Letter Enclosures Contain Proprietary Information Progress ... · Letter Enclosures Contain Proprietary Information Progress inergy ithhold in Accordance with 10 CFR 2.390(a)(6) Vice

Letter Enclosures Contain Proprietary Informationithhold 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 50.90TSC-201 1-01

U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001

Subject: Brunswick Steam Electric Plant, Unit Nos. 1 and 2Renewed Facility Operating License Nos. DPR-71 and DPR-62Docket Nos. 50-325 and 50-324Request for License Amendments - Addition of Analytical MethodologyTopical Report to Technical Specification 5.6.5, "CORE OPERATINGLIMITS REPORT (COLR)" and Revision to Technical Specification 2.1.1.2Minimum Critical Power Ratio Safety Limit

Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90, CarolinaPower & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc..is requesting a revision to the Technical Specifications (TS) for the Brunswick SteamElectric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendments: (I) reviseTS 5.6.5.b by replacing AREVA Topical Report ANF-524(P)(A), ANF Critical PowerMethodology'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 theNRC for determining core operating limits, (2) revise TS 2.1. 1, "Reactor Core SLs," byincorporating revised Safety Limit Minimum Critical Power Ratio (SLMCPR) values, and(3) revise the license condition in Appendix B, "Additional Conditions," of the operatinglicenses regarding an alternate method for evaluating SLMCPR values. An evaluation ofthe proposed license amendments is provided in Enclosure 1.

CP&L has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1), usingthe criteria in 10 CFR 50.92(c), and determined that this change involves no significanthazards considerations.

In accordance with 10 CFR 50.91(b), CP&L is providing a copy of the proposed licenseamendments to the designated representative for the State of North Carolina.

CP&L requests approval of the proposed amendments by March 1, 2013, in order tosupport reactor start-up following the Unit 2 refueling outage, which is currently scheduledto begin in March 2013. Once approved, the Unit 2 amendment shall be implemented

Progress Energy Carolinas, Inc.P.O. Boa 104298Southport, NC 28461

T > 910.457.3698

Document Control DeskBSEP 12-0031 / Page 2

prior to start-up from the 2013 Unit 2 refueling outage and the Unit 1 amendment shall beimplemented prior to start-up from the 2014 Unit 1 refueling outage.

No regulatory commitments are contained in this submittal. Please refer any questionsregarding this submittal to Mr. Lee Grzeck, Acting Supervisor - Licensing/RegulatoryPrograms, at (910) 457-2487.

I declare, tinder penalty of perjury. that the foregoing is true and correct. Executed onMarch 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 23. Typed Technical Specification Pages - Unit 14. Typed Technical Specification Pages - Unit 25. 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 inAccordance With 10 CFR 2.390)

7. AREVA Affidavit Regarding Withholding AREVA DocumentNo. 51-9175814-000, "Brunswick Unit I Cycle 19 SLMCPR Analysis WithSAFLIM3D Methodology (Proprietary Version)" from Public Disclosure

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

9. AREVA Document No. 51-9176407-000, "Brunswick Unit I Cycle 19SLMCPR Analysis With SAFLIM3D Methodology - Operability Assessment(Proprietary Version)" (Proprietary Information - Withhold from PublicDisclosure in Accordance With 10 CFR 2.390)

10. AREVA Affidavit Regarding Withholding AREVA DocumentNo. 51-9176407-000, "Brunswick Unit 1 Cycle 19 SLMCPR Analysis WithSAFLIM3D Methodology - Operability Assessment (Proprietary Version)"from Public Disclosure

11. AREVA Document No. 51-9177315-000, "Brunswick Unit I Cycle 19SLMCPR Analysis With SAFLIM3D Methodology - Operability Assessment(Nonproprietary Version)"


12. AREVA Document No. 51-9175787-000, "Brunswick Unit 2 Cycle 20SLMCPR Analysis With SAFLIM3D Methodology (Proprietary Version)"(Proprietary Information - Withhold from Public Disclosure inAccordance With 10 CFR 2.390)

13. AREVA Affidavit Regarding Withholding AREVA DocumentNo. 51-9175787-000, "Brunswick Unit 2 Cycle 20 SLMCPR Analysis WithSAFLIM3D Methodology (Proprietary Version)" from Public Disclosure

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

15. AREVA Document No. 51-9176342-000, "Brunswick Unit 2 Cycle 20SLMCPR Analysis With SAFLIM3D Methodology - Operability Assessment(Proprietary Version)" (Proprietary Information - Withhold from PublicDisclosure in Accordance With 10 CFR 2.390)

16. AREVA Affidavit Regarding Withholding AREVA DocumentNo. 51-9176342-000, "Brunswick Unit 2 Cycle 20 SLMCPR Analysis WithSAFLIM3D Methodology - Operability Assessment (Proprietary Version)"fiom Public Disclosure

