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Criticality Analysis of Shearon Harris Spent Fuel Racks W/ Ifba Fuel
	ML18005A852
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Issue date:	11/30/1988
From:	Boyd W, Fecteau M, Schmidt R; CAROLINA POWER & LIGHT CO.
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CRITICALITYANALYSIS OF SHEARON HARRIS SPENT FUEL RACKS WITH IFBA FUEL DUPLICATE November 1988 W. A. Boyd R. F. Schmidt M. W. Fecteau W. A. Bordogna BS'04240240 8904ii PDR ADQCK 05000400 P

PNU

TASLE OF CONTENTS 1.0 Introduction 1.1 1.2 Design Description Design Criteria

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1 2.0 Criticality Analysis 2.1 Reactivity Equivalencing 2.1.1 Analytical Methods 2.1.2 Reactivity Calculations 2.2 Infinite Multiplication Factor 2.2.1 Reactivity Calculations 2.3 Postulated Accidents

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o 6 3.0 Acceptance Criterion For Criticality...............,.............

8 Ibliography..................................................

'17 B

Table of Contents

LIST OF TABLES Table

1. Fuel Parameters Employed in Criticality Analysis Table

2. Shearon Harris Fuel Assembly Minimum IFBA rods vs Initial U' 'nrichment for Region 1 Spent Fuel Rack Table

3. Comparison of PHOENIX Isotopics Predictions to Yankee Core Measurements Table

4. Benchmark Critical Experiments PHOENIX Comparison Table

5. Data for U Metal and UO~ Critical Experiments 9

10 5

ll 12 13 List of Tables

LIST OF ILLUSTRATIONS Figure

1. Shearon Harris Spent Fuel Storage Cell Nominal Dimensions 15 Figure

2. Shearon Harris Fuel Assembly Minimum Number of IFBA Rods vs. Initial U' 'nrichment for Storage in Region 1 Spent Fuel Racks

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16 List of Illustrations

1.0 INTRODUCTION

The Shearon.Harris spent fuel rack'(SFR) design described

herein, referred to as Region 1, is designed on the basis of the currently accepted NRC guidance on spent fuel rack design.

The Region 1 spent fuel rack design is a poisoned rack, previously analyzed for the storage of Westinghouse 17x17 OFA and STD fuel assemblies with enrichments up to 4.2 w/o U' 'tilizing every storage location in the fuel rack array.

This criticality analysis has been performed to show that nominal 5.0 wlo 17x17 OFA and STD fuel assemblies with Integral Fuel Burnable Absorbers (IFBA's) can be stored in every cell location in the fuel rack and maintain Kefr 5 0.95.

This analysis is a supplement to the original analysis' 'nd does not replace the original analysis.

The fuel assembly IFBA's consist of a

neutron absorbing material (i.e.

gadolinium or boron) homogeneously mixed with the fuel pellet or applied as a thin coating on the outside of the fuel pellet.

As a result, the neutron ab-sorbing material is a non-removable or integral part of the fuel assembly once it is manufactured.

1.1 DESIGN DESCRIPTION The Region 1 spent fuel storage cell design is depicted schematically in Figure 1 with nominal dimensions given on the figure.,

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 assemblies and inserting neutron poison between assemblies.

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 multiplication factor (Ken) of the fuel assembly array will be less than 0.95 as recommended in ANSI 57.2-1983, and in Reference 1.

Introduction

2.0 CRITICALITYANALYSIS This section develops and describes the analytical techniques and models em-ployed to perform the criticality analyses for storage of spent fuel in the Shearon Harris spent fuel pool above a nominal 4.2 w/o U' 'ith Integral Fuel Burnable Absorbers (IFBA's).

Two analytical techniques are used to establish the criticality criteria for the storage of IFBA fuel in the fuel racks.

The first method uses reactivity equivalencing to establish the poison material loading required to meet the criticality limits.

The-poison material considered in the analysis is a zirconium diboride (ZrBz

) coating manufactured by Westinghouse.

