ML17226A117
Text
RBS USAR TABLE 15.0-1 14 RESULTS
SUMMARY
OF TRANSIENTS EVENTS APPLICABLE TO BWRs (For Original Rated Power 2894 Mwt) 14 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) 14 (For Original Rated Power 2894 Mwt) 14 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 Para g ra p h 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 Para g ra p h 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 RBS USAR TABLE 15.0-1A (Cont)
Revision 17 3 of 5 Para g ra p h 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 15.4REACTIVITY 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 122.89931,007987161.14(2) a 0 0 RBS USAR TABLE 15.0-1A (Cont)
Revision 17 4 of 5 Para g ra p h 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 472.49981,009984149.56(2) a 0 015.4.5 15.4-3 Fast opening of both main recirc
valves 353.79821,005978140.50(2) a 0 0 15.4.7Misplaced 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 RBS USAR TABLE 15.0-1B (Cont)
Revision 14 2 of 4 September 2001 TRANSIENT NAME ODYN PID A NALYSIS 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 RBS USAR TABLE 15.0-1B (Cont)
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 RBS USAR TABLE 15.0-1B (Cont)
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, o F 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-oF0.1892 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 (-)¢/
o F 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.A PRM simulated thermal power trip, % NBR Analysis set point (114 x 1.042) 118.8 34.Time constant, sec 7 35.Nuclear characteristics usedin 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 J o n g T o where:
t = Inertia time constant (sec)
J o = Pump motor inertia (lb-ft) n = Rated pump speed (rps) g = Gravitational constant (ft/sec)
T o = 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 Power5% Power Uprate1. 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 84.5x10 668.6 x10 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 425 425.76. Vessel dome pressure, psig 1,045 10557. Vessel core pressure, psig 1,056 10708. Turbine bypass capacity, % NBR 10 9.489. Core coolant inlet enthalpy, Btu/lb 529.9 531.210. Turbine inlet pressure, psig 960 1012 (6)11. Fuel lattice P8x8R GE8x8EB GE1112. Core average gap conductance, Btu/sec-ft 2-°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 RBS USAR TABLE 15.0-2A (Cont)
Revision 17 2 of 3 Original Power5% 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 Same24. Relief function response Time constant, sec 0.10 Same 25. Safety function delay, sec 0.0 Same26. 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. A PRM 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 Same37. 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, ft 3 3243.6 3,27539. 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 J o n t = gT o where: t = Inertia time constant (sec)
J o = Pump motor inertia (lb-ft) n = Rated pump speed (rps) g = Gravitational constant (ft/sec)
T o = 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 SectionTitleFailed Fuel RodsGE-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 1of1August1987TABLE 15.1-1SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATER,AUTO FLOW CONTROL (FIGURE 15.1-1)Time(sec) Event 0Initiate a 100°F temperature reduction in the feedwater system 5Initial effect of unheated feedwater starts to raise core power level but theautomatic flow control system automatically reduces core flow to maintaininitial steam flow 40Reactor variables settle into new steady state RBS USAR 1of1August1987TABLE 15.1-2SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATER,MANUAL FLOW CONTROL (FIGURE 15.1-2)Time(sec) Event 0Initiate a 100°F temperature reduction into the feedwater system 5Initial effect of unheated feedwater starts to raise core power level andsteam flow 10Turbine control valves start to open to regulate pressure 61.9Initiation of reactor scram on high simulated thermal power 73.0Narrow range (NR) sensed water level reaches Level 3 (L3) set point 73.2Trip of recirculation pump power source to low frequency MG speed; RPTinitiates due to Level 3 Trip (not included in simulation)>80(est)Wide range (WR) sensed water level reaches Level 2 (L2) set point