17. AREVA Document No. 51-9177316-000, "Brunswick Unit 2 Cycle 20SLMCPR Analysis With SAFLIM3D Methodology - Operability Assessment(Nonproprietary Version)"

18. AREVA Document ANP-3086(P), Revision 0, "Brunswick Unit I and Unit 2SLMCPR Operability Assessment Critical Power Correlation forATRIUM 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 AssessmentCritical Power Correlation for ATRIUM I OXM Fuel - Improved K-factorModel" from Public Disclosure

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


cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region IIATTN: Mr. Victor M. McCree, Regional Administrator245 Peachtree Center Ave, NE, Suite 1200Atlanta, GA 30303-1257

U. S. Nuclear Regulatory CommissionATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector8470 River RoadSouthport, NC 28461-8869

U. S. Nuclear Regulatory Commission (Electronic Copy Only)ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A)11555 Rockville PikeRockville. MD 20852-2738

cc (with Enclosures 1 through 5 only):

Chair - North Carolina Utilities CommissionP.O. Box 29510Raleigh, NC 27626-05 10

Mr. W. Lee Cox, ll, Section ChiefRadiation Protection SectionNorth Carolina Department of Environment and Natural Resources1645 Mail Service CenterRaleigh, NC 27699-1645

BSEP 12-0031Enclosure 1

Page 1 of 13

Evaluation of Proposed License Amendment Request

Subject: Request for License Amendments - Addition of Analytical MethodologyTopical Report to Technical Specification 5.6.5, "CORE OPERATINGLIMITS REPORT (COLR)" and Revision to TechnicalSpecification 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 doingbusiness as Progress Energy Carolinas, Inc., to amend Appendix A, TechnicalSpecifications (TS), of Renewed Facility Operating License Nos. DPR-71 and DPR-62for 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 ReportANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors withAREVA Topical Report ANP-10307PA, Revision 0, AREVA MCPR SafI/y LimitMethodology;for Boiling Kater Reactors, June 2011, in the list of analytical methods thathave 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 MinimumCritical Power Ratio (SLMCPR) values, and (3) revise the license condition inAppendix B, "Additional Conditions," of the operating licenses regarding an alternatemethod for evaluating SLMCPR values. These changes are needed to support the nextcycles of operation for BSEP Units I and 2 (i.e., Cycle 20 for BSEP, Unit 1, which isscheduled to begin April 2014 and Cycle 21 for BSER Unit 2, which is scheduled tobegin March 2013).

2.0 Detailed Description

Proposed Change 1:

TS 5.6.5.b identifies the analytical methods that should be used to determine coreoperating limits. The current TS states:

The analytical methods used to determine the core operating limits shall be thosepreviously reviewed and approved by the NRC, specifically those described in thefollowing documents:

BSEP 12-0031Enclosure I

Page 2 of 13

Currently, the SLMCPR protected by the reactor core critical power limits is determinedusing the NRC-approved analytical methodology described in the following AREVATopical 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 ReportANP- I 0307PA, Revision 0, AREVA MCPR Safety Limit MethodologyJbor Boiling WaterReactors, June 2011, in the list of analytical methodologies in TS Section 5.6.5.b thatmay be used for determining core operating limits which have been reviewed andapproved by the NRC. Topical Report ANP- 1030 7PA describes an improved AREVAmethodology for determining SLMCPR and incorporates a realistic fuel channel bowmodel. 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 illlicensing applications for boiling water reactors.

Proposed Change 2:

Using the analytical methods described in Topical Report ANP-10307PA, Revision 0, thetwo 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 beingrevised.

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 singlerecirculation 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 singlerecirculation loop operation.

The proposed amendments will revise the SLMCPR values in TS 2.1.1.2 for two loopoperation and single loop operation. The SLMCPR value for two recirculation loopoperation is being changed from > 1.11 (i.e., for both Unit 1 and Unit 2) to > 1.08, andfrom > 1.12 (i.e., for Unit 1) and > 1.13 (i.e., for Unit 2) to > 1.11 (i.e.. for both Unit Iand Unit 2) for SLO.

