ML17291A624

Rev 1, Cycle 10 Colr

ML17291A624
Person / Time
Site:	Columbia
Issue date:	12/31/1994
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17291A623	List: GO2-95-014, Forwards Cycle 10 Colr. Informs That Committee Has Reviewed & Approved Rev 1 to COLR 94-10 ML17291A622 ML17291A624
References
COLR-94-10, COLR-94-10-R01, COLR-94-10-R1, NUDOCS 9502010200
Download: ML17291A624 (43)

Text

941229 1498 CpLR 94 10 Remsion 1

ContmHed CopyNo WNP-2 Cycle 10

, Core Operating Lixnits Report December 1994 Washington Public Power Supply System 9502010200 950125 PDR ADOCK 05000397 P

PDR 4

941229 149$

WNP-2 Cycle 10 Coze Qperatiag Limits Repozt LIST F EFFECTlVE PA ES

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WNP-2 Cycle 10 Coic OpeiatlIlg L1101~ Report ST F EFFECTIVE PAGES

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WNP-2 Cycle 10 Core Operating Limits Report B E F

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

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1.0 INTR D CTI N AND SUhQvfARY This report provides the Average Planar Linear Heat Generation Rate (APLHGR) limits, the Minimum Critical Power Ratio (MCPR) limits, and the Linear Heat Generatioa Rate (LHGR) limits for WNP-2, Cycle 10 as required by Technical Specificatioa 6.9.3.1.

As zequized by Technical Specifications 6.9.3.2 and 6.9.3.3, these limits were determined using NRC-approved methodology and axe established so that all applicable limits of the plant safety analysis are met.

The thexmal limits for SPC fuel given in this report are documented in the "Cycle 10 Plant Transient Analysis" (Refezcnce 5.1.1), the "Cycle 10 Reload Analysis" (Reference 5.1.2) and the "Impzoved Reduced Plow MCPR Oyexatiug Limits for WNP-2 Cycle 10" (Refezence 5.1.9). The thezmal limitsdetexmined thxough the approved methodology aze modified for the GE11 and SVEA-96 LPAs as discussed below.

The WNP-2 Cycle 10 core includes four Siemens Power Corpoxation (SPC), four GE Nuclear Energy (GE), and four ABB Combustion Eugiaeexing Nuclear Ogexatioas (ABB CENO) Lead Puel Assemblies (LFAs). The SPC LFAs were inserted duziug the reload for Cycle 5. The GE and ABB CENO LPAs 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 axe not expected to be the most limitingfuel assemblies oa either a nodal or an assembly power basis.

The GE Nuclear Energy GE11= LPAs aze described in the "GE11 Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No. 2, Reload 5, Cycle 6" (Refexence 5.3.1).

This xefexence describes the design goals of the GE11 LPAs and provides support for monitozing the GE11 LPAs at thermal limits based on the SPC 8x8 reload fuel thezmal limits.

The ABB CENO SVEA-96 LPAs are described ia the "Supplemental Lead Fuel Assembly Licensing ReportSVEA-96 LPAs forWNP-2Summaxy" (Reference 5.3.2). The process for developing thermal limits for the SVEA-96 LPAs based upon the SPC 8x8 xeload fuel thermal limits is described in References 5.3.2 thxough 5.3.4 The MAPLEGRlimits for the GE11 LFAs are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-2]/I'81-7])is applied to account for the different number of fuel pins in the two designs.

The MAPLHGRlimits forthe SVEA-96 LPAs are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-2]/f100-4]) is apylied to account for the diffezent number of fuel pins in the two designs.

Fuzthexmoze, the MAPLHGR limits for the SVEA-96 LPAs are multiplied by the following constants:

(a) 1.04 to account for a diffezent estimation of the local power in the output from POWEIU'LEX compaxed to ABB CENO methods aad (b) 1.02 to account for a different estimation of exposure ia the output from POWERPLEX compared to ABB CENO methods.

The MCPR limitis the maximum of (a) the applicable exposure dependeat, fullpower and full flow MCPR limit, (b) the applicable exposure and power dependent MCPR limit, and (c) the flowdependent MCPR limitspecified in this reyoxt. This stipulation assures that the safety limit MCPR willnot be violated thzoughout the WNP-2 operating regime. Fullpower MCPR limits are specified to define oyexating limits at rated power and Qow.

