ML19341D385
| ML19341D385 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/25/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19341D381 | List: |
| References | |
| NUDOCS 8103050506 | |
| Download: ML19341D385 (3) | |
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.....l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIC _
RELATED TO AMENDMENT NOS. 66 AV 65 TO FACILITY OPERATING LICENSE N05. DPR-32 AND DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281 Introduction By letter dated May 15, 1980, Virginia Electric and Power Company (the licensee) requested amendments to the Surry Power Station, Unit Nos. 1 and 2 licensae which would change the Technical Specification limits for enrichment of new and spent fuel. The licensee proposed an enrichment of 4.1 weight percent U-235 in the new and spent fuel storage locations. The storage of increased enrichment fuel is necessary to permit the licensee's participation in a Department of Energy demonstration and evaluation program concerned with high burnup technology. This higher enrichment can permit a higher average discharge burnup of the fuel from the reactor.
We have not completed our review of the safety aspects of operating the reactor at an enrichment of 4.1 weight percent U-235 and extended burnup to 45,000 WD/hTU.
We have re~ iewed the safety aspects of storing 4.1% fuel v
in the new and spent fuel storage racks. We have also reviewed the opera-tion of the reactor with fuel enriched to 3.7% which is an insignificant increase over the current Technical Specification value. Since we are limiting the operating value to 3.7%, the limit on storage will also be 3.7% even though the analysis was done for 4.1%.
The reason for granting an increase to 3.7% from 3.6% is that the licensee is procuring fuel at 3.6% enrichment and fuel procured has a tolerance on enrichment which could sl.ightly exceed 3.6%
We are continuing
> nf the licensee's request to operate the reactor with an inc. -
vichment to 4.1% and burnups to 45,000 WD/MTU.
Discussion The new fuel storage racks consist of rows of storage boxes, each of which consist of a square stainless steel cylinder 1/8 inch thick with a 9-inch interior dimension.
Boxes are placed on 21-inch centers in the row and the rows are either 21 or 30 inches apart. The spent fuel storage racks consist of square cylinders of stainless steel having exterior dimensions of 9.12 inches and a wall thickness of 0.090 inches.
These boxes are placed in an array having 14 inch' center-to-center spacing.
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Calculations of the effective multiplicatien factor were made for both storage racks for the nominal (design) configuration and for normal (expected) and abnormal variations from the design.
Normal variations included off-center placement of fuel in the boxes, variations in box-to-box spacing, moderator density (temperature) and box wall thickne ;.
Abnormal configurations included the effect of dropping a bundle on top of the racks so that it entered a box and impacted on the stored fuel and dropping an assembly across the racks.
Dropping an sssembly between storage locations or immediately alongside the racks is precluded by the design of the racks.
Base case calculations were performed with the KEND-IV Monte-Carlo code.
Sensitivity studies were done with the EXTERMINATOR diffe!!cr. R.?ory code.
The KENO-IV code has been benchmarked against experiment and shown R be acceptably accurate.
Calculations were performed at the maximum fuel enrichment (4.1 weight percent U-235) in order t0 ensure maximum multiplication factors.
This evaluation applies to the storage of the fuel and any effects of the enrichment on operations will be evaluated in an acciaent analysis.
Evaluation Calculations of the effective multiplication factor for the fresh fuel storage racks with fresh 4.1 weight percent U-235 fuel show that the value is 0.47 for an infinite array on a 21.0 inch pitch and no moder-ation.
Calculations of the infinite lattice multiplication factor as a function of moderator density yields a value of 0.973 at a density of about 0.1 gm/cmd. This meets our acceptance criterion of 0.98 (Standard Review Plan 9.1.1) but in view of the small margin a series of calculations which more properly accounted for the actual geometry were performed.
In these calculations the geometry in the north-south direction was accurately modeled and infinite dimensions were assumed in the east-west direction.
Under thesq conditions the maximum multiplication factor occurred at about 0.07 gms/cma and was 0.90.
In view of the use of standard calculational techniques which have been benchm3rked and of the large margin to acceptance criteria in the properly modeled calculation we conclude that an acceptable assessment has been made of the criticality effect of storing fuel of 4.1 weight percent U-235 in the new fuel storage racks.
1 The calculation of multiplication factor for the storage of 4.1 weight percent U-235 fuel in the spent fuel storage racks results in a value of 0.924 for the nominal configuration and 0.931 for the worst case normal variation including calculational uncertainty. The value includ-ing the worst case abnormal configuration is 0.938 (for a pool tempera-
'i ture of 250*F (no credit taken for boron in the water). Based on the fact that standard calculational techniques which have been benchmarked are used, that calculational and physical uncertainties have been included, and that the results meet our acceptance criterion of 0.95 (standard Review Plan 9.1.2) we conclude that an acceptable assessment of the criticality of the spent fuel storage pool has been made.
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Based on our review we conclude that any number of* fuel assemblies of the standard Westinghouse 15x15 or 17x17 having enrichments no greater than 4.1 weight percent U-235 (51.7 grams of U-235 per cm.) may be stored in the spent fuel racks at the Surry Power Station.
- Further, the new fuel storage racks at Surry may be completely filled with similar assemblies with enrichment of 4.1 weight percent U-235 or less.
We have reviewed the safety aspects including accident analyses for operating the reactor at an enrichment of 3.7% U-235. This is an increase from the Technical Specification value of 3.6%.
This change is less than 3% and any consequences of this change on accidents are insigni-ficant. Since this enrichment change is insignificant and we have not approved operation of the reactor to burnups above those now allowed, we conclude that this change is acceptable..
ENVIRONMENTAL CONSIDERATIONS' We have determined that this action does not authorize a change in effluent types or total amounts nor and increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the action is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with this action.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assuran:e that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in complia'nce with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date: February 25, 1981
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