>80Recirculation 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 USARRevision141of1September2001TABLE 15.1-3SEQUENCE OF EVENTS FOR FEEDWATER CONTROLLER FAILURE,MAXIMUM DEMAND (FIGURE 15.1-3)Time(sec) Event14 0 Initiate simulated failure of 108 percent upper limit on feedwater flow at asystem design pressure of 1,065 psig32.3L8 vessel level set point initiates reactor scram and trips main turbine andfeedwater pumps32.4RPT actuated by stop valve position switches32.4Main turbine bypass valves opened due to turbine trip 33.8SRVs open due to high pressure 36Water 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 1of1August1987TABLE 15.1-5SEQUENCE OF EVENTS FOR INADVERTENTSAFETY/RELIEF VALVE OPENINGTime-sec Event 0Initiate opening of one SRV 0.5 (est)Relief flow reaches full flow 15 (est)System establishes new steady-state operation RBS USAR 1of1August1987TABLE 15.1-6SEQUENCE OF EVENTS FOR INADVERTENT RHRSHUTDOWN COOLING OPERATIONApproximateElapsed Time Event 0Reactor at states B or D (Appendix 15A) when RHR shutdown cooling inadvertently activated 0-10 minSlow rise in reactor power +10 minOperator may take action to limit power rise; Flux scram occurs if noaction 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 AreaConcentration Boundary ReleasedConcentration Percent Isotope (Ci/hr)
( Ci/cc) ECL (1) of ECL14 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 6.5-6Kr-83m 2.8-1 2.5-10 5.0-5 5.1-4Kr-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+1Xe-131m 2.9-3 2.7-12 2.0-6 1.44Xe-133m 4.9-2 4.5-11 6.0-7 7.5-3Xe-133 1.4+0 1.3-9 5.0-7 2.5-1Xe-135m 8.0+0 7.3-9 4.0-8 1.8+1Xe-135 4.3+0 3.9-9 7.0-8 5.6+0Xe-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 Conditions14 Rated power (%)
10 0.3 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 1of1August1987TABLE15.3-1SEQUENCEOFEVENTSFORTRIPOFONERECIRCULATIONPUMP(FIGURE15.3-1)
Time (sec)Event0Tripofonerecirculationpumpinitiated5Jetpumpdiffuserflowreversesinthetripped loop39.0Coreflowandpowerlevelstabilizeatnewequilibriumconditions.
1of1August1987TABLE15.3-2SEQUENCEOFEVENTSFORTRIPOFTWORECIRCULATIONPUMPS(FIGURE15.3-2)
Time (sec)Event0Tripofbothrecirculationpumpsinitiated4.2Vesselwaterlevel(L8)tripinitiatesscram,turbinetripandfeedwaterpumptrip4.3Turbinetripinitiatesbypassoperation5.8SRVsopenduetohighpressure11.2SRVsclose17.2Vesselwaterlevel(L2)setpointreached47.2(est)HPCSandRCICflowentersvessel(not simulated) 1of1August1987TABLE15.3-3SEQUENCEOFEVENTSFORFASTCLOSUREOFONEMAINRECIRCULATIONVALVE(FIGURE15.3-3)
Time (sec)Event 0Initiatefastclosureofonemainrecirculation valve 2Jetpumpdiffuserflowreversesintheaffected loop40(est)Coreflowandpowerapproachnewequilibrium conditions 1of1August1987TABLE15.3-4SEQUENCEOFEVENTSFORFASTCLOSUREOFTWOMAINRECIRCULATIONVALVES(FIGURE15.3-4)
Time (sec)Event0Initiatefastclosureofbothmainrecirculation valves5.15Vessellevel(L8)tripinitiatesscramandturbinetrip5.15Feedwaterpumpstrippedoff5.30Turbinetripinitiatesbypassoperation 6.46SRVsopenduetohighpressure12.31SRVsclose17.50VesselwaterlevelreachesLevel2setpoint 47.50(est)HPCSandRCICflowentersvessel(notsimulated) 1of1August1987TABLE15.3-5SEQUENCEOFEVENTSFORRECIRCULATIONPUMPSEIZURE(FIGURE15.3-5)
Time (sec)Event 0Singlepumpseizurewasinitiated.
0.8Jetpumpdiffuserflowreversesinseizedloop.3.11Vessellevel(L8)tripinitiatesreactorscram.3.11Vessellevel(L8)tripinitiatesturbineandfeedwaterpumptrips.3.30Turbinetripinitiatesbypassoperation.3.35Turbinetripinitiatesrecirculationpumps trip.4.85SRVsopenduetohighpressure.
10.2SRVsclose.
