ML20081A994

Proposed Tech Specs Re Fuel Enrichment Increase

ML20081A994
Person / Time
Site:	Crystal River
Issue date:	03/09/1995
From:	FLORIDA POWER CORP.
To:
Shared Package
ML20081A991	List: 3F0395-09, Updates Sholly Evaluation & Replaces Attachment 3 W/ Nonproprietary Version of Rept BAW-2209,Rev 1, Crystal River Unit 3 Spent Fuel Storage Pool B Criticality Analysis, Per Discussions W/Nrc Re FPC 950126 Application ML20081A994 ML20081B007
References
NUDOCS 9503150373
Download: ML20081A994 (18)

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l REPLACEMENT PAGES FOR APPENDIX A CRYSTAL RIVER UNIT 3 TECHNICAL SPECIFICATIONS I

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Spent Fuel Ass:mbly Storage 3.7.15 j'.

p 3.7 PLANT SYSTEMS 3.7.15. Spent. Fuel Assembly Storage

.LC0 3.7.15 The combination of initial enrichment and burnup of each spent fuel assembly stored in Storage Pool A and Storage Pool B shall-be within the' acceptable region of Figure 3.7.15-1, Figure 3.7.15-2, Figure 3.7.15-3, or stored in accordance with the FSAR.

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APPLICABILITY:

Whenever'any fuel assembly is stored in Storage Pool A or Storage Pool B of the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Requirements of the A.1

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NOTE---------

! C) not met.

LC0 3.0.3 is not applicable.

Initiate action to Immediately move the' noncomplying l

fuel assembly from Storage Pool A, or Storage Pool B.

t Crystal River Unit 3 3.7-30 Amendment No.

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_--_-_----__-_----------___--_---_------_--__A

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Spent Fuel Assembly. Storage 3.7.15.

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SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY-

'l SR 3.7.15.1.

Verify by administrative means the initial" Prior to i

enrichment and burnup of the ' fuel assembly storing the I

is in accordance with Figure 3.7.15-1, fuel. assembly Figure 3.7.15-2, figure 3.7.15-3, or in in Storage Pool ~

accordance with.the FSAR.

A or Storage Pool B.:

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Crystal River Unit 3 3.7-31 Amendment No.

'A Spent Fuel Assembly _ Storage 3.7.15 MINIMUM BURNUP REQUIRED FOR "A"

POOL STORAGE Minimum Burnup vs Initial Enrichment-L-

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r Figure 3.7.15-1 (page 1 of 3) t Burnup versus Enrichment Curve for l

r Spent Fuel Storage Pool A i

f Crystal River Unit 3 3.7-32 Amendment No.

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Spent Fuel Assembly Storage 3.7.15 MINIMUM BURNUP REQUIRED FOR REGION 1 0F "B" POOL,

Minimum Burnup vs Initial Enrichment Burned Fuel in-checkerboard Configuration with 5.0 wt% Fresh Fuel 35.0 30.0 o

ACCEPTABLEFOR STORAGE,lDJACENT TOFRI SHFUELHITH ENREHMENT.

j g 25.0 15 m NHLN All0VEAND F0THE

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20.0

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15.0 3

E 10.0 5.0 0.0 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Wila! Enrichment, wi% U235 t

Figure 3.7.15-2 (page 2 of 3)

Burnup versus Enrichment Curve for Spent Fuel Storage Pool B, Region 1 Crystal River Unit 3 3.7-33 Amendment No.

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Spent Fuel Ass:mbly Storage -

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MINIMUM BURNUP REQUIRED FOR i

REGION 2 0F "B" POOL Minimum Burnup vs Initial Enrichment I

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ACCEPTABLE FOR ST( RAGE i

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Initial Enrichment, w17. U235

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Figure 3.7.15-3 (page 3 of 3)

Burnup versus Enrichment Curve for Spent Fuel Storage Pool B, Region 2 d

Crystal River Unit 3 3.7-33A Amendment No.

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.D: sign Features--

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4.0 DESIGN FEATURES:

'4.1' Site

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The 4,738 acre site is charac'terized by'a 4,400 foot minimum exclusion-t radius' centered on the Reactor Building;-isolation from nearby-population centers; sound foundation for structures; an abundant supply L

of cooling water; an ample supply of emergency power; and favorable -

conditions of hydrology, geology, seismology, and meteorology.--

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 177 fuel assemblies.

