ML17291A686

Rev 2 to WNP-2 Cycle 10 Colr

ML17291A686
Person / Time
Site:	Columbia
Issue date:	03/08/1995
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17291A685	List: GO2-95-047, Forwards WNP-2 Cycle 10 COLR, COLR 94-10,Rev 2.Rev Incorporates Extended Single Loop Max Average Planar Linear Heat Generation Rate Limits for SPC 9x9 Reload & Lead Fuel Assembly Fuel ML17291A686
References
COLR-94-10, COLR-94-10-R02, COLR-94-10-R2, NUDOCS 9503140448
Download: ML17291A686 (43)

Text

9$0220 13:49 COLR 94-10, Revision 2 Controlled Copy No.

WNP-2 Cycle 10 Core Operating Limits Report

February, 1995 Washington Public Power Supply System

" 9503140448 950308 PDR ADOCK 05000397 P

PDR

950220 13:49 WNP-2 Cycle 10 Core Operating Limits Report LIST F EFFE TIVEPA ES

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9S0220 13:50 WNP-2 Cycle 10 Core Operating Limits Report I T F EFFE TIVEPAG gev~ii II 35 35a 36 37 1

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WNP-2

'ycle 10 Core Operating Limits Report TABLE F

NTENT 1.0 OD CTI N AND

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2.0 AVERA E PLANARL AR HEAT ENERATI N RATE APLH R LIMIT F R EINTE HNI AL SPE IFI ATI N 2

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IN TE HNI AL PE IFI ATI N 24......,....................

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E 35 Washington Nuclear-Unit 2 COLR 94-10, Revision 0

950221 12:18 1.0 INTR DU N AND

SUMMARY

This report provides the Average Planar Linear Heat Generation Rate (APLHGR) limits, the Minimum Critical Power Ratio (NICPR) 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),

the "Cycle 10 Reload Analysis" (Reference 5.1.2), the "Improved Reduced Flow MCPR Operating LimitsforWNP-2 Cycle 10" (Reference 5.1.9) and the "Extended Single Loop MAPLHGR Limits for SPC 9x9 Fuel in WNP-2" {Reference 5.1.12). 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 I.FAs 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 GE11 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 S.3. 1). This reference describes the design goals of the GE11 LFAs and provides support for monitoring the GE1) 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 ReportSVEA-96 LFAs for WNP-2Summary" (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 S.3.4 The MAPLHGR limits for the GEI I 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 MAPLHGRlimits 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 limitis the maximum of (a) the applicable exposure dependent, fullpower and full flow MCPR limit, {b) the applicable exposure and power dependent MCPR limit, and (c) the flowdependent MCPR limitspecified in this report, This stipulation assures that the safety limit MCPR willnot be violated throughout the WNP-2 operating regime.

Full power MCPR limits are specified to define operating limits at rated power and fiow. For the WNP-2 core, the Turbine Trip without Bypass event is limiting for operation at rated power and fiow. Power Washington Nuclear-Unit 2 COLR 94-10, Revision 2

950221 12:18 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 is calculated to be more severe than from full powex: conditions.

A flow dependent MCPR is specified to define operating limitsat other than rated flowconditions. The reduced flowMCPR limit provides bounding protection for the limiting Recirculation Flow Increase event (Reference 5.1.9).

The LHGR limits for the GE11 LFAs are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-21/[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 %NP-2 up to the 109% rod line which at fullpower 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 fiow 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 ELLLAoperation.

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 AVERA EPLANARLI ARHEAT ENERATI NRATE APLH R LIMIT F IN TE HNI L

PE IFI ATI 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.

Washington Nuclear-Unit 2 COLR 94-10, Revision 1

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Average Planar MAPLHGR Exposure 11.2 18,860 4.0 7.6 6.3 3.7 30,000 36,000 46,000 2.0 0

10,000 20,000 30,000 Average Planar Exposure (MWd/MT)

Maximum Average Planar Linear Heat Generation Rate fMAPLHGR)Versus Average Planar Exposure SPC 9x9-9X Reload Fuel and SPC Bx9 l.FA's Figure 2.3 40,000

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940610 13:46 3.0 MINIMUMCRITICALPOWER RATI CPR LIMITFOR USE IN TECHNICAL SPECIFICATI N 3 2.3 The MCPR limitfor use in Technical Specification 3.2.3 shall be:

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.lib.

