ML20212C099
| ML20212C099 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/31/1999 |
| From: | HOLTEC INTERNATIONAL |
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| References | |
| HI-992285, NUDOCS 9909210137 | |
| Download: ML20212C099 (39) | |
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'4 Hollec Center,555 Lincoln Drive West, Marlton, NJ 08053 HOLTEC INTERNATIONAL
- Dll8;%8;88 CRITICALITY SAFETY ANALYSIS OF THE CRYSTAL RIVER UNIT 3 POOL A FOR STORAGE OF 5% ENRICHED MARK B-11 FUEL IN CHECKERBOARD ARRANGEMENT WITH WATER HOLES FOR FLORIDA POWER CORPORA 10N Holtec Report No: HI-992285 Holtec Project No: 90825 Report Category: A Report Class: Safety Related August 1999 ES'*AN N S $o$o2
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TABLE OF CONTENTS
1.0 INTRODUCTION
AND
SUMMARY
. . . . . . . . l
. ........ . .. .. 1 l
2.0 ANALYTICAL BASES AND SPECIFICATIONS .. . . .. . . . 3
. 2.1 - Regulatory Requirements . . . . . . . . . .. .. ... . .. 3 2.2 Calculational Assumptions . .. ..... .. ... .. . .3 2.3 Fuel Assembly Specifications . . .. . .. .. .. 4 2.4 Storage Rack Design Specifications . ..... . ... .....4 3.0 ANALYTICAL METHODOLOGY , . . . ... 6 3.1 Computer Programs and Benchmarking . ... .. . ..6 4.0 RESULTS OF ANALYSIS . . . . . . . .. . . .. . .. 7 4.1 Manufacturing Tolerance Reactivity Effects . . . . . . .7 4.2 Reactivity Effect ofTemperature . . .. . .. .. 7
5.0 REFERENCES
. . .. .... ..... ... ...... ... . ... 8 LIST OF TABLES <
i i
Table 1
SUMMARY
OF CRITICALITY ANALYSIS FOR POOL' A '
CHECKERBOARD CONFIGURATION UNDER NORMAL AND ACCIDENT CONDITIONS . . .. . ... . ...... . 9 Table 2 FUEL CHARACTERISTICS USED IN CRITICALITY ANALYSES . . . . . , .. . .... . .. . .. . ..... 10 Appendix A Benchmark Calculations (Total 26 Pages)
Holtec Report HI-992285 Page i
A a
1.0 INTRODUCTION
AND
SUMMARY
The present analysis is an extension of two previous analyses (References 1 and 2),
intended to provide analyses of the checkerboar'd arrangement under normal and accident conditions. The arrangement considered is the storage location of fuel assemblies of the maximum permissible reactivity in a checkerboard pattea alternating with storage cells containing only water. The analyses were based on the Mark B-11 fuel which in the previous analysis
- was demonstrated to be the most reactive and to bound all other fuel types in use or proposed for use at Crystal River Unit 3. The accident scenario considered is the inadvertent placement of an extra assembly of the maximum reactivity into a cell intended to remain empty, creating a local cross-shaped arrangement of assemblies in a close packed array. Although credit for soluble boron normally in the pool water is permitted under accident conditions (References 5,6,7, and 10 CFR 50.68),
calculations were made, without soluble boron, for the second concurrent accident condition of a mis-placed fuel assembly of the highest reactivity. Results of the analysis confirm that the maximum reactivity for the postulated accident scenario with Mark B-11 fuel remains below the regulatory limit without requiring any soluble baron normally present. Thus, it is concluded that the Pool A storage racks can safely accept the most reactive fuel (5% enriched Mark B-11 fuel) well within the NRC guidelines and without credit for the soluble boron in the pool water. Funhermore, credit for the soluble boron would provide an additional very large sub-critical margin.
