ML12076A062

Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, Core Operating Limits Report (Colr) and Revision to Technical Specification 2.1.1.2

ML12076A062
Person / Time
Site:	Brunswick
Issue date:	03/06/2012
From:	Annacone M J Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 12-0031, TSC-2011-01
Download: ML12076A062 (34)

Text

Letter Enclosures Contain Proprietary Information ithhold in Accordance with 10 CFR 2.390(a)(6)

Progress inergy Vice President Brunswick Nuclear Plant MAR 0'6 SERIAL: BSEP 12-0031 10 CFR 50.90 TSC-201 1-01 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Request for License Amendments

-Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)" and Revision to Technical Specification 2.1.1.2 Minimum Critical Power Ratio Safety Limit Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90, Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc..is requesting a revision to the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendments: (I) revise TS 5.6.5.b by replacing AREVA Topical Report ANF-524(P)(A), ANF Critical Power Methodology'jbr Boiling. Water Reactors with AREVA Topical Report ANP- I 0307PA.Revision 0, AREVA MCPR Satety Limit Methodologyjfbr Boiling Water Reacors., June 2011, in the list of analytical methods that have been reviewed and approved by the NRC for determining core operating limits, (2) revise TS 2.1. 1, "Reactor Core SLs," by incorporating revised Safety Limit Minimum Critical Power Ratio (SLMCPR) values, and (3) revise the license condition in Appendix B, "Additional Conditions," of the operating licenses regarding an alternate method for evaluating SLMCPR values. An evaluation of the proposed license amendments is provided in Enclosure 1.CP&L has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1), using the criteria in 10 CFR 50.92(c), and determined that this change involves no significant hazards considerations.

In accordance with 10 CFR 50.91(b), CP&L is providing a copy of the proposed license amendments to the designated representative for the State of North Carolina.CP&L requests approval of the proposed amendments by March 1, 2013, in order to support reactor start-up following the Unit 2 refueling outage, which is currently scheduled to begin in March 2013. Once approved, the Unit 2 amendment shall be implemented Progress Energy Carolinas, Inc.P.O. Boa 104298 Southport, NC 28461 T > 910.457.3698 Document Control Desk BSEP 12-0031 / Page 2 prior to start-up from the 2013 Unit 2 refueling outage and the Unit 1 amendment shall be implemented prior to start-up from the 2014 Unit 1 refueling outage.No regulatory commitments are contained in this submittal.

Please refer any questions regarding this submittal to Mr. Lee Grzeck, Acting Supervisor

-Licensing/Regulatory Programs, at (910) 457-2487.I declare, tinder penalty of perjury. that the foregoing is true and correct. Executed on March 6,2012.Sincerely, ichael J. Annacone WRTM/wrm

Enclosures:

I. Evaluation of Proposed License Amendment Request.Marked-up Technical Specification and Operating License Pages -Unit 2 3. Typed Technical Specification Pages -Unit 1 4. Typed Technical Specification Pages -Unit 2 5. Marked-up Technical Specification Bases Pages -Unit 2 (For information only)6. AREVA Document No. 51-9175814-000, "Brunswick Unit I Cycle 19 SLMCPR Analysis With SAFLIM3D Methodology (Proprietary Version)" (Proprietary Information

-Withhold from Public Disclosure in Accordance With 10 CFR 2.390)7. AREVA Affidavit Regarding Withholding AREVA Document No. 51-9175814-000, "Brunswick Unit I Cycle 19 SLMCPR Analysis With SAFLIM3D Methodology (Proprietary Version)" from Public Disclosure

8. AREVA Document No. 51-9177317-000, "Brunswick Unit 1 Cycle 19 SLMCPR Analysis With SAFLIM3D Methodology (Nonproprietary Version)" 9. AREVA Document No. 51-9176407-000, "Brunswick Unit I Cycle 19 SLMCPR Analysis With SAFLIM3D Methodology

-Operability Assessment (Proprietary Version)" (Proprietary Information

-Withhold from Public Disclosure in Accordance With 10 CFR 2.390)10. AREVA Affidavit Regarding Withholding AREVA Document No. 51-9176407-000, "Brunswick Unit 1 Cycle 19 SLMCPR Analysis With SAFLIM3D Methodology

-Operability Assessment (Proprietary Version)" from Public Disclosure

11. AREVA Document No. 51-9177315-000, "Brunswick Unit I Cycle 19 SLMCPR Analysis With SAFLIM3D Methodology

-Operability Assessment (Nonproprietary Version)"

Document Control Desk BSEP 12-0031 / Page 3 12. AREVA Document No. 51-9175787-000, "Brunswick Unit 2 Cycle 20 SLMCPR Analysis With SAFLIM3D Methodology (Proprietary Version)" (Proprietary Information

-Withhold from Public Disclosure in Accordance With 10 CFR 2.390)13. AREVA Affidavit Regarding Withholding AREVA Document No. 51-9175787-000, "Brunswick Unit 2 Cycle 20 SLMCPR Analysis With SAFLIM3D Methodology (Proprietary Version)" from Public Disclosure

