ML20046B566

Application for Amends to Licenses DPR-42 & DPR-60, Incorporating Ref to Revised Methodologies Described in WCAP-13677 & NSPNAD-93003

ML20046B566
Person / Time
Site:	Prairie Island
Issue date:	07/29/1993
From:	Richard Anderson NORTHERN STATES POWER CO.
To:
Shared Package
ML20046B562	List: ML20046B561 ML20046B566 ML20046B580
References
NUDOCS 9308050168
Download: ML20046B566 (6)

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UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED July 29, 1993 Northern States Power Company, a Minnesota corporation, requests authorization for changes to Appendix A of the Prairie Island Operating License as shown on the attachments labeled Exhibits A, B, and C.

Exhibit A describes the proposed changes, reasons for the changes, and a significant hazards eval-uation.

Exhibits B and C are copies of the Prairie Island Technical Specifications incorporating the proposed changes.

This letter contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY

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By

' Rs/ger 0 Anderson Director Licensing,and Management Issues Onthish.

ay of L /ff3 before me a notary public in and for said County, persona 19y app 4/ared Roger 0 Anderson, Director of Licensing and Management Issues, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.

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'I Exhibit A i

Prairie Island Nuclear Generating Plant License Amendment Request Dated July 29, 1993 Evaluation of Proposed Changes to the Technical Specifications Appendix A of l

Operating License DPR-42 and DPR-60 i

Pursuant to 10 CFR Part 50, Sections 50 59 and 50.90, the holders of Operating-Licenses DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Specifications:

Background

i Generic Letter 88-16, dated October 4, 1988, was issued to encourage licensees to prepare changes to Technical Specifications related to cycle-specific parameters. The generic letter provided guidance for the relocation of certain cycle-dependent core operating limits from the Technical i

Specifications.

This would allow changes to the values of these cycle-dependent core operating limits without prior NRC approval (i.e., license amendment), so long as an NRC-approved methodology for the calculations is i

followed. A License Amendment Request, which requested the relocation of.the cycle-specific core operating limits from the Prairie Island Technical Specifications to a Core Operating Limits Report, was submitted to the NRC by letter dated November 17, 1989.

The resulting license amendment, issued by the NRC on March 13, 1990, revised the administrative controls section of the Technical Specifications to incorporate references to the NRC-approved calculation methodologies to be used to determine the operating limits contained in the Core Operating Limits Report.

Those changes to the administrative controls section ensure that.the calculation of the core operating limits in the Core Operating Limits Report will be performed in accordance with NRC-approved methodologies, Technical Specification Section 6.7.A.6.b states that the analytical methoos used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

Section 6.7.A.6.b then specifically references the documents that describe the approved methodologies.

Pronosed Changes and Reasons for Change The proposed changes to the Prairie Island Technical Specifications being requested by this license amendment request are described below, and the specific wording changes to Technical Specifications are shown in Exhibits B and C.

Pronosed changes to Technical Specification Section 6.7.A.6.b Westinghouse topical report WCAP-13677-A, "10 CFR 50.46 Evaluation Model Report: U-COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLo, Cladding Options", April 1993 and Northern States Power topical report n

NSPNAD-93003-A, " Transient Power Distribution Methodology", (latest approved version) are being added to Technical Specification Section 6.7.A.6.b.

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Exhibit a Page 2 of 3 Westinghouse Topical Report WCAP-13677, is currently under review by the NRC Staff.

Following NRC review and approval, Westinghouse will issue a revised topical report with the NRC Safety Evaluation Report, WCAP-136777-A.

We.

expect this to occur prior to December of this year. Northern States Power topical report NSPNAD-93003, dated April 1993, was reviewed and approved by the NRC Staff as documented in an NRC Safety Evaluation Report dated July 16, 1993. Northern States Power will issue a revised topical report.with the NRC Safety Evaluation Report, NSPNAD-93003-A.

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i This license amendment has been submitted to incorporate a reference to the j

revised methodologies described in WCAP-13677 and NSPNAD-93003 into the Prairie Island Technical Specifications.following their approval by the NRC Staff.

The revised methodologies must be incorporated into Section 6.7.A.6.b so the model revisions can be used in the determination of the core operating limits.

