ML16019A029

Issuance of Amendments Regarding the Adoption of Topical Report ANP-10298P-A, Revision 1

ML16019A029
Person / Time
Site:	Brunswick
Issue date:	02/09/2016
From:	Andrew Hon Plant Licensing Branch II
To:	William Gideon Duke Energy Progress
Hon A, NRR/DORL/LPL2-2
References
CAC MF5851, CAC MF5852
Download: ML16019A029 (25)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. William R. Gideon Site Vice President Brunswick Steam Electric Plant 8470 River Road, SE M/C BNP001 Southport, NC 28461 February 9, 2016

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING THE ADOPTION OF TOPICAL REPORT ANP-10298P-A, REVISION 1 (CAC NOS. MF5851 AND MF5852)

Dear Mr. Gideon:

The Commission has issued the enclosed Amendment Nos. 269 and 297 to Renewed Facility Operating License (RFOL) Nos. DPR-71 and DPR-62, respectively, for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2. These amendments are in response to your application dated February 19, 2015, as supplemented by letter dated November 5, 2015.

Duke Energy Progress, Inc. (Duke Energy, the licensee) requested an amendment to the Appendix A Technical Specifications (TSs), and to Appendix B, "Additional Conditions," of the BSEP RFOLs. The amendments revise TS 5.6.5.b related to the analytical methods used to determine core operating power limits.

The amendments to the TSs replace U.S. Nuclear Regulatory Commission (NRC)-approved Revision 0 of the AREVA critical power ratio (CPR) correlation Topical Report (TR)

(ANP-10298PA, "ACE/ATRIUM 10XM Critical Power Correlation," March 2010) with NRC-approved Revision 1 of the AREVA CPR correlation TR (ANP-10298P-A, "ACE/ATRIUM 10XM Critical Power Correlation," March 2014). Revision 1 of the AREVA TR updated and resolved inappropriate calculation assumptions of Revision 0 regarding a modelling parameter that characterizes the effect on CPR of radial fuel rod peaking distribution within a fuel bundle (K-factor). Adoption of the improved analytical methods accepted through Revision 1 of the AREVA TR obviated the need to maintain the Appendix B license condition that was issued as part of Amendment Nos. 262 and 290 for Units 1 and 2, respectively. This amendment removes the portions of the license conditions pertaining to the evaluation of the safety limit minimum CPR, setpoint, and core operating limit values.

The NRC staff has completed its review of the above information provided by the licensee and approved the request based on the enclosed safety evaluation (SE). The NRC staff has determined that its documented SE does not contain sensitive security-related information pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 2.390, "Public inspections, exemptions, requests for withholding." However, the NRC will delay placing the enclosed SE in the public document room for a period of 10 working days from the date of this letter to provide Duke Energy with the opportunity to comment on any sensitive aspects. If you believe that any information in the SE contains sensitive information, please identify such information line-by-line and define the basis pursuant to the criteria of 1 O CFR 2.390. After 10 working days, the enclosed SE will be made publicly available unless we hear from you.

Notice of issuance will be included in the Commission's biweekly Federal Register notice.

Docket Nos. 50-325 and 50-324

Enclosures:

1. Amendment No. 269 to DPR-71

2. Amendment No. 297 to DPR-62

3. Safety Evaluation cc w/enclosures: Distribution via Listserv Sincerely,

../

./

~

Andrew Hon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS, INC.

DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 269 Renewed License No. DPR-71

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Progress, Inc., dated February 19, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-71 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 269, are hereby incorporated in the license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Technical Specifications.

3.

Once approved, the amendment shall be implemented prior to start-up from the 2016 Unit 1 refueling outage.

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 9, 201 6 FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 269 RENEWED FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace pages 6 and 10 of Renewed Facility Operating License No. DPR-71 with the attached revised pages 6 and 10.

Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove 5.0-22 Insert 5.0-22 Replace the following page of Appendix B, "Additional Conditions," with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove App. B-2 Insert App. B-2 (c)

Transition License Conditions

1.

Before achieving full compliance with 1 O CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.

2.

The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 1 O CFR 50.48( c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3.

The licensee shall complete all implementation items, except item 9, listed in LAR Attachment S, Table S-2, "Implementation Items," of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 1801h day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.

B. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 269, are hereby incorporated in the license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 203 to Renewed Facility Operating License DPR-71, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 203. For SRs that existed prior to Amendment 203, including SRs with modified acceptance criteria and SRs whose frequency of Renewed License No. DPR-71 Amendment No. 269

3.