Proposed Change 3:

The Facility Operating License, Appendix B, "Additional Conditions," includes a licensecondition which requires the performance of a confirmatory evaluation for SLMCPR,


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 coreoperating limit values determined using the ANP- I 0298PA, ACE/ATRIUM 1 OXMCritical Power Correlation (i.e., TS 5.6.5.b.21), shall be evaluated with methodsdescribed in AREVA Operability Assessment CR 2011-2274, Revision 1 to verifythe values determined using the NRC-approved method remain applicable and thecore operating limits include margin sufficient to bound the effects of the K-factorcalculation issue described in AREVA Operability Assessment CR 2011-2274,Revision 1. The results of the evaluation shall be documented and submitted tothe 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 evaluationof SLMCPR with the methods described in AREVA Document ANP-3086(P), Revision 0,Brunswick Unit 1 and Unit 2 SLMCPR Operability Assessment Critical PowerCorrelation /br ATRIUM'I ]OXM Fuel - Inproved K-faIctor Model, provided inEnclosure 18. The revised license condition would read as follows:

Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and coreoperating limit values determined using the ANP-10298PA, ACE/ATRIUM IOXMCritical Power Correlation (i.e., TS 5.6.5.b.21), shall be evaluated to verify thevalues determined using the NRC-approved method remain applicable and thecore operating limits include margin sufficient to bound the effects of the K-factorcalculation issue described in AREVA Operability Assessment CR 2011-2274,Revision 1. SLMCPR shall be evaluated with methods described in AREVADocument ANP-3086(P), Revision 0, "Brunswick Unit 1 and Unit 2 SLMCPROperability Assessment Critical Power Correlation for ATRIUM 1OXM Fuel -Improved K-factor Model." Setpoint and core operating limit values shall beevaluated with methods described in AREVA Operability AssessmentCR 2011-2274, Revision 1. The results of the evaluation shall be documented andsubmitted to the NRC, for review, at least 60 days prior to startup of eachoperating cycle.

The license condition will retain the existing requirement to perform the evaluation ofsetpoint and core operating limit values using the methods described in AREVAOperability Assessment CR 2011-2274, Revision 1.

Enclosure 2 contains marked-up TS and Operating License pages for BSEP Unit 2indicating the proposed changes. Since TS 2.1.1.2, TS 5.6.5.b and the license conditionchange associated with this request are equivalent for BSEP Units I and 2, only mark-ups


Page 4 of 13

for Unit 2 are provided. Enclosures 3 and 4 provide typed versions of the Unit I andUnit 2 TS, respectively. These typed TS pages are to be used for issuance of theproposed amendments.

In addition, in support of the proposed TS changes, TS Bases Section 2.1.1 will berevised, after issuance of the amendments, to reflect application of AREVA TopicalReport ANP-10307PA, Revision 0. Mark-ups of the TS Bases changes for Unit') areprovided 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 aredocumented in topical reports which are reviewed by the NRC on a generic basis. TheNRC, in its safety evaluation for the approved topical report, defines the basis foracceptance in conjunction with any limitations and conditions on use of the topical report,as appropriate. In situations where a plant-specific license amendment request referencesa generic topical report, plant-specific applicability of the material presented in thetopical report is reviewed.

System Description/Applicable SafeOt Analysis'

The BSEP, Unit I and 2 cores consist primarily of ATRIUM-10 and ATRIUM 1OXM fuelassemblies, along with a small number of GEl 4 assemblies. Beginning with Unit 1Cycle 19 in Spring 2012, the Unit I core will no longer use any co-resident non-AREVAfuel. Beginning with Unit 2 Cycle 21 in Spring 2013, the Unit 2 core will also no longeruse any co-resident non-AREVA fuel.

Beginning with BSEP Unit 2 Cycle 21, CP&L intends to use the analytical methodologydescribed in Topical Report ANP-1 0307PA for determining the SLMCPR values for twoloop and single loop operation. Topical Report ANP-10307PA describes an improvedAREVA critical power methodology. NRC approval of Topical Report ANP-10307P isdocumented 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 ReportANP-10307PA will enable CP&L to implement the analytical methods described in thereport. SLMCPR is currently determined using the NRC-approved analyticalmethodology described in AREVA Topical Report ANF-524(P)(A), ANF Critical PowerMethodology/or Boiling Water Reactors (i.e., Item 11 in TS 5.6.5.b). This topical reportis listed in Brunswick Technical Specification 5.6.5.b as an analytical methodology thatmay be used to determine core operating limits.


Page 5 of 13

Plant Specilc Methodology Applicability Evaluation

The methodology described in ANP- 1 0307PA, Revision 0, is used to determine theSLMCPR such that at least 99.9 percent of the fuel rods in the core will not experiencedry-out during normal operation and anticipated operational occurrences if the coreMCPR is greater than or equal to the SLMCPR. Enclosures 6 and 12 summarize themethodology, inputs, and results supporting the BSEP Unit 1 Cycle 19 and Unit 2Cycle 20 SLMCPR values calculated using the ANP-l 0307PA methodology, respectively.The SLMCPR is determined using a statistical analysis that employs a Monte Carloprocess that perturbs key input parameters used in the MCPR calculation based on theiruncertainties. Table I in Enclosures 6 and 12 identifies these uncertainty inputs.