The various options for storage of spent fuel in Pool A from the three analyses may be summarized as follows:
Option 1 (Reference 1) used a generic assembly design intended to bound fuel then in storage (B&W Mark 9,10 and earlier designs). For this option, soluble boron credit is not needed for either normal or the accident condition of a mis-placed fresh 5%
assembly. Reference 1 also evaluated a checkerboard arrangement of fresh 5% fuel alternating with empty (water-filled) cells, obtaining a maximum kar of 0.833 including bias and uncertainties for the generic fuel design. The accident condition was not Holtec Report HI 992285 Page1
evaluated for the checkerboard case because of the very low reactivity observed for the normal checkerboard case. The checkerboard connguration has subsequently been re-evaluated as described in Option 3 below.
Option 2 (Reference 2) assumed partial credit for soluble boron (Reference 10 CFR 50.68) and based the calculations on the Mark B-ll fuel design at 5% initial enrichment (bounds the Mark B-10F and earlier designs). This analysis assumed a conservative 15%
loss in Boron of the absorber panel. Reference 2 documents that 5% enriched fuel can be safely stored in Pool A in a 3-out-of-4 pattern with the fourth cell containing spent fuel at a specined enrichment-burnup combination. For this case, 400 ppm soluble boron is required for normal storage conditions and 450 ppm under the accident condition of a mis-loaded fresh fuel assembly of 5% enrichment.
Ontion 3 is the 1-out-of-2 or checkerboard arrangement with fresh 5% enriched fuel alternating with cells containing only water. Reference 1 originally determined that the maximum kar for this condition with a generic 15x15 fuel design was 0.833 without any soluble boron. This case has been re-evaluated in the present study with the Mark B-11 fuel assembly without any soluble boron and conservatively assuming a 15% loss in boron from the absorber panels. Under normal storage conditions, the maximum kar is 0.8466, including bias and uncertainties. Under the postulated condition of a mis-loaded assembly of the highest permissible reactivity, the maximum kar, without any soluble boron, is 0.9208, including bias and uncertainties (see Table 1) .
These data confirm that all three options are acceptable for the specified conditions, with a maximum reactivity within the USNRC guidelines (0.95 maximum reactivity).
Holtec Report HI-992285 Page 2 i
I
4 2.0 ANALYTICAL BASES AND SPECIFICATIONS I
2.1 Regulatory Guidelines Applicable codes, standards and regulations (or sections thereof that may be pertinent)
~
include the following. i l
{
- General Design Criterion 62 - Prevention of Criticality in Fuel Storage and !
Handling.
- L.I. Kopp, " Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants", Memorandum to )
Timothy Collins. August 19,1998.
- NRC Letter of April 14,1978, to all Power Reactor Licensees - OT Position for Review and Ac::eptance of Spent Fuel Storage and Handling Applications.
- Regulatory Guide 1,13, Spent Fuel Storage Facility Design Basis (Proposed),
December 1981.
- USNRC Standard Review Plan, NUREG 0800, Sections 9.1.2, Spent Fuel Storage.
- ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.
. 10 CFR 50.68," Criticality Accident Requirements", February 1998.
2.2 Calculational Assumptions The evaluation of the storage capability of the Crystal River Unit 3 Spent Fuel Pool A is based on the following assumptions:
. The racks are assumed to contain the most reactive fuel authorized to be stored in the facility without any control rods or burnable poison.
. In Pool A, the boron carbide plates are conservatively assumed to be reduced by 15% in boron loading.
Holtec Repon HI-992285 Page
l Criticality safety analyses are based upon the assumption of an infinite radia!
I array of storage cells, i.e., no credit is taken for radial neutron leakege. In the
! axial direction, the finite length of the actual fuel assemblies is assumed.
1 1
Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.
. The analyses are intended to defime bounding conditions which assure that the
.acks remain suberitical in clean un-borated water including bias and uncertainties, evaluated for 95% probability at the 95% confid:nce level, i.e.,
the accident condition in which all soluble boron is assumed to be lost.
. The maximum calculated reactivity includes an allowance for calculational l and manufacturing uncertainties, statistically combined, such that the true kerr will be less than the limit with a 95% probability and a 95% confidence level.
l The total manufacturing uncertainty of 0.0097 Ak, previously evaluated for l Pool A in Reference 1, was assumed to be valid.
