ML20211E192
| ML20211E192 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/14/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211E182 | List: |
| References | |
| NUDOCS 8610220376 | |
| Download: ML20211E192 (5) | |
Text
v 18 4
UNITED STATES N
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20655 g
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SisPPORTING AMENDMENT NO. 92 TO FACILITY OPERATING LICENSE NO. DPR-72 FLORIDA POWER CORPORATION, ET AL.
CRYSTAL RIVER UNIT NO. 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 INTRODUCTION By letter dated December 10, 1985, Florida Power Corporation (FPC or the licensee) requested amendment to the Technical Specifications (TSs) appended to Facility Operating License No. DPR-72 for the Crystal River Unit No. 3 Nuclear Generating Plant (CR-3). The proposed amendment would increase the permitted enrichment for fuel to be stored in both Spent Fuel Pool B and the dry fuel storage rack at CR-3.
The current analyses and TSs permit a 3.5 weight percent uranium-235 enrichment for all storage areas. This amendment would increase the enrichment to 4.0 weight percent for Storage Pool B and the dry fuel storage rack only.
In support of the proposed amendment, the licensee has submitted two reports (Refs. 2 and 3) l Science Office of Black & Veatch (Southern Science) prepared by.the Southern The NRC staff has reviewed _the proposed amendment and prepared the following evaluation.
EVALUATION a.
Spent Fuel Pool B Four independent methods.of evaluation were used to provide confidence in the reference criticality calculations. These methods included the (1) CASM0-2E program, (2) AMPX-KEN 0 computer package with the NITAWL subroutine for performing uranium-238 resonance shielding calculations using the Nordheim integral treatment, and (3) NULIF-SNEID diffusion theory method of analysis.
The AMPX-KENO calculations were performed using both a 123 group cross section library and a more recently developed 27 group cross section library. The AMPX-KEN 0 computer codes are widely used in the calculation of spent fuel pool criticality where AMPX is used to generate the neutron cross section data and KEN 0 is used to perform the Monte Carlo criticality calculations.
The CASM0-2E code is a two-dimensional, transport theory code for calculating fuel assemblies that is used by a number of organizations to perform reactor core physics calculations. The NULIF-SNEID diffusion theory method of analysis is used by Southern Science. Appendix A of Reference 2 describes the benchmarking that has been performed by Southern Science for its AMPX-KEN 0 and CASM0-2E methods of evaluation. Based on our review of these benchmark calculations, we conclude that use of the AMPX-KEN 0 and CASM0-2E by Southern Science is acceptable for the criticality evaluation of Spent Fuel Pool B.
8610220376 861014 PDR ADOCK 05000302 P
d The reference calculation was perfonned for a Babcock & Wilcox (B&W) 15x15 fuel assembly for 4.0 weight percent uranium-235 enrichment and for nominal fuel assembly composition. The nominal Spent Fuel Pool B rack lattice spacing (center-to-center distance between rack locations) of 21.125 inches was used.
Pure water at a temperature of 20 C (density of 0.998 grams per cubic centimeter) was assumed. The lattice of storage locations is assumed to be infinite in all directions. All four independent calculational methods gave values of the effective multiplication factor, K that were in good agreement. Since CASM0-2E gave the highest value of K itYs, used by Southern Science as the reference value of K foraddedconIhv,atism. This reference value of K is 0.9221 and includestRYcalculationalbiasderivedfromthebenchmarkingrelufts. This evaluation of the reference case K is acceptable since a conservative value hasbeenobtainedwithappropriateTfbenchmarkedmethodsfora15x15B&Wfuel assembly with 4.0 percent enriched fuel at nominal rack qimensions and spent fuel pool water conditions.
The uncertainties treated included those due to rack lattice spacing, eccentricity of fuel assembly placement in a rack location, and fuel enrichment and density variation. The effect of these uncertainties when combined statistically at the 95/95 probability / confidence level is 10.0024 in reactivity. The uncertainty in the calculational methodology based on the benchmarking results is 0.0018 in reactivity on a 95/95 probability / confidence level. Statistically combining these two uncertainties gives an overall uncertainty of 0.0030 on a 95/95 probability / confidence level. Therefore, a K f 0.925 (0.9221 + 0.0030) is effunder the worst possible conservatively estimated to be the maximum K combinationofcalculationalandmechanical8k[ertaintieswitha95 percent probability at a 95 percent confidence level under normal conditions. This t6f[of0.925isacceptablesinceitmeetstheNRCstaff'scriterionof0.95for K
quantity.
The loss of pool cooling with a subsequent increase in the spent fuel pool temperature has been analyzed. The results presented show that reactivity decreases from its nominal value with an increase in pool temperature. A calculation simulating boiling within the fuel assemblies gave a decrease in reactivity from the reactivity for the pool water at nominal temperature.
