BSEP 15-0007, Request for License Amendments - Adoption of Topical Report ANP-10298P-A, Revision 1

Request for License Amendments - Adoption of Topical Report ANP-10298P-A, Revision 1

ML15075A021
Person / Time
Site:	Brunswick
Issue date:	02/19/2015
From:	William Gideon Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ANP-10298P-A, Rev 1, BSEP 15-0007, TSC-2015-01
Download: ML15075A021 (25)

Text

William R. Gideon DUKE Vice President ENERGY, Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 o: 910.457.3698 February 19, 2015 Serial: BSEP 15-0007 TSC-2015-01 10 CFR 50.90 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 - Adoption of Topical Report ANP-1 0298P-A, Revision 1 Ladies and Gentlemen:

Pursuant to the provisions of 10 CFR 50.90, Duke Energy Progress, Inc. (Duke Energy), hereby requests a revision to the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendments revise TS 5.6.5.b by replacing AREVA Topical Report ANP-1 0298PA, ACE/A TRIUM IOXM CriticalPower Correlation, Revision 0, March 2010, with AREVA Topical Report ANP-10298P-A, ACE/ATRIUM IOXM CriticalPower Correlation,Revision 1, March 2014, in the list of analytical methods that have been revised and approved by the NRC for determining core operating limits. Adoption of the topical report also results in the elimination of a license condition associated with ANP-1 0298PA Revision 0.

Adoption of the improved analytical methods accepted through Topical Report ANP-1 0298P-A, Revision 1, obviate the need to maintain the Appendix B license condition that was issued as part of License amendments 262 and 290 for Units 1 and 2, respectively. Accordingly, Duke Energy is requesting removal of the portion of the license condition pertaining to the evaluation of safety limit minimum critical power ratio (SLMCPR), setpoint, and core operating limit values using the methods described in AREVA Operability Assessment CR 2011-2274, Revision 1.

Further discussion of the proposed change is provided in Enclosure1.

Duke Energy 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, Duke Energy is providing a copy of the proposed license amendments to the designated representative for the State of North Carolina.

Duke Energy requests approval of the proposed amendments by February 19, 2016, to coincide with the start of the Unit 1 refueling outage. Once approved, the Unit 1 amendment shall be e--3i1

U.S. Nuclear Regulatory Commission Page 2 of 3 implemented prior to start-up from the 2016 Unit 1 refueling outage and the Unit 2 amendment shall be implemented prior to start-up from the 2017 Unit 2 refueling outage.

A list of regulatory commitments contained in this letter is provided in Enclosure 6.

Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on February 19, 2015.

Sincerely, William R. Gideon WRM/wrm

Enclosures:

1. Evaluation of Proposed License Amendment Request

2. Marked-up Technical Specification and Operating License Pages - Unit 1

3. Marked-up Technical Specification and Operating License Pages - Unit 2

4. Typed Technical Specification and Operating License Pages - Unit 1

5. Typed Technical Specification and Operating License Pages - Unit 2

6. List of Regulatory Commitments

U.S. Nuclear Regulatory Commission Page 3 of 3 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. Andrew Hon (Mail Stop OWFN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission ATTN: Ms. Michelle Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only)

Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1645 lee.cox@dhhs.nc.gov

BSEP 15-0007 Enclosure 1 Page 1 of 8 Evaluation of Proposed License Amendment Request

Subject:

Request for License Amendments - Adoption of Topical Report ANP-1 0298P-A, Revision 1 1.0 Summary Description This letter is a request by Duke Energy Progress, Inc. (Duke Energy), to amend the Appendix A, Technical Specifications (TS) and Appendix B, Additional Conditions, of Renewed Facility Operating License Nos. DPR-71 and DPR-62, for the Brunswick Steam Electric Plant (BSEP),

Unit Nos. 1 and 2, respectively.

The proposed changes: (1) revise TS 5.6.5.b by replacing AREVA Topical Report ANP-10298PA, ACE/A TRIUM IOXM Critical Power Correlation,Revision 0, March 2010, with Revision 1 of the same topical report, and (2) revise Appendix B, Additional Conditions, by removing the license condition issued by Amendment Nos. 262 and 290 for Units 1 and Unit 2, respectively.

These changes are needed to support use of the updated analytical methods contained in Topical Report ANP-1 0298P-A, Revision 1, for determining core operating limits for the Unit 1 and Unit 2 operating cycles beginning in 2016 and 2017, respectively.