Temperatures used for the analyses shall be those within the operating range l corresponding to the highest kerr.
2.3 Fuel Assembly Specifications .
l l
The fuel assembly design used for the analyses is the 15x15 Mark B-11 assembly containing 208 UO2 fuel rods,16 guide tubes and 1 instrument thimble. Design l Specifications for the fuel assembly analyzed is shown in Table 1. This fuel assembly l was evaluated for a maximum enrichment of 5.0%. The Mark B-11 fuel is the most reactive in Pool A, and was demonstrated in Reierence 2 to bound all other fuel concepts.
2.4 Storage Rack Design Specifications !
l i l The spent fuel storage racks in Pool A consists of stainless steel storage boxes on a 10.5 l
inch lattice spacing (2), bounded on four sides with 0.048 inch thick plates containing I boron carbide (natural enriched boron) clad with 0.060 inch thick steel plates. The B-10 j 2
loading is nominally 0.015 i 0.003 gms/cm which provides a minimum loading of 0.012 l 2
gms/cm A reduction of 15% in nominal B-10 loading was assumed as a conservative allowance to encompass possible degradation and other uncertainties including particle Holtec Report HI-992285 Page 4 l
l L
r ,
self-shielding.' A typical storage cell is pictorialjy described in Reference 2. The rack i design provides a flux-trap wamr-gap between absorber plates. A minor adjustment in the water-gap thickness was necessary for internal consictency, reducing the water-gap to 1.227 inch in thickness.
I i
j l
l Hollec Repon Hb992285 Page 5 1
3.0 ANALYTICAL METIIODOLOGY I
3.1 Computer Programs and Benchmarking l
In the evaluation of the two supplemental cases, the criticality analyses were performed with the KENO 5a computer package [Ref. 4] using the 238-group SCALE' cross-section l library and the NITAWL subroutine for U-238 resonance shielding effects (Nordheim ;
integral treatment). Benchmark Calculations are summarized in Appendix A. l In the geometric model used in the calculations, each fuel rod and its cladding were described explicitly in the KENO 5a model. Reflecting (zero neutron current) or periodic boundary conditions were used in radial direction which has the effect of creating an infmite array of storage cells in X-Y directions. The actual fuel assembly length was used in the axial direction, assuming a thick (30 cm) water reflector.
Monte Carlo (KEN 05a) calculations inherently include a statistical uncertainty due to the 1 random nature of neutron tracking. To minimize the statistical uncertainty of the KEN 05-calculated reactivity, a minimum of 1 million neutron histories in 1000 i generations of 1000 neutrons each, were accumulated in each calculation.
l
- SCAL I, is an acronym for Standard Computer Analyses for leicensing Evaluation.
Iloltec Report HI-992285 Page 6
~
4.0 RESULTS OF ANALYSES 4.1 Manufacturing Tolerance Reactivity Effects Reactivity effects of manufacturing tolerances in Pool A were previously evaluated [Ref.
I and 2] and it is assumed that the total value, i 0.0097 Ak, remains valid. With bias and KEN 05a statistical uncertainties included the total v~ertainty becomes 0.0100 Ak.
4.2 Reactivity Effect of Temperature In Pool A, the temperature coefficient of reactivity is negative and the highest reactivity occurs at 68 F, as determined in Reference 2.
I i
Holtec Report HI-992285 Page 7
m- .
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5.0 REFERENCES
- 1. S.E. Turner " Criticality Safety Evaluation of the Pool A Spent Fuel Storage Racks in Crystal River Unit 3 with Fuel of 5.0% Enrichment", Holtec International, HI-931111, December 1993.
- 2. S.E. Turner " Criticality Safety Evaluation of the Crystal River Unit 3 Fuel Storage Facilities with Partial Soluble Boron Credit", Holtec Intemational, HI-971794, December 1997.
