ML070860494
| ML070860494 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 03/21/2007 |
| From: | David Helker Exelon Nuclear |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| Download: ML070860494 (15) | |
Text
Exelon Nuclear www.exeloncorp.com 2 0 0 Exelon Way Kennett Square, PA 19348 uclear TS 6.9.1.12 March 21, 2007 US. Nuclear Regulatory Commission Attention: Document Control Desk Washington DC 20555 Limerick Generating Station, Unit 2 Facility Operating License No. NPF-85 NRC v
Subject:
Issuance of the Core Operating Limits Report For Reload 9, Cycle 10, Revision 3
Dear Sir/Madam:
Enclosed is a copy of the Core Operating Limits Report (COLR) for Limerick Generating Station, Unit 2, Reload 9, Cycle 10, Revision 3. Revision 3 of this report incorporates the revised cycle specific parameters resulting from the new core configuration implemented during the Limerick Generating Station, Unit 2 refueling outage.
This COLR is being submitted to the NRC in accordance with LGS, Unit 2 Technical Specifications (TS) Section 6.9.1.12.
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 J. Shea, Project Manager [LGS], USNRC
ExeIon Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 3 CORE OPERATING LIMITS REPORT FOR LIMERICK GENERATING STATION U M T 2 RELOAD 9, CYCLE 10 (This is a complete rewrite)
Prepared By:
Prepare r Reviewed By:
iuA-Date: ?/L ?/D ?
Michael R. Holines Independent Reviewer Page 1
Exeloar NucIerts Fuels 1.O Terms and Definitiuns 2.0 General Information 3.U MAPLHGR Limits 4.0 MCPR Limits 5.0 6.0 Control Rod Block Setpoints 7.O 8.0 Stability Protection Setpoints 9.0 Modes of Operation Linear Heat Generation Rate Limits Turbine Bypass Valve Parameters 10.0 Methodology 11.0 References Doc ID: COLR Limerick 2, Rev. 3 Page 4
5 6
7 9
1 1 12 13 13 14 14 Page 2
Exelon Nuclear Fuels rloc ID: COLR Limerick 2, Table 3-1 Table 3-2 Table 4-1 Table 4-2 Table 4-3 Table 5-1 Table 5-2 Table 5-3 Tabk 5-4 Table 6-1 Table 4-2 Table 7-1 MBPLHGR Versus Average Planar Exposure All Fuel Types MAPLMGR Single Loop Operation (SLO) Reduction Factor Operating Limit Minimum Critical Power Ratio (OLMCPR)
Power Dependent MCPR Limit Adjustments and Muttipfiers Flow Dependent MCPR Limits MCPR(F)
Linear Heat Generation Rate Limits LHGR Single Loop Operation (SLQ) Reduction Factor Power Dependent LHGR Multipkr LNGRFACIP)
Flow Dependent LHGR Multiplier LWGRFAC(F)
Rod Block Monitor Setpoints Reactor Coolant System Recirculation Flow Upscale Trip Turbine Bypass System Response Time Table 7-2 Table 8-1 Table 9-1 Modes of Operation Minimum Required Eypass Valves To LVaintain System Qperabiiity QPRM PBDA Trip Setpoints Page 6
6 7
8 8
9 9
10 10 I I 1 1 12 12 13 13 Page 3
ExeZon Nuclear Fuels 1.0 Terms And Definitions DOC ID: COLR Limerick 2, Rev. 3 ARTS BASE CASE DTSP EOOS End of Rated (EQR)
FFVVTR FWHQOS HTSP ICF ITSP LHGR LHGWAC(F)
LHGRFAC( P)
MCPR(F)
MELLLA QLMCPR OPRW PBDA W O O S APRM and RBM Technical Specification Analysis A case analyzed with Turbine Bypass System in service and Recirculation Pump Trip in service and Feedwater Temperature Reduction allowed (FFWTR includes feedwater heater 00s or final feedwater temperature reduction) at any point during the cycle in Dud Loop mode.
Rod Block Monitor Downscale Trip Setpoint Equipment Out of Service The cycle exposure at which reactor power is equat to 3458 MWth with recirculation system flow equal to 100%,, a11 control rods fufly withdrawn, all feedwater heating in service and equilibrium Xenon.