For the WNP-2 coze, the Turbine Trip without Bypass event is limiting for operation at xated power and fiow. Power dependent MCPR limits axe specified to define opexatiag limits at other thaa zated power Washington Nuclear-Unit 2 COLR 94-10, Revision 1

941229 14QS conditions.

For the WNP-2 core, the Feedwater Controller Failure event from zeduced power is calculated to be more severe than from fullpower conditions. A fiow dependent MCPR is slxeified to define operating limitsat other than rated flowconditions. The reduced flowMCPR limit provides bounding protection for the limiting Recizculation Flow Increase event (Refezence 5.1.9).

The LHGR limits for the GE11 LFAs are the same as for the SPC 8x8 zeload 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 LHGR limits for the SVEA-96 LFAs are taken dimply from Refezence 5.3.2.

The reload licensing analyses for this cycle provide opezating limits for Extended Load Line (ELl~) operation which extends the power and flow operating regime for WNP-2 up to the 109% zod line which at fullpower cozzesponds to 87% ofzated flow. The MCPR limits deflned in this zepozt aze applicable up to 100% of rated thermal power along and below the 109% zod line.

The minimum low for operation at rated power is 87% of zated fiow; the maximum is 106%.

Refezences 5.1.1 and 5.1.2 and the zefezences in Section 5.4 document the analyses in suppozt of ELLLAopezation.

Preparation, zeview and approval of this report weze performed in accordance with applicable Supply'System pzoceduzes.

The specific topical report revisions and supplements which describe the. methodology utilized in this cycle specific analysis aze zefezenced in Section 5.2.

2.0 AVERA EPLANARLINEARHEATGENIHMTINRATE APLH R LIMITSFOR U E IN TE CAL SPECIFICATI N 3 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 opezation and in Figuzes 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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940610 13:46 SPECIFICATl N 3 2.3t t

3.0 MINXMUM RlTI ALP WER RATI CPR LIMITFOR USE IN TECKVI AL'he 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. la and 3.1b and Figures 3.1 and 3.2a through 3.lib.

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

e 940614 l6:54 Table 3.1a WNP-2 Cycle 10 MCPR Operating Conditions Cycle Exposures s 4500 MWd/MTU SLMCPR = 1.07 Condition Limit NM" How Depcndcnt Power Dependents SPC 8xS SPC 9x9 SPC 9x9 SVBA-96 GB11 LFA LFA I

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1.28 1A4 Figure 3.1 Fig. 3.2a Fig. 33a Fig. 33a Fig. 3.2a Nssu~

Flow Dcpcndcnt Power Dependent'+

1.26 1.24 135 1A8 Figure 3.1 Fig. 3.4a Fig. 3.5a Fig. 3.5a Fig. 3Aa RFF Full Power Inoperable Flow Dcpcndcnt Power Dcpcndcntro SLOP NSS How Dcpcndcnt Power Dependents SLO TSSS Full Power Flow Dcpcndcnt Power Dcpcndcnà LO+ NSS RFI'ull Power Inoperable Flow Dcpcndcnt Power Dependent'+

1.2S 1.27 1.43 1.51 Figure 3.1 Fig. 3.10a Fig. 3.11a Fig. 3.11a Fig. 3.10a 186 136 1.36 1.9S None Fig. 3.2a Fig. 3.3a Fig. 3.3a Fig. 3.2a 1.56 1.36 1.36 1.98 None Fig. 3.4a Fig. 3.5a Fig. 3.5a Fig. 3.4a 1.56 1.36 1.36 1.98 None Fig. 3.10a Fig. 3.11a Fig. 3.11a Fig. 3.10 Washington Nuclear-Unit 2 COLR 94-10, Revision 0

940614 16M Table 3.1b WNP-2 Cycle 10 MCPR, Operating Conditions Cyde Exposures ) 4500 MWd/MTU SLMCPR = 1.07 SLMCPR = 1.07 FFTR Condition Limit NSS'o SPC 8xg SPC 9x9 SPC 9x9 SVBA-96 GEII LFA LFA SPC gxg SPC 9x9SPC 9x9 SVBA-9 GBII LFA LFA 1.30 1.27 1.44

