Nuclear Systems Volume II : Elements of Thermal Hydraulic Design book cover
2nd Edition

Nuclear Systems Volume II
Elements of Thermal Hydraulic Design




ISBN 9781482239584
Published December 14, 2021 by CRC Press
657 Pages 213 B/W Illustrations

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Book Description

This book provides advanced coverage of a wide variety of thermal fluid systems and technologies in nuclear power plants, including discussions of the latest reactor designs and their thermal/fluid technologies. Beyond the thermal hydraulic design and analysis of the core of a nuclear reactor, the book covers other components of nuclear power plants, such as the pressurizer, containment, and the entire primary coolant system.

Placing more emphasis on the appropriate models for small-scale resolution of the velocity and temperature fields through computational fluid mechanics, the book shows how this enhances the accuracy of predicted operating conditions in nuclear plants. It introduces considerations of the laws of scaling and uncertainty analysis, along with a wider coverage of the phenomena encountered during accidents.

FEATURES

  • Discusses fundamental ideas for various modeling approaches for the macro- and microscale flow conditions in reactors
  • Covers specific design considerations, such as natural convection and core reliability
  • Enables readers to better understand the importance of safety considerations in thermal engineering and analysis of modern nuclear plants
  • Features end-of-chapter problems
  • Includes a solutions manual for adopting instructors

This book serves as a textbook for advanced undergraduate and graduate students taking courses in nuclear engineering and studying thermal/hydraulic systems in nuclear power plants.

Table of Contents

1. Formulation of the Reactor Thermal Hydraulic Design Problem

2. Scaling of Two-Phase Flows in Complex Nuclear Reactor Systems

3. Single, Heated Channel Transient Analysis

4. Multiple Heated Channels Connected Only at Plena

5. Analysis of Interacting Channels by the Porous Media Approach

6. Analysis of Interacting Channels by the Subchannel Approach

7. Flow Loops

8. Steady State and Transient Analysis of Centrifugal Pumps

9. Thermal Analysis of Pressurizers

10. Thermal Analysis of Containments

11. Thermal Analysis of Steam Generators and Condensers

12. Fundamentals of Reactor Transient Simulation

13. Treatment of Uncertainties in Reactor Thermal Analysis

...
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Author(s)

Biography

Neil E. Todreas is a Professor Emeritus in the Departments of Nuclear Science and Engineering and Mechanical Engineering at the Massachusetts Institute of Technology. He held the Korea Electric Power Corporation (KEPCO) chair in nuclear engineering from 1992 until his retirement to part- time activities in 2006. He served an 8-year period from 1981 to 1989 as the Nuclear Engineering Department Head. Since 1975, he has been a codirector of the MIT Nuclear Power Reactor Safety summer course, which presents current issues of reactor safety significant to an international group of nuclear engineering professionals. His area of technical expertise includes thermal and hydraulic aspects of nuclear reactor engineering and safety analysis. He started his career at Naval Reactors working on the Shippingport reactor and surface nuclear vessels after earning a BEng and an MS in Mechanical Engineering at Cornell University. Following his ScD in Nuclear Engineering at MIT, he worked for the Atomic Energy Commission (AEC) on organic-cooled/heavy water-moderated and sodium-cooled reactors until he returned as a faculty member to MIT in 1970. He has an extensive record of service for government (Department of Energy (DOE), U.S. Nuclear Regulatory Commission (USNRC) and national laboratories), utility industry review committees including INPO, and international scientific review groups. He has authored more than 250 publications and a reference book on safety features of light water reactors. He is a member of the U.S. National Academy of Engineering and a fellow of the American Nuclear Society (ANS) and the American Society of Mechanical Engineers (ASME). He has received the American Nuclear Society Thermal Hydraulics Technical Achievement Award, its Arthur Holly Compton Award in Education, and its jointly conferred (with the Nuclear Energy Institute) Henry DeWolf Smyth Nuclear Statesman Award.

Mujid S. Kazimi was a Professor in the Departments of Nuclear Science and Engineering and Mechanical Engineering at the Massachusetts Institute of Technology. He was the first director of the Center for Advanced Nuclear Energy Systems (CANES) and the Tokyo Electric Power Company (TEPCO) chair professor in Nuclear Engineering both at MIT. Prior to joining the MIT faculty in 1976, Dr. Kazimi worked for a brief period at the Advanced Reactors Division of Westinghouse Electric Corporation and at Brookhaven National Laboratory. At MIT, he was Head of the Department of Nuclear Engineering from 1989 to 1997 and Chair of the Safety Committee of the MIT Research Reactor from 1998 to 2009. He served from 1990 to 2015 as the codirector of the MIT Nuclear Power Reactor Safety summer course. He was active in the development of innovative designs of fuel and other components of nuclear power plants and in analysis of the nuclear fuel cycle options for sustainable nuclear energy. He co-chaired the 3-year MIT interdisciplinary study on the Future of the Nuclear Fuel Cycle, published in 2011. He served on scientific advisory committees at the U.S. National Academy of Engineering and several other national agencies and laboratories in the United States, Japan, Spain, Switzerland, Kuwait, the United Arab Emirates and the International Atomic Energy Agency. He authored more than 200 articles and papers that have been published in journals and presented at international conferences. Dr. Kazimi earned a BEng at Alexandria University in Egypt and an MS and PhD at MIT, all in Nuclear Engineering. He was a fellow of the American Nuclear Society and the American Association for the Advancement of Science. Among his honors is the Technical Achievement Award in Thermal Hydraulics by the American Nuclear Society and membership in the U.S. National Academy of Engineering.

Mahmoud Massoud is an adjunct Professor at the Department of Mechanical Engineering, University of Maryland, College Park, where he has taught undergraduate and graduate level engineering courses since 1988. Dr. Massoud has been active in the Nuclear Industry as a Principal Engineer at the Exelon Corporation and as a Consulting Engineer since 1989. He has authored a textbook in engineering and over 20 journal and conference papers. Dr. Massoud earned his B.S. and M.S. in Mechanical Engineering with emphasis in Internal Combustion Engines from Tehran University. Subsequently, he earned his M.S. and Nuclear Engineer’s degree from MIT’s Mechanical and Nuclear Engineering Departments, respectively. He earned his Ph.D. in Nuclear Engineering from the University of Maryland College Park.