ML100990233

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Issuance of the Core Operating Limits Report for Cycle 14 Revision 8
ML100990233
Person / Time
Site: Limerick Constellation icon.png
Issue date: 04/02/2010
From: Mudrick C
Exelon Generation Co, Exelon Nuclear
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML100990233 (17)
v • d • e


Text

Limerick Generating Station www.exeloncorp.com Exelon.

3146 Sanatoga Road Nuclear Pottstown, PA 19464 T.S. 6.9.1.12 April 2, 2010 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Limerick Generating Station, Unit 1 Facility Operating License No. NPF-39 NRC Docket Nos. 50-352

Subject:

Issuance of the Core Operating Limits Report For Cycie14 Revision 8 Enclosed is a copy of the Core Operating Limits Report (COLR) for Limerick Generating Station (LGS). Unit 1 Cycle 14 Revision 8 of this report incorporates the revised cycle specific parameters resulting from the new configuration implemented during the LGS, Unit 1 refueling outage.

The COLR is being submitted to the NRC in accordance LGS, Unit 1 Technical Specification 6.9.12.

All licensing work bounds the final Measurement Uncertainty Recapture (MUR) power level and the current power level.

If you have any questions or require additional information, please do not hesitate to contact us.

Sincerely, h r H. Mudrick Plant Manager-LGS Exelon Generation Company, LLC

Enclosure:

Unit 1 COLR for Cycle 14 Rev. 8 cc: S. Collins, Administrator, Region I, USNRC E. DiPaolo, USNRC Sr. Resident Inspector, LGS P. Bamford, USNRC Project Manager for LGS R. R. Janati, PADEP-BRP A R,P

bcc: J. Grimes-KSA R. DeGregorio-KSA C. Mudrick-GML-5 E. Callan-GML-5 P. Gardner-GML-5 C. Hoffman-SSB 4-2 D. Helker-KSA J. Hunter III SSB 2-4 M. Murphy-PADEP BRP (SSB 2-4)

B. Miller-KSA

Exelon Nuclear Fuels N eocID: COLR Limerick 1, Rev. 8 CORE OPERATING LIMITS REPORT FOR LIMERICK GENERATING STATION UNIT 1 RELOAD 13 CYCLE 14 Prepared By: 6A'  ? //-*W*/*-c' I, Date:3 8j Brian C. Miller/Michael R. Holmes Reviewed By: X 4/ Date:

kndepende viewer Reviewed By: 4Py0s, Date: 31A /O10 F.A Reviewer Approved By: Date: "3 -f /0 James J. Tusar Manager - BWR Design Station Qualified <2d_

Reviewed By: Date: 3_..cj 0 Robert C. Potter Page I of 15

Rev. 8 Limerick I, Doc ID: COLR Fuels Exelon Nuclear Table of Contents Page 4

5 1.0 Terms and Definitions 6

2.0 General Information Rate Limits Planar Linear Heat Generation 7 3.0 Maximum Average 9

Power Ratio Limits 4.0 Minimum Critical Rate Limits 12 Linear Heat Generation 5.0 13 Rod Block Setpoints 6.0 Control Parameters .14 Bypass Valve 7.0 Turbine 14 Protection Setpoints 8.0 Stability 15 9.0 Modes of Operation 15 10.0 Methodology 11.0 References Page 2 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 List of Tables Page Table 3-1 MAPLHGR Versus Average Planar Exposure 6 Table 3-2 MAPLHGR Single Loop Operation (SLO) Reduction Factor 6 Table 4-1 Operating Limit Minimum Critical Power Ratio (OLMCPR) 7 Table 4-2 Power Dependent MCPR Limit Adjustments And Multipliers 8 Table 4-3 Flow Dependent MCPR Limits MCPR(F) 8 Table 5-1 Linear Heat Generation Rate Limits - U0 2 Rods 9 Table 5-2 Linear Heat Generation Rate Limits - Gad Rods 9 Table 5-3 Linear Heat Generation Rate Limits - Gad Rods 10 Table 5-4 LHGR Single Loop Operation (SLO) Multiplier 10 Table 5-5 Power Dependent LHGR Multiplier LHGRFAC(P) 10 Table 5-6 Flow Dependent LHGR Multiplier LHGRFAC(F) 11 Table 6-1 Rod Block Monitor Setpoints 12 Table 6-2 Reactor Coolant System Recirculation Flow Upscale Trip 12 Table 7-1 Turbine Bypass System Response Time 13 Table 7-2 Minimum Required Bypass Valves To Maintain System Operability 13 Table 8-1 OPRM PBDA Trip Setpoints 14 Table 9-1 Modes of Operation 14 Page 3 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 1.0 Terms and Definitions ARTS APRM and RBM Technical Specification Analysis BASE CASE A case analyzed with Turbine Bypass System in service and Recirculation Pump Trip in service and Feedwater Temperature Reduction allowed (FFWTR includes feedwater heater OOS or final feedwater temperature reduction) at any point in the cycle operation in Dual Loop mode.

