GO2-09-050, Supplemental Response to Request for Additional Information Regarding License Amendment Request Involving Core Operating Limits Report and Scram Time Testing
| ML091040762 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/19/2009 |
| From: | Gambhir S Energy Northwest |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GO2-09-050 | |
| Download: ML091040762 (17) | |
Text
j Sudesh K. Gambhir ENERGY Vice President, Technical Services P.O. Box 968, Mail Drop PE04 NORTHWEST Richland, WA 99352-0968 Ph. 509-377-8313 F. 509-377-2354 sgambhir@energy-northwest.com March 19, 2009 G02-09-050 U.S. Nuclear Regulatory Commission 10 CFR 50.90 ATTN: Document Control Desk Washington, D.C. 20555-0001
Subject:
COLUMBIA GENERATING STATION, DOCKET NO. 50-397 SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING
References:
- 1) Letter G02-08-108 dated July 16, 2008, SK Gambhir (Energy Northwest) to NRC, "License Amendment Request for Changes to Technical Specifications Involving Core Operating Limits Report and Scram Time Testing," (ADAMS Accession No. ML082250680).
- 2) Letter dated November 25, 2008, NRC to JV Parrish (Energy Northwest),
"Request for Additional Information Related to Request for Changes to Technical Specifications Involving COLR and Scram Time Testing,"
(ADAMS Accession No. ML083190804).
- 3) Letter G02-09-001 dated January 2, 2009, SK Gambhir (Energy Northwest) to NRC, "Response to Request for Additional Information (RAI)
Regarding License Amendment Request Involving Core Operating Limits Report and Scram Time Testing," (ADAMS Accession No. ML090230569)
Dear Sir or Madam:
By Reference 1, Energy Northwest requested changes to the Columbia Generating Station (Columbia) Technical Specifications involving the Core Operating Limits Report (COLR) and Scram Time Testing. Via Reference 2, the NRC requested additional information related to the Energy Northwest submittal.
With Reference 3, Energy Northwest provided a response that addressed five of the staffs six questions, as the information required to fully address one of the questions was not yet available. The complete response to the remaining question is provided as Attachment 1.
The information contained in this supplement does not impact the original no significant hazards determination. There are no new commitments contained in this letter. 400(
Ltw
SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 2 If you have any questions or require additional information,' please contact MC Humphreys at 509-377-4025.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the date of this letter.
IR*," ectfully, Sambhir Vice President, Technical Services.
Attachment:
Supplemental Response to Request for Additional Information cc: EE Collins, Jr. - NRC RIV NRC Senior Resident Inspector/988C RN Sherman - BPA/1 399 CF Lyon - NRC NRR WA Horin - Winston & Strawn
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 1 of 15 Supplemental Response to Request for Additional Information Question 1:
The application dated July 16, 2008, describes a transition from the current core with a full loading of ATRIUM-I0 to a full loading of GEl 4 fuel for the [Columbia Generating Station]
CGS. Accordingly, the application describes a transition to [Global Nuclear Fuels] GNF analytical methodologies (NEDC-33419P and NEDE-2401 1-P-A). Once the NRC staff approves the inclusion of these methodologies into the TS, the licensee calculates the core operating limits listed in TS 5.6.5.a and performs the plant accident analyses without further review by the NRC staff, consistent with Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits for Technical Specifications." Therefore, the NRC staff requests that the licensee provide additional information to allow the staff to review the impact of the transition on the safety analyses.
Specifically, for the affected analyses performed for the inftial transition core:
a) Please state the approved methodology and/or the computer codes used and the reference (e.g. topical report) documenting the methodology/computer codes; b) Please state if the analysis is in compliance with all applicable restrictions in the staff safety evaluation(s) approving the methodology; and c) Please provide the quantitative results of the figure(s) of merit (FOM) compared against the acceptance criteria.
For the LOCA analysis, provide a narrative showing that:
a) the limiting break location, break size, and single failure(s) were identified and used; b) the limiting power/flow conditions and axial power shape were used; and c) the legacy fuel analysis (MAPLHGR) will continue to be applicable through the transition cycles.
