ML20045C184

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Criticality Analysis of Prairie Island,Units 1 & 2 Fresh & Spent Fuel Racks.
ML20045C184
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 02/28/1993
From: Bell R, Newmyer W, Savage C
NORTHERN STATES POWER CO.
To:
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ML20045C138 List:
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NUDOCS 9306220170
Download: ML20045C184 (42)
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!

l CRITICALITY ANALYSIS OF THE PRAIRIE ISLAND UNITS 1 & 2 FRESH AND SPENT FUEL RACKS February 1993 Authored: me W. D. Nedmyer [

Core Design B QO Verified: f--M f) yh R. M. 5eil Core Design C Approved: -

, y C. R. Savage, Manager Core Design B t

9306220170 930228 "

PDR P ADOCK C5000282 a PDR tf

TABLE OF CONTENTS 1.0 Introduction ............................................... 1 1.1 Design Description ................,................. 2 1.2 Design Criteria ..................................... 2 2.0 Analytical Methods ................................. ........ 3 2.1 Criticality Calculation Methodology ........................ 3 2.2 Reactivity Equivalencing for Burr.up ....................... 4 3.0 Criticality Analysis of Fresh Fuel Racks .......................... 6 3.1 Full Density Moderation Analysis ......................... 7 3.2 Low Density Optimum Moderation Analysis .................. 8 4.0 Spent Fuel Rack Criticality Analysis for the Region 1 Configuration . . . 10 4.1 KENO Reactivity Calculations ........................... 10 4.2 Burnup Credit Reactivity Equivalencing ..................... 12 4.3 Sensitivity Analysis ..........,...................... 13 4.4 Soluble Boron Worth ................................ 13 5.0 Spent Fuel Rack Criticality Analysis for the Region 2 Configuration . . . 14 5.1 Burnup Credit Reactivity Equivalencing ..................... 14 6.0 Storage Configuration interface Requirements ................... 16 7.0 Discussion of Postulated Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 7.1 Fresh Fuel Storage Racks ............................. 17 7.2 Spent Fuel Storage Racks ............................. 17 8.0 Summary of Criticality Results ................................ 19 Bibliography .................................................. 37 Table of Contents i

1 I

1 LIST OF TABLES l

l Table 1. Fuel Parameters Employed in the Criticality Analysis ....... 20 l Table 2. Benchmark Critical Experiments [5,6] ................. 21 Table 3. Comparison of PHOENIX lsotopics Predictions to Yankee Core 5 Measurements ................................. 22 Table 4. Benchmark Critical Experiments PHOENIX Comparison - ....... 23 Table 5. Data for U Metal and UOr Critical Experiments ........... 24 Table 6. Prairie Island Spent Fuel Storage Minimum Burnup Requirements 26 List of Tables il

  • ~' - +

1 i

i LIST OF ILLUSTRATIONS Figure 1. Prairie Island Fresh Fuel Storage Rack Axial Layout ........ 27 Figure 2. Prairie Island Fresh Fuel Rack Radial Layout ............. 28 Figure 3. Prairie Island Spent Fuel Storage Cell Nominal Dimensions ... 29 Figure 4. Prairie Island Spent Fuel Rack Layout ................. 30 Figure 5. Prairie Island Sensitivity of K.ev to Water Density in the Prairie Island Fresh Fuel Racks ...... ................... 31 Figure 6. Prairie Island Region 1 Burned / Fresh Checkerboard Cell Layout . 32 Figure 7. Prairie Island Region 1 Minimum Burnup Requirements ...... 33 Figure 8. Prairie Island Region 1 Reactivity Sensitivities ........... 34 Figure 9. Prairie Island Region 1 Spent Fuel Pool Soluble Boron Worth .. 35 Figure 10. Prairie Island Region 2 Minimum Burnup Requirements ...... 36 t

r List of Illustrations lii

l l

1.0 INTRODUCTION

This report presents the results of a criticality re-analysis of the Prairie Island Units 1 & 2 Fresh and Spent Fuel Storage Racks. The fresh rack designs con-sidered in this analysis are existing arrays of unpoisoned racks, previously qualified for storage of 14x14 fuel assemblies with enrichments up to 4.27 w/o U'". The spent rack designs are existing arrays of poisoned racks, previously qualified for storage of 14x14 fuel assemblies with enrichments up to 4.27 w/o 2n The Prairie Island fresh and spent fuel racks are being reanalyzed to allow storage of all 14x14 fuel assemblies with enrichments up to 5.0 w/o U'". Only one type of spent fuel storage rack exists at Prairie Island. Two different storage configurations are analyzed as discussed below. The following storage configurations and enrichment limits are considered in this analysis:

Fresh Fuel Rack Storage of fuel assemblies with nominal enrichments up to 5.0 w/o in selected locations, with no requirements for burnup or burnable absorbers.

Spent Fuel Rack Storage of " burned" and " fresh" fuel assemblies in a 2x2 Region 1 checkerboard pattern. Fuel assemblies stored in " burned" cell (Checkerboard) locations must have an initial enrichment less than 2.5 w/o (nominal) or satisfy a minimum burnup requirement. Fuel assemblies stored in the " fresh" cell locations can have enrichments up to 5.0 w/o with no requirements for burnup or burnable absorbers.

Spent Fuel Rack Storage of fuel assemblies which satisfy a minimum burnup Region 2 (Close requirement as a function of enrichment.

Packed)

The Prairie Island Units 1 & 2 Fresh and Spent Fuel Rack criticality analyses are based on maintaining K.tv $ 0.95 under full water density conditions and 5 0.98 under low water density (optimum moderation) conditions. The Spent Fuel Rack criticality analysis is based on maintaining K.et 5 0.95. Fuel types being con-sidered in the analyses include the Westinghouse 14x14 STD and OFA designs  ;

and the Exxon 14x14 fuel assembly types currently in storage in the Prairie is-land spent fuel pool.

t introduction 1 i l

1.1 DESIGN DESCRIPTION The Prairie Island fresh fuel rack layout is shown in Figure 1 on page 27 and Figure 2 on page 28. The Prairie tsiand spent fuel storage cell design is de-picted in Figure 3 on page 29 and Figure 4 on page 30, with nominal dimensions provided on each figure. A schematic of the spent fuel checkerboard pattern of bumed and fresh storage cell is given in Figure 6 on page 32.

1.2 DESIGN CRITERIA Criticality of fuel assemblies in a fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between fuel assemblies for the fresh and spent fuel racks and also inserting neutron poison between fuel assemblies for the spent fuel racks.

The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective neutron multiplication f actor, K.,f, of the fuel assembly array will be less than 0.95 as recommended by ANSI 57.2-1983, ANSI 57.3-1983 and Reference 1, or less than 0.98 under low water density (optimum moderation) conditions as recommended by NUREG-0800. The optimum moderation condition applies only to the Fresh Fuel Storage Rack since this rack is used to store fuel in a dry configuration.

Introduction 2

2.0 ANALYTICAL METHODS 2.1 CRITICALITY CALCULATION METHODOLOGY The criticality calculation method and cross-section values are verified by comparison with critical experiment data for fuel assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps and low moderator densities.

The design method which insures the criticality safety of fuel assemblies in the fuel storage rack uses the AMPX*

  • system of codes for cross-section gener-ation and KENO Va for reactivity determination.

