ML20138F664

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Critical Safety Analysis of Crystal River Pool B Fuel Storage Rack W/Fuel of 4% Enrichment
ML20138F664
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Site: Crystal River 
Issue date: 12/10/1985
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BLACK & VEATCH
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ML20138F648 List:
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NUDOCS 8512160147
Download: ML20138F664 (29)
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i CRITICALITY SAFETY ANALYSIS OF THE 1

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TABLE OF CONTENTS Page 1.0 I N TR 00 0C T I ON.............................................

1 2.0 S UMMA R Y..................................................

2 I'

3.0 DESIGN BASES.............................................

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4.0 GE0 METRIC AND CALCULATIONAL M00ELS.......................

6 4.1 ' Re f e rence Fuel As s embly............................

6 4.2 ~ Re f e ren ce Sto ra ge Ar ray.............................

6 4.3 An a l yt i c al Me t h od s..................................

6 4.4 Calculational Bias and Uncertainty..................

8 5.0 REFERENCE SUBCRITICALITY AND MECHANICAL TOLERANCE VARIATIONS.....................................

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5.1 Nomi n al De s i gn Cas e.................................

9 5.2 Reactivity. Effects of Manuf acturing Tolerances......

9 5.2.1 Latti ce Pi tch Va ri ati on......................

9 5.2.2 Fuel Enrichment and Density Variation........

9 5.2.3 Summa ry of Tol erance Va ri ati ons............. 10 6.0 AB N ORMAL AND ACC IDENT COND IT IONS......................... 11 6.1 Temperature and Water Density Ef fects............... 11 6.2 Fuel Assembly Abnormally Located Outside Storage Rack 11 REFERENCES APPENDIX A BENCHMARK CALCULATIONS 11

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Page 1

SUMMARY

OF CRITICALITY' CALCULATIONS..........................

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2 FUEL ASSEMBLY DESIGN SPECIFICATIONS..........................

7 3

CALCULATED STATISTICAL VARIATIONS IN REACTIVIlY (MECHANICAL).................................................

10 4

EFFECT OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF S T OR A G E RA C K.................................................

11 a!

c LIST OF FIGURES 1

Maximum Reactivity (k, ) Including Uncertainties for Fuel of various Enrichments in the Pool "B"

Storage Racks.........

3 2

Effect of Lattice Spacing on Reactivity of 4.0% Enriched Fuel Assemblies..............................................

13

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iii

1.0 INTRODUCTION

The present pool "B"

standard spent fuel storage racks ir, tif, Cry:tal River Nuclear Power Plant are. licensed to ' store fuel of 3.5 wt.%

U-235 initial enrichment.

The previous criticality analysis, submitted in support of the present Technical Specification limit, documented a subcriticality margin substantially below the NRC limiting reactivity value Of 0.95 including all uncertainties.

' The evaluation reported here was prepared to justify the criticality safety'of an increase in the Technical Specification limit on fuel enrichment to 4.0% 1.n.the existing pool "B" fuel storage racks.

Results of the present evaluation confirm that.the maximum reactivity will be less than 0.95 including all uncertainties, with the racks fully loaded with fuel of 4.0 wt.% U-235 enrichment and flooded with unborated water at the temperature corresponding to the highest reactivity.

With fuel of 4.0 wt.% nominal enrichment, the U-235 loading is 52.75 1 0.87 grams per axial centimeter of fuel assembly, including tolerances on fuel density and enrichment, and the calculated maximum reactivity (k,, including uncertainties) is 0.9251. Extra-polation of the calculational results to higher enrichments indicate that the maximum reactivity remains below 0.95 even for an enrichment of 4.5%.

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2.0

SUMMARY

The criticality analyses of the Crystal River pool "B"

spent fuel ' storage racks under normal and abnormal conditions with fuel of 4.0% enrichment are summariz' d in Table 1 below.

e Table 1

SUMMARY

OF CRITICALITY CALCULATION'S Case k, or ak, Comment Normal Condition k, reference 0.9208-Section 5.1 Calculational bias 0.0013 Section 4.4 (Appendix A)

Uncertainties-Bias 0.0018 Section 4.4 Mechanical 10.0024 Section 5.2 (Table 3) i0.0030 Statistical combination Total 0.9221 1 0.0030 Maximum k, 0.9251 Abnormal and Accident Conditions Fuel element eccentric positioning negligible Section 5.2 Increased temperature or void negative Section 6.1 Assembly outside rack negligible Section 6.2

-Thus, a k, of 0.925 is conservatively estimated to be the maximum k, under the worst combination of calculational and mechanical uncertainties with a 95%

probability at a 95% confidence level under normal conditions.