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 269, are hereby incorporated into this license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Unit 1 - Technical Specifications -Appendices A and B Date of Issuance: June 26, 2006 Renewed License No. DPR-71 Amendment No. 269

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 SOM, 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. 269

Amendment Number 262 Brunswick Unit 1 Additional Conditions The fuel channel bow standard deviation component of the channel bow model 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.

App. B-2 Implementation Date Upon implementation of Amendment No. 262.

Amendment No. 269

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS, INC.

DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 297 Renewed License No. DPR-62

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Progress, Inc., dated February 19, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-62 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 297, are hereby incorporated in the license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Technical Specifications.

3.

Once approved, the amendment shall be implemented prior to start-up from the 2017 Unit 2 refueling outage.

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 9, 2016 FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 297 RENEWED FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace pages 6 and 10 of Renewed Facility Operating License No. DPR-62 with the attached pages 6 and 10.

Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove 5.0-22 Insert 5.0-22 Replace the following page of Appendix 8, "Additional Conditions," with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove App. 8-2 App. 8-3 Insert App. 8-2 (c)

Transition License Conditions

1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.

2. The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48( c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3. The licensee shall complete all implementation items, except Item 9, listed in LAR Attachment S, Table S-2, "Implementation Items," of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 1801h day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts (thermal).

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 297, are hereby incorporated in the license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 233 to Renewed Facility Operating License DPR-62, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 233. For SRs that existed prior to Amendment 233, Renewed License No. DPR-62 Amendment No. 297 M.

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(1)

Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (2)

Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (3)

Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders N.

The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

3.

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 297, are hereby incorporated into this license. Duke Energy Progress, Inc. shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Unit 2 - Technical Specifications - Appendices A and B Date of Issuance: June 26, 2006 Renewed License No. DPR-62 Amendment No. 297

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 SOM, 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. 297

Amendment Number 276 290 Additional Conditions Upon implementation of Amendment No. 276 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.3.3, in accordance with TS 5.5.13.c.(i), the assessment of CRE habitability as required by Specification 5.5.13.c.(ii),

and the measurement of CRE pressure as required by Specification 5.5.13.d, shall be considered met. Following implementation:

(a)

The first performance of SR 3.7.3.3, in accordance with Specification 5.5.13.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from June 11, 2004, the date of the most recent successful tracer gas test.

(b)

The first performance of the periodic assessment of CRE habitability, Specification 5.5.13.c.(ii), shall be within the next 9 months.

(c)

The first performance of the periodic measurement of CRE pressure, Specification 5.5.13.d, shall be within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test.

The fuel channel bow standard deviation component of the channel bow model 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 2 App. 8-2 Implementation Date As described in paragraphs (a), (b),

and (c) of this Additional Condition.

Upon implementation of Amendment No. 290 Amendment No. 297

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 269 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-71 AND AMENDMENT NO. 297 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-62 DUKE ENERGY PROGRESS. INC.

BRUNSWICK STEAM ELECTRIC PLANT. UNITS 1 AND 2 DOCKET NOS. 50-325 and 50-324

1.0 INTRODUCTION

By letter dated February 19, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15075A021) (Reference 1 ), as supplemented by letter dated November 5, 2015 (ADAMS Accession No. 15329A377), Duke Energy Progress, Inc. (Duke Energy, the licensee) submitted a license amendment request (LAR) for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2. The proposed amendments would revise the technical specifications (TSs) for the analytical methods to be used to determine the reactor core critical power limits. BSEP, Units 1 and 2, are General Electric boiling water reactors (BWR 4) with Mark I wet containments at a licensed power level of 2,923 megawatts thermal (MWt).

1. 1 Proposed Changes The license amendment requests proposed changes to TS 5.6.5.b, "CORE OPERATING LIMITS REPORT (COLR)." These changes are related to the correlation for determination of the critical power ratio (CPR) and can be summarized as follows:

The change to the Appendix A Technical Specifications would supersede Revision 0 of the AREVA CPR correlation (ANP-10298PA, "ACE/ATRIUM 10XM Critical Power Correlation," Revision 0, March 2010) for calculation of core CPR with Revision 1 of the AREVA CPR correlation (ANP-10298P-A, "ACE/ATRIUM 10XM Critical Power Correlation," Revision 1, March 2014). Note: Both revisions of these CPR correlations are U.S. Nuclear Regulatory Commission (NRC or the Commission)-reviewed and approved for Revision 0 (ADAMS Accession Nos. ML100670523 and ML100670225 (public); ML100670567 (non-public)) and Revision 1 (ADAMS Accession Nos. ML14072A355 (public) and ML14072A360 (non-public)).