Washington Nuclear-Unit 2 COLR 94-10, Revision 0

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940514 16:$4 Table 3.1b WNP-2 Cyde 10 MCPR, Operating Conditions Cyde Exposures > 4500 MWd/MTU SLMCPR = 1.07 SLMCPR = 1.07 FFTR Condition Limit Nssr>>

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Figure 3.1 Figure 3.1 Fig. 3.2b Fig. 3.3b Fig. 3.3b Fig. 3.2b Fig. 3.6 Fig. 3.7 Fig. 3.7 Fig. 3.6 Full Power 1.33 1.30 1.49 1.60 1.35 1.32 1.51 1.63 NSS>>

Flow Dependent Power Dependents Figurc 3.1 Figure 3.1 Fig. 3.4b Fig. 3.5b Fig. 3.5b Fig. 3.4b Fig. 3.8 Fig. 3.9 Fig. 3.9 Fig. 3.8 RFI'ullPower Inoperable Flow Dcpcndcnt Power Dependent<+

SLO(" Nss 1.38 1.35 1.61 1.68 Figure 3.1 Fig. 3.10b Fig. 3.11b Fig. 3.11b Fig. 3.10b Not Analyzed Full Power Flow Dcpcndcnt Power Dcpcndcnà SLO Vsss 1.56 1.36 1.36 1.98 None 1.56 1.36 1.36 1.98 None Fig. 3.2b Fig. 3.3b Fig. 3.3b Fig. 3.2b Fig. 3.6 Fig. 3.7 Fig. 3.7 Fig. 3.6 Full Power 1.56 1.36 1.36 1.98 1.56 1.36 1.36 1.98 SLOo) Nss Flow Dcpcndcnt Power Dependent'+

None Fig. 3.4b Fig. 3.5b Fig. 3.5b Fig. 3.4b None Fig. 3.8 Fig. 3.9 Fig. 3.9 Fig. 3.8 RFI'ullPower Inoperablc Flow Dcpcndcnt Power Dcpcndcn&

1.56 1.36 1.36 None 1.98 Fig. 3.10b Fig. 3.11b Fig. 3.11b Fig. 3.10b Not Analyzed Washington Nuclear-Unit 2 COLR 94-10, Revision 0

940614 16:S4 Notes for Tables 3.1a and 3.1b Note 1:

The scram insertion times must meet the requirements of Technical Specification 3.1.3.4.

The NSS MCPR values are based on the SPC transient analysis performed using the control rod insertion times shown below (defined as normal scram speed: NSS).

In the event that Surveillance 4.1.3.2 shows these scram insertion times have been exceeded, the MCPR limit shall be determined from the applicable Technical Specification Scram Speed (TSSS)

MCPR limits in Tables 3. 1a and b.

Position Inserted From Fully Withdrawn Notch 45 Notch 39 Notch 25 Notch 5

Slowest measured average control rod insertion times t

specified notches for all operable control rods for each grou of four control rods arranged in a two-by-two array (seconds) 0.380 0.720 1.600 2.950 Note 2:

Note 3:

For Single Loop Operation (SLO), the SLMCPR increases by 0.01.

The increase is included in the MCPR limits for SLO.

/

For the noted full power MCPR limits, the control rod withdrawal error (CRWE) event is limiting. The turbine trip without bypass (TI'NB) event is limiting for the remaining fullpower limits. CRWE analysis was performed with a nominal rod block monitor (RBM) setpoint of 1.06. Use ofthe nominal setpoint is in accordance with the methodology described in Reference 5.2.6, consistent with approved industry practice.

Note 4:

Power dependent MCPR limits are provided for core thermal powers greater than or equal to 25% of rated power at all core flows. The power dependent

-MCPR limits for core thermal powers less than or equal to 30% of rated power are subdivided by core flow. Limits are provided for core flows greater than 50% of rated flow and less than or equal to 50% of rated flow, respectively.

A step change in the power dependent MCPR limits occurs at 30% of rated power because direct scram on turbine throttle valve closure is automatically bypassed per Technical Specification 3.3.1.