Credible abnormal og accident conditions will not result in exceeding the ' limiting reactivity of 0.95.

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Results of calculations of the maximum reactivity (k, ) for enrichment are shown on Fig.1.

These data show that the reactivity' remains below the lim-iting value (keff of 0.95) even for an enrichment of 4.5%.

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3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 I

ENRICHMENT, WTI U-238 Fig.1 Maximum Reactivity (koo ) Including Uncertainties for Fuel of J

Various Enrichments in the Pool "B" Storage Racks.

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3.0 DESIGN BASES The objective in -the pool "B" spent fuel storage racks for the Crystal. River plant is to assure that a keff equal to or less than 0.95 is maintained with the racks fully. loaded with fui of the highest anticipated reactivity and flooded with unborated water at a temperature corresponding to the highest 7

reactivity.

The maximum calculated reactivity includes a margin for uncer-tainty in reactivity calculations and in mechanical tolerances, statistically combined, such that the true keff will be equal to or less than 0.95 with a 95% probability at a 95% confidence level.

Applicable codes, standards and, regulations or pertinent sections thereof

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1nclude the following:

o General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.

o NRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.

o USNRC Standard Review Plan, NUR EG-0800, Section 9.1.2, Spent Fuel Storage.

o Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis (proposed), December 1981.

o Regulatory Guide 3.41, Validation of Calculational Method for Nuclear Criticality Safety (and related ANSI N16.9-1975).

o. ANSI N210-1976, Design (bjectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.

o ANSI N18.2-1973, - Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants.

The design basis fuel assembly is a 15 x 15 array of fuel rods (Babcock &

Wilcox design) containing UO2 at a maximum uniform enrichment of 4.0% U-235 by weight, corresponding to 52.73. grams U-235 pec axial centimeter of fuel assembly.

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To assure' the true reactivity will always be less than the calculated reactiv-ity, the followin'g conservative assumptions were made.

o Moderator is pure, unborated water at a temperature corresponding to the highest reactivity.

~ Lattice of storage racks is infinite in ali directions i.e., no credit o

is taken for axial or radial neutron leakage (except in the considera-tion of certain abnormal / accident conditions).

o Neutron absorption in minor structural members is. neglected i.e., spacer grids are replaced by water.

G 5

4.0 GE0 METRIC AND CALCULATIONAL MODELS i

4.1 Reference Fuel Assembly i

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The reference design fuel assembly -is a standard Babcock & Wilcox 15 x 15 array of fuel rods, with 17 rods replaced by 16 control rod guide tubes and one instrument thimble.

Table 2 summarizes the fuel assembly design specifi-cations and expected range of significant fuel tolerances.

4.2 Reference Storage Array In the pool "B" fuel storage racks, the fuel assemblies are supported top and bottom on a 21.12510.0625 inch lattice spacing.

The analytical model assumed the fuel assemblies to be infinitely long, separated only by pure unborated water.

In the two-dimensional representations, each fuel rod and cladding were explicitly described and a zero current (white albedo) boundary condition

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applied at the centerline of the water space between assemblies, effectively creating an infinite array.

4.3 Analytical Methods Nuclear criticality analyses of the fuel array were performed with both the 2

3 CASM0-2El program and the AMPX -KENO computer package using the NITAWL sub-routine for U-238 resonance shielding effects (Nordheim integeral treat-ment). With AMPX-KENO, both the 123-group GAM-THERM 0S and the more recent 27-4 group SCALE cross-section libraries were used.

In addition, a four-group S

6 diffusion theory method of analysis (NULIF -SNEID ) was used to provide addi-tional confidence in the reference criticality calculations.

These four independent methods of analysis were in reasonably good agreement, although the CASMO-2E calculation yielded the highest value for reactivity, which, for conservatism, was adopted as the reference.

For the evaluation of manufac-turing tolerance effects, both CASM0-2E and the diffusion theory method of analysis (NULIF-SNEID) were used (Ref. 8) to calculate small incremental reactivity changes.

