ML20078G611

Criticality Safety Evaluation of Crystal River Unit 3 New Fuel Storage Vault W/Fuel of 5% Enrichment

ML20078G611
Person / Time
Site:	Crystal River
Issue date:	12/31/1993
From:	Sarah Turner HOLTEC INTERNATIONAL
To:
Shared Package
ML19311B716	List: 3F0195-05, TS Change Request 201 to License DPR-72,increasing Allowable Nomial Fuel Enrichment from 4.2 to 5.0 Weight Percent for Reload Fuel Assemblies.Proprietry B&W Fuel Co Rept BAW-2209P Encl.Proprietary Rept BAW-2209P Withheld Per 10CFR2.790 ML20078G604 ML20078G611 ML20078G616
References
HI-93110, NUDOCS 9502030183
Download: ML20078G611 (32)

Text

(-

et CRITICALITY SAFETY EVALUATION OF THE CRYSTAL RIVER UNIT 3 NEW FUEL STORAGE VAULT WITH FUEL OF 5% ENRICHMENT Prepared for the FIDRIDA POWER CORPORATION by i

Stanley E. Turner, PhD, PE December 1993 Holtec Project 21195 Holtec Report HI-93110 i

230 Normandy Circle 2060 Fairfax Ave.

Palm Harbor, FL 34683 Cherry Hill, NJ 08003 I

9502030183 950126 PDR ADOCK 05000302 i

p PDR

i I

TABLE OF CONTENTS

1.0 INTRODUCTION

1 2.0

SUMMARY

2 3.0 CRITICALITY ANALYSIS 3

[

3.1 Fuel Assembly Specifications.

3 3.2 New Fuel Storage Rack Design.

3 3.3 Analytical Methods.

3 3.4 Manufacturing Tolerances.

4 i

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4.0 ABNORMAL AND ACCIDENT CONDITIONS 5

5.0 REFERENCES

6 i

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I i

t 4

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t 1

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t i

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l List of Tables Table 1

SUMMARY

OF CRITICALITY SAFETY ANA* ISIS.

7 i

Table 2 FUEL ASSEMBLY SPECIFICATIONS 8

{

i List of Floures Fig. 1 New Fuel Storage Vault Configurations.

9 Fig. 2 Reactivity Variation with Moderator Density.

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1.0 INTRODUCTION

In a previous evaluationm, criticality safety analyses established that the New Fuel Storage Vault at Crystal River Unit 3 could safely accommodate fuel of 4.5% enrichment, with restrictions on the number of useable storage locations.

The present study is intended to extend that analysis to confirm the capability of the New Fuel Storage Vault to safely receive and store fuel of 5.0%

initial enrichment with the same restriction on useable storage I

locat_ans. A companion report, HI-931111, documents the capebility of Pool A to also accept fuel of 5.0% initial enrichment.

The New Fuel Storage Vault is intended for the receipt and storage of fresh fuel under normally dry conditions where the reactivity is very low.

To assure criticality safety under accident conditions and to conform to the requirements of General Design Criterion 62,

" Prevention of Criticality in Fuel Storage and Handling",

two separate criteria must be satisfied as defined in NUREG-0800, Standard Review Plan 9.1.1, "New Fuel Storage". These criteria are j

as follows:

e When fully loaded with fuel of the highest anticipated reactivity and flooded with clean unborated water, the maximum reactivity, including uncertainties, shall not exceed a k,,, of 0.95.

e With fuel of the highest anticipated reactivity in place and assuming the optimum hypothetical low density moderation, (i.e.,

fog or foam), the maximum reactivity shall not exceed a k,,, of 0.98.

Results of the present evaluation confirms that the New Fuel Storage Vault can safely accommodate fuel of 5.0% enrichment with the restriction that certain storage locations must remain empty of fuel.

These locations are the same as those defined in the previous evaluation.

_1 _

k

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2.0 SUMRRY i

The New Fuel Storage Vault nonnally provides a 6 x 11 cell array of storage locations arrange > on a 21.125 inch lattice spacing.

Results of the previous evaluation showed that it is necessary to blank-off, and keep empty of fuel, two rows of storage locations as indicated in Figure 1.

With the same restriction, the remaining S4 storage locations in the New Fuel Storage Vault can accommodate 5.0% enriched fuel within the cwo Regulatory guidelines identified above.

