ML18117A430
| ML18117A430 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/06/2018 |
| From: | Exelon Generation Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18117A343 | List:
|
| References | |
| TMI-18-047 | |
| Download: ML18117A430 (11) | |
Text
TMI-1 UFSAR CHAPTER 14 14-i REV. 24, APRIL 2018 CHAPTER 14 - SAFETY ANALYSIS TABLE OF CONTENTS SECTION TITLE 14.0 SAFETY ANALYSIS 14.1 CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS 14.1.1 ABNORMALITIES 14.1.2 ANALYSIS OF EFFECTS AND CONSEQUENCES 14.1.2.1 UNCOMPENSATED OPERATING REACTIVITY CHANGES 14.1.2.2 STARTUP ACCIDENT 14.1.2.3 ROD WITHDRAWAL ACCIDENT AT RATED POWER OPERATION 14.1.2.4 MODERATOR DILUTION ACCIDENT 14.1.2.5 COLD WATER ACCIDENT 14.1.2.6 LOSS OF COOLANT FLOW 14.1.2.7 STUCK-OUT, STUCK-IN, OR DROPPED CONTROL ROD ACCIDENT 14.1.2.8 LOSS OF ELECTRIC POWER 14.1.2.9 STEAM LINE BREAK 14.1.2.10 STEAM GENERATOR TUBE FAILURE 14.1.2.11 ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) 14.2 STANDBY SAFEGUARDS ANALYSIS 14.2.1 SITUATIONS ANALYZED AND CAUSES 14.2.2 ACCIDENT ANALYSES 14.2.2.1 FUEL HANDLING ACCIDENT 14.2.2.2 ROD EJECTION ACCIDENT 14.2.2.3 LARGE BREAK LOSS OF COOLANT ACCIDENT 14.2.2.3.1 IDENTIFICATION OF CAUSES 14.2.2.3.2 LARGE BREAK LOCA ANALYSIS 14.2.2.3.3 RESULTS - LARGE BREAK LOCA 14.2.2.3.4 ENVIRONMENTAL ANALYSIS OF LOSS OF COOLANT ACCIDENTS 14.2.2.3.5 POST ANALYSIS-OF-RECORD EVALUATIONS FOR LARGE BREAK LOCA 14.2.2.4 SMALL BREAK LOSS OF COOLANT ACCIDENT 14.2.2.4.1 IDENTIFICATION 14.2.2.4.2 SMALL BREAK LOCA ANALYSIS 14.2.2.4.3 RESULTS - SMALL BREAK LOCA 14.2.2.4.4 POST ANALYSIS-OF-RECORD EVALUATIONS FOR SMALL BREAK LOCA 14.2.2.5 MAXIMUM HYPOTHETICAL ACCIDENT 14.2.2.6 WASTE GAS TANK RUPTURE 14.2.2.7 LOSS OF FEEDWATER ACCIDENT 14.2.2.8 FUEL CASK DROP ACCIDENT 14.2.2.9 FEEDWATER LINE BREAK ACCIDENT 14.2.2.9.1 FEEDWATER LINE BREAK OUTSIDE CONTAINMENT 14.2.2.9.2 FEEDWATER LINE BREAK INSIDE CONTAINMENT
14.3 REFERENCES
TMI-1 UFSAR CHAPTER 14 14-ii REV. 24, APRIL 2018 LIST OF TABLES TABLE TITLE 14.0-1 EQUIPMENT AND RELATED SYSTEMS ASSUMED TO FUNCTION DURING ACCIDENT ANALYSIS 14.1-1 ABNORMALITIES AFFECTING CORE AND COOLANT BOUNDARY 14.1-2 UNCOMPENSATED REACTIVITY DISTURBANCES 14.1-3 STARTUP ACCIDENT PARAMETER 14.1-4
SUMMARY
OF STARTUP ACCIDENT ANALYSIS 14.1-5 ROD WITHDRAWAL ACCIDENT PARAMETERS 14.1-6
SUMMARY
OF ROD WITHDRAWAL ACCIDENT ANALYSIS 14.1-7 MODERATOR DILUTION ACCIDENT PARAMETERS 14.1-8
