ML14169A475

Issuance of Amendment No. 249, Revise Technical Specification 2.1.1.1 to Add a New Fuel Centerline Melt Temperature and Burnup Relationship Using Copernic Code

ML14169A475
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/09/2014
From:	Peter Bamford Plant Licensing Branch IV
To:	Entergy Operations
Bamford P
References
TAC MF2277
Download: ML14169A475 (12)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 9, 2014 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT 1 - ISSUANCE OF AMENDMENT RE:

REVISION TO TECHNICAL SPECIFICATION 2.1.1.1, REACTOR CORE SAFETY LIMITS (TAC NO. MF2277)

Dear Sir or Madam:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 249 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit 1 (AN0-1).

The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated June 11, 2013, as supplemented by letter dated December 11, 2013.

The amendment changes AN0-1 TS 2.1.1.1, to add a provision for the determination of the maximum local fuel pin centerline temperature using the NRC reviewed and approved COPERNIC fuel performance computer code.

A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, Peter J. Bamford, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313

Enclosures:

1. Amendment No. 249 to DPR-51

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY ARKANSAS, INC.

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 249 Renewed License No. DPR-51

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

A. The application for amendment by Entergy Operations, Inc. (EOI, the licensee),

dated June 11, 2013, as supplemented by letter dated December 11, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license 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.

Enclosure 1

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-51 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 249, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~~

Michael T. Markle:**Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-51 and Technical Specifications Date of Issuance: July 9, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 249 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 Replace the following pages of the Renewed Facility Operating License No. DPR-51 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Operating License REMOVE INSERT 3 3 Technical Specifications REMOVE INSERT 2.0-1 2.0-1

(5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6) EOI, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

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 1: 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 249, are hereby incorporated in the renewed license.

EOI shall operate the facility in accordance with the Technical Specifications.

(3) Safety Analysis Report The licensee's SAR supplement submitted pursuant to 10 CFR 54.21 (d),

as revised on March 14, 2001, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than May 20, 2014.

(4) Physical Protection EOI 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 combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Arkansas Nuclear One Physical Security Plan, Training and Qualifications Plan, and Safeguards Contingency Plan," as submitted on May 4, 2006.

Renewed License No. DPR-51 Amendment No. 249 Revised by letter dated July 18, 2007

Sls 2.0 2.0 SAFETY LIMITS (Sls) 2.1 Sls 2.1.1 Reactor Core Sls 2.1.1.1 In MODES 1 and 2, the maximum local fuel pin centerline temperature shall be:-:;; 5080- (6.5 x 10-3 x (Burnup, MWD/MTUtF) for TAC02 applications, :-:; 4642 - (5.8 x 1o-3 x (Burn up, MWD/MTUtF) for TACO 3 applications, and:-:;; 4901 °F, decreasing linearly by 13.7 oF per 10,000 MWD/MTU of burnup for COPERNIC applications.

2.1.1.2 In MODES 1 and 2, the departure from nucleate boiling ratio shall be maintained greater than the limits of 1.3 for the BAW-2 correlation, 1.18 for the BWC correlation, and 1.132 for the BHTP correlation.

2.1.1.3 In MODES 1 and 2, Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the Variable Low RCS Pressure-Temperature Protective Limits as specified in the Core Operating Limits Report, so that the safety limits are met.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained :-:; 2750 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed:

2.2.1 In MODE 1 or 2, if SL 2.1.1.1 or SL 2.1.1.2 is violated, be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 In MODE 1 or 2, if SL 2.1.1.3 is violated, restore RCS pressure and temperature within limits AND be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.3 In MODE 1 or 2, if SL 2.1.2 is violated, restore compliance within limits AND be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.4 In MODES 3, 4, and 5, if SL 2.1.2 is violated, restore RCS pressure to

-:; 2750 psig within 5 minutes.

2.2.5 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

AN0-1 2.0-1 Amendment No. ~.~. 249

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 249 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313

1.0 INTRODUCTION

By application dated June 11, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13162A736), as supplemented by letter dated December 11, 2013 (ADAMS Accession No. ML13347B236), Entergy Operations, Inc. (Entergy, the licensee),

requested changes to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit 1 (AN0-1 ). The supplemental letter dated December 11, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 1, 2014 (79 FR 18331).

The proposed changes add a provision toTS 2.1.1.1 for the determination of the maximum local fuel pin centerline temperature using the NRC reviewed and approved COPERNIC fuel performance computer code. This computer code is used for fuel rod design and analysis of natural, slightly enriched (up to 5 percent) uranium dioxide fuels and urania-gadolinia fuels with the advanced cladding material, M5. AN0-1 is currently authorized to use AREVA Inc.'s Mark B-HTP [high thermal performance] fuel, a fuel type that uses M5 cladding.

