ML17226A117
| ML17226A117 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 07/28/2017 |
| From: | Entergy Operations |
| To: | Office of Nuclear Reactor Regulation, Office of Nuclear Material Safety and Safeguards |
| Shared Package | |
| ML17226A087 | List:
|
| References | |
| RBG-47776, RBF1-17-0089 | |
| Download: ML17226A117 (103) | |
Text
RBS USAR TABLE 15.0-1 xo14 RESULTS
SUMMARY
OF TRANSIENTS EVENTS APPLICABLE TO BWRs (For Original Rated Power 2894 Mwt) 14mx Revision 25 1 of 4 Paragraph I.D.
Figure I.D.
Description Maximum Neutron Flux
(% NBR)
Maximum Dome Pressure (psig)
Maximum Vessel Pressure (psig)
Maximum Steam Line Pressure (psig)
Maximum Core Average Surface Heat Flux
(% of Initial)
Minimum Critical Power Ratio Frequency Category (1)
Duration of Blowdown No. of Valves First Blowdown Duration of Blowdown (sec) 15.1 DECREASE IN CORE COOLANT TEMPERATURE 15.1.1 15.1-1 Loss of feedwater heater, automatic flow control 112.3 1,046 1,085 1,035 106.07 (2) a 0
0 15.1.1 15.1-2 Loss of feedwater heater, manual flow control 121.0 1,060 1,099 1,047 113.88 0.12 a
0 0
15.1.2 15.1-3 Feedwater control failure, max demand 176.3 1,191 1,222 1,188 106.45 0.11 a
16 6
15.1.3 (3) 15.1-4 Pressure regulator fail - open 104.31 1,127 1,159 1,127 100.27 (2) a 16 5
15.1.4 Inadvertent opening of safety or relief valve See Text 15.1.6 RHR shutdown cooling malfunction decreasing temp See Text 15.2 INCREASE IN REACTOR PRESSURE 15.2.1 15.2-1 Pressure regulation downscale failure 160.8 1,186 1,219 1,182 102.69 0.09 a
16 7
RBS USAR TABLE 15.0-1 (Cont) 14 (For Original Rated Power 2894 Mwt) 14 Revision 14 2 of 4 September 2001 Paragraph I.D.
Figure I.D.
Description Maximum Neutron Flux
(% NBR)
Maximum Dome Pressure (psig)
Maximum Vessel Pressure (psig)
Maximum Steam Line Pressure (psig)
Maximum Core Average Surface Heat Flux
(% of Initial)
Minimum Critical Power Ratio Frequency Category (1)
Duration of Blowdown No. of Valves First Blowdown Duration of Blowdown (sec) 15.2.2 15.2-2 Generator load rejection, bypass on 189.3 1,191 1,219 1,185 102.63 0.08 a
16 6
15.2.2 15.2-3 Generator load rejection, bypass off 237.7 1,204 1,232 1,198 104.88 0.11 a
16 7
15.2.3 15.2-4 Turbine trip, bypass on 164.1 1,189 1,217 1,184 100.92 0.07 a
16 6
15.2.3 15.2-5 Turbine trip, bypass off 216.0 1,203 1,231 1,198 103.22 0.09 a
16 7
15.2.4 15.2-6 All MSIV closure 105.15 1,178 1,207 1,174 100.10 (2) a 16 5
15.2.5 15.2-7 Loss of condenser vacuum 168.7 1,190 1,217 1,184 100.90 (2) a 16 6
15.2.6 15.2-8 Loss of auxiliary power transformer 104.2 1,171 1,186 1,170 100.05 (2) a 16 5
15.2.6 15.2-9 Loss of all grid connections 121.08 1,187 1,211 1,182 100.03 (2) a 16 8
15.2.7 15.2-10 Loss of all feedwater flow 104.2 1,046 1,085 1,035 100.06 (2) a 0
0 15.2.8 Feedwater piping break See Table 15.0-3, event 15.6.6 15.2.9 Failure of RHR shutdown cooling See Text 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE
RBS USAR TABLE 15.0-1 (Cont) 14 (For Original Rated Power 2894 Mwt) 14 Revision 14 3 of 4 September 2001 Paragraph I.D.
Figure I.D.
Description Maximum Neutron Flux %
NBR Maximum Dome Pressure (psig)
Maximum Vessel Pressure (psig)
Maximum Steam Line Pressure (psig)
Maximum Core Average Surface Heat Flux
(% of Initial)
Minimum Critical Power Ratio Frequency Category (1)
Duration of Blowdown No. of Valves First Blowdown Duration of Blowdown (sec) 15.3.1 15.3-1 Trip of one recirculation pump motor 104.2 1,047 1,085 1,036 100.0 (2) a 0
0 15.3.1 15.3-2 Trip of both recirculation pump motors 104.2 1,168 1,182 1,165 100.0 (2) a 16 5
15.3.2 15.3-3 Fast closure of one main recirc valve 104.2 1,049 1,085 1,037 100.0 (2) a 0
0 15.3.2 15.3-4 Fast closure of two main recirc valves 104.2 1,175 1,188 1,171 100.12 (2) a 0
0 15.3.3 15.3-5 Seizure of one recirculation pump 104.2 1,167 1,184 1,164 100.14 (2) c 16 5
15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1.1 RWE - Refueling See Text b
15.4.1.2 RWE - Startup See Text b
15.4.2 RWE - At power See Text a
15.4.3 Control rod mis-operation See 15.4.1 and 15.4.2 15.4.4 15.4-3 Abnormal startup of idle recirculation loop 122.8 993 1,007 987 161.14 (2) a 0
0
RBS USAR TABLE 15.0-1 (Cont) xo14 (For Original Rated Power 2894 Mwt) 14mx Revision 25 4 of 4 Paragraph I.D.
Figure I.D.
Description Maximum Neutron Flux %
NBR Maximum Dome Pressure (psig)
Maximum Vessel Pressure (psig)
Maximum Steam Line Pressure (psig)
Maximum Core Average Surface Heat Flux
(% of Initial)
Minimum Critical Power Ratio Frequency Category (1)
Duration of Blowdown No. of Valves First Blowdown Duration of Blowdown (sec) 15.4.5 15.4-4 Fast opening of one main recirc valve 472.4 998 1,009 984 149.56 (2) a 0
0 15.4.5 15.4-5 Fast opening of both main recirc valves 353.7 982 1,005 978 140.50 (2) a 0
0 15.4.7 Misplaced bundle accident See Text b
15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5.1 15.5-1 Inadvertent HPCS pump start 104.2 1,046 1,085 1,035 100.12 a
0 0
15.5.3 BWR transients See appropriate Events in 15.1 and 15.2 (1) a = moderate b = infrequent c = limiting fault (2) 'CPR <0.12 (3) The pressure regulator failure-open is no longer a postulated single failure scenario; no Ovation TCPS single failure scenarios are postulated that could result in all valves open. This specific USAR section is maintained for historical purposes.
RBS USAR TABLE 15.0-1A RESULTS
SUMMARY
OF TRANSIENTS EVENTS APPLICABLE TO BWRs Revision 25 1 of 5 Paragraph I.D.
Figure I.D.
Description Maximum Neutron Flux %
original NBR Maximum Dome Pressure (psig)
Maximum Vessel Pressure (psig)
Maximum Steam Line Pressure (psig)
Maximum Core Average Surface Heat Flux
(% of Initial)
Minimum Critical Power Ratio Frequency Category (1)
Duration of Blowdown No. of Valves First Blowdown Duration of Blowdown (sec) 15.1 DECREASE IN CORE COOLANT TEMPERATURE 15.1.1 15.1-1 Loss of feedwater heater, automatic flow control 112.3 1,046 1,085 1,035 106.07 (2) a 0
0 15.1.2 (NOTES 6,8) 15.1-2 Loss of feedwater heater, manual flow control 121.0 1,060 1,099 1,047 113.88 0.12 a
0 0
15.1.2 (NOTES 1,8) 15.1-3 Feedwater control failure, max demand 176.3 1,191 1,222 1,188 106.45 0.11 a
16 6
15.1.3 (NOTE 10) 15.1-4 Pressure regulator fail - open 105.31 1,127 1,159 1,127 100.27 (2) a 16 5
15.1.4 Inadvertent opening of safety or relief valve See Text 15.1.6 RHR shutdown cooling malfunction decreasing temp See Text 15.2 INCREASE IN REACTOR PRESSURE 15.2.1 (NOTES 2,8) 15.2-1 Pressure regulation downscale failure 160.8 1,186 1,219 1,182 102.69 0.09 a
16 7
Revision 25 2 of 5 Paragraph I.D.
Figure I.D.
Description Maximum Neutron Flux %
original NBR Maximum Dome Pressure (psig)
Maximum Vessel Pressure (psig)
Maximum Steam Line Pressure (psig)
Maximum Core Average Surface Heat Flux
(% of Initial)
Minimum Critical Power Ratio Frequency Category (1)
Duration of Blowdown No. of Valves First Blowdown Duration of Blowdown (sec) 15.2.2 15.2-2 Generator load rejection, bypass on 189.3 1,191 1,219 1,185 102.63 0.08 a
16 6
15.2.2 (NOTES 3,8) 15.2-3 Generator load rejection, bypass off 237.7 1,204 1,232 1,198 104.88 0.11 a
16 7
15.2.3 15.2-4 Turbine trip, bypass on 164.1 1,189 1,217 1,184 100.92 0.07 a
16 6
15.2.3 (NOTES 4,8) 15.2-5 Turbine trip, bypass off 216.0 1,203 1,231 1,198 103.22 0.09 a
16 7
15.2.4 (NOTES 5,9) 15.2-6 All MSIV closure 118.9 1,229 1,262 1,228 100.1 a
9 15.2.5 15.2-7 Loss of condenser vacuum 168.7 1,190 1,217 1,184 100.90 (2) a 16 6
15.2.6 15.2-8 Loss of auxiliary power transformer 104.2 1,171 1,186 1,170 100.05 (2) a 16 5
15.2.6 15.2-9 Loss of all grid connections 121.08 1,187 1,211 1,182 100.03 (2) a 16 8
15.2.7 15.2-10 Loss of all feedwater flow 104.2 1,046 1,085 1,035 100.06 (2) a 0
0 15.2.8 Feedwater piping break See Table 15.0-3, event 15.6.6 15.2.9 Failure of RHR shutdown cooling See Text 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE
Revision 17 3 of 5 Paragraph I.D.
Figure I.D.
Description Maximum Neutron Flux %
original NBR Maximum Dome Pressure (psig)
Maximum Vessel Pressure (psig)
Maximum Steam Line Pressure (psig)
Maximum Core Average Surface Heat Flux
(% of Initial)
Minimum Critical Power Ratio Frequency Category (1)
Duration of Blowdown No. of Valves First Blowdown Duration of Blowdown (sec) 15.3.1 15.3-1 Trip of one recirculation pump motor 104.2 1,047 1,085 1,036 100.0 (2) a 0
0 15.3.1 15.3-2 Trip of both recirculation pump motors 104.2 1,168 1,182 1,165 100.0 (2) a 16 5
15.3.2 15.3-3 Fast closure of one main recirc valve 104.2 1,049 1,085 1,037 100.0 (2) a 0
0 15.3.2 15.3-4 Fast closure of two main recirc valves 104.2 1,175 1,188 1,171 100.12 (2) a 0
0 15.3.3 15.3-5 Seizure of one recirculation pump 104.2 1,167 1,184 1,164 100.14 (2) c 16 5
15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1.1 RWE - Refueling See Text b
15.4.1.2 RWE - Startup See Text b
15.4.2 (NOTES 7,8)
RWE - At power See Text a
15.4.3 Control rod mis-operation See 15.4.1 and 15.4.2 15.4.4 15.4-1 Abnormal startup of idle recirculation loop 122.8 993 1,007 987 161.14 (2) a 0
0
Revision 17 4 of 5 Paragraph I.D.
Figure I.D.
