Alamos National Laboratory (LANL). This code was named TRAC-P1A [10]. Following TRAC-P1A, the BWR version was developed. The BWR version of TRAC has been developed as a result of cooperation between General Electric Company and Idaho National Engineering Laboratory (INEL). Up through 1985, the development work at GE was jointly funded by GE, NRC, and EPRI under the REFILL-REFLOOD and FIST programs. At INEL (which has the main responsibility for the NRC version of TRAC-BWR), this work has led to the development of TRAC-BD1 and TRAC-BF1, while TRACB04 was the final product of the REFILL-REFLOOD and FIST programs at GE. TRAC development has continued at GE after the completion of these programs with the evolution of the TRACG code. The TRAC-BD1 code was developed at INEL for BWR LOCA analysis in 1979. The main work of TRAC-BD1 was composed of both TRAC fundamental model modification and BWR peculiar model installation. This study could be successfully analyzed. This code features a three-dimensional treatment of the BWR pressure vessel and a onedimensional treatment of fuel bundles and pipings. The thermal hydrodynamic model is a nonhomogeneous, nonequilibrium two-fluid formation for two-phase flow. This code is characterized by the coding system. By combining with components such as VESSEL, PIPE, BREAK, and FILL, the main flow structures can be modeled as modules. TRAC was further developed for the analysis of abnormal transients, LOCA and ATWS (anticipated transient without scram), and this version was named TRACG Version 2. TRACG Version 2 was modified to include the kinetic model and was developed to lead to TRACT by Toshiba. (b) RELAP (reactor excursion and leak analysis program) [11] Long-term transient behaviors for small LOCA of BWR/PWR can be analyzed with this analysis code. The original RELAP code was developed by the USNRC until 1996. The original RELAP was applied to the design basis accident (DBA) analysis. For the two-phase model, RELAP has a two-fluid model, a drift flux model, and a homogeneous model. The old version of RELAP4 was based on the homogeneous model, while RELAP5/MOD2 has the two-fluid model. So, among several versions of the RELAP code, RELAP05/MOD2 has been widely used for BE safety analysis. The calculation speed was improved by 3–4 times due to vector treatment. After 1996, RELAP development was continued by ISS (Innovative Systems Software, LLC.) and it has been extended to analyzing the capability for severe accident (SA) with the combination of SCDAP.

3.5.1.6 SA progression analysis code Generally, SA progression analysis has many technical difficulties in that the physical phenomenon itself is complicated. The phenomenon covers three-dimensional with multiphase flow (e.g., vapor-water two-phase or vapor-water-solid three-phase flow). The calculation time is long. The typical analysis outputs are FP source term (e.g.,

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aerosol concentration) as well as thermal-hydraulic data (e.g., temperature and pressure). By using the SA analysis code, the following FP behaviors and their models can be verified [12]. FP aerosol transport model FP grouping FP species treatment Aerosol sedimentation model Gaseous FP sedimentation model Aerosol progression model FP removal model by mitigation measures such as pool scrubbing and spray (a) MELCOR (methods for estimation of leakages and consequences of releases) [13,14] This code has been widely used as a SA progression analysis code. This code has international cooperation from USNRC under CSARP (Cooperative Severe Accident Research Program). MELCOR has been utilized mainly by the regulatory sections [13]. Generally, as for SA analysis code development, the following difficulties accompany it. (1) To calculate complicated phenomena (2) To consider the uncertainty (3) To consider mitigation of the uncertainty (4) To utilize the theory and experimental correlation (5) To incorporate most of the SA important phenomena in the code (6) To consider engineering compromise This code satisfied most of the abovementioned characteristics. This code is a fully integrated, engineering-level computer code developed by Sandia National Laboratories (SNL) for the US Nuclear Regulatory Commission to model the progression of severe accidents in nuclear power plants. A broad spectrum of severe accident phenomena in both BWR and PWR is treated in MELCOR in a unified framework. MELCOR applications include the estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This code is composed of many packages relating SA phenomena [14]. For example, the MP package treats the material properties, the COR package is an interaction between core and structures, and the RN package (radionuclide) targets FP release and delivery behaviors. Using this code, the Fukushima Daiichi nuclear accidents have been also analyzed up to the present. (b) MAAP (modular accident analysis program) [15] This code is also widely utilized as the SA progression analysis code and is mainly used by both utilities and vendors. This computational analysis code was developed at FAI Company (Fauske & Associates, LLC) under the IDCOR (Industry Degraded Core Rulemaking) program. After program completion, its right was removed to EPRI (Electric Power Research Institute) and the developments and refinements are still continued by MUG (MAAP Users Group). Using MAAP long-term thermal-hydraulic and/or FP behavior inside the RPV and PCV during the SA phase can be analyzed. MAAP has been widely used all over the world for PRA and AM management decisions. The main inputs are RPV and PCV pressure and temperature.

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The code model differences between MELCORE and MAAP are frequently compared [12]. As for the debris diameter and cooling performance, the calculated diameter by MAAP is estimated to be smaller than that of MELCOR and, thus, the debris superficial area enhances the coolability of the lower plenum. As for the RPV bottom vessel failure, the melt through temperature by MELCORE is 600 K lower than that of MAAP and thus the melt through timing is faster. With MAAP calculations, melt through timing of penetration is faster than that of the vessel bottom. As for the MCCI model difference, both the initial condition and debris heat transfer model are different, and thus PCV pressure, PCV temperature, hydrogen generation rate, and concrete invasion depth are differently calculated [16]. (c) THALES-2 (integrated severe accident analysis code THALES-2) [17] This code is an integrated severe accident analysis code that simulates core meltdown accident progression and transport of radioactive materials for the probabilistic safety assessment (PSA) of a nuclear power plant. This code can simulate a part of a level 3 PSA. The calculated release fraction of CsI and CsOH are calculated. The governing factors for the source term are judged and the efficiency of containment venting can be estimated. This code was developed by JAEA. Its first version was developed for PWR analysis, which uses THALES-P for primary system thermal hydraulics, THALES-M for core heat up and meltdown, and THALES-CV for containment temperature and pressure response. Several program libraries were also developed not only for THALES but also for general usage. A new analysis technique of hydraulics in the primary cooling system was developed and used in THALES-P aiming at accurate estimation of water level in a water-steam mixture and shorter computer time, which are necessary for core meltdown analysis. Using this code, the release fractions of CsI and CsOH to the environment can be calculated. (d) Other SA analysis codes As another computational code for SA, the SAMPSON code was developed at the IAE (Energy Research Institute) in Japan. Using this code, the Fukushima Daiichi nuclear accident progression was analyzed. Thermal-hydraulic behaviors from RPV to PCV can be analyzed continuously. In Europe, the ASTEC code has been jointly developed by IRSN (France) and GRS (Germany). ASTEC has been used mainly in EU countries.

3.5.1.7 Computational fluid dynamic (CFD) analysis code As far as BWR safety analysis is concerned, most flow environments are in a two-phase flow condition. But, the single-phase flow field is not missing. For example, the flowrelated structure inside a pipe is simulated with fine mesh. This mesh is divided 10,000 to 1,000,000. These single-phase activities are called CFD analysis. In the LWR field as well as in the FBR field, PHOENICS, STREAM, STAR-CD (STAR-CCM +), Abaquas, Nastran, Marc, RIAM-COMPACT, and FrontFlow are the typical codes. In some codes, the flow part of the structural analysis code is included. (a) Star-CD [18] Most of the SA environmental condition is the gas-liquid two-phase flow condition but the single-phase steam flow condition is also important. For example, FP behaviors in PCV have been analyzed by single-phase flow analysis. Recently, the STAR-CD code

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Fig. 3.5.4 Example of Star-CD noding [19]. has been frequently used. This model was developed by Siemens AG. The included full models are composed of combustion, heat problem, unsteady calculation, and turbulent effect. The dependence of the drywell cooler (DWC) heat removal performance on noncondensable species was discussed by analyzing the computational fluid code STAR-CD. And the DWC was found to be promising even under the wide range of SA conditions. The DWC casing which is the analytical object is shown in Fig. 3.5.4 [19].

3.5.1.8 Large-scale test facility for code verification and obtaining correlations During analysis code development, many integral and separate effect experiments were carried out cooperatively by the Japanese BWR utilities and vendors. These test facilities were constructed, aiming at both code verification and obtaining the correlation. In the nuclear thermal-hydraulic test, scaling compromise is not generally allowed. Major facilities have full-height test sections. The lineup of the test facilities is shown in the following. (a) ROSA III (Rig of safety assessment) [20] JAERI (Japan Atomic Energy Research Institute, currently JAEA) started fundamental research regarding the thermal hydraulic tests of LWR LOCA in 1970–73. This is the Phase I ROSA program (ROSA-I). The research was continued to the next phase (ROSA-II) in 1974–77, and the PWR large break of LOCA was investigated. Thus, the ROSA test facility was originally constructed for PWR LOCA clarification because JAERI safety research was limited to PWR safety clarification. During the 5 years since 1978, the ROSA-III test facility at JAERI was modified to simulate a LOCA in a BWR/6 plant with 848 fuel bundles and 24 jet pumps. Analysis of the ROSA-III break area spectrum tests in a recirculation line was accomplished using SAFER03 to assess the predictive capability of the code for a BWR LOCA [21]. Although

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ROSA has several small experimental facilities, the main test component is the LargeScale Test Facility (LSTF). LSTF is a full-height and 1/48 volumetrically scaled test facility for system integral experiments simulating the thermal-hydraulic responses at the fullpressure conditions of a 1100 MWe-class PWR during small-break LOCA (SBLOCAs) and other transients. After the ROSA-III project, ROSA was returned to clarify the PWR small LOCA and useful tests were continued. In the ROSA-IV program, various system integral tests have been conducted to certify the effectiveness of both AM measures in beyond design basis accidents (BDBAs) and improved safety systems in next-generation reactors. (b) FIST ABWR (full-height integral simulation test facility) [22] Before describing FIST, it is better to introduce the TLTA facility founded by GE. The Two-Loop Test Apparatus (TLTA), located at San Jose, CA, was a 1/624-volume-scaled, reduced-height test facility. TLTA has a full pressure/temperature (i.e., 1325 psia, 600 F), nonnuclear test facility emulating a BWR/4. The mission of TLTA was to study steam-line breaks and LOCAs and the validation of GE’s safety analysis methods. As the TLTA height was not full scale, a scaling compromise was needed. To overcome these defects, TLTA was reconfigured and renamed the Full-height Integral Simulation Test (FIST) facility. The design of FIST addressed the principal scaling compromises of the TLTA by extending the pressure vessel to full height (64 ft). This allowed full-size jet pumps and elimination of the volume-scale distortion in TLTA’s recirculation loops. (c) GIRAFFE (gravity-driven integral full-height test for passive heat removal) [23,24] The front view and schematic diagram of GIRAFFE are shown in Fig. 3.5.5. GIRAFFE consists of separate component vessels of the PCCS pool, RPV, SC, DW, and GDCS pool.

Fig. 3.5.5 GIRAFFE Loop and TRAC nodalization for GIRAFFE simulation [25].

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The design criterion of the GIRAFFE test facility is to simulate the long-term PCV pressure transient identical to a reference passive plant of SBR in real time. Flow paths following a LOCA in the SBWR are driven primarily by gravitational force. GIRAFFE has been constructed to maintain the full-height scale of the SBWR design. The scale of the test facility is 1/400 in volume and the height is of the full scale of 30 m. The PCCS unit consists of a steam box, vertical heat transfer tubes, and a water box. PCCS has three heat transfer tubes, which have dimensions of 0.051 m outer diameter to a 1/400 scaled volume. Dimensions of the heat transfer tube, the clearance between adjacent tubes, and the secondary side-flow cross-sectional area per tube is of the full scale of the original SBWR design. In the GIRAFFE testing programs, many thermal-hydraulic aspects have shown that the PCCS has a heat-removal capability sufficient to suppress the containment pressure well below the design pressure. As an extension of the development program, PCCS heatremoval capability under a severe accident condition has been also investigated. These tests have shown that PCCS is efficient even under a severe accident condition [23,24]. After PCCS development of SBWR, this facility was modified and the FP removal characteristics under a low flow rate of the PCV spray condition were investigated and their effects were confirmed [25,26]. Then, a similar type of safety facility (TIGER) was reconstructed by Toshiba. GIRAFFE influenced the worldwide test facility competition. After GIRAFFE, PANDA (PSI) and PUMA (Purdue University) were constructed and the PCCS characteristics were investigated in an international project. (d) ESTA (18° sector test apparatus) [2,27] ESTA is the BWR REFILL-REFLOOD test facility operated under atmospheric conditions, which mocks up an 18° sector of the BWR plant with full height from the jet pump bottom to the standpipe top as shown in Figs. 3.5.6 and 3.5.7. This facility was constructed by Toshiba. As this test section was divided into 18° portions of the full circular reactor, the main components within the shroud are 1/20 of the actual BWR size. The ESTA is an atmospheric facility. Steam generation is simulated by external steam injection from a boiler. The sector representation of the reactor core includes 25 full or partial fuel channels, each with a steam injector to simulate vapor generation in the bundle. Steam is also injected into bypass and guide-tube regions for simulating liquidflashing steam. All ECCSs are simulated. Using this facility, the following correlations were experimentally obtained. UTP CCFL and CCFL breakdown Core inlet orifice CCFL Bypass top CCFL Multichannel effect Bypass bottom CCFL Single-phase steam cooling heat transfer Mist flow heat transfer Spray heat transfer The cross section for ESTA is shown in Fig. 3.5.1. Obtained correlations were installed in SAFER03. (e) Other facilities Hitachi designed the TBL (two-bundle loop) for LOCA/ECCS design development. Hitachi also developed a natural convection heat transfer test facility for the development of a water wall during the SBWR design phase. Although other thermal hydraulic test facilities were constructed by the Central Research Institute of Electric Power Industry (CRIEPI), the description of the safety-related facilities is restricted in the chapter.

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Fig. 3.5.6 ESTA Loop.

3.5.2

BWR safety systems for severe accident

3.5.2.1 Passive safety concept (a) New safety movement to passive safety Nuclear power plants have high reliability in case of any serious accident. Certain amount of electrical power has to be continuously supplied even during an accident. To satisfy these requirements completely, the adoption of a passive system concept was proposed. This concept prevailed in the 1980s. The passive plant makes use of natural force for core cooling to simplify the safety system and to enhance reliability. To conflict, it was contradicted that the electrical power system was available with enforcement safety. After many turns and twists, passive safety still has a position for severe accident measures. (b) PIUS (process inherent ultimate safety) [28,29] In 1975, the development of PIUS started in Sweden (ASEA-ATOM) as the earlier representative of the passive reactor. The small-type PWR is submerged in the large pool containing boron water. Both upper and lower pipes are been connected to the pool water without valves. These nest-box-type two vessels are locked by the density difference so as to avoid mixing. The density lock was accomplished by honeycomb structures (diameter 55 m, height 1 m). In case of an accident, the density balance between the RPV and the pool

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Fig. 3.5.7 Schematic representation of ESTA [27]. is broken and boron water naturally enters into the core and thus fission is interrupted naturally. After that, the decay heat is cooled by natural circulation. In this way, the reactor was automatically stopped without electrical power. The RPV core is also cooled simultaneously. Control rods are not needed. Although the PIUS research activity was reported somewhat earlier than the following actions, it can be said that the passive safety concept was initiated from the PIUS studies.

3.5.2.2 Reinforcement for passive safety Since the construction of many LWR plants, initial plant troubles were detected and cleared. As the plant operational rate exceeded 80%, the adoption of a new plant design concept was raised. The design theme focused on the passivity. The passive design encompassed not only safety but also operational advantage. The US DOE requested the LWR vendors to design all passive reactors. To compete, the BWR vendor consociation presented a new plant design concept for SBWRs. The Simplified BWR (SBWR) makes use of natural force for core cooling and decay heat removal to simplify the safety system and to enhance its reliability. The SBWR was of medium size and its electrical power was 600 MW. The coolant is circulated by two-phase natural circulation. The long-term decay heat is cooled by the passive containment cooling system (PCCS).

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After the competition, one plant each for BWR and PWR was selected. The SBWR pursuit was for simplification but the BWR makers continued to design. The ESBWR (Economical SBWR) was presented by a GE and Hitachi joint group. But, these competitions were terminated since the Fukushima Daiichi Nuclear Accident in 2011. Nevertheless, the natural force for passive safety is not much selected. Natural circulation and condensation are possible technologies. The power density using a passive system is lower than that with electrical power. Up to the present, the part of the passive plant is limited. The long-term decay heat cooling by a passive safety system is the possible option. The normal system uses electricity. In this chapter, the design of the decay heat removal system is described.

3.5.2.3 Lineup of passive safety systems In the lineup of the decay heat removal systems, both existing passive systems and proposed next-generation plant systems are summarized. This system is almost the same as containment cooling system. (a) Isolation condenser (IC) As is shown in Fig. 3.5.8, the steam-line from RPV is separately divided with the main steam lines (MSLs). In case of accidents, all main steam isolation valves (MSIVs) are shut and the RPV steam is shut to the turbine through the IC line. Then, isolated steam is introduced to this line and this steam is condensed by the secondary shell side water, which is located in the upper drywell space. The RPV steam is condensed inside the tube and the condensate water returns to the RPV via the ICRL. The RPV water level is always maintained. This system is called the isolation condenser (IC). As a heat sink, water makeup for the secondary shell is possible (Fig. 3.5.8).

Fig. 3.5.8 Isolation condenser [30].

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Isolation condensers were installed in earlier BWR plants in Japan. The Turuga Unit 1 and Fukushima Unit 1 nuclear power plants installed the IC. This system is needed to locate the heavy water tank at higher elevation. As the plant mass center moves higher, the IC was not installed. Once the IC is activated, much steam is released into the plenum. It is easily noticed that much steam release from the blowout panel indicates its activation. It is reported that such information could not be well transferred to the operator in case of the Fukushima Daiichi nuclear accident in 2011. (b) Passive containment core cooling system (PCCS) [23] The passive containment cooling system was designed for the SBWR design. It removes decay heat following an accident without any electric power and any operational actions. It provides engineered safety features enhancing safety system reliability and plant simplification. This system is strongly influenced by the IC. Although its function is identical to that of the IC, PCCS is limited to decay heat removal only. The general function of the PCCS is described in the following. As the GDCS pool empties following a LOCA, a coolant must be supplied to keep the core covered. To ensure this, the SBWR RPV is surrounded by a suppression pool (S/P) so as to supply water to the RPV through the equalizing line (EQL). The PCCS is designed for low pressure. Decay heat removal following a reactor isolation or a LOCA is achieved by the isolation condenser (IC) and/or the PCCS. Fig. 3.5.9 shows the PCCS configuration. The PCCS consists of a steam supply line connected to the upper portion of the drywell, a vertical shell-and-tube single-path heat exchanger in a large water pool, a condensate drain line, and a vent line for noncondensable gas (denoted as the PCC vent line hereafter). During a LOCA, the high-pressure steam reactor coolant comes from the RPV to the drywell causes a pressure rise in the D/W. The pressure difference between the D/W and the suppression chamber (S/C) renders a steam-noncondensable mixture in the D/W that is absorbed into the PCCS heat exchanger through the steam supply line. The reactor steam is condensed inside the PCC tubes and the condensate drains to the GDCS pool through the condensate drain line by the gravitational force, and the noncondensable is designed to be vented to the S/C through the PCC vent line. It is important that the submergence of this noncondensable gas vent line is kept shallower than the horizontal vent line. No poweractuated active devices are required for the PCC to function. The large water pool is located outside the drywell and serves as heat sink for reactor decay heat. The water pool retains a sufficient amount of water to remove the decay heat for at least three days from the shutdown, which is called the 3-day walk-away period. Consequently, the PCCS offers high reliability of functioning due to no valve operation required. In addition, the PCCS has recently come to be expected to suppress the containment pressure in case of a severe accident as well as in a LOCA, and prevent containment failure without venting gases in the containment to the atmosphere. The Japanese vendor Toshiba has been taking the lead in the thermal-hydraulic research of the PCCS. The BWR vendors have carried out an experimental program as well as analytical work using a best estimate thermal-hydraulic code TRAC to develop the PCCS. The test used was GIRAFFE. Based on experimental and analytical studies, the mechanism of the passive heat removal has been clarified and the PCCS heat-removal performance following a LOCA has been assessed intensively. Fig. 3.5.10 shows the design confirmation test facility GIRAFFE by Toshiba (also shown in Fig. 3.5.5). Fig. 3.5.11 shows the horizontal tube PCCS systems. The heat-removal performance of the horizontal-type PCCS was investigated by the cooperation between Toshiba and JAEA.

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Fig. 3.5.9 PCCS of SBWR [23]. After the SBWR design completion, the PCCS concept has impacted on the BWR PCV design. The next-generation plant design was presented with many cases, and this decay heat-removal system has been included. (c) RCIC (reactor core isolation cooling system) In case of abnormal accidents such as pipe break, all MSIVs are closed and the RPV is isolated. Although normal operation is shut down, the RPV pressure still increases due to decay heat. The SRV opens and much steam is vented to the suppression pool and the water level decreases. As makeup water cannot be supplied into the RPV, this RCIC system is activated and water is supplied by the pump. Pump power is generated by a turbine, which is produced by the RPV steam. The makeup water is used for the CS tank and suppression pool (SP) water. In case of initiation only, electrical power is needed but once activated, this system opens without electrical power. In case of the Fukushima Daiichi nuclear accident in 2011, it is reported that the RCIC worked on the 2nd and 3rd Unit plants for 2– 3 days after the accident.

Nuclear reactor dynamics and thermal hydraulics of reactor core

Fig. 3.5.10 PCCS test facility (GIRAFFE).

Fig. 3.5.11 PCCS configuration and heat transfer exchanger (horizontal type).

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(d) GDCS (gravity-driven core cooling system) This system is designed by General Electric for the next-generation BWR (SBWR). The GDCS pool is placed above the reactor pressure vessel (RPV) and the gas space above the pool is common to the drywell (DW). The GDCS is designed to keep the core covered with water in any LOCA, assuming a single failure. However, if the RPV pressure does not decrease well, the GDCS cannot inject water by gravity. In order to activate the GDCS, RPV depressurization is required. The RPV depressurization valves (DPVs) promote GDCS injection by depressurization the RPV. The GDCS effect was confirmed by the GE test facility (GIST). (e) Utilization of normal systems for PCV cooling Apart from the next-generation reactors, for the AM of the existing BWR, new concept systems are difficult to modify. The utilization of the normal system is the best option. As for the BWR AM, if the PCV heat removal cannot be soon recovered by the residual heatrecovery system (RHR), alternate heat-removal systems such as a PCV spray by the fire pump or drywell cooler (DWC) are possible measures. Although a low flow-rate spray is considered to suppress the PCV pressure, combined usage of the DWC and spray is expected to reduce the PCV pressure. (e-1) Drywell cooler for SA [19] The drywell cooler (DWC) is a normal system and its heat exchanger is located in the PCV (Fig. 3.5.12). The DWC removes the heat released from the RPV by a fan or cooling coil. The PCV air is blown by a fan through a duct, while the in-coil water is cooled by RCW and RSW. Transferred heat is finally cooled by seawater. In case of the SA condition, air around the DWC has high humidity and condensation is expected to be predominant. As the BWR DWC heat-removal performance is favorable, it has been applied as a Japanese phase-II management strategy. Separate effect tests were conducted using a single DWC unit of a typical BWR plant under the SA condition. It was demonstrated that a noncondensable gas mixture with nitrogen and helium was constantly vented from the DWC casing by natural circulation due to the density difference between the DWC tube space and outside. A favorable steam condensation rate was maintained even under the highest assumed noncondensable gas condition. Obtained condensation data sets were reduced to empirical correlations for the DWC heat-removal model. The dependence of the DWC heat-removal performance on the noncondensable species was discussed by analyzing the computational fluid dynamic code of STAR-CD. In conclusion, the DWC was found to be promising even under a wide range of SA conditions.

Fig. 3.5.12 BWR drywell cooler [19].

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Spray nozzle

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Fig. 3.5.13 Low flow-rate spray shape (Narrow spray angle and large droplet).

High humidity condion by steam Atomizaon of spray droplet

(e-2) Low flow rate of PCV spray for SA [32] The PCV spray is designed to reduce the PCV vessel pressure by condensing steam using atomization. From the PCV spray nozzle, water issues with tiny droplet and behaves as the mist flow under normal conditions. It plays an important role in pressure reduction due to the high efficiency of the heat absorption rate. Under the low flow rate of the SA condition, however, spray mist was not well created. Although the spray angle becomes narrow and droplet diameter become large, it has been confirmed that PCV pressure could be suppressed lower by the vendors’ tests. Fig. 3.5.13 shows the low flow-rate spray patterns. (f) Other systems As for other passive safety systems, steam injection systems (SISs) were also applied to passive safety design. By using condensation of the shear layer between the water jet and steam flow, these systems behave like pumps and supply water without electricity. The SIS is a very simple and compact device. Research and developments for SI have been conducted and these are summarized in Section 5.6.

References [1] N. Abe, M. Katoh, H. Nagasaka, H. Aoki, TRAC-BD1 calculation of ESTA test, in: Tenth Water Reactor Safety Research Information Meeting, 1982. [2] H. Nagasaka, S. Itoya, S. Yokobori, N. Abe, Qualitative evaluation of ECCS capability during refill-reflood phase of BWR LOCA, J. Nucl. Sci. Technol. 28 (1991) 258–267. [3] ASME, Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer, ASME V&V 20, 2009. [4] S. Koshizuka, V&V importance at CAE (in Japanese), in: NAFMES Japan, 2013. [5] H. Nakamura, Development of thermal-hydraulic nuclear safety analysis code in Japan and overseas countries (in Japanese), RIST News (2011) 30–42. No. 51. [6] A. Omoto, T. Sugisaki, H. Nagasaka, SAFER code development of BWR LOCA, AESJ J. (1986) 950–956. [7] TLR-028: Toshiba Licensing Report of SAFER03. [8] J.W. Spore, et al, n.d. TRAC-BD1: An Advanced Best Estimate Computer Program for Boiling Water Reactor Loss of Coolant Accident Analysis, NUREG/CR-2178.

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[9] J.G.M. Andersen, J.C. Shaug, B.S. Shiralkar, TRAC development at General Electric, in: Proceedings of the USNRC 14th Water Reactor Safety Information Meeting, Vol. 5, 1987. [10] TRAC-PA1, An advanced best estimate computer program for PWR LOCA analysis, Los Alamos Scientific Laboratory Report, LA-7777-MS (NUREG/CR-0665), May 1979. [11] V.H. Ransom, et al., RELAP5/MOD2, EGG-SAAM-6377, EG&G Idaho Inc, 1984. [12] AESJ, Fission Product Behavior under Severe Accident, Research Committee on Fission Product Behavior under Severe Accident (in Japanese), 2021. [13] SNL MELCOR website, http://energy.sandia.gov/energy/nuclear-energy/nuclear-nergysafety-technologies/melcor/. [14] Y. Maruyama, Severe accident progressive code MELCOR (in Japanese), AESJ Fuel WG (2012). [15] Mitsubishi Heavy Industries, As for MAAP application (in Japanese), MHI-NES-1056, 2013. [16] Nishimura, et al., Transient and accident analysis of the typical LWR by using MAAP 5.01 and MELCOR 2.1, CRIEPI Report, L13006, 2014. [17] J. Ishikawa, K. Muramatsu, T. Sakamoto, Systematic source term analysis for level 3 PSA of a BWR Mark-II type containment with THALES-2 code, JAERI-Research 2005-021, 2005. [18] http://www.hpv.co.jp/product/software/star-cd-star-com/s. [19] S. Yokobori, T. Tobimatsu, M. Akinaga, M. Fukasawa, H. Nagasaka, Steady heat removal test by BWR drywell cooler under severe accident conditions, AESJ 2 (3) (2003) 230. [20] H. Nakamura, As for LSTF Plan (in Japanese), 2007. [21] S. Itoya, H. Nagasaka, K. Moriya, S. Miura, Overview of SAFER03 assessment studies, J. Nucl. Sci. Technol. (1988). [22] R. Martin, GE’s Integral Systems Tests SSTF, TLTA and FIST. http://www.linkedin.com/ pulse/ges-integral-systems-tests-sstf-tlta-fist-robert-martin. [23] S. Yokobori, T. Tobimatsu, K. Arai, H. Oikawa, Characteristic Features of Passive Containment Cooling System (PCCS) and its Related Removal Research (in Japanese), JSME No. 940-56, 1994, pp. 447–451. [24] S. Yokobori, T. Tobimatsu, T. Kurita, M. Akinaga, K. Arai, H. Oikawa, Heat removal mechanism of passive containment cooling system for ALWR, in: Proceeding of 12th International Heat Transfer Conference, 2002. [25] H. Oikawa, S. Yokobori, K. Arai, M. Akinaga, Research Activities on Passive Containment Cooling System for ALWR Application, in: NTAS98: First Korean-Japan Symposium on Nuclear Thermal Hydraulics and Safety, 1998. [26] H. Nagasaka, A. Watanabe, S. Yokobori, Fission product aerosol removal test by containment spray under accident management conditions, in: 3rd OECD Specialist Meeting on Nuclear Aerosols in Reactor Safety, Cologne, 1998. [27] H. Nagasaka, M. Katoh, S. Yokobori, 18 Degree Sector System Test (ESTA-II), in: Transaction of 12th Light Water Reactor Specialist Information Meeting, Washington, 1984, pp. 706–721. [28] K. Hannerz, T. Pederson, PIUS LWR design progress, in: IAEA Technical Meeting on Advances in LWR Technologies, Washington, 1996. [29] K. Tasaka, The Concept of Passive Reactor Safety Systems and their Future Needs (in Japanese), JSME No. 940-56, 1994, pp. 13–21. [30] NHK Special Meltdown III (in Japanese). [31] J.A. Beard, ESBWR Overview, 2007. [32] S. Yokobori, H. Nagasaka, A. Watanabe, T. Tobimatsu, M. Akinaga, Fission Product Aerosol Removal Tests by Low Flow Rate Containment Spray (in Japanese), JSME, 2000, pp. 429–430.

4

Fukushima Daiichi nuclear power plant accident and analysis evaluation Tadashi Narabayashi Tokyo Institute of Technology, Meguro, Tokyo, Japan

Chapter outline 4.1 4.2 4.3 4.4 4.5 4.6

Outline of accident 336 Event progress and analysis evaluation at Unit 1 337 Event progress and analysis evaluation at Unit 2 339 Event progress and analysis evaluation at Unit 3 341 Hydrogen explosion at Unit 4 343 Avoiding severe accidents at Fukushima Daini NPS 346 4.6.1 Overview of emergency response at Fukushima Daini NPS 346 4.6.2 Fukushima Daini Unit 1 response and station behavior 350

4.7 Lessons learned from Fukushima Daiichi accident 4.7.1 4.7.2 4.7.3 4.7.4 4.7.5

362

Causes of severe accidents and countermeasures 362 Measures for severe accidents installed in the United States and European NPPs 362 Filtered containment venting system 362 Special emergency heat removal system 367 Tsunami protection 367

4.8 New nuclear regulatory requirements in Japan

368

4.8.1 New nuclear regulatory requirements 368 4.8.2 Tsunami protection examples 368 4.8.3 Tornado protection examples 373

4.9 Example of compliance with new regulatory standards for PWRs that can be used as a reference for BWRs 374 4.10 BWR NPS to be reviewed for new requirements or restarting 378 4.11 Activities toward decommissioning Fukushima Daiichi 389 4.11.1 4.11.2 4.11.3 4.11.4 4.11.5

Current status of reactors at Units 1 through 4 389 Finding contaminated water leak path for leak shutdown from PCV 392 Isolation of groundwater flow from contaminated water 394 Contaminated water management 396 Preparation for fuel-debris removal 400

4.12. Important lessons learned from Fukushima Daiichi NPS accident References 404

Boiling Water Reactors. https://doi.org/10.1016/B978-0-12-821361-2.00005-2 Copyright © 2023 Elsevier Inc. All rights reserved.

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Boiling Water Reactors

Outline of accident

On March 11, 2011, Tokyo Electric Power Company’s Fukushima Daiichi Nuclear Power Station (NPS) was struck by a tsunami caused by the Great East Japan Earthquake, resulting in nuclear accidents in Units 1 through 4 [1,2]. With the aim of improving the safety of nuclear power plants (NPPs) worldwide, we summarize the lessons learned following a thorough analysis of the event and make specific proposals for improving the safety of such facilities. The author has been involved in investigating the causes of accidents and developing countermeasures for other NPPs in Japan as a member of the Committee for the Investigation of Nuclear Safety of the Atomic Energy Society of Japan. He is an advisory meeting members of the Nuclear and Industrial Safety Agency (NISA) and the Nuclear Regulation Authority (NRA) with regard to technical lessons learned from the Fukushima Daiichi NPS accidents, and a Safety Evaluation Member of the NISA for the other NPPs in Japan [3–5]. This information was submitted by NISA to the IAEA as a Japanese government report [6] in September 2011, indicating that the IAEA also shared detailed information at a fairly early stage. Many lessons can be learned from the Fukushima Daiichi NPS accident. First, if an isolation condenser (IC) had continued to operate, the accident would have terminated soon. Reactor core isolation cooling (RCIC) steam turbines also stopped because loss of battery power in Units 2 and 3, and temperature and pressure in each primary containment vessel (PCV) were so high that the accident management and water injection took too long. After the loss of emergency core cooling system (ECCS) and IC core cooling, fuels in the core melted. Leak of fission product and hydrogen began because of the damage to the O-ring seals between PCV upper flanges due to high temperature. Hydrogen explosion occurred in the upper floor in the reactor building at Units 1, 3, and 4. The New Regulatory Requirements, based on the concept of “defense in depth,” for Commercial Nuclear Power Reactors came into force on July 8, 2013. It is hoped that the lessons learned from this accident will improve the safety of nuclear power plants worldwide. Most of this chapter is reprinted from the Fukushima Nuclear Power Plant Accident and thereafter in the book “Energy Roadmaps of Japan” [2] published by Springer. Dr. Tadashi Narabayashi, author, would like to thank Springer for the permission to reprint. Fig. 4.1 compares flooded areas at each NPS. Although other NPSs such as Fukushima Daini, Onagawa, and Tokai Daini were also struck by the tsunami, they all were able to safely terminate operation, until the cooldown condition. The Fukushima Daini NPS succeeded in safe shutdown, even though Unit 1 was affected by water flooding through hatches and an emergency diesel generator (EDG) air intake. AC power was restored by changing the power cable and the seawater pump motors were replaced by bringing in new motors from the Toshiba Mie Works and Kashiwazaki-Kariwa NPS by helicopters. At the Fukushima Daiichi NPS, Unit 5 was brought under control by using EDG power from Unit 6 [2]. Fig. 4.2 shows the comparison of the flood damage to EDGs. At Units 1 through 4, there was a complete loss of both AC power from the EDGs and DC power, and this was the main cause of severe accidents, as summarized in Table 4.1 [2].

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Fig. 4.1 Comparison of flooded areas at each NPP [2,4].

Fig. 4.2 Comparison of flood damage to emergency diesel generators for Fukushima Daiichi and Daini NPS [2,4].

4.2

Event progress and analysis evaluation at Unit 1

At Unit 1, DC battery power was lost in the main control room. This caused motoroperated (MO) isolation valves to undergo fail-close action, thereby cutting off the isolation condenser (IC), as shown in Fig. 4.3. This was a fail-dangerous system under such situation. If the IC had continued to operate, the situation would soon have been brought under control [4].

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Table 4.1 Components damaged by tsunami at each unit of Fukushima Daiichi NPP [4]. #1

#2

#3

#4

#5

#6

DG

A:NG B:NG (T/B B1)

A:NG (B1) B:OK (FP/B 1F)

A:NG B:NG (T/B B1)

A:NG (T/B B1) B:OK (FP/B 1F)

Metal-clad switch Power center DC Buttery ECCS RCIC

NG (T/B B1) NG (T/B B1) NG (C/B B1) HPCI:NG IC: OK(FC)

NG (T/B B1) Barely (T/B B1) NG (C/B B1) NG RCIC: OK

NG (T/B B1) NG (T/B B1) OK (T/B BM1) HPCI:OK RCIC: OK

NG (T/B B1) Barely (T/B 1F) NG (C/B B1) (No Fuels in RPV)

A:OK->NG B:OK->NG (T/B B1) Water cooling NG (T/B B1) Barely (T/B 2F) OK (T/B BM1) -

A:OK->NG (R/B B1) Water cooling B:OK (DG/B 1F) Barely (R/B B2F) Barely (R/B B2F) OK (T/B BM1)

MO

MO

MO

MO

HPCS:OK (R/B B1)

MO MO

MO MO

MO MO

Fig. 4.3 Isolation condensers in Fukushima Daiichi Unit 1 [2,4].

After the loss of both the emergency core cooling system and IC core cooling, primary containment vessel (PCV) pressure increased at midnight of 0:00, March 12, as shown as an arrow mark and a circle in Fig. 4.4. Water-level measurement drifted because of water evaporation in the reference leg (Fig. 4.5). Radiation level increased at a turbine building (T/B). There was a hydrogen explosion after suppression chamber (S/C) wet venting.

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Fig. 4.4 Measured pressure and water level in RPV and CV of Unit 1 [4].

Fig. 4.5 MAAP analysis results compared with actual plant data for Unit 1 [2,6].

As shown in Fig. 4.5A, both Modular Accident Analysis Program (MAAP) code [7] analysis results and actual data suggest that depressurization of the reactor pressure vessel (RPV) began before its bottom failed.

4.3

Event progress and analysis evaluation at Unit 2

At Unit 2, reactor core isolation cooling (RCIC) continued to function for about 3 days. Fig. 4.6 shows that soon after the loss of RCIC water injection, the water level in the RPV decreased. The safety relief valve (SRV) was opened and seawater

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Boiling Water Reactors

Fig. 4.6 Measured pressure and water level in RPV and PCV of Unit 2 [2,4].

injection started. But, RPV pressure shows fluctuation due to water evaporation and metal-water reaction in core. Dry well (DW) pressure increased from 400 to 750 kPa, and PCV top flange leak began through silicon rubber O-ring. It was the initiation of severe contamination around the NPS. In the afternoon on March 15, wind blew toward Iidate village. Melted core relocation into a lower plenum caused the RPV bottom control rod drive (CRD) pipe failure and PCV pressure and radiation level increased (Fig. 4.7). The radiation level was measured by containment atmospheric monitoring system (CAMS) [8].

Fig. 4.7 MAAP analysis results compared with actual plant data for Unit 2 [2,9].

Fukushima Daiichi nuclear power plant accident and analysis evaluation

4.4

341

Event progress and analysis evaluation at Unit 3

At Unit 3, the steam turbine-driven isolated water injection system (RCIC) continued to operate even after the tsunami. However, due to the freshwater source shortage, it became impossible to inject water and RCIC tripped at 11:36 on March 12. Due to the shutdown of RCIC, the reactor water level dropped, as shown in Fig. 4.8. When this water-level drop was detected, the steam turbine-driven high-pressure core water injection system (HPCI) was started. This HPCI steam turbine sucks the main steam, and the pressure in the reactor pressure vessel drops rapidly due to the injection of cold water into the core, as shown in Fig. 4.9. As a result, HPCI was tripped by operator. The pressure in RPV increased again to the operating pressure of the SRV as a safety mode valve. For this reason, it became impossible to inject water into the core with the water injection pump driven by the diesel engine that had been prepared as mobile pump. The operator was connecting the car battery to try to activate the automatic depressurization system (ADS), and at around 9:00 on March 13 the ADS suddenly activated and the RPV pressure dropped sharply. As the inventory of water in the RPV was rapidly lost due to the operation of ADS and the core water injection was not enough, the reactor water level began to drop in MAAP analysis, and the top of active fuel (TAF) level was cut after 9:00. Around 10:40, the core damage was started, but it is pointed out that the actual core damage was earlier than the drift of the water-level gauge. It means the temperature in the PCV had already heated up and vaporizes water in the reference leg of water-level meter and the measured value of the water level is displayed higher than the analysis value by nearly 3 m. This suggests that it is necessary to adopt “back fill” system to supply water into the reference leg, as shown in

Fig. 4.8 Water-level trend in RPV of Unit 3 [9] (TAF ¼ 0 m).

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Boiling Water Reactors

Fig. 4.9 RPV pressure trend of Unit 3 [9].

Fig. 4.62. The measures implemented in the United States as the means of measuring the water level in the event of a severe accident. As shown in Fig. 4.10, the measured pressure in the S/C reached about 650 kPa at 12:00 on March 12, wet well (WW) vent was conducted three times. But due to degradation of the O-ring made by silicon rubber, the PCV top flange lost tightness for leak, causing steam containing hydrogen to leak for about 2 days. It is estimated that it

Fig. 4.10 Pressure trend in PCV of Unit 3 [9].

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had accumulated in the upper part of the building. At 11:01 on March 14, the violent hydrogen explosion that blew up black smoke blew off the upper part of the reactor building walls and the steel frame on the roof were deformed to the extent that they could not retain their original shape. Since then, the reactor pressure and the D/W pressure have been moving in parallel with almost the same pressure curve, some leak path of RPV or connected pipes were made. It is estimated that the D/W pressure drifted to the indicated value due to the impact of the hydrogen explosion. It is necessary to calibrate these instruments when they are put into inspection inside the reactor building to improve the accuracy of event analysis. It has been pointed out that since the piping of the digestive system had leaked to the condenser, almost no water was injected into the core until water was injected via the water supply system that was added later.

4.5

Hydrogen explosion at Unit 4

Fig. 4.11 shows hydrogen explosion/detonation in Units 1, 3 and 4. Upon the Unit 1’s explosion occurrence the blowout panel of Unit 2 was opened. Hydrogen in Unit 2 was released through the opened blowout panel and there was no explosion in that unit. The sound of the explosion was reported near the S/C of Unit 2. However, examination of the data showed that this was due to a hydrogen detonation in the reactor building R/B of Unit 4. Soon after this detonation, DW pressure in the Unit 2 decreased (Fig. 4.7). Fig. 4.12 shows trends in monitored radiation dose levels for Units 1–4, which can be compared with events illustrated in Fig. 4.13. It appears that the explosion occurred after venting operations. The radiation level increased soon after the Unit 2 PCV rupture on March 15. A loss of core cooling occurred because of the IC trip in Unit 1, and the RCIC steam turbine also tripped

Fig. 4.11 Hydrogen explosion/detonations occurred after vent operations (Units 1, 3, and 4) [2,9].

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Fig. 4.12 Monitored radiation levels for Fukushima Daiichi Units 1–4 [3].

Fig. 4.13 Chain of major events at Units 1 through 4 causing severe accidents at Fukushima Daiichi NPS [2].

owing to loss of battery power in Units 2 and 3. Suppression pool (S/P) temperature and pressure became so high that water injection actions for accident management took a long time. This was the reason for the chain of severe accidents in the four units of Fukushima Daiichi NPS, as shown in Fig. 4.13 [2]. Fig. 4.14 shows that the PCV top flange and hatches can act as leakage pathways. Hydrogen and FP flow upwards by way of stairways and hatches. Although there was no nuclear fuels in the reactor core of Unit 4, hydrogen flowed from Unit 3 through the stack line into Unit 4 and underwent reverse flow through the standby gas treatment system (SGTS) filters (Fig. 4.15) [2,9].

Fig. 4.14 Estimated leak path from PCV by over pressure and high temperature [2,9,10].

Fig. 4.15 H2 gas flow into Unit 4 reactor building from Unit 3 [2,6,9,10].

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Boiling Water Reactors

Fig. 4.16 Flow diagram of SGTS/HVAC and added hard vent system [2,4].

There was a strong hydrogen explosion occurred in the Unit 4 reactor building on March 14. The author pointed out to the NISA that the harden vent line might have acted as a means of hydrogen and fission products (FP) leakage through SGTS and HVAC lines (Fig. 4.16). As shown in Fig. 4.17, TEPCO confirmed that the SGTS filters were contaminated, and all MO valves were open because of the fail-open design in Units 3 and 4. Seat of the butterfly valves was made of neo-plane rubber and damaged by iodine. This might have caused hydrogen detonation in Unit 4, where there were no nuclear fuels in the reactor core, because hydrogen and FP could have flowed back into each room through the exhaust gas ducts. The vent lines of each NPP should have been independent of the SGTS/HVAC line.

4.6

Avoiding severe accidents at Fukushima Daini NPS

4.6.1 Overview of emergency response at Fukushima Daini NPS There are four nuclear plants at Fukushima Daini NPS. Three nuclear plants (Units 1, 2, and 4) lost all their ultimate heat sinks owing to damage from the tsunami on March 11, 2011, as shown in Table 4.2. Water was injected into the reactors by alternate measures, damaged cooling systems were restored with promptly supplied alternative equipment, fuel oil for engines was carried into the station, and then all reactors were brought to a cold shutdown state within 4 days.

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Gravity Dumper A System

Close Close

Close Close

Close

Close Close

Close

B System

(a) Fukushima Daiichi Unit 3

Close Connected with vent line from Unit3

Close Close

(b) Fukushima Daiichi Unit 4

Fig. 4.17 Results of SGTS valves open/close status and filter contamination [2,6].

Lessons learned from this experience were identified to improve emergency management, especially in the areas of strategic response planning, logistics, and functions supporting response activities continuing over a long period. It was found that continuous planning activities reflecting information from plant parameters and response action results were important, and that relevant functions in emergency response organizations should be integrated. Logistics were handled successfully, but many difficulties were experienced.

Table 4.2 Status of cooling system at Fukushima Daini NPS after tsunami [4]. System RHR (A) including cooling systems

Unit 1 RHR (A)

4

RHRC/RHRS (A, C)

✕

EECW (A)

✕

4

LPCS

RHR (B) including cooling systems

RHR(B)

4

RHRC/RHRS (B, D)

✕

EECW (B)

✕

Unit 2 Loss of power source and cooling system Submerge of power source and motor Submerge of power source and motor Loss of power source and cooling system Loss of cooling system Submerge of power source and motor Submerge of power source and motor

Unit 3

Unit 4

4

Loss of cooling system

4

Loss of cooling system

4

Loss of cooling system

4

Loss of cooling system

✕

✕

✕

Submerge of power source and motor Loss of cooling system

✕

Submerge of power source and motor Submerge of power source and motor Loss of cooling system

Submerge of power source and motor Submerge of power source and motor Loss of cooling system

4

4 ✕

✕

Loss of cooling system Submerge of power source Submerge of power source

4

✕

4

◯

Stand-by

4

◯

Stand-by

✕

◯

Operation

✕

Loss of cooling system Submerge of power source and motor Submerge of power source

RHR (C)

4

Loss of cooling system

4

Loss of cooling system

◯

Stand-by

4

Loss of cooling system

RWCU

4

Loss of cooling system

4

Loss of cooling system

4

Loss of cooling system

4

Loss of cooling system

MUWC (alternative water injection)

◯

Stand-by

◯

Stand-by

◯

Stand-by

◯

Stand-by

RCIC

◯

Stand-by

◯

Stand-by

◯

Stand-by

◯

Stand-by

◯, operable; 4, loss of function due to the loss of support systems; ✕, inoperable.

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Boiling Water Reactors

Therefore, their functions should be clearly established and improved by emergency response organizations. Supporting emergency responders in the aspects of their physical and mental conditions was important for sustaining continuous response. As a platform for improvement, the concept of the incident command system (ICS) was applied for the first time to a nuclear emergency management system, with specific improvement ideas such as phased approach in response planning and common operation pictures. Here, the process of cold shutdown of four reactors was analyzed, taking as an example the emergency response of Unit 1, which was the most severe condition when the emergency diesel generator was submerged [11].

4.6.2 Fukushima Daini Unit 1 response and station behavior 4.6.2.1 Response status at the time of tsunami arrival Unit 1 was in rated thermal operation when the earthquake occurred at 14:46 on March 11. The earthquake had its hypocenter in offshore of Sanriku, and caused automatic shutdown of the reactor at 14:48 of the same day. The reactor was confirmed to be subcritical at 15:00 of the same day. There were four lines for off-site power equipment at Fukushima Daini NPS (Tomioka line: two lines, Iwaido line: two lines). Excluding one of the Iwaido lines which was shutdown for inspection prior to earthquake occurrence, three lines were usable. Of these three lines, two shutdown: one of the Tomioka lines due to earthquake, and one of the Iwaido lines due to Shin Fukushima Substation equipment malfunction. The last of the Tomioka lines continued supplying power. After reactor automatic shutdown, the work management team stationed in an office near the main control room (MCR) (comprised of shift supervisor and operators, separate from the shift team in charge of operations) rushed to the MCR to support the shift team. Supporting personnel were also dispatched to the MCR from the emergency response center (ERC) at the power station. Operators focused on station monitoring/operation for response from that point onward, while also keeping close contact between the MCR and the ERC at the power station. Response was carried out after tsunami arrival (visual inspection after first wave arrival at 15:22 on March 11). These included manually fully closing the MSIV and manually activating the RCIC for reactor cooling water injection at 15:36 on March 11. Reactor depressurization via SRV was started at 15:55 of the same day. Reactor water-level control via RCIC and reactor pressure control via SRV were both carried out based on station parameters, at locations stipulated in the emergency operating procedure (EOP) (warning sign basis). As shown in Fig. 4.18, all emergency component cooling water system pumps were inoperable due to the tsunami impact (unusable due to water damage to certain motors and power sources), all ECCS pumps became inoperable. The site superintendent deemed the situation to be one falling under Article 10 of the Nuclear Emergency Act (loss of reactor heat removal function) at 18:33 on March 11 due to loss of reactor residual heat removal function caused by the above events.