The fuel-related power distribution uncertainty inputs used by the ANP-10307PAmethodology are calculated from separately determined uncertainty componentsdescribed in EMF-2 1 58(P)(A), Siemens Power Corporation Methodology for BoilingWater Reactors: Evaluation and [alidation o/ CASMO-4/MICROBURN-B2. The NRC-approved uncertainty components friom EMF-2158(P)(A) are shown to be applicable toBSEP 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 flowuncertainties) are identified in Table I of Enclosures 6 and 12, and are consistent with theNRC-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 isbased on AREVA fuel channel measurements. Channel fluence is an input to the channelbow 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 AREVAfuel channels made of Zircaloy-4 material. BSEP has not experienced indications ofabnormal channel bow with AREVA channels. Should indications of abnormal channelbow be experienced by BSEP, the channel bow model will be applied in a conservativemanner as described in ANP-10307PA. The BSEP Unit 2 Cycle 20 core includes a smallnumber of third-cycle GEI 4 fuel assemblies near the core periphery that have substantialcritical power margin. The channel bow model uncertainty applied to these GEl4 fuelassemblies was conservatively increased based on GE 14 specific channel bowuncertainty provided by Global Nuclear Fuel, consistent with the ANP-1 0307PAmethodology. The ANP-10307PA channel bow model is applicable to BSEP, because themodel accounts for channel fluence calculated specific to the BSEP reactor coreoperating conditions. The channel bow model uncertainty is applicable to BSEP., becauseBSEP has not experienced abnormal channel bow with AREVA fuel channels andchannel bow uncertainty specific to the remaining GEI 4 fuel channels is applied.

BSEP 12-003 1Enclosure I

Page 6 of 13

License Condition Evaluation

In conjunction with the issuance of License Amendment Nos. 257 and 285 to the BSEPUnit I and 2 Operating Licenses (i.e., Reference 4), the NRC included a license conditionin Appendix B, "Additional Conditions," of the operating licenses. This license conditionis intended to ensure the limits generated with the NRC-approved methods appropriatelybound the effects of the K-factor calculation issue described in AREVA OperabilityAssessment CR 2011-2274, Revision 1. The license condition currently states:

Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and coreoperating limit values determined using the ANP- I 0298PA, ACE/ATRIUM 1 OXMCritical Power Correlation (i.e., TS 5.6.5.b.2 1), shall be evaluated with methodsdescribed in AREVA Operability Assessment CR 2011-2274, Revision 1 to verifythe values determined using the NRC-approved method remain applicable and thecore operating limits include margin sufficient to bound the effects of the K-factorcalculation issue described in AREVA Operability Assessment CR 2011-2274,Revision 1. The results of the evaluation shall be documented and submitted tothe 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 toassess SLMCPR values determined using the ANF-524(P)(A) methodology may notalways remain appropriate to assess SLMCPR values determined using theANP- 10307PA methodology. ANP-3086(P), Revision 0, Brunswick Unit ] and Unit 2SLMCPR Operability Assessment Critical Power Correlation jbr ATRIUM I OXM Fuel -Improved K-/actor Model., describes a BSEP-specific methodology to verify SLMCPRvalues determined using the methods described in ANP- 100307PA are applicable andinclude margin sufficient to bound the effects of the K-factor calculation issue describedin AREVA Operability Assessment CR 2011-2274, Revision 1. ANP-3086(P),Revision 0, is provided in Enclosure 18. Evaluations performed consistent with themethods described in ANP-3086(P), Revision 0, have been performed and are provided inEnclosures 9 and 15. Based on these evaluations, it has been concluded that theSLMCPR results determined using the methods described in ANP-10307PA, andprovided in Enclosure 6 and 12, are applicable and include margin sufficient to bound theeffects of the K-factor calculation issue described in AREVA Operability AssessmentCR 2011-2274, Revision 1.

The proposed amendments will revise the current license condition by requiring theevaluation of SLMCPR with the methods described in AREVA DocumentNo. ANP-3086(P), Revision 0. The license condition will retain the existing requirementto perform the evaluation of setpoint and core operating limit values using the methodsdescribed in AREVA Operability Assessment CR 2011-2274, Revision 1.


Page 7 of 13

Confobrmance with Methodologv and Saety Evaluation Limitations

The NRC letter approving Topical Report ANP-10307PA states that the report isacceptable for referencing in licensing applications for boiling water reactors to the extentspecified and tinder the limitations delineated in the topical report and in the final NRCsafety evaluation. The final NRC safety evaluation concluded the topical report isacceptable for use for plant-specific licensing actions, without listing any limitations orconditions.

Upon approval of this license amendment application and incorporation of Topical ReportANP- I 0307PA, Revision 0, into the BSEP Unit 1 and 2 Technical Specifications, CP&Lwill implement the analytical methods described in the report and will conform with themethodology described in the topical report.

