ML20081A883

Criticality Safety Analysis of Crystal River Spent Fuel Storage Rack

ML20081A883
Person / Time
Site:	Crystal River
Issue date:	12/31/1983
From:	Gurley M, Sarah Turner FLORIDA POWER CORP.
To:
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ML20081A872	List: ML20081A870 ML20081A875 ML20081A883
References
SS-152, NUDOCS 8403060366
Download: ML20081A883 (19)

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3 55-152 CONTROLLED COPY NO. 2 Page 1 of 19 CRITICALITY SAFETY ANALYSIS OF THE CRYSTAL RIVER SPENT FUEL STORAGE RACK by S. E. Turner, Ph.D.

M. K. Gu rl ey

. December 1983 p E O 2 PDR,

1 SS-152 ~

Page 2 of 19 TABLE OF CONTENTS Page 1.0 INTR 000CTION................................................ 1 2.0 Subt%RY..................................................... 2 3.0 DESIGN 8ASES................................................ 3 4.0 GE0 PETRI C AND CALCULATIONAL >0DELS . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.1 Ref e rence Fu el As s emb1y. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.2 An a lyt i c a l Met h o d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.3 Cal cul ati onal Bi as and Uncertai nty. . . . . . . . . . . . . . . . . . 8 4.4 Reference Fuel Storage Cell . . . . . . . . . . . . . . . . . . . . . . . . . 8 5.0 REFERENCE SUBCRITICALITY AND MECHANICAL TOLERANCE VARIATIONS........................................ 9 5.1 Nomi n al D e s i gn Ca s e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 5.2 Boron Loading Variation............................. 9 5.3 Storage Cell Latti ce Pitch Vari ation. . . . . . . . . . . . . . . . 10 5.3.1 inner Water Thickness Vari ati ons. . . . . . . . . . 10 5.3.2 Outer (Flux-Trap) Water Thickness Variation................................. 10 5.4 Stainl ess-Steel Thickness Va ri ations . . . . . . . . . . . . . . .. 10 5.5 Fuel Enrichment and Density Vari ati on. . . . . . . . . . . . . . . 11 5.6 Absorber Width Tolerance Vari ati on. . . . . . . . . . . . . . . .. . 11 5.7 Summa ry of Stati sti cal Vari ati ons . . . . . . . . . . . . . . . . . . . 11 6.0 ABNOR F%L AND ACCIDE NT CONDITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 6.1 Eccentric Positioning of Fuel Assembly in Storage Rack........................................ 13 6.2 Temperature and Water Density Effects............... 13 6.3 Fuel Assembly Abnormally Located Outside Storage Rack........................................ 14 REFERENCES ii

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SS-152 Page 3 of 19 l LIST OF TABLES No. Page 1 SU KwnRY OF CRITI CALITY CALCULATIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2 FUEL ASSE FSLY DES IGN SPE CIF I CATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3 CALCULATED STATISTICAL VARIATIONS IN REACTIVITY

( ME C HA N I CA L ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4 EFFECT OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF S T O R A CE RA C K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 LIST OF FIGURES 1 Re f erence des i gn con fi gu rati on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2 Reactivity effect of separation between fuel assemblies (unpoisoned).................................................... 15 1

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S5-152 Page 4 of 19

1.0 INTRODUCTION

The current spent fuel storage racks in the Crystal River Nuclear Power Plant are licensed to store fuel of 3.3 wt.% U-235 initial enrichment. The previcus criticality analysis, submitted in support of the present Technical Specification limit on fuel enrichment, 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 enri chment in the existing spent fuel storage racks. Results of the present evaluation confirm that the maximum reactivity will be less than 0.95, includ-ing all uncertainties, with the racks fully loaded with fuel of 3.5 wt.% U-235 enrichment and flooded with unborated water at the temperature corresponding to the highest reactivity. With fuel of 3.5 wt.% nominal enrichment, the U-235 loading is 46.14 0.78 grams per axial centimeter of fuel assembly, including tolerances on fuel density and enrichment.

