ML072210499

Issuance of Amendments Topical Report References in Technical Specification for Core Operating Limits Report (Tac Nos. MD3171 and MD3172)

ML072210499
Person / Time
Site:	Salem
Issue date:	08/23/2007
From:	Richard Ennis NRC/NRR/ADRO/DORL/LPLI-2
To:	Levis W Public Service Enterprise Group
Ennis R, NRR/DORL, 415-1420
Shared Package
ML072210466	List: ML072210499
References
TAC MD3171, TAC MD3172
Download: ML072210499 (15)

Text

August 23, 2007 Mr. William Levis President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: TOPICAL REPORT REFERENCES IN TECHNICAL SPECIFICATION FOR CORE OPERATING LIMITS REPORT (TAC NOS.

MD3171 AND MD3172)

Dear Mr. Levis:

The Commission has issued the enclosed Amendment Nos. 284 and 267 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated September 26, 2006, as supplemented by letter dated May 14, 2007.

The amendments revise TS 6.9.1.9, Core Operating Limits Report (COLR), to remove the revision numbers and dates from the list of topical reports that contain the analytical methods used in the COLR. The Salem Unit 2 amendment also adds a new topical report to the list of COLR methods referenced in TS 6.9.1.9.

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/ra/

Richard B. Ennis, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosures:

1. Amendment No. 284 to License No. DPR-70

2. Amendment No. 267 to License No. DPR-75

3. Safety Evaluation cc w/encls: See next page

August 23, 2007 Mr. William Levis President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: TOPICAL REPORT REFERENCES IN TECHNICAL SPECIFICATION FOR CORE OPERATING LIMITS REPORT (TAC NOS.

MD3171 AND MD3172)

Dear Mr. Levis:

The Commission has issued the enclosed Amendment Nos. 284 and 267 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated September 26, 2006, as supplemented by letter dated May 14, 2007.

The amendments revise TS 6.9.1.9, Core Operating Limits Report (COLR), to remove the revision numbers and dates from the list of topical reports that contain the analytical methods used in the COLR. The Salem Unit 2 amendment also adds a new topical report to the list of COLR methods referenced in TS 6.9.1.9.

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/ra/

Richard B. Ennis, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosures:

1. Amendment No. 284 to License No. DPR-70

2. Amendment No. 267 to License No. DPR-75

3. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

PUBLIC RidsOgcRp RidsNrrLAABaxter LPL1-2 R/F RidsAcrsAcnwMailCenter RidsNrrDorlLpl1-2 RidsNrrDirsltsb RidsNrrLASLittle RidsRgn1MailCenter RidsNrrPMREnnis GHill (2), OIS RidsNrrDorlDPR FOrr Package Accession No.: ML072210466 Amendment Accession No: ML072210499 TS Accession Nos.: Amd.284/ML072390186; Amd.267/ML072390188 OFFICE LPL1-2/PM LPL1-2/LA LPL1-1/LA SRXB/BC ITSB/BC OGC LPL1-2/BC NAME REnnis ABaxter SLittle GCranston TKobetz BMizuno Hchernoff (GEMiller for)

DATE 8/14/07 8/15/07 8/15/07 8/9/07 8/20/07 8/23/07 8/23/07 OFFICIAL RECORD COPY

Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc:

Mr. Thomas Joyce Senior Vice President - Operations PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Dennis Winchester Vice President - Nuclear Assessment PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Robert Braun Site Vice President - Salem PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Carl Fricker Vice President - Operations Support PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. George Gellrich Plant Manager - Salem PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. James Mallon Manager - Licensing PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Steven Mannon Manager - Salem Regulatory Assurance PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Township Clerk Lower Alloways Creek Township Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Mr. Paul Bauldauf, P.E., Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415 Mr. Brian Beam Board of Public Utilities 2 Gateway Center, Tenth Floor Newark, NJ 07102 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector Salem Nuclear Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038

PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 284 License No. DPR-70 1.

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

A.

The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated September 26, 2006, as supplemented by letter dated May 14, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR), Chapter I; B.

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

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

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 Facility Operating License No. DPR-70 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 284, 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 prior to restart from the 19th refueling outage in fall 2008. Implementation shall include revision of the Core Operating Limits Report as described in Section 4.1 of Attachment 1 of the licensees application dated September 26, 2006.

