ML17311A325
| ML17311A325 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/30/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17311A323 | List: |
| References | |
| NUDOCS 9410120296 | |
| Download: ML17311A325 (18) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20555-0001 94i012029b 940930 PDR ADOCK 05000528 P
PDR-SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
82 TO FACILITY OPERATING LICENSE NO. NPF-41 AMENDMENT NO.
69 TO FACILITY OPERATING LICENSE NO.
NPF-51 AND AMENDMENT NO. 54 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 letter dAted February 18, 1994, the Arizona Public Service Company (APS'r the licensee) submitted a request for changes 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 Agricultural 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 proposed changes would allow credit to be taken for burnup of spent fuel assemblies in establishing storage locations within the spent fuel storage pool.
The current spent fuel storage pool is configured to store fresh fuel assemblies with a maximum radially average enrichment of 4.30 weight percent (w/o) U-235 in a two-out-of-four checkerboard array.
The proposed changes would allow for three distinct storage regions.
Region 1 would allow storage of fresh fuel assemblies with a maximum radially averaged enrichment equal to 4.30 w/o U-235 in a checkerboard configuration.
'Region 2 would allow storage of spent fuel assemblies in a three-out-of-four. confi'guration.
Region 3 would allow storage of spent fuel assemblies in every location (four-out-of-four configuration).
Allowable storage in Region 2 or 3 depends upon the initial assembly enrichment and the assembly
- burnup, as shown in TS Figure 5.6-1.
In each fueled location, a stainless steel L-shaped insert is used to position the fuel and maintain the minimum edge-to-edge spacing between assemblies.
The licensee supplemented their original amendment with a letter dated June 20, 1994.
This letter provided responses to a staff request for additional information.
The information was of a clarifying nature and did not affect the staff's original no significant hazards determination.
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- 2. 0 EVALUATION The analysis of the reactivity effects of fuel storage in the spent fuel storage racks was performed by ABB-Combustion Engineering (CE) using the two-dimensional discrete ordinates transport theory DOT-IV computer code, with four energy group neutron cross sections generated by the CEPAK code.
These codes have been previously used by CE for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments.
These experiments simulate the PVNGS fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment and assembly spacing.
In Harch 1992, the NRC issued Information Notice 92-21 and Supplement 1 concerning discrepancies that were discovered in spent fuel pool reactivity calculations.
The discrepancies were due to an overestimation of neutron absorption in the CEPAK generation of cross sections.
These discrepancies were found to exist only in regions containing a strong neutron absorber (poison).
Since neutron poison is not present, this problem does not exist for the PVNGS racks.
The staff concludes that the analytical methods used are acceptable and capable of predicting the reactivity of the PVNGS storage r acks with a high degree of confidence.
Region 1 will be comprised of three 9x8 storage
- racks, one 12x8 storage
- rack, and one 9x9 storage rack.
To prevent inadvertent storage of a fuel assembly in a cell required to be vacant, the cell blocking devices currently in place in every other storage cell location will remain to maintain a two-out-of-four checkerboard configuration.
Therefore, the configuration of Region 1 is identical with the current spent fuel pool configuration, and the previously approved criticality analysis remains applicable.
Region 2 will be comprised of three 9x8 storage racks and one 12x8 storage rack.
Since storage in Region 2 will be limited to a three-out-of-four storage arrangement, cell blocking devices will be employed in one out of every four storage cell locations to preclude the possibility of an unanalyzed assembly configuration.
Region 3 will be comprised of six 9x8 storage racks and two 12x8 storage racks.
Since fuel assemblies may be stored in every Region 3 cell location, no cell blocking devices will be installed in Region 3.
Cell blocking devices will also be placed along the Region 2
interface with Region 3 to eliminate the possibility of an unanalyzed arrangement of assemblies.
The modeling of Regions 2 and 3 included several conservative assumptions.
These assumptions neglected the reactivity effects of axial leakage, poison shims in the assemblies, structural grids, and soluble boron in the 68 'F pool water.
These assumptions tend to increase the calculated effective multiplication factor (k,<<) of the racks and are, therefore, acceptable.
The stored fuel assemblies were modeled as CE 16x16 assemblies with a nominal pitch of 0.506 inches between fuel rods, a fuel pellet diameter of 0.33 inches, and a
UOz density of 10.4 g/cc.
DOT-IV calculations were used to construct a curve of burnup versus initial enrichment for both Regions 2 and 3
(TS Figure 5.6-1) such that all points on
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the curve produce a k,<<value (without uncertainties or biases) of 0.93.
