ML20247F191

Safety Evaluation Supporting Amend 151 to License NPF-1

ML20247F191
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	03/17/1989
From:	Office of Nuclear Reactor Regulation
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ML20247F170	List: ML20247F165 ML20247F187 ML20247F191 ML20247F398
References
NUDOCS 8904030355
Download: ML20247F191 (4)

Text

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,aur' o UNITED STATES

~g 8" n NUCLEAR REGULATORY COMMISSION g ;E WASHINGTON, D. C. 20555 ,

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.151 TO FACILITY OPERATING LICENSE NO. NPF-1 PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFICPOWERANDLIGHTCDMPANY TROJAN NUCLEAR PLANT DOCKET NO. 50-344

1.0 INTRODUCTION

By letter dated March 1,1988, Portland General Electric Company (PGE) applied for an amendment to Facility Operating License NPF-1 of the Trojan Nuclear Plant to remove the maximum fuel enrichment limit specified in Technical Specifications (TS) Section 5.3.1 for reload fuel and to increase the enrichment limit for the new fuel storage racks specified in Technical Specification Section 5.6.1.1 to 4.5 weight percent U-235 (Ref.1). In support of this change, PGE submitted " Nuclear Criticality Analysis of 4.5 Weight Percent Uranium Enrichment in the New Fuel Storage -

Racks of the Trojan Nuclear Plant" (Ref. 2). This evaluation summarizes the results of core reanalysis of the new (unirradiated) fuel storage racks at Trojan and of operational considerations assuming 4.5 weight percent U-235 enriched Westinghouse standard 17x17 fuel. The spent fuel storage racks have previously been analyzed and approved for storage of l 4.5 weight percent U-235 in Amendment 88 to Facility Operating License No. NPF-1 dated June 8, 1984.

The Trojan new fuel storage rack consists of four rows of fuel storage cells with a nominal spacing of 21 inches center-to-center between cells.

The storage cells consist of 2 inch wide and 0.25 inch thick stainless steel angles in the four corners around the assemblies.

Although new fuel is normally stored in a dry (air) environment resulting in an extremely subcritical configuration, the NRC acceptance criteria for new fuel storage is that there is a 95 percent probability at a 95 percent confidence level (including uncertainties and biases) that k-effective of the fuel assembly array will be (1) no greater than 0.95 t when fully loaded and flooded with unborated water, and (2) no greater than 0.98 under conditions of optimum moderation if higher reactivities can be attained at achieiable underation conditions other than full density unborated water.

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-2 The PDQ-7 two-dimensional diffusion theory computer code was used to calculate the reactivities of the storage arrays. Neutron cross sections were generated by the CASM0-3 two-dimensional transport theory code.

CASMO-3 was also used to obtain transport theory to diffusion theory adjustment factors to account for possible inaccuracies to diffusion theory results at lower moderator densities. Cross sections genarated by CASMO-3 were used with the DIXY diffusion theory code to benchmark against a wide range of experimental data and were found to adequately reproduce the critical values.. The calculational bias was + 0.0004 and the uncertainty in the bias us 0.0023 rit the 95/95 probability / confidence level. Although the experimental conditions did not fully envelope those of the Trojan new fuel storage rack analyses, the absence of any observ-able trend in k-effective with any of the parameters examined (temperature, enrichment, water-to-fuel ratio, neutron leakage, water gap between assemblies) indicates that the calculational bias and uncertainty in the

~ bias obtained in the benchmarking is appropriate. In addition, since DIXY was used in the benchmarking rather than PDQ-7, one of the critical experiments was reanalyzed with CASM0-3 and PDQ-7. The results did not differ significantly from the result reported in the benchmark study, justifying the use of CASMO-3 and PDQ-7 for the Trojan analyses.

In addition to the calculational method uncertainty and bias, the maximum possible increase in k-effective calculated due to asymetric loading of assemblies in the storage racks and less than nominal spacing between storage locations was added to the calculated nominal k-effective.

Using these methods and assumptions, the nominal k-effective of the new fuel storage racks fully flooded with unborated water is calculated as 0.9378. The fuel is assumed to be the Westinghouse 17x17 standard fuel assembly design at a U-235 enrichment of 4.5 weight percent. Adding the appropriate 95/95 probability / confidence uncertainties and biases yields a value of 0.9403 for k-effective of the fully flooded storage racks.

The results of calculations assuming various amounts of water moderation ranging from 1 percent to 100 percent of full density water indicated that optimum moderation occurs when the racks are flooded with 100 percent dense water. Therefore, the staff's accet '.:nce criteria of 0.95 (fully flooded)and0.98(optimummoderation)aremet.

Postulated accident conditions such as the inadvertent drop of an assembly onto or within the storage facility would not cause a criticality accident because of the assumption of the double contingency principle. This states that it is unnecessary to assume two unlikely, independent, con-current events to ensure protection against a criticality accident.

Therefore, for accidents such as this, the absence of water in the new fuel storage pit can be assumed since assuming its presence would be a second unlikely event. Without water, any postulated assembly drop accident would result in a k-effective value significantly less than the staff's acceptance criterion of 0.95.

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. The staff, therefore, concludes that the Trojan new fuel storage racks meet the criticality requirements of General Design Criterion 62,

" Prevention of Criticality in Fuel Storage and Handling and that fuel assemblies of the Westinghouse standard 17x17 design having enrichnwnts no greater than 4.5 weight percent may be stored in the Trojan new fuel racks.

The deletion of the enrichment limit in the Fuel Design Specification 5.3.1 does not directly affect plant operating parameters or safety limits since there is no margin of safety or safety limit directly associated with fuel enrichment. Rather, the. safety margins are maintained by other i.

limits such as power, power distribution, reactivity coefficients, burnup, etc., which are verified to be acceptable with each reload safety eval-uation. Therefore, the deletion of the enrichment specification from TS 5.3.1 is acceptable.

3.0 CONTACT WITH STATE OFFICIAL The NRC staff has notified the Oregon Depart:nent of Energy of the proposed issuance of this amendment along with the proposed determination of no significant hazards consideration. No comments were received.

4.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 651.21, 51.32 and 51.35, an environmental assessment has been published (53 FR 43485) in the Federal Register on October 27, 1988. Accordingly, the Commission has determined that the issuance of this amendment will not result in any environmental impacts other than i those evaluated in the Final Environmental Statement.

5.0 CONCLUSION

The Commission has issued a Notice of Consideration of Issuance of Amend- '

ment to Faci 1Hy Operating License and Opportunity for Prior Hearing which was published in the Federal Register (53 FR 17807) on May 18, 1988.

No request for hearing or petition for leave to intervene was filed following this notice.

Wehaveconcluded,basedontheconsiderationsdiscussedabove,that(1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner,such (2) public activities will be conducted in compliance with the Connission's regu-latfuns,and(3)theissuanceoftheamendmentwillnotbeinimicaltothe commen defense and security or to the health and safety of the public.

PRINCIPAL CONTRIBUTORS:

L. Kopp T. Chan Dated: March 17, 1989

1 REFERENCES

1. Letter from D. Cockfield to U.S. Nuclear Regulatory Conmi".sion, License Change Application 165, dated March 1,1988.

2. Attachment to Letter from D. Cockfield to U.S. Nuclear Regulatory Connission, dated March 1, 1988.

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