14. AREVA Document No. 51-9177314-000, "Brunswick Unit 2 Cycle 20 SLMCPR Analysis With SAFLIM3D Methodology (Nonproprietary Version)" 15. AREVA Document No. 51-9176342-000, "Brunswick Unit 2 Cycle 20 SLMCPR Analysis With SAFLIM3D Methodology

-Operability Assessment (Proprietary Version)" (Proprietary Information

-Withhold from Public Disclosure in Accordance With 10 CFR 2.390)16. AREVA Affidavit Regarding Withholding AREVA Document No. 51-9176342-000, "Brunswick Unit 2 Cycle 20 SLMCPR Analysis With SAFLIM3D Methodology

-Operability Assessment (Proprietary Version)" fiom Public Disclosure

17. AREVA Document No. 51-9177316-000, "Brunswick Unit 2 Cycle 20 SLMCPR Analysis With SAFLIM3D Methodology

-Operability Assessment (Nonproprietary Version)" 18. AREVA Document ANP-3086(P), Revision 0, "Brunswick Unit I and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for ATRIUM 1OXM Fuel -Improved K-factor Model" (Proprietary Information

-Withhold from Public Disclosure in Accordance With 10 CFR 2.390)19. AREVA Affidavit Regarding Withholding AREVA Document ANP-3086(P), Revision 0, "Brunswick Unit 1 and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for ATRIUM I OXM Fuel -Improved K-factor Model" from Public Disclosure

20. AREVA Document ANP-3086(NP), Revision 0, "Brunswick Unit I and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for ATRIUM IOXM Fuel -Improved K-factor Model" (Non-Proprietary Version)

Document Control Desk BSEP 12-0031 / Page 4 cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Victor M. McCree, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U. S. Nuclear Regulatory Commission ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A)11555 Rockville Pike Rockville.

MD 20852-2738 cc (with Enclosures 1 through 5 only): Chair -North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-05 10 Mr. W. Lee Cox, ll, Section Chief Radiation Protection Section North Carolina Department of Environment and Natural Resources 1645 Mail Service Center Raleigh, NC 27699-1645 BSEP 12-0031 Enclosure 1 Page 1 of 13 Evaluation of Proposed License Amendment Request

Subject:

Request for License Amendments

-Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)" and Revision to Technical Specification 2.1.1.2 Minimum Critical Power Ratio Safety Limit 1.0 Summary Description This letter is a request by Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., to amend Appendix A, Technical Specifications (TS), of Renewed Facility Operating License Nos. DPR-71 and DPR-62 for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.The proposed changes: (1) revise TS 5.6.5.b by replacing AREVA Topical Report ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors with AREVA Topical Report ANP-10307PA, Revision 0, AREVA MCPR SafI/y Limit Methodology;for Boiling Kater Reactors, June 2011, in the list of analytical methods that have been reviewed and approved by the NRC for determining core operating limits, (2) revise TS 2. 1. 1, "Reactor Core SLs," by incorporating revised Safety Limit Minimum Critical Power Ratio (SLMCPR) values, and (3) revise the license condition in Appendix B, "Additional Conditions," of the operating licenses regarding an alternate method for evaluating SLMCPR values. These changes are needed to support the next cycles of operation for BSEP Units I and 2 (i.e., Cycle 20 for BSEP, Unit 1, which is scheduled to begin April 2014 and Cycle 21 for BSER Unit 2, which is scheduled to begin March 2013).2.0 Detailed Description Proposed Change 1: TS 5.6.5.b identifies the analytical methods that should be used to determine core operating limits. The current TS states: 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:

BSEP 12-0031 Enclosure I Page 2 of 13 Currently, the SLMCPR protected by the reactor core critical power limits is determined using the NRC-approved analytical methodology described in the following AREVA Topical Report listed in TS Section 5.6.5.b: ANF-524(P)(A), ANF Critical Power Methodologoyfor Boiling Water Reactors (i.e., Item 11 in TS Section 5.6.5.b)The proposed amendments will replace ANF-524(P

)(A) with AREVA Topical Report ANP- I 0307PA, Revision 0, AREVA MCPR Safety Limit MethodologyJbor Boiling Water Reactors, June 2011, in the list of analytical methodologies in TS Section 5.6.5.b that may be used for determining core operating limits which have been reviewed and approved by the NRC. Topical Report ANP- 1030 7PA describes an improved AREVA methodology for determining SLMCPR and incorporates a realistic fuel channel bow model. By letter dated June 14, 2011 (i.e., Reference 1), Topical Report ANP-10307PA, Revision 0, has been approved by the NRC and found acceptable for referencing ill licensing applications for boiling water reactors.Proposed Change 2: Using the analytical methods described in Topical Report ANP-10307PA, Revision 0, the two recirculation loop operation (TLO) and single recirculation loop operation (SLO)safety limit minimum critical power ratio (SLMCPR) values in TS .1. 1.2 are also being revised.TS 2.1.1.2 specifies the values for the SLMCPR. The current BSEP Unit I TS states: MCPR shall be > 1.11 for two recirculation loop operation or > 1.12 for single recirculation loop operation.

and the current BSEP Unit 2 TS states: MCPR shall be > 1.11 for two recirculation loop operation or > 1 .1 3 for single recirculation loop operation.

