LR-N08-0054, License Amendment Request (LAR) S08-03, Request for Change to Technical Specifications Refueling Operations - Containment Building Penetrations

License Amendment Request (LAR) S08-03, Request for Change to Technical Specifications Refueling Operations - Containment Building Penetrations

ML080660205
Person / Time
Site:	Salem
Issue date:	03/05/2008
From:	Braun R Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
(LAR)-S08-03, LR-N08-0054, TAC MB5488, TAC MB5489
Download: ML080660205 (15)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG NuclearLLC 10 CFR 50.90 MAR 0 5 2008 10 CFR 50.91 LR-N08-0054 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Gentlemen:

SALEM GENERATING STATION - UNIT 1 and UNIT 2 FACILITY OPERATING LICENSE NOS. DPR 70 and DPR-75 NRC DOCKET NOS. 50-272 and 50-311

Subject:

REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS REFUELING OPERATIONS - CONTAINMENT BUILDING PENETRATIONS

-LICENSE AMENDMENT REQUEST (LAR)-S08-03 -

References:

(1) Safety Evaluation Report from NRC to PSEG: "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendment Re:

Refueling Operations - Fuel Decay Time Prior to Commencing Core Alterations or Movement of Irradiated Fuel (TAC Nos. MB5488 and MB5489)", dated October 10, 2002 (2) Letter from PSEG to NRC: "Request for Changes to Technical Specifications, Refueling Operations - Relaxation of Requirements Applicable During Movement of Irradiated Fuel, Salem Nuclear Generating Station, Units 1. and 2, Facility Operating Licenses DPR-70 and'DPR-75, Docket Nos. 50-272 and 50-311", dated July 29, 2002 (3) Safety Evaluation Report from NRC to PSEG: "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendments Re:

Request For Relaxation of Technical Specification Requirements Applicable During Movement of Irradiated Fuel (TAC Nos. MB571 0 and MB5711)", dated September 16, 2004 (4) Letter from NRC to PSEG: "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Correction to Issuance of Amendment Nos. 263 and 245 (TAC Nos. MB571 0 and MB571 1)", dated November 2, 2004 95-2168 REV. 7/99

Document Control Desk 2 MAR 0 5 2008 LR-N08-0054 (5) Letter from PSEG to NRC: "Request for One-Time Change to Technical Specifications, Refueling Operations -Decay Time, LAR S07-06, Salem Nuclear Generating Station, Unit 2, Facility Operating License DPR-75, Docket No. 50-311", dated October 17, 2007 In accordance with the provisions of 10CFR50.90, PSEG Nuclear LLC (PSEG) hereby requests an amendment of the Technical Specifications (TS) for the Salem Nuclear Generating Station, Units 1 and 2. In accordance with I0CFR50.91(b)(1), a copy of this submittal has been sent to the State of New Jersey.

PSEG proposes to revise the surveillance requirement (SR) 4.9.4.2 for the verification of closure of the equipment hatch within one hour when the equipment hatch is to be open during the movement of irradiated fuel within the containment. SR 4.9.4.2 is being revised to clarify that this SIR is applicable to the inside equipment hatch door or an installed equivalent closure device allowed in accordance with Limiting Condition of Operation (LCO) 3.9.4.a. There are no safety consequences associated with this proposed change to SR 4.9.4.2 as discussed in Attachment 1. provides the existing TS pages marked-up to show the proposed changes.

PSEG has evaluated-_the -proposed-changes in accordance with _iOCFR50.91.(a)(_.),

using the criteria in IOCFR50.92(c), and has determined this request involves no significant hazards considerations. This amendment to the Salem TS meets the criteria of 10CFR51.22(c)(9) for categorical exclusion from an environmental impact statement.

PSEG requests approval of the proposed License Amendment by March 14, 2008 to be implemented within 1 day, to support Salem Unit 2 refueling outage 2R16 which commences on March 11, 2008. As a result of NRC inspection activities associated with the 2R16 outage, it was determined that the current wording of SIR 4.9.4.2 can only be complied with through the use of the inner equipment hatch door and not with an equivalent closure device allowed by LCO 3.9.4.a. Therefore, leaving the equipment hatch door open during fuel movement would only be allowed if the inside equipment hatch door can be installed within one hour.