- 3. R.M. Westfall, et. al., "NITAW'L-S: Scale Modular System for Performing Resonance Shielding and Working Library Production" in SCALE: A Modular Code System for Performing Standardized Comouter Analyses for Licensing Evaluation, NUREG/CR-0200 Rev 4,1990.
L.M. Petrie and N.F. Landers, " KEN 05a: An Improved Monte Carlo Criticality Program with Supergrouping" in SCALE: A Modular Code System for Performing Standardized Comouter Analyses for Licensinn Evaluation, NUREG/CR-0200 Rev 4,1990.
- 4. M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
- 5. NRC Letter of April 14,1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications.
- 6. L.I. Kopp, " Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants", Memorandum to Timothy l Collins, August 19,1998.
- 7. Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis (Proposed),
December 1981.
Holtec Repon HI-992285 Page 8
4 Table 1
SUMMARY
OF CRITICALITY ANALYSIS FOR POOL A CHECKERBOARD CONFIGURATION UNDER NORMAL AND ACCIDENT CONDITIONS Misloaded Fresh Normal Fuel Assembly Fuel Assembly ' Mark B-11 Mark B-11 Loading Pattern Checkerboard Checkerboard Enrichment, fresh (unburned) 5% 5%
Temperature 20 C 20 C Soluble Boron None none KEN 05a Calc. K-eft 0.8336 0.9078 Bias uncertainty 0.0012 KENO statistical uncertainty 0.0014 Reference 2 0.0097 Total Uncertainty @ 95%/95% i0.0100 i 0.0100 '
Calculational Bias 0.0030 0.0030 Maximum k-eff 0.8466 0.9208 :
Regulatory Limit 0.9500 0.9500 Notes:
- 1. The total uncertainty is a statistical combination of the manufacturing and ,
calculational uncertainties.
- 2. In the mistoaded fresh assembly accident, a single 5% enriched fresh assembly was placed into a ce!! intended to remain empty, i
Table 2 1
l 11oltec Report 111-992285 Page 9
e ,
FUEL CHARACTERISTICS USED IN CRITICALITY ANALYSES Mark B-11 Rod Matrix 15x15 No. Fuel Rods 208 No. Guide Tubes 16 No. Inst. Tubes 1 Fuel Rod Pitch (in) 0.568 l Fuel Pellet O.D. (in) 0.3622 i 0.3700 I Fuel Clad I.D. (in)
Fuel Clad O.D. (in) 0.4140 ,
Guide Tube I.D. (in) 0.5000' Guide Tube O.D. (in) 0.5280
' Inst. Tube I.D. (in) 0.4430 Inst. Tube O.D. (in) 0.4910 UO2Density (gm/cc) 10.522 Max. U-235 Enrichment 5.0% !
Notes: I
- 1. These values were taken from Reference 2. l
\
I l
Holtec Repon HI-992285 Page 10
. I i
I Appendix A Benchmark Calculations
..(total number of pages: 26, including this page) l i
Holtec Report HI-992285 Page: A-0
APPENDIX A: BENCHMARK CALCULATIONS A.1 INTRODUCTION AND
SUMMARY
Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP [A.1] is a continuous energy Monte Carlo code and KEN 05a [A.2] uses group-dependent cross sections. For the KENO 5a analyses reported here, the 238-group library was chosen, processed through the NITAWL-II [A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errors' (trends) that have been reported (e.g., [A.3 through A.5]) for calculations with collapsed cross section sets.
In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the "B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Odier parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.
Table A.1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.
One possible way of representing the data is through a spectmm index that incorporates all of the variations in parameters. KENO 5a computes and prints the " energy of the average lethargy causing fission". In MCNP, by utilizing the tally option with the identical 238-group energy structure as in KEN 05a, the number of fissions in each group may be collected and the energy of the average lethargy causing fission determined (post-processing).