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Fig. 4.18 Flooding status caused by tsunami [9].

4.6.2.2 Reactor cooling water injection and PCV cooling Reactor cooling water injection was initially performed solely via RCIC. Alternate water injection (introduced as AM measures, reflected in EOP) via the makeup water from condensate tank (MUWC) system took place as well from 00:00 on March 12. The RCIC was manually isolated at 04:58 on March 12 due to RCIC turbine drive steam pressure drop accompanying reactor depressurization. Reactor water level was adjusted with alternate water injection via MUWCs from that point onward, as shown in Fig. 4.19. Since S/C water temperature rose above 100°C at 05:22 on March 12 due to RCIC operation and opening of SRV, the site superintendent deemed the situation to be one falling under Article 15 of the Nuclear Emergency Act (loss of pressure suppression function). The flammability control system cooler began using the S/C cooling water drain line to inject cooling water (MUWCs) into the S/C. This took place from 06:20 of March 12 onward. At the same time, D/W spray (from 07:10 of the same day) and S/C spray (from 07:37 of the same day) were carried out as needed to cool the PCV. D/W and S/C spray via MUWCs were introduced as AM measures, and were reflected in the EOP. This emergency mitigation action suppressed the containment vessel temperature and pressure rise and ensured the recovery action time of the residual heat removal (RHR) system. Fig. 4.20 shows reactor cooling trends, securing uninterrupted water injection throughout the depressurization process with RCIC at high-pressure condition and MUWC at low-pressure condition was a critical factor for successful core cooling, as shown in Figs. 4.19–4.21.

352

Boiling Water Reactors

Fig. 4.19 Reactor cooling water injection and PCV cooling by RCIC and MUWC [11].

Fig. 4.20 Reactor cooling trends at Fukushima II-1 using RCIC and MUWC [11].

Fig. 4.21 Effect of alternative water injection measures employing NUWC [11].

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Since PCV pressure showed signs of rising due to loss of reactor heat removal function and reactor RHR function restoration was predicted to take some time, configuration of a line for PCV pressure resistance venting (one action left to open valve on the side of the S/C took place from 10:21 to 18:30 on March 12). This differed from the PCV pressure resistance venting after core damage AM measures. In this case, if reactor RHR function restoration is delayed, a line is assembled in advance to lower rising PCV pressure by continuing reactor cooling water injection to maintain core soundness while releasing steam into the atmosphere via S/C pool (same as other units). Since PCV pressure did not reach levels requiring PCV pressure resistance venting, this was not performed.

4.6.2.3 RHR restoration and reactor cold shutdown The ERC at the power station planned to check equipment damage status via field checks to be performed alongside postearthquake/tsunami response. This would allow the formation of restoration strategy and prioritization of work. However, the restoration unit could not be immediately sent into the field for various reasons. These include lack of field lighting; danger posed by large amounts of debris and sinkholes; continued aftershocks; tsunami alert still in effect; the standby/evacuation notification paging system being unusable during tsunami arrival; and mobile phones being unusable within buildings damaged by tsunami. Standby/evacuation procedures for personnel (e.g., messengers) distribution were stipulated and safety equipment prepared. When all this was completed, the restoration unit began field damage checks of the heat exchanger building, which was near the ocean. This was around 22:00 on March 11. Based on restoration unit field check results, the ERC at the power station decided on the policy that prioritized inspection/maintenance of various equipments within the heat exchanger building. These included RHR component cooling system pumps of system D (D), RHR seawater system pumps (B), and EDG cooling system pumps (B) (for RHR component cooling system pumps (D) and EDG cooling system pumps (B), motor was replaced). Kashiwazaki-Kariwa NPS was commissioned to perform emergency motor procurement at the same time. The Kashiwazaki-Kariwa NPS proactively performed support for Fukushima Daiichi and Daini NPSs during this disaster (e.g., procuring needed materials and equipment). The power panel that powered the motors for these pumps lost its function due to water damage. Therefore, the ERC at the headquarters was commissioned to perform emergency equipment (e.g., high-voltage power supply cars, mobile transformers, and cables) procurement by the ERC at the power station. These would connect power panels unaffected by tsunami and high-voltage power supply cars to motors. Of the usable power panels unaffected by tsunami, those at the rad-waste building were chosen for use. There were several reasons why this panel was chosen by the recovery team (all based on field status), despite said building being farthest from the heat exchanger building. These were fewest complex indoor cable layings; majority of installation routes follow aboveground straight roads; most compatible with manual installation of hard and heavy power cables in a short amount of time.

354

Boiling Water Reactors

Materials and equipment commissioned for procurement by the ERC at the headquarters and Kashiwazaki-Kariwa NPS gradually arrived at Fukushima Daini NPS by 06:00 on March 13. Transportation of these took longer than expected due to several factors. Those are included road status worsening due to disasters, and mobile phones used for the communication between transportation team and the ERC at the power station not working. As shown in Fig. 4.22, the total length of temporarily installed cables at all four plants was approximately 9 km. Installation of those cables was completed by 23:30 of March 13. That was accomplished by 200 personnel, comprised of employees (including those sent for support from the distribution department) and contractor workers. Cable installation work was first carried out at Unit 2, since its PCV pressure increase was fastest. This was the important data connection of PCV pressure monitoring and to use prediction for emergency response center (ERC) and the power station engineering team. However, since Unit 1 PCV pressure increase became faster than that of Unit 2 in the early hours of March 13, Unit 1 was given priority. Although later event progression showed that the Unit 1 PCV pressure increase was faster, the change in priority meant that restoration could be completed without requiring Unit 1 PCV venting, thus, allowing successful cold shutdown. Alongside cable installation, pump component status checks and motor installation were performed. Each pump was activated as soon as their preparations were completed, starting at 20:17 of March 13. Due to the activation of the RHR cooling system pumps of B system, shown such as (B), the site superintendent deemed the situation to have recovered from one to which Article 10 of the Nuclear Emergency Act applied (loss of reactor heat removal function) at 01:24 of March 14. The RHR cooling system pumps (B) were used for S/C cooling, which resulted in a gradual decrease of S/C water temperature. Since S/C water temperature dropped below 100°C at 10:15 on March 14, the site superintendent deemed the situation to be one that recovered from one to which Article 15 of the Nuclear Emergency Act applied (loss of pressure suppression function). The RHR cooling system pumps (B) were used to begin injection of S/C water into the reactor via low-pressure injection line at 10:05 on March 14. At the same time, emergency cooling was performed via a circulation line. Here, reactor water was sent to the S/C via SRV, where S/C water would be cooled by the RHR heat exchanger (B), before being injected into the reactor again via low-pressure injection line (S/C) ! RHR cooling system pumps (B) ! RHR heat exchanger (B) ! low-pressure injection line ! reactor ! SRV ! (S/C). These actions aimed toward early supply of reactor coolant alongside cooling by S/C. As a result, reactor coolant temperature dropped below 100°C at 17:00 of the same day, and it was confirmed that the reactor had entered cold shutdown. Since the signs of cooling function was also temporarily lost for the spent fuel pool (SFPs) at Fukushima Daini Units 1–4, but the limiting conditions of operation stipulated in the Technical Standards for Nuclear Reactor Facility (SFP water level: near overflow water level, water temp.: under 65°C) were able to be satisfied.

Fig. 4.22 Temporary cable installation work for emergency power supply [11].

356

Boiling Water Reactors Reactor Building Condensate Storage Tank

Sea

Heat Exchanger Building

Fig. 4.23 Cold shutdown achieved after emergency restoration [11].

Hydrogen concentration increase (hydrogen: approximately 5%, oxygen: approximately 2%) were seen with the increase of via gamma rays (CAMS) at 05:12 on March 16 (approximately 2 days after cold shutdown), the flammability control system was operated. This suppressed hydrogen/oxygen concentration below the flammable range (Fig. 4.23).

4.6.2.4 Continuous ERC planning activities ERC established immediately after the earthquake, as shown in Fig. 4.24. Organization of emergency response units maintained accountability by setting clear goals and reporting/visualizing situation to ERC on a constant basis. ERC personnel with operation background were dispatched to MCR. This allowed sift operators focus on operation and supervision, while maintaining communication between MCR and ERC through dispatched ERC personnel. Site ERC members had to stay at their posts as there weren’t any backup members. They did not leave the site to meet their families until April, devoting themselves to restoration activities, even in the case where their family were suffered in disaster. Due to the damage caused by the tsunami, the system that removed heat from the reactor completely lost its function. As an emergency measure until restoration, water was injected into the reactor using alternative equipment that was not the original emergency equipment. They urgently procured materials for the restoration of heat removal equipment, and urgently restored the functions with temporary equipment, as shown in Fig. 4.25. At the Fukushima Daini NPS, the plant was successfully shutdown by the 5th day after the disaster without damaging the fuel in the cores and without the need to vent

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Fig. 4.24 ERC established immediately after earthquake [11].

the containment vessel. This was the first important action to success reactor cold shutdown by transport the new pump’s motors and recovery of the seawater cooling pump system. It means that it led to the successful restoration of RHR system and attained cold shutdown of the reactor. To strengthening emergency management, it was clarified that the strategic planning function was one of the second important success factors in the Fukushima Daini Power Station, but rather more important than the strategy itself. Kawamura emphasized the importance of ICS. ICS is about how emergency organizations can function and act in response to changing circumstances in situations where it is not always clear how far the situation will expand due to natural disasters such as forest fires and hurricanes in the United States. It is an emergency management system (EMS) created from the viewpoint, and has been improved by reflecting the actual disaster response experience, including the response to Hurricane Katrina [11]. Focusing on the incident management system (IMS), Kawamura applied ICS to nuclear power accident emergency management system (EMS) to avoid severe accident, based on the critical success factors and issues extracted in the Fukushima Daini analysis. Under the ICS, he devised concrete measures such as strengthening the strategic planning function, common operation picture, and strategic design by the phased approach. Each of the above research results has an interrelationship as shown in Fig. 4.25, and it is thought that this will make resilience through collaboration between hardware and software effective and strengthen the fourth layer of defense in depth. Lessons learned from this experience were identified to improve emergency management, especially in the areas of strategic response planning, logistics, and functions supporting response activities continuing over a long period. It was found that continuous planning activities reflecting information from plant parameters and response

358

Boiling Water Reactors

kPa(abs)

Unit 1 PCV pressure trend

Unit 1

Rupture Disc Break 310kPa

S/P (kPa) S/P D/W

D/W (kPa)

Estimated time of Failure

(c) Plant parameters

(d) Increases in PCV pressure predict reaching limit

Fig. 4.25 Planning operation and restoration at ERC using ICS [11].

action results were important, and that relevant functions in emergency response organizations should be integrated. Logistics were handled successfully but many difficulties were experienced. In the conventional ERC group system, 12 functional groups were organized in parallel under the general manager, as shown in Fig. 4.24. In such system where each group develops quick activities in parallel according to a predetermined procedure, it was difficult to deal with situations beyond expectations, and the command system members were confused at the Fukushima Daiichi Power Station. Therefore, it is necessary to have a system that can flexibly respond to changes in the situation. Inspection of pumps for RHR cooling systems (RHRC: circulation pump, RHRS: seawater pump) and emergency engine cooling water (EECW) pump were

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important. Motors were replaced for pumps in RHRC and EECW. In order to restore the inundated electrical buses, temporary cable and high-voltage mobile power vehicles were deployed. Temporary cable was laid from survived power cubicles in the rad-waste building and Unit 3 heat exchanger building, as shown in Figs. 4.22 and 4.26. RCIC and MUWC pumps were used for core cooling injection water. And finally RHR system was activated by the cable connection to seawater pumps, decay heat was transferred to seawater, Units 1–4, as shown in Figs. 4.27 and 4.28, respectively. This means recovery of the ultimate heat sink from PCV to sea. The strategic planning function itself is important, not the prepared response strategy. At Fukushima Daini Power Station, they started responding under limited information, raised awareness of the situation through the system response, and gradually revised the strategy to respond to the situation. In the conventional ERC group system, this function is not clear. By analyzing Fukushima Diani’s response, it was also found that there are many issues in logistics activities and activities that support long-term emergency activities. It is the function that has not been clearly defined in the emergency response organization so far, and it is necessary to position it as an important function and strengthen it. In establishing these functions, Kawamura recommended [10] using the incident command system (ICS), which has been used for responding to natural disasters in the United States, for responding to nuclear accidents, as shown in Fig. 4.29.

Fig. 4.26 EOP for cable connection to activate RHR systems [11].

360

Boiling Water Reactors

Fig. 4.27 Success path to cold shutdown of Units 1, 2, and 4 [11].

Fig. 4.28 Success path to cold shutdown of Units 3 [11].

The second lesson from the Fukushima accident was that field activities were restricted for a certain period of the initial action. The proposed strategic planning based on a three-phase approach is as follows: Phase 1: Response strategy that can be carried out in a short time with permanent equipment (Estimated period: up to 12 h after the accident)

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Fig. 4.29 Strategic planning based on a three-phase approach [11].

Phase 2: Strategy to add means for ensuring safety while proceeding with restoration using portable equipment and materials on site (estimated period: up to 7 days after the accident) Phase 3: Strategy to introduce human and physical support from outside the company to ensure the continuity of safety assurance (estimated period: long term after 7 days from the accident) While clearly positioning the strategic planning function, other main functions were consolidated into the on-site operation function, external cooperation function, and activity support function, and a responsible organization was established. Considering the importance of cooperation with external organizations in the event of a nuclear disaster, the external cooperation function, which was the staff of the general manager in the original ICS, was organized as an important independent function, as shown in Fig. 4.30. In addition, the logistics function is shared by the head office countermeasures organization.

Fig. 4.30 Functional structure of an improved site emergency response organization with ICS concept applied [11].

362

Boiling Water Reactors

As a lesson learned from the Fukushima Daini Nuclear Power Station, it is important to collect and display information, share it, analyze it, and formulate strategies. Therefore, it was decided to appoint a person in charge of strategic planning to comprehensively carry out information analysis and strategy planning. It was devised to standardize information on emergency response resources (plant equipment, emergency equipment, etc.) as a common operation picture and share it on paper and electronic media, as shown in Fig. 4.31.

4.7

Lessons learned from Fukushima Daiichi accident

4.7.1 Causes of severe accidents and countermeasures NISA ordered that the licensees make a new independent vent line for filtered vent system. This is one of the lessons learned. There were no accident reports about the Fukushima Daiichi accident pointing out the vent system fault and potential risks. The causes of severe accidents and countermeasures are shown in Fig. 4.32. In the figure, (P) means protection and (R) resilience action. Corium cooling was very effective in achieving the cold shutdown cooling, even after the containment failure at the Fukushima Daiichi NPS.

4.7.2 Measures for severe accidents installed in the United States and European NPPs There are many good practices of countermeasures to prevent FP release in the world. Based on the DiD concept (Fig. 4.33), essential safety features were incorporated in the third layer for design basis accident (DBA) and prevention of simultaneous loss of all safety functions owing to common causes, such as a tsunamis. Mobile safety features for the fourth layer such as mobile fire pumps should be deployed for core and containment cooling or corium cooling (Fig. 4.34) [12].

4.7.3 Filtered containment venting system As shown in Fig. 4.35, after the Three Mile Island (TMI) Unit 2 and Chernobyl Unit 4 accidents, countries such as France, Germany, Switzerland, Finland, and Sweden decided to install filtered containment venting systems (FCVS) to protect against radioactive material exhaust (Figs. 4.36 and 4.37) [13]. Fig. 4.38 shows a schematic diagram of the FCVS installed in the Leibstadt NPP. Venting is automatically initiated when CV pressure reaches the pressure set for the rupture disk. An operator who wishes to vent early can easily open the vent valve using a hand wheel drive shaft [13]. In the Fukushima Daiichi NPP accidents, operators should have closed numerous valves in the SGTS system and then opened the vent valve with an air compressor and connecting tubes, because of the station blackout condition. If a FCVS had been installed in the Fukushima Daiichi NPPs, environmental contamination by FP could have been avoided. The decontamination factor is about 1000 for aerosols and 100 for I2.

CRD

Unit 7 14:10 Aug. 25

RPV

External Power

Remarks

Remarks

MSSRV RPV Press.

14:20 Receive power

GTG

RPV Level

Mobile Power

Remarks

PCV

Lost

LOCA D/W Press. S/P Press. D/W Temp. S/P Temp.

Emergency Power

Mobile M/C

Remarks

14:20 Receive power

Water Injection DEC

High Pressure Water Injection Remarks

Power DEC

Remarks

Lost Remarks

SFP Level

Plant Output Scrum Time of Scrum

Filtered water D/D Fire Pump

CSP

Remarks

Tripped Remarks

Remarks

CSP

MUWC Pump

CSP

CSP Filtered water

Low Pressure Water Injection

Remarks

Time of TAF PCV Vent

Fire Tank

Remarks

Fire Engine

S/P

S/P

14:02 Arrival of water injection team

Sea Water

Remarks

S/P

Pressure dropped Possibility of line break Remarks

Fire Engine

Remarks

Remarks

Filtered water Pure water tank Fire Tank Reservoir

Failure

Started

Standby

Fig. 4.31 Example of common operation picture for information sharing among responders in nuclear emergency [11].

Not Clear

Fig. 4.32 Causes of severe accidents and countermeasures [2].

Fig. 4.33 Concept of “defense in depth” to terminate accidents [2].

Fig. 4.34 Mobile safety features for severe accidents in DiD fourth layer [12].

Fig. 4.35 Filtered containment venting system [2,13].

Fig. 4.36 FCVS installed in Chooz NPP (PWR), France [2,13].

Fig. 4.37 FCVS in Leibstadt NPP (BWR), Switzerland [2,13].

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Boiling Water Reactors

Fig. 4.38 Schematic diagram of FCVS in Leibstadt NPP [2,13].

■ After the TMI-2 accidents, KKL back-fitted the DiD3 (additional C/V cooling) and DiD4 (mitigation of Sever Accident). DiD: Defense in Depth Steam

Heat Exchanger

Fuel rod

Grand water 0.0m SG SHER-Bunker

Suppression Pool

D/G

D/G

Two D/G for SEHR

River Grand water -42.0m

Fig. 4.39 Special emergency heat removal (SEHR) System [2,13].

After the TMI-2 accident in 1979, Kernkraftwerk Leibstadt (KKL) backfitted the Leibstadt NPP with additional CV cooling (DiD 3) and a mitigation system for severe accidents (DiD 4). The backfitted system was called the special emergency heat removal (SEHR) system, as shown in Fig. 4.39. This system was required by the Swiss

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regulatory body of Federal Nuclear Safety Inspectorate (ENSI) and Swiss Federal Office of Energy (HSK) in the late 1970s, shortly after the start of project planning, so it was the first backfitting in the present design of KKL (Kernkraftwerk Leibstadt AG).

4.7.4 Special emergency heat removal system Fig. 4.39 shows special emergency heat removal (SEHR). System has two trains of heat removal system. The system was installed to remove a minimum of 36.3 MW (estimated decay heat: 1% of nominal power). The system has two special EDGs and a huge underground well for water heat sink. The system is able to cool both the core and the CV, using the heat exchanger [13].

4.7.5 Tsunami protection When the Fukushima Daiichi NPS was attacked by the tsunami, all AC and DC power was lost because of damage to the EDGs, power center, metal clad switchgear, and seawater pump motors. At the Fukushima Daini NPS, AC power was able to restore seawater pumps by changing power cable and installing new pump motors. Therefore, it is very important to prevent the seawater flow into important areas. As shown in Fig. 4.40, at the Diablo Canyon NPS in California, United States, the seawater pump motors are equipped with waterproof hatch-type doors and snorkel air ventilation piping for pump motor cooling.

Fig. 4.40 Tsunami protection at Diablo Canyon NPS, United States [2,13].

368

4.8

Boiling Water Reactors

New nuclear regulatory requirements in Japan

4.8.1 New nuclear regulatory requirements A new nuclear regulatory body, the Nuclear Regulation Authority (NRA) was established on September 19, 2012. The NRA performed a complete review of safety guidelines and regulatory requirements [14]. On July 8, 2013, new regulatory requirements for commercial power reactors came into force. The requirements stipulate that all Japanese utilities conform to the regulatory requirement before restarting NPPs. Design requirements must treat natural phenomena, such as volcanoes, tornados, and forest wildfires (Fig. 4.41). The new regulatory guidelines (Fig. 4.42) require direct deployment of mobile power, mobile pumps, fire engine, and installation of tsunami protection. Design requirements should be prepared to protect against cable fire between the reactor building and main control room, and against internal inundation by waterproof areas of important safety components and systems. New design requirements for severe accident require measures such as mobile powers, mobile pumps, fire engines, and water tanks or reservoir to protect core damage and to protect containment vessel failure. A bunker-type underground building (Fig. 4.43) should be constructed to control reactor cold shutdown against intentional attack by an air craft, within a 5-years grace period after the approved of construction permission.

4.8.2 Tsunami protection examples Tsunami protection is very important in preventing to protect against severe accident such as what occurred at Fukushima Daiichi NPS. Fig. 4.44 shows tsunami protection concept based on the DiD. Embankment and seawalls are the primary tsunami

Fig. 4.41 Regulatory requirements comparison before and after Fukushima [2,14].

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Fig. 4.42 Enforcement of new regulatory requirements, from July 2013 [2,15].

Mobile equipment And power supply Car

Measures mainly For example, 100m distance using mobile Connector

equipment

R/B

Specialized safety facility Core Containment

Aux. Build.

Connector

Turbine Build.

Mobile equipment And power supply Car above 30 m sea

*System configuration is an example (Specialized safety facility)

Fig. 4.43 Measures against intentional aircraft crashes [2,14].

FCVS

370

Boiling Water Reactors

Reactor Building

Seawater Pump

Fig. 4.44 Tsunami protection concept based on the defense in depth [2,10].

Fig. 4.45 Tsunami protection measures at Tomari NPS (PWR) [2,10].

protection, waterproof walls and doors at entrances to reactor building are second, waterproof doors for EDG and pump rooms are third, and mobile gas-turbine cars on hill are fourth. Fig. 4.45 shows tsunami protection measures at Tomari NPS of Hokkaido Electric Power Company, including a large mobile gas-turbine generator and steam generator

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(SG) feedwater pump in a parking area of 31 m height from sea level. Tsunami-proof doors and balconies have been installed in the reactor and auxiliary buildings. Turbinedriven feed pumps for SG are protected against tsunami of 25 m height. The power center (P/C) in the main control room is protected against those of 20 m height. Fig. 4.46 shows tsunami protection measures at Hamaoka NPS of Chubu Electric Power Company. A large tsunami wall of 22 m height and 1.6 km length has already been constructed. There is a snorkel building for EWS pumps and a 4000 kVA (3.2 MW) gas-turbine generator at 25 m height. A cross-sectional view of tsunami measures and water reservoir is shown in Fig. 4.47. Fig. 4.48 shows examples of tsunami protection measures at NPSs, such as gasturbine generators, an oil tank, and snorkel building at Shimane NPS. At Kashiwazaki-Kariwa NPS, there is a large wall, door and balconies, mobile cooling car, and fire engines. Table 4.3 shows typical examples of measures that meet requirements at Shimane and Kashiwazaki-Kariwa NPS. Typical examples meeting requirements are shown in Table 4.3.

(a) Tsunami protection wall (22m x 1.6km)

(b) Complete tsunami protection wall construction

Tsunami protection door Tsunami protection wall

Seawater intake pump

Snorkel building for EWS pumps

Drain pump

(c) Cross sectional view of tsunami measures

Fig. 4.46 Tsunami protection measures at Hamaoka NPS (BWR) [17].

Watertight doors

GTG

(a) Snorkel building for EWS pumps

GTG

TP+40m

6.9kV

(b) 3.2 MW gas turbine generator installed at 40 m height.

Tsunami protection door (Approx. 1m thick)

Watertight door door Watertight

(c) Cross-sectional view of tsunami measures

(d) Emergency water storage tank Fig. 4.47 Tsunami protection measures under construction at Hamaoka NPS [17].

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Fig. 4.48 Examples of tsunami protection measures at NPSs [2].

4.8.3 Tornado protection examples Fig. 4.49 shows a tornado evaluation method and measures for seawater pumps. Tornado wind streamlines were calculated using a model designed by Dr. Fujita [18]. Based on the wind field, a missile analysis was conducted for both ground level and 40 m height in particular. Final missile speed was used to estimate its kinetic energy for shield strength to protect a seawater pump, reactor building, or other safety-related components.

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Boiling Water Reactors

Table 4.3 Typical examples meeting requirements [2]. Purpose

New requirement

Protection against flooding

A plant shall be able to withstand against design tsunami Supply of electricity

Ensuring redundancy of power supply

Ensuring redundancy of core cooling

Cooling function under high reactor pressure Depressurization function of reactor Cooling function under low reactor pressure

Reinforcement of safety measures applicable during severe accidents

4.9

Ultimate heat sink for prevention of severe accidents Prevention function for containment break due to excessive pressure Emergency response center

Typical examples which meet the requirements Installation of protection walls and doors against design tsunami Deployment of permanently installed or portable backup AC power supply, enhancement of permanently installed DC power supply, deployment of portable DC power supply, etc. Deployment of buttery for RCIC control, preparation of operation procedure, training, etc. Deployment of buttery for ADS actuation, preparation of operation procedure, training, etc. Permanently installed coolant injection equipment, portable coolant injection equipment, preparation of operation procedure, training, etc. Deployment of vehicle for backup ultimate heat sink, preparation of operation procedure, training, etc. Installation of filtered containment venting system, preparation of operation procedure, training, etc. Ensuring earthquake and tsunamiresistant emergency response center, radiation protection, logistics, etc.

Example of compliance with new regulatory standards for PWRs that can be used as a reference for BWRs

In September 2014, Sendai NPS Units 1 and 2 (Kyushu Electric Power Company) received NRA permission to upgrade their safety systems as required by the postFukushima regulations. This was followed by a construction permit and detailed design review by the NRA. Fig. 4.50 shows the complete cooling strategy for station blackouts (SBOs) to protect against severe accident (SA) for PWR. Fuels in the core are cooled by steam generators (SGs). A water supply for the secondary side of SG is

Fukushima Daiichi nuclear power plant accident and analysis evaluation

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A schematic diagram of DBT-77. In this simplified model, the core is divided into inner and outer portions. Vertical motions are concentrated inside the outer core while the inner core is assumed to rotate like solid discs stacked up into a cylinder.

(a) Tornado wind field (Dr. Fujita)

(b) Tornado streamlines from Dr. Fujita model

Height =40m

Grand

(c) Tornado trajectories for missile analysis

(d) Missile shield for sea water pump

Fig. 4.49 Tornado evaluation method and measures for seawater pumps [18].

used to cool the primary side of a U tube. The core is cooled by natural recirculation in the primary loop driven by gravity. In the inverted U tube, the outlet cold side head is larger than that of the inlet hot side. Coolant injection to the primary and secondary loops using ECCS pumps driven by diesel power supply car and containment vessel (CV) spray is possible, even if under SBO. Spent fuel water injection is possible because of easy access routes for fire engines. Fig. 4.51 shows an additional emergency water injection pump installed on seismic base isolation rubber on the top floor of a turbine building. Discharge piping is via a flexible bellows pipe. The pump can supply water to the core or containment spray as a severe accident resilience action. Personnel in the Genkai NPS are trained to be able to connect the flexible pipe within a very short time. There are numerous safety reinforcement measures at Genkai Units 3 and 4 (Fig. 4.52). It is very important to prevent CV failure from overpressure, over temperature, and hydrogen detonation. For this purpose, CV recirculation cooler and spray, passive autocatalytic recombiner (PAR), and igniter for hydrogen combustion with oxygen in CV are available. A water cannon to suppress radioactive material diffusion has already been deployed to meet requirements of intentional aircraft impact or

Primary Containment Vessel Supply water to make up for lowered water level

PCV

Fire Hydrant Water supply

Spent Fuel Pit

Fire Engine Cooler Cooling Pump (2 Units)

Fig. 4.50 Full cooling strategy for SBO to protect against SAs for PWR [15].

Fukushima Daiichi nuclear power plant accident and analysis evaluation

Fig. 4.51 Water injection pump [2].

Fig. 4.52 Safety reinforcement measures at Genkai Units 3 and 4 for SBO [2].

377

378

Boiling Water Reactors

Fig. 4.53 Mobile pumps deployed in Ikata NPS [2,16].

tornado-driven missiles on the reactor building and CV. These measures are based on the DiD concept and use diverse strategies. As shown in Fig. 4.53, many mobile pumps driven by diesel/gasoline engines are deployed at Ikata NPS. Such devices have already been deployed at all NPSs in Japan.

4.10

BWR NPS to be reviewed for new requirements or restarting

Following the design reviews of 12 PWRs, 9 BWRs with enhanced safety measures were in ongoing design review for restarting. As shown in Fig. 4.54, by using seawater cooling system, S/P water in S/C will be cooled and core injection system can remove decay heat from the core to the pool via a main steam SRV. The system can be operated using mobile generator and heat exchanger cars, even for a natural disaster such as a major earthquake or tsunami, or sudden flooding. It is very important to cool the S/P water by the connected mobile cooling car using plate-fin heat exchanger cooled by pumped-up seawater to maintain ultimate heat sink. Upon occurrence of a SA, filtered containment venting system (FCVS) was installed and vent gas with radioactive fission products is blown into the scrubbing pool through numerous venturi nozzles, as shown in Fig. 4.55. Mist in steam moves upward to a metal fiber filter through a multihole baffle plate. After the mist is removed by that filter, radioactive methyl iodine (CH3I) is captured on the surface of a molecular sieve or AgX, made from zeolite particles with silver coating [19,20]. Fig. 4.56 shows the FCVS pit at Hamaoka NPS of Chubu Electric Power Company and installing FCVS at Kashiwazaki-Kariwa NPS of TEPCO, respectively. Fig. 4.57 shows the FCVS visualization test facility at the Hokkaido University. An AgX filter is used at the downstream of the scrubbing pool and the metal fiver filter. The thickness of the AgX filter is very important parameter to obtain enough decontamination factor (DF). As shown in Table 4.4, a confirmation test was conducted in

Fukushima Daiichi nuclear power plant accident and analysis evaluation

Fig. 4.54 Core injection heat removal by S/P cooling system [2].

(a) Aircraft landing zone Metal Filter

Molecular Sieve / AgX Vent Gas Inlet Baffle Plate Vent Gas

Vent Gas Outlet Benturi Nozzle

(b) Internal structure of vent filter

(c) Outline and effect of FCVS

Fig. 4.55 Filtered containment venting system (FCVS) with silver zeolite [2,19].

379

380

Boiling Water Reactors

Fig. 4.56 Installation of FCVS [2,19].

Fig. 4.57 Development of high DF FCVS using silver zeolite at the Hokkaido University [2,19].

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Table 4.4 Test result of adsorption efficiency of CH3I under various bed depth [20]. Bed depth (mm)

Residence time (s)

Adsorption efficiency of CH3I (%)

50 75 100

0.246 0.369 0.492

99.967 >99.999 >99.999

Types of silver zeolite

AgX (Silver doped X-type zeolite)

Testing conditions: temperature: 130°C, pressure: 399 kPa, LV: 20 cm/s; relative humidity (RH): 95%, CH3I concentration: 1.75 mg/m3 (I-131).

€ inspection system for safety standards, the DF for the radioactive the German TUV iodine exceeds 10,000 at the bed depth (AgX filter thickness) greater than 75 mm [19]. Fig. 4.58 shows full cooling strategy for SBO to protect against SA for BWR [19]. Fuels in the core are cooled by direct water injection via ECCS line of feedwater line. Core injection, PCV spray, PCV head flange cooling, suppression pool cooling, and pedestal water injection using MUCW pumps are driven by the gas-turbine generator, the mobile power supply car, and the mobile ultimate heat sink car. Long-term evaporative cooling of feed and bleed is possible with a variety of water injection methods and filter vents.

Fig. 4.58 Full cooling strategy for SBO to protect BWR from SA [2,19].

382

Boiling Water Reactors

Steam turbine Water lubricant Pump

RCIC Pump Steam

1.8m

1.8m

TWL RCIC Pump

Water

Fig. 4.59 TWL-RCIC using a steam turbine-driven water-lubricated pump [2,15].

Fig. 4.59 shows a new high-pressure auxiliary core cooling (HPAC) system using turbine-driven water-lubricated (TWL) pump. The pump is compact and has a head of 900 m under 183 m3/h, and it also an oil-free water-lubricated pump. The required DC power supply is about half that of current RCIC pumps. These new types of high-pressure water injection devices will provide substantial diversity under SBO conditions. Some BWR utilities will install the TWL-RCIC pump. Research and development items in progress to improve safety and resolve problems are shown in Table 4.5. Table 4.5 R&D items in progress to improve safety and resolve problems [2,15]. Measurement value

Problems

Reactor water level

l

l

Dry well water level

l

Review items (BWR)

The water level could not be obtained due to vaporization of water in reference legs Multiplexing was ineffective because the same problem happened in all the places

l

The water level could not be obtained because the measurement

l

l

Review items (PWR)

Procedures to prevent vaporization of water and refill water in reference legs Multiplexing with different types of measuring methods

l

A measuring method to obtain the water level at every height

l

l

Reference legs are sealed in existing differential pressure type water level gauges Water level can be estimated from the temperature on the outlet of the reactor core The water level of the recirculation sump can be monitored.

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Table 4.5 Continued Measurement value

Problems

Review items (BWR)

points were limited

Dry well hydrogen concentration

Reactor building hydrogen concentration

l

l

Review items (PWR)

down to the bottom of a dry well

The measuring method was the sampling method that did not work due to loss of power and coolant

l

There was no hydrogen concentration meter in reactor buildings

l

A measuring method without sampling

The environmental effect should be investigated l

l

A measuring method for hydrogen concentration in reactor buildings

l

Installation of hydrogen concentration meters in containment vessels A measuring method without sampling Installation of hydrogen concentration meters in annulus

As shown in Fig. 4.60, probabilistic risk assessment (PRA) was carried out reflecting such safety measures for Hamaoka NPS. It was confirmed that the core damage frequency was reduced by about 3 orders of magnitude compared to before the measures; 2.9  105 before the countermeasure and 3.8  109 after the countermeasure. This is a provisional value, and the loss of decay heat removal function which was the highest risk among them can be easily moved to low-pressure core injection of mobile pump. If several SRV are opened, then the steam from the RPV will be discharged into the S/P via SRV, the decay heat of the fuel in the core is discharged to the S/P. However, when the S/P reaches the saturation temperature, the PCV pressure rises, leading to the loss of pressure suppression function of the containment vessel. Therefore, if FCVS is installed, the steam in the containment vessel can be vented to the atmosphere, which is UHS. The radioactive materials are removed via FCVS. In other words, the decay heat from fuels in the core removed with low-pressure feedwater injection by a mobile pump and discharged it in the atmosphere. This is called “Feed and Bleed” method to remove decay heat from RPV core to the atmosphere. Long-term evaporative cooling of feed and bleed is possible with a variety of water injection methods and FCVS reduce the risk of NPP dramatically.

Core Damage Frequency (Reactor / year)

384

Boiling Water Reactors Alternate Alternate Gas Alternate Low-Press. turbine Depress. Injection Power and System System 24H Battery

10-8

Alternate Shutdown Hamaoka NPP Total CDF: 1/1000 CRD 3.8X10-9 System

FCVS

10-3

Before Improve Safety

10-3

After Improve Safety

10-4

10-3 10-3

Loss of Interface Loss Loss of all Loss of Loss of water ATWS High/Low Ultimate System AC Power HPCI/ADS injection Press. Heat LOCA (TQUX) Injection (TB) for LOCA Sink (TQUV) (TW)

Total CDF

Fig. 4.60 The core damage frequency was reduced by about 3 orders of magnitude compared to before the measures [17].

As a result, the core damage frequency of loss of decay heat sequence called “TW” reduced almost 1/1000, and that of loss of all AC power called “TB” reduced by 108 due to the large gas-turbine power supply inside the seismic isolation building, the total core damage frequency decreased from 2.9  106/reactor-year to 3.8  109/reactor-year. Fig. 4.61 shows an installed corium shield as a partial core catcher to prevent the inflow of melted fuel into the D/W sump at Kashiwazaki-Kariwa NPS Units 6 and 7. The corium shield was made of zirconia (ZrO2), which has a melting point of 2700°C and is a very tough and heat-resistant ceramic. This zirconia will build a dike on the pedestal floor in the PCV to prevent the core melt from flowing into the drain sump in the unlikely event of a severe accident. As shown in Fig. 4.61, molten debris fell on the pedestal floor at Unit 2 of the Fukushima Daiichi NPS. Thus, even if the fuels or stainless-steel structure in the reactor core melts and falls down, the height of the corium shield is selected so that it will not overflow. The drain sump is the weak point in the pedestal floor of PCV, because thickness of the D/W sump bottom concrete wall on the steel liner of the PCV is only about 20 cm. The drain sump is necessary to measure condensed water flow rate to detect leakage of pipes in the PCV. The corium shield was installed at the Kashiwazaki-Kariwa NPS Units 6 and 7, to prevent molten core debris flow into the drain sump and to prevent molten core concrete interaction (MCCI). The design concept of corium shield is a simple core catcher to be used to catch corium and cool it by injecting water.

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Fig. 4.61 Installed corium shield as a partial core catcher to prevent the inflow of melted fuel into the drywell sump at Kashiwazaki-Kariwa NPS Units 6 and 7 [13].

Fig. 4.62 shows the principle of water-level indication drift due to evaporation of water in the reference leg under the high-temperature PCV condition during severe accident. At the Fukushima Daiichi NPS, water-level gauge reading errors occurred at Units 1–3 during the accident, and the water level was displayed as high as 4 m at the maximum. At Unit 1, although the water level in the RPV fell below the TAF, it was displayed as if the water level was rising and kept the level at the height of suction nozzle of jet pump. As a result, almost all the people misunderstood and thought that there was still water in the fuel zone and the fuel was partially submerged. Operators misunderstood that the reactor was full due to the indication error of the water-level gauge, and stopped the ECCS during the severe accident at TMI-2 in the United States. There is a plant in the United States equipped with a backfill system that injects CRD pump discharge water (14 MPa) from outside the PCV via a needle valve into the instrumentation pipe that recovers the reference leg head below it, before March 11. However, in the event of the loss of all AC power, the CRD pump will stop, so in Japan, Hokkaido University, TEPCO and Hitachi GE decided to inject water from a small plunger pump (1 kW) or an accumulator pressurized with nitrogen

386

Boiling Water Reactors

Fig. 4.62 Principle of water-level indication drift due to evaporation of reference leg in hightemperature PCV during severe accident [16].

gas. In addition, a multipoint heated thermocouple (HTC) water-level gauge that measures the temperature of a metal piece heated by a heater with a thermocouple and whose temperature drops when submerged, and a new time-domain reflectometry (TDR) water-level gauge that measures the reflection time of an electric pulse using the conductivity of water have been developed. These devices are planned to be put in a neutron instrumentation tubes to check the actual water level to prevent severe accidents as a “voluntary safety measures,” as shown in Fig. 4.63. The effectiveness of backfill was tested at the Hokkaido University [16], and after heating to 400°C in a constant temperature chamber to evaporate reference water in the reference leg, water was injected with a small plunger pump, and the water-level gauge was recovered, as shown in Fig. 4.64. It was confirmed that the water hammer after cold water injection did not occur. Fig. 4.65 shows a new-type water-level gauge that applies TDR [16]. The measurement method utilizes the physical characteristic that the velocity of the reflected pulse differs from the velocity of the incident pulse depending on the presence or the absence of water because of conductivity difference. The TDR probe is mounted in the cover tube of the local output area neutron detector (LPRM), and it comes into direct contact with the coolant in the core from the TAF to the bottom of active fuel (BAF). It was confirmed under high-temperature and high-pressure conditions of 7 MPa that a wide range of water-level measurement is possible, and water-level measurement is possible even in the event of a severe accident (SA). The water-level meter by the TDR method can continuously measure the actual water level. The number of penetrations of electric signals for TDR is only a few. On the other hand, in the case of a multipoint thermocouple method, the water level can be measured only at a certain interval, and a lot of penetrations would be necessary for each thermocouple.

Fukushima Daiichi nuclear power plant accident and analysis evaluation

Fig. 4.63 Improved backfill system and new water-level indicator for severe accident protection [8].

Fig. 4.64 Improved back fill test facility [16].

387

388

Boiling Water Reactors

Steam\Probe Top

PI TAF

Water/Steam

Fuel Channel (Two Phase Flow)

Bypass (Water Gap) Channel

Probe Outer Conductor

Probe Center Conductor Probe Bottom\ Water

Lower Plenum

LP RM Cover Tube/Reactor Penetration Flange/Bolt Seal

MI Cable/Soft Cable Electrical Connector

Fig. 4.65 Principle of TDR water-level indicator [8].

Fukushima Daiichi nuclear power plant accident and analysis evaluation

4.11

389

Activities toward decommissioning Fukushima Daiichi

4.11.1 Current status of reactors at Units 1 through 4 Fig. 4.66 shows the status of the reactors in Units 1 through 4 as of 2020. It is assumed that the reactor cores of Units 1 through 3 are melted and that some portion dropped to the pedestal floor of the CV. It is very important to survey the melted core debris distribution from the core to pedestal floor. Some of debris may be attached on surface of the control rod drive outer casing. Table 4.6 shows the temperature and the water injection flow rate. Units 1 through 4 are in a stable and safe condition because of controlled water cooling. About 4 ton/h of water are injected into the cores of Units 1 through 3, and temperature in the RPV is maintained at around 30°C–40°C. Temperature in the spent fuel pool (SFP) is maintained around 30°C [15,21].

Fig. 4.66 Current status of reactors in Units 1 through 4 as of 2020 [21,22].

Table 4.6 Temperature and water injection flow rate of core in Units 1 through 4.

RPV bottom temp. PCV internal temp. Fuel pool temp. Reactor cooling water injection volume

Unit 1 (°C)

Unit 2 (°C)

Unit 3 (°C)

Unit 4 (°C)

About 29

About 36

About 34

–

About 29

About 37

About 34

–

About 26

About 22

About 21

About 22

About 4.6 m3/h

About 4.5 m3/h

About 4.3 m3/h

–

Boiling Water Reactors

Radiation dose rate (mSv/year)

390

0.6

1.7

0.5 0.4 0.3 0.2 0.1 0 2011

2012

2013

2014 (year)

2015

2016

2017

2018

Fig. 4.67 Annual radiation dose at site boundary of Fukushima Daiichi NPS [23].

Fig. 4.67 shows the annual radiation dose change derived from the measurement of radioactive materials released from Units 1 through 4 at the site boundary of Fukushima Daiichi NPS. At the time of the accident, vast amount of radioactive materials were released, but the maximum annual dose is now 0.03 mSv/year equivalent to 1/70th of nondetected zone. Fig. 4.68 shows that the doze in the approximate half area of the Fukushima Daiichi NPS is less than 10 μSv/h (87.7 mSv/year), but the dose in the reactor building is higher than 100 μSv/h (877 mSv/year).

Fig. 4.68 Exposed dose distribution at Fukushima Daiichi NPS [15,21].

Fukushima Daiichi nuclear power plant accident and analysis evaluation

391

Just after the accident, contaminated water that was accumulated in the reactor and turbine buildings leaked into the plant harbor through an underground tunnel. The leak was stanched at the end of October 2014. The current concentration of radioactive materials in the seawater near the harbor is below the WHO’s standard for drinking water (Fig. 4.69). Table 4.7 shows the mid- and long-term roadmap, developed by the government and TEPCO. The decommissioning work will be undertaken in three phases. Phase 1 is fuel removal from SFP, Phase 2 is fuel debris removal, and Phase 3 is plant dismantling. Fig. 4.70 shows the Status of Fukushima Daiichi NPS Units 1 through 4. Unit 1: After the hydrogen explosion on March 12, 2011. To remove the spent fuel from the SFP, rubble should be removed before spent fuel removal. To prevent dust scattering during rubble removal, work to cover the entire building is in progress. Bq/L 107 106 105 104 103 100 10 1 0.1 0.01 3/11 6/9 9/7 12/6 2011

3/5

6/3 9/1 11/30 2/28 5/29 8/27 11/25 2/23 5/24 8/22 2014 2013 2012

Fig. 4.69 Concentration of radioactive materials in seawater at north side of water outlets in Units 5 and 6 [15,21,23]. Table 4.7 Mid-and-long-term roadmap [21].

392

Boiling Water Reactors

Fig. 4.70 Status of Fukushima Daiichi NPS Units 1 through 4 [15,21,23].

Unit 2: Its reactor building is undamaged because the blowout panel was opened owing to the hydrogen explosion in Unit 1. Therefore, the dose rate in the reactor building is very high. Considering this fact, small hole will be drilled and a crane-type machine will be used to remove the rubble and spent fuels without dismantling the building. Thus, process for removing SFP fuels is under study. Unit 3: Rubble removal was completed in October 2013. Installation of the temporary cover was installed and spent fuel removed from the SFP by remote-controlled handling equipment, as shown in Figs. 4.71 and 4.72. The spent fuel removal completed in February 2021. Unit 4: A cover structure was constructed, and fuel-handling machine and heavyduty ceiling crane were installed, and spent fuel removal from Unit 4 was completely ended by the end of December 2014. The cover structure is supported by a huge cantilever structure, which enables to support the handling of the fuel into transport cask with the ceiling crane.

4.11.2 Finding contaminated water leak path for leak shutdown from PCV Fig. 4.73 shows the estimated leak path of fission products from the PCV. After recirculating water injection to the core, melted core debris was cooled and the reactor reached cold shutdown status. However, contaminated water still leaked from the

Fukushima Daiichi nuclear power plant accident and analysis evaluation

393

Fig. 4.71 Tensile truss and manipulator in fuel-handling machine used to remove small rubble and spend fuels [23].

Fig. 4.72 Lifting up deformed upper tie plate handle by the manipulator [23].

damaged PCV, and flowed out of the reactor building. Fig. 4.74 shows the survey robot to discover contaminated water leakage in Unit 1 [15,21]. Fig. 4.75 shows the effort in Unit 2 to find the water-level in PCV. A water-level probe with thermocouple was inserted through X-53 penetration. This found a water depth 300 mm. This level was near the overflow height into the vent pipe, which indicated that the leakage path might exist at the vent pipe or the suppression chamber torus. This leak can be stopped using balloon plugs, which are used for main

394

Boiling Water Reactors

Reactor Building Operation Floor

Turbine Building RPV TIP PCV

MSIV

Pedestal S/P

Sand cushion

Fig. 4.73 Estimated leak path of fission products from PCV [2].

streamline plugging during the annual maintenance of the main steam isolation valve (MSIV) [8]. As shown in Fig. 4.76, video camera was used to find PCV water leak a location in the reactor building at Unit 3. The camera was inserted from the second floor and discovered leakage from the expansion joint of main streamline pipe D from the PCV in the MSIV room on the rector building first floor.

4.11.3 Isolation of groundwater flow from contaminated water There are several trenches for pipelines and power lines between buildings and pumps in the seaside area. Extremely contaminated water was discovered there. It seemed that groundwater flowed into buildings and mixed with contaminated water in the buildings, and then the mixed water leaked out from the buildings to the ground and the trenches. The building area was isolated from groundwater by introducing the frozen wall in the ground as shown in Fig. 4.77. The wall was built by frozen soil (fundamental measure (2) in Fig. 4.77). The frozen wall system began operation in May 2014. With this system, the volume of the groundwater flow rate of subdrain pump reduces the water level in the ground to prevent the grand water flowing into the buildings was reduced from 350 to 80 ton/day after frozen operation. As a result, the volume of the contaminated water in the buildings was also greatly reduced. The next fundamental countermeasure is to build an impermeable wall to prevent contaminated water from flowing into the sea (fundamental measure (1) in Figs. 4.77 and 4.78). Further, the groundwater pumpup system around the building was introduced

Fukushima Daiichi nuclear power plant accident and analysis evaluation

395

Fig. 4.74 Survey robot to discover contaminated water leakage in Unit 1 [2,15,21].

(fundamental measure (3) in Fig. 4.77). Those worked to greatly diminish groundwater flowing into the buildings and contaminated water leaking from the buildings to the sea. The land-side frozen wall composed of frozen soil (ice wall) surrounds the buildings of Units 1 through 4 to block the groundwater from getting contaminated (Fig. 4.79). After a small-scale demonstration test was succeeded in, the large ice wall construction began. There are 1500 bores of 30 m depth at 1 m intervals for installation of coolant-circulating pipes.

