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.
1. 10.1: Introduction - p. 2 2. 10.2: Special process models - p. 3 3. 10.3: Special components - p. 53 4. 10.4: Summary remarks - p. 68 5. Exercises and questions - p. 69 6. 11 The structure of system thermal-hydraulic code for nuclear reactor applications - p. 71 7. Foreword - p. 71 8. 11.1: Introduction to system codes - p. 71 9. 11.2: The requirements and the domain of simulation - p. 73 10. 11.3: Key features of system codes - p. 75 11. 11.4: The “nodalization” concept: Modeling of systems, components, selected phenomena, and aspects - p. 80 12. 11.5: The numeric solution methods - p. 93 13. 11.6: The relation between SYS TH code and containment - p. 100 14. 11.7: The relation between SYS TH code, component codes, and subchannel codes - p. 101 15. 11.8: Predicting break flow and choked flow - p. 102 16. 11.9: Predicting two-phase flow in horizontal pipes including stratification - p. 118 17. 11.10: The use of flow regime maps - p. 131 18. 11.11: Developing and validating closure relations - p. 133 19. 11.12: Predicting CCFL - p. 135 20. 11.13: Modeling of selected phase change occurrences - p. 137 21. 11.14: Modeling of pressure wave propagation - p. 176 22. 11.15: Modeling of reflooding in system codes - p. 185 23. 11.16: Upscaling capabilities of system codes - p. 231 24. 11.17: Predictive capabilities of SYS TH codes - p. 234 25. 11.18: Drift flux, two fluids, and TIA in system codes - p. 238 26. Exercises and questions - p. 248 27. 12 An overview of computational fluid dynamics and nuclear thermal hydraulics applications - p. 251 28. Foreword - p. 251 29. Part 1. Computational fluid dynamics for nuclear thermal hydraulics: The current overview - p. 252 30. Part 2. Insights into computational fluid dynamics for nuclear power plant applications - p. 265 31. 13 Verification and validation of system thermal-hydraulic computer codes and scaling and uncertainty evaluation of calculated code results - p. 349 32. Foreword - p. 349 33. 13.1: Introduction - p. 350 34. 13.2: Verification and validation of thermal-hydraulic system computer codes - p. 350 35. 13.3: Scaling of thermal-hydraulic phenomena and relation to system code validation - p. 364 36. 13.4: Current uncertainty methods - p. 378 37. 13.5: Summary remarks - p. 424 38. Exercises and questions - p. 426 39. 14 Best estimate plus uncertainty approach for accident analysis - p. 427 40. 15 Applications of SYS TH codes to nuclear reactor design and accident analysis - p. 475 41. Foreword - p. 475 42. 15.1: Introduction - p. 476 43. 15.2: Accident scenarios and phenomena - p. 487 44. 15.3: The cross-link between accident scenarios and phenomena - p. 498 45. 15.4: The characterization of phenomena and accident scenarios - p. 502 46. 15.4.1: Background - p. 502 47. 15.4.2: The LBLOCA and the IBLOCA - p. 503 48. 15.4.3: The SBLOCA class - p. 518 49. 15.4.4: The LOFW - p. 531 50. 15.4.5: The SBO - p. 535 51. 15.4.6: The LOOSP - p. 539 52. 15.4.7: The steam generator tube rupture - p. 543 53. 15.4.8: The MSLB and the FWLB - p. 547 54. 15.4.9: The RIA—Reactivity excursion class of accident scenarios - p. 553 55. 15.4.10: The ATWS - p. 557 56. 15.4.11: The AM procedures and selected related accident scenarios - p. 560 57. 15.4.12: Additional accident scenarios and related topics - p. 564 58. 15.5: Outcomes from the performed process - p. 610 59. 15.6: Conclusions - p. 612 60. Exercises and questions - p. 613 61. 16 Thermal hydraulic aspects of key nuclear accidents - p. 615 62. 17 Instrumentation in nuclear thermal hydraulics - p. 665 63. Foreword, background, and introductory remarks - p. 665 64. 17.1: Basic measurement techniques - p. 669 65. 17.2: Flow rate measurements - p. 671 66. 17.3: Measuring velocity - p. 676 67. 17.4: Void fraction measurements - p. 679 68. 17.5: Droplet and film thickness - p. 690 69. 17.6: Tomography - p. 691 70. 17.7: Pressure measurement - p. 693 71. 17.8: Temperature measurement - p. 697 72. 17.9: Issues and challenges in nuclear thermal hydraulic instrumentation - p. 709 73. 17.10: Summary - p. 715 74. Exercises and questions - p. 715 75. List of symbols and abbreviations - p. 717 76. Questions and answers from Chapters 10 to 17 - p. 745 77. References - p. 799 78. Index - p. 947
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.