PSEG developed the 2R16 outage schedule for the replacement of the steam generators on the use of an equivalent closure device to meet the requirements of SR 4.9.4.2 through demonstrating that this equivalent closure device can be closed within one hour accommodating any obstructions associated with the equipment being moved or installed in the equipment hatch to support the steam generator replacement. With SR 4.9.4.2 only being applicable to the inside equipment hatch door, the inside equipment hatch door would need to be installed during the movement of fuel inside containment. This is due to the inability to move equipment in support of the steam generator replacement from the hatch and close the equipment hatch within one hour.

The impact to the current 2R16 outage schedule would be approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> of critical path time; therefore, extending the start up from the refueling outage by 2-1/2 2

Document Control Desk 3 MAR 0 52008 LR-N08-0054 days. Based upon the recent identification of this compliance issue and the significant impact to the 2R1 6 outage schedule PSEG is requesting that the proposed change be approved on an exigent basis in accordance with 10CFR50.91 (a)(6)(vi).

Ifyou have any questions or require additional information, please do not hesitate to contact Mr. Brian Thomas at (856) 339-2022.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on 3 _510_8 (Date)

Sincerely, Robert C. Braun Site Vice President Salem Generating Station Attachments: 2 C .- Mr.-S. -Collins,-Administrator -- Region-I - -------

U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis, Project Manager - Salem Unit I and Unit 2 U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem Unit I and Unit 2 Mr. P. Mulligan Bureau of Nuclear Engineering PO Box 415 Trenton, New Jersey 08625

Attachment I LAR S08-03 LR-N08-0054 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS REFUELING OPERATIONS - CONTAINMENT BUILDING PENETRATIONS Table of Contents

1. DE S C R IPT IO N ................................................................................................ 1

2. PROPOSED CHANGE ................................................................................... I

3. BA C KG R O UND .................................................................................................. 1

4. TECHNICAL ANALYSIS .................................................................................. 2

5. REGULATORY SAFETY ANALYSIS ............................................................... 4 6.' ENVIRONMENTAL CONSIDERATION ............................................. ...... I............. 7

7. R E FER E NC ES ................................................ ............................................... 8

Attachment I LAR S08-03 LR-N08-0054 I. DESCRIPTION PSEG proposes to revise the surveillance requirement (SR) 4.9.4.2 for the verification of closure of the equipment hatch within one hour when the equipment hatch is to be open during the movement of irradiated fuel within the containment. SR 4.9.4.2 is being revised to clarify that this SR is applicable to the inside equipment hatch door or an installed equivalent closure device allowed in accordance with Limiting Condition of Operation (LCO) 3.9.4.a.

2. PROPOSED CHANGE Technical Specification (TS) SR 4.9.4.2 would be revised as follows:

4.9.4.2 Once per refueling prior to the start of movement of irradiated fuel assemblies within the containment building, verify the capability to install close, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the equipment hatch inside door or an equivalent closure device. Applicable only when the equipment hatch is open during movement of irradiated fuel in the containment building.

No further clarifications of the TS bases associated with proposed changes to SR

. 4.9.4.2 were-determined-to -be necessary ........ ...-..

3. BACKGROUND On July 29, 2002, PSEG submitted a request for amendment (Reference 7.2) to allow the containment equipment hatch to remain open during the movement of irradiated fuel within containment. The NRC approved the amendment request on September 16, 2004 as Amendment 263 for Salem Unit 1 and Amendment 245 for Salem Unit 2 (Reference 7.3). The change to allow the equipment hatch to remain open during fuel movement was based on a revised analysis of the fuel handling accident (FHA) analysis using the selective implementation of alternate source term (AST) methodology. The selective implementation of the FHA AST analysis was approved by the NRC as Amendments 251 and 232 for Salem Unit Nos. 1 and 2, respectively, on October 10, 2002 (Reference 7.1).

