ML100200644

Issuance of Amendments Relocation of Certain Technical Specification Requirements Associated with Refueling Operations

ML100200644
Person / Time
Site:	Salem
Issue date:	02/17/2010
From:	Richard Ennis Plant Licensing Branch 1
To:	Joyce T Public Service Enterprise Group
Ennis R, NRR/DORL, 415-1420
References
TAC ME1279, TAC ME1280
Download: ML100200644 (26)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 17, 2010 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear P.O. Box 236, N09 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: RELOCATION OF CERTAIN TECHNICAL SPECIFICATION REQUIREMENTS ASSOCIATED WITH REFUELING OPERATIONS (TAC NOS. ME1279 AND ME1280)

Dear Mr. Joyce:

The Commission has issued the enclosed Amendment Nos. 293 and 277 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2.

These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated April 9,2009, as supplemented by letter dated October 23,2009.

The amendments relocate TS requirements pertaining to communications during refueling operations (TS 3/4.9.5), manipulator crane operability (TS 3/4.9.6), and crane travel (TS 3/4.9.7) to the Technical Requirements Manual.

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

Sincerely, Richard B. Ennis, 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. 293 to License No. DPR-70

2. Amendment No. 277 to License No. DPR-75

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 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. 293 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 April 9, 2009, as supplemented by letter dated October 23, 2009, 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 (2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 293, 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 l!/iC?-AI

~'mld K. Chernoff, chi Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and the Technical Specifications Date of Issuance: February 17, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 293 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 4

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 IX IX 3/4 9-5 3/4 9-5 3/4 9-6 3/4 9-6 3/49-7 3/4 9-7

4 (1)

Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at a steady state reactor core power level not in excess of 3459 megawatts (one hundred percent of rated core power).

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 293, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(3)

Deleted Per Amendment 22, 11-20-79 (4)

Less than Four Loop Operation PSEG Nuclear LLC shall not operate the reactor at power levels above P-7 (as defined in Table 3.3-1 of Specification 3.3.1.1 of Appendix A to this license) with less than four (4) reactor coolant loops in operation until safety analyses for less than four loop operation have been submitted by the licensees and approval for less than four loop operation at power levels above P-7 has been granted by the Commission by Amendment of this license.

(5)

PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, and as approved in the NRC Safety Evaluation Report dated November 20, 1979, and in its supplements, subject to the following provision:

PSEG Nuclear LLC 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. 293

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION 3/4.9.2 INSTRUMENTATION 3/4.9.3 DECAY TIME 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3/4.9.5 DELETED 3/4.9.6 DELETED 3/4.9.7 DELETED 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION All Water Levels Low Water Level 3/4.9.9 DELETED 3/4.9.10 WATER LEVEL -

REACTOR VESSEL 3/4.9.11 STORAGE POOL WATER LEVEL 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN....

3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS 3/4.10.3 PHYSICS TESTS 3/4.10.4 NO FLOW TESTS PAGE 3/4 9-1 3/4 9-2 3/4 9-3 3/4 9-4 3/4 9-5 3/4 9-6 3/4 9-7 3/4 9-8 3/4 9-8a 3/4 9-10 3/4 9-11 3/4 9-12 3/4 10-1 3/4 10-2 3/4 10-3 3/4 10-4 SALEM -

UNIT 1 IX Amendment No. 293

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UNIT 1 3/4 9-5 Amendment No. 293

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UNIT 1 3/4 9-6 Amendment No. 293

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK SALEM -

UNIT 1 3/4 9-7 Amendment No. 293

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 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. 277 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) April 9, 2009, as supplemented by letter dated October 23, 2009, 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 (2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 277, 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 i{(~,~:1 Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and the Technical Specifications Date of Issuance: February 17, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 277 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 4

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 IX IX 3/4 9-5 3/49-5 3/4 9-6 3/49-6 3/49-7 3/49-7

4 (2)

Technical Specifications and Environmental Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 277, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(3)

Special Low Power Test Program PSE&G shall complete the training portion of the Special Low Power Test Program in accordance with PSE&G's letter dated September 5, 1980 and in accordance with the Commission's Safety Evaluation Report "Special Low Power Test Program~,

dated August 22, 1980 (See Amendment No. 2 to DPR-75 for the Salem Nuclear Generating Station, Unit No.2) prior to operating the facility at a power level above five percent.