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SS-152 Page 5 of 19 2.0

SUMMARY

The criticality analyses of the Crystal River spent fuel storage rack under normal and abnormal conditions with fuel of 3.5% enrichment are summa-rized in Table 1 below.

l Table 1 SUF9%RY OF CRITICALITY CALCULATIONS i

Case k, o r a k, Comment Normal Condition k ,, reference 0.9347 Section 5.1 Calculational bias +0.000 Section 4.3 (Ref. 5)

Uncertainties Bias t0.003 Section 4.3 Cal cul ati onal 0.0036 Section 5.1 Mechanical t0.0097 Section 5.7 (Table 3) 0.0108 Statistical combination Total 0.9347 0.0108 Maximum k, 0.9455 Abnormal and Accident Conditions Fuel element eccentric positioning negative Section 6.1 Increased temperature or void negative Section 6.2 Assembly outside rack negligible Section 6.3 Thus, a k, of 0.946 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 or accident conditions will not result in exceeding the limiting reactivity of 0.95.

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1 55-152 Page 6 of 19 3.0 DESIGN BASES The objective in the spent fuel storage racks for the Crystal River plant is to assure that a k rr e equal to or less than 0.95 is maintained with the racks fully loaded with fuel of the highest anticipated reactivity and flooded with unborated water at a temperature corresponding to the highest reactiv-ity. The maximum calculated reactivity includes a margin for uncertainty 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% proba-bility at a 95% confidence level.

Applicable codes, standards and regulations or pertinent sections thereof include the following:

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

e USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage.

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

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

e ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.

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

e ANSI /ANS-57.2-1983, Design Requi rements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.

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SS-152 Page 7 of 19 The design basis fuel assembly is a 15 x 15 array of fuel rods (Babcock &

Wilcox design) containing U02 at a maximum uniform enrichment of 3.5% U-235 by wei ght , corresponding to 46.14 grams U-235 per axial centimeter of fuel assembly.

To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made, e Moderator is pure, unborated water at a temperature corresponding to the highest reactivity, e Lattice of storage racks is infinite in all directions; i .e. , no credit is taken for axial or radial neutron leakage (except in the consideration of certain abnormal / accident conditions).

e Neutron absorption in minor structural members is neglected; i.e.,

spacer grids are replaced by water, e Pure zirconium is used for cladding, control rod guide tubes, and instrument thimbles; i.e., higher neutron absorption of alloying materials in Zircaloy is neglected.

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SS-152 Page 8 of 19 4.0 GE0KTRIC AND CALCULATIONAL K)DELS 4.1 Reference Fuel Assembly The reference design fuel assembly, illustrated in Fig.1, is a 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 _ Analytical Methods Nuclear criticality analyses of the high-density spent fuel storage rack were performed with the AMPX -KEN 1 02 computer package, using the 123-group GAM-THERMOS cross-section set and the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treatment). AWX-KEN 0 has been exten-sively benchmarked against a number of critical experiments (e.g., Refs. 3, 4, and 5). In the geometric description in tne KEN 0 input, each fuel rod and guide tube was explicitly represented.

The reference criticality calculation was supplemented by two independent methods of analysis in order to verify the reference calculation and to pro-vide greater assurance that the true reactivity will be less than the limiting value of 0.55. These two independent methods are (1) AM3X-KENO calculations using the more recent 27-group SCALE cross-section set,6 and (2) the CASFO code,7 a twc dimensional transport theory code based on capture yrobabilities.

For investigation of small reactivity effects (e.g. , mechanical toler-ances), a four-group diffusion / blackness theory method of analysis (NULIF-

' CHR00-PDQ97) was used (Ref. 8) to calculate small incremental reactivity

. changes.