The 227 energy -group cross-section library that is the common starting point for all cross-sections used for the benchmarks and the storage rack is generated from ENDF/B-V* data. The NITAWL* program includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry. The Nordheim Integral Treatment is used. Energy and spatial weighting of cross-sections is performed by the XSDRNPM* program which is a one-dimensional Sn transport theory code. These multigroup cross-section sets are then used as input to KENO Va which is a three dimensional Monte Carlo theory program designed for reactivity calculations.

A set of 44 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and uncertainty. The benchmark experiments cover a wide range of ge-ometries, materials, and enrichments, ranging from relatively low enriched (2.35, 2.46, and 4.31 w/o), water moderated, oxide fuel arrays separated by various materials (B4C, aluminum", steel, water, etc) that simulate LWR fuel shipping and storage conditions to dry, harder spectrum, uranium rneta! cylinder arrays at high enrichments (93.2 w/o) with various interspersed materials (Plexiglas and air).

Comparison with these experiments demonstrates the wide range of applicability of the method. Details of the experiments are provided in References 5 through

9. Table 2 on page 21 summarizes these experiments.

The highly enriched benchmarks show that the criticality code sequence can correctly predict the reactivity of a hard spectrum environment, such as the optimum moderation condition often considered in fresh rack and shipping cask analyses. However, the results of the 12 highly enriched benchmarks are not incorporated into the criticality method bias because the enrichments are well above any encountered in commercial nuclear power applications. Basing the method bias solely on the 32 low enriched benchmarks results in a more ap-propriate and more conservative bias.

Analytical Metheds 3

The 32 low enriched, water moderated experiments result in an average KENO Va K.if of 0.9933. Comparison with the average measured experimental K.ft of 1.0007 results in a method bias of 0.0074. The standard deviation of the bias value is 0.0013 AK. The 95/95 one-sided tolerance limit factor for 32 values is 2.20. Thus, there is a 95 percent probability with a 95 percent _ confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0029 ,

AK. This KENO Va bias and uncertainty are consistent with the previous '

Westinghouse bias and uncertainty calculated for KENO IV"*. ,

2.2 REACTIVITY EQUIVALENCING FOR BURNUP L Storage of spent fuel assemblies with initial enrichments higher than that shown to be acceptable by the methodology described in Section 2.1 is achievable by means of the concept of reactivity equivalencing. Reactivity equivalencing is -

predicated upon the reactivity decrease associated with fuel depletion. A series '

of reactivity calculations are performed to generate a set of enrichment-burnup ordered pairs which all yield an equivalent K.vf when the fuel is stored in the Prairie Island Units 1 & 2 spent fuel racks.

  • The data points on the reactivity equivalence curve are generated with a trans-port theory computer code, PHO ENI X"". PHOENIX is a depletable, two-dirnensional, multigroup, discrete ordinates, transport theory code. - A 25 energy group nuclear data library based on a modified version of the British WIMS""

library is used with PHOENIX.

A study was done to examine fuel reactivity as a function of time following

, discharge from the reactor. Fission product decay was accounted for using ' >

CINDER"". CINDER is a point-depletion computer code used to determine fission product activities. The fission products were permitted to decay for 30 years af ter discharge. The fuel reactivity was found to reach a maximum at approxi-mately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after discharge. At this time, the major fission product pol- l son, Xe'", has nearly completely decayed away. ' Furthermore, the fuel reactivity #

was found to decrease continuously from.100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 30 years following dis-charge. Therefore, the most reactive time for a fuel assembly after discharge-from the reactor can be conservatively approximated by removing the Xe'".

r The PHOENIX code has been validated by comparisons with experiments where -

the isotopic fuel composition has been examined following discharge from a - '

reactor. In addition, an extensive set of benchmark critical experiments has been i analyzed with PHOENIX. Comparisons between measured and predicted uranium ,

and plutonium isotopic fuel compositions. are shown in Table 3 on page 22. '

The measurements were made on~ fuel discharged from Yankee Core 5"*. The data in Table 3 on page 22 shows that the agreement between PHOENIX pred- '

ictions and measured isotopic compositions is good.

The agreement between reactivities computed with PHOENIX and the results of 81 critical benchmark experiments is summarized in Table 4 on page 23. Key parameters describing each of the 81 experiments are given in Table 5 on page Analytical Methods 4

t

24. These reactivity comparisons again show good agreement between exper-iment and PHOENIX calculations.

Uncertainties associated with the burnup reactivities computed with PHOENIX are accounted for in the development of the individual reactivity equivalence limits. An uncertainty is applied to the PHOENIX calculational results which starts at zero for zero burnup and increases linearly with burnup, passing through 0.01 AK at 30,000 MWD /MTU. This bias is considered to be very conservative and is based on consideration of the good agreement between PHOENIX pred- '

ictions and measurements and on conservative estimates of fuel assembly re-activity variances with depletion history. Additional information concerning the specific uncertainties included in each of the Prairie Island burnup credit limits j is provided in the individual sections of this report. '

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L l

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Analytical Methods 5

- . _. ~ . . . . - - . - _- _

4 3.0 CRITICALITY ANALYSIS OF FRESH FUEL RACKS This section describes the analytical techniques and models employed to per-form the criticality analysis for the storage of fresh _ fuel in the Prairie Island Units 1 & 2 Fresh Fuel. Storage Racks.

Since the fresh fuel racks are normally maintained in a dry condition, the criticality analysis will show that the rack K.tv is less than 0.95 for the accidental full water density flooding scenario and less than 0.98 for the accidental' low i water density (optimum moderation) flooding scenario. The c-iticality method-ology employed in this analysis is discussed in Section 2 of this report.

t The following assumptions were used to develop the KENO model'for the stor - t age of fresh fuel in the Prairie.lsland Units 1 & 2 Fresh Fuel Storage Rack under full density and low density optimum moderation conditions:

1. The limiting fuel assembly type is assumed for each condition.
2. The fuel assembly is modeled at its most reactive point in life. ,
3. All fuel rods contain uranium dioxide at an enrichment of 5.00 w/o (nominal) and 5.05 w/o (worst case) over the entire length of each rod. ,
4. The fuel pellets are modeled at 96% of . theoretical density without dishing or chamfering to bound the maximum fuel assembly loading. This as-sumption results in conservative calculations of reactivity for all fuel as-semblies in use at Prairie Island, including those which contain gadolinium ,

fuel rods.

5. No credit is taken for any natural or enriched axial blankets. This as- t sumption results in equivalent or conservative calculations of reactivity for all fuel assemblies in use at Prairie Island including those with annular -

pellets at the fuel rod ends.

6. No credit is taken for any U* or U* in the fuel, nor is any credit taken for the build up of fission product poison material.
7. .No credit is taken for any spacer grids or spacer sleeves.
8. No credit is taken for any burnable absorber in the fuel rods. l
9. For both the- full density and optimum moderation cases, there is no boron in the water. ,
10. Fuel rods are modelled with a fuel stack height-which is infinitely long for .

the full density moderation scenario and 144 inches long for the optimum I moderation scenario.

l l

1 Criticality Analysis of Fresh Fuel Racks 6 -l

_ _ _ _ _ _ _ - __ _ _ _ __ _ ~ . _ _ . . . . _ ,_ . _ _ - -

- _ . . _ . m . ._ . .