Description Maximum Neutron Flux %
original NBR Maximum Dome Pressure (psig)
Maximum Vessel Pressure (psig)
Maximum Steam Line Pressure (psig)
Maximum Core Average Surface Heat Flux
(% of Initial)
Minimum Critical Power Ratio Frequency Category (1)
Duration of Blowdown No. of Valves First Blowdown Duration of Blowdown (sec) 15.4.5 15.4-2 Fast opening of one main recirc valve 472.4 998 1,009 984 149.56 (2) a 0
0 15.4.5 15.4-3 Fast opening of both main recirc valves 353.7 982 1,005 978 140.50 (2) a 0
0 15.4.7 Misplaced bundle accident See Text b
15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5.1 15.5-1 Inadvertent HPCS pump start 104.2 1,046 1,085 1,035 100.12 a
0 0
15.5.3 BWR transients See appropriate Events in 15.1 and 15.2 (1) a = moderate b = infrequent c = limiting fault (2) CPR <0.12
Revision 25 5 of 5 RBS USAR TABLE 15.0-1A (Cont)
Notes:
(1) The Feedwater Controller Failure case has been re-analyzed at 3039 MWt core power with normal and reduced feedwater temperature.
Results of the re-analyses are reported in Table 15.0-1B.
(2) Pressure Regulator Downscale Failure case has been re-analyzed at 3039 MWt core power with normal and reduced feedwater temperature. The results of the reduced feedwater temperature case are reported in Table 15.0-1B. The normal feedwater temperature case is bounded by the reduced feedwater temperature.
(3) The Generator Load Rejection with no Bypass option case has been re-analyzed at 3039 MWt core power with normal and reduced feedwater temperature, at full and partial arc turbine control valve (TCV) options. Results of the re-analyses are reported in Table 15.0-1B. The reduced feedwater temperature case is bounded by the normal feedwater temperature, and the partial arc is bounded by the full arc TCV mode of operation.
(4) The Turbine Trip with no Bypass option case has been re-analyzed at 3039 MWt core power with normal and reduced feedwater temperature, at full and partial arc turbine control valve (TCV) options. Results of the re-analyses are reported in Table 15.0-1B. The reduced feedwater temperature case is bounded by the normal feedwater temperature, and the partial arc is bounded by the full arc TCV mode of operation.
(5) DELETED (6) The Loss of Feedwater Heater event is re-analyzed at 3100 MWt core power (2% over 3039 MWt). The resulting delta CPR is 0.11.
The results of the re-analyses are reported in Table 15.0-1B. This event is described in detail in Section 15.1, and also in Appendix 15B.
(7) The RWE event is re-analyzed at 3039 MWt core power. The resulting delta CPR at full power is 0.16. The event is also described in detail in Section 15.4.
(8) This event is re-analyzed at 3091 MWt (100% TPO rated core power) per Reference 10, App. E. The results of the re-analyses are reported in Appendix 15B.
(9) This MSIV closure with position scram event is re-analyzed at 3091 MWt consistent with Cycle 19. The results of the re-analyses are reported in Section 15.2.4. The calculated change in Minimum Critical Power Ratio is bounded by the load rejection without bypass and turbine trip without bypass. As such, the MSIV closure event with position scram need not be evaluated each reload.
Note that the blowdown through the safety relief valves had not ended by the end of the simulation (~8 seconds after start of the event), but pressure is decreasing rapidly.
(10) The pressure regulator failure-open is no longer a postulated single failure scenario; no Ovation TCPS single failure scenarios are postulated that could result in all valves open. This specific USAR section is maintained for historical purposes.
RBS USAR TABLE 15.0-1B Summary of Events Analyzed at Power Uprate Conditions Revision 24 1 of 4 ANALYSIS ID ANALYSIS NAME TRANSIENT OUTPUT FILE NAMES (.*)
ODYN PID SUB EVENTS POWER
(%)
FLOW
(%)
STEAM FLOW
(%)
100P107F STANDARD 000E6_E00000_T02_ODYNV09_LRNBP 00264 7SRVOS 100.0 107.0 100.0 100P107F STANDARD 000E6_E00000_T03_ODYNV09_TTNBP 001A1 7SRVOS 100.0 107.0 100.0 100P107F STANDARD 000E6_E00000_T04_ODYNV09_FWCF 0019C 7SRVOS 100.0 107.0 100.0 100P107F STANDARD 000E6_E00000_T05_ODYNV09_PRFDS 00305 7SRVOS 100.0 107.0 100.0 100P107F (1)
STANDARD 00165_E00000_T05_ODYNV09_PRFDS 00A65 7SRVOS 100.0 107.0 88.2
Revision 14 2 of 4 September 2001 TRANSIENT NAME ODYN PID ANALYSIS ID EXPOSURE Mwd/st PEAK FLUX (N)
% ref PEAK FLUX (Q/A)
% init MAX QFUEL OFUEL PU MAX NET REACT $
DCPR G1136 DCPR B684W LRNBP 00264 100P107F E00000 429.70 116.77 0.59 0.76 0.1931 0.1191 TTNBP 001A1 100P107F E00000 407.36 114.49 0.53 0.75 0.1822 0.1022 FWCF 0019C 100P107F E00000 316.77 111.81 0.00 0.69 0.1430 0.0753 PRFDS 00305 100P107F E00000 145.65 104.81 0.00 0.29 0.0987 0.0464 PRFDS (1) 00A65 100P107X E00000 146.80 105.71 0.00 0.29 0.1180 0.0505
Revision 14 3 of 4 September 2001 G1136 B684W TRANSIENT NAME ODYN PID ANALYSIS ID EXPOSURE Mwd/st DCPRB DCPRA DCPRB DCPRA LRNBP 00264 100P107F E00000 0.2079 0.1256 TTNBP 001A1 100P107F E00000 0.1982 0.1086 FWCF 0019C 100P107F E00000 0.1574 0.0852 PRFDS 00305 100P107F E00000 0.1120 0.0566 PRFDS (1) 00A65 100P107F E00000 0.1318 0.0597
Revision 24 4 of 4 TRANSIENT NAME ODYN PID ANALYSIS ID EXPOSURE Mwd/st PEAK FLUX Q/A %
init PEAK DOME PRESSURE RATE psi/sec PEAK PRESSURE DOME psig PEAK PRESSURE P(V) psig PEAK PRESSURE P(SL) psig MIN DELTA P(UCL) psi MIN DELTA P(SSV) psi MIN DELTA P(ECL) psi LRNBP 00264 100P107F E00000 116.77 301.0 1269.8 1296.4 1265.6 78.6 203.6 TTNBP 001A1 100107F E00000 114.49 319.6 1268.2 1295.1 1264.1 79.9 204.9 FWCF 0019C 100P107F E00000 111.81 329.8 1244.2 1267.8 1242.0 107.2 232.2 PRFDS 00305 100P107F E00000 104.81 114.1 1255.8 1284.0 1253.2 91.0 216.0 PRFDS (1) 00A65 100P107X E00000 105.71 108.9 1249.3 1277.0 1247.7 98.0 223.0 (1) Second PRFDS case run with Reduced Feedwater Temperature
RBS USAR 1 of 3 August 1987 TABLE 15.0-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS
- 1. Thermal power level, MWt Warranted value Analysis value 2,894 3,015
- 2. Steam flow, lb/hr Warranted value Analysis value 12.45 x 10 6
13.07 x 10 6
- 3. Core Flow, lb/hr 84.5 x 10 6
- 4. Feedwater flow rate (1), lb/sec Warranted value Analysis value 3,458 3,631
- 5. Feedwater temperature, oF 425
- 6. Vessel dome pressure, psig 1,045
- 7. Vessel core pressure, psig 1,056
- 8. Turbine bypass capacity, % NBR 10
- 9. Core coolant inlet enthalpy, Btu/lb 529.9
- 10. Turbine inlet pressure, psig 960
- 11. Fuel lattice P 8X8R
- 12. Core average gap conductance, Btu/sec-ft 2-oF 0.1892
- 13. Core leakage flow, %
11
- 14. Required MCPR operating limit First core Reload core 1.18 1.19
- 15. MCPR safety limit First core Reload core 1.06 1.07
- 16. Doppler coefficient (-)¢/
oF Analysis data (2) 0.132
- 17. Void coefficient (-)¢/% rated voids Analysis data for power increase events (2)(4)
Analysis data for power decrease events (2) 14.0 4.0
- 18. Core average rated void fraction, %
(2) 42.53
- 19. Scram reactivity Analysis data (2)(4)
Fig. 15.0-2
- 20. Control rod drive speed, position versus time Fig. 15.0-3
- 21. Jet pump ratio, M 2.47
RBS USAR 2 of 3 August 1987 TABLE 15.0-2 (Cont)
- 22. SRV capacity, % NBR @ 1,210 psig Manufacturer Quantity installed 109.4 Crosby 16
- 23. Relief function delay, sec 0.40
- 24. Relief function response Time constant, sec 0.10
- 25. Safety function delay, sec 0.0
- 26. Safety function response Time constant, sec 0.2
- 27. Set points for SRVs Safety function, psig Relief Function, psig 1175, 1185, 1195 1205, 1215 1125, 1135, 1145 1155
- 28. Number of valve groupings simulated Safety function, no.
Relief function, no.
5 4
- 29. SRV reclosure Set point - both modes (% of set point)
Maximum safety limit (used in analysis)
Maximum operational limit 98 89
- 30. High flux trip, % NBR Analysis set point (122 x 1.042) 127.2
- 31. High pressure scram set point, psig 1,095
- 32. Vessel level trips, ft above bottom of separator skirt bottom Level 8 - (L8), ft Level 4 - (L4), ft Level 3 - (L3), ft Level 2 - (L2), ft 5.88 4.03 1.94
(-)2.86
- 33. APRM simulated thermal power trip, % NBR Analysis set point (114 x 1.042) 118.8
- 34. Time constant, sec 7
- 35. Nuclear characteristics used in ODYN simulations (4)
End of equilibrium cycle (EOEC)
- 36. Recirculation pump trip delay, sec 0.14
- 37. Recirculation pump trip inertia time constant for analysis, sec (3)
Max 5.0 Min 3.0
RBS USAR 3 of 3 August 1987 TABLE 15.0-2 (Cont)
- 38. Total steam line volume, ft 3
3,275
- 39. Pressure set point of recirculation pump trip - psig (nominal) 1,135 (1)
Includes control rod drive flow (2)
Applies only for events analyzed using model described in Reference 1 to Section 15.1.
(3)
The inertia time constant is defined by the expression:
t = 2 Jo n g To where:
t = Inertia time constant (sec)
Jo = Pump motor inertia (lb-ft) n = Rated pump speed (rps) g = Gravitational constant (ft/sec)
To = Pump shaft torque (lb-ft)
(4)
The transient analyses for RBS are based on end of equilibrium cycle (EOEC) nuclear parameters for a pre-control cell core design. These analyses results described in Chapter 15 are bounding for limiting transients relative to the expected performance of the plant at end of cycle 1 conditions for the control cell core.
RBS USAR TABLE 15.0-2A INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS(10)
Revision 17 1 of 3 Original Power 5% Power Uprate
- 1. Thermal power level, MWt Warranted value Analysis Value 2,894 3,015 3039
- 2. Steam Flow, lb/hr Warranted value Analysis Value 12.45 x 10 6
13.07 x 10 6
13,199 x 10 6
- 3. Core Flow, lb/hr 84.5 x 10 6
68.6 x 10 6 -
90.4 x 10 6 (5)
- 4. Feedwater flow rate (1), lb/sec Warranted value Analysis value 3,458 3,631 3666.4
- 5. Feedwater temperature, °F 425 425.7
- 6. Vessel dome pressure, psig 1,045 1055
- 7. Vessel core pressure, psig 1,056 1070
- 8. Turbine bypass capacity, % NBR 10 9.48
- 9. Core coolant inlet enthalpy, Btu/lb 529.9 531.2
- 10. Turbine inlet pressure, psig 960 1012 (6)
- 11. Fuel lattice P8x8R GE8x8EB GE11
- 12. Core average gap conductance, Btu/sec-ft2-°F 0.1892 0.3657 (7)
- 13. Core leakage flow, %
11 13.9 (7)
- 14. Required MCPR operating limit First core Reload core 1.18 1.19 1.32
- 15. MCPR safety limit First core Reload core 1.06 1.07 1.10
- 16. Doppler coefficient (-)¢/F Analysis data (2) 0.132 0.139 (7)
- 17. Void coefficient(-)¢/% rated voids Analysis data for power increase events (2)(4)
Analysis data for power decrease events (2) 14.0 4.0 9.94 (7)
- 18. Core average rated void fraction, %
(2) 42.53 40.01 (7)
- 19. Scram reactivity, Analysis data (2) (4)
Fig. 15.0-2 Same
- 20. Control rod drive speed, position versus time Fig. 15.0-3 Same
- 21. Jet pump ratio, M 2.47 2.48
Revision 17 2 of 3 Original Power 5% Power Uprate
- 22. Installed SRV capacity, % NBR Manufacturer Quantity installed 109.4 (8)
Crosby 16 100.2 (8) same Same
- 23. Relief function delay, sec 0.40 Same
- 24. Relief function response Time constant, sec 0.10 Same
- 25. Safety function delay, sec 0.0 Same
- 26. Safety function response Time constant, sec 0.2 Same
- 27. Set points for SRVs Safety function, psig Relief function, psig 1175, 1185, 1195 1205, 1215 1125, 1135, 1145, 1155 1231, 1241, 1246 1163, 1173, 1183
- 28. Number of valve groupings simulated Safety function, no.