Final Feedwater Temperature Reduction Feedwater Heaters Out of Service Rod Block Monitor High Trip Setpoint Increased Core Flow Rod Btock Monitor Intermediate Trip Setpoint Linear Heat Generation Rate ARTS LHGR thermal I imit f'low dependent adjustments and multipliers ARTS LHGR thermal limit power dependent adjustments and multipliers Rod Block Monitor Low Trip Setpoint Maximum Average Planar Linear Hear Generation Rate Minimum Critical Power Ratio ARTS MCPR thermal limit power dependent adjustments and rntaitipfiers ARTS MCPR thermal I imit flow dependent adjustments and multipliers Maximum Extended Load Line Limit Analysis Operating Limit Minimum Critical Power Ratio Oscillation Power Range Monitor Period Based Detection Algorithm Recircutation Pump Trip Out of Service Page 4
Exelon Nucllear Fuels Doc ID: COLR Limerick 2, Rev. 3 SLhf CPR SLU Single Loop Operation TBVOOS Safety Limit Minimum Critical Poccer Ratio Turbine Bypass Valves Out of Service 2.0 Generat Information This report provides the following cycte-specific parameter limits for Limerick Generating Station Unit 2 Cycle 10:
0 0
0 0
0 0
0 a
Maximum Average Planar Linear Heat Generation Rate (MAPLHCR)
Minimum Critical Power Ratio (MCPR)
Single b o p Operation (SLU) MCPR adjustment ARTS MCPR thermal limit adjustments and multipliers (MCPR(P) or MCPR(F))
ARTS LHGR thermal Iirnit multipfiers (LHGWAC(P) or LWCIZFAC(F))
Rod Block Monitor (RHM) setpoints MAPLHCR single loop operation reduction factor LMCR single loop operation reduction factor Linear Heat Generation Rate (LMCR)
Turbine Bypass Valve parameters Reactor Coolant System Recirculation Flow Upscale Trips Oscillation Power Range Monitor Period Based Detection Algorithm Trip Setpoints These values have been determined using NRC-approved methodology (Reference 61, and are established such that all applicable limits of the plant safety analysis are met.
This report is prepared in accordance with Technicat Specification 6.9. f -9 of Reference 1. Preparation of this report was performed in accordance with Exeion Nuclear, Nuclear Fuels T & M hF-AB-120-3600.
The data presented in this report is valid for all licensed operating domains on the operating map, including:
o o
o Maximum Extended Load Line Limit down to 81% of rated core ffow during fult power operat ion Increased Core Ffow (ICF) up to i IO% of rated core flow Final Feedwater Temperature Reduction (FEWTR) up to 105°F during cycle extension operation Feedwater Heater Out of Service (FWHOOS) up to 60°F feedwater temperature reduction at any time during the cycle prior to cycle extension.
Further information on the cycle specific analyses for Limerick 2 Cycle 10 and the associated operating domains discussed above is available in Reference 2, Page 5
Exelan Nuclear Fuels h c ID: GOLR Limerick 2, Rev. 3 3, I Technical Specification Section 3.2, I 3.2 Description The limiting MMI.GR value for the most limiting lattice (excfuding natural uranium) of each fuel type as a function of average planar exposure is given in Table 3-1 (References 2 and 9), The Iimiting MAPLHGR value is the same far all fuel types in the Limerick Unit 2 Cycle 10 core.
For single loop operation, a reduction factor is used which is shown in Table 3-2 (Reference 2).
MAPFAC(P) and MAFFAC(F) are 1.O for at1 power and flow conditions. The power and flow dependent multipliers for MMLHCR have been removed and replaced with LHGRFAC(P) and LHGRFAC(F;).
TABLE 3-1 MAPLHG-R Versus Average Planar Exposure All Fuel Types (References 2 and 9) 0.0 12.82 I
TABLE 3-2 NAPLHGR Single Loop Operation (SLO) Reduction Factor (Reference 2)
SLO Reduction Factor Page 6
Exelon Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 3 4.0 MCPR Limits
- 4. I 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 I0 fuel types and operating domains. Table 4-1 includes treatment of these MCPR limits for SLQ. Bounding MCPR values are also provided for inoperable Recirculation 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 Management Report or an associated Engineering Change Request.
ARTS provides or power and flow dependent thermal limit adjustments and multipliers, which allow for a mare reliable administration of the MCPR thermal limit. The flow-dependent adjustment MCPR(F) is sufficiently generic to apply to ail fuel types and operating domains (Reference 2). Xn addition, there are also two sets of power-dependent: MCPR multipliers for use with the Turbine Bypass Valves in service and TBVOOS conditions (References 2 and 3).
Section 7.0 contains the conditions for Turbine Bypass Valve Operability. These adjustments are provided in Tabfes 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, The,WCPR(P) curves are independent of recirculation pump trip operability (Reference 3).
