ML12299A092

Issuance of Amendment No. 243, Revise Technical Specification 2.0, Safety Limits, to Revise Safety Limit Minimum Critical Power Ratio Values of a Cycle-Specific Calculation

ML12299A092
Person / Time
Site:	Cooper
Issue date:	11/09/2012
From:	Wilkins L E Plant Licensing Branch IV
To:	O'Grady B J Nebraska Public Power District (NPPD)
Wilkins L E
References
TAC ME8853
Download: ML12299A092 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 9, 2012 Mr. Brian J. O'Grady Vice President-Nuclear and CNO Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321 COOPER NUCLEAR STATION -ISSUANCE OF AMENDMENT RE: REVISION OF TECHNICAL SPECIFICATIONS

-SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (TAC NO. ME8853)

Dear Mr. O'Grady:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 243 to Renewed Facility Operating License No. DPR-46 for the Cooper Nuclear Station (CNS). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated May 30, 2012, as supplemented by letters dated October 3 and 31, 2012. The amendment modifies TS Section 2.0, "Safety Limits," by revising the two recirculation loop and single recirculation loop safety limit Minimum Critical Power Ratio (MCPR) values to reflect results of a cycle-specific calculation.

Specifically, the amendment revises the safety limit in TS 2.1.1.2 by changing the value of MCPR for two-loop operation from.:: 1.10 to .:: 1.11 and the value of MCPR for single-loop operation from.:: 1.12 to .:: 1.13. 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. ;;;IY,___

Lynnea E. Wilkins, Project Manager Plant Licensing Branch IV DiVision of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

1. Amendment No. 243 to DPR-46 2. Safety Evaluation cc w/encls: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 243 License No. DPR-46 The U.S. Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Nebraska Public Power District (the licensee), dated May 30,2012, as supplemented by letters dated October 3 and 31,2012, 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; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; 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 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-46 is hereby amended to read as follows: Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 243, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. The license amendment is effective as of its date of issuance and shall be implemented prior to startup from Refueling Outage RE27. FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-46 and Technical Specifications Date of Issuance:

November 9, 2012 ATIACHMENT TO LICENSE AMENDMENT NO. RENEWED FACILITY OPERATING LICENSE NO. DOCKET NO. Replace the following pages of the Renewed Facility Operating License No. DPR-46 and Appendix A Technical Specifications with the enclosed revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Renewed Facility Operating License REMOVE INSERT 3 3 Technical Specifications REMOVE INSERT 2.0-1 2.0-1

-(5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility. This 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 2419 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 243, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection 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 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 contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Cooper Nuclear Station Safeguards Plan," submitted by letter dated May 17, 2006. NPPD shall fully implement and maintain in effect all provisions of the approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The NPPD CSP was approved by License Amendment No. 238. (4) Fire Protection The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Cooper Nuclear Station (CNS) Updated Safety Analysis Report and as approved in the Safety Evaluations dated November 29, 1977; May 23, 1979; November 21, 1980; April 29, 1983; April 16, 1984; June 1,1984; January 3,1985; August 21,1985; April 10, 1986; September 9, 1986; November 7, 1988; February 3, 1989; August 15,1995; and July 31, 1998, 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. Amendment No. 243 2.0 SLs 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow: THERMAL POWER shall be 25% RTP. 2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow: MCPR shall be 1.11 for two recirculation loop operation or > 1.13 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1337 psig. SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods. 2.0-1 Amendment No. 243 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 243 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

By letter dated May 30,2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 12157 A206), and supplemented by letters dated October 3 and 31, 2012. (ADAMS Accession Nos. ML 12285A356 and ML 12312A064), Nebraska Public Power District (NPPD, the licensee) requested an amendment to the Cooper Nuclear Station (CNS) Technical Specification (TS) Section 2.0, "Safety Limits." The amendment would revise two recirculation loop and single recirculation loop Safety Limit Minimum Critical Power Ratio (SLMCPR) values to reflect results of a cycle-specific calculation. Specifically, the amendment would revise the safety limit (SL) in TS 2.1.1.2 by changing the value of Minimum Critical Power Ratio (MCPR) for two-loop operation (TLO) from::: 1.10 to .::: 1.11 and the value of MCPR for single-loop operation (SLO) from::: 1.12 to .::: 1.13. Portions of the letters dated May 30 and October 3, 2012, contain sensitive unclassified non-safeguards information (proprietary) and, accordingly, those portions have been withheld from public disclosure.

