ML17291A214
| ML17291A214 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 06/30/1994 |
| From: | WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | |
| Shared Package | |
| ML17291A213 | List: |
| References | |
| COLR-94-10, COLR-94-10-R, COLR-94-10-R00, NUDOCS 9407190061 | |
| Download: ML17291A214 (75) | |
Text
940514 17:05 COLR 94-10, Revision 0 Controlled Copy No.
WNP-2 Cycle 10 Core Operating Limits Report June 1994 washington Public Power Supply System 9407190061 '744711 PDR ADOCK 05000397
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940614 )7:N WNP-2 Cycle 10 Core Operating Limits Report LI T F FE
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940514 17:09 WNP-2 Cycle 10 Core Operating Limits Report T F EFFE VE PA
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WNP-2 Cycle 10 Core Operating Limits Report ABLE F NTENT Paae 1.0 INTR D CTI N AND S ARY ........................... 1 2.0 VERA E PLANA LINEAR HEAT E RA RATE APL R LIMIT F R EINTE LSPE IFI ATI N 2, ............ 2 3.0 ALP WER RA PR R E L E A N 2 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 4.0" L R HEA ENERATI N RATE H R L R EIN TE AL E IFI N 24........................... 29 5.0 REEF~RBN S ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 35 Washington Nuclear-Unit 2 COLR 94-10, Revision 0
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9405l6 10:22 1.0 D This report provides the Average Planar Linear Heat Generation Rate {APLHGR) limits, the Minimuin'Critical Power Ratio'MCPR) limits; and the Linear Heat Generation Rate (LHGR) limits. for WNP-2, Cycle 10 as required by Technical Specification 6.9.3.1. As required-by Technical Specifications 6.9.3.2 and 6.9.3.3, these limits were determined using NRC-approved methodology and are established so that all applicable limits of the plant safety analysis are met. The thermal limits for SPC fuel given in this report are documented in the "Cycle 10 Plant Transient Analysis" (Reference 5.1.1) and the "Cycle 10 Reload Analysis" (Reference 5.1.2). The thermal limits determined through the approved methodology are modified for the GE11 and SVEA-96 LFAs as discussed below.
The WNP-2 Cycle 10 core includes four Siemens Power Corporation (SPC), four GE Nuclear Energy (GE), and four ABB Combustion Engineering Nuclear Operations (ABB CENO) Lead Fuel Assemblies (LFAs). The SPC LFAs were inserted during the reload for Cycle 5. The GE and ABB CENO LFAs were inserted at the beginning of Cycle 6 and were designed to be compatible with the reload fuel utilized in Cycle 6. The LFAs are loaded in core locations which analysis has shown to have sufficient thermal margin such that the LFAs are not expected to be the most limiting fuel assemblies on either a nodal or an assembly power basis. The GE Nuclear Energy GEll LFAs are described in the "GE11 Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No. 2, Reload 5, Cycle 6" (Reference 5.3. 1). This reference describes the design goals of the GE11 LFAs and provides support for monitoring the GE11 LFAs at thermal limits based on the SPC 8x8 reload fuel thermal limits.
The ABB CENO SVEA-96 LFAs are described in the "Supplemental Lead Fuel Assembly Licensing Report SVEA-96 LFAs for WNP-2 Summary" (Reference 5.3.2). The process for developing thermal limits for the SVEA-96 LFAs based upon the SPC 8x8 reload fuel thermal limits is described in References 5.3.2 through 5.3.4 The MAPLHGR limits for the GE11 LFAs are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-2]/[81-7J) is applied to account for the different number of fuel pins in the two designs. The MAPLHGR limits for the SVEA-96 LFAs are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-2]/[100-4]) is applied to account for the different number of fuel pins in the two designs. Furthermore, the MAPLHGR limits for the SVEA-96 LFAs are multiplied by the following constants: (a) 1.04 to account for a different estimation of the local power in the output from POWERPLEX compared to ABB CENO methods and (b) 1.02 to account for a different estimation of exposure in the output from POWERPLEX compared to ABB CENO methods.
The MCPR limit is the maximum of (a) the applicable exposure dependent, full power and full flow MCPR limit, (b) the applicable exposure and power dependent MCPR limit, and (c) the flow dependent MCPR limit specified in this report. This stipulation assures that the safety limit MCPR will not be violated throughout the WNP-2 operating regime. Full power MCPR limits are specified to define operating limits at rated power and flow. For the WNP-2 core, the Turbine Trip without Bypass event is limiting for operation at rated power and flow. Power dependent MCPR limits are specified to define operating limits at other than rated power conditions. For the WNP-2 core, the Feedwater Controller Failure event from reduced power Washington Nuclear-Unit 2 COLR 94-10, Revision 0
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is calculated to be more severe than from full power conditions. A flow dependent MCPR is specified to define operating limits at other than rated flow conditions. The reduced flow MCPR limit provides bounding protection for the limiting Recirculation Flow Increase event.
