ML20246N970
| ML20246N970 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/28/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246N966 | List: |
| References | |
| NUDOCS 8909110010 | |
| Download: ML20246N970 (9) | |
Text
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" *4 o UNITED STATES d
NUCLEAR REGULATORY COMMISSION n
WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
RELATED TO AMENDMENT N05. 90 AND '83 TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET N05. 50-282 AND 50-306
1.0 INTRODUCTION
By letter cated April _6, 1989 (Ref. 1), Northern States Power Company (NSP),
the licensee, requested a change to Facility Operating License Nos. DPR-42 and DPR-60 which would change Specifications 5.6.A and 5.3.A.2 and add a new Specification 3.8.E and associated Bases to the Technical Specifications for-the Prairie Island Nuclear Generating Plant, Units 1 and 2 (PING).
The proposed changes would permit the reload of fuel assemblies with enrichments up to 4.25 weight percent (w/o) Uranium-235 and the storage of such assemblies prior to and subseauent to loading in the PING reactors.
The increase in fuel enrichment is needed to support longer fuel cycles for the two plants.
2.0 EVALUATION 2.1 Analytical Methods The analyt'ical methods used in the criticality analysis of the PING fuel storage racks use the AMPX (Ref. 2) system of codes for neutron cross section generation and the. KENO-IV (Ref. 3) Monte Carlo computer code for reactivity determination.
The 227 energy group neutron cross section library used in the AMPX system of codes is based on ENDF/B-V (Ref. 4).
For processing the 227 group neutron cross section library to obtain multigroup cross sections for evaluating criticality experiments and the PING fuel storage racks, the NITAWL code (Ref. 2) is used to provide the self-shielded resonance cross sections.
NITAWL uses the Nordheim Integral Treatment for the resonances.
The XSDRNPM weightingofcrosssection9.codeisusedtoperformtheenergyandspatial (Ref. 2) one-dimensional 5 The multigroup neutron cross sections generated for a particular configuration are then input to the KEND-IV (Ref. 3) Monte Carlo code to evaluate the criticality of the critical experiments and PING fuel storage racks.
These methods were benchmarked by Westinghouse, the vendor performing the analysis of the fuel storage racks, by analyzing 33 critical experiments.
These experiments covered water moderated, uranium-oxide fuel arrays separated by various materials that simulate light water reactor (LWR) fuel shipping and 8909110010 890828 PDR ADOCK 05000282 P
FDC
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storage conditions _(Ref. 5) to critical experiments using highly enriched uranium metal cylindrical arrays with various interspersed materials (Ref. 6).
The results of the analysis of these critical experiments are:
(1) the average calculated effective multiplication factor (k of the critical experiments is 0.992, (2) the standard deviation of the b$)value is 0.0008 delta-k, and (3) the 95 percent probability with a 95 percent confidence level (95/95 probability / confidence level) uncertainty in reactivity of the analytical methods is 0.0018 delta-k.~
- The reactivity equivalencing performed to determine the fuel assembly discharge burnup as a function of initial fuel enrichment and poison gap size was performed with the PHOENIX (Ref. 8) and CINDER (Ref. 9) computer codes.
A 25 energy' group nuclear data library based on the British WIMS (Ref. 10) code is used with.
PHOENIX.
The PHOENIX code is a depletable, two-dimensional, multigroup, discrete ordinate, transport theory code.
The CINDER coc'e is a point-depletion computer code to determine fission product activities.
Using these two codes,.
Westinghouse determined that, as a function ci discharge burnup, fuel assemblies had their maximum reactivity at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after discharge from a reactor. This time of maximum reactivity occurs at the point in time when the major fission product poison Xenon-135 has almost completely decayed away.
Therefore, toe most reactive point in time ~ for a fuel assembly discharged from the reactor is conservatively approximated by Westinghouse by removing the Xenon-135 from the calculations.
The PHOENIX code was validated by Westinghouse by analyzing extensive benchmark critical experiments and by analyzing the isotopic composition-that has been measured for fuel discharged from a reactor (Ref. 11).
The data that has been analyzed indicate good agreement with measurements for both the critical experiments and the isotopic data. Westinghouse includes an additional bias of 500 MWD /MTU to account for uncertainty associated with burnup dependent reactivities.
We conclude that the use of the AMPX system of codes and the KENO-IV Monte Carlo code for the PING fuel storage racks is acceptable because the results obtained for the critical experiments are satisfactory for these codes that are widely used by the industry for fuel storage rack analyses.
In addition, we conclude that the use of the PH0ENIX and CINDER codes is acceptable because the results obtained for the critical experiments and isotopic compositions are satisfactory for these codes.