This method of reactivity equivalencing has been accepted by the NRC and used for numerous other spent fuel storage pools which take credit for burnup.
The NRC criticality acceptance criterion for fuel storage is that k,> be no greater than 0.95, including all uncertainties at a 95X probabi)Yty/
95K confidence level.
Therefore, the reactivity effects due to uncertainties in minimum center-to-center pitch, eccentric positioning of assemblies, minimum monolith thickness, temperature variations, minimum L-insert thickness, assembly enrichment, and assembly burnup were obtained as well as a
methodology uncertainty and bias.
These were applied to the nominal value of 0.93 to obtain a final k << of 0.944 for the spent fuel racks.
This meets the NRC criterion of no greasier than 0.95 and is, therefore, acceptable.
Host abnormal storage conditions will not result in an increase in the k, < of the racks.
However, it is possible to postulate
- events, such as an assembly drop on top of a rack or between a rack and the pool walls or the misloading of an assembly, with a burnup and enrichment combination outside of the acceptable area in TS Figure.5.6-1, which could lead to an increase in reactivity.
However, for such events, credit may be taken for the presence of 2150 ppm of boron in the pool water required by TS 3.9. 13, since the staff does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident (double contingency principle).
The reduction in k,<< caused by the boron more than offsets the reactivity addition caused by credible accidents.
Therefore, the staff criterion of k,<< no greater than 0.95 for any postulated accident is met.
The following TS changes have been proposed as a result of the requested amendment.
The staff finds these changes acceptable.
(1)
TS 3.9. 13 and its associated Bases are added to require a minimum boron concentration of 2150 ppm in the spent fuel storage pool with a 7-day surveillance interval whenever fuel assemblies are in the pool.
This is acceptable since this boron concentration more than offsets the reactivity addition caused by any credible accident and the staff considers a 7-day surveillance interval to be a reasonable interval to verify boron concentration.
(2)
TS 5.3. 1 is, modified to delete the footnote requiring that "no fuel with an enrichment greater than 4.0 weight percent U-235 shall be stored in a high density mode in the spent fuel storage facility." This is acceptable since the criticality analysis crediting fuel assembly burnup has shown that initial assembly enrichments up to 4.30 w/o can be stored in the three-region pool, depending on the associated assembly burnup.
(3)
TS 5.6. 1 is modified to clarify the uncertainties margin as the 95K probability/95N confidence level, to establish a nominal 9.5-inch center-to-center distance between adjacent storage cell locations, to define the three regions of the spent fuel pool, and to incorporate TS Figure 5.6-1.
This is acceptable since the criticality analysis has shown that the NRC acceptance criterion of k,<< no greater than 0.95 is met for these conditions.
0
Based on the review described
- above, the staff finds that the criticality aspects of the proposed changes to the PVNGS spent fuel pool TS are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling.
Although the PVNGS TS have been modified to specify the above-mentioned fuel as acceptable for storage
'in the fresh or spent fuel racks, evaluations of reload core designs (using any enrichment) will, of course, be performed on a
cycle-by-cycle basis as part of the reload safety evaluation process.
Each reload design is evaluated to confirm that the cycle core design adheres to the limits that exist in the accident analyses and TS to ensure that reactor operation is acceptable.
Additionally, the staff reviewed the proposed change for heat load considerations.
The licensee's heat load calculations were based on a full pool with 1300 fuel assemblies, and were found to be acceptable in the NRC Safety Evaluation Report.
The maximum number of fuel assemblies that can be stored in the proposed three-region configuration is 1054 fuel assemblies.
The actual loading pattern will therefore have a lower decay heat than assumed in the calculations for a full pool.
The heat load calculations are bounding for the three-region configuration and are, therefore, acceptable.
The staff also questioned the licensee regarding the potential adverse effects to the consequences of the fuel handling accident outside containment (this TS amendment allows a closer spacing of fuel assemblies, in certain regions of the spent fuel pool, than was allowed previously - thereby potentially
.allowing more fuel assemblies to be damaged on a fuel, handling accident).
The licensee stated that the original'icensing basis allowed for spent fuel to be loaded in either a 4x4 array or a checkerboard
- array, depending on the use of boraflex poison.
Therefore, a fuel handling accident was assumed to occur with maximum loading of the pool.
Section 15.7.4 of the UFSAR, "Radiological Consequences of Fuel Handling Accidents," states:
The fuel assemblies are stored within the spent fuel rack at the bottom of the spent fuel pool.
The top of the rack extends above the tops of the stored fuel assemblies.
