Text

Supplement (Reduced Scope) - Request for Changes to Technical Specifications, Refueling Operations - Decay Time License Amendment Request (LAR) S08-01
	ML081790153
Person / Time
Site:	Salem
Issue date:	06/17/2008
From:	Braun R; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LR-N08-0135
	Download: ML081790153 (16)
	v • d • e

Robert Braun Site Vice President-Salem PSEG Nuclear LLC Salem Nuclear Generating Station P.O. Box 236, M/C S-05, Hancocks Bridge, NJ 08038 tel: 856.339.1998 fax: 856.339.2956 email: Robert.Braun@pseg.com 0 PSEG Nuclear L.L. C.

10 CFR 50.90 JUN

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1 2008 LR-N08-0135 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 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:

References:

SUPPLEMENT (REDUCED SCOPE) - REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS, REFUELING OPERATIONS -

DECAY TIME LICENSE AMENDMENT REQUEST (LAR) S08-01 (1) Letter from PSEG to NRC: "Request for Changes to Technical,,

Specifications, Refueling Operations -Decay Time, LAR S08-01, Saýlem Nuclear Generating Station, Unit 1 and Unit 2, Facility Operating;License DPR-70 and DPR-75, Docket Nos. 50-272 and 50-311", dated March 11, 2008 In Reference 1, PSEG Nuclear LLC (PSEG) submitted License Amendment Request, (LAR) S08-01, proposing revisions to the requirements for fuel decay time prior to commencing movement of irradiated fuel. TS 3/4.9.3 "Decay Time" would be (1) revised to allow fuel movement to commence at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months after the reactor is subcritical between October 1 5 th and May 1 5 th, and (2) relocated to the Salem UFSAR, or Technical Requirements Manual (TRM). Currently, TS 3/4.9.3 requires a fuel decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to fuel movement between October 1 5 th and May 1 5 th.

Subsequent to the submittal of Reference 1, PSEG and the NRC staff have discussed the LAR to provide additional clarification. Based on these discussions PSEG proposes to limit the scope of the LAR to: revise the TS to allow fuel movement to commence at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months after the reactor is subcritical between October 15 th and May 15 th (Item 1 above). The previous request to relocate the TS to the UFSAR or TRM (Item 2 above) is no longer being proposed.

,4o

Document Control Desk Page 2 LR-N08-0135 Specifically, in Attachment 1 of Reference 1, the request to relocate in Section 2 is no longer being proposed, along with the supporting discussion and analysis in Sections 3.2 and 4.2. Accordingly, Section 5.0, the 10 CFR 50.91 (a) no significant hazards considerations evaluation, has been revised and is attached to this submittal (Attachment 1). The revised limited scope of this amendment request continues to meet the criteria of 10 CER 51.22(c)(9) for categorical exclusion from an environmental impact statement. provides the existing TS pages marked-up to show the proposed changes based on the revised limited scope. Attachment 3 provides the existing TS Bases pages marked-up to reflect the associated changes to the TS (for information only).

In accordance with 10 CFR 50.91 (b)(11), a copy of this submittal has been sent to the State of New Jersey.

PSEG requests approval of the proposed License Amendment by September 30, 2008 to be implemented within 30 days, to support Salem Unit 1 refueling outage 1 Rig9.

If you have any questions or require additional information, please do not hesitate to contact Mr. Jeff Keenan at (856) 339-5429.

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

Executed on (Date)

Sincerely, Robert C. Braun Site Vice President Salem Generating Station Attachments: 3