Relief function, no.
5 4
3 3
- 29. SRV reclosure Set point-both modes (% of set point)
Maximum safety limit (used in analysis)
Maximum operational limit 98 89 Same Same
- 30. High flux trip, % NBR Analysis set point (122 x 1.042) 127.2 122.0
- 31. High pressure scram set point, psig 1,095 1,125
- 32. Vessel level analysis trips, ft above bottom of separator skirt Level 8 - (L8), ft Level 4 - (L4), ft Level 3 - (L3), ft Level 2 - (L2), ft 5.88 4.03 1.94
(-)2.86 Same Same Same Same
- 33. APRM simulated thermal power trip, % NBR Analysis set point (114 x 1.042) 118.8 115.0
- 34. Time constant, sec 7
6.6 (9)
- 35. Nuclear characteristics used in ODYN simulations (4)
End of equilibrium cycle (EOEC)
Same as Cycle 7/8 core loading
- 36. Recirculation pump trip delay, sec 0.14 Same
- 37. Recirculation pump trip inertia time constant for analysis, sec (3)
Max 5.0 Min 3.0 Max 6.0 (3)
Same
- 38. Total steam line volume, ft3 3243.6 3,275
- 39. Pressure set point of recirculation pump trip - psig (nominal) 1,135 1157
Revision 17 3 of 3 Notes:
(1) Includes control rod drive flow (2) Applicable only for events analyzed using model described in Reference 1 to Section 15.1.
(3) The inertia time constant is defined by the expression:
2 Jon t
=
gTo where:
t = Inertia time constant (sec)
Jo = Pump motor inertia (lb-ft) n = Rated pump speed (rps) g = Gravitational constant (ft/sec)
To = Pump shaft torque (lb-ft)
The 6 second inertia characteristic has been conservatively assumed in T-G trip RPT analysis for several reload cycles (per OPL-3).
(4) The transient analyses for RBS are based on end of equilibrium cycle (EOEC) nuclear parameters for a pre-control cell core design. These analyses results described in Chapter 15 are bounding for limiting transients relative to the expected performance of the plant at end of cycle 1 conditions for the control cell core.
(5) Transients were performed at the core flow range of 81% to 107% of rated or 68.6 to 90.4 Million lb-hr (rated core flow is 84.5 Mlb/hr).
(6) Turbine inlet pressure is measured at Turbine Stop Valve (TSV) inlet conditions.
(7) Values taken from PANACEA/ODYN/CRNC at increased core flow (ICF) condition (100%P/107%F).
(8) The Safety and Relief valve setpoints were obtained from OPL-3 at uprate conditions, where the capacities are based at a reference pressure of 1080 psig. The capacities originally developed for this USAR table were based on a reference pressure of 1210 psig. The pre-uprate OPL-3 information exchange has been shifted to the 1080 psig reference pressure for several cycles.
(9) The 6.6 value for the SIP time constant has been in use for several reload cycles (per OPL-3).
(10) The input parameters and initial conditions for the cycle-specific analyses are given in Attachment B to Appendix 15B.
RBS USAR Revision 8 1 of 1 August 1996 TABLE 15.0-3
SUMMARY
OF ACCIDENTS Section Title Failed Fuel Rods GE-Calculated Value NRC Worst-Case Assumption 15.3.3 Seizure of one recirculation pump None 15.3.4 Recirculation pump shaft break None 8
15.4.9 Rod drop accident
<770 770**
8 15.6.2 Instrument line break None None 15.6.4 Steam system pipe break outside containment None None 15.6.5 LOCA within RCPB None 100%
15.6.6 Feedwater line break-outside containment None None 15.7.1.1 Main condenser gas treatment system failure N/A N/A 15.7.3 Liquid radwaste tank failure N/A N/A 15.7.4 Fuel handling accident
<125 125 15.7.5 Cask drop accident None None 15.8 ATWS
- Special event still under negotiation 8
- See Appendix 15B2.3 for reload core conditions 8
RBS USAR 1 of 1 August 1987 TABLE 15.1-1 SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATER, AUTO FLOW CONTROL (FIGURE 15.1-1)
Time (sec)
Event 0
Initiate a 100°F temperature reduction in the feedwater system 5
Initial effect of unheated feedwater starts to raise core power level but the automatic flow control system automatically reduces core flow to maintain initial steam flow 40 Reactor variables settle into new steady state
RBS USAR 1 of 1 August 1987 TABLE 15.1-2 SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATER, MANUAL FLOW CONTROL (FIGURE 15.1-2)
Time (sec) Event 0
Initiate a 100°F temperature reduction into the feedwater system 5
Initial effect of unheated feedwater starts to raise core power level and steam flow 10 Turbine control valves start to open to regulate pressure 61.9 Initiation of reactor scram on high simulated thermal power 73.0 Narrow range (NR) sensed water level reaches Level 3 (L3) set point 73.2 Trip of recirculation pump power source to low frequency MG speed; RPT initiates due to Level 3 Trip (not included in simulation)
>80(est)
Wide range (WR) sensed water level reaches Level 2 (L2) set point
>80 Recirculation pumps trip off due to Level 2 RPT
>110(est)
HPCS/RCIC flow enters vessel (not simulated)
>120(est)
Reactor variables settle into limit cycle.
RBS USAR Revision 14 1 of 1 September 2001 TABLE 15.1-3 SEQUENCE OF EVENTS FOR FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND (FIGURE 15.1-3)
Time (sec)
Event
- 14 0
Initiate simulated failure of 108 percent upper limit on feedwater flow at a system design pressure of 1,065 psig 32.3 L8 vessel level set point initiates reactor scram and trips main turbine and feedwater pumps 32.4 RPT actuated by stop valve position switches 32.4 Main turbine bypass valves opened due to turbine trip 33.8 SRVs open due to high pressure 36 Water level dropped to low water level setpoint (Level 2)
>66 (est)
RCIC and HPCS flow into vessel (not simulated) 14*
RBS USAR Revision 25 1 of 1 TABLE 15.1-4 SEQUENCE OF EVENTS FOR PRESSURE REGULATOR FAILURE -
OPEN TO 130% (FIGURE 15.1-4)
(NOTE 1)
Time (sec)
Event 0
Simulate steam flow demand to 130 percent 0.5 Main turbine bypass fully opens 8
Turbine control valves wide open 19 Low turbine inlet pressure trip initiates main steam isolation 19.5 MSIV closure initiates reactor scram 22.0 Vessel water level reaches L3 set point.
Recirculation pumps trip to low frequency M/G sets.
25.5(est)
SRVs open 26(est)
Vessel water level reaches L2 set point.
Recirculation pumps trip due to Level 2 RPT signal.
30.5 SRVs close 41.18 Group 1 SRVs open again to relieve decay heat 46.18 Group 1 SRVs close again
>50 (est)
HPCS and RCIC flow enters vessel (not simulated)
NOTE 1: The pressure regulator failure-open is no longer a postulated single failure scenario; no Ovation TCPS single failure scenarios are postulated that could result in all valves open. This specific USAR section is maintained for historical purposes.
RBS USAR 1 of 1 August 1987 TABLE 15.1-5 SEQUENCE OF EVENTS FOR INADVERTENT SAFETY/RELIEF VALVE OPENING Time-sec Event 0
Initiate opening of one SRV 0.5 (est)
Relief flow reaches full flow 15 (est)
System establishes new steady-state operation
RBS USAR 1 of 1 August 1987 TABLE 15.1-6 SEQUENCE OF EVENTS FOR INADVERTENT RHR SHUTDOWN COOLING OPERATION Approximate Elapsed Time Event 0
Reactor at states B or D (Appendix 15A) when RHR shutdown cooling inadvertently activated 0-10 min Slow rise in reactor power
+10 min Operator may take action to limit power rise; Flux scram occurs if no action is taken
RBS USAR Revision 14 1 of 1 September 2001 TABLE 15.2-1 SEQUENCE OF EVENTS FOR PRESSURE REGULATION DOWNSCALE FAILURE (FIGURE 15.2-1)
Time (sec)
Event 0
Simulate zero steam flow demand to main turbine and bypass valves 0
Turbine control valves start to close
14 0.86 Neutron flux reaches high flux scram set point and initiates a reactor scram 2.07 Reactor pressure reaches high pressure setpoint and initiates recirculation pump trip 2.64 SRVs open
13
- Shown for information only. Starting with Cycle 10 this event is no longer required to be analyzed.
13 14
RBS USAR 1 of 1 August 1987 TABLE 15.2-2 SEQUENCE OF EVENTS FOR GENERATOR LOAD REJECTION WITH BYPASS (FIGURE 15.2-2)
Time (sec)
Event
(-)0.015 Turbine-generator detection of loss of (approx.)
electrical load 0
Turbine-generator load rejection sensing devices trip to initiate TCV fast closure and main turbine bypass system operation 0
Turbine control valve (TCV) fast closure initiates scram trip and RPT 0.07 TCVs closed 0.10 Turbine bypass valves start to open 1.30 SRVs open due to high pressure 7.66 SRVs close
RBS USAR Revision 14 1 of 1 September 2001 TABLE 15.2-3 SEQUENCE OF EVENTS FOR GENERATOR LOAD REJECTION WITH FAILURE OF BYPASS (FIGURE 15.2-3)
Time (sec)
Event
(-)0.015 T-G detection of loss of electrical load (approx.)
0 T-G load rejection sensing devices trip to initiate TCV fast closure 0
Turbine bypass valves fail to operate 0
TCV fast closure initiates scram trip and RPT
14 0.08 TCVs closed 1.53 SRVs open due to high pressure 14
RBS USAR 1 of 1 August 1987 TABLE 15.2-4 SEQUENCE OF EVENTS FOR TURBINE TRIP WITH BYPASS (FIGURE 15.2-4)
Time (sec)
Event 0
Turbine trip initiates closure of main stop valves 0
Turbine trip initiates bypass operation 0.01 Main turbine stop valves reach 90% open position and initiate reactor scram trip and RPT 0.10 Turbine stop valves close 0.10 Turbine bypass valves start to open to regulate pressure 1.34 SRVs open due to high pressure 7.60 SRVs close
RBS USAR Revision 14 1 of 1 September 2001 TABLE 15.2-5 SEQUENCE OF EVENTS FOR TURBINE TRIP WITH FAILURE OF BYPASS (FIGURE 15.2-5)
Time (sec)
Event 0
Turbine trip initiates closure of main stop valves 0
Turbine bypass valves fail to operate 0.01 Main turbine stop valves reach 90% open position and initiate reactor scram trip and RPT 0.10 Turbine stop valves close
14 1.58 SRVs open due to high pressure 14
RBS USAR Revision 24 1 of 1 TABLE 15.2-6 SEQUENCE OF EVENTS FOR CLOSURE OF ALL MSIVs (FIGURE 15.2-6)
Time (sec)
Event 0
Initiate closure of all MSIVs 0.45 MSIVs reach 85% open 0.45 MSIV position trip scram initiated 3.6 SRVs open due to high pressure
>8 (est)
SRVs close
RBS USAR Revision 21 1 of 1 TABLE 15.2-7 RADIOLOGICAL CONSEQUENCES OF MSIV CLOSURE Restricted Maximum Area Concentration Boundary Released Concentration Percent Isotope (Ci/hr)
( Ci/cc)
14 I-131 1.2-4(2) 1.1-13 2.0-10 5.6-2 I-132 1.0-3 9.4-13 2.0-8 4.7-3 I-133 1.5-3 1.4-12 1.0-9 1.4-1 I-134 1.1-3 1.0-12 6.0-8 1.7-3 I-135 1.3-3 1.2-12 6.0-9 2.0-2 Br-83 4.1-6 3.7-15 9.0-8 4.1-6 Br-84 1.3-6 1.2-15 8.0-8 1.5-6 Br-85 7.1-8 6.5-17 1.0-9 6.5-6 Kr-83m 2.8-1 2.5-10 5.0-5 5.1-4 Kr-85m 5.8-1 5.3-10 1.0-7 5.3-1 Kr-85 1.4-2 1.3-11 7.0-7 1.9-3 Kr-87 1.2+0 1.1-9 2.0-8 5.7+0 Kr-88 1.7+0 1.6-9 9.0-9 1.8+1 Xe-131m 2.9-3 2.7-12 2.0-6 1.44 Xe-133m 4.9-2 4.5-11 6.0-7 7.5-3 Xe-133 1.4+0 1.3-9 5.0-7 2.5-1 Xe-135m 8.0+0 7.3-9 4.0-8 1.8+1 Xe-135 4.3+0 3.9-9 7.0-8 5.6+0 Xe-138 1.6+0 1.5-9 2.0-8 7.3+0 Total percent of ECL = 55.4 (1) Effluent Concentration Limits (airborne) in unrestricted areas from 10CFR20, Appendix B, Table 2 Column 1.