In the July 29, 2002 amendment request, PSEG proposed changes to TS 3.9.4 to incorporate a revision to the LCO 3.9.4.a to allow the inner equipment hatch door or an equivalent closure device to remain open during the movement of fuel within containment provided it was capable of being closed. Even though containment closure is not credited in the dose calculations approved by Amendments 251/232, PSEG committed to implement administrative controls to provide reasonable assurance that containment closure, as a defense-in-depth measure, can be reestablished promptly (within one hour) following a FHA to limit the releases much lower than assumed in the dose calculations. To ensure that these administrative controls were demonstrated prior to movement of irradiated 1

Attachment I LAR S08-03 LR-NOB-0054 fuel within the containment, a new SR 4.9.4.2 was included in this request for amendment to verify the ability to install the equipment hatch within one hour.

However, due to an oversight, the proposed SR wording specifically referenced only the hatch inside door, not the equivalent closure device.

As a result of NRC inspection activities associated with the 2R16 outage, it was determined that the current wording of SR 4.9.4.2 can only be complied with through the use of the inner equipment hatch door and not with an equivalent closure device allowed by LCO 3.9.4.a. Therefore, leaving the equipment hatch door open during fuel movement would only be allowed if the inside equipment hatch door can be installed within one hour.

PSEG developed the 2R.16 outage schedule for the replacement of the steam generators on the use of an equivalent closure device to meet the requirements of SR 4.9.4.2 through demonstrating that this equivalent closure device can be closed within one hour accommodating any obstructions associated with the equipment being moved or installed in the equipment hatch to support the steam generator replacement. With SR 4.9.4.2 only being applicable to the inside equipment hatch door, the inside equipment hatch door would need to be installed during the movement of fuel inside containment. This is due to the inability to move equipment in support of the steam generator replacement from the hatch and close the equipment hatch within one hour. The impact to the current_2R_16_outage-schedule would be-app roximately_60-hours-ofcritical path _.

time; therefore, extending the start up from the refueling outage by 2-1/2 days.

Based upon the recent identification of this compliance issue and the significant impact to the 2R16 outage schedule PSEG is requesting that the proposed change be approved on an exigent basis.

4. TECHNICAL ANALYSIS Each Unit 1 and Unit 2 containment at SalemNuclear Station is equipped with an equipment hatch. Technical Specification (TS) 3.9.4.a requires that during the movement of irradiated fuel assemblies within containment, the associated inner equipment hatch door is capable of being closed and secured with at least four bolts or an equivalent closure device installed and capable of being installed.

The applicable design basis event is the Fuel Handling Accident inside containment. During movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a fuel handling accident, involving dropping of a spent fuel assembly resulting in the rupture of the cladding of all the fuel rods in the assembly. In the re-analysis of this design basis event approved in Amendments 251/232, airborne activity resulting from the initiating event (FHA) is assumed to be released to the environment over a two hour time period via the open equipment hatch and the plant vent.

Following reactor shutdown, decay of the short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. Following sufficient decay time, the primary success path for mitigating the FHA no longer 2

Attachment I LAR S08-03 LR-NO8-0054 includes the functioning of the active containment systems. Therefore, water level and decay time are the primary success paths for mitigating a FHA.

Calculations were performed to determine atmospheric dispersion factors (X/Q's) at the Salem Nuclear Generating Station (SNGS) control room (CR) air intake due to the FHA releases from the Containment Equipment Hatch and Plant Vent (PV).

Analyses were performed to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ) and Control Room (CR) doses due to a FHA occurring in the containment building with containment equipment hatch and Personnel Locks opened. The FHA analysis was performed using the AST, the guidance in the Regulatory Guide 1.183, Appendix B, and TEDE dose criteria. The results of these calculations are within the regulatory acceptance criteria and approved in Amendments 251 and 232.

In the July 29, 2002 amendment request PSEG proposed changes to TS 3.9.4 to incorporate a revision to LCO 3.9.4.a to allow the inner equipment hatch door or an equivalent closure device to remain open during the movement of fuel within containment provided it was capable of being closed. Even though containment closure is not credited in the dose calculations approved by Amendments

-..... 25i1/232 ,PSE G__c-ommitted._toimplement .admi.nistrative controIs to provide . ...

reasonable assurance that containment closure, as a defense-in-depth measure, can be reestablished promptly (within one hour) following a FHA to limit the releases much lower than assumed in the dose calculations. To ensure that these administrative controls were demonstrated prior to movement of irradiated fuel within the containment, a new SR 4.9.4.2 was included in this request for amendment to verify the ability to install the equipment hatch within one hour.

