ML090230569
| ML090230569 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/02/2009 |
| From: | Gambhir S Energy Northwest |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GO2-09-001 | |
| Download: ML090230569 (23) | |
Text
ENERGY NORTHWEST Sudesh K. Gambhir Vice President, Technical Services P.O. Box 968, Mail Drop PE04 Richland, WA 99352-0968 Ph. 509-377-8313 F. 509-377-2354 sgambhir@energy-northwest.com January 2, 2009 G02-09-001 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001
Subject:
COLUMBIA GENERATING STATION, DOCKET NO. 50-397 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 J. V. 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).
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.
The Energy Northwest response to the Reference 2 request for additional information is provided in Attachment 1. Some of the information requested, related to specific values determined by the accident and transient analyses, is not yet available for the initial transition core. The expected date for submittal of this additional information is March 19, 2009.
Enclosed with this letter is a report (Enclosure 2) that provides the requested information on the mixed core analysis. The report, GNF S-0000-0092-8136P Revision 0, "GE14 Thermal Hydraulic Compatibility With Columbia Legacy Fuel," contains information that is proprietary to Global Nuclear Fuels - Americas (GNF-A). The proprietary information is requested to be withheld from public disclosure in accordance with 10 CFR 9.17(a)(4) and 10 CFR 2.390(a)(4). An affidavit attesting to the proprietary nature of the information is provided in. A non-proprietary version of the document is also enclosed (Enclosure 3).
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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 2 There is one new commitment contained in this letter which is identified in Attachment 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.
R ctfully,.
Vice President, Technical Services Attachments:
- 1. Response to Request for Additional Information
- 2. List of Regulatory Commitments
Enclosures:
- 1. Affidavit for the Request to Withhold Information
- 2. GNF S-0000-0092-8136P, Revision 0, "GEI4 Thermal Hydraulic Compatibility With Columbia Legacy Fuel," October 2008 (proprietary version)
- 3. GNF S-0000-0092-8136, Revision 0, "GE14 Thermal Hydraulic Compatibility With Columbia Legacy Fuel," October 2008 (non-proprietary version) cc:
EE Collins, Jr. - NRC RIV NRC Senior Resident Inspector/988C RN Sherman - BPA/1 399 CF Lyon - NRC NRR WA Horin - Winston & Strawn
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 1 of 16 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-1 0 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 initial 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).
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 2 of 16
Response
Anticipated Transient Without Scram (ATWS)
Methodology Comply With All SCode(s)
Staff Approval Applicable Used Restrictions And Conditions ISCOR09 Note 1 Yes PANAC1 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 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. O0by 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-24011-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 & II (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.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 3 of 16 The following table illustrates the results of the figures of merit (FOM) (** data to be supplied when analysis has been completed) compared to the acceptance criteria.
ATWS FOM Comparisons to Criteria Parameter Result Acceptance Criteria Peak Vessel Pressure (psig)
- 1500 Peak Suppression Pool
<204.5 Temperature (OF)
Peak Containment Pressure
- 45.0 (psig)
Peak Cladding Temperature (OF)
_2200 Peak Local Cladding Oxidation
<17
(%)
American Society of Mechanical Engineers (ASME) Overpressure Methodology Comply With All Applicable I Code(s)
Staff Approval Restrictions And Used 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 I -
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.
Wafford (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..
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 4 of 16 The following table illustrates the results of the FOM (** data to be supplied when analysis has been completed) 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 (1)
(psig)
(psig)
MSIV Closure (Flux Scram) 1325 1375
- ICF (HBB) 1325_1375 MSIV Closure (Flux Scram) 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 Conditions ISCOR09 Note 1 Yes PANAC1 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-2401 1-P Rev. 0 by the May 12, 1978 letter from D. G.
Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable,
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 5 of 16 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 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.
The following tables** and associated data** will be provided when analysis has been completed for the initial transition core.
'Limiting Pressurization Events OLMCPR(1) Summary Table Application Condition / Exposure Range Option B Option A Equipment in Service BOC to MOC MOC to EOC End-of-Cycle Recirculation Pump Trip (EOC-RPT) Out-of-Service (OOS)
BOC to MOC MOC to EOC Turbine Bypass Valve (TBV) OOS BOC to MOC MOC to EOC EOC-RPT OOS + TBV OOS BOC to MOC MOC to EOC 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).