The change to Appendix B eliminates the additional conditions that were required for Revision 0 of the correlation but are not required for Revision 1 of the AREVA correlation due to the improvement made to Revision 1 of the AREVA CPR correlation.

This safety evaluation (SE) documents the basis for the NRC staff's determination that the proposed amendments are acceptable.

2.0 REGULATORY EVALUATION

The Atomic Energy Commission issued Facility Operating License No. DPR 62 for BSEP, Unit 2, December 27, 1974, and Facility Operating License No. DPR-71 for BSEP, Unit 1, September 8, 1976. The BSEP Updated Final Safety Analysis Report (UFSAR) states that the General Design Criteria (GDC) for Nuclear Power Plants listed in Part 50 to Title 10 of the Code of Federal Regulations (10 CFR 50), Appendix A, as amended July 7, 1971, were used as the basis for an audit of the design features. The plant GDC is discussed in Section 3 of the UFSAR, "Conformance with Nuclear Regulatory Commission General Design Criteria."

The regulatory requirements and guidance documents that the NRC staff used to review the application are as follows:

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 4.2, "Fuel System Design," and Section 4.4, "Thermal and Hydraulic Design." These sections provide guidance to ensure conformance with GDC 10 and GDC 12 by satisfying the required thermal margins (departure from nucleate boiling ratio (DNBR) for pressurized-water reactors and CPR for boiling-water reactors) to protect fuel integrity and preserve major barriers to the release of radioactivity in normal operation, operational transients, and design-basis accidents.

GDC 10, "Reactor design." 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 (AOOs).

GDC 12, "Suppression of reactor power oscillations." The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed. (Note: To comply with GDC 12, the scram limits are established using methodologies established to assure that a scram is initiated prior to reaching the minimum critical power ratio (MCPR) safety limit.)

10 CFR 50.36(c)(5), "Technical specifications," contains many categories, including administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting, necessary to assure operation of the facility in a safe manner. Each licensee is required to submit reports to the Commission pursuant to approved TSs. The TSs, in turn, 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, the TSs require the analytical methods used to determine the core operating limits to be those previously reviewed and approved by the NRC.

3.0 BACKGROUND

The fuel rod cladding, an important barrier to the release of radioactivity, is protected during steady state and anticipated operational occurrences. Thermal protection of the cladding is obtained by ensuring sufficient coolant flows over the fuel rods and by maintaining the required margin to the onset of transition boiling, especially for the peak fuel rods in the thermally limiting fuel bundle. This assurance is built into the fuel cycle design by the use of an NRG-approved critical power correlation and methodology. Critical power correlations generally use a modelling parameter to account for factors such as radial power distribution within a fuel bundle, lattice nuclear design, average exposure, and void fraction. The structure of the CPR and the related modelling parameter depend on the fuel vendor. For AREVA fuel, the methodology and modelling parameter are called ACE and the K-factor. For General Electric Hitachi fuel, the CPR correlation is GEXL, and the modelling parameter is referred to as the R-factor.

Since the AREVA CPR correlation depends on the fuel assembly design, and there are several AREVA assembly designs (e.g., ATRIUM 9B, ATRIUM 10XP, and ATRIUM 10XM), for clarification, ACE in the AREVA correlation is followed by the type of the fuel assembly design.

ACE/ATRIUM 10 and ACE/ATRIUM 10XM refer to the AREVA CPR for ATRIUM 10 and ATRIUM 1 OXM assembly designs, respectively.

The complexity of CPR correlations requires a massive amount of input data for determination of the dryout location in the core and the CPR value. Furthermore, the solution method for CPR requires a multi-step iteration. The CPR solution must also account for the effects of fuel rod power distributions, fuel bundle and fuel channel geometry, grid spacer, and lattice characteristics on the fuel bundle critical power. These effects are culminated into the modeling parameter referred to as the R-factor for GE GEXL and the K-factor for ACE/ATRIUM correlations (mentioned above). Since the ATRIUM 1 OXM assembly design introduces the variable spacer pitch, the impact of local spacer effects and assembly geometry on critical power are accounted for by two different sets of parameters. The first set is a series of constants, one for each rod in the assembly called the additive constants, and another set to provide for modeling of a design feature specific to the ACE/ATRIUM 10XM bundle.