The proposed amendments will revise the SLMCPR values in TS 2.1.1.2 for two loop operation and single loop operation.

The SLMCPR value for two recirculation loop operation is being changed from > 1.11 (i.e., for both Unit 1 and Unit 2) to > 1.08, and from > 1.12 (i.e., for Unit 1) and > 1.13 (i.e., for Unit 2) to > 1.11 (i.e.. for both Unit I and Unit 2) for SLO.Proposed Change 3: The Facility Operating License, Appendix B, "Additional Conditions," includes a license condition which requires the performance of a confirmatory evaluation for SLMCPR, BSEP 12-0031 Enclosure I Page 3 of 13 setpoint, and core operating limit values that have been determined using AREVA Topical Report ANP-10298PA, "ACE/ATRIUM 1OXM Critical Power Correlation." The license condition currently states: Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and core operating limit values determined using the ANP- I 0298PA, ACE/ATRIUM 1 OXM Critical Power Correlation (i.e., TS 5.6.5.b.21), shall be evaluated with methods described in AREVA Operability Assessment CR 2011-2274, Revision 1 to verify the values determined using the NRC-approved method remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. The results of the evaluation shall be documented and submitted to the NRC, for review, at least 60 days prior to startup of each operating cycle.The proposed amendments will revise this license condition by requiring the evaluation of SLMCPR with the methods described in AREVA Document ANP-3086(P), Revision 0, Brunswick Unit 1 and Unit 2 SLMCPR Operability Assessment Critical Power Correlation

/br ATRIUM'I ]OXM Fuel -Inproved K-faIctor Model, provided in Enclosure

18. The revised license condition would read as follows: Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and core operating limit values determined using the ANP-10298PA, ACE/ATRIUM IOXM Critical Power Correlation (i.e., TS 5.6.5.b.21), shall be evaluated to verify the values determined using the NRC-approved method remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. SLMCPR shall be evaluated with methods described in AREVA Document ANP-3086(P), Revision 0, "Brunswick Unit 1 and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for ATRIUM 1OXM Fuel -Improved K-factor Model." Setpoint and core operating limit values shall be evaluated with methods described in AREVA Operability Assessment CR 2011-2274, Revision 1. The results of the evaluation shall be documented and submitted to the NRC, for review, at least 60 days prior to startup of each operating cycle.The license condition will retain the existing requirement to perform the evaluation of setpoint and core operating limit values using the methods described in AREVA Operability Assessment CR 2011-2274, Revision 1.Enclosure 2 contains marked-up TS and Operating License pages for BSEP Unit 2 indicating the proposed changes. Since TS 2.1.1.2, TS 5.6.5.b and the license condition change associated with this request are equivalent for BSEP Units I and 2, only mark-ups BSEP 12-0031 Enclosure 1 Page 4 of 13 for Unit 2 are provided.

Enclosures 3 and 4 provide typed versions of the Unit I and Unit 2 TS, respectively.

These typed TS pages are to be used for issuance of the proposed amendments.

In addition, in support of the proposed TS changes, TS Bases Section 2.1.1 will be revised, after issuance of the amendments, to reflect application of AREVA Topical Report ANP-10307PA, Revision 0. Mark-ups of the TS Bases changes for Unit') are provided in Enclosure 5, for information only, and do not require NRC approval.3.0 Technical Evaluation In general, methodologies or computer codes used to support licensing basis analyses are documented in topical reports which are reviewed by the NRC on a generic basis. The NRC, in its safety evaluation for the approved topical report, defines the basis for acceptance in conjunction with any limitations and conditions on use of the topical report, as appropriate.

In situations where a plant-specific license amendment request references a generic topical report, plant-specific applicability of the material presented in the topical report is reviewed.System Description/Applicable SafeOt Analysis'The BSEP, Unit I and 2 cores consist primarily of ATRIUM-10 and ATRIUM 1OXM fuel assemblies, along with a small number of GEl 4 assemblies.

Beginning with Unit 1 Cycle 19 in Spring 2012, the Unit I core will no longer use any co-resident non-AREVA fuel. Beginning with Unit 2 Cycle 21 in Spring 2013, the Unit 2 core will also no longer use any co-resident non-AREVA fuel.Beginning with BSEP Unit 2 Cycle 21, CP&L intends to use the analytical methodology described in Topical Report ANP-1 0307PA for determining the SLMCPR values for two loop and single loop operation.

Topical Report ANP-10307PA describes an improved AREVA critical power methodology.