'64 192 1.29 1.46 168 co Flow Dcpcndcnt Power Dependent+

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 1.33 1.30 1.49 1.60 1.35 1.32 1.51 1.63 NSS"'low Dcpcndcnt Power Dependent Figure 3.1 Figure 3.1 Fig. 3Ab 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 Dcpcndent Power Dcpcndcn&

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 1.56 1.36 1.36 1.98 1.56 1.36 1.36 1.98 Flow Dcpendcnt Power Dcpcndcnà None Fig. 3.2b Fig. 33b Fig. 3.3b Fig. 3.2b None Fig. 3.6 Fig. 3.7 Fig. 3.7 Fig. 3.6 SLO TSSS Full Power 186 196 1.36 1.98 186 1.36 136 1.98 Flow Dcpendcnt Power Dependent+

SLO+ NSS None Fig. 3.4b Fig. 3.5b Fig. 3.5b None Fig. 3.4b Fig. 3.8 Fig. 3.9 Fig. 3.9 Fig. 3.8 RFF Full Power inoperable Flow Dcpcndcnt Power Dependent'+

1.56 1.36 1.36 None 1.98 I

Fig. 3.10b Fig. 3.11b Fig. 3.11b Fig. 3.10b Not Analyzed Washington Nuclear-Unit 2 COLR 94-10, Revision 0

NoMfor Tables 3.1a and 3.lb 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 shaQ be determined from the applicable Technical Specification Scram Speed (TSSS)

MCPR limits in Tables 3. la and b.

Position Inserted From Fully Withdrawn Notch 45 Notch 39 Notch 25 Notch 5 Slowest measured average control rod insertion times qmcified notches for all operable control rods for each grou of four contxol rods arranged in a two-by-two array (seconds) 0.380 0.720 1.600 2.950 Note 2:

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

The increase is included in the MCPR limits for SLO.

Note 3:

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, reslxctively.

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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Power 30%

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941229 14:38

'5:0 'MFHR HE 5.1 R

rts for Curient C cle 5.1.1 5.1.2 5.1.3 5.1.4 EMF-94-095, "WNP-2 Cycle 10 Plant Transient Analysis," Siemens Power Corporation, June 1994.

EMF-94-096, "WNP-2'ycle 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 YUFresk, 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 Corporation, to RA Vopalensky, Supply System, June 14, 1994.

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.

RDW:94-092, "WNP-2 Cycle 9 Core Operating Limits Report - GE11 Lead Use Assemblies," Letter Born 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.

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.

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 )

5.2 Licen in T ical rt in Technical S ification 6

.3 2 5.2.1 5.2.2 ANF-1125(P)(A) and Supplements 1

and 2,

"ANFB Critical Power.

Correlation," Advanced Nuclear Fuels Corporation, April 1990.

"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 1

940523 11:03

'.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 5.2.5 5.2.6 ANF-913(P)(A), Volume 1, Revision 1 and Volume 1, Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for BoiTing Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.

'I ANF-CC-33(P)(A), Supplement 2, "EGDPZ: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K, Heatup Option,," Advanced Nuclear Fuels Corporation, January 1991.

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 THERhiEX: 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 E Nuclear En r d ABB C m u tion En ineerin Nuclear e

tion Lead Fuel A sembl 5.3.1 "GEll 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

5.4 5.3.2 5.3.3 5.3.4 UK 90-126, "Supplemental Lead Fuel Assembly Licensing ReportSVEA.9$.

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 &om CG Schon, ABB Combustion'Engineering Nuclear Operations, to RA Vopalensky, Supply System, June 15, 1994.

f r xten Li imi An L

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 ofAdjustable 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 GE-NE-189-34-0392, "Jet Pump Sensing Line Vibration Test for Washington Nuclear Project 2," GE Nuclear Energy, March 1992.

NEDE-24222, "Assessment ofBWR Mitigation of ATWS, Vol. II (MlREG 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.

.5.4.7 5.4.8 5.4.9 GE Report 22A7104, Revision 0, "Dynamic Load ReportFuel Vertical Support," GE Nuclear Energy, June 30, 1982.

"Fuel Lift Non-Proprietary Letter," Letter from DM Kelly, GE Nuclear Energy, to WC Wolkenhauer, Supply System, February 15, 1993.

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