The modeling parameter K-factor is related to the peaking factors of the rod of interest and its immediate neighbors. This approach is consistent with an industrywide practice of using a single axially averaged K-factor (or R-factor) value. However, it turned out that this assumption was inappropriate as (a) it allows downstream conditions above the location of dryout to non-physically influence the critical power, and (b) it provides equal weighting to all axial locations (low power regions, as well as regions far from the location of dryout). Both of these problems were found to be capable of influencing the predicted results in a non-conservative manner. These errors are compensated for by current licensing conditions in Appendix B, "Additional Conditions," which were implemented when Revision 0 of the ACE/ATRIUM 10XM correlation was approved for use in March 2010. These errors are now corrected in Revision 1 of the ACE/ATRIUM 1 OXM correlation, and thus, obviate the need for the associated license conditions. The correction of the errors reduces the CPR margin by a small amount as demonstrated later in this section.

4.0 TECHNICAL EVALUATION

4.1 Update to TS 5.6.5.b to Adopt ANP-10298P-A "ACE/ATRIUM 10XM Critical Power Correlation." Revision 1. March 2014 While approving AREVA's TR ANP-10298P-A, Revision 1, methodology for generic application, the NRC staff also identified two limitations and conditions (L&Cs) for plant-specific applications:

(1) The ACE/ATRIUM 10XM methodology may only be used to perform evaluations of AREVA ATRIUM 10XM fuel design. The ACE/ATRIUM 10XM correlation may also be used to evaluate the performance of the co-resident fuel in mixed cores as discussed in Section 3.4 of this SE.

(2) The ACE/ATRIUM 10XM 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 5.

The NRC staff evaluated this request to ensure both of these L&Cs are satisfied for BSEP plant-specific application of Revision 1 of the ACE/ATRIUM 1 OXM methodology as follows:

(1) L&C 1: The licensee in Reference 1 states that, "Duke Energy will only use the ANP-10298P-A, Revision 1, methodology to perform evaluations of the AREVA A TRI UM 1 OXM fuel design and to evaluate the performance of co-resident fuel in mixed cores as discussed in the licensee's associated Safety Evaluation (i.e., ADAMS Accession Number ML14072A355)." The reporting requirements of TS 5.6.5 for the COLR are revised accordingly in this amendment.

(2) L&C 2: The licensee in Reference 1 states that Revision 1 of the correlation will be not be used outside the range of applicability of Table 2-1 of ANP-10298P-A, Revision 1, which identifies the ACE/ATRIUM 10XM 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 core monitoring system. The restriction on design local peaking is also implemented in AREVA automation tools.

In support of the switch from ACE/ATRIUM 10XM, Revision 0, to ACE/ATRIUM 10XM, Revision 1, the licensee submitted an LAR in Reference 1. To assure compliance with GDC 1 O and GDC 12, as well as both L&Cs, a request for additional information (RAI) was generated (Reference 2), requesting the licensee perform an analysis that (a) selects the limiting AOOs, and (b) applies conservative inputs and assumptions to demonstrate the effect of the new features of the AREVA correlation on the !\\CPR (the reduction in CPR that occurs during a transient event). Of the analyses that were requested in the RAI, the following AOOs were included as acceptable examples: feed-water controller failure, load rejection with no steam bypass to the main condenser (load rejection no bypass (LRNB)) at rated power, pressure regulator failure - closed, and turbine trip with no bypass (TTNB) at rated power. These events lead to reactor pressurization.

In response to the RAI (Reference 4), the licensee identified the following two limiting events that result in reactor pressurization: LRNB and TTNB at rated power. The licensee in Reference 3 states that the basis for selecting these pressurization events is the fact that in the past licensing analyses for end of full power cycle exposure and rated power (2,923 MWt), the TTNB and the LRNB had been the BSEP CPR limiting fast transient events. In addition to these two events, the licensee also included the analysis of a third event, which represents control rod withdrawal error (CRWE) at rated power. As requested in the RAI, the licensee also selected a conservative set of inputs and assumptions consisting of the following:

(a) reactor operating at rated power when the event occurs, (b) vessel flow rate is at 104.5 percent of its rated flow, (c) reactor scram speed is at its technical specification licensing basis exposure, and (d) pressure in the vessel head steam dome, as well as the feedwater temperature, being at their nominal values.