The supplemental letters dated October 3 and 31,2012, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 7,2012 (77 FR 47127).

2.0 REGULATORY EVALUATION

2.1 System Description In its letter dated May 30,2012, the licensee provided the following system description:

CNS is a boiling water reactor (BWR) of General Electric BWR4 design, with a Mark 1 containment.

The design of the BWR core and fuel is based on a proper combination of design variables, such as moderator-to-fuel volume ratio, core power density, thermal-hydraulic characteristics, fuel exposure level, nuclear characteristics of the core and fuel, heat transfer, flow distribution, void content, bundle power, and operating pressure.

The CNS Cycle 28 core has 180 GNF2 Enclosure 2

-and 368 GE14 fuel assemblies, and will be licensed by approval of the Cycle 28 Core Operating Limits Report (COLR). Cycle 28 is scheduled to end the middle of November 2014. 2.2 Regulatory Requirements Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The NRC's regulatory requirements related to the content of the TSs are contained in Section 50.36, "Technical specifications," of Title 10 of the Code of Federal Regulations (10 CFR), which requires that the TSs include items in the following specific categories:

(1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

However, the regulation does not specify the particular requirements to be included in TSs. The regulations in 10 CFR 50.36(c){1

)(i){A) state, in part, that Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity.

For the limiting value of critical power ratio (CPR) correlations, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis for Nuclear Power Plants," (SRP), Chapter 4.4, "Thermal and Hydraulic Design," states, in part, that The limiting (minimum) value of ... CPR correlations is to be established such that at least 99.9 percent of the fuel rods in the core will not experience a ... boiling transition during normal operation or ACOs [antiCipated operational occurrences].

The information in the SRP describes adherence to General Design Criterion (GDC) 10, "Reactor design," in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, which state that The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

Regarding the connection between the SRP review guidance and GDC 10, in its letter dated May 30, 2012, the licensee stated, in part, that As part of a reload core design, cycle specific transient analyses are performed to determine the required SLMCPR and the change in Critical Power Ratio (CPR) [ilCPR] for specific transients.

To ensure that adequate margin is maintained, a design requirement based on a statistical analysis was selected in that moderate frequency transients caused by a single operator error or equipment malfunction

-3 [i.e., anticipated operational occurrences as stated in GOC 10] shall be limited such that, considering uncertainties in manufacturing and monitoring the core operating state, at least 99.9% of the fuel rods would be expected to avoid boiling transition.

The lowest allowable transient MCPR limit which meets the design requirement is termed the fuel cladding integrity SLMCPR. CNSls construction predated the issuance of the GOCs in Appendix A 1 to 10 CFR Part 50. CNS is designed to conform to the proposed general design criteria (GOC) published in the Federal Register on July 11, 1967 (32 FR 10213), except where commitments were made to specific 1971 GOC. The Atomic Energy Commission accepted CNS's conformance with the proposed GOC. CNS's conformance to the draft GOC is specified in Appendix F to the CNS Updated Safety Analysis Report (USAR). CNS's USAR Appendix F discussion of Criterion 6, "Reactor Core Design, of Group II, Protection by Multiple Fission Product Barriers" is as follows: The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified.

The core design, together with reliable process and decay heat removal systems shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power. In its letter dated May 30,2012, the licensee explained the tie between the SLMCPR and Criterion 6 of USAR Appendix F as follows: Using the sum of maximum LlCPR and cycle specific SLMCPR to determine the OLMCPR preserves compliance with Criterion 6 of the CNS USAR Appendix F, and the equivalent GOC 10, CNS continues to meet Criterion 6 from the CNS USAR Appendix F. 3.0 TECHNICAL EVALUATION 3.1 Proposed TS Changes Current TS 2.1.1.2 states that MCPR shall be 1.10 for two recirculation loop operation or 1.12 for Single recirculation loop operation.

The 1967 Proposed GDC as described in the CNS Updated Safety Analysis Report, Appendix F, are the licensing basis for CNS; however, the NRC staff concluded in its 1973 Safety Evaluation Report for CNS that the intent of the 1971 Final Rule for 10 CFR Part 50, Appendix A, had also been met.