12.7VesselwaterlevelreachesLevel2setpoint.42.7(est)HPCS/RCICflowentersthevessel(notsimulated).
__________________________
- 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 8Revision 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 *
- 81. 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 RBS USAR 8__________________________________
NOTE: See Appendix 15B for reload core conditions 8Revision 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 k eff 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
- 2. Number of damaged rods 100% Power Event
Low Power Event (gap release)
850 GE 8x8
50 GE 8x8
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
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 />
- 2. Condenser Iodine Release
Fractions Per RG 1.183, Appendix C
97% Elemental, 3% Organic
RBS USAR TABLE 15.4-11 (Cont) 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 ft 3
C. Dispersion Data
- 1. EAB X/Q Data 0-2 hrs 7.51E-04 sec/m 3 2. LPZ X/Q Data 0-8 hrs 7.79E-05 sec/m 3 8-24 hrs 5.23E-05 sec/m 3 1-4 days 2.21E-05 sec/m 3 4-30 days 6.40E-06 sec/m 3
- 3. Control Room X/Q Data Main Air Intake 0-2 hrs 3.02E-3 sec/m 3 2-8 hrs 2.47E-3 sec/m 3 8-24 hrs 1.05E-3 sec/m 3 1-4 days 9.01E-4 sec/m 3 4-30 days 6.74E-4 sec/m 3 D. Control Room Parameters
- 1. Free Air Volume 188,000 ft 3
- 2. Unfiltered In-leakage Rate 300 cfm
- 3. Outside Air Ventilation Rate 1700 cfm
Aerosol 0% 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+0 3 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 10 2
RBS USAR TABLE 15.4-13 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 RBSUSAR 1of1 August1987RBSUSARTABLE15.5-1SEQUENCEOFEVENTSFORINADVERTENTSTARTUPOFHPCS(FIGURE15.5-1)
Time (sec)Event 0SimulateHPCScoldwaterinjection 3FullflowestablishedforHPCS 7Depressurizationeffectstabilized RBS USAR *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) Event0 Guillotine break of one main steam line outside primary containment 0.5 High steam line flow signal initiates closure of (Approx.) MSIVs <1.0 Reactor begins scram 5.5 MSIVs fully closed 1060 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 AssumptionI. Data and assumptions used to estimate radioactive source from postulated accident. 1. Power Level 3100 MWt 2. Maximum Pre-accident Spike Iodine Concentration 4 µCi/gm DE I-131 3. Maximum Equilibrium Iodine Concentration 0.2 µCi/gm DE I-131 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 hotstandby 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 RBS USAR TABLE 15.6-2 (Cont)
Revision 17 2 of 2 Description of Input/Assumption Design Basis Input and/or AssumptionIII. Dispersion Data 1. EAB X/Q Data 0-2 hrs 6.33E-04 sec/m
- 32. LPZ X/Q Data 0-8 hrs 8-24 hrs 1-4 days 4-30 days 7.57E-05 sec/m 3 5.08E-05 sec/m 3 2.13E-05 sec/m 3 6.24E-06 sec/m
- 33. 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/m 3 1.08E-03 sec/m 3 4.57E-04 sec/m 3 3.50E-04 sec/m 3 2.58E-04 sec/m 3IV. Control Room Parameters 1. Free Air Volume 188,000 ft
- 32. 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 113 3 11 RBS USAR ____________________________