The conservative sets of inputs and assumptions, along with the corrections in the CPR correlation, guarantee the prediction of a very conservative !\\CPR. The licensee's methodology to perform !\\CPR analysis consists of running multiple software packages and computer codes to obtain the required input data and perform the transient analysis. Reference 3 describes this methodology and provides the analysis where the licensee demonstrates that the ACE/ATRIUM 1 OXM correlation is not used outside the range of applicability defined by the range of critical power correlation (the test data prescribed in Table 2-1 of Reference 5), thus satisfying L&C 2.

The NRC staff audited the calculation package supporting the RAI response (Reference 3),

which outlines the sequences for using various computer codes and software packages. Most of these software packages are automated. In summary, the AUTO_HTBAL automates the heat balance calculation to determine vessel pressure, enthalpy, and mass flow rate. The heat balance analysis is performed in steady state and transients. Subsequently, the !\\CPR is calculated, which requires the use of three computer codes: COTRANSA2, XCOBRA, and XCOBRA-T. The COTRANSA2 is a systemwide code for BWR nuclear steam supply system, simulating the reactor system response to specific transients such as LRNB and TTNB. The output data produced by the COTRANSA2 code is used as input in the XCOBRA code to obtain the hot channel flow rate as a function of the radial peaking factor.

The licensee in Reference 4 has summarized the results of the analyses obtained from the audited supporting calculation package, which uses Revision 1 of the ACE/ATRIUM 10XM CPR correlation. As was requested in the RAI, the results are compared with Revision 0 of the correlation to show the trend. The tabulated results of Reference 4 are presented in Table 1 below. Table 1 includes three limiting events with respect to CPR, namely, load rejection with no bypass, turbine trip with no bypass, and control rod withdrawal error. The least limiting event is the control rod withdrawal error at full power, and the most limiting event is turbine trip with no bypass. Correction of the errors in Revision O of the correlation has resulted in a loss of margin to the MCPR safety limit, with the largest loss of such margin corresponding to turbine trip with no bypass. Also, as seen from Table 1, the largest loss of margin is on the order of 4 percent, and the smallest loss of margin on the order of only 1 percent. Therefore, it can be concluded that while the correction of errors leads to a loss of margin in ~CPR, the magnitude of the margin loss for the most limiting event is only 4 percent, which can be viewed as negligible, given the extent of conservatism applied in the input data, assumptions, and analytical methodology.

Table 1 Comparison of ACPRs between Revision 0 and Revision 1 of the ACE/ATRIUM 10XM Critical Power Correlation for BSEP Limiting Events Event t:iCPR t:iCPR Delta ACE Revision 0 ACE Revision 1 Load Reject with No Bypass 0.297 0.288

-0.009 at 100% Rated Power Turbine Trip with No Bypass 0.302 0.290

-0.012 at 100% Rated Power Control Rod Withdrawal Error 0.271 0.262

-0.009 at 100% Rated Power This comparison shows that even in the most limiting transient, a reasonable margin to transition boiling, exists, and the fuel cladding integrity has been maintained via the efficient nucleate boiling mechanism.

Since the application of Revision 1 of the ACE/ATRIUM 10XM correlation has maintained the margin to transition boiling despite the application of conservative methodology using conservative sets of input data, it can be concluded that the GDC 10 requirement for the protection of cladding as an important fission product barrier is maintained. Additionally, the proposed change does not affect the reactor scram to be initiated prior to reaching the minimum critical power ratio safety limit, which satisfies the GDC 12 guideline.

From the audit of the analysis performed to support the licensee's RAI response, the NRG staff confirmed that the licensee has applied Revision 1 of the correlation to ACE/ATRIUM 10XM fuel and co-resident fuel, thus complying with the first NRG-imposed L&C requirement. The review of Table 2-1 of Reference 5 against the data of the Reference 3 analysis shows that the correlation has not been used outside the range of applicability. Hence, the licensee has demonstrated its adherence to the NRC's second L&C requirement.

4.2 Change to Appendix B to Eliminate the Additional Conditions Associated with ANP-10298PA. "ACE/ATRIUM 10XM Critical Power Correlation," Revision 0.