The LHGR limits for the GE11 LFAs are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-2]/[81-7]) is applied to account for the different number of fuel pins in the two designs. The LHGR limits for the SVEA-96 LFAs are taken directly from Reference 5.3.2.
The reload licensing analyses for this cycle provide operating limits for Extended Load Line (ELLLA) operation which extends the power and flow operating regime for WNP-2 up to the 109% rod line which at full power corresponds to 87% of rated flow. The MCPR limits defined in this report are applicable up to 100% of rated thermal power along and below the 109% rod line. The minimum flow for operation at rated power is 87% of rated flow; the maximum is 106%. References 5.1.1 and 5.1.2 and the references in Section 5.4 document the analyses in support of ELLLA operation.
,Preparation, review and approval of this report were performed in accordance with applicable Supply System procedures. The specific topical report revisions and supplements which describe the methodology utilized in this cycle specific analysis are referenced in Section 5.2.
2.0 AV RA E A LINEAR AT E TI NRATE APLH R LIMIT F U E INTE A PE IFI A N,2 The APLHGRs for use in Technical Specification 3.2.1 shall not exceed the, limits shown in Figures 2.1, 2.2, 2.4, and 2.5.when in two-loop operation and in Figures 2.1, 2.3, 2.4, and 2.5 when in single loop operation. The limits for each fuel type as a function of Average Planar Exposure are provided for the SPC reload fuel, the SPC LFAs, the SVEA-96 LFAs, and the GE11 LFAs.
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9406 l0 13:46 3.0 MINIMXJMCRITICALPOWER RATI CPR LIMITF R U E IN TECHNI AL SPECIFICATI N 3.2.3 The MCPR limit for use in Technical Specification 3.2.3 shall be:
1 Greater than or equal to the greater of the limits determined from Tables 3.1a and 3.1b and Figures 3.1 and 3.2a through 3.11b.
Washington Nuclear-Unit 2 COLR 94-10, Revision 0
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9405l4 l6:54 Table 3.la
%NP-2 Cycle 10 MCPR Operating Conditions Cyde Exposures ~ 4500 MWd/MTU SLMCPR ~ 1.07 SPC 8xg SPC 9x9 SPC 9x9 SVEA-96 Condition Limit GB11 LFA LFA Nssu~
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Full Power 1.26 1.24 1.35 1.48 Flow Dcpcndcnt Figure 3.1 Power Dependent" Fig. 3.4a Fig. 3.5a Fig. 3.5a Fig. 3.4a Nsst>>
RPI' Full Power 1.28 1.27 1.43 1.51 Inoperable Flow Dcpcndent Figure 3.1 Power Dependent+ Fig. 3.10a Fig. 3.1la Fig. 3.11a Fig. 3.10a SLO+ NSS Full Power 1.56 1.36 1.36 1.98 Flow Dcpcndcnt None Power Dependent + Fig. 3.2a Fig. 3.3a Fig. 3.3a Fig. 3.5L SLO TSSS Full Power 1.56 136 1.36 1.98 Flow Dcpcndcnt None Power Dcpendc>>P Fig. 3.4a Fig. 3.$ a Fig. 3.$ a Fig. 3.4a SLO+ NSS RFl'ull Power 156 1.36 136 1.98 lnopcrable Flow Dcpendcnt None Power Dependents Fig. 3.10a Fig. 3.11a Fig. 3.11a Fig. 3.10a Washington Nuclear-Unit 2 COLR 94-10, Revision 0
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Table 3.1b WNP-2 Cycle 10 MCPR Operating Conditions Cycle Exposures ) 4500 MVd/MTU SLMCPR = 1.07 SLMCPR = 1.07 FFIR SPC 8x8 SPC 9x9 SPC 9x9 SVEA-96 SPC 8x8 SPC 9x9SPC 9x9 SVEA-96 Condition Limit GE11 LFA LFA GE11 LFA LFA NW>>
Full Power 1.30 1.27 1.44 1.54 1.32 1.29 1.46 1.58 Flow Dependent Figure 3.1 Figurc 3.1 Power Dependent" Fig. 3.~J) Fig. 3.3b Fig. 3.3b Fig. 3.2b Fig. 3.6 Fig. 3.7 Fig. 3.7 Fig. 3.6 TSSS<>>
