ML17310B235
| ML17310B235 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 04/19/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17310B234 | List: |
| References | |
| NUDOCS 9404220178 | |
| Download: ML17310B235 (10) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
4 TO FACILITY OPERATING LICENSE NO.
NPF-41 AMENDMENT N0.60 TO FACILITY OPERATING LICENSE NO.
NPF-51 AND AMENDMENT NO. 4 TO FACILITY OPERATING LICENSE NO.
NPF-74 ARIZONA PUBLIC SERVICE COMPANY ET AL.
PALO VERDE NUCLEAR GENERATING STATION UNIT NOS.
1 2
AND 3 DOCKET NOS.
STN 50-528 STN 50-529 AND STN 50-530
- 1. 0 INTRODUCTION By application dated October, 26,
- 1993, as supplemented by letter dated
- March 28, 1994, Arizona Publ'ic Service Company (APS or the licensee) requested an amendment to the Technical Specifications (TS) for the Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3 (Appendix A to Facility Operating License Nos.
NPF-41, NPF-51, and NPF-74, respectively).
The Arizona Public Service Company submitted this request on behalf of itself, the Salt River Project A'gricultural Improvement and Power District, Southern California Edison
- Company, El Paso Electric Company, Public Service Company of New
- Mexico, Los Angeles Department of Water and
- Power, and Southern California Public Power Authority.
The additional information contained in the supplemental letter dated March 28,
- 1994, was clarifying in nature and thus within the scope of the initial notice and did not affect the NRC staff's proposed no significant hazards consideration determination.
2.0 DISCUSSION The licensee is requesting a revision to TS 5.3. 1 for PVNGS Units 1, 2, and 3
that will increase the maximum allowable fuel enrichment from 4.05 weight percent U-235 to 4.30 weight percent U-235.
There was no change requested to the current 52,000 HWD/HTU burnup.
The licensee provided supplemental information at the request of the NRC to bring TS 5.3. 1 into conformance with Generic Letter 90-02, Supplement 1 and to clarify the assumptions used in the Fuel Handling Accident Analysis.
A previous submittal to increase the fuel enrichment from 4.0 to 4.05 weight percent U-235 was granted March 9, 1988.
That submittal included a Combustion Engineering letter dated Hay 27,
- 1987, which verified that the original analyses of the new fuel, spent fuel (except as noted below) and the 9404220i78 9'404i9 PDR ADOCK 05000528 P
III 4k
intermediate
- racks, as well as the fuel elevator, fuel upender, and transfer machine were all performed for 4.30 weight percent U-235 fuel.
Since the k,ff values based on the storage of 4.30 weight percent U-235 fuel meet the NRC acceptance criteria of no greater than 0.95 for fully.flooded (unborated) conditions and 0.98 for optimum moderation conditions, the PVNGS storage facilities were found acceptable for.the storage of 4.05 weight percent U-235 fuel with one exception.
The spent fuel racks with neutron poison (boron) boxes in the cells were only analyzed for a maximum enrichment of 4.0 weight percent U-235 using the high density mode.
Therefore, a footnote that states, "No fuel with an enrichment greater than 4.0 weight percent U-235 shall be stored in high density mode in the spent fuel storage facility," was added to TS 5.3. 1.
The spent fuel storage racks are made up of 17 individual modules, twelve 8x9 assembly arrays, four Bxl2 assembly
- arrays, and a 9x9 assembly array.
The storage racks are stainless steel honeycomb structures with rectangular storage cells.
A cell blocking device is used in every other storage rack location which produces a checkerboard array of fueled and non-fueled locations.
In each fueled location, a stainless steel "L" shaped insert is used to position the fuel and maintain the minimum edge-to-edge spacing.'etween assemblies.
The new fuel storage racks consist of a 15x6 assembly array divided into two compartments which are separated from each other by a two-foot concrete wall.
In the long direction there is a minimum 9-inch edge-to-edge separation between assemblies and in the shorter perpendicular direction there is a
minimum 22-inch edge-to-edge separation between assemblies.
The new fuel storage rack design encloses each assembly in a stainless steel box with a minimum thickness equal to 0. 10 inches.
- 3. 0 EVALUATION The licensee has requested a technical specification change for a maximum fuel enrichment to be increased from 4.05 weight percent U-235 to a radially averaged 4.30 weight percent enrichment.
As stated in the discussion section
- above, the original analyses of the PVNGS spent fuel pools with high densi.ty storage confi'guration supported a maximum enrichment limit of 4.0 weight percent U-235.
The checkerboard storage configuration was analyzed with-a maximum uniform-enrichment of 4.30 weight percent U-235.
In addition, the dry new fuel storage racks were originally analyzed -for a maximum uniform enrichment..of.4.30 weight. percent 'U-235.
S ent Fuel Stora e Racks The two dimensional discrete ordinates transport cod'e DOT,-IV was used to determine the spatial solution and multiplication factor in the spent fuel storage racks.
The CEPAK lattice program was used to calculate the four neutron energy group cross-sections for fuel, water, and steel regions.