396

Boiling Water Reactors

Monitoring results Height (OP. m)

33.7

11.920

33.7

10.690

33.5

Water depth 0.30m

8.100

33.5

OFF

6.430

33.5

OFF

6.230

33.6

OFF

6.030

35.0

OFF

35.8

ON

5.830 *0.35m from the bottom of PCV

5.630 *0.15m from the bottom of PCV

Fig. 4.75 Water-level checking at bottom of PCV of Unit 2 [2].

MSIV Room Piping penetration part

Water

Reactor Building 2 nd floor Air conditioner room

Seen from the PCV interior

Main steam pipe process radiation monitor

OP. 15.280m

Feedwater piping

Feedwater piping penetration part OP. 13.270

Containment Vessel water Level

Grating

OP. 10.261

Main steam piping

Main steam piping penetration part

MSIV piping penetration part Cross-section schematic diagram

Leakage location Reactor Building 1

st

floor MSIV (Cross- section)

Fig. 4.76 Investigation to find leakage location in Unit 3 [2].

4.11.4 Contaminated water management Contaminated water management is presently the most important and urgent issue at Fukushima Daiichi NPS. Fig. 4.80 shows the total contaminated water processing system with the cesium removal system; Simplified Active Water Retrieve and Recovery System (SARRY), the desalination system; KURION and the multinuclide removal system; and ALPS. The ALPS was designed to remove all radionuclides except tritium. The first

Fukushima Daiichi nuclear power plant accident and analysis evaluation

Pump-up Water level

397

Fundamental Measure(3) Pump up groundwater through sub-drains and groundwater drains. [Bypass] Reactor Building

Pump-up

Sub drain

Turbine Building

Pump-up

Drain Pump-up well

Pump-up

Sea surface

Sub drain Keeping water away from the contamination source

Fundamental Measure (2) Install the land-side water-shielding frozen walls to prevent water flowing into buildings

Groundwater drain ン

Fundamental Measure (1) Install the sea-side water-shield walls to prevent leakage Existing seawall

Coolant

Coolant circulation ( to crate frozen soil) Permeable Layer

Pipes

Frozen soil

Frozen soil

Aquiclude Layer

Fig. 4.77 Measures to prevent contaminated water flowing-out to sea by frozen-soil wall and subdrain pumpup [21,23].

Fig. 4.78 Impermeable wall to prevent contaminated water from flowing into sea [21,23].

398

Boiling Water Reactors

Fig. 4.79 Landside impermeable “frozen-soil wall” [21,23].

Fig. 4.80 Contaminated water processing system with SARRY, KURION, and ALPS [21,23].

system consists of three independent lines. Total treatment capacity was 750 ton/day. To increase total processing capacity, a second system was constructed. A high-performance ALPS was designed to reduce the volume of secondary waste, and the processing capacity increased to 2000 ton/day including the first installed system. The ALPS has the important role of water decontamination. Fig. 4.81 shows the capabilities of the three water processing systems for removing radioactive materials. The cesium and strontium concentration level at the ALPS outlet is reduced to 1/108 of the original concentration.

Fukushima Daiichi nuclear power plant accident and analysis evaluation

399

Fig. 4.81 Capabilities of various water processing systems to remove radioactive materials [21,23].

Fig. 4.82 Tritium treated water stored about 1,300,000 m3 as of March 10, 2022 [24].

ALPS to remove various radioactive materials including strontium, except tritium. Therefore, water processed with the ALPS is now called “tritium treated water,” and the safely stored in the storage tanks, as shown in Fig. 4.82. Currently, tritium treated water stored about 1,300,000 m3 in total. It is planned to be diluted to less than 1/100 with seawater and released into the ocean and has undergone rigorous review by IAEA experts, as shown in Fig. 4.83. Diluting to 1/100 or less with seawater fully satisfies the WHO drinking water tritium concentration standard.

400

Boiling Water Reactors

Fig. 4.83 Overview of a diluting ocean discharge facility for ALPS treated water [24].

4.11.5 Preparation for fuel-debris removal To find melted debris under the RPV, a transformer-type robot was developed by Hitachi GE Nuclear Energy, Ltd. [20,21]. As shown in Fig. 4.84, the robot investigated debris inside the pedestal and estimates the location of fuel debris dropped from the core in Unit 2. Fig. 4.85 shows that the robot moved into the pedestal using X-6 large penetration. The robot that moved like a snake move in the penetration pipe fixed on the CRD exchanges rail. After passing the penetration pipe, the robot transforms its

Transformer Type Robot for Investigation debris at pedestal

Small cr awler ①Status investigation on CRD rail and opening in pedestal

ＰＣＶ Shape-changing robot

RPV

②Status/circumstance investigation on the inside of pedestal

③CRD/Platform investigation

CRD

Platform X-6 penetration

CRD exchanging rail

④Pedestal bottom investigation ⑤*Investigation of personnel access opening at pedestal bottom

Opening in pedestal

Personnel access opening

Fig. 4.84 Transformer-type robot for investigating debris at the pedestal in Unit 2 [2].

Fukushima Daiichi nuclear power plant accident and analysis evaluation

401

Fig. 4.85 Robot investigation route into the pedestal using X-6 large penetration [2].

shape from a snake to “E” shape vehicle. The vehicle has a radiation proof CCD camera at the center of the front frame of “E.” TEPCO has a plan to investigate in several steps. The preliminary investigation will identify an access route into the pedestal via the CRD rail and survey the status of PCV inside the pedestal. A full-scale investigation for identifying the location of fuel debris will then follow. Fig. 4.86 shows (a) survey deposit images result in the Unit 2’s PCV at lower part of the CRD housing and (b) on the pedestal floor. As shown in Fig. 4.86a, there were PIP/LPRM cable cannot be seen due to deposits, suggesting the drop off of several CRDs [25]. As shown in Fig. 4.87, it was confirmed that the high-level contamination in the shared exhaust stack for Units 1 and 2 dismantling and fuel-debris deposit retrieval by gripped and moved [23]. In 2022, a robot arm developed by Veolia Nuclear Solutions Ltd, in the United Kingdom will be used for bringing out sample debris from pedestal in the PCV of Unit 2, as shown in Fig. 4.88. Nagoya University and TOSHIBA Corporation released to the press the results of the muon radiography to find fuel debris [26]. Fig. 4.88 shows the layout of the muon detector imaging plate and the measurement area from the core in the RPV to the RPV bottom. Fig. 4.89 compares the muon radiography results of Units 2 and 5. No fuel was found in the core and the lower plenum at the RPV bottom in Unit 2. Almost all fuels might melt and fell down through the hole created by CRD drop off, and on the pedestal concrete floor, as shown in Fig. 4.90.

4.12

Important lessons learned from Fukushima Daiichi NPS accident

There are numerous effective countermeasures for prevent the FP release and tsunami in the world. TEPCO and NISA should consider such systems.

402

Boiling Water Reactors

Fig. 4.86 Survey deposit image result in the Unit 2 PCV at lower part of the CRD housing and on the pedestal floor [25].

Fig. 4.87 Survey the high-level contamination in the shared exhaust stack dismantling for Units 1 and 2 and fuel-debris deposit was confirmed to retrieval by gripped and moved [23].

Fukushima Daiichi nuclear power plant accident and analysis evaluation

403

Fig. 4.88 Robot arm to bring out sample debris from Unit 2 pedestal (developed by Veolia Nuclear Solutions Ltd, in the United Kingdom) [25].

Fig. 4.89 Muon radiography to find fuel debris [26].

The Fukushima Daiichi NPS accident could have been quickly brought under control if adequate countermeasures had been installed, such as waterproof doors and mobile power sources. In Europe, as a result of lessons learned from the TMI-2 and Chernobyl accidents, het removal systems and FCVS have already been installed.

404

Boiling Water Reactors

Fig. 4.90 Comparison of muon radiography results of Units 2 and 5 [26].

Using the lessons learned from this analysis of the Fukushima Daiichi accident, the author wishes to contribute to achieve first-class nuclear safety worldwide. In response to the Fukushima Daiichi NPS disaster, measures for the safety of Japanese NPPs have been strengthened or are being greatly bolstered. It has been pointed out that DiD in Japan before Fukushima accident was “Level 3” and now it is enhanced “Level 4.” Therefore, the author submits proposals as shown below to provide scientific and technological support. The lessons can be incorporated in measures taken by institutions and government agencies, thereby enhancing the safety of the many nuclear power plants in operation worldwide. (1) Enhance seismic electric device to prevent loss of external power by earthquake, SF6 gas insulated switchgear (GIS) and flexible insulators should be installed for transmission line. (2) For SBO caused by wetting of EDG, P/C, DC battery, I&C and cell phones, waterproof door or hatches, and mobile power should be installed on hills. (3) To prevent core meltdown by loss of water injection, the diversification of water injection and heat sinks is very important. (4) To prevent loss of containment function by overheating damage, CV cooling and FCVS should be installed independently from the hard vent systems. (5) Pedestal water injection during a severe accident is very important to avoid PCV bottom failure that causes water pollution around the reactor building.

Sendai Nuclear Power Station Units 1 and 2 of Kyushu Electric Power Company, Inc. received the NRA permission in September 2014 and Units 1 has restarted since August 2015. At March 2022, safety design reviews passed of 12 PWRs and 5 BWRs with enhanced safety measures are ongoing for restarting 7 NPPs. Five PWRs and 5 BWRs are still safety design reviews. The plant licensing period of over 40 years operation should be determined based on technical aging assessment.

References [1] T. Narabayashi, K. Sugiyama, Fukushima Daiichi NPPs accidents caused by the TohokuPacific Ocean Earthquake and Tsunami, AESJ Atoms 53 (6) (2011) 387–400. [2] T. Narabayashi, Fukushima nuclear power plant accident and thereafter, in: K. Kato, M. Koyama, Y. Fukushima, T. Nakagaki (Eds.), Energy Technology Roadmaps of Japan, Springer, 2016, pp. 57–119.

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[3] H. Nakajima, T. Narabayashi, Start from Level 7, Nikkei Science, 2011. July, (in Japanese) http://www.nikkei-science.net/modules/journal/index.php?vol¼201107. [4] NISA, Technical knowledge of Fukushima-Daiichi NPP’s accidents and countermeasure (Interim report), 2012 (in Japanese) http://www.meti.go.jp/press/2011/03/20120328009/ 20120328009.html. [5] NISA, JNES, The 2011 Pacific Coast of Tohoku Pacific Earthquake and the Seismic Damage to the NPPs, 2011. [6] Report of Japanese Government, The Accident at TEPCO’s Fukushima Nuclear Power Stations, IAEA Ministerial Conference on Nuclear Safety, June 2011. https://www. iaea.org/report-japanese-government-iaea-ministerial-conference-nuclear-safety-acci dent-tepcos-fukushima-nuclear-power-stations. [7] EPRI, Modular Accident Analysis Program (MAAP), EPRI, 2012. https://www.epri.com/ research/products/000000000001025795. [8] T. Hayashi, M. Tanigawa, G. Nakamur, About an application and the constitution in the true Reactor core water level measurement system using TDR, 16th Academic lecture symposium, Japan Society of Maintenology, 329–335, 2019. [9] TEPCO, Fukushima Nuclear Accident Analysis Report, June 20, 2012. [10] T. Narabayashi, Lessons learned from the Fukushima Daiichi Nuclear Power Plant Accident, Turbulence, Heat and Mass Transfer 7, Begell House, Inc., 2012. www.gepr.org/en/ contents/20121231-01/KeynoteDrNarabayashi-THMT-12r2.pdf. [11] S. Kawamura, T. Narabayashi, Improved nuclear emergency management system reflecting lessons learned from the emergency response at Fukushima Daini nuclear power station after Great East Japan earthquake, J. AESJ (In Japanese) 15 (2) (2016) 84–96. [12] M. Gavrilas, et al., Safety Features of Operating Light Water Reactors of Western Design, CNES, 2000. [13] T. Narabayashi, Lessons of Fukushima-Daiichi NPP’s Accidents for Achievement of the 1st Class Safety in the World, Fukushima Severe Accident Dose Management & Global Lessons Learned in Occupational Dose Reduction, 2012 International ISOE ALARA Symposium, Fort Lauderdale, Florida, Jan, 2012. [14] Nuclear Regulation Authority, Nuclear Regulation Authority Enforcement of the New Regulatory Requirements for Commercial Nuclear Power Reactors, July 8, 2013. http://www.nsr.go.jp/data/000067212.pdf. [15] T. Narabayashi, Lessons Learned from the Fukushima Daiichi Accident to Establish Resilience Technology for Nuclear Plants Based on the Defense in Depth Philosophy (Invited Plenary), North American ISOE ALARA Symposium, Ft. Lauderdale, Florida USA, Jan. 2015. [16] T. Narabayashi, Y. Kura, Development of water level gage drift recovery system by using back fill system during severe accident, in: The 23rd National Symposium on Power and Energy Systems, A133, 2018. [17] Materials presented by Chubu Electric Power at the Atomic Energy Science Council Nuclear Subcommittee in Shizuoka Prefecture, 2021. [18] Y. Eguchi, S. Sugimoto, H. Hattori, H. Hirakuchi, Tornado pressure retrieval from Fujita’s engineering model, DBT-77, in: Proceedings of the 6th International Conference on Vortex Flows and Vortex Models (ICVFM Nagoya), Nov, 2014. [19] T. Narabayashi, Edit, FVCS working group report in JSME, Filtered Containment Venting System, ERC Publishing Co, Ltd, 2018. [20] T. Narabayashi, et al., Development of High Efficiency Filtered Containment Venting System by using AgX, Short Paper of ICMST-Kobe, 2014.

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[21] N. Yamashita, Activities towards the Decommissioning of Fukushima-Daiichi (TEPCO), ICMST Kobe, 2014. [22] METI, The Update of Fukushima Daiichi NPS, Agency for Natural Resources and Energy, METI, Mar, 2021. [23] METI, Current Status of Fukushima Daiichi Nuclear Power Station-Efforts for Decommissioning and Contaminated Water Management, Agency for Natural Resources and Energy, METI, April, 2020. [24] Radiological Impact Assessment Report Regarding the Discharge of ALPS Treated Water into the Sea (Design stage) [Overview], Nov, 2021. [25] Fukushima Daiichi Nuclear Power Station Unit 2 Reactor Containment Vessel Internal Investigation Results, IRID/TEPCO Holdings, June, 2018. [26] Nagoya Univ, Press release of muon radiography results at Fukushima Daiichi, 2015. https://www.youtube.com/watch?v¼qw_445aOefs. http://www.nagoya-u.ac.jp/about-nu/ public-relations/researchinfo/upload_images/20150320_esi.pdf. (March 20, 2015).

5

BWR innovations

Tetsushi Hinoa, Shinichi Morookab, Kenichi Yoshiokac, Tadashi Narabayashid, Michitsugu Morie, and Chikako Iwakif a Hitachi, Ltd., Hitachi, Ibaraki, Japan, bFormer Waseda University, Shinjuku, Tokyo, Japan, c Toshiba Energy Systems & Solutions, Corp., Kawasaki, Kanagawa, Japan, dTokyo Institute of Technology, Meguro, Tokyo, Japan, eGraduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan, fToshiba Energy Systems & Solutions, Corp., Yokohama, Kanagawa, Japan

Chapter outline 5.1 Trans-uranic (TRU) burner reactor and reduced-moderation water reactor

408

5.1.1 TRU burner reactor 408 References 422 5.1.2 Reduced-moderation light water reactor 423 References 433

5.2 Design innovation of BWR and high-pressure BWR 434 5.2.1 Introduction 434 5.2.2 Objective of LSBWR design 435 5.2.3 Safety system and PCV concept 447 5.2.4 Module fabrication and construction 449 5.2.5 Summary of design innovation of LSBWR, LLBWR, and high-pressure BWR 454 References 455

5.3 Power uprate in BWR

456

5.3.1 Current status and trend of reactor power uprates 456 5.3.2 Reactor thermal power and electric power 458 5.3.3 Reactor power uprate with constant rated reactor thermal power operation 459 5.3.4 Relationship between reactor thermal power and electric power outputs 461 5.3.5 Issues and safety in constant rated reactor thermal power operation 462 5.3.6 Experiences in BWR operation with constant rated reactor thermal power operation 463 5.3.7 Power uprate with equipment modification 464 References 472

5.4 Post-BT standard for BWR power plant

473

5.4.1 Introduction 473 5.4.2 Standard for the assessment of fuel integrity under anticipated operational occurrences 475 References 477

5.5 Core catcher

479

5.5.1 Overview of core melt stabilization and cooling 479 5.5.2 Core catcher of EU-ABWR 482 5.5.3 Core catcher for the existing BWR 487 References 491

5.6 Steam injector

492

5.6.1 Introduction 493 5.6.2 Principle and application of SI 493 Boiling Water Reactors. https://doi.org/10.1016/B978-0-12-821361-2.00006-4 Copyright © 2023 Elsevier Inc. All rights reserved.

408

Boiling Water Reactors 5.6.3 5.6.4 5.6.5 5.6.6 5.6.7 5.6.8

SI analysis model 495 Visualized fundamental tests 497 Application of steam jet-type SI to PCIS 503 Application of water jet type SI to RLP 506 Simplified feed water system by SI 513 Steam injector (SI) pump-up water system to refill pool for passive containment cooling isolation condenser (PCC/IC); SIPOWER 523 References 531

5.7 Built in upper internal control rod drives (CRDs) for ABWR-III

533

5.7.1 Introduction of merits and technical tasks for internal CRD 533 5.7.2 Plant concepts of ABWR-III 534 5.7.3 Power devices for the internal CRD 536 5.7.4 Internal CRD’s mechanism 542 5.7.5 Evaluation of ABWR-III conditions 551 5.7.6 Two-phase flow and structural integrity 560 5.7.7 LOCA and pressure transient analysis 562 5.7.8 Aseismic analysis results 564 5.7.9 Summary 564 References 566

5.1

Trans-uranic (TRU) burner reactor and reduced-moderation water reactor

5.1.1

TRU burner reactor

Tetsushi Hino Hitachi, Ltd., Hitachi, Ibaraki, Japan One of the problems with nuclear power generation is the accumulation of radioactive waste of the long-lived TRUs generated as by-products of the fission of uranium fuel. Hitachi [1] is developing a nuclear reactor that can burn TRU fuel and is based on a BWR design that is already in use in commercial reactors. Achieving the efficient fission of TRUs requires that the spectrum of neutron energies in the nuclear reactor be modified to promote nuclear reactions by these elements. By taking advantage of one of the features of BWRs, namely that their neutron energy spectrum is more easily controlled than that of other reactor types, the new reactor combines effective use of resources with a reduction in the load on the environment by using TRUs as fuel that can be repeatedly recycled to burn these elements up.

BWR innovations

5.1.1.1

409

Introduction

Nuclear power generation has an important role to play in both energy security and reducing emissions of carbon dioxide. One of its problems, however, is the accumulation of radioactive waste from the long-lived transuranium elements (TRUs) generated as by-products of the fission of uranium fuel. The TRUs include many different isotopes with half-lives ranging from hundreds to tens of thousands of years or more. As a result, the radiotoxicity of TRU-containing radioactive waste (a measure of the intensity of radiation in terms of the combined effects of all radioactive isotopes in the waste on the people) takes around 100,000 years to fall to a level equivalent to that of naturally occurring uranium ore. If, on the other hand, the TRUs could be burned up to eliminate them from the radioactive waste, this time could be reduced to a few hundred years [2]. Work is progressing on the research and development of nuclear reactors capable of not only eliminating TRUs from waste but also reducing the consumption of uranium by burning TRUs as fuel. Sodium-cooled fast reactors, which use sodium (Na) to cool the fuel, are one example. Hitachi, meanwhile, is working on the development of its resource-renewable boiling water reactor (RBWR) based on the design of a boiling water reactor (BWR) that is already in use in commercial reactors.

5.1.1.2

RBWR concept

(a) Plutonium breeding reactor: The original concept on which the RBWR is based was the plutonium generation boiling water reactor (PGBR) proposed by Takeda and others from Hitachi in 1988 [3]. The PGBR produces fissile plutonium (Puf), a TRU fuel, for the generation of nuclear power containing plutonium-239 (239Pu) and plutonium-241 (241Pu), from uranium-238 (238 U), a nonfissile isotope that makes up more than 99% of natural uranium (U). Puf breeding means that more Puf is created during the burning of the fuel than is provided in the initial fuel as a start-up neutron source. As it was generally assumed at the time that this could only be achieved using sodium-cooled fast reactors, the PGBR’s ability to breed plutonium in a light water reactor made it a ground-breaking proposal. To achieve the breeding of plutonium, it is necessary to promote the absorption of neutrons by 238 U to transmute it into 239Pu. For it to be used in a power reactor, it also needs to sustain a fission chain reaction in the nuclear reactor. That is, a higher number of neutrons than in a conventional BWR is required to maintain the simultaneous transmutation and fission chain reactions. Typically, the higher the energy of the neutrons that trigger fission, the more it encourages neutron absorption by 238 U and the higher, in relative terms, the number of neutrons emitted by the fission reaction. Accordingly, breeding plutonium requires that the mean neutron energy in the reactor be raised. In a BWR, the heat from the fuel rods causes the water (coolant) that flows through the core of a nuclear reactor to boil, thereby removing heat from the fuel rods. Meanwhile, the coolant also acts to moderate the high energy of the neutrons generated by the fission reaction (reduce it down to a low level) through repeated collisions between neutrons and the hydrogen nuclei in the coolant. To minimize this effect, the PGBR reduces the proportional

410

Boiling Water Reactors

volume of coolant in the reactor core by having narrower coolant channels between fuel rods. Calculations have also demonstrated that the PGBR can achieve high neutron energy and enable Pu breeding despite being a light water reactor by taking advantage of BWR characteristics whereby the boiling of coolant to form steam reduces the density of hydrogen nuclei. (b) Intrinsic safety: Nuclear reactors need to exhibit inherent safety, which means that, when the output of a reactor increases due to some external factor, owing to the nature of its design means it automatically acts to reduce that output. The main mechanisms for inherent safety in a conventional BWR are the effect whereby fission is suppressed by the greater absorption of neutrons as the fuel temperature increases (Doppler effect), and the effect whereby higher temperatures promote more coolant boiling, which raises the proportional volume of steam (void fraction), reduces the moderation of neutrons, and thereby also suppresses fission. This latter effect is a consequence of the fact that, in a conventional BWR, fission is primarily triggered by low-energy neutrons (around 0.1 eV). This is because, at low energies, the fission of fissile 235U and 239, 241Pu tends to become less likely as the neutron energy increases. An indicator that represents this relationship between void fraction and the likelihood of fission is called the void reactivity coefficient. The void reactivity coefficient has a negative value in situations where an increase in void fraction makes fission less likely, as in a conventional BWR. This relationship reverses at high energies of 100 keV or more, where the higher the neutron energy the more likely fission is to occur. Accordingly, when the proportion of fission reactions triggered by high-energy neutrons becomes large, as in the case of PGBR, which is intended for plutonium breeding, the effect whereby a higher void ratio automatically acts to reduce reactor output becomes attenuated. Although the Doppler effect still works automatically to reduce output, the PGBR has a positive void reactivity coefficient. Takeda and his team continued their work, and in 1995 proposed the concept of an RBWR actinide recycler (RBWR-AC) that combined plutonium breeding with a negative void reactivity coefficient [4]. The name of the RBWR derives from its ability to recycle not only plutonium but also the other TRUs as fuel. A feature of the RBWR-AC is that Puf is added as a start-up neutron source to the fuel in two separate zones (two-zoned core) (see Fig. 5.1.1.1). Because the increase in void fraction as the reactor output rises acts to reduce the probability that neutrons will collide with and bounce off hydrogen nuclei, the probability of the neutrons leaving the fuel increases. The idea behind the two-zoned core was to suppress fission by amplifying this effect whereby an increase in void fraction results in greater neutron leakage (see Fig. 5.1.1.2). (c) Multirecycling: Another feature of the RBWR-AC is its multirecycling capability whereby it can repeatedly recycle the TRUs contained in its own spent fuel (see Fig. 5.1.1.3). The TRUs consumed by the fission reactions that produce the heat needed to generate electric power are created by the operation of the reactor itself through the transmutation of depleted uranium. This means that the operating cycle of the RBWR-AC can be maintained simply by replenishing it with depleted uranium. Being the by-product left over after the manufacture of the enriched uranium (which contains a high proportion of the fissile 235U isotope) used to fuel conventional commercial light water reactors, depleted uranium is a plentiful resource that can provide the basis of a long-term energy supply. Multirecycling requires that the isotopic composition of TRUs in the fuel be kept the same before and after burning them in the reactor to avoid the following problem. If the proportion of fissile isotopes after use is low, then the amount of these isotopes will

Fig. 5.1.1.1 RBWR concept. (Apart from the core, the RBWR is largely the same as conventional BWR systems. To achieve effective burning of TRUs, the gaps between fuel rods through which the coolant flows are narrower on the RBWR. Also, the TRUs are contained in two separate zones to ensure inherent safety.)

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Boiling Water Reactors

Fig. 5.1.1.2 Inherent safety of RBWR. (A rise in thermal output leads to more coolant boiling and greater neutron leakage, thereby suppressing fission.)

Fig. 5.1.1.3 Fuel cycle for RBWR-AC. (Each operation cycle of the RBWR-AC forms its own TRUs to replace those consumed during operation while maintaining the same isotopic composition of TRUs before and after use.)

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diminish each time the fuel is recycled, ultimately resulting in the loss of criticality in the reactor. There is also the risk of compromising reactor design and operation criteria, such as a change in the fuel composition causing the void reactivity coefficient to become positive. In addition to having narrower coolant channels, the RBWR-AC is able to satisfy a variety of criteria as it runs through repeated operation cycles by modifying the coolant flow rate to adjust the neutron energy spectrum during operation and maintain a constant isotopic composition of TRUs in the fuel before and after use. (d) TRU burner: One of the advantages of nuclear power generation is that it has a higher energy density than thermal and other forms of power generation, meaning that it requires less fuel to produce the same amount of electric power. On the other hand, when TRUs are used to fuel a nuclear reactor, the percentage of the initial fuel load burned up in each operation cycle is only in the single digits or low teens. Accordingly, the fuel needs to be recycled many times to burn up a large amount of TRUs. Takeda and his team took advantage of the characteristics of the RBWR that allow it to repeatedly recycle fuel while keeping its isotopic composition of TRUs constant and proposed the concept of employing a TRU burner (RBWR-TB), meaning a reactor that could reduce the amount of TRUs by burning them up [5,6]. Although the RBWR-TB shares the RBWR-AC’s characteristic of maintaining a constant isotopic composition of TRUs in the fuel before and after use, the process of burning the fuel reduces the absolute quantity of TRUs. The operation cycle is repeated by making up for this loss of TRUs by supplying fuel from another RBWR-TB. That is, the concept behind the RBWR-TB is to run the operation cycle with progressively fewer reactors until all of the TRUs are burned up except for those loaded into the final reactor (see Fig. 5.1.1.4). This presents a scenario under which the fuel is first used for long-term energy production in RBWR-ACs, during which time the TRU fuel is maintained at a constant level. Subsequently, once alternative nonnuclear forms of energy production become available, RBWR-TBs are then used to burn up the TRUs and transition away from nuclear power generation without leaving behind long-lived radioactive waste. (e) Feasibility study by US universities: Between 2007 and 2011, three US universities (Massachusetts Institute of Technology, University of Michigan, and University of California, Berkeley) conducted feasibility studies of RBWR reactors under research contracted to the Electric Power Research Institute, Inc. (EPRI) [7]. Though some differences between the analysis results obtained by Hitachi and the universities need to be evaluated further, the analyses collectively indicated that the RBWRs appeared to achieve their design objectives. Part of the contracted study included the proposal of another TRU reactor, the RBWR-TB2, for comparison with sodium-cooled fast reactors. An RBWR-TB2 operates in parallel with a conventional light water reactor and burns the TRUs in its spent fuel (see Fig. 5.1.1.5). An RBWR-TB2 recycles fuel repeatedly, loading a mixture of fuels comprising the TRUs in both its own and the light water reactor’s spent fuel. The operation of the RBWR-TB2 serves to minimize the buildup of excess TRUs.

5.1.1.3

RBWR specifications

(a) Plant overview: The rated thermal output, power output, reactor pressure vessel diameter, and core pressure of the RBWR are the same as the advanced BWR (ABWR), the latest commercial

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Boiling Water Reactors

Fig. 5.1.1.4 Fuel cycle for RBWR-TB. (The TRUs consumed during operation are replenished from the spent fuel of other RBWR-TBs. The fuel is recycled through progressively fewer reactors until all the TRUs are burned up apart from those in the final reactor.)

BWR (see Fig. 5.1.1.6). The core has 720 fuel assemblies and 223 Y-profile control rods. As the fuel assemblies for the RBWR-AC, RBWR-TB, and RBWR-TB2 all have roughly the same size, their cores can be swapped between each other by exchanging fuel assemblies. The next section describes the latest specifications for each reactor type [8]. (b) Core fuel configuration: Fig. 5.1.1.7 shows the fuel assemblies for the RBWR-AC, RBWR-TB, and RBWR-TB2. The fuel assemblies for the RBWR-AC have upper and lower TRU zones (with heights of 280 and 193 mm), sandwiched between the upper, central, and lower depleted uranium zones (with heights of 70, 520, and 280 mm). Neutron absorber zones, meanwhile, are located above and below the upper and lower depleted uranium zones. These are provided to enhance the output suppression effect by absorbing neutrons when a rise in output causes the void fraction of the coolant to increase, thereby increasing the leakage of neutrons from the fuel zones (TRU and depleted uranium zones). The heights of the depleted uranium and TRU zones are determined so as to maintain the same isotopic composition of TRUs before and after use. Because the requirement for TRU breeding is lower on the RBWR-TB and RBWR-TB2 TRU burner reactors than it is on the RBWR-AC, their fuel assemblies do not have the lower depleted uranium zone. As in the case of the RBWR-AC, they have neutron absorber

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Fig. 5.1.1.5 Fuel cycle for RBWR-TB2. (The TRUs consumed during operation are replenished from the spent fuel of light water reactors. The RBWR-TB2 operates in parallel with a light water reactor to minimize the buildup of excess TRUs.)

zones above and below the fuel zones. The heights of the TRU and depleted uranium zones for the RBWR-TB are also determined so as to maintain the same isotopic composition of TRUs before and after use, as in the case of the RBWR-AC. In the case of the RBWR-TB2, when the isotopic composition of TRUs in the fuel supplied from light water reactors is the same for each operation cycle, the heights of the TRU and depleted uranium zones are determined such that the isotopic composition that results after combining the RBWR-TB2’s own spent fuel will be the same for each operation cycle. Along with adjusting the heights of the TRU and depleted uranium zones in the RBWR-AC, RBWR-TB, and RBWR-TB2, the balance between consumption and creation of TRUs is also adjusted by modifying the fuel rod diameter and gaps between fuel rods to adjust the neutron energy spectrum (see Fig. 5.1.1.8). In the case of the RBWR-AC, where not only the isotopic composition of TRUs but also their quantity needs to be kept constant before and after use, it is necessary to increase the mean neutron energy by having the lowest proportion of coolant of the three designs so that the ratio of fissile TRU isotopes before and after use (breeding ratio) will be above 1.0. After the RBWR-AC, it is the RBWR-TB that has the next lowest proportion of coolant. This is because the breeding ratio of fissile TRU isotopes needs to be increased somewhat to keep the relative reduction in fissile TRU

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Boiling Water Reactors

Fig. 5.1.1.6 Basic specifications of RBWR plant. (The rated thermal output, power output, reactor pressure vessel diameter, and core pressure of the RBWR are the same as in an ABWR. A follower is fitted to the top of each control rod to prevent water from entering the space vacated when the control rod is withdrawn.) isotopes and nonfissile TRU isotopes the same in the RBWR-TB, where the isotopic composition needs to be kept constant even as the quantity decreases as the fuel is burned. The fuel configuration of the RBWR-TB2, on the other hand, has a higher proportion of coolant than the RBWR-TB because it is supplied with TRUs in spent fuel from light water reactors with a high proportion of fissile isotopes.

Fig. 5.1.1.7 RBWR fuel assembly. (The height of the TRU and depleted uranium zones, the fuel rod diameter, and the gap between fuel rods are adjusted to ensure that the multirecycling of TRUs can be performed in a way that suits the purpose of each reactor design. As the RBWR-TB and RBWR-TB2 TRU burner reactors have less need for TRU formation, their fuel assemblies do not include the lower depleted uranium zone.)

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Boiling Water Reactors

Fig. 5.1.1.8 Proportionate volume of coolant in RBWR and conventional BWR reactors. (Since the RBWR-AC and -TB need to continue the operation cycle without consuming fissile materials other than those contained in the fuel discharged from themselves, their water-to-fuel volume ratios are set lower than those of the RBWR-TB2 and conventional BWR.)

5.1.1.4

RBWR core characteristics

Burning uranium in a conventional BWR forms both fissile and nonfissile TRU isotopes (see Fig. 5.1.1.9). When using mixed oxide (MOX) fuel that contains plutonium and uranium, the fuel ends up with more of the nonfissile isotopes of plutonium and other TRUs than it had when loaded because the fissile isotopes of plutonium are consumed, whereas the nonfissile isotopes are not. Calculations have demonstrated that the RBWR-TB and RBWR-TB2 TRU burner reactors are able to consume both fissile and nonfissile TRU isotopes at more than twice the rate they are produced by a conventional BWR. Fig. 5.1.1.10 shows the dependence of the neutron capture rate and fission reaction rate on neutron energy in the nuclear reactor of an RBWR-TB, where the neutron capture rate is the proportion of cases in which a transmutation reaction occurs due to the capture of a neutron and the fission reaction rate is the proportion of cases in which fission occurs [8]. Neutron capture occurs for a wide range of energies, from low energies of around 10–1 eV up to high energies on the order of 106 eV. Nonfissile isotopes such as 240Pu and americium-241 (241Am) are transmuted by neutron capture into fissile 241Pu and 242Am, respectively. Also, the direct fission of isotopes like 240Pu and 241 Am occurs at high neutron energies in the vicinity of 106 eV. By using this wide distribution of neutron energies to achieve the fission of nonfissile TRUs, both by first transmuting them into fissile TRUs and also through direct fission by high-energy neutrons, the RBWR can burn up nonfissile TRUs in the same proportion as fissile TRUs and maintain the same isotopic composition of TRUs before and after use.

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Fig. 5.1.1.9 Rate of formation and consumption of TRUs. (A negative value indicates that the consumption of TRUs results in a reduction in their quantity.) The value for conventional BWRs is obtained from Y. Ando, H. Takano, Evaluation of isotopes present in spent fuel from light water reactors, JAERI-Research 99-004 (1999) (in Japanese) [9].

Table 5.1.1.1 lists the isotopic composition of TRUs before and after use and the core characteristics of the RBWR-AC, RBWR-TB, and RBWR-TB2. When the RBWRAC and RBWR-TB are operated to keep the same isotopic composition of TRUs in the fuel before and after use, the quantity of TRUs increases in the case of the RBWR-AC and decreases in the case of RBWR-TB. When an RBWR-TB2 is loaded with fuel in which the TRUs in its own spent fuel and the spent fuel from a light water reactor are in the same proportion, and under conditions in which the isotopic composition of TRUs remains the same in each cycle, the quantity of TRUs decreases as the fuel is burned. Calculations have also demonstrated that all of the reactor types have a negative void reactivity coefficient [8].

5.1.1.5

Progressive introduction of RBWR [10]

In response to public expectations, the intention is the development of the RBWR in stages, in step with progress on the development of fuel cycle technologies. Accordingly, Hitachi-GE started work on the development of an RBWR that provides flexibility in planning the use of plutonium by introducing the use of tight lattice fuel in existing BWRs. Fig. 5.1.1.11 shows the concept behind the introduction of the RBWR. The square lattice RBWR can be implemented through the replacement or

420

Boiling Water Reactors

Fig. 5.1.1.10 Dependence on neutron energy of neutron capture and fission reactions in RBWR-TB core. (Nonfissile TRUs are transmuted into fissile TRUs by the capture of a neutron.)

new installation of components that can easily be achieved, such as the fuel bundles and control rods from the existing BWRs. It is also anticipated that existing reprocessing and mixed-oxide (MOX) fuel manufacturing techniques will be used. The square lattice RBWR will also facilitate future recycling and make better use of resources because, as it consumes plutonium, it also leaves the spent fuel with a higher proportion of fissionable plutonium isotopes than does the current MOX fuel.

Table 5.1.1.1 Change in isotopic composition of TRUs during burning in reactor. This considers the case when spent fuel is left for three years after removal from the reactor to allow radiation and heat generation to diminish. RBWR-AC

RBWR-TB

RBWR-TB2

Isotope

When fuel loaded

Three years after removal

When fuel loaded

Three years after removal

When fuel loaded

Three years after removal

TRUs discharged from light water reactor

Np237 Pu238 Pu239 Pu240 Pu241 Pu242 Am241 Am242m Am243 Cm244 Cm245 Cm246 Cm247 Cm248 Cm249 Puf (t) TRU (t)

0.4 2.9 43.5 36.3 5.1 5.1 3.6 0.2 1.3 1.1 0.4 0.1 0 0 0 1.94 3.99

0.4 2.9 43.5 36.3 5.1 5.1 3.6 0.2 1.3 1.1 0.4 0.1 0 0 0 1.96 4.03

0.1 4.7 9.5 39.5 4.4 25.4 4.7 0.2 4.7 4.1 1.2 1 0.2 0.2 0.1 1.14 8.18

0.1 4.7 9.5 39.6 4.4 25.4 4.7 0.2 4.7 4 1.2 1 0.2 0.2 0.1 1.06 7.62

1.9 6.3 27.7 38.5 5.5 9.6 5.4 0.2 2.4 1.8 0.5 0.2 0 0 0 2.06 6.2

1.4 6.7 25.5 40.1 5.4 10.1 5.4 0.2 2.4 2 0.6 0.2 0 0 0 1.74 5.63

6.7 2.8 48.8 23 7 5 4.7 0 1.5 0.5 0 0 0 0 0 0.32 0.58

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Boiling Water Reactors

Fig. 5.1.1.11 Concept behind introduction of RBWR. (The aim is to work toward the practical application of hexagonal lattice RBWRs in conjunction with progress in other areas like reprocessing and fuel manufacturing technologies by building up experience with the use of tight lattice fuel and high-energy neutrons in light-water reactors while also contributing to progress on the fuel cycle.)

While facilitating the early implementation of the use of tight lattice fuel and highenergy neutrons in light-water reactors through this step-by-step verification and also contributing to the progress of the fuel cycle, the aim is to move on to the practical application of hexagonal lattice RBWRs with their superior resource sustainability in conjunction with progress in other areas like reprocessing and fuel manufacturing technologies. Continuous work to improve reactor core analysis techniques and verify their applicability to RBWR is also in progress, by collaborating with research institutions in Japan, the UK, and the US [11–13].

References [1] T. Hino, M. Ohtsuka, K.M. Masayoshi, Light water reactor system designed to minimize environmental burden of radioactive waste, Hitachi Rev. 63 (9) (2014) 602. [2] Working Group on Technologies for Separation of Nuclides and Nuclear Transmutation, Nuclear Science and Technology Committee, Ministry of Education, Culture, Sports, Science and Technology. http://www.mext.go.jp/b_menu/shingi/gijyutu/gijyutu2/070/index. htm. (in Japanese). [3] R. Takeda, et al., A conceptual core design of plutonium generation boiling water reactor, in: Proc. of the 1988 International Reactor Physics Conference 3, 1988. p. 119.

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[4] R. Takeda, et al., General features of resource-renewable BWR (RBWR) and scenario of long-term energy supply, in: Proc. of International Conference on Evaluation of Emerging, Nuclear Fuel Cycle Systems 1, 1995. p. 938. [5] R. Takeda, et al., BWRS for long-term energy supply and fissioning almost all transuranium, in: Proc. of GLOBAL 2007, 2007. p. 1725. [6] R. Takeda, et al., RBWRs for fissioning almost all uranium and transuraniums, Trans. Am. Nucl. Soc. 107 (2012) 853. [7] EPRI, Technical Evaluation of the Hitachi Resource-Renewable BWR (RBWR) Design Concept, EPRI Technical Report 1025086, 2012. [8] T. Hino, et al., Core designs of RBWR (resource-renewable BWR) for recycling and transmutation of transuranium elements—an overview, in: Proc. of ICAPP 2014, Paper 14271, 2014. [9] Y. Ando, H. Takano, Evaluation of isotopes present in spent fuel from light water reactors, in: JAERI-Research 99-004, 1999 (in Japanese). [10] K. Kito, T. Hino, K. Matsumura, M. Matsuura, Hitachi’s vision for nuclear power and development of new reactors, Hitachi Rev. 69 (4) (2020) 564. [11] T. Hino et al., Core design and analysis of axially heterogeneous boiling water reactor for burning transuranium elements, Nucl. Sci. Eng., 187, pp. 213–239 (Jan. 2017). [12] T. Hino, et al., Core design of RBWR (resource-renewable boiling water reactor) and benchmark calculation of core analysis tools, in: ICAPP 2019, May 2019. [13] B.A. Lindley et al., Development of a core design capability for innovative boiling water reactor designs for burning transuranic isotopes using WIMS/PANTHER, Ann. Nucl. Energy, 123, pp. 162–171 (Jan. 2019).

5.1.2

Reduced-moderation light water reactor

Shinichi Morookaa and Kenichi Yoshiokab a

Former Waseda University, Shinjuku, Tokyo, Japan, bToshiba Energy Systems & Solutions, Corp., Kawasaki, Kanagawa, Japan

5.1.2.1

Introduction [1–4]

A reduced-moderation spectrum water reactor is an attractive reactor that can realize multirecycling of plutonium (Pu) and even breeding or high conversion cycles based on light water reactor (LWR) technology. However, because this reactor needs to use the tight fuel lattice to induce the fast neutron spectrum, there are several technical issues related to the tight fuel lattice, such as the reduction of heat transfer performance and the void reactivity coefficient that tends to be positive. The term reduced moderation is derived from the reactor physical characteristic that the moderation of fast neutrons is suppressed as low as possible. This feature

424

Boiling Water Reactors

is different from that of the current LWRs, where fast neutrons are sufficiently moderated by water, which is a moderator and coolant. In addition, the use of fast neutrons is a way to increase the conversion ratio to Pu. To achieve the concept of the reducedmoderation spectrum water reactor, it is necessary to reduce the amount of water, which is a neutron moderator, as much as possible. The fuel assembly is different from that of the current LWR.1 However, for parts other than the core, basically the same system and equipment as the current LWR can be used. Research and development of this reactor are being actively promoted by LWR manufacturers and national research institute, and several different concepts are proposed. Japan Atomic Energy Agency (JAEA) and Hitachi have been collaborating on research and development of reduced-moderation water reactors aiming for a high conversion ratio [1]. From the viewpoint of the economy, Toshiba has developed “the cost-reduced low-moderation spectrum BWR” targeting the high burnup and compatibility with the current BWR1 [1,2,4]. In this chapter, the cost-reduced low-moderation spectrum BWR conducted by S. Morooka et al. with the support of the Institute of Applied Energy (IAE) and Ministry of Economy Trade and Industry (METI) is described [2–4].

5.1.2.2

Research and development of the cost-reduced low-moderation spectrum BWR

(a) Target of the design [1–8]: The concept of this BWR development is to significantly reduce the construction cost by modifying only the core of the ABWR and reusing the rest of the ABWR system. In addition, by adopting an innovative fuel system named BARS (BWR with an Advanced Recycle System) to the fuel cycle side, the utilization of uranium resources and reduction of radioactive wastes were enhanced. We called it the cost-reduced low-moderation spectrum BWR because of the improved economic efficiency. Fig. 5.1.2.1 shows the BARS concept. The spent fuels from the conventional LWRs are recycled as MOX fuels manufactured by the oxide dry-processing and the vibro-packing fuel fabrication. Fig. 5.1.2.1 BARS (BWR with an advanced recycle system) concept [3].

1

The current BWR fuel assembly is explained in “Section 3.2”.

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(b) Core description [1–8]: Figs. 5.1.2.2–5.1.2.5 show the drawing of this low-moderation ABWR. The triangle lattice fuel assembly is used to reduce the moderation of neutron and obtain a fast neutron spectrum. Square flow channel was used for compatibility with the current ABWR. In addition, a large fuel assembly is used to eliminate the bypass flow path as much as possible and reduce the amount of water in the core. By doing so, it is possible to reduce the time for refueling during periodic inspections, which leads to a reduction in power generation costs. Fig. 5.1.2.2 Vertical view of BARS [3].

Fig. 5.1.2.3 BARS core layout [3].

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Boiling Water Reactors

Fig. 5.1.2.4 Vertical view of fuel assembly [3].

Fig. 5.1.2.5 Cross-sectional view of BARS fuel [3] concept. It is well known that the void reactivity coefficient in the reduced-moderation BWRs has the tendency to be positive in harder neutron energy spectrum. In designing the reduced-moderation BWRs, the key point is how to make the void reactivity coefficient negative. In the current LWR, when core power increases, the steam increases. Because the neutron moderation effect of the steam is smaller than that of the water, the core power decreases. This is called a self-limiting or a self-regulating characteristic of nuclear fission (inherent safety of nuclear reactor). In the low-moderation BWR, there is less water and more steam, so even if the steam increases, the core power is less likely to decrease.

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Then, a new core concept named the streaming channel has been applied to improve the void reactivity coefficient. As shown in Fig. 5.1.2.2, the core of this BWR consists of partial fuel assemblies and normal fuel assemblies. The space above the partial fuel assemblies is called the neutron streaming channel. The streaming channel consists of the cavity can and water gap region around it as shown in Fig. 5.1.2.4. When void fraction increases due to the increase in the core power or the decrease in the core flow, the void fraction in the water gap increases. The streaming channel can increase the axial leakage of neutrons. This results in the negative void reactivity coefficient under the event of the increased void fraction. (c) Thermal–hydraulic test of tight lattice bundle [3,9–11]: Because it is necessary to reduce the amount of water, which is a neutron moderator and cooling fluid, as much as possible in order to achieve the fast-neutron spectrum, the tight lattice fuel assembly (the triangle lattice fuel assembly) is used. Therefore, there is a technical issue for decreasing the heat transfer performance and increasing the pressure drop. Fig. 5.1.2.6 shows the cross-sectional view of test assemblies. The effects of the flow distribution effect and the radial peaking were added to Arai’s correlation [12] for the critical power of tight lattice bundle. Fig. 5.1.2.7 shows the comparisons between the measured critical power of 7-rod and 14-rod, and the modified Arai’s correlation [9]. Fig. 5.1.2.8 shows the comparisons between the measured pressure drop of 7-rod and 14-rod, and Morooka’s prediction method [10,11]. It was found from the comparisons that we could obtain the correlations for evaluating the important thermal-hydraulics performance of the tight lattice core. (d) Critical experiments on reduced-moderation LWR: Critical experiments for a reduced-moderation LWR were done by NCA (Toshiba nuclear critical assembly) which is described in Section 3.3.1.3 [13].2 The void coefficient of reactivity of reduced-moderation LWR tends to be positive. This trend can be improved by applying a streaming channel, which is a new concept, to reduced-moderation LWR. Fig. 5.1.2.9 shows a typical reduced-moderation LWR fuel assembly and streaming channel. The streaming channel consists of a void region, structural material, and a water gap. The void fraction in the void region is 100%, and the rated void fraction of a typical reduced-moderation LWR plant is normally 60%. The void fraction in the water gap increases with increasing reactor power. More neutron leakage from the core make the void coefficient of reactivity negative. Void reactivity is defined as the k-eff (effective neutron multiplication factor) change caused by the change of the void fraction. The k-eff change can be obtained by evaluating the change in k-inf (infinite neutron multiplication factor) and neutron leakage. The k-inf

Fig. 5.1.2.6 Cross-sectional view of Test assemblies [3].

2

The study had been conducted under the sponsorship of the Ministry of Education, Science and Technology in Japan.

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Boiling Water Reactors

Fig. 5.1.2.7 Calculated critical power of 7-rod and 14-rod data compared with measured data [9].

Fig. 5.1.2.8 Calculated pressure drop of 7-rod and 14-rod data compared with measured data [9].