The SLMCPR results determined using the ANP-1 0307PA methodology provided inEnclosures 6 and 12 support two loop operation SLMCPR values of 1.07 and 1.06, andsingle loop operation SLMCPR values of 1.09 and 1.08, for BSEP Unit I Cycle 19 andBSEP Unit 2 Cycle 20, respectively. These results support the requested TS SLMCPRfor 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 beingrequested to accommodate small cycle-to-cycle variations.

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

4.0 Regulatory Evaluation

4.1 Applicable Regulatory Requirements/Criteria

10 CFR 50.36, "Technical specifications," paragraph (c)(1), requires that power reactorfacility TS include safety limits for process variables that protect the integrity of certainphysical barriers that guard against the uncontrolled release of radioactivity. The fuelcladding integrity SLMCPR is established to assure that at least 99.9% of the fuel rods inthe core do not experience boiling transition during normal operation and anticipatedoperational 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 SLMCPRfrom the TS.

10 CFR 50, Appendix A, General Design Criterion (GDC) 10 requires that the reactorcore and associated coolant, control, and protection systems be designed with appropriatemargin to assure that specified acceptable fuel design limits are not exceeded during anycondition of normal operation, including the effects of AQOs. To ensure compliance withGDC 10, CP&L has performed the plant-specific SLMCPR analyses using


Page 8 of 13

NRC-approved methodologies as prescribed in NUREG-0800, "Standard Review Planfor the Review of Safety Analysis Reports for Nuclear Power Plants," Section 4.4. TheSLMCPR ensures that sufficient conservatism exists in the operating limit MCPR suchthat, in the event of an AOO, there is a reasonable expectation that at least 99.9 percent ofthe fuel rods in the core will avoid boiling transition for the power distribution within thecore 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 ReportANP- 1 0307PA states that the report is acceptable for referencing in licensing applicationsfor boiling water reactors to the extent specified and under the limitations delineated inthe topical report and in the final NRC safety evaluation. The final NRC safetyevaluation concluded the topical report is acceptable for use for plant-specific licensingactions, without listing any limitations or conditions.

Since the proposed amendments for BSEP, Units I and 2 are the first licensee request touse 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 isbeing revised to replace AREVA Topical Report ANF-524(P)(A), ANF Critical PowerAiethodology/br Boiling Water Reactors with AREVA Topical Report ANP- I 0307PA,Revision 0, AREVI MCPR Sa/ety Limit Methodology/br Boiling Water Reactors, June2011, in the list of analytical methods that have been reviewed and approved by the NRCfor determining core operating limits. Second, the Safety Limit Minimum Critical PowerRatio (SLMCPR) values contained in Technical Specification (TS) 2.1.1.2 are beingrevised 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 bothUnit I and Unit 2) for single loop recirculation operation. Finally, a license condition inAppendix B, "Additional Conditions" of the operating licenses regarding evaluatingSLMCPR values is being updated to ensure SLMCPR values determined using themethodology described in AREVA Topical Report ANP- 10307PA are evaluated with analternate method to AREVA Operability Assessment CR 2011-2274, Revision 1, which isappropriate for verifying SLMCPR values detennined using the ANP-1 0307PAmethodology remain applicable, while maintaining the existing methodology used toevaluate setpoint and core operating limit values.


Page 9 of 13

CP&L has evaluated whether or not a significant hazards consideration is involved withthe 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 theindividual precursors to that accident. The proposed license amendments do notinvolve any plant modifications or operational changes that could affect systemreliability or performance, or that could affect the probability of operator error. Assuch, the proposed changes do not affect any postulated accident precursors. Sinceno individual precursors of an accident are affected, the proposed licenseamendments do not involve a significant increase in the probability of a previouslyanalyzed event.

The consequences of an evaluated accident are determined by the operability ofplant systems designed to mitigate those consequences. The basis for the SLMCPRcalculation is to ensure that during normal operation and during anticipatedoperational occurrences, at least 99.9 percent of all fuel rods in the core do notexperience transition boiling if the safety limit is not exceeded.

The proposed SLMCPR values have been determined using NRC-approvedmethods discussed in AREVA Topical Report ANP-10307PA, Revision 0, AREM4A'JCPR Sqaety Limit Methodology for Boiling Water Reactors, June 2011. Tosupport 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-approvedanalytical methods identified in Technical Specification 5.6.5.b. Replacing AREVATopical Report ANF-524(P)(A), ANF Critical Power Methodology for BoilingWater Reactors with the analytical methods described in TopicalReport ANP-10307PA in Technical Specification 5.6.5.b does not alter theassumptions of accident analyses. Furthermore, establishing a two recirculationloop 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 least99.9 percent of all fuel rods in the core do not experience transition boiling), whilethe revised license condition ensures that SLMCPR, setpoint, and core operatinglimit values determined using the NRC-approved AREVA methodologies remainapplicable and the core operating limits include margin sufficient to bound theeffects of the K-factor calculation issue described in AREVA OperabilityAssessment CR 2011-2274, Revision 1. Based on these considerations, the


Page 10of 13

proposed changes do not involve a significant increase in the consequences of apreviously analyzed accident.

