Generic assessment of delayed reactor coolant pump trip during small break loss-of-coolant accidents in pressurized water reactors

목차 (Table of Contents)

• CONTENTS
• Abstract = ⅱ
• Table of Contents = ⅲ
• List of Figures = ⅳ
• List of Tables = ⅴ
• 1.0 Introduction = 1
• 2.0 Summary and Conclusions = 1
• 3.0 Background = 2
• 3.1 Bulletins 79-05A, 79-06A, 79-06B = 2
• 3.2 Analysis Results = 3
• 3.3 Bulletins 79-05C, 79-06C = 6
• 3.3.1 Short-Term Actions = 6
• 3.3.2 Long-Term Actions = 7
• 3.3.3 Schedule = 7
• 3.4 Summary of Responses to Bulletins Short-Term Items 1-4 and Long-Term Item = 7
• 4.0 Analysis Methods and Results = 8
• 4.1 Previous Analytical Studies = 8
• 4.2 General Description of Thermal-Hydraulic Phenomena With Delayed Pump Trip = 9
• 4.2.1 Case 1 : Immediate Pump Trip = 9
• 4.2.2 Case 2 : No Pump Trip = 9
• 4.2.3 Case 3 : Delayed Pump Trip = 10
• 4.3 General Modeling Methods = 10
• 4.3.1 Two-Phase Equilibrium Modeling = 10
• 4.3.2 Homogeneous Modeling = 10
• 4.3.3 Heterogeneous(Phase Separation) Modeling = 10
• 4.3.4 ECC Injection = 13
• 4.3.5 Pump Performance and Discharge = 13
• 4.3.6 Hot Leg and Steam Generator Hydraulic Behavior = 19
• 4.3.7 Quench Behavior During Core Recovery = 20
• 4.3.8 Steam Superheat and Core Fluid Quality = 20
• 4.3.9 Initial Condition Uncertainty = 23
• 4.4 Analysis Results = 23
• 4.4.1 Break Size-Pump Trip Delay Time = 23
• 4.4.2 Break Location = 27
• 4.4.3 Continuous Pump Operational = 28
• 4.4.3.1 All Pumps Operating = 28
• 4.4.3.2 One Pump per Loop for CE 2x4 Plants = 28
• 4.4.4 Plant Types = 28
• 4.4.5 HPI Operational Status = 29
• 4.4.6 Best-Estimate Analyses = 29
• 4.5 Staff Audit Calculations = 30
• 4.6 Experimental Verification = 32
• 4.6.1 Present Status = 32
• 4.6.2 Future Plans = 32
• 5.0 Impact on Non-LOCA Transients = 32
• 5.1 Transients Which Produce Primary System Depressurization = 33
• 5.2 Vendor Analyses = 33
• 6.0 Staff Review and Conclusions = 34
• 7.0 Requirements for Pump Trip = 36
• 7.1 Pump Trip Criteria = 36
• 7.1.1 Low Pressure = 36
• 7.1.2 Low Pump Current = 36
• 7.1.3 Safety Injection Actuation = 36
• 7.1.4 Time Delay = 36
• 7.2 Short-Term Requirements = 37
• 7.2.1 Control Room Operators = 37
• 7.2.2 Westinghouse-Designed Plants = 37
• 7.2.3 Combustion Engineering-Designed Plants = 37
• 7.2.4 Babcock and Wilcox-Designed Plants = 38
• 7.2.5 Training Guidelines and Emergency Procedures = 38
• 7.3 Long-Term Requirements = 39
• 7.3.1 Schedule = 39
• 7.3.2 Automatic Reactor Coolant Pump Trip Requirement = 39
• 8.0 References = 40
• Appendix A - Failure of Natural Circulation During the Three Mile Island Accident = A-1
• Appendix B - Office of Inspection and Enforcement Bulletins 79-05A, 79-06A, and 79-06B = B-1
• Appendix C - Letter from Babcock and Wilcox to All B&W Operating Plants on Revised Operating Guidelines for Reactor Coolant Pumps Upon Receipt of ESFAS Actuation = C-1
• Appendix D - Office of Inspection and Enforcement Bulletins 79-05C and 79-06C = D-1