This model has been used previously with good results and is normally used only to evaluate trends and small incremental reactivity effects that would otherwise be lost in the XENO statistical variation. Where pos-sible, trends calculated by AFPX-KENO and by diffusion / blackness theory were compared and found to be in good agreement-5

SS-152 ~

, Page 9 of 19 10.50'

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FUEL ASSEMBLY: 15 x 15 Square Array 208 Fuel Rods 16 Control Rod Thimbles I in stru m e n t Tube l

Fig. 1 Reference design configuration.

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55-152

.Page 10'of 19 Table 2 FUEL ASSEltiLY DESIGN SPECIFICATIONS Fuel Rod Data Outside dimension, in. 0.430 Cladding thickness, in. ,

0.0265 Cladding material Zr-4 Pellet diameter, in. 0.369 UO2 density, g/cm 3 10.420 t 0.166 Enrichment, wt. % U-235 3.50 t 0.02 Fuel Assembly Data Number of fuel rods 208 (15 x 15 array)

Fuel rod pitch, in. O.568 -

Control rod guide tube Number 16 0.0., in. 0.530-Thickness, in. 0.016 Materi al Zr-4 -

Instrument thimtle Number 1 0.0.,in. 0.493

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g/ axial cm of assembly 46.14 0.78 j .g

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SS-152 Page 11 of 19 4.3 Calculational Bias and Uncertainty Results of benchmark calculations 5 on a series of critical experiments indicate a calculational bias of 0, with an uncertainty of t0.003 (95% proba-bility at a 95% confidence level). In addition, a small correction in the calculational bias may be necessary to account for the water-gap thickness be-tween fuel assemblies in the Crystal River spent fuel rack compared to the corresponding thickness in the benchmark critical experiments. Based upon the correlation developed in Ref. 5, the correction for water-gap thickness in the Crystal River spent fuel storage rack indicates a small overprediction. For conservatism, the overprediction is neglected and the net calculational bias is taken as 0.000 t 0.003.

4.4 Reference Fuel Storage Cell The nominal spent fuel storage cell model used for the criticality anal-yses is shown in Fig. 1. The rack is composed of boron carbide absorber material sandwiched between two 0.060-in. stainless-steel boxes. The fuel assemblies are centrally located in each storage cell on a nominal lattice spacing of 10.5 in. The outer water space constitutes a flux-trap between the two boron carbide absorber plates that are essentially opaque (black) to thermal neutrons. The absorber has a nominal thickness of 0.075 in. and a B-10 areal density of 0.015 t 0.003 gram per cm 2. For two-dimensional X-Y analysis, a zero current (white albedo) boundary condition was applied in the axial di rection and at the centerline through the outer water space (flux-trap) on all four sides of the cell, effectively creating an infinite array of storage cells.

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SS-152 Page 12'of 19 )

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5.0 REFERENCE SUBCRITICALITY AND MECHANICAL TOLERANCE VARIATIONS 5.1 Nominal Design Case Under normal conditions, with nominal dimensions, the calculated k, is 0.9347 0.0020 (la with 300 generations of 500 neutrons each or a total of 150,000 histories) for the nominal case. With a one-sided tolerance factor 9 of 1.80, corresponding to 95% probability at a 95% confidence limit with 300 generations, the maximum deviation of k, is t0.0036.

An independent KENO check calculation was made using the more recent 6

27-group SCALE cross-section library developed by ORNL for criticality safety analysis. This calculation yielded a k, of 0.925 0.004 which, when cor-rected for a 0.01 ok analytical bias, confirms the reference calculation with the 123-group GAM-THERM 0S library. A second independent calculation with the CASFO code yielded a k, of 0.939, which also r.onfirms the reference reactiv-ity calculation.

