Boiling Water Reactors

Boiling Water Reactors

1st Edition - July 1, 2022

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  • Editors: Yasuo Koizumi, Koji Nishida, Shinichi Morooka, Michitsugu Mori
  • Paperback ISBN: 9780128213612

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Description

Boiling Water Reactors, Volume Four in the JSME Series on Thermal and Nuclear Power Generation compiles the latest research in this very comprehensive reference that begins with an analysis of the history of BWR development and then moves through BWR plant design and innovations. The reader is guided through considerations for all BWR plant features and systems, including reactor internals, safety systems and plant instrumentation and control. Thermal-hydraulic aspects within a BWR core are analyzed alongside fuel analysis before comparisons of the latest BWR plant life management and maintenance technologies to promote safety and radiation protection practices are covered. The book's authors combine their in-depth knowledge and depth of experience in the field to analyze innovations and Next Generation BWRs, considering prospects for a variety of different BWRs, such as High-Conversion-BWRs, TRU-Burner Reactors and Economic Simplified BWRs.

Key Features

  • Written by experts from the leaders and pioneers in nuclear research at the Japanese Society of Mechanical Engineers
  • Includes real examples and case studies from Japan, the US and Europe to provide a deeper learning opportunity with practical benefits
  • Considers societal impacts and sustainability concerns and goals throughout the discussion
  • Explores BWR plant design, thermal-hydraulic aspects, the reactor core and plant life management and maintenance in one complete resource

Readership

Nuclear engineers and researchers focusing on BWRs; BWR plant designers and operators; regulators; post graduate students of nuclear engineering; thermal energy engineers and researchers; national labs; government officials and agencies in power and energy policy and regulations