. 2.2 Spent Fuel Racks The PING spent fuel storage racks have a nominal 9.5 inch center-to-center spacing.
Boraflex strips are encased in each wall of a storage cell.
The Boraflex contains the neutron absorber Boron-10 at a loading of 0.040 grams per square centimeter.
The spent fuel racks can provide storage for 1582 fuel assemblies.
Total storage is limited to 1386 fuel assemblies to allow insertion of a shipping cask in the spent fuel pool.
The PING spent fuel racks are currently licensed to store fuel assemblies which do not exceed 39 grams of Uranium-235 per axial centimeter of fuel assembly.
This corresponds to an enrichment of about 3.9 w/o Uranium-235 for the 14x14 fuel assemblies used at PING.
The present submittal addresses the following issues; (1) the maximum enrichment fresh fuel assembly that can be stored in the spent fuel racks;
3-4 (2) the maximum enrichment fuel assembly that can be stored in the spent fuel racks with poison gaps (gaps in the Boraflex) with credit for fuel burnup; and (3) the maximum enrichment fuel assembly that can be stored in a modified checkerboard loading using 3-out-of-4 storage locations with poison gaps in the Boraflex.
2.2.1 Maximum Enrichment Fresh Fuel Assembly for Spent Fuel Racks Analyses were performed to determine the maximum enrichment fresh fuel assembly that could be stored in the spent fuel racks. The analyses were performed for WestinghouseOFAfuelassemblieswhichgivealargerk[f than does the Westinghouse STD fuel assembly when both fuel assembli8 have the same Uranium-235 enrichment.
Other assumptions that were made are:
1.
All fuel rods contain uranium dioxide fuel with the same Uranium-235 enrichment over the entire ler.gth of each fuel rod.
2.
No credit is taken for the Uranium-234 or Uranium-236 in the fuel or for burnup.
3.
The spent fuel pool water is at a temperature of 68 F with a conservative value of 1.0 gm/cc for the water density.
4.
No credit is taken for spacer grids or sleeves.
5.
The storage cells from an infinite array in the lateral dimensions.
The axial dimension is taken to be finite.
6.
A minimum poison material loading of 0.040 grams of Boron-10 per square centimeter is used for the Boraflex.
7.
No credit is assumed for the soluble boron in the pool water.
Westinghouse considered the worst case location of a fuel assembly in the, pent fuel racks and the worst case dimensions of the storage racks based on construction tolerances.
Based on this worst case analysis, Westinghouse determined that an enrichment of 4.07 w/o Uranium-235 results in a k of l
0.9331.
To this worst case k biases are added to account for the Mases in e
thecalculationalmethodand$Nsonparticleself-shielding.
These biases add 0.0093 delta-k to the worst case k In addition, uncertainties at the 95/95 probability /confidencelevelarea8b[dtok to account for the uncertainty in the method and in the KEND-IV Monte Car 18 uncertainty on the worst case f
k These uncertainties add 0.0053 delta-k to the worst case k'If.
The c8bfe.cted keff, including biases and uncertainties, is 0.9477.
We conclude that the unrestricted storage of fresh fuel assemblies (0FA and STD)
I enriched to 4.07 w/o Uranium-235, if no poison gaps exists in the spent fuel racks, is acceptable because the k including biases and uncertainties, of the spent fuel racks is less than N, staff criterion of 0.95 and because suitably conservative analysis assumptions have been made.
i l
Westinghouse analyzed accidents that could increase the reactivity of the spent fuel racks.
For accident conditions, the staff position allows credit for the soluble boron in the spent fuel pool water.
A heavy load dropped on top of the 1
spent fuel racks was postulated.
The methodology used by Westinghouse is the same as was previously used by Quadrex (Ref. 7).
The same assumptions that were used to establish the 4.07 w/o fresh fuel assambly enrichment limit were used to analyze the heavy load drop except that the model is infinite in all directions.
Calculations were performed with and without soluble boron as a function of water to uranium dioxide volume ratio.
Additional uncertainties of 0.005 delta-k are added for the no boron result and 0.010 delta-k to the boron result for conservatism.
The results show that the maximum spent fuel rack reactivity will be less than 0.90, including uncertainties at a 95/95 probability / confidence level, with 1800 ppm of baron in the spent fuel pool water.
We conclude that a heavy load drop on the spent fuel storage racks loaded with fresh fuel assemblies enriched to 4.07 w/o Uranium-235 is acceptable because k
is less than the staff criterion of 0.95 when credit is taken for 1800 ppm f
of[olubleboroninthespentfuelpoolwater.