2.0 Detailed Description Currently, reactor core critical power limits are determined using the analytical methodology described in AREVA Topical Report ANP-10298PA, ACE/ATRIUM IOXM CriticalPower Correlation,Revision 0, March 2010.

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:

21. ANP-10298PA, ACE/ATRIUM 1OXM Critical Power Correlation, Revision 0, March 2010.

The proposed amendments update TS 5.6.5.b.21 as follows:

21. ANP-10298P-A, ACE/ATRIUM 1OXM Critical Power Correlation, Revision 1, March 2014.

Appendix B, Additional Conditions, of each unit's Facility Operating License includes the following license condition:

Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and core operating limit values determined using the ANP-10298PA, ACE/ATRIUM 1OXM 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

BSEP 15-0007 Enclosure 1 Page 2 of 8 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 I and Unit 2 SLMCPR OperabilityAssessment CriticalPower Correlation for ATRIUM IOXM 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 proposed amendments delete this condition.

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.

The purpose of this license amendment request is to seek approval for and implement the analytical methods described in Topical Report ANP-1 0298P-A, Revision 1. Core operating limits are currently determined using the NRC-approved analytical methodology described in AREVA Topical Report ANP-1 0298PA, ACE/ATRIUM IOXM CriticalPower Correlation, Revision 0, March 2010 (i.e., TS 5.6.5.b.21).

Backgqround On April 8, 2011, the NRC issued Amendment Nos. 257 and 285 to the Renewed Facility Operating Licenses for BSEP, Units 1 and 2, respectively. The amendments changed the BSEP Unit 1 and 2 TS to support the transition to ATRIUM 1OXM fuel and associated core design methodologies. In conjunction with the issuance of these amendments, a license condition was included in Appendix B, Additional Conditions, of the Facility Operating Licenses to ensure that the core operating limits generated with the NRC-approved methods in ANP-10298PA, Revision 0, appropriately bound the effects of an inappropriate ANP-1 0298PA, Revision 0, K-factor calculation assumption using the methods described in AREVA Operability Assessment CR 2011-2274, Revision 1. K-factor is a modelling parameter that characterizes the effect on critical power ratio of radial fuel rod peaking distribution within a fuel bundle.

Subsequently, on March 1, 2013, the NRC issued Amendment Nos. 262 and 290 to the Renewed Facility Operating Licenses for BSEP, Units 1 and 2, respectively. These amendments replaced ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors, with AREVA Topical Report ANP-1 0307PA, Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, June 2011, in TS 5.6.5.b. The methods described in AREVA Operability Assessment CR 2011-2274, Revision 1, to assess Safety Limit Minimum Critical Power Ratio (SLMCPR) values determined using the ANF-524(P)(A) methodology were not always appropriate to assess SLMCPR values determined using the ANP-10307PA methodology. Therefore, these amendments also updated the Appendix B license condition to

BSEP 15-0007 Enclosure 1 Page 3 of 8 use the methods described in ANP-3086(P), Revision 0, Brunswick Unit I and Unit 2 SLMCPR OperabilityAssessment Critical Power Correlationfor A TRIUM I OXM Fuel - Improved K-factor Model, to verify SLMCPR values determined using the methods described in ANP-10307PA and ANP-10298PA, Revision 0, are applicable and include margin sufficient to bound the effects of the inappropriate ANP-10298PA, Revision 0, K-factor calculation assumption.

The NRC Safety Evaluation for Amendment Nos. 262 and 290 for BSEP, Units 1 and 2, respectively, included a note stating that a generic application to use ANP-10298PA, Revision 0, Supplement 1 P, which addresses the K-factor model issue, was under review by the NRC staff.

The NRC note goes on to explain that Duke Energy used the plant-specific ANP-3086(P),

Revision 0 methodology in the license condition to avoid a linked amendment with ANP-10298PA, Revision 0, Supplement 1P.

On March 31, 2014, the NRC issued a final safety evaluation that concluded AREVA Topical Report ANP-1 0298PA, Revision 0, Supplement 1P, Revision 0, is acceptable for referencing in licensing applications. The safety evaluation defined the basis for the NRC's acceptance of the topical report. AREVA subsequently incorporated Topical Report ANP-10298PA, Revision 0, Supplement 1 P, Revision 0, into the previously accepted Topical Report ANP-1 0298PA, Revision 0, to create Revision 1 of the topical report. The NRC's March 31, 2014, letter acknowledges Revision 1 of Topical Report ANP-1 0298PA to be the accepted version of the topical report. AREVA has incorporated the NRC's approval letter and safety evaluation into the revised topical report and published this as ANP-1 0298P-A, Revision 1. By letter dated November 24, 2014, the NRC determined that ANP-1 0298P-A, Revision 1, is acceptable for referencing in licensing applications.