Heat generated by 4.0 percent enriched fuel in Spent Fuel Pool B will not differ significantly from that generated by 3.3 percent enriched fuel for the same core power density. Therefore, the existing spent fuel pool cooling system will be able to maintain a temperature of 129'F when the fuel assemblies from 16 successive refuelings (944 spent fuel assemblies) are stored in the pools and to maintain a spent fuel temperature of 140 F or less when a fuel core is discharged to the spent fuel pools in addition to the 944 assemblies noted above. This cooling capability was previously reviewed and approved by the NRC staff as documented in the Safety Evaluation transmitted to the licensee by letter (Reference 4) dated November 17, 1980, from Robert Reid, NRC, to J. A. Hancock, Florida Power Corporation.
A fuel assembly accidentally positioned outside the rack cannot be located any closer than 10 inches from another fuel assembly and will, therefore, have a negligible reactivity effect. An assembly dropped on top of the rack will be about 20 inches away from other fuel assemblies and will have a
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, negligible reactivity effect. Moreover, for the accident conditions analyzed, credit may. be taken for the soluble boron present in the water which would reduce K significantly below the criterion of 0.95. We concludethatcredibikfaccident configurations will not lead to a reduction in the margin to criticality for Spent Fuel Pool B.
Irradiation of 4.0 percent enriched fuel will not appreciably increase the quantity of radioactive species in the spent fuel assemblies over that for fuel enriched to 3.3 or 3.5 percent. Therefore, the radioactive doses both within and outside the plant resulting from an accidental drop of a permitted load over Spent Fuel Pool B will not change over those previously analyzed and approved by the NRC staff.
b.
Dry Fuel Storage Rack The new configuration of the dry fuel storage rack consists of a 6x11 array of storage cells with a center-to-center spacing of 21.125 inches. This storage cell spacing is identical to that of the storage rack in Spent Fuel Pool B.
Therefore, the calculation performed for the Spent Fuel Pool B rack for K as a function of enrichment and the associated uncertainties are direck applicable to the dry fuel storage rack when flooded with pure water The maximum K is 0.944 for at a density of 1 gram per cubic centimeter.
the15x15B&Wfuelassemblywithauranium235enrichmento?d.5 percent.
f This K meets our criterion of 0.95 and is, therefore, acceptable.
eff The dry fuel storage rack was also analyzed for low-density water such as may occur for fog or mist. The analysis was performed with the KEN 0 Monte Carlo code using the 123 group neutron cross scution library. The results of preliminary analyses indicated that if all locations were filled with fuel, the criticality criterion would not be net for the 4.5 weight percent enriched fuel. A new configuration was selected such that the original 66 storage locations arranged in a 6x11 array were reduced to 54 storage locations arranged in three 3x6 arrays. Two rows of storage locations (rows 4 and 8), each containing six storage locations, would be blocked to prevent fuel from being placed within. The analysis of this new fuel storage configuration showed that the maximum K occurred at a water density of ff TheR for this configuration is 0.941.
0.075 grams per cubic centimeter.
Including uncertainties at least at a 95 M probability / confidence level gives a maximum K of 0.952 for fuel having a 4.5 weight percent uranium-235 enrichment. This*kf of 0.952 is acceptable since it meets the NRC staff's criterion of 0.98 fNfthis quantity.
Since the NRC staff's criteria for the storage of 4.5 weight percent uranium-235 enrichment fuel are met, the storage of 4.0 weight percent uranium-235 enrichment fuel in the dry storage rack is, therefore, acceptable.
On the basis of our review described above we conclude that fuel of the B&W IEx15 design having enrichment no greater than 4.0 weight percent uranium-235 may be safely stored in Spent Fuel Pool B and in the dry fuel storage rack, and the revised TSs are. acceptable.
. ~.
. ENVIRONMENTAL CONSIDERATION This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
CONCLUSION We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health a'd safety of the public.
n Dated:
October 14, 1986 Principal Contributors:
D. Fieno, N. Wagner, B. Mozafari 4
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REFERENCES 1.
Letter from E. C. Simpson (Florida Power Corporation) to H. R. Denton (NRC),3F1285-05, December 10, 1985.
2.
" Criticality Safety Analysis of the Crystal River Pool B Fuel Storage Rack with Fuel of 4% Enrichment," SSA-160, Southern Science Office of Black & Veatch, September 1985.
3.
" Criticality Safety Analysis of the New-Fuel Storage Vault With Fuel of
. 4.5% Enrichment", Southern Science Office of Black & Veatch (undated).
4.
Letter from Robert Reid (NHC) to J. A. Hancock (Florida Power Corporation) November 17, 1980.
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