Calculations were made with the 27-group NITAWL-KENO-Sa code package, a three-dimensional Monte Carlo analytical technique, using the configuration illustrated in Figure 1.

Results of the criticality safety analyses are summarized in Table 1 for the two accident conditions.

Figure 2 :lllustrates the variation in k,,,

with moderator density and shows that the peak reactivity (optimum moderation) occurs at 7.5% moderator density.

The.naximum reactivity at 7.5% moderator density is a 0.978, including uncertainties, which is within the Regulatory limit of 0.98, thus confirming the acceptability of the MTV for 5.0% fuel.

In the flooded condition (clean unborated water), the storage locations are essentially isolated from each other (neutronically).

Under these conditions and with fuel of 5.0% enrichment, the maximum reactivity, including all known uncertainties, is 0.948 which is less than the limiting value of 0.95, thus confirming the acceptability of the NFV for 5.0% fuel.

l be-3.O CRITICALITY ANALYSES i

3.1 Fuel Assembly Specifications The reference design fuel assembly is a standard Babcock & Wilcox 16 x 15 array of fuel rods, with 17 rods replaced by 16 control f

guide tubes and one instrument thimble.

Table 2 summarizes the I

fuel assembly design specifications and expected range of significant fuel tolerances.

1 3.2 New Fuel Storage Rack Design l

The racks in the New Fuel Storage Vault include steel lead-ins, although there is no steel in the active fuel region.

The storage locations are arranged in eleven rows of six cells each, located on a 21 1/8 i 1/16 inch lattice spacing, as illustrated in Figure 1.

Normally, fuel is stored in the dry condition with very low reactivity.

3.3 Analytical Methods The criticality analyses were made with the three-dimensional Monte-Carlo code package NITAWL-KENO 5a(2) using the 27-group SCALE

• cross-section library (3) and the Nordheim integral treatment for U-238 resonance shielding effects. Benchmark calculations, presented in Appendix A, indicate a bias of 0.0103 i O.0018 for the NITAWL-KENO-Sa code package, at the 95% probability, 95% confidence level").

In the calculational model, each fuel rod, cladding, or guide tube were explicitly described.

The model also used the standard concrete reflector option available in KENO-5a to describe the

• " SCALE' is an acronym for Standardized Computer Analysis for Licensing Evaluation, a standard cross-section set developed by ORNL for the USNRC.

i concrete walls of the NFV.

Monte Carlo (KENO-Sa) calculations inherently include a statistical i

uncertainty due to the random nature of neutron tracking.

To minimize the statistical uncertainty of the KENO-calculated reactivity, a minimum of 500,000 neutron histories in 1000 generations of 500 neutrons

each, were accumulated in each calculation.

For the flooded case and at optimum low-density moderation, confir-= tory calculations were made with 2,500,000 neutron histories.

Arthermore, because of the close approach to the limiting react.vity, check calculations for the flooded case and at optimum low-density moderation were made with the 218 neutron group library.

Results of these calculations are as follows:

CASE 7.5% Mod. Dens.

Flooded 27-groups 0.9648 i 0.0006 0.9345 i 0.0008 1

218-groups 0.9623 1 0.0010 0.9320 0.0012

(

The 218-group calculations resulted in a slightly lower k,,, than the reference 27-group library, thus confirming the reference calculations.

3.4 Manufacturing Tolerances Tne reactivity uncertainties associated with various manufacturing tolerances were calculated by the difference between KENO-Sa calculations, each with the nominal value and a second calculation with each value set at the maximum tolerance.

Results are tabulated below:

Ak Uncertainty Tolerance

@ 7% Mod Deng flooded i 1/8 in lattice spacing i 0.0015 i 0.0007 i 0.02 in % enrichment

• 0.0013 i 0.0014 i

i 0.166 gg/cc i 0.0015 1 0.0015

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h*

Statistical Combination i 0.0025 t 0.022 l

I 4.0 Abnonnal and Accident Conditions Normally, the new fuel storage vault is dry with a very low reactivity.

The two limiting criticality criteria are accident conditions and no other safety concerns have been identified.

Under the double contingency principle of ANSI-N16-1975, endorsed by the April 1978 USNRC position statement, it is not necessary to consider the simultaneous occurrence of independent accident conditions.

f l i

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5.0 REFERENCES

1 1.

S.E.