The second method uses the fuel assembly infinite multiplication factor to establish a reference reactivity.

The reference reactivity point is compared to a fuel assemblies peak reactivity to determine its acceptability for storage in the spent fuel racks.

2.1 REACTIVITY EQUIVALENCING Spent fuel storage above a nominal 4.2 w/o in the Shearon Harris Region 1 spent fuel storage

racks, is achievable by means of the concept of reactivity equiv-alencing.

The concept of reactivity equivalencing is predicated upon the reac-tivity decrease associated with the addition of IFBA fuel -rods and fuel depletion.

A series of reactivity calculations are performed to generate a set of IFBA rod number versus enrichment ordered pairs which all yield the equivalent KaH when the fuel is stored in the spent fuel racks, The fuel burnup used in the reactivity calculation is that burnup which yields the highest equivalent Kerf when the fuel is stored in the spent fuel racks.

Fuel assembly depletions performed in PHOENIX and the Westinghouse licensed core design codes show that for the number of IFBA rods per assembly considered in this analysis, the maximum reactivity occurs at zero burnup.

Although the boron concentration in the IFBA rods decreases with fuel depletion, the fuel assembly reactivity decreases more rapidly resulting in a maximum fuel rack reactivity at zero burnup.

The following assumptions were used for the IFBA rod assemblies in the PHOENIX models:

Criticality Analysis

1~

Calculations for spent fuel racks similar to the Region 1 racks analysis herein have shown that the W 17x17 OFA fuel assemblies yields a

larger Ke<< than does tQe W 17x17 Standard fuel assembly when both fuel assem-blies have the sane Unrichment.

Thus, only the W 17x17 OFA fuel assembly was analyzed for Region 1

(See Table 2 for fuel parameters).

2.

The moderator is pure water at a temperature of 68'F.

A conservative value of 1.0 gm/cm's used for the density of water.

3.

No credit is taken for any spacer grids or spacer sleeves.

4.

The IFBA absorber material is a zirconium diboride (ZrBz

) coating on the fuel pellet.

5.

Each IFBA rod has a minimum poison material loading of 0.0015 grams B-10 per inch.

6.

The B-10 loading is reduced by 25 percent in each IFBA rod to conservatively model a minimum poison length of 108 inches.

Figure 2 shows the constant Ke<<contour generated for the Shearon Harris spent fuel racks.

Note in Figure 2 the endpoint at 0 IFBA rods where the enrichment is 4.2 w/o and at 48 IFBA rods where the enrichment is 5.0 w/o.

The inter-pretation of the endpoint data is as follows:

the reactivity of the spent fuel racks containing fuel with 48 IFBA rods which has an initial enrichment of 5.0 w/o is equivalent to the reactivity of the spent fuel racks containing fresh fuel having an initial enrichment of 4.2 w/o.

It is important to recognize that the curve in Figure 2 is based on a constant rack reactivity for that region and not on a constant fuel assembly reactivity.

The data in Figure 2 is also provided as Table 2.

2.'1.1 ANALYTICALMETHODS The data points on the reactivity equivalence curve are calculated with a

transport theory computer code, PHOENIX' '.

PHOENIX is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code.

A 25 energy group nuclear data library based on a modified version of the British WIMS'

'ibrary 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 after discharge.

The fuel reactivity was found to reach a

maximum at approximately 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after discharge.

At this point in time, the major fission product poison, Xe,

has nearly completely decayed away.

Fur-

thermore, the fuel reactivity was found to decrease continuously from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Criticality Analysis

to,30 years following discharge.

Therefore, the most reactive point in time for a

fuel assembly after discharge from the reactor can be conservatively ap-proximated by removing the Xe' '

The PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been examined following discharge from a reac-tor.

In addition, an extensive set of benchmark critical experiments has been analyzed with PHOENIX.

Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 3.

The measure-ments were made on fuel discharged from Yankee Core 5' The data in Table 3

shows that the agreement between PHOENIX predictions 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.