PDocument Control Desk Page 3 LR-N08-0135 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 1 and Unit 2 U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem Unit 1 and Unit 2 Mr. P. Mulligan Bureau of Nuclear Engineering PO Box 415 Trenton, New Jersey 08625 Supplement, LAR S08-01 LR-N08-0135 SUPPLEMENT (REDUCED SCOPE) - REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS, REFUELING OPERATIONS - FUEL DECAY TIME LICENSE AMENDMENT REQUEST (LAR) S08-01 PSEG Nuclear LLC (PSEG) submitted License Amendment Request (LAR) S08-011, proposing revisions to the requirements for fuel decay time prior to commencing movement of irradiated fuel. TS 3/4.9.3 "Decay Time" would be (1) revised to allow fuel movement to commence at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months after the reactor is subcritical between October 15th and May 15 th, and (2) relocated to the Salem UFSAR, or Technical Requirements Manual (TRM). Currently, TS 3/4.9.3 requires a fuel decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to fuel movement between October 1 5th and May 1 5th Subsequent to the submittal of LAR S08-01, PSEG and the NRC staff have discussed the LAR to provide additional clarification. Based on these discussions PSEG proposes to limit the scope of the LAR to: revise the TS to allow fuel movement to commence at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months after the reactor is subcritical between October 15 th and May 1 5 th (Item 1 above). The previous request to relocate the TS to the UFSAR or TRM (Item 2 above) is no longer being proposed.

Specifically, in Attachment 1 of LAR S08-01, the request to relocate in Section 2 is no longer being proposed, along with the supporting discussion and analysis in Sections 3.2 and 4.2. Accordingly, Section 5.0, the 10 CFR 50.91(a) no significant hazards considerations evaluation, has been revised (below). The revised limited scope of this amendment request continues to meet the criteria of 10 CFR 51.22(c)(9) for categorical exclusion from an environmental impact statement.

5.0 Regulatory Safety Analysis 5.1 Basis for proposed no siginificant hazards consideration determination As required by 10 CFR 50.91(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 10 CFR 50.92 are met for this amendment request are indicated below:

Letter from PSEG to NRC: "Request for Changes to Technical Specifications, Refueling Operations -

Decay Time, LAR S08-01, Salem Nuclear Generating Station, Unit 1 and Unit 2, Facility Operating License DPR-70 and DPR-75, Docket Nos. 50-272 and 50-311", dated March 11, 2008 1 of 6 Supplement, LAR S08-01 LR-N08-0135

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

Response

No.

The proposed license amendment would allow fuel assemblies to be removed from the reactor core and be stored in the Spent Fuel Pool (SFP) in less time after subcriticality than currently allowed by the Technical Specifications.

Decreasing the decay time of the fuel affects the radionuclide make-up of the fuel to be offloaded as well as the amount of decay heat that is present from the fuel at the time of offload. The accident previously evaluated that is associated with the proposed license amendment is the fuel handling accident. Allowing the fuel to be offloaded in less time after subcriticality using actual heat loads does not impact the manner in which the fuel is offloaded. The accident initiator is the dropping of the fuel assembly. Since earlier offload does not affect fuel handling, there is no increase in the probability of occurrence of a Fuel Handling Accident (FHA). The time frame in which the fuel assemblies are moved has been evaluated against the 1 OCFR50.67 dose limits for members of the public, licensee personnel and control room. Additionally, the guidance provided in Reg.

Guide 1.183 was used for the selective application of Alternative Source Term.

All dose limits are met with the reduced core offload times; and significant margin is maintained, as the minimum decay time prior to movement of fuel for the FHA analysis is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Therefore, the proposed license amendment does not significantly increase the probability of occurrence or the consequences of accidents previously evaluated.

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

Response

No.

The proposed license amendment would allow core offload to occur in less time after subcriticality which affects the radionuclide makeup of the fuel to be offloaded as well as the amount of decay heat that is present from the fuel at the time of offload. The radionuclide makeup of the fuel assemblies and the amount of decay heat produced by the fuel assemblies do not currently initiate any accident. A change in the radionuclide makeup of the fuel at the time of core offload or an increase in the decay heat produced by the fuel being offloaded will not cause the initiation of any accident. The accident previously evaluated that is associated with fuel movement is the fuel handling accident; no new accidents are introduced. There is no change to the manner in which fuel is being handled or in the equipment used to offload or store the fuel. The effects of the additional decay heat load have been analyzed. The analysis demonstrates that the existing Spent Fuel Pool cooling system and associated systems under worst-2 of 6

Attachrnent 1 Supplement, LAR S08-01 LR-N08-0135 case circumstances would maintain licensing limits and the integrity of the Spent Fuel Pool.