BWR innovations

429

Fig. 5.1.2.9 An example of a fuel assembly and a streaming channel of a reducedmoderation LWR. From Fig. 1 of K. Yoshioka, T. Kikuchi, S. Gunji, H. Kumanomido, I. Mitsuhashi, T. Umano, M. Yamaoka, S. Okajima, M. Fukushima, Y. Nagaya, T. Mori, T. Kitada, T. Takeda, Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments, J. Nucl. Sci. Technol. 52 (2015) 282–293. change is focused in the experiments, because the neutron leakage of a commercial reactor is different from that of NCA. The k-inf measurement in critical experiments is difficult. Here, instead of k-inf, the “finite neutron multiplication factor”, k* [14] is defined. The k* is the ratio of neutron production rates to absorption rates at a position. The k* is equal to k-inf in the case of no neutron-leakage condition. k* is expressed by the following equation:

k ∗ ðr Þ ¼

Sð r Þ Að r Þ

(5.1.2.1)

In this equation, S(r) and A(r) denote the neutron production rates and the neutron absorption rates at a fuel cell position. The fuel cell includes the fuel pellet, cladding, and moderator. The void reactivity in the fuel cell, Δρcell, can be expressed by the following equation: Δρcell ¼ 1=k∗ðr Þ ðvoidÞ  1=k∗ðr Þ ðno voidÞ:

(5.1.2.2)

The void reactivity was measured in three core configurations as shown in Fig. 5.1.2.10. Case 1 is a simulated condition where the test assembly is filled with cold water. Cases 2 and 3 are simulated 0% and 60% of void fraction. The void

Stainless Steel rod

㻿

1

㻿

1

Stainless Steel rod

Unit䠖mm

Driver Zone (235U enrichment 2wt.% UO2)

Driver Zone (235 U enrichment 2wt.% UO2 )

15

2

2

A 1 spacer 3

4 5

4 5

㻿

6

6

㻿

Void simulated zone (0V or 60V) 235

:

:235U enrichment 4.9wt.% UO2 (96 rods) 1

6

:Measured Rods

:Water hole (4 holes)

:235U enrichment 4.9wt.% UO2 (96 rods) 1

(235U enrichment 4.9wt.% UO2 6 rods)

60 % or 0% voidsimulated polystyrene

:235U enrichment 3.9wt.% UO2 (82 rods)

U enrichment 3.9wt.% UO2 (86 rods)

600

3

6 :Measured Rods

(235U enrichment 4.9wt.% UO2 6 rods) S :Stainless Steel Rods (4 rods) 15

:Water hole (4 rods) A 1 spacer

70

20cm 60%Void

340

End of fuelactive length

Stainless Steel rod

Case 2 0 % void condition and Case 3 60 % void condition Stainless Steel rod

Case1 cold condition

0%Void

End of a fuel rod

0% void simulated polystyrene plate or 60% void simulated polystyrene form

Fig. 5.1.2.10 Core configurations for void reactivity distribution measurements. From Fig. 9 of K. Yoshioka, T. Kikuchi, S. Gunji, H. Kumanomido, I. Mitsuhashi, T. Umano, M. Yamaoka, S. Okajima, M. Fukushima, Y. Nagaya, T. Mori, T. Kitada, T. Takeda, Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments, J. Nucl. Sci. Technol. 52 (2015) 282–293.

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Measurement(Δk*/k*(%))

-8.0 -7.0

Cold→60%V

-6.0 -5.0 -4.0 -4.0

-5.0

-6.0

-7.0

-8.0

Calculation(Δk*/k*(%))

Fig. 5.1.2.11 Comparison between continuous-energy Monte Carlo calculations with measurements for the distributions of the void reactivity in a fuel cell. From Fig. 13 of K. Yoshioka, T. Kikuchi, S. Gunji, H. Kumanomido, I. Mitsuhashi, T. Umano, M. Yamaoka, S. Okajima, M. Fukushima, Y. Nagaya, T. Mori, T. Kitada, T. Takeda, Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments, J. Nucl. Sci. Technol. 52 (2015) 282–293.

fraction of cases 2 and 3 was simulated by polystyrene plates. Void reactivity is measured at some positions. Fig. 5.1.2.11 compares measurement data with calculation on void reactivity effect. Continuous-energy Monte Carlo calculation code, MVP [15], and JENDL-3.3 [16] were used to investigate the measured void reactivity. The MVP calculation considered three-dimensional geometry. The calculations show good agreements with measurements within the measurement uncertainty. On the other hands, critical experiments on streaming channel were also conducted in NCA [17].c Fig. 5.1.2.12 shows the test configurations. A stainless-steel streaming channel and an aluminum streaming channel were used for case 1 and case 2, respectively. A streaming channel was installed at the core center. Polyethylene plates simulated the water gap of a streaming channel. The thicknesses of these polyethylene plates were 5, 10, and 15 mm, which simulate the void fractions of 70%, 35%, and 0% for the 2-cm-thick water gap during the hot operation condition. The void fraction in the case with no polyethylene is 100% in the water gap. Void reactivity ρv was derived from the difference between the reactivity of the 100%-void case and another void case based on the following equation: ρv ¼ ρ ðV Þ  ρ ð100%Þ,

(5.1.2.3)

where V is 70%, 35%, or 0%.

c

A part of this study has been conducted under the sponsorship of the Ministry of Economy, Trade and Industry, Japan.

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3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 3 3

3 3 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

5

4.9 wt% fuels

3

3 wt% fuels

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5

5 5 3 3

3 3 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 3 3

3 3 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

Streaming channel (Stainless steel or aluminum) Polyethylene plate (0, 5, 10, 15mm)

2 wt% fuels Fig. 5.1.2.12 Core configuration for the streaming effect measurement. From Fig. 8 of K. Yoshioka, M. Yamaoka, K. Hiraiwa, T. Kitada, Streaming effect of void reactivity in LWR critical experiments with streaming channel, Nucl. Sci. Eng. 195 (2021) 101–117.

Fig. 5.1.2.13 shows the comparison of calculation and measurement of the void reactivity. The MCNP code [18] was used to analyze the measurement results for the void reactivity of a streaming channel. The JENDL-4.0 [19] and ENDF/B-VIII.0 [20] nuclear data libraries were used for these analyses. The MCNP calculation considered three-dimensional geometry. The trend of the calculations with both libraries agreed well with that of the measurements.

BWR innovations

300

2400

Stainless Steel

100 -100 -300 -500 -700 -900

JENDL-4.0 ENDF/B-VIII.0

-1100 -1100-900 -700 -500 -300 -100 100 300 500 Measurements(pcm)

Calculations(pcm)

Calculations(pcm)

500

433

2200

Aluminum

2000 1800 1600 1400 1200 1000

JENDL-4.0 ENDF/B-VIII.0

800 800 1000 1200 1400 1600 1800 2000 2200 2400 Measurements(pcm)

Fig. 5.1.2.13 Comparison of the C/M of void reactivity. From Fig. 13 of K. Yoshioka, M. Yamaoka, K. Hiraiwa, T. Kitada, Streaming effect of void reactivity in LWR critical experiments with streaming channel, Nucl. Sci. Eng. 195 (2021) 101–117.

References [1] T. Okubo, S. Morooka, R. Takeda, Progress and issues on development of reducedmoderation water reactors; aiming at Multiple recycling and breeding of Plutonium based on Light Water Reactor technology, J. Atom. Energy Soc. Jpn. 48 (7) (2006) 484–489 (in Japanese). [2] S. Morooka, Development of cost-reduced low-moderation boiling water reactor, in: JAERI-Review 2005-029, 2005 (in Japanese). [3] Y. Yamamoto, K. Hiraiwa, S. Morooka, N. Nobuaki, Critical power performance of tight lattice bundle, JSME Int. Ser. B 47 (2) (2004) 344–350. [4] S. Morooka, Development of Low-moderation Spectrum BWR, J. Jpn. Soc. Energy Resour. 23 (6) (2002) 395–398 (in Japanese). [5] K. Hiraiwa, K. Yoshioka, Y. Yamamoto, M. Akiba, Y. Yamaoka, N. Abe, J. Mimatsu, Experimental study on reduced moderation BWR with advanced recycle system (BARS), in: PHYSOR 2004—The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments Chicago, Illinois, April 25–29, 2004. [6] K. Hiraiwa, M. Yamaoka, N. Abe, Y. Yamamoto, I. Mitsuhashi, S. Morooka, J. Mimatsu, A. Inoue, Study on reduced-moderation spectrum BWR with an Advanced Recycle System, in: PHYSOR 2002, Seoul, Korea, October 7–10, 2002. [7] K. Hiraiwa, Y. Yamamoto, K. Yoshioka, M. Yamaoka, A. Inoue, J. Mimatsu, BARS: BWR with Advanced Recycle System, in: Proceeding of Advanced Reactors with Innovative Fuels Second Workshop, Chester, United Kingdom 22–24 October 2001 Hosted by British Nuclear Fuels Limited (BNFL), 2001, pp. 275–286. [8] M. Yamaoka, Y. Yamamoto, N. Abe, K. Hiraiwa, T. Yokoyama, Study on fast spectrum BWR core for actinide recycle, in: 9th International conference on nuclear engineering (ICONE-9), Nice Acropolis (France), 2001. [9] Y. Yamamoto, M. Akiba, S. Morooka, K. Shirakawa, N. Abe, Thermal hydraulics performance of tight lattice bundle, JSME Int. Ser. B 49 (2) (2006) 334–342.

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[10] S. Morooka, Y. Yamamoto, K. Shirakawa, Pressure drop of tight lattice rod bundle, Trans. Atom. Energy Soc. Jpn. 2 (3) (2003) 301–306 (in Japanese). [11] S. Morooka, Y. Yamamoto, K. Shirakawa, Study for pressure drop of rod bundle with tight lattice array, Trans. JSME Ser. B 72 (715) (2006). No. 05-0625 (in Japanese). [12] K. Arai, S. Tsunoyama, S. Yokobori, K. Yoshimura, Critical power characteristics of a high conversion boiling water reactor, in: Proceeding of IAEA Technical Committee on Technical and Economical Aspects of High Converters, Germany, IAEA-TECDOC638 (1990). [13] K. Yoshioka, T. Kikuchi, S. Gunji, H. Kumanomido, I. Mitsuhashi, T. Umano, M. Yamaoka, S. Okajima, M. Fukushima, Y. Nagaya, T. Mori, T. Kitada, T. Takeda, Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments, J. Nucl. Sci. Technol. 52 (2015) 282–293. [14] M. Ueda, Determination of multiplication factors using experimental lattice parameters of cluster-type fuel lattices, J. Nucl. Sci. Technol. 12 (1975) 229–242. [15] Y. Nagaya, K. Okumura, T. Mori, M. Nakagawa, MVP/GMVP II: general purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multi group methods, in: JAERI 1348, Japan Atomic Energy Agency, Tokyo, 2005. [16] K. Shibata, T. Kawano, T. Nakagawa, O. Iwamoto, J. Katakura, T. Fukahori, S. Chiba, A. Hasegawa, T. Murata, H. Matsunobu, T. Ohsawa, Y. Nakajima, T. Yoshida, A. Zukeran, M. Kawai, M. Baba, M. Ishikawa, T. Asami, T. Watanabe, Y. Watanabe, M. Igashira, N. Yamamuro, H. Kitazawa, N. Yamano, H. Takano, Japanese evaluated nuclear data library version3 revision-3: JENDL-3.3, J. Nucl. Sci. Technol. 39 (2002) 1125–1136. [17] K. Yoshioka, M. Yamaoka, K. Hiraiwa, T. Kitada, Streaming effect of void reactivity in LWR critical experiments with streaming channel, Nucl. Sci. Eng. 195 (2021) 101–117. [18] J.F. Briesmeister, MCNP: A General Monte Carlo N-particle Transport Code, Version 4C, in: Technical Report LA-13709-M, Los Alamos National Laboratory, 2000. [19] K. Shibata, et al., JENDL-4.0: A New Library for Nuclear Science and Engineering, J. Nucl. Sci. Technol. 48 (2011) 1. [20] D.A. Brown, et al., ENDF/B-VIII.0: The 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data, Nucl. Data Sheets 148 (2018) 1.

5.2

Design innovation of BWR and high-pressure BWR Tadashi Narabayashi Tokyo Institute of Technology, Meguro, Tokyo, Japan

5.2.1

Introduction

In Japan, increasing unit capacity of nuclear power plants is being promoted in order to utilize economies of scale and further enhance safety and reliability. As a result, more than 50 nuclear power plants have played an important role in power generation.

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However, due to the accident at the Fukushima Daiichi Nuclear Power Station caused by the Great East Japan Earthquake on March 11, 2011, only 10 units could be restarted, and examinations for other restarts have been prolonged. After the Chernobyl nuclear accident in 1984, Toshiba started developing a simplified BWR (SBWR) in cooperation with GE and Hitachi, and after 1995, started developing a small modular reactor [1–6]. Nuclear power generation is a stable basic power source that does not emit CO2 on the premise of ensuring safety, and has recently been reevaluated as an attractive option from the viewpoint of energy security and environmental protection. Factors such as the recent sluggish power demand, power grid capacity limits, and initial investment limits to avoid risks do not favor large-scale plant output. In order to globalize nuclear power generation to mitigate the greenhouse effect, we need a small modular reactor (SMR) that can be easily adopted in any country and can be modularized and manufactured in factories with short construction periods. The concept of the reactor introduced in this section has a simplified BWR (LSBWR) configuration with a low output, long operating cycle, and comprehensive safety features, which was presented in 1999 at the annual meeting of the JSME [7] and ICONE11 [8] by Narabayashi et al. BWRs have been designed and operated at an operating pressure of 1000 psig (about 7 MPa) [4,5]. In this section, the operating pressure and thermal efficiency are also examined as the ideal form of SMR and next-generation BWR. To be economically competitive, the long operating simplified BWR (LSBWR) design [8] includes system and structural simplifications, modularity for short construction times, and increased availability. Comprehensive safety features are not intended to be evacuated by reliable equipment or systems such as extensive RPV inventory, lower core layout, molten core in-vessel retention (IVR) features, and hybrid ECCS including passive features [9]. IVR is very important, because in case of Fukushima Daiichi NPP accident, the molten core penetrated the lower flange of RPV and fell down to the pedestal floor, causing severe contamination on the surface of components, wall, and floor in the PCV, and making decommissioning work difficult. The concept proposed here is to provide flexibility for different site conditions and power demands, reduce investment risk and promote public acceptance. Finally, the author also introduces a new LLBWR (load-following and long-operating symbiotic BWR for renewable energy), which uses a reactor internal recirculation pump (RIP) for load follow with fluctuating renewable energy and facilitates for stable grid control.

5.2.2

Objective of LSBWR design

The future of nuclear power generation looks uncertain because of increasing competition with other sources of power generation, such as green energy. For mitigating greenhouse effect, nuclear power plants should have merits of stable operation and high-capacity factor, and should be easily adopted in any country required in order to globalize nuclear power generation. The nuclear power generation is generally

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recognized as an attractive option from the viewpoints of energy security and environment protection. The LSBWR design has the following objectives [8]: – –

Economically competitive with other source of power generation. Comprehensive safety feature with no evacuation.

To meet the power output targets, the SMR power range has been chosen from 50 to 300 MWe. To overcome the economical demerit of small-scale reactors, the LSBWR has taken into account the following technologies: – – – –

Simplification of systems by combination of a direct cycle, natural circulation, and hybrid safety systems. Simplification of structures by applying ship hull structure for a building. Modular fabrication method for a short construction period. Improvement in availability by long operation cycle (2–7 years).

The LSBWR safety approach was taken into account for the currently available recommendations for future reactor concepts. For comprehensive safety, no evacuation is achieved by highly reliable equipment and systems as follows: – – – –

large RPV inventory bottom located core layout IVR cooling capability hybrid safety system including passive ECCS and PCV cooling

5.2.2.1 Natural circulation core concept From the perspective of regulation change and the infrastructure of fuel fabrication and handling, the following two-stage approaches are being studied for long cycle core development of LSBWR: – –

The near-term development: Using the conventional core design, long cycle operation is pursued. The long-term development: Considering future deregulation and infrastructure changes, special core design for super-long cycle core is developed.

Though super-long cycle core will be achieved within the limit of uranium enrichment of 20%, it is difficult to develop a fuel cladding material from conventional zircalloy. Years of experience and development of Zircaloy fuel cladding have confirmed that it can withstand irradiation under low neutron flux, it is similar that of naval vessels used under normal combustion conditions, within seven years. Therefore, conventional core designs are therefore based on Zircaloy cladding.

5.2.2.2 Conceptual design of long cycle core of LSBWR A plant design was further simplified from SBWR to fit for a small reactor. The plant concept was named the LSBWR. The power density of the core was decreased and the fuel length was shortened. The decrease in the power density resulted in simplification in the coolant circulation system of the BWR because the natural circulation was high

BWR innovations

437

enough for the core with such a low power density and low pressure drop with shortened fuel. In general, BWRs has a high natural circulation capability in the RPV because of large difference in the coolant density between the inside and the outside of the core shroud. The low power density core resulted in the improvement not only in design simplification but also in the availability of the plant. The low power density lengthened refueling intervals, and consequently enhanced the availability of the plant. As in the previous example, the effective full power year (EFPY) 3 cycles length exceeds 97% in the low power density core with standard 45GWd/t BWR fuel. Major characteristics of the reactor are summarized in Table 5.2.1 together with a cross-sectional view of the reactor with core configuration as shown in Fig. 5.2.1. From the perspectives of simplification and safety, LSBWR reactor system is designed by employing highly reliable equipment design and systems such as. – – – –

large RPV inventory bottom core configuration IVR capability passive ECCS and PCV cooling

Fig. 5.2.2 shows LSBWR reactor internals and configuration. Distinctive features of LSBWR are follows. (a) Natural circulation core cooling: Natural circulation core cooling is applied for eliminating a recirculation pump, which results in high reliability in operation. For attaining natural circulation core cooling, the fuel length is shortened to 2.2 m from conventional 3.7 m to decrease pressure drop. (b) Innovative internal upper-entry CRD and related reactor internals configuration: In BWR reactor design history, lower-entry CRD had been applied for over 30 years. It is difficult to design upper-entry CRD [9] because of two-phase flow and separators and dryers above the upper plenum. The innovative upper-entry CRD was developed to be mounted above the core in the reactor as shown in Fig. 5.2.3. The guide chimney was

Table 5.2.1 Major characteristics of LSBWR. Thermal output Electric output Reactor pressure Steam temperature Core flow rate Steam flow rate Core power density Operating cycle Length Active fuel length Equivalent diameter Number of fuel assemblies No. of control rods

900 MWth 300 MWe 7 MPa, 9 MPa 286 °C 2.7  103 kg/s 0.5  103 kg/s 39 kW/L Cycle 3 years 2.0 m 3.8 m 484 121

Fig. 5.2.1 Core configuration of LSBWR [8].

BAF: Bottom of Active Fuel TAF: Top of Active Fuel GDCS: Gravitationally-Driven Cooling System

Fig. 5.2.2 LSBWR reactor concept [8].

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Lift Coil

Gravity driven Scrum by de-latch

Fig. 5.2.3 Schematic drawing of gravity-driven scrum mechanism [9].

Fig. 5.2.4 Layout corelation of CRD and chimney [8]. designed to have two functions as a CR guide and a two-phase flow path above the core that was separated from the control rod (CR) from core flow. This design had the advantage of being free from flow-induced vibration (FIV) of CR. To avoid an interference of CRDs and separators, offset square layout design between CRDs and separators was adopted as shown in Fig. 5.2.4.

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Boiling Water Reactors

(c) Internal upper-entry CRD: Internal upper-entry CRD was based on electromagnet coupling and motor-driven technologies, as shown in Section 5.7. The internal upper-entry CRD consisted of a CR driving motor, latch mechanism for scram, a position indicator, and an electromagnet coupling, as shown in Fig. 5.2.3. The electromagnet coupling was used to transfer signal and the electric power without direct contact between outside and inside of RPV. All these electric coils were shielded by ceramics that could withstand high temperature. This innovative internal upper-entry CRD was developed under the METI sponsored program. (d) Separator and dryer: Gravitational mist separation was studied as an option for eliminating separator. The gravitational mist separation is performed by low mist velocity and suitable traveling distance. As shown in Fig. 5.2.5, a cylindrical type dryer is located at the top of the RPV. This cylindrical-type dryer was studied from the perspectives of internals simplification and easy fuel handling. It was aimed to provide a large flow passage to effectively remove moisture in steam flow.

On the other hand, Tokyo Tech also developed a flat box-type separator concept by using 3D-CAD that allows the CRD to be taken out from the upper flange of the RPV Fig. 5.2.5 Cylindrical dryer.

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Fig. 5.2.6 Flat dryer [10].

due to the connection between the fuel assembly, control rods, and guide chimney as shown in Fig. 5.2.6 [10]. The width of separator element is approximately 22 cm, and the distance between the CRD drive rod pitch is 1ft. (30.48 cm), so it is structurally possible to place a flat-type dryer between the rods of the CRD. It was confirmed by CFD analyses that the passing flow velocities through the punching metal holes could be made uniform by adjusting the hole diameters. In Figs. 5.2.6 and 5.2.7, it can be seen that the drive shaft of the CRD passes from the upper end of the fuel, along the outside of the guide chimney BOX, through the separator and the flat plate dryer, to the CRD at the top of RPV flange, which confirms that the connection is possible. You can see that it is possible to move the control rods in the reactor up and down with the CRD. The CRD’s drive shaft connects multiple drive shafts with a coupling like the current PWR. If the drive shaft is coaxial and the center rod is pushed, the grip of each of the 4 m drive shafts can be released. This is a mechanism similar to the PWR and CR [8]. As a result of the abovementioned innovative design, following effects were observed for the LSBWR. – – –

The configuration of the natural circulation core and the innovative reactor internals resulted in large water inventory above the core to lead large safety margin against the loss of coolant inventory. RPV and PCV heights were shortened by the internal upper-entry CRD and the guide chimney above the reactor core. CRD drive rods penetrated the steam dryer and CRDs located above a top flange of the RPV.

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Fig. 5.2.7 CRD rods and CR [8].

5.2.2.3 Examination of plant operating pressure and plant thermal efficiency The operating pressure of most BWRs in the world is about 7 MPa. Increasing the operating pressure increases the efficiency of the steam turbine, which in turn increases the thermal efficiency of the plant. If a natural recirculation flow in a core is used, the core flow rate will decrease, thus it seems necessary to reduce the core thermal output density. Therefore, Toshiba tried to examine optimizing the operating pressure based on the previous knowledge of the operating pressure and limit output of the fuel loaded in the core. Although the output density of the core is reduced in LSBWR for long-time cycle operation, Toshiba developed high-performance fuel bundle by using twisted tapes for the ferrule-type spacer, the critical power could be improved by about 20%. Therefore, Toshiba investigated the increase in the operating pressure of the BWR and the thermal efficiency of the plant. (a) Thermal efficiency analysis of plant: Fig. 5.2.8 shows the schematic system diagram of the current ABWR. The ratedoperating pressure of ABWR (hereinafter abbreviated as reactor pressure) is 7.17 MPa, which raises the reactor pressure by 0.14 MPa compared to 7.03 MPa of BWR/5. A moisture

Reactor feedwater pump turbine Moisture separater heater

ABWR Reactor pressure vessel

7624t/h 2.4MPa 415t/h 216oC 186oC

1358MW

HP Turbine Turbine-driven feedwater pump 9MW

7646t/h

1.3MPa 566t/h

Low-pressure condensate pump 3MW

LP Turbine

Generator

Condenser

AESJ

0.4MPa 209t/h

2.2MPa

0.21MPa 198t/h

0.1MPa 216t/h

CF/CD

0.05MPa 313t/h

o

156 C 8.7MPa

High-pressure heaters High-pressure drain tank Reactor Internal Pump 10units 13MW

97oC 139oC 117oC Moter-driven Low-pressure feedwater heaters pump High-pressure drain pump 4.8MW

Fig. 5.2.8 Heat-balance-of-plant diagram of current ABWR [11].

A,B,C 3 series X 4 stage =12units

75oC

49oC 2.8MPa

42oC 0.29MPa

Low-pressure drain tank Low-pressure drain pump 0.9MW

High-pressure condensate pump 5.7MW

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Fig. 5.2.9 Analysis results of reactor pressure effects on plant thermal efficiency [7].

Thermal Efficiency (%)

separator heater (MSH) was added as the reactor pressure increased, and a drain pump-up system was adopted to collect drain water from the feed water heater. The plant thermal efficiency has been increased from 33.40% to 34.54% (at the generating end). The main steam flow rate is 7640 t/h, the turbine is TC6F-52 “(a reheat turbine) with 52in blades at the final stage, and the main steam pressure at the turbine stop valve (NSV) inlet is 6.77 MPa. The steam exhausted from the turbine is dehumidified and heated by the MSH, and sent to the low-pressure turbine. The exhausted steam from the low-pressure turbine is condensed by the main turbine condenser and becomes condensate. It is heated to 216°C (regeneration) by the steam extracted from each stage of turbine with a low-pressure feedwater heater and a high-pressure feedwater heater, and finally feedwater is supplied to the RPV. The thermal output of the core is 3926 MWth, and the electric output (power generation side) of 1356 MWe is obtained. Based on this ABWR heat balance, the plant thermal efficiency was analyzed using a detailed turbine model with the reactor pressure as a parameter. Based on this ABWR heat balance, the plant thermal efficiency was analyzed using a detailed turbine model with the reactor pressure as a parameter. Analysis results of reactor pressure effects on plant thermal efficiency are shown in Fig. 5.2.9. The thermal efficiency is increased from 34.4% to 35.4%. (b) Improvement of critical power of fuel bundle: The BWR fuel bundle tests were performed in General Electric’s (GE’s) ATLAS heat transfer facility. Fig. 5.2.10 [5] shows the critical power, normalized with respect to value at 1000 psia, vs pressure (GETAB, 1973) [4]. The critical power performance of the bundle increases almost linearly with inlet subcooling and monotonically with mass flux. As shown in Fig. 5.2.10, normalized critical power at 1000 psia drops monotonically with increasing pressure within the range 800 < p < 1400 psia. As shown Fig. 5.2.11, the peak critical power is around 4 MPa. Therefore, if the operating pressure of the reactor is increased, the critical power should be decreased. So, when the operating pressure is increased in order to improve the thermal efficiency of turbine system of the plant, the critical power gives the limit of thermal output of the reactor core which can be safely removed from the fuel rod assembly. The critical power is the output at which the liquid film on the surface of the fuel rod dries (liquid film dry out) in the annular mist flow region. If the core pressure is increased from the current 7 MPa to 9 MPa, the limit output will decrease by about 15%. Therefore, if the thermal efficiency is improved, the core thermal output as well as the electric output should be reduced. Therefore, in order to achieve both thermal efficiency and economy of the BWR plant, it is necessary to improve the limit output of the fuel at high pressure. In order to set the target pressure to 9 MPa and to obtain the 37 High-pressure BWR

36

35

34

ABWR

6

7 8 9 Plant Operating Pressure (MPa)

10

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Pressure (MPa) 5

Normalized critical power kWc(P)/kWc(1000)

1.1

6

7

10

9

8

11

No. of Rods 1.0

0.9

12

Corner Peaking

16

1.00

16

1.16-3

16

1.24-1

16

1.23-1

0.8 Bands represent spread of data for four presentative test assemblies and two flow rates

1psia=0.07MPa 0.7 600

800

1000

1200

1400

1600

1800

Pressure (psia)

Relative Critical Power (-)

Fig. 5.2.10 Critical power dependency on pressure-normalized with value at 1000 psia (GETAB, 1973) [4].

1.2

DH=0.015m 1.0

G=1356kg/m2s

Analysis

v

0.8

Data

G=678kg/m2s 0.6 0.4

G=339kg/m2s

0.2 0.0 0

7.2

14.4

Pressure (MPa) Fig. 5.2.11 Critical power dependency on pressure [5].

21.6

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Boiling Water Reactors

Fig. 5.2.12 Configuration of test bundles [7]. (a) Nominal case, (b) improved spacer case.

Improved Spacer

3.7m

(a)

(b)

Fig. 5.2.13 Fuel spacers [7]. (a) Ferrule spacer, (b) improved spacer.

limit output equivalent to 7 MPa, a test was conducted using a simulated fuel assembly using a 4  4 subbundle. Fig. 5.2.12 shows the outline of the test piece used. Fig. 5.2.12a shows the current fuel simulation assembly of the ferrule-type fuel spacer shown in Fig. 5.2.13a. In Fig. 5.2.12b, based on the current simulated fuel assembly, the second, third, and fourth spacers from the most downstream are changed to the twisted tape type-improved spacer named cyclone spacer as shown in Fig. 5.2.13b. When the improved spacer was used, the critical power was increased by about 20% compared to the conventional spacer. Fig. 5.2.14 shows the result of the relative critical output of 9 MPa estimated using test results. The critical power decreases as the reactor operating pressure is further increased to 9 MP, along the thick line in Fig. 5.2.14. It was shown the critical power ratio becomes 1.0 at 9.0 MPa, and almost equal to the value of 7 MPa of the current fuel by adopting the improved spacer. In conclusion, the operating pressure increase for economic improvement is limited to 9 MPa. (c) Merits of high-pressure BWR and technical tasks to be examined: When the reactor pressure is raised to 9 MPa, the thermal efficiency of the plant increases by about 1% and the electrical output increases by about 3% (about 40 MWe). This is a great economic advantage as the nuclear power plant is operated for quite a long period. The wall thicknesses of the pressure vessel and the main steam pipe might get slightly thicker in the 9-MPa nuclear plant than that in the 7-MPa plant, but the diameters

Relative Critical Power (-)

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Fig. 5.2.14 Critical power measurement results (normalized with critical power of normal 8  8 fuel in 7 MPa) [7].

1.4

Improved spacer predicted line

1.2

1.0

0.8

6

7

8

9

10

Pressure (MPa)

of those might be reduced because the specific volume of steam at 9 MPa is smaller than that at 7 MPa by 7.7%. Since the saturation temperature of the reactor water rises by 17°C from 286°C to 303°C when the operating pressure is raised from 7 to 9 MPa, it is necessary to study hydrochemistry such as material corrosion. It is also necessary to increase the discharge pressure of water supply pumps and ECCS pumps from the current value of 9 MPa to 11 MPa, which is not technically a big concern. As for safety evaluations, when the elevated reactor operating pressure is introduced, it is of course required to perform all required safety analyses especially concerning the effectiveness of the improved spacer. Since the 9-MPa specification matches the heat output to 7 MPa (Table 5.2.1), the steam flow rate decreases as the steam enthalpy increases. The pipe diameter and RPV wall thickness will be examined in the future, but a 1% increase in thermal efficiency will lead to a large increase in electricity generation revenue, so it is thought that the fuel verification test cost can be sufficiently absorbed. In addition, it will be possible to achieve high operating pressure in the future by relying on the operating results of the reactor operating pressure, such as from 7.04 MPa (BWR/5) to 7.17 MPa (ABWR).

5.2.3

Safety system and PCV concept

LSBWR safety system featured the application of the passive safety system for the emergency coolant injection and the containment cooling. The emergency coolant injection system consisted of the depressurization valve (DPV) and the gravity-driven core cooling system (GDCS) was able to achieve high reliability by flooding the reactor core following an accident since the reactor core was placed at the bottom of the RPV by adopting the internal upper-entry CRD. This measure is a very important measure against severe accidents and is called IVR (In Vessel Retention). The IVR specifically fills the lower drywell with gravity suppression pool water to cool the bottom flange of the RPV to prevent melting and melt penetration. A top-mounted CRD eliminates the 10 m space below the RPV. As shown in Section 5.7, the top-mounted CRD enable to remove the CRD housing tube from the bottom of the RPV and the space at the bottom of the RPV is reduced, greatly reducing the amount of water required and the water filling time shorten. In this way IVR is greatly facilitated. Three Mile Island Nuclear Power Plant Unit 2 (TMI-2) in the United States achieved IVR by depositing zirconia, which is a zircaloy cladding oxide, on the inner surface of the bottom of the RPV and allowing water to flow in, preventing RPV bottom melt through, molten-core concreate interaction (MCCI), containment vessel contamination, and local town contamination.

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The containment vessel has double steel walls of inner and outer wall with ship hull structure, as shown in Fig. 5.2.15. The gap between the inner and the outer wall was filled with cooling water that is boiled off to the atmosphere to cool the PCV passively during an accident. The concept of double wall containment vessel cooling system is also used for the drywell cooling during normal operation and therefore the drywell arrangement is simplified without drywell cooling component used in the current BWR containment. When cooling water in the PCCS pool above PCV is exhausted, external pool or seawater will be easily supplied by gravitational force and therefore the highly reliable and long-term PCV cooling would be achieved. The performance of the safety system was analyzed for a feedwater line break accident. The analysis was performed using the TRAC code incorporated with the heat transfer models for the natural convection cooling and the steam condensation cooling with a noncondensable gas. The heat transfer models had been developed to estimate the heat transfer coefficients in the containment space and the containment wall coolant channels. The analysis results for the containment pressure transient and the heat removal transient are shown in Fig. 5.2.16. After reaching peak value during the blowdown phase, the containment pressure decreases while the GDCS coolant flow is sufficient to suppress steam production in the reactor core. The containment pressure begins to increase after 3 h since the GDCS flow decreases and the steam is produced by the decay heat. Even if the containment pressure begins to increase, it will be suppressed by the containment wall cooling and is maintained well below the design pressure during 24 h. The heat removal rate of the containment wall cooling becomes almost comparable with the decay heat after 24 h as shown in Fig. 5.2.16. The condensate produced by the containment wall cooling flows from the dry well to the RPV through the GDCS injection line, and the reactor core remains covered with cooling water.

Stiffener (smallbeam) Girder (largebeam) Stain less steel plate Inner PCV wall

Stain less steel plate Outer PCV wall

Water Girder

Fig. 5.2.15 Double PCV walls of the inner and the outer wall with ship hull structure.

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Fig. 5.2.16 TRAC code analysis result of feedwater line break accident [8]. (a) PCV pressure response, (b) PCCS cooling performance.

5.2.4

Module fabrication and construction

Aiming a short construction period and high production quality, module fabrication and construction were studied for the LSBWR. In the system, module means not only system equipment, but also building structure. As the LSBWR was a small-scale plant, it is possible to fabricate, transport, and construct in a one piece whole plant if the site transport condition allows. So far, module construction methods were generally applied to some component assemblies such as a pump and its supporting structures. Since a reactor building was usually a reinforced concrete structure, it was impossible to fabricate component modules with the building module. In the shipbuilding industry, ship hull structure is applied for a large size ship such as a 500,000-ton class. Though the ship hull structure is lighter than the reinforced concrete structure, it has enough strength and appropriate characteristics to apply for a nuclear reactor building. By using this ship hull structure, it is possible to fabricate modules containing RPV and PCV components and parts of the building at a shop at the same time. In the LSBWR building design, the reactor building and the turbine building were combined into one building. Because the LSBWR is small and lighter than medium- or large-size plants, it was possible to mount the turbine system on the upper part of the reactor building. The one building arrangement would reduce the building volume, and the seismic isolation rubber structure for aseismic design would be installed. By applying a seismic isolation structure to the whole building, it enables standardized the LSBWR, such as building modules and reactor vessel, internals and safety cooling systems, regardless of various site-specific seismic conditions. General arrangement of the LSBWR is shown in Fig. 5.2.18. The LSBWR was developed by Heki, Narabayashi, and others at Toshiba in early 2000s, and they collaborated with Aritomi of Tokyo Institute of Technology. After that, in 2018, Narabayashi moved to Tokyo Institute of Technology and installed a reactor internal pump (RIP) to change core flow rate and thermal output. By installing the RIP as an ABWR, the core flow rate changes in response to the change in pump speed, and the void fraction in the core changes, so that the heat output of the core will change rapidly.

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Many SMRs employ natural core circulation, but in the case of BWRs, the core output will depend on control rod operation. Since the ball screw of the control rod drive mechanism requires accuracy, it is necessary to prevent the ball screw from being worn due to the load-following operation. It is more reliable to use RIP for load-following operation. Therefore, in load-following and long-operating symbiotic BWR for renewable energy (LLBWR), it is possible to add a load-following function to control the power output in response to fluctuations in the electrical output of renewable energy. Thus, LLBWR can cooperate with renewable energy. In combination with renewable energy, it would enable achieving carbon neutrality in the 2050s.

5.2.4.1 Ship hull structure for reactor building Ship building technology employs advanced automation and remarkably improved assembly lines as a result of competition in the severe international market. The hull structure of a large ship is constructed simply with steel plates as shown in Fig. 5.2.15. The basic structure of the hull consists of steel plates, girders (large beams), and stiffeners (small beams) as shown in Fig. 5.2.17. Almost the entire process, including the receipt of materials, forming large blocks, the welding process, and the removal of distortion after welding, is performed automatically on an assembly line. The large blocks are transferred to a shipbuilding dock where they are assembled to shape a hull. When most of the construction work is completed, the hull is launched and shipped to the construction site. Ordinary ships have ‘single’ steel plate lattice structure. But the ‘double’ steel plate lattice structure, that is, two steel plates sandwiched together, is

Fig. 5.2.17 LSBWR building design using ship hull structure [8].

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also used for the latest tankers to prevent oil leaks. Ordinary walls and floors of a reactor building can be designed using the single steel plate lattice structure. For places where radiation shielding is required, concrete is filled into the double steel plate lattice structure as one of the various methods. Since supporting brackets for pipings and foundations for machines can be welded to walls or floors directly, the installation work will be simple compared with that in conventional reactor buildings.

5.2.4.2 General arrangement of LSBWR and LLBWR’s building design Improvements had been made to the LSBWR and LLBWR for passive safety measures. For the electric output of a nuclear reactor, 800 MWe was selected from the range of 600–1000 MWe, which was in high demand in Japan and around the world. The internal upper-entry CRD was used to strengthen the direct cooling of the lower part of the reactor core to facilitate the in-vessel retention (IVR) cooling in the event of a severe accident, and the operating pressure was set at 9 MPa to improve the economical-thermal efficiency. The passive safety system was chosen, as it has the isolation condenser (IC), the water wall-type PCCS, and has the ability to perform cold shutdown without operating the ECCS. As the LSBWR and the LLBWR had the GDCS and the PCCS, a raised suppression pool, the GDCS pool, and the PCCS pool were located above the reactor core. These unique building concepts resulted in remarkable reduction in building construction cost by reducing the building volume reduction and the standardized shop fabrication.

5.2.4.3 Construction methodology and evaluation The plant construction methodology was evaluated taking into consideration the characteristics of the LSBWR concept, and various kinds of construction methods were studied. From the ABWR construction experience, it took 48 months from the rock inspection to the commercial operation for a standard ABWR construction including a concrete structure building. This 48-month construction period is shortened to 35.5 months by applying a ship hull structure building [3]. (In our nuclear power industry and general construction industry, the period required for construction work is called the construction period.) It was estimated that the construction period of the LSBWR would be 20–30 months when the hull structure building was introduced. This construction period showed remarkable reduction when compared with the conventional construction period. In module fabrication design, each module block was designed to be divided into modules within 1000 tons so that they can be handled by the largest 1000-ton cranes available today. This fabrication procedure would be applied to the LSBWR and the LLBWR. The shortest construction period of 22 months was attained by one-piece construction method. In the module construction method, the whole plant in one module was transported to the site by a barge as shown in Fig. 5.2.18 (LLBWR). This method suitable for a coast side, a big riverside, and a lakeside that have possible approach routes by barge transportation. The divided module construction period is

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(a)

(b)

Fig. 5.2.18 Building design of LSBWR and LLBWR mounted on barge ship. (a) Shipping of LSBWR by barge ship LLBWR [8], (b) installed integrated module reactor [10].

Fig. 5.2.19 Feature of ship hull structure [8].

estimated to be approximately 30 months. The construction periods mentioned above are influenced by the fuel loading timing and the system inspection timing. The internals of LLBWR and their layouts are explained below using Fig. 5.2.19. This reactor uses a core arrangement of 300 MW of LSBWR, but the fuel length is 4 m, which is the same as the conventional BWR. If the fuel length is 2 m, it must be heated to the same void ratio at half-length of fuels, so the linear power density must be increased. If the length is 4 m, the load capacity of uranium can be increased without increasing the line output density, so it can be used for long-term cycle operation. If six RIPs are used, the rated output of 800 MW can be obtained by using six RIPs, which is almost the same power output per RIP as the ABWR’s rated electrical output of 1356 MW as follows: ð1356 MW=10Þ  6 ¼ 814 MW

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If 20 units of LLBWRs are built in response to the decommissioning of thermal power plants, the load following capacity will be (800–300)  20 units ¼ 10,000 MW ¼ 10 GW of adjustment power by the load-following operation is implemented to respond to fluctuations in the electrical output of renewable energy such as solar and wind power generation due to weather, and to rapidly change the electrical output of nuclear power plants to generate the required power. It will be possible to stabilize the system of the power grid together with pumped-storage power plants of 10 GW without no emission of CO2. By installing the built-in CRD, the core position was lowered and a guide chimney was installed at the core outlet to enhance natural recirculation. With the low-pressure loss separator, the separators can be arranged in a square grid. Since the width of the dryer element is sufficiently narrower than the width [1 ft. (30.48 cm)] of the core pitch of the flat plate dryer, the drive shaft guide pipe of the CRD can be passed through. In this way, the drive shaft can be connected from the grip part of the control rod drive shaft at the same height as the upper end of the fuel to the CRD installed on the upper part of the RPV by adding the four drive shafts with each length of about 4 m. As with the PWR, the drive shafts are connected each other by using the push rod installed in the center of the coaxial rod in order to attach and detach the shaft. Development of LLBWR using RIP for symbiotic BWR for renewable energy is possible by this feasibility study, building layout of LLBWR is almost the same as shown in Fig. 5.2.20. Steam turbine, generator, and condenser are installed in the same reactor building of ship hull structure. Isolation condensers (ICs) are installed in the IC pool. Decay heat is removed from both IC pool water and water in the ship hull structure. The primary containment vessel (PCV) is naturally cooled through “a water wall”, filled with water in the double walls of the ship hull structure used in the suppression pool’s double walls as shown in Fig. 5.2.21 [10, 12]. The specific volume, the volume per electric power output (m3/kWe) the specific weight, and the weight per electric power output (ton/kWe) of the building of the Fig. 5.2.20 Development of LLBWR using RIP for symbiotic BWR for renewable energy [10].

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Fig. 5.2.21 Building layout of LLBWR [10].

Fig. 5.2.22 Comparisons of specific volume and weight of building of LSBWR and ABWR.

LSBWR are compared with those of the ABWR building in Fig. 5.2.22. Those of the LSBWR are approximately reduced to three-fourth and three-fifth of the ABWR, respectively. These results show the effects of the innovative concept such as the one-piece building arrange, the ship hull structure building, and system simplification. It is expected that these result in effective cost reduction in spite of the economical demerits of scale down.

5.2.5

Summary of design innovation of LSBWR, LLBWR, and high-pressure BWR

Considering requirements of next-generation reactors, a long-cycle operation simplified-BWR (LLBWR) concept is proposed. The major features of LSBWR are as follows:

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(1) Long-cycle-operation core using uranium fuels. (2) Simplified system and component as well as passive systems. (3) Combined building concept with ship hull structure by modular fabrication.

These concepts have potential to reduce construction cost and to increase capacity factor after starting operation. Safety feature of the LSBWR and LLBWR make it possible to eliminate an evacuation plan requirement even in the case of a severe accident. Further researches and developments are underway.

References [1] Y. Ronen (Ed.), High Conversion Water Reactors, CRC Press, 1990. [2] M.D. Carelli, D.V. Paramonov, C.V. Lombardi, M.E. Ricotti, N.E. Todreas, E. Greenspan, J. Vujic, R. Yamazaki, K. Yamamoto, A. Nagano, G.L. Fiorini, K. Yamamoto, T. Abram and H. Ninokata, IRIS, International New Generation Reactors, Proceedings of the ICONE-8, Baltimore, USA, (April 2–6, 2000). [3] M. Kanno, T. Maruyama, T. Yamakawa, Study of application of ship structure to nuclear buildings, in: International Workshop on Utilization of Nuclear Power in Oceans N’ocean 2000, Tokyo, Japan, Feb. 21–24, 2000, pp. 84–94. [4] NEDO-10958, General Electric BWR thermal analysis basis (GETAB), 1977. [5] R.T. Lahey Jr., F.J. Moody, The Thermal-Hydraulics of a Boiling Water Nuclear Reactors, in: ANS, 1979. [6] Y. Yamamoto, T. Mitsutake, S. Morooka, R. Yoshioka, Critical power dependency on pressure, in: International Seminar on Subchannel Analysis 3 (ISSCA3), 1995, pp. 63–72. [7] T. Narabayashi, Y. Yamamoto, S. Morooka, K. Shirakawa, Y. Asanuma and S. Yajima, Examination of plant operating pressure and plant thermal efficiency for BWR, Proc. Ann. Meet. JSME, Vol. 4, pp.335–336, (Jul. 1999). (In Japanese). [8] H. Heki, M. Nakamaru, T. Maruyama, K. Hiraiwa, K. Arai, T. Narabayashi, M. Aritomi, Development of Small Simplified Reactor, in: ICONE11–36100, Tokyo, Japan, 2003. [9] T. Narabayashi, M. Sato, T. Kameda, S. Kawano, T. Hagiwara, C. Iwaki, T. Tokumasu, N. Kobayashi, K. Araoka, T. Terashima, R. Sugawara, S. Ishizato, Development of Internal CRD for Next Generation BWR, in: ICONE13-50783, Beijing, China, May 2005. [10] T. Narabayashi, T. Vien, H. Takahashi, H. Kikura, New useful SMR for load follow with fluctuating renewable energy and enhance facilitates for stable grid control, in: The 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS 13), Taichung, Taiwan, 2022. [11] S. Ohmori, M. Mori, T. Narabayashi, M. Nakamaru, Y. Asanuma, M. Yasuoka, Development of steam injector feedwater heater system, in: Proceedings of International Conference on Nuclear Engineering, ICONE-8582, Baltimore, USA, 2000. [12] T. Narabayashi, Proposals that contribute to the innovation of SMR, such as LBB, Seismic Isolation Structures, Ship Hull Structures, and Application of General Industrial Products, IAEA Technical Meeting on Codes and Standards, Design Engineering and Manufacturing of Components for, Small Modular Reactors, 10–13 May 2022 (Virtual Event), Ref. No. EVT2103861.

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Power uprate in BWR Michitsugu Mori Graduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan

5.3.1

Current status and trend of reactor power uprates

In the basic concept of maximizing the use of existing nuclear power facilities on the premise of maintaining and improving safety, the advanced use of light water reactors, such as power enhancement, has already been implemented, especially in Europe and the United States to maximize the use of reactor power and increase the electrical output. Compared to the expansion of new power plants, there is less risk in terms of licensing and investment, which encouraged many countries to implement reactor power uprate as an effective means of reducing greenhouse gases and ensuring energy security. There are two types of power uprate to gain larger electric power more efficiently, which have been carried out to date: one is to increase electric power output without changing equipment and rated thermal power generated by fission in a core, which is called the constant rated reactor thermal power operation, and the other is to uprate the licensed rated thermal power of the reactor with changing equipment to achieve a large increase in electric power output.

5.3.1.1 Benefits and safety of constant rated reactor thermal power operation Constant rated reactor thermal power operation without facility modification to increase electric output is an operation method that enables more effective use of resources and facilities than rated electric power operation, and is implemented worldwide. In rated electric power operation, the reactor thermal power varies according to the plant efficiency, which varies with the circulating water (seawater) temperature. For example, when the seawater temperature decreases and the plant efficiency increases in winter, the plant thermal power is reduced from the rated reactor thermal power of 100%, and the electric power output is kept constant at the rated electric power of 100%. In other words, in the electric power rated operation, the allowable thermal margin is not used for power generation; on the contrary, in constant rated reactor thermal power operation, power can be generated up to the uprated electric power of 105% at the licensed rated reactor thermal power of 100% without facility modification. This allowable electric power margin can be deduced from Appendix 1 of the “Regulations Concerning the Installation, Operation, etc., of Nuclear Reactors for Commercial Power Generation” as amended by the Nuclear Regulatory Authority of Japan Regulation No. 3 of 2020 [1], which states that the prior notification is

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required in 3 rector cooling system facilities 6 (2) for “Modification of steam turbines involving changes in rotational speed or the rated electrical output increase more than 5%.” In other words, unless the uprate is more than 105% of the rated electrical output, the prior notification is not required and utilities can implement the modification on its own initiative. In the safety assessment of nuclear reactors on the Nuclear Reactor Regulation Law, safety is verified based on the thermal power of the reactor at the installation licensing, and the same applies to the approval of the security regulations, where the thermal power of the reactor is set as the operating limit. Electric power generated by steam turbines and generators is not subject to the reactor safety regulations. Unless the licensed reactor thermal power is changed, there will be no safety issues on the reactor side even if the electric power output is changed, and there is no need to change the framework of the maintenance code on safety regulations for operation and management.

