ML060800366

Issuance of the Core Operating Limits Report for Reload 8, Cycle 9, Revision 2

ML060800366
Person / Time
Site:	Limerick
Issue date:	03/20/2006
From:	David Helker Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML060800366 (15)

Text

Exelon Nuclear www.exeloncorp.com 200 Exelon Way Kennett Square, PA 19348 TS 6.9.1.12 March 20,2006 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington DC 20555 Limerick Generating Station, Unit 2 Facility Operating License No. NPF-85 NRC: fh.&&Lh 5~-35.?

Subject:

Issuance of the Core Operating Limits Report For Reload 8, Cycle 9, Revision 2

Dear SirlMadam:

Enclosed is a copy of the Core Operating Limits Report (COLR) for Limerick Generating Station, Unit 2, Reload 8, Cycle 9, Revision 2. Revision 2 of this report incorporates the adjusted off-rated thermal limit multipliers for the revised Power Load Unbalance Analysis.

This COLR is being submitted to the NRC in accordance with LGS, Unit 2 Technical Specifications (TS) Section 6.9.1.1 2. If you have any questions, please do not hesitate to contact us. Very truly yours, David P. Helker Manager - Licensing and Regulatory Affairs Exelon Generation Company, LLC Enclosure cc: S. J. Collins, Regional Administrator, Region I, USNRC S. Hansell, USNRC Senior Resident Inspector, LGS T. Valentine, Project Manager [LGSJ, USNRC Exelon Nuclear Fuels Dac ID: COLR Limerick 2, Rev. 2 CORE OPERATING LIMITS REPORT FOR LIMERICK GENERATING STATIOS UNIT 2 RELOAD 8, CYCLE 9 (This is a complete rewrite) Preparer Approved By: Date: 3i//33/0b 1 J. J. Tusar anager ~ BWR Design (GNF) ! Page 1 Exelan Nuclear Fuels 1 .0 2.0 3 .0 4.0 5 .o 6.0 7.0 8.0 9.0 Terms and Definitions General ~~~o~nat~on MAFLHGR Limits nilCPR Limits Linear Heat Generation Rate Limits Control Rod Block Setpoints Turbine Bypass Valve Parameters Stability Protection Setpoints Modes of Operation 10.0 Methodology 1 1 .0 References Dac ID: COLR Limerick 2, Rev. 2 Page 3 5 6 7 9 11 12 13 13 13 14 Page 2 Exeton Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 2 Page Tabte 3- t ?IIAPLHGR Versus .4verage Planar Exposure Aft Fuel Types Table 3-2 MAPLHGR Single Loop Operation (SLO) Reduction Factor Table 4- 1 Operating Limit

~~jiimL~m Critical Power Ratio (OLMCPR) Table 1-2 Table 4-3 Power Dependent hfCPR Limit Adjustments and

~~l~ipiiers Flow Dependent MCPR Limits R;ICPR(F}

Table 5-1 Table 5-2 Table 5-3 Linear Heat Generation Rate Limits LHGR Single Loop Operation (SLO) Reduction Factor Power Dependent.

LHGR Mulripfier LHCRFAC(P)

Table 5-4 Flow Dependent LHGR Multiplier LHGRFAC(F)

'fable 6- 1 Rod Block Monitor Setpoints Table 6-2 Reactor Coolant System Recirculation Flow Upscale Trip Table 7-1 Table 7-2 Tabie 8- I Turbine Bypass System Response Time ~in~nlu~ Required Bypass Valves To Maintak System Operability OPRM PBDA Trip Setpoints Table 9- 1 Modes af Operation 6 6 7 8 8 9 9 10 1.0 11 11 f2 12 13 13 Page 3 Exelon Nuclear Fuels 1.0 Terns And Definitions Doc ID: COLR Limerick 2, Rev. 2 ARTS BASE CASE DTSP EOUS End of Rated (EOR) FFWTR WHOOS HTSP XCF ITSP LHCR LHGRF AC(F) LHGRFAC(P)

LTSP MAPLHGR MCPR MCPRCP) MCPR(F) MELLLA BLMCPR OPRM PBDA I-WTOOS APRM and RBM Technical Specification Analysis A caSe analyzed with Turbine Bypass System in service and Recirculation Pump Trip in service and Feedwater Temp~rature Reduction aliowed (EFWTR inclrides feedwater heater 00s or final feedwater temperature reduction) at any point during the cycle in Dual Loop mode. Rod Block Monitor Downscafe Trip Setpoint Equipment Out of Service The cycle exposure at which reactor power is equaf to 3458 MWth with recirculation system flow equal to lO@%, all control rod3 fully withdrawn, all feedwater heating in service and equl~ibr~~ni Xenon. Final Feedwater Temperature Reduction Feedwater Heaters Out of Service Rod BIock Monitor High Trip Setpaint Increased Core Flow Rod BIock Monitor Intermediate Trip Setpoint Linear Weat Generation Rare ARTS LHCR thermal limit flow dependent adj~stmen~s and multipliers ARTS LHGR thermal limit power dependent adjustments and muItipliers Rod Block Monitor LQW Trip Sexpaint Maximum Average Planar Linear Heat Generation Rsre ~~ni~i~i~ Critical Power Ratio ARTS hr*CPR thermal limit pouer dependent adjustments and mul~ip[~~~