Full Power 1.33 1.30 1.49 1.60 1.35 1.32 1.51 1.63 Flow Dependent Figurc 3.1 Figure 3.1 Power Dependent+ Fig. 3.4b Fig. 3.5b Fig. 3.5b Fig. 3.4b Fig. 3.8 Fig. 3.9 Fig. 3.9 Fig. 3.8 Nss~>>
RPT Full Power 1.38 1.35 1.61 1.68 Not Analyzed Inoperable Flow Dcpcndent Figure 3.1 Power Dependent" Fig. 3.10b Fig. 3.lib Fig. 3.lib Fig. 3.10b SLO+ NSS Full Power 1.56 1.36 1.36 1.98 1.56 1.36 1.36 1.98 Flow Dependent None None Power Dependent+ Fig. 3.~J) Fig. 3.3b Fig. 3.3b Fig. 3.2b Fig. 3.6 Fig. 3.7 Fig. 3.7 Fig. 3.6 SLO~
Full Power 1.56 1.36 1.36 1.98 1.56 1.36 1.36 1.98 Flow Dcpcndcnt None None Power Dependent+ Fig. 3.4b Fig. 3.5b Fig. 3.5b Fig. 3.4b Fig. 3.8 Fig. 3.9 Fig. 3.9 Fig. 3.8 SLO+ NSS Rpl'ull Power 1.56 1.36 1.36 1.98 Not Analyzed Inoperable Flow Dcpcndcnt None Power Dependent Fig. 3.10b Fig. 3.11b Fig. 3.11b Fig. 3.10b Washington Nuclear-Unit 2 COLR 94-10, Revision 0
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Power s 30% Two Loop and Single 2.8 Loop Operation SPC 8x8 Power GE11 LFA SVEA-96 LFA 2.6 25% 2.50 2.73 All Core Flows 30% 2.40 2.63
'E Power > 30%
2.4 Core Flow a 50%
Power a 30% SPC 8x8 g 22 Power GE'11 LFA SVEA-96 LFA SPC 8x8 30% 1.45 1.68 Power GE11 LFA SVEA-96 LFA o 2.0 2.10 2.33 100 1.28 1.51 25%
CL 30% 2.00 2.23 g 1.8 SVEA-96 LFA 1.6 1.4 SPC 8x8 GE11 LFA 20% 30% 40% 50% 60% 70% 80% S0% 100% 1 10%
Percent of Rated Power Reduced Power MCPR Operating Limit Versus Percent of Rated Power NSS, RPT inoperable SPC 8x8, GE11 LFA, SVEA-96 LFA Cycle Exposures c 4500 MWd/MT Figure 3.10a
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~ Power GE11 LFA SVEA-96 LFA 22 SPC Sx8 30% 1.46 1.76 Power GE11 LFA SVEA-96 LFA 100% 1.38 1.68 o 2.0 25% 2.10 2.40 K 30% 2.00 2.30 LL 1.8 SVEA-96 LFA p
1.6 1.4 SNP SxS GE11 LFA 1.2 20% 30% 40% 50% 60% 70% 80% 90% 100% 110%
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Washington Nuclear-Unit 2 COLR 94-10, Revision 0
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940623 i i:03 S.D I~IEPE EN 5.1 e o fr urren C cle 5.1.1 'MF-94-095, "WNP-2 Cycle 10 Plant Transient Analysis," Siemens Power Corporation, June 1994.
5.1.2 EMF-94-096, "WNP-2 Cycle 10 Reload Analysis," Siemens Power Corporation,'une 1994.
5.1.3 SPCWP-94-041, "Licensing Results Supporting Section 3/4.2 of the WNP-2 Technical Specifications for Cycle 10," Letter from YU Fresk, Siemens Power Corporation, to RA Vopalensky, Supply System, April 1, 1994.
5.1.4 SPCWP-94-062, "STAIF Stability Results in Support of WNP-2 Cycle-10,"
Letter from YU Fresk, Siemens Power Corporation, to RA Vopalensky, Supply System, June 14, 1994.
5.1.5 SPCWP-94-042, "Licensing Results Supporting Section 2.1 of the WNP-2 Technical Specifications for Cycle 10," Letter from YU Fresk, Siemens Power Corporation, to RA Vopalensky, Supply System, April 1, 1994.
5.1.6 SPCWP-94-068, "SPC Comments on WNP-2 Cycle 10 Draft COLR," Letter from YU Fresk, Siemens Power Corporation, to RA Vopalensky, Supply System, June 23, 1994.
5.1.7 RDW:94-092, "WNP-2 Cycle 9 Core Operating Limits Report - GE11 Lead Use Assemblies," Letter from RD Williams, GE Nuclear Energy, to. DL Whitcomb, Supply System, June 21, 1994.
5.1.8 ABBWP-94-040, "SVEA-96 Lead Fuel Assembly Treatment in WNP-2 Cycle 10 Core Operating Limits Report," Letter from CG Schon, ABB Combustion Engineering Nuclear Operations, to RA Vopalensky, Supply System, June 15, 1994.