DTSP Rod Block Monitor Downscale Trip Setpoint EOOS Equipment Out of Service EOR End of Rated. The cycle exposure at which reactor power is equal to rated thermal power with recirculation system flow equal to 100%, all control rods fully withdrawn, all feedwater heating in service and equilibrium Xenon.

FFWTR Final Feedwater Temperature Reduction FWHOOS Feedwater Heaters Out of Service HTSP Rod Block Monitor High Trip Setpoint ICF Increased Core Flow ITSP Rod Block Monitor Intermediate Trip Setpoint Kp Off-rated power dependent OLMCPR multiplier LHGR Linear Heat Generation Rate LHGRFAC(F) Off-rated flow dependent LHGR multiplier LHGRFAC(P) Off-rated power dependent LHGR multiplier LTSP Rod Block Monitor Low Trip Setpoint MAPLHGR Maximum Average Planar Linear Heat Generation Rate MAPLHGR(F) Off-rated flow dependent MAPLHGR multiplier MAPLHGR(P) Off-rated flow dependent MAPLHGR multiplier MCPR Minimum Critical Power Ratio MCPR(F) Off-rated flow dependent OLMCPR MCPR(P) Off-rated power dependent OLMCPR Page 4 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 MELLLA Maximum Extended Load Line Limit Analysis OLMCPR Operating Limit Minimum Critical Power Ratio OPRM Oscillation Power Range Monitor RBM Rod Block Monitor RPTOOS Recirculation Pump Trip Out of Service SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation TBVOOS Turbine Bypass Valves Out of Se9rvice 2.0 General Information This report is prepared in accordance with Technical Specification 6.9.1.9 of Reference 1. Power and flow dependent limits are listed for various power and flow levels. Linear interpolation is to be used to find intermediate values.

The data presented in this report is valid for all licensed operating domains on the operating map, including:

" Maximum Extended Load Line Limit down to the minimum licensed core flow during full power operation

  • Increased Core Flow (ICF) up to 110% of rated core flow
  • Final Feedwater Temperature Reduction (FFWTR) up to 105. 1°F during cycle extension operation

Page 5 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 3.0 Maximum Average Planar Linear Heat Generation Rate Limits 3.1 Technical Specification Section 3.2.1 3.2 Description The limiting MAPLHGR value for the most limiting lattice (excluding natural uranium) of each fuel type as a function of average planar exposure is given in Table 3-1. The limiting MAPLHGR value is the same for all fuel types in Limerick Unit 1 Cycle 14. For single loop operation, a multiplier is used which is shown in Table 3-2. The power and flow dependent multipliers for MAPLHGR have been removed and replaced with LHGRFAC(P) and LHGRFAC(F), therefore MAPLHGR(P) and MAPLHGR(F) are equal to 1 (Reference 2).

Table 3-1 MAPLHGR Versus Average Planar Exposure All Fuel Types (Reference 2)

Average Planar Exposure MAPLHGR Limit (GWD/ST) (kW/ft) 0.0 12.82 14.51 12.82 19.13 12.82 57.61 8.00 63.50 5.00 Table 3-2 MAPLHGR Single Loop Operation (SLO) Multiplier (Reference 2)

SLO Multiplier 0.80 Page 6 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 4.0 Minimum Critical Power Ratio Limits 4.1 Technical Specification Section 3.2.3 4.2 Description Table 4-1 is derived from the Reference 2 analyses and is valid for all Cycle 14 fuel types and operating domains. Table 4-1 includes treatment of these MCPR limits for all conditions listed in Section 9.0, Modes of Operation. The cycle exposure that represents EOR is given in the latest verified and approved Cycle Management Report or an associated Engineering Change Request.

