ML20141B033
| ML20141B033 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/12/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20141B032 | List: |
| References | |
| NUDOCS 9706230247 | |
| Download: ML20141B033 (8) | |
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,j NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N05.129 AND 121 TO FACILITY OPERATING LICENSE NOS. OPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
By letter dated July 28. 1995, as revised Februa; _ 21, 1997, the Northern States Power Company (NSP or the licensee) requested amendments to the Technical 5)ecifications (TS) appended to Facility Operating License Nos.
DPR-42 and )PR-60 for the Prairie Island Nuclear Generating Plant (PI), Unit Nos. 1 and 2.
The proposed amendments would allow credit for soluble boron in spent fuel criticality analyses. The request is based on the Nuclear Regulatory Commission (NRC) ap3roval of the Westinghouse Owners Group generic methodology for crediting soluale boron given in Topical Report WCAP-14416-NP-A " Westinghouse Spent Fuel Rack Criticality Analysis Methodology," Revision 1 November 1996.
i 2.0 EVALUATION The PI spent fuel storage racks were analyzed using the Westinghouse methodology (Ref.1) which has been reviewed and a) proved by the NRC (Ref. 2).
This methodology takes partial credit for soluble acron in the fuel storage pool criticality analyses and requires conformance with the following NRC acceptance criteria for preventing criticality outside the reactor:
shall be less than 1.0 if fully flooded with unborated water, which k,,,ludes an allowance for uncertainties at a 95% probability, 95%
1) inc confidence (95/95) level as described in WCAP-14416-NP-A: and 2) k,,, shall be less than or equal to 0.95 if fully flooded with borated water, which includes an allowance for uncertainties at a 95/95 level as described in WCAP-14416-N' 4 The analysis of the reactivity effec'. of fuel storage in the PI spent fuel racks was performed with the three-dimensional Monte Carlo code, KENO-Va. with neutron cross sections generated with the NITAWL-II and XSDRNPM-S codes using the 227 group ENDF/B-V cross-section library.
Since the KENO-Va code package l
does not have burnup capability, depletion analyses and the determination of l
small reactivity increments due to manufacturing tolerances were made with the l
two-dimensional transport theory code PHOENIX-P, which uses a 42 energy group 9706230247 970612 PDR ADOCK 05000282 P
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.- nuclear data library. The analytical methods and models used in the reactivity analysis have been benchmarked against experimental data for fuel assemblies similar to those for which the PI racks are designed and have been found to adequately reproduce the critical values. This experimental data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include close proximity storage and strong neutron absorbers. The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the PI storage racks with a high degree of confidence.
The PI spent fuel storage rack design has previously been qualified for storage of various 14 x 14 fuel assembly types with maximum enrichments up to 5.0 weight percent (W/o) U-235. The maximum enrichment is based on a nominal value of 4.95 w/o U-235 alus a manufacturing tolerance of 0.05. The spent fuel rack Boraflex absor)er panels were considered in this previous analysis.
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')I spent fuel racks have been reanalyzed to allow storage of all 14 x 14 assemblies with nominal enrichments up to 4.95 w/o U-235 using credit for eckerboarding, burnup, burnable absorbers, and soluble boron.
For the nominal storage cell design. no credit was taken for the presence of Boraflex panels in the storage racks.
In addition. the moderator was assumed to be pure water at a temperature of 68 *F and a density of 1.0 gm/cc and the array was assumed to be infinite in lateral extent.
Uncertainties due to tolerances in fuel enrichment and density, storage cell inner diameter, storace cell pitch, stainless steel thickness, assembly position, calculational uncertainty, and method] logy bias mcertainty were accounted for. These uncertainties were approariately determined at the 95/95 probability / confidence level. A metaodology bias (determined from benchmark calculations) as well as a reactivity bias to account for the effect of the normal range of s)ent fuel pool water temperatures (50 *F to 150 *F) were included. These Jiases and uncertainties meet the previously stated NRC requirements and are, therefore, acceptable.