(2) 1.2-4 = 1.2x10-4 14
RBS USAR 1 of 1 August 1987 TABLE 15.2-8 TYPICAL RATES OF DECAY FOR CONDENSER VACUUM Estimated Vacuum Cause Decay Rate
- 1.
Failure or isolation of
<1 in Hg/min steam jet air ejectors
- 2.
Loss of sealing steam to Approximately 1 to 2 in shaft gland seals Hg/min
- 3.
Opening of vacuum breaker Approximately 2 to 12 in valves Hg/min
- 4.
Loss of one or more cir-Approximately 4 to 24 in culating water pumps Hg/min
RBS USAR 1 of 1 August 1987 TABLE 15.2-9 SEQUENCE OF EVENTS FOR LOSS OF CONDENSER VACUUM (FIGURE 15.2-7)
Time (sec)
Event
-3.0 (est)
Initiate simulated loss of condenser vacuum at 2 in of Hg/sec 0.0 (est)
Low condenser vacuum main turbine trip actuated 0.08 Main turbine trip initiates RPT and scram 1.34 SRVs open due to high pressure 5.0 Low condenser vacuum initiates MSIV closure 5.6 Low condenser vacuum initiates bypass valve closure 7.60 SRVs close 8.13 Group 1 SRVs open again to relieve decay heat 14.03 Group 1 SRVs close again 17.42 Group 1 SRVs open again to relieve decay heat 23.45 Group 1 SRVs close again
RBS USAR 1 of 1 August 1987 TABLE 15.2-10 TRIP SIGNALS ASSOCIATED WITH LOSS OF CONDENSER VACUUM Vacuum (in of Hg)
Protective Action Initiated 27 to 28 Normal vacuum range 20 to 23 Main turbine trip (stop valve closures) 7 to 10 Main steam isolation valve (MSIV) closure and bypass valve closure
RBS USAR 1 of 1 August 1987 TABLE 15.2-11 SEQUENCE OF EVENTS FOR LOSS OF NORMAL AND PREFERRED STATION SERVICE TRANSFORMERS (FIGURE 15.2-8)
Time (sec)
Event 0
Loss of normal and preferred station service transformers occurs.
0 Recirculation system pump motors are tripped.
0 Feedwater and condensate pumps are tripped.
2.00 Reactor scram and closure of MSIV occur due to loss of power to the solenoids.
5.10 SRVs open due to high pressure 10.12 SRVs close 12.10 Group 1 SRVs cycle open and close on pressure 21 Vessel water level reaches Level 2 set point
>45 (est)
HPCS and RCIC flow enters vessel (not simulated)
RBS USAR 1 of 1 August 1987 TABLE 15.2-12 SEQUENCE OF EVENTS FOR LOSS OF ALL GRID CONNECTIONS (FIGURE 15.2-9)
Time (sec)
Event
(-)0.015 Loss of grid causes T-G to detect a loss (approx.)
of electrical load.
0 Turbine control valve fast closure is initiated.
0 T-G power-load unbalance (PLU) trip initiates main turbine bypass system operation.
0 Recirculation system pump motors are tripped.
0 TCV fast closure initiates a reactor scram trip.
0.07 TCVs closed.
0.11 Turbine bypass valves open.
1.33 SRVs open due to high pressure.
2.00 Closure of MSIV due to loss of power.
9.03 SRVs close.
19 (est)
Vessel water level reaches Level 2 set point.
>45(est)
HPCS and RCIC flow enters vessel (not simulated).
RBS USAR 1 of 1 August 1987 TABLE 15.2-13 SEQUENCE OF EVENTS FOR LOSS OF ALL FEEDWATER FLOW (FIGURE 15.2-10)
Time (sec)
Event 0
Trip of all feedwater pumps initiated.
2.33 Vessel water level reaches level 4 and initiates recirculation flow runback.
4.88 Feedwater flow decays to zero.
8.71 Vessel water level (L3) trip initiates scram trip and recirculation pumps trip to low frequency M/G set.
24(est)
Vessel water level reaches Level 2.
24(est)
Recirculation pumps trip due to Level 2 RPT signal.
>50(est)
HPCS and RCIC flow enters vessel (not simulated).
RBS USAR Revision 24 1 of 1 TABLE 15.2-14 14 13 SEQUENCE OF EVENTS FOR FAILURE OF RHR SHUTDOWN COOLING Configuration for Activity C1 (a)
Time (sec)
Event 13 0
Reactor is operating at 100.3 percent rated when loss of offsite power occurs initiating power plant shutdown.
0 Concurrently loss of Division power (i.e., loss of one diesel generator) occurs.
600 Suppression pool cooling initiated to prevent overheating from SRV actuation.
1465 Controlled depressurization initiated (100°F/hr) using selected safety/relief valves.
8900 Blowdown to approximately 100 psig completed.
8900 Personnel are sent in to open RHR shutdown cooling suction valve; this fails.
9200 ADS valves are opened to complete blowdown to suppression pool, and RHR pump discharge is redirected from pool to vessel via LPCI line.
Alternate shutdown cooling path has now been established.
11,600 Cold Shutdown achieved (200 degrees F RPV Temperature) 16,146 Maximum Suppression Pool Temperature (183.1 degrees F) 14
RBS USAR Revision 24 1 of 1 TABLE 15.2-14a 14 SEQUENCE OF EVENTS FOR FAILURE OF RHR SHUTDOWN COOLING Configuration for Activity C1 (b)
Time (sec)
Event 0
Reactor is operating at 100.3 percent rated power when loss of offsite power occurs initiating plant shutdown.
0 Concurrently loss of Division power (i.e., loss of one diesel generator) occurs.
600 Suppression pool cooling initiated to prevent overheating from SRV actuation.
1465 Controlled depressurization initiated (100°F/hr) using selected safety/relief valves.
8900 Blowdown to approximately 100 psig completed.
8900 Personnel are sent in to open RHR shutdown cooling suction valve; this fails.
9200 ADS valves are opened to complete blowdown to suppression pool, and RHR pump discharge is redirected from pool to vessel via LPCI line.
Alternate shutdown cooling path has now been established.
13,368 Maximum Suppression Pool Temperature (177.9 degrees F) 40,586 Cold Shutdown achieved (200 degrees F RPV Temperature) 14
RBS USAR Revision 17 1 of 1 TABLE 15.2-15 INPUT PARAMETERS FOR EVALUATION OF FAILURE OF RHR SHUTDOWN COOLING Initial Conditions
14 Rated power (%)
100.3 Suppression pool water volume (ft
- 3) 1.228E5 RHR Hx constant (Btu/sec/°F) 390 Vessel pressure (psia) 1072 Vessel temperature (°F) 553 Primary coolant inventory (lbm) 4.598E5 Pool temperature (°F) 100 Service water temperature (°F) 95 Vessel heat capacity (Btu/lbm/°F) 0.123 HPCS flow rate (lbm/sec) 676.8 Maximum at 0 psid, vessel to drywell pressure difference LPCI flow rate per loop (lbm/sec) 686.1 Maximum at 0 psid, vessel-to-drywell pressure difference LPCI Pump Heat (HP) 700 HPCS Pump Heat (HP) 2500 14
1 of 1 August 1987 TABLE 15.3-1 SEQUENCE OF EVENTS FOR TRIP OF ONE RECIRCULATION PUMP (FIGURE 15.3-1)
Time (sec)
Event 0
Trip of one recirculation pump initiated 5
Jet pump diffuser flow reverses in the tripped loop 39.0 Core flow and power level stabilize at new equilibrium conditions.
1 of 1 August 1987 TABLE 15.3-2 SEQUENCE OF EVENTS FOR TRIP OF TWO RECIRCULATION PUMPS (FIGURE 15.3-2)
Time (sec)
Event 0
Trip of both recirculation pumps initiated 4.2 Vessel water level (L8) trip initiates scram, turbine trip and feedwater pump trip 4.3 Turbine trip initiates bypass operation 5.8 SRVs open due to high pressure 11.2 SRVs close 17.2 Vessel water level (L2) set point reached 47.2(est)
HPCS and RCIC flow enters vessel (not simulated)
1 of 1 August 1987 TABLE 15.3-3 SEQUENCE OF EVENTS FOR FAST CLOSURE OF ONE MAIN RECIRCULATION VALVE (FIGURE 15.3-3)
Time (sec)
Event 0
Initiate fast closure of one main recirculation valve 2
Jet pump diffuser flow reverses in the affected loop 40(est)
Core flow and power approach new equilibrium conditions
1 of 1 August 1987 TABLE 15.3-4 SEQUENCE OF EVENTS FOR FAST CLOSURE OF TWO MAIN RECIRCULATION VALVES (FIGURE 15.3-4)
Time (sec)
Event 0
Initiate fast closure of both main recirculation valves 5.15 Vessel level (L8) trip initiates scram and turbine trip 5.15 Feedwater pumps tripped off 5.30 Turbine trip initiates bypass operation 6.46 SRVs open due to high pressure 12.31 SRVs close 17.50 Vessel water level reaches Level 2 set point 47.50(est) HPCS and RCIC flow enters vessel (not simulated)
1 of 1 August 1987 TABLE 15.3-5 SEQUENCE OF EVENTS FOR RECIRCULATION PUMP SEIZURE (FIGURE 15.3-5)
Time (sec)
Event 0
Single pump seizure was initiated.
0.8 Jet pump diffuser flow reverses in seized loop.
3.11 Vessel level (L8) trip initiates reactor scram.
3.11 Vessel level (L8) trip initiates turbine and feedwater pump trips.
3.30 Turbine trip initiates bypass operation.
3.35 Turbine trip initiates recirculation pumps trip.
4.85 SRVs open due to high pressure.
10.2 SRVs close.
12.7 Vessel water level reaches Level 2 set point.
42.7(est)
HPCS/RCIC flow enters the vessel (not simulated).
- Based on a l.0-foot RWL increment.
1 of 1 August 1987 RBS USAR TABLE 15.4-l SEQUENCE OF EVENTS FOR ROD WITHDRAWAL ERROR IN POWER RANGE Elapsed Time Event 0
Core is operating on thermal limits with a typical control rod pattern.
0 Operator selects and withdraws a single rod or gang of rods continuously.
~1 sec The local power in the vicinity of the withdrawn rod (or gang) increases. Total core power output increases.
~4* sec RWL blocks further withdrawal.
~25 sec Core stabilizes at slightly higher core power level.