Safety Evaluation t

The proposed administrative change to Technical Specification Section 6.7.A.6.b incorporates references to revised core analysis methodology reviewed and approved by the NRC Staff.

Because the proposed change is a&ninistrative in nature and because the revised methodology referenced in the change will have prior NRC review and apprcval, Nort hern States Power believes-there is reasonable assurance that the health and st,fety of the public will not be adversely affected by the proposed Technical Specification changes.

Determination of Sicnificant Hazards Considerations The proposed changes to the Operating Licen.te have been evaluated to determine whether they constitute a significant hazards consideration as required by 10

-l CFR Part 50, Section 50.91 using the standards provided in Section 50.92.

This analysis is provided below:

1.

The proposed amendment will not involve a significant increase in the probability or consecuences of an accident nreviously evaluated.

The proposed administrative change to Technical Specification Section 6.7.A.6.b incorporates references to revised core analysis methodology reviewed and approved by the NRC Staff.

Because the proposed change is administrative in nature and because the revised methodology referenced in the change will have prior NRC review and approval, the proposed change will not involve a significant increase in the probability'or consequences of an accident previously evaluated.

2.

The proposed amendment will not create the possibility of a new or different kind of accident from any accident nreviously analyzed.

i As stated above, the proposed change does not contribute in any way to the probability or consequences of an accident. No safety-related equipment, safety function, or plant operations will be altered as a result of the proposed changes.

The cycle-specific core operating limits will be calculated using the revised NRC-approved methods and submitted to the NRC.

The Technical Specifications will continue to require operation

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1 within the required core operating limits and appropriate actions will be taken when or if limits are exceeded.

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

3.

The proposed amendment will not involve a significant reduction in the marrin of safety.

The margin of safety is not affected by the addition of references to NRC i

approved core analysis methodology to the Technical Specifications. The.

margin of safety provided by the current Technical Specifications remains' t

unchanged.

The Technical Specifications continue to require operation i

within the core limits obtained from NRC-approved reload design methodologies. The actions to be taken when or if limits are violated l

remain uncl.anged.

Therefore, the proposed changes are administrative in nature and do not impact the operation of the plant in a manner that involves a reduction in the margin of safety.

Based on the evaluation described above, and pursuant to 10 CFR Part 50, j

Section 50.91, Northern States Power Company has determined that operation of the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request does not involve any significant hazards considerations as defined by NRC regulations in 10 CFR Part 50, Section 50.92.

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Environmental Assessment Northern States Powr.r has evaluated the' proposed changes and determined that:

1.

The changes do rot involve a significant hazards consideration, 2.

The changes do no: involve a significant change in the types or l

significant increase in the amounts of any effluents that may be released offsite, or 3.

The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51 Section 51.22(c)(9).

Therefore, pursuant to 10 CFR Part 51 Section 51.22(b), an environmental assessment of the proposed changes is not required.

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i Exhibit B Prairie Island Nuclear Generating Plant j

License Amendment Request Dated July 29, 1993 J

i Proposed Changes Marked Up i

On Existing Technical Specification Page i

Exhibit B consists of the existing Technical Specification page with the f

proposed changes highlighted on that page.

The existing page affected by this License Amendment Request is listed below:

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Pare TS-6.7-5 l

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TS.6.7-5 i

REV 92 2/11/91 XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase j

II", May, 1981 WCAPil3677 -W"10iCTR? S07461 EYalusEic h? M73el?Rup3r tS EfCOBRX/~ TRAC 2f1hopiOppey]bjis?@ppil)L1993PlenumjInj e c tidnJod61[D@dh7 Cladding (Opti NSPNADf93003fAU"Tiansleht?P6pe6D17,tir11sutiisifiMe:thado145 %

(latestf approved l version)

The core operating limits shall be determined such that all c.

applicable limits (e.g.,

fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be supplied upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.

B. REPORTABLE EVENTS P

The following actions shall be taken for REPORTABLE EVENTS:

a.

The Commission shall be notified by a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the Operations -

Committee and the results of this review shall be submitted to the Safety Audit Committee and the Vice President Nuclear Generation.

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