6

r-Table 2 FUEL ASSEMBLY DESIGN SPECIFICATIONS Fuel Rod Data Outside dimension, in.

0.430 Cladding thickness, in.

0.0265 Cladding material Zr-4 Pellet diameter, in.

0.369 3

U0 density, g/cm

- 10.420 t 0.166 2

Enrichment, wt.% U-235 4.00 1 0.02 Fuel Assembly Data Number of fuel _ rods 208 (15 x 15 array)

Fuel rod pitch, in.

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

0.530 Thickness, in.

0.016 Material Zr-4 Instrument thimble Numbcr 1

0.0.,in.

0.493 Thickness, in.

0.026 Material Zr-4 U-235 g/ axial cm of assembly 52.73 0.87 7

4.4 Calculational Bias and Uncertainty Results of benchmark calculations of the 123-group AMPX-KEN 0 calculations have been previously published,7 and indicate a bias of 0.000!0.003 (95% proba-bility at a 95% confidence level) for the critical experiments analyzed.

Similar benchmark calculations for the 27-group SCALE ~ cross-section set with AMPX-KEN 0 and for the CASMO-2E code are described in Appendix A.

The 27-group AMPX-KEN 0 benchmarking indicated a bias of 0.0106 1 0.0048, which is similar to that reported by ORNL amd Westinghouse.9 The CASM0-2E computer code, a 8

two-dimensional, multigroup transport theory code for fuel assemblies is used both as a primary method of analysis and as a means of evaluating small reac-tivity increments associated with manufacturing tolerances.

CASM0-2E bench-marking (Appendix A) resulted in a calculational bias of 0.0013 t 0.0018 (95%/95%).

In fuel rack analyses, the higher, hence more conservative, of the reactivity values (including uncertainties) calculated by either CASM0-2E or AMPX-KEN 0 is generally 'used for the reference storage cell infinite multipli-cation factor.

8 i

r 5.0 REFERENCE SUBCRITICALITY~ AND MECHANICAL TOLERANCE VARIATIONS 5.1 Nominal Design Case Under normal conditions, with nominal dimensions and composition, the k, values calculated by the four methods of analysis are as follows.

Maximum k Analytical Method Bias-Corrected k_

(95%/95%)*

CASM0-2E 0.9221 i 0.0018 0.9239 27-group AMPX-KEN 0 0.9150 t 0.0089 0.9239 123-group AMPX-KEN 0 0.91 0.0065 0.918 NULIF/SNEID 0.9089 (for information only)

The statistical uncertainty in the AMPX-KEN 0 calculations includes a one-sided 10 tolerance factor of 1.924 corresponding to 95% probability at a 95% confi-dence level with 50,000 neutron histories in 100 generations. For the nominal design case, the CASM0-2E calculation ' yields the highest (most conservative) reactivity and is therefore used as the reference reactivity.

5.2 Reactivity Effects of Manufacturing Tolerances 5 '.2.1 Lattice Pitch Variation The' fuel assemblies are positioned by aluminum guide boxes at the top and by special extruded blocks at the bottom, maintaining an effective lattice spacing of 21.125 i 0.0625 inches.

At this large lattice spacing, the re-activity is insensitive to the tolerance in spacing (see Fig. 2). Similarly, the reactivity is insensitive to the small tolerance (t0.0625 inch) in eccen-tricity of the fuel assembly location within a storage location.

5.2.2 Fuel Enrichment and Density Variation The design maximum enrichment is 4.00 1 0.02 wt.% U-235.

CASM0-2E calcula-tions of the sensitivity to small enrichment variations yielded a coefficient 9

of 0.0041 ak per 0.1 wt.% U-235 at the design enrichment, in the range from 3.9% to 4.0% enrichment.

For the tolerance on U-235 enrichment of t0.02 in wt.%, the uncertainty on k, is 0.0008 ak.

i I

3 Calculations-were made with the maximum U02 fuel density (10.586 g/cm ),

resulting in an uncertainty in reactivity of !0.0023 ak over the expected r

tolerance in UO2 densities.

5.2.3 Summary of Tolerance Variations i

Calculated reactivity increments from mechanical and fabrication tolerances are summarized in Table. 3.