2. Does the proposed change create the possibility of a new or different kind ofaccident from any accident previously evaluated?

Response: No

Creation of the possibility of a new or different kind of accident requires creatingone or more new accident precursors. New accident precursors may be created bymodifications of plant configuration, including changes in allowable modes ofoperation. The SLMCPR is a TS numerical value calculated for two recirculationloop operation and single recirculation loop operation to ensure at least99.9 percent of all fuel rods in the core do not experience transition boiling if thesafety limit is not exceeded. SLMCPR values are calculated using NRC-approvedmethodology identified in the TS. The proposed SLMCPR values and the AREVAmethodology being added to TS do not involve any new modes of plant operationor any plant modifications and do not directly or indirectly affect the failure modesof any plant systems or components. Therefore, the proposed changes do not createthe possibility of a new or different kind of accident from any accident previouslyevaluated.

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 ofthe fuel rods do not experience transition boiling during normal operation andanticipated 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 ofNRC-approved analytical methods identified in Technical Specification 5.6.5.b,revision of the SLMCPR values in Technical Specification 2.1.1.2 usingNRC-approved methodology, and confirmation that the SLMCPR,. setpoint, andcore operating limit values remain applicable and the core operating limits includemargin sufficient to bound the effects of the K-factor calculation issue described inAREVA Operability Assessment CR 2011-2274, Revision 1, will ensure that thecurrent level of fuel protection is maintained by continuing to ensure that the fueldesign safety criterion is met (i.e., that no more than 0.1 percent of the rods areexpected to be in boiling transition if the MCPR Safety Limit is not exceeded).


Page 11 of 13

Meeting the fuel design criterion that at least 99.9 percent of all fuel rods in thecore do not experience transition boiling and establishing core operating limitsbased on the proposed SLMCPR values, to ensure that the SLMCPR is notexceeded, ensures the margin of safety required by the fuel design criterion ismaintained. Therefore, the proposed amendments do not result in a significantreduction in the margin of safety.

Based on the above, CP&L concludes that the proposed amendments present nosignificant 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 regulationsand requirements continue to be met. CP&L has determined that the proposed changesdo not require any exemptions or relief fronm regulatory requirements, other than theTechnical Specifications, and do not affect conformance with any General DesignCriterion (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 StationUnits I and 2," dated November 1973, BSEP meets the intent of the General DesignCriteria (GDC), published in the Federal Register on May 21, 1971, as Appendix A to10 CFR Part 50. The proposed changes do not affect compliance with the intent of theGDCs. 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 bedesigned with appropriate margin to assure that specified acceptable fuel designlimits are not exceeded during any condition of normal operation, including theeffects of anticipated operational occurrences.

To ensure compliance with GDC 10, CP&L performs plant-specific critical power limitanalyses using NRC-approved methodologies. The MCPR Safety Limit ensures thatsufficient conservatism exists in the operating limit MCPR such that, in the event of ananticipated operational occurrence, there is a reasonable expectation that at least99.9 percent of the fuel rods in the core will avoid boiling transition for the powerdistribution within the core including uncertainties.

10 CFR 50.36(c)(5) states that the Technical Specifications will include administrativecontrols that address the provisions relating to organization and management, procedures,record keeping, review and audit, and reporting necessary to assure operation of thefacility in a safe manner. The Core Operating Limits Report (COLR) is required as a part


Page 12 of 13

of the reporting requirements specified in the Brunswick Technical SpecificationsAdministrative Controls section. The Technical Specifications require the core operatinglimits to be established prior to each reload cycle, or prior to any remaining portion of areload cycle, and to be documented in the COLR. In addition, it requires the analyticalmethods used to determine the core operating limits to be those that have been previouslyreviewed and approved by the NRC, and specifically to be those described in TeclmicalSpecification 5.6.5.b. The proposed amendments ensure that these requirements are met.

In conclusion, based on the considerations discussed above, (1) there is reasonableassurance that the health and safety of the public will not be endangered by operation inthe proposed manner, (2) such activities will be conducted in compliance with theCommission's regulations, and (3) the issuance of the amendments will not be inimical tothe 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 anddo not change a requirement with respect to installation or use of a facility componentlocated within the restricted area, as defined in 10 CFR 20, Standards for ProtectionAgainst Radiation, and do not change an inspection or surveillance requirement. Theproposed amendments do not involve: (i) a significant hazards consideration, (ii) asignificant change in the types or significant increase in the amounts of any effluent thatmay be released offsite, or (iii) a significant increase in individual or cumulativeoccupational radiation exposure. Accordingly, the proposed amendments meet theeligibility 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 orenvironmental assessment need be prepared in connection with the proposedamendments.