Within 31 days following completion of the power ascension testing program outlined in Chapter 13 of the Final Safety Analysis Report, PSE&G shall perform a boron mixing and cooldown test using decay heat and Natural Circulation.

PSE&G shall submit the test procedure to the NRC for review and approval prior to performance of the test.

The results of this test shall be submitted to the NRC prior to starting up following the first refueling outage.

(4)

Initial Test Program PSE&G shall conduct the post-fuel-loading initial test program (set forth in Chapter 13 of the Final Safety Analysis Report, as amended) without making any major modifications of this program unless modifications have been identified and have received prior NRC approval.

Major modifications are defined as:

(a)

Elimination of any test identified in Chapter 13 of the Final Safety Analysis Report, as amended, as essential; (b)

Modification of test objectives, methods or acceptance criteria for any test identified in Chapter 13 of the Final Safety Analysis Report, as amended, as essential; (c)

Performance of any test at a power level different by more than five percent of rated power from there described; and Amendment No. 277

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION 3/4.9.2 INSTRUMENTATION 3/4.9.3 DECAY TIME 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3/4.9.5 DELETED 3/4.9.6 DELETED 3/4.9.7 DELETED 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION All Water Levels Low Water Level 3/4.9.9 DELETED 3/4.9.10 WATER LEVEL -

REACTOR VESSEL 3/4.9.11 STORAGE POOL WATER LEVEL 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN....

3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS 3/4.10.3 PHYSICS TESTS 3/4.10.4 NO FLOW TESTS PAGE 3/4 9-1 3/4 9-2 3/4 9-3 3/4 9-4 3/4 9-5 3/4 9-6 3/4 9-7 3/4 9-8 3/4 9-9 3/4 9-11 3/4 9-12 3/4 9-13 3/4 10-1 3/4 10-2 3/4 10-4 3/4 10-5 SALEM -

UNIT 2 IX Amendment No. 277

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK SALEM -

UNIT 2 3/4 9-5 Amendment No. 277

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK SALEM -

UNIT 2 3/4 9-6 Amendment No. 277

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK SALEM -

UNIT 2 3/4 9-7 Amendment No. 277

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 293 AND 277 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 April 9, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML091340092), as supplemented by letter dated October 23,2009 (ADAMS Accession No. ML093070506), 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 requested changes would relocate certain TS requirements associated with refueling operations to the Salem Technical Requirements Manual (TRM).

Specifically, the proposed amendment would relocate TS requirements pertaining to communications during refueling operations (TS 3/4.9.5), manipulator crane operability (TS 3/4.9.6), and crane travel (TS 3/4.9.7) to the TRM.

The supplement dated October 23, 2009, 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) staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 25,2009 (74 FR 42929).

2.0 REGULATORY EVALUATION

2.1 Regulatory Discussion The Commission's regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications."

This regulation requires that the TSs include items in the following five specific categories:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LeOs); (3) surveillance requirements; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TSs.

Enclosure

-2 On July 22, 1993 (58 FR 39132), the Commission published a "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (Final Policy Statement) which discussed the criteria to determine which items are required to be included in the TSs as LCOs. The criteria were subsequently incorporated into the regulations by an amendment to 10 CFR 50.36 (60 FR 36953, July 19, 1995). Specifically, 10 CFR 50.36(c)(2)(ii) requires that a TS LCO be established for each item meeting one or more of the following criteria:

Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4 A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

As discussed in the Federal Register notice for the final rule dated July 19, 1995 (60 FR 36955):

LCOs that do not meet any of the criteria, and their associated actions and surveillance requirements, may be proposed for relocation from the technical specifications to licensee-controlled documents, such as the FSAR [Final Safety Analysis Report]. The criteria may be applied to either standard or custom technical specifications.

