ML17103A509

Issuance of Amendments Allowing the Use of the Copernic Fuel Performance Code

ML17103A509
Person / Time
Site:	Oconee
Issue date:	05/11/2017
From:	Hall J Plant Licensing Branch II
To:	Teresa Ray Duke Energy Carolinas
Hall JR, DORL/LPLII-1, 415-4032
References
CAC MF8158, CAC MF8159, CAC MF8160
Download: ML17103A509 (17)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 11, 2017 Mr. Thomas D. Ray Site Vice President Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752

SUBJECT:

OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3- ISSUANCE OF AMENDMENTS ALLOWING THE USE OF THE COPERNIC FUEL PERFORMANCE CODE (CAC NOS. MF8158, MF8159, AND MF8160)

Dear Mr. Ray:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment Nos. 403, 405, and 404 to Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55, for the Oconee Nuclear Station, Units 1, 2, and 3, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated July 21, 2016.

These amendments revise TS 2.1.1.1, "Reactor Core SLs [Safety Limits]," to specify the limit on maximum local fuel pin centerline temperature based on the use of the COPERNIC fuel performance code. The amendments also add the COPERNIC code topical report as an NRC reviewed and approved fuel performance code to the list of topical reports in TS 5.6.5, "Core Operating Limits Report (COLR)."

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

T. Ray If you have any questions, please contact me at 301-415-4032, or via email at Randy.Hall@nrc.gov.

Sincerely, James R. Hall, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287

Enclosures:

1. Amendment No. 403 to DPR-38

2. Amendment No. 405 to DPR-47

3. Amendment No. 404 to DPR-55

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 403 Renewed License No. DPR-38

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

A. The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. DPR-38, filed by Duke Energy Carolinas, LLC (the licensee), dated July 21, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.

Enclosure 1

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 403, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~-f.Au.-~

Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-38 and the Technical Specifications Date of Issuance: May 11 , 2 o1 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 405 Renewed License No. DPR-47

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

A. The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. DPR-47, filed by Duke Energy Carolinas, LLC (the licensee), dated July 21, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.

Enclosure 2

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.8 of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows:

8. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 405, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-47 and the Technical Specifications Date of Issuance: May 11, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 404 Renewed License No. DPR-55

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

A. The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility), Renewed Facility Operating License No. DPR-55, filed by Duke Energy Carolinas, LLC (the licensee), dated July 21, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act}, and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.

Enclosure 3

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 404, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-55 and the Technical Specifications Date of Issuance: May 11 , 2o1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 403 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO. 405 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 404 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 OCONEE NUCLEAR STATION. UNITS 1. 2, AND 3 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages Licenses Licenses License No. DPR-38, page 3 License No. DPR-38, page 3 License No. DPR-47, page 3 License No. DPR-47, page 3 License No. DPR-55, page 3 License No. DPR-55, page 3 2.0-1 2.0-1 5.0-27 5.0-27

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

8. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 403 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 1l1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-38 Amendment No. 403

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 405 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,-r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-47 Amendment No. 405

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 404 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ~1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-55 Amendment No. 404

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 -----------------------------------------NOTE------------------------------------------

Following transition to the COPERNIC Fuel Performance Methodology the TACO and GDTACO Fuel Performance Methodologies are not applicable.

In MODES 1 and 2, for U02 fuel, the maximum local fuel pin centerline temperature for TACO applications shall be ~ 4656 -

3 2 (5.8 x 10- x (Burnup, MWD/MTU)) - 709.041chil - 786.62(chi) +

1087.07(chi)3 °F where chi is the quantity oxygen-to-uranium ratio minus 2.0. For Gadolinia fuel, the local fuel pin centerline temperature for GDTACO applications shall be~ 4656 - (6.5 x 10- x 3

(Burnup, MWD/MTU)) °F. For COPERNIC applications the maximum 3

local fuel pin centerline temperature shall be s 4901 - (1.37 x 10- x (Burnup, MWD/MTU)) °F. Operation within these limits is ensured by compliance with the Axial Power Imbalance Protective Limits as specified in the Core Operating Limits Report.

2.1.1.2 In MODES 1 and 2, the departure from nucleate boiling ratio shall be maintained greater than the limit of 1.18 for the BWC correlation, 1.19 for the BWU correlation, and 1.132 for the BHTP correlation.

Operation within these limits is ensured by compliance with the Axial Power Imbalance Protective Limits and RCS Variable Low Pressure Protective Limits as specified in the Core Operating Limits Report.