NOTE: 1.9E+01 = 1.9x10 1 Revision 17 1 of 1 TABLE 15.6-3 MAIN STEAM LINE BREAK ACTIVITY RELEASE D 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 1411 1411* 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) (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 0.2µCi/gm DE I-131 0.2 5 14113 3 1114* 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 AssumptionsI. 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 1310II. 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 1 50 scfh =
2.5 2 cfm* 25 minutes - 30 days 0 scfh D. Engineered Safety Features Leakage 1 gpm
- 1 50 scfh was converted based on a maximum DW temperature of 330û 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 = 1.16x10 6 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 ft 3 B. Containment 1.19E+06 ft 3 C. Annulus
- 3.57E+05 ft 3 D. Auxiliary Building
- 1.16E+06 ft 3 E. Control Room 1.88E+05 ft 3 F. Suppression Pool
- 1.25E+05 ft 3*Only 50% of the auxiliary building volume w as credited in the actual analysis (values listed are actual volumes) ** The Actual analysis conservatively assumed a volume of 120,000 ft 3.***The annulus volume in this analysis was 1.0 ft 3 to minimize mixing of the annulus atmosphere reflecting disabling of the annulus mixing system. 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 RBS USAR ____________________
NOTE: 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 14 13 10 BWR CORE INVENTORIES AST Group Isotope Core Inventory (CI/MWT) 5 Co-58 2.61E+02 5 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 = 6.81x10 3
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 RBS USAR 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 RBSUSAR1of1August1987TABLE15.6-8SEQUENCEOFEVENTSFORFEEDWATERLINEBREAKOUTSIDECONTAINMENT Time (sec)Event 0Onefeedwaterlinebreaks 0+Feedwaterlinecheckvalvesisolatethereactorfromthebreak5(approx)Reactorscramonlowwaterlevel
<30HPCSandRCICstartonlow-lowwaterlevel,L2,andareexpectedtomaintainthewaterlevelabovethelow-low-lowlevel,L1,trip,andeventuallyrestoreittothenormalelevation1to 2hrNormalreactorcooldownprocedureestablished RBSUSAR1of1August1987TABLE15.7-1SEQUENCEOFEVENTSFORMAINCONDENSEROFFGASTREATMENTSYSTEMFAILURE ApproximateElapsedTime Event0secEventbegins-systemfails.0secNoblegasesarereleased.<1minArearadiationalarmsalertplant
personnel.<1minOperatoractionsbeginwith:1.Initiationofappropriatesystem isolations2.Manualscramactuation 3.Assuranceofreactorshutdown cooling.1hrSteamjetairejectorisshutdown.
RBSUSAR1of2August1987TABLE15.7-2GASEOUSRADWASTESYSTEMFAILURE-PARAMETERSTABULATEDFORPOSTULATEDACCIDENTANALYSISDesignBasis AssumptionsI.DataandassumptionsusedtoestimateradioactivesourceforpostulatedaccidentsA.Powerlevel NAB.Burnup NAC.Fueldamage NoneD.ReleaseofactivityTable15.7-3E.Iodinefractions NA1.Organic NA2.Elemental NA3.Particulate NAF.Reactorcoolantactivitybeforetheaccident NAII.Dataandassumptionsusedtoestimateactivityreleased A.Containmentleakrate(%/day)
NAB.Secondarycontainmentleak rate(%/day)
NAC.Valvemovementtimes NAD.Absorptionandfiltration efficiencies NA1.Organiciodine NA2.Elementediodine NA3.Particulateiodine NA4.Particulatefission products NAE.Recirculationsystem parameters NA1.Flowrate NA2.Mixingefficiency NA3.Filterefficiency NAF.Contaimentvolumes NAG.Allotherpertinentdataandassumptions None RBSUSARNOTE:8.58-4=8.58x10
