ML20150F259

Application for Amends to Licenses DPR-42 & DPR-60,changing Peaking Factors Per Upper Plenum Injection Best Estimate LOCA Program

ML20150F259
Person / Time
Site:	Prairie Island
Issue date:	07/05/1988
From:	Musolf D NORTHERN STATES POWER CO.
To:
Shared Package
ML20150F258	List: ML20150F254 ML20150F259 ML20150F265 ML20150F277 ML20150F281
References
NUDOCS 8807180204
Download: ML20150F259 (6)

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- UNITED STATES NUCLEAR REGUIATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO. 50 282 306

' REQUEST FOR AMENDMENT TO OPERATING LICENSE DPR-42 & DPR 60 LICENSE AMENDMENT REQUEST DATED July 5,- 1988 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, C, D and E. Exhibit A describes the proposed changes, describes the reasons for the changes, and contains a significant hazards' eval-uation. Exhibits B and C are copies of the Prairie Island Technical SI ecifications incorporacing the proposed changes. Exhibit D and E are reports supporting the requested changes.

-This letter contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY By -

9 David Musolf -'

Manager Nuclear Support Services On thi day.of_ 14 / N before me a notary public in and for said Coudly, ;$rsonally appeared David Musolf, Manag-er Nuclear Support Services, and being first duly sworn'acknowl-edged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents there-of, and that to the best of his knowledge, information, and be-lief the statements made in it are true and that is is not inter-posed for delay.

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MARQA K LaCORE NOTARY FUBLIC-MINNESOTA HEhN!F1J COUNTv My Comm:w haies Sept 14 !W3 w-:::::::::.wswwwww:::::,::::a 8807180204 080705 PDR ADOCK 05000282 P PDC

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(f Exhibit A Prairie Island Nuclear Generating Plant License Amendment Request Dated July 5,1988 Evaluation of Proposed Changes to the Technical Specifications, Appendix A of Operating Licenses DPR 42 and DPR-60 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:

PEAKING FACTOR CHANGES Proposed Changes a) Delete Figure 3.10 8 "Acceptable Values of F (F33) q and Fag (Fq )" and remove from Table of Contents page TS.x.

b) Change Peaking Factor limits as shown on pages TS.3.10-1 and 2.

c) Modify the Bases discussion on the relationship between ECCS analysis and peaking tactors on page TS.3.10 9.

Reason for Changes These changes are being submitted to increase the power distribution peaking factors specified in the Technical Specifications.  ;

The application of the Westinghouse Large-Break LOCA Best-Estimate Methodology justifies the raising of peaking factors to provide additional margin for the core designer and operation of the plant.

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u Safety Eva?uation and Determination of Significant Hazarus Considerations The proposed change to the Operating License has been evaluated to determine whether it constitutes a significant hazards consideration as required by 10 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 consequences of an accident previously evaluated.

Penking factors are used as input in the transient analyses, large-break LOCA analysis, small-break LOCA analysis and fuel handling accident analysis.

The transient analyses have used the currently proposed peaking factors beginning with the analysis of Unit 1 Cycle 12. This analysis, "Safety Evaluation of Increased Peaking Factors, NSPNAD 8705," was submitted to the NRC Staff as Exhibit D to License Amendment Request dated April 13, 1987. It was approved by the NRC Staff on July 8, 1987. Subsequent Cycles have been analyzed with these peaking factor limits and found to meet all acceptance criteria. An updated transient analysis is enclosed as Exhibit D.

The large-break LOCA analyses supporting this change is enclosed as Exhibit E. This analysis meets the requirements of 10 CFR Part 50.46 and Appendix K per the guidance provided in SECY 83 472. This methodology is currently being reviewed by the NRC Staff. The methodology requires an estimate of the peak cladding temperature at the 95 percent probability level and assurance that this value is lower tnan the value calculated using the required features of Appendix K.

As discussed in the proposed Bases page TS.3.10-9, the peak linear heat generation rate must be below 14.2 kw/ft at the 95% probability level.

SECY 83-472 states "We interpret the 95 percent probability level of the peak cladding temperature to mean that in 95 out of 100 LOCAs of a particular size, che peak cladding temperature would be below this value." We will comply with this requirement by analyzing each cycle's expected peak linear heat generation rate to ensure at the 95%

probability level it is below 14.2 kw/ft. This analysis is for Westinghouse fuel. Unit 2 Cycle 12 will be operating with 1/3 core of Exxon fuel until March of 1989. As shown in Figure 1 the twice burned Exxon fuel will never be limiting. This figure plots the maximum average assembly power (P-bar) for Westinghouse and Exxon assemblies at a given time in core life. The estimated exposure for Unit 2 Cycle 12 is 8 GWD/MTU in September 1988. Since the fuel designi are similar and the Westinghouse fuel is more limiting, the referenced analysis will bound the Exxon fuel also.

The large break analysis evaluated several power shapes and is no longer a basis for the shape of the K(z) curve. However, the small break analysis is based on the K(z) curve in the Technical Specifications and therefore, no change has been proposed to the K(z) curve.

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• The small break LOCA analyses supporting the proposed peaking factors was submitted to the NRC Staff as Exhibit F to License Amendment Request dated January 13, 1986. It was approved by the NRC Staff on April 3, 1986.

Previous fuel handling accidents have assumed the radial peaking factor to be 1.65 as required by 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." The proposed the hot rod peaking factor is 1.70. However, the LOCA analysis assumed a maximum assembly peaking factor (P-bar) of 1.50. This restricts the maximum assembly power to 1.5 times the average power and thus, the existing fuel handling accident analysis is bounding.

The proposed Bases refer to 15.8 kw/ft being associated with an Fq-2.50, as shown below:

kw/ft generated -

(core average power)(1.015)(0.974)(F l_

in the fuel rod (total number of feet of fuel rods) q 15.8 kw/ft - (1650000)(1.02)(0.974)(2.5)(1.0015)

(121)(179)(12) where: 1,650,000 - Core power in kw 1.02 - Appendix K requirement to account for calorimetric uncertainty 0.974 - Fraction of the fission energy released inside the cladding 1.0015 - Accounts for the axial shrinkage in the fuel stack 121 - Number of assemblies in the core 179 - Number of fuel rods per assembly 12 - Number of feet of fuel rod per rod

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

This change in Technical Specifications limits will not create a new or different kind of accident.

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3. T.te proposed amendment will not involve a significant reduction in the margin of safety.

As shown by the conclusions of the analyses, these changes will not result in a significant reduction in the plant's margin ot' safety.

.The Commission has provided guidance (March 6, 1986 Federal Register) concerning the application of the standards in 10 CFR 50.92 for determining whether a significant hazards consideration exists by providing certain examples of amendments that will likely be found to involve no significant hazards considerations. The changes to the Prairie Island Technical Specifications proposed in this amendment request are representative of NRC example (vi): because they involve changes which either may result in some change to the probability or consequences of a previously analyzed accident or may change in some way a safety margin, but where the results of the change are clearly within 10 CFR 50 requirements and previous NRC guidance. Based on this guidance end 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 effluent 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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