ML111010234

Issuance of Amendments Regarding Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5

ML111010234
Person / Time
Site:	Brunswick
Issue date:	04/08/2011
From:	Farideh Saba Plant Licensing Branch II
To:	Annacone M Carolina Power & Light Co
Saba F, NRR/DORL/LPL2-2, 301-415-1447
References
TAC ME3856, TAC ME3857
Download: ML111010234 (44)

Text

OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 8, 2011 Mr. Michael J. Annacone, Vice President Brunswick Steam Electric Plant Carolina Power & Light Company Post Office Box 10429 Southport, North Carolina 28461 SUB~IECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING ADDITION OF ANALYTICAL METHODOLOGY TOPICAL REPORT TO TECHNICAL SPECIFICATION 5.6.5 (TAC NOS. ME3856 AND ME3857)

Dear Mr. Annacone:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 257 to Renewed Facility Operating License No. DPR-71 and Amendment No. 285 to Renewed Facility Operating License No. DPR-62 for Brunswick Steam Electric Plant (BSEP), Units 1 and 2, respectively. The amendments are in response to your application dated April 29, 2010, as supplemented by letters dated June 9, July 22, July 29, September 29, October 12, November 9, November 18, and December 16, 2010, and April 6, 2011. The amendments revise the BSEP, Units 1 and 2 Technical Specification 5.6.5.b by adding the AREVA topical report, ANP-10298PA, "ACE/ATRIUM 10XM Critical Power Correlation," Revision 0, March 2010, to the list of analytical methods that have been reviewed and approved by the NRC for determining core operating limits. The amendments change the BSEP, Units 1 and 2 Technical Specifications to support transition to ATRIUM 10XM fuel and associated core design methodologies.

The NRC staff reviewed the licensee's application for amendment and the licensee's submitted supplements. In addition, the NRC staff reviewed the licensee's calculations and documents supporting the proposed amendments during an audit at the AREVA's offices located in Bethesda, Maryland. As a result of addition of an alternate method to the approved topical report for ACEA TRIUM 10XM methodology, the licensee submitted a letter dated April 6, 2011,

which included the licensee's submittal of AREVA's operability assessment for BSEP, Unit 2, Cycle 20 (Condition Report (CR) 2011-2274) and resulted into addition of a license condition in Appendix B, "Additional Conditions" of the operating licenses for BSEP, Units 1 and 2. This license condition is described in Section 4 of the NRC staffs safety evaluation (SE).

The NRC staff has determined that its documented SE contains proprietary information pursuant to Title 10 of the Code ofFederal Regulations (10 CFR), Section 2.390, "Public inspections, exemptions request for withholding." Accordingly, the NRC staff has prepared a redacted, nonproprietary version. However, we will delay placing the nonproprietary SE in the public document room for a period of 10 working days from the date of this letter to provide you with the Document transmitted herewith contains sensitive unclassified information. When separated from Enclosure 4, this document is decontrolled.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY* PROPRIETARY INFORMATION M. Annacone

-2 opportunity to comment on any proprietary aspects. If you believe that any information in the enclosure is proprietary. 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 nonproprietary SE will be made publicly available. Copies of the proprietary and nonproprietary versions of the SE are enclosed.

A notice of issuance will be included in the NRC's biweekly Federal Register notice.

Sincerely, Farideh E. Saba, Senior 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. 257 to License No. DPR-71

2. Amendment No. 285 to License No. DPR-62

3. Safety Evaluation (Nonproprietary Information)

4. Safety Evaluation (Proprietary Information) cc w/enclosures 1, 2, 3, and 4: Addressee cc w/enclosures 1, 2, and 3: Distribution via ListServ OFFICIAL USE ONLY PROPRIETARY INFORMATION

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 257 Renewed License No. DPR-71

1.

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

A.

The application for amendment filed by Carolina Power &Light Company (the licensee), April 29, 2010, as supplemented by letters dated June 9, July 22, July 29, September 29, October 12, November 9, November 18, and December 16, 2010, and April 6, 2011, 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 Title 10 of the Code ofFederal Regulations (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:

The Technical Specifications contained in Appendix A. as revised through

\\

Amendment No. 257, are hereby incorporated in the license. Carolina Power &

Light Company shall operate the facility in accordance with the Technical Specifications.

3.

The license is also amended by changes to the Additional Conditions contained in Appendix B, as indicated in the attachment to this license amendment; and Section 3 of Renewed Facility Operating License No. DPR-71 is hereby amended to read as follows:

3.

Additional Conditions The Additional Conditions contained in Appendix B. as revised through amendment No. 257. are hereby incorporated into this license. Carolina Power &

Light Company shall operate the facility in accordance with the Additional Conditions.

4.

This license amendment is effective as of the date of its issuance and shall be implemented prior to start-up from the 2012 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

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Douglas. Broaddus. Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License.

Technical Specifications and Additional Conditions Date of Issuance: April 8, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 257 RENEWED FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following pages of Renewed Operating License DPR-71 with the attached revise pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages 4

4 8

8 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 Page Insert Page 5.0-22 5.0-22 Replace the following page of the 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 Page Insert Page B-2 B-2

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 257, are hereby incorporated in the license. Carolina Power &Light Company 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 performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 203.

(a)

Effective June 30, 1982, the surveillance requirements listed below need not be completed until July 15, 1982. Upon accomplishment of the surveillances, the provisions of Technical SpeCification 4.0.2 shall apply.

Specification 4.3.3.1, Table 4.3.3-1. Items 5.a and 5.b (b)

Effective July 1,1982, through July 8,1982, Action statement "a" of Technical Specification 3.8.1.1 shall read as follows:

ACTION:

a. With either one offsite circuit or one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.A. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4 within two hours and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter; restore at least two offsite circuits and four diesel generators to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(3)

Deleted by Amendment No. 206.

D.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Physical Security Plan, Revision 2,u and "Safeguards Contingency Plan, ReVision 2," submitted by letter dated May 17, 2006, and "Guard Training and Qualification Plan, Revision 0,"

submitted by letter dated September 30,2004.

Renewed License No. DPR-71 Amendment No. 257

3.

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 257, are hereby incorporated into this license. Carolina Power & Light Company shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION IRAJ 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. 257 1

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

20.

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

21.

ANP-10298PA, ACE/ATRIUM 10XM Critical Power Correlation, Revision 0, March 2010.

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 Amendment No. 257 I

257 Amendment Number Additional Conditions Safety Limit Minimum Critical Power Ratio (SlMCPR), setpoint, and core operating limit values determined using the ANP-10298PA, ACE/ATRIUM 10XM Critical Power Correlation (Le., TS 5.6.5.b.21), shall be evaluated with methods described in AREVA Operability Assessment CR 2011-2274, Revision 1 to verify the values determined using the NRC-approved method remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment C R 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.

Implementation Date Upon implementation of Amendment No. 257.

Brunswick Unit 1 App. B-2 Amendment No. 257

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 285 Renewed License No. DPR-62

1.

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

A.

The application for amendment filed by Carolina Power & Light Company (the licensee), April 29, 2010, as supplemented by letters dated June 9, July 22, July 29, September 29, October 12, November 9, November 18, and December 16, 2010, and April 6, 2011, 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 Title 10 of the Code of Federal Regulations (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.

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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 fonows:

The Technical Specifications contained in Appendix A. as revised through Amendment No. 285. are hereby incorporated in the license. Carolina Power &

light Company shall operate the facility in accordance with the Technical Specifications.

3.

The license is also amended by changes to the Additional Conditions contained in Appendix Bf as indicated in the attachment to this license amendment; and Section 3 of Renewed Facility Operating license No. DPR-62 is hereby amended to read as follows:

3.

Additional Conditions The Additional Conditions contained in Appendix B. as revised through amendment No. 285, are hereby incorporated into this license. Carolina Power &

light Company shall operate the facility in accordance with the Additional Conditions.

4.

This license amendment is effective as of the date of its issuance and shall be implemented prior to start-up from the 2011 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

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Douglas A. Broaddus, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License.

Technical Specifications and Additional Conditions Date of Issuance: April 8, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 285 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following pages of Renewed Operating License DPR-62 with the attached revise pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages 3

3 8

8 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 Page Insert Page 5.0-22 5.0-22 Replace the following page of the 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 Page Insert Page B-2 B-2

as sealed neutron sources for reactor startup. sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source, and special nuclear materials without restriction to chemical of physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)

Pursuant to the Act and 10 CFR Parts 30 and 70 to posses, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Brunswick Steam Electric Plant, Unit Nos. 1 and 2, and H. B. Robinson Steam Electric Plant, Unit No.2 (6)

Carolina Power &Light Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Report dated November 22, 1977, as supplemented April 1979, June 11, 1980, December 30, 1986, December 6, 1989, July 28, 1993, and February 10, 1994 respectively, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

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. 285, are hereby incorporated in the license.

Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.

Renewed License No. DPR-62 Amendment No. 285

3.