Each. fue1~ assembly.

shall consist of a matrix of Zircaloy-4 clad fuel rods with an

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initial composition of natural or slightly enriched uranium dioxide (U0 ) as fuel material, with a maximum enrichment of 5.0 l-2 weight percent U-235.

Limited substitutions of-stainless steel filler rods for fuel rods, in'accordance with approved applications of fuel rod configurations, m1y be used.. Fuel-assemblies shall be limited to those fuel-designs that have been-analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.

Each fuel rod shall have a nominal active fuel length of' 144 inches and shall contain a maximum total weight of 2253 grams uranium.

1 4.2.2 CONTROL RODS The reactor core shall contain 60. safety and regulating (including.

extended life CCNTROL RODS) and 8 AXIAL POWER SHAPING (APSR) rods.

Except for the extended life CONTROL RODS, the CONTROL RODS shall i

contsin a nominal 134 inches of absorber material.

The extended life CONTROL RODS shall contain a nominal 139 inches of absorber material. The nominal values of absorber material shall be 80 l

percent silver, 15 percent indium, and 5 percent cadmium.

Except for extended life CONTROL RODS, all CONTROL RODS shall be clad with stainless steel tubing. The extended life CONTROL RODS shall be clad with Inconel. The APSRs shall contain a nominal 63 inches of absorber material at their lower ends. The absorber material for the APSRs shall be 100 % Inconel.

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(continued) l Crystal River Unit 3 4.0-1 Amendment No.

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e Design Features 4.0 4.0' DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel stcrage racks'are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; b.

k s 0.95 if fully flooded with unborated water,-

wNfch includes an allowance for uncertainties as described in Section 9.6 of the FSAR; c.

A nominal 10.6 inch center to center distance between fuel assemblies placed in Region 1 of the B pool; d.

A nominal 9.17 inch center to center distance between fuel assemblies placed in Region 2 of the B pool; and e.

A nominal 10.5 inch center to center distance between fuel assemblies placed in the A pool.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; l

b.

k s 0.95 if fully flooded with unborated water, wN[ch includes an allowance for uncertainties as described in Section 9.6 of the FSAR; c.

k s 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as i

described in Section 9.6 of the FSAR; and d.

A nominal 21.125 inch center to center distance between fuel assemblies placed in the storage racks.

(continued)

Crystal River Unit 3 4.0-2 Amendment No.

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l BASES (INFORMATION ONLY)

e Spent Fuel Assembly Storage 3.7.15 B 3.7 PLANT-SYSTEMS i

B 3.7.15 Spent Fuel Assembly Storage BASES BACKGROUND This document describes the Bases for the Spent Fuel Assembly Storage which imposes storage requirements upon irradiated and unirradiated fuel assemblies stored in the fuel storage pools containing high density racks. The storage areas, which are part of the Spent Fuel System, governed by this Specification are:

a.

Fuel storage pool "A" and b.

Fuel storage pool "B".

In general, the function of the storage racks is to support and protect new and spent fuel from the time it is placed in the storage area until it is shipped offsite.

Spent fuel is stored underwater in either fuel storage pool A or B.

Only fuel pool A has the capability to store failed fuel in containers.

Spent fuel pool A features high density poison storage racks with a 10 1/2 inch center-to-center distance capable of storing 542 assemblies.

Fuel pool A is capable of storing fuel with enrichments up to 5.0 weight percent U-235 (Ref. 1) without exceeding the criticality criteria of Reference 3 providing the feel has sufficient burnup.

Spent fuel pool B also contains high density racks sept

.d into 2 regions. The racks in Region I have a 10.60 i. u center-to-center spacing capable of storing 174 asse A The high density racks in Region 2 have 9.17 inch cent center distance capable of storing 641 assemblies.

14 pool B is capable of storing fuel with enrichments up

.0 weight percent U-235 (Ref. 2) without exceeding the criticality criteria of Reference 3, providing the fuel has sufficient burnup and required storage configuration.