-Revised TS 2.1.1.2 would state that MCPR shall be 1.11 for two recirculation loop operation or 1.13 for single recirculation loop operation. NRC Staff Evaluation The SLMCPR is determined using a process that statistically convolutes various uncertainties associated with the nuclear and thermal design methods and the fuel manufacturing process. The licensee stated that increases were needed to the SLMCPR values contained in the TS to ensure unrestricted full power operation for an upcoming operating cycle. The updated SLMCPR values were calculated using NRC-approved.

Global Nuclear Fuels-Americas A) reload licensing methods. These include methods described in the following licensing topical reports: NEDE-24011 P-A, General Electric Standard Application for Reactor Fuel, latest applicable revision (proprietary) NEDC-32601 P-A. Methodology and Uncertainties for Safety Limit MCPR Evaluations, August 1999 (proprietary) NEDC-32694P-A, Power Distribution Uncertainties for Safety Limit MCPR Evaluations, August 1999 (proprietary), and NEDO-10958-A.

General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application (non-proprietary).

In light of the fact that the licensee is using NRC-approved methods to calculate the proposed SLMCPR values. the NRC staff reviewed the proposed change to ensure that the generic methods are appropriately applied to CNS. In a request for additional information (RAI) dated September 5,2012 (ADAMS Accession No. ML 12235A252), the NRC staff requested information from the licensee (1) to confirm that Studsvik Scandpower inputs are appropriate in the GNF-A SLMCPR calculative process; (2) to identify cycle design changes that caused the SLMCPR to change; and (3) to confirm the applicability of relevant statistical databases to CNS specifically.

In Section 2.8 of Enclosure 1 to its letter dated May 30,2012, the licensee stated that it requested the application of General Electric Thermal Analysis Basis (GETAB) power distribution methodology and uncertainties, and that a separate enclosure provided the basis for application of this methodology and these uncertainties for the GARDEL core monitoring system. The NRC staff reviewed the applicable uncertainty terms, which relate to core monitoring capabilities under various equipment availability scenarios, and they did not appear to be directly comparable.

In response to the NRC staff's RAI. by letter dated October 3,2012, the licensee identified GARDEL reported uncertainty values, which enabled the staff to confirm that they were, in fact, less than the GETAB values used in the SLMCPR analysis.

Because the GARDEL uncertainties associated with the core monitoring system are less than those assumed in the SLMCPR analysis.

the SLMCPR value is conservative relative to the actual capability of the GARDEL system. Based on this consideration, the NRC staff concludes that the licensee's

-5 use of the GETAB uncertainties in the SLMCPR analysis is appropriate when operating with the GAROEL system. In its review, the NRC staff observed that there was a significant change in assumed batch fraction from the current cycle to the cycle for which the increased SLMCPR was proposed.

Based on this observation, in its RAI dated September 5,2012, the NRC staff requested that the licensee explain the basis for the batch fraction change and its effect on the SLMCPR value. In its RAI response dated October 3, 2012, the licensee stated that an increase in batch fraction was planned to improve full power capability at end of rated conditions.

The licensee also stated that, while the batch fraction does not directly affect the SLMCPR, there is an impact based on the change in radial power distribution due to the increased fresh fuel loading. The NRC staff determined that this assertion is reasonable, since the SLMCPR is affected by the changes in the pin-by-pin and bundle-to-bundle power distributions, and increasing the batch fraction would reasonably be expected to change the bundle-to-bundle power distribution.

Based on this consideration, the NRC staff concludes that the licensee has acceptably addressed its concern related to the increase in batch fraction.

Fuel channel bowing has been shown historically to have an adverse effect on the SLMCPR prediction.

This tendency has been attributed to several different causes, including control blade shadow corrosion and differences between predicted and actual fluence gradients on the fuel channel inner and outer surfaces.

Several notifications in accordance with 10 CFR Part 21 have been made to the NRC discussing these issues, and the various SLMCPR calculative methods include provisions to address channel bow issues. Accordingly, the CNS licensee discussed fuel channel bowing in its SLMCPR increase request. The licensee stated that the GNF-A analyses include an uncertainty to account for an increase in channel bow due to control blade shadow corrosion-induced channel bow. The licensee also stated that channel bow has not been an issue at CNS, and that current practice at CNS is to place a new fuel channel on each fresh fuel assembly when it is initially loaded into the core. By letter dated October 3,2012, the licensee stated that the R-factor uncertaint y 2 used in the SLMCPR analysis is based on a standard deviation in the core average bow, with which PANACEA 3D core simulator calculations confirm that the CNS cycle designs comply. This check is routinely performed as a part of the NRC-approved reload licensing core design and verification process. Because the licensee provided information to confirm that the channel bow value at CNS is appropriate for the analysis values assumed in the SLMCPR determination, and that this value would be confirmed on a cycle-specific basis, the NRC staff determined that the licensee has shown that channel bow is acceptably treated in the CNS SLMCPR calculation.