Additional Conditions The Additional Conditions contained in Appendix 8, as revised through Amendment No. 285, are hereby incorporated into this license. Carolina Power &Light Company shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION IRAJ J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Unit 2 - Technical Specifications - Appendices A and 8 Date of Issuance: June 26,2006 Renewed License No. DPR-62 Amendment No. 285 I

5.6 Reporting Requirements 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.

ANPw10298PA, ACE/ATRIUM 10XM Critical Power Correlation, Revision 0, March 2010.

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. 285

276 Amendment Number Additional Conditions Implementation Date 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.

285 Safety Limit Minimum Critical Power Ratio Upon implementation of (SLMCPR). setpoint, and core operating limit Amendment No. 285.

values determined using the ANP-10298PA, ACE/ATRIUM 10XM Critical Power Correlation (i.e., TS 5.6.5.b.21), shall be evaluated with methods described in AREVA Operability Assessment CR 2011-2274, Revision 1 to verify the values determined using the NRC-approved method remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. The results of the evaluation shall be documented and submitted to the NRC, for review, at least 60 days prior to startup of each operating cycle.

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

OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 257 AND 285 TO RENEWED FACILITY OPERATING LICENSES NOS. DPR-71 AND DPR-62 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324

1.0 INTRODUCTION

By application dated April 29, 2010 (Reference 1), as supplemented by letters dated June 9, July 22, July 29, September 29, October 12, November 9, November 18, and December 16, 2010, and April 6, 2011 (References 2 through 10, and Reference 39), Carolina Power & Light Company (CP&L, the licensee), requested license amendments to revise the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2.

The proposed license amendments will revise BSEP TS 5.6.5.b by adding the AREVA topical report, ANP-10298PA, "ACE/ATRIUM 10XM Critical Power Correlation," Revision 0, March 2010 (Reference 11), to the list of analytical methods that have been reviewed and approved by the Nuclear Regulatory Commission (NRC) for determining core operating limits. The proposed amendments would change the BSEP TSs to support transition to ATRIUM 1 OXM fuel and associated core design methodologies.

BSEP Units 1 and 2 are General Electric (GE) boiling water reactors (BWRs) BWRl4 design.

Both BSEP units are currently operating at 120 percent of originally licensed thermal power at extended power uprate and maximum extended load line limit analysis conditions. Each unit reactor core is composed of 560 fuel assemblies. In each cycle, approximately 40 percent of the irradiated fuel assemblies are replaced with new fuel. On March 27,2008, the NRC issued Amendments Nos. 246 and 274 for transition from GE fuel to A TRIUM-1 0 fuel for BSEP Units 1 and 2, respectively (Reference 12). The first transition from GE fuel to ATRIUM-10 fuel occurred in spring 2008 and spring 2009, when the licensee loaded 248 and 238 fresh ATRIUM 10 fuel assemblies in BSEP Units 1 and 2, respectively.

The licensee plans to implement this amendment loading the ATRIUM 10XM fuel design in the BSEP Unit 2 during the spring 2011 refueling outage, beginning with Cycle 20. The BSEP Unit 1 amendment supporting transitioning to ATRIUM 10XM will be implemented during the spring 2012 refueling outage, beginning with Cycle 19.

The supplements dated June 9, July 29, September 29, October 12, November 9, November 18, and December 16, 2010, and April 6, 2011, provided additional information that clarified the application, did not expand the scope of the original Federal Register notice, did not change the OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 2 NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 10, 2010 (75 FR 48372), and did not expand the scope of the original license amendment request (LAR).

2.0 REGULATORY EVALUATION

2.1 Background

The use of the ATRIUM 10XM fuel design requires the addition of the ACE/ATRIUM 10XM Critical Power Correlation, ACE/ATRIUM methodology, to the list of analytical methods specified in BSEP TS 5.6.5.b and reviewed and approved by the NRC for determining core operating limits. Topical report ANP-10298PA (Reference 11) describes a new correlation (ACE/ATRIUM 10XM) developed by AREVA to predict the critical power for BWRs to ensure that reactors using ATRIUM 10XM fuel remain within required safety limits during steady-state operation and anticipated operational occurrences (AOOs).

By letter dated April 15, 2010, AREVA provided to the NRC the results of its evaluations performed for the ATRIUM 10XM fuel design to demonstrate how the NRC-approved fuel licensing criteria defined in ANF-89-98(P)(A), Revision 1, Supplement 1 (Reference 13) are met. The AREVA's evaluation results are presented in ANP-2899(P), Revision 0 (Reference 14) provided as an enclosure to the April 15, 2010, letter. The NRC staff has determined that, in accordance with the process described in ANF-89-98(P)(A}, Revision 1, Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," the ATRIUM 10XM new fuel design has satisfied the ANF-89-98 (P)(A) acceptance criteria.

The licensee also requested another amendment (Reference 15), in support of BSEOP, Units 1 and 2 cores transmittal to ATRIUM 10XM, that would add, if approved, BAW-10247PA, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," Revision 0, April 2008 to the list of analytical methods specified in BSEP TS 5.6.5.b. The NRC-approved topical report, BAW-10247PA, Revision 0 (Reference 16), "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," describes a fuel performance code, RODEX4, and best-estimate thermal-mechanical evaluation methodology for fuel rods of BWRs. The NRC staffs determination on this amendment request will be documented in a separate safety evaluation.

2.2 Regulatorv Requirements and Guidance Documents The NRC staff reviewed the LAR to evaluate the applicability of the ACE/ATRIUM 10XM Critical Power Correlation to the BSEP TSs, confirm that the use of this methodology is within NRC-approved ranges of applicability, and verify that the results of the analyses are in compliance with the requirements of the following General Design Criteria (GDC) specified in Appendix A to Title 10 of the Code ofFederal Regulations (10 CFR), Part 50:

• GDC-10, "Reactor design," requiring the reactor design (reactor core, reactor coolant system (RCS), control and protection systems) to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including anticipated operational occurrences (AOOs).

• GDC-12, "Suppression of reactor power oscillations," requiring that power oscillations that can result in conditions exceedillg specified acceptable fuel design limits are not possible, or can be reliably and readily detected and suppressed.

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- 3 GDC-15, "Reactor coolant system design," requiring the RCS and associated auxiliary, control, and protection systems to be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including ADOs.

GDC-20, "Protection system functions," requiring the protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of ADOs and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

GDC-25, "Protection system requirements for reactivity control malfunctions," requiring the protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

GDC-26, "Reactivity control system redundancy and capability," requiring two independent reactivity control systems of different design principles be provided, one of which is capable of holding the reactor subcritical under cold conditions.

GDC-27, "Combined reactivity control system capability," requiring the reactivity control systems to be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system (ECCS), of reliably controlling reactivity changes under postulated accident conditions.

GDC-28, "Reactivity limits," requiring the reactivity control systems to be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding, nor (2) sufficiently disturb the core, its support structures, or other reactor pressure vessel internals to impair significantly the capability to cool the core.

GDC-35, "Emergency core cooling, II requiring a system to provide abundant emergency core cooling to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (2) clad metal-water reaction is limited to negligible amounts.

The Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP, NUREG-OBOO), Section 4.2, "Fuel System Design," provides regulatory guidance for the review of fuel rod cladding materials and the fuel system. In addition, the SRP provides guidance for compliance with the applicable GDC in Appendix A to 10 CFR Part 50. According to SRP Section 4.2, the fuel system safety review provides assurance that:

The fuel system is not damaged as a result of normal operation and ADOs, Fuel system damage is never so severe as to prevent control rod insertion when it is

required, The number of fuel rod failures is not underestimated for postulated accidents, and Coolability is always maintained.

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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 staff on a generic basis. The NRC staff 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.

A generic topical report describing a methodology or computer code does not provide the full justification for each plant-specific application. In situations where a plant-specific LAR references a topical report that has not been previously applied, the licensee submits a plant-specific analysis to demonstrate applicability of the topical report.

The licensee, in its application dated April 29, 2010 (Reference 1), requested changes to the TSs to support the addition of the ACE/ATRIUM 10XM Critical Power Correlation to the list of analytical methods specified in BSEP Units 1 and 2 TS 5.6.5.b for determining core operating limits at BSEP. By References 2 through 8, the licensee submitted information to demonstrate compliance with the staff limitations and conditions imposed for application of the ACE/ATRIUM methodology, and to demonstrate the applicability of the AREVA codes and methods for BSEP Unit 2 at extended power uprate conditions.

The BSEP Units 1 and 2 cores contain 560 fuel assemblies that consist of both GE14 and ATRIUM-10 assemblies. The licensee plans to load the ATRIUM 10XM fuel design in BSEP Units 1 and 2 in spring of 2012 and 2011, respectively. The proposed Cycle 20 core for BSEP Unit 2 will consist of 226 fresh ATRIUM 10XM assemblies, 238 irradiated ATRIUM-10 assemblies, and 96 irradiated GE14 assemblies.

For each operating cycle, the core design evaluation ensures that greater than 99.9 percent of the fuel rods in the reactor core avoid boiling transition (BT) during plant operation, if the safety limit is not exceeded. Thermal margin performance calculations for each type of fuel design in the core are determined using an applicable critical power ratio (CPR) correlation. The safety limit minimum critical power ratio (SLMCPR) is defined as the minimum value of CPR which ensures that less than 0.1 percent of the fuel rods in the core are expected to experience BT during normal operation or an AOO.