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 2.1.2 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 />.

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

~ 2750 psig within 5 minutes.

OCONEE UNITS 1, 2, & 3 2.0-1 Amendment Nos.4o3,4os, & 404

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

(7) DPC-NE-3000-P-A, Thermal Hydraulic Transient Analysis Methodology; (8) DPC-NE-2005-P-A, Thermal Hydraulic Statistical Core Design Methodology; (9) DPC-NE-3005-P-A, UFSAR Chapter 15 Transient Analysis Methodology :

( 10) BAW-10227-P-A, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel; (11) BAW-10164P-A, RELAP 5/MOD2-B&W-An Advanced Computer Program for Light Water Reactor LOCA and non-LOCA Transient Analysis; and (12) DPC-NE-1006-P-A, Oconee Nuclear Design Methodology Using CASM0-4/SIMULATE-3 (Revision 0, May 2009).

(13) BAW-10231 P-A, COPERNIC Fuel Rod Design Computer Code, FRAMATOME ANP, January 2004.

The COLR will contain the complete identification for each of the Technical Specifications referenced topical reports used to prepare the COLR (i.e.,

report number, title, revision number, report date or NRG SER date, and any supplements).

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 System (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

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

5.6.6 Post Accident Monitoring (PAM) and Main Feeder Bus Monitor Panel (MFPMP)

Report When a report is required by Condition B or G of LCO 3.3.8, "Post Accident Monitoring (PAM) Instrumentation" or Condition D of LCO 3.3.23, "Main Feeder Bus Monitor Panel," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring (PAM only),

the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

OCONEE UNITS 1, 2, & 3 5.0-27 Amendment Nos. 403 , 4os, & 404

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 403 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO. 405 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-47 AND AMENDMENT NO. 404 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DUKE ENERGY CAROLINAS, LLC OCONEE NUCLEAR STATION. UNITS 1, 2. AND 3 DOCKET NOS. 50-269. 50-270. AND 50-287

1.0 INTRODUCTION

By application dated July 21, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16209A223), Duke Energy Carolinas, LLC (Duke, the licensee),

requested changes to the Technical Specifications (TSs) for the Oconee Nuclear Station, Units 1, 2, and 3 (ONS). The proposed amendments would add the COPERNIC Fuel Performance Code to TS 2.1.1.1 and TS 5.6.5.

On October 8, 2009, the U.S. Nuclear Regulatory Commission (NRC) issued Information Notice (IN) 2009-23, "Nuclear Fuel Thermal Conductivity Degradation" (ADAMS Accession No. ML091550527), which discusses the impact of irradiation on fuel thermal conductivity. The IN states that:

It is well understood that irradiation damage and the progressive buildup of fission products in fuel pellets result in reduced thermal conductivity of the pellets.

Thermal performance codes approved by the NRC before 1999 do not include this reduction in thermal conductivity with increasing radiation because earlier test data were inconclusive as to the significance of the effect.

TAC03, "Fuel Pin Thermal Analysis Computer Code, B&W Fuel Company, October 1989 (ADAMS Accession No. ML15040A369) was approved by the NRC prior to 1999, and in late 2009, AREVA issued a Deviation Determination under Title 10 of the Code of Federal Enclosure 4

Regulations (10 CFR) Part 21, which concluded that certain fuel performance codes do not adequately account for fuel thermal conductivity degradation (TCD) with burn up. It was determined that the previously approved codes may under predict fuel temperatures as the result of TCD with burnup. Subsequently, AREVA provided the necessary information for Duke to use to evaluate the impact of the TCD issue on operating limits and take the necessary actions.

The proposed changes to the ONS TSs would add a provision to TS 2.1.1.1 for the determination of the maximum local fuel pin centerline temperature using the NRC reviewed and approved "COPERNIC Fuel Rod Design Computer Code" (ADAMS Accession No. ML0200701S8). The proposed changes would also add the COPERNIC code to the list of NRG-approved topical reports in ONS TS S.6.S.

The COPERNIC code is used for fuel rod design and analysis of natural, slightly enriched (up to S percent) uranium dioxide fuels and urania-gadolinia fuels with the advanced cladding material, MS. ONS Units 1, 2, and 3 are currently authorized to use AREVA lnc.'s Mark 8-High Thermal Performance (HTP) fuel, a fuel type that uses MS cladding. Use of mixed oxide (MOX) fuel is not a part of the license amendment request (LAR) and, therefore, it is not reviewed or evaluated as part of this Safety Evaluation (SE).