In order to determine the enrichment required to maintain k less than 1.0 y
with no credit for soluble boron or Boraflex, two configuraflons were analyzed. The first was all cell storage in which fuel assemblies with sufficiently low nominal enrichments 'were stored in every cell location. The second involved checkerboarding in which assemblies with nominal enrichments u) to 4.95 weight percent (w/o) U-235 were stored in the center of a 3x3 caeckerboard configuration. The criticality analyses considered all the fuel types currently stored in the spent fuel pool and in use at PI. These include the Optimized Fuel Assembly (OFA) and Vantage Plus designs currently being used, and the Westinghouse Standard (STD) and Exxon (currently Siemens Power Corporation) fuel types used in the past at PI and currently Mored in the spent fuel pool. The reactivity of the Westinghouse OFA design was found to be equivalent to the Westinghouse Vantage Plus fuel assemblies and the Westinghouse STD design was found to bound the reactivity of the Exxon fuel assemblies.
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l i The nominal k,,, for the all cell storage configuration was determined to be l
0.96914 for Westinghouse OFA fuel enriched to 1.87 w/o U-235 and 0.96799 for Westinghouse STD fuel enriched to 1.77 w/o U-235 with no credit for soluble boron or Boraflex. The 95/95 k,, was then determined by adding the temperature and methodology bia,ses and the statistical sum of independent tolerances and uncertdf aties to the nominal k,,d.99947 and 0.99893 values, as described in of for the OFA This resulted in a 95/95 k,,,hese values are less than 1.0 and Reference 1.
snd STO fuel types, respectively. Since t
. ere determined at a 95/95 probability / confidence level, they meet the NRC w
criteria for precluding criticalf cy and are acceptable.
The soluble ooron credit calculations assumed the all cell storage configuration moderated by water borated to 200 p)m. As previously described.
the individual tolerances and uncertainties and tie temperature and methodology biases were added to the calculated nominal k,,, to obtain a 95/95 value. 'ne resulting 95/95 k was 0.93505 and 0.94070 for OFA and STD fuel assembly types, respectively.,,,Since k is less than 0.95 with 200 ppm of boron and uncertainties at a 95/95 prob,a,bility/ confidence level, the NRC acceptance criterion for precluding criticality is satisfied. There' ore, storage of fuel assemblies with nominal enrichments up to 1.87 w/o U-235 and 1.77 w/o U-235 is acceptable for Westinghouse OFA or STD fuel assembly types, respectively, in all cells of the PI spent fuel storage racks with credit for the presence of 200 ppm boron in the water. This is well below the minimum spent fuel pool boron concentration value of 1800 ppm required by TS 3.8.E.2 and is, therefore, acceptable.
The concept of reactivity equivalencing due to fuel burnup was used to achieve the storage of 0FA assemblies with enrichments higher than 1.87 w/o U-235 and STD fuel assemblies with enrichments higher than 1.77 w/o U-235. The NRC has previously accepted the use of reactivity equivalencing predicated upon the reactivity decrease associated with fuel de)letion. This analysis also includes spent fuel decay time credit, whic1 results from the radioactive decay of isotopes in the spent fuel to daughter isotopes. The loss in reactivity due to the radioactive decay of the spent fuel results in re&ing the minimum burrap needed to meet the reactivity requirements.
In the decay time methodology, the fission )roduct isotopes are frozen at the concentrations existing at the time of disclarge from the core, except for Xe-135, which is removed. These calculations are performed at different discharge burnups. The fuel is depleted using a high soluble boron letdown curve to enhance the buildup of plutonium making the fuel more reactive in the spent fuel storage racks. Credit is taken only for the decay of actinides, one of the major contributors being the decay of Pu-241 to Am-241.
Calculations by Westinghouse from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown (at which time the ma;ior fission product Xe-135 has essentially decayed away) to 30 years following shutdown have shown that decay of the fission products has the effect of continuously reducing the reactivity of the spent fuel. However, no credit for fission p nduct decay is used in the decay time credit. Based on l
these conservative assumptions, the KRC concludes that the proposed use of decay tima credit is acceptable.