5.2 Boron Loading Variation The boron carbide absorber plates are nominally 0.075 in, thick with a B-10 areal density of 0.015 g/cm2. For the manufacturing tolerance of 0.003 grams 8-10 per cm2 in areal density, the calculated reactivity uncertainty is 0.0068 ak as calculated by diffusion / blackness theory. The trend calcu-lated by AMPX-KENO is slightly less (30.0048 ok) than with diffusion / blackness theo ry. However, for conservatism, the higher uncertainty value (0.0068 ak) was used for the reactivity uncertainty associated with the tolerance in B-10 areal density.

• Because of limitations on the geometrical representation of the absorber width permitted in CASFO, a correction of 0.025 ak, determined by dif fu-sion/ blackness theory, was necessary and is included in the value reported.

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55-152 Page 13 of Ig 5.3 Storage Cell lattice Pitch Variation The storage cell lattice spacing between fuel assemblies is approximately 10.5 in. A decrease in lattice spacing by decreasing the outer flux-trap water thickness increases reactivity, although decreasing the inner water thickness (between the fuel and the inner stainless-steel box) results in an almost negligible change in reactivity. Both of these effects have been evaluated for the independent design tolerances.

5.3.1 Inner Water Thickness Variations The inner stainless-steel box dimension , 8.937 in. defines the inner water thickness between the fuel and the inside box. For a tolerance limit of 0.032 in., the calculated uncertainty in reacti vity is 0.0001 ak, with k, decreasing as the inner stainless-steel box dimension (and derivative lattice spacing) increases.

5.3.2 Outer (Flux-Trap) Water Thickness Variation The outer (flux-trap) water thickness is nominally 1.173 1/16 in.,

which results in an uncertainty of 0.0064 ak due to the tolerance in flux-trap water thickness. Decreasing the flux-trap thickness increases reactiv-ity.

4 5.4 Stainless-Steel Thickness Variations The nominal thickness is 0.060 in. for both the inner and outer stain-less-steel boxes. The maximum positive reactivity effect of '_ the expected stainless-steel thickness tolerance variation ( 0.005 in.) was calculated to be ;0.0006 ak (by diffusion / blackness theory, since the reactivity increment is too small to be calculated by AWX-KENO).

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S5-152 Page 14 of 19 5.5 Fuel Enrichment 'and Density Variation The design maximum enrichment is 3.50 t 0.02 wt.% U-235. _ Calculations of the sensitivity to small enrichment variations by diffusion / blackness theory f ielded a coefficient of 0.0059 ak per 0.1 wt.% U-235 at the design enrich-ment, in the range from 3.4% to 3.6% enrichment. For the tolerance on U-235 enrichment of t0.02 in wt.%, the uncertainty on k, is 0.0012 ok.

3 Calculations were made with the maximum 002 fuel density (10.586 g/cm ),

resulting in an uncertainty in reactivity of 10.0018 a k over the expected tolerance in UO2 densities.

5.6 Absorber Width Tolerance Variation The reference storage cell design (Fig.1) uses a nominal absorber width of 6.687 t 0.0625 in. A positive i ncremer.t in reactivity occu rs for a decrease in absorber width. For the width tolerance of 0.0625 in., the maximum calculated reactivity increment is :0.0013 4 k. Increasing the absorber width decreases reactivity.

5.7 Summary of Statistical Variations Calculated reactivity increments from mechanical and fabrication toler-ances are summarized in Table 3.

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55-152 Page 15'of 19 Table 3 CALCULATED STATISTICAL VARIATIONS IN REACTIVITY (ECHANICAL)

Incremental Case Tolerance Reactivity, ok Baron concentration *0.003 g B-10/cm2  ; 0.0068 Lattice pitch Inner water thickness $1/32 in. T 0.0001 Outer water thickness t1/16 in. T 0.0064 i

SS tolerance t0.005 i n.