NRC approval of Topical Report ANP-10307P is documented in a safety evaluation issued by letter dated June 14, 2011 (i.e., Reference 1).Approval of this license amendment request and the incorporation of Topical Report ANP-10307PA will enable CP&L to implement the analytical methods described in the report. SLMCPR is currently determined using the NRC-approved analytical methodology described in AREVA Topical Report ANF-524(P)(A), ANF Critical Power Methodology/or Boiling Water Reactors (i.e., Item 11 in TS 5.6.5.b).

This topical report is listed in Brunswick Technical Specification 5.6.5.b as an analytical methodology that may be used to determine core operating limits.

BSEP 12-0031 Enclosure I Page 5 of 13 Plant Specilc Methodology Applicability Evaluation The methodology described in ANP- 1 0307PA, Revision 0, is used to determine the SLMCPR such that at least 99.9 percent of the fuel rods in the core will not experience dry-out during normal operation and anticipated operational occurrences if the core MCPR is greater than or equal to the SLMCPR. Enclosures 6 and 12 summarize the methodology, inputs, and results supporting the BSEP Unit 1 Cycle 19 and Unit 2 Cycle 20 SLMCPR values calculated using the ANP-l 0307PA methodology, respectively.

The SLMCPR is determined using a statistical analysis that employs a Monte Carlo process that perturbs key input parameters used in the MCPR calculation based on their uncertainties.

Table I in Enclosures 6 and 12 identifies these uncertainty inputs.The fuel-related power distribution uncertainty inputs used by the ANP-10307PA methodology are calculated from separately determined uncertainty components described in EMF-2 1 58(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and [alidation o/ CASMO-4/MICROBURN-B2.

The NRC-approved uncertainty components friom EMF-2158(P)(A) are shown to be applicable to BSEP Unit 1 and Unit 2 in Enclosure 3 of Reference

3. Plant related uncertainty inputs (i.e., feedwater flow rate, feedwater temperature, core pressure and total core flow uncertainties) are identified in Table I of Enclosures 6 and 12, and are consistent with the NRC-approved plant uncertainties reported from Topical Report NEDC-32601 P-A, Methodoloqy and Uncertainties.for Safety Limit AICPR Evaluations.

ANP-10307PA incorporates a realistic fuel channel bow model. Model uncertainty is based on AREVA fuel channel measurements.

Channel fluence is an input to the channel bow model, and is calculated based on BSEP-specific reactor core operating conditions.

AREVA fuel loaded in the BSEP Unit I and 2 reactor cores is channeled with AREVA fuel channels made of Zircaloy-4 material.

BSEP has not experienced indications of abnormal channel bow with AREVA channels.

Should indications of abnormal channel bow be experienced by BSEP, the channel bow model will be applied in a conservative manner as described in ANP-10307PA.

The BSEP Unit 2 Cycle 20 core includes a small number of third-cycle GEI 4 fuel assemblies near the core periphery that have substantial critical power margin. The channel bow model uncertainty applied to these GEl4 fuel assemblies was conservatively increased based on GE 14 specific channel bow uncertainty provided by Global Nuclear Fuel, consistent with the ANP-1 0307PA methodology.

The ANP-10307PA channel bow model is applicable to BSEP, because the model accounts for channel fluence calculated specific to the BSEP reactor core operating conditions.

The channel bow model uncertainty is applicable to BSEP., because BSEP has not experienced abnormal channel bow with AREVA fuel channels and channel bow uncertainty specific to the remaining GEI 4 fuel channels is applied.

BSEP 12-003 1 Enclosure I Page 6 of 13 License Condition Evaluation In conjunction with the issuance of License Amendment Nos. 257 and 285 to the BSEP Unit I and 2 Operating Licenses (i.e., Reference 4), the NRC included a license condition in Appendix B, "Additional Conditions," of the operating licenses.

This license condition is intended to ensure the limits generated with the NRC-approved methods appropriately bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. The license condition currently states: Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and core operating limit values determined using the ANP- I 0298PA, ACE/ATRIUM 1 OXM Critical Power Correlation (i.e., TS 5.6.5.b.2 1), shall be evaluated with methods described in AREVA Operability Assessment CR 2011-2274, Revision 1 to verify the values determined using the NRC-approved method remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. The results of the evaluation shall be documented and submitted to the NRC, for review, at least 60 days prior to startup of each operating cycle.The methods described in AREVA Operability Assessment CR 2011-2274, Revision I to assess SLMCPR values determined using the ANF-524(P)(A) methodology may not always remain appropriate to assess SLMCPR values determined using the ANP- 10307PA methodology.

ANP-3086(P), Revision 0, Brunswick Unit ] and Unit 2 SLMCPR Operability Assessment Critical Power Correlation jbr ATRIUM I OXM Fuel -Improved K-/actor Model., describes a BSEP-specific methodology to verify SLMCPR values determined using the methods described in ANP- 100307PA are applicable and include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. ANP-3086(P), Revision 0, is provided in Enclosure

18. Evaluations performed consistent with the methods described in ANP-3086(P), Revision 0, have been performed and are provided in Enclosures 9 and 15. Based on these evaluations, it has been concluded that the SLMCPR results determined using the methods described in ANP-10307PA, and provided in Enclosure 6 and 12, are applicable and include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1.The proposed amendments will revise the current license condition by requiring the evaluation of SLMCPR with the methods described in AREVA Document No. ANP-3086(P), Revision 0. The license condition will retain the existing requirement to perform the evaluation of setpoint and core operating limit values using the methods described in AREVA Operability Assessment CR 2011-2274, Revision 1.