SUMMARY
OF MODERATOR DILUTION ACCIDENT ANALYSIS 14.1-9 LOSS OF COOLANT FLOW ACCIDENT PARAMETERS 14.1-10 LOCKED ROTOR ACCIDENT PARAMETERS 14.1-11
SUMMARY
OF LOSS OF COOLANT FLOW ACCIDENT ANALYSIS 14.1-11a DELETED 14.1-12 NATURAL CIRCULATION CAPABILITY 14.1-13 DROPPED ROD ACCIDENT PARAMETERS 14.1-14 LOSS OF LOAD TRANSIENT PARAMETERS AND RESULTS 14.1-14a LOSS OF ELECTRICAL LOAD PARAMETERS AND RESULTS 14.1-15 LOSS OF ALL A-C POWER EVENT (STATION BLACKOUT) RADIOLOGICAL ANALYSIS PARAMETERS AND RESULTS 14.1-15a DELETED 14.1-16 STEAM LINE BREAK PARAMETERS 14.1-17 DELETED
TMI-1 UFSAR CHAPTER 14 14-iii REV. 24, APRIL 2018 LIST OF TABLES (cont'd) 14.1-18 RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE BREAK ACCIDENT IN CONJUNCTION WITH ACCIDENT-INDUCED STEAM GENERATOR TUBE LEAK (REM) 14.1-19 DELETED 14.1-20 STEAM GENERATOR TUBE FAILURE PARAMETERS 14.1-21
SUMMARY
OF STEAM GENERATOR TUBE FAILURE ANALYSIS 14.1-22 COLD WATER ACCIDENT PARAMETERS 14.1-23 DELETED 14.1-24 DELETED 14.2-1 SITUATIONS ANALYZED AND CAUSES 14.2-2 RADIOACTIVE RELEASE FOR THE FUEL HANDLING ACCIDENT 14.2-3 FUEL HANDLING ACCIDENT PARAMETERS AND RESULTS 14.2-4 FISSION PRODUCT INVENTORIES FOR THE CORE, THE AVERAGE ASSEMBLY, AND THE REACTOR COOLANT SYS 14.2-4a FISSION PRODUCT CORE INVENTORY FOR FHA IN REACTOR BUILDING AND LOCA 14.2-5 RADIOACTIVE RELEASE FOR THE POSTULATED FUEL HANDLING ACCIDENT AND DOSE RESULTS (IN THE REACTOR BUILDING) 14.2-6 ROD EJECTION ACCIDENT PARAMETERS (INITIAL CYCLE) 14.2-7 NOMINAL VALUES OF INPUT PARAMETERS FOR ROD EJECTION ACCIDENT ANALYSIS 14.2-8 COMPARISON OF SPACE-DEPENDENT AND POINT KINETICS RESULTS ON THE FUEL ENTHALPY (INITIAL CYCLE) 14.2-9
SUMMARY
OF ROD EJECTION ACCIDENT ANALYSIS 14.2-10 REACTOR VESSEL PARAMETERS 14.2-11 ENVIRONMENTAL EFFECTS OF ROD EJECTION ACCIDENT 14.2-11 REACTOR VESSEL INTERNALS - DISSIMILAR METALS
TMI-1 UFSAR CHAPTER 14 14-iv REV. 24, APRIL 2018 LIST OF TABLES (cont'd) 14.2-13 ASSUMPTIONS AND RESULTS OF CONTROL ROD TEMPERATURE ANALYSIS 14.2-14 LOSS OF COOLANT ACCIDENT ANALYSIS 14.2-15 DELETED 14.2-16 SEQUENCE OF EVENTS OF STUCK OPEN PORV ACCIDENT 14.2-17 DELETED 14.2-18 DELETED 14.2-19 DELETED 14.2-20 ENVIRONMENTAL DOSES RESULTING FROM MHA 14.2-21 WASTE GAS TANK INVENTORY 14.2-22 KEY INPUT PARAMETERS FOR LOFW ANALYSIS 14.2-23 DELETED 14.2-24 DELETED 14.2-25 NOBLE GAS AND IODINE GAP ACTIVITY (10 FUEL ASSEMBLIES) 120 DAY DECAY 14.2-26 SITE BOUNDARY DOSES 14.2-27 LOW PRESSURE INJECTION FLOW VS CORE FLOOD/LPI NOZZLE PRESSURE 14.2-28 HIGH PRESSURE INJECTION FLOW VERSUS HPI NOZZLE PRESSURE 14.2-29