Figures A.1 and A.2 show the calculated k, for the benchmark critical experiments as a function of the " energy of the average lethargy causing fission" for MCNP and KENO 5a, respectively (UO2 fuel cf.y). The scatter in the data (even for comparatively minor variation Small but observable trends (errors) ha te been reported for calculations with the 27-group and 44-group collapsed librari:s, These errors are probably due to the use of a single collapsing spectrum when the spectruni should be different for the various cases analyzed, as evidenced by the spectrum indices.
?
Appendix A, Page 1 Holtec Report Hi-992285
4 2
in critical parameters) represents experimental error in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL eriticals.
Linear regression analysis of the data in Figures A.1 and A.2 show that there are no trends, i as evidenced by very low values of the correlation coefficient (0.13 for MCNP and 0.21 for KEN 05a). The total bias (systematic error, or mean of the deviation from a (, of exactly 1.000) for the two methods of analysis are shown in the table below.
Calculational Bias of MCNP and 13NO5a McNP 0.0009i0.0011 l KENO 5a 0.0030 i 0.0012 The bias and standard error of the bias were derived directly from the calculated (, values in Table A.1 using the following equations', with the standard error multiplied by the one-sided K-factor for 95% probability at the 95% confidence level from NBS Handbook 91
[A.18] (for the number of cases analyzed, the K-factor is -2.08 or slightly more than 2).
i=f k, (A.1) 2 A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.
3 These equations may be found in any standard text on elementary statistics, for example, reference [A.6](or the MCNP manual) and is the same methodology used in MCNP and in KEN 05a.
Appendix A, Page 2 Holtec Report Hi-992285
i ki k,' -( . k,)2 f, 0* =
n (n-1)
Blas = 1 - k
- K og (A.3) l where ki are the calculated reactivities of n critical experiments; o, is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); K is the one-sided multiplier for 95% probability at the 95% confidence level (NBS Handbook 91 i
[A.18]). l 1
The larger of the systematic errors (truncated bias) were used to evaluate the maximum k,n values for the rack designs. KENO 5a has a slightly larger systematic error than MCNP, but
' both result in greater precision than published data [A.3 through A.5] would indicate for collapsed cross section sets in KENO 5a (SCALE) calculations, i
)
A.2 Effect of Rnrichment i
The benchmark critical experiments include those with enrichments ranging from 2.46 w/o ;
to 5.74 w/o and therefore span the enrichment range for rack designs. Figures A.3 and A.4 i show the calculated k,y values (Table A.1) as a function of the fuel enrichment reported for I the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.026 for MCNP and 0.379 for KENO 5a). Thus, there are no corrections to the bias for the various enrichments.
As further confirmation of the absence of any trends with enriclunent, a typical configuration I was calculated with both MCNP and KENO 5a for various enrichments. The cross-comparison ;
of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41.
Results of this comparison, shown in Table A.2 and Figure A.5, confum no significant difference in the calculated values of k,y for the two independent codes as evidenced by the !
45' slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence l
of an enrichment effect (trend) in the bias.
1 Appendix A, Page 3 Holtec Report HI-992285 i -
i A.3 Effect of % Londino Several laboratories have performed critical experiments with a variety of thin absotter panels similar to the Boral panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment), the reactivity worth of the absorbers in the PNL tests is very low and any signiscant errors that might exist in the treatment of strong thin absorbers could not be revealed. 1 Table A.3 lists the subset of experiments using thin neutron absorbers (from Table A.1) and shows the reactivity worth (Ak) of the absorber.'
No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.
To further confirm the absence of a significant trend with i B concentration in the absorber, a cross-comparison was made with MCNP and KENO 5a (as suggested in Reg. Guide 3.41).
Results are shown in Figure A.6 and Table A.4 for a typical geometry. These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 45* line) within an expected 95% probability limit. ,
i A.4 Miscellaneous and Minar Parameters A.4.1 Reflector Material and Snneinen PNL has performed a number of critical experiments with thick steel and lead reflectors.5 Analysis of these critical experiments are listed in Table A.5 (subset of data in Table A.1). '
There appears to be a small tendency toward overprediction of k, at the lower spacing, although there are an insufficient number of data points in each series to allow a quantitative '
determination of any trends. The tendency toward overprediction at close spacing means that l
the rack calculations may be slightly more conservative than otherwise.