T 0.0006 Fuel enrichment to.02% U-235 t 0.0012 Fuel density t0.166 g/cm 3 t 0.0018 l 1

Absorber width t1/16'in.  ; 0.0013 Statistical combination to.0097 (root-mean-square of reactivity increments) I O

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. SS-152 Page 16 of 19 6.0 ABNORMAL AND ACCIDENT CONDITIONS 6 .1 Eccentric Positioning of Fuel Assembly in Storage Rack The fuel assembly is normally located ist the center of the storage rack cell. Nevertheless, calculations were made with adjacent fuel assemblies moved into the corner of the storage rack cell (four-assembly cluster at closest approach) resulting in a negative reactivity effect (0.0033 Ak) . Fuel assembly bowing will produce a small negative reactivity effect locally.

Thus, the nominal case, with the fuel assembly positioned in the center of the storage rack cell, yields the maximum reactivity.

6 .2 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 reacti vity effects calculated by diffusion / blackness theory) .

Introducing voids in the water internal to the storage cell (to simulate boili:ig) 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 TEMPERATURE AND V0ID ON CALCULATED REACTIVITY OF STORAGE RACK Ak a Case Comment 40*F (Reference) 0 Maximum water density 104 F (40 C) -0.005 o (H20) = 0.992 212*F (100 C) -0.017 p (H2 0) = 0.958 212 F with 50". void -0.232 Simulates boiling 13 L

55-152 Page 17'of 19 l 6.3 Fuel Assembly Abnormally Located Outside Storage Rack To investigate the possible reactivity effect of a fuel assembly abnor-mally located outside the rack, calculations were made for unpoisoned assem-blies separated only by water. Figure 2 shows the results of these calcu-lations. From these data, the reactivity (k.) will be less than 0.95 for any water-gap spacing greater than ~5 in, in the absence of any neutron-absorbing material other than water between assemblias.

A fuel assembly accidentally installed outside the rack cannot be posi-tioned closer than 6 in, from a rack module. With this minimum separation, the reactivity 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.

For a drop on top of the rack, the fuel assembly will come to rest hori-zontally on top of the rack with a minimum separation distance greater than 6 in. Consequently, fuel assembly drcp accidents will not result in an increase in reactivity above that calculated for the infinite nominal design storage rack.

l In both cases of a fuel assembly abnormally located outside the rack, local neutron leakage would further reduce the reacti vity effect of the abnormally-located fuel assembly. Furthermcre, 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 the criticality safety of the Crystal River spent fuel stor-age racks.

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. SS-152 Page 18 of 19

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SS-152 Page 19 of 19 REFERENCES

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

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

2. L. M. Petrie and N. F. Cross, KENO-IV, An Improved Nnte Carlo Criticality Program, ORNL-4938, Oak Ridge National Laboratory, November 1975.

3. S. R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.29 w t ". U235 Enriched U02 Rods in Water with Fixed Neutron Poisons, NUREG/CR-0073, Battelle Pacific Northwest Laboratories, May 1978, with errata sheet issued by the USNRC August 14, 1979.

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. S. E. Turner and M. K. Gurley, Evaluation of AWX-KENO Benchmark Calcu-lations for High Density Spent Fuel Storage Racks, Nuclear Science and Engineering,80(2): 230-237, February 1982.

6. R. M. Westfall et al., " SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200, 1979.

7. A. Ahli n , M. Edenius, CASMO - A Fast Transport Theory Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604,1977.

CASMO-2E Nuclear Fuel Assembly Analysis, Application Users Manual, Rev. A, Control Data Corporation, 1982.

8. W. A. Wittk opf , NULIF - Neutron Spectrum Generator, Few-Group Constant Generator and Fuel Depletion Code, BAW-426, The Babcock & Wilcox Company, August 1976.

S. E. Turner, CONROD - A Southern Science blackness theory routine to calculate equivalent diffusion-theory constants using transmission proba-bilities to effectively impose a transport-theory boundary condition at the surface of strong absorbers (unpublished).

W. R. Cadwell, PDQ-7 Reference Manual, WAPD-TM-678, Bettis Atomic Power Laboratory, January 1967

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