1. 10.1: Introduction - p. 2 2. 10.2: Special process models - p. 3 3. 10.3: Special components - p. 53 4. 10.4: Summary remarks - p. 68 5. Exercises and questions - p. 69 6. 11 The structure of system thermal-hydraulic code for nuclear reactor applications - p. 71 7. Foreword - p. 71 8. 11.1: Introduction to system codes - p. 71 9. 11.2: The requirements and the domain of simulation - p. 73 10. 11.3: Key features of system codes - p. 75 11. 11.4: The “nodalization” concept: Modeling of systems, components, selected phenomena, and aspects - p. 80 12. 11.5: The numeric solution methods - p. 93 13. 11.6: The relation between SYS TH code and containment - p. 100 14. 11.7: The relation between SYS TH code, component codes, and subchannel codes - p. 101 15. 11.8: Predicting break flow and choked flow - p. 102 16. 11.9: Predicting two-phase flow in horizontal pipes including stratification - p. 118 17. 11.10: The use of flow regime maps - p. 131 18. 11.11: Developing and validating closure relations - p. 133 19. 11.12: Predicting CCFL - p. 135 20. 11.13: Modeling of selected phase change occurrences - p. 137 21. 11.14: Modeling of pressure wave propagation - p. 176 22. 11.15: Modeling of reflooding in system codes - p. 185 23. 11.16: Upscaling capabilities of system codes - p. 231 24. 11.17: Predictive capabilities of SYS TH codes - p. 234 25. 11.18: Drift flux, two fluids, and TIA in system codes - p. 238 26. Exercises and questions - p. 248 27. 12 An overview of computational fluid dynamics and nuclear thermal hydraulics applications - p. 251 28. Foreword - p. 251 29. Part 1. Computational fluid dynamics for nuclear thermal hydraulics: The current overview - p. 252 30. Part 2. Insights into computational fluid dynamics for nuclear power plant applications - p. 265 31. 13 Verification and validation of system thermal-hydraulic computer codes and scaling and uncertainty evaluation of calculated code results - p. 349 32. Foreword - p. 349 33. 13.1: Introduction - p. 350 34. 13.2: Verification and validation of thermal-hydraulic system computer codes - p. 350 35. 13.3: Scaling of thermal-hydraulic phenomena and relation to system code validation - p. 364 36. 13.4: Current uncertainty methods - p. 378 37. 13.5: Summary remarks - p. 424 38. Exercises and questions - p. 426 39. 14 Best estimate plus uncertainty approach for accident analysis - p. 427 40. 15 Applications of SYS TH codes to nuclear reactor design and accident analysis - p. 475 41. Foreword - p. 475 42. 15.1: Introduction - p. 476 43. 15.2: Accident scenarios and phenomena - p. 487 44. 15.3: The cross-link between accident scenarios and phenomena - p. 498 45. 15.4: The characterization of phenomena and accident scenarios - p. 502 46. 15.4.1: Background - p. 502 47. 15.4.2: The LBLOCA and the IBLOCA - p. 503 48. 15.4.3: The SBLOCA class - p. 518 49. 15.4.4: The LOFW - p. 531 50. 15.4.5: The SBO - p. 535 51. 15.4.6: The LOOSP - p. 539 52. 15.4.7: The steam generator tube rupture - p. 543 53. 15.4.8: The MSLB and the FWLB - p. 547 54. 15.4.9: The RIA—Reactivity excursion class of accident scenarios - p. 553 55. 15.4.10: The ATWS - p. 557 56. 15.4.11: The AM procedures and selected related accident scenarios - p. 560 57. 15.4.12: Additional accident scenarios and related topics - p. 564 58. 15.5: Outcomes from the performed process - p. 610 59. 15.6: Conclusions - p. 612 60. Exercises and questions - p. 613 61. 16 Thermal hydraulic aspects of key nuclear accidents - p. 615 62. 17 Instrumentation in nuclear thermal hydraulics - p. 665 63. Foreword, background, and introductory remarks - p. 665 64. 17.1: Basic measurement techniques - p. 669 65. 17.2: Flow rate measurements - p. 671 66. 17.3: Measuring velocity - p. 676 67. 17.4: Void fraction measurements - p. 679 68. 17.5: Droplet and film thickness - p. 690 69. 17.6: Tomography - p. 691 70. 17.7: Pressure measurement - p. 693 71. 17.8: Temperature measurement - p. 697 72. 17.9: Issues and challenges in nuclear thermal hydraulic instrumentation - p. 709 73. 17.10: Summary - p. 715 74. Exercises and questions - p. 715 75. List of symbols and abbreviations - p. 717 76. Questions and answers from Chapters 10 to 17 - p. 745 77. References - p. 799 78. Index - p. 947
