ML20043F722
| ML20043F722 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/06/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20043F715 | List: |
| References | |
| NUDOCS 9006180041 | |
| Download: ML20043F722 (4) | |
Text
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WASHINGTON, D. C, 20656
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0s.12T AND lll TO f-FACILITY OPERATING LICENSE NOS. NPF-4 AND NPF-7 1
VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION. UNITS NO. 1 AND NO. 2 DOCKET NOS. 50-338 AND 50-339
1.0 INTRODUCTION
By letter dated September 30, 1988, as supplemented August 18, 1989, the Virginia Electric and Power Company (the licensee) requested a change to the Technical Specifications (TS) for the North Anna Power Station, Units No. I and 2 (NA-182). The proposed change would increase the allowable enrichment of fuel assemblies irradiated at NA-1&2 to 4.3 weight percent (w/o) U-235. An
.' increase in the current NA-182 Technical Specifications (TS) limit of 4.1 w/o U-235 to 4.3 w/o U-235 would allow an increase in batch average discharge burnup to levels approaching the currently licensed limit of 45,000 Megawatt Days per Metric' Ton Uranium (MWD /MTU). The enrichments currently used limit the batch average burnup to a value from 38 000 MWD /MTU to 42,000 MWD /MTU depending.onthenumberoffuelassembliesloadedeachcycle. An increase in the enrichment limit would result in significant fuel cost cycle savings and enhance fuel: management plans to increase batch average discharge burnups.
.The safety impact for operation of NA-182 with high burnup fuel was oreviously addressed by the licensee in-letters to the NRC dated December 4,1980, March 6 and 26,.1981 and July 24, 1981.
By letter dated April 9, 1984, the NRC apprcved a
operation of NA-182 to a batch discharge of 45,000 MWD /MTU. A generic impact of extended ~burnup on the design and operation of Westinghouse fuel was addressed in WCAP-10125-P-A, " Extended Burnup Evaluation of. Westinghouse Fuel," dated December 1985.
In addition, the NRC made an independent assessment of the environmental and economic impacts of the use of extended burnup fuel in light water power reactors. This assessment was dated February 1988 and entitled
" Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors,"
- Pacific Northwest Laboratory, NUREG/CR-3009. The overall findings of NUREG/CR-3009 were that no significant adverse effects would be generated by increasing the present batch-average burnup level to values of 50,000 MWD /MTU
-or above, as long as the maximum rod average burnup of any rod is no greater than 60,000 MWD /MTV. Since the findings of these evaluations provided in.
NUREG/CR-3009 concerning the impact of extended burnup fuel are valid for an enrichment of 4.3 w/o U-235, and since the NA-1&2 spent fuel storage facility is currently licensed to 4.3 w/o U-235 (NA-1&2 Amendment Nos. 61 and 45 issued 9006180041 900606 PDR ADOCK 05000338 P
4h December 21,1984),- the license's submittal addresses only the impact of increased enrichment on the requirements for the currently approved new fuel storage racks at NA-1&2.
- The specific 10 CFR Part 50 Appendix A General Design Criteria for new fuel storage facilities are listed in Section 9.1.1 of the Standard Review Plan (NUREG-0800). Since no >hysical modifications are being made to the current NA-182 new fuel racks, tie licensee's analysis only addresses the impact of the increased enrichment on the requirement of suberiticality under normal and postulatedabnormalrackconditions(GeneralDesignCriterion62). The highest effective multiplication factor (K-effective)* allowable by Section 9.1.1 of NUREG-0800 for all conditions is less than 0.95 fully flooded and 0.98 for optimum low density moderation.
The August 18, 1989 letter provided additional information concerning the licensee's environmental assessment uf extending the current limits on enrichment.
The additional information did not alter in any way the staff's initial determi-nation of no significant hazards consideration as noticed in the Federal Register on November 16, 1988 (53 FR 46163).
2.0 DISCUSSION The new fuel storage area at NA-182 consists of nine parallel rows of storage racks with a total capacity of 126 fuel assemblies. Each storage location consists of a square 9 inch (inside measure) stainle s steel box, 165 inches tall with walls 1/8 inch thick. The storage area walls and floor are concrete.
A steel grating at the top prevents accidental placement of an assembly between storage cans. The storage area is normally dry.