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_ To determine the amount of soluble boron required to maintain k.:,s0.95 for storage of fuel-assemblies with enrichments higher than 1.87 mo U-235 and 1.77 w/o U-235, a series of reactivity calculations were performed to gene ~ rate a set of enrichment versus fuel assembly discharge burnup (for different decay times) ordered pairs which all yield an equivalent k,,, when stored in the PI spent fuel storage racks. These are shown in TS Figures 3.8-1 and 3.8-2 and represent combinations of fuel enrichment and discharge burnup which yield the same rack k as the rack loaded with 1.87 w/o and 1.77 w/o fuel (at zero burnup)far,,bFAandSTDassembliesinallcelllocations. Uncertainties cssociated with burnup credit include a reactivity uncertainty of 0.01 Ak at 30,000 MWD /MTU applied linearly to the burnup credit requirement to account l
for calculational and depletion uncertainties and 4% on the calculated burnup to account for burnup measurement uncertainty. The NRC staff concludes that l
these uncertainties are acce) table. The amount of additional soluble boron, above the 200 ppm required a)ove, that is needed to account for these uncertainties is 200 ppm.and 250 ppm for the OFA and STD fuel types, respectively. This results in a total soluble boron credit of 400 ppm and 450 i
ppm for 0FA and STD fuel, respectively. This is well below the minimum spent i
fuel pool boron concentration value of 1800 ppm required by TS 3.8.E.2 and is, therefore, acceptable.
The nominal k,0FA fuel assembly (current PI fuel type) at the centerfor the 3x3 4.95w/oU-235 l
l surrounded by 0FA fuel enriched to 1.30 w/o U-235 was determined to be 0.96157, and 0.95918, if surrounded by STD fuel enriched to 1.20 w/o U-235 with no credit for soluble boron or Boraflex.
The 95/95 k,, was then l
determined by adding the temperature and methodology biase,s and the l
statistical sum of independent tolerances and uncertainties to the nominal k,,,
i values as described in Ref.1. This resulted in a 95/95 t of 0.99983 and 0.99944 forafresh0FAassemblysurroundedbythe0FAandNDfueltypes, respectively. Since these values are less than 1.0 and were determined at a 95/95 probability / confidence level, they are acceptable. These results indicate that the PI spent fuel racks will remain subtritical when cells are loaded in a 3x3 checkerboard configuration with a 4.95 w/o U-235 0FA fuel assembly at the center surrounded by any combination of 1.30 w/o U-235 0FA assemblies or 1.20 w/o U-235 STD assemblies.
The soluble boron credit calculations for the 3x3 checkerboard storage configuration assumed the assemblies were moderated by water borated to 250 ppm for surrounding 0FA fuel and 300 ppm for surrounding STD fuel. As previously described, the individual tolerances and uncertainties and the temperature and methodology biases were added to the calculated nominal k,,, to obtain a 95/95 value. The resulting 95/95 k,,, was 0.94134 and 0.93466 for surrounding 0FA and STD fuel assembly types, respectively. Since k,,, is less than 0.95 with credit for boron and uncertainties at a 95/95 probability / confidence level, the NRC acceptance criterion for precluding criticality is satisfied. Therefore, storage of fresh 0FA fuel assemblies with nominal enrichments up to 4.95 w/o U-235 in the center of a 3x3 checkerboard surrounded by 1.30 w/o U-235 0FA or 1.20 w/o U-235 STD fue assemblies is acceptable in the fuel storage racks with credit for the presence of 250 ppm and 300 ppm boron in the s tor, respectively.
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. Storage of surrounding fuel assemblies with enrichments higher than 1.30 w/o U-235 and 1.20 w/o U-235 for the OFA and STD fuel types in the 3x3 checkerboard configuration is achievable by means of reactivity equivalencing.
as described above. TS Figures 5.6-3 and 5.6-4 show the constant k,, contours
'as a function of assembly average burnup, for different decay times,, generated for the 3x3 checkerboard storage configuration. The amount of additional boron needed to account for the uncertainties associated with burnup credit was determined to be 350 ppm for surrounding 0FA assemblies and 450 ppm for
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surrounding STD assemblies. This results in a total soluble boron credit of 600 ppm (0FA) and 750 ppm (STD).