5.3.1.2 Possibilities and issues on constant rated reactor thermal power operation A steam turbine control (governor) valve is located at the steam turbine inlet in the main steam line to adjust the steam flow rate into the steam turbine. The steam turbine control valve is designed and manufactured with a sufficient margin so that 100% of the rated steam flow fully enters the steam turbine, and the rated steam flow of 100% can be achieved without fully opening the steam turbine control valve. In the case of constant rated electric power operation in the higher-efficiency winter season, the main steam flow rate, steam flow rate at the steam turbine inlet, and feedwater flow rate are lower than that in summer season, but the electric power output, main steam pressure, and feedwater temperature are constant. On the other hand, in the case of constant rated reactor thermal power operation, the steam turbine output, that is, the rotation power of steam turbine and the electric power output become larger and may exceed the maximum value throughout the year. Effects of constant rated reactor thermal power operation varies unit by unit of the plant; however, under the ideal condition of the steam turbine system such that no contamination exists in the heat exchanger tubes, the maximum electric power that can be obtained by operating the reactor at rated thermal power is expected to be 101%–108% of the rated electric power [2]. Operating a nuclear power plant in such a way as to generate energy most efficiently while ensuring safety is beneficial from the perspective of effective energy utilization. As will be discussed later, it is possible to uprate the reactor power by improving the accuracy of feedwater flow measurement without changing the reactor design or without increasing the number of fuel rods in the reactor core, which is economically advantageous.

5.3.1.3 Current status of reactor power uprate with equipment modification There are currently three categories of reactor power uprates with equipment modifications as follows [3]:

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Measurement uncertainty recapture (MUR/MU)-type power uprates; Stretch (S)-type power uprates; Extended (E)-type power uprates.

The US Nuclear Regulatory Commission (NRC) approved the stretch (S)-type power uprate in 1977, followed by the extended (E) type in 1988. Measurement uncertainty recapture (MU)-type was approved in 1999. A total number of 170 units of the S-, E-, and MU-type uprates have been approved by March 10, 2020 [3]. The MU-type power uprate is an uprate in which the power increases by recapturing the uncertainty assumed in the safety analysis where the measurement uncertainty of the reactor power is improved. In the safety analysis for the case of an emergency core cooling system (ECCS) working, the initial reactor power is set to 102% of the rated reactor power, because of considering 2% error in the measurement of feedwater flow rate, etc. It is applied to the licensing process to recapture this uncertainty, compress the margin, and uprate the rated power operation from 100% to less than 102%. The power uprate due to the recapture of the measurement uncertainty requires more accurate measurement system of the feedwater flow rate to calculate more accurate reactor power. Leading edge feedwater flow measurement devices, which are exhibited in Table 5.3.2, are applied to compress the uncertainty in the power level and to ensure the safety shut down of the reactor under expected accident conditions. The S-type power uprate uses the existing design margins of the plant equipment and therefore do not require significant plant modifications. The US NRC defines the S-type power uprate as a power uprate of less than 7% of the originally approved reactor thermal output [3]. While this is within the design capacity of the plant and remains in the category of stretch power uprates, the actual value of power uprate that a plant can achieve varies plant by plant and depends on the operating margins included in the design of the particular plant. The S-type power uprate typically involves changes in equipment set points, but do not involve major changes to the plant. The E-type power uprate is larger than the S-type power uprate. In the United States, NRC defines an E-type uprate as a power uprate of 7% or more, and uprates as high as 20% have been approved [3]. The E-type power uprate requires significant changes to major balance-of-plant equipment, including high-pressure turbines, condensate pumps, motors, main generators, and transformers.

5.3.2

Reactor thermal power and electric power

The relationship between reactor thermal power and electric power output is briefly described. The steam generated in a BWR core is supplied to the steam turbine as shown in Fig. 5.3.1, and electricity is generated by the generator installed in tandem with the turbine main shaft by rotating the rotor blades of the steam turbine. The steam then leaves the turbine and enters the condenser, where it is cooled and condensed back to water, which is fed to the reactor by the feedwater pump. The condensation of the steam reduces its volume, the phenomenon of which results in a vacuum state in a condenser. The condenser vacuum (level) indicates how close the internal pressure of the condenser is to the vacuum.

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Fig. 5.3.1 Path of steam from turbine and condenser to feed water pump and cooling by circulating water (sea water) in condenser with arrangement of generator connected tandem to turbine.

5.3.3

Reactor power uprate with constant rated reactor thermal power operation

The BWR thermal power output is obtained by subtracting the energy of the incoming feedwater from the energy of the outgoing steam from the reactor. In other words, BWR thermal power is expressed as {(energy of steam flowing out of reactor) (energy of feedwater flowing in) – (others)}, which almost equal to {(steam flow rate  specific enthalpy of steam)  (feedwater flow rate  specific enthalpy of feedwater)}. In BWRs, the others are seal-water enthalpy of control rod drives, enthalpy of input and output reactor clean-up water, enthalpy of recirculation-pump heating, and heat loss from the system [4]. The heat balance in BWR mentioned above can be expressed by the following equation [5]:
 Q ¼ W s  hs  W fw  hfw  W cr  hcr + W cu  ðhcui  hcuo Þ  W rec  Δhpump + Qloss where, Q: Reactor thermal power output. Ws: Flowrate of main steam. Wfw: Flowrate of feedwater. Wcr: Flowrate of control rod drive system. Wcu: Flowrate of reactor water clean-up system. Wrec: Flowrate of recirculation system. hs: Specific enthalpy of steam.

(5.3.1)

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hfw: Specific enthalpy of feedwater. hcr: Specific enthalpy of control rod drive system. hcui: Specific enthalpy of reactor water clean-up system inlet. hcuo: Specific enthalpy of reactor water clean-up system outlet. Δ hpump: Recirculation pump heating. Qloss: Heat loss.

For reference, the PWR thermal power output is expressed as {(steam-specific enthalpy generated from SG  steam flow rate)  (SG feedwater-specific enthalpy  feedwater flow rate) + (enthalpy released from SG as blowdown water)}[6]. In constant rated reactor thermal power operation, the reactor thermal power is kept at the “rated thermal power,” which is the licensed maximum value allowed by the permit to install a nuclear reactor and the electric power is changed as long as efficiency allows to match the rated thermal power. Here, the reactor water level in a BWR is kept constant at a normal level by the three-element control with feedwater flow, steam flow, and reactor water level. In constant rated electric power operation, the electric power is kept at the “rated electric power,” which is the output value that can be used to generate a constant electric power throughout the year, and the thermal power is changed to match the rated electric power, so the difference between the maximum rated thermal power and the necessary thermal power to generate the rated electric power is not used for power generation. In general, the thermal efficiency decreases in summer season as shown in Fig. 5.3.2, since the seawater temperature is higher and the condenser vacuum is lower in summer than in winter season [2]. That is, even with the same thermal output and the same steam turbine inlet pressure, the thermal energy of the steam consumed by the steam turbine is greater in winter than in summer seasons because of the higher

Fig. 5.3.2 Effect of condenser cooling water (seawater) temperature on thermal efficiency.

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condenser vacuum in winter than in summer season, so that more electricity is generated in winter, as can be seen from Fig. 5.3.2 and the following equation: Thermal efficiency ð%Þ ¼

5.3.4

Electrical output Reactor thermal output

(5.3.2)

Relationship between reactor thermal power and electric power outputs

In the case of constant rated reactor thermal power operation, the electric power output is almost the same as the rated electric power in summer due to lower thermal efficiency, though depending on the seawater temperature, but it will exceed the rated electric power in other seasons, especially in winter season. Fig. 5.3.3 shows the concept of the variation of reactor thermal and electric power outputs with seasons. The figures on the left-hand side of Fig. 5.3.3 show the reactor thermal power (top) at constant electric power operation, and the figure on the bottom shows the variation of condenser vacuum. After the electric power reaches 100% rated power operation, the reactor thermal power is operated below the reactor rated thermal power due to increase in the thermal efficiency with increase in the condenser vacuum from summer to winter seasons, and the thermal power that enables to generate electric power beyond the 100% of rated electric power is not used to generate power and is dissipated. The figures on the right-hand side of Fig. 5.3.3 show the variation of the electric power output (top) during constant rated reactor thermal power operation. After the reactor thermal power reaches 100% of reactor rated power operation, the increase in thermal efficiency according to the increase in the condenser vacuum, shown in

Fig. 5.3.3 Variations in electric output by season.

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the bottom figure, is used to generate electric power and the reactor is operated above the rated electric power of 100% [2].

5.3.5

Issues and safety in constant rated reactor thermal power operation

In the transition from constant rated electric power operation to constant rated thermal power operation, the pressure difference between the steam turbine inlet and outlet becomes larger in winter than in summer because the condenser vacuum increases due to the lower seawater temperature, as described in Fig. 5.3.3. Effects of constant rated reactor thermal power operation vary by plant; however, it spans about 101%–108% of the rated electrical output, and the annual facility utilization rate, which is defined in Eq. (5.3.3), is about 101%–105% [2]. Annual facility utilization rate ¼

Amount of electricity generated Turbine nominal output  Calendar time (5.3.3)

When the rated reactor thermal power operation is implemented, no safety problem is considered to occur for the power generation unless the thermal power of the rated operation is changed. The maximum values exceeding the rated condition are expected for the output and rotational powers of the steam turbine. Therefore, the issues to be confirmed are listed with regard to the integrity and safety of the steam turbine and generator at uprate conditions exceeding the rated electric power in a BWR [2]. (1) Increasing the rotating speed of the steam turbine shall not require any change to the conventional assessment that there is no need to consider the impact of turbine missile events by internal flying objects in the nuclear power plant. With regard to the turbine-missile phenomenon, which is assumed to occur when the power output of a steam turbine exceeds its rated electrical power, the impact of a turbine-missile event on the reactor and the spent fuel storage pool is currently being assessed under a hypothetical rotating speed of the steam turbine up to 120% of the rated speed for a BWR. It was confirmed that these assessments were based on a sufficient margin. (2) The increase in the output of the steam turbine shall not cause any problems in the integrity of the steam turbine equipment. With regard to the detailed design of steam turbine equipment, it has been confirmed that the current detailed design is acceptable, as the manufacturer’s strength design encompasses the maximum operational electrical output condition. In practice, typical installations were built with sufficient strength margins, so that constant rated thermal power operation will not cause strength problems. (3) The increase in electrical output shall not cause any problems with the integrity of the electrical equipment (generator, main transformer). The generator’s and steam turbine’s main shafts are designed under the same conditions, so there shall be no strength problems even during constant rated thermal power operation. The generator and main transformer temperatures can be kept below the temperature limit even at maximum rated operating power, so there shall be no integrity problems even at constant rated reactor thermal power operation.

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(4) The state of increasing electrical output shall be monitored and controlled. The currently installed equipment should be confirmed that their possible monitoring and control ranges shall cover the ranges of the measurement and control over the constant rated thermal power operation. On evaluating above, it shall be considered that the required performances and strengths against possible issues in the turbine missile event, steam turbine, and electrical equipment including their measurement and control performances on constant rated reactor thermal power operation were all within the current design margins, where no safety issues shall remain. Therefore, the constant rated reactor thermal power operation can be carried out safely without changing the currently installed equipment and devices.

5.3.6

Experiences in BWR operation with constant rated reactor thermal power operation

Constant rated reactor thermal power operations were actively practiced in many countries including Japan for both BWR and PWR. The results of the constant rated reactor thermal power operation at TEPCO’s Kashiwazaki-Kariwa NPP (Nuclear Power Plant) in Japan are shown in Table 5.3.1 [7]. Kashiwazaki-Kariwa NPP has five BWR-5 units with 1100-MWe rated power and two ABWR units with 1356-MWe rated power. In February 2011, the temperature of the seawater at Kashiwazaki-Kariwa NPP was around 10°C, which is much lower than the design value of seawater intake temperature of around 40°C. The total power output of the four units shown in Table 5.3.1 were increased to about 113 MWe, saving the fuel cost of fossil-fuel-fired power plant generation and its CO2 emissions.

Table 5.3.1 Operation status of TEPCO’s Kashiwazaki-Kariwa Nuclear Power Station (February 2011). Items

Unit

Reactor type

Rated power (MWe)

1 5 6 7

BWR BWR ABWR ABWR

1100 1100 1356 1356

Start of commercial operation

Capacity factor (February 2011) (%)

Sep. 1985 Apr. 1992 Nov. 1998 Jul. 1999

102.4 102.1 103.6 101.8

Extracted and edited from Niigata Prefecture Disaster Prevention Bureau, Nuclear Safety Division, Status at Kashiwazaki-Kariwa Nuclear Power Station for February 2011, 2011 (in Japanese).

464

5.3.7

Boiling Water Reactors

Power uprate with equipment modification

The Nuclear Regulatory Commission (NRC) of the United States categorized the power increase licenses into three types as described in Section 5.3.1.3. The first is the MUtype (measurement uncertainty recapture), which targets a power increase of 2% or less by improving the measurement uncertainty of the ultrasonic flowmeter used as the reactor feedwater flowmeter, and the second is the S-type (stretch), which targets a power increase of 7% without major modification within the performance range of the current plant. There is also the E-type (extended), which has a track record of approving output increases of up to 20%, requiring modification of major equipment. In the United States, since the 1970s when the use of power enhancement began, the NRC has approved 170 power enhancements by January 2021. This has resulted in an output enhancement of about 24 GMWt or about 8 GWe [3]. This power generation capacity is equivalent to building about eight new reactors with an electrical uprated output of 1000 MWe, and has contributed to the reduction of greenhouse gases at a low cost.

5.3.7.1 Uprate by measurement uncertainty recapture By improving the measurement uncertainty of the reactor feedwater flowmeter, power uprate with the MU-type can achieve a reactor power increase to about 2% without major changes in reactor equipment or safety measures [8]. Fig. 5.3.4 illustrates an example of heat balance in a BWR [5]. The reactor thermal power can be calculated almost exactly from the difference between the products of the feedwater flow rate and the main steam flow rate by the respective enthalpies. In Fig. 5.3.4, the others only account for less than 1% of the enthalpy possessed by the feedwater and main steam [9]. In the heat balance equation shown in the figure, the product of feedwater flow rate (Wfw) and feedwater-specific enthalpy (hfw) has a large effect on the reactor thermal power (Q). In the current safety analysis of nuclear reactors, this uncertainty is taken into account as about 2%, and the safety analysis is performed with 102% output as the initial setting value. Therefore, if the measurement accuracy error of the feedwater flow rate is reduced, then that amount can be used to increase the power. In other words, if the measurement uncertainty of the feedwater flow rate is reduced to 1% or less, then safety can be ensured even if operation is continued at 101%, and the reactor power can be increased by improving the measurement uncertainty, while ensuring safety without changing the reactor equipment or safety measures at all. Fig. 5.3.5 illustrates this concept of uprate mentioned above. Currently, the safety analysis is performed with the initial value of 102% power, taking into account the feedwater flow rate error and other errors of 2%, to ensure safety. As shown in the right-hand side of Fig. 5.3.5, if the feedwater flow measurement error is improved to 0.5%, the 1.26% improvement in feedwater flow measurement error can be used to increase the power without changing the initial reactor thermal power of 102% in the safety analysis [10]. The uprate based on this error analysis is the MU-type uprate, which has already been introduced in Europe, the United States, and other countries.

Fig. 5.3.4 Example of heat balance in a BWR to evaluate the reactor thermal power.

Fig. 5.3.5 Uprate of reactor thermal power by improving the accuracy of feedwater flow measurement in MU (Measurement Uncertainty Recapture).

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Boiling Water Reactors

Therefore, if the measurement accuracy error of the feedwater flow rate is reduced, then that amount enables us to make reactor power uprate. In other words, if the measurement uncertainty of the feedwater flow rate is reduced to 1% or less, then safety can be ensured even if operation is continued at 101%. In other words, by improving the measurement uncertainty, the reactor power can be increased while ensuring safety without changing the reactor equipment or safety measures at all.

5.3.7.2 High accuracy leading edge flowmeter (LEFM) for nuclear reactor feedwater measurement in MU Currently, flow nozzles and orifice flowmeters are installed as feedwater flowmeters; on the other hand, ultrasonic flowmeters are considered for MU-type reactor power uprate by improving the measurement accuracy of feedwater flow rate. There are three types of ultrasonic flowmeters for MU-type uprate [8]: Chordal-type LEFM (leading edge flow meter); external-type LEFM; CROSSFLOW-type LEFM, as shown in Table 5.3.2, are currently applied for MU uprate. Chordal-type LEFM, which is licensed by USNRC for MU uprate, is an ultrasonic time-of–flight (TOF) flowmeter. The mean flow velocity profiles along four chordal paths are measured on four or eight measuring lines of the pipe cross section, and the flow rate is obtained using the Gaussian quadrature method and corrected by the profile factor (PF). Chordal-type LEFM is reported to be less dependent on flow rate (Reynolds number) and has a stable profile factor, since Chordal-type flowmeter method is based on the Gaussian quadrature method and the flow velocity profile can be well approximated by the seventh-order (¼2n  1, n ¼ 4) equation for four measuring lines. Manufacturer’s nominal error is 0.3%–0.5%. The external LEFM is an ultrasonic TOF flowmeter that measures the ultrasonic transit time difference and then determines the mean flow velocity on a measuring line. It is easy to install because it is externally mounted clamp-on flowmeter. Flowrate is obtained from the transit time difference by correcting with PF without the measurement information on the flow velocity profile. It is necessary to clarify the flow rate (Reynolds number) dependence of PF to guarantee highly accurate measurement. Manufacturer’s nominal error is 1%. CROSSFLOW ultrasonic flowmeter, assuming that the characteristics such as eddies are retained in the flow direction of the pipe, emits ultrasonic waves perpendicular to the flow at two distanced cross sections in the pipe to identify the flow cross sections at two measuring points by correlation method, and measures the travel time between the two cross sections at measuring points. The flow rate is then estimated from the travel time between the two cross sections using assumed velocity profile and correction factor PF. The dependence of PF on the flow velocity (Reynolds number) needs to be clarified. The NRC staff has suspended its approval of WEC’s topical report on the CROSSFLOW flowmeter for new and future use until the staff’s concerns are resolved [8,11].

Table 5.3.2 Ultrasonic flowmeters currently applied or under consideration for reactor feedwater flowmeter in MU (Measurement Uncertainty Recapture).

FM Type

Chordal-type LEFMa

External-type LEFMa

Crossflow-type FM

The Chordal-type LEFM is the time-of-flight (TOF) ultrasonic flowmeter with 4 or 8 chordal paths. Measuring ultrasonic transit time differences on the measurement lines along chordal paths in a pipe cross section, the mean velocity profile over chordal paths using Gaussian quadrature integration is obtained and then converted to flowrate using the Profile Factor (PF). The Chordal-type flowmeter was licensed by US-NRC for MU.

The external-type LEFM is a clamp-on type TOF ultrasonic flowmeter. Since the difference in ultrasonic transit time between the upstream and downstream directions of ultrasonic waves in the same path is proportional to the mean velocity of the fluid, the PF, which is a correction factor for the velocity profile dependent on the Reynolds number, is necessary to determine the flow rate in the pipe.

The Cross-flow flowmeter uses, providing that the features of the flow cross section such as turbulent eddies are maintained, a correlation method to identify these features of flow cross section at two pipe distanced cross sections, and the flow velocity is obtained from the traveling time with the length along the pipe between the two pipe cross sections. The velocity profile in the pipe is assumed on the flow velocity and the a flow rate is estimated using by correction factor PF.

Ultrasonic-Doppler flowvelocity-profile measurement-type FM

Measuring arrangement

Measuring principle

a

Leading edge flow meter.

Ultrasonic-Doppler flowvelocity-profile measurement flowmeter can measure the true flow rate by integrating the flow velocity profile obtained by repeating the Doppler-pulse signal without using PF, which has a significant impact on the measurement value and error. The flow rate can be measured with high accuracy regardless of the Reynolds number (flow velocity profile) varying without using PF by measuring from clampingon points of multiple ultrasonic flowmeters, for instance, at 90° each on the pipe.

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Boiling Water Reactors

5.3.7.3 Inevitable issues on the accuracy of the PF in high accuracy ultrasonic flowmeters and new-concept flowmeter possibility Currently, the PF is generally determined and extrapolated with a Reynolds number of 106 or lower, which is almost an order of magnitude smaller than that for the actual feedwater flowmeter of a nuclear power plant. Extrapolation of experimental values at low flowrates to estimate one at high flowrates (high Reynolds number) is not warranted. Experiments at high Reynolds numbers of 106 or more are possible using a national standard loop that can reproduce the actual feedwater flowrate Reynolds number of a nuclear power plant. For example, the National Institute of Advanced Industrial Science and Technology (AIST, Japan) operates a large standard loop to achieve higher flowrate experimental conditions with a Reynolds number of approximately 1.7  107 corresponding to BWR and PWR [9]. Ultrasonic flowmeters are based on the principle of the transit time difference method, which utilizes the phenomenon that the propagation speed of an ultrasonic pulse is affected by the velocity of the fluid [12]. By measuring the transit time of the pulse in the flow, the flow rate can be expressed in a simple generalized form as follows [13]: Q¼K

t1  t2 t1  t2

(5.3.4)

where t1 and t2 are the transit time along the flow and against the flow, respectively. K is a factor to convert this quantity to volumetric flow rate and is called “profile factor.” However, except for K, the right-hand side of Eq. (5.3.4) contains no information other than time. Consequently, K must be a function of many factors and parameters related to measurement configurations as well as flow configurations such as acoustic path length, cross section and velocity profile of the flow, etc. On the other hand, in order to eliminate the aforementioned errors in the PF evaluation, direct integration of actual flow-velocity profiles in a pipe is necessary to obtain the true flow rate. In order to obtain the true flow rate exactly, a threedimensional flow map must be obtained in a time-dependent manner, and the true flow rate Q(t) can be obtained by the following equation: ð ðð QðtÞ ¼ V ðtÞds ¼ V x ðr, θ, tÞrdrdθ

(5.3.5)

where Vx is the axial component, r and θ are the radial and angular position, respectively, and s is an aerial element perpendicular to the tube axis of the flow in the pipe cross section [14]. Comparing the definition with Eq. (5.3.4) for ultrasonic flowmeter, it could be easily seen that all the important factors and parameters such as velocity profiles are squeezed into the profile factor, K. The integral in Eq. (5.3.5) is an area integral, whereas the integral in Eq. (5.3.4) is a line integral (line average propagation velocity).

BWR innovations

469

These differences in the process of determining the flow rate and the simplification of unknown influencing factors collectively should be pessimistic when it comes to the accuracy of the flow measurement. Therefore, the instrument is calibrated very carefully at the standard loop, especially as a function of Reynolds number and upstream conditions such as bends and elbows. It appears that the interpolation of K-values against Reynolds numbers is generally accepted and seems to work, but the extrapolation is not generally accepted and there seems to be no guarantee of reliability [13]. The concepts mentioned above, in order to eliminate the aforementioned errors in the PF evaluation, were adopted in the ultrasonic-Doppler flow-velocity-profile measurement flowmeter [14] and also shown in Table 5.3.2. It is a flowmeter that differs from three other flowmeters in measurement principles. It enables to directly measure flowrates by integrating the measured velocity profiles, eliminating the PF. Since continuously repeated Doppler pulses are applied to measure velocity profiles on the measuring lines, where microreflectors such as microbubbles and microparticles are required as reflectors in the fluid. The calibration tests were carried out with this flowmeter at the national standard loops of four countries as shown in Table 5.3.3, in which the nominal errors were within 0.5% for water [15]. It was partially commercialized as a general flowmeter [16].

5.3.7.4 Recent implementation and issues of uprates in the United States With regard to the impact in the power-uprated BWR, especially by S and E types, the impacts for the NSSS (Nuclear Steam Supply System) affect the margins on nuclear and thermal designs. The nuclear and thermal margins of the core and fuel are expected to be smaller during normal operation, abnormal transients during operation, and accidents due to the power uprate. As for the BOP (Balance of Plant), the residual heat of the fuel increases due to the power uprate, and then the residual heat removal system in normal operation and the emergency core cooling system in an accident may be affected. As the main steam flow rate increases with the increase in power output, the turbine waste heat increases and the amount of heat exchanged in the condenser increases. Therefore, the temperature at the outlet of the condenser cooling water (seawater) and the circulating water system are affected. Increased power affects structural materials. An increase in neutron flux will accelerate the embrittlement of the RPV and in-core structures. In addition, radioactive corrosion products and radioactive materials such as iodine may increase. As the main steam flow rate increases due to the increase in thermal power, the maximum pressure applied to the reactor pressure boundary increases when the main steam isolation valve is closed, which may affect the capacity of the main SRV (steam relief valve). As the load increases due to the increase in power output, there may be an increase in stress on the turbine rotor, etc., and an impact on the generator system equipment. Increased feedwater and condensate flow velocities may also increase the structural vibration, accelerate material erosion, and FAC (flow-accelerated corrosion). The

470

Boiling Water Reactors

Table 5.3.3 Calibration test results in the national standard loops by ultrasonic-Doppler flowvelocity-profile measurement flowmeter. The calibration tests that carried out at the national standard loops in four countries exhibited in references [13,15] were summarized in Table 5.3.3. National standard NIST 1st (U.S.A) 100A NIST 2nd (U.S.A) 100A NMIJ (Japan) 400A

NMi (Netherlands) 150A

CENAM (Mexico) 100A

CENAM (Mexico) 200A a

Fluid

Reynolds number

Deviation in meter (%)

Expanded uncertainty (%)

Water Water Water Water Water Water Water Water Water Water Water Kerosene Kerosene Kerosene Water Water Water Water Water Water Water Water Water Water Water

4.0E+ 05 2.7E+ 06 8.3E+ 05 7.2E+ 05 6.0E+ 05 2.2E+ 06 1.7E+ 06 1.1E+ 06 1.8E+ 05 1.3E+ 05 8.6E+ 04 8.5E+ 04 6.2E+ 04 4.1E+ 04 4.2E+ 05 6.2E+ 05 3.1E+ 05 6.4E+ 05 2.1E+ 05 5.2E+ 05 1.3E+ 06 6.6E+ 05 1.1E+ 06 4.5E+ 05 8.9E+ 05

0.03 0.59 0.19 0.02 0.06 0.4 0.3 0.1 0.12 0.01 0.17 0.33 0.17 0.54 0.17 0.13 0.12 0.14 0.03 0.11 0.08 0.09 0.18 0.23 0.08

0.2 0.4

0.4 0.1 0.3 0.5 0.6 0.5 0.1 0.6 0.3 0.21 k ¼ 2.52a νeff ¼ 7 p ¼ 95.45%

0.16 k ¼ 2.52a νeff ¼ 7 p ¼ 95.45%

k is a coverage factor based on t-distribution for νeff ¼ 7 degrees of freedom and p is a level of confidence of approximately 95%.

increase in feedwater and condensate flow rate due to increased output may affect feedwater and condensate pump capacity, feedwater heater performance, feedwater and condensate piping size, etc. [17]. Overcoming the above-mentioned challenges, many reactors in the United States have carried out power uprates. Implementations of uprates in the United States since 2010 are shown in Table 5.3.4, which was extracted and edited from [3]. Since 2010, uprates in the United States are as follows: 27 MU-type uprates and 14 E-type uprates with a total uprate output of 1364.6 MWt and an uprate rate of 1.02%–1.07% by MU; the latter with a total uprate output of 196.5 MWt and an uprate rate of 11.9%–17.0%. The S-type uprate was last uprated at Millstone 3 in December 2008, with an uprate

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Table 5.3.4 Uprate implementation in the United States since 2010. Plant

Uprate (%)

Uprate MWt

Date approved Mo./Day/Yr.

Uprate type

Prairie Island 1 Prairie Island 2 LaSalle 1 LaSalle 2 Surry 1 Surry 2 Limerick 1 Limerick 2 Harris 1 McGuire 1 McGuire 2 Braidwood 1 Braidwood 2 Byron 1 Byron 2 Fermi 2 Catawba 1 Columbia Gen. St. Peach Bottom 2 Peach Bottom 3 Hope Creek 1 Farley 1 Farley 2 Watts Bar 2 Oconee 1 Oconee 2 Oconee 3 Point Beach 1 Point Beach 2 Nine Mile Point 2 Turkey Point 3 Turkey Point 4 St. Lucie 1 Grand Gulf 1 St. Lucie 2 Monticello Peach Bottom 2 Peach Bottom 3 Browns Ferry 1 Browns Ferry 2 Browns Ferry 3

1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.7 1.7 1.6 1.6 1.6 1.6 1.6 1.7 1.02 1.66 1.66 1.6 1.7 1.7 1.4 1.64 1.64 1.64 17 17 15 15 15 11.9 13.1 11.9 12.9 12.4 12.4 14.3 14.3 14.3

27 27 57 57 41 41 57 57 48 58 58 58.4 58.4 58.4 58.4 56 58 58 65 65 62 46 46 48 42 42 42 260 260 521 344 344 320 510 320 229 437 437 494 494 494

08/18/10 08/18/10 09/16/10 09/16/10 09/24/10 09/24/10 04/08/11 04/08/11 05/30/12 05/16/13 05/16/13 02/07/14 02/07/14 02/07/14 02/07/14 02/10/14 04/29/16 05/11/17 11/15/17 11/15/17 04/24/18 10/09/20 10/09/20 10/21/20 01/26/21 01/26/21 01/26/21 05/03/11 05/03/11 12/22/11 06/15/12 06/15/12 07/09/12 07/18/12 09/24/12 09/13/12 08/25/14 08/25/14 08/14/17 08/14/17 08/14/17

MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU MU E E E E E E E E E E E E E E

Extracted and edited from U.S. NRC, n.d. Types of Power Uprates, https://www.nrc.gov/reactors/operating/licensing/ power-uprates/type-power.html (Accessed on 15 September 2021).

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Boiling Water Reactors

output of 239 MWt and an uprate rate of 7%. The MU-type only needs to improve the accuracy of the feedwater flowmeter and does not require any changes to the equipment. The power uprate by extended equipment (E-type) is larger than that of the normal stretch type. For uprates involving equipment changes, the higher output uprates of the E-type tend to be chosen over the S-type. The E-type uprate is approved up to 20% increase in nuclear thermal power, in which high-pressure turbine, condenser pump, main plant equipment, and transformer must be modified. In the BWR power uprate by MUR, the impact on the fuel design will not be significant and there are not major changes to the core design, which will require slight changes in the number of fuel bundles increase and in fuel core loading patterns. On the other hand, contrast to the situation in Japan, Europe and the United States have gained the power generation capacity of several large-scale nuclear power plants without new plant constructions, reducing emission of Carbon dioxide.

References [1] “Regulations Concerning the Installation, Operation, etc. of Nuclear Reactors for Commercial Power Generation” as amended by the Nuclear Regulation Authority of Japan Regulations No. 3 of 2020 (in Japanese). [2] METI, Agency of Natural Resources and Energy, Nuclear and Industrial Safety Subcommittee, Nuclear Reactor Safety Committee, subcommittee140319, resources 4–1.pdf, “Safety of Constant Rated Thermal Power Operation”, December 7, 2001 (in Japanese). [3] U.S. NRC, Types of Power Uprates, https://www.nrc.gov/reactors/operating/licensing/ power-uprates/type-power.html (Accessed on 15 September 2021). [4] Tokyo Electric Power Company, Application for permission to install the reactor, the unit 3 at Fukushima Dai-ni (second) nuclear power station (the complete set), 1998 (in Japanese). [5] M. Mori, et.al., Industrial applications of new type flow-metering system by ultrasonicDoppler profile-velocimetry, in: Fourth International Symposium on Ultrasonic Doppler Methods for Fluid Mechanics and Fluid Engineering, Hokkaido University, Sapporo, Japan, September 6–8, 2004. [6] Hokkaido Electric Power Company, Application for permission to install the reactors at Tomari nuclear power station (the units 1 and 2), 1986 (in Japanese). [7] Niigata Prefecture Disaster Prevention Bureau, Nuclear Safety Division, Status at Kashiwazaki-Kariwa Nuclear Power Station for February 2011, 2011 (in Japanese). [8] K. Okamoto, et al., Technical study on nuclear reactor safety assessment for power uprates, Final Report of the Special Technical Committee on Technical Study on Nuclear Reactor Safety Assessment for Power Uprates, J. Atom. Energy Soc. Jpn. 50 (12) (2008) 772–784 (in Japanese). [9] N. Furuichi, A survey of a power uprate in a commercial nuclear power plant and a water flow facility for calibration standard, AIST Bull. Metrol. 3 (4) (2005) (in Japanese). [10] M. Mori, Fluid dynamics contribution to nuclear energy, Nagare 35 (2016) 21–26 (in Japanese). [11] NRC Regulatory Issue Summary 2007–24, September 27, 2007. [12] Endres+Hauser Flowtec AG, Flow Handbook, Reinach/BL, Switzerland, 2004. p. 125. [13] M. Mori, Y. Takeda, On the traceability of accuracy of ultrasonic flowmeter, JSME J. Power Energy Syst. 5 (3) (2011) 401–405.

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[14] M. Mori, et al., Development of a novel flow measuring system using ultrasonic velocity profile measurement, Exp. Fluids 32 (2002) 153–160. [15] M. Mori, et al., Series of calibration tests at national standard loops and industrial application experiences of new type flow-metering system by ultrasonic pulse-Doppler profile-velocimetry, in: Sixth Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS 6), Okinawa, Japan, November 24–27, 2008. N6P1147. [16] H. Yao, Hybrid ultrasonic flowmeter, Fuji New Report 77 (6) (2004) 432–433. [17] Atomic Energy Society of Japan, Special Technical Committee on Technical Study and Evaluation of Reactor Power Uprate, Report on the evaluation of the technical study on the safety of reactor power uprate, October 2007 (in Japanese).

5.4

Post-BT standard for BWR power plant Shinichi Morooka Former Waseda University, Shinjuku, Tokyo, Japan

5.4.1

Introduction

The onset of boiling transition (BT) under the steady-state operation and the anticipated operational occurrences (AOOs) are not allowed in the current design of BWR core to avoid the integrity of nuclear fuel.a BT leads to decrease in heat transfer performance from the fuel rod to the coolant and increase in the excessive fuel cladding temperature. However, the following new important results were obtained in a joint study with Japanese BWR electrical companies and manufacturers [1–3]. “Even if BT occurs at AOOs, rewet occurs due to decrease in the reactor power due to scram and increase in reactor void fraction, and the fuel temperature returns to that of nucleate boiling as shown in Fig. 5.4.1. In short, because the dryout time and peak cladding temperature are small, the short post-BT operation during AOOs does not cause serious damage to the fuel.” In addition, Nuclear Power Engineering Corporation (NUPEC)b has done many post-BT tests using full-scale simulated BWR fuel assemblies [4–8]. It was proved from the experimental results that even if BT occurred during AOOs, the rod temperature rise after BT was suppressed by rewetting. The following advantages can be expected when the short post-BT operation during AOOs (post-BT criteria) is permitted [9–12]:

a

The current thermal–hydraulic design is introduced in Section 3.2.3.

b

Proving test of NUPEC is introduced in Section 3.2.6.

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Boiling Water Reactors

Fig. 5.4.1 Rapid flow coast-down test result (ICPR (Initial Critical Power Ratio) ¼ 1.20, 4  4 rod bundle) [3].

BWR innovations

– –

475

Operating limit of minimum critical power ratio (OLMCPR), the margin of BT occurrence, can decrease. The core power at rated operation can increase and the core design can be optimized. Equipment for avoiding BT during AOOs can be reduced. For example, in advanced BWR (ABWR), the core coolant flow rate during pump trip event may decrease rapidly due to the small inertial force of reactor internal pump (RIP). This may lead to the onset of BT. Therefore, a motor generator (MG) set has been installed to avoid BT. By adopting the post-BT criteria, MG set can be eliminated.

These advantages increase the economic efficiency and the safety margin. In view of the above findings, the Standards Committee of Atomic Energy Society of Japan (AESJ) has published “Standard for assessment of fuel integrity under anticipated operational occurrences in BWR power plants” (post-BT standard) in June 2003 [13]. This standard was reviewed by the Nuclear Safety Commission of Japan, and the assessment report titled “Post-Boiling Transition Fuel Integrity” was published in the Guidelines of Nuclear Safety Commission of Japan [14,15].

5.4.2

Standard for the assessment of fuel integrity under anticipated operational occurrences

This standard [13] provides the evaluation method of fuel integrity during post-BT operation under AOOs. In order to judge the fuel integrity and the criteria of fuel integrity, a reliable method for predicting the change of rod temperature under post-BT operation was required. These two items have been developed in the joint study with Japanese BWR electrical companies and manufacturers.

5.4.2.1 The method for predicting the change of rod temperature during post-BT operation In order to predict the change of rod temperature under post-BT operation, BT onset time, heat transfer coefficient between coolant and the fuel cladding after BT (post-BT heat transfer coefficient), and rewetting time are needed. The method is described below. (a) Prediction of BT onset time [16]: GEXL [GE Critical Quality (xc) Boiling Length (LB)] correlation,c which is the current design correlation for critical power prediction under steady state and AOOs, has good prediction ability for the BT onset time. Therefore, GEXL was recommended to predict BT onset time. (b) Post-BT heat transfer coefficient [13,17,18]: From the viewpoint of the conservative prediction of post-BT heat transfer coefficient, the modified Dougal-Rohsenow and the Groeneveld 5.9 were recommended.

c

GEXL is explained in Section 3.2.3.

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Boiling Water Reactors

(c) Rewetting time: Many rewetting correlations for predicting the reflooding phenomena during loss of coolant accident (LOCA) were reported so far. However, no reliable correlation was available for rewetting phenomena encountered under the operating conditions during AOOs. Therefore, the joint study with Japanese BWR electrical companies and manufacturers has been challenged and developed the flowing two rewetting correlations. – Correlation 1 [1,10,13]: Correlation 1 is the empirical rewetting correlation developed based on the rewetting data for the annulus flow channel with a single heater rod and steady state. The resulting correlation is xrewet ¼ mðPÞAðG∗ Þq}EQ + BðG∗ Þ ð 1 LB +λ q}ðzÞdz q}EQ ¼ LB λ
 G∗ ¼ G=Go, G0 ¼ 1389 kg= m2 s where xrewet is rewetting quality, G is the mass flux (kg/(m2 s), LB is the boiling length 00 00 (m), P is the pressure (MPa), qEQ is the equivalent uniform axial heat flux (W/m2), q (z) 2 is the heat flux from cladding surface (W/m ), and λ is the axial location of boiling starting point (m). mðPÞ ¼

–

– –

104 , 6:9 < P < 7:8 ðMPaÞ f1  0:204ðP  6:86Þg

where A (G*) and B (G*) are as shown in Refs. [10,13]. Correlation 2 [9,10,13,19]: Correlation 2, which was developed by Kudo et al., is recommended in this standard. Correlation 2 has been developed based on the following experimental results: Critical quality for BT onset and the rewetting quality under quasi-steady BWR transients are almost identical. There is the difference between the rewetting quality and the critical quality. The difference is expressed by

xrewet ¼ xc + Δx ( )     Dw F hfg Gðx  xc Þ G0 ΔT w  ΔT 0 dx  Δt0 Δx ¼ 0:0635 dt q} G ΔT 0 u0lf t¼tm where Dw is the thermal equivalent diameter (m), F is the local peaking factor, G is the mass flux (kg/(m2s)), G0 is equal to 1356 (kg/(m2s)), hfg is the latent heat of vaporization (J/kg), tm is the time to reach the maximum (x  xc) (s), u0lf is the standard value of propagation velocity of liquid film ¼ 1 (m/s), x is the cross-sectional averaged flow quality, xc is the critical quality, Δx is the quality deviation from a time that quality reduces to the critical quality to the time of transient rewetting, ΔTw is wall superheat (K), ΔT0 represents superheat of three-layer interface among the solid, water, and vapor at the front of liquid film ¼30 (K), and dx=dt is the cross-sectional quality time derivative.

BWR innovations

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(d) Qualification for the method of predicting change of rod temperature during post-BT operation The method for predicting change of rod temperature during post-BT operation compared with many post-BT test data during the simulated AOOs to qualify this method. Simulated transient conditions are the load rejection without turbine bypass event and rapid flow coast-down event (for example, all pump trip event of advanced BWR). Transient tests have been done in the joint study with Japanese BWR electrical companies and manufacturers, NUPEC and Japan Atomic Energy Agency (JAEA) using 4  4, high burnup 8  8, 9  9A, and 9  9B simulated fuel assemblies [13]. A comparison between measurement of peak cladding temperature (PCT) and dryout duration is given in Refs. [9,13]. It was found from these comparisons that the prediction method could be obtained the conservative prediction for dryout duration and PCT.

5.4.2.2 The criteria of fuel integrity after BT [9,10,13,20] The following tests have been conducted to develop the criteria for fuel integrity after the short-BT operation: – – –

Test under short-BT duration using the irradiated fuel at the Halden Reactor Project of BWR loop. Tests for fuel cladding deformation under short-BT duration using the nonirradiated fuel rod with the zirconium liner of BWR. Fuel cladding annealing tests after short-BT duration.

Based on the results of these tests, the criteria of fuel integrity after short-BT duration were investigated and judged. Criteria of fuel integrity and reuse of a fuel bundle after post-BT operation are given in Refs. [9,13]. In the post-BT standard, the criteria of fuel integrity of a fuel bundle after post-BT operation in terms of both peak cladding temperature (PCT) and dryout duration are as follows: l

l

PCT should not exceed 800°C (1073 K), Dryout duration should be within 100 s.

Post-BT study has been conducted by Government and private companies for many years. As a result, the standard of AESJ could be established and this is Japan’s original standard. The authors sincerely hope that this standard can contribute to improvements in safety, reliability, and efficiency of nuclear power plants.

References [1] S. Muto, T. Anegawa, S. Morooka, S. Yokobori, Y. Takigawa, S. Ebata, Y. Yoshimoto, S. Suzuki, An experimental study on rewetting phenomena in transient conditions of BWRs, Nucl. Eng. Des. 120 (1990) 311–321. [2] T. Enomoto, et al., Study on transient behavior of rod surface temperature after boiling transition, in: 2nd International Topical Meeting on Nuclear Plant Thermal Hydraulics and Operations, Tokyo, 1986, 4–51.

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[3] N. Abe, S. Miura, S. Ebata, S. Morooka, T. Anegawa, Post-BT analysis by best-estimate thermal–hydraulic code TRACG, in: ICONE-7, Tokyo, Japan, ICONE-7157, April 1999. [4] T. Mitsutake, M. Kitamura, K. Kamimura, N. Abe, S. Morooka, J. Kimura, Y. Masuhara, Proving test for critical power performance of BWR 9  9 Type-A fuel assembly under transient conditions and evaluation of transient analysis code, in: ICONE-7. Tokyo Japan, ICONE-7156, 1999. April 19–23. [5] A. Inoue, H. Hayashi, M. Kitamura, T. Mitsutake, S. Morooka, J. Kimura, A. Hoshide, N. Saitoh, N. Abe, K. Arai, S. Ebata, S. Komura, S. Nakamura, Proving test on thermal–hydraulic performance of high burnup 8  8 fuel assembly for BWR, J. AESJ 40 (10) (1998) 784–797 (in Japanese). [6] Y. Tsukuda, H. Hayashi, K. Kamimura, T. Hattori, H. Kaneko, S. Morooka, T. Mitsutake, M. Akiba, N, Abe, M. Warashina, Y. Masuhara, J. Kimura, A. Tanabe, Y. Nishino, K. Isaka and R. Suzuki, Proving test on thermal–hydraulic performance of BWR fuel assemblies, Trans. AESJ, 1(4) (2002) 384–403 (in Japanese). [7] M. Akiyama, A. Inoue, M. Ohishi, S. Morooka, A. Hoshide, T. Ishizuka, K. Yoshimura, Study on post-BT heat transfer in a full scale BWR (8  8) rod bundle, Nucl. Eng. Des. 117 (1989) 341–347. [8] Y. Tsukuda, H. Hayashi, K. Isaka, H. Kaneko, T. Mitsutake, M. Akiba, N. Abe, S. Morooka, Y. Yasuhara, R. Suzuki, Y. Yuji, BWR 9  9 fuel assembly thermal–hydraulics tests(1) New formula for post boiling transition heat transfer correlation and rewet correlation, in: ICONE-10. Arlington, VA, ICONE10–22556, April 14–18, 2002. [9] K. Mishima, S. Mizokami, Y. Kudo, S. Komura, S. Morooka, K. Nishida, The past, present, the future for fuel integrity study under Post Boiling Transition of BWR Power Plants, J. AESJ 46 (5) (2004) 332–338 (in Japanese). [10] T. Hara, S. Mizokami, Y. Kudo, S. Komura, Y. Nagata, S. Morooka, Current status of the post boiling transition research in Japan, J. Nucl. Sci. Technol. 40 (10) (2003) 852–861. [11] S. Morooka, Y. Kudo, A. Hotta, Heat removable limit for nuclear fuel, Trans. JSME 75 (758) (2009) 1890–1895 (in Japanese). [12] S. Morooka, Y. Kudo, A. Hotta, Heat removable limit for nuclear fuel, Heat TransferAsian Res. 39 (7) (2010) 482–491. [13] AESJ, Standard for assessment of fuel integrity under anticipated operational occurrences in BWR power plants, in: AESJ-SC-P002-2003 (2003-6) (in Japanese), 2003. [14] Nuclear Safety Commission of Japan, Guideline for Nuclear Safety Commission (rev. 12), Taisei Publishing Co. Ltd., 1999 (in Japanese). [15] Nuclear Regulatory Authority of Japan. https://www.nsr.go.jp/law_kijyun/sonota/ anzenshinsa.html (Accessed on 13 June 2022) (in Japanese). [16] GE, General Electric BWR Thermal Analysis Basis (GETAB) Data; Correlation and design application, GE Report, NEDO-10958, 1973. [17] T. Iguchi, C. Iwaki, Y. Anoda, Experimental results of BWR post-CHF tests—Critical heat flux and post-CHF heat transfer coefficients, in: JAERI-Research 2001-060, 2002 (in Japanese). [18] D.C. Groeneveld, Post-dryout heat transfer at reactor operating conditions, in: Proceeding of Topical Meeting Water Reactor Safety, CONF-730304. U. S. Atomic Energy Commission, 1973. [19] Y. Kudo, T. Hara, Development of a phenomenological rewetting correlation in a fuel assembly under BWR transient conditions, Trans. AESJ 2 (2) (2003) 121–129 (in Japanese). [20] S. Komura, T. Hara, BWR fuel performance in case of short period boiling transition, Trans. AESJ 2 (2) (2003) 187–195 (in Japanese).

BWR innovations

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Core catcher Chikako Iwaki Toshiba Energy Systems & Solutions, Corp., Yokohama, Kanagawa, Japan

Nomenclature D d g H hfg K L S Q u Vv ΔP, ΔPl ΔPh ρd ρv ρr

hydraulic diameter (m) fin diameter (m) gravitational acceleration (m/s2) height of channel (m) latent heat of vaporization (J/kg) coefficient of shape loss () flow pass length (m) fin pitch (m) heat flux (W/m2) velocity (m/s) volumetric flow rate of vapor (m/s) pressure loss (Pa) hydro-static head (Pa) mean density in downcomer (kg/m3) vapor density (kg/m3) mean density of rising flow (kg/m3)

Greek symbols λ ϕ

2

coefficient of friction loss () two phase flow multiplier ()

5.5.1

Overview of core melt stabilization and cooling

In the event of severe accident (SA) where the molten corium and reactor pressure vessel (RPV) fails, there is a possibility of the release of molten corium into the reactor cavity. Stabilization and cooling of molten corium is an important issue to prevent containment failure and release of fission products caused by penetration of the basemat, overpressure, or serious damage to the internal structure. As accident management of the existing BWR plant, measures are taken to prevent molten core-concrete interaction (MCCI) by injecting water into the debris and filling the pedestal with water before the pressure vessel is damaged. By prefilling with water, the molten core becomes particles in water when it falls, and the molten core

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Boiling Water Reactors

expands as a particulate bed, which is expected to improve debris cooling performance. However, since there is uncertainty in the debris cooling performance only by cooling the upper surface of the debris, various structures for promoting the cooling of the debris have been proposed. As one of them, a system has been developed to prevent concrete erosion and cool debris passively when the debris falls by laying a refractory material with a high melting point on the bottom of the pedestal and the cylindrical part, which are supposed to come into contact with debris as shown in Fig. 5.5.1. The other one is core catcher presently adopted in different designs. Each core catcher design involves passive safety features such as gravity flooding and natural convection cooling to increase reliability, which depend on the specific reactor designs. Currently, several concepts of core catcher have been developed and applied to the actual plants. One of the representative PWR designs is a crucible-type catcher that has been applied practically to the VVER-1000 reactors built in China (Tyanwan) and India (Kudankulam) (Fig. 5.5.2) [1]. The VVER-1000 core catcher design mainly consists of lower plate, an intake covering RPV lower head, and crucible-type core catcher installed under the RPV and partially filled with sacrificial material (SM). The SM is arranged in the form of a honeycomb of large cells, which promotes rapid corium spreading and increases the interaction between the SM and the corium. In the event of SA after RPV melt-through, the molten corium relocates to the core catcher and the heat flux and temperature of the molten corium are reduced by mixing with SM. The crucible is cooled from outside by flooding the reactor cavity by water, water also enters into the surface of the melt. The heat from the top surface of the melt is temporarily removed by radiation and then covering the surface of the melt by water. The

Fig. 5.5.1 Debris cooling system with refractory material.