The 5% maximum enrichment limit is actually a nominal value.

The tolerance of fuel supplied by the enrichment facility is 10.013 weight percent. Thus, it is possible to have fuel with an initial enrichment slightly in excess of the stated limit. Tolerance variations are included in the criticality analysis and are acceptable.

(continued)

Crystal River Unit 3 B 3.7-72 Revision No. 01

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' Spent Fuel. Assembly ' Storage _

l B 3.7.15-BASES BA'KGROUND

. Both of the spent fuel pools are constructed.of -

C (continued) reinforced-concrete and lined with stainless steel plate.

They are located in the. fuel handling area of the auxiliary building (Ref. 2).-

New fuel storage requirements are addressed in Section 4.0,

" Design Features".

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s-APPLICABLE The function of the spent fuel storage racks are to support-.

SAFETY ANALYSES and protect spent fuel assemblies from the time they are placed in the pool until they are shipped offsite. The spent fuel assembly storage LCO was derived from the need to establish limiting conditions on fuel storage to assure sufficient safety margin exists to prevent inadvertent.

criticality. The spent fuel assemblies are stored entirely underwater in a configuration that has been shown to result.

in a reactivity of less than 0.95 under worse case conditions (Ref. I and 2). The spent fuel assembly

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enrichment requirements in this LC0 are required to ensure inadvertent criticality does not occur in the spent fuel e

pool.

Inadvertent criticality within the' fuel storage area could -

result in offsite radiation doses exceeding 10 CFR 100

limits, j

The spent fuel assembly storage satisfies Criterion 2 of the 1

NRC Policy Statement.

LC0 Limits on the irradiated fuel assembly storage in high density racks were established to ensure the assumptions of the criticality safety analysis of the spent fuel pools is maintained.

Limits on initial fuel enrichment and burnup for spent fuel stored in pool A have been established. Two limits are defined:

1.

Initial fuel enrichment must be less than or equal to l

5.0 weight percent U-235, and i

(continued)

I Crystal River Unit 3 B 3.7-73 Revision No. 01 t

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LCO' 2.

For spent fuel;with initial' e' richment' less:than or-i n

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to'3.5 weight percent, fuel burnup must be.within:the'-

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limits'specified in Figure 3,7.15-1:

(Figure 3.7.15 :#

" presents required fuel assembly.burnup as aLfunction:

of-initial enrichment.)::.

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i Fuel enrichment limits are based on avoiding inadvertenti g.

criticality in.the spent fuel pool.> TheLCR-3 spent fuet '

s storage system was initially designed.to-a maximum

.i enrichment ~ of 3.5 weight. percent.

Enrichments of up to~ 5.01 l

weight percent are permissible'for~ storage in spent fuel'

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pool A as long as the fuel. burnup is sufficient to' limit the

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worst case reactivity in the storage pool to less than 0.95..

1 Fuel burnup reduces the reactivity of-the fuel due to-the accumulation of fission. product poisons. Reference 1--

documents that the required burnup varies linearly.as. a -

function of enrichment with.10500 megawatt days per metric ton uranium (Mwd /mtU) required for fuel with 5.0 weight

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percent enrichment and 0 burnup required for 3.5 weight:

percent enriched fuel.

Similar. types of restrictions:have been established for.

i Pool B.

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Initial fuel enrichment must be's 5.0 weight j

percent'U-235,

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'i 2.

For Region 1, fuel with initial : enrichment 15.0 -

weight percent.and 1 2.08 weight percent fuel burnup must be within the. limits specified in 1

Figure 3.7.15-2 and' arranged in.a required-l checkerboard' configuration with new fuel or 1

burned fuel of $~ 5.0 weight percent, and 3.

For spent fuel with initial enrichment 15.0 i

weight percent'and-2 1.63 weight. percent in-

.i Region 2, fuel burnup must be within the limits specified in Figure 3.7.15-3.

(Figure.3.7.15-3 j

presents required fuel assembly burnup as a" e

i function of initial enrichment.).