3.1 FULL DENSITY MODERATION ANALYSIS In the KENO model for the full' density moderation analysis, the moderator is pure water at a temperature of 68'F. A conservative value of 1.0 gm/cmis- '

used for the density of water. The fuel array is infinite in lateral.(x and y) and axial (vertical) extent which precludes any neutron leakage from the_ array.

The fuel assembly parameters assumed in the full- density analysis are based on the Westinghouse 14x14 OFA design (see Table 1 on page 20 for fuel pa -

rameters). Under fully flooded conditions and at the enrichment level considered in this evaluation, the OFA is the most reactive fuel type of the available 14x14 designs in use or storage at Prairie island.

The maximum K.tv under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting; from the manufacturing ,

process in addition to asymmetric positioning of . fuel assemblies with - the.

storage cells. Studies of asymmetric positioning of fuel essemblies within the ..

storage cells have shown that symmetrically placed fuel assemblies yield con-servative results in rack K.vi. Due to the relatively large cell spacing, the small tolerance on the cell center-to-center spacing is not considered since it will have an insignificant effect on the fuel rack reactivity. The most conservative, or

" worst case", KENO model of the fresh fuel storage racks contains no fuel rack .

steel with symmetrically placed fuel assemblies. Furthermore, the fuel enrichment is assumed to be 5.05 w/o U* to conservatively account for ,

enrichment variability.

Based on the analysis described above, the following equation'is used to de-velop the maximum K.tv for the Prairie island Units 1 & 2 Fresh Fuel Storage  ;

Racks:

K.eva K orsi + Bm.thoo + /[(ks)* wor., + (ks)'m.tnoo ]

where: ,

Kw rst = worst case KENO K.fi that includes material, mechanical and enrichment tolerances Bm.inoo = method bias determined from benchmark critical comparisons ks.or t = 95/95 uncertainty in the worst case KENO K.vf

  • ksmrinoo = 95/95 uncertainty in the method bias ,

Substituting calculated values in the order listed above, the result is:

K.tr = 0.9150 + 0.0074 + /[(0.0083)' + (0.0029)' ] = 0.9312 .

Since K.tv is less than 0.95 including uncertainties at a 95/95 probability confi-dence level, the acceptance criteria for criticality is met.

t Criticality Analysis of Fresh Fuel Racks 7

?

' ' * - ' ' d z- _-w. ._-__._m_ _ _ _ _ _ _ _ _.______m__ . _ _ - - _ _ _ _ _ _ _

3.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS For the low density optimum moderation analysis, the fuel array is finite in all directions. The " worst case" assumptions from the full density analysis are i used in modeling the entire fresh fuel rack array. Only 74 specific cells of the '

i 88 available storage locations are utilized as depicted in Figure 2 on page 28.

Fuel rods are modeled with a nominal active fuel length of 144 inches. Concrete walls and floor are modeled. Under low water density conditions, the presence of concrete is conservative because neutrons are reflected back into the fuel array more efficiently than they would be with just low density water. The area above the fresh fuel rack is filled with water at the optimum moderation density.

The fuel assembly parameters assumed in the low density analysis are based l on the Westinghouse 14x14 STD design (see Table 1 on page 20 for fuel pa-rameters). Under optimum moderation conditions, experience has shown that the 14x14 STD design is the most reactive fuel type of the available 14x14 designs. 1 This is because the STD fuel assembly contains a higher uranium loading than the other assemblies, and when optimum moderation conditions are present, higher loadings result in higher reactivity.

Analysis of the Prairie Island Units 1 & 2 Fresh Fuel Racks shows that the maximum rack K.ve under - low density moderation conditions occurs at 0.06 gm/cm' water density. The K.ve of the fresh rack at' O.06 gm/cm' water density is 0.9539 with a 95 percent probability and 95 percent confidence level oncer- ..

tainty of 10.0071. Figure 5 on page 31 shows the fresh fuel rack reactivity as a function of water density in the range where the optimum moderation peak occurs. Over the remainder of the density range between 0.'10 gm/cm' and 1.00 gm/cm' (fully flooded), the rack reactivity will be less than the values calculated for the optimum moderation peak and for the full density condition.

The following equation is used to develop the maximum K.ve for the Prairie is-

  • land Units 1 & 2 Fresh Fuel Storage Racks under low water density optimum moderation conditions:

K.ve = Kopomum + Bm.inoo + (((k s)'d... + (ks)*m.inoo ]

where:

l Kopumum = maximum KENO K.tv with low density optimum moderation ,

Bm.inoo = method bias determined from benchmark critical comparisons k s e... = 95/95 uncertainty in the base case KENO K.ve i

Criticality Analysis of Fresh Fuel Packs 8

1

)

l l

k sm.ined = 95/95 uncertainty in the method bias -l Substituting calculated values in the order listed above, the result is:

K.vf = 0.9539 + 0.0074 + (((0.0071)' + (0.0029)*. ] = 0.9690 Since K.it is .less than 0.98 including uncertainties . at a 95/95 probability / confidence level, the acceptance criteria for criticality is met. .

b l

r 0

4 Criticality Analysis of Fresh Fuel Racks 9

l 4.0 SPENT FUEL RACK CRITICALITY ANALYSIS FOR THE REGION 1 CONFIGURATION This section develops and describes the analytical techniques and models em-ployed to perform the criticality analysis and reactivity equivalencing evaluations for the Prairie Island Units 1 & 2 spent fuel racks Region 1 configuration using a checkerboard arrangement of burned and fresh fuel assemhiles. In this con-figuration, each fresh location is separated from the nearest adjacent fresh fuel location by a burned fuel assembly (see Figure 6 on page 32 for layout).

Section 4.1 describes the KENO reactivity calculations for the Region 1 check-erboard arrangement of burned and fresh fuel assemblies. For the KENO analy-sis, the " burned" fuel assemblies are represented by fresh fuel assemblies with a low enrichment. Section 4.2 describes the PHOENIX reactivity equivalencing analysis which establishes. the minimum burnup requirements of the " burned" fuel assemblies in the checkerboaru. Section 4.3 presents the results of calcu-lations performed to show the reactivity sensitivity of variations in enrichment, center-to-center spacing, and BORAFLEX poison loading. Finally, Section 4.4 discusses soluble boron worth calculations in the Region 1 configuration.

4.1 KENO REACTIVITY CALCULATIONS The following assumptions are used to develop the nominal case KENO model l for checkerboard storage of burned and fresh fuel assemblies in the Prairie is-land spent fuel rack (see Figure 6 on page 32 for layout).

t

1. The fuel assembly parameters relevant to the criticality analysis are based on the Westinghouse 14x14 OFA design (see Table 1 on page 20 for fuel parameters). Calcuktions for spent fuel rack analyses herein have shown that the Westinghouse OFA fuel assemblies yield a larger K.ft than the Westinghouse STD fuel assemblies for the same U* enrichment.
2. All burned fuel locations contain uranium dioxide at an enrichment of 2.50 w/o (nominal) and 2.55 w/o (worst case) over the entire length of each fuel rod. -
3. All fresh fuel locations contain uranium dioxide at an enrichment of 5.00 w/o (nominal) and 5.05 w/o (worst case) over the entire length of each fuel rod.
4. The fuel pellets are modeled at 96% of theoretical density without dishing i or chamfering to bound the maximum fuel assembly loading. This as-sumption results in conservative calculations of reactivity for all fuel as-Spent Fuel Rack Criticality Analysis for the Region 1 Configuration 10

. -. - _ - _ ~ _ _ . . - _ _ _ _ _ _ _

semblies in use at Prairie island, including those which contain gadolinium ,

fuel rods.