ARTS MCPR thermal limit flow dependmt adjustments and multiplier<

~~~~~m~rn Extended Load Lint: Limit Analysis Operating Limit ~*~~imum Critical Power Ratio Oscillatim Pocter Range Moniror Period Based Detection Algorithm Recirctiiation Pump Trip Out of Service Page 4 Exelon Nucltear Fuels Doc ID: COLR Limerick 2, Rev, 2 SLMCPR Safety Limit M~nimum Critical Power Ratio SLO Single Loop Operation TBVUOS Turbine Bypass Valves Out of Service This reprt provides the following cycle-specific parameter limits for Limerick Generating Station Unit 2 Cycle 9: Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

Minimum Critical Power Ratio (MCPR) Single Loop Operation (SLO) NCPR adjust~~nt ARTS MCPR thermal limit adjustments and multipliers (MCPR(Pf or ILICPR(F))

ARTS LHGR thermal f imit multipliers (LWCWAC(P) or LHGRFAC(F))

Rod Bfock Monitor (RBM) setpoints MMLWGR single loop operation reduction factor LHCR singfe foop operation reduction factor Linear Heat Generation Rare (LHGR) Turbine Bypass Valve parameters Reactor Coolant System Recircuiation Flow Upscale Trips Oscillation Power Range Monitor Period Based Detection Algorithm Trip Setpoints These values have been determined using NRC-approved me&hodology (Reference 6), and are established such that all applicable fimits of the plant safety analysis are met This report is prepared in accordance with Technical S~~~fication 6.9. I .9 of Reference I. Preparation of this report was performed in accordance with Exefon Nuclear, Nuclear Fuel Management T&RM NF-AB-I 20- 3600. The data presented in this report is valid for all iicensed operating domains on the operating map, including:

0 0 0 Maximum Extended Load Line Limit dctwn to 81% of rated core flow during full power operation Increased Core Flow (ICF) up to I IO% of rated core flow Find Feedwater Temperature Reduction (FFWTR) up to 105°F during cycle extension operation Feedwater Hearer Out of Service (FWWOOS) up to 60°F feedwater temperature reduction at any time during the cpcte prior to cycle extension.

Further inf~~at~on on the cycle specific analyses for Limerick 2 Cycle 9 and the associated operating domains discussed above is available in Reference

2. Page 5 Exelon Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 2 3,O MAPLHGR Limits 3.1 Technical Specification Section 3.2. I 3.2 Description The limiting MAF'LHGR value for the most limiting Iattice

{excluding natural uranium) of each fuel type as a function of average planar exposure is given in Table 3-1 (Reference 2). The limiting MAPLHCR value is the same for alt fuel types in the Limerick Unit 2 Cycle 9 core. For single loop operation, a reduction factor is used which is shown in Table 3-2 (Reference 2). TABLE 3-1 MAPLHCR Versus Average PIanar Exposure All Fuel Types (Reference

2) MAPLHGR Limit TABLE 3-2 &WLHGR SingIe I,mp Operation (SLO) Reduction Factor (Reference

2) StU Reduction Factor Page 6 Exeh Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 2 4.0 MCPR Limits 4.1 Technical Specification Section 3.2.3 4.2 Description Table 4-1 is derived from the Reference 2 anaXyses and is valid for ail Cycle 9 fuel types and operating domains.

Table 4-1 includes treatment of theiie MCPR limits for SLO. Bounding MCPR values are afso ~rov~~d for inoperable Recircufation Pump Trip or inoperable Steam Bypass System. These two options represent the Equipment Out of Service conditions.

The cycle exposure that represents EOR is given in the latest verified and approved Cycle Mana~e~en~

Report or an associated Engineering Change Request.

ARTS pratides for power- and ~low-de~nd~nt thermal limit adjustments and muttiplicrs.

which allow for a more reliable adminis~rati~n of the MCPR thermal limit. The flow- dependent adjustment MCPR(F) is sufficiently generic to apply to alf fuel types and operating domains (References 2 and S). In addition, &ere are also two sets of powerdependent MCPR muftipliers for use with the Turbine Bypass Valves in service and TBVOOS conditions (References 8 and 11). Section 7.0 contains the conditions for Turbine Bypass Valve Operabiiity. These adjustments are provided in Table 4-2 and 4-3. The OLMCPR IS determined for a given power and flow condition by evaluating the power-dependent MCPR and the ftow- dependent MCPR and selecting the greater of the two. The MCPR(P) curves are inde~ndei~~

of refircutation pump trip operability (Reference 8). TABLE 4-1 Operating Limit ~~i~i~~rn Critical Power Ratio (ULMCPR)' (Reference

2) ' This tabfe is valid for ail C>de 9 fuef types. ' When Tau does not equal 0 or f , determine QLMCPR 'c ia linear interpolation.