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(continued) f

' Crystal River Unit 3 8 3.7-74 Revision No. 01

Spent Fuel Assembly Storage B 3.7.15 BASES LC0 The analysis for spent fuel pool A and the new fuel storage (continued)

. racks includes allowance for a tolerance of 5.0 1 0.02 weight percent enrichment. The analysis for Region 1 and Region 2 of spent fuel pool B includes allowance for a tolerance of 5.0 1 0.05 weight percent enrichment. The uncertainty allowance in the measured burnup is included in the administrative controls by requiring an additional 5%

margin in the estimated vs required burnup. ~This allowance for burnup is conservatively assumed to be the same as the measurment uncertainty allowance used in determining Fan.

The LC0 requires administrative controls, as specified in applicable fuel handling procedures, to be used in storing fuel in accordance with the FSAR and Figures 3.7.15-1, 3.7.15-2 and 3.7.15-3. This is acceptable since this assures fuel is stored in configurations which meet the requirements of the safety analysis.

The LC0 requires new fuel to be stored in Region 1 of spent fuel pool B in a checkerboard configuration with burned fuel which meets t.he requirements of Figure 3.7.15-2. The checkerboad configuration ds defined by storage of new fuel in diagonally adjacent locaitions. Burned fuel or a vacant location is immediately adjacent (face-to-face) with new fuel or fuel not meeting the burnup requirements of Figure 3.7.15-2.

APPLICABILITY

'In general, limiting fuel enrichment of stored fuel prevents inadvertent criticality in the storage pools.

Inadvertent criticality is dependent on whether fuel is stored in the pools and is completely independent of plant MODE.

Therefore, this LCO is applicable whanever any fuel assembly is stored in high density fuel storage locations.

ACTIONS M

Required Action A.1 is modified by a Note indicating LC0 3.0.3 does not apply. Since the design basis accident of concern in this Specification is an inadvertent criticality, (continued)

Crystal River Unit 3 8 3.7-75 Revision No. 01

[a Spxnt Fuel Assembly Storage B 3.7.15 BASES ACTIONS

/L1 (continued) and since the possibrlity or consequences of this event are independent of plant MODE, there is no reason to shutdown the plant if the LC0 or Required Actions cannot be met.

When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with Figure 3.7.15-1, Figure 3.7.15-2, Figure 3.7.15-3, or the FSAR, immediate action must be taken to make the necessary fuel assembly movement (s) to bring the configuration into compliance. The Immediate Completion Time underscores the necessity of restoring spent fuel pool irradiated fuel loading to within the initial assumptions of the criticality analysis.

The ACTIONS do not specify a time limit for completing movement of the affected fuel assemblies to their correct location.

This is not meant to allow an unnecessary delay in resolution, but is a reflection of the fact that the complexity of the corrective actions is unknown.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS Verification by administrative means that initial enrichment and burnup of fuel assemblies in accordance with Figure 3.7.15-1, Figure 3.7.15-2, and Figure 3.7.15-3 is required prior to storage of spent fuel in storage pool A or pool B, (as applicable). This surveillance ensures that fuel enrichment limits, as specified in the criticality' safety analysis (Ref. 2), are not exceeded. The surveillance Frequency (prior to storage in high density region of the fuel storage pool) is appropriate since the ir,itial fuel enrichment and burnup cannot change after removal from the core.

REFERENCES 1.

Criticality Safety Evaluation of the Pool A Spent Fuel Storage Racks in Crystal River Unit 3 with Fuel of 5.0% Enrichment, S.E. Turner, Holtec Report HI 931111, December 1993.

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(continued)

Crystal River Unit 3 B 3.7-76 Revision No. 01

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- Spent Fuel Assembly-Storage-ic B 3.7.15' 7

.BASESL.

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. REFERENCES

2. -

Crystal River Unit 3 Spent Fuel Storage Pool B

-(continued)

Criticality ' Analysis, W. A. Wittkopf, L. A. Hassler, B&W G

. Fuel Company, BAW-2209P, October 1993.f l

3.

NUREG 0800, Standard Review Plan, Section 9.1.1 and 9.1.2, Rev.2,-July 1981.--

4.

10 CFR 100.

5.

CR-3 FSAR, Section 9.6,. Revision-11.

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Crystal River Unit 3 8 3.7-76A Rev sion No. 01

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ATTACHMENT 1

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s CRITICALITY SAFETY EVALUATION OF THE.