FOR THE NUCLEAR REGULATORY COMMISSION

/ra/ (Emiller for)

Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and the Technical Specifications Date of Issuance: August 23, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 284 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Replace the following page of Facility Operating License No. DPR-70 with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert Page 4 Page 4 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 6-24 6-24 6-24a 6-24a

PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 267 License No. DPR-75 1.

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

A.

The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated September 26, 2006, as supplemented by letter dated May 14, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR), Chapter I; B.

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

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

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 Facility Operating License No. DPR-75 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 267, 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 prior to restart from the 16th refueling outage in spring 2008. Implementation shall include revision of the Core Operating Limits Report as described in Section 4.1 of of the licensees application dated September 26, 2006.

FOR THE NUCLEAR REGULATORY COMMISSION

/ra/ Ed Miller for Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and the Technical Specifications Date of Issuance: August 23, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 267 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Replace the following page of Facility Operating License No. DPR-75 with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert Page 4 Page 4 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 6-24 6-24 6-24a 6-24a

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 284 AND 267 TO FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311

1.0 INTRODUCTION

By letter dated September 26, 2006, as supplemented by letter dated May 14, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession Nos.

ML062780237 and ML071410454, respectively), PSEG Nuclear LLC (PSEG or the licensee) submitted a request for changes to the Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2, Technical Specifications (TSs). The proposed amendments would revise TS 6.9.1.9, Core Operating Limits Report (COLR), to remove the revision numbers and dates from the list of topical reports that contain the analytical methods used in the COLR. For Salem Unit 2, the proposed amendment would also add a new topical report to the list of COLR methods referenced in TS 6.9.1.9.

The supplement dated May 14, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on November 7, 2006 (71 FR 65143).

2.0 REGULATORY EVALUATION

In an effort to avoid TS changes for every fuel reload cycle that results in changes to the cycle-specific parameter limits, licensees have relocated the cycle-specific core operating parameters from the TSs to the COLR, which is a licensee-controlled document. Generic Letter (GL) 88-16, Removal of Cycle-Specific Parameter Limits From Technical Specifications, dated October 3, 1988, provides guidance for the preparation of license amendment requests to relocate cycle-specific TS information to the COLR. The guidance in GL 88-16 states that licensees shall identify (in the Administrative Controls, Reporting Requirements section of the TSs), the previously-approved analytical methods used to determine the core operating limits by identifying the topical report number, title, and date (or identify the staffs safety evaluation (SE) report for a plant-specific methodology by NRC letter and date).

In a letter dated December 15, 1999 (ADAMS Accession No. ML993540351), the NRC staff accepted a method proposed by Siemens Power Corporation of referencing approved topical reports. The proposed method would allow licensees to use current topical reports to support limits in the COLR without having to submit an amendment request for the facility operating license each time a revision to the topical report is approved by the NRC. This method would allow the references to approved topical reports in the TSs to be cited using the report number and title. The citation in the COLR would include specific information for each of the TS references to topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). This method of referencing was subsequently incorporated into the Standard TSs (i.e., NUREGs 1430 through 1434) via NRC approval of TS Task Force (TSTF)

Traveler 363, Revise Topical Report References in ITS 5.6.5, COLR. As discussed in the letter dated December 15, 1999, revisions to the topical reports listed in the COLR would be controlled under the requirements contained in Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR).

For Salem Unit 2, the proposed amendment would revise the COLR topical report references in TS 6.9.1.9 to add a reference to WCAP-10054-P-A, Addendum 2, "Addendum to the Westinghouse Small Break ECCS [emergency core cooling system] Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model."

This Westinghouse small break loss-of-coolant accident (SBLOCA) methodology was generically approved by the NRC in an SE transmitted by letter dated August 12, 1996 (Reference 3). The NRC staffs review of the acceptability of this proposed change to TS 6.9.1.9 focused on the applicability of WCAP-10054-P-A, Addendum 2, to Salem Unit 2, and that the SBLOCA analysis results satisfy the requirements specified in 10 CFR 50.46(b).