1 of 1 August 1987 RBS USAR TABLE 15.4-2 SEQUENCE OF EVENTS FOR ABNORMAL STARTUP OF IDLE RECIRCULATION PUMP (FIGURE 15.4-l)
Time (sec)
Event 0
Start pump motor 1.26 Jet pump diffuser flows on started pump side become positive 2.73 Pump motor at full speed and drive flow at about 25% of rated 21.5 (est)
Last of cold water leaves recirculation drive loop 22.0 Peak value of core inlet subcooling 50.0 (est)
Reactor variables settle into new steady state
1 of 1 August 1987 RBS USAR TABLE 15.4-3 SEQUENCE OF EVENTS FOR FAST OPENING OF ONE RECIRCULATION VALVE (FIGURE 15.4-2)
Time (sec)
Event 0
Simulate failure of single loop control 0.97 Reactor APRM high flux scram trip initiated 3.5 (est)
TCVs start to close upon falling turbine pressure 7.9 (est)
TCVs closed; turbine pressure below pressure regulator set points
>lOO (est)
Reactor variables settle into new steady state
1 of 1 August 1987 RBS USAR TABLE 15.4-4 SEQUENCE OF EVENTS FOR FAST OPENING OF TWO RECIRCULATION VALVES (FIGURE 15.4-3)
Time (sec)
Event 0
Initiate failure of master controller 1.0 Reactor APRM high flux scram trip initiated 4.0 (est)
TCVs start to close upon falling turbine pressure 8.0 (est)
TCVs closed; turbine pressure below pressure regulator set points
>lOO (est)
Reactor variables settle into new steady state
1 of 1 August 1987 RBS USAR TABLE 15.4-5 SEQUENCE OF EVENTS FOR MISPLACED BUNDLE ACCIDENT
- 1.
During core loading operation, a bundle is placed in the wrong location.
- 2.
Subsequently, the bundle intended for this location is placed in the location of the previous bundle.
- 3.
During core verification procedure, the two errors are not observed.
- 4.
Plant is brought to full power operation without detecting misplaced bundle.
- 5.
Plant continues to operate throughout the cycle.
RBS USAR NOTE: Core conditions are assumed to be normal for a hot, operating core at EOC.
8
- See Appendix 15B for reload core conditions 8
Revision 8 1 of 1 August 1996 TABLE 15.4-6 INPUT PARAMETERS AND INITIAL CONDITIONS FOR FUEL BUNDLE LOADING ERROR
8 Input Parameters Initial Conditions
- 8
- 1.
Power, % rated 100
- 2.
Flow, % rated 100
- 3.
MCPR operating limit (est) 1.18
- 4.
MLHGR operating limit, kw/ft 13.4
- 5.
Core exposure End of cycle
8 NOTE: See Appendix 15B for reload core conditions 8
Revision 8 1 of 1 August 1996 TABLE 15.4-7
8 RESULTS OF WORST FUEL BUNDLE LOADING ERROR ANALYSIS (INITIAL CORE) 8
- 1.
MCPR limit 1.18
- 2.
MCPR with misplaced bundle 1.08 3.
CPR for event 0.10
- 4.
LHGR limit 13.4
- 5.
LHGR with misplaced bundle 14.7 6.
LHGR for event 1.3
1 of 1 August 1987 RBS USAR TABLE 15.4-8 SEQUENCE OF EVENTS FOR ROD DROP ACCIDENT Approximate Elapsed Time (sec)
Event Reactor is operated at 50 percent rod density pattern Maximum worth control rod blade becomes decoupled from the CRD Operator selects and withdraws the CRD of the decoupled rod either individually or along with other control rods assigned to the RCIS group Decoupled control rod sticks in the fully inserted or an intermediate bank position 0
Control rod becomes unstuck and drops to the drive position at the nominal measured velocity plus three standard deviations
<1 Reactor goes on a positive period and the initial power increase is terminated by the Doppler coefficient
<1 APRM 120 percent power signal scrams reactor
<5 Scram terminates accident
1 of 1 August 1987 RBS USAR TABLE 15.4-9 INPUT PARAMETERS AND INITIAL CONDITIONS FOR ROD WORTH COMPLIANCE CALCULATION Input Parameters Initial Conditions
- 1.
Reactor power, % Rated 0.0
- 2.
Reactor flow, % Rated 0.0
- 3.
Core average exposure, MWd/t 0.0
- 4.
Control rod fraction Approx. 0.50
- 5.
Average fuel temperature, C 286
- 6.
Average moderator temperature C 286
- 7.
Xenon state None
(1) The following assumptions were made to ensure that the rod worths were conservatively high for the BPWS:
- a. BOC 1, 0.0 GWD/St average exposure
- b. Hot startup
- c. No xenon
- d. Rod groups l-6 withdrawn
- e. Sequence A 1 of 1 August 1987 RBS USAR TABLE 15.4-10 INCREMENT WORTH OF THE MOST REACTIVE ROD USING A BANK POSITION WITHDRAWAL SEQUENCE (l)
Core Control Banked Control Condition Rod At Rod Drops Increase (MWD/T)
Group Notch (I,J)
From-To In keff 0.0 7
04 (24,49) 00-08 0.0012 0.0 7
08 (24,49) 00-12 0.0032 0.0 7
12 (24/49) 00-48 0.0079 0.0 7
48 (24,49) 00-48 0.0005
RBS USAR TABLE 15.4-11 14 8 8 14 Revision 23 1 of 2 CONTROL ROD DROP ACCIDENT RADIOLOGICAL CONSEQUENCE ANALYSIS PARAMETERS Description of Input/Assumption Design Basis Input and/or Assumption A. Data and assumptions used to estimate radioactive source from postulated accident
- 1. Power Level 3100 MWt
- 3. Total Rods in core GE 8x8 GE 9x9 GE 10x10 38,688 (62 rods per assembly) 46,176 (74 rods per assembly) 57,408 (92 rods per assembly)
- 4. Number of assemblies damaged Design Basis - Maximum Fuel Damage (based on 8x8)
Limited CRDA (Based on 8x8) 850/62 = 13.7 50/62 = 0.8 Note: For the CRDA scenario GE8 fuel is bounding. This is confirmed each reload cycle.
- 5. Core Activity available for release Table 15.4-11A
- 6. Radial Peaking Factor 2.00
- 7. Assumed % fuel melt Design Basis - Maximum Fuel Damage Limited CRDA 100%
0%
- 8. Gap Activity Release Fractions Per RG 1.183, Table 3 and Appendix C 10% noble gases, 10% iodines, 12% alkali metals
- 9. Fuel Melt Release Fractions Per RG 1.183, Appendix C 100% noble gases, 50% iodines 10.Fuel Release Duration Design Basis-Maximum Fuel Damage Limited CRDA Instantaneous 10 sec. burst B. Data and assumptions used to estimate activity released to the environment.
- 1. Condenser Leak Rate Design Basis CRDA Limited CRDA Per RG 1.183, Appendix C 1% per day for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 4000 cfm for 20 minutes, 1% per day for next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
- 14 *8 8* 14*
Revision 22 2 of 2 Description of Input/Assumption Design Basis Input and/or Assumption
- 3. Condenser Radioactive Decay During Holdup Credited
- 4. Condenser Volume 106,460 ft3 C. Dispersion Data
- 1. EAB X/Q Data 0-2 hrs 7.51E-04 sec/m3
- 2. LPZ X/Q Data 0-8 hrs 7.79E-05 sec/m3 8-24 hrs 5.23E-05 sec/m3 1-4 days 2.21E-05 sec/m3 4-30 days 6.40E-06 sec/m3
- 3. Control Room X/Q Data Main Air Intake 0-2 hrs 3.02E-3 sec/m3 2-8 hrs 2.47E-3 sec/m3 8-24 hrs 1.05E-3 sec/m3 1-4 days 9.01E-4 sec/m3 4-30 days 6.74E-4 sec/m3 D. Control Room Parameters
- 1. Free Air Volume 188,000 ft3
- 2. Unfiltered In-leakage Rate 300 cfm
- 3. Outside Air Ventilation Rate 1700 cfm
Elemental and Organic 0%
Limited CRDA Aerosol 99%
Elemental and Organic 98%
- 5. Time for Control Room Ventilation Isolation per Operator Action Design Basis CRDA Not credited Limited CRDA 20 minutes
- 6. Emergency Mode Recirculation Rate (Post-isolation Mode) 2000 cfm
- 7. Control Room Breathing Rates and Occupancy Factors Per RG 1.183
- 14 *2 2* 14*
RBS USAR TABLE 15.4-11A Revision 22 1 of 1 RBS CRDA CORE ACTIVITY Isotope EOC Core Inventory (Ci/MWt)
Kr-85 3.66E+02 Kr-85m 7.02E+03 Kr-87 1.35E+04 Kr-88 1.89E+04 Rb-86 6.31E+01 I-131 2.70E+04 I-132 3.92E+04 I-133 5.52E+04 I-134 6.06E+04 I-135 5.17E+04 Xe-133 5.26E+04 Xe-135 1.99E+04 Cs-134 6.11E+03 Cs-136 2.00E+03 Cs-137 3.95E+03
RBS USAR Revision 22 1 of 1 TABLE 15.4-12 CRDA ACTIVITY RELEASED TO THE ENVIRONMENT (CURIES)
Isotope 100% Power Event Kr-85 1.61E+01 Kr-85m 3.08E+02 Kr-87 5.93E+02 Kr-88 8.30E+02 Rb-86 3.33E-05 I-131 5.93E+00 I-132 8.61E+00 I-133 1.21E+01 I-134 1.33E+01 I-135 1.14E+01 Xe-133 2.31E+03 Xe-135 8.74E+02 Cs-134 3.22E-03 Cs-136 1.05E-03 Cs-137 2.08E-03
- 14 14*
NOTE: 1.26E+02 = 1.26 x 102
- 14 14*
Revision 22 1 of 1 CONTROL ROD DROP ACCIDENT RADIOLOGICAL CONSEQUENCES Receptor Regulatory Limit (TEDE)
Design Basis Event Dose (TEDE)
EAB 6.3 1.0 4.91 LPZ 6.3 0.4 0.51 Control Room 5
4.4 1.30
- 2 2*
RBS USAR 1 of 1 August 1987 RBS USAR TABLE 15.5-1 SEQUENCE OF EVENTS FOR INADVERTENT STARTUP OF HPCS (FIGURE 15.5-1)
Time (sec)
Event 0
Simulate HPCS cold water injection 3
Full flow established for HPCS 7
Depressurization effect stabilized
- See also Section 6.3.3.7.7.
Revision 10 1 of 1 April 1998 TABLE 15.6-1 SEQUENCE OF EVENTS FOR STEAM LINE BREAK OUTSIDE CONTAINMENT*
Time (sec)
Event 0
Guillotine break of one main steam line outside primary containment 0.5 High steam line flow signal initiates closure of (Approx.)
<1.0 Reactor begins scram
5.5 MSIVs fully closed
10
60 SRVs open on high vessel pressure. The valves open and close to maintain vessel pressure at approximately 1,000 psi
190 RCIC and HPCS would initiate on low water level, L2, (RCIC considered unavailable, HPCS assumed single failure and therefore not available)
800 ADS receives signal to initiate on low water level, L1; ADS bypass timer starts
1300 All ADS timers timed out. ADS valves are actuated initiating rapid depressurization of vessel
1400 Reactor water level above core begins to drop slowly due to loss of steam through the SRVs; reactor pressure still at approximately 1,000 psi
1420 LPCS system initiates injection
~1450 LPCI system initiates injection
~1600 Reactor vessel water level recovers back to initial level: no fuel rod heatup and no fuel rod failure.
10
RBS USAR Revision 17 1 of 2 TABLE 15.6-2 MAIN STEAM LINE BREAK RADIOLOGICAL CONSEQUENCE ANALYSIS PARAMETERS Description of Input/Assumption Design Basis Input and/or Assumption I. Data and assumptions used to estimate radioactive source from postulated accident.
- 1. Power Level 3100 MWt
- 4. Noble Gas Source Term Based on 310,000 µCi/sec at 30 minutes, corrected to time equal zero.
- 5. Alkali Metals Reactor coolant activity design concentration ratioed to account for 102% power.
II. Data and assumptions used to estimate activity released to the environment.
- 1. Mass Release Note: This data corresponds to that calculated for initial licensing of the plant.
Analyses demonstrate these values bound the hot standby conditions for Power Uprated conditions.
Steam, 11,620 lbm Liquid, 68,942 lbm
- 2. Iodine Carryover Fraction 4%
- 3. Break Isolation Time 5.5 seconds
- 4. Building Release Rate Instantaneous ground level release with no credit for plateout, holdup, or dilution.
- 5. Iodine Species Release Fractions to Environment Per RG 1.183, Appendix D, 95% Aerosol, 4.85 % Elemental, 0.15% Organic
- 6. Activity Released to Environment Table l5.6-3
Revision 17 2 of 2 Description of Input/Assumption Design Basis Input and/or Assumption III. Dispersion Data
- 1. EAB X/Q Data 0-2 hrs 6.33E-04 sec/m3
- 2. LPZ X/Q Data 0-8 hrs 8-24 hrs 1-4 days 4-30 days 7.57E-05 sec/m3 5.08E-05 sec/m3 2.13E-05 sec/m3 6.24E-06 sec/m3
- 3. Control Room X/Q Data
- 0-2 hrs 2-8 hrs 8-24 hrs 1-4 days 4-30 days
- Since no credit is taken for the ESF filter trains, the X/Q values assumed in this analysis were based on the Main Air Intake.