Key parameters describing each of the 81 experiments are given in Table 5.

These reactivity comparisons again show good agreement between experiment and PHOENIX calculations.

An uncertainty associated with the IFBA dependent reactivity computed with PHOENIX is accounted for in the development of the IFBA loading requirements.

A bias of approximately 0.005 Bk (4 IFBA rods) at 5.0 w/o is considered very conservative since comparison between PHOENIX results and the experiments and the licensed core design methods show very good agreement.

2.'l.2 REACTIVITYCALCULATIONS The equivalent Ke~r for the storage of spent fuel in the fuel racks is determined using the methods described in the original criticality report' The reference conditions for this are defined by the zero IFBA intercept point in Figure 2.

The KENO-IV' 'omputer code was used to calculate the storage rack multi-plication factor with an equivalent fresh fuel enrichment of 4.2 w/o and no IFBA's.

The KENO calculation for the nominal case resulted in a

Ker~ of 0.9207 with a 95 percent probability/95 percent confidence level uncertainty of +0.0046.

The maximum K~~r under normal conditions was determined with a "worst case" KENO model which included mechanical and material tolerances in addition to asymmetric positioning of fuel assemblies within the storage cells.

The maxi-mum Ken for the Shearon Harris spent fuel storage racks was 0.9448 including method biases and uncertainties at a

95/95 probability/confidence level.

This analysis is discussed in detail in Reference 7.

Criticality Analysis

2,2 INFINITE MULTIPLICATIONFACTOR To store fuel assemblies in the Shearon Harris spent fuel racks which do not meet the IFBA assumptions specified in Section 2.1, and therefore cannot use the IFBA number curve in Figure 2, an infinite multiplication factor for a nominal fresh 4.2 w/o fuel assembly was determined.

The infinite multiplication factor, or Koo, is used as a reference criticality reactivity point which eliminates the need to specify an acceptable enrichment versus number of IFBA rods corre-lation.

The fuel assembly K~

depletion calculations are performed using the Westinghouse licensed core design codes.

These codes include TURTLE' 'nd PHOENIX-P' The following assumptions were used to develop the infinite multiplication factor model:

1.

The fuel assembly contains the highest enrichment authorized, is at its most reactive point in life and no credit is taken for any burnable absorbers in the assembly.

A Westinghouse 17x17 OFA fuel assembly was analyzed (See Table 1 for fuel parameters).

2.

All fuel rods contain uranium dioxide at an enrichment of 4.2 w/o Uver the infinite length of each rod.

3.

The fuel array is in the Shearon Harris reactor geometry and is.infinite in the lateral and axial extent.

4.

The moderator is pure water at a temperature of 68'.

A conservative value of 1.0 gm/cm's used for the density of water.

Calculation of the infinite multiplication factor resulted in a reference K~ of 1.470.

As a result all fuel assemblies placed in the Shearon Harris spent fuel racks which do not qualify to use the enrichment versus number, of. IFBA rods curve in Figure 2 must have a reference reactivity less than or equal to the above value.

2.2,1 REACTIVITy CALCULATIONS Using the previous analysis and results described in the original criticality report' and an additional uncertainty for the reference reactivity, the fol-lowing equation is used to verify that the reference K~ of 1.470 results in a

maximum Keff 5 0.95 for the Shearon Harris Region 1 spent fuel storage racks:

Keff - Kworst

+

Bmetrtod

+ Bpsrt

+ t/ [ (ks) worst

+ (ks) method

+ (ks) rr ]

where:

Criticality Analysis

Kworst worst case KENO Korr that includes material tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal Bmethod method bias determined from benchmark critical comparisons Bpsrt ksworst ksmethod bias to account for poison particle self-shielding 95/95 uncertainty in the worst case KENO Koo 95/95 uncertainty in the method bias ksrr uncertainty in.reactivity equal to 0.2 w/o in fuel enrichment to account for enrichment and Ks calculational uncertainties.