ARTS provides for power- and flow-dependent thermal limit adjustments and multipliers, which allow for a more reliable administration of the MCPR thermal limit. The flow-dependent adjustment MCPR(F) and" power-dependent adjustment MCPR(P) are sufficiently generic to apply to all fuel types and operating domains. The MCPR(P) values are independent of recirculation pump trip operability (Reference 3). MCPR(P) and MCPR(F) are independent of Scram Time Option. 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 flow-dependent MCPR and selecting the greater of the two.

When the actual Scram speed falls between Option B (Tau = 0) and Option A (Tau = 1), linear interpolation shall be used to determine MCPR limits.

Table 4-1 Operating Limit Minimum Critical Power Ratio (OLMCPR)

All Fuel Types (Reference 2)

SCRAM Cycle Exposure Time < EOR - 2725 > EOR - 2725 EOOS Combination Option MWd/ST MWd/ST B 1.34 1.39 BASE A 1.37 1.42 B 1.44(l) 1.44"1 BASE SLO A 1.44(o) 1.44 B 1.38 1.43 TBVOOS A 1.41 1.46 B 1.44") 1.45 TBVOOS SLO A 1.440) 1.48 B 1.40 1.46 RPTOOS A 1.51 1.63 B 1.44(l) 1.48 RPTOOS SLO A 1.53 1.65 OLMCPR limit set by the Single Loop Operation Recirculation Pump Seizure Event (Reference 2).

Page 7 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 TABLE 4-2 Power Dependent MCPR Limit Adjustments And Multipliers (Reference 2)

Core Core Thermal Power (% of Rated)

EoOS Flow 0 I 25 1 <30 >301 45 1 60 1100 Combination (% of Operating Limit MCPR, Operating Limit MCPR rated) MCPR(P) Multiplier, Kp Base *60 2.66 2.66 2.44 1.481 1.280 1.150 1.000

> 60 3.39 3.39 2.93 Base SLO *60 2.68 2.68 2.46 1.481 1.280 1.150 1.000

> 60 3.41 3.41 2.95 RPTOOS *60 2.66 2.66 2.44 1.481 1.280 1.150 1.000

> 60 3.39 3.39 2.93 1 11 RPTOOS < 60 2.68 2.68 2.46 1.481 1.280 1.150 1.000 SLO > 60 3.41 3.41 2.95 T<OO 60 3.07 3.07 2.63 TBVOOS > 60 4.5 4.54$ 4.0 4.54 3.77 1.481 1.280 1.150 1.000 TBVOOS < 60 3.09 3.09 2.65 SLO > 60 4.56 4.56 3.79 1.481 > 1.280 1.150 1000 TABLE 4-3 Flow Dependent MCPR Limits MCPR(F)

(References 2 and 12)

Flow MCPR(F)

(% rated) Limit 0.0 1.7073 30.0 1.53 79.06 1.25 110.0 1.25 Page 8 of 15

Exelon Nuclear Fuels Doe ID: COLR Limerick I, Rev. 8 5.0 Linear Heat Generation Rate Limits 5.1 Technical Specification Section 3.2.4 5.2 Description The LHGR limit is an exposure dependent value. Table 5-1 provides the exposure dependent LHGR limit for all U0 2 pins for all bundles in the Cycle 14 core. Tables 5-2 and 5-3 provide the bounding, exposure dependent LHGR limit for Gad rods in the Cycle 14 core. The LHGR SLO multiplier is shown in Table 5-4.

ARTS provides for power and flow-dependent thermal limit multipliers, which allow for a more reliable administration of the LHGR thermal limits. There are two sets of flow-dependent LGHR multipliers for dual-loop and single-loop operation (References 2 and 5). In addition, there are also two sets of power-dependent LHGR multipliers for use with the Turbine Bypass Valves in service and TBVOOS conditions (Reference 2). Section 7.0 contains the conditions for Turbine Bypass Valve Operability. Thermal limit monitoring must be performed with the more limiting LHGR limit. resulting from the power- and flow-biased calculation. The LHGRFAC(P) curves are independent of recirculation pump trip operability (Reference 2).