SUMMARY
OF ANALYSIS INPUT VALUES
TMI-1 UFSAR CHAPTER 14 14-v REV. 24, APRIL 2018 LIST OF FIGURES FIGURE TITLE 14.1-1 STARTUP ACCIDENT FROM 10-9 RATED POWER USING A 1.5 PERCENT delta-k/k ROD GROUP; HIGH PRESSURE REACTOR TRIP IS ACTUATED 14.1-2 STARTUP ACCIDENT FROM 10-9 RATED POWER USING ALL RODS WITH A WORTH OF 10 PERCENT delta-k/k; HIGH FLUX REACTOR TRIP IS ACTUATED 14.1-3 PEAK THERMAL POWER VS ROD WITHDRAWAL RATE FOR A STARTUP ACCIDENT FROM 10-9 RATED POWER 14.1-4 PEAK NEUTRON POWER VERSUS ROD WITHDRAWAL RATE FOR A STARTUP ACCIDENT FROM 10-9 RATED POWER 14.1-5 PEAK THERMAL POWER VERSUS MODERATOR COEFFICIENT FOR A STARTUP ACCIDENT USING A 1.5 PERCENT delta-k/k ROD GROUP AT 1.09 X 10-4 (delta-k/k)/S FROM 10-9 RATED POWER 14.1-6 PEAK THERMAL POWER VERSUS MODERATOR COEFFICIENT FOR A STARTUP ACCIDENT USING A 1.5 PERCENT delta-k/k ROD GROUP AT 1.09 X 10-4(delta-k/k)/S FROM 10-9 RATED POWER 14.1-7 PEAK THERMAL POWER VERSUS DOPPLER COEFFICIENT FOR A STARTUP ACCIDENT USING ALL RODS AT 7.25 X 10-4 (delta-k/k)/S FROM 10-9 RATED POWER 14.1-8 PEAK THERMAL POWER VERSUS MODERATOR COEFFICIENT FOR A STARTUP ACCIDENT USING ALL RODS AT 7.25 X 10-4 (delta-k/k)/S FROM 10-9 RATED POWER 14.1-9 ROD WITHDRAWAL ACCIDENT FROM RATED POWER USING A 1.5 PERCENT delta-k/k ROD GROUP AT 1.09 X 10-4 (delta-k/k)/S; HIGH FLUX REACTOR TRIP IS ACTUATED 14.1-10 PEAK PRESSURE VERSUS ROD WITHDRAWAL RATE FOR A ROD WITHDRAWAL ACCIDENT FROM RATED POWER 14.1-11 PEAK PRESSURE VERSUS TRIP DELAY TIME FOR A ROD WITHDRAWAL ACCIDENT FROM RATED POWER USING A1.5 PERCENT delta-k/k ROD GROUP 14.1-12 PEAK PRESSURE VERSUS DOPPLER COEFFICIENT FOR A ROD WITHDRAWAL ACCIDENT FROM RATED POWER USING A 1.5 PERCENT delta-k/k ROD GROUP
TMI-1 UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE CHAPTER 14 14-vi REV. 24, APRIL 2018 14.1-13 PEAK PRESSURE VERSUS MODERATOR COEFFICIENT FOR A ROD WITHDRAWAL ACCIDENT FROM RATED POWER USING A 1.5 PERCENT delta-k/k ROD GROUP 14.1-14 MAXIMUM NEUTRON AND THERMAL POWER FOR AN ALL-ROD WITHDRAWAL ACCIDENT FROM VARIOUS INITIAL POWER LEVELS 14.1-15 PEAK FUEL TEMPERATURE IN AVERAGE ROD AND HOT SPOT FOR AN ALL-ROD WITHDRAWAL ACCIDENT FROM VARIOUS INITIAL POWER