2.2.2 Maximum Enrichment Fuel Assemblies with Poison Saps and Credit for Discharge Burnup Analyses were performed to determine the maximum reactivity effect of gaps in the Boraflex.
These analyses were also for the purpose of bounding expected shrinkage of the Boraflex, according to discussions with the licensee.
Westinghouse determined that some of the enrichment and poison gap sizes exceeded the criterion on k of being less than or equal to 0.95.
The k li (1) by taking credit for fuel b8fbup;mit can bemetinanyone8ffthree ways:
(2) by taking credit for soluble boron in the pool water for fuel assembly misloadings; and (3) by a checkerboard loading of the fuel assemblies.
The PH0ENIX (Ref. 8) and CINDER (Ref. 9) codes were used to determine the amount of fuel burnup required for a given fuel assembly enrichment and maximum assumed.
poison gap size using the method of reactivity equivalency.
In this method reactivity calculations are performed to generate a set of fuel assembly enrichment and associated fuel assembly discharge burnup for an assumed poison gap size which will yield k equal to or less than 0.95, including all uncertaintiesatthe95/95$fbbability/confidencelevel.
Curves of fuel assembly discharge burnup as a function of initial fuel assembly enrichment in weight percent Uranium-235 for poison gap sizes of 0, 2 and 4 inches which meet the spent fuel storage rack criterior were generated.
The results of the analysis show, as an example, that a fuel assembly, with an initial enrichment of 4.27 w/o Uranium-235 and with a 4 inch gap in the boraflex poison centered at the fuel assembly midplane throughout every spent fuel storage location, would meet the spent fuel rack k criterion if the discharge burnup was at least 4,272 eff MWD /MTU.
for fresh Thus,usingtheAMPX/ KENO-IVcodetodeterminethestoragerackk'fbthe fuel and the PHOENIX / CINDER codes to provide the discharge burnup reactivity equivalency analysis, the licensee has determined the storage requirements in the spent fuel pool when credit for discharge burnup is taken.
In the analyses the k of 0.95 includes all applicable biases and uncertainties atthe95/95probabilk/confidencelevel.
The analyses include the same assumptions as discussed in the previous section except that the axial gaps in the poison panels are positioned in the axial center of the active fuel in all poison panels in the spent fuel racks.
4
- We conclude that the storage of 0FA and STD fuel assemblies in the Prairie Island spent fuel pool is acceptable for the combination of initial fuel enrichment, fuel assembly discharge burnup, and potential poison gap size discussed in the Westinghousereportbecausethecriterienofk'hfobability/confidencelevel,is being equal to or less than 0.95, including all uncertainties at the 95/95 met.
2.2.3 Modified Checkerboard Loaded Spent Fuel Racks Westinghouse performed analyses to show that fresh fuel assemblies with an enrichment of 4.27 w/o Uranium-235 and 4 inch gaps in the poison panels could be stored in a modified loading of the spent fuel racks such that 3-out-of-4 storage locations were occupied by fresh fuel and the fourth location was empty.
Except for enrichment, loading pattern and poison gaps the assumptions are the same as those discussed in Section 2.2.1.
The axial gaps (4 inches) in the poison panels are positioned in the axial center of the active fuel in all poison panels in the spent fuel racks.
Westinghouse considered the worst case location of a fuel assembly in the spent fuel racks and the worst case dimensions of the storage racks based on construction tolerances.
Based on this worst case analysis, Westinghouse determined that k was equal to 0.9016.
Biases are added to this worst case s$$ftoaccountf8bf k
biases in the calculational method and poison particle shielding. These biases add 0.0093 delta-k to the worst case k Finally,uncertaintiesatthe95/95 probability /confidencelevelare*Ndedto k
to account for uncertainty in the method and in the KENO-IV Monte Carlo u8Nrtaintyontheworstcasek These uncertainties add 0.0049 delta-k to the worst case k,ff.
Thecorre8Ndkeff, including all biases and uncertainties is 0.9158.
We conclude that the storage of fresh fuel assemblies (OFA and STD) enriched to 4.27 w/o Uranium-235 in a modified checkerboard array using only 3 of every 4 locations is acceptable because the k including biases and uncertainties, of this spent fuel rack configuration il fe,ss than the staff criterion of 0.95 f
and because suitably ccaservative analysis assumptions have been made.
2.2.4 Fuel Misloading and Credit for Soluble Boron The modified checkerboard loading pattern without physically blocked unused locations and the taking (
redit for a fuel assembly's discharge burnup means that fuel misloading errors re now possible for the PING spent fuel racks.
The fuel misloading event is not discussed as such in the licensee's submittal.