Reactor Pressure Vessel Molten core

Refractory material

Particulate bed

Concrete

BWR innovations

481

Fig. 5.5.2 Principle layout of core catcher of Tyanwan (a) and Kudankulam NPPs (b) [1] (1: RPV, 2: reactor cavity, 3: lower plate, 4: cantilever girder, 5: core catcher vessel, 6: SM).

capacity of core catcher holds the expected total mass of the molten corium and is sufficient for heat transfer from the molten pool through the walls of the vessel to the cooling water. The other one is the core catcher with melt spreading developed for the European Pressurized Reactor (EPR) [2,3]. In the concept of the European EPR reactor, the melt relocating from the RPV spreads as a layer over a large horizontal area and is cooled by water flooding. The core catcher design involves an external molten corium retention system to temporarily hold molten corium released from the RPV, a melt plug in the lower part of the retention system, melt discharge channel, and spreading compartment for the stabilization of the spread melt (Fig. 5.5.3). In the event of SA after RPV melt-through, the molten corium is first collected in the external molten corium retention system while mixing with the sacrificial layer of the inner cavity wall to pretreat the composition and temperature of the corium. After penetration of the melt plug, the

Fig. 5.5.3 Plane (a) and elevation (b) views of the EPR core catcher design [2].

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Boiling Water Reactors

corium flows onto a large-surface spreading area through an inclined melt discharge channel. The corium is cooled by water flowing in cooling channels under the spreading area and top flooding. A horizontal cooling structure consisting of an array of steel blocks that form parallel rectangular channels is provided. The system reliability depends on melt diffusion behavior of low viscosity melt into the spreading compartment, therefore analysis code with model of the diffusion phenomenon has been developed and used to properly to evaluate the EPR design. The core catcher, called the basemat-internal melt arrest and coolability (BiMAC) device has been designed for Economic Simplified Boiling Water Reactor (ESBWR) by GEH [4]. It consists of a series of inclined pipes arranged side by side, forming a jacket that can be effectively and passively cooled by the natural circulation of water. A supporting steel plate covers the BiMAC to have a drywell floor which is large enough to capture the molten core, approximately 90 m3. The available full volume below the BiMAC cover is approximately 400% of the full molten corium volume, this prevents any corium from coming in direct contact with the lower drywell liner. The effectiveness of BiMAC concept has been confirmed by fundamental analytical considerations of separate-effects experiments on critical heat flux and twophase pressure drop in inclined pipes. Toshiba developed the core catcher of the European Advanced Boiling Water Reactor (EU-ABWR) and developed a device that can be installed in the existing BWR after Fukushima Daiichi accident. In the basic concept, inclined cooling channels is placed at the lower drywell floor and the molten corium is gathered in a round basin covered with refractory layers. After the molten corium falls into the core catcher, it is cooled by both the inclined cooling channels and the flooded water over the molten corium. The following section provides a more detailed explanation of core catcher of European Advanced Boiling Water Reactor.

5.5.2

Core catcher of EU-ABWR

5.5.2.1 Concept of core catcher A core catcher developed by Toshiba for EU-ABWR, which is shown in Fig. 5.5.4, has the passive containment cooling system (PCCS) and refractory layers to protect the containment integrity and avoid early containment venting in the event of SA. The core catcher is located on the bottom floor of the lower dry well. The core catcher, in combination with PCCS, cools the top and bottom surfaces of the molten corium from RPV to stabilize it. A large amount of steam produced by cooling the core melt is released into the dry well, which condenses in the wet well and the condensate is returned to the core catcher. The configuration of the core catcher is shown schematically in Fig. 5.5.5. It consists of a round basin, axisymmetrically arranged inclined cooling channels, an annulus riser, an annulus downcomer, and a water chamber. The axisymmetric structure of the inclined cooling channel is intended to provide uniform accumulation of molten

BWR innovations

483

Dry well Wet well NC gases Refractory layers

Fig. 5.5.4 Passive safety system with PCCS (Passive Containment Cooling System) and core catcher [5].

About 4m 2.5 m

Condensate from PCCS Vent Pipe Fusible Valve

Pedestal Round Basin Wall

Refractory Layer Molten Pool Solidified Debris

Annulus Riser Annulus Downcomer

Cooling Channel Basemat Concrete Water Chamber

Fig. 5.5.5 Schematic representation of Core catcher [5].

corium. Effective cooling can be achieved by increasing the surface-area-to-volume ratio of molten corium. The refractory layer is installed on the upper surface of the steel basin to prevent damage to the basin due to overheating, jet collisions, and heat intrusion. The condensate line in Fig. 5.5.4 and the condensate from PCCS in Fig. 5.5.5 are the same line.

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Boiling Water Reactors

Passive flappers are connected to the vent pipes that connect the dry well to the wet well. Cooling water is initially supplied from the suppression pool via the passive flooder, but in the long term it is supplied by PCCS as a condensate drain. There is a fusible valve at the end of each passive flooder, which causes the temperature of the lower drywell space to rise due to an RPV failure and opens when the melting temperature of the fusible material is reached. The melting temperature of the fusible material is high enough to avoid unnecessary open operation in the design-based LOCA (above 450 K) and low enough to avoid overheating of the containment (less than 540 K). After the molten corium is released into the lower dry well, the space of the lower dry well heats up rapidly, opening the flexible valve of the passive flooder in minutes and allowing the water in the suppression pool to flow into the annulus downcomer of the core catcher. After the core catcher is flooded, the molten dermis is passively cooled from the top and bottom by the flooded water and the bottom inclined cooling channel, respectively. The cooling water in the inclined channel boils due to the heat from the molten corium, and the resulting steam flows through the riser into the flooded water pool. At this time, the fluid in the inclined channel and in the riser is two phase of steam and water, while the fluid in the downcomer is water. Accordingly, as shown in Fig. 5.5.5, natural circulation is established by density difference between single-phase fluid in the downcomer and two-phase fluid in the inclined channels and riser. This natural circulation can continuously supply cooling water to the inclined channels, which contributes to the prevention of the failure of the core catcher due to overheating.

5.5.2.2 Performance evaluation test (a) Test facility: The thermal–hydraulic behavior and coolability of the core catcher was investigated using a test facility that simulated the length and height of the actual cooling channel, which affects natural circulation. Fig. 5.5.6 shows the diagram of natural circulation test facility. The cooling channel as a test section is a fan-shaped 1/64 sector model with the actual length of the core catcher. The inclination angle of the cooling channel is fixed at 10° and the height is almost 2.5 m. An electrical heater is installed at the upper side of the cooling channel to simulate heat flux from the molten corium. The power of the heater is parametrically controlled to investigate natural circulation behavior dependence on various parameters. The test facility has many instrumentations such as temperature, pressure, and flow rate to evaluate thermal–hydraulic behavior. In the experiment, in order to match the boundary conditions with the actual conditions, a predetermined heat flux is supplied to the upper part of the cooling channel by the electric heater, and the subcooling of the cooling channel inlet is adjusted. As shown in Table 5.5.1, the parameter range of test boundary conditions was set to cover the reference value, that is, the actual conditions in the event of SA. The heat flux was derived from thermal hydraulics analysis of the melt pool [6]. The representative pressure condition of the actual plant is 0.5 MPa while tests were conducted at atmospheric condition. Therefore, the heat flux was significantly adjusted to match the dimensionless numbers that characterize the natural circulation behavior with the actual plant values. The dimensionless numbers that should be considered are the subcooling number (Sc), the Froude number (Fr), and the Zuber number (Zu).

BWR innovations

485

Fig. 5.5.6 Schematic representation of natural circulation test facility [5].

Table 5.5.1 Test boundary conditions [5]. Parameters

Position

Test

Reference

Pressure (MPa) Heat flux (kW/m2)

Outlet tank Cooling channel Riser Inlet tank

0.1 55–120 0–315 0–3

0.5 105 105 1

Subcooling (K)

(b) Evaluation model of natural circulation: Thermal–hydraulic conditions in the cooling channel can be estimated by fundamental equations of the pressure balance and energy balance considering closed loop as shown in Fig. 5.5.7. Here, subscript i expresses local position of heated area and subscript j expresses local position of coolant flow loop. First, the driving force is the difference in hydraulic head pressure between upward flow region (in the cooling channel and the riser) and flow in the downcomer. Hence, the pressure balance in the circulation flow path can be expressed by the following equation: ΔPh ¼ ΔPl

(5.5.1)

where ΔPh is hydrostatic head and Δ Pl is pressure loss. Δ Ph is expressed by the following equation: ΔPh ¼ H  g  ðρd  ρr Þ

(5.5.2)

where H is height of channel, g is gravitational acceleration, ρd is mean density in downcomer, and ρr is the mean density of rising flow. The pressure loss, ΔPl is expressed by the following equation:

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Boiling Water Reactors

Fig. 5.5.7 Evaluation model of natural circulation [5]. ΔPl ¼

m X ρ

l

i¼1

2

 u2l,i

Li λ  ϕ2 + K i Di i i

 (5.5.3)

where ρl is liquid density, ul is flow velocity, L is length of flow pass, D is hydraulic diameter, λ is friction loss coefficient, ϕ2 is two-phase flow multiplier, and K is shape loss coefficient. The coolant closed loop of core catcher consists of a cooling channel, riser, down pipe, water supply line, and water chamber, so all of these pressure drops are added together. Next, the thermal energy given to the cooling channel of the core catcher is transferred to the coolant, resulting in the generation of steam. In the saturated state, the energy consumed for vaporization can be used to calculate the vapor flow rate at any position in the cooling channel. As the steam-water coolant flows through the cooling channel, it receives thermal energy and increases the amount of steam until it is discharged from the riser. Therefore, the steam flow rate is obtained by integrating the generated steam flow rate from the inlet of the channel to an arbitrary position as expressed by the following equation: V v,j ¼

Xj k¼1


 Qv,k = hfg  ρv

(5.5.4)

where Vv is vapor volumetric flow rate, Qv, is heat flux, hfg is latent heat of vaporization, and ρv is vapor density. Thermal–hydraulic conditions at any positions in the cooling channel, natural circulation flow rate, and pressure loss can be evaluated by solving Eqs. (5.5.1)–(5.5.4). (c) Test results: It has been confirmed that boiling starts by input heat simulating molten corium from the upper surface of the cooling channel, and a natural circulating flow of the coolant is generated in the circulation flow path. The test results of the natural circulation flow rate are shown in Fig. 5.5.8. As the amount of heating increases, the driving force due to the difference in density increases, so the natural circulation flow rate as well the cooling capacity increase. From this, it was confirmed that the heat transfer surface was sufficiently cooled and the temperature did not rise at any position under the actual plant conditions. The figure also shows the natural circulation flow rate calculated using Eqs. (5.5.1)–(5.5.4). Although the driving force increases as the amount of heating increases, the pressure loss increases due to the increase in the void ratio in the flow path. Due to these two effects, increase of the natural circulation flow rate decreases as the amount of heating increases. Including these tendencies, it can be seen that the analysis is in good agreement with the test results. Natural circulation flow rate prediction is important for the evaluation of cooling capacity, and it has been shown that such a simple one-dimensional evaluation can be used to evaluate it with sufficient accuracy.

487

Natural Circulation Flow Rate (kg/s)

BWR innovations 孵孳

孴學

孴孳

Test results

學

Analysis results 孳 孳季 季 季 季

孵孳孳季 季 季 季 季 季 季 季 季 季 季 季 季 季 季 季 季 孷孳孳

Total Amount of Supply Heat (kW)

Fig. 5.5.8 Test results [5].

5.5.3

Core catcher for the existing BWR

5.5.3.1 Concept of core catcher After the Fukushima Daiichi accident, it was considered whether core catcher based on the technology developed for EU-ABWR can be applied to the existing BWR plants. The existing plants did not have the space to install inclined cooling channels, therefore a flat and high-thermal-conductivity core catcher with a nontilted cooling channel was developed to reduce the overall height of the structure. The composition of the flat core catcher is schematically shown in Fig. 5.5.9a. It has horizontal basin plates and flat cooling channels with a lot of fins. The cooling channel of the flat core catcher has a modular assembly structure as shown in Fig. 5.5.9b so that it can be easily transported and installed in the lean space under the RPV of the existing plant. Since the temperature of the plate tends to rise due to the retention of vapor near the top of the plate compared to the case where there is an inclination, copper fins of good heat conduction are set on the top of plates by constantly keeping them in contact with the cooling water to improve

Fig. 5.5.9 Flat and high-thermal-conductivity core catcher [8]. (a) Schematic representation of flat core catcher, (b) module image of core catcher.

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Boiling Water Reactors

cooling performance. Since the tip of the fins is separated from the bottom surface of the cooling flow path in order to prevent heat conduction to the water supply flow path, there is a space under the cooling flow path without fins. There is little knowledge about the two-phase flow in the complicated rectangular horizontal path, therefore natural circulation characteristics and heat transfer behavior were evaluated by a mock test.

5.5.3.2 Performance evaluation test (a) Test section and test condition: A test section that simulates part of a flat core catcher was set up in a natural circulation test facility similar to the test facility shown in Fig. 5.5.6. The length of the test section is same as the inner diameter of RPV pedestal of the existing plants, i.e., 3.5 m. The height of the cross-sectional area of cooling channel is 47 mm and the width is 110 mm. Copper fins with a diameter of 30 mm are set on a steel top plate, and a total of 342 fins are placed on 3  114 in-line arrangement. The inclination of the cooling channel is fixed at 0°. The heat flux from molten core is simulated by electrical heating, and controlled parametrically to investigate various natural circulation conditions. The heat flux from the molten corium is simulated by an electric heater installed at the top of the test section and the test is performed with varying heater outputs of up to 250 kW/m2, including the evaluated heat flux of the molten corium. The test was conducted under atmospheric pressure conditions. The test facility has many instruments for understanding thermal–hydraulic behavior: fluid temperature, mass flow rate, pressure drop, and temperature of fins and top plate. The test section has three visualization windows in the flow direction and the internal flow pattern can be observed. (b) Natural circulation characteristics: When heat is supplied to the upper surface of the test section, boiling begins after a while and a circulating flow is formed despite long horizontal channel. The natural circulation occurs due to boiling at an extremely low heat flux (20 kW/m2), and natural circulation also occurs under the maximum heat flux conditions (250 kW/m2). Fig. 5.5.10 shows the circulating mass flux relative to the heat flux with pressure drop in the cooling channel. The mass flux increases with the increase in heat flux, reaches the maximum at 50 kW/m2, and starts to decrease. When the heat flux increases, the driving force due to the density difference by steam generation while two-phase pressure loss increases. The natural circulation flow rate is determined by these two effects. Since the driving force due to the head is smaller in the horizontal channel than in the inclined channel, the peak value of the flow rate can be observed. (c) Flow pattern: The photos taken through a window located 3.3 m from the inlet of the cooling channel (near the outlet) are shown in Fig. 5.5.11. Since steam is generated by the heat from the upper part as it flows in the cooling channel, the void fraction is highest near the outlet. Under minimum heat flux conditions of 20 kW/m2, the vapor flows near the top plate and the water flows in the lower part of the channel. The vapor–liquid phase is completely separated and it is classified as a stratified flow. Under conditions of heat flux of 50 kW/m2, the vapor–liquid interface is disturbed due to increase of void fraction and flow rate, but the vapor and liquid flows separately. Therefore, flow pattern can be classified as a stratified wavy flow. Under maximum heat flux conditions of 250 kW/m2, intermittent froth flow was observed in which the visualization window was intermittently covered with a liquid film due to further increase in turbulence by large vapor velocity.

BWR innovations

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Fig. 5.5.10 Test results of mass flux and pressure drop [8].

䣖䣱䣲䢢䣲䣮䣣䣶䣧

䣈䣫䣰

(a)

(b)

(c)

Fig. 5.5.11 Flow pattern at heat flux: (a) 20 kW/m2, (b) 50 kW/m2, and (c) 250 kW/m2 [9].

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Boiling Water Reactors

By comparing these flow patterns with the conventional flow pattern map developed with horizontal pipe [10], it is confirmed that the transition conditions of the flow pattern are almost the same even though the structure is included inside the flow channel. (d) Temperature of the top plate: The surface temperatures of the top plate and fins are measured by a lot of thermocouples to evaluate the cooling performance. It has been confirmed that under the condition that the heat flux is 200 kW/m2 or less, all the temperatures are maintained at 110 °C, almost same as the saturation temperature and structure of core catcher is satisfactorily cooled by the boiling heat transfer on the surface of the fins. Under the maximum heat flux condition of 250 kW/m2, all temperatures from the inlet to 2.5 m were 110 °C, but at 3.4 m, the temperatures of the top plate reached a maximum of about 290 °C. At 3 m from the inlet, the temperature fluctuated from 130 °C to 240 °C for each test, so it is suggested that the flow pattern and heat transfer mode change at around this position. In conclusion, even at the maximum heat flux, the maximum temperature in the cooling channel was maintained below 300 °C, which is below the melting point of copper fins (about 1080 °C), so it has been clarified that this core catcher can remove heat from the molten corium. (e) Natural circulation calculation model: Calculation of natural circulation flow rate is indispensable for the evaluation of cooling performance. In this core catcher, the natural circulation flow rate can be obtained by taking the abovementioned basic formula of energy and pressure balance. However, in this core catcher, the pressure loss calculation of two-phase flow in the cooling channel becomes an issue because of complicated configuration of the cooling channel. The cross section of cooling channel is shown schematically in Fig. 5.5.12. The top of the fin is not in contact with the bottom to prevent heat conduction to the lower coolant flow path. The fluid tends to flow easily in the region where there is no fin (Area B), because of small resistance the flow velocity in the cooling channel is not uniform. Here, assuming that the fluid flows in Areas A and B, the following equation holds for each pressure loss. ΔPA ¼ ΔPB

(5.5.5)

The pressure loss of Area A, ΔPA can be estimated by using pressure loss correlation of perpendicular flow around circular tube [11], as expressed by the following equation: ðl  S ρuA 2 N xdx ΔPA ¼ ϕ2 0:105 S  d 2 l 0

(5.5.6)

where l is cooling channel length, S is fin pitch, d is fin diameter, uA is velocity in Area A, and N is number of fin. On the other hand, the pressure loss of Area B, Δ PB is expressed by following equation:

Area A

Area B

Fig. 5.5.12 Cross section of cooling channel [9].

BWR innovations

ΔPB ¼ ϕ2

491

ðl

λx ρuB 2 dx 2 0 D

(5.5.7)

where uB is velocity in Area B. From these equations, the distribution of flow rate of the Areas A and B can be determined, and the pressure loss of the cooling channel can be obtained. Calculate total pressure drops other than cooling channel, including all of flow path components, riser, downcomer, tank, pipes, etc., and calculate the natural circulation flow rate to be equal to the driving force due to density difference. Since the flow velocity in the cooling channel is determined once the natural circulation flow rate is determined, the flow pattern is determined with reference to the abovementioned flow pattern map. By applying the corresponding heat transfer model, the temperature of the top plate can be calculated [9].

References [1] V.B. Khabensky, V.S. Granovsky, S.V. Bechta, V.V. Gusarov, Severe accident management concept of the VVER-1000 and the justification of corium retention in a crucibletype core catcher, Nucl. Eng. Technol. 41 (5) (2009) 561–574. [2] M. Fischer, The severe accident mitigation concept and the design measures for core melt retention of the European Pressurized Reactor (EPR), Nucl. Eng. Des. 230 (2004) 169– 180. [3] F. Bouteille, G. Azarian, D. Bittermann, J. Brauns, J. Eyink, The EPR overall approach for severe accident mitigation, Nucl. Eng. Des. 236 (2006) 1464–1470. [4] NEDO-33201, Revision 6, ESBWR Design Certification Probabilistic Risk Assessment, 2010. [5] T. Kurita, T. Tobimatsu, M. Tahara, K. Aoki, Y. Kojima, Heat removal capability of corecatcher with natural circulation, in: Proceedings of the 21st International Conference on Nuclear Engineering ICONE21, Paper 16,635, 2013. [6] R. Hamazaki, T. Nakagawa, M. Tahara, M. Komuro, T. Tobimatsu, Y. Suzuki, T. Kurita, Evaluation on core melt retention in core catcher of Toshiba’s EU-ABWR, in: Proceedings of Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP) 2011, Paper 11,414, 2011. [7] T. Kurita, T. Tobimatsu, M. Tahara, M. Yamada, Y. Kojima, Flow characterization and heat transfer of core-catcher in passive safety system, in: Proceedings of the 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference ICONE20POWER2012, Paper 55,129, 2012. [8] D. Takeyama, C. Iwaki, Y. Onitsuka, M. Tahara, Study on thermal-hydraulics characteristics of the flat and high-thermal-conductivity core-catcher, in: Proceedings of the 26th International Conference on Nuclear Engineering ICONE26, Paper 82,046, 2018. [9] D. Takeyama, C. Iwaki, Y. Onitsuka, M. Tahara, Development of thermal-hydraulics model for two-phase flow in the horizontal rectangular finned channel, in: Proceedings of the 27th International Conference on Nuclear Engineering ICONE27, Paper 1650, 2019. [10] Y. Taitel, N. Lee, A.E. Dukler, Transient gas–liquid flow in horizontal pipes; modeling the flow pattern transition, AIChE J. 24 (5) (1978) 921–934. [11] E. Nishikawa, S. Ishigai, Structure of gas flow and its pressure loss in tube banks with tube axes normal to flow, Trans. Japanese Soc. Mech. Eng. 43 (373) (1977) 3310–3319.

492

Boiling Water Reactors

5.6

Steam injector Tadashi Narabayashia, Chikako Iwakib, and Michitsugu Moric a

Tokyo Institute of Technology, Meguro, Tokyo, Japan, bToshiba Energy Systems & Solutions, Corp., Yokohama, Kanagawa, Japan, cGraduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan

Nomenclature A Ci D H h k m P Q r T U,V,W v z Γ ε τ Φ

flow area (m2) shear coefficient of interface between gas and liquid phases () diameter (m) heat transfer coefficient () specific enthalpy (kJ/kg) turbulent energy (m2/s2) mass flow rate (kg/s) pressure (MPa) heat flux (m3/s) radial coordinate in cylindrical geometry (m) temperature (°C) velocity (m/s) specific volume (m3/kg) axial coordinate in cylindrical geometry (m) production rate () rate of turbulent energy dissipation () shear stress dissipation energy (kJ/m3/s)

Subscripts D g i l lg N O S sat sub T W

diffuser or discharge port gas phase interface between gas and liquid phases liquid phase phase change nozzle over flow port steam inlet port saturation subcooling throat water inlet port

BWR innovations

5.6.1

493

Introduction

Next-generation reactors of various types [2] are under development around the world, along which innovative passive and simple devices are studied extensively. The most interesting of these devices is the steam injector (SI) [1], which is an active pump but does not require external power such as a motor a turbo-machinery. Thus, the SI is characteristically categorized as a passive pump. Narabayashi et al. [3–5], Iwaki et al. [10,11], and Cattadori et al. [7] investigated SI application to the nextgeneration BWR. The SI is a simple, compact, and passive device for water transportation. It might be applicable to the passive core injection system (PCIS) or the primary recirculation loop (PLR) system [6]. Analytical and experimental studies were conducted on large-scale SIs for the next-generation BWR. A water jet in the large-scale SI might be affected by twodimensional effect of radial heat transport enhanced by turbulence. Therefore, analyses of the characteristics of the SI were conducted for both small- and large-scale injectors using the newly developed separate-two-phase flow model that had the capability to analyze the multidimensional turbulent two-phase flow [1,12,13]. In order to check the feasibility of such large-scale SIs, a separate-two-phase flow model was installed in the PHOENICS Code [14]. Two scale-model tests were conducted for both the SI-PCIS and the SI-PLR to confirm the newly developed two-phase flow model. In the 1/2 scale SI-PCIS tests, the discharge pressure of approximately 8 MPa was attained from 7 MPa steam and 0.4 MPa water sources. In the 1/5 scale SI-PLR tests, the discharge pressure of 12.5 MPa was attained from 3 MPa steam and 7 MPa water sources. Both the test results were in good agreement with the analytical results, which confirmed the feasibility of the newly developed analytical two-phase model. It was also demonstrated that the SI fully functioned to pump up and transport water. The SIs will help to simplify next-generation BWRs and to enhance the reliability of those since the SIs do not need external power such as the large motor of the current RLP pump in the PCV. Another advantage of the SI is enthalpy recovery. The thermal enthalpy of the steam that drives the SI is recovered as heat energy for the temperature rise of the feed water into the core. In the SI-RLP application, the feed water heating effect is obtained because the enthalpy of the steam is condensed and condensed on the surface of the water jet. Air-cooled motor loss heat energy escaped into the atmosphere. However, the energy conversion efficiency is slightly lower than that of motor driven pumps [6,13].

5.6.2

Principle and application of SI

Fig. 5.6.1 shows the concept of the PCIS proposed by Narabayashi et al. [2,8]. It has the SI. The system can start up and inject water into the RPV without AC power, merely by opening steam and water supply valves.

494

Boiling Water Reactors

Steam

Water

SI-PCIS

Start up drain

to S/P

Fig. 5.6.1 Concept of PCIS using steam injector [4].

As soon as steam is supplied to the SI, a supersonic steam jet blows onto the cold water surface in a mixing nozzle as shown in Fig. 5.6.2. As shown in Fig. 5.6.2, there are three principle types of the SI: (a) central water jet type, (b) central steam jet type, and (c) double coaxial type, which attained a higher discharge pressure than supply steam pressure and got a JSME technical award. As Narabayashi et al. measured and analyzed [4], when normal temperature water was injected from the water jet nozzle at the axial center and steam was supplied to the annular steam nozzle arranged outside the water jet nozzle in the mixing nozzle, the steam became supersonic flow in the mixing nozzle and accelerated the water jet,

Steam

Mixing Nozzle

Water

Water Jet

Diffuser

Throat

(a)

Water Mixing Nozzle

Steam

Water Jet

(b)

Diffuser

Throat

(c)

Fig. 5.6.2 Principle of the steam injector [4]. (a) central water jet type, (b) central steam jet type, (c) double coaxial type (JSME award).

BWR innovations

495

producing high-speed water flow at the throat. Then, very high-speed water flow was formed along the central axis in the mixing zone and the throat. Just before the throat, the inflowing steam was totally condensed by the water. The kinetic energy of the high-speed water was converted to potential energy of pressure in the diffuser at the downstream of the throat. The applications of the SI to both the PCIS and the SI-PLR need high-pressure and large flow rate SIs. As soon as steam is supplied, a supersonic steam jet enters and condenses on the surface of the cold water jet in the mixing nozzle. After start-up, the discharge check valve opens under the influence of the high discharge pressure. Then, the drain check valve is closed by the near-vacuum pressure in the mixing nozzle. The SI is also able to pump up water to the PCC/IC pool of the SBWR by using discharged atmospheric pressure steam (SIPOWER: SI pool water refill) [9], which helps to make up the PCC/IC pool for vaporized coolant water as shown in Fig. 5.6.3. Dial coaxial-type SI is used for SI-driven feedwater heater (SI-FWH) and SI-driven jet pump for primary loop recirculation (SI-PLR) as shown in Fig. 5.6.3.

5.6.3

SI analysis model

Analyses of the two-phase flow in the SI were carried out for both the designs of the central jet type and the central steam type, as shown in Fig. 5.6.2, using the PHOENICS code. The two-phase flow analysis models used in the PHOENICS code were based on the results of the visual fundamental tests performed at Toshiba [3]. According to the observation in the tests, a steam jet or a water jet was formed in both the designs, which had a clear interface axisymmetrically along the central axis. Therefore, the separate-two-phase flow model between the steam jet and the water

Fig. 5.6.3 SI applications for the next-generation BWR with internal CRD [13].

496

Boiling Water Reactors

jet was adopted. In the case of the central steam nozzle design, as shown in Fig. 5.6.2b, the interface was determined at a void fraction of 0.5 in the cell mesh by radial and axial scanning during the calculation sweep. This new model and technique were used to analyze the SI-PCIS test results. The calculation in the case of the central water jet design was easier than in the case of the central steam design, because the water jet was cylindrical as shown in Fig. 5.6.2a. The diameter of the water jet was determined so as to achieve pressure continuation between the phases for each cell by using moving mesh functions. This model was used to analyze the SI-FWH, as shown in Fig. 5.6.3 after verification with fundamental tests. The SI achieved very high heat flux on the very small heat transfer area of the water jet surface by direct condensation [1]. It started up easily within a few seconds only when the air in the mixing nozzle was blown out through the overflow drain port and then steam condensation started in the nozzle [12,13]. Direct condensation onto the surface of the water jet is assumed for the steam (a), the water (b), and the interphase between the phases (c) model. The heat transfer coefficient for direct condensation is provided for each interface cell surface using the concept of ideal condensation based on the Clausius-Clapeyron equation and Bernoulli’s theorem for the vapor phase [1,4]. (a) Steam:

 ! div ρg U g ¼ m_ gl  ! ∂P div ρg U g V g  Γvg grad V g ¼  ∂r  !


 ∂P + C i W g  W i W g  W i div ρg U g W g  Γvg grad W g ¼  ∂z  ! DP div ρg U g hg  Γvg grad hg ¼  +Φ Dz  ! div ρg U g kg  Γkg grad kg ¼ Gkg  ρg εg  ! εg Gk  C2 ρg ε2g =kg div ρg U g εg  Γεg grad εg ¼ C1 kg g

(5.6.1) (5.6.2) (5.6.3) (5.6.4) (5.6.5) (5.6.6)

(b) Water:

 ! div ρl U l ¼ m_ gl  ! ∂P div ρl U l V l  Γ vl grad V l ¼  ∂r

(5.6.7) (5.6.8)

BWR innovations

 !


 ∂P div ρl U l W l  Γvl grad W l ¼  + C i W g  W i W g  W i ∂z + m_ gl W g

497

(5.6.9)

 ! div ρl U l hl  Γ gl grad hl ¼ 0

(5.6.10)

 ! div ρl U l kl  Γ kl grad kg ¼ Gkl  ρl εl

(5.6.11)

 ! εg div ρl U l kl  Γεl grad εl ¼ C1 Gkl  C2 ρl ε2l =kl kg

(5.6.12)

(c) Interface between the phases: The direct condensation onto the surface of the water jet was assumed. The heat transfer coefficients of direct condensation are given for each interface cell surface, by using the ideal condensation based on the Clausius-Clapeyron equation and Bernoulli’s theorem by Isshiki [1,4]:

H ci ¼

pffiffiffi 3=2 
 1=2 2 hlg = vg vg  vl fT sat ðT sat  T li Þg

(5.6.13)

m_ gl ¼ H ci ðT sat  T li ÞAi =hlg

(5.6.14)

 ! div ρl U l hl  Γ hl grad hl ¼ m_ gl hg

(5.6.15)

Interfacial velocities between the phases are given for each interface cell surface by

 W li ¼ τg W g + τg W l = τg + τg

(5.6.16)


 τg ¼ vg  vgT =Δrg

(5.6.17)

τl ¼ ðvl  vlT Þ

(5.6.18)

The abovementioned analysis models had been verified against the test results [7]. The steam injector treats very high heat flux through a very small heat transfer area by direct condensation. The steam injector starts up easily within a few second, only if the air in the mixing nozzle is blown out through the overflow drain port and the steam condensation starts in the nozzle.

5.6.4

Visualized fundamental tests

5.6.4.1 Test apparatus and measurements Figs. 5.6.4 and 5.6.5 show schematic diagram of the visual SI tests [4,11]. The test SI was modified based on a HilioPACTM by Helios research corporation established by Nicodemus [9], which consisted of a water jet nozzle, a mixing nozzle, an overflow

498

Boiling Water Reactors

Steam Water

Water Nozzle Windows for Visualization P1

Polycarbonate Annulus

Laser Prove

P2

Pitot Tube DP Cell

Mixing Nozzle

P3 Thermocouple

P4 P5

Throat Diffuser Fig. 5.6.4 Visual steam injector test devices [4].

Ast Z O

L=195mm

Y

DWO=9.4mm

At

X

Fig. 5.6.5 Coordinate system for visual test device [4].

BWR innovations

499

drain port, and a diffuser. The mixing nozzle has a stacked structure in which ring-shaped polycarbonate plates were piled-up via an O-ring, and a water jet flow can be visualized at the center of the mixing nozzle in the SI. A water jet was introduced from the central water jet nozzle while a steam jet was blown into the mixing nozzle through the annular gap between the water jet nozzle and the mixing nozzle, as shown in Figs. 5.6.4 and 5.6.5. Pressure, temperature, and a mass flow rate at the inlet and the outlet of the SI were measured. The radial velocity distribution of the water flow in the SI was measured at four axial locations with the transversally movable cylindrical pitot tube by Narabayashi et al. [4]. The radial velocity distribution of the steam flow was measured at four axial locations with the laser Doppler velocimetry (LDV). The radial temperature distribution of the flow was also measured at axial four locations with the transversally movable thermocouples [10,11]. An axial pressure distribution was also measured. These are depicted in Fig. 5.6.4. Super-sonic steam velocity was measured using the laser Doppler velocimetry (LDV) by Narabayashi et al. [4], and water velocity was measured using the pitot tube by Iwaki et al. [10,11]. The LDV system to measure super-sonic steam velocity produced by KANOMAX consisted of 1 W laser source, transmitting optics, a fiber optic probe, receiving optics, and a high-speed signal processor up to 150 MHz. The frequency shifting technique was used to decrease the observed Doppler frequency. As the distance between the fringes of the laser beam was 5.27 μm, the diameter of the detected droplets had to be much less than the laser beam fringe distance. Therefore, mist droplets moving with the steam flow were used as the tracer particles of the steam flow. In order to measure the radial temperature distributions in the mixing nozzle, the welded tensioned wire thermocouples were used [10]. The thermocouples consisted of butt-welded chromel and alumel wires of a diameter of 0.3 mm fixed on a stainless bow plate on the traverse device. Water velocity distributions were measured using the tensioned cylindrical pitot tube that could withstand high-speed water jet flow in the mixing nozzle because there was no tracer particle in the water jet. The outer diameter of the cylindrical pitot tube was 0.5 mm and the inner diameter was 0.2 mm. The tube had a pressure hole with a diameter of 0.2 mm at its center. We made a cylindrical-type pitot tube that traverses across the high-speed water jet by applying high-tension to the traverse device like a bowstring. The total pressure can be measured at the front of the cylinder at 0 degrees where the water jet impinges, and the static pressure at the side of the cylinder at 90 degrees. These two pressures can be measured with a differential pressure transducer connected with two thin stainless pipes. Since the differential pressure corresponds to the lead pressure (ρu2/2) in hydrodynamics theory, the flow velocity can be obtained by calculating the water density from the water temperature..

5.6.4.2 Test results Fig. 5.6.6 shows visualization results in the mixing nozzle. There was a clear water jet at the center of the mixing nozzle. The surface of the water jet was very smooth at the

500

Boiling Water Reactors

(a)

(b)

Fig. 5.6.6 Visualization of SI test results at the mixing nozzle [4]. (a) Visual mixing nozzle of SI, (b) visualized water jet at the center.

upstream, but minute waves on the surface and entrainments appeared in the downstream. By assuming a cylindrical water jet based on the visualized test, CFD analysis meshes was modeled by assuming the cylindrical water jet surface as the gas-liquid interface, it is possible to evaluate the momentum and energy transport through each mesh. Therefore, we measured the temperature distribution and velocity distribution using a very thin wire thermocouple and a very thin cylindrical Pitot tube. Steam velocity was measured by the laser Doppler velocimetry (LDV). In CFD analysis, moving mesh is used to converge so that the interfacial pressures of steam and water jets are the same, and the diameter of the water jet can be obtained. As shown in the analysis results of Figs. 5.6.18 and 5.6.19, the measurement results and the CFD analysis results showed good agreement. Fig. 5.6.7 shows steam velocity distributions measured with the LDV. The bulk region steam velocities at the axial positions of x/L ¼ 0.05 and x/L ¼ 0.38 were around 450 m/s except for area close to the water jet, which is a supersonic velocity of Mach 1.04. Fig. 5.6.8 shows the effect of inlet steam pressure on the steam velocity. As the inlet pressure increases, the steam velocity increases slightly. It suggests that an

BWR innovations

501

Fig. 5.6.7 Steam velocity distributions measured with LDV [4].

Steam velocity (m/s)

500 400 300 200

Water Jet boundary at x/L=0.38 Water jet boundary at x/L=0.05

100

x/L=0.05 x/L=0.38

0 0.0

0.2

0.4

0.6 y/R(x)

0.8

1.0

Fig. 5.6.8 Effect of inlet steam pressure on steam velocity [4].

Steam velocity (m/s)

500 400 300 200

Water Jet boundary at x/L=0.05 Case 1: Ps=0.13MPa Case 2: Ps=0.14MPa Case 3: Ps=0.15MPa x/L=0.38

100 0 0.0

0.2

0.4

0.6 y/R(x)

0.8

1.0

increase in the critical flow due to higher inlet steam pressure results in an increase in the supersonic velocity. Fig. 5.6.9 shows the water jet velocity distributions measured with the very thin cylindrical pitot tube. Those are contrasted with the flow in a pipe: the velocity is the highest at the center and zero at the wall. Large shear stress was axially exerted on the surface of the water jet and the water jet was accelerated. Fig. 5.6.10 shows radial temperature distributions in the mixing nozzle. A large temperature fluctuation near the water jet surface suggests that the turbulent thermal energy eddy transport in the center of the water jet. A comparison of velocity distribution with temperature distribution in the water jet is shown in Fig. 5.6.11. It suggests that the turbulent Prandtl number is nearly constant in the water jet. In general, the Taylor-Prandtl analogy would hold when the Prandtl number is almost equal to 1.0 in a pipe flow. Furthermore, dimensional similarity rule seems to be applicable to the water jet in the SI.

502

Boiling Water Reactors

Fig. 5.6.9 Water jet velocity measured with pitot tubes [4].

30

Water velocity (m/s)

25

20 15 10 5 0

Fig. 5.6.10 Radial temperature distributions [4].

x/L=0.33

Case 1: Ps=0.13MPa Case 2: Ps=0.14MPa Case 3: Ps=0.15MPa

x/L=0.52

Case 1: Ps=0.13MPa Case 2: Ps=0.14MPa Case 3: Ps=0.15MPa

-0.2

-0.4

0.0 y/R(x)

0.4

0.2

70

50 x/L=0.52

40 30 20 0 -10

-5

x/L=0.52

10

0 5 y(mm)

15

40

70 60 Temperature (oC)

Fig. 5.6.11 Comparison of the velocity distribution with the temperature distribution in water jets [4].

x/L=0.33

x/L=0.33

35

Water velocity at x/L=0.52

50

30

40

25

30

20 -1.0

Water Temperature at x/L=0.33

x/L=0.33 x/L=0.52

20 15

-0.5

0.0 y/R(x)

0.5

Water velocity (m/s)

Temperature (oC)

60

BWR innovations

5.6.5

503

Application of steam jet-type SI to PCIS

Fig. 5.6.11 shows the concept of the SI-driven PCIS (SI-PCIS) [5,7,8,13]. It is simply necessary to open water supply valve and steam supply valve to start up the SI. The SI is started up when steam is injected into water. As soon as steam is supplied, the supersonic steam jet flow is formed and condenses on the surface of the cold water jet in the mixing nozzle. When the SI starts up, the discharge check valve opens due to the high discharge pressure in the diffuser. The drain check valve is then closed by near-vacuum pressure in the mixing nozzle. This is a highly reliable start-up mechanism that does not require electricity for motor and only requires the operation of a passive check valve. By supplying steam and water, steam condensation occurs immediately in the steam injector, the inside of the mixing nozzle becomes sub-atmospheric pressure, and when the check valve is closed, it becomes a stable operating condition. The SI can be operated without AC power even during station blackout. It thus possesses great merits in terms of the passive safety of nuclear power plants. The SI for the actual SI-PCIS needs the discharge flow rate of about 6.9 kg/s, and the maximum discharge pressure of 8 MPa by 7 MPa steam, as shown in Fig. 5.6.12 of the SI operating range map.

5.6.5.1 High-pressure tests and analysis High-pressure tests for the PCIS were carried out by using a high-pressure steam test facility at Toshiba, as shown in Fig. 5.6.13. The 1/2 scale PCIS model shown in Fig. 5.6.14 was installed in the facility. The nozzle dimension was optimized by PHOENICS code analyses. The high-pressure steam from the pressure vessel was generated from deaerated boiling water in the pressure vessel using the 1 MWe heater bundle. As shown in Fig. 5.6.15, the 1/2 scale PCIS model attained a discharge pressure of about 8 MPa with 7 MPa steam and 0.4 MPa water supplies, which is enough to inject water into the RPV. The test model was started up for 9 s after the steam supply valve was opened. Pressure in the overflow port #1 was almost vacuum, and the overflow port #2 was at about 0.1 MPa. The discharge pressure had a peak of about 8 MPa after 28 s. The peak was caused by the two-phase flow from the pressure vessel Fig. 5.6.12 SI operating range map [22].

Pressure(MPa)

15

10

SI-PLR

1/5 Scale

PCIS(RCIC) 5 1/1

1/9.2 1/7.1

SIPOWER

0 1

10

SI-FWH 100

㻲㼘㼛㼣㻌㼞㼍㼠㼑㻌㻔䡐㻛㼔䠅

1/2

1/1

1000

Pressure vessel #1

Pressure vessel #2 5F 4F Model SI for PLR 3F

Water 2F Steam

Overflow drain

1F

680mm

Fig. 5.6.13 High-pressure steam test facility.

Fig. 5.6.14 1/2 Scale PCIS model.

BWR innovations

505

Fig. 5.6.15 High-pressure steam test result.

overflow into the steam supply line, which was detected by the water level transmitter of the pressure vessel. Fig. 5.6.16 shows the analysis results of the PCIS tests. The analysis boundary conditions of an inlet steam pressure, inlet water pressure, and a discharge flow rate are the same as those of the test conditions. The velocity vectors and the void contours in Fig. 5.6.16 show that the supersonic velocity steam jet accelerates water near the mixing wall. The steam is condensed just before the throat, causing a sudden rise in pressure. As shown in Fig. 5.6.16, the discharge pressure in the analysis was 10.2 MPa, which is higher than that in the test data plotted with solid circles. The SI-PCIS test results of the discharge pressure were lower than the analysis results. This shows that the design of the diffuser has room for improvement. Another technical task is that the overflow drain has a sufficiently low flow rate. A large amount of supersonic steam jet is required to pressurize the water to a high pressure. If this Fig. 5.6.16 Analytical result for PCIS test.

506

Boiling Water Reactors

amount of overflow drain is large, the cooling water will be wasted. This problem occurs when the steam injector supplies water from the steam locomotive’s water tank to the high-pressure boiler, so it is necessary to optimize the nozzle shape using advanced CFD analysis. If the water temperature rises too high, the saturation pressure inside the mixing nozzle will also rise and it will no longer suck in low pressure water in a water tank. This is the working limit of the steam injector.

5.6.6

Application of water jet type SI to RLP

5.6.6.1 Confirmation of analysis method (a) Results of analysis of visual test at low pressure: Conditions for the low-pressure visual test are shown in Table 5.6.1 and Fig. 5.6.17. The predicted cross-sectional velocity distribution at z/L ¼ 0.05 in the mixing zone (L ¼ 195 mm) is presented in Fig. 5.6.18 along with the measured velocities with the LDV. The steam velocity measured by LDV and the steam velocity obtained by analysis are in good agreement. Predicted cross-sectional temperature distributions at z ¼ 58 mm and z ¼ 101 mm are compared with the measured results with the thermocouple in Fig. 5.6.19. The measured temperatures in the steam flow region are lower than that predicted near the surface of the water jet. It is probably because of the waving motion of the water jet surface. (b) Results of analysis of small-scale test at high pressure: Based on the visual tests at low pressure, small-scale test results at high pressure were analyzed. Predicted and measured discharge pressures and temperatures are summarized in Table 5.6.2. As shown in Figs. 5.6.20 and 5.6.21, the predicted axial steam velocity becomes over 800 m/s in the mixing zone. The axial water velocity largely rises there by the shearing force of the supersonic steam flow and the axial momentum import due to the condensation of the supersonic velocity steam. Thus, the axial velocity of the water jet at the surface is faster than that at the center as shown in Fig. 5.6.21. These results approve the validity of the newly developed separate-two-phase flow model to analyze the flow in the SI which was installed in the PHOENICS code. As the analysis results for the low-pressure visualized model test showed good agreement with the test data, analysis for the small-size high-pressure model test was done. Table 5.6.1 Boundary conditions of low-pressure visual-test and small-scale high-pressure test [4]. Boundary

Parameter

Low-pressure visual-test

Small-scale highpressure test

Inlet steam port

Pressure (MPa) Quality Pressure (MPa) Temperature (°C) Controlled flow rate (kg/s)

0.132 1.0 0.17 14.3 1.26

5.5 1.0 1.60 12.0 0.83

Feedwater port Discharge port

BWR innovations

507

Steam Inlet Press: 0.132MPa Quality: 1.0

Water Inlet Press: 0.170MPa Temp.: 14.3OC

Metal Wall Momentum: Friction Energy: Adiabatic

Discharge Outlet Water Flowrate: 1.26kg/s

Overflow Outlet If OF Press > 0.1MPa Then, Overflow drain

Fig. 5.6.17 Boundary conditions for visualized test model [4].

Fig. 5.6.18 Analysis results of velocity for the visualized fundamental test results measured by LDV [4].

600

Velocity (m/s)

500 400 300

LDV data

200

Analysis

100

0

0

2

4 6 8 10 12 Radial distance r (mm)

14

16

Fig. 5.6.19 Analysis results of temperature distributions for the visualized fundamental test results [4].

508

Boiling Water Reactors

Table 5.6.2 Analytical results for small-size high-pressure model test [4].

Discharge pressure Discharge temperature

Analysis

Test

9.1 MPa 159°C

9.0 MPa 157°C

1000m/s

Pressure (MPa)

Fig. 5.6.20 Analysis results of velocity vectors and pressure contour for small-scale highpressure steam injector [1].

Velocity (m/s)

Steam velocity of small size SI

Steam velocity of 1/5.5 model Steam velocity of 1/1 size

Water velocity of 1/5.5 model Water velocity of 1/1 size

Distance from inlet of mixing nozzle Z (mm) Fig. 5.6.21 Cross-sectional velocity distribution along mixing zone in low-pressure visual-test and high-pressure 1/5.5 scaled model [4]. Analysis conditions are shown in Table 5.6.2. Fig. 5.6.22 shows the pressure distributions of the analysis results and measured data along the wall. The analysis data show good agreement with the test data. The pressure in the mixing nozzle was about 0.5 MPa and the pressure increased from 2 to 7 MPa over an axial distance of 20 mm through the throat. Comparisons between the analysis results and test results of discharge flow pressures and temperatures are shown in Table 5.6.2.

BWR innovations

509

14

1/5.5 Model

12

Pressure(MPa)

1/1 size

Fig. 5.6.22 Analysis results of pressure distributions for largescale steam injector [4].

10 8

Water pressure

6 4

Steam pressure *1/5/5 model: z’=5.5*z

2 0 -400 -200

0

200

400

600

800 1000 1200 1400

Distance from inlet of mixing nozzle Z (mm)

5.6.6.2 Scale-up examination of SI for application to RLP The scale-up examination of the SI to apply SI to the RLP was performed analytically. The analyses of the 1/5.5 small-scale model and the 1/1 full scale model and were performed. Analysis conditions are presented in Table 5.6.3. In order to avoid the mesh scale effect caused by the differences of component size, similar meshes arrangements were adopted in both the models. As shown in Fig. 5.6.21, the axial-temperature distributions are the almost the same in both the cases. The analysis result of the 1/5.5 scale model is depicted for an axial length multiplied by 5.5 in the figure. The radial temperature distributions at z ¼ 250 mm and z ¼ 632 mm were also found to be in good agreement with 1/5/5 and 1/1 scale. Analysis results of the pressure distribution are shown in Fig. 5.6.22 using the 5.5 times axial length for the 1/5.5 model. Both the results are almost the same except that the pressure of the 1/1 scale model is a little lower by 0.5 MPa than that of 1/5.5 model in the discharge zone. The results of Fig. 5.6.22 indicates that the steam supply of 3 MPa can pump up and elevate the water flow from 7 to 12.4 MPa by the SI, which suggest that the large recirculation pump in the current design BWR can be replaced with the SI-driven jet pump (SIDJP). As shown in Fig. 5.6.23, the longitudinal temperature distribution of steam and water temperature of the two steam injectors with different scales of 1/5.5 and 1/1, and the radial temperature distribution of the water jet are in very good agreement, Table 5.6.3 Analysis conditions of SI scale-up examination for application of SI to RLP [4]. Design parameter

1/1 Model

1/5.5 Model

Water jet length (mm) Water nozzle diameter (mm) Throat diameter (mm) Steam pressure (MPa) Water pressure (MPa) Water temperature (°C) Discharge flow rate (kg/s)

650 25.0 27.0 3 7 40 61.1

118 4.55 4.91 3 7 40 2.0 (¼ 1/5.52)

510

Boiling Water Reactors

300

400 300

Temperature (OC)

Steam temp. of 1/5.5 Model

200

Steam temp. of 1/1size

200

Z=632mm Z=250mm

100

*1/5/5 model: z’=5.5*z

Water temp. of 1/5.5 Model Water temp. of 1/1 size

100

Temp. of 1/5.5 Model Temp. of 1/1 size

0 0 0 10 20 30 40 50 -400 0 400 800 1200 Distance from inlet of mixing nozzle Z (mm) Radial distance r (mm) Fig. 5.6.23 Cross-sectional temperature distribution in mixing zone in low-pressure visual-test [4].

it can be seen that similarity-low is established. As a result, it is expected that the specified performance will be obtained even when a full-size SI-PLR is manufactured and operated in the future.