Approval of this license amendment request and the incorporation of Topical Report ANP-1 0298P-A, Revision 1, will enable Duke Energy to implement the analytical methods described in the report. Core operating limits are currently determined using the NRC-approved analytical methodology described in AREVA Topical Report ANP-10298PA, Revision 0 (i.e.,

TS 5.6.5.b.21). Adoption of the improved K-factor models accepted through Topical Report ANP-10298P-A, Revision 1, resolves the inappropriate ANP-10298PA, Revision 0, K-factor calculation assumption and obviates the need to maintain the Appendix B license condition that was issued as part of License amendments 262 and 290 for Units 1 and Unit 2, respectively.

Accordingly, Duke Energy is requesting removal of the portion of the license condition pertaining to the evaluation of setpoint, and core operating limit values using the methods described in AREVA Operability Assessment CR 2011-2274, Revision 1, and the evaluation of SLMCPR values using the methods described in ANP-3086(P), Revision 0.

Conformance with Methodology and Safety Evaluation Limitations The NRC identified two limitations and conditions on use of the ANP-10298P-A, Revision 1 methodology. These limitations and conditions, and the demonstration that BSEP complies with the conditions, follow.

Limitation and Condition 1:

The ACE/ATRIUM 10XM methodology may only be used to perform evaluations of AREVA ATRIUM 1OXM fuel design. The ACE/ATRIUM 1 OXM correlation may also be

BSEP 15-0007 Enclosure 1 Page 4 of 8 used to evaluate the performance of the co-resident fuel in mixed cores as discussed in Section 3.4 of this SE.

Duke Energy will only use the ANP-10298P-A, Revision 1, methodology to perform evaluations of the AREVA ATRIUM 1 OXM fuel design and to evaluate the performance of co-resident fuel in mixed cores as discussed in the associated Safety Evaluation (i.e., ADAMS Accession Number ML14072A355).

Limitation and Condition 2:

ACE/ATRIUM 1OXM correlation shall not be used outside the range of applicability defined by the range of the test data prescribed in Table 2-1 of Reference 2.

Reference 2 is ANP-1 0298PA, Revision 0, Supplement 1 P, Revision 0, incorporated in ANP-10298P-A, Revision 1.

Table 2-1 of ANP-1 0298P-A, Revision 1 identifies the ACE/A OXM correlation range of applicability for conditions of mass flow rate, pressure, inlet subcooling, and design local peaking. This range of applicability is unchanged relative to ANP-10298PA, Revision 0. The restrictions on these conditions are implemented in AREVA engineering guidelines. The restrictions on range of applicability for mass flow rate, pressure, and inlet subcooling are also implemented in AREVA engineering computer codes, which include the BSEP POWERPLEX III core monitoring system. The restriction on design local peaking is also implemented in AREVA automation tools.

4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. Duke Energy has determined that the proposed changes do not require any exemptions or relief from 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 1 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, Duke Energy 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

BSEP 15-0007 Enclosure 1 Page 5 of 8 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 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 Technical 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.

4.2 Precedent On March 31, 2014, the NRC issued a final safety evaluation that concluded that AREVA Topical Report ANP-1 0298PA, Revision 0, Supplement 1 P, Revision 0, is acceptable for referencing in licensing applications. AREVA subsequently incorporated Topical Report ANP-1 0298PA, Revision 0, Supplement 1 P, Revision 0, into the previously accepted Topical Report ANP-1 0298PA, Revision 0, to create Revision 1 of the topical report. The NRC's March 31, 2014, letter acknowledges Revision 1 of Topical Report ANP-1 0298P-A to be the accepted version of the topical report.

Since the proposed amendments for BSEP, Units 1 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 revise TS 5.6.5.b by replacing AREVA Topical Report ANP-10298PA, ACE/ATRIUM IOXM CriticalPower Correlation,Revision 0, March 2010, with AREVA Topical Report ANP-1 0298P-A, ACE/ATRIUM 1OXM Critical Power Correlation, Revision 1, March 2014, in the list of analytical methods that have been approved by the NRC for determining core operating limits. The proposed license amendments also eliminate a license condition in Appendix B, Additional Conditions,that ensures Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and core operating values are evaluated using alternate methods to bound effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. The updated methodology described in AREVA Topical Report ANP-1 0298P-A, ACE/A TRIUM IOXM Critical Power Correlation, Revision 1, March 2014, has resolved the K-factor calculation issue and the Appendix B license condition is no longer needed.