Currently, reactor core critical power limits are determined using the NRC-approved analytical methodologies described in topical reports: EMF-2209(P)(A), "SPCB Critical Power Correlation" (Reference 17); EMF-2245(P)(A), "Application of Siemens Power Corporation's ANF Critical Power Correlations to Co-Resident Fuel" (Reference 18); and ANF-524(P)(A}, "ANF Critical Power Methodology for Boiling Water Reactors" (Reference 19). Approval of this LAR and the incorporation of topical report, ANP-1 0298PA (Reference 11) will enable CP&L to implement the analytical methods described in this report, and in conformance with the limitations and conditions described in the topical report and its associated NRC safety evaluation.

As a result of addition of an alternate method to the approved topical report for ACE/ATRIUM 10XM methodology (Reference 16), the licensee submitted a letter dated April 6, 2011 (Reference 39), which included the licensee's submittal of AREVA's operability assessment for BSEP, Unit 2, Cycle 20 (Condition Report (CR) 2011-2274), the licensee's responses to the NRC staffs RAls regarding this CR, and addition of a license condition for the ACE/ATRIUM 10XM requested amendments (Reference 1).

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- 5 The NRC staff has reviewed the LAR (Reference 1) in conjunction with the supplemental letters (References 2 through 8), the responses to the staffs requests for additional information (RAls)

(References 9, and 10), and the licensee's submittal dated April 6, 2011 (Reference 39) to (1) evaluate the acceptability of the BSEP transition to ATRIUM-1 OXM fuel, (2) evaluate the use of the associated AREVA methodologies for licensing applications, and (3) confirm there are adequate technical basis for the proposed TS changes. In addition, the staff conducted a regulatory audit at the AREVA office in Bethesda, Maryland on November 3,4 and 5, 2010, and reviewed the BSEP-specific safety analyses, calculation notebooks and associated fuel transition methodologies.

3.1 AREVA Methodologies and Computer Codes As indicated in the license amendment request (Reference 1), the licensee performed licensing analyses using a variety of AREVA methodologies and computer codes as described below.

The NRC staff evaluated the applicability of these codes and methods specifically to BSEP Units 1 and 2.

3.1.1 AREVA Methodologies Critical Power Correlation Methodologies The SLMCPR for all fuel types in the BSEP Unit 2 Cycle 20 core is determined using the methodology described in topical report, ANF-524(P)(A) (Reference 19). This topical report provides the basis for the methodology for determining the operating safety limit for the minimum critical power ratio (MCPR) that ensures that 99.9 percent of the fuel rods are protected from BT during normal operation and AOOs. The methodology consists of a series of Monte Carlo calculations in which the variables affecting the probability of BT are determined for each Monte Carlo trial. The analysis was performed with a power distribution that conservatively represents expected reactor operating states that could both exist at the MCPR operating limit and produce an MCPR equal to the SLMCPR during an AOO.

The BSEP Unit 2 Cycle 20 SLMCPR analysis used the ACE/ATRIUM 10XM critical power correlation additive constants and additive constant uncertainty for the ATRIUM 10XM fuel described in Reference 11. The ACE/ATRIUM 10XM correlation is used to accurately predict assembly critical power for the ATRIUM 10XM fuel design. The correlation provides an accurate prediction of the limiting rod. The impact of local spacer effects and assembly geometry on critical power is accounted for by two different sets of parameters. The first is a set of constants, one constant for each rod in the assembly called additive constants listed in Table 5-2 of Reference 11, and the second a set of parameters that provides for modeling of design-specific axial effects including spacers within the critical power correlation. For comparison of correlation predictions to experimental data, an experimental critical power ratio (ECPR) is defined as the ratio of calculated critical power to the measured critical power. The ECPR distribution associated with the ACE/ATRIUM 10XM correlation is adequately represented with a normal distribution. The range of applicability of the ACE/ATRIUM 10XM correlation is listed in Table 2-1 of Reference 11 and is reproduced below.

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- 6 Table 3.1 Range of Applicability of ACE/ATRIUM-10and ACE/ATRIUM 10XM Correlations

((

))1 The ANF-524(P)(A) methodology is modified slightly for use with the ACE correlation form due to the ((

)). The modifications concern the treatment of channel bow variation along the length of the fuel channel. ((

)) The key difference between the SPCS (Reference 17) and ACE correlations that must be accounted for in the safety limit methodology is ((

)). The impact of the modifications is that they ((

))

((

))

((

1 The information in ((

)) contained proprietary information and as such has been redacted in this nonproprietary version of the SE.

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-7

))

A licensing condition that describes the performance of the supplemental evaluation applicable to ATRIUM 10XM core operating limits is included in Appendix B, "Additional Conditions," of the Operating Licenses for BSEP, Units 1 and 2 (See Section 4 of this safety evaluation). This license condition ensures the limits generated with the NRC-approved methodology appropriately bounds the effects of the K-factor calculation.

The NRC staff identified two limitations and conditions on the use of the ACE/ATRIUM 10XM correlation.

Limitation and Condition 1 states:

Since ACE/ATRIUM 10XM was developed from test assemblies designed to simulate ACE/ATRIUM 10XM fuel, the methodology may only be used to perform evaluations for fuel of that type without further justification.

The licensee will apply ACE/ATRIUM 10XM critical power correlation to ATRIUM 10XM fuel in BSEP, Unit 2 for Cycle 20 and in BSEP Unit 1 for Cycle 19. BSEP, Unit 2. Cycle 20 is expected to consist of 226 fresh ATRIUM 10XM fuel assemblies, 238 once-burned ATRIUM-10 fuel assemblies, and 96 twice-burned GE14 assemblies. The licensee will continue to apply the SPCB correlation to the GE14 fuel design.

The BSEP, Unit 1 core will be loaded with ATRIUM 10XM fuel assemblies for Cycle 19 and will contain the third reload of AREVA fuel; therefore, Unit 1 is not expected to contain any GE14 fuel.

Limitation and Condition 2 states:

ACE/ATRIUM 10XM should not be used outside its range of applicability defined by the range of the test data from which it was developed and the OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 8 additional justifications provided by AREVA in this submittal. This range is listed in Table 2-1 of Reference 1.2 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.

3.1.2 Reactor Analyses Methodology and Computer Codes The NRC approved the use of AREVA fuel ATRIUM-10 and core design methodologies to determine BSEP core operating limits with the issuance of License Amendments 246 and 274 for BSEP, Units 1 and 2, respectively (Reference 12).

XCOBRAlXCOBRA-T (XN-NF-84-105(P)(A)}: XCOBRA predicts the steady-state thermal-hydraulic performance of BWR cores at various operating conditions and power distributions. It is used to evaluate pressure drops, channel and bypass flow distributions, and MCPRs, as well as the hydraulic compatibility of fuel designs. XCOBRA-T predicts the transient thermal-hydraulic performance of BWR cores during postulated system transients and is used to evaluate the change in CPR (l1CPR) for the limiting fuel bundles in the core. As documented in XN-NF-84-105(P)(A) (Reference 20), the XCOBRA-T code has been approved by the NRC for use in BWR licensing applications. The use of the steady-state XCOBRA code has been accepted by the NRC staff (Reference 21) based on approval of XCOBRA-T and the similarity of the thermal-hydraulic models between the codes. The BSEP licensing analysis (Reference 2) shows that the core thermal-hydraulic conditions during steady-state and transient conditions are within the NRC-approved range of the code. Therefore, the NRC staff concludes that the application of XCOBRA and XCOBRA-T for the BSEP core thermal-hydraulic calculations is acceptable.

COTRANSA2 ANF-913(P)(A): COTRANSA2 is a BWR system transient simulation code, which includes an axial one-dimensional neutronics model that captures the effects of axial power shifts associated with the system transients (Reference 22). It is used to evaluate key reactor system parameters during core-wide BWR transient events. These parameters, such as power, flow, pressure, and temperature, are provided as boundary conditions to the hot channel analyses in XCOBRA-T and XCOBRA codes for determining CPRs for limiting transients. As documented in ANF-913(P)(A), the code has been generically approved by the NRC to analyze system responses to fast transients in BWRs (Reference 22).

The NRC approval of COTRANSA2 is subject to the limitations set forth in the safety evaluations for the methodologies described and approved for XCOBRA-T (Reference 20). COTRANSA2 is approved to perform the system analysis of the following fast AOO and anticipated transient without scram (ATWS) events: (1) load rejection without bypass, (2) turbine trip without bypass, (3) feedwater controller failure maximum demand, (4) pressure regulator downscale failure, (5) ATWS main steam isolation valve closure and pressure regulator failure open, and (6) the American Society of Mechanical Engineers (ASME) overpressurization analysis. The NRC staff, upon reviewing the results from the transient analyses, concludes that the licensee's use of COTRANSA2 to perform analYSis of fast transient events for the BSEP is acceptable.