CRYSTAL RIVER UNIT 3 NEW FUEL STORAGE VAULT i

WITH FUEL OF 5% ENRICHMENT f

Prepared for the

,i FLORIDA POWER CORPORATION i

by

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Stanley E. Turner, PhD, PE i

December 1993 i

i Holtec Project 21195 Holtec Report HI-93110 i

'I 230 Normandy Circle 2060 Fairfax Ave.'

Palm Harbor, FL 34683 Cherry Hill,'NJ 08003

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TABLE OF CONTENTS

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1.0 INTRODUCTION

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'2.0

SUMMARY

2 3.0 ' CRITICALITY ANALYSIS.

3 ~~

t 3.1 Fuel' Assembly Specifications.

3-3.2 New Fuel Storage Rack Design.

3 3.3 Analytical Methods..

3 3.4 Manufacturing Tolerances'. -

4 4

4.0 ABNORMAL AND ACCIDENT CONDITIONS 5

4

5.0 REFERENCES

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-List'of Tables Table 1 SUMARY OF ' CRITICALITY SAFETY ANALYSIS.

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. Table 2 FUEL ASSEMBLY SPECIFICATIONS

' 8 List of Fioures f

Fig. 1 New Fuel Storage Vault Configurations.

9.

Fig. 2 Reactivity Variation with Moderator Density.

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1.0 3 INTRODUCTION j

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Ini: a? previou's :(evaluation criticality l safety:E analyses:

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. established that the' New Fuel Storage-Vault at. Crystal River. Unit 3':

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.could_ safely accommodate fuel of 4.5%' enrichment, with restrictions -

j on:the, number ofl useable storageLlocations. '.The present study is~

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'inten'ded'to extend:that analysis to confirm the: capability'of the o

'New Fuel. Storage. Vault. to ' safely - receive and st!orel fuel of L 5.~0% -

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initial 5 enrichment with the same ; restriction.'on :. useable storagen

b

' locations. A companion report, HI-931111,: documents the capabilityJ of Pool A to also a'ecept fuel'of 5.0% initial enrichment.

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

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very low.

To assure criticality safety under accident condition's l

and to-conform to the requirements of General-Design Criterion 62,

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" Prevention of, Criticality in Fuel Storage and-Handling",- ~ two.

l separate criteria' must < be satisfied as' idefined.in NUREG-0800,

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Standard Review Plan 9.1.1, "New Fuel Storage". These criteria areL

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i as follows:

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when. fully. loaded with fuel of the highest anticipated:.

reactivity and flooded with clean unborated water, the maximum j

reactivity, including uncertainties, - shall not exceed'a k,,, of f i

O.95.

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l With fuel of the highest anticipated reactivity in place':'and ~

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assuming ' the optimum hypothetical lowidensity moderation,

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-(i.e., fog or foam)', ' the maximum : reactivity shall not exceedz

l a k,,, of.0.98.

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

fuel.

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

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SIDeGLRY -

' The New Futi Storage Vaukt normally prmides a 6 x 11 cell array of

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storage ' locations L arranged on a : 21.125. inch lattice. ~ spacing.

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.Results of the previous evaluation showed thatlit is;necessary to i

o-blank-of f, and keep ~ empty of fuel,.. two' rows of storage locations as

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' indicated in Figure 1.-

With the same restriction,: the remaining 54

'l storage. locations in the New Fuel -Storage Vault caniaccommodate f

5.0% enriched fuel within the two Regulatory guidelines identified above.

4 Calculations were 'made with the 27-group NITAWL-KENO-Sa code ~

package, a. three-dimensional'~ Monte Carlo analytical technique,-

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using.the configuration illustrated in Figure 1.

Results of The.

j criticality. safety analyses are summarized in Table 1 forithe two.

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accident conditions.

Figure'2 illustrates the variation in k,7 with moderator density and shows that the peak reactivity. (optimum j

moderation) occurs at 7.5% moderator ' density.

The. maximum.

.l reactivity at 7.5% moderator density is a O'.978, including.

3 uncertainties, which is within the Regulatory limit of 0.98, thus-li confirming the acceptability of the NFV for 5.0% fuel.