In its RAI dated September 5, 2012, the NRC staff requested that the licensee clarify its statement regarding its channel bowing practices.

By letter dated October 3, 2012, the licensee confirmed that all fuel in the CNS core received a new fuel channel when it was inserted as a fresh fuel assembly.

The staff concluded that the licensee's response was acceptable because 2 The R-factor uncertainty is one of the uncertainty values that is convoluted with others to determine the SLMCPR. The R-factor uncertainty is a component of the GNF-A method to account for, among others, the channel bow uncertainty.

GNF-A contends that other components of the R-factor uncertainty are proprietary.

-it clarified that the current practice discussed in the license amendment request is applicable to all fuel in the core. NRC Staff Conclusion Based on the following considerations, the NRC staff concludes that The licensee performed its SLMCPR calculations using NRC-approved methods; The licensee confirmed the applicability of generic uncertainty values for the CNS core design and operating configuration; and The licensee provided additional information to confirm that its calculational treatment of channel bow is appropriate relative to CNS-specific operating practice and predicted design values. Based on the above, the NRC staff concludes that the licensee's proposed revision to the SLMCPR value is acceptable.

The SLMCPR will continue to provide assurance that 99.9 percent of the fuel rods in the core will not exceed the critical power ratio, and that fuel cladding integrity will be maintained under conditions of normal operation and with appropriate margin for anticipated operational occurrences, consistent with GDC 10 and CNS UFSAR Appendix F, Design Criterion

6. Additionally, the changes will not impact the licensee's compliance with the regulatory requirements of 10 CFR 50.36(c)(1

)(i)(A). FINAL NO SIGNIFICANT HAZARDS CONSIDERATION The NRC's regulations in 10 CFR 50.92 state that the NRC may make a final determination that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not: (1) involve a significant increase in the probability different kind of accident from any accident previously evaluated; or (3) involve a Significant reduction in a margin of safety. As required by 10 CFR 50.91 (a), an evaluation of the issue of no significant hazards consideration is presented below: Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No. Miscellaneous radioactive sources are not part of any transient or accident analysis.

The proposed changes conform to the Nuclear Regulatory Commission's (NRC's) regulatory guidance regarding the content of plant TS as identified in 10 CFR 50.36 and NRC publication NUREG-1432.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

-7 Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No. The proposed change relocates the requirements for leak checking miscellaneous radioactive material sources to a licensee controlled document subject to the controls of 10 CFR 50.59. This change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Hence, the proposed change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. Does the proposed amendment involve a significant reduction in a margin of safety? Response:

No. The proposed change relocates the requirements for leak checking miscellaneous radioactive material sources to a licensee controlled document subject to the controls of 10 CFR 50.59. This change does not alter any safety margins. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied.

Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91. STATE CONSULTATION In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment.

The State official had no comments. 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 efnuents 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

-8 significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on August 7,2012 (77 FR 47127). 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.

7.0 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 Contributor:

B. Parks Date: November 9, 2012 November 9,2012 Mr. Brian J. O'Grady Vice President-Nuclear and CNO Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321 COOPER NUCLEAR STATION -ISSUANCE OF AMENDMENT RE: REVISION OF TECHNICAL SPECIFICATIONS

-SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (TAC NO. ME8853)

Dear Mr. O'Grady:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 243 to Renewed Facility Operating License No. DPR-46 for the Cooper Nuclear Station (CNS). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated May 30, 2012, as supplemented by letters dated October 3 and 31, 2012. The amendment modifies TS Section 2.0, "Safety Limits," by revising the two recirculation loop and single recirculation loop safety limit Minimum Critical Power Ratio (MCPR) values to reflect results of a cycle-specific calculation.

Specifically, the amendment revises the safety limit in TS 2.1.1.2 by changing the value of MCPR for two-loop operation from 1.10 to 1.11 and the value of MCPR for single-loop operation from 1.12 to 1.13. 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, IRA! Lynnea E. Wilkins, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

1. Amendment No. 243 to DPR-46 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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