TABLE 5-1 Linear Heat Generation Rate Limits - U0 2 Rods All Fuel Types (Reference 8)

Peak Pellet Exposure LHGR Limit (GWD/ST) (kW/ft) 0.00 13.40 14.51 13.40 57.61 8.00 63.50 5.00 TABLE 5-2 Linear Heat Generation Rate Limits - Gad Rods Fuel Types 2530, 2882, 3035, 3038, 3041, 3273, 3274, 3275, 3272, 3040, 3042, 3039, and 2883 (Reference 8)

Peak Pellet Exposure LHGR Limit (GWD/ST) (kW/ft) 0.00 11.76 12.08 11.76 54.21 7.02 59.98 4.39 Page 9 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 TABLE 5-3 Linear Heat Generation Rate Limits - Gad Rods Fuel Type 3271 (Reference 8)

Peak Pellet Exposure LHGR Limit (GWD/ST) (kW/ft) 0.00 12.00 12.17, 12.00 54.59 7.16 60.39 4.48 TABLE 5-4 LHGR Single Loop Operation (SLO) Multiplier (Reference 2)

SLO Multiplier' 0.80 TABLE 5-5 Power Dependent LHGR Multiplier LHGRFAC(P)

(Reference 2)

Core Core Thermal Power (% of rated)

EOOS Flow Combination (% of 0 25 1 <30 1 >30 100

_ rated) LHGRFAC(P) Multiplier Base *_60 0.485 0.485 0.490 0.6340 1.0000

> 60 0.434 0.434 0.473 1______

< 60 0.485 0.485 0.490

> 60 0.434 0.434 0.473 RPTOOS

  • 60 0.485 0.485 ,0.490 0.6340 1.0000

> 60 0.434 0.434 0.473 RPTOOS SLO 60 0.485 0.485 04901.0000

> 60 0.434 0.434 0.473 TBVOOS 660 0.463 0.463 0.490

> 60 0.352 0.352 0.386 TBVOOS SLO >: 60 60 0.352 0.352 0.463 0.463 0.386 0.490

>V S0.6340 1.0000 1Applied through Table 5-6 Page 10 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 TABLE 5-6 Flow Dependent LHGR Multiplier LHGRFAC(F)

(References 2 and 5)

Core* Flow (%'of rated)

EOOS Combination 0 44.074 70 80 110 LHGRFACPF) Multiplier- -

Dual Loop 0.5055 0.973.1..00 singl1e Loop *0.5055 0.80 0.8,.8 0 0.80

. ........ .

. *L05

1. - -Oso,*:.. . . :

Page 11 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 6.0 Control Rod Block Setpoints 6.1 Technical Specification Section 3.3.6 6.2 Description The ARTS Rod Block Monitor provides for power-dependent RBM trips. Technical Specification LimitingCondition for Operation number 3.3.6 requires control rod block instrumentation channels to 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 trip setpoints/allowable values and applicable RBM signal filter time constant data are shown in Table 6-1. The Reactor Coolant System Recirculation Flow Upscale Trip is found in Table 6-2 of this COLR. These setpoints are set high enough to allow full utilization of the enhanced ICF domain up to 110% of rated core flow.

TABLE 6-1 Rod Block Monitor Setpoints' (References 2 and 7)

Power Level Nominal Trip Setpoint' Aliowable Value LTSP 121.5% 121.5%

ITSP 116.5% 116.5%

HTSP 111.0% 111.7%

DTSP 5.0% 2.0%

TABLE 6-2 Reactor Coolant System Recirculation Flow Upscale Trip (Reference 7)

Nominal Trip Setpoint 113.4%

Allowable Value 115.6%

' Based on a cycle-specific rated RWE MCPR limit less than or equal to the minimum cycle OLMCPR. The values provided assume the Rod Block Monitor filter time constant between 0.1 seconds and 0.55 seconds is used.

Page 12 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick I, Rev. 8 7.0 Turbine Bypass Valve Parameters 7.1 Technical Specification Sections 3.7.8 and 4.7.8.c 7.2 Description The operability requirements for the steam bypass system are found in Tables 7-1 and 7-2, If these requirements cannot be met, the MCPR, MCPR(P) and LHGRFAC(P) limits for inoperable Steam Bypass System, known as Turbine Bypass Valve Out Of Service (TBVOOS), must be used.