Table 3 CALCULATED STATISTICAL VARIATIONS IN REACTIVITY (MECHANICAL)

Incremental Case Tolerance Reactivity, ak Lattice pitch t 0.0625 negligible Eccentricity i 0.0625 negligible

~j, Fuel enrichment t 0.02% U-235 1 0.0008 3

Fuel density i 0.166 g/cm 1 0.0023 Statistical combination t 0.0024 (root-mean-square of reactivity increments) 6 I'

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6.0 ABNORMAL AND ACCIDENT CONDITIONS-T 6.1 -

Temperature and Water Density Effects Increasing temperature from the nominal 40*F (water density of 1.000) is calculated to monotonically decrease reactivity, as indicated in Table 4 (incremental reactivity effects calculated by diffusion / blackness theory).

Introducing voids in the water internal to the storage cell.- (to simulate boiling) decreased reactivity, as shown in the table.

Voids due to boiling will not occur in the outer (flux-trap) water region.

Table 4 EFFECT OF TEMPERATLRE AND VOID ON CALCULATED REACTIVITY OF STORAGE RACK Case Ak_

Comment (H 0) = 0.998 20'C-0 Reference, p 2

(H 0) = 0.992 40'C

-0.0015 p

2 (H 0) = 0.983 60*C

-0.0022 p

2 (H 0) = 0.972 80 C

-0.0054 p

2 (H 0) = 0.958 100'C

-0.0075 p

2 (H O) = 0.943 120*C

-0.0111 p

2 120*C

-0.0799 Simulates boiling

+ 20% void 6.2 Fuel Assembly Abnormally Located Outside Storage Rack To investigate the possible reactivity effect of a fuel assembly abnormally

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-located outside the rack, calculations were made for unpoisoned assemblies

  • By 1-dimensional diffusion theory (cylindricized geometry) normalized to the reference design CASMO calculation (k,= 0.9208) at 21.125 inches lattice spacing.

11

I separated only by water. Figure 2 shows the results of these calculations.

From these data, the reactivity (k, ) will be less than 0.95 for any lattice spacing greater than ~15 inches ( ~6.5-inch water gap) in the absence of any neutron-absorbing material other than water between assemblies. Figure 2 also illustrates the insensitivity to spacing in the region of the reference spacing (21.125 inches).

A fuel assembly accidentally positioned outside the rack - cannot be located closer than 10 inches from another fuel assembly.

With this separation, the rectivity effect of such an accident will be negligible, as indicated in Fig.

2, even' without credit for the neutron-absorbing material between the rack and the abnormally-located fuel assembly.

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For a drop on top of the rack, the fuel assembly will come. to rest hori-zontally on top of the rack with a centerline separation of ~20 inches (water gap of ~15 inches).

Consequently, fuel assembly drop accidents will not result in a significant increase in reactivity above that calculated for the infinite nominal design storage rack.

Local leakage effects would further reduce reactivity.

In both cases of a fuel assembly abnormally located outside the rack, local neutron leakage would further reduce the reactivity effect of the abnormally-located fuel assembly.

Furthermore, soluble boron is normally present in the spent fuel pool (for which credit is permitted under accident conditions) and would reduce the maximum k, to substantially less than 0.95. Consequently, it is concluded that the postulated accident conditions will not adversely affect r-the criticality safety of the pool "B" spent fuel storage racks.

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o 0.90 0.90 10 12 14 16 18 20 22 LATTICE SPACING, INCHS Fig. 2 Effect of Lattice Spacing on Reactivity of 4.0% Enriched Fuel Assemblies 13

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REFERENCES 1.a A. Ahlin, M. Edenius, H. Haggblom, "CASMO - A Fuel Assembly Burnup Program," AE-RF-76-4158, Studsvik report (proprietary).

1.b.

A. Ahlin. and M. Edenius, "CASMO - A Fast Transport Theory Depletion

- Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604,1977.

1.c.

M.

Edenius et-al.,

"CASMO Benchmark Report," Studsvik/RF-78-6293, Aktiebolaget Atomenergi, March 1978.

2.

Green, Lucious, Petrie, Ford, White. Wri ght, "PSR-63/AMPX-1 (code package), AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," ORNL-TM-3706, Oak Ridge National

. Laboratory, March 1976.

3..

L.

M.

Petrie and N.

F.