5. No credit is taken for any natural or enriched axial blankets. This as-sumption results in equivalent or conservative calculations of reactivity for all fuel assemblies in use at Prairie Island including those with annular  :

pellets at the fuel rod ends.

6. No credit is taken for any U* or U** in the fuel, nor is any credit taken for the build up of fission product poison material.
7. No credit is taken for any spacer grids or spacer sleeves.
8. No credit is taken for any burnable absorber in the fuel rods.
9. The moderator is pure water (no boron) at a temperature of 68*F. A limiting value of 1.0 gm/cm* is used for the density of water to conservatively bound the range of normal spent fuel pool water temperatures.
10. The minimum Boraflex poison material loading of 0.040 grams B" per square centimeter is used throughout the array. A four inch mid plane gap in the Boraflex absorber panels is assumed in the both the nominal and worst case ,

calculations to conservatively account for material degradation effects.

11. The array is infinite in lateral (x and y) extent and finite in axial (vertical) !

ext ent. This allows neutron leakage from only the axial direction. A re-flector of clean, full density, water is also placed at the top and bottom of the array to conservatively provide neutron reflection. ,

12. All available storage cells are loaded with fuel assemblies. Fuel assemblies are arranged in a checkerboard pattern of burned and fresh fuel locations, as depicted in Figure 6 on page 32.

The KENO calculation for the nominal case resulted in a K.ft of 0.9042 with a 95 percent probability /95 percent confidence level uncertainty of 10.0054. The nominal case result can be compared to the worst case result to determine the ,

relative impact of applying worst case assumptions. The nominal case is also used as the center point for the sensitivity analyses.

The maximum K.ve under normal conditions arises from consideration of me-chanical and material tolerances resulting from the manufacturing process. For the Prairie Island spent fuel racks Region 1 configuration, the sheet metal and Boraflex absorber tolerances are considered along with construction tolerances related to the cell I.D. and cell spacing. This resulted in a reduction of the center to center spacing and Boraflex material dimensions to their minimum  !

values. The water gap between cells is also reduced from a nominal dimension of 0.752" to a " worst case" dimension of 0.562". The assemblies are sym-metrically positioned within the storage cells since experience has shown that _ ,

centered fuel assemblies yield equal or more conservative results in rack K.et relative to asymmetric (non-centered) positioning. Furthermore, fuel enrichments are increased by 0.05 w/o U*" to conservatively account for enrichment vari-Spent Fuel Rack Criticality Analysis for the Region 1 Configuration 11

- .. --- ~ _ -.. ~..- - - - -_-

ability. Enrichments are assumed to be 2.55 w/o U " for the assemblies _ stored in the burned cell positions and 5.05 w/o U'" for assemblies stored in the fresh cell positions. Thus,' the most conservative or worst case" KENO.model of the Prairie Island spent fuel storage rack Region 1 configurations assumes minimum center to center spacings, minimum Boraflex material dimensions,.. minimum water gap thickness, and symmetrically placed fuel assemblies at their maximum authorized enrichments.

Based on the assumptions described above, the following equation is used to -

develop the maximum K.ef for the Prairie Island Units 1 & 2 spent fuel storage -

rack Region 1 configuration:

. K.vf = Kworsi + Dm.thoo + Bpari + /[ (ks)* worst + (ks)*m.inoo ]

where:

Kwort = worst case KENO K.vt that includes material, mechanical and enrichment tolerances Bm.iwoo = method bias determined from benchmark critical comparisons B,.ri r method bias to account for Boraflex poison particle self-shielding j k swor. = 95/95 uncertainty in the worst case KENO K.ft ,

ksm.thoo = 95/95 uncertainty in the method blas Substituting calculated values in the order listed above, the result is:

K.vf = 0.9307 + 0.0074 + 0.0010 + (((0.0049)* + (0.0029)* ] = 0.9448 i

Since K.fr is less than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met for the checkerboard arrangement, Region 1, of burned fuels cells at a nominal '

enrichment of 2.5 w/o and fresh fuel cells at a nominal enrichment of 5.0 w/o.

U'".

4.2 BURNUP CREDIT REACTIVITY EQUIVALENCING Storage of burned fuel assemb!!es with in the Prairie Island Region 1 config-uration is achievable by means of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with fuel depletion. For burnup credit, a serios of reactivity calcu-lations are performed to ' generate a set of enrichment-fuel assembly discharge i burnup ordered pairs which all yield an equivalent K.tv when stored in the' spent fuel storage racks.

Figure 7 on page 33 shows the constant K.ve contour generated for the Pralrie'  :

Island Region 1 configuration. Note the endpoints at 0 MWD /MTU where the enrichment is 2.5 w/o, and at 28,670 MWD /MTU where the enrichment is 5.0 w/o.

Spent Fuel Rack Criticality Analysis for the Region 1 Configuration 12

.- . . . . - . - . . . . . . . . ~

i The interpretation of this endpoint data is as follows: the reactivity'of the spent 1 fuel rack containing 5.0 w/o U'" fuel at 28,670 MWD /MTU burnup is equivalent to ' the reactivity of: the rack containing ' fresh. fuel having an initial nominal enrichment of 2.5 w/o. The burnup credit curve shown in Figure 7 on page 33 includes a reactivity uncertainty of 0.0096 AK, consistent with the minimum  ;

burnup requirement of 28,670 MWD /MTU at 5.0 w/o. l It is important to recognize that the curve in Figure 7 on page 33 is based on calculations of constant rack reactivity. In this way, the environment of the ,

storage rack and its influence on assembly reactivity is implicitly considered. ,

For convenience, the data from Figur,e 7 on page 33 is also provided in >

Table 6 on page 26. The tabulated values have been conservatively reported  ;

to allow the use of linear interpolation between the provided data points (the tabulated data may not match the figure).

4.3 SENSITIVITY ANALYSIS To show the dependence

  • of K.ft on fuel and storage cells parameters as _ re-quested by the NRC , the variation of the K.ft with respect to the following -

parameters was developed using the PHOENIX computer code:

1. Fuel enrichment, with a 0.50 w/o U'" delta about the nominal case enrichment. For this sensitivity, both the " fresh" and " burned" fuel as- ,

sembly enrichments were adjusted simultaneously. ' Note that both " fresh" and " burned" fuel assemblies are at zero burnup.

2. Center-to-center spacing of storage cells, with a half inch. delta about the nominal case center-to-center spacing.
3. Poison loading, with a 0.01 gm-B"/cm' delta about the nominal case poison loading.  ;

Results of the sensitivity analysis are shown in Figure 8 on page 34.

4.4 SOLUBLE BORON WORTH PHOENIX calculations were performed to evaluate the reactivity benefits 'of soluble boron for the Region 1 rack configuration. Results of these calculations are provided in Figure 9 on page 35. As the curves show, the presence of soluble boron in the Prairie Island spent kel pool provides substantial reactivity margin.

Spent Fuel Rack Criticality Analysis for the Region 1 Configuration 13 a

5.0 SPENT FUEL RACK CRITICALITY ANALYSIS FOR THE REGION 2 CONFIGURATION f

This section develops and describes the analytical techniques and models em--

ployed to extend the previous reactivity equivalencing evaluations

  • performed for the Prairie Island Units 1 & 2 spent fuel racks' Region 2 configuration as-

. suming close packed storage. Close packed storage is based on all cells having the same storage requirements and limits. With this type of storage, all cells can be utilized and no special cell position requirements are imposed.