The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.
5' Parallel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not pertinent to the Holtec rack design. A lead reflector is also not directly pertinent, but might be used in future designs.
. lioltec Report HI-992285 Appendix A, Page 4
p .
A.4.2 Fuel Pellet Diameter and Lattice Pitch 4
The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.4% to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for BWR fuel. Thus, the critical experiments analyzed provide a reasonable representation of power reactor fuel. Based on the data in Table A.1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs.
A.4.3 Soluble Boron Concentratinn Effects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP (and one KENO 5a) calculations are shown in Table A.6. Analyses of the very high boron concentration cgeriments (> 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative.
A.5 MOX Fuel The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for UO2fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table A.7. Results of these analyses are generally above a k,, of 1.00, indicating that when Pu is present, both MCNP and KEN 05a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP. It may be noted that for the larger lattice spacings, the KENO 5a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO 5a. It is also possible that the overprediction in k,y for both codes may be due to a small m' adequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated k,y over a wide range of the spectral index (energy of the average lethargy causing fission)..
Appendix A, Page 5 Hollec Report HI-992285
4 A6 References
[A.1] J.F. Briesmeister, Ed., "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).
[A.2] SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NLREG-CSD-2/U2/RS, Revision 5, Oak Ridge National Laboratory, September 1995.
[A.3] M.D. DeHart and S.M. Bowman, " Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460)
Oak Ridge National Laboratory, September 1994.
[A.4] W.C. Jordan et al., " Validation of KENOV.a", CSD/TM-238, Martin Marietta Energy Systems, Inc., Oak Ridge National Labora:ory, December 1986.
[A.5) O.W. Hermann et al., " Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated.
[A.6] R.J. Larsen and M.L. Marx, An Introduction to Mathematical Statistics and its Applications, Prentice-Hall,1986.
[A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BA.W-1484-7, Babcock and Wilcox Company, July 1979.
[A.8] G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW- 1 1645-4, Babcock & Wilcox Company, November 1991. l
[A 9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development j and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company April 1984.
? Appendix A, Page 6 Holtec Report HI-992285 i
l
[A.10] J.C. Manaranche et al., " Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods," Trans. Am.
Nucl. Soc. 33: 362-364 (1979).
[A.11] S.R. Bierman and E.D. Clayton, Criticality Experiments with 2
Suberitical Clusters of 2.35 w/o and 4.31 w/o "U Enriched 2 UO Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.
[A.12] S.R. Bierman et al., Criticality Experiments with Suberitical Clusters 2
of 2.35 w/o and 4.31 w/o "U Enriched UO 2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December,1981.
[A.13] S.R. Bierman et al., Critical Separation Between Suberitical Clusters 2
of 4.31 w/o "U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.
[A.14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.
[A.15] B.M. Durst et al., Critical Experiments with 4.31 wt %2 "U Enriched UO2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.
[A.16] S.R. Bierman, Criticality Experiments with Fast Test React.or Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981.
[A.17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.
[A.18] M.G. Natrella, Exnerimental Statictics, National Bureau of Standards, Handbook 91, August 1963.
1 Appendix A, Page 7 Holtec Report Hi-992285 l
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1 Table A.2 7
COMPARISON OF MCNP AND KENO 5a CALCULATED REACTIVITIES ,
FOR VARIOUS ENRICHMENTS '
k,, i la Enrichment MCNP KENO 5a t
3.0 0.8465 0.0011 0.8478 i 0.0004 l 3.5 0.8820 i 0.0011 0.8841 i 0.0004 3.75 0.9019 i 0.0011 0.8987 0.0004 !
4.0 0.9132 0.0010 0.9140 0.0004 4.2 0.9276 i 0.0011 i 0.9237 i 0.0004 4.5 0.9400 0.0011 0.9388 0.0004 l
l
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7 Based on the GE 8x8R fuel assembly.