Turner,

" Criticality Safety Analysis of the New-Fuel Vault with fuel of 4.5% Enrichment", un-numbered SSA report, October 1985.

2.

R.M.

Westfall, et.

al.,

"NITAWL-S: Scale System Module for Performing Resonance Shielding and Working Library Production" in SCALE:

A Modular Code System for performina Standardized Computer Analyses for Licensino EvaluatioL., NUREG/CR-0200, 1979.

L.M. Petrie and N.F. Landers," KENO Va. An Improved Monte Carlo Criticality Program with Supergrouping" in Scale: A Modular Code System for oerformina Standardized Comouter Analyses for Licensina Evaluation, NUREG/CR-0200, 1979.

3.

R.M.

Westfall et al.,

" SCALE:

A Modular Code System for performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200, 1979.

4.

M.G.

Natrella, Experimental Statistics National Bureau of Standards, Handbook 91, August 1963, i.

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Table 1

SUMMARY

OF CRITICALITY SAFETY ANALYSES NEW FUEL VAULT - 5.0% ENRICHED FUEL (Under Accident Conditions) i Optimum Flooded Moderation Condition Temperature for analysis 20*C (68'F) 20*C (68'F) l Reference k, (KEN 05a) 0.9648 0.9345 Calculational bias, 8k 0.0103 0.0103 Uncertainties In the Bias")

i 0.0018 i 0.0018

' KENO Statistics")

i 0.0010 1 0.0014 Lattice spacing i 0.0015 i 0.0007 Fuel enrichment t 0.0013 i 0.0014 Fuel density i 0.0015 i 0.0015 Statistical combination i 0.0029 i 0.0032 of uncertainties (2) j i

Total 0.9751 i 0.0029 0.9448 i 0.0032 i

Maximum Reactivity (k,,,)

0.9780 0.9480 Regulatory Limit 0.98 0.95 l

i U)

With one-sided factor for 95%/95% tolerance (NBS Handbook 91).

(2)

Square root of sum of squares.

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TABLE 2 FUEL ASSEMBLY SPECIFICATIONS Fuel Rod Data l

Outside dimension, in.

0.430 l

Cladding ID, in.

0.377 Cladding thickness, in.

0.0265 Cladding material Zr-4 Pellet diameter, in.

0.369 t

3 UO, density, g/cm 10.420 i 0.166 Enrichment, wt.% U-235 5.0 t 0.02 Fuel Assembly Data l

Number of fuel rods 208 (15x15 array)

Fuel rod pitch, in.

0.568 Control rod guide tube Number 16 0.D.,

in.

0.530 Thickness, in.

0.016 M terial Zr-4 a

h Instrument thimble Number 1

0.D.,

in.

0.493 Thickness, in.

0.026 Nbterial Zr-4 I

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J APPENDIX A 1

BENCHNARK CAIfUIATIONS J

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by

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Stanley E. Turner, PhD, PE HOLTEC INTERNATIONAL November, 1993 b

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TABLE OF CONTENTS

1.0 INTRODUCTION

AND

SUMMARY

A-1 2.0 NITAWL-KENO 5a BENCHMARK CALCUIATIONS A-2 3.0 CASMO3 BENCHMARK CALCUIATIONS A-4 A=5 4.0 WORKER ROUTINE 5.0 CIDSE-PACKED ARRAYS A-6

6.0 REFERENCES

A-7

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i 1

e s-8

bb

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4 List of Tables Table 1 RESULTS OF 27-GROUP (SCALE) NITAWL-KENO 5a.

A-9 CALCULATIONS OF B&W CRITICAL EXPERIMENTS Table 2 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a

. A-10 l

CALCULATIONS OF FRENCH and BNWL CRITICAL EXPERIMENTS Table 3 RESULTS OF CASMO3 AND NITAWL-KEN 05a..... A-11 BENCHMARK (INTERCOMPARISON) CALCULATIONS Table 4 Intercomparison of WORKER-NITAWL-KEN 05a.