As discussed in the licensee's letter dated October 23, 2009, the proposed amendment would relocate certain TS requirements related to refueling operations to the Salem TRM, which is a licensee-controlled document. The TRM is described in Section 13.5, "Plant Procedures," of the Salem Updated Final Safety Analysis Report (UFSAR). Specifically, UFSAR Section 13.5.4 reads as follows:

13.5.4 Technical Requirements Manual (TRM)

The Technical Requirements Manual (TRM) contains technical requirements and/or supporting information (e.g., tables and component lists) which were once contained in the SGS [Salem Generating Station] Technical Specifications (TS)

(Le., Appendix A of the SGS Facility Operating License). Removal of the TS and information is approved by the NRC through individual TS amendments. The TRM is intended to provide operational guidance and requirements for various

-3 plant conditions, actions, and testing similar to TS, however, these requirements are in accordance with licensing commitments. These changes add the TRM into the scope of procedures to be processed through the Station Qualified Reviewer (SQR) process and reviewed by PORC [Plant Operations Review Committee].

Future changes to the relocated requirements and supporting information are processed in accordance with section 17.2 of the UFSAR, and are subject to a 10 CFR 50.59 Review. All non-editorial changes are reviewed by PORCo The TRM is comprised of an index, the individual specification and bases. The manual is intended to provide a single location for the relocated TS items as a convenience for operations and other station personnel. The individual sections of the TRM contain the relocated licensing commitments which are subject to the provisions of 10 CFR 50.59 described above, and are controlled in accordance with the applicable established procedure process.

Nuclear Energy Institute (NEI) guidance document NEI 98-03, Revision 1, "Guidelines for Updating Final Safety Analysis Reports" (ADAMS Accession No. ML003779028) lists the following methods of controlling the TRM on page 7 of Appendix A:

The TRM or other licensee controlled document is explicitly "incorporated by reference" into the UFSAR. Under this approach, the referenced document is subject to the change control requirements of 10 CFR 50.59 and the update/reporting requirements of 10 CFR 50.71 (e), e.g., periodic submittal of change pages, etc.

The TRM or other licensee controlled document is treated in a manner consistent with procedures fully or partially described in the UFSAR. Under this approach, the referenced document is maintained on-site in accordance with licensee administrative processes, and changes are evaluated using 10 CFR 50.59.

Regulatory Guide (RG) 1.181, "Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71 (e)" dated September 1999 (ADAMS Accession No. ML992930009), states that Revision 1 of NEI 98-03 provides methods that are acceptable to the NRC staff for complying with the provisions of 10 CFR 50.71 (e).

2.2 Current TS Requirements and Equipment Description TS 3/4.9.5: Refueling Operations - Communications LCO 3.9.5 requires that direct communications be maintained between the control room and personnel at the refueling station during core alterations. As discussed in the TS Bases, the requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions. The licensee's application dated April 9, 2009, stated that the components covered by this LCO include radios and associated power and transmission equipment necessary to establish and maintain communications between the control room and the refueling station.

-4 TS 3/4.9.6: Refueling Operations - Manipulator Crane Operability LCO 3.9.6 provides the operability requirements for the manipulator crane and auxiliary hoist.

As discussed in the licensee's application dated April 9, 2009, the manipulator crane is a bridge and trolley crane with a vertical mast. The bridge spans the reactor cavity and runs on rails set into the floor along the edge of the reactor cavity. The bridge and trolley motions are used during refueling operations with the reactor head removed to position the vertical mast over a fuel assembly in the reactor core. Inside the manipulator crane, a long tube with a pneumatic gripper on the end is lowered down from the mast to grip fuel assemblies. A winch mounted on the trolley raises the gripper tube and fuel assembly up into the mast tube. The fuel, while inside the mast tube, is transported to its new position. The manipulator crane is used to move individual fuel assemblies into, out of, and between positions in the reactor core during refueling operations. The manipulator crane is limited to moving a single fuel assembly at a time by its physical configuration and lift capacity limits. An overload cutoff device in the manipulator crane prevents it from applying excessive lifting force in the event it is inadvertently engaged with an object other than a fuel assembly during lifting operations.

