ML062140500

Issuance of Amendments Changes to Technical Specifications for Containment Isolation Valves

ML062140500
Person / Time
Site:	Salem
Issue date:	08/31/2006
From:	Stewart Bailey NRC/NRR/ADRO/DORL/LPLB
To:	Levis W Public Service Enterprise Group
SABA F, NRR/DORL/LPLI-2, 415-1453
Shared Package
ML062140496	List: ML062140500
References
TAC MC8544, TAC MC8545
Download: ML062140500 (14)

Text

August 31, 2006 Mr. William Levis Senior Vice 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: CHANGES TO TECHNICAL SPECIFICATIONS FOR CONTAINMENT ISOLATION VALVES (TAC NOS. MC8544 AND MC8545)

Dear Mr. Levis:

The Commission has issued the enclosed Amendment Nos. 274 and 255 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station (Salem), Unit Nos.

1 and 2, respectively. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated September 26, 2005, as supplemented by letter dated June 28, 2006.

These amendments revise the Salem, Units No. 1 and 2, TSs to eliminate certain Surveillance Requirements (SRs) for containment isolation valves. The amendments delete SR 4.6.3.1.1 and SR 4.6.3.1 for Salem Unit Nos. 1 and 2, respectively. These SRs require a complete valve stroke and stroke time measurement when a valve is returned to service after maintenance, repair, or replacement work. These changes are intended to minimize unnecessary testing and plant transients. Other Salem TS containment isolation valve SRs will ensure that the valves remain operable.

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/

Stewart N. Bailey, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosures:

1. Amendment No. 274 to License No. DPR-70

2. Amendment No. 255 to License No. DPR-75

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

August 31, 2006 Mr. William Levis Senior Vice President & Chief Nuclear Officer PSEG Nuclear LLC-X04 Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: CHANGES TO TECHNICAL SPECIFICATIONS FOR CONTAINMENT ISOLATION VALVES (TAC NOS. MC8544 AND MC8545)

Dear Mr. Levis:

The Commission has issued the enclosed Amendment Nos. 274 and 255 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station (Salem), Unit Nos.

1 and 2, respectively. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated September 26, 2005, as supplemented by letter dated June 28, 2006.

These amendments revise the Salem, Units No. 1 and 2, TSs to eliminate certain Surveillance Requirements (SRs) for containment isolation valves. The amendments delete SR 4.6.3.1.1 and SR 4.6.3.1 for Salem Unit Nos. 1 and 2, respectively. These SRs require a complete valve stroke and stroke time measurement when a valve is returned to service after maintenance, repair, or replacement work. These changes are intended to minimize unnecessary testing and plant transients. Other Salem TS containment isolation valve SRs will ensure that the valves remain operable.

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/

Stewart N. Bailey, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosures:

1. Amendment No. 274 to License No. DPR-70

2. Amendment No. 255 to License No. DPR-75

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

PUBLIC LPL1-2 R/F RidsNrrPMSBailey RidsNrrLACRaynor RidsNrrDorlLpl1-2 RidsOgcRp RidsAcrsAcnwMailCenter CAnderson, Rgn-I RidsNrrDorl WBeckner GHill (2) RidsNrrDirsItsb Package Accession Number: ML062140496 Amendment Accession Number: ML062140500 OFFICE LPL1-1/PM LPL1-2/LA CPTB/BC(A) OGC LPL1-2/BC(A)

NAME SBailey:rsa CRaynor TLiu SHamrick BPoole DATE 8/29/06 8/15/06 8/21/06 8/28/06 8/30/06 OFFICIAL RECORD COPY

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

Mr. Dennis Winchester Township Clerk Vice President - Nuclear Assessment Lower Alloways Creek Township PSEG Nuclear Municipal Building, P.O. Box 157 P.O. Box 236 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Mr. Paul Bauldauf, P.E., Asst. Director Mr. Thomas P. Joyce Radiation Protection Programs Site Vice President - Salem NJ Department of Environmental PSEG Nuclear Protection and Energy P.O. Box 236 CN 415 Hancocks Bridge, NJ 08038 Trenton, NJ 08625-0415 Mr. George H. Gellrich Mr. Brian Beam Plant Support Manager Board of Public Utilities PSEG Nuclear 2 Gateway Center, Tenth Floor P.O. Box 236 Newark, NJ 07102 Hancocks Bridge, NJ 08038 Regional Administrator, Region I Mr. Carl J. Fricker U.S. Nuclear Regulatory Commission Plant Manager - Salem 475 Allendale Road PSEG Nuclear - N21 King of Prussia, PA 19406 P.O. Box 236 Hancocks Bridge, NJ 08038 Senior Resident Inspector Salem Nuclear Generating Station Mr. Darin Benyak U.S. Nuclear Regulatory Commission Director - Regulatory Assurance Drawer 0509 PSEG Nuclear - N21 Hancocks Bridge, NJ 08038 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. James Mallon Manager - Licensing 200 Exelon Way, KSA 3-E Kennett Square, PA 19348 Mr. Steven Mannon Manager - Regulatory Assurance P.O. Box 236 Hancocks Bridge, NJ 08038 Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 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. 274 License No. DPR-70

