Thermal Hydraulics of Water-Cooled Nuclear Reactors reviews flow and heat transfer phenomena in nuclear systems and examines the critical contribution of this analysis to nuclear technology development. With a strong focus on system thermal hydraulics (SYS TH), the book provides a detailed, yet approachable, presentation of current approaches to reactor thermal hydraulic analysis, also considering the importance of this discipline for the design and operation of safe and efficient water-cooled and moderated reactors.
Part One presents the background to nuclear thermal hydraulics, starting with a historical perspective, defining key terms, and considering thermal hydraulics requirements in nuclear technology. Part Two addresses the principles of thermodynamics and relevant target phenomena in nuclear systems. Next, the book focuses on nuclear thermal hydraulics modeling, covering the key areas of heat transfer and pressure drops, then moving on to an introduction to SYS TH and computational fluid dynamics codes.
The final part of the book reviews the application of thermal hydraulics in nuclear technology, with chapters on V&V and uncertainty in SYS TH codes, the BEPU approach, and applications to new reactor design, plant lifetime extension, and accident analysis.
This book is a valuable resource for academics, graduate students, and professionals studying the thermal hydraulic analysis of nuclear power plants and using SYS TH to demonstrate their safety and acceptability.
1. 1: Introduction - p. xx 2. 2: A historical perspective of nuclear thermal-hydraulics - p. 40 3. 3: Parameters and concepts - p. 87 4. 4: Role of thermal-hydraulics in nuclear power plants: Design and safety - p. 141 5. Part Two: Thermal-hydraulics connections: The principles of thermodynamics and nuclear systems - p. 166 6. 5: Balance equations - p. 166 7. Abstract - p. 167 8. Chapter foreword - p. 173 9. 5.4 The time-averaged single-phase equations - p. 184 10. 5.5 The space-time-averaged 1-D equations for single-phase flows - p. 188 11. 5.6 The various approaches to two-phase flow modeling - p. 196 12. 5.7 Time-averaged two-fluid equations - p. 203 13. 5.8 The space averaged 1-D equations for two-phase flow - p. 208 14. 5.9 The time-space averaged 1-D equations for two-phase flow - p. 221 15. 5.10 The 1-D two-fluid equations as used in system codes - p. 222 16. 5.11 Some simplified two-phase models: HEM drift-flux model - p. 223 17. 5.12 Characteristic velocities and hyperbolicity - p. 231 18. 5.13 Transport of interfacial area and polydispersion models - p. 234 19. 5.14 3-Field models in system codes - p. 238 20. 5.15 Summary remarks - p. 241 21. 6: Target phenomena in nuclear thermal-hydraulics - p. 244 22. Abstract - p. 245 23. Acknowledgments - p. 245 24. Chapter foreword - p. 245 25. 6.5 Thermal-hydraulic phenomena related to advanced water-cooled reactors - p. 332 26. 6.6 Thermal-hydraulic phenomena for SCWRs - p. 344 27. 6.7 Conclusions - p. 351 28. Part Three: Nuclear thermal hydraulics modeling - p. 356 29. 7: Heat transfer in nuclear thermal hydraulics - p. 356 30. Abstract - p. 357 31. Acknowledgments - p. 357 32. Chapter foreword - p. 357 33. Part 1: Principles and modeling - p. 360 34. Part 2: Boiling Heat Transfer - p. 420 35. 8: Pressure drops in nuclear thermal-hydraulics: Principles, experiments, and modeling - p. 492 36. 9: Constitutive equations - p. 547 37. 10: Special models for nuclear thermal hydraulics - p. 594 38. 11: The structure of system thermal-hydraulic (SYS-TH) code for nuclear energy applications - p. 638 39. 12: An overview of computational fluid dynamics and nuclear applications - p. 727 40. Part Four: Applications of thermal-hydraulics in nuclear technology - p. 829 41. 13: Verification and validation of system thermal-hydraulic computer codes, scaling and uncertainty evaluation of calculated code results - p. 829 42. 14: Best-Estimate Plus Uncertainty (BEPU) approach for accident analysis - p. 903 43. 15: Applications of SYS TH codes to nuclear reactor design and accident analysis - p. 950 44. 16: Thermal-hydraulics aspects of key nuclear accidents - p. 1097 45. Subject Index - p. 1153 46. Author Index - p. pge1
Thermal Hydraulics of Water-Cooled Nuclear Reactors reviews flow and heat transfer phenomena in nuclear systems and examines the critical contribution of this analysis to nuclear technology development. With a strong focus on system thermal hydraulics (SYS TH), the book provides a detailed, yet approachable, presentation of current approaches to reactor thermal hydraulic analysis, also considering the importance of this discipline for the design and operation of safe and efficient water-cooled and moderated reactors.