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING-Page 6 of 16
- Off-Rated Power Dependent 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 Group flow flow flow flow 1
EOC 2
EOC 3
EOC 4
EOC Notes:
Application Group 1: Equipment in Service Application Group 2: End of Cycle - Recirculation Pump Trip Application Group 3: Turbine Bypass Valve (TBV) OOS Application Group 4: EOC-RPT OOS + TBV OOS (EOC-RPT) Out-of-Service (OOS)
- MCPRf Limits for GE14 and Atrium10 30% F 50% F 70% F 110% F
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 7 of 16 Loss of Coolant Accident (LOCA)
Methodology Comply With All Methodelog StApplicable U
Code(s)
Staff Approval Restrictions And Used Conditions ISCOR09 Note 1 Yes LAMB08 NEDE-20566P-A, "General Electric Analytical Model Yes for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K," September 1986 SAFER04/
NEDE-23785-1 -PA, "The GESTR-LOCA and SAFER Yes GESTR08 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, SAFERJGESTR 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 Program Yes for Transient Analysis of a Single Channel," Revision 2, July 2002
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 8 of 16 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 SAFERIGESTR-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, October1984 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 GE14 Result Acceptance Criteria
- 1) Licensing Basis
< 1710 OF
< 2200 OF PCT
- 2) Maximum Local
< 1%
< 17%
Oxidation
- 3) Core Wide Metal-
< 0.1%
< 1.0%
Water Reaction
- 4) Coolable Parameters 1 AND 2 Satisfied by:
Geometry PCT < 2200 OF AND Maximum Local Oxidation
< 17%
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 9 of 16 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:
High Pressure Core Spray (HPCS) emergency diesel generator
" Low Pressure Core Spray (LPCS) emergency diesel generator
" Low Pressure Coolant Injection (LPCI) 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.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 10 of 16 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 GEl 4 fuel; consequently, the mixed core does not invalidate the legacy fuel MAPLHGRs.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 11 of 16 Stability Methodology Comply With All Applicable U
Code(s)
Staff Approval Restrictions And Used Conditions ISCOR09 Note 1 Yes PANACl 1 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 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-2401 1-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.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 12 of 16 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 acceptable to use TRACG04 for the DIVOM application. The use of TRACG04 will be documented per the utility 10 CFR 50.59 evaluation process.
4-The final SE for NEDE-33213P is expected to be published by the NRC in January 2009. If the issuance of the SE is delayed, the utility 10 CFR 50.59 evaluation process will be applied.
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). The associated data (**) for this table will be provided when analysis has been completed.
OPRM Setpoint Versus OLMCPR OPRM Amplitude OLMCPR(SS)
OLMCPR(2PT)
Setpoint 1.05 1.06 1.15
@45% flow
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 13 of 16 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 Intercepts for Normal (NFWT) and Minimum (MFWT)
Feedwater Temperature and associated decay ratios will be provided in tables similar to the following ** when the analysis has been completed.
- Intercepts for Normal Feedwater Temperature Highest Core Decay Channel Power ()
Flow ()RatioChne Decay Ratio
- Intercepts for Minimum Feedwater Temperature Highest Core Decay Channel Power (%)
Flow (%)
Ratio Decay Ratio
"**/
The associated BSP Regions for NFWT and MFWT stability regions will be provided on respective power/flow maps when the analysis has been completed.
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 14 of 16 Question 2:
The application states that the Westinghouse SVEA-96 fuel may be re-introduced from the discharged fuel in the spent fuel pool, if failed fuel replacement is required. What is the licensing basis that permits the GNF methodologies to model the SVEA-96 fuel?
Response
Energy Northwest recognizes that there is no current licensing basis at Columbia that permits use of GNF methodologies to model SVEA-96 fuel. As mentioned in the application dated July 16, 2008, to the U.S. Nuclear Regulatory Commission (NRC) (Agencywide Document Access and Management System (ADAMS) Accession No. ML082250680), the potential to re-insert Westinghouse fuel for such an occurrence as a fuel failure would warrant using the Westinghouse analyses to establish the appropriate fuel thermal limits, and as such, maintaining these documents in the Columbia Technical Specifications would preclude the need for a TS change in the event these documents were needed. In the postulated fuel failure replacement scenario, any re-insertion of fuel would be conducted along with a core redesign and re-analysis, and if GNF methodologies were to be employed to model SVEA-96 fuel, then that specific analyses would be submitted to the NRC for prior review and approval.
Question 3:
Please provide the mixed core analysis demonstrating the thermal-hydraulic compatibility of ATRIUM-10 and GEI4.for CGS.
Response
Enclosures 2 and 3 contain the report for the proprietary and non-proprietary versions of the "GEl 4 Thermal Hydraulic Compatibility With Columbia Legacy Fuel."
Question 4:
The following statement in the application, 'The approach discussed in Reference 7.2 was adopted into the BWR/4 STS and resulted in simplifying the scram time LCO for plants utilizing the GNF methodology," seems to imply that this approach may not be applicable to plants not using the GNF methodology. Please clarify this statement.
Response
Energy Northwest did not mean to imply that vendors other than GNF are incapable of following the approach of the Standard Technical Specifications (STS) in regards to modeling scram timing. What Energy Northwest intended was that the standard approach of GESTAR-II is aligned to support scram timing as represented in the STS. Columbia
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 15 of 16 proposes to align with the STS such that no additional analysis will be required with the transition to GNF methodology to ensure that scram timing is properly modeled.