BSEP 15-0007 Enclosure 1 Page 6 of 8 Duke Energy 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:

1. 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 only involve an update to a currently-approved methodology for determining core operating limits. As such, 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.

AREVA Topical Report ANP-10298P-A, ACE/ATRIUM IOXM Critical Power Correlation, Revision 1, March 2014, is being adopted to resolve a previously identified concern with the calculation of the K-factor, which is a modelling parameter that characterizes the effect on critical power ratio of radial fuel rod peaking distribution within a fuel bundle. Adoption of AREVA Topical Report ANP-10298P-A, ACE/ATRIUM 1OXM CriticalPower Correlation, Revision 1, also eliminates the need to perform a confirmatory evaluation as described in the Appendix B license condition issued as part of License Amendments 262 and 290 for Units 1 and 2. Therefore, the license condition is being eliminated.

The adoption of AREVA Topical Report ANP-1 0298P-A, ACE/A TRIUM IOXM Critical Power Correlation,Revision 1, March 2014, continues to ensure that the SLMCPR, setpoint, and core operating limit values determined using NRC-approved methods continue to satisfy the acceptance criteria that at least 99.9 percent of all fuel rods in the core do not experience boiling transition. Based on these considerations, the proposed change does 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 proposed amendments do neither.

BSEP 15-0007 Enclosure 1 Page 7 of 8 Core operating limit values are calculated using NRC-approved methodology identified in the TS. AREVA Topical Report ANP-1 0298PA, Revision 0, is an NRC-approved methodology listed in TS 5.6.5.b for determining core operating limits. Replacing the analytical methodology described in Topical Report ANP-1 0298PA, Revision 0, with the methodology contained in ANP-10298P-A, Revision 1, will ensure that (1) core operating limits are no longer affected by the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1, and (2) the current level of fuel protection is maintained by continuing to ensure that the fuel design safety criterion is met (i.e., that at least 99.9 percent of all fuel rods in the core do not experience boiling transition if the MCPR Safety Limit is not exceeded).

The update of AREVA analytical methodology does not involve any new modes of plant operation or any plant modifications and does 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 ensures that at least 99.9 percent of the fuel rods do not experience boiling transition during normal operation and anticipated operational occurrences, if the SLMCPR is not exceeded. Topical Report ANP-10298PA is listed as an NRC-approved analytical method in Technical Specification 5.6.5.b. Replacing the analytical methodology described in Topical Report ANP-10298PA, Revision 0, with the methodology contained in ANP-10298P-A, Revision 1, will ensure that (1) core operating limits are no longer affected by the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1, and (2) 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).

Meeting the fuel design criterion that at least 99.9 percent of all fuel rods in the core do not experience boiling transition and establishing core operating limits 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, Duke Energy has concluded 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.

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 ProtectionAgainst 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

BSEP 15-0007 Enclosure 1 Page 8 of 8 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

1. Letter from Pedro Salas (AREVA NP Inc.) to U.S. Nuclear Regulatory Commission Document Control Desk, Request for Review and Approval of AN-10298PA Revision 0, Supplement IP, Revision 0, "Improved K-factor Model for ACE/ATRIUM IOXM Critical Power Correlation,"dated December 21, 2011, ADAMS Accession Number ML11363A121.

2. Letter from Sher Bahadur (USNRC) to Pedro Salas (AREVA NP Inc.), Final Safety Evaluation by the Office of Nuclear Reactor for Topical Report ANP-10298PA, Revision 0, Supplement IP, Revision 0, "Improved K-Factor Model for ACE/ATRIUM IOXM CriticalPower Correlation"(TAC No. ME7936), dated March 31, 2014, ADAMS Accession Number ML14183A734.

3. 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 ML111010234.

4. Letter from Christopher Gratton (USNRC) to Michael J. Annacone (CP&L), Issuance of Amendments RegardingAddition of Analytical Methodology Topical Reports to Technical Specification 5.6.5 and revision to Minimum Critical Power Ratio Safety Limit (TAC Nos. ME8135 and ME8136), dated March 1, 2013, ADAMS Accession Number ML13037A551.

5. Letter from Aby S. Mohseni (USNRC) to Pedro Salas (AREVA NP Inc.), Verification Letter of the Approval Version of AREVA Inc. Topical Report ANP-10298P-A, Revision 1, "ACE/ATRIUM 1OXM Critical Power Correlation"(TAC No. MF4433), dated November 24, 2014, ADAMS Accession Number ML14321A129.