. A-12 and CASMO3 Calculations at Various Temperatures Table 5 Reactivity Calculations for Close-Packed

. A-13

[

Critical Experiments i

List of Floures Fig. 1 COMPARISON OF CASMO AND KEN 05A CAIEULATIONS.. A-14 AT VARIOUS ENRICHMENTS IN REPRESENTATIVE FUEL STORAGE RACK Fig. 2 COMPARISON OF CASMO3 AND KEN 05A A=15 TEMPERATURE DEPENDENCE 11 8

l.0 INTRODUCTION AND

SUMMARY

The objective of this benchmarking study is to verify both the NITAWL-KENO 5a *8) methodology with the 27-group SCALE d

cross-section library and the CASMO3 code (3) for use in criticality safety calculations of high density spent fuel storage racks. Both calculational methods are based upon transport theory and have been benchmarked against critical experiments that simulate typical spent fuel storage rack designs as realistically as possible.

Results of these benchmark calculations with both methodologies are consistent with corresponding calculations reported in the literature.

Results of the benchmark calculations show that the 27-group (SCALE) NITAWL-KEN 05a calculations consistently under-

.I predict the critical eigenvalue by 0.0103 i 0.0018 8k (with a 95%

probability at a 95% confidence level) for critical experiments (')

that are as representative as possible of realistic spent fuel storage rack configurations and poison worths.

Extensive benchmarking calculations of. critical experi-ments with CASMO3 have also been reported (5) giving a mean k,,, of 1.0004 i 0.0011 for 37 cases.

With a K-factor of 2.14(') for 95%

probability at a 95% confidence level, and conservatively neglect-ing the small overprediction, the CASMO3 bias then becomes 0.0000 i 0.0024. CASMO3 and NITAWL-KEN 05a intercomparison calculations of infinite arrays of poisoned cell configurations (representative of typical spent fuel storage rack designs) show very good agressent, confirming that 0.0000 i 0.0024 is a reasonable bias and uncertain-ty for CASMO3 calculations.

Reference 5 also documents good agreement of heavy nuclide concentrations for the Yankee core isotopics, agreeing with the measured values within experimental error.

A-1

P*

The benchmark calculations reported here confirm that T

either. the 27-group (SCALE) NITAWL-KEN 05a or CASNO3 calculations are acceptable for criticality analysis of high-density spent fuel storage racks.

Where possible, reference calculations for storage rack designs should be performed with both code packages to provide independent verification.

CASNO3, however, is not reliable when large water gaps ( > 2 or 3 inches) are present.

2.0 NITAWL-KEN 05a BENCID' ARK CALCULATIONS I

Analysis of a

series of Babcock & Wilcox critical l

experiments *, including sons with absorber panels typical of a l

poisoned spent fuel rack, is summarized in Table 1, as calculated with NITAWL-KEN 05a using the 27-group SCALE cross-section library and the Nordheim resonance integral treatment in NITAWL.

Dancoff

]

factors for input to NITAWL were calculated with the Oak Ridge SUPERDAN routine (from the SCALE (2) system of codes). The mean for these calculations is 0.9899 i 0.0028 (1 e standard deviation of f'

the population).

With a one-sided tolerance factor corresponding to 95% probability at a 954 confidence level *, the calculational bias is + 0.0103 with an uncertainty of the mean of i 0.0018 for the sixteen critical experiments analyzed.

similar calculational deviations have been reported by ORNd73 for some 54 critical experiments (mostly clean criticals without strong absorbers), obtaining a mean bias of 0.0100 i 0.0013 i

(95%/95%).

These published results are in good agreement with the results obtained in the present analysis and lend further credence to the validity of the 27-group NITAWL-KEN 05a calculational model for use in criticality analysis of high density spent fuel storage r

racks.

No abnormal deviations in k,,, with intra-assembly water i

gap, with absorber panel reactivity worth, with enrichment or with poison concentration were identified with the 27 group SCALE library, comparable to those previously observed

• with the 123-group GAN-THERNOS cross-section library.

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A-2

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Additional benchmarking calculations were also made foc r

a series of French critical experiments

• at 4.75% enrichment and for several of the BNWL criticals with 4.26% enriched fuel.

l Analysis of the French criticals (Table 2) showed a tendency to f

overpredict the reactivity, a result also obtained by ORNdW.

The calculated k,,,

values showed a trend toward higher values with decreasing core size.

In the absence of a significant enrichment I

effect (see Section 3 below), this trend and the overprediction is l

uttributed to a small inadaquacy in NITAWL-KENO 5a in calculating neutron leakage from very small assemblies.

l Similar results were observed for the BNWL series of' critical experimentsW, which are also small assemblies (although

[

significantly larger than the French criticals).