Previous criticality analyses" performed for Prairie Island developed burnup credit curves for the spent fuel racks with 0, 2", and 4" mid-plane Boraflex gaps.

These curves represent lines of equivalent reactivity for each condition. The burnup credit curves were limited to enrichments below 4.27 w/o. In Section 5.1, the burnup credit curve for the 4" mid-plane Boraflex gap is extended using  ;

a reference equivalence reactivity from the previous analysis.

5.1 BURNUP CREDIT RIACTIVITY EQUIVALENCING Storage of fuel assemblies in the Region 2 configuration of the spent fuel racks was previously analyzed for Prairie Island with four inch mid-plane Boraflex gaps". A burnup credit curve was reported for U'" enrichments from 3.87 w/o to 4.27 w/o. The minimum discharge burnup required for a 3.87 w/o assembly was previously found to be 2000 MWD /MTU. The reactivity of such an assembly will be used as the basis for assemblies with enrichments between 4.27 w/o and 5.00 w/o U'". To calculate the burnup credit, a series of reactivity calculations are performed to generate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield a K.n equivalent to the 3.87 w/o U'" assembly with 2000 MWD /MTU burnup.

Figure 10 on page 36 shows the constant Ken contour generated for close packed storage in the Prairie island Region 2 spent fuel racks. Note the endpoints at 2000 MWDIMTU where 'he enrichment is 3.87 w/o, and at 10,640 MWD /MTU where the enrichment is 5.0 w/o. The interpretation of this endpoint data is as-follows: the reactivity of the spent fuel rack containing 5.0 w/o U'" fuel at 10,640 MWD /MTU burnup is equivalent to the reactivity of the rack containing fuel having an enrichment of 3.87 w/o at 2000 MWD /MTU burnup. The burnup credit curve shown in Figure 10 on page 36 includes a reactivity uncertainty of 0.0036 AK, consistent with the minimum burnup requirement of 10,640 MWD /MTU at 5.0 w/o.

It is important to recognize that the curve in Figure 10 on page 36 is based on '

calculations of constant rack reactivity. In this way, the environment of the t i

Spent Fuel Rack Criticality Analysis for the Region 2 Configuration 14 P

v 4 -

e e --r-- - -r-w - -~ - - _ . - - - - - _ _ _ - - - - - - - - _ _ _ - - _ - - - - . - - - - - - _ _ - - - - - -

storage rack and its ' influence on assembly reactivity is implicitly considered.

For convenience, the data from Figure 10 on page 36 is also provided on Table 6 on page 26. The. tabulated values have been conservatively reported to allow the use of linear interpolation between the provided data points (the ,

tabulated data may not match the figure). l s

i

+

s 1

i

! Spent Fuel Rack Criticality Analysis for the Region 2 Configuration 15 l

6.0 STORAGE CONFIGURATION INTERFACE REQUIREMENTS The Prairie Island Units 1 & 2 spent fuel racks are split into two different configurations designated as Region 1 and Region 2. The Region 1 configuration is for checkerboard storage, where neighboring cells have different requirements and limits. The Region 2 configuration is for close packed storage, where all.

cells share the same storage requirements and limits. A schematic of the Region 1 checkerboard pattern of burned and fresh storage celis is given in Figure 6 on page 32.

For the Region 1 configuration, the burned / fresh checkerboard zone can be po-sitioned anywhere within the spent fuel racks, but the interface boundaries shared with the Region 2 areas must be aligned as follows:

Region 1 checkerboard The boundary between the Region 1 checkerboard configuration next to the zone and the neighboring Region 2 close packed Region 2 close packed zone can be either separated by a vacant row configuration of cells or the interf ace must be configured such that there is a one row carryover of t' ) pattern of burned assemblies from the checkerboard zone into the first row of the close packed zone.

Figure 6 on page 32 lilustrates the interface configuration.

By employing the above boundary configuration controls, the pattern of fuel assemblies at the interface will not be more reactive than the patterns allowed on either side of the boundary.

Storage Configuration interface Requirements 16

7.0. DISCUSSION OF POSTULATED ACCIDENTS 7.1 FRESH FUEL STORAGE RACKS Under normal cond!tions, the fresh fuel racks are maintained in a dry environ-ment. The introduction of water into the fresh fuel rack area is the worst case accident scenario. The full density and low density optimum moderation cases analyzed in this report are the bounding accident situations which result in the most conservative fuel rack K.ir. ,

Other accidents can be postulated which would cause some reactivity increase (i.e., dropping a fuel assembly between the rack and wall or on top of the rack).

For these other accident conditions, the absence of a moderator in the fresh fuel storage racks can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event. Since the normal,: dry fresh fuel rack reactivity is less than 0.62 (see Figure 5 on page 31), there is sufficient reactivity margin to the 0.95 limit to cover the above postulated accidents.

i 7.2 SPENT FUEL STORAGE RACKS Most accident conditions will not result in an increase in K.if of the rack. Ex-amples are:

Loss of cooling systems Reactivity decreases since loss of cooling-causes an increase in temperature, which causes a decrease in water density, which results in decreased reactivity.

Fuel assembly dropped The rack structure pertinent for criticality is not horizontally on top of excessively deformed and the dropped assembly rack which comes to rest horizontally on top of the rack has sufficient water separating it from the active fuel height of stored assemblies to preclude neutronic interaction.

Fuel assembly dropped Design of spent fust rack is such that it between rack modules precludes the insertion 'of fuel assembly in other than prescribed locations.

Discussion of Postulated Accidents 17

- - . - y - , , , . .

.- . - - - . - - -. - - - -- .- ~ _ ~ . _ - - . _ ~ -- - - - _ _ .

l Fuel assembly dropped Design of spent fuel rack is such that it between rack modules precludes the insertion of fuel assembly in other and wall than prescribed locations.

However, accidents can be postulated which would increase reactivity. For these accident conditions, the double contingency principle of ANSI N16.1-1975 can  !

be applied. This states that one is not required to assume two unlikely, inde-pendent, concurrent events to ensure protection against a criticality accident.

Thus, for these postulated accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

A postulated accident that would increase reactivity beyond the analyzed con- >

dition in the Region 1 configuration is the misloading of a fuel assembly into a position for which the restrictions on location, enrichment, or burnup are not satisfied. To estimate the reactivity impacts of such an occurrence in the Re-gion 1 configuration, a calculation was run to determine the impact of mis-loading a number of fresh, 5.0 w/o fuel assemblies into burned cell locations such that 3 out of 4 locations would contain fresh, 5.0 w/o fuel assemblies.

The calculations show that reactivity would increase by only 0.04 AK. This result is conservative since, for simplification reasons, the calculational models actu-ally simulate the misloading of a single, fresh, 5.0 w/o fuel assembly into every 4 cell combination.

The worth of soluble boron in the Prairie Island spent fuel pool Region 1 con-figuration is shown in Figure 9 on page 35. As the curve shows, the presence of soluble boron in the pool water reduces rack reactivity significantly and is.

more than sufficient to offset the positive reactivity impact of a mistoad acci-dent. To bound the 0.04 AK reactivity increase in Region 1, it is conservatively estimated that 300 ppm will be required.