LEVELS 14.1-16 PUMP STARTUP FROM 50 PERCENT POWER AND 50 PERCENT FLOW 14.1-17 PERCENT REACTOR COOLANT FLOW AS A FUNCTION OF TIME AFTER LOSS OF PUMP POWER 14.1-18 DNB RATIO VERSUS TIME FOR A 4 PUMP COASTDOWN ACCIDENT FROM 102% OF RATED POWER 14.1-19 DNB RATIO VERSUS TIME FOR A LOCKED ROTOR ACCIDENT FROM 102 PERCENT OF RATED POWER 14.1-20 0.46 PERCENT delta-k/k ROD DROP FROM RATED POWER WITH AUTOMATIC RUNBACK TO 60 PERCENT DEMAND IN 12s 14.1-21 0.36 PERCENT delta-k/k ROD DROP FROM RATED POWER WITH AUTOMATIC RUNBACK TO 60 PERCENT DEMAND IN 12s 14.1-22A DOUBLE-ENDED RUPTURE OF 24 IN STEAM LINE BETWEEN STEAM GENERATOR AND STEAM STOP VALVE (with Feedwater Isolation) 14.1-22B DOUBLE-ENDED RUPTURE OF 24 IN STEAM LINE BETWEEN STEAM GENERATOR AND STEAM STOP VALVE (with Feedwater Isolation) 14.1-23 RCS HOT LEG PRESSURE (psia) RESPONSE FOR LOSS OF ALL AC POWER (STATION BLACKOUT) 14.1-24 DROPPED ROD W/ BAYONET FAILURE - RCS PRESSURE RESPONSE 14.1-25 DROPPED ROD W/ BAYONET FAILURE - REACTOR POWER RESPONSE 14.1-26 DROPPED ROD W/ BAYONET FAILURE - PRIMARY TEMPERATURE
RESPONSE
14.2-1 PEAK NEUTRON POWER VARIATION WITH EJECTED CONTROL ROD WORTH
TMI-1 UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE CHAPTER 14 14-vii REV. 24, APRIL 2018 14.2-2 PEAK THERMAL POWER AS A FUNCTION OF EJECTED CONTROL ROD WORTH 14.2-3 PEAK ENTHALPY OF HOTTEST FUEL ROD VERSUS EJECTED CONTROL ROD WORTH 14.2-4 EFFECT ON PEAK NEUTRON POWER OF VARYING THE DOPPLER COEFFICIENT FOR AN EJECTED ROD WORTH OF 0.56 PERCENT delta-k/k AT 10-9 RATED POWER AND 0.46 PERCENT delta-k/k AT RATED POWER 14.2-5 EFFECT ON PEAK THERMAL POWER OF VARYING THE DOPPLER COEFFICIENT FOR AN EJECTED ROD WORTH OF 0.56 PERCENT delta-k/k AT 10-9 RATED POWER AND 0.46 PERCENT delta-k/k AT RATED POWER 14.2-6 EFFECT ON PEAK NEUTRON POWER OF VARYING THE MODERATOR COEFFICIENT FOR AN EJECTED ROD WORTH OF 0.56 PERCENT delta-k/k AT 10-3 RATED POWER AND 0.46 PERCENT delta-k/k AT RATED POWER 14.2-7 EFFECT ON PEAK THERMAL POWER OF VARYING THE MODERATOR COEFFICIENT FOR AN EJECTED ROD WORTH OF 0.56 PERCENT delta-k/k AT 10-3 RATED POWER AND 0.46 PERCENT delta-k/k AT RATED POWER 14.2-8 EFFECT ON PEAK THERMAL POWER OF VARYING THE TRIP DELAY TIME FOR AN EJECTED ROD WORTH OF 0.56 PERCENT delta-k/k AT 10-9 RATED POWER AND 0.46 PERCENT delta-k/k AT RATED POWER 14.2-9 PERCENT CORE EXPERIENCING DNB AS A FUNCTION OF EJECTED CONTROL