In this case l

(Table 2),

the calculated mean k,,,

was 0.9959 i 0.0013 (1 e population standard deviation).

Because of the small size of the 1

BNWL critical experiments (compared to the B&W criticals used to determine the KEN 05a bias) and the absence of any significant t

enrichment effect, the results also suggest a small inadequacy of l

NITAWL-KEN 05a in treating large neutron leakage from very small l

~

assemblies.

Since the analysis of high-density spent fuel storage l

racks generally does not entail neutron leakage, the observed inadequacy of NITAWL-KEN 052 is not significant.

Furthermore, omitting results of the French and BNWL critical experiment f

analyses from the determination of bias is conservative since any leakagc that might enter into the analysis would tend to result in overprediction of the reactivity.

i 1

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1

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3.0 CASMO3 BENCHMARK CALCULATIONS

]

The CASMO3 code is a multigroup transport theory code i

utilizing transmission probabilities to accomplish two-dimensional calculations of reactivity and depletion for BWR and PWR fuel assemblies.

As such, CASMO3 is well-suited to the criticality analysis of spent fuel storage racks, since general practice is to treat the racks as an infinite medium of storage cells, neglecting leakaga effects.

i CASMO3 is a modification of the CASMO-2E code and has been extensively benchmarked against both mixed oxide and hot and cold critical experiments by Studsvik Energiteknik s)

Reported t

analyses (5) of 37 critical experiments indicate a mean k,,, of 1.0004 i 0.0011 (10).

To independently confirm the validity of CASMQ:

(and to investigate any effect of enrichment),

a series of calculations were made with CASMO3 and with NITAWL-KENO 5a on identical poisoned storage cells representative of high-density

/

(

spent fuel storage racks.

Results of these. intercomparison calculations * (shown in Table 3 and in Figure 1) are within the normal statistical variation of KEN 05a calculations and confirm the i

bias of 0.0000 1 0.0024 (95%/95%) for CASMO3.

Since two independent methods of analysis would not be expected to have the same error function with enrichment, results of the intercomparison analyses (Table 3) indicate that there is no significant effect of fuel enrichment over the range of enrich-ments involved in power reactor fuel.

i Intercomparison between analytical methods is a technique endorsed by Reg. Guide 3.41, " Validation of Calculational Methods for Nuclear Criticality Safety".

A-4 4

r.'

I

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A second series of CASMO3 and KEN 05a intercomparison calculations consisting of five cases from the-BAW critical experiments were analyzed for the central cell only. The calculat-ed results, also shown in Table 3,

indicate a, mean difference within the 95% confidence limit of the KEN 05a calculations.

This i

lands further credence to the recommended bias for CASMO3.

4.O WORKER ROUTINE The WORKER routine was obtained from ORNL and is intended to interpolate the hydrogen scattering matrices for temperature in order to correct for the deficiency noted in NRC Information Notice 91-66 (October 18, 1991). Benchmark calculations were made against CASMO3, based on the assumption that two independent methods of analysis would not eFhibit the same error.

Results of these calculations, shown in Table 4,

confirm that the trend with temperature obtained by both codes are comparable.

This agreecent establishes the validity of the WORKER routine, in conjunction with l

NITAWL-KEN 05a, in calculating reactivities at temperatures between 20'C and 120'C.

The deficiency in the NITAWL hydrogen scattering matrix l

at temperatures above 20 *C does not appear except in the presence of a large water gap where the scattering matrix is important.

However, the absolute value of the km from CASMO3 is not reliable l

in the presence of a large water gap, although the relative values should be acc7 rate.

In the calculations shown in Table 4 and in Figure 2, the absolute reactivity values differ somewhat but the trends with temperature are sufficiently in agreement to lend credibility to the WORKER routine over the temperature range from 20'C to 120'C.

k l

A-5

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4

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l l*

)a 5.0 CLOSE-PACKED ARRAYS

,. ~

The BAW close-packed series of critical experimentent) intended to simulate consolidated fuel, were analyzed with NITAWL-KEN 05a.

Results of these analyses, shown in Table 5, suggest a slightly higher bias than that for fuel with normal lattice spacings.

similar results'were obtained by ORNLU33 Becauae j

there are so few cases available for analysis, the maximum bias for close-packed lattices may be taken as 0.0155, including uncertain-ty.