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

A. The application for amendment filed by PSEG Nuclear LLC (the licensee) dated September 26, 2005, as supplemented by letter dated June 28, 2006, 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. 274, 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 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Brooke D. Poole, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: August 31, 2006

ATTACHMENT TO LICENSE AMENDMENT NO. 274 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 Page Insert Page 4 4 Replace the following page of the Appendix A, Technical Specifications, with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Page Insert Page 3/4 6-12 3/4 6-12

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. 255 License No. DPR-75

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

A. The application for amendment filed by PSEG Nuclear LLC (the licensee) dated September 26, 2005, as supplemented by letter dated June 28, 2006, 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. 255, 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 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Brooke D. Poole, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: August 31, 2006

ATTACHMENT TO LICENSE AMENDMENT NO. 255 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Replace the following page of the 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 Page Insert Page 4 4 Replace the following page of the Appendix A, Technical Specifications, with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Page Insert Page 3/4 6-14 3/4 6-14

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 274 AND 255 TO FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 PSEG NUCLEAR 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, 2005, as supplemented by letter dated June 28, 2006, PSEG Nuclear LLC (PSEG or the licensee) requested amendments to the Facility Operating Licenses for the Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. The amendments would revise the Technical Specifications (TSs) to eliminate certain Surveillance Requirements (SRs) on the containment isolation valves. The SRs require a timed full-stroke test prior to returning a valve to service following maintenance or repair. The licensee stated that this test is not always required since certain types of maintenance would not affect a valves performance parameters (e.g., ability to stroke or stroke time), and that performance of this test on some valves requires a plant shutdown. Thus, the purpose of the proposed amendments is to minimize unnecessary testing and plant transients while continuing to assure the necessary quality of the components.

2.0 REGULATORY EVALUATION

The requirements for the content of plant TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36. This regulation requires, in part, that the TSs include SRs to assure that the necessary quality of systems and components is maintained. The Nuclear Regulatory Commission (NRC or the Commission) has provided further guidance on the content of TSs in NUREG-1431, Revision 3, Standard Technical Specifications, Westinghouse Plants.

The requirements for inservice testing are found in 10 CFR 50.55a(f), which specifies testing requirements for pumps and valves. The NRC provided further guidance in NUREG-1482, Revision 1, Guidelines for Inservice Testing at Nuclear Power Plants.

3.0 TECHNICAL EVALUATION

The licensee proposed to delete the requirement that each containment isolation valve be demonstrated to be operable by performance of a cycle test and verification of isolation time prior to returning the valve to service after maintenance, repair, or replacement work is performed on the valve or its associated actuator, control or power circuit. This requirement is

specified in SR 4.6.3.1.1 and SR 4.6.3.1 for Salem Unit No. 1 and Unit No. 2, respectively.

The licensee stated that performance of these SRs typically requires the associated system to be removed from service and, in some instances, closure of containment isolation valves cannot be performed safely without removing the plant from service.

Salem Unit No. 1 SR 4.6.3.1.4 and Salem Unit No. 2 SR 4.6.3.4 require that the isolation time of each power-operated or automatic containment isolation valve be determined to be within its limit when tested pursuant to SR 4.0.5. SR 4.0.5 includes requirements for inservice testing of pumps and valves in accordance with Section Xl of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. The NRC staff provided guidelines and recommendations to licensees for developing and implementing programs for the inservice testing of pumps and valves at commercial nuclear power plants in NUREG-1482, Revision 1.