7.0 References

I: Letter from Robert A. Nelson (NRC) to Pedro Salas (AREVA NP Inc.), "FinalSafety Evaluation for AREVA NP, Inc. Topical Report ANP-10307P, Revision 0,'AREVA MCPR [Minimum Critical Power Ratio] Safety Limit Methodology forBoiling Water Reactors' (TAC No. ME2914)," dated June 14, 2011, ADAMSAccession Number ML 1 140A 125.

2. ANP-3086(P), Revision 0, Brunswick Unit I and Unit 2 SLMCPR OperabilityAssessment Critical Power Correlation for ATRIUM 1 OXM Fuel - ImprovedK-ftictor Model, February 2012.


Page 13 of 13

3. Letter from William Jefferson, Jr. (CP&L) to U.S. Nuclear RegulatoryCommission Document Control Desk, "Response to Additional InformationRequest Supporting License Amendment Requests for Addition of AnalyticalMethodology Topical Reports to Technical Specification 5.6.5 (NRC TACNos. 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 TopicalReport to Technical Specification 5.6.5 (TAC Nos. ME3856 and ME3857)," datedApril 8, 2011, ADAMS Accession Number MLI 11010234.


Marked-up Technical Specificationand Operating License Pages - Unit 2

SLs2.0

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:1.08MCPR shall be _> 444 for two recirculation loop operation or 4-4 forsingle recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of activeirradiated 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 hours:

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 Requirements5.6

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 forBoiling Water Reactors - Neutronic Methods for Design andAnalysis.

7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology forBoiling Water Reactors: Application of the ENC Methodology toBWR Reloads.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology forBoiling Water Reactors: Evaluation and Validation ofCASMO-4/MICROBURN-B2.

9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology forBoiling Water Reactors, THERMEX: Thermal Limits MethodologySummary 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 Method•. egy for Boiling W.tr

Boiling Water ReaetS.Reactors, Revision

0, June 2011 12. ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Programfor Boiling Water Reactor Transient Analyses.

13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient inBoiling Water Reactors.

14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation'sCritical 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 TransferCoefficients.

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 SolutionsLicensing Basis Methodology for Reload Applications.

(continued)

Brunswick Unit 2 5.0-21 Amendment No. 274 I

AmendmentNumber Additional Conditions

276 Upon implementation of Amendment No. 276adopting TSTF-448, Revision 3, the determinationof control room envelope (CRE) unfiltered airinleakage as required by SR 3.7.3.3, in accordancewith TS 5.5.13.c.(i), the assessment of CREhabitability as required by Specification 5.5.13.c.(ii),and the measurement of CRE pressure as requiredby Specification 5.5.13.d, shall be considered met.Following implementation:

(a) The first performance of SR 3.7.3.3, inaccordance with Specification 5.5.13.c.(i),shall be within the specified Frequency of6 years, plus the 18-month allowance ofSR 3.0.2, as measured from June 11, 2004,the date of the most recent successful tracergas test.

(b) The first performance of the periodicassessment of CRE habitability, Specification5.5.13.c.(ii), shall be within the next 9 months.

(c) The first performance of the periodicmeasurement of CRE pressure,Specification 5.5.13.d, shall be within18 months, plus the 138 days allowed bySR 3.0.2, as measured from the date of themost recent successful pressuremeasurement test.

Implementation Date

As described inparagraphs (a), (b),and (c) of thisAdditional Condition.

SLMCPR shall be evaluated withmethods described in AREVAdocument ANP-3086(P),Revision 0, Brunswick Unit 1 andUnit 2 SLMCPR OperabilityAssessment Critical PowerCorrelation for ATRIUM 1OXMFuel - Improved K-factor Model.Setpoint and core operating limitvalues shall be evaluated withmethods described in AREVAOperability Assessment CR2011-2274, Revision 1.

Safety Limit Minimum Critical Power Ratio(SLMCPR), setpoint, and core operating limit7c/ues determined using the ANP-10298PA,

Ck ATRIUM 1OXM Critical Power CorrelationiTS 5.6.5.b.21), shall be evaluated we#1

Upon implementation ofAmendment No. 2-85.

(•,•'• ,• 2911 22•"-o7-4" D•,• Re iiR "4"to verifylhe val es determined using the NRC-approved

method emain applicable and the core operatinglimits inc de margin sufficient to bound the effectsof the K-f ctor calculation issue described inAREVA O rability Assessment CR 2011-2274,Revision 1. The results of the evaluation shall bedocumented and submitted to the NRC, for review,at least 60 days prior to startup of each operatingcycle.

Brunswick Unit 2 App. B-2 Amendment No. 285 1


Typed Technical Specification Pages - Unit I

SLs2.0


2.1 SLs



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 Ž! 1.08 for two recirculation loop operation or Ž 1.11 forsingle recirculation loop operation.