2 Reference 1 in this condition is ANP-10298PA, which is Reference 11 of this safety evaluation.

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- 9 SLMCPR Methodology (ANF-524(P)(A)): The SLMCPR is imposed to protect at least 99.9 percent of the fuel rods in the core from boiling transition during steady-state and transient conditions (Reference 19). The NRC approved the tropical report, ANF-524(P)(A), which identifies the fuel-and nonfuel-related uncertainties and the statistical process used to determine an MCPR safety limit. Compliance to each restriction in the NRC staff's safety evaluation approving topical report, ANF-524(P)(A) is demonstrated in ANP-2637, Revision 3, "Boiling Water Reactor Licensing Methodology Compendium," (Reference 23). As discussed in further detail in Section 3.3 of this safety evaluation, the staff finds that the appropriate values of the MICROBURN-B2 uncertainties are used for the BSEP SLMCPR. Based on the approved CPR correlation, the staff considers the SLMCPR analysis acceptable.

CASMO-4/MICROBURN-B2 (EMF-2158(P)(A>>: The two principal computer programs for BWR nuclear design and analysis used by AREVA are CASMO-4 and MICROBURN-B2. The CASMO-4 code is a two-dimensional multi-group transport theory code used to calculate the lattice physics constants of BWR fuel assemblies. The MICROBURN-B2 code is a two-group nodal code used for the three-dimensional simulation of the nuclear and thermal-hydraulic conditions in BWR cores. The MICROBURN-B2 code determines core-wide nodal neutron flux, fission power, and coolant density distributions; reactivity parameters; nodal exposure and nuclide density distributions; control rod patterns; channel inlet flow distributions; and fuel thermal performance parameters such as linear heat generation rate (LHGR), axial planar LHGR, and CPR. These results are used to design fuel cycles, to assess safety margins, and to monitor operating reactor cores.

Section 5.2.3 of BAW-10247PA (Reference 16) and Section 5 of ANF-524(P)(A) (Reference 19) describe the application of power distribution measurement uncertainties (Le., radial and axial) by these methodologies. The radial and axial power uncertainties are calculated from uncertainty components as described in EMF-2158(P)(A) (Reference 24). Three of the uncertainty components used to calculate these power distribution uncertainties are determined using traversing incore probe (TIP) measurements. These uncertainty components are: (1) the deviation between the CASM0-4/MICROBURN-B2 (C4/MB2) calculated TIP response and the measured TIP response on a radial (~T'ij), nodal (~T'ijk) and planar (~T'planar) bases, (2) TIP measurement uncertainty on a radial (~Tmij), nodal (~Tmijk) and planar (~TmPlanar) bases, and (3) synthesis uncertainty on a radial (~Sij), and nodal (~Sijk) basis. The licensee has shown that all three uncertainty components identified above are bounded by the values reported in Sections 9.4 and 9.5 of EMF-2158(P)(A), and the net calculated TIP distribution uncertainty components

(~Tijk' ~Tij and ~Tplanar) are also bounded by the values reported in Section 9.4 of EMF 2158(P}(A). Details of the calculation process and results of the power distribution uncertainties are discussed in Section 3.3.5 of this safety evaluation.

The licensee has shown compliance with the NRC-approved methodology for power distribution measurement uncertainties and, therefore, the NRC staff has determined that the licensee's use of the EMF-2158(P}(A) methodology is acceptable.

Stability Methodology (NEDO-32465-A): BSEP has implemented the BWR Owners Group long-term solution, Option III as their licensing basis stability protection methodology (Reference 25). To support Option III, the licensee uses the RAMONA5-FA system analysis code to generate the delta over initial CPR versus oscillation magnitude relationship, which provides the CPR performance during reactor instability (Reference 26). RAMONA5-FA is a coupled neutronic thermal-hydraulic three-dimensional transient model for the purpose of determining the relative change in ACPR and the hot channel oscillation magnitude on a plant-specific basis.

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- 10 Backup stability protection (BSP) analyses are performed in anticipation of the long-term Option III solution becoming unavailable and the oscillation power range monitoring system being declared inoperable. The BSP analysis is a prevention approach in which certain areas on the power-to-f1ow map, where instability is likely based on decay ratio calculations, are excluded from operation. The calculations to support BSP are performed using the NRC-approved STAIF code (References 27 and 28). Compliance to each restriction in the NRC's safety evaluation approving use of the STAIF code is demonstrated in' Reference 23.

Therefore, the staff finds that the licensee's use of the STAIF code in support of BSEP licensing application is acceptable.

EXEM BWR-2000 Loss-of-Coolant Accident (LOCAl Methodology (EMF-2361 (P}(A): The AREVA methodology for showing compliance with 10 CFR Part 50, Appendix K, is referred to as the EXEM BWR-2000 evaluation model. This model was reviewed and approved by the NRC staff in Reference 29. The EXEM BWR-2000 methodology employs three primary codes. The reactor system and hot channel response is evaluated with RELAX (Reference 29); fuel assembly heatup during the LOCA is analyzed with HUXY (Reference 30), which incorporates approved cladding swelling and rupture models (Reference 31); and stored energy and fuel characteristics are determined with RODEX2 (Reference 32). Compliance with each restriction in the staff's safety evaluations approving these codes is demonstrated in Reference 23.

Therefore, the staff finds that the licensee's use of the EXEM BWR-2000 methodology and associated code systems in support of the BSEP licensing application is acceptable.

Methods and Codes Summary The licensee evaluated compliance with the restrictions specified in the safety evaluations of the NRC-approved AREVA topical reports that are identified in the LAR (Reference 1). Based on the review of the licensee's evaluation, the staff concludes that the licensee adequately demonstrated conformance to the restrictions and conditions in the identified safety evaluations.

Further, the licensee performed plant-specific analyses of the limiting licenSing basis events with the AREVA codes and methods to demonstrate acceptability of the use of those codes and methods. The analyses show that the results meet the applicable criteria (Sections 3.2 and 3.3 of this safety evaluation). Therefore, the NRC staff concludes that the application of the NRC-approved AREVA codes and methods for the BSEP licensing analysis is acceptable.

3.2 ATRIUM 10XM Fuel Design The licensee plans to load ATRIUM 10XM fuel design in BSEP Unit 2 during the spring 2011 refueling outage and in Unit 1 during the 2012 refueling outage. The NRC-approved topical report, ANF-89-98(P)(A), Revision 1 t Supplement 1 (Reference 13) describes a process and criteria that allows Siemens Power Corporation (now AREVA NP) to be used to demonstrate that NRC review and approval of changes or improvements to existing BWR fuel designs are not required.

3.2.1 Fuel Thermal-Mechanical Design The ATRIUM 10XM fuel design, as described in the submittal of ANP-2899(P), Revision 0 (Reference 14) and BSEP 10-0118, Enclosure 1 (Reference 6), shares many of the same features of the ATRIUM-9 and -10 fuel designs that were used in BWR plants. The ATRIUM 10XM fuel bundle has the same basic geometry as the currently approved ATRIUM-10 fuel bundle design. The geometry consists of a 1 Ox1 0 fuel lattice with a square internal water OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 11 channel that displaces a 3x3 array of fuel rods. The ATRIUM 10XM incorporates additional key design features relative to ATRIUM-10 fuel:

((

))

The fuel pellet can be either U02 or U02-Gd20 3. The fuel rods are made with zircaloy 2 cladding.

The fuel bundle is encased in a channel box of identical material and dimensions as of the ATRIUM-10 fuel bundle design.

The licensee uses the approved generic fuel rod design methodology (Reference 13) and fuel performance code RODEX4 (Reference 16) to evaluate the thermal and mechanical performance of the ATRIUM 10XM fuel design. The RODEX4 code was approved to a peak rod average burnup of 62 gigawatt days per metric ton of uranium.

3.3 Transition Core Approach Each of the BSEP units has 560 assemblies. The Unit 2 Cycle 20 core will contain 226 fresh ATRIUM 10XM assembles, 238 once-burned ATRIUM-10 assemblies and 96 twice-burned GE 14 assemblies. The NRC staff reviewed the licensee's evaluation of the mixed-core configuration of the BSEP Unit 2 Cycle 20 core to predict the thermal-hydraulic performance, hydraulic compatibility, thermal margin performance, critical power performance, and the impact on core design and licensing analysis (Reference 4). The conclusion of this evaluation is also applicable to the BSEP Unit 1 transition to ATRIUM 10XM fuel.

3.3.1 Hydraulic Compatibility The licensee reported the results of thermal-hydraulic analyses in accordance with NRC-approved AREVA thermal-hydraulic methodology (References 13, 33, and 34). The methodology and constitutive relationships used in the licensee's calculation of pressure drop in BWR fuel assemblies are presented in Reference 34 and are implemented in the XCOBRA code. The XCOBRA code predicts steady-state thermal-hydraulic performance of the fuel assemblies of BWR cores at various operating conditions and power distributions. Hydraulic compatibility, as it relates to the relative performance of the ATRIUM 10XM, ATRIUM-10, and GE14 fuel designs, has been evaluated. Analysis for mixed cores with ATRIUM 10XM, ATRIUM-10, and GE14 fuel assemblies were performed to demonstrate that the thermal-hydraulic design criteria are satisfied for transition core configurations.