BSEP 15-0007 Enclosure 2 Marked-up Technical Specification and Operating License Pages - Unit 1

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.

~~Revision21. 0 , March ACE/ATRIUM 1OXM Critical Power Correlation,

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Brunswick Unit 1 5.0-22 Amendment No4) I

Amendment Additional Conditions Implementation Number Dat._ee mehremain a licable and the 6ore//,

opring limits* dcude margin s flcent to //

Th fuelfehannte bow standard deviation Upon iplementation f uc deritsed byinR OpeANP-1030Ay AR WSt.PRe (i.e.,TSall evaluat1) be to ve r me 201 in Lt boweMeimuN Critical-pove eLeMide dcoLmpnt R) terSafety ointhe canne modeaigIi Amendment No. 262 chaneueegrditto the n tc MPowrRai shall be ievased byithme rthodo channelsflunc dechanne with fluenegradin Do ANP teneaet gradints outs thannel) lenevs 0,gradetbunswc ifthIhand nelý measu r dtbse, ability whent app cled wer boundsen of meaent the dtb o PoeralRation shall be incrmeased byntheuratioeo as melculation ialuelsudesihalluevaluatedient cHichte l ncertainty isAN- ousiethod determined.

erbouds in thR Ameasremnaabi measremnt whe applied atab ase, Assessrmen rto Bareunwc RUnito1 TSApp. . .11) t Amnmn Thecfuel chanel bowertaintyais deviterionedompeetaino component ofith cannel bowmoe Amendment No.262

BSEP 15-0007 Enclosure 3 Marked-up Technical Specification and Operating License Pages - Unit 2

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

") 29*8 p-h. 20. BAW-1 0247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.

21 . -1 29P ACE/ATRIUM 10XM Critical Power Correlation,

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation,a a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Brunswick Unit 2 5.0-22 Amendment No.A ) I

Amendment Number Additional Conditions Imolementation Date 276 Upon Implementation of Amendment No. 276 As described in adopting TSTF-448, Revision 3, the determination paragraphs (a), (b),

of control room envelope (CRE) unfiltered air and (c) of this inleakage as required by SR 3.7.3.3, in accordance Additional Condition.

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 tesL (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.

Amendment Number Additional Conditions Implementation Date 290 The fuel channel bow standard deviation Upon implementation of component of the channel bow model uncertainty Amendment No. 290 used by ANP-1 0307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors (i.e.,

TS 5.6.5.b.1 1) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.

Brunswick Unit 2 App. B-3 Amendment No. 290 1

BSEP 15-0007 Enclosure 4 Typed Technical Specification and Operating License Pages - Unit 1

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.

21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Brunswick Unit 1 5.0-22 Amendment No. xxx I

Amendment Additional Conditions Implementation Number Date 262 The fuel channel bow standard deviation Upon implementation of component of the channel bow model Amendment No. 262.

uncertainty used by ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors (i.e., TS 5.6.5.b.11) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.

Brunswick Unit 1 App. B-2 Amendment No. xxx I

BSEP 15-0007 Enclosure 5 Typed Technical Specification and Operating License Pages - Unit 2

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.

21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Brunswick Unit 2 5.0-22 Amendment No. xxx I

Amendment Number Additional Conditions Implementation Date 276 Upon implementation of Amendment No. 276 As described in adopting TSTF-448, Revision 3, the determination paragraphs (a), (b),

of control room envelope (CRE) unfiltered air and (c) of this inleakage as required by SR 3.7.3.3, in accordance Additional Condition.

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.

290 The fuel channel bow standard deviation Upon implementation of component of the channel bow model uncertainty Amendment No. 290 used by ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors (i.e.,

TS 5.6.5.b. 11) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.

Brunswick Unit 2 App. B-2 Amendment No. xxx I

BSEP 15-0007 Enclosure 6 List of Regulatory Commitments The following table identifies the actions in this document to which the Brunswick Steam Electric Plant has committed. Statements in this submittal, with the exception of those in the table below, are provided for information purposes and are not considered commitments. Please direct questions regarding these commitments to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

Commitment Completion Date Duke Energy will only use the ANP-1 0298P-A, Revision 1, N/A methodology to perform evaluations of the AREVA ATRIUM 1OXM fuel design and to evaluate the performance of co-resident fuel in mixed cores as discussed in the Safety Evaluation associated with NRC approval of ANP-10298P-A, Revision 1 (i.e., ADAMS Accession Number ML14072A355).