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

Katt

= 0.9306

+ 0.0083

+ 0.0014

+ s/ [(0.0041)*

+ (0.0018)'

(0.0085)'

= 0.9499 Since Keff is less than 0.95 including uncertainties at a

95/95 probability/confidence level, the acceptance criteria for criticality is met with fuel that has a reference reactivity less than or equal to 1470.

2.3 POSTULATED ACCIDENTS Most accident conditions will not result in an i@crease in Ksrt of the rack. Ex-amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a

fuel assembly on top of the rack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel height of stored assemblies which precludes interaction).

However, accidents can be postulated which would increase reactivity (i.e.,'r dropping a fuel assembly between the rack and pool wall).

For these accident conditions, the double contingency principle of ANSI N16,1-1975 is applied.

This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.

Thus, for 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.

The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by about 30 percent hK.

Thus, for postulated accidents, should there be a reactivity increase, Kott would be less than or equal to 0,95 due to the effect of the dissolved boron.

Criticality Analysis

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1

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4

,<.4e optimum moderation accident is not a problem.

The presence of poison plates removes the conditions necessary for optimum moderation so the Ke<<

continually decreases as moderator density decreases from 1.0 gm/cm'o 0,0 gm/cm' Criticality Analysis

3.0 ACCEPTANCE CRITERION FOR CRITICALITY The neutron multiplication factor in spent fuel pool shall be less than or equal to 0.95, including all uncertajnties, under all conditions.

The analytical methods employed herein conform with ANSI N18.2-1973, "Nu-clear 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, "Spent Fuel Storage";

and the NRC guidance, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications".

Acceptance Criterion For Criticality

e Table 1.

Fuel Parameters Employed in Criticality Analysis Parameter W 17x17 OFA W 'l7x17 STANDARD Number of Fuel Rods per Assemb I y 264 264 Rod Zirc-4 Clad O.D.

(inch)

Clad Thickness (inch)

Fuel Pellet O.D.(inch) 0.360 0.0225 0.3088 0.374 0.0225 0.3225 Fuel Pellet Density (0 of Theoretical) g6 g6 Fuel Pellet Dishing Factor Rod Pitch (inch) 0.0 0.496 0.0 0.496 Number of Zirc-4 Guide Tubes 24 24

'uide Tube O.D.

(inch)

Guide Tube Thickness (inch) 0.474 0.016 0.482 0.016 Number of Instrument Tubes Instrument Tube O.D.

(inch) 0.474 0.482 Instrument Tube Thickness (inch) 0.016 0.016

Table 2.

Shearon Harris Fuel Assembly Minimum IFBA rods vs Initial Unrichment for Region 1 Spent Fuel Rack Initial U'

'nrichment IFBA Rods in Assenbly 4.2 4,4 12 4.6 24 4.8 36 5.0 48

Table 3.

Comparison of PHOENIX Isotopics Predictions to Yankee Core 5

Measurements Quantity (Atom Ratio)

U235/U U236/U U238/U PU239/U PU240/U PU241/U PU242/U P U239/U238 Mass(PU/U)

F ISS-PU/TOT-P U

% Difference

-0.67

-0.28

-0.03

+3.27

+3.63

-7.01

-0.20

+3.24

+1.41

-0.02 11

Table 4.

Benchmark Critical Experiments PHOENIX Comparison Description of Experiments Number of Experiments PHOENIX Kerr Using Experiment 8ucklings UO2 Al clad SS clad Borated H20 14 19 0.9947 0.9944 0.9940 Subtotal 40 0.9944 U-Metal Al clad 41 1.0012 TOTAL 81 0.9978 12

Table 5.

