ML17320A521
| ML17320A521 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/04/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17320A520 | List: |
| References | |
| NUDOCS 8305170109 | |
| Download: ML17320A521 (5) | |
Text
~ ~ p.l\\ AEoy UNITED STATES IE.:
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 o
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
~ ~
RELATED TO AMEHDMEHT NO.
- 73. TO FACILITY OPERATING L!CENSE NO.
DPR-58 AND AMEt<DMEHT NO.
55 TO FACILITY OPERATING LICENSE NO.
DPR 74 INDIANA AHD MICHIGAN ELECTRIC COMPANY DONALD C.
COOK NUCLEAR PLAt<T UNIT NOS.
1 AHD 2 DOCKET HOS.
50-315 AtlD 50-316 lA0' QwQ
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INTRODUCTION By letter dated February 28, 1983, Indiana and Michigan Electric Company has requested amendments to Facility Operating License IIo. DPR-58 for D.
C.
Cook Unit 1 and Facility Operating License No.
DPR-74 for D.
C.
Cook Unit 2.
This proposed change will increase the maximum reload fuel enrichment for Unit 1 from 3.5 weight percent U-235 to 4.0 weight percent U-235.
The submittal includes analyses of the effects of the higher enrichment on the criticality aspects of both,the new and spent fuel racks at the D.
C.
Cook Units.
These changes will permit the storage of the Westinghouse 15xl5 Optimized Fuel Assembly (OFA) design fuel in the D.
C.
Cook new fuel storage area and spent fuel pool.
The effect of the proposed change on plant operating characteristics is still under staff review.
- However, we have completed our review of those portions of the amendment request which relate to the storage of the new fuel in the new fuel vaults and to new fuel storage in the.spent fuel pool in preparation for loading into the reactor, if operation with such reload is subsequently'uthorized by the NRC.
There is no change in the total number of assemblies authorized for storage in the spent fuel storage pools.
DISCUSSIOI~
~A1 i
II h
d The Westinghouse 15x15 OFA enrichment will not exceed 4.0 weight percent U-235.
Therefore, the proposed Technical Specification changes limit the maximum reload fuel enrichment for Westinghouse reload fuel in D.
C.
Cook
'nit 1 to 4.0 weight percent U-235.
However, the criticality aspects of the storage of Westinghouse 17x17 standard fuel assemblies with 4.5 weight percent U-235 fuel in the new fuel racks and Westinghouse 15x15 OFAs with 4.05 weight percent U-235 fuel in the spent fuel pool have been analyzed using the KENO-IV Monte Carlo computer code for reactivity determination with neutron cross sections generated by the AMPX code package.
A 218 energy energy group cross section library is generated from ENDF/B-IV data.
I These codes have been benchmarked against a set of 27 critical experiments in the range. of pellet diameters, water-to-fuel ratios and U-235 enrichments that encompasses the D.
C.
Cook design..
This benchmar king led to the conclusion that the calculational model is capable of determining the multiplication factor (k ff) of the new and spent fuel racks to within 1.3 percent with a 95 percent probability at the 95 percent confidence level.
New Fuel Stora e Rack Anal sis The criticality of fuel assemblies in the new fuel storage rack is prevented by maintaining a minimun separation of 21 inches between assemblies.
Al.though new fuel is normally stored in a dry 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) that k ff of the fuel assembly array eff will be; (1) no greater than 0.95 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 achievable moderation conditions other than full'ensity unborated water.
In addition to the calculational method uncertainty mentioned previously, uncertainties and biases due to mechanical tolerances such as stainless steel thickness, cell inner diameter, center-to-center
- spacing, and asymmetric assembly position are included either by using worst case initial conditions or by performing sensitivity studies to obtain the appropriate values.
Credit is taken for the neutron absorption in the full length stainless steel angle irons at the corners of each fuel assembly.
Using these methods and assumptions, the nominal k ff of the new fuel storage eff racks fully flooded with unborated water is calculated as 0.9189.
The fuel is assumed to be the >lestinghouse.:17x17 standard fuel assembly design at a U-235 enrichment of 4.5 weight percent.
This fuel is similar neutronically to the 15x15 standard fuel assembly design.
Any additional neutr onic differences between the standard and OFA 15x15 assemblies are conservatively accounted for by use of the higher enrichment in the calculations, i.e., 4.5 rather than 4.0 weight percent U-235.
Adding the appropriate 95/95 probability/confidence
uncertainties a'nd biases'ields a value of 0.9343 for the multiplication factor.
This meets our acceptance criterion of 0.95.
Physically achievable water densities which could be caused by fire fighting operations such as sprinklers or fog nozzles are considerably too low (much less than,0.01 gm/cc) to yield keff values higher than full density water, and 'boiling between cells is prevented by the rack design.
