Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors

Volume 1: Foundations and Principles

2nd Edition - July 29, 2024
Editors: Francesco D'Auria, Yassin A. Hassan
Language: English
Paperback ISBN:
9 7 8 - 0 - 3 2 3 - 8 5 6 0 6 - 5
eBook ISBN:
9 7 8 - 0 - 3 2 3 - 8 5 6 0 7 - 2

Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 1, Foundations and Principles includes all new chapters which delve deeper into the topic, adding context a… Read more

Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors

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Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 1, Foundations and Principles includes all new chapters which delve deeper into the topic, adding context and practical examples to help readers apply learnings to their own setting. Topics covered include experimental thermal-hydraulics and instrumentation, numerics, scaling and containment in thermal-hydraulics, as well as a title dedicated to good practices in verification and validation. This book will be a valuable reference for graduate and undergraduate students of nuclear or thermal engineering, as well as researchers in nuclear thermal-hydraulics and reactor technology, engineers working in simulation and modeling of nuclear reactors, and more.

In addition, nuclear operators, code developers and safety engineers will also benefit from the practical guidance provided.

Cover image
Title page
Table of Contents
Copyright
Dedication
List of contributors
Contributors for volumes 1, 2 and 3
Foreword
Glossary
Preface to the first edition of the book
Preface to the second edition of the book
Acknowledgments (for the past) and wishes (for the future)
1. Introduction
Foreword
1.1. Introduction
Glossary
Exercises and questions
2. A historical perspective of nuclear thermal hydraulics
Foreword
2.1. Introduction
2.2. System thermal hydraulics history and trends
2.3. Perspectives for system thermal hydraulics
2.4. Conclusions
Exercises and questions
3. Parameters and concepts in nuclear thermal hydraulics
Foreword
3.1. General remarks
3.2. Concepts involved in two-phase flow development
3.3. Concepts involved in heat transfer developments
3.4. Concepts involved in target phenomena
3.5. Analytical tools
3.6. Verification and validation
3.7. Concepts connected to design and licensing bases
3.8. General safety concepts
3.9. Concepts related to deterministic safety assessment
3.10. Concepts connected to actual accidents
3.11. Concepts related to other uses of thermal-hydraulic analysis
Exercises and questions
4. Role of thermal hydraulics in nuclear power plant design and safety
Foreword
4.1. Introduction
4.2. The thermal-hydraulic analysis envelope
4.3. Demonstration of adequacy of thermal-hydraulic analysis tool
4.4. Final remarks and connections with other chapters of the book
Exercises and questions
5. Balance equations
Foreword
5.1. Introduction to balance equations
5.2. The principles of thermodynamics and mechanics
5.3. Various approaches to single-phase modeling
5.4. Time-averaged single-phase equations
5.5. Space-time averaged 1-D equations for single-phase flows
5.6. Various approaches to two-phase flow modeling
5.7. Time-averaged two-fluid equations
5.8. Space averaged one-dimensional equations for two-phase flow
5.9. Time-space-averaged one-dimensional equations for two-phase flow
5.10. One-dimensional two-fluid equations in system codes
5.11. Simplified two-phase models: homogeneous equilibrium drift-flux model
5.12. Characteristic velocities and hyperbolicity
5.13. Transport of interfacial area and polydispersion models
5.14. Three-field models in system codes
5.15. Summary remarks
Exercises and questions
6. Target phenomena in nuclear thermal hydraulics
Dedication
Foreword
6.1. Introduction
6.2. Classification of thermal-hydraulic phenomena for separate effects tests
6.3. Characterization of thermal-hydraulic phenomena for separate effects tests
6.4. Thermal-hydraulic phenomena for integral facilities (pressurized water reactors, boiling water reactors, and Russian pressurized water-cooled and water-moderated energy reactors)
6.5. Thermal-hydraulic phenomena related to advanced water-cooled reactors
6.6. Thermal-hydraulic phenomena for supercritical water-cooled reactors
6.7. Phenomena in the CANDU and RBMK research reactors
6.8. Phenomena in core subchannels and containments
6.9. Conclusions
Exercises and questions
7. Heat transfer in nuclear thermal hydraulics
Foreword
Part 1: Principles and modeling
Part 2: Boiling heat transfer
8. Pressure drops in nuclear thermal hydraulics—Principles, experiments, and modeling
Foreword
8.1. Introduction
8.2. Pressure drop components
8.3. Overview of pressure drop models and related phenomena
8.4. Relations among pressure drops, phenomena, and accident scenarios
8.5. Modeling pressure drops and applications in thermal-hydraulic system codes
8.6. Measurement techniques
8.7. Envisaged developments in pressure drops
8.8. Conclusions
Exercises and questions
9. Constitutive equations
Foreword
9.1. Introduction
9.2. Two-fluid equations and constitutive equations
9.3. Flow regime map
9.4. Interfacial heat and mass transfer
9.5. Interfacial drag
9.6. Wall heat transfer
9.7. Wall friction
9.8. Two-fluid three-field equations and additional constitutive equations
9.9. Validation of constitutive equations
9.10. Summary remarks
Exercises and questions
List of symbols and abbreviations
Questions and answers from Chapters 1 to 9
References
Index

Francesco D'Auria

Francesco D’Auria is a Professor at the University of Pisa, Italy, and Founder and Head of the University’s San Piero a Grado Nuclear Research Group. He was Founder and Editor-in-Chief of the J. Science & Technology of Nuclear Installations (Hindawi) and has been a member of the Advisory Editorial Board of the J. Nuclear Engineering and Design (Elsevier) since 1990. He has been Chairman and/or Organizer of almost 100 sessions in Conferences since 1982 and was recently General Chairman of the ANS NURETH-15 Conference held in Pisa in 2013. He is a Member of the Technical Review Committee of the JANSI (Japan Nuclear Safety Institute), Member of various international groups in IAEA and OECD/NEA, Chair of the ENS TopSafe Conferences of 2008 and 2012 and forthcoming 2017, as well as Foreign Member of the Academy of Science in Buenos Aires. He received several recognitions including the gold medal ‘N. A. Dollezhal’ from NIKIET in Russia. D’Auria is co-author of more than 900 papers (more than 150 of which have been published in International Journals) and more than 1000 reports.

Affiliations and expertise

Professor, University of Pisa, Italy; Founder and Head, University’s San Piero a Grado Nuclear Research Group

Yassin A. Hassan

Dr. Yassin Hassan is the Department Head and Professor at Texas A&M University. He received his master’s and doctorate in nuclear in nuclear engineering from the University of Illinois, and a master’s in mechanical engineering from the University of Virginia. He is a fellow of the American Association for the Advancement of Science, the American Society of Mechanical of Mechanical Engineers and American Nuclear Society (ANS). In 2007, Hassan was sworn in as a part-time technical judge to the Atomic Safety and Licensing Board Panel of the U.S. Nuclear Regulatory Commission. He has authored more than 490 refereed publications in technical journals and conference proceedings (195 technical journals and 295 conferences) and more than 380 summaries in American Nuclear Society transactions (the largest for any ANS member since its establishment in 1950). He also serves as editor-in-chief of the Journal of Nuclear Engineering and Design and is associate editor of the Journal of Verification, Validation and Uncertainty Quantification. He is elected to National Academy of Engineering in 2019.

Affiliations and expertise

Professor and Department Head, Texas A&M University, USA

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