The calculations were performed with explicit modeling of ATRIUM 10XM, ATRIUM-10 and GE14 assemblies for several power-to-flow conditions, rated and off-rated, and for bottom-,

middle-, and top-peaked axial power distributions. Four core configurations were analyzed to address the relative assembly comparisons and core hydraulic compatibility evaluations. The OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 12 four core configurations are (1) ATRIUM-10 coresident with GE14 for Cycle 19, (2) first transition core for Cycle 20 fuel loading of ATRIUM 10XM, ATRIUM 10, and GE14, (3) second transition loading with ATRIUM 10XM and ATRIUM-10, and (4) full core of ATRIUM 10XM. Results of the calculations listed in Reference 4 and those presented during the regulatory audit indicate that core average results and the difference between ATRIUM 10XM, ATRIUM-10, and GE14 results at rated power are within the range considered compatible. Similar agreement exists at off-rated power levels. ((

)) The staff concludes that the licensee's hydraulic compatibility analysis provides reasonable assurance that the introduction of ATRIUM 10XM fuel into the BSEP units will not significantly impact the core flow distribution.

3.3.2 Thermal Margin Performance Thermal margin analyses were performed in accordance with the thermal hydraulic methodology based on AREVA's XCOBRA code. The calculation of CPR, which is a measure of thermal margin performance is established by means of an empirical correlation based on results of BT test programs. CPR values for ATRIUM 10XM are calculated with the NRC-approved ACE/ATRIUM 10XM critical power correlation (Reference 11) while the CPR values for the ATRIUM-10 and GE14 fuel are calculated using the NRC-approved SPCB critical power correlation (Reference 17). Fuel assembly design features are incorporated in the CPR calculation through the K-factor in the ACE correlation and through the F-effective term for the SPCB correlation. The K-factors and F-effective terms are based on local power peaking factors that are functions of assembly void fraction and exposure. Analysis results (Reference 4) indicate that the introduction of ATRIUM 10XM in the BSEP units will not cause thermal margin problems for the co-resident fuel designs.

The NRC staff concludes that there is no adverse impact on thermal margin performance due to the mixed core configuration at the BSEP units.

3.3.3 Safety Limit Minimum Critical Power Ratio (SLMCPR) Analysis The SLMCPR is defined as the minimum value of the CPR which ensures that less than 0.1 percent of the fuel rods in the core are expected to experience BT during normal operation or an ACO. The SLMCPR is determined using the methodology in Reference 20 using a power distribution that conservatively represents expected reactor operating states that exist at the MCPR operating limit and produce an MCPR equal to the SLMCPR during an AOO. The BSEP Unit 2 Cycle 20 SLMCPR analysis uses ACE/ATRIUM 10XM critical power correlation additive constants and additive constant uncertainty for ATRIUM 10XM fuel described in Reference 11.

For the ATRIUM-10 fuel SLMCPR analysis, the SPCB critical power correlation additive constants and related uncertainty are used as per Reference 17. The SLMCPR analysis explicitly includes channel bow. The channel bow local peaking uncertainty is a function of the nominal and bowed local peaking factors and the standard deviation of the channel bow.

The SLMCPR analysis results support a two-loop operation SLMCPR of 1.11 with 0.092 percent of the rods in BT (Reference 8). For single-loop operation, the SLMCPR value is 1.13 with 0.076 percent of the rods in BT (Reference 8).

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- 13 3.3.4 Core Design and Licensing Analysis Fuel cycle design and fuel management calculations for the Cycle 20 operation of BSEP Unit 2 are performed in accordance with the NRC-approved methodology, EMF-2158(P)(A)

(Reference 24). The CASMO-4 lattice depletion code is used to generate nuclear data including cross sections and local power peaking factors. The MICROBURN-B2 three-dimensional core simulator code utilizes the pin power reconstruction model to determine the thermal margins.

The ACE correlation is used for the ATRIUM 10XM fuel assemblies while the coresident ATRIUM-10 and GE14 fuel assemblies are monitored with the SPCB correlation. The core neutronic design includes control blade depletion, explicit neutronic treatment of the spacer grids, explicit modeling of PLFR plenums, and explicit modeling of the water rod flow.

Control rod patterns are developed to be consistent with conservative margin to thermal limits.

The fuel cycle design demonstrates adequate hot excess reactivity and cold shutdown margin throughout the cycle. Fuel assembly thermal-mechanical limits for ATRIUM 10XM, ATRIUM-10, and coresident fuel are verified and monitored for each mixed core designed by AREVA. The thermal-mechanical limits established by the vendor of the coresident fuel are applied for that fuel in mixed (transition) cores. AREVA performed design and licensing analyses to demonstrate that the core design meets the limits during steady-state and AOO conditions. The NRC staff finds the approach acceptable.

3.3.5 Radial and Axial Power Distribution Measurement Uncertainties As stated on Page 9-1 of topical report, EMF-2158(P){A) (Reference 24), the AREVA methodology for measuring the power distribution in a BWR reactor and the procedure by which the uncertainty associated with the measurement of a BWR power distribution would be determined, was originally described in XN-NF-80-19(P)(A), Vol. 1, Supplements 3 & 4 (Reference 35). Section 5 of Reference 35 and Section 9 of Reference 24 together provide a very detailed description of the analyses and calculations performed to determine the TIP uncertainty components. The NRC staff requested more details on the TIP uncertainty components listed in the two tables on Page 4 of Enclosure 1 of Reference 15. The licensee provided details of these TIP uncertainty measurements in Reference 9 and during the regulatory audit conducted in November 2010.

Figures 1, 2, and 3 on Pages 11 through 13 of Enclosure 1 of Reference 15 present 177 database points. Each database point is calculated using a TIP flux map consisting of measurements obtained from 21 axial levels at 31 radial core locations. Except for the size of the data population, the detailed equations provided by References 24 and 35 and listed in Table 17.1 of Reference 9 are the same for both the database points and the final uncertainty components. Each database point is based on 651 local TIP readings (i.e., 21 times 31) obtained at a core operating state characterized by its core power, core void fraction, and core power-to-f1ow ratio, whereas the TIP uncertainty components are based on 115,227 local TIP readings. The 177 TIP flux maps were obtained from BSEP Units 1 and 2 core operating states from March of 2000 through February of 2010, from cores consisting entirely of 9x9 GE13 fuel, mixed cores of GE13 and 10x10 GE14 fuel, GE14 fuel alone, and mixed cores of GE14 and 10x10 ATRIUM-10 fuel.

The BSEP C4/MB2 benchmark completed by AREVA to incorporate explicit water rod, PLFR plenum, and spacer model options was not available at the time the TIP statistics presented in the CP&L letter, BSEP 10-0057 (Reference 2) were calculated; however, none of these changes materially affect C4/MB2 calculated TIP distributions. The licensee has since recalculated TIP OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 14 statistics based on the latest AREVA benchmark. The database values presented in the CP&L letter, BSEP 10-0057 that are dependent on C4/MB2 calculated TIP response are plotted against the values calculated based on the latest AREVA benchmark in Figure 17.1 of Reference 9. The results demonstrate the explicit water rod PLFR plenum and spacer model options have no impact on the TIP statistics. The TIP uncertainty component values and trends based on the latest C4/MB2 benchmark (Le., incorporating explicit water rod, PLFR plenum, and spacer model options) are provided as Table 17.2 and Figures 17.2, 17.3 and 17.4 of Reference 9.

The D-Lattice (BSEP) uncertainty component values identified in Sections 9.4 and 9.5 of EMF-2158(P)(A) bound the BSEP-specific uncertainty component values shown in Table 17.2 and reproduced below in Table 3.3. This result demonstrates that the uncertainties applied by the BAW-10247PA and ANF-524(P)(A) methodologies, and determined in accordance with the EMF-2158(P)(A) methodology, are applicable to BSEP Units 1 and 2.

The NRC staff concludes that the licensee's evaluation of the TIP database for previous cycles including both BSEP units has demonstrated that uncertainties documented in EMF-2158(P)(A) for D-Lattice plants (BSEP Units 1 and 2) remain conservative and none of the features of ATRIUM 10XM design will have any impact on the accuracy of the methodology to predict TIP response.

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- 15 Table 3.3 Updated TIP Component Uncertainties Component Reference 15 Value (%)

Latest Benchmark Value (%)

EMF-2158 Value (0 Lattice (%)

Nodal TIP Measurement Uncertainty, OTmiik 1.90 1.90

((

))

Radial Tip Measurement Uncertainty, oTmij 1.25 1.25

((

))

Planar TIP Measurement Uncertainty, oTm planar 1.97 1.97

((

))

Nodal Deviation between measured TIP Readings a.nd Calculated by C4/MB2, oT Uk 4.47 4.44

((

))

Radial Deviation between Measured TIP Readings a.nd Calculated by C4/MB2, oT jj 2.07 2.07

((

))

Planar Deviation between Measured TIP Readings a.nd Calculated by C4/MB2, oT planar 2.58 2.58

((

))

Nodal Synthesis Procedure Uncertainty, oSij 0.22 0.21

((

))

Radial Synthesis Procedure Uncertainty, OSijk 1.79 1.68

((

))

Net Nodal Calculated TIP Distribution Uncertainty, oTjjk Calculation method provided in Reference 2 instead of proprietary value

((

))

((

))

Net Radial Calculated TIP Distribution Uncertainty, oTH

((

))

((

))

Net Planar Calculated TIP Distribution Uncertainty, oT planar

((

))

((

))

3.3.6 Transition Core Summary In the AREVA thermal-hydraulic methodology, each fuel type is explicitly modeled. Therefore, the impacts of the differences in mechanical design on geometry and loss coefficients are explicitly accounted for. The critical power performance of each fuel type is also explicitly modeled using the applicable critical power correlation for each fuel design. Limits are established for each fuel type and operation within these limits is verified by the core monitoring system during plant operation. Therefore, the NRC staff concludes that the licensee's treatment of transition cores is acceptable.