Additional information on the operability of the turbine bypass system can be found in Reference 10.

TABLE 7-1 Turbine Bypass System Response Time (Reference 4)

Maximum delay time before start of bypass valve opening following generation of the turbine bypass valve flow signal 0.11 sec Maximum time after generation of a turbine bypass valve flow signal for bypass valve position to reach 80% of full flow 0.31 sec (includes the above delay time)

TABLE 7-2 Minimum Required Bypass Valves To Maintain System Operability (Reference 4)

Reactor Power 1 No. of Valves in Service P > 25% 7 Page 13 of 15

Exelon Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 8.0 Stability Protection Setpoints 8.1 Technical Specification Section 2.2.1 8.2 Description The Limerick 1 Cycle 14 OPRM Period Based Detection Algorithm (PBDA) Trip Setpoints for the OPRM System are found in Table 8-1. These values are based on the cycle specific analysis documented in Reference 2. The Cycle 14 OPRM PBDA trip setpoints specified in Table 8-1 require a minimum OLMCPR value of 1.34 (See Section 4.0 MCPR Limits). The setpoints provided in Table 8-1 are bounding for all modes of operation shown in Table 9-1.

TABLE 8-1 OPRM PBDA Trip Setpoints (Reference 2)

PBDA Trip Amplitude Corresponding Maximum Confirmation Count Trip Setting

  • 1.12 14 9.0 Modes of Operation TABLE 9-1 Modes of Operation (References 2 and 5)

EOOS Options Operating Region' Base, Option A or B Yes Base SLO, Option A or B Yes TBVOOS, Option A or B Yes TBVOOS SLO, Option A or B Yes RPTOOS, Option A or B Yes RPTOOS SLO, Option A or B Yes TBVOOS and RPTOOS, Option A or B No TBVOOS and RPTOOS SLO, Option A or B No Operating Region refers to operation on the Power to Flow map with or without FFWTR.

Page 14 of 15

Exeion Nuclear Fuels Doc ID: COLR Limerick 1, Rev. 8 10.0 Methodology The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described' in the following documents:

1. "General Electric Standard Application for Reactor Fuel", NEDE-2401 1-P-A-16, October 2007 and U.S.

Supplement NEDE-240 11-P-A- 16-US, October 2007.

2. "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications", NEDO-32465-A, Rev. 0, August 1996.

11.0 References I. "Technical Specifications and Bases for Limerick Generating Station Unit 1", Docket No. 50-352, License No. NPF-39.

2. "Supplemental Reload Licensing Report for Limerick 1 Reload 13 Cycle 14", Global Nuclear Fuel Document No. 0000-0104-1095-SRLR, Revision 0, January 2010.
3. "GE 14 Fuel Design Cycle-Independent Analyses for Limerick Generating Station Units 1 and 2", GE-NE-L12-00884-00-01P, March 2001.
4. "Final Resolved OPL-3 Parameters for Limerick Unit 1 Cycle 14", TODI ES0900029, December 17, 2009.
5. "ARTS Flow-Dependent Limits with TBVOOS for Peach Bottom Atomic Power Station and Limerick Generating Station", GENE Document NEDC-32847P, June 1998.
6. "General Electric Standard Application for Reactor Fuel", NEDE-2401 1-P-A-16, October 2007 and U.S.

Supplement NEDE-2401 1-P-A-16-US, October 2007.

7. "GE NUMAC PRNM Setpoint Study", LE-0107, Rev. 0, May 1, 2000. Including Minor Revision OC, April 24, 2008.
8. "Fuel Bundle Information Report for Limerick 1 Reload 13 Cycle 14", Global Nuclear Fuel Document No.

0000-0104-1095-FBIR, Revision 0, January 201f0.

9. Deleted.
10. "Tech Eval Stop Valve Load Limit Documentation", Exelon Document IR 917231 Assignment 7, November 11, 2009.
11. "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications",

NEDO-32465-A, Rev. 0, August 1996.

12. "Removal of MCPR(F) Low Flow Correction in NEDC-32847P", GE Nuclear Energy Letter NSA-02-080, February 2002.

Page 15 of 15

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