Therefore, should a postulated accident occur which causes a reactivity increase, Kort will be maintained less than or equal to C.95 due to the effect of dissolved ,

boron.

P 1

Discussion of Postulated Accidents 18 ,

8.0

SUMMARY

OF CRITICALITY RESULTS The acceptance criteria for criticality requires the effective neutron multiplication f actor, Keve, in the fresh fuel storage rack to be less than or equal to 0.95, in-cluding uncertainties, under flooded conditions and less than or equal to 0.98, including uncertainties, under optimum moderation conditions. For the spent fuel racks, K.ve must be maintained less than 0.95, including uncertainties, for all conditions.

This report shows that the acceptance criteria for criticality is met for the Prairie Island Units 1 & 2 fresh and spent fuel Region 1 and 2 storage racks for the storage of Westinghouse 14x14 fuel assemblies and the Exxon 14x14 fuel assembly types currently in storage in the Prairie Island spent fuel pool with the following configurations and enrichment limits:

Fresh Fuel Rack Storage of assemblies with nominal enrichments up to 5.0 w/o in selected locations as shown in Figure 2 on page 28.

Region 1 Storage of " burned" and " fresh" fuel assemblies in a 2x2 Checkerboard checkerboard pattern as shown in Figure 6 on page 32. Fuel Configuration assemblies stored in " burned ' cell locations must _ have an initial enrichment less than 2.5 w/o (nominal) or satisfy the minimum burnup requirements of Figure 7 on page 33. Fuel assemblies stored in the " fresh" cell locations can have nominal enrichments up to 5.0 w/o, with no requirements for a minimum accumulated burnup.

Region 2 Close Storage of fuel assemblies which satisfy the Region 2 Packed minimum burnup requirements shown in Figure 10 on page Configuration 36. For these assemblies, there are no special placement restrictions.

The analytical methods employed herein conform with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

Section 5.7, Fuel Handling System; ANSI 57.2-1983, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2; ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety";

NRC Standard Review Plan, Section 9.1.2, S p e nt Fuel Storage"; and ANSI 57.3-1983, " Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants."

Summary of Criticality Results 19

Table 1. Fuel Parameters Employed in the Criticality Analysis Parameter W 14x14 STD W 14x14 OFA Number of fuel Rods per Assembly ,

179 179 Rod Zirc-4 Clad 0.D. (i nch) 0.422 0.400 Clad Thickness (i nch) 0.0243 0.0243 Fuel Pellet 0.D. (inch) 0 3659 0 3444 Fuel Pellet Density ,

(% of Theoretical) 96 96 Fuel Pellet Dishing Factor 0.0 0.0 -

Rod Pitch (inch) 0 556 0 556 Number of Zirc-4 Guide Tubes 16 16 Guide Tube 0.D. (i nch) 0 539 0 526 Guide Tube Thickness (inch) 0.017 0.017 Number of Instrument Tubes 1 1 Instrument Tube 0.D. (inch) 0.422 0 399 Instrument Tube Thickness (inch) 0.0240 0.0235 l

1 Summary of Criticality Results 20 l

l

t Critical General Enrichment Separating Soluble Peasured KENO Reactivity Nurrber Description U235 (w/o) Reflector Material Baron (ppm) Keff (Keff +/- One Sigma) g 1 UO2 Rod Lattice 2.4B water water O 1.0002 0.9960 +/- 0.0024 g 3 2 D02 Rod Lattice 2.48 water water 1037 1.0001 0.9914 +/- O.On19 m R 3 UO2 Rod Lattice 2.46 Water water 764 1.0000 0.9943 +/- 0.0019 ,ro

< 4 UO2 Rod Lattice 2.4S water B40 pins O O.9999 0.9871 +/- 0.0022 0 5 UO2 Rod Lattice 2.48 water B4C pins 0 1.0000 0.9902 +/- 0.0022 to 6 U02 Rod Lattice 2.46 water B4C pins 0 1.0097 0.9948 +/- 0.0021 @

1 7 UO2 Rod Lattice 2.46 water B4C pins O O.9998 0.9888 +/- 0.0021 n O. 8 U02 Rod Lattice 2.46 water B4C pins 0 1.0083 0.9973 +/- 0.0021 g C 9 UO2 Rod Lattice 2.46 water water O 1.0030 0.9968 +/- 0.0021 na E 10 UO2 Rod Lattice 2.48 water water 143 1.0001 0.9973 +/- 0.0021 5-

< 11 002 Rod Lattice 2.40 water stainless steel 514 1.0000 0.9992 +/- 0.0020 O y 12 UO2 Rod Lattice 2.48 water stainless steel 217 1.0000 1.0031 +/- 0.0021  %

m 13 UO2 Rod Lattice 2.46 water borated aluminum 15 1.0000 0.9939 +/- 0.0022 o-h 14 002 Rod Lattice 2.48 water borated aluminum 92 1.0001 0.9882 +/- 0.0022 E.

15 UO2 Rod Latttee 2.48 water borated aluminum 395 0.9998 0.9854 +/- 0.0021 m 16 UO2 Rod Lattice 2.46 water borated aluminum 121 1.0001 0.9848 +/- 0.0022 g 17 002 Rod Lattice 2.48 water borated aluminum 487 1.0000 0.9392 +/- 0.0021 $

18 U02 Rod Lattice 2.46 water borated aluminum 197 1.0002 0.9944 +/- 0.0022 y 19 UO2 Rod Lattice 2.46 water borated aluminum 634 1.0002 0.9956 +/- 0.0020 2 20 U02 Rod Lattice 2.48 water borated aluminum 320 1.0003 0.9893 +/- 0.0022 3 21 UO2 Rod Lattice 2.46 water borated aluminum 72 0.9997 0.9900 +/- 0.0022 22 002 Rod Lattice 2.35 water borated aluminum O 1.0000 0.9980 +/- 0.0024 2 23 002 Rod Lattice 2.35 water stainless steel 0 1.0000 0.9933 +/- 0.0022 'c3 24 U02 Rod Lattice 2.35 water water O 1.0000 0.9920 +/- 0.0024 25 U02 Rod Lattice 2.35 water stainless steel 0 1.0000 0.9877 +/- 0.0022 26 U02 Rod Lattice 2.35 water borated aluminum O 1.0000 0.9912 +/- 0.0022 27 U02 Rod Lattice 2.35 water B4C O 1.0000 0.9921 +/- 0.0021 28 UO2 Rod Lattice 4.31 water stainless steel O 1.0000 0.9988 +/- 0.0023 29 U02 Rod Lattice 4.31 Water water O 1.0000 0.9963 +/- 0.0025 30 UO2 Rod Lattice 4.31 water stainless steel 0 1.0000 0.9950 +/- 0.0026 31 UO2. Rod Lattlee 4.31 water borated aluminum O 1.0000 0.9952 +/- 0.0025 32 UO2 Rod Lattice 4.31 water borated aluminum O 1.0000 1.0006 +/- 0.0024 33 U-metal Cylinders 93.2 bare air O 1.0000 0.9988 +/- 0.0023 34 U-metal Cylinders 93.2 bare ate 0 1.0000 1.0082 +/- 0.0025 35 U-metal Cylinders 93.2 bare air O 1.0000 0.9935 +/- 0.0024 36 U-metal Cylinders 93.2 bare air O 1.0000 0.9982 +/- 0.0028 37 U-metal Cylinders 93.2 bare air O 1.0000 0.9916 +/- 0.0025 38 U-metal Cylinders 93.2 bare air O 1.0000 0.9922 +/- 0.0025 39 U-metal Cylinders 93.2 bare plexiglass 0 1.0000 0.9972 +/- 0.0025 3 40 U-metal Cy1Inders 93.2 paraffin plexiglass 0 1.0000 0.9973 +/- 0.0029 41 U-metal Cylinders 93.2 bare plexiglass 0 1.0000 1.0019 +/- 0.0027 42 U-metal Cylinders 93.2 paraffin plexiglass 0 1.0000 1.0103 +/- 0.0025 43 U-metal Cylinders 93.2 paraffin plexiglass 0 1.0000 1.0021 +/- 0.0028 44 U-metal Cylinders 93.2 paraffin plexiglass 0 1.0000 1.0022 +/- 0.0029

i l

Table 3. Comparison of PHOENIX isotopics Predictions to Yankee Core 5 Measurements I

