ML20070L777

Application for Amends to Licenses DPR-42 & DPR-60,revising Tech Specs 5.3.A.1 to Allow Zircaloy-4 Filler Rods,Stainless Steel Filler Rods or Open Water Channels to Be Substituted for Fuel Rods in Assemblies

ML20070L777
Person / Time
Site:	Prairie Island
Issue date:	03/13/1991
From:	Parker T NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20070L778	List: ML20070L777 ML20070L781
References
GL-90-02, GL-90-2, NUDOCS 9103200239
Download: ML20070L777 (7)

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Northern States Power Company 414 Nicollet Mali Minneapohs, Minnesota $$4011927 Telephone (612) 330-5500 March 13, 1991 10 CFR Part 50 Section 50.90 U S Nuclear r'gulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PIANT Docket Nos. $0 282 License Nos. DPR 42 50 306 DPR 60 l

Licence Amendment Request Dated March 13, 1991 l

Fuel Asr.emb1v Design Features Chnnr.es j

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Attached is a request for a change to the Technical Specifications Appendix A of the Operating Licenses, for the Prairie Island Nuclear Generating Plant. This reque.it is submitted in accordance with the provisions of 10 CFR Part 50, Section l

50.90, The Prairie Island Technical Specifications include design requirements for fuel assemblies in Section 5, Design Features.

On a plant-specific basis, the NRC Stafi has approved changes to these requirements that provide flexibility for improved fuct performance by permitting timely removal of fuel rods that are found to be Icaking during a refueling outage or are determined to be probable sources of future Icakage.

Because improvements in fuel performance will result in lower occupational radiation exposure and plant radiological releases, this alternative was made cvailabb to all plants as a line item Technical Specification improvement by Generic Letter 90 02, " Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications. The proposed changes to Technical Specification 5.3.A.1 are being submitted in response to Generic Letter 90 02.

Exhibit A contains a description of the proposed chem 3, the reasons for requesting the changes and the supporting safety evaluatic,n/significant hazards determination. Exhibit B contains current Prairie Island Technical Specification pages marked up to show the proposed changes.

Exhibit C contains the revised Technical Specification pages.

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i Northem States Power Company gg p March 13, 1991 Page 2 Please contact us if you have any questions related to this License Amendment Request.

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1NG Thomas M Parker Manager Nuclear Support Services c:

Regional Administrator III, NRC NRR Project Manager, NRC Senior Resident Inspector, NRC MPCA Attn: J V Perman J E Silberg Attachments:

Affidavit Exhibit A Evaluation of Proposed Changes to the Technical Specifications Exhibit B Proposed Changes Marked Up on Existing Technical Specification Pages Exhibit C Revised Technical Specifi;ation Pages l

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0 U'11TED STATES NUCLEAR REGULATORY COMMISSION l

NORTHERN STATES P NER COMPANY t

PRAIRIE ISIN1D NtICLEAk dNERATING PIANT DOCKET No. 50 282 t

50 306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR 42 6 DPR.60 I

LICENSE AMENDMENT REQUEST DATED March 13, 1991 i

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 proponed changes, reasons for the changes, and a significant hazards evaluation. Exhibits B and C are copies of the Prairie Island Technical Specifications incorporating i

the proposed changes.

This letter contains no restricted or other defense information.

NORTHERN STATES P WEE-MPANY

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sy Thomas M P'arker Manager Nuclear Support Services on this /d day of 7EN4eb

/64/ before me a notary public in and for said County, personally appeared Thomiis'M Parker, Manager Nuclear Support Services, 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 l

statements made in it are true and that it is not interposed for delay, A'u 0 s.wrNd y

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Exhibit A Prairie Island Nuclear Generating Plant License Amendment Request Dated March 13, 1991 Evaluation of Proposed Changes to the Technical Specifications Appendix A of Operating License DPR 42 and DPR 60 Pursuant to 10 CPR 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:

EI.qposed Chrmgta Revise specification 5.3.A.1, as shown on paSe TS.S.3 1 in Exhibit B, to allow Zircaloy 4 filler rods, stainless steel filler rods or open water channels to be substituted for fuel rods in fuel assemblies.

Reason For Changes The Prairie Island Technical Specifications include design requirements for fuel assemblies in Section 5, Design Features.

On a plant. specific basis, the NRC Staff has approved changes to these requirements that provide flexibility for improved fuci performance by permitting timely temoval of fuel rods that are found to be Icaking during a refueling outage or are determined to be probable sources of future leakage.

Because improvements in fuel performance will result in lower occupational radiation exposure and plant radiological releases, this alternative was made available to all plants as a line item Technien1 Specification improvement by Generic Letter 90 OL.

l Generic Letter 90 02 provided specific guidance fo'

..e modification of the fuel assembly design fo etures specification.

Technical Specification 5.3.A.1 has been modified in acccrdance with this guidance.