BSEP 12-0031 Enclosure I Page 7 of 13 Confobrmance with Methodologv and Saety Evaluation Limitations The NRC letter approving Topical Report ANP-10307PA states that the report is acceptable for referencing in licensing applications for boiling water reactors to the extent specified and tinder the limitations delineated in the topical report and in the final NRC safety evaluation.

The final NRC safety evaluation concluded the topical report is acceptable for use for plant-specific licensing actions, without listing any limitations or conditions.

Upon approval of this license amendment application and incorporation of Topical Report ANP- I 0307PA, Revision 0, into the BSEP Unit 1 and 2 Technical Specifications, CP&L will implement the analytical methods described in the report and will conform with the methodology described in the topical report.The SLMCPR results determined using the ANP-1 0307PA methodology provided in Enclosures 6 and 12 support two loop operation SLMCPR values of 1.07 and 1.06, and single loop operation SLMCPR values of 1.09 and 1.08, for BSEP Unit I Cycle 19 and BSEP Unit 2 Cycle 20, respectively.

These results support the requested TS SLMCPR for BSEP Units I and 2 of 1.08 for two loop operation and 1.11 for single loop operation.

More conservative values than supported by the results in Enclosures 6 and 12 are being requested to accommodate small cycle-to-cycle variations.

No plant hardware or operational changes are required with the proposed license amendments.

4.0 Regulatory

Evaluation

4.1 Applicable

Regulatory Requirements/Criteria 10 CFR 50.36, "Technical specifications," paragraph (c)(1), requires that power reactor facility TS include safety limits for process variables that protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.

The fuel cladding integrity SLMCPR is established to assure that at least 99.9% of the fuel rods in the core do not experience boiling transition during normal operation and anticipated operational occurrences (AOOs). Thus, the SLMCPR is required to be contained in TS.The proposed amendments to the BSEP, Unit 1 and 2 TS do not remove the SLMCPR from the TS.10 CFR 50, Appendix A, General Design Criterion (GDC) 10 requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of AQOs. To ensure compliance with GDC 10, CP&L has performed the plant-specific SLMCPR analyses using BSEP 12-0031 Enclosure 1 Page 8 of 13 NRC-approved methodologies as prescribed in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 4.4. The SLMCPR ensures that sufficient conservatism exists in the operating limit MCPR such that, in the event of an AOO, there is a reasonable expectation that at least 99.9 percent of the fuel rods in the core will avoid boiling transition for the power distribution within the core including uncertainties.

4.2 Precedent

By letter dated June 14, 2011 (i.e., Reference 1), Topical Report ANP-10307PA, Revision 0, has been approved by the NRC. The NRC letter approving Topical Report ANP- 1 0307PA states that the report is acceptable for referencing in licensing applications for boiling water reactors to the extent specified and under the limitations delineated in the topical report and in the final NRC safety evaluation.

The final NRC safety evaluation concluded the topical report is acceptable for use for plant-specific licensing actions, without listing any limitations or conditions.

Since the proposed amendments for BSEP, Units I and 2 are the first licensee request to use the topical report, no precedent exists for this licensing action.5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration The proposed license amendments involve three related activities.

First, TS 5.6.5.b is being revised to replace AREVA Topical Report ANF-524(P)(A), ANF Critical Power Aiethodology/br Boiling Water Reactors with AREVA Topical Report ANP- I 0307PA, Revision 0, AREVI MCPR Sa/ety Limit Methodology/br Boiling Water Reactors, June 2011, in the list of analytical methods that have been reviewed and approved by the NRC for determining core operating limits. Second, the Safety Limit Minimum Critical Power Ratio (SLMCPR) values contained in Technical Specification (TS) 2.1.1.2 are being revised for two recirculation loop operation from > 1.11 (i.e., for Units I and 2) to > 1.08, and firom > 1.12 (i.e., for Unit 1) and > 1.13 (i.e., for Unit 2) to > 1.11 (i.e., for both Unit I and Unit 2) for single loop recirculation operation.

Finally, a license condition in Appendix B, "Additional Conditions" of the operating licenses regarding evaluating SLMCPR values is being updated to ensure SLMCPR values determined using the methodology described in AREVA Topical Report ANP- 10307PA are evaluated with an alternate method to AREVA Operability Assessment CR 2011-2274, Revision 1, which is appropriate for verifying SLMCPR values detennined using the ANP-1 0307PA methodology remain applicable, while maintaining the existing methodology used to evaluate setpoint and core operating limit values.

BSEP 12-003 1 Enclosure I Page 9 of 13 CP&L has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92,"Issuance of amendment.," as discussed below: I. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident.