1.42E-03 sec/m3 1.08E-03 sec/m3 4.57E-04 sec/m3 3.50E-04 sec/m3 2.58E-04 sec/m3 IV. Control Room Parameters
- 1. Free Air Volume 188,000 ft3
- 2. Unfiltered In-leakage Rate 300 cfm
- 3. Outside Air Ventilation Rate 1700 cfm
- 4. Emergency Mode Filtered Intake/Unfiltered Inleakage Rate (1700 cfm ventilation rate +
300 cfm inleakage rate) 2000 cfm
11 3 3 11
RBS USAR NOTE: 1.9E+01 = 1.9x101 Revision 17 1 of 1 TABLE 15.6-3 MAIN STEAM LINE BREAK ACTIVITY RELEASED TO ENVIRONMENT O.2Ci/gm DE* I-131 4Ci/gm DE I-131 Isotope Case (Ci)
Case (Ci)
I-131 1.70E+00 3.40E+01 I-132 2.51E+01 5.02E+02 I-133 2.27E+01 4.53E+02 I-134 3.97E+01 7.94E+02 I-135 2.19E+01 4.37E+02 Cs-134 5.42E-03 5.42E-03 Cs-136 3.51E-03 3.51E-03 Cs-137 1.40E-02 1.40E-02 Kr-85m 5.22E-02 5.22E-02 Kr-85 1.63E-04 1.63E-04 Kr-87 1.79E-01 1.79E-01 Kr-88 1.79E-01 1.79E-01 Xe-133 6.58E-02 6.58E-02 Xe-135 1.95E-01 1.95E-01
14 11 14
11
- Dose Equivalent
RBS USAR Revision 17 1 of 1 TABLE 15.6-4 MAIN STEAM LINE BREAK ACCIDENT RADIOLOGICAL CONSEQUENCES EAB Dose LPZ Dose Regulatory Limit Case (REM TEDE)
(REM TEDE) 4µCi/gm DE* I-131 1.4 0.2 25 0.2µCi/gm DE I-131
<0.1
<0.1 2.5 Control Room Dose Regulatory Limit Case (REM TEDE)
(REM TEDE) 4µCi/gm DE I-131 2.2 5
14 11 3 3 11 14
- Dose Equivalent
RBS USAR Revision 18 1 of 2 TABLE 15.6-5 LOSS-OF-COOLANT ACCIDENT RADIOLOGICAL CONSEQUENCES ANALYSIS PARAMETERS Design Basis Input Description of Input/Assumption and/ or Assumptions I.
Data and assumptions used to estimate radioactive source from postulated accident
14 A. Power level 3,100 MWt B. Core Activity available for release Table 15.6-6 C. Gap Activity Release Fractions Per Table 1 of RG 1.183 D. Release fission product species and chemical form Per RG 1.183, Section 3.5
13 10 II.
Release Rates A. Primary Containment Leakage Rate 0-24 hours 0.325 volume % per day 1-30 days 0.179 volume % per day B. Secondary Containment Bypass Leakage Rate 0-24 hours 580,000 cc/hr @ Pa (0.341 cfm) 1-30 days 319,000 cc/hr @ Pa (0.188 cfm)
C. Main Steam Line Leakage 0-25 minutes 150 scfh = 2.52 cfm*
25 minutes - 30 days 0 scfh D. Engineered Safety Features Leakage 1 gpm 150 scfh was converted based on a maximum DW temperature of 330 F. The pressure assumed was 7.6 psig (power uprate reports show that the drywell pressure is 22.8 psia @ 121 seconds decreases steadily to ~19 psia at 10 minutes).
III. Dispersion Data A. EAB X/Q Data See Table 15.6-5A B. LPZ X/Q Data See Table 15.6-5A C. Control Room X/Q Data See Table 15.6-5A IV.
Control Room Parameters A. Unfiltered In-Leakage Rate 300 cfm B. Outside Air Ventilation Rate Actual 2000 cfm Assumed 2000-300 = 1700 cfm C. Filter Initiation Time 20 minutes D. CR ESF Iodine Filter Efficiency Elemental/Organic (Charcoal) 98%
Particulate (HEPA) 99%
E. CR Breathing Rates and Occupancy Factors Per RG 1.183
RBS USAR Table 15.6-5 (Cont) 1.16E+06 = 1.16x106 Revision 18 2 of 2 V.
Standby Gas Treatment Parameters A. Positive Pressure Period 30 minutes B. Flow Rates Annulus 2,500 cfm Auxiliary Building 10,000 cfm C. SGTS Iodine Filter Efficiency Elemental/Organic (Charcoal) 90%
Particulate (HEPA) 99%
VI.
Building Volumes A. Drywell 2.36E+05 ft3 B. Containment 1.19E+06 ft3 C. Annulus***
3.57E+05 ft3 D. Auxiliary Building*
1.16E+06 ft3 E. Control Room 1.88E+05 ft3 F. Suppression Pool**
1.25E+05 ft3 Only 50% of the auxiliary building volume was credited in the actual analysis (values listed are actual volumes)
- The Actual analysis conservatively assumed a volume of 120,000 ft3.
VII. Containment Mixing Data A. Blowdown Data (Drywell Containment) 0-10 minutes 4.74E+05 cfm 10 minutes +
0 cfm B. Hydrogen Mixing Data (Drywell Containment) 0-25 minutes 0 cfm 25 minutes - 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> 600 cfm 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> - 30 days (Infinite mixing) 1.0E+08 cfm C. Steaming Data (Drywell Containment) 0-25 minutes 0 cfm 25 minutes - 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> 3000 cfm 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> - 30 days (Infinite mixing)
(Included in Hydrogen Mixing)
VIII. Misc. Data A. Dose Conversion Factors Based on Federal Guidance Report 11 & 12 B. Off-Site Breathing Rates Based on RG 1.183, Section 4.1.3 A. Drywell Plateout Coefficients Elemental 1.01hr-1 Particulate RADTRAD default -
Powers (10) Model D. ESF Leakage - Halogen Flashing Fraction 0.10
11 10 11 13 14
2.55E-4 = 2.55x10-4 Revision 22 1 of 1 Table 15.6-5A X/Q VALUES USED IN LOCA ANALYSIS Release Point EAB*
LPZ MCR SGTS/Containment 0-2 hours 6.05E-4 7.49E-5 2.55E-4 2-8 hours 6.05E-4 7.49E-5 1.92E-4 8-24 hours 6.05E-4 5.02E-5 8.09E-5 1-4 days 6.05E-4 2.10E-5 6.22E-5 4-30 days 6.05E-4 6.13E-6 5.09E-5 Turbine Building 0-2 hours 7.51E-4 7.79E-5 4.66E-4 2-8 hours 7.51E-4 7.79E-5 3.83E-4 8-24 hours 7.51E-4 5.23E-5 1.67E-4 1-4 days 7.51E-4 2.21E-5 1.27E-4 4-30 days 7.51E-4 6.40E-6 9.33E-5
- The 0-2 hour values conservatively apply for the duration of the accident to ensure the maximum 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose is calculated as required per RG 1.183.
RBS USAR Revision 22 1 of 2 TABLE 15.6-6
Co-60 4.50E+02 1
Kr-85 3.66E+02 1
Kr-85m 7.02E+03 1
Kr-87 1.35E+04 1
Kr-88 1.89E+04 3
Rb-86 6.31E+01 4
Sr-89 2.54E+04 4
Sr-90 2.91E+03 4
Sr-91 3.20E+04 4
Sr-92 3.47E+04 6
Y-90 3.08E+03 6
Y-91 3.28E+04 6
Y-92 3.49E+04 6
Y-93 4.04E+04 6
Zr-95 4.78E+04 6
Zr-97 4.98E+04 6
Nb-95 4.80E+04 5
Mo-99 5.13E+04 5
Tc-99m 4.51E+04 5
Ru-103 4.19E+04 5
Ru-105 2.90E+04 5
Ru-106 1.61E+04 5
Rh-105 2.67E+04 4
Sb-127 2.92E+03 4
Sb-129 8.77E+03 4
Te-127 2.95E+03 4
Te-127m 3.92E+02 4
Te-129 8.62E+03 10* 13* 14*
RBS USAR NOTE: 6.81E+03 = 6.81x103 Revision 22 2 of 2 Table 15.6-6 (Cont)
AST Group Isotope Core Inventory (CI/MWT) 4 Te-129m 1.28E+03 4
Te-131m 3.92E+03 4
Te-132 3.84E+04 2
I-131 2.70E+04 2
I-132 3.92E+04 2
I-133 5.52E+04 2
I-134 6.06E+04 2
I-135 5.17E+04 1
Xe-133 5.26E+04 1
Xe-135 1.99E+04 3
Cs-134 6.11E+03 3
Cs-136 2.00E+03 3
Cs-137 3.95E+03 4
Ba-139 4.92E+04 4
Ba-140 4.74E+04 6
La-140 4.88E+04 6
La-141 4.49E+04 6
La-142 4.33E+04 7
Ce-141 4.49E+04 7
Ce-143 4.15E+04 7
Ce-144 3.69E+04 6
Pr-143 4.06E+04 6
Nd-147 1.80E+04 7
Np-239 5.42E+05 7
Pu-238 1.10E+02 7
Pu-239 1.28E+01 7
Pu-240 1.66E+01 7
Pu-241 5.26E+03 6
Am-241 6.55E+00 6
Cm-242 1.51E+03 6
Cm-244 7.62E+01
- 13 *11 11* 13*
Revision 22 1 of 1 TABLE 15.6-7
- 11 11* *13 *10 LOSS-OF-COOLANT ACCIDENT RADIOLOGICAL CONSEQUENCES Release Descriptions EAB LPZ MCR Containment/Secondary Containment 4.0 2.2 0.46 Secondary Containment Bypass/MSIV 13.4 6.2 3.12 ESF Liquid Leakage 0.4 0.6 0.16 Total 17.8 9.0 3.7 Regulatory Limit 25.0 25.0 5.00
- 14 10* 13* 14*
RBS USAR 1 of 1 August 1987 TABLE 15.6-8 SEQUENCE OF EVENTS FOR FEEDWATER LINE BREAK OUTSIDE CONTAINMENT Time (sec)
Event 0
One feedwater line breaks 0+
Feedwater line check valves isolate the reactor from the break 5 (approx)
Reactor scram on low water level
<30 HPCS and RCIC start on low-low water level, L2, and are expected to maintain the water level above the low-low-low level, L1, trip, and eventually restore it to the normal elevation 1 to 2 hr Normal reactor cooldown procedure established
RBS USAR 1 of 1 August 1987 TABLE 15.7-1 SEQUENCE OF EVENTS FOR MAIN CONDENSER OFF GAS TREATMENT SYSTEM FAILURE Approximate Elapsed Time Event 0 sec Event begins - system fails.
0 sec Noble gases are released.
< 1 min Area radiation alarms alert plant personnel.
< 1 min Operator actions begin with:
1.
Initiation of appropriate system isolations 2.
Manual scram actuation 3.
Assurance of reactor shutdown cooling.
1 hr Steam jet air ejector is shut down.
RBS USAR 1 of 2 August 1987 TABLE 15.7-2 GASEOUS RADWASTE SYSTEM FAILURE - PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSIS Design Basis Assumptions I.
Data and assumptions used to estimate radioactive source for postulated accidents A.
Power level NA B.
Burnup NA C.
Fuel damage None D.
Release of activity Table 15.7-3 E.
Iodine fractions NA 1.
Organic NA 2.
Elemental NA 3.
Particulate NA F.
Reactor coolant activity before the accident NA II.
Data and assumptions used to estimate activity released A.
Containment leak rate (%/day)
NA B.
Secondary containment leak rate(%/day)
NA C.
Valve movement times NA D.
Absorption and filtration efficiencies NA 1.
Organic iodine NA 2.
Elemented iodine NA 3.
Particulate iodine NA 4.
Particulate fission products NA E.
Recirculation system parameters NA 1.
Flow rate NA 2.