2.0 REGULATORY EVALUATION

The following explains the applicability of the 10 CFR Part SO, Appendix A, General Design Criteria (GDC) to ONS. The construction permits for the ONS Units 1, 2, and 3 were issued by the Atomic Energy Commission (AEC) on November 6, 1967. The operating licenses were issued on February 6, 1973, October 6, 1973, and July 19, 1974, respectively. The plant design criteria are discussed in the ONS Updated Final Safety Analysis Report (UFSAR), Chapter 3.1, "Conformance With NRC General Design Criteria," with more details given in the applicable UFSAR sections. These plant-specific "principal design criteria" were based on the proposed AEC General Design Criteria. The AEC published the final rule that added 10 CFR Part SO, Appendix A, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 32SS) on February 20, 1971, with the rule effective on May 21, 1971. As explained in an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "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 10 CFR Part SO, Appendix A, GDC to plants with construction permits issued prior to May 21, 1971.

Therefore, the design criteria which constitute the licensing bases for ONS Units 1, 2, and 3 are those identified in the ONS UFSAR.

As discussed in the UFSAR, the licensee for ONS has made some changes to the facilities over the life of the units, and in making those changes, has committed to certain GDCs from 10 CFR Part SO, Appendix A. The extent to which the Appendix A GDCs have been invoked can be found in specific sections of the UFSAR and in other ONS licensing basis documentation, such as license amendments.

Based on a review of UFSAR Section 3.1 and the licensee's submittal dated July 21, 2016, the NRC staff identified GDC 10, "Reactor design," as being applicable to ONS and to the proposed amendment. GDC 1O specifies that during any condition of normal operation, including the

effects of Anticipated Operational Occurrences (AOOs), there shall be appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-water Reactor] Edition," (Standard Review Plan (SRP)), Section 4.2, "Fuel System Design" (ADAMS Accession No. ML070740002), provides guidance for the NRC staff's review of fuel system design, including analysis. One of the design bases for fuel rod failure is the prevention of overheating and subsequent melting of fuel pellets, which is the focus of this LAR. This aspect of SRP Section 4.2 references the analysis requirements of 10 CFR 50.34 and 10 CFR 50.46 for its regulatory basis.

Section 50.34 includes requirements for the analysis and evaluation of the design and performance of structures, systems, and components (SSCs) of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions, as well as the adequacy of these SSCs (including the fuel) for prevention of accidents and mitigation of their consequences.

Section 50.46 requires each pressurized water reactor and boiling water reactor to be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following a postulated loss of coolant accident conforms to acceptance criteria set forth in the regulation. Acceptance criteria in 10 CFR 50.46 establish both fuel system design limits and core cooling requirements. 10 CFR 50.46(b)(1) requires that the calculated fuel element maximum cladding temperature shall not exceed 2200°F.

The subject LAR proposes changes to the TSs, the contents of which are controlled by the requirements in 10 CFR 50.36, "Technical specifications." TSs are required to include items in the following categories related to plant operation: (a) safety limits, limiting safety system settings, and limiting control settings; (b) limiting conditions for operation; (c) surveillance requirements; (d) design features; and (e) administrative controls. The proposed changes would revise the safety limit for fuel centerline melt temperature.

NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications" (ADAMS Accession No. ML031130447), concluded that it is essential to safety that plants are operated within the bounds of cycle-specific parameter limits. The specific values of these parameter limits may be calculated by licensees through the use of Core Operating Limits Reports (COLR), as incorporated into a facility's TSs. Licensees may calculate cycle-specific parameter limits, provided that these limits are determined using NRG-approved methodologies and are consistent with limits and assumptions described in the plant's Final Safety Analysis Report. The NRC approved the incorporation of the COLR into the ONS TSs in license amendments issued on January 26, 1989 (ADAMS Accession No. ML012000403).

3.0 TECHNICAL EVALUATION

3.1 COPERNIC Code In an April 18, 2002, SE of the COPERNIC fuel performance code topical report (BAW-10231 P)

(ADAMS Accession No. ML020070158), the NRC staff concluded that 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 NRC SE.

The April 18, 2002, SE addressed several major areas of the COPERNIC code, including Maximum Fuel Pin Centerline Temperature, Cladding Corrosion and Hydriding Models, Irradiation Creep, High Stress Creep Model, Fuel Rod Internal Pressure, and Clad Strain. Each of these areas is evaluated for the subject LAR, as discussed below.