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In the flooded condition (clean unborated ' water), the storage l

locations are essentially isolated from each other (neutronically).

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Under these conditions and with fuel of~5.0% enrichment, the maximum reactivity, including all known. uncertainties', is 0.948 I

which is less than the limiting value of 0.95, thus confirming the i-l acceptability of the NFV for 5.0% fuel.

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3.0 CRITICALITY ANALYSES 3.1 Fuel Assembly Specifications

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

Table 2 summarizes the fuel assembly design specifications and expected range. of significant fuel tolerances.

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3.2 New Fuel Storage Rack Design I

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

The storage locations are arranged in eleven rows of six cells each, located on.

a 21 1/8 1/16 inch lattice spacing, as illustrated in Figure 1.

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

3.3 Analytical Methods The criticality analyses were made with the three-dimensional Monte-Carlo code package NITAWL-KEN 05a(2) using the 27-group SCALE

• cross-section library (3) and the Nordheim integral treatment for U-238 resonance shielding effects. Benchmark calculations, presented j

in Appendix A, indicate a bias of 0.0103 i O.0018 for the NITAWL-KENO-Sa code package, at the 95% probability, 95% confidence-

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level (').

l In the calculational model, each fuel rod, cladding, or guide tube.

were explicitly described.

The model also used the standard

'l concrete reflector option available-in KENO-5a to describe the l

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

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

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uncertainty. due - to the random 1 nature of neutron ' tracking'.

To y

x minimize -the-statistical-~ uncertainty: of the KENO-calculated creactivity,- a minimum-of 500,000 neutron histories-iin 1000 generations of 500' neutrons each,, were accumulated in--each calculation'.

For the - flooded. case and. at ' optimum - low-density moderation,-. confirmatory calculations were. made with ' 2,500;000 neutron histories.

Furthermore, because 'of: the close approach tio the limiting reactivity, check calculations for the flooded case and at ~ optimum low-density-iroderation were ' made with the 218 -

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neutron group library..

Results of these calculations are' asD follows:

CASE 7.5%' Mod. Dana.

Flooded' 27-groups 0.9648't 0.0006 0.9345

• 0.0008-218-groups 0.9623 t 0.0010 0.9320-g 0.0012 The 218-group calculations' resulted in a slightly lower k, than..

g the reference 27-group library, thus confirming the reference calculations..

L 3.4 Manufacturing Tolerances The reactivity uncertainties associated with various manufacturing tolerances were calculated by the. difference' between KENO-Sa calculations, each with the nominal value and a,second calculation l

with each value set at the maximum tolerance.

Results. are tabulated below:

Ak' Uncertainty ~

Tolerance e 76 Mod Dens' flooded:

i 1/8 in lattice spacing i.0.0015 t.0.0007 i 0.02 in % enrichment t 0.0013 i 0.0014 i 0.166 gg/cc i 0.0015 t 0.0015 _.___L

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Statisti~ cal Combination

.* 0.0025 t 0.022

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4.0 Abnormal and-Accident Conditions l

i Normally,. the new fuel storage vault is dry with a very low reactivity.

The two limiting criticality criteria'are accident l

- conditions. and no other safety. concerns. have been. identified.

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Under.the. double contingency principle of ANSI-N16-1975, endorsed.

by the April.1978 USNRC position statement, it is not necessary to consider: the simultaneous occurrence of independent accident ~

conditions.

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5.O REPERENCES 1.

S.E.

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

2.

R.M.

Westfall, et.

al.,

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

A Modular Code System for cerformina Standardized

.4

' Comouter Analyses for Licensina Evaluation, NUREG/CR-0200,.

1979.

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

3.

R.M.

Westfall et al.,

" SCALE:

A Modular Code - System - for

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

4.

M.G.

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

d !

l 4

r

' g '.-

i Li

.j Table

SUMMARY

OF CRITICALITY: SAFETY ANALYSES-NEW FUEL VAULT - 5.0%' ENRICHED ~ FUEL

'(Under Accident: Conditions)

~

Optimum Flooded!