Table of Contents

  • 1. History of BWR Development

    1.1 Nuclear Energy Development in Japan

    1.1.1 Primary energy supply

    1.1.2 Electric power generation

    1.1.3 Nuclear power generation

    1.1.4 Nuclear power generation legislation

    1.2 Establishment and Realization of BWR Technologies

    1.2.1 Established stage

    1.2.2 Realizing stage

    1.3 Improvement and Standardization program in Japan

    1.3.1 Technology importation

    1.3.2 First improvement and standardization program

    1.3.3 Second improvement and standardization program

    1.3.4 Third improvement and standardization program

    1.4 Improvement of System and Construction

    1.4.1 Reduction of construction period of BWRs

    1.4.2 Improvement of ABWR system and construction

    1.5 Construction Experience and Operation Performance

    1.5.1 Introduction period

    1.5.2 Improvement and standardization programs of Japan

    1.5.3 Recent status

    2. Features of BWR Plant

    2.1 Introduction

    2.2 Reactor

    2.2.1 Overview

    2.2.2 Reactor system

    2.2.3 Reactivity control system

    2.2.4. Core monitoring system

    2.3 Reactor Coolant System and Connected Systems

    2.3.1 Overview

    2.3.2 Nuclear boiler system

    2.3.3 Reactor recirculation system

    2.3.4 Reactor water clean-up system

    2.3.5 Residual heat removal system

    2.3.6 Leak detection system

    2.4 Engineered Safety Features

    2.4.1 Overview

    2.4.2 Containment system

    2.4.3 Emergency core cooling system

    2.5 Instrumentation and Controls

    2.5.1 Introduction

    2.5.2 Overall architecture (example of ABWR)

    2.5.3 Major control systems and auxiliary control systems

    2.5.4 Safety systems

    2.5.5 Process computer system

    2.5.6 Human machine interface

    2.6 Electric Power

    2.6.1 Overview

    2.6.2 Function

    2.6.3 Configuration/Main equipment (example of ABWR)

    2.7 Auxiliary System

    2.7.1 Overview

    2.7.2 Fuel pool cooling and clean-up system

    2.7.3 Reactor building cooling water system

    2.7.4 Reactor building service water system

    2.7.5 Turbine building cooling water system

    2.7.6 Turbine building service water system

    2.7.7 Makeup water condensate system

    2.7.8 Instrument air system

    2.7.9 High pressure nitrogen gas supply system

    2.7.10 Sampling system

    2.7.11 Heating ventilating and air conditioning system

    2.8 Steam and Power Conversion Systems

    2.8.1 Overview

    2.8.2 Steam and power conversion systems

    3. Nuclear Reactor Dynamics and Thermal-Hydraulics of Reactor Core and Fuel Assembly

    3.1 Reactor Internals and Coolant Flow Paths in Reactor Pressure Vessel

    3.1.1 Unique basic characteristics of BWR core

    3.1.2 Reactor core support structure and other reactor internals

    3.1.3 Coolant flow paths and BWR operating map

    3.2 Advances of Reactor Core and Fuel Assembly

    3.2.1 High burnup fuel design

    3.2.2 MOX fuel design

    3.2.3 Thermal-hydraulics design

    3.2.4 Enhancement of critical power

    3.2.5 Countermeasure and cause of fuel rod failure

    3.2.6 Proving test on thermal-hydraulic performance of BWR fuel assemblies

    3.3 Advances in Reactor Core and Fuel Assembly Analysis

    3.3.1 Nuclear analysis in BWR

    3.3.2 Thermal-hydraulic system analysis codes and subchannel analysis codes

    3.4 Advances in Containment Vessel Design

    3.4.1 Thermal-hydraulics of severe accident

    3.4.2 Accident management for BWR

    3.5 Advances of Safety Analysis Code and Safety Systems

    3.5.1 Various BWR analysis codes

    3.5.2 BWR safety systems for sever accident

    4. Fukushima Daiichi (Fukushima I) Nuclear Power Plant Accident and Analysis Evaluation

    4.1 Outline of Accident

    4.2 Event Progress and Analysis Evaluation at Unit 1

    4.3 Event Progress and Analysis Evaluation at Unit 2

    4.4 Event Progress and Analysis Evaluation at Unit 3

    4.5 Hydrogen Explosion at Unit 4

    4.6 Avoiding Severe Accidents at Fukushima Daini NPS

    4.6.1 Overview of emergency response at Fukushima Daini NPS

    4.6.2 Fukushima Daini Unit 1 response and station behavior

    4.7 Lessons Learned from Fukushima Daiichi Accident

    4.7.1 Causes of severe accidents and countermeasures

    4.7.2 Measures for severe accidents installed in western NPPs

    4.7.3 Filtered containment venting system

    4.7.4 Special emergency heat removal system

    4.7.5 Tsunami protection

    4.8 New Nuclear Regulatory Requirements in Japan

    4.8.1 New nuclear regulatory requirements

    4.8.2 Tsunami protection examples

    4.8.3 Tornado protection examples

    4.9 Example of Compliance with New Regulatory Standards for PWRs that can be used as a Reference for BWRs

    4.10 BWR NPS to be Reviewed for New Requirements or Restarting

    4.11 Activities towards Decommissioning Fukushima Daiichi

    4.11.1 Current status of the reactors at Units 1 through 4

    4.11.2 Finding contaminated water leak path for leak shut-down from PCV

    4.11.3 Isolation of ground-water flow from contaminated water

    4.11.4 Contaminated water management

    4.11.5 Preparation for fuel debris removal

    4.12 Important Lessons Learned from Fukushima Daiichi NPS Accident

    5. BWR Innovations

    5.1 Trans-Uranic (TRU) Burner Reactor and Reduced-Moderation Water Reactor

    5.1.1 TRU burner reactor

    5.1.2 Reduced-Moderation light water reactor

    5.2 High Pressure BWR

     