Similar information was provided for the Grand Gulf amendment request for a similar transition (ADAMS Accession No. ML082470361).
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 2 of 15
Response
Anticipated Transient Without Scram (ATWS)
Methodology Comply With All ICode(s) Staff Approval Applicable S ARestrictions And Used Conditions ISCOR09 Note 1 Yes PANACl I NEDE-30130-P-A, "Steady-State Nuclear Yes Methods," April 1985 - Note 2 ODYN09 NEDC-24154P-A, "Qualification of the One- Yes Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement I -
Volume 4)," Revision 1, February 2000 STEMP04 Note 3 N/A TASC03 NEDC-32084P-A, "TASC-03A, A Computer Yes Program for Transient Analysis of a Single Channel," Revision 2, July 2002 NOTES:
1 - The ISCOR code is not approved by name. However, the Safety Evaluation (SE) supporting approval of NEDE-24011-P Rev. 0 by the May 12, 1978 letter from D.G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and Loss of Coolant Accident (LOCA) applications is consistent with the approved models and methods.
2 - The physics code PANACEA provides inputs to the transient code ODYN. The use of PANAC Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE),
Subject:
"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC No.
MA6481), November 10, 1999.
3 - The STEMP code uses fundamental mass and energy conservation laws to calculate the suppression pool heat up.. The use of STEMP was noted in NEDE-24222, "Assessment of BWR Mitigation of ATWS, Volume I & I! (NUREG-0460 Alternate No. 3)," December 1, 1979. The code has been used in ATWS applications since that time. There is no formal NRC review and approval of STEMP.
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 3 of 15 The following table illustrates the results of the FOM compared to the acceptance criteria.
ATWS FOM Comparisons to Criteria Parameter Result Acceptance Criteria Peak Vessel Pressure (psig) 1449 _<1500 Peak Suppression Pool 189.4 _<204.5 Temperature (OF)
Peak Containment Pressure 12.4 <45.0 (psig)
Peak Cladding Temperature (0F) 1542 _<2200 Peak Local Cladding Oxidation Insignificant(1) *517
(%)
(1) The calculated peak cladding temperature is less than 1600°F and therefore cladding oxidation is insignificant compared to the acceptance criteria and is not explicitly calculated.
American Society of Mechanical Engineers (ASME) Overpressure Methodology Comply With All.
/ Code(s) Staff Approval Applicable Used Restrictions And Conditions PANAC1 1 NEDE-30130-P-A, "Steady-State Nuclear Yes Methods," April 1985 - Note 1 ODYN09 NEDC-24154P-A, "Qualification of the One- Yes Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement 1 - Volume 4)," Revision 1, February 2000 NOTE:
1 - The physics code PANACEA provides inputs to the transient code ODYN. The use of PANAC1 1 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE),
Subject:
"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC No.
MA6481), November 10, 1999.
SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 4 of 15 The following table illustrates the results of the FOM compared to the acceptance criteria.
ASME Overpressure FOM Comparisons to Criteria TS ASME Steam Line Dome Vessel Dome Vessel Event Pressure Pressure Pressure Pressure Limit (2)
(psig) (psig) (psig) Limit ) (psig)
(psig)
MSIV Closure (Flux Scram) 1301 1305 1341 1325 1375
- ICE (HBB)
MSIV Closure (Flux Scram) 1296 1300 1328 1325 1375
- ELLLA (HBB)
NOTES:
1 - Technical Specification 2.1.2 Reactor Coolant System Pressure Safety Limit.
2 - GESTAR II, Revision 16, Section S.3 Vessel Pressure ASME Code Compliance Model.