However, the proposed SR wording specifically referenced only the hatch inside door, not the equivalent closure device.

As documented in the safety evaluation report (SER) for Amendments 263/345, the NRC stated on page 14, "that the NRC staff has reviewed the proposed administrative controls and finds that they are adequate to ensure the ability to establish containment closure in a timely manner in the unlikely event of a FHA.

Also the use of equivalent closure devices and the minimum four-bolt closure requirement on the containment equipment hatch are already part of the Salem licensing bases and are not being changed by these amendments." In reviewing proposed SR 4.9.4.2, the NRC stated in the SER that, "this new SR sets the requirement to verify that during the movement of irradiated fuel that the equipment hatch can be installed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in the event of an FHA. This verification provides the requisite assurance that the equipment hatch is able to be closed in the event of an FHA. Although closing the equipment hatch is not necessary to meet the requirements of 10CFR50.67, the NRC staff has determined that these measures are an important element of defense-in-depth 3

Attachment I LAR S08-03 LR-N08-0054 that serves to manage the consequences of an FHA, further reducing the release."

On October 17, 2007, PSEG submitted a one-time license amendment request for Salem Unit 2 to revise the fuel decay time from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br />. With this request, PSEG submitted a revised FHA dose analysis. This revised dose analysis was performed using the regulatory requirements of RG 1.183 consistent with the methods approved by the NRC in Amendments 251/232.

This revised dose analysis also assumes that the radiological activity is released through either the open Containment Equipment Hatch or the plant vent consistent with the previous analysis. Therefore, approval of the October 17, 2007 request will not change the conclusion presented in this proposed amendment request.

5.0 Regulatory Safety Analysis 5.1 Basis for proposed no significant hazards consideration determination As required by 10CFR50.91(a), PSEG provides its analysis of the no significant hazards consideration. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated;

2. Create the possibility of a new or different kind of accident from any previously analyzed; or

3. Involve a significant reduction in a margin of safety.

The determinations that the criteria set forth in 10CFR50.92 are met for this amendment request are indicated below:

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

Response: No.

An alternate source term calculation has been performed for Salem Nuclear Station that demonstrates that offsite and control room dose consequences of a postulated fuel handling accident remain within the limits provided sufficient decay has occurred prior to the movement of irradiated fuel without taking credit for containment closure. Fuel movement is allowed provided that irradiated fuel has undergone the required decay time. This alternate source term calculation for a fuel handling accident inside containment was approved in Amendment Nos. 251 and 232 for Salem Unit Nos. 1 and 2, respectively and, therefore, is already part of the current Salem licensing basis.

4

Attachment I LAR S08-03 LR-N08-0054 The proposed amendment would revise surveillance requirement (SR) 4.9.4.2 to state that this SR is applicable to the inner equipment hatch door or an equivalent closure device allowed by TS 3.9.4.a. SR 4.9.4.2 demonstrates the ability to close the equipment hatch within one hour when the equipment hatch remains open during the movement of irradiated fuel within containment. An equivalent closure device is already specifically allowed by the Salem Unit Nos. 1 and 2 TSs. That allowance was incorporated into the Salem TSs by Amendment Nos. 217 and 199 for Salem Unit Nos. 1 and 2, respectively and, therefore, is already part of the current Salem licensing basis. This amendment request simply clarifies that if an equivalent closure device is installed in lieu of the equipment hatch inside door, the same restrictions and administrative controls apply to ensure closure will take place within one hour following a Fuel Handling Accident inside containment.

This amendment does not alter the methodology of the FHA or equipment used directly in fuel handling operations. The equipment hatch is not an accident initiator.

Actual fuel handling operations are not affected by the proposed changes.

Therefore, the probability of a Fuel Handling Accident is not affected with the proposed amendment. No other accident initiator is affected by the proposed changes.