Cross," KENO-IV, An Improved Monte Carlo Criticality Program,"

0RNL-4938, Oak Ridge National Laboratory, November 1975.

4.

- R. ' M. Westf all et al., " SCALE: ~ A Modular Code System for. Performing

-Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200, 1979.

5.

- W. A. Wittkopf, "NULIF - Neutron Spectrum Generator, Few-Group Constant -

Generator and Fuel Depletion Code," BAW-426, The Babcock & Wilcox Company, August 1976.

6.

S. E. Turner, SNEID is a one-dimensional diffusion theory routine developed by Black & Veatch and veri fied by comparison with PDQ07 one-dimensional calculattons.

7.

S. E. Turner and M.

K.

Gurley, " Evaluation of AMPX-KEN 0 Benchmark Calculations for High Density Spent Fuel Storage Racks," Nuclear Science and Engineering, 80(2): 230-237, February 1982.

8.

R. M. Westfall and J. R. Knight, " Scale System Cross-section Validation

^

with Shipping-cask Critical Experiments," ANS Transactions, Vol. 33,

p. 368, November 1979.

-9.

B. F. Cooney et al., " Comparisons of Experiments and Calculations for LWR Storage Geometries," Westinghouse NES, ANS Transactions, Vol. 39, p. 531,- November 1981.

10.

- M. G. Natrella, Experimental Statistics National Bureau of Standards, Handbook 91, August 1963..

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APPENDIX A BENCHMARK CALCULATIONS

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1.0 INTRODUCTION

AND

SUMMARY

The objective of this benchmarking study is-to verify both I

the AMPX (NITAWL)

KENO methodology with the 27-group SCALE 2

3 cross section library and the CASMO-2E code for use in criti-cality calculations of high density spent fuel storage racks.

Both calculational methods are based on transport theory and have been benchmarked against critical experiments that, as realisti-cally as

possible, simulate typical spent fuel storage rack designs.

Results of these benchmark calculations with both methodologies are consistent with corresponding calculations

-reported in the literature and with the requirements of Regulatory Guide 3.41, Rev.

1, May 1977.

Results of these benchmark calculations show that the 27-group (SCALE) AMPX-KENO calculations consistently underpredict the critical eigenvalue by 0.0106 i 0.0048 Ak (with a 95% proba-bility at a

95%

confidence level) for critical experiments selected to be representative of realistic spent fuel storage rack configurations and poison worths.

Similar calculations by Westinghouse suggest a bias of 0.012 i 0.0023, and the results of ORNL analyses of 54 relatively " clean" critical experiments show a bias of 0.0100 i 0.0013.

Similar calculations with CASMO for clean critical experi-ments resulted in a bias of 0.0013 i 0.0018 (95%/95%).

CASMO and AMPX-KENO intercomparison calculations of infinite arrays of poisoned cell configurations show very good agreement and suggest that a bias of 0.0013 i 0.0018 is the reasonably expected bias and uncertainty for CASMO calculations.

  • Validation of Calculational Methods for Nuclear Criticality.

Safety.

See Also ANSI N16.9-1975.

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O The benchmark calculations recorted here indicate that either the 27-group (SCALE) AMPX-KENO or CASMO calculations are acceptable for criticality analysis of high density spent fuel storage racks.

The preferred methodology, however, is to perform independent calculations with both code packages and to utilize the higher, more conservative value for the reference design infinite multiplication factor.

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2.0 AMPX (NITAWL)-KENO BENCHMARK CALCULATIONS i

i 7,

Analysis of a series of B&W critical experiments,4 which

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include some with absorber plates typical of a spent fuel rack, 1

  1. h is summarized in Table 1 as calc'ulated with AMPX-KENO using the 27--group SCALE cross -section library and the Nordheim resonance

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. integral treatment in NITAWL.

The mean (and standard deviation of the mean) for these calculations is 0.9894 1 0.0019.

With a 5

one-sided tolerance factor (K

2.502), corresponding to 95%

=

probability at a 95% confidence level, the calculational bias is

+0.0106 with an uncertainty of.t0.0048.

6 Similar calculational deviations reported by Westinghouse are also shown in Table 1 and suggest a bias of 0.012 0.0023 (95%/95%).