Anticipated Operational Occurrences (AOOs) Including Off-Rated Limits Methodology Comply With All Applicable I Code(s) Staff Approval Restrictions And Used Rsrcin n Conditions ISCOR09 Note 1 Yes PANACi 1 NEDE-30130-P-A, "Steady-State Nuclear Yes Methods," April 1985, Note 2 ODYN09 NEDC-24154P-A, "Qualification of the One- Yes Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement 1 -
Volume 4)," Revision 1, February 2000 TASC03 NEDC-32084P-A, "TASC-03A, A Computer Yes Program for Transient Analysis of a Single Channel," Revision 2, July 2002 NOTES:
1 - The ISCOR code is not approved by name. However, the SE supporting approval of NEDE-24011-P Rev. 0 by the May 12, 1978 letter from D.G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer
SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 5 of 15 code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods.
2 - The physics code PANACEA provides inputs to the transient code ODYN. The use of PANAC Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from' S.A. Richards (NRC) to G.A. Watford (GE),
Subject:
"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC No.
MA6481), November 10, 1999.
Limiting Pressurization Events OLMCPR(1) Summary Table Application Condition I Exposure Range Option B Option A Equipment in Service BOC to MOC 1.30 1.33 MOC to EOC 1.39 1.42 End-of-Cycle Recirculation Pump Trip (EOC-RPT) Out-of-Service (OOS)
BOC to MOC 1.36 1.47 MOC to EOC 1.46 1.63 Turbine Bypass Valve (TBV) OOS BOC to MOC 1.36 1.39 MOC to EOC 1.44 1.47 EOC-RPT OOS + TBV OOS BOC to MOC 1.41 1.52 MOC to EOC 1.50 1.67 NOTE:
1 - The Operating Limit Minimum Critical Power Ratio (OLMCPR) values presented apply to rated power operation based on the two-loop operating safety limit Minimum Critical Power Ratio (MCPR).
SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 6 of 15 Power Dependent MCPR Limits for GE14 and Atrium10 MCPRp Below Pbypass Kp Multiplier Above Pbypass 25% P 25% P 30% P 30% P App. Exp. <50% >50% <50% >50% 30% P 45% P 60% P 85% P 100% P flow flow flow flow 1 EOC 2.24 2.24 2.15 2.15 1.483 1.280 1.150 1.072 1.000 2 EOC 2.24 2.24 2.15 2.15 1.483 1.280 1.150 1.072 1.000 3 EOC 3.12 3.28 2.69 2.89 1.483 1.280 1.150 1.072 1.000 4 EOC 3.12 3.28 2.69 2.89 1.483 1.280 1.150 1.072 1.000 NOTES:
Application Group 1: Equipment in Service Application Group 2: End of Cycle - Recirculation Pump Trip (EOC-RPT) Out-of-Service (OOS)
Application Group 3: Turbine Bypass Valve (TBV) OOS Application Group 4: EOC-RPT OOS + TBV OOS F
Flow Dependent MCPR Limits for GE14 and Atrium10 30% F 90% F 108.5% F 1.65 1.25 1.25
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 7 of 15 Loss of Coolant Accident (LOCA)
Methodology Comply With All Methodolog StApplicable Code(s)
U Staff Approval Restrictions And Used Conditions ISCOR09 Note 1 Yes LAMB08 NEDE-20566P-A, "General Electric Analytical Yes Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K," September 1986 SAFER04/ NEDE-23785-1.-PA, "The GESTR-LOCA and Yes GESTR08 SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 1, GESTR-LOCA - A Model for the Prediction of Fuel Rod Thermal Performance," Revision 1, October 1984 NEDE-23785-1-PA, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 2, SAFER - Long Term Inventory Model for BWR Loss-of-Coolant Analysis," Revision 1, October 1984 NEDE-23785-1-PA, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 3, SAFER/GESTR Application Methodology," Revision 1, October 1984 NEDC-32950P, "Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model,"
January 2000 as reviewed by letter from S. A.