    5.3 Power Uprate in BWR

    5.3.1 Current status and trend of reactor power uprates

    5.3.2 Reactor thermal power and electric power

    5.3.3 Reactor power uprate with constant rated reactor thermal power operation

    5.3.4 Relationship between reactor thermal power and electric power outputs

    5.3.5 Issues and safety in constant rated reactor thermal power operation

    5.3.6 Experiences in BWR operation with constant rated reactor thermal power operation

    5.3.7 Power uprate with equipment modification

    5.4 Post-BT Standard for BWR Power Plant

    5.4.1 Introduction

    5.4.2 Standard for assessment of fuel integrity under anticipated operational occurrences

    5.5 Core Catcher

    5.5.1 Overview of core melt stabilization and cooling

    5.5.2 Core catcher of EU-ABWR

    5.5.3 Core catcher for the existing BWR

    5.6 Steam Injector

    5.6.1 Introduction

    5.6.2 Principle and application of SI

    5.6.3 SI analysis model

    5.6.4. Visualized fundamental tests

    5.6.5 Application to SI-PCIS

    5.6.6 Characteristic analysis for SI-PLR by using water jet type SI

    5.6.7 Simplified feed water system

    5.6.8 SI pump up system for PCC/IC pool water refill (SIPOWER)

    5.7 Built in Upper Internal Control Rod Drives (CRD) for Next Generation BWR

    5.7.1 Introduction of merits and technical tasks for internal CRD

    5.7.2 Plant concepts

    5.7.3 Power devices for the internal CRD

    5.7.4. Internal CRD`s mechanism

    5.7.5 Evaluation of BWR conditions

Product details

  • No. of pages: 400
  • Language: English
  • Copyright: © Elsevier 2022
  • Published: July 1, 2022
  • Imprint: Elsevier
  • Paperback ISBN: 9780128213612

About the Editors

Yasuo Koizumi

Koizumi, Yasuo is a research promotor and an invited researcher at the University of Electro-Communications at present. He had been an invited researcher of the Japan Atomic Energy Agency for five years before now. He received his PhD degree 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 stayed at the Idaho National Engineering Laboratory from 1981 through 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 he had been in the Japan Atomic Energy Agency since then. His research is focused in 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 on the microscale. Since his research field is closely related to energy systems, he has great interest in thermal and nuclear power stations and energy supply in society.

Affiliations and Expertise

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

Koji Nishida

Nishida, Koji received a Doctor of Engineering degree in 1987 from Kobe University for his study on convective film boiling heat transfer. He entered Hitachi Research Laboratory where he started researching thermal hydraulics of boiling water reactors (BWRs). He was engaged in developing high burn-up fuel bundles and high performance next generation BWRs including the SMR, Small Modular Reactor. 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. At present, he is doing research on severe accidents and safety systems for pressured water reactors.

Affiliations and Expertise

Institute of Nuclear Safety System, Mihama, Fukui, Japan

Shinichi Morooka

Morooka, Shinichi is an emeritus professor of Waseda University. He graduated from the Doctor course of Mechanical Engineering at Waseda University in 1977. He received Dr. Eng. degree from Waseda University in 1980. His research field is Thermal-hydraulics of Nuclear Power Plant. He worked at Toshiba Corporation in the thermal-hydraulics R&D Center of nuclear power plants for about 30 years. He has a great deal of experience in developing components for actual nuclear power plants. He came back to Waseda University as a professor in 2010. He is an emeritus professor of Waseda University. Now, he optimizes the heat transfer performance for Light Water Reactor components using Computed Fluid Dynamics code and experimental technologies. Target Components are Nuclear Fuel, Separator system, Steam Generator, so on. He constructs flow mechanism, develops an original simulation code based on flow mechanisms and predicts the heat transfer performance of fuel assembly.

Affiliations and Expertise

Emeritus Professor of Waseda University, Shinjuku, Tokyo, Japan

Michitsugu Mori

Mori, Michitsugu is currently an invited guest professor of the Graduate School of Engineering, Hokkaido University and a guest researcher of Japan Atomic Energy Agency (JAEA). He researched the quenching cooling process of a fuel rod during reactivity-initiated accidents at Department of Nuclear Engineering, Graduate School of Engineering, Tohoku University, Japan and was awarded 1981 with Dr. of Eng. He researched a modular gas-cooled reactor (MGR) Brayton cycle and components at the Department of Nuclear Science & Engineering of Massachusetts Institute of Technology (MIT), USA from 1987 to 1989. He joined the R & D Center of Tokyo Electric Power Company (TEPCO), and then researched the LWR thermal-hydraulics and advanced measurement technologies including next generation reactors. He became a full professor at Hokkaido University in 2012 and performed the experiments and simulations on the nuclear system safety, e. g., reactor core injection system, plant transient and debris behaviors, and next generation reactors. He was the president of JSMF, JSME and AESJ Board of Directors, the vice-presidents of HTSJ and AESJ, and currently is the fellows of JSME and AESJ, and the honorary members of JSME and HTSJ.

Affiliations and Expertise

Invited Guest Professor of Graduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan, and Guest Researcher of JAEA, Tokai, Ibaraki, Japan

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