The limiting case is the misloading of fresh fuel assemblies with an enrichment of 4.27 w/o Uranium-235 in the racks if 4 inch axially centered poison gaps exist in every racks' poison panels.
In such a misloading case credit may be taken i
for the soluble boron in the pool water.
The licensee calculated that 250 ppm of soluble boron in the water, including a 10 percent conservatism in the equal to 0.950.
The limiting solubleboronconcentration,wouldyieldak'Mennocreditistakenforthe case also shows that k i
solubleboroninthew$Nr.sequalto0.9810 Thus, the spent fuel pool racks would remain l
subcritical even without taking credit for soluble boron in the water in the event of the limiting case misloading event.
The PING Technical Specifications will require a minimum soluble boron content of 500 ppm.
A Surveillance
Requirement was added to the Technical Specifications.
Based on these requirements, we conclude that administrative procedures are acceptable for the storage of fuel assemblies at PING in a 3 of every 4 configuration or for the storage of fuel assemblies with discharge burnups equal to or greater than 5,000 MWD /MTU.
In addition, we conclude that credit may be taken for the soluble boron in the water for spent fuel rack misloading events.
2.3 New Fuel Racks The licensee has provided an analysis of the criticality of the new fuel storage racks for both the fully flooded and low hydrogenous moderation conditions.
The analyses used the same calculational methods and neutron cross section libraries that were used in the analyses for the spent fuel racks.
The analyses were based on the following assumptions:
1.
The fuel assembly containing the highest enrichment authorizeo, is at its most reactive point in life, and no credit is taken for any burnable poison in the fuel rods.
2.
All fuel rods contain uranium dioxide at an enrichment of 4.27 w/o Uranium-235 over the entire length of each rod.
3.
No credit is taken for any Uranium-234 or Uranium-236 in the fuel, nor is any credit taken for the buildup of fission products.
4.
No credit is taken for any spacer grids or spacer sleeves.
For the fully flooded condition the water is taken to be at 68 F and at a density of 1.0 gm/cc.
The analysis was performed for Westinghouse OFA fuel assemblies in an array that is infinite in both the lateral and axial dimensions.
Westinghouse considered the worst case location of a fuel assembly in the new fuel racks and the worst case dimensions of the racks based on construction tolerances.
Based on this worst case analysis, Westinghouse determined that an enrichment of 4.27 w/o Uranium-235 results in a k of 0.8736.
To this worst case k a bias of 0.0083 delta-k is added to ac8$unt for the bias in the calcul![$onalmethod.
In addition uncertainties at the 95/95 probability /
to account for the uncertainty in the method confidencelevelareaddedtok(yontheworstcasekThe correefdb These uncertainties andintheMonteCarlouncertai8 k($ffcriterionof, including biases add 0.0076 delta-k to the worst case k and uncertainties, is 0.8895.
This k is well below the e
0.95 for the fully flooded case.
A similar analysis was performed for the case of low density hydrogenous moderation.
For this case it was determined that a Westinghouse STO fuel assembly was more reactive than an 0FA fuel assembly. Westinghouse determined that a full 8x11 array of fuel assemblies would exceed the staff criterion of k
less than or equal to 0.98.
Westinghouse determined that a 5x11 array of f8bfassembliesatthelowdensityofmaximumreactivityof0.075gm/ccwould meet the criterion.
The three deleted rows of fuel storage locations will be modified to prevent storage of fuel assemblies.
The calculated k was 0.9794 including the worst case k of 0.9634, a bias of 0.0083 8[$ta-k and anuncertaintyata95/95probabiN(y/confidencelevelof0.0077 delta-k.
4
- We conclude that the storage of new STD and 0FA fuel assemblies having a maximum enrichment of 4.27 w/o Uranium-235 in no more than a 5x11 array of new fuel storage locations is acceptable because it will meet the staff criterion on k((oflessthanorequalto0.95forthefullyfloodedcaseandofkthan or equal to 0.98 for the low o
II 2.4 Technical Specifications The Technical Specifications for this proposed license amendment were reviewed and discussed with the licensee.
As a result of these discussions a number of wording changes were made to the Specifications.
All of these changes were agreed to with the licensee.
Based on our review, we found the following Specifications to be acceptable.
1.
Specification 3.8.E and Bases Statements This Specification provides limitations on the storage of a fuel assembly with a burnup of less than 5000 MWD /MTU.
The first limitation is that the concentration of boron in the pool shall be equal to or greater than 500 ppm.
The other limitation specifies that storage of such fuel shall be in a 3 of every 4 array with the fourth storage location empty.