March 2010 Approval of this LAR and the incorporation of TR ANP-10298P-A, Revision 1, updates the analytical methodology used in the determination of core operating limits. Adoption of Revision 1 provides improved K-factor models; resolves the inappropriate ANP-10298PA, Revision 0, K-factor calculation assumptions; and obviates the need to maintain the Appendix B license condition that was issued as part of License Amendment Nos. 262 and 290 for Units 1 and 2, respectively. Accordingly, the licensee's request to remove the portion of the license condition pertaining to the evaluation of setpoints and core operating limit values using the methods described in AREVA Operability Assessment CR 2011-2274, Revision 1, and the evaluation of safety limit minimum critical power ratio (SLMCPR) values using the methods described in ANP-3086(P), Revision 0, "Brunswick Unit 1 and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for ATRIUM 10XM Fuel - Improved K-factor Model," are consistent with the improved methodology and added margin of safety of the proposed amendments.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the North Carolina official was notified on February 5, 2016 of the proposed issuance of the amendments. The state official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on April 28, 2015 (80 FR 23603). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The NRC staff accepts the licensee's proposed LAR as described in Section 1.1 of this SE. The staff finds that the licensee used methods consistent with regulatory requirements and guidance identified in Section 2.0 above. The staff also concluded, based on the considerations discussed above, that: (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) no plant hardware or operational changes are required with this TS change. Therefore, the proposed change is acceptable.

8.0 REFERENCES

1. Duke Energy letter to NRC, "Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Request for License Amendments -Adoption of Topical Report ANP-10298P-A, Revision 1," dated February 19, 2015 (ADAMS Accession No. ML15075A021).

2. NRC e-mail to Duke Energy, "Brunswick Steam Electric Plant, Units 1 and 2 -

Request for Additional Information Related to License Amendment Request -

Adoption of Topical Report ANP-10298P-A, Revision 1," dated September 22, 2015 (ADAMS Accession No. ML15266A530).

3. Regulatory Audit Report Regarding License Amendment Request to Revise Technical Specifications for the Analytical Methods to be Used to Determine the Reactor Core Critical Power Limits for the Brunswick Steam Electric Plant, Units 1 and 2, dated February 4, 2016 (ADAMS Accession No. ML16034A508).

4. Duke Energy letter to NRC, "Brunswick Steam Electric Plant, Units Nos. 1 and 2, Response to Request for Additional Information Regarding License Amendment Request to Adopt Topical Report ANP-10298P-A, Revision 1," dated November 5, 2015 (ADAMS Accession No. ML15329A377).

5. NRC letter to AREVA NP Inc., "Verification Letter of the Approval Version of AREVA Inc. Topical Report ANP-10298P-A, Revision 1, 'ACE/ATRIUM 1 OXM Critical Power Correlation,'" dated November 24, 2014 (ADAMS Accession No. ML14321A129).

Principal Contributor: Fred Forsaty Date: February 9, 2016

The NRC staff has completed its review of the above information provided by the licensee and approved the request based on the enclosed safety evaluation (SE). The NRC staff has determined that its documented SE does not contain sensitive security-related information pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 2.390, "Public inspections, exemptions, requests for withholding." However, the NRC will delay placing the enclosed SE in the public document room for a period of 10 working days from the date of this letter to provide Duke Energy with the opportunity to comment on any sensitive aspects. If you believe that any information in the SE contains sensitive information, please identify such information line-by-line and define the basis pursuant to the criteria of 10 CFR 2.390. After 10 working days, the enclosed SE will be made publicly available unless we hear from you.

Notice of issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA/

Andrew Hon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324

Enclosures:

1. Amendment No. 269 to DPR-71

2. Amendment No. 297 to DPR-62

3. Safety Evaluation cc w/enclosures: Distribution via Listserv DISTRIBUTION:

PUBLIC: (No PDC/Listserv for 10 working days)

LPL2-2 R/F RecordsAmend RidsNrrLABClayton Resource RidsACRS_MailCTR Resource RidsNsirDpr Resource RidsNrrDorlDpr Resource RidsNrrPMBrunswick Resource RidsNrrDorllpl2-2 Resource DJohnson, NSIR RidsNrrDssSrxb Resource RidsRgn2MailCenter Resource RidsNrrDssStsb Resource FForsaty, NRR ADAMS A ccess1on N ML16019A029 o.:

• b

• 1 1y e-ma1 OFFICE DORL/LPLll-2/PM DORL/LPLll-2/LA NRR/DSS/SRXB/BC*

NRR/DSS/STSB/BC NAME AH on (LRonewicz for) BClayton CJackson RElliott DATE 2/02/2016 2/03/2016 1/07/2016 2/02/2016 OFFICE NRR/DSS/SNPB/BC OGC - NLO w/edits NRR/DORL/LPLll-2/BC NRR/DORL/LPLll-2/PM NAME JDean BHarris BBeasley AH on DATE 2/04/2016 2/08/2016 2/09/2016 2/09/2016 OFFICIAL RECORD COPY