I 5.2 Licen i ical Re 'n Technical S i cati 5.2.1 ANF-1125(P)(A) and Supplements 1 and 2, "826 B Critical Power Correlation," Advanced Nuclear Fuels Corporation, April 1990.
5.2.2 "NRC Approval of ANFB Additive Constants for 9x9-9X BWR Fuel," Letter from RC Jones, NRC, to RA Copeland, Advanced Nuclear Fuels Corporation, November 14, 1990.
Washington Nuclear-Unit 2 COLK 94-10, Revision 0
5.2.3 ANF-524(P)(A), Revision 2 and Supplements 1 and 2, "Advanced Nuclear Fuels Critical Power Methodology for Boiling Water Reactors," Advanced Nuclear Fuels Corporation, November 1990.
5.2.4 ANF-913(P)(A), Volume 1, Revision 1 and Volume 1, Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.
5.2.5 ANF-CC-33(P)(A), Supplement 2, "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K, Heatup Option, Advanced Nuclear Fuels Corporation, January 1991.
5.2.6 XN-NF-80-19(P)(A), Volume 1, Supplements 3 and 4, "Advanced Nuclear Fuels Methodology for Boiling Water Reactors: Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodology," Advanced Nuclear Fuels Corporation, November 1990.
5.2.7 XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, Inc., June 1986.
5.2.8 XN-NF-80-19(P)(A), Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, Inc., January 1987.
5.2.9 XN-NF-85-67(P)(A), Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, Inc.,
September 1986.
5,2.10 ANF-89-014(P)(A), Revision 1 and Supplements 1 and 2, "Generic Mechanical Design for Advanced Nuclear Fuels 9x9-IX and 9x9-9X Reload Fuel,"
Advanced Nuclear Fuels Corporation, October 1991.
5.2.11 XN-NF-81-22(P)(A), "Generic Statistical Uncertainty Analysis Methodology,"
Exxon Nuclear Company, Inc., November 1983.
5.2.12 NEDE-24011-P-A-6, "General Electric Standard Application for Reactor Fuel," GE Nuclear Energy, April 1983.
5.3 Nuclear E er d ABB sti n En 'neerin ucl on d Fue R
5.3.1 "GE11 Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No. 2, Reload 5, Cycle 6," GE Nuclear Energy, December 1989.
Washington Nuclear-Unit 2 COLR 94-10, Revision 0
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%0523 I l:03 5.3.2 UK 90-126, "Supplemental Lead Fuel Assembly Licensing Report SVEA-96 LFAs for WNP-2 Summary," ABB Atom, January 1990.
5.3.3 . ATOF-91-120, "Assembly Treatment in WNP-2 Cycle 7 Core Operating Limits Report," Letter from WR Harris, ABB Atom, to DLWhitcomb, Supply System, May 1, 1991.
5.3.4 ABBWP-94-039, "WNP-2 SVEA-96 Lead Fuel Assembly Operating Limit MCPR," Letter from CG Schon, ABB Combustion Engineering Nuclear Operations, to RA Vopalensky, Supply System, June 15, 1994.
5.4 Re for the Extended ad Line Limit Anal si LLLA Ct 5.4.1 "Reactor Vessel Internals Evaluation Task Report for WNP-2 Power Uprate Project," GE Nuclear Energy, April 1993 (DRAFT):
5.4.2 "WNP-2-Power Uprate Containment Response Evaluation Input to Engineering Report," GE Nuclear Energy, January 19, 1993 (DRAFT).
5.4.3 GE-NE-189-69-1092, "Effects of Adjustable Speed Drive on Reactor Internal Vibration at the WNP-2 Nuclear Power Plant," GE Nuclear Energy, October 1992.
5.4.4 GE-NE-189-34-0392, "Jet Pump Sensing Line Vibration Test for Washington Nuclear Project 2," GE Nuclear Energy, March 1992.
5.4.5 NEDE-24222, "Assessment of BWR Mitigation of ATWS, Vol. II {NUTMEG 0460, Alternate No. 3)," General Electric Company, December 1979.
5.4.6 "Washington Nuclear Project Unit 2 System Evaluation Report for Power Upzate Reactor Recirculation Control System," GE Nuclear Energy, February 1, 1993.
5.4.7 GE Report 22A7104, Revision 0, "Dynamic Load Report Fuel Vertical Support," GE Nuclear Energy, June 30, 1982.
5.4.8 "Fuel Lift Non-Proprietary Letter," Letter from DM Kelly, GE Nuclear Energy, to WC Wolkenhauer, Supply System, February 15, 1993.
5.4.9 93-PU-0054, "ELLLA Related Power Uprate Task Reports," Letter from DM Kelly, GE Nuclear Energy, to WC Wolkenhauer, Supply System, June 3, 1993.
Washington Nuclear-Unit 2 COLR 94-10, Revision 0
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