Therefore, the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response

No.

The margin of safety pertinent to the proposed changes is the dose consequences resulting from a fuel handling accident. The shorter decay time prior to fuel movement has been evaluated against 10 CFR 50.67 and all limits continue to be met. All dose limits are met with the reduced core offload times; and significant margin is maintained, as the minimum decay time prior to movement of fuel for the FHA analysis is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Decay heat-up calculations performed prior to the refueling outage, as part of the Integrated Decay Heat Management (IDHM) Program, ensure that planned spent fuel transfer to the SFP will not result in maximum SFP temperature exceeding the design basis limit of 1490 F (with both heat exchangers available) or 1800 F (with one heat exchanger alternating between the two pools). As stated above, the changes in radionuclide makeup and additional heat load do not impact any safety settings and do not cause any safety limit to not be met. In addition, the integrity of the Spent Fuel Pool is maintained.

The time frame in which the fuel assemblies are moved has been evaluated against the 10 CFR 50.67 dose limits for members of the public, licensee personnel and control room. Additionally, the guidance provided in Reg. Guide 1.183 was used. Calculations performed conclude that expected dose limits following a Fuel handling Accident are met with the proposed decay time prior to commencing fuel movement.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on this review, it is concluded that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, PSEG proposes that a finding of "no significant hazards consideration" is justified.

3 of 6

Attachment I Supplement, LAR S08-01 LR-NO8-0135 5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50 Appendix A, General Design Criteria 5--Sharing of structures, systems, and components.

GDC 5 requires that structures, systems, and components important to safety shall not be shared among nuclear power units, unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

In the unlikely event that a SFP heat exchanger (HX) becomes unavailable after core off-load begins, the sharing of one SFP heat exchanger between units has been adequately evaluated. In the worst case scenario, Operations has adequate time to cross-connect the available heat exchanger to the refuel pool before 180°F is reached; the time to make the cross-connect has been appropriately addressed with procedural controls. When the HX is aligned to the non-refuel pool, temperature is rapidly brought down, so the time needed to swap back again to the non-refuel pool is much greater. In addition, Operations has the expected pool heat-up rates prior to fuel off-load; they monitor actual heat-up rates, so they can anticipate required actions.

10 CFR 50 Appendix A, General Design Criteria 61--Fuel storage and handling and radioactivity control.

The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

The changes proposed by the amendment request do not reduce the existing UFSAR requirements for meeting GDC 61. The heat removal capability of the SFP cooling system is maintained for normal, abnormal and accident conditions.

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.

4 of 6 Supplement, LAR S08-01 LR-N08-0135 NRC Regulatory 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 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 that are acceptable to the NRC staff for performing design basis radiological analyses using an AST. This guidance supersedes corresponding radiological analysis assumptions provided in other regulatory documents when used in conjunction with an approved AST and the TEDE criteria provided in 10 CFR 50.67.

This application and the supporting analyses comply with this guidance as it applies to a Fuel Handling Accident.

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.

USNRC Branch Technical Position ASB 9-2, Residual Decay Heat for Light-Water Reactors for Long-Term Cooling, Revision 2 of July 1981.

BTP ASB 9-2 uses a conservative approach for calculating fuel element decay heat, and is applied to this amendment without scaling factors or other adjustments.

Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors".

RG 1.183 supersedes corresponding radiological assumptions provided in other regulatory guides and standard review plan chapters when used in conjunction with an approved alternative source term and the TEDE provided in 10 CFR 50.67.

5 of 6 Supplement, LAR S08-01 LR-N08-0135 10 CFR 100, "Determination of Exclusion Area, Low Population Zone and Population Center Distance".

10 CFR 100.11 provides criteria for evaluating the radiological aspects of reactor sites. A footnote to 10 CFR 100.11 states that the fission product release assumed in these evaluations should be based on a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission products. A similar footnote appears in 10 CFR 50.67. In accordance with the provisions of 10 CFR 50.67(a), PSEG applied the dose reference values in 10 CFR 50.67 (b) (2) in the analyses in lieu of 10 CFR 100 for the Fuel Handling Accident.