Mixing efficiency NA 3.
Filter efficiency NA F.
Contaiment volumes NA G.
All other pertinent data and assumptions None
8.58-4 = 8.58x10-4 Revision 3 2 of 2 August 1990 TABLE 15.7-2 (Cont)
Design Basis Assumptions III.
Dispersion data
- 3 A.
EAB distance (m) 894 B.
X/Qs for EAB (sec/m-)
8.58-4 3*
IV.
Dose data A.
Method of dose calculation NA B.
Dose conversion assumptions Reg. Guides 1.98 and 1.109 C.
Peak activity concentrations in containment NA D. Doses Table 15.7-4
RBS USAR 1 of 1 August 1987 TABLE 15.7-3 EQUIPMENT FAILURE RELEASE ASSUMPTIONS RELEASE FRACTIONS ASSUMED FOR DESIGN BASIS ANALYSIS Noble Particulate Equipment Piece Gases Daughters Radioiodine Holdup pipe 1.00 1.00 NA Prefilter NA 0.01 NA Charcoal adsorbers 1.00 0.01 NA Steam jet air ejector (1-hr release) 1.00 1.00 NA
RBS USAR Revision 14 1 of 1 September 2001 TABLE 15.7-4 GASEOUS RADWASTE SYSTEM FAILURE SYSTEM RUPTURE (DESIGN BASIS ANALYSIS) OFFSITE RADIOLOGICAL EFFECTS Dose at Exclusion Area Boundary Skin Whole Body Noble Gases (rem)
(rem)
Charcoal bed 6.0-1 6.8-1 Holdup pipe 1.2-2 2.3-2 Steam jet air ejector 3.8-1 5.1-1 (1-hr release)
Particulates
- 14 Prefilter 5.6-2 Charcoal bed 1.9-4 Holdup pipe 3.2-4 Steam jet air ejector 2.1-3 (1-hr release) 14*
TOTAL 9.9-1 1.3+0 NOTE:
6.0-1 = 6.0x10-1
RBS USAR 1 of 1 August 1987 TABLE 15.7-5 ACCIDENT ANALYSIS ASSUMPTIONS RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE (RELEASE TO ATMOSPHERE) 1.
Design bases activities of Section 11.1, and flows and fractions of primary coolant activities of Section 11.2 are used to develop source terms.
2.
Noble gases are not considered due to constant venting of radwaste system tanks.
3.
Halogen partition factor 0.001 4.
Exclusion area boundary
/Q Breathing Rate (sec/m3)
(m3/sec) 0-2 hr 8.58-4 3.47-4 NOTE:
8.58-4 = 8.58x10-4
RBS USAR TABLE 15.7-6 LIQUID RADWASTE SYSTEM TANKS HALOGEN INVENTORIES Br-83 Br-84 Br-85 I-131 I-132 I-133 I-134 I-135 (Ci)
(Ci)
(Ci)
(Ci)
(Ci)
(Ci)
(Ci)
(Ci)
NOTE:
8.4-4 = 8.4x10-4 Revision 14 1 of 1 September 2001 Floor drain collector tanks 2A, B, and C (total) 8.4-4 1.8-4 1.1-5 5.1-3 9.9-3 4.2-2 4.2-3 1.7-2
- 14 Waste collector tanks 1A, B, C, and D (total) 3.6-1 7.2-2 3.6-3 3.6+0 3.3+0 2.0+1 1.9+0 7.6+0 Recovery sample tanks 4A, B, C, and D (total) 1.9-4 1.5-5 6.8-8 2.7-3 1.8-3 1.4-2 6.2-4 4.8-3 14*
Phase separator and backwash tanks 6A, B, and 7 2.1+1 4.6+0 2.7-1 1.6+3 6.2+2 1.6+3 1.1+2 4.4+2 Regenerant waste tanks 3A and B 3.2+1 7.0+0 4.0-1 4.0+2 2.8+2 2.2+3 2.2+2 7.0+2 Waste and regenerant evaporators EV-1 and 2 1.5+1 7.4-1 4.0-3 5.4+3 3.4+2 1.1+4 3.8+1 9.6+2
- 14 Total in all tanks 6.8+1 1.2+1 6.8-1 7.4+3 1.2+3 1.5+4 3.7+2 2.1+3 14*
RBS USAR 1 of 1 August 1987 TABLE 15.7-7 OFFSITE DOSES RESULTING FROM LIQUID RADWASTE SYSTEM TANKS RUPTURE Whole Body Thyroid Gamma Beta Exclusion area boundary (rem)
(rem)
(rem) 0-2 hr 5.1+0 4.0-3 1.8-3 NOTE:
4.0-3 = 4.0x10-3
RBS USAR 1 of 1 August 1987 TABLE 15.7-8 ACCIDENT ANALYSIS DATA LIQUID RADWASTE TANK RUPTURE RELEASE TO GROUNDWATER Regenerant Waste Evaporator Feed rate 25.7 gpm Concentration factor 42.8 Total operating volume 4,200 gal Feed stream (regenerant waste tank)
See Table 15.7-9 Nearest Municipal Surface Water Supply -
Bayou Lafourche, Louisiana Travel time 8.72 yr Dilution factor 1.72+10 NOTE:
1.72 +10 = 1.72x1010
RBS USAR 1 of 1 August 1987 TABLE 15.7-9 REGENERANT WASTE TANK INVENTORY Isotope Ci/cc Isotope Ci/cc Na-24 5.4-2 Ag-110m 1.4-3 P-32 3.9-3 Te-129m 8.1-3 Cr-51 1.2-1 Te-131m 5.6-3 Mn-54 1.4-3 Te-132 2.4-1 Mn-56 5.3-2 Ba-139 1.1-1 Fe-55 2.1-2 Ba-140 2.0-1 Fe-59 1.8-3 Ba-141 3.0-2 Co-58 1.2-1 Ba-142 1.7-2 Co-60 1.1-2 La-142 3.2-2 Ni-63 2.1-5 Ce-141 3.7-3 Ni-65 3.1-4 Ce-143 1.8-3 Cu-64 1.4-1 Ce-144 8.6-4 Zn-65 4.2-3 Pr-143 4.8-3 Zn-69m 1.0-2 Nd-147 3.5-4 Sr-89 7.3-2 W-187 2.7-2 Sr-90 5.6-3 Np-239 3.8+0 Sr-91 3.1-1 Br-83 3.7-1 Sr-92 1.6-1 Br-84 8.3-2 Y-91 8.7-3 Br-85 4.7-3 Y-92 2.0-1 I-131 4.9+0 Y-93 9.1-2 I-132 3.5+0 Zr-95 1.0-3 I-133 3.1+1 Zr-97 2.4-4 I-134 2.6+0 Nb-95 9.8-4 I-135 8.5+0 Nb-98 6.6-3 Rb-89 2.4-3 Mo-99 3.8-1 Cs-134 3.8-3 Tc-99m 2.7-1 Cs-136 2.3-3 Tc-101 3.2-2 Cs-137 1.0-2 Tc-104 4.6-2 Cs-138 5.5-2 Ru-103 2.7-3 Ru-105 1.8-2 Ru-106 3.8-4 NOTE:
5.4-2 = 5.4x10-2
RBS USAR 1 of 1 August 1987 TABLE 15.7-10 RADWASTE EQUIPMENT FAILURE ACCIDENT RADIOACTIVITY CONCENTRATIONS AT BAYOU LaFOURCHE WATER SUPPLY Final Activity Fraction of Maximum Isotope (Ci/cc)
Permissible Concentrations*
I-129 1.5-22 2.4-15 Sr-90 1.1-11 3.8-05 Y-90 1.1-11 5.7-07 Ru-106 2.4-15 2.4-10 Cs-134 5.0-13 5.6-08 Cs-137 2.1-11 1.0-06 Ce-144 9.5-16 9.5-11 Pm-147 1.0-15 5.1-12 Mn-54 3.2-15 3.2-11 Fe-55 5.6-12 7.0-09 Co-60 8.7-12 1.7-07 Zn-65 1.3-15 1.3-11 Ag-110m 5.9-16 2.0-11 Ni-63 4.9-14 1.6-09 TOTAL 4.0-05
- Maximum permissible concentrations are from 10CFR20, Appendix B, Table II. column 2.
NOTE 1.5-22 = 1.5x10-22
RBS USAR Revision 21 1 of 1 TABLE 15.7-11 FUEL HANDLING ACCIDENT RADIOLOGICAL CONSEQUENCE ANALYSIS PARAMETERS Design Basis Input Description of Input/Assumption and/or Assumption I.
Data and assumptions used to Estimate radioactive source from postulated accident.
- 1. Power Level 3100 MWt 2.
Number of damaged rods (GE 9x9) 150 3.
Total number of rods in core 46,176 (Limiting GE 9x9 Fuel)
- 4. Core Activity available for release Table 15.7-11A
- 5. Radial peaking factor 2.00
- 6. Gap Activity Release Fractions RG 1.183
- 7. Release fission product species and chemical form RG 1.183, Appendix B
- 8. Decay time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> II.
Data and assumptions used to Estimate Activity released to the environment.
- 1. Building Release Rate 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> linear release rate
- 2. Halogen Decontamination Factor 200 III. Dispersion Data
- 1. EAB X/Q Data 0-2 hrs 8.58E-04 sec/m3
- 2. LPZ X/Q Data 0-8 hrs 1.13E-04 sec/m3 8-24 hrs 7.89E-05 sec/m3 1-4 days 3.65E-05 sec/m3 4-30 days 1.21E-05 sec/m3
- 3. Control Room X/Q Data 0-20 mm 1.62E-03 sec/m3 20 min-8 hrs 4.05E-04 sec/m3 8-24 hrs 3.00E-04 sec/m3 1-4 days 1.01E-04 sec/m3 4-30 days 1.62E-05 sec/m3 IV.