The proposed license amendments do not involve any plant modifications or operational changes that could affect system reliability or performance, or that could affect the probability of operator error. As such, the proposed changes do not affect any postulated accident precursors.

Since no individual precursors of an accident are affected, the proposed license amendments do not involve a significant increase in the probability of a previously analyzed event.The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences.

The basis for the SLMCPR calculation is to ensure that during normal operation and during anticipated operational occurrences, at least 99.9 percent of all fuel rods in the core do not experience transition boiling if the safety limit is not exceeded.The proposed SLMCPR values have been determined using NRC-approved methods discussed in AREVA Topical Report ANP-10307PA, Revision 0, AREM4 A'JCPR Sqaety Limit Methodology for Boiling Water Reactors, June 2011. To support use of Topical Report ANP-10307PA, Revision 0, by BSEP, Units I and 2, this NRC-approved analytical method is being added to the list of NRC-approved analytical methods identified in Technical Specification 5.6.5.b. Replacing AREVA Topical Report ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors with the analytical methods described in Topical Report ANP-10307PA in Technical Specification 5.6.5.b does not alter the assumptions of accident analyses.

Furthermore, establishing a two recirculation loop SLMCPR value of> 1.08 and a single recirculation loop SLMCPR value of> 1.11 ensures that the acceptance criteria continues to be met (i.e., at least 99.9 percent of all fuel rods in the core do not experience transition boiling), while the revised license condition ensures that SLMCPR, setpoint, and core operating limit values determined using the NRC-approved AREVA methodologies remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. Based on these considerations, the BSEP 12-00331 Enclosure I Page 10of 13 proposed changes do not involve a significant increase in the consequences of a previously analyzed accident.2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors.

New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation.

The SLMCPR is a TS numerical value calculated for two recirculation loop operation and single recirculation loop operation to ensure at least 99.9 percent of all fuel rods in the core do not experience transition boiling if the safety limit is not exceeded.

SLMCPR values are calculated using NRC-approved methodology identified in the TS. The proposed SLMCPR values and the AREVA methodology being added to TS do not involve any new modes of plant operation or any plant modifications and do not directly or indirectly affect the failure modes of any plant systems or components.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?Response:

No The SLMCPR provides a margin of safety by ensuring that at least 99.9 percent of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences if the MCPR Safety Limit is not exceeded.Replacing the analytical methodology described in Topical Report ANF-524(P)(A) with the methodology described in Topical Report ANP-10307PA in the list of NRC-approved analytical methods identified in Technical Specification 5.6.5.b, revision of the SLMCPR values in Technical Specification 2.1.1.2 using NRC-approved methodology, and confirmation that the SLMCPR,. setpoint, and core operating limit values remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1, will ensure that the current level of fuel protection is maintained by continuing to ensure that the fuel design safety criterion is met (i.e., that no more than 0.1 percent of the rods are expected to be in boiling transition if the MCPR Safety Limit is not exceeded).

BSEP 12-0031 Enclosure I Page 11 of 13 Meeting the fuel design criterion that at least 99.9 percent of all fuel rods in the core do not experience transition boiling and establishing core operating limits based on the proposed SLMCPR values, to ensure that the SLMCPR is not exceeded, ensures the margin of safety required by the fuel design criterion is maintained.

Therefore, the proposed amendments do not result in a significant reduction in the margin of safety.Based on the above, CP&L concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable

Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. CP&L has determined that the proposed changes do not require any exemptions or relief fronm regulatory requirements, other than the Technical Specifications, and do not affect conformance with any General Design Criterion (GDC) differently than described in the Updated Final Safety Analysis Report (UFSAR).As stated in the NRC's "Safety Evaluation of the Brunswick Steam Electric Station Units I and 2," dated November 1973, BSEP meets the intent of the General Design Criteria (GDC), published in the Federal Register on May 21, 1971, as Appendix A to 10 CFR Part 50. The proposed changes do not affect compliance with the intent of the GDCs. In particular, the intent of GDC 10, "Reactor design," continues to be met.GDC 10 states: The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

To ensure compliance with GDC 10, CP&L performs plant-specific critical power limit analyses using NRC-approved methodologies.

The MCPR Safety Limit ensures that sufficient conservatism exists in the operating limit MCPR such that, in the event of an anticipated operational occurrence, there is a reasonable expectation that at least 99.9 percent of the fuel rods in the core will avoid boiling transition for the power distribution within the core including uncertainties.

10 CFR 50.36(c)(5) states that the Technical Specifications will include administrative controls that address the provisions relating to organization and management, procedures, record keeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. The Core Operating Limits Report (COLR) is required as a part BSEP 12-003 1 Enclosure I Page 12 of 13 of the reporting requirements specified in the Brunswick Technical Specifications Administrative Controls section. The Technical Specifications require the core operating limits to be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and to be documented in the COLR. In addition, it requires the analytical methods used to determine the core operating limits to be those that have been previously reviewed and approved by the NRC, and specifically to be those described in Teclmical Specification 5.6.5.b. The proposed amendments ensure that these requirements are met.In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.6.0 Environmental Considerations A review has determined that the proposed amendments are administrative in nature and do not change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, Standards for Protection Against Radiation, and do not change an inspection or surveillance requirement.