Therefore, the proposed change does not involve a significant increase in the probability or 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 involve addition or modification to any plant system, structure, or component. The proposed amendment would revise surveillance requirement (SR) 4.9.4.2 to state that this SR is applicable to the inner equipment hatch door or an equivalent closure device allowed by TS 3.9.4.a. SR 4.9.4.2 demonstrates the ability to close the equipment hatch within one hour when the equipment hatch remains open during the movement of irradiated fuel within containment. Having the equipment hatch open during movement of irradiated fuel in containment does not create the possibility of a new accident. Closure of the equipment hatch can be accomplished by either the equipment hatch inside door, or an equivalent closure device already specifically allowed by the Salem Unit Nos. 1 and 2 TSs. An equivalent closure device may be installed as an alternative to installing the Containment Equipment Hatch inside door. Any equivalent closure device will satisfy the requirements of Technical Specification 314.9.4.a and the associated TS Bases. If an equivalent closure device is installed in lieu of the equipment hatch inside door, the same 5

Attachment I LAR S08-03 LR-N08-0054 restrictions and administrative controls apply to ensure closure will take place within one hour following a fuel handling accident inside containment.

The proposed amendment does not create the possibility of a new or different kind of accident than any previously evaluated.

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

Response: No.

An alternate source term calculation has been performed for Salem Nuclear Station that demonstrates that offsite and control room dose consequences of a postulated fuel handling accident remain within the limits provided sufficient decay-has occurred prior to the movement of irradiated fuel without taking credit for containment closure. Fuel movement is allowed provided that irradiated fuel has undergone the required decay time. This alternate source term calculation for a fuel handling accident inside containment was approved in.Amendment Nos. 251 and 232 for Salem Unit Nos. 1 and 2, respectively and, therefore, is already part of the current Salem licensing basis.

The proposed change to SR 4.9.4.2 does not alter the FHA analysis approved by Amendments 251 and 232. This amendment request simply clarifies that if an

-..... equivalent-closu re-device is-installed-in-lieu-oflhe-eq uipment-hatch-inside-door; .

the same restrictions and administrative controls apply to ensure closure will take place within one hour following a fuel handling accident inside containment.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Reguirements/Criteria NRC Requlatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors".

The NRC's traditional methods for calculating the radiological consequences of design basis accidents are described in a series of regulatory guides and Standard Review Plan (SRP) chapters. That guidance was developed to be consistent with the TID-14844 source term and the whole body and thyroid dose guidelines stated in 10 CFR 100.11.

Many of those analysis assumptions and methods are inconsistent with the ASTs and with the total effective dose equivalent (TEDE) criteria provided in 10 CFR 50.67. This guide provides assumptions and methods 6

Attachment I LAR S08-03 LR-N08-0054 that are acceptable to the NRC staff for performing design basis radiological analyses using an AST.

Title 10, Code of Federal Regulations, Part 50 Section 67, "Accident Source Term".

10 CFR 50.67 permits licensees to voluntarily revise the accident source term used in design basis radiological consequences analyses. This document is part of a 10 CFR 50.90 license amendment application and evaluates the consequences of a design basis fuel handling accident as previously described in the Salem UFSAR.

10 CFR 50 Appendix A, General Design Criteria 19, Control Room PSEG has applied the guidelines provided by 10 CFR 50.67 and RG 1.183, which is consistent with the current requirements of GDC 19 for the Fuel Handling Accident.

5.3 Conclusion In conclusion, based on the considerations discussed above,

--- {(heIts-rea-s -6ableassuiaiie thtlýt-e-h-lth-an-d -f~ty- df-thte-pblitcwill..

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

6.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10CFR51.22(b), an evaluation of this license amendment request has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10CFR51.22(c)(9) of the regulations.

PSEG has concluded that implementation of this amendment will have no adverse impact upon the Salem units; neither will it contribute to any significant additional quantity or type of effluent being available for adverse environmental impact or personnel exposure. The change to SR 4.9.4.2 does not introduce any new effluents or significantly increase the quantities of existing effluents. As such, the change cannot significantly affect the types or amounts of any effluents that may be released offsite.

It has been determined there is:

1. No significant hazards consideration, 7

LAR S08-03 LR-N08-0054

2. No significant change in the types, or significant increase in the amounts, of any effluents that may be released offsite, and ,

3. No significant increase in individual or cumulative occupational radiation exposures involved.

Therefore, this amendment to the Salem TS meets the criteria of IOCFR51.22(c)(9) for categorical exclusion from an environmental impact statement.