TABLE 4-1 Operating timit Itlliaimum Criticat Power Ratio (OLMCPR)'
(Reference 2)
' This table is valid far all Cycle I0 fuel types.
' OLMCPR itmit set ha the Single Loop Operatio11 Recirculation Pump Seimre hrraljsis (Refkrence 2,)
4 When Tau does not equal 0 or L, determine OLMCPR via linear interpolation.
Page 7
Exelion Nuelear Fucis Dac ID: COLR Limerick 2, Rev. 3 TABLE 4*2 Power Dependent MCPR Limit Adjustments And Multipliers (References 2 and 3)
TABLE 4-3 Mow Dependent MCPR Limits MGPR(F)
(Reference 2)
Page 8
Exelon Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 3 5.0 Linear Heat Generation Rate Limits 5.1 Technical Specification Section 3.2.4 5.2 Description The LHGR is m exposure dependent value. Due to the proprietary nature 5f these values only the maximum t i 0 2 LHGR for each fuel type is listed in Table 5-I. For single loop operation, a reduction factor is used which is shown in Table 5-2 (Refaence 2).
ARTS provides for power-and flow-dependent thermal limit multipliers, which allow for a m5re reliable administration of the LHGR thermal limits. There are two sets of flaw-dependent LGWR rnultipf iers for dual-loop and single-loop operation (References 2 and 5). In addition, there are ailso two sets of power-dependent LHGR multipliers for use with the Turbine Bypass Valves in service and TBVOOS conditions (Reference 3). Section 7.0 contains the conditions for Turbine Bypass Valve Operability, The LHGR multipliers are shown in Tables 5-3 through 5-4.
Thermal iimit monitoring must be performed with the more limiting LHGR limit resulting from the power-and flow-biased calculations. The LHGRFAC(P) curves are independent of recirculation pump trip operability (Reference 3).
TABLE 5-1 Linear Heat Generation Rate Limits (References 8 and 10)
TABLE 5-2 LHGR Single Loop Operatian @LO) Reduction Factur (References 2)
SLO Reduction Factor' 0.80 I AppIied through Table 5-4 Page 9
Exelon Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 3 TABLE 5-3 Power Dependent LHGR Itlultiplier LHGIWACfP)
(Reference 3)
TBVOOS TBVUQS SLQ TABLE 5-4 Flow Dependent LHGR Multiplier LHGRIF'AC(F)
(References 2,3, and 9)
EUOS Combination 1
0.80 0.80 Single Loop 0.5055 0.80 0.80 I~-
Page It0
Exelon Nuclear Fuels DQC ID: COLR Limerick 2, Rev. 3 6.0 Control Rod Block Setpoints 6.1 Technical Specification Section 3.3.6 6.2 Description Technical Specification Limiting Condition for Operation number 3.3.6 requires control rod block instrumentation channels shall be OPERABLE with their trip setpoints consistent with the values shown in the Trip Setpoint column of Technical Specification TabIe 3.3.6-2. The Reactor Coolant System Recirculation Flow Upscafe Trip is a cycfe-specific value and as such is found in Table 6-2 of this CQLR. Table 6-2 Iists the Nominal Trip Setpoint and Allowable Vattie, These setpoints are set high enough to allow full utilization of the enhanced ICF domain up to 1 lU% of rated cure flow. Additionatfy, the ARTS Rod Bfrtck,Monitor provides for power-dependent Rlt3,V trips. The trip setpointdatlowable values and applicable RBM signal filter time constant data are shown in Table 6-1.. These values are for use with Technical Specification 3.3.6.
TABLE 6-1 Rod BIock Monitor Setpoints' (References 2 and 7)
Alfowabie Value TABLE 6-2 Reactor Coolant System Recirculation Flow Upscafe Trip (Reference 7)
Yiominal Trip Setpoint 113.4%
These setpoints (with Rod Block Moniior fIlter time constant hetueen 0.1 seconds and 0.55 seconds) itre based on a cyclespecific rated RWE MCPR limit of 1.30, which is less than or equal to the minimum cycle OLMCPR (see COLR references 2 and 7).
This is the MCPR limit (given THEFCVAL POWER is 230.0% and < 90%) below which the RBM is required to he OPERABLE (see COLR reference 2)
This is the MCPR h i t : (given THERMAL POWER is 2 90%) hcfow which the RBM is rcquired to he OPERABLE (see COLR reference 2).