Richards (NRC) to J. F. Klapproth (GE), "General Electric Nuclear Energy (GENE) Topical Reports NEDC-32950P and NEDC-32084P Acceptability Review," May 24, 2000 NEDE-23785P-A, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 3 Supplement 1, Additional Information for Upper Bound PCT Calculation,"
Revision 1, March 2002 Note 2 TASC03 NEDC-32084P-A, "TASC-03A, A Computer Yes Program for Transient Analysis of a Single Channel," Revision 2, July 2002
SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 8 of 15 NOTES (for LOCA table):
1 - The ISCOR code is not approved by name. However, the SE supporting approval of NEDE-24011-P Rev. 0 by the May 12, 1978 letter from D.G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods.
2 - The ECCS-LOCA evaluations are in compliance with the one SE restriction upon the SAFER/GESTR-LOCA methodology. This SE restriction is described within NEDE-23785-1 -PA Rev. 1, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 3, SAFER/GESTR Application Methodology, October 1984, as "NRC will require a demonstration that the PCT value calculated by the licensing method always exceeds the upper-bound value." All ECCS-LOCA evaluations explicitly calculate both the licensing Peak Cladding Temperature (PCT) and the upper bound PCT, and all evaluations report a licensing PCT, which exceeds the upper bound PCT. Note that the 1,600°F restriction on Upper Bound PCT in NEDE-23785-1-PA Rev. 1, Vol. 3, October 1984 was removed by NEDE-23785P-A Rev. 1, Vol. 3 Supplement 1, March 2002.
The following table quantifies the ECCS-LOCA evaluation results and compares those results to the 10 CFR 50.46 acceptance criteria. The results show that all 10 CFR 50.46 acceptance criteria are satisfied.
GE14 SAFER/GESTR-LOCA Results Parameter GEI4 Result Acceptance Criteria
- 1) Licensing Basis < 1710°F < 2200°F PCT
- 2) Maximum Local < 1% 5 17%
Oxidation
- 3) Core Wide Metal- < 0.1% < 1.0%
Water Reaction
- 4) Coolable Parameters 1 AND 2 Satisfied by:
Geometry PCT < 2200°F AND Maximum Local Oxidation
< 17%.
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 9 of 15 Parameter GE14 Result Acceptance Criteria
- 5) Core Long Term Satisfied by: Core temperature Cooling EITHER acceptably low Core reflooded above AND Top of Active Fuel Long-term decay heat OR removed.
Core reflooded to the elevation of jet pump suction and 1 core spray system in operation.
LOCA Specific Question For the LOCA analysis, also provide a narrative showing that:
a) the limiting break location, break size, and single failure(s) were identified and used.
Response
The ECCS-LOCA evaluations evaluate all potentially limiting single failures including:
- Low Pressure Core Spray (LPCS) emergency diesel generator.
The ECCS-LOCA evaluations identify the HPCS emergency diesel generator as the limiting single failure.
The ECCS-LOCA evaluations examine all potentially limiting break locations including:
" Recirculation suction line.
- Core spray line.
" Feedwater line.
" Low pressure coolant injection line.
" Steam line (both inside and outside containment).
The ECCS-LOCA evaluations examine break sizes ranging from a double-ended guillotine break of recirculation suction line to small breaks for which no core heat up is predicted. The break sizes are selected to satisfy the 10 CFR 50, Appendix K requirement for the use of the Moody Slip Flow Model'with a discharge coefficient of 1.0, 0.8, and 0.6. Additional break sizes are evaluated to fully characterize the PCT versus break size response and to ensure that the limiting break size is identified.
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 10 of 15 The limiting large break is identified as the double-ended guillotine break of the recirculation suction line and the limiting small break is identified as a 0.07 ft2 break of the recirculation suction line. The 0.07 ft2 break is identified as the overall most limiting break.
LOCA Specific Question For the LOCA analysis, also provide a narrative showing that:
b) the limiting power/flow conditions and axial power shape were used.
Response
The potentially limiting power and flow conditions are evaluated including the rated power/flow point and the point on the Extended Load Line Limit Analysis (ELLLA) rod line with highest power and lowest flow. Calculations are performed at these power/flow points using both nominal and 10 CFR 50 Appendix K conditions. The rated power/flow point is identified as the limiting power and flow condition.