3.0 TECHNICAL EVALUATION

3.1 Removal of Revision Numbers and Dates from TS 6.9.1.9 Topical Report References For Salem Units 1 and 2, TS 6.9.1.9.b references the topical reports, previously reviewed and approved by the NRC, that describe the analytical methods used to determine the core operating limits. Currently, the topical report references in TS 6.9.1.9.b include the topical report numbers, titles, dates (of the report or the NRC SE approving the report) and revision numbers as applicable. The proposed amendment would remove the revision numbers and dates from TS 6.9.1.9.b such that each topical report would be referenced only by the report number and title. The Salem COLR currently references TS 6.9.1.9 rather than specifically identifying the topical reports that are used. The licensees application dated September 26, 2006, indicated that as part of the implementation of the requested amendment, the COLR format will be revised to specifically identify the topical reports including the dates and revisions (i.e., information that is currently in TS 6.9.1.9.b).

Section 4.3, Nuclear Design, Section 4.4, Thermal and Hydraulic Design, and Section 4.5, Reload Analyses, of the Salem Updated Final Safety Analysis Report (UFSAR) discuss the COLR and the associated methods of evaluating the core operating limits. Since the COLR methods are described in the UFSAR, proposed changes to the methods (e.g., licensee desiring to use a newer NRC-approved revision of a topical report) would be subject to the provisions of 10 CFR 50.59. Specifically, 10 CFR 50.59(c)(2)(viii), requires that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90, prior to implementing a proposed change, if the change would [r]esult in a departure from a method of evaluation described in the Final Safety Analysis Report (as updated) used in establishing the design basis or in the safety analyses. The definition in 10 CFR 50.59(a)(2) states that departure from a method of evaluation described in the Final Safety Analysis Report (as updated) used in establishing the design basis or in the safety analyses means: (i) changing any of the elements of the method described in the FSAR unless the results of the analysis are conservative or essentially the same; or (ii) changing from a method described in the FSAR to another method unless that method has been approved by the NRC for the intended application.

Based on its review, the NRC staff finds that: (1) the proposed TS changes are consistent with TSTF-363; (2) the proposed amendment does not alter the current TS 6.9.1.9.b requirement that core operating limits be determined using analytical methods previously approved by the NRC; and (3) future revisions to the COLR, to use updated revisions of the referenced topical reports, will be adequately controlled under the provisions of 10 CFR 50.59. Based on these considerations, the NRC staff concludes that the proposed removal of revision numbers and dates from TS 6.9.1.9.b for Salem Units 1 and 2 are acceptable.

3.2 Addition of WCAP-10054-P-A, Addendum 2, to TS 6.9.1.9 for Salem Unit 2 The licensees SBLOCA analysis of record for Salem Units 1 and 2 was completed using Topical Report WCAP-10054-P-A, Revision 1, Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code, which is referenced in TS 6.9.1.9.b.3 for both Salem Units. In support of the replacement of the Salem Unit 2 steam generators (SGs), planned for the spring 2008 refueling outage, Westinghouse reanalyzed the SBLOCA for Salem Unit 2. The analysis modeled the replacement SGs and used the methodology described in WCAP-10054-P-A, Addendum 2, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model." The COSI model provides increased condensation efficiencies for safety injection flows to the faulted and intact reactor coolant system loops, thereby resulting in lower calculated peak cladding temperature (PCT). Because the COSI model represents a change in ECCS evaluation methods, the proposed amendment would add a reference to WCAP-10054-P-A, Addendum 2, to TS 6.9.1.9.b for Salem Unit 2. Note, the SBLOCA analysis performed to support the replacement of the Salem Unit 2 SGs is not applicable to Salem Unit 1. As such, the licensee did not request implementation of the COSI model for Salem Unit 1.

As discussed above, WCAP-10054-P-A, Addendum 2, was generically approved by the NRC in an SE transmitted by letter dated August 12, 1996 (Reference 3). Section 3.0 of the SE indicated that the NRCs approval was based on the following limitations:

(1) the model will be used within the experimental range of injection jet velocities (i.e.,

0.1 feet/sec (ft/sec) to 30 ft/sec); and (2) the model will be applied within the pressure range of 550 to 1200 pounds per square inch atmospheric (psia).