The computer modeling of the storage racks was performed in three-dimension (3-D) to minimize unnecessary conservatism and uncertainty. All K-effective-l calculations were performed with the Monte-Carlo program KENO V.a within the i
SCALE package. The SCALE package automatically processes cross sections through NITAWL and BONAMI to create a set of resonance self-shielded cross sections for use by KENO. This code is widely used for nuclear criticality analysis and is acceptable. Because all calculations for this analysis were made using a discrete pin representation, no spatial self shielding was performed prior to the KEN 0 execution. The cross section set chosen was the-27 group ENDF/B-IV data contained in the SCALE package.
Sufficient neutron histories were run for each case to limit the statistical uncertainties in the K-effective to less than 0.4% delta K/K.
The base condition gg the analysis consisted of a fully loaded storage area of 126 fresh 4.3 w/o U enriched >: n-ilic ;= tared. nominally in the storage cans. Fuel assembly dimensions and material data are provided below:
- K-effective is tee ratio of neutrons from fissions in each generation to the total number lost by absorption and leakage in the preceding generations.
To achieve criticality in a finite system, K-effective must equal 1.0.
l I
NORTH ANNA 17 X 17 FUEL ASSEMBLY DATA 235 Fuel Enrichment 4.3 w/o U Assembly Pitch 8.466 in.
Pellet Diameter 0.3225 in.
Diametral Gap 0.0065 in Clad Thickness 0.0225 in Clad 0.D.
0.3740 in.
Pellet Material 955 th. dens. UO 2
Clad Material Zircaloy-4 Fuel Rod Pitch 0.4960 in.
Active Fuel Length 144.0 in, i
Fuel Rods / Assembly 264 Guide Tubes / Assembly 25 Guide Tube Material Zircaloy-4 Guide Tube 0.D.
0.482 in.
Guide Tube I.D.
0.450 in.
Several fuel assembly and rack components have been neglected in this model-for simplicity and conservatism. Assembly top and bottom nozzles (SS-304), grids (Inconel), sleeves (SS-304), and all storage rack structural materials other than the storage can itself were modeled as void or moderator regions. These omissions are all conservative from a criticality standpoint because steel and Inconel are both strong neutron absorbe The air regions in the storage areaweremodeledaswatervaporat10~gs. gram per cubic centimeter (g/cc).
Normal air humidity variations from dry conditions to heavy fog can result in water densities ranging from 0 to.0025 g/cc.
In addition, fire or a pipe break can result in flooding of the storage area by foam or water of many possible densities. To allow for these conditions, the air storage area were assigned water densities ranging from 10~6 regions in the g/cc to 0.998 g/cc.
Eccentric assembly positioning or a seismic event can lead to small assembly pitch changes. -Assuming the rack does not deform leads to a maximum pitch change for any two assemblies of 0.57 inches. Although any pitch changes are likely to be random, the effect of pitch reduction on K-effective ha,e been conservatively determined by reducing the pitch of all the storage: locations by 0.5 and 1.0 inch.
A dropped assembly could result in the fuel being compacted within the storage cell. To conservatively model this accident, the fuel pellet diameter of all assemblies in the rack was increased 105. Calculations were performed assuming no change in assembly height and with a change in assembly) height which preserves the total fuel volume (both at 95% theoretical density UO The compaction 2
effect was determined at two moderator densities.
In the compaction model the fuel was assumed.to contact and radially expand the clad (i.e., clad thickness waspreserved).
d 4
The results of the analysis for the worst-case normal configuration (including humidity changes and When all uncertainties (95/95)pitchchanges)showedthatK-effectiveis0.572.
are statistically combined and added, the result is 0.587.
The results of the analysis for the worst-case abnormal configuration (worst-case normal configuration plus maximum difference caused by accident condition) showed that K-effective is 0.884 Whenalluncertainties(95/95)are statistically combined and added, the result is 0.914 3.0 EVALUATION Tg35results discussed above indicate that for a fuel enrichment of 4.3 w/o l
V
, the NA-1&2 new fuel storage area meets the criticality limit of K-effective less than 0.95 fully flooded and 0.98 for optimum low density moderation, the criticality specifications set forth in the Standard Review Plan (NUREG-0800) and the requirements of 10 CFR Part 50 Appendix A General Design Criteria 6?.
Finally, the safety analysis approved in the NA-1&2 Amendments No. 61 and 45 issued December 21, 1984 provide the same assurance for spent fuel.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on June 6, 1990 (55 FR 23154).
Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of these amendments will not have a significant effect on the quality of the human environment.
-5.0 CONCLUSI0fl 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 the anendments will not be inimical to the comon defense and security or to the health and safety of the public.
l Date:
June 6, 1990 Principal Contributor:
Leon Engle
.