3.4 Plant-Specific Reload Safety Analyses Reload licensing analyses in support of BSEP's fuel transition are performed using NRC-approved generic methodologies for BWRs. The reload licensing analyses are performed for the potentially limiting events and other events are identified as disposition events. The results of the analyses are used to establish the BSEP TS Core Operating Limits Report limits and ensure that that the design and licensing criteria are met (Reference 8).

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- 16 A summary of disposition of events is listed in Tables 2.1 and 2.2 of Reference 8. The objective of the disposition of events is to identify the limiting events which must be analyzed to support operation at BSEP with the introduction of ATRIUM 10XM fuel. Events and analyses identified as potentially limiting are either evaluated generically for the introduction of ATRIUM 10XM fuel or on a cycle-specific basis.

The sections below list the limiting events and a short description of the analyses and results.

3.4.1 Anticipated Operational Occurrences Section 3.2 of this safety evaluation lists the major codes used in the thermal limits analyses, neutronics methodology, and critical power calculations. The limiting exposure for rated power pressurization transients is typically at the end of full power when the control rods are fully withdrawn. To provide additional margin to the operating limits earlier in the cycle, analyses were also performed to establish operating limits at a near end-of-cycle (NEOC) exposure of 16,700 megawatt days per metric ton of uranium. Analyses were also performed to support extended cycle operation with final feedwater temperature reduction and power coastdown. The Sections below provide brief descriptions of a few selected AOO analyses performed for the BSEP units.

Load Rejection No Bvpass (LRNB): The load rejection causes a fast closure of the turbine control valves. The resulting compression wave in the steam lines into the vessel creates a rapid pressurization. Pressurization causes a decrease in voids and a rapid increase in power.

The turbine control valve closure causes a reactor scram. LRNB analyses are performed for a range of powerlflow conditions to support generation of the thermal limits. Results are used to generate the NEOC and end-of-the cycle licensing basis (EOCLB) operating limits for both TS scram speed (TSSS) and nominal scram speed (NSS) insertion times.

Turbine Trip No Bypass (TTNB): The turbine trip causes a closure of the turbine stop valves.

The resulting compression wave travels through the steam lines into the vessel and creates a rapid pressurization. The increase in pressure causes a decrease in core voids, which in turn causes a rapid increase in power. The closure of the turbine stop valves also causes a reactor scram. The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core. The LRNB analyses for previous cycles have shown that the consequences of the TTNB event are bounded by those of the LRNB event. The licensee's TTNB analysis for Cycle 20 has shown that the LRNB event remains bounding.

Feedwater Controller Failure (FWCF): The increase in feedwater flow due to a failure of the feedwater control system results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. The water level continues to rise and eventually reaches the high water level trip point.

The high water level trip causes the turbine stop valves to close in order to prevent damage to the turbine from excessive liquid inventory in the steam line. The valve closures create a compression wave that travels to the core causing a void collapse and subsequent rapid power excursion. The closure of the turbine stop valves also initiates a reactor scram. FWCF analyses are performed for a range of power/flow conditions to generate the thermal limits. Reference 9 lists the FWCF analyses results that are used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times.

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- 17 Loss of Feedwater Heating (LFWH): The LFWH event analysis supports an assumed 100 degrees Fahrenheit (OF) decrease in the feedwater temperature. The result is an increase in core inlet subcooling, which reduces voids, thereby increasing core power and shifting the axial power distribution toward the bottom of the core. The axial power shift and increase in core power causes the voids to build up in the bottom of the core, acting as negative feedback to the increased subcooling effect. The negative feedback moderates the core power increase. The increase in core thermal power event does not result in a corresponding increase in steam flow because some of the added power is used to overcome the increase in inlet subcooling. The increase in steam flow is accommodated by the pressure control system via the turbine control valves (TCVs) or the turbine bypass valves, so no pressurization occurs. The licensee performed a cycle-specific analysis according to Reference 36 methodology to determine the change in MCPR for an LFWH event. The NRC staff finds the results acceptable.

Control Rod Withdrawal Error (CRWE): The CRWE transient is an inadvertent reactor operator initiated withdrawal of a control rod. This withdrawal increases local power and core thermal power, lowering the core MCPR. The CRWE is typically terminated by control rod blocks initiated by the rod block monitor (RBM). The CRWE event was analyzed assuming no xenon and allowing credible instrumentation out-of-service in the RBM system. The CRWE analysis has demonstrated that, in addition to supporting the standard filtered RBM setpoint reductions, the 1 percent strain and centerline melt criteria are met for both ATRIUM 10XM and ATRIUM-10.

Equipment Out-of-Service Scenarios (EOOS):

The following EOOS scenarios are supported for the BSEP Unit 2 Cycle 20 operation:

Feedwater heater out-of-service (FHOOS) - This scenario assumes a final feedwater temperature reduction (FFTR) of 110.3 of at rated power and steam flow. The FFTR causes an increase in core inlet subcooling that can change the axial power shape and core void fraction.

The steam flow for a given power level decreases since more power is required to increase the enthalpy of the coolant to saturated conditions. The FWCF is analyzed to ensure that appropriate FHOOS operating limits are established.

Other EOOS scenarios analyzed are:

Turbine bypass valves out-of-service (TBVOOS) - Analysis of the FWCF are performed to establish the TBVOOS operating limits.

Combined FHOOS and TBVOOS - Operating limits for this combination are established using the FWCF results.

One safety/relief valve out-of-service (One SRVOOS) - All operating limits support operation with one SRVOOS.

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- 18 One main steam isolation valve out-of-service (One MSIVOOS) - Operation with one MSIVOOS is supported for operation up to 70 percent rated power operation. Operation with one MSIVOOS has no impact on the other non-pressurization events evaluated to establish power-dependent operating limits. Therefore, the power-dependent operating limits applicable to base case operation with all MSIVs in service remain applicable for operation with one MSIVOOS for power levels less than or equal to 70 percent of rated.

The slow flow runup analyses were performed to support operation with one MSIVOOS.

Single-loop operation 3.4.2 Core Hydrodynamic Stability The BSEP is currently operating under the requirements of the reactor stability long-term Option III solution approved by the NRC staff in GE licensing topical report, NEOO-32465-A (Reference 25). The stability based operating limit MCPR (OLMCPR) is provided for two conditions as a function of oscillation power range monitor (OPRM) amplitude setpoint as listed in Table 4.3 of Reference 8. The two conditions evaluated are for a postulated oscillation at 45 percent core flow steady-state operation and following a two-recirculation pump trip (2PT) from the limiting full power operation state point. The Cycle 20 power and flow dependent limits provide adequate protection against violation of SLMCPR for postulated reactor instability as long as the operating limit is greater than or equal to the specific value for the selected OPRM setpoint.

AREVA performed calculations for the relative change in ilCPR as a function of the calculated hot channel oscillation magnitude (HCOM). These calculations were performed using the RAMONA5-FA code in accordance with Reference 26 methodology. RAMONA5-FA is a coupled neutronic thermal-hydraulic three-dimensional transient model for determining the relationship between the relative change in ilCPR and the HCOM on a plant-specific basis. The stability based OLMCPRs were calculated using the most limiting of the calculated change in relative ilCPR for a given oscillation magnitude or the generic value provided in Reference 25.

In cases where the OPRM system is declared inoperable for BSEP Unit 2, BSP is provided. The BSP curves are evaluated using STAIF (Reference 37) to determine endpoints that meet decay ratio criteria for BSP base minimal Region I (scram region) and BSP base minimal Region II (control entry region). Analyses are performed to support operation with both nominal and reduced feedwater temperature conditions (both FFTR and FHOOS). The BSP endpoints are provided in Table 4.4 of Reference 8.

Based on the information provided by the licensee and discussed above and based on the information presented during the audit in November 2010, the NRC staff finds that the stability analysis and evaluation performed in support of the LAR provides reasonable assurance that the proposed transition in fuel and methods will not adversely impact BSEP ability to satisfy GOC 10 and 12.

3.4.3 Emergency Core Cooling System (ECCS) Performance The ECCS is designed to mitigate postulated LOCAs cased by ruptures in primary system coolant piping. The ECCS performance under all LOCA conditions and the evaluation model must satisfy the requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix K.