ROD WORTH AT RATED POWER, BOL 14.2-10 LARGE BREAK ANALYSIS CODE INTERFACES 14.2-11 HOT SPOT CLAD TEMPERATURE VS TIME FOR A 36 IN ID, DOUBLE-ENDED, HOT LEG PIPE RUPTURE AND VARIABLE QUENCH COEFFICIENT 14.2-12 LBLOCA LIMIT CASE (BOL) - REACTOR VESSEL UPPER PLENUM PRESSURE 14.2-13 LBLOCA LIMIT CASE (BOL) - BREAK MASS FLOW RATE 14.2-14 LBLOCA LIMIT CASE (BOL - 2.506-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK UNRUPTURED LOCATION 14.2-15 LBLOCA LIMIT CASE (BOL - 4.264-FT) - HOT CHANNEL FUEL AND CLAD TEMPERTURE AT PEAK UNRUPTURED LOCATION 14.2-16 LBLOCA LIMIT CASE (BOL - 6.021-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK UNRUPTURED LOCATION
TMI-1 UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE CHAPTER 14 14-viii REV. 24, APRIL 2018 14.2-17 LBLOCA LIMIT CASE (BOL - 7.779-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK UNRUPTURED LOCATION 14.2-18 LBLOCA LIMIT CASE (BOL - 9.536-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK UNRUPTURED LOCATION 14.2-19 LBLOCA LIMIT CASE (BOL - 2.506-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK RUPTURED LOCATION 14.2-20 LBLOCA LIMIT CASE (BOL - 4.264-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK RUPTURED LOCATION 14.2-21 LBLOCA LIMIT CASE (BOL - 6.021-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK RUPTURED LOCATION 14.2-22 LBLOCA LIMIT CASE (BOL - 7.779-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK RUPTURED LOCATION 14.2-23 LBLOCA LIMIT CASE (BOL - 9.536-FT) - HOT CHANNEL FUEL AND CLAD TEMPERATURE AT PEAK RUPTURED LOCATION 14.2-24 LOCA LIMIT AXIAL POWER SHAPES 14.2-25 SBLOCA COMPARISON OF RCS PRESSURE (0.01 - 0.06 FT2 BREAKS) 14.2-26 SBLOCA COMPARISON OF RCS PRESSURE (0.07 - 0.10 FT2 BREAKS) 14.2-27 SBLOCA COMPARISON OF RCS PRESSURE (0.15 - 0.75 FT2 BREAKS) 14.2-28 SBLOCA COMPARISON OF REACTOR VESSEL COLLAPSED LIQUID LEVEL (0.01 - 0.06 FT2 BREAKS) 14.2-29 SBLOCA COMPARISON OF REACTOR VESSEL COLLAPSED LIQUID LEVEL (0.07 - 0.10 FT2 BREAKS) 14.2-30 SBLOCA COMPARISON OF REACTOR VESSEL COLLAPSED LIQUID LEVEL (0.15 - 0.75 FT2 BREAKS) 14.2-31 SBLOCA COMPARISON OF PEAK CLAD TEMPERATURES (0.01 - 0.06 FT2 BREAKS) 14.2-32 SBLOCA COMPARISON OF PEAK CLAD TEMPERATURES (0.07 - 0.10 FT2 BREAKS) 14.2-33 SBLOCA COMPARISON OF PEAK CLAD TEMPERATURES (0.15 - 0.75 FT2 BREAKS)