5.6.6.3 High-pressure tests using scale models Narabayashi et al. [13] conducted characteristic evaluation tests using the scaled models of the SI plus the SI-driven jet pump (SIDJP), under the same pressure conditions as the actual plant. (a) Test models: The central water jet-type SI was used for high-pressure test. The central steam jet-type SI is stable under the condition of high-pressure steam and low-pressure water supply. The central water jet-type SI is suitable for high-pressure inlet water because feed water is pumped up by condensate pump and water jet is accelerated without contacting a mixing nozzle wall in the SI, and obtain large pressure gain. The 1/5 scale of SI-PLR model used to drive jet pump for primary loop recirculation model is shown in Fig. 5.6.26. The nozzle dimensions were optimized by using the PHOENICS. Table 5.6.4 Test conditions of high pressure tests using scale models [13]. 1/5 SI-PLR model Type of SI Jet length (mm) Jet nozzle diameter (mm) Steam pressure (MPa) Water pressure (MPa) Water temperature (°C) Discharge flow rate (kg/s)

Central water jet 135 (1/5) 5.0 (1/2) 3 7 30 2.4(1/52)

BWR innovations

511

Handwheel Stem for Water Jet Nozzle Water Port

712 mm

712mm

Steam Port

Pressure Taps

Water Jet Nozzle

Overflow Port #1 Overflow Port #2

Discharge Port

(a)

(b)

Fig. 5.6.24 The 1/5 scale SI-PLR model [13]. (a) Schematic cross section, (b) 1/5 scale SI-PLR test model. (b) Test conditions: Table 5.6.4 shows the test conditions of high-pressure tests using scaled models The 1/5 scale SI-PLR model is shown in Fig. 5.6.24. The accuracy of the pressure sensors used in this test was 0.1%. The orifice flow meters and temperature measurement system had 1% accuracy. Demineralized water was used. In the test facility, the 3-MPa steam from the pressure vessel was supplied and deaerated boiling water in the pressure vessel using 1-MWe heater bundle. (c) Test results: Fig. 5.6.25 shows the comparison of test results and analysis results of the 1/5 scale SI-PLR model. A discharge pressure of 12.5 MPa was achieved with 3 MPa steam and 7 MPa water. The test results showed a good agreement with the analysis results. The test

(a)

(b)

Fig. 5.6.25 High-pressure test results of the 1/5 scale SI-PLR [13]. (a) Pressure trend after start up. (b) Pressure distribution along Z axis.

512

Boiling Water Reactors

unit was started up 20 s after the steam supply valve was opened. The operation was very stable after start-up, which demonstrated the feasibility of the application of the SI-PLR to next-generation BWRs including SMR. Fig. 5.6.26a shows color contour and flow vector of pressure obtained from CFD analysis. Pressure increased sharply by the diffuser through the throat at the exit of the mixing nozzle and rises as kinetic energy is converted to pressure energy. Attention must be paid to the erosion caused by the high-speed jet in the throat, and to the cavitation caused by the compression and disappearance of fine bubbles in the diffuser. Fig. 5.6.26b shows the image of the SI-PLR applied to the next-generation BWRs as the PLR pump. The SI-PLR is composed six Sis in a casing. The nozzle shape is the same as the CFD analysis.

(a)

Steam

Actuators for steam flowrate control

Discharge to drive Jet pump

Feedwater

Startup overflow drain

(b) Fig. 5.6.26 Application image of SI-PLR to next BWRs as a PLR pump [13]. (a) Analysis result of SI-PLR’s pressure contour and velocity vectors. (b) SI-PLR pump of integrated six full-scale SI in a casing.

BWR innovations

5.6.7

513

Simplified feed water system by SI

The steam injector is a passive jet pump that has no movable part, and can pump up and boost water just by supplying steam as already stated [1,4,12]. The steam condenses on the water jet, and water single-phase flow is discharged. The enthalpy of the incoming steam condenses on the surface of the water jet and raises the temperature of the water, thus increasing the enthalpy of the exiting water, as shown in Eq. (5.6.1). Enthalpy is an important thermodynamic parameter used in steam turbines, etc., so please refer to any thermal engineering book. The energy equation of the supplied steam and the water jet, and the discharged water flow describes the boost of the water flow. It also indicates that the supplied water jet is heated up, which suggests that the SI works as a heat exchanger. The direct condensation of the steam on the water jet seems quite effective; when steam extracted from a turbine is used in the SI to pump up condensate from the condenser, the system would work as the heat exchanger as well as a booster pump. The multistage SI system to boost and heat up feed water for the BWR RPV was proposed as shown in Fig. 5.6.27 [15–18]. It consists of four serial SIs to boost water flow sufficiently. The final stage has also deairing function. Since the SI does not have movable parts and super effective heat exchange ability, the system is quite compact, which results in the simplification of the feed water heater (FWH) system of a BWR to reduce the volume of the FWH system greatly. The system diagram of balance-of-plant (BOP) of the ABWR are shown in Fig. 5.2.8. Fig. 5.6.28 shows application image of SI-FWHs to next BWRs as a feedwater heater integrated six full-scale 4 stage SI in a casing, instead of large 12 low-pressure feedwaters (FWHs). One low-pressure FWH size is about 2 m in diameter and 13 m in length and installed as necked heaters in three main turbine condensers, as shown in Fig. 5.6.29. When the FWH system of the ABWR, 4 stages times 3 series need 12 heaters as shown in Fig. 5.6.30a. As shown in Fig. 5.6.30b, they will be replaced with the six parallel four-stage SI feedwater heater (SI-FWH) system as shown in Fig. 5.6.30b or the integrated six full-scale SI in a casing as shown in Fig. 5.6.28, it would be simplified and greatly volumetrically reduced as shown in Fig. 5.6.31.

First Stage Second Stage Third Stage Final Stage with Jet deaerator

Water Jet

Fig. 5.6.27 Four-stage steam injector system.

514

Boiling Water Reactors Recirc 4th stage 3rd stage 0.2MPa Steam 0.4MPa Steam Steam

2nd stage 1st stage 0.1MPa 0.05MPa Steam Steam

Feedwater 2nd stage Startup Drain

1st stage Startup Drain

Fig. 5.6.28 Application image of SI-FWH to next BWRs as a feedwater heater integrated six full-scale SI in a casing [13].

Fig. 5.6.29 Main turbine condenser with neck heaters.

BWR innovations

515

0.4MPa

0.21MPa

209t/h

198t/h

2.2MPa

0.1MPa 216t/h

Extracked steam 0.21MPa 0.1MPa

0.05MPa 313t/h

Buffer tank

117oC

139oC

97oC

Low-pressure heaters A,B,C 3 series X 4 stage =12units

75oC

0.4MPa

0.05MPa

42oC 0.29MPa

49oC 2.8MPa

Low-pressure drain tank Low-pressure drain pump 0.9MW

(a)

42oC High-pressure condensate pump 5.7MW

Booster pump Multi-stage steam injector

High-pressure condensate pump

(b)

Fig. 5.6.30 Simplification of low pressure heaters. (a) Low-pressure feedwater heaters (LPFH), (b) SI-FWH.

Recirc Steam

0.4MPa Steam

Buffer Tank

0.21MP a S #4th Jet Deaerator

#3rd SI

0.05MPa Steam

0.1MP a St #2nd SI

#1st SI

Feed Water

Discharge

Fig. 5.6.31 3D CAD design of simplified steam injector feedwater heaters (SI-FWH) in a turbine condenser.

5.6.7.1 Scaled model tests of simplified feedwater system Experimental work to develop the SI-FWH was performed by Toshiba and TEPCO using scaled-down apparatuses based on the analyses [14–17]. The outline of the facility is presented in Fig. 5.6.32. In the tests, the 1/7 scale model and the 1/5 scale model were used based on the scaling analogy. Table 5.6.5 summarize the test results of the 1/7- and the 1/5-scaled test model in accordance with the supposed full-scale target values and the ABWR conditions. The diameter and the length of the SI were determined to obtain the target values of the discharge pressure and the temperature, and the discharge water flow rate from the supplied steam pressure, the temperature and the flow rate, the supplied water pressure, the temperature and the flow rate based on the analyses [15–18]. The ratio of the diameter and the length of the 1/5 model to those of the 1/7 model is the square root of 1/5 and 1/7. The test results of the outlet temperatures and the pressures at each stage in the 1/5 scale model were almost the same as those of the 1/7 scale model, and

516

Boiling Water Reactors

Steam

Steam

Water

1st-Stage Steam Injector (b)

(a)

Fig. 5.6.32 Multistage steam injector test facility. (a) First-stage steam injector, (b) first- and second-stage steam injector.

Table 5.6.5 Test results of pressures and temperatures for SI-FWH. Case Model scale 2nd-Stage size (mm) Steam pressure (MPa)

Feedwater flow rate (kg/s) Inlet jet flow pressure (MPa)

Exit pressure (MPa) Each stage inlet temperature (°C)

Exit Temp. (°C)

Stage Dia. Length #1 #2 #3 #4 #1 #1 #2 #3 #3 #1 #2 #3 #4 #4

ABWR

Target

Test model

– 2m 15 m 0.05 0.11 0.21 0.40 400  3 – – – – 42 75 97 117 139

1/1 455 4900 0.05 0.11 0.21 0.40 200  6 1.6 0.9 0.6 0.4 42 75 97 117 139

1/7 64 693 0.05 0.11 0.21 0.40 4.0 1.6 – – 0.44 42

1/5 91 980 – 0.11 0.21 – 8.0 – 0.9 – 0.41 – 75

117 142

116 –

close to the target values. The size of the full-scale SI-FWH is determined to supply 7600 ton/h of feed water by six multistage Sis. Therefore, 1/1 scale one multistage SI rated flow needs to supply 200 kg/s of feed water, 1/5 scale test model needs to supply 8 kg/s feed water, and 1/7 scale model needs to supply 4 kg/s feedwater as shown in Table 5.6.5. The flow rate was adjusted using coaxial inlet nozzle at first-stage SI.

BWR innovations

517

5.6.7.2 Analysis for improving SI-FWH It is not efficient and economically unbearable to carry out many experimental works to improve and verify the SIs. Improving the SIs to apply to the FWH was analytically examined. (a) CFD analysis model: Analyses were performed using the Star-CD code in which the separate-two-phase flow model for the steam and the water jet flow in the SI [文献] was installed. The analyzed FWH was composed of fourth-stage SIs and the configurations of the first three stages are presented in the form of the 3D-CAD in Fig. 5.6.32. The water jet part is indicated by blue color in the figure. The performance improvement of the SI for the FWH enhances the BWR thermal efficiency. The performance of the first-stage SI has a great effect on the SI-FWH system efficiency since it is driven by the lowest pressure steam of 0.05 MPa extracted from a turbine. Therefore, the effect of the shape of the steam inlet on the performance improvement was examined by CFD analysis, such as enlarging the inlet radius of the inlet bell mouth of the first-stage SI and also extending the length of bell mouth nozzle, as shown in Fig. 5.6.33. Fig. 5.6.34 illustrates analysis mesh distributions for the SIs. The pressure and the temperature boundary conditions of the analysis are also included in the figure, which were derived from the FWH and the turbine conditions of the ABWR. 3rd-stage

2nd-stage

1st-stage Water Jet Large-diameter bell-mouth Enlarged and extended inlet and mixing nozzle bell-mouth Water Jet

Flow control needle

(a) (a)Enlarged Original bell-mouth bell-mouth

(b) Extended bell-mouth

Fig. 5.6.33 3D-CAD of multistage steam injector’s analysis model from first to third stage.

518

Boiling Water Reactors

Fig. 5.6.34 Mesh distributions and analysis boundary conditions [19]. (b) Analysis results: Analyses were parametrically carried out for the first-stage SI to get higher discharge temperature by varying the inner diameter and the length of the inlet bell mouth of the first stage. When the bell mouth was applied for the mixing nozzle adopted in the first design of the steam nozzle inlet, the outlet temperature of the first-stage SI was 51.7°C. The analyses were parametrically carried out, varying the inner diameter of the inlet bell mouth, and the diameter and the length of the mixing nozzle through the first-stage steam injector. A series of the parametric survey was performed to seek better configuration of the first-stage SI to increase the outlet temperature. When the bell mouth was applied for mixing nozzle adopted in the last design to enlarge the steam: – First case of the analysis: The result of first case of the analysis result is shown in Figs. 5.6.35 and 5.6.36. As shown in Fig. 5.6.35a, pressure at each stage is almost the same as the supplied steam from the turbine extracted steam port, except in a mixing nozzle in the first stage. The analysis result of the water temperature in Fig. 5.6.35b was 67°C. The first-stage outlet water temperature was calculated (Eq. 5.6.19) by using the inlet steam flow rate and water flow rate in the first-stage mixing nozzle. It was also 67˚C. ml hl + mg hg  , Tmix ¼
ml + mg Cpl

(5.6.19)

here m is mass flow rate (kg/s), h is enthalpy (kJ/kg), and Cpl is specific heat. The value was less than that of the current ABWR’s feedwater heater. As shown in Fig. 5.6.35b, the analysis results of steam temperature distribution curve in the first stage had a valley in the mixing nozzle from 80°C to 70°C at the bottom.

BWR innovations

519

Fig. 5.6.35 Temperature and pressure distributions of steam and water in first- to third-stage SIs in original design case [19]. (a) Pressure distribution, (b) temperature distribution.

Fig. 5.6.36 Temperature, pressure, and axial velocity contour from the first to the third-stage SI: Case (a). (a) Axial velocity contour, (b) pressure contour, (c) temperature contour.

520

Boiling Water Reactors

As shown in Fig. 5.6.35b, the analysis results of steam temperature distribution curve in the first stage had a valley in the mixing nozzle from 80°C to 70°C. The reason why the steam temperature decreased to 70˚C was that the steam velocity in the mixing nozzle was too high to decrease static pressure, according to the polytropic expansion line [4] and steam temperature also decreased as the saturation temperature of the steam pressure decreases, forming some moisture in the two-phase flow area, as shown axial velocity peak in Fig. 5.6.35. Large bell mouth and extended mixing zone case: The inlet bell mouth was enlarged and the mixing zone was prolonged in order to improve the heating-up efficiency of the first-stage SI. As shown in analysis results in Fig. 5.6.36, there is still a steam velocity peak and the steep valley-shaped temperature drop in the latter half of the mixing zone, and the improvement in the exit temperature of the first-stage SI is not sufficient at 69° C, as shown in Fig. 5.6.37. Improvement is only 2°C increase in the discharge water temperature. With the improved bell mouth, which had an enlarged inlet diameter and lengthened length of the mixing nozzle for the first stage, the temperature rose at the outlet of the first stage increased only 2°C. Then, we improved the inner diameter of the downstream in the 1st-stage mixing nozzle and the water nozzle of the second stage were enlarged as shown in Fig. 5.6.38. As shown in the temperature contour in Fig. 5.6.39 and the axial temperature distribution in Fig. 5.6.40b, the outlet water temperature of the first stage rose to remarkable temperature of 75˚C. The steam velocity in Fig. 5.6.40a is reduced to a maximum of 250 m/s, and the static pressure and its saturation temperature were successfully increased. As a result, the multi-stage SI feedwater heater achieved almost the same feedwater heating performance as the current ABWR feedwater heater.

5.6.7.3 Transient test result of the first stage When the SI-FWH is introduced, it should withstand abnormal conditions as well as steady-state conditions. It was tested experimentally how the multistage SIs for the FWH respond to the abnormal conditions. The multistage steam injectors proved capable of performing transient operation of steam pressure and feed water as shown in Fig. 5.6.41. Even if one of the steam injectors tripped due to a drop in the flow rate of steam supply, according to the test findings, they could restart and resume the operation. 400

130

Steam axial-velocity peak

120 300

Temperature ( o C)

Velocity(m/s)

110 200

Steam

100

0

Water

100 90

Steam

80 70

69°C

60

Water

-100

1st-stage -200 –500

(a)

0

500

2nd-stage 1000

1500

50

3rd-stage 2000

Distance from the 1st-stage water jet nozzle (mm)

1st-stage

40 –500

2500

(b)

0

500

2nd-stage 1000

1500

3rd-stage 2000

2500

Distance from the 1st-stage water jet nozzle (mm)

Fig. 5.6.37 Temperature and pressure distributions of steam and water in the first- to the third-stage SIs in first-stage enlarged-extended bell mouth and prolonged mixing zone case. (a) Axial-velocity distribution, (b) axial temperature distribution.

BWR innovations

521

Fig. 5.6.38 Enlarged and extended bell mouth, and prolonged and widened mixing zone of firststage SI [19]. (a) Extended large bell-mouth, (b) improved diameter mixing nozzle.

Extended large bell-mouth

Improved thick-dia. mix. nozzle

(a) Extended large bell-mouth

Improved thick-dia. mix. nozzle

(b) Extended large bell-mouth

Improved thick-dia. mix. nozzle

(c)

Fig. 5.6.39 Comparison of the improved effects on steam injectors from the first to the thirdstage SI: Case (c) [19]. (a) Axial-velocity contour, (b) pressure contour, (c) temperature contour.

522

Boiling Water Reactors 130

400

Steam axial-velocity peak was decreased

120

V e lo c ity (m /s )

300

110 100

Temperature (°C)

200

Water 100 0

1stt

2nd-stage 3rd-stage

-100

90

Steam

80 75oC

70 60

Water

1stt

2nd-stage 3rd-stage

50

-200 -500

(a)

40

0

500

1000

1500

2000

Distance from the waterjet nozzle (mm)

-500

2500

(b)

0

50 0

1000

1500

2000

2500

Distance from the waterjet nozzle (mm)

Fig. 5.6.40 Temperature and pressure distributions of steam and water in first- to third-stage SIs in enlarged and extended bell mouth and prolonged and widened mixing zone case [19]. (a) Axial-velocity distribution, (b) axial temperature distribution.

Mass flow rate (kg/s), Steam Pressure (MPa) 㻢㻚㻜 㻡㻚㻜 㻠㻚㻜 㻟㻚㻜

Mass flow rate Steam pressure

㻞㻚㻜 㻝㻚㻜 㻜㻚㻜 㻜

(a)

㻡㻜

㻝㻜㻜

㻝㻡㻜

㻞㻜㻜

Time (sec)

㻞㻡㻜

㻟㻜㻜

㻟㻡㻜

(b)

Fig. 5.6.41 Performance of multistage SIs in transient conditions [19]. (a) Transient test condition, (b) transient test result.

5.6.7.4 Advantages of SI introduction to ABWR in volume and mass reduction It is expected that the volume and the mass of the ABWR FWH system could be reduced greatly by introducing the SI-FWH system as already shown in Figs. 5.6.29–5.6.31. The bird’s eye views of the FWH layout of the current ABWR and the SI-FWH introduction case in the turbine are presented in Fig. 5.6.42. The volume and the mass of those derived from the figures and design sheets are illustrated in Fig. 5.6.43, which indicates that the FWH system could be reduced approximately to a third in volume and mass by introducing the simplified SI-FWH system to the ABWR. It might result in a significant reduction in construction cost, shortening of a construction period, and the reduction of maintenance and repair work significantly. It is obvious that the SI-FWH system has an economic advantage. The simplicity of the system also enhances the reliability and the certainty of the ABWR.

BWR innovations

523

3.5m

C

C B

B A

A

(a)

(b)

Fig. 5.6.42 Comparison of volume and weight for the feedwater system between ABWR and simplified plant. (a) Current ABWRs turbine building layout, (b) simplified SI-FWH.

500

ABWR

ABWR

400

Simplified 300

by S I

Simplified by SI

200 100 0 Volume (m3)

Weight (ton)

Fig. 5.6.43 Comparison of volume and weight for the feedwater system between ABWR and SI-FWH plant.

5.6.8

Steam injector (SI) pump-up water system to refill pool for passive containment cooling isolation condenser (PCC/IC); SIPOWER

5.6.8.1 Concept of SIPOWER The steam injector (SI) can operate at very low steam pressure such as atmospheric pressure (0.1 MPa), and discharge water at 0.5 MPa. As soon as steam is supplied to the SI, it starts up autonomously, and then a discharge check vale is opened to discharge boosted water and the drain check valve is closed autonomously because of the boosted pressure in the diffuser. The SI is very reliable and extremely structurally

524

Boiling Water Reactors

simple. One useful application of the SI is introduction into the simplified boiling water reactor (SBWR) [9,21]. The SBWR has the passive containment cooling (PCC) isolation condenser (IC) system. This system is called the PCCS/IC. When a loss-of-coolant accident (LOCA) happens, steam in the contain vessel (CV) and that generated in the reactor pressure vessel (RPV) flows into the IC. The IC is cooled by water in the PCC/IC pool. If the LOCA continues, boiling starts in the pool and the water level in the pool gradually decrease and finally the water in the pool might boil off if it is not made for boiling water. It is desirable to have water make-up system from the outside of the reactor building to the pool. One idea is to apply the SI to the pump to refill the PCCS/IC pool. Since the SI works autonomously and passively without external power such as electrical power, it is most suitable for the PCC/IC. The SI is able to pump up water to refill the PCC/IC pool from the water tank outside the reactor building by only using steam generated by the PCCS/IC. Moreover, the IC also takes a role of heat sink which releases the heat generated in the RPV to the outside of the reactor building since the SI is an excellent heat exchanger. Here, this SI-applied PCC/IC system is called the SIPOWER. Fig. 5.6.44 shows the conceptual schematics of the SIPOWER. The secondary steam from the PCC/IC pool is supplied to the SI. The SI pumps up water from the water tank outside the reactor building. The SIPOWER is a simple and compact safety injection system (SIS), which rationalizes the layout in a reactor building. It reduces the area and volume of the PCC/IC pool to 39%. According to analysis and test results, the SBWR grace period is prolonged from 3 days to 30 days by using the SIPOWER [9, 22].

Fig. 5.6.44 A conceptual schematic of the steam injector application to the SIPOWER for PCCS [9]. PCC/IC POOL

WATER TANK

RPV

S/C

SI

BWR innovations

525

5.6.8.2 Evaluation of PCC/IC pool water level transient by SIPOWER The performance and the effectiveness of the SIPOWER applied to the SBWR during the LOCA of DBA were confirmed by the analyses with the Toshiba steam injector analysis code (TOSIA) [12]. As shown in Table 5.6.6, three cases are described here [9]. Case 1 for the base design of SBWR. Case 2 for the SI-applied SBWR. Case 3 for the SI-applied SBWR with the reduced PCC/IC pool. In terms of performance, the water level transient in the PCC/IC pool is important which decides the grace time defined as a time period to allow inaccessibility. This grace period is quite important from the perspective of inherent safety of a nuclear reactor. Even if some accident happens in the reactor, the safety of the reactor is maintained autonomously without any auxiliary assistance from the outside and sufficient grace time is guaranteed. The conditions in the three cases are summarized in Table 5.6.7. As shown in Fig. 5.6.45, in the SBWR, the cross-sectional area of the PCC/IC pool is 825 mm2 and the water level in the pool is 4.4 mm. It is assumed that the pool is filled with water at 30°C. In the third case, the cross-sectional area of the PCC/IC pool is reduced to 320 mm2 of 30% of the SBWR PCC/IC pool. The water flow rate to the steam injector is about 3 kg/s which is determined by the static water head of 20 m from the PCC/IC pool to the SI and the SI water discharge pressure. When the water level in the PCC/IC pool exceeds the upper limit, the SI is stopped passively one by one in the analyses.

Table 5.6.6 Analysis conditions for confirming performance and effectiveness of SIPOWER during LOCA [9]. Case no.

Pool area (m2)

Initial water (m)

SIPOWER On/Off

Grace period (days)

Case 1

825 (100%) 825 (100%) 320 (30%)

4.4

Off

3.1

4.4

On

Over 30

4.4

On

Over 30

Case 2 Case 3

Table 5.6.7 Typical analytical conditions for SIPOWER [9]. CASE no.

Pool area (m2)

Initial water level (m)

SIPOWER On/Off

Grace period (days)

Case 1 Case 2 Case 3

825 (100%) 825 (100%) 320 (30%)

4.4 4.4 4.4

Off On On

3.1 Over 30 Over 30

526

Boiling Water Reactors

Fuel transport channel PCCS Pool

RPV

S/P Core

Fig. 5.6.45 PCC/IC pool volume and layout of original DOE SBWR [9,21].

POOL X 4.4m SIPOWER OFF

8

10

10

9

9

8

8

7

7 STEAM FLOW RATE (kg/S)

6

6

5

5

4

4 TOP OF HX TUBE

3

2

2 0 0

1

2

3

4

5

6

7

8

7

(b)

5

SI 1 UNIT PCC POOL LEVEL

4 3

3 TOP OF HX TUBE

2 0

TIME AFTER LOCA (days)

SI 2 UNITS

4

0 9 10 11 12 13 14 15

8 6

5

1

9 7

STEAM FLOW RATE (kg/S)

6

1

PCC POOL LEVEL

1

(a)

3

10

POOL 320m 2 X 4.4m SIPOWER ON

SI FLOW (kg/s) 0

1

2

3

4

5

6

7

8

2

FLOW RATE (kg/s)

825m 2

WATER LEVEL (m)

WATER LEVEL (m)

9

FLOW RATE (kg/s)

10

1

0 9 10 11 12 13 14 15

TIME AFTER LOCA (days)

Fig. 5.6.46 The effect of SIPWER on pool water level transient [9]. (a) Case 1, (b) Case 2.

The steam flowrate from the RPV to the PCC/IC and the PCC/IC pool water level transient are presented in Fig. 5.6.46. In Case 1, shortly after the initiation of the LOCA, the steam generated by the decay heat in the RPV flows into the PCC/IC. The inflow steam flow rate gradually decreases with time as the decay heat gradually decreases. After the initiation of the steam inflow, the water in the PCC/IC pool is heated, and then boiling starts and the water level in the poll gradually decreases with time. The water level drops to the top of the heat transfer tubes in the IC, which is the lower limit in the PCC/IC pool. The water level further decreases monotonously with time. The grace time in this case is 3 days. So, refilling of the pool is required from a safety point of view after 3 days. In Case 2, just after the initiation of the LOCA, steam flows into the PCC/IC, and after a short delay the water level in the pool starts to decrease because of boiling.

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However, the generated steam in the PCIS/IC pool flows to the SI. Two SIs are installed in the present case. The SIs initiate pumping up water in the water tank outside of the reactor building and deliver it to the PCC/IC pool in the reactor building as shown in Fig. 5.6.46b. After a while, the water level in the PCC/IC pool turns to recover the initial level. Then, one of two SIs is stopped by passive level switch. The remaining SI continues to refill the PCC/IC pool and the water level in the pool is maintained above the lower limit of the water level in the pool although the level gradually decreases. The grace time in this case is 30 days, which has been extended from 3 days by the introduction of the SIs. In Case 3 the volume of the PCC/IC pool is decreased to 30% of the original SBWR PCC/IC pool. The water level in the PCC/IC pool is maintained above the lower limit by SI refiling as in Case 2. Two SIs continues operation. The grace time in this case is 30 days, the same as in Case 2. The adoption of the SIPOWER in the SBWR results in a significant reduction in the volume of PCC/IC providing the grace time of 30 days.

5.6.8.3 Full-scale mock-up test to confirm feasibility of SIPOWER The feasibility and demonstration tests were conducted and water was successfully pumped up to the higher pool. The test program consisted of two steps: The feasibility of the SIPOWER was confirmed experimentally [9]. The confirmation tests were composed of two steps: (1) Fundamental performance tests under steady-state conditions. (2) Characteristic confirmation tests under unsteady conditions.

Fig. 5.6.47 shows the external view of the SIPOWER test facility. The facility has the height same as that of the SIPOWER of the SBWR and consists of large full-height

Fig. 5.6.47 SIPOWER test facility.

528

Boiling Water Reactors

water tank, full-height PCC/IC and the pool, one full-scale SI, and SI discharge line. The full-height water tank can simulate the water static-head from 3 to 20 m. About 110 kPa atmospheric pressure steam is sucked and supplied to the test SI. The full-scale test SI is the special order for the SI of the SIPOWER. The SI discharge line is designed to simulate the discharge head to deliver water into the PCC/IC pool [9]. Test parameters in the fundamental performance tests are presented in Table 5.6.8. The test conditions covered the actual working conditions of the SBWR SIPOWER system. It was confirmed in the tests that the SI worked sufficiently by low-pressure steam supply such as atmospheric pressure steam, and discharged water at 0.45 MPa. The discharge pressure could deliver water to the PCC/IC pool 35 m above the SI. The SI worked over an inlet water static head of 3.5 m from the water tank outside the reactor building. The SI could deliver water at a static head of 38 m or more when water from the water tank was at a static head of 10 m or more. When temperature of water from the water tank was below 30°C, SI discharge static head exceeded 30 m. From these results, it was confirmed that the SI for the SIPOWER works sufficiently under steady-state conditions. After the fundamental performance tests under steady-state conditions, characteristic confirmation tests under unsteady conditions were conducted. In the tests, SI inlet water pressure and temperature from the water tank outside the reactor building, and the back pressure of the SI outlet were varied. One of test results are presented in Fig. 5.6.48. Steam is initially at 0.10 MPa and inlet water static head from the water tank is at 18 m. The SI is in the operating state at Table 5.6.8 Test conditions in fundamental performance tests under steady state conditions [9]. Test parameter

Test range

Steam pressure (kPa) SI inlet water head (m) IC pool water temperature (°C)

100–112 3.5–20 20–40

DISCHARGE WATER HEAD

30

0.50 0.40

FEEDWATER HEAD

20

0.30 0.20

10 STEAM PRESS. 0

0.10 0

0

10

20

30

40 50 TIME (SEC)

Fig. 5.6.48 SIPOWER feasibility and demonstration test result.

60

STEAM PRESSURE(MPa)

WATER HEAD (m)

40

BWR innovations

529

around 6 s. Static head is imposed on the SI water inlet. Just after starting, steam pressure becomes slightly below 0.10 MPa of atmospheric pressure because of steam condensation in the mixing nozzle in the SI since the part is cold. After the part is warmed by condensation, pressure returns to 0.10 MPa. SI discharge pressure jumps up to 0.32 MPa, and then gradually rise to 0.45 MPa. These results imply that the SI in the SIPOWER system will be fully operational in 28 s. The start-up time is quite short, and the operating condition is very stable.

5.6.8.4 Air-purge analysis in PCC/IC pool for SIPOWER By reducing the PCCS pool volume, SIPOWER has advantages of promotion of earthquake proof and reducing costs. However, SI performance is affected greatly by noncondensable gas. Noncondensable gas degrades condensation heat transfer on the surface of the water jet as well as causes operation instability by mixing into the water jet. In order to evaluate this system, it is important to predict concentration of noncondensable gas in the PCCS pool. After LOCA occurrence, concentration of noncondensable gas, that is, air which is filled generally in the PCCS pool, changes by purging due to the generation of large amount of steam. Therefore, air purge phenomena after LOCA was analyzed by using CFD code considering mixing of two kinds of gas, and the applicability of innovative simplified safety system by utilization of SI was evaluated from the perspective of concentration of noncondensable gas in the PCCS pool [20]. (a) PCC/IC system configuration: The conceptual figures of both types of PCC/IC system are shown in Fig. 5.6.49. Both types consist of steam line connected to the RPV or PCV, heat exchanger tubes, drain line

PCCS Pool

Exhaust line

Exhaust line Steam line

PCCS Pool

Steam line

Drain line

Drain line

Fig. 5.6.49 PCC/IC pool volume comparison between vertical and horizontal heat exchanger [20].

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Boiling Water Reactors

Fig. 5.6.50 PCC pool model of CFD [20].

Flow-out boundary (Exhaust pipe)

Y Z

X Flow-in boundary (Steam generation)

and vent line connected to the suppression pool. Assume that the heat exchanger has a cylindrical outer dimension, the installation height will be lower if it is installed horizontally. Pool volume is reduced to 1/3 compared to vertical installation. In actual design, when the PCCS heat exchanger is installed horizontally, the pool volume above the heat exchanger is about 800 m3, which is about 1/3 of that of the vertical type. Fig. 5.6.50 shows the mesh model for horizontal-type heat exchanger of PCCS condition. As shown in Fig. 5.6.51, the analysis results show that the air in pool is completely purged in 2600 s for a vertical heat exchanger tube and 1800 s for a horizontal tube, as shown in Fig. 5.6.51b. In other words, if the SI start SIPOWER after 30 min, SIPOWER will almost certainly start. Steam should flow through the SI in advance, and the pool’s passive float-type level switch should be used to detect a drop in water level and start supplying water to the SI. It was started dozens of times, but we have not experienced nonoperation.

5.6.8.5 Summary of SIPOWER The SI is a simple, compact passive pump. It could operate without AC power, a large motor, or a turbo-machinery. The SI can operate under the condition of very low steam pressure, such as atmospheric pressure, and it discharges water at 0.5 MPa (40 m water head above the SI). The SIPOWER for the PCCS makes use of the excellent characteristics of the SIS under a very low steam pressure and will rationalize reactor building layout by reducing the area and volume of the PCC/IC pool to 39%. Major points to be stressed are as follows: (1) By using the full-scale steam injector system, SIPOWER feasibility has been confirmed. (2) The SI started up easily and discharged water with a pressure of 30 m or higher by atmospheric steam and 10 mAq water supply. (3) By using the SIPOWER, PCC/IC pool area will be reduced to about 40% and the 30 days grace period time will be attained. (4) Reactor building volume can be greatly reduced by the SIPOWER.

BWR innovations

(a)

531

1.0

(b)

0.9

0.8

0.8

0.7

0.7

0.6

0.6

0.5

0.5

0.4

0.4

0.3

0.3

0.2

0.2

0.1

0.1

0.0

(a) t=30s

1.0

0.9

1.0

0.0

(a) t=30s

0.9

1.0

0.8 0.7 0.6

0.9

0.5

0.4 0.3

0.8

0.2 0.1 0.0

(b) t=120s

0.7

(b) t=120s 1.0

1.0

0.9

0.95 0.8

0.9

0.7

(c) t=180s

(d) t=2700s

1.0

(c) t=180s

1.0

0.95

0.95

0.9

0.9

(d) t=1800s

Fig. 5.6.51 Steam mass ratio in (a) vertical-type PCCS [20] and (b) horizontal-type PCCS [20].

References [1] N. Isshiki, Heat Transfer Engineering, Fundamental Mechanical Engineering Complete Book, Vol. 9, Morikita Publishing, 1967, 1973 (in Japanese). [2] K. Hannerz, Making progress on PIUS design and verification [advanced LWRs], Nucl. Eng. Int. 33 (412) (1988) 29–31. [3] T. Narabayashi, C. Iwaki, H. Nei, Thermal-hydraulics study on steam injector for nextgeneration reactor, in: Proceedings of International Conference on New Trends in Nuclear Systems Thermal-hydraulics, Pisa, Italy, vol. 1, 1994, pp. 653–661. [4] T. Narabayashi, W. Mizumachi, M. Mori, Study on two-phase flow dynamics in steam injectors, Nucl. Eng. Des. 175 (1997) 147–156.

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[5] T. Narabayashi, T. Ishiyama, H. Miyano, H. Nei, A. Shioiri, Feasibility and application on steam injector for next-generation reactor, in: Proc. of Int. Conf. on Nucl. Eng, (ICONE-1), 1991, pp. 23–28. [6] T. Narabayashi, H. Miyano, H. Nei, O. Ozaki, W. Mizumachi, A. Shioiri, Feasibility study on steam injector driven jet pump for next-generation reactor, in: ‘92 International conference on design and safety of advanced nuclear power plants, vol. 4, 1992. pp. 36.2/ 1 /7. [7] G. Cattadori, L. Galbiati, L. Mazzocchi, P. Vanini, A single-stage high-pressure steam injector for the ALWR, Trans. Am. Nucl. Soc. 69 (1993) 549–550. [8] T. Narabayashi, C. Iwaki, Y. Kudo, W. Mizumachi, A. Shioiri, T. Tanaka, M. Mori, A. Horie, Y. Machida, Analytical study on large scale steam injector to next-generation BWR, in: Proceedings of the 3rd International Conference on Nuclear Engineering (ICONE-3), 1994, pp. 877–883. [9] T. Narabayashi, C. Iwaki, W. Mizumachi, A. Shioiri, Feasibility study on steam injector driven system for SBWR PCC/IC pool water refill, in: International Topical Meeting of Advanced Reactors Safety (ARS’94), 1994, pp. 1102–1109. [10] C. Iwaki, T. Narabayashi, Study on heat and momentum transfer mechanisms in steam injectors for next-generation reactors, Proc. Winter Ann. Mtg. ASME HTD 294 (1994) 19–30. [11] C. Iwaki, T. Narabayashi, H. Nei, Experimental study on heat and momentum transfer mechanisms in steam injectors for next generation reactors, in: 2nd International Conference on Multiphase Flow, 4, 1995, pp. 262–268. [12] T. Narabayashi, H. Nei, O. Ozaki, A. Shioiri, W. Mizumachi, Study on high performance of steam injectors (1st report: Formulation of operating mechanism and characteristic analysis TOSIA Code), JSME J. B 62 (597) (1996) 155–162 (In Japanese). [13] T. Narabayashi, M. Mori, M. Nakamaru, S. Ohmori, Study on two-phase flow dynamics in steam injectors II. High-pressure tests using scale-models, Nucl. Eng. Des. 200 (2000) 261–271. [14] PHOENICS (Parabolic Hyperbolic or Elliptic Numerical Integration Code Series) Overview, CHAM Technical Report: TR 001, http://www.cham.co.uk/phoenics/d_polis/d_ docs/tr001/tr001.htm#what. [15] S. Ohmori, M. Mori, T. Narabayashi, M. Nakamaru, Y. Asanuma, M. Yasuoka, Development of steam injector feedwater heater system, in: Proceedings of International Conference on Nuclear Engineering, ICONE-8582, Baltimore, USA, 2000. [16] M. Mori, S. Ohmori, T. Narabayashi, Development of simplified steam injector feedwater system-large scale model tests and design improvement by CFD, in: Proceedings of International Conference on Nuclear Engineering, ICONE11-36488, Tokyo, Japan, 2003. [17] M. Mori, T. Narabayashi, S. Ohmori, Research and development program of innovative simplified and severe-accident-free BWR by high-performance steam injector system, in: Proc. of ICAPP ‘03, Cordoba, Spain, Paper 3294, 2003. [18] S. Ohmori, T. Narabayashi, K. Okamoto, C. Iwaki, M. Mori, Development of nuclear power plant by using Innovative simplified high-performance steam injector technology (numerical flow analysis and visualization measurement), JSME J. B 74 (5742) (2008) 1287–1296 (In Japanese). [19] T. Narabayashi, Y. Asanuma, C. Iwaki, M. Mori, S. Ohori, Development of multi-stage steam injector for feedwater heaters in simplified nuclear power plant, JSME Int. B, 49 (2) (2006).

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[20] C. Iwaki, T. Narabayashi, S. Ohmori, M. Mori, Development of simplified nuclear power plant using high-efficiency steam injectors-(4) air-purge analysis for water makeup system, in: ICONE13-50625, 2005. [21] J.D. Duncan, SBWR, a simplified boiling water reactor, Nucl. Eng. Des. 109 (1–2) (September–October 1988) 73–77. https://www.sciencedirect.com/science/article/abs/ pii/0029549388901434#!. [22] T. Narabayashi, S. Ohmori, M. Nakamaru, M. Mori, Y. Asanuma, Development of steam injector for next-generation reactors, S03/2, in: Transactions of the 15th International Conference, on Structural Mechanics in Reactor Technology, (SMiRT-15), Seoul, Korea, 1999.

5.7

Built in upper internal control rod drives (CRDs) for ABWR-III Tadashi Narabayashia, Chikako Iwakib, and Michitsugu Moric a

Tokyo Institute of Technology, Meguro, Tokyo, Japan, bToshiba Energy Systems & Solutions, Corp., Yokohama, Kanagawa, Japan, cGraduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan The development program of the internal CRD system with heatproof ceramic coils for a next-generation BWR [1] is named “ABWR-III” project in this book. Evaluations and several tests and analysis have been completed by one of the selected offers for the technical developments of the Institute of Applied Energy founded by METI (Ministry of Economy, Trade and Industry) of Japan [2,3]. Development of built-in upper internal CRD for ABWR-III, including structural integrity & LOCA analysis were conducted as a joint study between TOSHIBA and TEPCO and the University of Tokyo.

5.7.1

Introduction of merits and technical tasks for internal CRD

An internal CRD using a heatproof ceramic-insulated coil was developed also for marine reactor [4,5]. The built-in upper internal CRD for ABWR-III had a competitive and higher performance compared with ABWR (see Chapter 2). In case of a power output of 1700MWe ABWR-III, the internal CRDs are installed in the RPV whose size is equivalent to the 1356-MWe ABWR, and no space is required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a dry well. This brings

534

Boiling Water Reactors

further advantages of elimination of reactivity-induced accidents (RIA) caused by CR withdrawing under the pressure boundary broken, and easy in-vessel retention (IVR) by vessel bottom cooling in case of a severe accident. In order to develop the 1700-MWe ABWR-III as a next-generation BWR, many new advanced technologies were examined, such as (1) heatproof ceramic-insulated coil’s technologies, (2) RIP tube internal structures without shroud, and (3) computed fluid dynamics (CFD) technologies [2,3,6–11]. The internal CRD consists of (1) a heat-resistant motor for normal adjustment of CR position, (2) heat-resistant solenoid drive latch mechanism for gravity-driven scrum operation, and (3) an electromagnetic power coupling for signal and power transmission from outside of RPV. These new devices need heat-resistant magnet coils used in high-pressure and high-temperature coolant in RPV. Therefore, the technical development for the internal CRD is performed focusing on the ceramic-insulated coil at approximately 600°C or more. The technical development items are as follows: (1) Development of heat-resistant motor and driving mechanism and latch magnet for gravitydriven scram. (2) Development of the ceramic-insulated heat- and radiation-resistant coil. (3) Durability test of a ball bearing for the internal CRD in the high-pressure and hightemperature reactor coolant under BWR condition. (4) Evaluation of structural integrity and flow instability due to two-phase flow of core exit condition.

5.7.2

Plant concepts of ABWR-III

The next-generation 1700-MWe ABWR-III, the internal CRDs are installed in a RPV whose size is equivalent to the 1356-MWe ABWR, as shown in Figs. 5.7.1 and 5.7.2. Major components in the RPV are the double coaxial dryer, low pressure-drop separators, the internal CRD, the magnet coupling power connector, guide chimneys, core fuels installed in shroud-less core structure, and eight large RIPs with the RIP-tube core flow meter. Major specifications are shown in Table 5.7.1. As shown in Fig. 5.7.3, there is a space of 10 m under the RPV of the current ABWR. The height is needed to withdraw the CRD from the bottom of the RPV. The internal CRDs eliminate penetration via a bottom flange of RPV and enable to lower installation level of RPV in a dry well, as shown in Fig. 5.7.4. These advantages realize a compact PCV and reduced volume of a reactor building. Further advantages are elimination of RIA caused by control rod (CR) withdrawing under pressure boundary broken, and easy IVR with a small volume of water for vessel bottom cooling in case of a severe accidents, as shown in Fig. 5.7.5.

Double co-axial dryer Low pressure-drop separators

Internal CRD Magnet-coupling power connector CRD guide Chimney

RIP tude flow meter Shroud-less core

Eight large RIP

1.5m

3.7m

7.1m

4.4m

3.7m

2.3m 21m

3.3m

5.8m

Fig. 5.7.1 3D-CAD view of the next generation BWR.

5.8m

Fig. 5.7.2 Dimensions for major internals.

536

Boiling Water Reactors

Table 5.7.1 Major specification.

Electrical output (MWe) Thermal output (MWth) Main steam flow (103 kg/s) RPV inner diameter (m) Type of core Number of fuel bundles Number of CRDs Number of separators Dryer flow area (m2) Rated core flow (104 kg/s) Number of RIP

Current ABWR

ABWR-III

1356 3926 2.11 7.11 C-lattice 872 205 349 58 1.45 10

1700 4922 2.64 7.11 C-lattice 988 257 218 78 1.57 8 (1.25 x Diameter)

10m 10m CRD

Fig. 5.7.3 Space under the RPV of the current in a PCV.

As mentioned above, the internal CRDs have great advantages of elimination of RIA caused by CR withdrawing under the CRD pressure boundary broken, enlarge the water inventory above the core, and easy in-vessel retention (IVR) by vessel bottom cooling in case of a severe accident.

5.7.3

Power devices for the internal CRD

5.7.3.1 Magnet coupling power connector In order to realize the internal CRD system, electric power must be sent via RPV pressure boundary and some kind of power connectors should be used for the annual maintenances of the CRDs and fuel shuffling in a core. Therefore, the magnet coupling power connector was developed.

BWR innovations

537

10m

ABWR's RPV Next-Generation BWR's RPV Main steam line Suppression pool

Fig. 5.7.4 Lower installation level of RPV ABWR.

NWL NWL

TAF

TAF

HCU IVR Cooler

IVR Cooler (b)

(a) Fig. 5.7.5 Simplification of bottom of the RPV. (a) Current ABWR, (b) ABWR-III.

538

Boiling Water Reactors

Gap Iron core

Iron core Secondary coil

Primary coil

Magnet flux

Pressure boundary

Fig. 5.7.6 A principle of the magnet coupling power connector.

Table 5.7.2 Power loss fractions in the magnet coupling power connector. Position

Coil (%)

Can (%)

Subtotal (%)

Primary Secondary Total

56.9 2.1 59.0

22.1 18.9 41.0

79.0 21.0 100.0

Fig. 5.7.6 shows the principle of the magnet coupling power connector. There is a pressure boundary and induction-eddy current loss in the pressure can as shown in Table 5.7.2 and Fig. 5.7.7. A mock-up test section of the magnet coupling power connector for hightemperature test is shown in Fig. 5.7.8. Fig. 5.7.9 shows the results of magnetic-field analysis for (a) maximum secondary current and (b) zero secondary current. Fig. 5.7.10 shows an FEM heat transfer analysis result of temperature distributions in the test section under the rated power transfer condition of 50 W. The surface temperature of water-shield cans was supposed to be 300°C. The maximum temperature was approximately 360°C in the primary coil, and was low enough to the allowable maximum temperature of 600°C. As shown in Fig. 5.7.11, power transfer efficiency of 33%, under the condition of 110 V, 260 W three-phase power conditions, was obtained by using a test apparatus shown in Fig. 5.7.12. When the temperature rises to about 300°C, the electrical resistance of the coil rises and copper loss increases. These measured and analyzed data were used to improve the magnet coupling power connector. The estimated power transfer efficiency at high temperature is about 20%. The rated internal CRD’s power is approximately 50 W. Therefore, the primary power of the

Fig. 5.7.7 The schematic concept of magnet coupling power connector installed in an RPV.

Internal CRD

Power pipe

Power pipes nozzle

Magnetcoupling power connector Primary Secondary One unit

for signal for power

Fig. 5.7.8 Test section of a magnet coupling power connector for high-temperature test.

Fig. 5.7.9 Results of magneticfield analysis. (a) Max secondary current, (b) zero secondary current.

(a)

(b)

70

350

60

300

50

250

40

200

30

150

20

100

10

50

0 0

20

40

60

80

100

120

3-Phase output power (W)

Power transfer efficiency (%)

Fig. 5.7.10 FEM heat transfer analysis result of temperature distributions (rated power conditions).

0 140

Input voltage (V)

Fig. 5.7.11 Three-phase power transfer test result. Power Nozzle via RPV

Heat-Proof Motor

For Signal

Power Piping

Output signal (V)

For Motor

0

0.1

0.2

0.3

Time (sec)

0.4

0.5

Control Rod

Fig. 5.7.12 Magnet coupling integral-test apparatus for three-phase power transfer and CR signal transfer.

BWR innovations

541

each CRD will be approximately 250 W and the total power of 257 CRDs will be only 64 kW. The power is very low compared with the CRD pump of the current ABWR.