The proposed changes will allow Zircaloy 4 filler rods, stainless steel filler l

rods or open water channels to be substituted for fuel rods in fuel assemblics as long as that replacement is justified by cycle specific reload analyses performed using NRC approved methodology.

Prior to each fuel cycle an analysis is performed to ensure that, with each reload of fuel, all core design safety criteria are met.

This analysis is performed using the NRC approved methodology listed in Technical Specification 6,7.A 6 b.

In the case where fuel assemblies are repaired or fuel rods replaced,.

appropriate safety analysis will be conducted in conjunction with the normal

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reload analysis, verifying that all applicable core safety limits for' fuel-l rods in the vicinity of the missing or substituted rods are still met.

By l

modeling based on the exact substitution, an accurate and complete safety analysis can be performed, and conformance_with established safety margins will be ensured.

These analyses and the core reload changes are reviewed by Northern States Power as required by 10 CFR Part 50, Section 50.59.

If a change to Technical Specifications or an unreviewed safety question is

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tahlbit A Page 2 M 3 identified, then appropriate changes and analyses will be provided to the NRC i

for review and approval.

The qualifier *using an NRC approved methodology" is included in the proposed specification to ensure that the effects of fuel rod removal / replacement will be analyzed or evaluated on a cycle specific basis usi'ng the same NRC approved methodology and design limits that apply to any reload core.

The requiremenc to report fuel rod replacement for more than 30 rods in the core or 10 rods in any fuel assembly per refueling is included to ensure the NRC is advised of abnormal fuel performance.

The reporting threshobi yriteria is consistent with the guidance provided in Generic Letter 90 02.

The proposed changes would require fuel rod replacement in excess of these numbers to be reported within 30 days af ter cycle startup.

Enfety Fvnluntion and Determination of Sinnificant Hazards Considerations The proposed changes to the Operating 1.icense have been evaluated to determine i

whether they constitute a significant hazards consideration as required by 10 CPR Part 50, Section 50.91 using the standards provided in Section 50.02.

This analysis is provided below:

1 The proposed amendcont will not involve a significant increase in 5.

the probability or conseguences of an accident previously evaluated.

The proposed modifications to the Design Features section of the prairie j

1sland Technical Specifications require that operation with fuel assemblies that have been repaired or have had fuel rods replaced, be justified by cycle specific reload analyses using an NRC approved j

methodology.

This justification will be provided by the performance of an appropriate safety analysis in conjunction with the normal reload j

analysis, verifying that all applicable core safety limits for fuel rods in the vicinity of the missing or substituted reds are still met.

By j

modeling based on the exact substitution, an accurate-and complete safety analysis will be performed, and conformance with established safety margins will be r sured. These analyses and the core reload changes will 2

be revi s ed as required by 10 CFR part 50, Section 50.59.

The evaluations and analyses described above will provide adequate asaurance that the repair of a fuel assembly by the replacement or removal of one or more fuel rods w'.11 not significantly affect the probability or j

consequences of an accident previously 6 valuated.

2.

The proposed amendment will not create the possibility of a new or di f ferent kind of accident from any accident oreviousiv analvred.

There are no new failure modes or mechanisms associated with the proposed changes.

The proposed char.ges do not involve any modification in t

operational limits 'While fuel assemblies containing a substitute rod or vacancies represent a change in the physical core configuration, it is a physical change which is no more significant than, for example, using fuel i

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of a different enrichment from a pre rious cycle. All such changes will be accounted for by the reload analysis as described above.

Given successful completion of such an analysis, it is not possible to create a new or different kind of accident and the accident eaalyses presented in the Updated Safety Analysis Report will remain bounding.

3.

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

As stated above, the proposed changes require that operation with fuel assemblies that have been repaired or have had fuel rods replaced, be justified by cycle specific reload analyses using an NRC approved methodology.

If the physical parameters of the reload core are evaluated as being within previously defined acceptance criteria, then a reduction in the mars n of safety is precluded.

Therefore, the proposed changes i

will not result in any reduction in the plant's margin of safety.

The Commission has provided guidance concerning the application of the standards in 10 CFR $0.92 for der rmining whether a significant hazards consideration exists by providin certain examples of amendments that will o

likely be found to involve no significant hazards considerations. These examples were published in the Federal Register on March 6, 1986.

The changes to the Prairie Island Technical Specifications proposed above are equivalent to NRC example (vi), because they involve changes which eith r may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan.

Based on this guidance and the reasons discussed above, we have concluded that the proposed changes do not involve a significant hazards consideration.

Environmental Assessment This license amendment request does not change effluent types or total ct' fluent amounts nor does it involve an increase in power level. Therefore, this change will not result in any significant environmental impact.

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I Exhibit B l

Prairie Island Nuclear Generating Plant License Amendment Roquest Dated March 13 1991 Proposed Changes Marked Up On Existing Technical Specification Pagen l.

Exhibit B consists of an existing Technical trecification page with the proposed changes written on that page. The existing page affected by this License Amendment Request is listed below:

TS.5.3.1 L.

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