Question 5:
Please provide the actual scram time table used in the transient analysis.
Response
The actual scram time table used in the transient analysis is shown in the table below.
Columbia is implementing the Option B methodology as described in GESTAR II.
CRD Control TS Scram Time Option B Scram Fraction (Sec)
Time (Sec) 0 0.200 0.200 5
0.490 0.324 20 0.900 0.694 50 2.000 1.459 90 3.500 2.535 100 3.875 2.804 Question 6:
Please provide the CGS scram time performance data from past cycles to justify the proposed scram time table. Also, based on historical CGS data, how many control rods would typically be considered slow based on the proposed scram time table?
Response
On April 19, 2005, Energy Northwest requested a change to the frequency of control rod scram time testing via a license amendment request (ADAMS Accession No. ML051190270). In that request Energy Northwest identified that there had been 4958 scram time tests performed from the beginning of 1995 to April 2005 with no control rods being declared "slow" as a result of the scram time failing to meet the two-by-two average time Technical Specification (TS) or Core Operating Limits Report (COLR) acceptance criteria. The NRC acknowledged in the Safety Evaluation for Amendment 194, issued September 29, 2005, that "the control rod insertion (scram) time test results at Energy Northwest have shown the control rod scram rates to be highly reliable" (ADAMS Accession No. ML051600123).
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 16 of 16 From 1995 to the present, 6835 scram time tests have been performed at normal operating pressures. No failures or declaration of "slow" control rods were required as a result of the scram time failing to meet the current licensing basis requirements of two-by-two average time for TS or COLR acceptance criteria.
Application of the proposed scram time table to the above described Columbia historical data would have resulted in one control rod being declared "slow" from a scram time test conducted on December 3, 2006. This control rod was not declared "slow" at the time of the test due to the two-by-two averaging criteria being met. Hence, the declaration of a "slow" control rod would be an infrequent occurrence at Columbia, with only one such example in the past 13 years of operation encompassing 6835 scram time tests.
Comparing the results of the 6835 tests to the new TS criteria indicate that the average control rod scram times are significantly faster in that there are approximately nine standard deviations between the average times and the proposed TS criteria. The following table is a summary of the results of the 6835 control rod scram time tests.
Time in Notch Notch Notch Notch seconds Position 45 Position 39 Position 25 Position 05 Proposed TS scram time 0.528 0.866 1.917 3.437 criteria Average 0.312 0.610 1.322 2.424 time Standard Diation 0.021 0.029 0.059 0.108 Deviation III
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING Page 1 of I LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Energy Northwest in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
SCHEDULED TYPE COMPLETION (check one)
DATE COMMITMENT (if required)
ONE-TIME CONTINUING ACTION COMPLIANCE The requested information for anticipated transient without scram (ATWS), abnormal X
3/19/2009 operation occurrence (AOO), American Society of Mechanical Engineers (ASME)
Code overpressure, and stability analyses performed for the initial transition core is not yet available. This information will be provided by a separate letter when it is available. (The expected date of submittal of this additional information is March 19, 2009.)
I
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING LICENSE AMENDMENT REQUEST INVOLVING CORE OPERATING LIMITS REPORT AND SCRAM TIME TESTING AFFIDAVIT FOR THE REQUEST TO WITHHOLD INFORMATION Affidavit for the request to withhold information for GNF S-0000-0092-8136P, Revision 0, dated October 2008.
Global Nuclear Fuel - Americas AFFIDAVIT I, Anthony P. Reese, state as follows:
(1) I am Reload Licensing Manager, Fuel Engineering, Global Nuclear Fuel-Americas, LLC
("GNF-A"), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in the attachment, "GE14 Thermal Hydraulic Compatibility With Columbia Legacy Fuel" dated October 2008.
GNF proprietary information is identified by a dotted underline inside double square brackets. ((This sentence is an example. (3) In each case, the superscript notation
{3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.
(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), andPublic Citizen Health Research Group
- v. FDA, 704F2dl280 (DC Cir. 1983).
(4)
Some examples of categories of information which fit into the definition of proprietary information are:
- a.
Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over other companies;
- b.
Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c.
Information which reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A;
- d.
Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
Affidavit Page 1 of 3
The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.
(5)
To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A. Access to such documents within GNF-A is limited on a "need to know" basis.
(7)
The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.
Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The information identified in paragraph (2) is classified as proprietary because it contains details of GNF-A's fuel design and licensing methodology.
The development of the methods used in these analyses, along with the testing, development and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to GNF-A or its licensor.
Affidavit Page 2 of 3
(9)
Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities.
The information is part of GNF-A's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.
The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A.
The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.
GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.
I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.
Executed on this 30th day of October, 2008.
Antho'ry-l. Reese Reload Licensing Manager, Fuel Engineering Global Nuclear Fuel - Americas, LLC Affidavit Page 3 of 3