Part One presents the background to nuclear thermal hydraulics, starting with a historical perspective, defining key terms, and considering thermal hydraulics requirements in nuclear technology. Part Two addresses the principles of thermodynamics and relevant target phenomena in nuclear systems. Next, the book focuses on nuclear thermal hydraulics modeling, covering the key areas of heat transfer and pressure drops, then moving on to an introduction to SYS TH and computational fluid dynamics codes.
The final part of the book reviews the application of thermal hydraulics in nuclear technology, with chapters on V&V and uncertainty in SYS TH codes, the BEPU approach, and applications to new reactor design, plant lifetime extension, and accident analysis.
This book is a valuable resource for academics, graduate students, and professionals studying the thermal hydraulic analysis of nuclear power plants and using SYS TH to demonstrate their safety and acceptability.
1. 1: Introduction - p. xx 2. 2: A historical perspective of nuclear thermal-hydraulics - p. 40 3. 3: Parameters and concepts - p. 87 4. 4: Role of thermal-hydraulics in nuclear power plants: Design and safety - p. 141 5. Part Two: Thermal-hydraulics connections: The principles of thermodynamics and nuclear systems - p. 166 6. 5: Balance equations - p. 166 7. Abstract - p. 167 8. Chapter foreword - p. 173 9. 5.4 The time-averaged single-phase equations - p. 184 10. 5.5 The space-time-averaged 1-D equations for single-phase flows - p. 188 11. 5.6 The various approaches to two-phase flow modeling - p. 196 12. 5.7 Time-averaged two-fluid equations - p. 203 13. 5.8 The space averaged 1-D equations for two-phase flow - p. 208 14. 5.9 The time-space averaged 1-D equations for two-phase flow - p. 221 15. 5.10 The 1-D two-fluid equations as used in system codes - p. 222 16. 5.11 Some simplified two-phase models: HEM drift-flux model - p. 223 17. 5.12 Characteristic velocities and hyperbolicity - p. 231 18. 5.13 Transport of interfacial area and polydispersion models - p. 234 19. 5.14 3-Field models in system codes - p. 238 20. 5.15 Summary remarks - p. 241 21. 6: Target phenomena in nuclear thermal-hydraulics - p. 244 22. Abstract - p. 245 23. Acknowledgments - p. 245 24. Chapter foreword - p. 245 25. 6.5 Thermal-hydraulic phenomena related to advanced water-cooled reactors - p. 332 26. 6.6 Thermal-hydraulic phenomena for SCWRs - p. 344 27. 6.7 Conclusions - p. 351 28. Part Three: Nuclear thermal hydraulics modeling - p. 356 29. 7: Heat transfer in nuclear thermal hydraulics - p. 356 30. Abstract - p. 357 31. Acknowledgments - p. 357 32. Chapter foreword - p. 357 33. Part 1: Principles and modeling - p. 360 34. Part 2: Boiling Heat Transfer - p. 420 35. 8: Pressure drops in nuclear thermal-hydraulics: Principles, experiments, and modeling - p. 492 36. 9: Constitutive equations - p. 547 37. 10: Special models for nuclear thermal hydraulics - p. 594 38. 11: The structure of system thermal-hydraulic (SYS-TH) code for nuclear energy applications - p. 638 39. 12: An overview of computational fluid dynamics and nuclear applications - p. 727 40. Part Four: Applications of thermal-hydraulics in nuclear technology - p. 829 41. 13: Verification and validation of system thermal-hydraulic computer codes, scaling and uncertainty evaluation of calculated code results - p. 829 42. 14: Best-Estimate Plus Uncertainty (BEPU) approach for accident analysis - p. 903 43. 15: Applications of SYS TH codes to nuclear reactor design and accident analysis - p. 950 44. 16: Thermal-hydraulics aspects of key nuclear accidents - p. 1097 45. Subject Index - p. 1153 46. Author Index - p. pge1