Reactor steam dome pressure shall be <_ 1325 psig.

I

2.2 SL Violations




Brunswick Unit 1 2.0-1 Amendment No. I








10. XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Codefor BWR Transient Thermal-Hydraulic Core Analysis.

11. ANP-10307PA, AREVA MCPR Safety Limit Methodology forBoiling Water Reactors, Revision 0, June 2011.

12. ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Programfor Boiling Water Reactor Transient Analyses.

13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient inBoiling Water Reactors.




17. EMF-2292(P)(A), ATRIUM TM-10: Appendix K Spray Heat TransferCoefficients.


19. NEDO-32465-A, Reactor Stability Detect and Suppress SolutionsLicensing Basis Methodology for Reload Applications.

(continued)



Typed Technical Specification Pages - Unit 2








10. XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Codefor BWR Transient Thermal-Hydraulic Core Analysis.

11. ANP-10307PA, AREVA MCPR Safety Limit Methodology forBoiling Water Reactors, Revision 0, June 2011.

12. ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Programfor Boiling Water Reactor Transient Analyses.

13. ANF-1 358(P)(A), The Loss of Feedwater Heating Transient inBoiling Water Reactors.




17. EMF-2292(P)(A), ATRIUM TM-10: Appendix K Spray Heat TransferCoefficients.


19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions

Licensing Basis Methodology for Reload Applications.

(continued)


SLs2.0


2.1 SLs



THERMAL POWER shall be •< 23% RTP.

2.1.1.2 With the reactor steam dome pressure Ž> 785 psig and core flow _> 10%rated core flow:

MCPR shall be > 1.08 for two recirculation loop operation or _> 1.11 forsingle recirculation loop operation.



Reactor steam dome pressure shall be < 1325 psig.

I

2.2 SL Violations




Brunswick Unit 2 2.0-1 Revision No. I


Marked-up Technical Specification Bases Pages - Unit 2(For information only)

MCPRB 3.2.2

BASES

REFERENCES

ANP-10307PA,"AREVA MCPRSafety LimitMethodology forBoiling WaterReactors"

-I--'

1. UFSAR Section 4.4.2.1.

2. AN^ 521 (P)(A), "ANP Critical Po•eF Methodology f)o Beoling

Watcr Rczictor5."

3. UFSAR, Chapter 4.



I

6. (Deleted.)

7. XN-NF-80-19(P)(A) Volume 3, "Exxon Nuclear Methodology forBoiling Water Reactors, THERMEX: Thermal Limits MethodologySummary Description," (as identified in the COLR).

8. EMF-2158(P)(A), "Siemens Power Corporation Methodology forBoiling Water Reactors: Evaluation and Validation ofCASMO-4/MICROBURN-B2," (as identified in the COLR).

9. ANF-913(P)(A) Volume 1, "COTRANSA2: A Computer Programfor Boiling Water Reactor Transient Analyses," (as identified in theCOLR).

10. XN-NF-84-105(P)(A), "XCOBRA-T: A Computer Code for BWRTransient Thermal-Hydraulic Core Analysis," (as identified in theCOLR).

11. 10 CFR 50.36(c)(2)(ii).

Brunswick Unit 2 B 3.2.2-4 Revision No. 62 I

Reactor Core SLsB 2.1.1

BASES

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuelclad 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 fueldesign criteria. SL 2.1.1.3 ensures that the reactor vessel water level isgreater than the top of the active irradiated fuel in order to preventelevated 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 potentialVIOLATIONS for radioactive releases in excess of 10 CFR 50.67, "Accident Source

Term," limits (Ref. 5). Therefore, it is required to insert all insertablecontrol rods and restore compliance with the SLs within 2 hours. The2 hour Completion Time ensures that the operators take prompt remedialaction and also ensures that the probability of an accident occurringduring this period is minimal.

I

REFERENCES 1. EMF-2209(P)(A), "SPCB Critical Power Correlation," (as identifiedin the COLR).

ANP-10307PA,"AREVA MCPRSafety LimitMethodology forBoiling WaterReactors"

2. EMF-2245(P)(A), "Application of Siemens Power Corporation'sCritical Power Correlations to Co-Resident Fuel," (as identified inthe COLR).

3. ANF 524(P)(A), "ANF Critical PcReW8 Methodelog, for Bo~ilingWAtcr 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


AREVA Affidavit Regarding WithholdingAREVA Document No. 51-9175814-000,

"Brunswick Unit I Cycle 19 SLMCPR AnalysisWith 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. McCoyNOTARY PUBLIC, STATE OF WASHINGTONMY COMMISSION EXPIRES: 1/14/2016

Documents

Letter Enclosures Contain Proprietary Information Progress ... · Letter Enclosures Contain Proprietary Information Progress inergy ithhold in Accordance with 10 CFR 2.390(a)(6) Vice