In addition, events such as the inadvertent drop of a'n assembly between the outside periphery of the rack and the pit wall 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, concurrent 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 f value very much less than our acceptance criterion of 0.95.
Me, there-fore, conclude that fuel assemblies of the Westinghouse (OFA) 15xl5 design having enrichments no greater than 4.0 weight percent may be stored in the D.
C.
Cook new fuel racks.
Spent Fuel Stora e Rack Anal sis The criticality of fuel assemblies in the spent fuel storage rack is prevented by maintaining a minimm separation of 10.5 inches 'between assemblies and by inserting the neutron absorber, Boral, between assemblies.
Although spent fuel is normally stored in borated pool water containing approximately 2000 ppm boron, the HRC acceptance criterion for spent fuel storage is that there is a
.95 percent probability at a
95 percent confidence level (including uncertainties) that 'k ff of the fuel assembly array will be less than 0.95 when fully flooded with unborated water.
-In addition to the calculational'ethod 'uncertainty mentioned previously, uncertainties and biases due to mechanical tolerances, thermal conditions, and
'84C particle self-shielding are included either by using worst case initial
'conditions or by performing sensitivity studies to obtain the appropriate values.
Credit is taken for the neutron absorption in full length structural materials and in solid materials added specifically for neutron absorption.
However, for conservatism, the minimum poison loading is assumed in these cases.
Using these methods and assumptions, the nominal k
of the spent fuel racks'ully eff flooded with unborated water is calculated as 0.9284.
The fuel is assumed to be Mesting'house 15xl5 OFA at a higher than expected enrichment of 4.05 weight percent, U-235.
The temperature of the water is taken as that which yields the largest reactivity.
A conservative value of 1.0 gm/cc is'sed for the density of the water.
Adding the appropriate 95/95 probability/confidence uncertainties and biases yields a value of 0.9482 for the multiplication factor.
This meets our acceptance criterion of 0.95.
Postulated events such as the inadvertent drop of an assembly between the outside periphery of the rack and the pool wall would not cause a criticality accident because of the assumption of the double
.contingency principle.
In other words, for accident conditions, the presence of soluble boron in the storage pool can be assumed and would result in a keff value very much less than our acceptance criterion of 0.95.
We, therefore, concluded that fuel assemblies of the Mestinghouse (OFA) 15xl5 design having enrichments no greater than 4.0 weight percent may be stored in the 'D.
C.
Cook spent fuel pool.
t/ew Fuel Handlin The new fuel arriving at the site is typically placed in the new fue1 vault for initial storage.
Before insertion in the reactor co e th f 1
he spent fuel pool.
Upon insertion in the spent fuel pool, the fuel becomes sslightly contaminated due to the environment.
Subsequent t.th f
1 equen o.
e spent fuel pool from the ool f torage of new fuel and upon determining that a new f 1
bl b
ue assem y must be removed lin is re poo or examination, measurements, or for other reasons 1
h d
, specia an-be done in g
qu'red.
From discussions bith the licensee
. this fuel h
dl'll accordance with plant procedures concerning handling of components having a low level of activity.
We find this acceptable.
~ Conclusions Based on our review, we conclude that the storage racks meet the requirements of General Design Criterion 62 as regards criticality.
- Also, we conclude that any number of Westinghouse (OFA) 15x15 fuel assemblies of maximum enrichment no greater than 4.0 weight percent U-235 may be stored in the new and spent fuel
'acks of D. C.
Cook Units 1 and 2.
These conclusions are based on the following considerations:
1.
State-of-the-art calculational methods which have been verified by comparison with experiment have been used.
2.
Conservative assumptions have been made about the enrichment of the fuel to be stored and the pool conditions.
3.
Credible accidents have been considered.
4.
Suitable uncertainties have been considered in arriving at the final value of the multiplication factor.
5.
The final effective multiplication factor value meets our acceptance criterion.
We also conclude that the proposed changes to Section 5 of the D.
C.
Cook Technical Specifications adequately account for the reload fuel enrichment increase and are, therefore, acceptable.
'e Environmental. Consideration Me have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power Ieve1 and will not result in any significant enviroanental impact.
Having made this determination, we have further concluded that the amendments involve an action which is insignificant frcm the standpoint of environmental impact and, pursuant to 10 CFR
$51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion Me have concluded,,based on the considerations discussed aoove, that:
(I) because the amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated, do not create the possibility of an accident of a type different from any evaluated previously, and do not involve a significant reduction in a margin of safety, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance 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 compliance with the Commission's regulations and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
Nay 4, lg83 Principal. Contributor:
L. Kopp