Single loop operation is investigated and shown to be bounded by two-loop operation.
Likewise, feedwater temperature reduction is also examined and shown to be bounded by LOCA at rated feedwater temperature. Increased core flow is also dispositioned as bounded by LOCA at rated core flow.
Both mid peaked and top peaked axial power shapes are evaluated by the ECCS-LOCA evaluations. The limiting axial power shape is identified as mid peaked for large breaks and the limiting axial power shape for small breaks is identified as top peaked.
LOCA Specific Question For the LOCA analysis, also provide a narrative showing that:
c) the legacy fuel analysis (MAPLHGR) will continue to be applicable through the transition cycles.
Response
The legacy fuel MAPLHGRs remain applicable through the transition cycles because they are unaffected by the new SAFER/GESTR-LOCA evaluation. Thermal hydraulic compatibility is demonstrated for the legacy fuel and the GE14 fuel; consequently, the mixed core does not invalidate the legacy fuel MAPLHGRs.
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 11 of 15 Stability Methodology Comply With All MCode(s) Staff Approval Applicable UCds SRestrictions And Used Conditions ISCOR09 Note 1 Yes PANAC1 I NEDE-30130-P-A, "Steady-State Nuclear Yes - Note 2 Methods," April 1985 ODYSY05 NEDC-32992P-A, "ODYSY Application for Stability Licensing Calculations," July 2001.
NEDE-33213P, "ODYSY Application for Yes Note 4 Stability Licensing Calculations Including Y Option I-D and II Long Term Solutions," May 2007 - Note 4.
Detect & NEDO-32465-A, "Reactor Stability Detect and Suppress Suppress Solutions Licensing Basis Methodology / Methodology for Reload Applications," August TRACG04 1996.
Yes - Note 3 NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology, including Supplement 1,"
November 1995.
NOTES:
1 - The ISCOR code is not approved by name. However, the SE supporting approval of NEDE-24011-P Rev. 0 by the May 12, 1978 letter from D.G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods.
2 - The physics code PANACEA provides inputs to the stability codes ODYSY and TRACG. The use of PANAC1 1 was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE),
Subject:
"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC No.
MA6481), November 10, 1999.
3 - TRACG was approved for use in the DIVOM analysis in NEDO-32465-A. The TRACG version at that time was TRACG02. TRACG04 has been shown to provide the same results in DIVOM applications as TRACG02. It is thus deemed
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 12 of 15 acceptable to use TRACG04 for the DIVOM application. The use of TRACG04 will be documented per CGS's 10 CFR'50.59 evaluation process.
4 - The final safety evaluation for NEDE-33213P was issued by the NRC on February 23, 2009 via letter from S.L. Bone (NRC) To D. Coleman (BWROG Chair),
Subject:
"Final Safety Evaluation for Boiling Water Reactors Owners' Group (BWROG) Licensing Topical Report (LTR) NEDE-33213P, "Application for Stability Licensing Calculations Including Option I-D and II Long Term Solutions" (TAC No. MD5743)," (ADAMS Accession No. ML090230577).
For the stability Long-Term Solution (LTS) Option III, the figure of merit (FOM) is the operating limit minimum critical power ratio (OLMCPR). The OLMCPR should be greater than or equal to the stability-based operating limit given the Oscillation Power Range Monitor (OPRM) Trip Amplitude Setpoint. In addition, the FOM associated with Backup Stability Protection (BSP) regions is the modification of the power/flow map to exclude regions on the power/flow map susceptible to thermal-hydraulic instability oscillations.
Stability-based OLMCPR results for specific OPRM trip set points will be provided for the transition core when the analysis has been completed. The evaluation provides adequate protection against violation of the safety limit minimum critical power ratio for the two postulated reactor instability events, Steady State (SS) and 2 Recirculation Pump Trip (2PT), as long as the plant OLMCPR is equal to or greater than the OLMCPR(SS) and OLMCPR(2PT) for the selected OPRM amplitude setpoint. The table below provides the OPRM Setpoint Versus OLMCPR values calculated with a plant- and cycle-specific DIVOM slope in accordance with GE-NE-0000-0028-9714-RI (Plant-Specific Regional Mode DIVOM Procedure Guideline, June 2005).