The auxiliary hoist is currently only used to facilitate latching and unlatching the control rod drives for individual rod cluster control assemblies (RCCAs) in the reactor core during refueling operations. The auxiliary hoist is not used to lift and move fuel assemblies.

LCO 3.9.6 states that:

The manipulator crane and auxiliary hoist shall be used for movement of control rods or fuel assemblies and shall be OPERABLE with:

a.

The manipulator crane used for movement of fuel assemblies having:

1.

A minimum capacity of 3250 pounds, and

2.

An overload cut off limit less than or equal to 2850 pounds

b.

The auxiliary hoist used for movement of control rods having:

1.

A minimum capacity of 700 pounds, and

2.

A load indicator which shall be used to prevent lifting loads in excess of 600 pounds As discussed in the T8 Bases, the operability requirements ensure that the manipulator crane and auxiliary hoist will only be used for movement of fuel assemblies and control rods (respectively) and have sufficient load capacity to perform these refueling operations. In addition, the operability requirements ensure that core internals and the reactor pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

-5 TS 3/4.9.7: Refueling Operations - Crane Travel-Fuel Handling Area LCO 3.9.7 requires that loads in excess of 2200 pounds are prohibited from travel over fuel assemblies in the storage pool. The TS is applicable when fuel assemblies are in the storage pool. The licensee's application dated April 9, 2009, stated that "PSEG would also conservatively assume this LCO to be applicable with fuel assemblies in the fuel transfer pool during spent fuel cask loading operations." The 2200 pound weight is the nominal combined weight of a single fuel assembly, an RCCA, and a fuel handling tool. The 2200 pound limit was established in Salem Unit Nos. 1 and 2, Amendment Nos. 77 and 51, dated March 31,1987 (ADAMS Accession No. ML011660319). As discussed in the NRC staff's safety evaluation (SE) for these amendments, the 2200 pound limit was established by the licensee to meet the guidance in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."

As discussed in the licensee's application dated April 9, 2009, the two cranes capable of carrying heavy loads over the storage pool or fuel transfer pool are the fuel handling crane and the cask handling crane. Both cranes are located in the fuel handling building (FHB). Individual fuel assemblies are handled in the fuel storage pool or fuel transfer pool by means of a special tool suspended from the fuel handling crane hook. The fuel handling crane, by design, can only handle a single fuel assembly and integral RCCA at one time. The cask handling crane is a bridge and trolley crane used to: (1) transfer new fuel containers from the truck bay to the laydown area near the new fuel storage area; and (2) move spent fuel casks from the transfer pool to the decontamination pit and to the truck bay. The cask handling crane bridge rails are located such that the cask handling crane can only travel over the fuel transfer pool, decontamination pit, new fuel handling area, and receiving bay. It cannot carry loads over the spent fuel pool where long-term storage of irradiated fuel takes place.

3.0 TECHNICAL EVALUATION

The licensee's submittal stated that the proposed relocation of the three TSs to the TRM is justified because the associated LCOs do not meet any of the criteria in 10 CFR 50.36(c)(2)(ii).

The licensee also stated the proposed relocations are consistent with NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," Revision 3, dated June 2004, because the Standard TSs [STS] do not contain refueling operations TS requirements for communications, manipulator crane operability, or crane travel. The licensee stated, once relocated to the TRM, future changes to the TRM requirements would be controlled under the provisions of 10 CFR 50.59.

The NRC staff agrees that the requirements proposed for relocation to the TRM from the Salem TSs are not within the scope of requirements included in NUREG-1431. The staff evaluated the proposed relocation of the Salem TS requirements against the criteria of 10 CFR 50.36(c)(2)(ii) as discussed below in SE sections 3.1 through 3.4.