l The reactivity decrease associated with the presence of gadolinium (GAD) burnable absorbers imbedded in the UO fuel pellet was also analyzed to 2
determine the allowable storage of fuel assemblies with enrichments higher than 1.30 w/o U-235 and 1.20 w/o U-235 for the surrounding 0FA and STD fuel 4
ty>es. in the 3x3 checkerboard configuration. The credit for the presence of e
GA) is based on matching the reactivity of these assemblies to an equivalent enrichment of fresh assemblies without GAD. The assemblies with equivalent enrichment are put in a 3x3 checkerboard configuration and the enrichment for the assemblies surrounding the center location (fresh 4.95 w/o U-235 0FA assembly with varying number of GAD rods) is determined so that the new 3x3 i
j limit. TS Figures 5.6-5 through checkerboard configuration meets the 0.95k,, fuel enrichment and discharge 5.6-12 represent combinations of allowable burnup of the surrounding assemblies. The uncertainties associated with GAD
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credit include 3% for manufacturing and 10% fer calculational uncertainties.
These are acceptable based on current calcubtionC methods. The amount of additional soluble boron needed to accour.t for these uncertainties was determined to be 150 ppm for the OFA fuel assembly type since GAD is only in the center assembly location which is an 0FA assembly. This results in a soluble boron credit of 400 ppm and 450 ppm for the surrounding 0FA and STD fuel assembly types, respectively, and a maximum total credit of 750 ppm based on STD fuel.
Although most accidents will not result in a reactivity increase, two accidents can be postulated for each storage configuration which would increase reactivity beyond the analyzed conditions. The first would be a loss I
of fuel pool cooling system and a rise in pool water temperature from 150 F to 240 F.
The second would be a misload of an assembly into a cell for which the restrictions on location, enrichment, burnup, decay time, or GAD credit are not satisfied.
Calculations have shown that the misload assembly accident for a 3x3 checkerboard configuration in which a fresh 4.95 w/o U-235 0FA fuel assembly is olaced into an incorrect cell results in the highest reactivity increase. T1e reactivity increase is 0.05891 Ak, which is equivalent to an additional 550 ppm of soluble boron.
However, for such events, the double contingency principle can be applied. This states that the assumption of two unlikely. independent concurrent events is not required to ensure protection against a criticality accident. Therefore, the minimum amount of boron required by TS 3.8.E.2.a (1800 ppm) is more than sufficient to cover any accident and the presence of the additional boron above the concentration required for normal conditions and reactivity equivalencing (750 ppm maximum) i e
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can be assumed as a realistic initial ccndition since not assuming its presence would be a second unlikely event.
The NRC Safety Evaluation (SE) for crediting soluble boron (Ref. 2) states that potential events that could dilute the spent fuel pool soluble boron to i
the concentration required to maintain the 0.95 k,,,f these dilution events limit should be identified.
In addition, the available time span o should be cuantified to show that sufficient time is available to enable adequate detection and suppression of any dilution event.
Deterministic dilution event calculations were performed for P1 in order to define the dilution times and volumes necessary to dilute the spent fuel pool from the minimum TS boron concentration of 1800 ppm to a soluble boron concentration where a k,,, of 0.95 would be approached (750 ppm). The various initiating events considered included dilution from chemical and volume control system (CVCS) holdup tanks, CVCS monitor tanks, reactor water makeup tanks, CVCS blender, demineralized water system component cooling water, aerated water, resin flush / fill system, fire protection system, and the reverse osmosis system, and other events that may affect the boron concentration of the pool, such as seismic events, random pipe breaks. loss of offsite power, and the effects of the spent fuel pool demineralizer.
Both the small and the large PI spent fuel pools were considered.
An evaluation of these sources has shown that the only credible scenarios involve the transfer of unborated water from the reactor water makeup system to the spent fuel pool cooling or cleanup systems. The reactor water makeup system is capable of supplying the approximately 345,000 gallons of water l
necessary to dilute the pool from 1800 ppm to 750 ppm at a rate of approximately 80 gam if the inservice reactor water makeup tank is repeatedly replenished from tie water treatment system.