A dropped fuel assembly could not strike more than one fuel assembly in the storage rack.
Impact could occur only between the ends of the involved fuel assemblies, with the lower end fitting of the dropped fuel assembly impacting against the upper end fitting of the stored fuel assembly....
Horizontal impact of a fuel assembly could result from a dropped fuel assembly falling in the horizontal position....
The worst case bundle impact results from the horizontal drop....
As a result of the fuel assembly
- drop, no more than four rows of fuel rods would fail.
The staff reviewed the UFSAR to ensure that the new configuration was bounded by the current analysis.
Neither the individual fuel assembly characteristics nor the rack construction are changed by this TS amendment, only the location of assemblies within the spent fuel pool.
The fuel pool rack construction precludes more than one assembly from being impacted in a fuel handling
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accident.
Therefore, the UFSAR analysis conclusion regarding the worst scenario for a dropped assembly (in which the horizontal impact of a fuel assembly on top of the spent fuel assembly damages fuel rods in the dropped
,assembly but does not impact fuel in the stored assemblies) continues to be limiting, and is, therefore, acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arizona State official was notified of the proposed issuance of the amendments.
The State official had no comments.
- 4. 0 ENVIRONMENTAL CONSIDERATION The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements.
The NRC staff has determined that the amendments involve 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 the amendments involve no significant hazards considera-
- tion, and there has been no public comment on such finding (59 FR 17593).
Accordingly, the amendments meet 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 the amendments.
5.0 CONCLUSION
The Commission has 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, (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 Contributors:
L. Kopp A. Dummer B. Holian Date:
September 30, 1994
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 2055&0001 September 21, 1994 MEMORANDUM TO:
Sholly Coordinator FROH:
SUBJECT:
Linh N. Tran, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation RE(VEST FOR PUBLICATION IN BIWEEKLY FR NOTICE NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE (TAC NOS. H87244, M87245 AND H87246)
Arizona Public Service Com an et al.
Docket Nos.
STN 50-528 STN 50-529 and STN 50-530 Palo Verde Nuclear Generatin Station Unit Nos.
1 2 and 3
Marico a Count Arizona Date of a lication for amendments:
August 5, 1993 Brief descri tion of amendments:
The amendments change the phrase "Pressurizer Pressure - Wide Range" to "Reactor Coolant System Pressure Wide Range" in item 4 of TS Table 3.3-10 and item 4 of Table 4.3-7.
These amendments will clarify the instrumentation required and eliminate potential confusion between the reactor coolant system pressure instruments and the pressurizer pressure instruments.
Date of issuance:
September 21, 1994 Effective date:
September 21, 1994 68, F cilit 0 eratin License Nos.
NPF-41 NPF-51 and NPF-74:
Amendments revised the Technical Specifications.
Date of initial notice in FEDERAL REGIS ER:
September 29, 1993 (58 FR 50962)
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 21, 1994.
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Sholly Coordinator No significant hazards consideration comments received:
No.
Local Public Document Room location:
Phoenix Public Library, 12 East McDowell Road,
- Phoenix, Arizona 85004 Linh N. Tran, Project Manager Project Directorate IV-2'ivision of Reactor Projects III/IV Office of Nuclear Reactor Regulation
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'P, Sholly Coordinator September 23, 1994 No significant hazards consideration comments received:
No.
Local Public Document Room location:
Phoenix Public Library, 12 East HcDowell Road,
- Phoenix, Arizona 85004 Original signed by:
Linh N. Tran, Project Hanager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation STRIBU ON:
Docket PDIV-2 Reading File BHolian LTran DFoster-Curseen Sholly Coordinator (Orig+1)
OGC (015B18)
DOCUHENT NAHE:
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PDIV-2 PH PDIV-2 PH PDIV-2 D
. NAME DFoster-Curseen LTran:
BHolian+
TOuay DATE
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Sholly Coordinator September 21, 1994 No significant hazards consideration comments received:
No.
Local Public Document Room location:
Phoenix Public L'ibrary, 12 East HcDowell Road,
- Phoenix, Arizona 85004 Original signed by:
Linh N. Tran, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation DISTRIBUTION:
Docket File PDIV-2 Reading File BHolian LTran DFoster-Curseen Sholly Coordinator (Orig+1)
OGC (015B18)
DOCUMENT NAME:
PV87244.BWI DRPW LA DFoster-Curseen PDIV-2 PH LTran:
PDIV-2 PH PDIV-2 0 BHoli an+
'Quay DATE Q/94 3 /
l/94 OFFICIAL RECORD COPY
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