7.0 REFERENCES

7.1 Safety Evaluation Report from NRC to PSEG: "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendment Re:

Refueling Operations - Fuel Decay Time Prior to Commencing Core Alterations or Movement of Irradiated Fuel (TAC Nos. MB5488 and MB5489)", dated October 10, 2002 7.2 Letter from PSEG to NRC: "Request for Changes to Technical Specifications, Refueling Operations - Relaxation of Requirements Applicable During Movement of Irradiated Fuel, Salem Nuclear Generating Station, Units 1 and 2, Facility Operating Licenses DPR-70 and DPR-75, Docket Nos. 50-272 and 50-311", dated July 29, 2002

. 7.3.SafetyEvaluation- Report..from .NR1CtoPSEG:_"Salem__N-uclear.-_ _

Generating Station, Unit Nos. 1 and 2, Issuance of Amendments Re:

Request For Relaxation Of Technical Specification Requirements Applicable During Movement of Irradiated Fuel (TAC Nos. MB5710 and MB5711)", dated September 16, 2004 7.4 Letter from NRC to PSEG: "Salem Nuclear Generating Station, Unit Nos. 1 and 2, Correction to Issuance of Amendment Nos. 263 and 245 (TAC Nos. MB5710 and MB5711)", dated November 2, 2004 7.5 Letter from PSEG to NRC: "Request for One-Time Change to Technical Specifications, Refueling Operations -Decay Time, LAR S07-06, Salem Nuclear Generating Station, Unit 2, Facility Operating License DPR-75, Docket No. 50-311", dated October 17, 2007 7.6 PSEG Calculation S-C-ZZ-MDC-1920 7.7 PSEG Salem Units 1 and 2, Updated Final Safety Analysis Report 7.8 PSEG Salem Units 1 and 2, Technical Specifications 7.9 10 CFR 50.67, "Accident Source Term" 7.10 Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" 7.11 NUMARC 93-01, Revision 3, " Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", July 2000 8

LAR S08-03 LR-N08-0054 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License DPR-70 and DPR-75 are affected by this change request:

DPR-70, Salem Unit 1 Technical Specification Page 3/4.9.3 3/4 9-4 DPR-75, Salem Unit 2 Technical Specification Page 3/4.9.3 3/4 9-4 1

REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment hatch inside door is capable of being closed and held in place by a minimum of four bolts, or an equivalent closure device installed and capable of being closed,

b. A minimum of one door in each airlock is capable of being closed

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by the Containment Purge and Pressure-Vacuum Relief Isolation System.

Note: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of irradiated fuel in the. containment building. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by a manual or automatic containment isolation valve at least once per 7 days.

4.9.4.2 Once per refueling prior to the start of movement of irradiated fuel assemblies within the containment building, verify the capability to i*sta# close, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the equipment hatch inside door or an equivalent closure device. Applicable only when the equipment hatch is open during movement of irradiated fuel in the containment building.

4.9.4.3 Verify, once per 18 months, each required containment purge isolation valve actuates to the isolation position on a manual actuation signal.

SALEM - UNIT 1 3/4 9-4 Amendment No.26-3

REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment hatch inside door is capable of being closed and held in place by a minimum of four bolts, or an equivalent closure device installed and capable of being closed,

b. A minimum of one door in each airlock is capable of being closed

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by the Containment Purge and Pressure-Vacuum Relief Isolation System.

Note: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY:. During movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of irradiated fuel in the containment building. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by a manual or automatic containment isolation valve at least once per 7 days.

4.9.4.2 Once per refueling prior to the start of movement of irradiated fuel assemblies within the containment building, verify the capability to install close, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the equipment hatch inside door or an equivalent closure device. Applicable only when the equipment hatch is open during movement of irradiated fuel in the containment building.

4.9.4.3 Verify, once per 18 months, each required containment purge isolation valve actuates to the isolation position on a manual actuation signal.

SALEM - UNIT 2 3/4 9-4 Amendment No.246