The proposed amendments do not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.

7.0 References

I: Letter from Robert A. Nelson (NRC) to Pedro Salas (AREVA NP Inc.), "Final Safety Evaluation for AREVA NP, Inc. Topical Report ANP-10307P, Revision 0,'AREVA MCPR [Minimum Critical Power Ratio] Safety Limit Methodology for Boiling Water Reactors' (TAC No. ME2914)," dated June 14, 2011, ADAMS Accession Number ML 1 140A 125.2. ANP-3086(P), Revision 0, Brunswick Unit I and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for ATRIUM 1 OXM Fuel -Improved K-ftictor Model, February 2012.

BSEP 12-0031 Enclosure I Page 13 of 13 3. Letter from William Jefferson, Jr. (CP&L) to U.S. Nuclear Regulatory Commission Document Control Desk, "Response to Additional Information Request Supporting License Amendment Requests for Addition of Analytical Methodology Topical Reports to Technical Specification 5.6.5 (NRC TAC Nos. ME3856, ME3857, ME3858, and ME3859)," dated November 18, 2010, ADAMS Accession Number ML 103330242.

4. Letter from Farideh E. Saba (USNRC) to Michael J. Annacone (CP&L),"Issuance of Amendments Regarding Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5 (TAC Nos. ME3856 and ME3857)," dated April 8, 2011, ADAMS Accession Number MLI 11010234.

BSEP 12-0031 Enclosure 2 Marked-up Technical Specification and Operating License Pages -Unit 2 SLs 2.0 2.0 SAFETY LIMITS (SLS)2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%rated core flow: THERMAL POWER shall be _ 23% RTP.2.1.1.2 With the reactor steam dome pressure _> 785 psig and core flow L 10%rated core flow: 1.08 MCPR shall be _> 444 for two recirculation loop operation or 4-4 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be _< 1325 psig.2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.Brunswick Unit 2 2.0-1 Revision No. 254 1 Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. XN-NF-80-19(P)(A)

Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis.7. XN-NF-80-19(P)(A)

Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads.8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2.

9. XN-NF-80-19(P)(A)

Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description.

ANP-10307PA, AREVA MCPR 10. XN-NF-84-105(P)(A)

Volume 1, XCOBRA-T:

A Computer Code Safety Limit for BWR Transient Thermal-Hydraulic Core Analysis.Methodology for 11. AN^ 52.(P)(A), AN, Critical Pow..c egy for Boiling W.tr Boiling Water ReaetS.Reactors, Revision 0, June 2011 12. ANF-913(P)(A)

Volume 1, COTRANSA2:

A Computer Program for Boiling Water Reactor Transient Analyses.13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.17. EMF-2292(P)(A), ATRIUMTM-10:

Appendix K Spray Heat Transfer Coefficients.

18. EMF-CC-074(P)(A)

Volume 4, BWR Stability Analysis -Assessment of STAIF with Input from MICROBURN-B2.

19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications.(continued)

Brunswick Unit 2 5.0-21 Amendment No. 274 I Amendment Number Additional Conditions 276 Upon implementation of Amendment No. 276 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.3.3, in accordance with TS 5.5.13.c.(i), the assessment of CRE habitability as required by Specification 5.5.13.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.13.d, shall be considered met.Following implementation: (a) The first performance of SR 3.7.3.3, in accordance with Specification 5.5.13.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from June 11, 2004, the date of the most recent successful tracer gas test.(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.13.c.(ii), shall be within the next 9 months.(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.13.d, shall be within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test.Implementation Date As described in paragraphs (a), (b), and (c) of this Additional Condition.

SLMCPR shall be evaluated with methods described in AREVA document ANP-3086(P), Revision 0, Brunswick Unit 1 and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for ATRIUM 1OXM Fuel -Improved K-factor Model.Setpoint and core operating limit values shall be evaluated with methods described in AREVA Operability Assessment CR 2011-2274, Revision 1.Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and core operating limit 7c/ues determined using the ANP-10298PA, Ck ATRIUM 1OXM Critical Power Correlation iTS 5.6.5.b.21), shall be evaluated we#1 Upon implementation of Amendment No. 2-85.2911 Re iiR "4"to verifyl he val es determined using the NRC-approved method emain applicable and the core operating limits inc de margin sufficient to bound the effects of the K-f ctor calculation issue described in AREVA O rability Assessment CR 2011-2274, Revision 1. The results of the evaluation shall be documented and submitted to the NRC, for review, at least 60 days prior to startup of each operating cycle.Brunswick Unit 2 App. B-2 Amendment No. 285 1 BSEP 12-0031 Enclosure 3 Typed Technical Specification Pages -Unit I SLs 2.0 2.0 SAFETY LIMITS (SLS)2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%rated core flow: THERMAL POWER shall be 23% RTP.2.1.1.2 With the reactor steam dome pressure _> 785 psig and core flow _> 10%rated core flow: MCPR shall be ! 1.08 for two recirculation loop operation or 1.11 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be <_ 1325 psig.I 2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.Brunswick Unit 1 2.0-1 Amendment No. I Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. XN-NF-80-19(P)(A)

Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis.7. XN-NF-80-19(P)(A)

Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads.8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2.