Data for U Metal and UO~ Critical Experiments (Part 10f 2)

Case Number Cel 1

Type A/0 H2D/U U-235" Ratio Fuel Density (G/CC)

Pellet 01ameter (CM)

Material Clad Clad Clad OD Thickness (CM)

(CM)

Lattice Pitch (CM) 8-10 PPM 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Square Square Square Square Square Square Square Square Square Square Square Square Square Square Squat e Square Square Square Square Square Square Square Square Squar e Square Square Square Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa

1. 328 1.328

1. 328

1. 328 1.328 2.734 2.734 2.734 2.734 2.734 2.734 2.734 2.734 3.745 3.745 3.745 3.745 3.745 3.745 3.745 3.745 4.069 4.069 4.069 2.490 3.037 3.037 4.069 4.069 4.069 4.069 2.490 2.096 2.096 2.096 2.096 2.096

1. 307

1. 307 3.02 3.95 4.95 3.92 4.89 2.88 3.58 4.83

2. 18 2.92 3.86 7.02 8.49 10.38 2.50 4.51 2.50 4.51 4.51 4.51 4.51 4.51 4.51 4.51 2.55 2.55

2. 14 2.84 2.64

8. 16 2.59 3.53 8.02 9.90 2.84 2.06 3.09

4. 12

6. 14 8.20 1.01 1.51 2.02 7.53 7.53 7.53 7.52 7.52

10. 53

10. 18

10. 27

10. 37

9. 46 9.46 9.46

10. 24 9.28 9.28 9.45 9.45 9.45 9.45

10. 24

10. 38

18. 90

18. 90 1.5265 1.5265

1. 5265

.9855

.9728

.9728 9728

.7620

.7544

.7544 7544

.7544

1. 1278

1. 1278 1.0297

1. 1268

1. 1268 1.0297

1. 5240

1. 5240 A 1 um1num A 1 um 1 num A l um 1 num A 1 um 1 num Aluminum A 1 um 1 num A 1um 1 num Aluminum SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS"304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 Aluminum SS-304 SS"304 SS-304 SS-304 SS-304 SS-304 A 1 um1num A 1 umi num Aluminum Aluminum A 1 um1num Aluminum Aluminum A 1 umi num A 1 umi num

1. 6916

1. 1506

.8594

.B594

.8594

.8600

1. 2090

1. 2060

1. 1701

1. 2701

1. 2701 1.2060

1. 6916 1. 6916

1. 6916

1. 6916 1. 6916

1. 6916

.07110

.04085

.04060

. 08130

.07163

. 07163.

. 07163

. 08130

.07112

.07112 2.2050'.3590

2. 5120 1.5580 1.6520

1. 5580

1. 6520 1.8060 1.0287

1. 1049

1. 1938

1. 4554 1.5621

1. 6891

1. 0617

1. 2522 1.0617 1.2522 1.2522 1.2522 1.2522 1.2522 1.2522

1. 2522 1.5113

1. 5113

1. 4500

1. 5113 1.5550

2. 1980 1.5550

1. 6840

2. 1980

2. 3810 1.5113

2. 1737 2.4052 2.6162

2. 9891 3.3255

2. 1742 2.4054

2. 6162 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 456.0 709.0 1260.0 1334.0 1477.0 0.0 3392.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 1677.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 13

Table 5.

Data for U Metal and UO~ Critical Experiments (Part 2 of 2)

Case Cell Number Type A/0 U-235 H2D/U Ratio Fuel Pellet Clad Clad Lattice Density D1ameter Material DD Thickness Pitch 6-10 (G/CC)

(CM)

Clad (CM)

(CM)

PPM 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa Hexa 1.307

1. 307

1. 160

1. 040 1.040 1.040 1.040 1.040 1.307

1. 307

1. 307 1.307 1.307

1. 160

1. 160 1.040 1.040 1.040 1.040 1.040

1. 040

1. 310

1. 310 1:159

1. 159 1.312 1.312 3.01 4.02 1.01 1.51 2.02 3.01 4.02 1.01 1.51 2.02 3.01 4.02 1.00

1. 52 2.02 3.02 4.02 1.52 2.02 3.02 4.02 1.00

1. 52 2.02 3.02 4.02

1. 33

1. 58 1.83 2.33 2.83 3.83 2.02 3.01 2.02 3.01 2.03 3.02

18. 90 18.90

18. 90

18. 90 18.90

18. 90

18. 90 18.88

18. 88 18.88 18.88

'8.