A monthly Surveillance Requirement is added to Table 4.1-2b for the pool boron concentration. The Bases statements for this Specification are consistent with the requirements of the Specification.
This Specification and associated Bases are acceptable because they are based on the safety analysis for the spent fuel racks.
2.
Specification 5.3.A.2 This Specification has been changed to allow a maximum fuel assembly enrichment of 4.25 w/o Uranium-235 in the reactor core.
This change is acceptable because it is ansistent with the safety analysis for the new and spent fuel racks and becaur. cycle specific reload safety analyses will be performed for the maximum allowed enrichment fuel to be placed in the reactor core to ensure that all applicable safety criteria are met.
3.
Specification 5.3.A.1 l
This change deletes the reference to the total weight of uranium in the core.
lhis change is not related to the safety analysis for the new and spent fuel racks rather it is an editorial change.
This is acceptable because the cxact weight of uranium in the core has little significance to the safety of the i
reactor.
4.
Specification 5.6.A This Specification has been changed to allow a maximum enrichment of 4.25 w/o Uranium-235 for the new fuel storage pit and racks.
This change is acceptable because it is supported by the safety analysis.
4
+
g.
. 2.5 Design Basis Accident Analysis Relative to Fuel Burnup l
. We have evaluated the potential impact of. fuel burnup on the radiological basis
!)
accident for the Prairie Island Nuclear Generating Plant.
By letter dated December 28, 1983, the Commission issued Amendment Nos. 67 and.61 that evaluates off-site doses from potential radiological consequences for fuel burnup up to 55 GWD/MTU.
The evaluation concludes that the extended burnup will not result ~
in higher doses from those previously analyzed for postulated accidents nor will
. doses exceed the dose guidelines of 10 CFR 100.11.
This conclusion is still applicable for this amendment since the licensee does not intend to irradiated
- fuel assemblies above the 55 GWD/MTU limit.
- 3. O
SUMMARY
Based on the review described above, the proposed Technical Specification modifications.are acceptable and fuel assemblies having initial enrichments up to.4.25 weight' percent Uranium-235 may be operated in the reactors and safely stored inithe new fuel storage pit if the requirements of the Technical Specifications are met.
The Surveillance Requirement for Specification 3.8.E requires that the spent fuel pool boron concentration be verified monthly.
In addition,-the new fuel storage racks will be modified to preclude the storage of more than an array of 5x11 new fuel assemblies by deleting a 3x11 array of storage cells.
4.0 ~ ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51.35, and environmental assessment and finding.of no significant impact have been prepared and published in the Federal. Register on. August 28, 1989 (54 FR 35543)
Accordingly, based upon the environmental assessment, we have determined that the issuance of the amendment will not have a significant effect on the quality of the human
. environment.
5.0 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 not be conducted in compliance with the Commission's regulations and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
D. Fieno D. Dilanni Dated:
August 28, 1989 L _
,y 1
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6.0 REFERENCES
1.
Letter ~ from David Musolf (NSP) to Director of Nuclear Reactor Regulation, i.
dated April 6,.1989.
L 2.
N. M. Greene, "AMPX:
A Modular Code System for Generating Coupled Multigroup Neutron - Gamma Libraries from ENDF/B," 0RNL/TM-3706, March 1976.
3.
L. M. Petrie and N. F. Cross, " KENO-IV:
An Improved Monte Carlo Criticality Program," ORNL-4938, November 1975.
'4.
W. E. Ford III, "CSRL-V:
Processed ENFF/B-V 227-Neutron Group and Pointwise Cross Section Libraries for Criticality Safety, Reactor and Shielding Studies," ORNL/CSD/TM-160, June 1982.
5.
M. N. Baldwin, " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel," BAW-1484-7, July 1979.
5.
J. T. Thomas, " Critical Three-Dimensional Arrays of U(93.2) Metal Cylinders," Nuclear Science and Engineering, Volume 52, pages 350-3259, 1973.
7.
A. J. Elliott and K. Wong, " Licensing Report for Prairie Island Nuclear Generating Plant Units 1 and 2 Spent Fuel Cask Drop Evaluation,"
QUAD-1-83-017, October 1984.
8.
A. J. Harris, "A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors," WCAP-10106, June 1982.
9.
T. R.-England, "A One-Point Depletion and Fission Product Program,"
WAPD-TM-334, August 1962.
10.
Askew,-J. R., Fayers, F.
J., and Kernshell, P.
B., "A General Description of the Lattice Code WIMS," Journal of British Nuclear Energy Society, 5, pp. 564-584, 1966.
11.
J. B. Meleham, " Yankee Core Evaluation Program Final Report,"
WCAP-3017-6094, January 1971.
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