The data base for both fast and thermal neutron cross-sections for the CEPAK program were derived from several
- sources, mainly ENDF/B-IV.
These codes are
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widely used in the nuclear industry and have been benchmarked to adequately reproduce the critical values.
The group dependent poison cross-sections, if poison inserts were present in the array, were generated by a 123-group, P-3, S-8 XSDRNPH calculation.
The staff concludes that the resulting set of four-group poison cross-sections properly account for epithermal self-shielding.
Therefore, the staff finds the use of these codes acceptable.
The geometric buckling was supplied to CEPAK from DOT X-Y transport solution for a fuel assembly in the rack environment.
The staff concludes that the geometric buckling calculated, in this manner is indicative of the neutron environment of the fuel assembly in the spent fuel rack and is, therefore, acceptable.
The PVNGS spent fuel racks were modeled with a nominal pitch of 9.515 inches.
In each fueled location, a stainless steel "L" shaped insert with a nominal thickness equal to 0. 175 inches was modeled.
Periodic boundary conditions were employed to correctly account for the arrangement of fuel locations with the steel inserts and the non-fueled locations.
The modeling of the spent fuel rack assumed no axial leakage, no poison shims present in the assemblies, no grids, and no soluble boron in 68 degree Fahrenheit water.
The nominal.pitch of fuel cells modeled for the 16xl6 assembly is 0.506 inches.
The original analysis of the PVNGS spent fuel racks assumed that the assembl.ies were uniformly enriched to 4.30 weight percent U-235, the pellet diameter was 0.325 inches, the stack density of uranium dioxide in each rod was 10.061 g/cc.
The calculated k,<< value for the original nominal configuration was 0.88980.
The updated analysis included the reactivity effects of the radially averaged versus uniformly enriched fuel, a higher stack density (10.41 g/cc),
and a
larger pellet diameter (0.33 inches).
The calculated reactivity effects in delta k,<< units was 0.00200 for the radially averaged 4.30 weight percent fuel, 0.00400 for the higher stack density, and 0.00200 for the larger pellet diameter.
Thus the calculated k,<< value for the new nominal configuration was found to be. 0.89780. (0.88980
+ 0.00200
+ 0.00400
+ 0.00200).
The delta k <<-.for uncertainties in del.ta k,<< units was 0.00455 for the minimum cenfer.,to center. pitch, 0.00942 for the eccentric positioning of assemblies, 0.00184 for, the minimum monolith thickness, 0.00442 for temperature vari'ati'ons, 0.00150 for the minimum,L-insert thickness, 0.00350'or.
the, assembly enrichment, and 0;00714 for the methodology uncertainty.
The square root of the sum of the uncerta'inties squared is equal to 0.01407 delta k,<< units.
The calculation methodology has a bias equal to -0.00197 delta k~ units.
Therefore, the final k << value for the PVNGS spent fuel racks is equal to 0.90990 (0.89780
+ 0.01407 - 0.00197).
The design basis of k,<< value for spent fuel storage racks is 0.95.
Therefore, sufficient margin between the final calculated k,<< value and the design basis was demonstrated.
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New Fuel Stora e Racks The new fuel storage array was modeled using KENO by assuming the entire array of 4.30 weight percent U-235 assemblies were enclosed by a tight fitting two-foot thick concrete reflector.
The analysis considered hydrogenous moderation ranging from foam conditions (equivalent water densities of 0.0'/cc to 0. 10 g/cc) up to full flood conditions.
The resulting k,<< versus water density data exhibits two peaks; one from 0.06 to 0.07 g/cc water density, and one at full flood condition.
The nominal calculated k,<< value at 0.0625 g/cc water density was less than 0.80.
The nominal calculated k << value at full flood conditions was less than 0.90.
Both of these calculated k,<< values are below the design basis of 0.98 under conditions of optimum moderation and 0.95 when flooded.
Therefore, sufficient margin between the final calculated k,<<
values and the design bases were demonstrated.
Evaluation Summar The licensee has proposed a change to Section 5.3. 1 of the Technical Specifications.
The analysis submitted by the licensee were performed. using-well established methods with conservative assumptions.
The results were within staff limits for the maximum k,<< for the various scenarios
.analyzed.
Since no analysis for high density storage in the spent fuel racks was presented, the current checkerboarding restriction on the spent fuel storage racks will continue to apply.
Based on the above review and evaluation, the staff concludes that fuel assemblies having radially averaged enrichments up to 4.30 weight percent U-235 and 52,000 MWD/MTU burnup may be safely stored in the new and spent fuel storage racks at PVNGS.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arizona State official was notified of the proposed issuance of the amendment.
The State official had no comments.
- 5. 0 ENVIRONMENTAL CONSIDERATION Pursuant to: 10 CFRY51.21;" 51.32, and 51.35, an environmental assessment and finding of no significant-impact was published in the FEDERAL REGISTER on April 12, 1994 (59 FR'7402).
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.
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- 6. 0 CONCLUSION The Commission has concluded, based on the considerations discussed
- above, that:
(I) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: T. Polich Date:
April 19, 1994
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