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 13 of 15 OPRM Setpoint Versus OLMCPR OPRM Amplitude OLMCPR(SS) OLMCPR(2PT)
Setpoint 1.05 1.244 1.184 1.06 1.267 1.205 1.07 1.291 1.228 1.08 1.324 1.259 1.09 1.367 1.300 1.10 1.413 1.344 1.11 1.450 1.379 1.12 1.470 1.399 1.13 1.490 1.417 1.14 1.510 1.437 1.15 1.514 1.440 OLMCPR Off-rated Acceptance OLMCPR Rated Power Criteria @45% flow OLMCPR BSP end points are calculated in accordance with OG-002-0119-260, "Backup Stability Protection (BSP) for Inoperable Option III Solution, July 2002," for the transition core using ODYSY05. The power/flow points and associated decay ratios are provided in Table 1 and Table 2. The stability scram and controlled entry region boundaries for normal and minimum feedwater temperature are plotted on a power/flow map in Figures 1 and 2 respectively.
Table 1 - Intercepts for Normal Feedwater Temperature (NFWT)
Core Decay Highest Power (%) Flow (%)Ratio Channel Decay Ratio 63.8 40.0 < 0.796 < 0.429 33.8 23.8 0.793 0.365 71.8 50.0 < 0.797 < 0.412 25.1 23.8 0.786 0.346
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 14 of 15 Table 2 - Intercepts for 'Minimum Feedwater Temperature (MFWT)
Highest Power () Flow ()RatioChne Core Decay Channel Decay Ratio 67.4 44.3 0.795 0.395 28.5 23.7 0.795 0.285 71.8 50.0 < 0.793 < 0.374 24.5 23.4 0.796 0.372 Figure I - BSP Regions for NFWT
--- NauralC culation Line II I III II I
- etonded Opraling Doafm I I
.5 -.. . , -.... ... -* - - ...J .- .
-- 3SP Scram R gon Boundary I 110.00~
-- ti BSP Comroled 9my PRegon oundar-y Wo~o I I I I. . . . ..
4----
. . ... . . .I . . . .I.. . .
I I
- - L L-------------- ------ ------ --------------------------
lF I IIII II 70.00- L - -- - - I I-60.00 IL E
41 50.00 ,-
I
-4 ----
I
/ I I ~ II I I 40.00
- I I I II I I I I I I I II I I I I 30.00 I- _ . . . . I. . . J . . I . . . . . .. I 20.00
L ------- -j-- -i-- l .-- 4---1-------
10.00 I- - -- - I I- - I -- I 0.00 0 10 20 30 40 50 60 70 90 90 100 110 120 Core Flow (V Figure 1 BSP Regions for NFWT
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REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Attachment Page 15 of 15 Figure 2 - BSP Region3 for MFWT I IW.UU
-- e Natura Circulation Line I I I i i I
g I
I I
aE xtande: Operating Domairn I I ,
100.00
)--- BSP Scram Region Boundary . "
SP Controlod E6i" 13-- Fegon Boundary 90,00 I I IIIIII L S- - L - - ---------------
- --- - -- - -- i -- - - - ------- ----------------- - ------- ----.
I I 1 I I IIII I
... .- . -I . - . ..- ----II.----. - I
-- --- IIII
- I--------------.
-A' IL I I I I-I-I I II .-.-.- I III 50.00.
40.00 I I I I ! I I I 11 I II 30.00 20.00
-I--------------
l------------------------I--- -
10.00 I I I -I I I I I I I I I I I I. I I I I I 0.00 0 10 20 30 40 50 60 70 s0 90 100 110 120 Core Flow (%9 Figure 2: BSP Regions for 1IFWT