3.1 Evaluation of Proposed TS Relocations against Criterion 1 Criterion 1 applies to:

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

-6 The functions of the equipment associated with TSs 3/4.9.5,3/4.9.6, and 3/4.9.7, are described above in SE Section 2.2. None of this equipment is considered instrumentation used to detect degradation of the reactor coolant boundary. Therefore, the NRC staff concludes that none of the subject LCOs meets Criterion 1.

3.2 Evaluation of Proposed TS Relocations against Criterion 2 Criterion 2 applies to:

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Evaluation of Criterion 2 with respect to TS 3/4.9.5 With respect to TS 3/4.9.5, the equipment used for communications between the control room and personnel at the refueling station during core alternations is not considered a process variable, design feature, or operating restriction of a design basis accident or transient condition.

Therefore, the NRC staff concludes that LCO 3.9.5 does not meet Criterion 2.

Evaluation of Criterion 2 with respect to TS 3/4.9.6 With respect to TS 3/4.9.6, the following design basis accidents (DBAs) discussed in the Salem UFSAR could involve the manipulator crane:

1) UFSAR Section 15.3.3 - Inadvertent Loading of a Fuel Assembly into an Improper Position

2) UFSAR Section 15.4.6 - Fuel Handling Accident As discussed in UFSAR Section 15.3.3, fuel assembly loading errors are prevented by administrative procedures implemented during core loading. As discussed above in SE Section 2.2, the operability requirements in LCO 3.9.6 ensure that the manipulator crane and auxiliary hoist will only be used for movement of fuel assemblies and control rods and have sufficient load capacity to perform these refueling operations. In addition, the operability requirements ensure that core internals and the reactor pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

These operability requirements have no bearing with respect to an administrative error related to the inadvertent loading of a fuel assembly into an improper position. Therefore, the NRC staff concludes that LCO 3.9.6 does not meet Criterion 2 for the Inadvertent Loading of a Fuel Assembly into an Improper Position DBA.

As discussed in Salem UFSAR Section 15.4.6, a fuel handling accident (FHA) "is defined as dropping of a spent fuel assembly onto the spent fuel pit floor in the fuel handling building or inside containment resulting in the rupture of the cladding of all the fuel rods in the assembly despite many administrative controls and physical limitations imposed on fuel handling operations." As discussed in the licensee's application dated April 9, 2009, for an FHA postulated to occur in containment, the fuel assembly would be dropped by the manipulator crane. The manipulator crane cannot physically access the FHB, so it plays no role in the FHA postulated to occur there.

-7 The important process variables, design features, and initial conditions for the FHA consequence analysis are the fuel design, the peak fuel burn-up, the decay time, the number of fuel pins damaged, and the water depth available to scrub the postulated release. The Salem FHA analysis assumes all 264 fuel rods in a peak power fuel assembly are breached as a result of the drop. The failure of all fuel rods is a conservative assumption that provides a bounding source term for determining the potential radiological consequences resulting from any level of damage to a single fuel assembly.

The LCO associated with the manipulator crane specifies that the crane be operable with an adequate lifting capacity to handle fuel assemblies and an overload cut-off to protect the reactor vessel internals and reactor vessel from excessive uplift force. Since the FHA assumes the manipulator crane drops a fuel assembly, the minimum capacity of the crane is not an initial condition associated with the design-basis FHA. Similarly, operation of the overload cutoff is not an initial condition of the design-basis FHA because the assumption that all fuel pins are damaged in a single assembly bounds the potential damage to a fuel assembly that could result from excessive uplift forces. Therefore, the NRC staff concludes that LCO 3.9.6 does not meet Criterion 2 for an FHA.

Based on the above, the NRC staff concludes that LCO 3.9.6 does not meet Criterion 2.

Evaluation of Criterion 2 with respect to TS 3/4.9.7 With respect to TS 3/4.9.7, the only applicable DBA is a FHA postulated to occur in the FHB. As discussed in SE Section 2.2, the two cranes associated with LCO 3.9.7 are the fuel handling crane and the cask handling crane.

