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.
Please Note: This is an On Demand product, delivery may take up to 11 working days after payment has been received.
Table of Contents
1. History of BWR development 2. Features of BWR Plant 3. Nuclear Reactor Dynamics and Thermal-Hydraulics of Reactor Core and Fuel Assembly 4. Fukushima Daiichi (Fukushima I) Nuclear Power Plant Accident and Analysis Evaluation 5. BWR Innovations
Authors
Koji Nishida Institute of Nuclear Safety System, Mihama, Fukui, Japan.
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.
Shinichi Morooka Emeritus Professor of Waseda University, Shinjuku, Tokyo, Japan.
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.
Michitsugu Mori Invited Guest Professor of Graduate School of Engineering, Hokkaido University, Sapporo, Hokkaido, Japan, and Guest Researcher of JAEA, Tokai, Ibaraki, Japan.
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.
Yasuo Koizumi The University of Electro-Communications, Chofu, Tokyo, Japan.
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.