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- 19 For a BWR, a LOCA may occur over a wide spectrum of break locations and sizes. Because of significant variations in responses over a break spectrum, an analysis covering the full range of break sizes and locations is performed to identify the limiting break characteristics. Regardless of the initiating break characteristics, the LOCA event is separated into three phases the blowdown phase, the refill phase, and the reflood phase. During the blowdown phase of a LOCA, there is a net loss of coolant inventory, an increase in fuel cladding temperature due to core flow degradation, and the core may become fully or partially uncovered depending on the break size. During the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant inventory due to the activation of the core sprays that provide core cooling.

The low pressure and high pressure coolant injection systems supply coolant to refill the lower portion of the reactor vessel. During the reflood phase, when the coolant inventory has increased, the cooling is provided above the mixture level by entrained reflood liquid. The ECCS must be designed such that the plant response to a LOCA meets the acceptance criteria specified in 10 CFR 50.46(b).

The evaluation model used for the BSEP LOCA analysis is the NRC-approved EXEM BWR-2000 LOCA analysis methodology described in Reference 29. The EXEM BWR-2000 employs three major computer codes, RELAX, HUXY, and RODEX2, to evaluate the system and fuel response during all phases of a LOCA. RELAX (Reference 29) is used to calculate the system and hot channel response during the blowdown, refill and reflood phases of the LOCA.

The HUXY code (Reference 38) is used to perform heatup calculations for the entire LOCA, and calculates the peak clad temperature (PCT) and local clad oxidation at the axial plane of interest. RODEX2 (Reference 32) is used to determine fuel parameters (such as stored energy) for input to the other LOCA codes. A complete analysis for a given break size starts with the specification of fuel parameters using RODEX2. RODEX2 is then used to determine the initial stored energy for both the blowdown analysis (RELAX hot channel) and the heatup analysis (HUXY). This is accomplished by ensuring that the initial stored energy in RELAX and HUXY is the same or higher than that calculated by RODEX2 for the power, exposure, and fuel design being considered.

The LOCA break spectrum analysis is performed for a full core of ATRIUM 10XM fuel. Table 3.2 provides a summary of reactor initial conditions used in the break spectrum analysis.

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- 20 Table 3.2 Initial Conditions for Break Spectrum Analysis and Heatup Analysis Parameter

((

))

((

))

Reactor power (% of rated) 102 102

((

))

((

))

(( ))

Reactor power (MW(th>>

2981.5 2981.5

((

))

((

))

((

))

((

))

I

((

))

((

))

Steam flow rate (100Ib/hr) 13.1 13.1 Steam dome pressure (psia) 1048.9 1048.7 Core inlet enthalpy (Btu/lb) 527.7 522.4 ATRIUM 10XM hot assembly MAPLHGR (kw/ft) 13.1 13.1

((

))

((

))

((

))

Mid-and Top-Mid-and Top-Rod average power distributions

Peaked, Peaked, Figure 4.6*

Figure 4.7*

• Reference 5 The 2 percent uncertainty increase in rated power is required by 10 CFR Part 50, Appendix K.

The analyses are performed at ((

))

The licensee stated that the break spectrum analyses are applicable to ((

)) The break characteristics identified in the LOCA break spectrum analysis are used in the subsequent fuel type specific LOCA heat up analysis (Enclosure 2 (ANP-2943(P>> to Reference 5) to determine maximum average planar linear heat generation rate (MAPLHGR) limits for the appropriate fuel type. The NRC staff finds that it is reasonable to conclude that the BSEP LOCA break spectrum analysis for full core ATRIUM 10XM fuel ((

))

In conformance with regulatory requirements, the LOCA analyses are performed assuming that all offsite power supplies are lost instantaneously and that only safety grade systems and components are available. In addition, per regulatory requirements the most limiting single failure of ECCS equipment is assumed in the LOCA analysis. The term "most limiting" refers to the ECCS equipment failure that produces the greatest challenge to event acceptance criteria (10 CFR 50.46(b>>. The potential limiting single failures identified in the BSEP Updated Final OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 21 Safety Analysis Report are: DC power (SF-BATT), diesel generator, low-pressure coolant injection valve (SF-LPCI), and high-pressure coolant injection system (SF-HPCI). The licensee reviewed the accident scenarios and demonstrated that the SF-BATT and SF-LPCI injection valve failures are the limiting failures, as the other single failures result in as much or more ECCS capacity.

The licensee has performed a complete spectrum analysis of break sizes that include double-ended guillotine (DEG) with discharge coefficients from 1.0 to 0.4, split breaks with areas between full pipe area and 0.05 tf and break locations (recirculation and non-recirculation pipes). As discussed above, the single failures considered in the recirculation line break analyses are SF-BATT and SF-LPC!.

The results of the LOCA break spectrum analysis show that the limiting recirculation line break is the 0.8 DEG break in the pump discharge piping with an SF-LPCI single failure and top-peaked axial power shape when operating at 102 percent rated core power and ((

)). Detailed results are provided in (Enclosure 1 (ANP-2941 (P>> to Reference 5).

For single loop operation (SLO), a multiplier less than one is applied to the MAPLHGR limits to ensure that the SLO LOCA results are bounded by the two-loop operation LOCA results. In the SLO analysis, the decrease in the MAPLHGR limit is achieved by applying this factor to the radial peaking factor. Local power distributions for the BSEP ATRIUM 10XM neutronic designs are used in the heatup analysis ANP-2943(P). The initial conditions used for the LOCA heatup analysis are listed in the second column of Table 3.2, with exception that the rod average power distribution is a top-peaked axial as shown in Figure 4.5 of ANP-2943(P).

The EXEM BWR-2000 evaluation model is applied to confirm the acceptability of the ATRIUM 10XM MAPLHGR limit for BSEP Units 1 and 2. The analysis results are listed below.

The acceptance criteria of 10 CFR 50.46 are met for operation at or below the ATRIUM 10XM MAPLHGR limit specified in Figure 2.1 of ANP-2941 (P).

PCT 1871 OF < 2200 oF.

Local cladding oxidation thickness 0.99 percent < 17 percent

• Total hydrogen production 0.46 percent <1 percent Cool able geometry, satisfied by meeting all of the criteria Long term core cooling satisfied by demonstrating the core flooded to top of active fuel or core flooded to the jet pump suction elevation.

The MAPLHGR limit is applicable for ATRIUM 10XM full cores as well as transition cores containing ATRIUM 10XM fuel The licensee performed BSEP-specific LOCA analysis based on an NRC-approved methodology. The initial conditions, break spectrum, and power profiles selected for LOCA analysis are consistent with the NRC-approved topical report, which covers sufficiently limiting scenarios to determine the PCT. The NRC staff finds that the 10 CFR 50.46 acceptance criteria are met and the ECCS performance is acceptable. Based on the above, the staff finds the LOCA analyses performed in support of the LAR acceptable.

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- 22 3.4.4 Anticipated Transient Without Scram (ATWS) Events An A TWS is defined as an AOO followed by failure of the scram function of the protection system required by GDC-20. The NRC staff reviewed the licensee's ATWS analysis to ensure that (1) the peak vessel bottom pressure is less than the ASME service level C limit of 1500 pounds per square inch gauge (psig); (2) the PCT is within the 10 CFR 50.46 limit of 2200 of; (3) the peak suppression pool temperature is less than the design limit (220 of for BSEP); and (4) the peak containment pressure is less than the containment design pressure (62 psig for BSEP). Since AREVA does not have a generically approved long-term ATWS containment evaluation methodology, the NRC staff reviewed the licensee's long term evaluation for ATRIUM 10XM introduction.

The A TWS overpressurization analyses were performed at 100 percent power for both 99 percent flow and 104.5 percent flow. The MSIV closure and pressure regulator failure open (PRFO) events were evaluated. Failure of the pressure regulator in the open position causes the turbine control and turbine bypass valves to open such that steam flow increases until the maximum combined steam flow limit is attained. The system pressure decreases until the low pressure setpoint is reached, resulting in the closure of the MSIVs. The resulting pressurization wave causes a decrease in core voids and an increase in core pressure thereby increasing the core power.

The ATWS overpressurization analyses are presented in Table 7.3 and Figures 7.5 through 7.8 of Reference 8 show the response of various reactor plant parameters during the limiting PRFO event, the event which results in the maximum vessel pressure.. The maximum lower plenum pressure and the maximum dome pressure are both below the ASME limit of 1500 psig. The peak pressure results are adjusted to address NRC concerns associated with the void-quality correlation and Doppler effects. The effects of exposure-dependent thermal conductivity degradation were included in the analysis. The results demonstrate that the ATWS maximum vessel pressure limit of 1500 psig is not exceeded.

Relative to the 10 CFR 50.46 acceptance criteria (Le., PCT and cladding oxidation). the consequences of an A TWS event are bound by those of the limiting LOCA event. Based on fuel performance analyses conducted for BWR A TWS events, the NRC staff finds that there is reasonable assurance that 10 CFR 50.46 criteria will not be challenged during ATWS, and therefore finds it acceptable.