Section 4.2 of Attachment 1 of the licensees application dated September 26, 2006, addressed the above limitations as follows:

Salem Unit 2 RSG [replacement steam generator] SBLOCA analysis complies with the above limitations. Internal guidance used by the Westinghouse analysts provides assurance that the SBLOCA analysis satisfies all restrictions and/or requirements imposed by the Nuclear Regulatory Commission (NRC). PSEG and Westinghouse have ongoing processes to assure that the values and ranges of the SBLOCA analyses parameter inputs conservatively bound the values and ranges of the as-operated plant for those parameters. Following implementation of the proposed change, SBLOCA analyses will continue to demonstrate compliance to 10 CFR 50.46 using NRC approved methods and plant-specific input assumptions consistent with the conditions of NRC approval of the analytical methods.

The NRC staff concludes that use of WCAP-10054-P-A, Addendum 2, is acceptable for Salem Unit 2 because: (1) the Salem Unit 2 SBLOCA analysis was performed within the limitations cited in the SE generically approving WCAP-10054-P-A, Addendum 2; and (2) PSEG and Westinghouse have ongoing processes to assure that the values and ranges of the SBLOCA analyses parameter inputs conservatively bound the values and ranges of the as-operated plant for those parameters.

The NRC staff reviewed the results of the Salem Unit 2 SBLOCA analysis, submitted by the licensee to support the proposed amendment, to confirm that the results satisfy the requirements specified in 10 CFR 50.46(b). Specifically, 10 CFR 50.46(b)(1) through (b)(5) provide the following acceptance criteria for ECCS cooling performance following postulated LOCAs:

(1)

Peak cladding temperature - The calculated maximum fuel element cladding temperature shall not exceed 2200 EF.

(2)

Maximum cladding oxidation - The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

(3)

Maximum hydrogen generation - The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(4)

Coolable geometry - Calculated changes in core geometry shall be such that the core remains amenable to cooling.

(5)

Long-term cooling - After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

In its submittal dated May 14, 2007, the licensee provided the results of the Salem Unit 2 SBLOCA analyses performed in accordance with the Westinghouse COSI SBLOCA methodology. The results are provided in the following table along with the acceptance criteria of 10 CFR 50.46(b).

TABLE 1 - SBLOCA ANALYSIS RESULTS - SALEM UNIT 2 Parameter Limiting Break Sizes 10 CFR 50.46 Acceptance Criteria 2-inch 3-inch 4-inch PCT 910 EF 987 EF 964 EF

< 2200 EF (10 CFR 50.46(b)(1))

Maximum local oxidation 0.00 %

0.01 %

0.01 %

< 17.0 %

(10 CFR 50.46(b)(2))

Total hydrogen generation

< 1 %

< 1 %

< 1 %

< 1.0 %

(10 CFR 50.46(b)(3))

The NRC staff concludes that the results of the Salem 2 SBLOCA analyses (as shown in the table above) meet the acceptance criteria of 10 CFR 50.46(b)(1) through (b)(3). Meeting these criteria provides reasonable assurance that, at the current licensed power level (3459 megawatts thermal), the Salem Unit 2 core will be amenable to cooling as required by 10 CFR 50.46(b)(4). The capability of Salem Unit 2 to satisfy the long-term cooling requirements of 10 CFR 50.46(b)(5) is unaffected by this amendment.

3.3 Technical Evaluation Conclusion

Based on the evaluation discussed in SE Section 3.1 and 3.2, the NRC staff concludes that the proposed amendment is acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). 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) 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.

7.0 REFERENCES

1.

Letter, Thomas P. Joyce (PSEG) to NRC, Request for Amendment to Technical Specifications to Revise Topical Report References in Technical Specification 6.9.1.9, Core Operating Limits Report, dated September 26, 2006.

2.

WCAP-10054-P-A, Addendum 2, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, July 1997.

3.

Letter from Robert C. Jones (NRC), to N.J. Liparulo (Westinghouse Electric Corporation), WCAP-10054-P, Addendum 2, Revision 1, NOTRUMP SBLOCA Using the COSI Steam Condensation Model, (TAC No. M90784), dated August 12, 1996.

4.

Letter, Thomas P. Joyce (PSEG) to NRC, Response to Request for Additional Information Request for License Amendment - Topical Report References in Technical Specification 6.9.1.9, Core Operating Limits Report, dated May 14, 2007.

Principal Contributors: F. Orr R. Ennis Date: August 23, 2007