Page 11
Exelon Nuclear Fuels Doc ID: COLR Limerick 2, Rev. 3 7.0 7.1 7 2 Turbine Bypass Valve Parameters Technical Specification Section 3.7.8 and 4.7.8.C Description The operability requirements for the steam bypass system for use in Technical Specifications 3.7.8 and 4.7.8.C are found in Tables 7-1 and 7-2. If these requirements cannot be met, the MCPR, MCPR(P) and LHGRFAC(P) limits far inoperable Steam Bypass System, known as Turbine Bypass Valve Out Of Service, must be used.
TABLE 7-1 (Reference 4)
Turbine Bypass System Response Time Maximum delay time before start of bypass vaIve opening 1
following generation of the turbine bypass vatve flow signai
- 0. I 1 sec Maximum time after generation of a turbine bypass valve trow signal for bypass valve position to reach 80% of full flow (includes the above delay time) 0.31 see TABLE 7-2 Minimum Required Bypass Valves Ta Maintain System Operability f References 4)
Reactor Power I
NO. of Valves in Service Page 12
Exelan Nuclear Fuels 8.0 Stability Protection Setpoints 8.1 Technical Specification Section 2.2. I.2.F 8.2 Description The Limerick 2 Cycle 10 OPRM Perioi Baset Doc ID: COLR Limerick 2, Rev. 3 Detection lgorithm (PBDA) Trip Setpoints for the O P M System for use in Technical Specification 2.2.1 are found in Table 84. These values are based an the cycle specific analysis documented in Reference 2. The setpoints provided in Table 8-f are bounding for all modes of operation shown in Tabk 9-1.
TABLE 8-1 OPRM PBDA Trip Setpointd (Reference 2) 9.0 Modes Of Operation TABLE 9-1 Modes of Operation (Reference 2,3 and 5)
I The station has conservatively decided ta maintain the PDSA Trip Amplitude at 1.12 with a Corresponding Maximum Confirmation Count Trip Setting of 14 unrif such time where these changes do not introduce a Uiiit difference nt Limerick. This decision was agrccd upon by Operaticins and Sitc Engineering
' Operating Region refers to operirlt-rn on the Power to Flow nap with or svithout FWTR.
Page 13
Exelon Nuclear Fueis Doc ID: COLR Limerick 2, Rev. 3 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 document:
- 1. General Electric Standard Application for Reactor Fuel, NEDE-2401 I-P-A-IS, September 2005 and U.S. Supplement LNEDE-2401 1-P-A-15-US, Revision 15, September 2005.
11.O References 1.
- 2.
- 3.
- 4.
- 5.
- 6.
- 7.
- 8.
- 9.
- 10.
Technical Specifications and Bases for Limerick Generating Station Unit 2, Docket No. 50-353, License No. NPF-85.
Supplemental Reload Licensing Report for Limerick Generating Station Unit 2 Reload 9 Cycle 1 o, Global. Nttctear Fuel Document No. Q004)-0054-5796-SRLR, Revision 0, January 2007.
GEl4 Fuel Design Cycle-Independent Analyses for Limerick Generating Station Units I and 2, GE-NE-L 1 2-Qo884-OO-O I P, Revision 0, March 200 I.
OPL-3 Transient Protection Parameters Verification for Reload Licensing Analyses for Limerick 2 Reload 9 Cycle lo, Exefon Document No. TODI 06003 1 1 Revision 0, November 20%.
ARTS Flow-Dependent Limits with TBVOUS for Peach Bottom Atomic Power Station and Limerick Generating Station, GENi Document NEDC-328479, Revision 0, June 19%.
Generat Electric Standard Application for Reactor Fuel, GNF Document NEDE-240 1 1 -P-A-15, September 2005 and U.S, SuppIement NEDE-240 t I -P-A-I 5-US, Revision 1 5, September 2005, Power Range Neutron Monitoring System Setpoint Calculations Limepic k Generating Station, Units f &
2 Mod. ho. P#224, Exelon Calculation: LE-0107, Revision 0 {including minor revision 0A>,
March 2000.
Fuel Bundle Information Report for Limerick Generating Station Unit 2 Reload 9 Cycle fO, Global Nuclear Fuel Document No. 0000-0054-5796-ERXR, Revision 0, January 2007.
Suppfementaf Reload Licensing Report for Limerick Generating Station Unit 2 Reload 8 Cycle 9, Global Nuclear Fuel Document No. 0000-003 1 -7705-SER, Revision 0, January 2005.
Fuel Bundle Information Report for Limerick Generating Station Unit 2 Reload 8 Cycle 9+ Global Nuclear Fuel Document No. oo00-003 1-7705-FBIR.
Revision 0. Sanuaw 2005.