2.0 REGULATORY EVALUATION

The license amendment request includes changes to the TS, the contents of which are controlled by requirements in Title 10 of the Code of Federal Regulations (1 0 CFR)

Section 50.36, "Technical specifications." TSs are required to include items in the following five categories related to plant operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. The proposed change would revise the safety limit for fuel centerline melt temperature.

Enclosure 2

Appendix A to 10 CFR Part 50 lists the General Design Criteria (GDC), which are provided to establish minimum requirements for the principal design criteria for a nuclear power plant. In particular, GDC 10, "Reactor design," provides the criterion that during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs) there shall be appropriate margin to assure that the specified acceptable fuel design limits (SAFDLs) are not exceeded. By the application dated June 11, 2013, the licensee stated that AN0-1 was not licensed to the 10 CFR Part 50, Appendix A, GDCs but that it was "designed to be consistent with the GDCs applicable to this application."

The construction permit for AN0-1 was issued by the Atomic Energy Commission (AEC) on December 6, 1968, and an operating license was issued on May 21, 1974. The AN0-1 operating license was issued based on compliance with the proposed GDC published by the AEC in the Federal Register (32 FR 10213) on July 11, 1967 (hereinafter referred to as "draft GDC"). The AEC published the final rule that added Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971 (hereinafter referred to as "final GDC" or "GDC"). In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, with the subject, "SECY-92-223-Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes AN0-1. Draft GDC 6 contains similar language to final GDC 10 regarding reactor core design and SAFDLs. The NRC staff also reviewed the most recently submitted AN0-1 safety analysis report (SAR) (ADAMS Accession No. ML14035A373), and notes that the licensee has incorporated the current GDC 10 into the SAR, section 1.4.6.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," (SRP) Section 4.2, "Fuel System Design," provides NRC guidance for the review of fuel system design, including analysis. One of the design bases for fuel rod failure is the overheating and subsequent melting of fuel pellets, which is the focus of the current license amendment request. This aspect of SRP 4.2 relies on the 10 CFR 50, Appendix A, GDCs as well as 10 CFR 50.34 analysis requirements for its regulatory basis.

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes The licensee's existing TS 2.1.1.1 states the following:

In MODES 1 and 2, the maximum local fuel pin centerline temperature shall be s 5080- (6.5 x 10-3 x (Burnup, MWD/MTU) °F) for TAC02 applications and

5 4642- (5.8 X 1o- X (Burnup, MWD/MTU) °F) for TACO 3 applications.

3

The proposed new wording of TS 2.1.1.1 would be as follows:

In MODES 1 and 2, the maximum local fuel pin centerline temperature shall be 3

S 5080- (6.5 x 10* x (Burnup, MWD/MTU) °F) for TAC02 applications, 3

S 4642- (5.8 x 10* x (Burnup, MWD/MTU) °F) for TACO 3 applications, and

S 4901 °F, decreasing linearly by 13.7 °F per 10,000 MWD/MTU of burnup for COPERNIC applications.

In both the existing and proposed TSs, MWD/MTU means megawatt-days per metric ton of uranium and oF reflects degrees Fahrenheit.

3.2 COPERNIC Code By letters dated April 18, 2002, and June 14, 2002 (ADAMS Accession Nos. ML020070158 and ML021360461, respectively), the NRC staff concluded that the use of the COPERNIC code was acceptable for referencing in licensing applications, to the extent specified and under the limitations delineated in the associated topical report and the associated NRC safety evaluation.

The associated topical report is specified as BAW-1 0231 P, "COPERNIC Fuel Rod Design Computer Code." 1 These two approval letters applied to a version of BAW-10231 specifically addressing the advanced cladding material, M5. By letter dated January 14, 2004 (ADAMS Accession No. ML040150701), the NRC staff approved a later version of BAW-10231 P that contained a section applicable to the use of COPERNIC with mixed-oxide (MOX) fuel. As will be explained later in this safety evaluation, the NRC staff is not authorizing the use of MOX fuel at AN0-1 with this amendment.

3.3 Maximum Local Fuel Pin Centerline Temperature Limit The melting point of nuclear fuel pellets constitutes an SAFDL as defined in 10 CFR Part 50, Appendix A, GDC 10. The purpose of the maximum local fuel pin centerline temperature limit is to ensure that fuel centerline melting will not occur in normal operation or AOOs. Traditionally, it has been assumed that fuel failure will occur if centerline melting takes place. According to SRP Section 4.2, this criterion "was established to assure that axial or radial relocation of molten fuel would neither allow molten fuel to contact the cladding nor produce local hot spots."

The SRP further states that "the assumption that centerline melting results in fuel failure is conservative." Licensees who use M5-clad fuel meet GDC 10 by demonstrating, through analysis with COPERNIC, that the temperature will not exceed the fuel centerline melt limit in normal operation or AOOs based on cycle-specific fuel parameters.