3.1.1 Maximum Fuel Pin Centerline Temperature 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 fuel pin centerline temperature limit is to ensure that fuel centerline melting will not occur as a result of 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 centerline melt limit, as presented in COPERNIC, decreases linearly with fuel burnup. In its April 18, 2002, SE, the NRC staff stated that there was good agreement in the fuel temperature predictions between the COPERNIC and NRC audit codes, and concluded that that the thermal conductivity model in the COPERNIC code was acceptable. The licensee used the COPERNIC code to calculate a best-estimate limiting fuel melt temperature of 4901°F for new fuel, decreasing linearly by 13. 7°F per 10,000 megawatt-days per metric ton of uranium (MWD/MTU) of burnup. The NRC staff independently confirmed that the licensee correctly applied the COPERNIC code in calculating the fuel pin centerline temperature limit. Therefore, the licensee's proposed change is consistent with the requirements of GDC 10 and 10 CFR 50.46.

3.1.2 Cladding Corrosion and Hydriding Models The fuel cladding corrosion models in COPERNIC are used to predict corrosion during normal operation and are used as input to loss-of-coolant accident (LOCA) analyses, and account for clad thinning in mechanical analyses. Framatome Cogema Fuels (FCF), now known as Framatome ANP, has provided M5 corrosion data from four foreign plants with significant operating history, which have experienced a maximum rod-average burnup of 40 gigawatt-days per metric ton of uranium (GWd/MTU) (ADAMS Accession Nos. ML003671021 and ML993320125). In its April 18, 2002, SE for the COPERNIC topical report, the NRC staff reviewed the examination of the M5 data from 28 fuel rods from 10 plants that showed that the corrosion level for M5 cladding is less than the 100 micron oxide thickness limit for corrosion at rod-average burnups of 62 GWd/MTU. The corrosion rate predicted by the COPERNIC M5 corrosion model has been demonstrated to be similar to in-reactor data for M5 cladding, up to a rod-average burnup of 62 GWd/MTU.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change to TS 2.1.1.1, allowing the use of the COPERNIC code in calculating cladding corrosion at ONS, is consistent with 10 CFR 50.34 requirements and is acceptable.

3.1.3 Irradiation Creep The M5 creep model in COPERNIC has been modified from that provided in the M5 topical report approved by the NRC (ADAMS Accession Nos. ML003671021 and ML993320125);

however, in its April 18, 2002, SE, the NRC staff found that the differences in predicted creepdown are not significant between the two models, particularly within the first 500 days of irradiation when fuel-clad gap closure takes place. The COPERNIC creep model has added a thermal creep component and an irradiation creep component, with the latter component being the most dominant for in-reactor creep(> 95 percent creep is irradiation creep). A comparison of the M5 creep model to the COPERNIC creep model for the in-reactor creepdown data showed a relatively good fit, as discussed in the NRC's April 18, 2002, SE for the COPERNIC topical report.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change to TS 2.1.1.1, allowing the use of the COPE RN IC code in calculating irradiation creep at ONS, is consistent with 10 CFR 50.34 requirements and is acceptable.

3.1.4 High Stress Creep Model As the NRC staff found in its April 18, 2002, SE, FCF has implemented a high stress model for M5 creep in COPERNIC. The main application of this model in the code is to determine stress relaxation when the fuel and cladding are in hard contact. For the situation of hard fuel-cladding contact, the fuel strain determines the total elastic and uniform plastic cladding strain. The high stress creep model is used to determine the rate of stress relaxation and the ratio of plastic to elastic strain. The NRC staff reviewed this model in its April 18, 2002, SE for the COPERNIC code topical report and concluded that the COPERNIC creep model is conservative and is acceptable.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change to TS 2.1.1.1, allowing the use of the COPERNIC code in calculating stress creep at ONS, is consistent with 10 CFR 50.34 requirements and is acceptable.