88

18. 88

1. 5240

.9830

19. 050 19.050 19.050

19. 050 19.050

19. 050

1. 5240

,1. 5240

1. 5240

.9830

.9830 A 1 umi num A 1 umi num A 1uminum Aluminum A l um1 num A 1um 1 num A 1 um 1 num A lum1num Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum Aluminum A lumi num Aluminum Aluminum Aluminum Al um 1 num A'lum 1 num A lum 1 num Aluminum Aluminum Aluminum A 1 um 1 num A 1um1num Aluminum A 1umi num Alumi num A 1 umi num A 1 umi num Aluminum Aluminum Aluminum Aluminum 1.6916

1. 6916

1. 6916 1.6916 1.6916

1. 6916

1. 1506

1. 1506 2.0574 2.0574 2.0574 2.0574 2.0574 2.0574

1. 6916 1.6916

1. 1506

.07112

. 07112

.07112

. 07112

.07112

.07620

.07620.

.07620

.07112

. 07112

.07112

.07112 2.9896 3.3249

2. 1742 2.4054

2. 6162 2.9896 3.3249

2. 1742 2.4054

2. 6162 2.9896 3.3249 1.4412 1.5926

1. 7247 1.9609

2. 1742

'.5926 1.7247 1.9609

2. 1742

1. 4412 1.5926

1. 7247

1. 9609

2. 1742 2.8687 3.0086

3. 1425 3.3942 3.6284 4.0566

2. 6160

2. 9900

2. 6160

2. 9900 1.7250

1. 9610 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 14

7.5" BORAFLEX

1. 33" 8.75 DETATL

~CELL CENTER TO CENTER (10.5" )~

.100" GAP T

0.075

'ZFYER CELL MALL 0.035

%UPPER 0.075" BORArLEX (0. 020 Qf-B10/cm2)

Figure 1.

Shearon Harris Spent Fuel Storage Cell Nominal Dimensions

48 40 36

~ 32 4J w28 Cl CD IX

~ 24 tQ ACCEPTAB I

I I

LE L

I I

--r I

I

~ 20 CD E

16 I

I NOT ACCEPTABLE 12 J.

I I

4.2 4.3 I

1 I

44 4J 46 47 U~~~ ENRICHMENT (W/0) 4.8 I

T' I

I I

I 4.9 5.0 Figure 2.

Shearon Harris Fuel Assembly Minimum Number of IFBA Rods vs.

Initial U' 'nrichment for Storage in Region T Spent Fuel Racks

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 ApplicationsApril 14, 1978.

2.

L. M, Petrie and N. F. Cross, KENO IV--An Improved Monte Carlo Criticality

'Program, ORNL-4938, November 1975.

3.

A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors, WCAP-10106, June 1982.

4.

Askew, J.

R., Fayers, F. J., and Kemshell, P. BA General Description of the Lattice Code

WINS, Journal of British Nuclear Energy

Society, 5,

pp.

564-584, 1966.

5.

England, T.

R., CINDER -

A One-Point Depletion and Fission Product

Program, WAPD-TM-334, August 1962.

6.

Melchan, J.

B., Yankee Core Evaluati on Program Final

Report, WCAP-3017-6094, January 1971.

7.

Boyd, W. B., Fecteau, M. W., Savin, N., Criticality Analyis of Shearon Harris Spent Fuel Racks January 1987.

8.

Altomare, S.,

Barry, R.

F., The TURTLE 24.0 Diffusion Depletion

Code, WCAP-7758-A, February 1975.

9.

Nquyen, T. Q., et alQualification of the PHOENIX-PIANC Nuclear Design System for Pressurized Water Reactor

Cores, WCAP-11597-A, November 1987.

Bibliography 17

'~1 I.

~

ML18005A852

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1.0 INTRODUCTION

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