Washington Nuclear-Unit 2 COLR 94-10, Revision 0

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940610 13:46 4.0 LINEARHEAT GENERATION RATE GR LIhQT FOR USE IN TECHNICAL SPECIFICATION 3.2.4 The LHGR limitfor use in Technical Specification 3,2.4 shall not exceed the values shown in Figures 4.1 through 4.5.

Washington Nuclear-Unit 2 COLR 94-10, Revision 0

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~a i 941229 14:38 5.0 REFEMiJJlCBS 5.1 R

rts for Current C cle 5.1.1 EMF-94-095,~ "WNP-2 Cycle 10 Plant Transient Analysis," Siemens Power Corporation, June 1994.

EMF-94-096, "WNP-2 Cycle 10 Reload Analysis,"

Siemens Power Corporation, June 1994.

SPCWP-94-041, "Licensing Results Supporting Section 3/4.2 of the WNP-2 Technical Specifications for Cycle 10," Letter from YUdesk, Siemens Power Corporation, to RA Vopalensky, Supply System, April 1, 1994.

SPCWP-94-062, "STAIF Stability Results in Support of WNP-2 Cycle 10,"

Letter from YU Fresk, Siemens Power Coqmration, 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 YUFresk, Siemens Power Corporation, to RA Vopalensky, Supply System, April 1, 1994.

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.

5.1.8 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.

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.

5.1.10 SPCWP-94-108, "Improved Reduced Flow MCPR Operating Limits for WNP-2 Cycle 10," Letter from YUFresk, Siemens Power Corporation, to RA Vopalensky, Supply System, December 12, 1994.

RDW:94-162, "WNP-2 Cycle 10 Core Operating Limits ReportRev. 1,"

Letter from RD Williams, GE Nuclear Energy, to DL Whitcomb, Supply System, December 14, 1994.

5.1.11 NFBWR-94-055, "SVEA-96 LFA Flow-Dependent MCPR Limits," Letter from CG Schon, ABB CENO Fuel Operations, to R Vopalensky, Supply System, December 22, 1994.

Washington Nuclear-Unit 2 COLR 94-10, Revision 1

5.1.12 SPCWP-95-015, "Extended Single Loop MAPLHGRLimits for SPC 9x9 Fuel in WNP-2," Letter from YU Fresk, Siemens Power Corporation, to RA Vopalensky, Supply System, February 21, 1995.

5.2 i ensin To ical Re in Technical

'ficati n 3 2 ANF-1125(P)(A) and Supplements 1

and 2,

"ANFB 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

-35a-COLR 94-10, Revision 2

r

'V 940529 11:03 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-BCalculation Methodology," Advanced Nuclear Fuels Corporation, November 1990.

,5.2.7 5.2.8 5.2.9 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.

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.

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 E Nucl embl r Ener nd ABB Combus io En ineerin uclear e

tion Lead Fuel Re rt 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

4

% ~

94062%] l:03 5.4 5.3.2 5.3.3 5.3.4 UK90-126, "Supplemental Lead Fuel Assembly Licensing ReportSVEA-96 LFAs for WNP-2Summary,," ABB Atom, January 1990.

ATOF-91-120, "Assembly Treatment in WNP-2 Cycle 7 Core Operating LimitsReport," Letter from WR Harris, ABBAtom, to DLWhitcomb, Supply System, May 1, 1991.

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.

f r t Ex en Lo d Line Limit An i

LLLA 5.4.1 5.4.2 5.4.3 "Reactor Vessel Internals Evaluation Task Report for WNP-2 Power Uprate Project," GE Nuclear Energy, April 1993 (DRAFT).

"WNP-2 Power Uprate Containment Response Evaluation Input to Engineering Report," GE Nuclear Energy, January 19, 1993 (DRAFT).

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 5.4.5 5.4.6 5.4.7 5.4.8 5.4.9 GE-NE-189-34-0392, "Jet Pump Sensing Line Vibration Test for Washington Nuclear Project 2," GE Nuclear Energy, March 1992.

NEDE-24222, "Assessment of BWR Mitigation of ATWS, Vol. II (NUREG 0460, Alternate No. 3)," General Electric Company, December 1979.

"Washington Nuclear Project Unit 2 System Evaluation Report for Power UprateReactor Recirculation Control System,"

GE Nuclear

Energy, February 1, 1993.

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