-4Revision32of2August1990TABLE15.7-2(Cont)DesignBasis AssumptionsIII.Dispersiondata3A.EABdistance(m) 894B.X/QsforEAB(sec/m-)
8.58-4 3IV.DosedataA.Methodofdosecalculation NAB.DoseconversionassumptionsReg.Guides1.98and1.109C.Peakactivityconcentrationsin
containment NAD.DosesTable15.7-4 RBSUSAR1of1August1987TABLE15.7-3EQUIPMENTFAILURERELEASEASSUMPTIONSRELEASEFRACTIONSASSUMEDFORDESIGNBASISANALYSISNobleParticulateEquipmentPieceGases Daughters RadioiodineHolduppipe1.001.00 NA Prefilter NA 0.01 NA Charcoal adsorbers1.000.01 NASteamjetair ejector(1-hr
release)1.001.00 NA RBSUSARRevision141of1September2001TABLE15.7-4GASEOUSRADWASTESYSTEMFAILURESYSTEMRUPTURE(DESIGNBASISANALYSIS)OFFSITERADIOLOGICALEFFECTSDoseatExclusionAreaBoundarySkinWholeBodyNobleGases (rem)(rem)Charcoalbed 6.0-1 6.8-1Holduppipe 1.2-2 2.3-2Steamjetairejector 3.8-1 5.1-1(1-hrrelease)
Particulates14 Prefilter-5.6-2Charcoalbed
-1.9-4Holduppipe
-3.2-4Steamjetairejector
-2.1-3(1-hrrelease) 14TOTAL 9.9-1 1.3+0NOTE:6.0-1=6.0x10
-1 RBSUSAR1of1August1987TABLE15.7-5ACCIDENTANALYSISASSUMPTIONSRADIOACTIVELIQUIDWASTESYSTEMLEAKORFAILURE(RELEASETOATMOSPHERE)1.DesignbasesactivitiesofSection11.1,andflowsandfractionsofprimarycoolantactivitiesofSection11.2areusedtodevelopsourceterms.2.Noblegasesarenotconsideredduetoconstantventingofradwastesystemtanks.3.Halogenpartitionfactor0.0014.Exclusionareaboundary/QBreathingRate (sec/m 3)(m 3/sec)0-2hr 8.58-4 3.47-4NOTE:8.58-4=8.58x10
-4 RBSUSARTABLE15.7-6LIQUIDRADWASTESYSTEMTANKSHALOGENINVENTORIESBr-83Br-84Br-85I-131I-132I-133I-134I-135(Ci)(Ci)(Ci)(Ci)(Ci)(Ci)(Ci)(Ci) NOTE:8.4-4=8.4x10
-4Revision141of1September2001Floordraincollector tanks2A,B,andC(total)8.4-41.8-41.1-55.1-39.9-34.2-24.2-31.7-214Wastecollectortanks1A,B,C,andD(total)3.6-17.2-23.6-33.6+03.3+02.0+11.9+07.6+0Recoverysampletanks4A,B,C,andD(total)1.9-41.5-56.8-82.7-31.8-31.4-26.2-44.8-3 14Phaseseparatorandbackwash tanks6A,B,and72.1+14.6+02.7-11.6+36.2+21.6+31.1+24.4+2Regenerantwaste tanks3AandB3.2+17.0+04.0-14.0+22.8+22.2+32.2+27.0+2Wasteandregenerant evaporatorsEV-1and21.5+17.4-1 4.0-3 5.4+3 3.4+2 1.1+4 3.8+1 9.6+214Totalinalltanks6.8+11.2+16.8-17.4+31.2+31.5+43.7+22.1+3 14 RBSUSAR1of1August1987TABLE15.7-7OFFSITEDOSESRESULTINGFROMLIQUIDRADWASTESYSTEMTANKSRUPTUREWholeBodyThyroidGammaBetaExclusionareaboundary (rem)(rem)(rem)0-2hr5.1+04.0-31.8-3NOTE:4.0-3=4.0x10
-3 RBSUSAR1of1August1987TABLE15.7-8ACCIDENTANALYSISDATALIQUIDRADWASTETANKRUPTURERELEASETOGROUNDWATERRegenerantWasteEvaporatorFeedrate25.7gpmConcentrationfactor 42.8Totaloperatingvolume4,200galFeedstream(regenerantwastetank)SeeTable15.7-9NearestMunicipalSurfaceWaterSupply-BayouLafourche, LouisianaTraveltime8.72yrDilutionfactor 1.72+10NOTE:1.72+10=1.72x10 10 RBSUSAR1of1August1987TABLE15.7-9REGENERANTWASTETANKINVENTORY IsotopeµCi/cc IsotopeµCi/ccNa-245.4-2 Ag-110m 1.4-3 P-32 3.9-3 Te-129m 8.1-3Cr-511.2-1 Te-131m 5.6-3Mn-541.4-3 Te-132 2.4-1Mn-565.3-2 Ba-139 1.1-1Fe-552.1-2 Ba-140 2.0-1Fe-591.8-3 Ba-141 3.0-2Co-581.2-1 Ba-142 1.7-2Co-601.1-2 La-142 3.2-2Ni-632.1-5 Ce-141 3.7-3Ni-653.1-4 Ce-143 1.8-3Cu-641.4-1 Ce-144 8.6-4Zn-654.2-3 Pr-143 4.8-3Zn-69m1.0-2 Nd-147 3.5-4Sr-897.3-2 W-187 2.7-2Sr-905.6-3 Np-239 3.8+0Sr-913.1-1 Br-83 3.7-1Sr-921.6-1 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+0Zr-951.0-3 I-133 3.1+1Zr-972.4-4 I-134 2.6+0Nb-959.8-4 I-135 8.5+0Nb-986.6-3 Rb-89 2.4-3Mo-993.8-1 Cs-134 3.8-3Tc-99m2.7-1 Cs-136 2.3-3Tc-1013.2-2 Cs-137 1.0-2Tc-1044.6-2 Cs-138 5.5-2Ru-1032.7-3Ru-1051.8-2 Ru-1063.8-4NOTE:5.4-2=5.4x10