Holtec Report HI-992285 Appendix A, Page 14
- 1
. 1
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\
Table A.3 MCNP CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS Energy of Ak Average Ref.
Worth of Lethargy Experknent Absorber k, Calculated of Fission A.13 PN1-2615 Boral Sheet 0.0139 0.9994 i 0.0012 0.1165 A.7 B&W-1484 Core XX 0.0165 1.0008i0.00ll 0.1724 A.13 PNI-2615 1.62% Boron-stee! 0.0165 0.9996 i 0.0012 0.1161 A.7 B&W-1484 Core XIX 0.0202 0.9961 i 0.0012 0.2103 A.7 B&W-1484 Core XXI 0.0243 0.9994 i 0.0012 0.1544 A.7 B&W-1484 Core XVII 0.0519 0.9962 i 0.0012 0.2083 A.11 PNL-3602 Boral Sheet 0.0708 0.9941 i 0.0011 0.3135 A.7 B&W-1484 Core XV 0.0786 0.9910 i 0.0011 0.2092 A.7 B&W-1484 Core XVI 0.0845 0.9935 i 0.0010 0.1757 A.7 B&W-1484 Core XIV 0.1575 0.9953 i 0.0011 0.2022 A.7 B&W-1484 Core XIII 0.1738 1.0020 i 0.0011 0.1988 A.14 PNL-7167 Expt 214R flux trap 0.1931 0.9991 i 0.0011 0.3722 4,,
Hollec Report Hi-992285 Appendix A, Page 15 r
r i
1
~
l i
i Table A.4 COMPARISON OF MCNP AND KENO 5a CALCULATED REACTIVITIES' FOR VARIOUS ' B LOADINGS k, 10 2
B, g/cm2 MCNP KENO 5a 0.005 1.0381 i 0.0012 1.0340 0.0004 l 0.010 0.9960 i 0.0010 0.9941 i 0.0004 0.015 0.9727 0.0009 0.9713 0.0004 0.020 0.9541 0.0012 0.9560 0.0004 0.025 0.9433 0.0011 0.9428 i 0.0004 0.03 0.9325 0.0011 0.9338 i 0.0004 !
0.035 0.9234 i 0.0011 0.9251 i 0.0004 0.04 0.9173 0.0011 0.9179 0.0004 l
l l
l l
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Based on a 4.5% enriched GE 8x8R fbel assembly.
Holtec Report HI-992285 Appendix A, Page 16
e Table A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORS' Separation, Ref. Case E, wt% cm MCNP L KEN 05a b A.I1 Steel 2.35 1.321 0.9980 i 0.0009 Reflector 0.9992 i o.0006 2.35 2.616 0.9968 i 0.0009 0.9964 t0.0006 2.35 1 3.912 0.9974 i 0.0010 0.9980 i 0.0006 2.35 =
0.9962 i 0.0008 0.9939 i 0.0006 j
A.11 Steel 4.306 1.321 0.9997 i 0.0010 Reflector 1.0012 i 0.0007 4.306 2.616 0.9994 i 0.0012 0.9974 i 0.0007 4.306 3.405 0.9969 i 0.0011 0.9951 i 0.0007 4.306 =
0.9910 i 0.0020 0.9947 i 0.0007 i
A.12 lead 4.306 !
0.55 1.0025i0.00ll Reflector 0.9997 i 0.0007 4.306 1.956 1.0000 i 0.0012 0.5985 i 0.0007 4.306 5.405 0.9971 1 0.0012 0.9946 i 0.0007 l
Arranged in order ofincreasing reflector-fuel spacing.