QLMCPR limit set by the Single Loop Operation Recircutation Pump Seizure Analysis (Reference 2.) i Page 7 Exelon Nuclear Fuels Doc ID: COLR Limerick 2, Rev, 2 TABLE 4-2 Pawer Dependent MCPR Limit Adjustments And Multipliers (References 2,8 and 11) Ftaw (n of 0 25 <30 =mt&waw I. 193 1.193 - 1.193 Base I .150 1.300 - I .300 - 1.300 1.300 _I_ Base SLO I .oa - RPTOOS 1.300 - 1.370 1.193 ____ 1.193 1 .00( __I_ 1 .m RPTOOS SLO TBVOUS TSVOOS StO 1.370 P 1.370 - 1.193 - TABLE 4-3 Flow Dependent MCPR Limits R/ICPR( F) (References 2,s and 5) ! Page 8 Exelon Nuclear Fuels Dac ID: COLR Limerick 2, Rev. 2 5.0 Linear Heat Generation Rate Limits 5. I Technical Specification Section 3.2.4 5.2 Description The LHGR is an exposure dependent vatue. Due to tfae proprietary nature of these values only the maximum U02LHGR for each fuel type is listed in Table S-1 . For single bop operation, a reduction factor is used which is shown in Table 5-2 (Reference 2). ARTS provides For power- and fl#w-d~~nden~ thermal limit multipliers, which allow for a more reliable administration of the LHGR thermal limits. There are

~MKI sets of flow- dependent LGHR multipliers for dual-loop and single-loop operation (References 2, 3, and 5). In addition, there itre also two sets of power-dependent LHGR multipliers for use with the Turbine Bypass Valves in service and TBVOOS conditions (References 8 and 1 1). Section 7.0 contains the conditions for Turbine Bypass Vatve Qpcrability. The LHGR multipliers are shown in Tables 5-3 through 5-4. Thermal limit monitoring must be performed with the more limiting LHGR limit resulting from the power-and flow-biascd calculation. The LHGWAC(P) curves are independent of recirculation pump trip operability (Reference 8). TABLE 5-1 Linear Heat Generation Rate Limits (Reference

9) TABLE 5-2 LHGR Siagk Loap Operation

{SLO) Reduction Factor (Reference 21 SLO Reduction Facror' Page 9 Exelon Nuclear Fuels Doc ID: GOLR Limerick 2, Rev. 2 TABLE 5-3 Power Dependent LHCR Multiplier LHGRFACCP) (References 2,5 and 8) TBVOOS SLO TABLE 5-4 Flow Dependent LHGR MuItipIier L~G~A~C~~ (References 2,s and 5) Page 10 Exelon Nuclear Fuels 6.0 Control Rod Block Setpoints

6.1 Technical

Specification Section 3.3.6 6.2 Description Doc ID: COLR Limerick 2, Rev. 2 Technical S~~~ficatjon Limiting Condition for Operation number 33.6 requires control rod block ins~~~~tation channels shatl be OPERABLE with their trip setpoints consistent with the values shown in the Trip Setpoint column of Technical Specification Table 3.3.6-2. The Reactor Cooiant System Recirculation Flaw Upscale Trip is a cycle-specific value and as such is found in Table 5-2 of this COLR. Table 6-2 lists the Nominal Trip Setpoint and Altowable Value. These setpoints are set high enough to ailow futt utilkation of the enhanced ICF domain up to 1 fO% of ritted core flow. Additionally, the AEUS Rod Block Monitor provides for ~~er-de~~d~nt RBM trips. The trip Se~~in~~aII~wa~le values and applicable RBM signal filter time constant data. are shown in Table 6-1. These values &re for use with 'rechnicat Specification

3.3.6. TABLE

6-1 Rod Block Monitor Setpoints' (References 2 and 7) TABLE 6-2 Reactor Coolant System Recirculation Ftotv Upsate Trig (Referenee

7) ' These setpints (kith Rod Block Monitor filter rim constant &r~.een 0. I seconds and 0.55 xxoncts) are based an a cycle-specific rated RWE MCPR limit t\hich is Ie$s than or equat to the ~~ini~u~~

cyck ULMCPR (.see COLR references 2 2nd 3. Page 11 Exeion Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 2 7.0 Turbine Bypass Valve Parameters

7. I Technical Specification Section 3.7.8 and 4.7.8.C 7.2 Description The operability r~q~iremen~s for the steam bypass cystem for use in Technical Specifications 3.7.8 and 4.7.8.C are found in Tables 7-1 and 7-2. If these r~q~ire~en~~

cannot be met, the hifCPR, MCPR(P) and LWCRFAC(P) limits for inopefahfe Steam Bypass System, known as Turbine Bypass Valve Out Of Service, must be used. TABLE 7-* Turbine Bypass SysEem Kespnse Time ~Re~e~~~~e

4) ~~x~rnu~ time after generation of a turbine bypass valve Row signal for bypass valve position to reach 805% of fufl flow TABLE 7-2 mu^ Required Bypass Valves To ~~t~~~ System Operability (References 4 and 8) Page 12 Exelon NucIear Fuels Doc 1D: CULR Limerick 2, Rev. 2 8.0 Stability Protection Setpoints

8.1 Technical

Specification Section 2.2.1 2.F 8.2 Description The Limerick 2 Cycle 9 OPRM Period Based Detection Algorithm (PBDA) Trip Setpoints for rhe OPRM System for use in Technical Specification 2.2.1 are found in Table 8-1. These values are based on the cycle specific analysis documented in Reference

10. The setpoints provided in Table 8- 1 are bounding for a11 modes of operation shown in Table 9- I . TABLE 8-1 OPRM PBDA Trip Setpoints

9.0 Modes

Of Operation TABLE 9-1 Miodes of Operation (Reference 2,3 and 5) The anaIytical