-2 RBSUSAR1of1August1987TABLE15.7-10RADWASTEEQUIPMENTFAILUREACCIDENTRADIOACTIVITYCONCENTRATIONSATBAYOULaFOURCHEWATERSUPPLYFinalActivityFractionofMaximum Isotope (µCi/cc)PermissibleConcentrations*
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*Maximumpermissibleconcentrationsarefrom10CFR20,AppendixB,TableII.column2.NOTE1.5-22=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 AssumptionI.Data and assumptions used to Estimate radioactive source from postulated accident.1. Power Level3100 MWt2.Numberofdamagedrods(GE 9x9) 1503.Totalnumberofrodsincore46
, 176 (Limiting GE 9x9 Fuel)4. Core Activity available for release Table 15.7-11A5. Radial peaking factor 2.006. Gap Activity Release FractionsRG 1.183
- 7. Release fission product species and chemical form RG 1.183, Appendix B8. Decay time24 hoursII. 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
rate2. Halogen Decontamination Factor200III. Dispersion Data1. EAB X/Q Data 0-2 hrs 8.58E-04 sec/m
- 32. LPZ X/Q Data 0-8 hrs 1.13E-04 sec/m 3 8-24 hrs 7.89E-05 sec/m 3 1-4 days 3.65E-05 sec/m 3 4-30 days 1.21E-05 sec/m
- 33. Control Room X/Q Data 0-20 mm 1.62E-03 sec/m 3 20 min-8 hrs 4.05E-04 sec/m 3 8-24 hrs 3.00E-04 sec/m 3 1-4 days 1.01E-04 sec/m 3 4-30 days 1.62E-05 sec/m 3IV. Control Room Parameters1. Free Air Volume 188,000 ft
- 32. Unfiltered In-leakage Rate300 cfm
- 3. Outside Air Ventilation Rate1700 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 = 1.90x10 2 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 1.3 6E+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 CONSEQUENCESRegulatory LimitFHA Dose Receptor (REM TEDE)(REM TEDE)
EAB 6.3 2.6 LPZ 6.3 0.4 Control Room 5 1.7131082 2 138 10 RBSUSAR1of1August1987TABLE15.7-14MAXIMUMCONDENSATESTORAGETANKINVENTORY IsotopeµCi/cc IsotopeµCi/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 2.4-3NOTE:3.2-13=3.2x10
-13 RBSUSAR1of2August1987TABLE15.7-15CONDENSATESTORAGETANKRUPTUREACCIDENTRADIOACTIVITYCONCENTRATIONSATBAYOULAFOURCHEWATERSUPPLYFinalActivityFractionofMaximum Isotope (µCi/cc)PermissibleConcentration*
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 RBSUSAR2of2August1987TABLE15.7-15(Cont)FinalActivityFractionofMaximum Isotope (µCi/cc)PermissibleConcentration*
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*Maximumpermissibleconcentrationsarefrom10CFR20,AppendixB,TableII,Column2NOTE:8.0-17=8.0x10
-17 RBSUSAR1of1August1987TABLE15.7-16CONDENSATESTORAGETANKRUPTUREACCIDENTRADIOACTIVITYCONCENTRATIONSATWELL56(GROUNDWATER)FinalActivityFractionofMaximum Isotope (µCi/cc)PermissibleConcentration*
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*Maximumpermissibleconcentrationsarefrom10CFR20,AppendixB,TableII,Column2NOTE:1.3-19=1.3x10