The centerline melt limit, as presented in COPERNIC, decreases linearly with fuel burnup. The original phrasing of the TS limit proposed by Entergy in the submittal dated June 11, 2013, was unclear in indicating that the limit was a linear function of fuel burnup. By letter dated December 11, 2013, in response to an NRC staff question, Entergy clarified the proposed limit to indicate that the functional dependence on burn up is linear, rather than a step function. The NRC staff independently confirmed that the proposed limit of ":S 4901 °F, decreasing linearly by A non-proprietary version of the most recent NRC-approved version of BAW-10231, labeled BAW-10231-NP, Revision 1, (same title) is available at ADAMS Accession Nos. ML042930240 and ML042930247.

13.7 oF per 10,000 MWD/MTU of burnup for COPERNIC applications" is the same, with the appropriate unit conversions, as that presented in the approved COPERNIC topical report.

3.4 Applicability of COPERNIC to AN0-1 The NRC staff's safety evaluation approving the use of COPERNIC code states, in part, that the "COPERNIC computer code is an improved fuel performance code for fuel rod design and analysis of natural, slightly enriched (up to 5 percent) uranium dioxide fuels and urania-gadolinia fuels with the advanced cladding material, M5." The only condition on the use of COPERNIC listed in the SE requires that "Licensees that reference this [COPERNIC] topical report still need to meet 10 CFR 51.52, 'Environmental effects of transportation of [spent] fuel and waste'-

Table S-4." In the technical evaluation supporting its license amendment request, the licensee stated that "the environmental effects analyses of transportation in accordance with 10 CFR 51.52, is performed independent of COPERNIC analyses." The licensee is responsible for compliance with all NRC regulations, including 10 CFR 51.52, unless an approved exemption is applicable and authorized by the NRC. Thus, this approval of the use of COPERNIC, as proposed by the licensee, does not authorize transport, nor otherwise impact the licensee's obligations regarding compliance with the requirements of 10 CFR 51.52.

By letter dated September 12, 2005 (ADAMS Accession No. ML052380280), AN0-1 was authorized by the NRC to use AREVA's Mark B-HTP fuel with M5 cladding. As such, the use of COPERNIC as a fuel performance code for AN0-1 is appropriate. However, in its license amendment request, Entergy provided a reference to Revision 1 of BAW-10231P-A, including with that reference the ADAMS accession number for the NRC staff's safety evaluation of Chapter 13 of the topical report, titled "MOX Application Methodology." BAW-1 0231 P-A, Revision 1, contains additional methodologies for computing the performance of MOX fuel assemblies. While BAW-1 0231 P-A, Revision 1, contains a chapter that addresses the use of MOX fuel, this license amendment does not authorize the use of MOX fuel at AN0-1. The NRC staff notes that AN0-1 TS 4.2.1, "Fuel Assemblies," states, in part, the following:

The reactor shall contain 177 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or M5 clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2 ) as fuel material. ..

This TS ensures that the use of MOX fuel at AN0-1 would not be allowed, without prior NRC review and approval via an additional license amendment request.

3.5 NRC Staff Evaluation Summary Based on the above, the NRC staff concludes that the proposed TS changes incorporating the use of COPERNIC into TS 2.1.1.1 are acceptable. This is based on the following: (1) Entergy has already been authorized for M5 cladding use by a separate license amendment, (2)

COPERNIC has been previously approved by the staff for use with fuel with M5 cladding, (3) the NRC staff has verified independently that the license amendment request proposes to use COPERNIC to the extent specified and under the limitations delineated in the approved topical report and corresponding safety evaluation, and (4) existing TS controls ensure that any use of MOX fuel would require a separate license amendment. In addition, the proposed change revises the TS safety limit for fuel centerline melt temperature in a manner appropriate to the

type of fuel and associated analysis code, and therefore the NRC staff concludes that the proposed change meets the requirements of 10 CFR 50.36(c)(1 ), regarding safety limits.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes 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 amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on April1, 2014 (79 FR 18331). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission 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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: R. Anzalone M. Hardgrove P. Bamford Date: July 9, 2014

ML14169A475 *via memo OFFICE NRR/DORULPL4-1/PM NRR/DORULPL4-1/LA NRR/DSS/SNPB/BC* NRR/DSS/STSB/BC NAME PBamford JBurkhardt JDean REIIiott DATE 6/23/14 6/19/14 04/16/2014 6/24/14 OFFICE NRR/DSS/SRXB/BC* OGC- NLO NRR/DORULPL4-1/BC NRR/DORULPL4-1/PM NAME CJackson CKanatas MMarkley PBamford DATE 04/16/2014 7/7/14 7/9/14 7/9/14