This would conservatively encompass all but one of the cases i

measured.

1

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ie:

1*

6.0 REFERENCES

TO APPENDIX A

~

1.

Green, Lucious, Petrie, Ford, White, and Wright, "PSR /NITAWL-1 (code package) NITAWL Modular Code System For Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B", ORNL-TM-3706, Oak Ridge National Laboratory, November 1975.

2.

R.M. Westf all et. al., " SCALE: A' Modular System for Performing Standardized Computer Analysis for Licensing Evaluation",

NUREG/CR-0200, 1979.

3.

A.

Ahlin, M.

Edenius, and H.

Haggblom, "CASMO A Fuel Assembly Burnup Program", AE-RF-76-4158, Studsvik report.

A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for IMR Analysis", ANS Transactions, Vol. 26,

p. 604, 1977.

"CASMO3 A Fuel Assembly Burnup Program, Users Manual",

Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986 i

4.

M.N.

Baldwin et al., " Critical Experiments Supporting Close Proximity Water-Storage of Power Reactor Fuel", BAW-1484-7, The Babcock & Wilcox Co., July 1979.

5.

M. Edenius and A. Ahlin, "CASMO3: New Features, Benchmarking, and Advanced Applications", Nuclear Science and Enaineerina, 100, 342-351, (1988) 6.

M.G.

Natrella, Ernerimental Statistics, National Bureau of Standards, Handbook S1, August 1963.

7.

R.W. Westfall and J. H. Knight, " SCALE System Cross-section Validation with Shipplag-cask Critical Experiments", -&HS, Transactions, Vol. 33, p. 368, November 1979 8.

S.E.

Turner and M.K.

Gurley,

" Evaluation of NITAWL-KENO Benchmark Calculations for High Density Spent Fuel Storage Racks",

Nuclear Science and Enaineerina, 80(2):230-237, February 1982.

A-7

(.

I

(.

9.

J.C. Manaranche, et. al., " Dissolution and Storage Experiment with 4.75% U-235 Enriched UO Rods", Nuclear Technoloav, Vol.

50, pp 148, September 1980.g 10.

A.M.

Hathout, et.

al.,

" Validation of Three Cross-section Libraries Used with the SCALE System for Criticality Analy-sis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981.

1 11.

S.R.

Bierman, et.

al.,

" Critical separation between sub-critical Clusters of 4.29 Wt. % 2D Enriched UO Rods in Water 2

with Fixed Neutron Poisons",

Battelle Pacific Northwest Laboratories, NUREG/CR/0073, May 1978 (with August 1979 errata).

i 12.

G.S. Hoovler, et al., " Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins", BAW-1645-4, Babcock & Wilcox Company (1981).

13.

R.M. Westfall, et al., " Assessment of Criticality Computation-al Software for the U.S.

Department of Energy Office of Civilian Radinactive Waste Management Applications",Section I

6, Fuel Const911ation Applications, ORNL/CSD/TM-247 (undated).

I I

e O

A-8

(

i.:

N Table 1

(~

RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a CAICUIATIONS OF B&W CRITICAL EXPERIMENTS Experiment Calculated u

Number k,,,

I 0.9922 1 0.0006 II 0.9917 i 0,,0005 III 0.9931 1 0.0005 IX 0.9915 i 0.0006 X

0.9903 i 0.0006 XI 0.9919 i 0.0005 XII 0.9915 i 0.0006 XIII 0.9945

, i 0.0006 XIV 0.9902 i 0.0006 I

XV O.9836 1 0.0006 XVI 0.9863 1 0.0006 XVII 0.9875 1 0.0006 XVIII 0.9880 i 0.0006 XIX 0.9882 1 0.0005 XX 0.9885 1 0.0006 XXI 0.9862 i 0.0006 Mean 0.9897 i 0.00070)

Bias (95%/95%)

0.0103 1 0.0018 "I

Standard Deviation of the Mean, calculated from the k,,, values.