Control Room Parameters
- 1. Free Air Volume 188,000 ft3
- 2. Unfiltered In-leakage Rate 300 cfm
- 3. Outside Air Ventilation Rate 1700 cfm
- 5. Control Room Breathing Rates and Occupancy Factors RG 1.183
RBS USAR Revision 22 1 of 1 TABLE 15.7-11A FUEL HANDLING ACCIDENT CORE ACTIVITY AT REACTOR SHUTDOWN (i.e., Decay Time = 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />)
Isotope EOC Core Inventory (Ci/MWt)
I-131 2.70E+04 I-132 3.92E+04 I-133 5.52E+04 I-134 6.06E+04 I-135 5.17E+04 Kr-85 3.66E+02 Kr-85m 7.02E+03 Kr-87 1.35E+04 Kr-88 1.89E+04 Xe-133 5.26E+04 Xe-135 1.99E+04
RBS USAR NOTE: 1.90E+02 = 1.90x102 Revision 22 1 of 1 TABLE 15.7-12 FUEL HANDLING ACCIDENT ACTIVITY RELEASED TO ENVIRONMENT Isotope Gap Activity
[Ci]
(t=0 hrs)
Gap Activity
[Ci]
(t=24 hrs)
Released to Environment
[Ci]
Kr-83m
- N/A
- N/A
- N/A Kr-85 7.37E+02 7.37E+02 7.37E+02 Kr-85m 7.07E+03 1.72E+02 1.72E+02 Kr-87 1.36E+04 2.83E-02 2.83E-02 Kr-88 1.90E+04 5.44E+01 5.44E+01 Kr-89
- N/A
- N/A
- N/A I-131 4.35E+04 3.99E+04 2.00E+02 I-132 3.95E+04 2.85E+01 1.43E-01 I-133 5.56E+04 2.50E+04 1.25E+02 I-134 6.10E+04 3.50E-04 1.75E-06 I-135 5.21E+04 4.20E+03 2.10E+01 I-136
- N/A
- N/A
- N/A Xe-131m
- N/A
- N/A
- N/A Xe-133m
- N/A
- N/A
- N/A Xe-133 5.30E+04 4.64E+04 4.64E+04 Xe-135m
- N/A
- N/A
- N/A Xe-135 2.00E+04 3.21E+03 3.21E+03 Xe-137
- N/A
- N/A
- N/A Xe-138
- N/A
- N/A
- N/A
- 13 *10 10* 13*
RBS USAR Revision 21 1 of 1 TABLE 15.7-13 FUEL HANDLING ACCIDENT RADIOLOGICAL CONSEQUENCES Regulatory Limit FHA Dose Receptor (REM TEDE)
EAB 6.3 2.6 LPZ 6.3 0.4 Control Room 5
1.7
13
10
8 2 2 13
8 10
RBS USAR 1 of 1 August 1987 TABLE 15.7-14 MAXIMUM CONDENSATE STORAGE TANK INVENTORY Isotope Ci/cc Isotope Ci/cc Br-83 3.2-13 Cs-134 7.2-6 I-129 6.3-17 Cs-136 3.8-6 I-131 3.7-4 Cs-137 1.8-5 I-132 1.9-4 Ba-137m 1.7-5 I-133 5.3-4 Ba-139 2.3-19 I-135 2.6-6 Ba-140 1.8-4 Sr-89 7.2-5 Ba-141 5.8-77 Sr-90 5.9-5 La-140 1.4-4 Sr-91 8.6-6 La-141 7.2-10 Sr-92 1.3-11 La-142 1.3-18 Y-90 3.3-5 CE-141 5.9-6 Y-91m 5.7-6 Ce-143 7.8-7 Y-91 1.7-5 Ce-144 8.5-7 Y-92 4.4-9 Pr-143 4.4-6 Y-93 3.0-6 Pr-144 8.5-7 Zr-95 9.9-7 Nd-147 3.3-7 Zr-97 3.7-8 Pm-147 8.3-10 Nb-95m 9.0-9 Na-24 2.0-5 Nb-95 1.0-6 P-32 1.1-5 Nb-97m 3.6-8 Cr-51 3.5-4 Nb-97 4.0-8 Mn-54 4.6-6 Mo-99 2.7-4 Mn-56 6.2-12 Tc-99m 2.6-4 Fe-55 6.5-5 Ru-103 2.6-6 Fe-59 5.1-6 Ru-105 1.9-9 Co-58 3.4-4 Rh-103m 2.6-6 Ni-63 6.5-8 Rh-105m 1.9-9 Ni-65 2.9-14 Rh-105 6.1-10 Cu-64 3.3-5 Rh-106 4.0-7 Zn-65 1.3-5 Te-129m 7.8-6 Zn-69m 2.8-6 Te-129 7.9-6 Ag-110m 4.0-6 Te-131m 2.2-6 Ag-110 8.0-8 Te-131 4.5-7 W-187 2.4-5 Te-132 1.8-4 Np-239 2.4-3 NOTE:
3.2-13 = 3.2x10-13
RBS USAR 1 of 2 August 1987 TABLE 15.7-15 CONDENSATE STORAGE TANK RUPTURE ACCIDENT RADIOACTIVITY CONCENTRATIONS AT BAYOU LAFOURCHE WATER SUPPLY Final Activity Fraction of Maximum Isotope (Ci/cc)
Permissible Concentration*
Br-83 8.0-17 2.7-11 I-129 1.6-20 2.6-13 I-131 9.3-8 3.1-1 I-132 4.6-8 5.8-3 I-133 1.3-7 1.3-1 I-135 6.5-10 1.6-4 Sr-89 1.8-8 6.0-3 Sr-90 1.5-8 5.0-2 Sr-91 2.2-9 3.1-5 Sr-92 3.3-15 4.7-11 Y-90 8.3-9 4.1-4 Y-91m 1.4-9 4.7-7 Y-91 4.3-9 1.4-4 Y-92 1.1-12 1.9-8 Y-93 7.4-10 2.5-5 Zr-95 2.5-10 4.1-6 Zr-97 9.4-12 4.7-7 Nb-95m 2.3-12 7.5-7 Nb-95 2.6-10 2.6-6 Nb-97 1.0-11 1.1-8 Mo-99 6.8-8 3.4-4 Tc-99m 6.5-8 1.1-5 Ru-103 6.4-10 8.0-6 Ru-105 4.7-13 4.7-9 Ru-106 1.0-10 1.0-5 Rh-103m 6.4-10 6.4-8 Rh-105 1.5-13 1.5-9 Te-129m 2.0-9 6.6-5 Te-129 2.0-9 2.5-6 Te-131m 5.6-10 9.4-6 Te-132 4.5-8 1.5-3 Cs-134 1.8-9 2.0-4 Cs-136 9.6-10 1.1-5 Cs-137 4.6-9 2.3-4 Ba-140 4.6-8 1.5-3 La-140 3.6-8 1.8-3 La-141 1.8-13 6.0-8 Ce-141 1.5-9 1.6-5 Ce-143 2.0-10 4.9-6 Ce-144 2.1-10 2.1-5
RBS USAR 2 of 2 August 1987 TABLE 15.7-15 (Cont)
Final Activity Fraction of Maximum Isotope (Ci/cc)
Permissible Concentration*
Pr-143 1.1-9 2.2-5 Nd-147 8.4-11 1.4-6 Pm-147 2.1-13 1.0-9 Na-24 4.9-9 2.4-5 P-32 2.8-9 1.4-4 Cr-51 8.7-8 4.4-5 Mn-54 1.1-9 1.1-5 Mn-56 1.6-15 1.6-11 Fe-55 1.6-8 2.0-5 Fe-59 1.3-9 2.2-5 Co-58 8.5-8 8.5-4 Co-60 8.8-9 1.8-4 Ni-63 1.6-11 5.4-7 Ni-65 7.2-18 7.2-14 Cu-64 8.3-9 2.8-5 Zn-65 3.2-9 3.2-5 Zn-69m 7.0-10 9.9-6 Ag-110m 1.0-9 3.4-5 W-187 5.9-9 8.4-5 Np-239 6.0-7 6.0-3 Total 5.2-1
- Maximum permissible concentrations are from 10CFR20, Appendix B, Table II, Column 2 NOTE:
8.0-17 = 8.0x10-17
RBS USAR 1 of 1 August 1987 TABLE 15.7-16 CONDENSATE STORAGE TANK RUPTURE ACCIDENT RADIOACTIVITY CONCENTRATIONS AT WELL 56 (GROUNDWATER)
Final Activity Fraction of Maximum Isotope (Ci/cc)
Permissible Concentration*
I-129 1.3-19 2.1-12 Sr-90 4.9-8 1.6-1 Y-90 4.9-8 2.4-3 Ru-106 5.0-13 5.0-8 Cs-134 2.8-10 3.1-5 Cs-137 1.5-8 7.6-4 Ce-144 1.5-13 1.5-8 Pm-147 3.7-13 1.8-9 Mn-54 1.8-12 1.8-8 Fe-55 5.4-9 6.8-6 Co-60 1.0-8 2.0-4 Zn-65 5.2-13 5.2-9 Ag-110m 2.4-13 7.9-9 Ni-63 6.2-11 2.1-6 Total 1.6-1
- Maximum permissible concentrations are from 10CFR20, Appendix B, Table II, Column 2 NOTE:
1.3-19 = 1.3x10-19
RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
FUEL HANDLING ACCIDENT - (INSIDE CONTAINMENT) PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSIS DURING TYPE C LEAK RATE TESTING TABLE 15.7-17 REVISION 17 THIS TABLE HAS BEEN DELETED
RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
FUEL HANDLING ACCIDENT - (INSIDE CONTAINMENT DURING TYPE C LEAK RATE TESTING) RADIOLOGICAL EFFECTS TABLE 15.7-18 REVISION 17 THIS TABLE HAS BEEN DELETED
RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
FUEL HANDLING ACCIDENT - (INSIDE CONTAINMENT) PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSIS WITH CONTAINMENT AIR LOCKS OPEN TABLE 15.7-19 REVISION 17 THIS TABLE HAS BEEN DELETED
RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORT
FUEL HANDLING ACCIDENT - (INSIDE CONTAINMENT WITH CONTAINMENT AIR LOCKS OPEN) RADIOLOGICAL EFFECTS TABLE 15.7-20 REVISION 17 THIS TABLE HAS BEEN DELETED
- 11 11* 14*
Revision 14 1 of 1 September 2001
- 10 Table 15.8-1 Initial Conditions for ATWS Analysis Parameter Units Value
- 14 Core Power Mwt (% Rated) 3039 Dome Pressure psig 1055 Core Flow Mlbm/hr (% Rated) 68.4 (81)
Steam Flow Mlbm/hr 13.199 Feedwater Flow Mlbm/hr 13.169 RPV Water Level ft above separator skirt 4.2 Core Average Void Fraction 47 Feedwater Enthalpy BTU/lbm 402.6 Initial Void Reactivity Coefficient cents/%
-11.2 Initial Suppression Pool Temperature
°F 100
- 11 Initial Suppression Pool Volume At Minimum Water Level ft3 135,500 11*
10*
- 10 Table 15.8-2 ATWS - Equipment Performance Characteristics Name Unit Value Revision 14 1 of 2 September 2001 Nominal Closure Time of MSIV
- 14 sec 4.0 Relief Valve System Capacity (Relief mode, all 16 SRVs assumed to be in service)
% Rated Steam Flow/
Number of Valves 101/16 Relief Valve Setpoint Range psia 1178-1198 Relief Valve and Sensor Time Delay sec 0.4 Relief Valve Opening Time sec 0.15 Relief Valve Closure Time Delay sec 0.3 SLCS Injection Location Standpipe Sodium Pentaborate Solution Concentration in the Storage Tank
- 12
% by weight 9.5 Nominal Boron 10 Enrichment atom %
60 Nominal SLCS Boron Injection rate 12*
gpm 41.2 SLCS Initiation Method Manual RCIC Flow Rate gpm 600 RCIC Start/Stop Levels L2/L8 ATWS High Pressure RPT Setpoint, Upper Analytical Limit (UAL) psig 1180 10* 14*
- 10 Table 15.8-2 (Cont)
ATWS - Equipment Performance Characteristics Name Unit Value Revision 14 2 of 2 September 2001
- 14 ATWS Dome Pressure Sensor and Logic Time Delay 14*
sec 0.1 ATWS Low Water RPT Setpoint L2 Recirculation Pump System Inertia Constant sec 7
RHR Pool Cooling Capacity (each)
BTU/sec 390 Setpoint for Low Water Level Closure of MSIV L1 Setpoint for Low Steamline Pressure Closure of MSIV psig 860 Service Water Temperature
°F 95 10* *14 *12 12* 14*
RBS USAR Revision 14 1 of 1 September 2001
- 10 Table 15.8-3 Typical Sequence of Events of ATWS Main Steamline Isolation Valve Closure Event Event Time (sec)
Main steam line isolation valves begin to close.
Control rods do not insert in response to reactor protection system logic.
- 14 0
Main steam line isolation valves closed.
4 Peak Neutron Flux.
4.0 ATWS high pressure setpoint reached.
4.25 Recirculation pumps trip.
4.39 Main steam safety/relief valves begin to lift in relief mode.
4.46 Peak vessel pressure reached.
4.71 Peak Heat Flux.
4.81 Suppression pool temperature reaches 110°F.
Operator initiates level reduction and inhibits the automatic depressurization system logic.
29 Operator initiates the standby liquid control system.
124 Boron from Standby liquid control system reaches the reactor core.
230 Operator initiates RHR in the suppression pool cooling mode.
600 Peak suppression pool temperature achieved.
1480 Hot shutdown achieved (neutron flux < 1% of Nuclear Boiler Rated).
1513 10* 14*
RBS USAR 10 Table 15.8-4 Summary of Peak Results 1
Revision 25 1 of 1 14 11 Event 11 Neutron Flux Surface Heat Flux Bottom Head Pressure Fuel cladding Temperature 2
Suppression Pool Temperature
% Rated
@ Time (sec)
% Rated
@ Time (sec) psig
@ Time (sec)
F
@ Time (sec)
F
@ Time (sec)
MSIV Closure GE-11 GE-8 293 4
131 4.8 1293.9 4.7 736 1185 5
35 177.3 1513 Pressure Regulator Failures - Maximum Demand (3)
GE-11 GE-8 358 21.9 146 25.3 1295.7 25.2 1586 1431 66 66 174.6 1220 Inadvertent Open Relief Valve GE-11 GE-8 100 0.0 100 0.0 1082.2 0.0 145.2 5380 1 Results are representative of a typical reload cycle.
2 Cladding temperature only calculated for the MSIV closure and pressure regulator - failure downscale cases as these cases result in the highest surface heat flux.
11 10 11 14 3 The pressure regulator failure-open is no longer a postulated single failure scenario; no Ovation TCPS single failure scenarios are postulated that could result in all valves open. This specific USAR section is maintained for historical purposes.