Quantity (Atom Ratio)  % Difference U235/U -0.67 U236/U -0.28 U238/U -0.03 Pu239/U + 3.27 Pu240/U + 3.63 Pu241/U -7.01 Pu242/U -0.20 Pu239/U238 +3.24 Mass (Pu/U) + 1.41 FiSS-Pu/ TOT-Pu -0.02 Summary of Criticality Results 22

. . . - . . m. . . . . _ . - _ _ . - . -

Table 4. Benchmark Critical Experiments PHOENIX Comparison j l

Description of Number of ' PHOENIX k.n Using Experiment l Experiments Experiments Bucklings .

t UO j Al clad 14 0.9947 SS clad 19 0.9944-  !

Borated H2O 7 0.9940 Subtotal 40 0.9944 U-Metal Al clad 41 1.0012 t

TOTAL 81 0.9978 ,

i

[

, l I

Summary of Criticality Results 23 l

~

l

Fuel Pellet Clad Clad Lattice

[ Case Cell A/O H20/U Density Diameter Material OD Thickness Pitch Boron a 3 Number Type U-235 Ratio (G/CC) (CN) Clad (CM) (CM) (CM) PPM E 3 ------------------------------------------------------------------------------------------------------ *

! R 1 Hexa 1.328 3.02 7.53 1.5265 Aluminum 1.6916 .07110 2.2050 0.0 ui

< 2 Hexa 1.328 3.95 7.53 1.5265 Aluminum 1.6916 .07110 2.3590 0.0 S 3 Hexa 1.328 4.95 7.53 1.5265 Aluminum 1.6916 .07110 2.5120 0.0 C g 4 Hexa 1.328 3.92 7.52 .9855 Aluminum 1.1506 .07110 1.6580 0.0 5 Hexa 1.328 4.89 7.52 .9855 Aluminum 1.1506 .07110 1.6520 0.0 3 6 Hexa 1.328 2.88 10.53 .9728 Aluminum 1.1506 .07110 1.5580 0.0 0

g-e_s 7 Hexa 1.328 3.58 10.53 .9728 Aluminum 1.1506 .07110 1.6520 0.0 A

' C 8 Hexa 1.328 4.83 10.53 .9728 Aluminum 1.1506 .07110 1.8060 0.0 i 9 Square 2.734 2.18 10.18 .7620 S5-304 .8594 .04085 1.0287 - 0. 0 E

[ 10 Square 2.734 2.92 10.18 .7620 55-304 .B594 .04085 1.1049 0.0 h ,

y 11 Square 2.734 3.86 10.18 .7620 SS-304 .8594 .04085 1.1938 0.0 - '

= 12 Square 2.734 7.02 10.18 .7620 55-304 .8594 .04085 1.4554 0.0 $

  • 55-304 .8594 1.5621 13 Square 2.734 8.49 10.18 .7620 .04085 0.0 CL 14 Square 2.734 10.38 10.18 .7620 5S-304 .8594 .04085 1.6891 0.0 C 15 Square 2.734 2.50 10.18 .7620 SS-304 .B594 .04085 1.0617 0.0 O is Square 2.734 4.51 10.18 .7620 SS-304 .8594 .04085 1.2522 0.0 g 17 Square 3.745 2.50 10.27 .7544 SS-304 .8600 .04060 1.0617 0.0 *2.

18 Square 3.745 4.51 10.37 .7544 S5-304 .8600 .04000 1.2522 0.0  %

19 Square 3.745 4.51 10.37 .7544 SS-304 .8000 .04060 1.2522 0.0 og, 20 Square 3.745 4.51 10.37 .7544 SS-304 .8000 .04060 1.2522 456.0 m 21 Square 3.745 4.51 10.37 .7544 S5-304 .8600 .04060 1.2522 709.0 y 22 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1260.0 e 23 Square 31745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1334.0 g 24 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1477.0 m 25 Square 4.069 2.55 9.46 1.1278 SS-304 1.2090 .04000 1.5113 0.0 3 26 Square 4.089 2.55 9.46 1.1278 SS-304 1.2090 .04000 1.5113 3392.0 58 27 Square 4.069 2.14 9.46 1.1278 SS-304 1.2090 .04060 1.4500 0.0 _

28 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 0.0 y 29 Square 3.037 2.64 9.28 1.1268 SS-304 1.1701 .07163 1.5550 0.0 2 30 Square 3.037 8.16 9.28 1.1268 55-304 1.2701 .07183 2.1980 0.0 ,

31 Square 4.069 2.59 9.45 1.1268 SS-304 1.2701 .07183 1.5550 0.0 g 32 Square 4.069 3.53 9.45 1.1268 SS-304 1.2701 .07103 1.6840 0.0 +

33 Square 4.069 8.02 9.45 1.1268 SS-304 1.2701 .07163 2.1980 0.0 3 34 Square 4.069 9.90 9.45 1.1268 SS-304 1.2701 .07163 2.3810 0.0 35 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 1677.0 36 Hexa 2.096 2.06 10.38 1.5240 Aluminum 1.0916 .07112 2.1737 0.0 g 37 Hexa 2.093 3.09 10.38 1.5240 Aluminum 1.6916 .07112 2.4052 0.0 h 38 Hexa 2.096 4.12 10.38 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 39 Hexa 2.096 6.14 10.38 1.5240 Aluminum 1.6916 .07112 2.9891 0.0 40 Hexa 2.096 8.20 10.38 1.5240 Aluminum 1.6916 .07112 3.3255 0.0 41 Hexa 1.307 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 42 Hexa 1.307 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0

i Fuel Pellet Clad Clad Lattice

[ Case Cell A/O H20/U Density Diameter Material DD Thickness Pitch Boron h 3 Nurnber Type U-235 Ratto (G/CC) (CM) Clad (CM) (CM) (CM) PPM E g ______________________________________________________________________________________________________ m

$ 43 Hexa 1.307 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 ut

< 44 Hexa 1.307 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 1 45 Hexa 1.307 4.02 18.90 1.5240 Aluminum 1.8916 .07112 3.3249 0.0 C Hexa Aluminum g

2.