(

A-9 4

i ie

+

..i*

I~

j Table 2 RESULTS GF 27-GROUP (SCALE) NITAWL-KEN 05a CATEUIATIONS i

0F FRENCH and BNWL CRITICAL EXPERIMENTS j

French Experiments Separation critical Calculated Distance, cm Height, cm k,g l

l 0

23.8 1.0302 i 0.0008 2.5 24.48 1.0278 i 0.0007 5.0 31.47 1.0168 i 0.0007 r

10.0 64.34 0.9998 i 0.0007 f

BNWL Experiments Calculated i

Case Expt. No.

k,g I

No Absorber 004/032 0.9942 1 0.0007 i

SS Plates (1.05 B) 009 0.9946 i 0.0007 SS Plates (1.62 B) 011 0.9979 i 0.0007 SS Plates (1.62 B) 012 0.9968 i 0.0007 SS Plates 013 0.9956 i 0.0007

[

SS Plates 014 0.9967 i 0.0007 Er Plates 030 0.9955 i 0.0007 Nean 0.9959 i 0.0013 l

A - 10 i

i 0

e Table 3 I

RESULTS OF CASMO3 AND NITAWL-KEN 05a BENCHMARK (INTERCOMPARISON) CAIEULATIONS Enrichment")

k*

G Wt. % U-235 NITAWL-KENO 5a )

CASMO3 l8kl 2.5 0.8376 i 0.0010 0.8386 0.0010 3.0 0.8773 i 0.0010 0.8783 0.0010 3.5 0.9106 i 0.0010 0.9097 0.0009 4.0 0.9367 i 0.0011 0.9352 0.0015 4.5 0.9563 i 0.0011 0.9565 0.0002 5.0 0.9744 1 0.0011 0.9746 0.0002 Mean 0.0008 i

Expt. No.(3)

XIII 1.1021 i O.0009 1.1008 0.0013 l

XIV 1.0997 i 0.0008 1.1011 0.0014 XV 1.1086 i O.0008 1.1087 0.0001

~

XVII 1.1158 i 0.0007 1.1163 0.0010 XIX 1.1215 i 0.0007 1.1237 0.0022 Mean 0.0012 08 Infinite array of assemblies typical of high-density spent fuel storage racks.

"3 k, from NITAWL-KEN 05a corrected for bias.

(3)

Central cell from BAW Critical Experiments A - 11

,(

=. -

Table 4 1

I Intercomparison of WORKER-NITAWL-KEN 05a and CASMO3 Calculations at various Temperatures Temperature CASM03 W-N-KEN 05a(*3 i

4*C 1.2276 1.2345 1 0.0014 17.5*C 1.2322 1.2328 i C.0015 25'C 1.2347 1.2360 1 0.0013 50*C 1.2432 1.2475 i 0.0014 75'c 1.2519 1.2569 i 0.0015 120*C 1.2701 1.2746 i 0.0014

• Corrected for bias

[

i 4

e A - 12

(

a

e Table 5 I

Reactivity Calculations for Close-Packed i

critical Experiments I

Calc.

BAW Pin Module Boron Calculated No.

Expt.

Pitch Spacing conc.

k,,,

No.

cm cm ppm KS01 2500 Square 1.792 1156 0.9891 1 0.0005 l

1.4097 l

KS02 2505 Square 1.792 1068 0.9910 1 0.0005 1.4097 1

KS1 2485 Square 1.778 886 0.9845 i 0.0005 Touching KS2 2491 Square 1.778 746 0.9849 i 0.0005 Touching KT1 2452 Triang.

1.86 435 0.9845 i 0.0006 Touching KT1A 2457 Triang.

1.86 335 0.9865 1 0.0006 Touching KT2 2464 Triang.

2.62 361 0.9827 i 0.0006 Touching KT3 2472 Triang.

3.39 121 1.0034 i 0.0006 Touching A - 13 i

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1.00 E

j I

0.95 y

s

/

z s

!t 0.9e s

0 Casts)

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KEPO 5e

-~~~~

l 0.05 4

~

3 0.80 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 i

FUEL ENRICHtENT, WTs U-235 I

Ftgi l

,, y,gelsk% Tao **gp CALCULAT ONS i

" ^ " *

• g RACK I

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A - 14 j

s.

i F

o

-e-l 1.28

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1.27' o

i

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i

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1.2e;

,/'

.Y e

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• - 1.25 ~

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i 5

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1.24 b ).

1.23

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1.22.

g Te mp e r a tu r e. Doorses C

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Fig. 2 COtPARISON OF CAST 10-3 and KEN 05.

TEt1PERATURE DEPENDENCE 1

(,

A - 15 6

f l

P' g

l l

ATTACHMENT 2 I