TMI-1 UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE CHAPTER 14 14-ix REV. 24, APRIL 2018 14.2-34 CFT LINE BREAK SYSTEM PRESSURE 14.2-35 CFT LINE BREAK COLLAPSED LIQUID LEVELS 14.2-36 CFT LINE BREAK MIXTURE LEVELS 14.2-37 CFT LINE BREAK HOT CHANNEL CLAD TEMPERATURES 14.2-38 HPI LINE BREAK SYSTEM PRESSURE (OPERATOR ACTION) 14.2-39 HPI LINE BREAK COLLAPSED LIQUID LEVELS (OPERATOR ACTION) 14.2-40 HPI LINE BREAK MIXTURE LEVELS (OPERATOR ACTION) 14.2-41 HPI LINE BREAK HOT CHANNEL CLAD TEMPERATURES (OPERATOR ACTION) 14.2-42 HPI LINE BREAK SYSTEM PRESSURE (NO OPERATOR ACTION) 14.2-43 HPI LINE BREAK COLLAPSED LIQUID LEVELS (NO OPERATOR ACTION) 14.2-44 HPI LINE BREAK MIXTURE LEVELS (NO OPERATOR ACTION) 14.2-45 HPI LINE BREAK HOT CHANNEL CLAD TEMPERATURES (NO OPERATOR ACTION) 14.2-46 SBLOCA SPECTRUM - PCT VERSUS BREAK SIZE 14.2-47 CRAFT 2 NODING DIAGRAM FOR SMALL BREAK 14.2-48 BREAK SPECTRUM-AVERAGE SYSTEM VOID FRACTION WITH THE RC PUMPS OPERATIVE AND 2 HPI PUMPS 14.2-49 PRESSURE VS TIME - SMALL BREAKS WITH EMERGENCY FEEDWATER 14.2-50 PRESSURIZER LEVEL VS TIME - SMALL BREAKS WITH EMERGENCY FEEDWATER 14.2-51 PRESSURIZER LEVEL VS TIME FOR SMALL BREAK IN PRESSURIZER 14.2-52 0.01 FT2 COLD LEG BREAK WITH NO EFW, 2 HPIS & STUCK PORV AT 20 MIN-NODE 14 PRESSURE VS TIME 14.2-53 0.01 FT2 COLD LEG BREAK WITH NO EFW, 2 HPIS & STUCK PORV AT 20 MIN-PRESSURIZER LIQUID LEVEL
TMI-1 UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE CHAPTER 14 14-x REV. 24, APRIL 2018 14.2-54 0.01 FT2 COLD LEG BREAK WITH NO EFW, 2 HPIS & STUCK PORV AT 20 MIN-UPPER PLENUM LIQUID LEVEL 14.2-55 0.01 FT2 COLD LEG BREAK WITH NO EFW, 2 HPIS & STUCK PORV AT 20 MIN-PORV LEAK FLOW 14.2-56 0.01 FT2 COLD LEG BREAK WITH NO EFW, 2 HPIS & STUCK PORV AT 20 MIN-PORV LEAK FLOW QUALITY 14.2-57 0.01 FT2 COLD LEG BREAK WITH NO EFW, 2 HPIS & STUCK PORV AT 20 MIN-COLD LEG BREAK FLOW 14.2-58 0.01 FT2 CLOLD LEG BREAK WITH NO EFW, 2 HPIS & STUCK PORV AT 20 MIN-COLD LEG BREAK LEAK FLOW QUALITY 14.2-59 SYSTEM PRESSURE VS TIME - SMALL BREAKS W/O EMERGENCY FEEDWATER 14.2-60 PRESSURIZER LEVEL VS TIME - CLASS 3 BREAKS W/O EMERGENCY FEEDWATER 14.2-61 INTEGRATED DIRECT DOSE FOLLOWING MHA WITH 3-1/2 FT REACTOR BULDING WALL THICKNESS 14.2-62 DELETED 14.2-63 BORON DILUTION AND PLATEOUT
TMI-1 UFSAR CHAPTER 14 14-xi REV. 24, APRIL 2018 APPENDICES 14 A DESIGN REVIEW FOR CONSIDERATION OF EFFECTS OF PIPING SYSTEM BREAKS OUTSIDE CONTAINMENTS 14 B DELETED 14 C DELETED 14 D DELETED 14 E DELETED