-19 RIVER BEND STATION UPDATED SAFETY ANALYSIS REPORTFUEL 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 REPORTFUEL 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 REPORTFUEL 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 REPORTFUEL HANDLING ACCIDENT - (INSIDE CONTAINMENT WITH CONTAINMENT AIR
LOCKS OPEN) RADIOLOGICAL EFFECTS TABLE 15.7-20 REVISION 17 THIS TABLE HAS BEEN DELETED RBSUSAR111114Revision141of1September200110Table15.8-1InitialConditionsforATWSAnalysis Parameter Units Value14CorePowerMwt(%Rated) 3039DomePressure psig 1055CoreFlowMlbm/hr(%Rated)68.4(81)SteamFlow Mlbm/hr 13.199FeedwaterFlow Mlbm/hr 13.169RPVWaterLevelftaboveseparator skirt 4.2CoreAverageVoid Fraction---47FeedwaterEnthalpy BTU/lbm 402.6InitialVoidReactivity
Coefficient cents/%-11.2InitialSuppressionPool
TemperatureF 10011InitialSuppressionPool VolumeAtMinimumWater
Level ft 3 135,500 1110 RBSUSAR10Table15.8-2ATWS-EquipmentPerformanceCharacteristics Name Unit ValueRevision141of2September2001NominalClosureTimeof MSIV14 sec 4.0ReliefValveSystemCapacity(Reliefmode,all 16SRVsassumedtobein
service)%RatedSteamFlow/NumberofValves 101/16ReliefValveSetpoint
Range psia 1178-1198ReliefValveandSensor TimeDelay sec 0.4ReliefValveOpeningTime sec 0.15ReliefValveClosureTime
Delay sec 0.3SLCSInjectionLocation
---StandpipeSodiumPentaborate SolutionConcentrationin theStorageTank12%byweight 9.5NominalBoron10 Enrichmentatom%60NominalSLCSBoron Injectionrate 12gpm 41.2SLCSInitiationMethod
---ManualRCICFlowRate gpm 600RCICStart/StopLevels
---L2/L8ATWSHighPressureRPTSetpoint,UpperAnalytical Limit(UAL) psig 1180 1014 RBSUSAR10Table15.8-2(Cont)ATWS-EquipmentPerformanceCharacteristics Name Unit ValueRevision142of2September200114ATWSDomePressureSensorandLogicTime
Delay 14sec 0.1ATWSLowWaterRPT Setpoint---L2RecirculationPump SystemInertiaConstant sec 7RHRPoolCooling Capacity(each)
BTU/sec 390SetpointforLowWater LevelClosureofMSIV
---L1SetpointforLow SteamlinePressure ClosureofMSIV psig 860ServiceWater
TemperatureF 95 1014121214 RBSUSARRevision141of1September200110Table15.8-3TypicalSequenceofEventsofATWSMainSteamlineIsolationValveClosureEvent EventTime(sec)Mainsteamlineisolationvalvesbegintoclose.Controlrodsdonotinsertinresponse toreactorprotectionsystemlogic.14 0 Mainsteamlineisolationvalvesclosed.
4PeakNeutronFlux.
4.0ATWShighpressuresetpointreached.
4.25Recirculationpumpstrip.
4.39Mainsteamsafety/reliefvalvesbegintoliftinreliefmode.
4.46Peakvesselpressurereached.
4.71PeakHeatFlux.
4.81Suppressionpooltemperaturereaches110F.Operatorinitiateslevelreductionandinhibits theautomaticdepressurizationsystemlogic.
29Operatorinitiatesthestandbyliquidcontrol
system.124BoronfromStandbyliquidcontrolsystem reachesthereactorcore.
230OperatorinitiatesRHRinthesuppressionpool coolingmode.
600Peaksuppressionpooltemperatureachieved.
1480Hotshutdownachieved(neutronflux<1%of NuclearBoilerRated).
1513 1014 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 29341314.81293.9 4.7 736 1185 5 35 177.31513 Pressure Regulator
Failures - Maximum Demand (3) GE-11 GE-8 35821.914625.31295.7 25.2
1586 1431 66 66 174.61220 Inadvertent Open Relief
Valve GE-11 GE-8 1000.01000.01082.2 0.0
--- --- ---
--- 145.25380 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.