~~~~5ds used to determine the core operating limit% Fhaii be thoxe p~~~~o~~l~

rskizated and approved by the NRC, specifically those defcribed in the following document:

1. "General Eirctric Standard ~ppIica~ion for Reactor Fuel", NEDE-2401 I-P-A-14.

June ZOO0 and US, Supplement 3'EBE-2401 I-P-A- 14-'LrS, Junt: 1OOO. Operating R~IO~I refers EO trperatron an the Poser to FIoa map with or M ithout FEWTR. Page 13 Exelon Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 2 11.0 References

1. "Technical Specifications and Bases for Limerick Generating Station Unit 3", Docket No. 50-353, License No. NPF-85. 2. "Su~p~emen~l Reload Licensing Repon for Limerick Generating Station Unit 2 Reload 8 Cycle 9'*, Global Nuclear Fuel Document No. ~~00~ 1-7705-SRLR, Revision 0, January 2005. 3. "GE13 Fuel Design C~~Ie-Inde~~de~~

Analyses for Limerick Generating Station Units I and X', GE-NE- L 1240884-00-01 P, March 2001. 4, "CPL-3 Transient Protection Parameters Verification for Reload Licensing Anafyses for Limerick 2 Reload 8 Cycle Y', TOD104-00254

5. "ARTS Flow-Dependent Limits with TBVOOS for Peach Bottom Atomic Power Station and Limerick Generating Station", GEM? Document MEDG-32847P, June 1998. 6. "General Electric Standard Application for Reactor Fuet", NEDE-2401 1 +-A- 14, June 2ooO and U.S. Supplement

%DE-240 1 1 -P-A- I4-US, June 2OOO. 7. "Power Range Neutron

~on~toring System Setpoinr Calculations Limerick Generating Station. Units 1 &2 Mod. No. PO0224', LE-0107, Rev. 0, March 2ooO. 8. "Limerick I and 2 Off-Rated Analyses Below the PLU Power Level", GE Nuclear Document No. GE-NE- ~~~~~?-~2~3-R0, March 2005. 9. '*Fuel Bundle Infomation Repon for Limerick Generating Station Unit 2 Reload 8 Cycle 9", Global Nuclear FueI Document Xo. ~-~3 i -7705-FBB2, January 2005. 10. "Limerick 2 Cycle 9 Option III Stability Analysis".

GE Nuclear Energy Document No. eE~E~-~~7-1 of5-RO, March 2005. I 1. "Application Of The Limerick Unit 1 and Unit 2 Power Dependent Limits for X 5S% FLU Power Level", GE Nuclear Energy ~un~~nt No. ~E~E-~-~~-OI WRO, December 2005. Page 14