Holtec Report HI-992285 A Ppendix A, Page 17
l I
i i
Table A.6 i
CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated k,
. Boron Concentration, Reference Experiment ppm MCNP KENO 5a A.9 B&W-1810 1337- 1.0023 i 0.0010 -
A.9 B&W-1810 1899 1.0060 i 0.0009 -
A.8 B&W-1645 886 0.9970 i 0.0010 0.9924 i 0.0006 A.15 PNL-4267 0 0.9974 i 0.0012 -
A.15 PNL-4267 2550 1.0057 i 0.0010 -
Holtec Report HI-992285' Appendix A, Page 18
Table A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP KENO Reference Case" '
Q EALP" h EALF" PNI-5803 MOX Fuel . Exp. No. 21 1.0041 1 0.0011 0.9171
[A.16] 1.0046io.0006 0.8868 MOX Fuel . Exp. No. 43 1.0058 i 0.0012 0.2968 1.0036 i 0.0006 0.2944 MOX Fuel . Exp. No.13 1.0083 i 0.0011 0.1655 0.9989 1 0.0006 0.1706 MOX Fuel - Exp. Na. 32 1.0079 1 0.0011 0.1139 0.9966 i 0.0006 0.1165 WCAP- Saxton @ 0.52" pitch 3385 0.9996 i 0.0011 0.8665 1.0005i0.0006 0.8417 lA.17] Saxton @ 0.56* pitch 1.0036i0.00ll 0.5289 1.0047 i 0.0006 0.5197 Saxton @ 0.56" pitch borated 1.0008 1 0.0010 0.6389 NC NC Saxton @ 0.79" pitch 1.0063 1 0.0011 0.1520 1.0133 1 0.0006 0.1555 Arranged in order ofincreasing lattice spacing.
EALF is the energy of the average lethargy causing fission.
Hollec Report HI-992285 Appendix A, Page 19
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FIGURE A.1 MCNP CALCULATED k-eff VALUES for VARIOUS VALUES OF THE SPECTRAL INDEX l
1 i
Holtec Report HI-992285 E" ^~
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4 FIGURE A.2 KEN 05a CALCULATED k-eff VALUES FOR VARIOUS VALUES OF THE SPECTRAL INDEX
- .s Page A-21 Holtec Report Hi-992285
Fr s Unear Regression with Correlation Coefficlent of 0.03 1.010 o l 0
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2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 6.0 l
Enrichment, wt SE U-235 FIGURE A.3 MCNP CALCULATED k-eff VALUES j
i AT VARIOUS U-235 ENRICHMENTS s
Holtec Report HI-992285 L
r ,
l .
+
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l l
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--- Unear Regression with Correlation Coefficient of 0.38 1.010 1.005 .
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I fiiiiiii;.iiiiiiisi 3.0 3.5 4.0 4.5 5.0 5.5 6.0 Enrichment. wt E U-235 FIGURE A.4 KENO CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS 1
Holtec Report HI-992285 E ^^
s l
- j 1 i l
. 1 i
i O.94 --
- 0 4.Sn E
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t
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0 0.90 ~ =
03.75m E
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c -
@ 3.EsE E 0.88 ' _
0.86 _
s d 3.Os E 0.84 . iiii iisi .. . .... ii., ,,,, ,,,, ,,,, ,,,, ,,,, ,,,,
0.84 0.86 0.88 0.90 0.92 0.94 ,
i WCNP k-off Calculations 1
I FICURE A.5 COMPARISCH OF WCHP AND KEN 05A CALCULATIONS FOR VARIOUS FUEL ENRICHWENTS I
Hollec Report HI-992285 Page A-24 1
l j
o
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0.900 0.920 0.940 0.980 0.980 1.000 1.020 1.040 Reactivity Calculated with KEN 05a FIGURE A.6 COMPARISON OF WCNP AND KEN 05a CALCULATIONS FOR VARIOUS BORON-10 AREAL DENSITIES Holtec Report HI-992285 -
1 1
9 FLORIDA POWER CORPORATION i CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 1
ATTACHMENT F CRITICALITY SAFETY ANALYSIS OF THE WESTINGHOUSE SPENT FUEL STORAGE RACKS IN POOL B OF CRYSTAL RIVER UNIT 3 LICENSE AMENDMENT REQUEST # 239, REVISION 0
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