NUREG-0800, Standard Review Plan, Section 15.7.4, "Radiological Consequences of Fuel Handling Accidents".

The SRP Section 15.7.4 describes the radiological effects of a postulated Fuel Handling Accident. The SRP does not directly refer to the guidance of RG 1.183 or 10 CFR 50.67. Instead, it refers to regulatory documents, which are superseded by the selective application of the Alternative Source Term for the Fuel Handling Accident.

5.3 Conclusion The FHA dose analyses were performed in accordance with AST and TEDE guidelines provided in Regulatory Guide 1.183 and 10 CFR 50.67. The SFP Cooling Capacity calculations were performed applying acceptable NRC guidance and conservatism aspects resulting in assurance that the design basis limits for SFP heat removal are maintained. Use of the IDHM Program will ensure that SFP temperature limits are not exceeded.

The doses are less than the TEDE criteria set forth in RG 1.183 and are a small fraction of the dose criteria in 10 CFR 50.67.

In conclusion, based on the considerations discussed above, (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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

6 of 6 Supplement, LAR SOB-01 LR-N08-0135 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 9-3 DPR-75, Salem Unit 2 Technical Specification Page 3/4.9.3 3/4 9-3

REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least:.

a.

1-0G--80 hours p:lfi.clabl:.i thrP.G4'C0:

b.

168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months APPLICABILITY:

Specification 3.9.3.a -

From October 1 5 th through May 1 5 th, during movement of irradiated fuel in the reactor pressure vessel.

Specification 3.9.3.b From May 1 6th through October 1 4 th, during movement of irradiated fuel in the reactor pressure vessel.

ACTION:

With the reactor subcritical for less than the required time, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical as required by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

SALEM -

UNIT 1 3/4 9-3 Amendment No.

REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least:

a.

8-0 hours f;*

~Ql cbccgh *su 2Q1

b.

168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months APPLICABILITY:

Specification 3.9.3.a From October 1 5 th through May 1 5th, during movement of irradiated fuel in the reactor pressure vessel.

Specification 3.9.3.b From May 1 6 th through October 1 4 th, during movement of irradiated fuel in the reactor pressure vessel.

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ACTION:

With the reactor subcritical for less than the required time, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical as required by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

SALEM -

UNIT 2 3/4 9-3 Amendment No.

Supplement, LAR S08-01 LR-NO8-0135 PROPOSED CHANGES TO TS BASES PAGES (INFORMATION ONLY)

The following Technical Specifications Bases for Salem Unit 1 and Unit 2, Facility Operating License No. DPR-70 and DPR-75 are affected by this change request:

Salem Unit I Technical Specification Page B 3/4.9.3 B 3/4.9.1b Salem Unit 2 Technical Specification Page B 3/4.9.3 B 3/4.9.1b and lc

3/4.9 REFUELING OPERATIONS BASES In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.

In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible.

In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions. Once actions have been initiated, they must be continued until the boron concentration is restored.

The restoration time depends on the amount of boron that must be injected to reach the required concentration.

The Surveillance Requirement (SR) ensures that the coolant boron concentration in the RCS, and connected portions of the refueling canal, the fuel storage pool and the refueling cavity, is within the COLR limits.

The boron concentration of the coolant in each required volume is determined periodically by chemical analysis.

Prior to reconnecting portions of the refueling canal, the fuel storage pool or the refueling cavity to the RCS, this SR must be met per SR 4.0.4. If any dilution activity has occurred while the cavity or canal was disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.

A minimum frequency of once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable amount of time to verify the boron concentration of representative samples.

The frequency is based on operating experience, which has shown 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to be adequate.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products.

The 444-80-hour decay time L

S,08--01, is consistent with the assumptions used in the fuel handling accident analyses and the resulting dose calculations using the Alternative Source Term described in Reg. Guide 1.183.

SALEM -

UNIT I B 3/4 9-1b Amendment No.2-4k