The initial condition of a design-basis FHA in the FHB, as discussed in Salem UFSAR Section 15.4.6, is the dropping of a single spent fuel assembly onto the spent fuel pit floor resulting in the rupture of the cladding of all the fuel rods in the assembly despite many administrative controls and physical limitations imposed on fuel handling operations. The 2200 pound weight limit in LCO 3.9.7 is the nominal combined weight of a single fuel assembly, an RCCA, and a fuel handling tool. As discussed in the TS Bases, the LCO ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel.

assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. As such, the potential dropping of a single fuel assembly is consistent with the assumptions associated with the maximum radiological consequences for a design-basis FHA.

However, as noted above, the initial conditions for a design-basis FHA do not credit administrative controls or physical limitations imposed on fuel handling operations. Therefore, the load limit in LCO 3.9.7 is not considered an initial condition of a design-basis FHA. Based on the above, the !\\IRC staff concludes that LCO 3.9.7 does not meet Criterion 2.

3.3 Evaluation of Proposed TS Relocations against Criterion 3 Criterion 3 applies to:

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

-8 The functions of the equipment associated with TSs 3/4.9.5,3/4.9.6, and 3/4.9.7, are described above in SE Section 2.2. None of this equipment includes any systems structures, or components (SSCs) that function to mitigate any DBA or transient. Therefore, the NRC staff concludes that none of the subject LCOs meets Criterion 3.

3.4 Evaluation of Proposed TS Relocations against Criterion 4 Criterion 4 applies to:

A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The functions of the equipment associated with TSs 3/4.9.5,3/4.9.6, and 3/4.9.7, are described above in SE Section 2.2. None of this equipment includes any SSCs shown to be significant to public health and safety through operating experience or probabilistic risk assessments.

Therefore, the NRC staff concludes that none of the subject LCOs meets Criterion 4.

3.5 Technical Evaluation Conclusion

Based on the evaluation in SE Sections 3.1 through 3.4, the NRC staff finds that LCOs 3.9.5, 3.9.6, and 3.9.7 do not meet the criteria in 10 CFR 50.36(c)(2)(ii) requiring inclusion in the TSs.

As discussed in SE Section 2.1, the licensee proposes to relocate TSs 3/4.9.5, 3/4.9.6, and 3/4.9.7, to the Salem TRM. Based on the description of the TRM in UFSAR Section 13.5.4, future changes to relocated TS requirements will be subject to the provisions of 10 CFR 50.59.

As such, the NRC staff finds that there is reasonable assurance that future changes to the relocated requirements will be made in a manner that continues to protect public health and safety.

Based on the above findings, the NRC staff concludes that relocation of TSs 3/4.9.5, 3/4.9.6, and 3/4.9.7 to the Salem TRM is acceptable. The licensee has also proposed to revise the associated TS index pages to reflect relocation of TSs 3/4.9.5,3/4.9.6, and 3/4.9.7. These changes are administrative in nature and, therefore, are acceptable.

PSEG's application dated April 9, 2009, provided revised TS Bases pages to be implemented with the associated TS changes. These pages were provided for information only and will be revised in accordance with the Salem TS Bases Control Program.

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 a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. 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

-9 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 (74 FR 42929). 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) 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 Contributors: S. Jones R. Ennis Date: February 17, 2010

February 17, 2010 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear P.O. Box 236, N09 Hancocks Bridge, NJ 08038 SUB.JECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: RELOCATION OF CERTAIN TECHNICAL SPECIFICATION REQUIREMENTS ASSOCIATED WITH REFUELING OPERATIONS (TAC NOS. ME1279 AND ME1280)

Dear Mr. Joyce:

The Commission has issued the enclosed Amendment Nos. 293 and 277 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2.

These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated April 9, 2009, as supplemented by letter dated October 23, 2009.

The amendments relocate TS requirements pertaining to communications during refueling operations (TS 3/4.9.5), manipulator crane operability (TS 3/4.9.6), and crane travel (TS 3/4.9.7) to the Technical Requirements Manual.

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

Sincerely,

/raJ Richard B. Ennis, 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. 293 to License No. DPR-70

2. Amendment No. 277 to License No. DPR-75

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

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