In addition, ORNL has analyzed some 54 critical experiments using the same methodology, obtaining a mean bias of 0.0100 1 0.0013 (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 AMPX-KENO calculational model for use in criticality analysis of high density spent fuel storage racks.

No trends in k,gg with intra-assembly water-gap, with absorcer plate reactivity worth, or with soluble poison concentration were identified.

  • Significantly large trends in k,gg with water-gap 8 "for AMPX-C absorber plate reactivity worth have been reported 4

KENO calculations with the 123-group GAM-THERMOS library.

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Variance analysis of the data -in Table 1 suggests the possibi,lity that an unknown factor.may be causing a slightly larger variance than might be expected from the Monte Carlo statistics alone.-

However, such a factor, if one truly exists, is too small to be resolved on the basis of critical-experiment data presently available.

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Table 1 RESULTS OF 27-GROUP (SCALE) AMPX-KENO CALCULATIONS f'

.OF B&W CRITICAL EXPERIMENTS f

4 Westinghouse

,. Experiment

. Calculated Calculated-Meas.

Number k,pp-o k,pp I

0.9889 t0.0049 0.008

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'II 1.0040' A0.0037 0.012 L

'III 0.9985

  • 0.,0046 0.008 IX 0.9924
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- X 0.9907

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-0.008 XI 0.9989

  • 0.0044

+0.002 XII 0.9932 i0.0046 0.013 XIII 0.9890~

  • 0.0054 0.007 XIV 0.9830
  • 0.0G38 0.013 XV 0.9852
  • 0.0044 0.016 XVI 0.9875 10.0042 0.015 XVII 0.9811
  • 0.0041 0.015

'XVIII 0.9784 i0.0050 0.015

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XIX 0.9888 i0.0033 0.016 XX 0.9922

  • 0 0048 0.011 XXI 0.9783 i0.0039 0.017

- Mean 0.9894 i0.0011(1) 0.0120 i 0.0010 Bias 0.0106 i0.0019(2) 0.0120 i 0.0010 Bias (95%/95%)

-0.0106

-10.0048 0.0120 i 0.0023 Maximum Bias 0.0154 0.0143 7..

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~ (1) Calculated from individual standard deviations.

(2)- Calculated from k gg values.

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3. 0' CASMO-2E BENCHMARK CALCULATIONS 3.1 General The CASMO-2E code is a multigroup transport theory code utilizing transmission probabilities to accomplish two-dimen-sional calculations of reactivity and depletion for BWR and PWR fuel assemblies.

As such, CASMO is well suited to the criti-cality analysis of spent fuel storage rackt.,

since general practice is to treat the racks as an infinite medium of storage cells, neglecting leakage effects.

9 CASMO is closely analogous to the EPRI-CPM code and has been extensively benchmarked against hot and cold critical experiments by Studsvik Energiteknik.3 Reported analyses of 26 critical experiments indicate a mean k gg of 1.000 t 0.0037 e

(1a)._

Yankee Atomic has also reported results of extensive 10 benchmark calculations with CASMO.

Their analysis of 54 Strawbridge and Barry critical experiments using the reported buckling indicates a mean of 0.9987 i 0.0009 (la), or a bias of 0.0013 i

0.0018 (with 95%

probability at a

95%

confidence level).

Calculations were repeated for seven of the Strawbridge and Barry experiments, selected at random, yielding a mean k,gg of 0.9987 i 0.0021 (la),

thereby confirming that the cross section library and analytical methodology being used for the present calculations are the same as those used in the Yankee analyses.

Thus, the expected bias for CASMO in the analysis of

" clean" critical experiments is 0.0013 i 0.0018 (95%/95%).

3.2 Benchmark Calculations 4

CASMO-2E benchmark-calculations have also been made for the B&W series of critical experiments with absorber

plates, simulating high density spent fuel storage racks.
However, CASMO, as an assembly code, cannot directly represent an entire A-7

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core.

configuration

  • without introducing uncertainty due to reflector constants and the appropriateness of their spectral weighting.

For this reason, the poisoned cell configurations of the central assembly, as calculated by CASMO, were benchmarked against corresponding calculations with the 27-group (SCALE)

AMPX-KENO code package.

Results of this comparison are.shown in Table 2.

10 Yankee has attempted such calculations using CASMO-generated constants in a two-dimensional _four-group PDQ model, obtaining a mean k,gg of 1.005 for eleven (11) poisoned cases and 1.009 for five (5) unpoisoned~ cases.