5.7.3.2 Magnet coupling signal connector As shown in Fig. 5.7.12, signal transfer test via a magnet coupling signal connecter was also conducted. Diagram of the magnet-coupling signal connector test circuit and signal monitor are shown in Fig. 5.7.13. There were two lead switches driven by magnets on the CRD shaft, one was for making rotation signal and the other was for making an initial position signal at the full insertion of a control rod. As shown in Fig. 5.7.14, the measured signal transfer efficiency of the magnet-couple signal connector was very good under signal frequency of 25–200 Hz. The signal could be transferred even at 1 kHz via two stainless steel cans of several mm thickness and a water

Magnet-Coupling Signal Connector

Shaft

Magnet Lead Switch 1 Rotation

Signal Circuit

Signal Monitor

Initial Position Lead Switch 2

Fig. 5.7.13 Diagram of the magnet-coupling signal connector test circuit and signal monitor.

Voltage Transfer Efficiency (%)

60 50 40 30 20 10 0

0

100

200

300

400

500 600

700

Signal Frequency (Hz) Fig. 5.7.14 Measured signal transfer efficiency vs frequency.

800

900 1000

542

Boiling Water Reactors

Fig. 5.7.15 Measured signal via the coupling.

Output Signal (V)

Motor Rotation 4 3 2 1 0 -1 -2 -3 -4

0

0.1

0.2

SW1: ON SW1: OFF

0.3

0.4

0.5

Time (sec)

gap of 5 mm. Fig. 5.7.15 shows the switching signal of on/off on a monitor scope during the shaft rotation.

5.7.4

Internal CRD’s mechanism

5.7.4.1 Latch mechanism for scram operation and lift a control rod Fig. 5.7.16 shows the schematic drawing of the internal CRD. The internal CRD consists of a latch mechanism and the rotation mechanism driven by the heatproof motor. The latch mechanism consists of a combination of ball-nut and a holder sleeve lifted with a lift coil through a linear magnet coupling. The heatproof motor through a radial magnet coupling also drives the combination ball-nut. Scram operation is attained by only turning power off in the lift coil. Then, the holder sleeve is dropped and the combination ball-nut is divided into three pieces so that a ball screw shaft that lifts a control rod (CR) is dropped by the gravity force, as shown in Fig. 5.7.17a. When the lift coil power is turned on, the holder sleeve is lifted upward and the three divided ball-nut pieces are combined into one ball-nut as shown in Fig. 5.7.17b. After this action, the heatproof motor drives the ball screw shaft and the CR is moved upward and downward smoothly. As shown in Figs. 5.7.18 and 5.7.19, a trial device of the latch mechanism was made. It was confirmed that the device showed very smooth operation of latch and scram by using the dummy weight. Based on the above examination, mock-up CRD was made to confirm the latch mechanism and heatproof motor drive action, as shown in Fig. 5.7.20. The maintenance is easy for overhaul of casing and latch mechanism as shown Fig. 5.7.21.

5.7.4.2 Development of heatproof motor The internal CRD has a heatproof motor inside a CRD casing to drive a control rod with fine motion control. The type of the motor is a three-phase induction type as shown in Fig. 5.7.25. A planetary gear is used to slowdown the rotational speed and drive the ball-nuts by using a pair of samarium-cobalt magnet via pressure

BWR innovations

543

240mmOD Lift coil

Linear magnet coupling Latch mechanism

Radial magnet coupling Planet gear Heatproof motor

Fig. 5.7.16 Schematic representation of the internal CRD.

(a)

(b)

Fig. 5.7.17 Latch mechanism of the internal CRD. (a) Scram, (b) latch to lift CR.

boundary. The magnet is very strong and heat/radiation proof. The motor’s power is three phases AC power of 50 Hz. This frequency is suitable for supplying the AC power via the magnet coupling power connector. Three-dimensional magnetic-field analysis was conducted as shown in Figs.5.7.22 and 5.7.23. The maximum magnetic field was approximately 1.5 T at the stator tee and rotor yoke. The value is low enough to suppress the motor loss. Especially, the magnetic field of the stator yoke is approximately 1.0 T. That contributes to the low-power loss and it was concluded the motor is

544

Boiling Water Reactors

Fig. 5.7.18 3D-CAD for checking the mechanisms. (a) Latch mechanism, (b) heatproof motor.

Fig. 5.7.19 Trial device of the latch mechanism.

BWR innovations

545

Fig. 5.7.20 Developed mockup of CRD. (a) 3D-CAD, (b) casing, (c) latch mechanism.

Fig. 5.7.21 Structure for overhaul of casing and latch mechanism for easy maintenance.

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Boiling Water Reactors

Fig. 5.7.22 A cross-sectional view of a heatproof motor.

Fig. 5.7.23 3D magnetic-field analysis.

feasible. As shown in Fig. 5.7.24, ball bearing high-temperature durability test was conducted at 300°C. After conducting a heat resistance test for 6 h in a constant temperature chamber at 350°C, as shown in Fig. 5.7.25, it was operated with a motor characteristic test device, and it was confirmed that the initial performance was obtained.

5.7.4.3 Ceramics coil radiation durability test Many kinds of ceramic insulators such as Al2O3 or Al2O3-SiO2 were examined. Fig. 5.7.26 shows some of the examples of the coaxial structure of a ceramic-insulated heatproof coil. To prevent the oxidation of copper, the copper is covered with stainless steel cladding. The surface of the stainless steel clad is coated with inorganic polymer and ceramic insulator, as shown in Fig. 5.7.27a.

BWR innovations

547

Fig. 5.7.24 Bearing high-temperature durability test at 300°C.

Fig. 5.7.25 A proto-type proof motor assemblage, heatproof test, and performance test. (a) Rotor assembly with cylindrical shaft, (b) stator assembly, (c) heatproof test and performance test was OK. Copper Stainless Steel Inorganic Polymer Ceramic Insulator

Fig. 5.7.26 Coaxial structure of a ceramic-insulated coil.

548

Boiling Water Reactors

Fig. 5.7.27 Neutron radiation data for heatproof coil in the Yayoi reactor. (a) Radiation test coil, (b) stater coil for motor.

A heat-cycle test from room temperature to 800°C were done over 10 times for each coil specimens and showed good robustness. Fig. 5.7.27b shows the ceramicinsulated stator coils for heatproof motor. The γ-ray radiation test was also conducted by using a high-temperature chamber, installed at Professor Terai’s Lab. in the University of Tokyo, as shown in Fig. 5.7.28. The test specimen coils showed good insulator resistance even under 800°C. The chamber can heat coils with a ceramics bobbin up to 1000°C. The control circuit of the constant temperature chamber and the thyristor were taken out of the γ -ray irradiation chamber to carry out a high-temperature heating test. This is because the thyristor is installed on the bottom area in the constant temperature chamber (Fig. 5.7.29), it cannot be heated immediately soon after the start of high γ-ray irradiation. 1E+11 1E+10 1E+9 1E+8 1E+7 1E+6

0

200

400

600

800

1000

Fig. 5.7.28 The γ-ray radiation test was also conducted by using a high-temperature chamber.

BWR innovations

549

Fig. 5.7.29 High-temperature chamber.

Fig. 5.7.30 Mock-up radiation hole of Yayoi reactor.

Fig. 5.7.30 shows the mock-up of radiation hole of Yayoi reactor and Fig. 5.7.32 shows the test specimen coil for radiation, respectively. Fig. 5.7.31 shows the test result of the neutron radiation data in the Yayoi reactor in the University of Tokyo. The resistance of ceramics insulator kept 1 GΩ under 2000 W of the Yayoi reactor’s thermal output.

550

Boiling Water Reactors

Fig. 5.7.31 Mock-up of Yayoi reactor radiation hole.

Fig. 5.7.32 Radiation test coil.

5.7.4.4 Neutron flux at the internal CRD In order to estimate the neutron flux at the position of the internal CRD in an actual BWR, two-dimensional neutron transport code DOT3.5 was used, as shown in Fig. 5.7.33. The analysis result showed each energy level of neutron as follows: Fast Neutron Flux:106 n/s/cm2. Middle Neutron Flux: 106 n/s/cm2. Thermal Neutron Flux: 106 n/s/cm2.

BWR innovations

551

Shield RPV Concrete 15.2

A A B B C E C D D G E F F G H H I K J J L K M

\

14

Separator 12

Shroud head

Internal CRD

M K L

N O P

8

Height (m)

Guide Chimney

10

J K

Q Q

R

N

O

S

T

6

N N

N

U

L

R P

M

V

4 Core

M N

2 0

S Q P O U S RT Q P O NL N

0

V R P O O Q

2

4

Radius (m) Radius (m)

Contour Levels (n/s/cm2) A: 1E-08 B: 1E-07 C: 1E-06 D: 1E-05 E: 1E-04 F: 1E-03 G: 1E-02 H: 1E-01 I: 1E+00 J: 1E+01 K: 1E+02 L: 1E+03 M: 1E+04 N: 1E+05 O: 1E+06 P: 1E+07 Q: 1E+08 R: 1E+09 S: 1E+10 T: 1E+11 U: 1E+12 V: 1E+13

5.5

Fig. 5.7.33 The analysis result of neutron flux distribution in a ABWR-III reactor, by using the 2D neutron transport analysis code DOT3.5.

The measured neutron flux doze was almost the same as the position of the internal CRD in a designed ABWR-III. The flux level was not so high, and the radiation test at the Yayoi reactor was valid.

5.7.5

Evaluation of ABWR-III conditions

5.7.5.1 Durability test of ball bearing In order to evaluate the robustness of the internal CRD, a durability test was conducted at the R&D center of TEPCO. The test using high-pressure and high-temperature water with crud was done and showed good robustness for even 60 years equivalent rotation test by Goto et al., ICONE13 [11].

552

Boiling Water Reactors

The water quality in the nuclear reactor is one of the factors that influence the corrosion resistance of the material. In order to decide the test condition, environment of water quality for the BWR was estimated. An analysis for the BWR plant influenced by quality of water was done using a model of the primary coolant system. This system has the modeling that divided some areas in the primary coolant section. The parameters, such as the amount of radiation, the flow velocity of the coolant, the time to stay within same area, etc., are inputted at each area. This analysis is done along the flow path. Especially, the parameters the distribution of void fraction and the coefficient of bubble-water shift for hydrogen and oxygen are considered because of two-phase flow with boiling in the reactor core. It was used in the area which was divided with flow path for jet pump-type BWR. The most of hydrogen and oxygen are generated by the radiation of the reactor core. It shifts to the vapor phase with boiling, and the part of vapor is dissolved to the coolant in the reactor of the BWR plant. Therefore, it is important for the evaluation of the shift speed from the liquid phase of hydrogen and oxygen to the vapor phase at treating the two-phase flow. This constant of shift speed deviates from the Henry constant that gives a rate of distributed gas–liquid in a static state, and it is usually handled as a parameter in the evaluation of the model. The value is set to be suitable for the calculation. The measured hydrogen and the density of oxygen in the plant the same amount of the main steam are used. Fig. 5.7.34 shows an example of analysis of the oxygen distribution in the primary coolant of the BWR-4-type plant [5]. In this case the density of hydrogen is 0.8 ppm, the density of the reactor coolant dissolved oxygen is in the range 20–50 ppb at the top part of the fuel in the reactor. It seems low dissolved oxygen comparatively. It seems that the examinations should be carried out at the real condition of a BWR, high pressure and high temperature, with an environment of water quality referring to the existing research case, etc. In order to evaluate the endurance and robustness of the bearing for internal CRD, an experimental apparatus was designed and placed at the R&D Center of Tokyo Electric Power Company (TEPCO). The test specimens of the ball bearing and pin/roller were supplied by TOSHIBA. The test under high pressure and high temperature was started in the fiscal year 2003 [6–11].

Fig. 5.7.34 The analysis result of the density of dissolved oxygen (0.8 ppm in BWR/4).

Dissolved Oxygen 100 ppb > 50 ~100ppb

20 ~ 50ppb 0 ~ 20ppb

BWR innovations

553

Based on the environmental investigation (neutron and water quality), the experimental condition was set as same high pressure (7 MPa) and high temperature (286°C) of the current BWR. In addition, the author added iron cruds into water to confirm the robustness of the ball bearing in high-density crud conditions. Three kinds of test bearing were selected under the high-pressure and hightemperature water. (a) Test Materials for the test bearing: Three kinds of test bearing were selected under high-pressure and high-temperature water referring to the results of Nunokawa et al. [5], and using results of ABWR. Table 5.7.3 shows the kind and the materials of the test piece. (b) Types of ball bearing for the test: The test piece of the ball bearing is a thrust type of rolling bearing, and it is composed with six balls that are separated with the retainer at equal intervals between a rotating ring and a flat ring. Fig. 5.7.35 shows the outline shape, and Fig. 5.7.37 shows the photograph of the typical externals of the ball bearing test piece. The test piece of the roller/pin bearing is used in the FMCRD of ABWR, and it is composed of three roller/pins that are maintained in the holder ring. Fig. 5.7.36 shows the outline shape, and Fig. 5.7.38 shows the photograph of the typical externals of the roller/pin bearing test piece. (c) Test condition: Based on the environmental investigation (neutron and water quality) and compared with current BWR, the experimental conditions were selected as follows: (1) Pressure and temperature: 7 MPa and 286°C, these are same as the current BWR.

Table 5.7.3 Type and materials of test pieces. Type

Ball/roller material

Retainer/pin material

(1) Ball bearing (2) Roller/pin bearing type A (3) Roller/pin bearing type B

SUS440C Stellite #3 Ni alloy

Graphite Alloy #25 Alloy #25 + CrN coating

φ52mm

Rotating Ring

35mm

Retainer

Flat Ring

Ball

Fig. 5.7.35 Outline shape of the ball bearing test piece.

554

Boiling Water Reactors

φ52mm

Pin Roller

Rotating Ring

35mm

Holder Ring

Fig. 5.7.36 Outline shape of the roller/pin bearing test piece.

Fig. 5.7.37 Photograph of the ball bearing test piece.

Fig. 5.7.38 Photograph of the roller/pin test piece.

BWR innovations

555

(2) Load weight of bearing test piece: The total weight of the internal CRD with latch mechanism, screw shaft, and control rod is assumed to be about 1600 N. When the bearing of 170 mm in the outside diameter is composed of 20 balls, the load that supports each ball is 80 N/piece. Therefore, it is assumed that the maximum weight is 80 N for each ball or one roller/pin. The total load of the ball bearing test piece is 480 N because it is composed of six balls, and one of the roller/pin bearing test piece is 240 N because it is composed of 3 roller/pins. (3) Rotational speed: The rotational speed of internal CRD is assumed to be about 150 rpm. The rotational speed in this examination is set to be 560 rpm because the shape ratio of outside diameter is 150 and 40 mm. (4) Rotational numbers: The total rotational numbers at 60-year operation of internal CRD is assumed to be 2.7  106 rotational numbers. Therefore, it is set at 2.7  105 rotational numbers of the 1/10 in life time (total distance ¼ 31 km) at the screening examination that evaluates the material characteristic of the test piece, and 3.0  106 rotational numbers (total distance ¼ 339 km) at the endurance and robustness tests in consideration of the design margin. (5) Iron cruds: Cladding powder was added to the test water to check the robustness of the bearings. Two types of iron oxide were selected as cladding, and the total amount was 25 g and 50 g. This amount is a rather stringent condition with a very high density compared to the clad density of the primary reactor water in a current BWR. (d) Endurance and robustness test facility and test procedure: – Evaluation items: The evaluation items of the high-pressure and high-temperature water examination are as follows: (1) Evaluation of wear loss in weight: The weight and the size of the bearing for the test were measured as changes of the test piece before and after the examination. (2) Observation of the sliding side: A roughness of the of the sliding surfaces of the bearings for the tests were observed by scanning electron microscope (SEM), etc. The experimental apparatus was designed and produced taking into consideration the shape of the bearing test piece and the test condition. The outline of the flow diagram of the experimental apparatus is shown in Fig. 5.7.39 and the external photograph of it under the bearing durability test apparatus is shown in Fig. 5.7.40. – Screening test: The test piece of the ball bearing and two kinds of roller/pin bearing, types A and B, were installed in the experimental apparatus and screening tests were exanimated. Table 5.7.4 shows the examination cases. The experimental data depended on the rotational numbers were measured because the amount of wear loss for roller/pin type bearing test pieces were larger than that for ball bearing test piece. Fig. 5.7.41 shows the appearance of examination test result. (e) Endurance and Robustness tests: The endurance and robustness tests were carried out following the result of screening test. For ball bearing test pieces, we adopted a material that showed less wear in a screening test. Fig. 5.7.42 shows the external appearance photos before and after the ball bearing test.

556

Boiling Water Reactors

TrA 1

SV-1

PIA 1

SI 1

SX 1

M 1

Magnet Rotator

SE-1

AG-1

Magnet Rotator Shaft

PG

1

PE-1

HV-4

TICA 1 TICA 1A

Test Bearing

CW FSW 101

TE-1

Upper Heater

TE-1A SCR 1A

TICA 1B

BV-1

Cooling Water Tank

Middle Heater

TE-1B SCR 1B

UC-1

TICA 1C

Lower Heater

TE-1C

Recirculation Line

SCR 1C

Driving Shaft Pressure Vessel

PRV-1

F-1

Load Weight HV-3 HV-2

P-1

HV-1

Pressurize Pump

T-1

Fig. 5.7.39 Flow diagram of the experimental apparatus.

(f) Screening test results: Fig. 5.7.43 shows the appearances of specimen after screening test performed under the condition of 286°C high-temperature water. All of the specimens revealed smooth movement even after the test and revealed no sign of galling. The values of weight change normalized by the applied load during test were plotted in Fig. 5.7.44. As it can be seen from the figure, the weight loss of ball bearing is about 10 times smaller than that of other roller/ pin bearing. The weight loss for the combination of CrN-coating pin and Ni alloy roller bearing is about 1/2 compared with the combination of Co-based alloy (alloy #25 pin and stellite #3 roller) bearing. The ball bearing showed higher wear resistance than other roller/pin bearing, therefore the ball bearing was exanimated for the endurance test to confirm the durability of the internal CRD, and to evaluate the wear loss by the influence of the different load weight. (g) Endurance and robustness test result: The specimen could be removed smoothly even after the endurance and robustness tests and no sign of galling was observed. The experimental results showed that the ball bearing

BWR innovations

557

Fig. 5.7.40 Bearing durability test apparatus.

Table 5.7.4 Examination test cases. Rotational numbers (2)

Load weight (N)

2.7  105 3.0  l06 2.7  l05

480 480 240

3.0  l06

240

3.0  l06

480

3.0  l06

480

Roller/pin bearing type A

2.7  103 1.76  105

240 240

Roller/pin bearing type B

2.7  103 6.75  104

240 240

1.35  105

240

Bearing type Ball bearing

Examination case Screening test Endurance test Screening test (load dependence) Endurance test (load dependence) Robustness test (iron cruds 25 g) Robustness test (iron cruds 50 g) Screening test Screening test (rotational numbers dependence) Screening test Screening test (rotational numbers dependence) Screening test (rotational numbers dependence)

Fig. 5.7.41 Appearance of examination test result (The bearing test piece is setting to the experimental apparatus).

Before

After

Fig. 5.7.42 Photographs of the examples of the ball bearing test piece before and after the endurance test. (Flat ring, retainer/ball, rotating ring from left side).

Flat Ring

Retainer

(a)

Rotating Ring

Rotating Ring

Roller / Pin

Ball

(b)

Fig. 5.7.43 The appearances of specimen after screening test at 286°C high-temperature water under the condition of 2.7  105 rotational numbers. (a) Ball Bearing, (b) roller/pin bearing.

BWR innovations

559

Fig. 5.7.44 The values of weight change normalized by the applied load after screening test at 286°C of high-temperature water under the condition of 2.7  105 rotational numbers. Fig. 5.7.45 The total weight loss of ball bearing as a function of rotational numbers for different applied load after the test at 286°C hightemperature water.

had robustness for iron cruds. Fig. 5.7.45 shows the total weight loss of ball bearing as a function of rotational numbers for different applied load. The rotational numbers of 2.7  105 and 3  106 correspond to the sliding distance of 31 km, which is the screening condition, and 339 km, which is the endurance and robustness tests condition. The weight loss increased with the rotational numbers and the applied load. The rate of weight loss vs rotational numbers decreased with increasing rotational numbers. It is considered that the applied stress on the contacting surface between the ball and the flat ring decreased with increase in the area of contacting surface due to the wear loss. The screening test results showed that wear loss of the ball bearing type is much smaller than other two pin/roller bearing type. Even after the endurance and robustness tests, the ball bearing could be removed smoothly and no abnormal wear phenomena were observed on the specimens. Fig. 5.7.46 shows the typical SEM images of wear scars on the surface of the flat test ring after the endurance test.

560

Boiling Water Reactors

Wear Scars

200μm

20μm

Fig. 5.7.46 Typical SEM image of wear scans on the surface of the flat ring of the ball bearing after the endurance test at 286°C high-temperature water under the load weight of 480 N.

The SEM observation of the test ring and ball surfaces after test has been revealed to be quite smooth. The graphite retainer is considered to react as the solid lubricant and to reduce the wear loss of the ball bearing during the test under high-temperature water. It is concluded from the experiments described above that the ball bearing made of SUS440C ball and graphite retainer is the candidate material for internal CRD. The endurance and robustness tests were examined in order to confirm the durability of the bearing for the internal CRD. The durability of the ball bearing for the internal CRD was performed in the high-pressure and high-temperature reactor water of current BWR conditions. The experimental results confirmed the durability of rotational numbers for the operation length of 60 years. The test results also showed good robustness even in high-density crud conditions, compared with the current BWR.

5.7.6

Two-phase flow and structural integrity

The visualized test apparatus for the two-phase FIV was installed at the NuclearPower Facility of the University of Tokyo, as shown in Fig. 5.7.47 and the highpressure FIV test was done around the internal CRD that simulates two-phase flow from the core flows upward through the guide chimney, as shown in Fig. 5.7.48. Thus, two-phase flow instability and fluid-induced vibrations (FIV) were evaluated. The apparatus succeeded to visualize the vortex fluctuation of the flow reduction part, as shown in Fig. 5.7.49. It was confirmed that at high pressure of 7 MPa, the acceleration signal of FIV due to the two-phase flow reduces to a level that was not a problem even in the reduced flow section, as shown in Fig. 5.7.50.

Void fraction measuring section

Air-water separating section

Visible section

15D 0.05m

Test section 1.5m

Downstream

5m

5D 0.1m

Fluid developing section Air-water mixing section

Tank

(a)

Pump 60-600L/min

Upstream

N2 30-300L/min

Internal CRD

(b)

RPV

Core

(c)

(d)

Fig. 5.7.47 Two-phase FIV test facility and a memorial photo of the cooperation team key members. (a) Visualized test facility at the University of Tokyo, (b) visualized test result in shrinking flow, (c) shrinking flow area at bottom of internal CRD in ABWR-III, (d) cooperation team key members.

562

Boiling Water Reactors

477mm

Void fraction probe

φ 97mm

Pressure sensor

750mm

φ 74mm

Acceleraon sensor

Pressure sensor

Acceleraon sensor

Fig. 5.7.48 High-pressure Two-phase FIV test facility using BEST loop at Toshiba.

t=36ms

t=24ms 30

20

20

y (mm)

y (mm)

30

10 0

10 0

-10

-10

-20 -30 -30 -20 -10 0 10 20 30 x (mm)

-20 -30 -30 -20 -10 0 10 20 30 x (mm)

30

t=30ms

30

20

y (mm)

y (mm)

2

t=42ms

20

10 0

Velocity Contour V (m/s) 3

10 0

-10

-10

-20 -30 -30 -20 -10 0 10 20 30 x (mm)

-20 -30 -30 -20 -10 0 10 20 30 x (mm)

1

0

Fig. 5.7.49 Visualized test for the two-phase PIV measurement by the University of Tokyo.

5.7.7

LOCA and pressure transient analysis

The internal structure of the ABWR-III is much different from the current ABWR. Therefore, LOCA and the pressure transient analysis were conducted by using the TRAC code, as shown in Fig. 5.7.51. The model includes the RIP tube for shroud-less low elevation core structure and CR guide chimney. Fig. 5.7.52 shows the LOCA

Acceleration in the contraction flow path (m/sec2)

BWR innovations

563

1.0 Test condition Contraction 4inch to 3 inch 100% rated Flow

0.8

7MPa, 0deg. 7MPa, 90deg. 5MPa 3MPa 1MPa

0.6

0.4

0.2

0

0

2.5

5.0

7.5

10.0 Quality (%)

12.5

15.0

17.5

20.0

Fig. 5.7.50 High-pressure FIV test result at flow contraction.

21.0m Vessel SRV TCV

Separator/Dryer 15.0m MS line

MSIV

TBV 12.0m 9.7m

FW line

CRD

RIP Tube 6.0m

1.5m Channel 0m 0m

RIP 3.0m 3.6m

Fig. 5.7.51 TRAC code analysis model for ABWR-III.

564

Boiling Water Reactors

Fig. 5.7.52 LOCA and the transient analysis model for the TRAC code. (a) Water level during LOCA, (b) pressure transient after the MSIV closure.

analysis results of water level in the RIP tube and outside of the short shroud. Though, the ECCS flow rate was the same as that of current ABWR of 1356 MWe, the results showed the full-term flooding over the top of active fuel (TAF). Fig. 5.7.52b showed the pressure transient analysis when all the MSIVs were shut by some signals. The ABWR-III’s power is up-rated by 25% (1700MWe) compared with the current ABWR. The analysis was conducted under the condition of total SRV flow rate, which was the same as that of current ABWR of 1356 MWe, the pressure transient peak is only 2.5% higher than that of the current ABWR.

5.7.8

Aseismic analysis results

Seismic analysis of ABWR-III was carried out. Fig. 5.7.53 shows the seismic analysis model, the analysis results in each typical seismic frequency deformation mode are shown in Fig. 5.7.54: fuel primary vibration mode, secondary vibration mode, CR guide primary mode, CR guide primary mode to third vibration mode. It was confirmed that there was no support deformation, the lateral displacement of the CR was small, and there was no problem with the insert-ability of the control rods.

5.7.9

Summary

The internal CRD using a heatproof ceramic-insulated coil was successfully developed. No space would be required for CRDs and CRD exchange under RPV of 1700-MWe ABWR-III. These advantages realize a compact PCV and reduced volume of a reactor building. The internal CRDs eliminate penetration via a bottom flange of

BWR innovations

565

5.8m

RPV

21

20

Stabilizer Dryer K4 19

1.5m

3.7m

7.1m

4.4m

3.7m

2.3m 21m

3.3m

1 8 17

CRD support 16 grid K3 15

14 support 13 skirt 12 11 10K2 9 8 Upper 7 core 6 grid 5 Core 4 support 3 K1 plate 2

38

Separator 37 36 Shroud head 35 34 33 32 31 30 29 28 27 26 25 24 23 22

Internal CRD CR guide chimney Core/Fuel

1

5.8m

Fig. 5.7.53 Aseismic analysis model for ABWR-III [10].

RPV

Fuel Bundle 1st mode (4.8Hz)

2nd mode(18.3Hz)

3rd mode (19.0Hz)

(k1,k2,k3=107 kN/m) Fig. 5.7.54 Aseismic analysis results for ABWR-III [10].

RPV and lower installation level of RPV in a dry well. The author conducted the following evaluations and the several tests and analysis have been completed by one of the selected offers for the advertised technical developments of IAE founded by METI.

566

Boiling Water Reactors

(1) Plant concept and reducing the height of reactor building has great merits. (2) The latch and scram tests of the internal CRD by using the magnetic coupling power connector and heatproof motor were succeeded.

The mock-up of radiation hole of Yayoi reactor and a test capsule for radiation. The neutron flux doze is almost the same as the position of the internal CRD in an actual ABWR-III. The internal CRD using a heatproof ceramic-insulated coil had been developed. The internal CRDs are installed in an RPV whose size is equivalent to the 1356MWe ABWR, and no space would be required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. The internal CRDs will eliminate penetrations via a bottom flange of RPV, and will enable to lower installation level of RPV in a dry well. The author conducted the following evaluations and the results showed the 1700-MWe ABWR-III concept is feasible as a next-generation BWR. In the ABWR-III project, the following results were successfully obtained: (1) (2) (3) (4)

Plant concept to reduce the height of reactor building. Internal CRD latch mechanism and heatproof motor. Aseismic design analysis of the shroud-less RPV internal structure with the RIP tube. LOCA and pressure transient analysis.

References [1] T. Narabayashi, T. Yamamoto, M. Sato, N. Kobayashi, T. Kameda, T. Tokumasu, S. Kawano, T. Hagiwara, M. Mori, S. Ohmori, T. Terai, H. Madarame, Y. Morimoto, Development of Internal CRD for Next Generation BWR, ICONE11-36508, 2003. [2] T. Narabayashi, et al., “Built-in CRD” 2004 Review Committee Materials, Innovative Practical Nuclear Technology Development Proposal Public Offering METI Project, March 13, 2004. [3] T. Narabayashi, et al., “Built-in CRD” 2004 Review Committee Materials, Innovative Practical Nuclear Technology Development Proposal Public Offering METI Project, March 10, 2005. [4] Ishida, et al., C41 Annual mtg. of AESJ, 2000. [5] H. Nunkokawa, et al., Development of Ball Bearing in High Temperature Water for Invessel Type Control Rod Drive mechanism of Advanced marine reactor, in: JAERI-Tech 2001-040, 2001. [6] T. Narabayashi, et al., “Development of Next Generation BWR with Internal CRD”, International Congress on Advances in Nuclear Power Plants: ICAPP03, Spain, May. 2003. [7] T. Narabayashi, et al., Development of Internal CRD for Next Generation BWR, Kyoto, in: GENES4/ANP2003, Sep 2003. [8] Y. Morimoto, Fluctuation of Void Fraction and Pressure Drop during Vertical Two-Phase Flow with Contraction, in: GENES4/ANP2003, Sept. 2003. [9] Y. Morimoto, Pulsation of two-phase flow through a vertical pipe with contraction, in: Asia Pacific Vibration Conference (APVC2003), Australia, Sep 2003. [10] T. Narabayashi, M. Sato, T. Kameda, S. Kawano, T. Hagiwara, C. Iwaki, T. Tokumasu, N. Kobayashi, K. Araoka, T. Terashima, R. Sugawara, S. Ishizato, M. Mori, S. Ohmori,

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S. Goto, T. Terai, H. Nishimura, A. Sawada, H. Madarame, Y. Morimoto, Development of Internal CRD for Next Generation BWR, ICONE13-50783, 2005. [11] S. Goto, S. OHMORI, M. Mori, S. Kawano, T. Narabayashi, S. Ishizato, Development of Internal CRD for Next Generation BWR, Endurance and robustness tests of ball-bearing materials in high-pressure and high-temprature water, ICONE13-50927, 2005.

Index Note: Page numbers followed by f indicate figures and t indicate tables. A ABWR. See Advanced boiling water reactor (ABWR) AC. See Atmospheric control system (AC) Accident management, BWR, 304–314, 304t core damage sequence, 307, 308t defense, depth, 305, 305t existing, 312 ex-vessel phenomena, RPV failure, 310–312, 310–311f international event scale (INES), 306–307, 306t, 310f in-vessel phenomena, 308–310, 311f, 313f operated, planned plants, 312–314, 313t, 313f selection, measures, 307, 308t Accident-tolerant fuels (ATF), 222–223 ADS. See Automatic depressurization system (ADS) Advanced boiling water reactor (ABWR), 55, 62 Advanced thermal reactor (ATR), 10 Advisory Committee on Reactor Safeguards (ACRS), 14 AEC. See US Atomic Energy Commission (AEC) Allgemeine Elektrizit€ats-Gesellschaft (AEG), 45 American Society of Mechanical Engineers (ASME), 13 ANL. See US Argonne National Laboratory (ANL) Aseismic analysis, 564 ASME. See American Society of Mechanical Engineers (ASME) Atmospheric control system (AC), 95 Atomic Energy Basic Law (AEBL), 14 Atomic Energy Commission (AEC), 36–37 Automatic depressurization system (ADS), 103, 341–342 Auxiliary normal transformer (ANT), 126

Auxiliary standby transformer (AST), 125–126 Auxiliary system, 135 fuel pool cooling and cleanup (FPC) system, 135–137, 136f heating ventilating and air conditioning (HVAC) system, 145–148, 147–148f high-pressure nitrogen gas supply (HPIN) system, 143, 144f instrument air (IA) system, 141–143, 142f makeup water condensate system (MUWC), 140–141, 141f reactor building cooling water (RCW) system, 137–138, 138f reactor building service water (RSW) system, 138f, 139 sampling system (SAM), 143–145 turbine building cooling water (TCW) system, 139, 140f turbine building service water system (TSW), 140, 140f B Boiling transition (BT), 473–475, 474f Boiling water reactors (BWRs), 55, 60, 61f analysis codes, 315–325 BE code, 320 best estimate code, evaluation model code, 315–317, 316–317f computational fluid dynamic (CFD), 322–323, 323f EM code, 318–319, 318f, 319t large-scale test facility, 323–325, 324f, 326–327f nuclear, 315, 316t SA progression, 320–322 verification and validation (V&V), 317 core characteristics, 168–172 configuration, 171–172, 171–172f, 171t negative void reactivity, 169–170, 169–170f

570

Boiling water reactors (BWRs) (Continued) design innovation, high-pressure BWR, 434–435 established stage, 16–18 Argonne National Laboratory, United States, 4f, 16–18 existing, 487–491 concept, 487–488, 487f performance evaluation test, 488–491, 489–490f fuel assemblies, void fraction measurement tests, 248, 248t fuel failures, 218–223, 219–221f, 222t NPS new requirements, 378–388, 379–382f, 381–383t, 384–388f nuclear analysis 3D core calculation analysis, 260–262, 261–263f 2D lattice calculation, 252–260, 252–253f, 255–260f validation, measurements, 262–266, 264–267f power plant, BT, 473–475, 474f power uprate, 456–458 constant rated reactor, 456–457 current status, equipment modification, 457–458 possibilities, issues, 457 realizing stage, 19–34 ABWR development, international cooperation, 29–30, 30f early stage, GE’s BWR development, 19, 20–21f further stage, GE’s BWR development, 19–29, 22f, 23–24t, 25–28f next BWR development, 26f, 30–34, 32t safety systems, severe accident, 326–333 passive safety concept, 326–327, 327f passive safety systems, 328–333, 328f, 330–333f reinforcement, passive safety, 327–328 BT. See Boiling transition (BT) C C&I. See Control and instrumentation (C&I) systems Circulating water pumps (CWPs), 150 Condensate demineralizer (CD), 150

Index

Condensate filter (CF), 150 Condensate storage pit (CSP), 103 Containment atmospheric monitoring system (CAMS), 339–340 Containment system, 90–102, 91f heat removal system, 97–99, 99–100f primary containment isolation system (PCIS), 94–95 primary containment vessel (PCV), 92–94, 93f primary containment vessel gas control systems, 95–97 atmospheric control system (AC), 95–96, 96f flammability control system (FCS), 96–97, 97–98f secondary containment, 100–101 standby gas treatment system (SGTS), 101–102, 101–102f Control and instrumentation (C&I) systems, 62 Control rod (CR), 19, 533–534 Control rod drive (CRD), 339–340, 533–534 power devices, 536–542, 538t, 538–540f Control rod guide tubes (CRGTs), 66 Coolant flow paths, BWR operating map, 175–176, 177f Core monitoring system, 74, 74f Crud-induced localized corrosion (CILC), 233 D Decommissioning Fukushima Daiichi, 389–401 contaminated water management, 396–399, 398–400f current status, reactors, Units 1 through 4, 389–392, 389t, 389–393f, 391t fuel-debris removal, 400–401, 400–403f groundwater flow, contaminated water, 394–395, 396–398f PCV, estimated leak path, 392–394, 394–396f Design base accidents (DBAs), 62 Digital I&C system, ABWR plant, 110–111, 111f auxiliary control system (ACS), 113–115, 116f

Index

human-machine interface (HMI), 123–124, 123–124f major control systems, 112–113, 112f, 114f process computer system, 120–122, 122f safety system logic and control system (SSLC), 115–120, 117–119f, 121f Digital technology, 109, 109t Digital trip module (DTM), 120 DOE. See US Department of Energy (DOE) E EBWR. See Experimental boiling water reactor (EBWR) ECCS. See Emergency core cooling system (ECCS) Economical improvement step I, 182–186, 183–185f, 187f step II, 186–188 step III, 188–190, 189f EDGs.. See Emergency diesel generators (EDGs) Electrical power distribution systems, 62 Electric power supply system, ABWR, 125–126, 125f configuration/main equipment, 126–134 AC instrumentation, 132–134, 134f auxiliary medium-voltage distribution buses, 127–129 DC, 130–132 emergency diesel generators (EDGs), 129–130, 131t grid connection, 126–127 transformers, 127 function, 126 Emergency core cooling system (ECCS), 14, 102–108, 103t, 104f, 336 high-pressure core flooder system, 105–108 low-Pressure Flooder (LPFL), 106–110 reactor core isolation cooling (RCIC) system, 103–106 Emergency diesel generators (EDGs), 62, 336 Emergency operating procedure (EOP), 350 Engineered safety features, 90–91 ERDA. See US Energy Research and Development Administration (ERDA) European Advanced Boiling Water Reactor (EU-ABWR), 314, 482–486

571

concept, 482–486, 483f performance evaluation test, 484–486, 485–487f, 485t Evaluation, ABWR-III conditions, 551–560 durability test, ball bearing, 551–560, 553t, 557t Event progress, analysis evaluation unit 1, 337–339, 338–339f unit 2, 339–340, 340f unit 3, 341–343, 341–342f Experimental boiling water reactor (EBWR), 18 F Fast breeder reactor (FBR), 5 Filtered containment venting system (FCVS), 378 Fine motion control rod drive (FMCRD), 10 Flammability control system (FCS), 95 Flood damage, EDGs, 336, 338t Flooded areas, NPS, 336, 337f Flow-induced vibration, 206–211, 206t, 207–211f Flow pattern, heat transfer, BWR fuel bundle, 212–213, 213–216f FMCRD. See Fine motion control rod drive (FMCRD) Fuel assemblies (FAs), 63 Fuel integrity, 475–477 core catcher, core melt stabilization, 479–482, 480–481f criteria, 477 Fuel rod failure, countermeasures, 223–244, 224t, 227–229f, 231–240f, 242f Fukushima Daiichi accident, 362–367, 401–404, 404f filtered containment venting system, 362–367, 365–366f measures, severe accidents installed, 362, 364f severe accidents, countermeasures, 362, 363–364f special emergency heat removal system, 366f, 367 tsunami protection, 367, 367f Fukushima Daini NPS, avoiding severe accidents, 346–362

572

Fukushima Daini NPS, avoiding severe accidents (Continued) emergency response, 346–350, 346–347f, 348–349t response, station behavior, 350–362 ERC planning activities, 356–362, 357–361f reactor cooling water injection, PCV cooling, 351–353, 352f response status, tsunami arrival, 350, 351f RHR restoration, reactor cold shutdown, 353–356, 355–356f G Gas-cooled reactor (GCR), 5 General Electric Company (GE), 18 Generator disconnecting switch (GDS), 126 Generator load switch (GLS), 126 Generator transformer (GT), 125–126 Gland steam condenser (GSC), 150 Gland steam evaporator (GSE), 150 H High burnup fuel design, 178–191, 178t High-pressure auxiliary core cooling (HPAC), 382 High pressure condensate pumps (HPCPs), 150 High-pressure core flooder (HPCF), 102–103 High-pressure core injection (HPCI), 9, 341 High-pressure core splay system (HPCS), 9 High-pressure drain pumps (HPDPs), 150 High-pressure drain tank (HPDT), 150 High-pressure feedwater heaters (HP FWHs), 150 Hydrogen explosion, unit 4, 342–345f, 343–346 I Internal control rod drive (CRD) mechanism, 542–551 ceramics coil radiation, 546–549, 547–554f, 556–565f heatproof motor, 542–546, 546–547f latch, 542, 543–545f neutron flux, 550–551 Isolation condenser system (ICS), 26

Index

J Japan Atomic Power Company (JAPC), 9 K Kernkraftwerk Leibstadt (KKL), 366–367 L Large-Scale Test Facility (LSTF), 323–324 Leak detection system, 87–89, 88f Light-water reactor (LWR), 16 Local power range monitors (LPRMs), 74 Long operating simplified BWR (LSBWR) design module fabrication, construction construction methodology, evaluation, 451–454 LSBWR and LLBWR’s building design, 451, 452–454f ship hull structure, 450–451, 450f, 452f objective, 435–447, 437t, 438–440f conceptual design, 436–441, 441–442f module fabrication, construction, 449–454 natural circulation, 436 safety system, PCV concept, 447–448, 448–449f thermal efficiency, 442–447, 443–447f Loss of coolant accident (LOCA), 30, 91 pressure transient analysis, 562–564 Loss of off-site power (LOOP), 91 Low conductivity waste treatment system (LCW), 76 Low-pressure condensate pumps (LPCPs), 150 Low-pressure core flooder (LPCF), 102–103 Low-pressure drain pumps (LPDPs), 150 Low-pressure drain tank (LPDT), 150 Low-pressure feedwater heaters (LP FWHs), 150 LWR. See Light-water reactor (LWR) M Magnet coupling signal connector, 541–542, 541–542f Main control room (MCR), 123 Main steam isolation valves (MSIVs), 75 Maximum linear heat generating rate (MLHGR), 64

Index

Minimum critical power ratio (MCPR), 64 Ministry of International Trade and Industry (MITI), 38 Modular Accident Analysis Program (MAAP), 339 Moisture separator reheaters (MSRs), 150 Motor control centers (MCCs), 129 MOX core design, 197–199, 198f MOX fuel assemblies, 61 LWR units, 191–192, 191f assembly design, 195–197, 196–197f design, 191–199 Multistage steam injector test facility, 515–516, 516t, 516f N New nuclear regulatory requirements, Japan, 368–373, 368–369f tornado protection examples, 373, 374t, 375f tsunami protection examples, 368–372, 370–373f New regulatory standards, 374–378, 376–378f NRC. See US Nuclear Regulatory Commission (NRC) Nuclear and Industrial Safety Agency (NISA), 336 Nuclear boiler (NB) system, 76–82 feedwater (FDW) system, 81–82, 81f main steam (MS) system, 76–81, 77–80f Nuclear energy development, Japan ABWR improvement, construction, 53–54 all-weather construction method, 54 large block-module construction method, 53–54 electric power generation, 4, 4–5f first improvement, standardization program, 37f, 39–42, 40t, 41f countermeasure for SCC, 39–42 improved primary containment vessel/ improved PCV, 42 improvement, standardization programs, Japan, 56–57 introduction period, BWR, 55–56 legislation, 13–15 nuclear power generation, 5–12, 6–8t, 10–11f, 11–13t

573

primary energy supply, 2–3, 2f reduction of construction period, BWRs, 50–52, 50f computer management, IT technologies, 51–52, 52f large block, module technologies, 51, 52f second improvement, standardization program, 42–44 fuel-core design improvement, CRD, 43–44 status, 57–58 technology importation, 36–39, 37f third improvement, standardization program, 43–46f, 44–48, 47t ABWR development program, 46–48 Nuclear Power Engineering Corporation (NUPEC), 45, 473 Nuclear power plant (NPP), 16, 336 Nuclear Regulation Authority (NRA), 15, 336, 368 Nuclear Regulatory Commission (NRC), 464–472 Nuclear thermal-hydraulic stability, 202–205, 203–205f O Operational improvement, 179–182, 180–182f P Passive containment cooling system (PCCS), 482, 483f PCIS. See Primary containment isolation system (PCIS) Pellet Clad Interaction (PCI), 43, 179 Plant concepts, ABWR-III, 534–536, 535–537f, 536t Plutonium (Pu) fuel basic information, 192, 192–193f utilization, 193–195, 194–195f Preconditioning interim operating recommendations (PCIOMR), 179 Pressurized light-water reactors (PWRs), 3 Primary containment isolation system (PCIS), 120 Primary containment vessel (PCV), 9, 75, 336 Primary containment vessel boundary (PCVB), 94 Probabilistic risk assessment (PRA), 383

574

R Resource-renewable boiling water reactor (RBWR) concept, 409–413, 411–412f, 414–415f core characteristics, 418–419, 420f progressive introduction, 419–422, 422f specifications, 413–417, 416–418f RCIC. See Reactor core isolation cooling system (RCIC) Reactivity control system, 71–74 control rod drive system, 71–72, 73f standby liquid control system (SLC), 72–74, 73f Reactor coolant pressure boundary (RCPB), 76 Reactor cooling systems, 60 Reactor core isolation cooling system (RCIC), 9, 336 Reactor core support structure, 172–174, 173–174f Reactor feedwater pumps (RFPs), 150 Reactor feedwater pump turbines (RFP-Ts), 150 Reactor internal pump (RIP), 30 Reactor power uprate, 459–461, 460f equipment modification, 464–472 implementations uprates, US, 469–472, 471t LEFM, 466–467, 467t ultrasonic flowmeters, 468–469, 470t uncertainty recapture, 464–466, 465f Reactor pressure vessel (RPV), 62, 339 Reactor recirculation system, 82–83, 82–84f Reactor system, 62–70, 63f control rods (CRs), control rod drive (CRD), 69–70, 70–72f core, fuel, 69 reactor internals, 66–69, 66–68f reactor pressure vessel (RPV), 64–65, 65f Reactor thermal power issues, safety, 461f, 462–463 electric power, 458, 459f, 461–462 Reactor water cleanup (CUW) system, 83–85, 85f Recirculation internal pumps, 75–76 Reduced-moderation light water reactor, 423–424 low-moderation spectrum BWR, 424–432, 424–433f

Index

Reinforced concrete containment vessel (RCCV), 10 Reliability improvement (1970s), 179 Remote multiplexing unit (RMU), 120 Research Safeguards Committee (RSC), 14 Residual heat removal system, 86–87, 86f RPV. See Reactor pressure vessel (RPV) S Safety logic units (SLUs), 120 Safety relief valves (SRVs), 75, 339–340 Safety system logic and control (SSLC), 129 SI analysis model, 495–497 SI-FWH analysis, 517–520, 517–519f, 521–522f ABWR FWH system, 522, 523f pump-up water system, 523–530 air-purge analysis, 529–530, 529–531f full-scale mock-up test, 527–529, 527–528f, 528t PCC/IC pool, 525–527, 525t, 526f SIPOWER, 523–524, 524f transient test result, 520–521, 522f Spent fuel pool (SFP), 389 Standby gas treatment system (SGTS), 92, 344 Standby liquid control system (SLC), 64 Start-up range neutron monitors (SRNMs), 74 Steam injector (SI), 493 principle, application, 493–495, 494–495f steam jet-type application, 503–506, 503f high-pressure tests, 503–506, 504–505f test results, 495f, 499–502, 500–502f visualized fundamental tests, 497–502, 498f water jet-type application, 506–512, 506t, 507–508f, 508t confirmation analysis, 506–508 high-pressure tests, scale models, 510–512, 510t, 511–512f scale-up examination, 509–510, 509t, 509–510f simplified feed water system, 513–522, 513–515f Steam jet air ejector (SJAE), 150 Steam, power conversion systems, 149–164, 149f circulating water system, 157, 158f

Index

condensate, feedwater system, 158–159, 160–161f condenser, 157 extraction steam system, 152–153, 153f feedwater heater drain, vent system, 154–156, 156f main, auxiliary steam system, turbine bypass system, 151–152, 152f off-gas system, 159–164, 164f turbine generator, 150–151 turbine gland steam system, 154, 155f Stress corrosion cracking (SCC), 9, 30 Systems, structures, and components (SSCs), 60 T TEPCO’s Nuclear Power Station, 463t Thermal hydraulics correlations, full-scale BWR fuel assemblies, 249 design, reactor core, 199–201, 201f performance, BWR fuel assembly, 247 severe accidents, 297–303 FP aerosol behaviors, 303, 309f initiation, fuel melt, 298–300, 298f, 299t, 300–302f melting jet structure, behaviors, 302–303, 306f melting relocation, RPV, 301–302, 303f water-zircaloy reaction, 300–301

575

subchannel analysis code, 281–297, 282–283f, 287f, 291f system analysis code, 270–281, 272f, 274f, 277f TMI. See US Three Mile Island (TMI) Tokyo Electric Power Company (TEPCO), 9 Trans-uranic (TRU) burner reactor, 409, 419f, 421t Turbine bypass valve (TBV), 150 Two-Loop Test Apparatus (TLTA), 324 Two-phase flow, structural integrity, 560–561 U Uninterruptible power supply (UPS) system, 126 US Argonne National Laboratory (ANL), 18 US Atomic Energy Act, 14 US Atomic Energy Commission (AEC), 14 US Department of Energy (DOE), 14 US Energy Research and Development Administration (ERDA), 14 US Nuclear Regulatory Commission (NRC), 14 US Three Mile Island (TMI), 39 W Westinghouse Electric Corporation (WEC), 16