In addition to the short-term vessel overpressure and PCT analysis, the long-term suppression pool performance must be evaluated for acceptability during A TWS. Fuel design differences may impact the power and pressure excursion experienced during the A TWS event. This, in turn, impacts the amount of steam discharged to the suppression pool and containment. The licensee stated that ((

))

((

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- 23

))

The results show that [(

)) Therefore, the NRC staff concludes that the introduction of A TRI UM 10XM fuel ((

))

The NRC staff concludes that the licensee has demonstrated that the required systems are installed at BSEP and that they will continue to meet the requirements of 10 CFR SO.62. In addition, the NRC staff reviewed the information submitted by the licensee related to A TWS and concludes that the licensee adequately accounted for the effects of the proposed fuel and methodology transition on ATWS. Therefore, the NRC staff finds the proposed LAR acceptable with respect to A TWS.

3.S Summary and Conclusion The NRC staff has reviewed the LAR (Reference 1), in conjunction with the supplemental information (References 2 through 8), the responses to the staffs requests for additional information (References 9 and 10), and the licensee's submittal dated April 6, 2011 (Reference 39) to evaluate the acceptability of the BSEP transition to ATRIUM 10XM fuel with AREVA safety analysis and core design methodologies. Based on its review, the NRC staff has determined that the licensee provided adequate technical basis to support the proposed TSs changes. Specifically, the NRC staff finds the licensee has demonstrated that (1) BSEP complies with the staff limitations and conditions imposed for application of the topical reports, (2) AREVA codes and methods are applicable for BSEP (3) the BSEP-specific safety analysis results based on the AREVA methodology meet the applicable licensing criteria, and (4) the proposed TSs changes are acceptable.

4.0 LICENSE CONDITION The following license condition shall be included in Appendix B, "Additional Conditions," of the Operating Licenses (OLs) for Brunswick Steam Electric Plant (BSEP), Units 1 and 2 to ensure the limits generated with the NRC-approved methods appropriately bound the effects of the K-factor calculation:

Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and core operating limit values determined using the ANP-10298PA, ACE/ATRIUM 10XM Critical Power Correlation (Le., TS S.6.S.b.21), shall be evaluated with methods described in AREVA Operability Assessment CR 2011-2274, Revision 1 to verify the values determined using the NRC-approved method remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. The results of the evaluation shall be documented and submitted to the NRC, for review, at least 60 days prior to startup of each operating cycle.

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5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the state of North Carolina official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

S The amendments change a requirement with respect to installation or use of a facility component 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 NRC has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (75 FR 48372).

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 amendments.

7.0 CONCLUSION

The NRC has 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 NRC'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.

8.0 REFERENCES

1. Letter BSEP 10-0052 from Michael J. Annacone to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324 Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, Core Operating Limits Report (COLR)," April 29, 2010.

2. Letter BSEP 10-0071 from Michael J. Annacone to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324 Brunswick Unit 2 Cycle 20 Fuel Cycle Design Report ANP-2936(P)," June 9,2010.

3. Letter BSEP 10-0083 from Michael J. Annacone to NRC, "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, Additional Information Supporting Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," July 22,2010.

4. Letter BSEP 10-0093 from Michael J. Annacone to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324 Brunswick Unit 2 Cycle 20 Thermal-Hydraulic Design Report," July 29,2010.

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- 25

5. Letter BSEP 10-0112 from Michael J. Annacone to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324 Brunswick Unit 2 Cycle 20 Loss of Coolant Accident Analysis Reports, (a) ANP-2941 (P) Revision 0 "Brunswick Units 1 and 2 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel," (b) ANP-2943(P) Revision 0, "Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM 10XM Duel,"

September 29, 2010.

6. Letter BSEP 10-0118 from Michael J. Annacone to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324, Additional Information Supporting License Amendment Requests - ATRIUM 10XM Fuel Rod Thermal and Mechanical Evaluation for Brunswick Unit 2 Cycle 20," (AREVA ReportANP-2950P), October 12, 2010.

7. Letter BSEP 10-0126 from Michael J. Annacone to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324, Additional Information Supporting License Amendment Requests - Mechanical Design Report for Brunswick ATRIUM 10XM Fuel Assemblies (AREVA ReportANP-2948(P), Revision 0), November 9,2010.

8. Letter BSEP 10-0126 from Michael J. Annacone to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324, Additional Information Supporting License Amendment Requests - Brunswick Unit 2 Cycle 20 Reload Safety Analysis," (AREVA Report ANP 2956(P), Revision 0), November 9,2010.

9. Letter BSEP 10-0133 from William Jefferson, Jr. to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324, Response to Additional Information Report Supporting License Amendment Requests for Analytical Methodology Topical Reports to Technical Specification 5.6.5 (NRC TAC Nos. ME3856, ME3857, ME3858, and ME3859),

November 18, 2010.

10. Letter BSEP 10-0141 from William Jefferson, Jr. to NRC "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324, Response to Additional Information Report Supporting License Amendment Requests for Analytical Methodology Topical Reports to Technical Specification 5.6.5 (NRC TAC Nos. ME3856, ME3857, ME3858, and ME3859),

December 16,2010.

11. ANP-1 0298PA, Revision 0, "ACE/ATRIUM 10XM Critical Power Correlation," AREVA NP Inc., March 2010

12. Letter from NRC (F. Saba) to B. Waldrep (C P & L), "Brunswick Steam Electric Plant Units 1 and 2 -Issuance of Amendment to Support Transition to AREVA Fuel and Methodologies," US NRC, March 27, 2008. (Adams Accession No. ML080870478)

13. ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, May 1995.

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- 26

14. ANP-2899P, Revision 0, "Fuel Design Evaluation for ATRIUM 10XM BWR Reload Fuel,"

AREVA NP Inc., April 2010.

15. Letter BSEP 10-0057 from Michael J. Annacone to NRC, "Brunswick Steam Electric Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324, Request for License Amendments - Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, Core Operating Limits Report (COLR)," April 29, 2010.

16. BAW-10247PA, Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP Inc., April 2008.

17. EMF-2209(P)(A) Revision 3, "SPCB Critical Power Correlation," AREVA ANP, Inc.,

September 2009.

18. EMF-2245(P)(A) Revision 0, "Applications of Siemens Power Corporation's Power Correlations to Co-Resident Fuel," Siemens Power Corporation, August 2000.

19. ANF-524(P)(A) Revision 2, "Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors," Advanced Nuclear Fuels Corporation, April 1989.

20. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company, February 1987.

21. XN-NF-80-19(P)(A), Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.

22. ANF-913(P)(A) Volume 1 Revision 1 Supplements 2, 3 and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.

23. ANP-2637 Revision 3, "Boiling Water Reactor licenSing Methodology Compendium,"

AREVA NP Inc., March 2010.

24. EMF-2158(P)(A) Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation, October 1999

25. NEDO-32465-A, Class 1, Licensing Topical Report, "Reactor Stability and Suppress Solutions Licensing Basis Methodology for Reload Applications," BWR Owners Group, August 1986.

26. BAW-10255(P)(A) Revision 2, "Cycle Specific DIVOM Methodology Using the RAMONA5-FA Code," AREVA NP Inc., May 2008.

27. EMF-CC-074(P)(A), Volume 1, "STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain," and Volume 2, "STAIF - A Computer Program for BWR OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 27 Stability Analysis in the Frequency Domain - Code Qualification Report," Siemens Power Corporation, July 1994.

28. EMF-CC-074(P)(A), Volume 4, Revision 0, "BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation, August 2000.

29. EMF-2361 (P)(A), Revision 0, "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP, May 2001.

30. XN-CC-33(A), Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual," Exxon Nuclear Company, November 1975.

31. XN-NF-82-07(P)(A), Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, November 1982.

32. XN-NF-81-58(P)(A), Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984.

33. XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors, Application of the ENC Methodology to BWR reloads," Exxon Nuclear Company, June 1986.

34. XN-NF-79-59(P)(A), "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies," Exxon Nuclear Company, November 1983.

35. XN-NF-80-19(P)(A), Vol. 1, Supplement 3 &4," Benchmark Results for the CASMO 3G/MICROBURN-B Calculation Methodology," Exxon Nuclear Company, November 1990.

36. ANF-1358(P)(A) Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005.

37. EMF-CC-074(P) Volume 4 Revision 0, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation, November 1999.

38. XN-CC-33(A) Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual," Exxon Nuclear Company, November 1975.

39. Letter BSEP 11-0040 from Michael J. Annacone to NRC, "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 Additional Information Supporting License Amendment Request to Add Analytical Methodology ANP 10298PA to Technical Specification 5.6.5, Core Operating Limits Report (COLR)"

(TAC Nos. ME3856 and ME3857), April 6, 2011 Principal Contributors: Mathew Panicker Shih-Liang Wu Date: April 8, 2011 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION M. Annacone

-2 opportunity to comment on any proprietary aspects. If you believe that any information in the enclosure is proprietary, 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 nonproprietary SE will be made publicly available. Copies of the proprietary and nonproprietary versions of the SE are enclosed.

A notice of issuance will be included in the NRC's biweekly Federal Register notice.

Sincerely, IRA!

Farideh E. Saba, Senior 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. 257 to License No. DPR-71

2. Amendment No. 285 to License No. DPR-62

3. Safety Evaluation (Nonproprietary Information)

4. Safety Evaluation (Proprietary Information) cc w/enclosures 1, 2, 3, and 4: Addressee cc w/enclosures 1,2, and 3: Distribution via ListServ DISTRIBUTION:

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