Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors

Volume 2: Modelling

2nd Edition - July 25, 2024
Editors: Francesco D'Auria, Yassin A. Hassan
Language: English
Paperback ISBN:
9 7 8 - 0 - 3 2 3 - 8 5 6 1 0 - 2
eBook ISBN:
9 7 8 - 0 - 3 2 3 - 8 5 6 1 1 - 9

Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 2, Modelling includes all new chapters which delve deeper into the topic, adding context and practical… 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 2, Modelling 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)
10. Special models for nuclear thermal hydraulics
Foreword
10.1. Introduction
10.2. Special process models
10.3. Special components
10.4. Summary remarks
Exercises and questions
11. The structure of system thermal-hydraulic code for nuclear reactor applications
Foreword
11.1. Introduction to system codes
11.2. The requirements and the domain of simulation
11.3. Key features of system codes
11.4. The “nodalization” concept: Modeling of systems, components, selected phenomena, and aspects
11.5. The numeric solution methods
11.6. The relation between SYS TH code and containment
11.7. The relation between SYS TH code, component codes, and subchannel codes
11.8. Predicting break flow and choked flow
11.9. Predicting two-phase flow in horizontal pipes including stratification
11.10. The use of flow regime maps
11.11. Developing and validating closure relations
11.12. Predicting CCFL
11.13. Modeling of selected phase change occurrences
11.14. Modeling of pressure wave propagation
11.15. Modeling of reflooding in system codes
11.16. Upscaling capabilities of system codes
11.17. Predictive capabilities of SYS TH codes
11.18. Drift flux, two fluids, and TIA in system codes
Exercises and questions
12. An overview of computational fluid dynamics and nuclear thermal hydraulics applications
Foreword
Part 1. Computational fluid dynamics for nuclear thermal hydraulics: The current overview
Part 2. Insights into computational fluid dynamics for nuclear power plant applications
13. Verification and validation of system thermal-hydraulic computer codes and scaling and uncertainty evaluation of calculated code results
Foreword
13.1. Introduction
13.2. Verification and validation of thermal-hydraulic system computer codes
13.3. Scaling of thermal-hydraulic phenomena and relation to system code validation
13.4. Current uncertainty methods
13.5. Summary remarks
Exercises and questions
14. Best estimate plus uncertainty approach for accident analysis
Foreword
14.1. Introduction
14.2. Bases and background
14.3. Elements and structure of BEPU for the accident analysis
14.4. BEPU within safety-licensing demonstration
14.5. Specific achievements and additional perspectives
14.6. Conclusions
Exercises and questions
15. Applications of SYS TH codes to nuclear reactor design and accident analysis
Foreword
15.1. Introduction
15.2. Accident scenarios and phenomena
15.3. The cross-link between accident scenarios and phenomena
15.4. The characterization of phenomena and accident scenarios
15.5. Outcomes from the performed process
15.6. Conclusions
Exercises and questions
16. Thermal hydraulic aspects of key nuclear accidents
Foreword
16.1. Introduction
16.2. An overview of nuclear accidents
16.3. The NPP systems and the nuclear events
16.4. The thermal hydraulics aspects of the nuclear events
16.5. Conclusions
Exercises and questions
17. Instrumentation in nuclear thermal hydraulics
Foreword, background, and introductory remarks
17.1. Basic measurement techniques
17.2. Flow rate measurements
17.3. Measuring velocity
17.4. Void fraction measurements
17.5. Droplet and film thickness
17.6. Tomography
17.7. Pressure measurement
17.8. Temperature measurement
17.9. Issues and challenges in nuclear thermal hydraulic instrumentation
17.10. Summary
Exercises and questions
List of symbols and abbreviations
Questions and answers from Chapters 10 to 17
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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