9. XN-NF-80-19(P)(A)

Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description.

10. XN-NF-84-105(P)(A)

Volume 1, XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis.11. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0, June 2011.12. ANF-913(P)(A)

Volume 1, COTRANSA2:

A Computer Program for Boiling Water Reactor Transient Analyses.13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.17. EMF-2292(P)(A), ATRIUM T M-10: Appendix K Spray Heat Transfer Coefficients.

18. EMF-CC-074(P)(A)

Volume 4, BWR Stability Analysis -Assessment of STAIF with Input from MICROBURN-B2.

19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications.(continued)

Brunswick Unit 1 5.0-22 Amendment No. I BSEP 12-0031 Enclosure 4 Typed Technical Specification Pages -Unit 2 Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. XN-NF-80-19(P)(A)

Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis.7. XN-NF-80-19(P)(A)

Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads.8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2.

9. XN-NF-80-19(P)(A)

Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description.

10. XN-NF-84-105(P)(A)

Volume 1, XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis.11. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0, June 2011.12. ANF-913(P)(A)

Volume 1, COTRANSA2:

A Computer Program for Boiling Water Reactor Transient Analyses.13. ANF-1 358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model.17. EMF-2292(P)(A), ATRIUM T M-10: Appendix K Spray Heat Transfer Coefficients.

18. EMF-CC-074(P)(A)

Volume 4, BWR Stability Analysis -Assessment of STAIF with Input from MICROBURN-B2.

19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications.(continued)

Brunswick Unit 2 5.0-21 Amendment No. I SLs 2.0 2.0 SAFETY LIMITS (SLS)2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%rated core flow: THERMAL POWER shall be < 23% RTP.2.1.1.2 With the reactor steam dome pressure > 785 psig and core flow _> 10%rated core flow: MCPR shall be > 1.08 for two recirculation loop operation or _> 1.11 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be < 1325 psig.I 2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.Brunswick Unit 2 2.0-1 Revision No. I BSEP 12-0031 Enclosure 5 Marked-up Technical Specification Bases Pages -Unit 2 (For information only)

MCPR B 3.2.2 BASES REFERENCES ANP-10307PA,"AREVA MCPR Safety Limit Methodology for Boiling Water Reactors"-I--'1. UFSAR Section 4.4.2.1.2. AN^ 521 (P)(A), "ANP Critical Methodology f)o Beoling Watcr Rczictor5." 3. UFSAR, Chapter 4.4. UFSAR, Chapter 6.5. UFSAR, Chapter 15.I 6. (Deleted.)

7. XN-NF-80-19(P)(A)

Volume 3, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," (as identified in the COLR).8. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).9. ANF-913(P)(A)

Volume 1, "COTRANSA2:

A Computer Program for Boiling Water Reactor Transient Analyses," (as identified in the COLR).10. XN-NF-84-105(P)(A), "XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," (as identified in the COLR).11. 10 CFR 50.36(c)(2)(ii).

Brunswick Unit 2 B 3.2.2-4 Revision No. 62 I Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs.SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria.

SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CFR 50.67, "Accident Source Term," limits (Ref. 5). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.I REFERENCES

1. EMF-2209(P)(A), "SPCB Critical Power Correlation," (as identified in the COLR).ANP-10307PA,"AREVA MCPR Safety Limit Methodology for Boiling Water Reactors" 2. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," (as identified in the COLR).3. ANF 524(P)(A), "ANF Critical PcReW8 Methodelog, for Bo~iling WAtcr Dcactors," (as identified in the COLR).4. ANP-10298PA, "ACE/ATRIUM 1OXM Critical Power Correlation," (as identified in the COLR).5. 10 CFR 50.67.Brunswick Unit 2 B 2.1.1-4 Revision No. 73 1 BSEP 12-0031 Enclosure 7 AREVA Affidavit Regarding Withholding AREVA Document No. 51-9175814-000,"Brunswick Unit I Cycle 19 SLMCPR Analysis With SAFLIM3D Methodology (Proprietary Version)" firom Public Disclosure AFFIDAVIT STATE OF WASHINGTON

)) ss.COUNTY OF BENTON )1. My name is Alan B. Meginnis.

I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary.

I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.3. I am familiar with the AREVA NP information contained in the report 51-9175814-000, "Brunswick Unit 1 Cycle 19 SLMCPR Analysis With SAFLIM3D Methodology (Proprietary Version)," dated February 2012 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA NP's research and development plans and programs or their results.(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.A /SUBSCRIBED before me this ao-.day of 2012.Susan K. McCoy NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 1/14/2016