However, the worst scenario would require continued manual actions on the part of plant personnel for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for tN. dilution to occur. Therefore, a dilution event large enough to result in a significant reduction in the spent fuel pool boron concentration would involve such a large water volume turnover and would occur l
over such a long time period that it should be readily detected via level alarms and/or visual inspections and terminated by plant personnel before a dilution sufficient to approach the 0.95 k,,, limit could occur. The weekly l
spent fuel pool boron concentration sampling requirement specified in TS Table l
4.1-2B will provide assurance that smaller and less readily identifiable boron concentration reductions are not taking place.
Additionally, the criticality analysis for the PI spent fuel racks also showed that k,/,the pool were completely filled with unborated water.would remain less tha even 1 Thus, even if the pool were diluted to zero ppm, which would take significantly more water than evaluated above, the racks would be expected to remain subtritical.
l The TS changes proposed as a result of the revised criticality analysis are consistent with the changes stated in the NRC SE for WCAP-14416 (Ref. 2).
Westinghouse submitted a revised Topical Report WCAP-14416-NP-A, which incorporated the changes stated in the NRC SE (Ref 2). Also, since the staff
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disagreed with the proprietar finding of the original WCAP-14416.
--Westinghouse's revised Topicak Report was submitted as a nonproprietary version. Based on this consistency with the approved methodology and on the above evaluation, the~ staff finds these TS changes acce) table.
The proposed associated Bases changes adequately describe these TS clanges and are also acceptable.
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2.1 Su mary Based on the review described above the staff finds the criticality aspects of the proposed PI license amendment request are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling. The analysis assumed credit for soluble boron, as allowed by WCAP-14416-NP-A. but no credit for the Boraflex neutron absorber panels.
The following storage configurations and U-235 enrichment limits were determined to be acceptable:
1)
Westinghouse 14x14 0FA fuel assemblies with nominal enrichments no greater than 1.87 w/o U-235 and Westinghouse 14x14 STD and Exxon 14x14 fuel assemblies with nominal enrichments no greater than 1.77 w/o U-235 can be stored in any cell location.
Fuel assemblies with initial nominal enrichments greater than these must satisfy a minimum burnup and decay time requirement as shown in TS Figures 3.8-1 and 3.8-2.
2)
Westinghouse 14x14 0FA assemblies with nominal enrichments no greater i
than 4.95 w/o U-235 can be stored in the center of a 3x3 checkerboard configuration. The surrounding fuel assemblies must have an initial nominal enrichment no greater than 1.30 w/o U-235 for Westinghouse 14x14 l
OFA fuel assemblies and 1.20 w/o U-235 for Westinghouse 14x14 STD and l
Exxon 14x14 fuel assemblies.
Fuel assemblies with initial nominal enrichments greater than these must satisfy a minimum burnu) and decay time requirement as shown in TS Figures 5.6-3 and 5.6-4.
T1e enrichment I
limits of the surrounding fuel assemblies may be increased by crediting GAD burnable absorber in the center assembly as shown in TS Figures 5.6-4 through 5.6-12.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Minnesota State off.ial was notified of the proposed issuance of the amendments. The State off. 1al had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21. 51.32 and 51.35, an environmental assessment and finding of no significant impact has been prepared and published in the federal Register on June 11. 1997 (62 FR 31852).
8-l Accordingly, based upon the environmental assessment. the Commission has
-determined that the proposed action will not have a significant effect on the quality of the human environment.
5.0 CONCLUSIM 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 amendments will not be inimical to the comon l
defense and security or to the health and safety of the public.
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6.0 REFERENCES
l 1.
Newmyer. W. D., Westinghouse Electric Corp.. " Westinghouse Spent Fuel i
Rack Criticality Analysis Methodology " WCAP-14416-NP-A. Rev.1.
November 1996.
i 2.
Collins. T. E.. NRC letter to T. Greene. Westinghouse Owners Group.
l Acceptance for Referencing of Licensing Topical Report WCAP-14416-P.
" Westinghouse Spent Fuel Rack Criticality Analysis Methodology."
l (TAC Ni ri93254). October 25. 1996.
Principal Contributor: L. Kopp Date: June 12, 1997 1
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