As discussed in Section 4.4.2 of NUREG-1482, the NRC staff concluded that, as long as certain limitations on the maintenance are met, a stroke test need not be performed if the test is not practical, but a confirmatory test is to be performed at the first opportunity allowed by plant conditions. The effect of Salem Unit No. 1 SR 4.6.3.1.1 and Salem Unit No. 2 SR 4.6.3.1 is to require a valve stroke test even if the maintenance meets the limitations in NUREG-1482. The result is unnecessary plant transients to perform the stroke test.

The licensee stated that post-maintenance testing ensures that equipment meets all applicable SRs before restoring the equipment to operable status. For maintenance activities that could adversely affect isolation time, SR 4.0.1 still requires the isolation time to be determined to be within limits in accordance with Salem Unit No. 1 SR 4.6.3.1.4 and Salem Unit No. 2 SR 4.6.3.4 before the affected valve could be restored to OPERABLE status. The licensee stated that deleting Salem Unit No. 1 SR 4.6.3.1.1 and Salem Unit No. 2 SR 4.6.3.1 would provide flexibility in determining the appropriate post-maintenance test based on the work performed and, when appropriate, a timed valve stroke test will be performed to ensure component operability following maintenance activities. The licensee committed to put administrative controls in place to implement the recommendations in Section 4.4.2 of NUREG-1482, Revision 1.

Based on its review of the information provided by the licensee, the NRC staff finds that the deletion of Salem Unit No. 1 SR 4.6.3.1.1 and Salem Unit No. 2 SR 4.6.3.1 is acceptable because the remaining SRs are sufficient to ensure operability of the containment isolation valves following maintenance. The NRC staff also notes that the proposed change is consistent with NUREG-1431, Revision 3. Further, the licensee has committed to put administrative controls in place to implement the recommendations in Section 4.4.2 of NUREG-1482, Revision 1. Therefore, the NRC staff finds the proposed change acceptable.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The Commissions regulations in 10 CFR 50.92 state that the Commission 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 or consequences of an accident previously evaluated; (2) create the possibility of a new or 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), the licensee provided its analysis of the issue of no significant hazards consideration in its September 26, 2005, amendment request. The NRC staff reviewed the licensees analysis and, based on its review, it appeared that the three standards of 10 CFR 50.92(c) were satisfied. Therefore, the NRC staff proposed to determine that the amendment request involves no significant hazards consideration, and published its proposed determination in the Federal Register for public comment on July 18, 2006 (71 FR 40739):

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment would revise the Technical Specification (TS)

Surveillance Requirements (SRs) for containment isolation valves, consistent with NUREG-1431, Standard Technical Specifications, Westinghouse Plants.

SRs are not initiators to any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The equipment specified in the Limiting Conditions for Operation is still required to be operable and capable of performing the accident mitigation functions assumed in the accident analysis. By performing the analysis, valve operability is maintained. This equipment will continue to be tested in a manner and at a frequency to give confidence that the equipment can perform its intended safety function. As a result, the proposed SR changes do not significantly affect the consequences of any accident previously evaluated.

Therefore, the proposed changes do not involve a significant increase in the probability or radiological consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated in the Updated Final Safety Analysis Report. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. Specifically, no new hardware is being added to the plant as part of the proposed change, no existing equipment is being modified, and no significant changes in operations are being introduced (only certain post-maintenance testing is eliminated leaving operation functions unchanged).

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed changes will not alter any assumptions, initial conditions, or results of any accident analyses. The proposed changes do not affect the operational limits or the physical design of the containment isolation valves. The containment isolation valves will remain capable of performing their design function. Unnecessary testing and associated plant transients will be minimized by the proposed changes. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

On the basis of the above evaluation, the NRC staff concludes that the proposed amendments meet the three criteria of 10 CFR 50.92. Therefore, the staff has made a final determination that the proposed amendments do not involve a significant hazards consideration.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendments. By letter dated December 9, 2005, the State commented that instead of deleting the SRs, the SRs should be revised as follows:

Each containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test and verification of isolation time, unless the conditions of Section 4.4.2 of NUREG-1482, Revision 1, are met.

The NRC staff notes that the wording proposed by the State is the functional equivalent of the change proposed by PSEG, due to the requirements of the remaining SRs discussed in Section 3.0 of this Safety Evaluation. For this reason, after further discussions between the State and PSEG, the State official withdrew the comment during a conference call with the NRC staff on April 19, 2006.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The 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 (71 FR 40739). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

The 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.

Principal Contributor: Stewart Bailey Date: August 31, 2006