46 47 Hexa 1.160 1.160 1.01 1.51 18.90 18.90 1.5240 1.5240 Aluminum 1.6916 1.6916

.07112

.07112 2.1742 2.4054 0.0 0.0

(+

y 48 Hexa 1.160 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 0 E 49 Hexa 1.160 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 g W 50 Hexa 1.160 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 51 Hexa 1.040 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 5

[ 52 Hexa 1.040 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0 h

  • 53 Hexa 1.040 2.02 18.90 1.5240 Aluminum 1.S916 .07112 2.6162 0.0 -

b 54 Hexa 1.040 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 $

55 Hexa 1.040 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 C.

50 Hexa 1.307 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 C 57 Hexa 1.307 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 0 58 Hexa 1.307 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 o 59 Hexa 1.307 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 2.

l 60 Hexa 1.307 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 [---

61 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 m_

i 62 Hexa 1.160 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 m l 63 Hexa 1.160 3.02 18.90 . 9 f. *30 Aluminum 1.1506 .07112 1.9609 0.0 y 64 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 m 65 Hexa 1.160 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 5-66 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 $

j 67 Hexa 1.160 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 Q l 68 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 m 69 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 70 Hexa 1.040 1.33 18.90 19.050 Aluminum 2.0574 .07620 2.8687 0.0 71 Hexa 1.040 1.58 18.90 19.050 Aluminum 2.0574 .07620 3.0086 0.0

[

72 Hexa 1.040 1.83 18.90 19.050 Aluminum 2.0574 .07620 3.1425 0.0 g 73 Hexa 1.040 2.33 18.90 19.050 Aluminum 2.0574 .07620 3.3942 0.0 g 74 Hexa 1.040 2.83 18.90 19.050 Aluminum 2.0574 .07620 3.6284 0.0

  • 75 Hexa 1.040 3.83 18.90 19.050 Aluminum 2.0574 .07820 4.0566 0.0 M 76 Hexa 1.310 2.02 18.88 1.5240 Aluminum 1.6916 .07112 2.6160 0.0 77 Hexa 1.310 3.01 18.88 1.5240 Aluminum 1.6916 .07112 2.9900 0.0 78 Hexa 1.159 2.02 18.88 1.5240 Aluminum 1.6916 .07112 2.8160 0.0 g 79 Hexa 1.159 3.01 18.88 1.5240 Aluminum 1.6916 .07112 2.9900 0.0 Ut 80 Hexa 1.312 2.03 18.88 .9830 Aluminum 1.1506 .07112 1.7250 0.0 81 Hexa 1.312 3.02 18.88 .9830 Aluminum 1.1506 .07112 1.9610 0.0

Table 6. Prairie Island Spent Fuel Storage Minimum Burnup Requirements Region Enrichment Burnup (w/o) (MWD /MTU)

Region 1 25 o 30 7400 4.0 18800 50 28800 Region 2 3.87 2000 4.07 3800 4.27 5200 4.40 6400 4.60 7900 4.80 9300 5 00 10800 Note: The minimum burnup requirements in this table have been conservatively reported to allow the use of linear interpolation between burnups.

Summary of Criticality Results 26

i 1

- )

i I

sc.t s ,p r o-

~

5 s s vis w 0' h, ,.,. ip+.,, .. .,j

- - - - - - - - - - ~ ~ ~ - - - -

Ob WO"

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.at ,,,y. irs ,,,, _ . , , , i f S'

  • 4 %I fd*Latt 1

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l .3a- t -

Figure 1 Prairie Island Fresh Fuel Storage Rack Axial Layout Summary of Criticality Results 27

i I

i l

A

.  :+ .

M

,M 7777777. .  ::

s I

a w I e t l

_ _ X X X X X X X _. _ _

l 9

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- - - - - - - - - -

  • I s,51", .

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10,, 10 SPACES @ l'- Y' = 17'- 6"

,10" s n n <

l 19' - T' \ ,

1

\ #

i mm.

h l Available Blocked l Cell Cell l

I Figure 2 Prairie Island Fresh Fuel Rack Radial Layout Summary of Criticality Results 28

1 d--

\

8.20" .

d

, 0.752"!

- 1

-}-

l S.27" ,

)

, il

,. ,p ,

=

O.090" Inner Casing cell Center to Center (9.5")

0.125" M r 0.024" Outer CasL'q Figure 3 Prairie Island Spent Fuel Storage Cell Nominal Dimensions Summary of Criticality Results 29

- o a

co M

/ t N ih 3'

~ . .

~ ~ ~

~ ~

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E

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~

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r Figure 4 Prairie island Spent Fuel Rack Layout Summary of Criticality Results 30

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REGION 1 CHECKERBOARD PATIERN BE E EB EE B gggljji.

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Region 2 BumedFuel: Must satisfy the minimum burnup Requirements of Figure 10.

Note: The Region 1 and Region 2 zones can attematively be seperated by a single row of vacant cells on each adjacent face.

Figure 6 Prairie Island Region 1 Burned / Fresh Checkerboard Cell Layout Summary of Criticality Results 32 s-

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Figure 7 Prairie Island Region 1 Minimum Burnup Requirements Summary of Criticality Results 33

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.03 .05 BORAFLEX B 10 04 LOADlHG(g/ce) 2 Figure 8 Prairie Island Region 1 Reactivity Sensitivities Summary of Criticality Results 34

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MMINAL U-235 DRiotENT (f 0)

Figure 10 Prairie Island Region 2 Minimum Burnup Requirements Summary of Criticality Results 36

BIBLIOGRAPHY

1. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications., April 14, 1978.
2. W. E. Ford lli, CSRL-V: Processed ENDFIB-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORNLICSDITM-160, June 1982.
3. N. M. Greene, AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDFIB, ORNLITM-3706, March 1976.
4. L. M. Petrie and N. F. Landers, KENO Va-- An Improved Monte Carlo Criticality Program With Supergrouping, NUREGICR-0200, December 1984.
5. M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, July 1979. ,
6. S. R. Bierman and E. D. Clayton, Criticality Separation Between Subcritical  ;

Clusters of 2.35 wt% 235U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2438, October 1977.

7. S. R. Bierman and E. D. Clayton, Crlucality Separation Between Subcritical Clusters of 4.29 wt% 235U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2615, August 1979.
8. S. R. Bierman and E. D. Clayton, Criticality Experiments with Subtritical Clusters of 2.35 wt% and 4.31 wt% 235U Enriched UO2 Rods in Water at a Water-to-Fuel Volume Ratio of 1.6, PNL-3314 July 1980.
9. J. T. Thomas, Critical Three-Dimensional Arrays of U(93.2) Metal Cylinders, Nuclear Science and Engineering, Volume 52, pages 350-359,1973.
10. D. E. Mueller, W. A. Boyd, and M. W. Fecteau (Westinghouse NFD), Qualification of KENO Calculations with ENDFIB-V Cross Sections.

American Nuclear Society Transactions, Volume 56, pages 321-323, June 1988.

11. A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Bolling Water Reactors, WCAP-10106, June 1982.
12. Askew, J. R., Fayers, F. J., and Kemshell, P. B., A General Description of the Lattice Code WIMS, Journal of British Nuclear Energy Society, 5, p p.

564-584, 1966.

i Bibliography 37

v '

l 1

l

13. Engfand, T. R., CINDER - A One-Point Depletion and Fission Product Program, WAPD-TM-334, August 1962.

i

14. Melehan, J. B., Yankee Core Evaluation Program Final Report, _

WCAP-3017-6094, January 1971.  ;

1

15. Boyd, W. A., et al., Criticality Analysis of Pralife Island Units 1 and 2 Fuel )

Racks,, March 1989.

l q

i

~

i 2

b Bibliography 38

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