3.1.5 Fuel Rod Internal Pressure As a part of the COPERNIC topical report application, FCF had provided an example of the COPERNIC results from a fuel rod internal pressure analysis of the Mark B fuel design. As part of its April 18, 2002, SE for that topical report, the NRC staff performed confirmatory analysis using the FRAPCON-3 steady-state fuel rod performance code from NUREG/CR-6534, Volume 1, "FRAPCON-3: Modifications to Fuel Rod Material Properties and Performance Models for High-Burnup Application" (ADAMS Accession No. ML092950544). The staff's confirmatory analysis showed that FRAPCON-3 predicted slightly greater pressures than COPERNIC, except near the end-of-life where the two predictions were nearly identical. The staff concluded that this was acceptable because most fuel today is licensed for rod-average burn ups greater than 55 GWd/MTU, where the two codes predict similar results.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change to TS 2.1.1.1, allowing the use of the COPE RN IC

code in calculating fuel rod internal pressure at ONS, is consistent with 10 CFR 50.34 requirements and is acceptable.

3.1.6 Clad Strain As the NRC staff stated in its April 18, 2002, SE, SRP section 4.2 suggests a one percent strain limit (elastic plus uniform plastic) on fuel cladding for normal operation and anticipated transients. In general, anticipated transients provide the greatest prediction of clad strain and are, therefore, used for the one percent strain analysis. In its review of the COPE RN IC code topical report, the NRC staff performed confirmatory analyses using FRAPCON-3. A comparison of the COPERNIC and FRAPCON-3 results showed that the two codes were generally in agreement. In the SE for the COPERNIC code topical report, the NRC staff concluded that the method for analyzing clad strain used in the COPERNIC code is acceptable.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, the NRC staff concludes that the proposed change to TS 2.1.1.1, using the COPERNIC code for calculating clad strain at ONS, is consistent with 10 CFR 50.34 requirements and is acceptable.

3.1.7 Applicability of COPERNIC to ONS (Limitations and Conditions)

The NRC staff's April 18, 2002, SE approving the COPERNIC code topical report 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 the COPERNIC code provided in the SE is that "Licensees that reference this topical report still need to meet 10 CFR 51.52, 'Environmental effects of transportation of fuel and waste' - Table S-4." This regulation is not applicable to ONS Units 1, 2, and 3, which were licensed prior to February 4, 1975.

Therefore, based on the NRC's prior approval of the COPERNIC code topical report, and the NRC staff's confirmation that the code is applicable to the ONS fuel design, the staff concludes that the proposed changes to TS 2.1.1.1 and TS 5.6.5, allowing the use of the COPERNIC code at ONS, are consistent with 10 CFR 50.34 requirements and with GL 88-16 recommendations and are acceptable.

3.1.8 Summary Based on the above evaluation, the NRC staff determined that the proposed TS changes incorporating the use of the COPERNIC code into ONS TS 2.1.1.1 and TS 5.6.5 meet the requirements of 10 CFR Part 50, Appendix A, GDC 10, 10 CFR 50.34, and 10 CFR 50.46 regarding safety limits, and are consistent with the guidance of GL 88-16; therefore the staff finds the proposed changes acceptable. The staff's finding is based on the following conclusions: (1) ONS is authorized to use fuel with M5 cladding by a prior license amendment, (2) the COPERNIC code has been previously approved by the staff for use with fuel with M5 cladding, and (3) the use of the COPERNIC code at ONS is not limited or precluded by the limitations and conditions of the NRG-approved SE for the COPERNIC code topical report.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments on April 7, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding, which was published in the Federal Register on February 14, 2017 (82 FR 10593). 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.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: F. Forsaty Date: May11,2017.

T. Ray

SUBJECT:

OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3- ISSUANCE OF AMENDMENTS ALLOWING THE USE OF THE COPERNIC FUEL PERFORMANCE CODE (CAC NOS. MF8158, MF8159, AND MF8160)

DATED: MAY 11, 2017.

DISTRIBUTION:

PUBLIC LPL2-1 R/F RidsACRS_MailCTR Resource RidsNrrDssSrxb Resource RidsNrrDorlDpr Resource RidsNrrDorlLp2-1 Resource RidsNrrDssSrxb Resource RidsNrrLAKGoldstein Resource RidsNrrPMOconee Resource RidsRgn2MailCenter Resource FForsaty, NRR/DSS/SRXB ADAMS Access1on No.: ML17103A509 BWI ML17103A513 *b>Y memo dae t d OFFICE DORL/LPL2-1 /PM DORL/LPL2-1 /LA

' OGC DSS/SRXB/BC DORL/LPL2-1 /BC DORL/LPL2-1 /PM NAME RH all KGoldstein JWachutka EOesterle MMarklev RHall AK/ett for DATE 04/27/17 04/21/17 05/08/17 02/17/2017 05/11/17 05/11/17 OFFICIAL RECORD COPY