Thus, Yankee benchmark calculations suggest that CASMO itends to slightly overpredict reactivity.

=s4 1'

A-8 L_

Table 2 RESULTS OF CASMO BENCHMARK (INTERCOMPARSION)

CALCULATIONS

~

k_(1}

'B&W Experiment No.III AMPX-KENO (2)

CASMO ak XIX 1.1203 i 0.0032 1.1193 0.0010 XVII-1.1149 i 0.0039 1.1129 0.0020 XV 1.1059 i 0.0038 1.1052 0.0007 3

Interpol'ated 1.1024 i 0.0042 1.1011 0.0013 XIV 1.0983 1 0.0041 1.0979 0.0004 XIII 1.0992 i 0.0034 1.0979 0.0013 Mean i 0.0038 0.0011 i 0.0006 BWR Fuel Rack 0.9212 i 0.0027 0.9218

-0.006 (1) Infinite array of central assemblies of 9-assembly B&W criti-cal configuration (Ref. 4)

(2) k, from AMPX-KENO corrected for bias of 0.0106 Ak.

(3) Interpolated from Fig. 28 of Reference 4 for soluble boron concentration ~at critical condition.

Since. the difference is well within the normal KENO statistical variation, these calculations confirm the validity of CASMO calculations for the typical high density poisoned spent fuel rack configurations.

The differences shown in Table 2 are also consistent with a bias of 0.0013 i 0.0018 determined in Section 3.1 above as the expected bias and uncertainty of CASMO calcula-tions.

A-9

i REFERENCES 1.

Green, Lucious, Petrie, Ford, White, Wright, PSR-63/AMPX-1 (code package),

AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B, ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.

L.

M.

Petrie and N.

F.

Cross, KENO-IV, An Improved Monte Carlo Criticality Program, ORNL-4938, Oak Ridge National Laboratory, November 1975.

2.

R.

M.

Westfall et al.,

SCALE:

A Modular Code System for Performing Standardized Computer Ar.alyses for Licensing Evaluation, NUREG/CR-0200, 1979.

W.

E.

Ford, III et al.,

A 218-Neutron - Group Masta Cross Section Library for Criticality Safety Studies,. ORNL/TM-4, 1976.

A Fuel Assembly 3.

A.

Ahlin, M.
Edenius, H.

Haggblom, CASMO Burnup Program, AE-RF-76-4158, Studsvik report (proprietary).

A Fast Transport Theory A.

Ahlin and M.

Edenius, CASMO Depletion Code for LWR Analysis, ANS Transactions, Vol. 26,
p. 604, 1977.

M.

Edenius' et al.,

CASMO Benchmark

Report, Studsvik/RF-78/6293, Aktiebolaget Atomenergi, March 1978 CASMO-2E Nuclear Fuel Assembly Analysis, Application Users Manual, Rev. A, Control Data Corporation, 1982.

4.

M.

N.

Baldwin et al.,

Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, The Babcock & Wilcox Company, July 1979.

5.

M.

G.

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

6.

B.

F.

Cooney et al.,

Comparisons of Experiments and Calculations for LWR Storage Geometries, Westinghouse NES, ANS Transactions, Vol. 39, p. 531, November 1981.

7.

R.

M.

Westfall and J.

R.

Knight, Scale System Cross-Section Validation with Shipping-Cask Critical Experiments, ANS Transactions, Vol. 33, p.

368, November 1979.

8.

S.

E.

Turner and M.

K.

Gurley, Evaluation of AMPX-KENO Benchmark Calculations for High Density Spent Fuel Storage
Racks, Nuclear Science and Engineering, 80(2):

230-237, February 1982.

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REFERENCES (Continued) 9.

The EPRI-CPM Data Library, ARMP Computer Code Manuals, Part II, Chapter 4,

CCM3, Electric Power Research Institute, November 1975.

10.

E.

E.

Pilat, Methods for the ' Analysis of Boiling Water Reactors, Lattice Physics, Yankee Atomic Electric Co.,

YAEC 1232, December 1980.

11.

L.

E.

Strawbridge and R.

F.

Barry, Criticality-Calculations for Uniform, Water-Moderated Lattices, NSE 23, 58, September 1965.

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