~

Moderation Condition-1 Temperature for. analysis 20*C (68'F) 20*C (68'F)

Reference k,

~(KEN 05a) 0.9648 0.9345-f Calculational bias, 8k 0.0103 0.0103 t

Uncertainties.

l In the-Bias")

t 0.0018 0.0018-KENO Statistics")

i 0.0010-t 0.0014

. Lattice spacing ~

t 0.0015

• 0.0007 i

Fuel enrichment t 0.0013 t 0.0014 Fuel density t 0.0015 t 0.0015-Statistical combination i 0.0029 i 0.0032 of uncertainties")-

l Total 0.9751-i 0.0029 0.9448 t 0.0032 i

Maximum Reactivity (k,,,)

0.9780 0.9480 t

Regulatory Limit

- 0.98 0.95 "3

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

"3 Square root of sum of squares.

i i

i b.i

f

.a i

TABLE 2 FUEL' ASSEMBLY SPECIFICATIONS L

i I

- FP'l Rod Data 3

Outside dimension, in.

0.430 Cladding ID, in.

0.377 i

Cladding thickness, in.

0.0265

)

Cladding material Zr-4 l

Pellet diameter, in.

0.369-3 UO. density, g/cm 10.420 t 0.166 2

Enrichment, wt.% U-235 5.0 0.02 i

Fuel Assembly Data

}

Number of fuel rods 208 (15x15 array)

Fuel rod pitch, in.

0.568-l Control rod guide tube Number 16-O.D.,-in.

0.530

~i Thickness, in.

0.016 Material Zr-4 i

Instrument thimble i

Number 1-0.D.,

in.

0.493 Thickness, in.

0.026 Material Zr-4 lo

.I 1

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)

i

I Not To Scale

~

~ ~ _

t ;O. c e y; n;.

.,a' &

> > " ~,.

N.

3

]

'5I FUEL ASSEMBLY SPACING = 21.125 in.

t 4

P

^

!,lk~,.

y~

1. 4

.g "g

7

, - {,.

C d

~...

.l o !.

s; a.

r

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1) N,

tt,7 m l.

W,...-

If unum uma men m

F 12 Ft. O in.

y

> :.t.. ik.Qq Wll +

v:$vi

%,? >

6%;. L.-.

'/

-"r

'.5 Ft. CONCRETE D' s

N.;.L..yg.,lp."J, i

f 7

Fig. 1 New Fuel Storage Vault Configuration 9.

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W;

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G J

3 0.85 4

a s.

/

)

+-

i e.ee f

r 0.75

)

(

/

0. 70.'

ye O.65 -

to 100 Percen t Modere tor De n e t ty, g/cc F tg. 2 Re ac t tv t ty Ver to t ton w tth Modere tor Do n e L ty 6

l t 1 APPENDIX A BENCHMARK CAIEUIATIONS.

bY Stanley E. Turner, PhD, PE BOLTEC INTERNATIONAL' November, 1993 1

9 e

d 4

9 9

l..

)

l 1

4 1

j i

TABLE OF CONTENTS J

5 A-1

1.0 INTRODUCTION

AND

SUMMARY

t 2.0 NITAWL-KENO 5a BENCHMARK CAI4ULATIONS A-2 r

3.0 CASMO3 BENCHMARK CAICULATIONS A-4 A=5 4.0 WORKER ROUTINE 5.0 CLOSE-PACKED ARRAYS

-A-6 A-7

6.0 REFERENCES

i i

F 4

i L

k

+

9 i

i

?

e t

h P

t 5

h List of Tables Table 1 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a...

A-9 CAICUIATIONS OF B&W CRITICAL EXPERIMENTS j

Table 2 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a A-10 i

CALCUIATIONS OF FRENCH and BNWL CRITICAL i

EXPERIMENTS l

Table 3 RESULTS OF CASMO3 AND NITAWL-KENO 5a.

A-11 BENCHMARK (INTERCOMPARISON).CAICUIATIONS Table 4 Intercomparison of WORKER-NITAWL-KEN 05a.

A-12 and CASMO3 Calculations at Various i

Temperatures Table 5 Reactivity Calculations for Close-Packed -.. A-13 Critical Experiments L

List of Floures Fig. 1 COMPARISON OF CASMO AND KEN 05A CAICUIATIONS... A-14 AT VARIOUS-ENRICHMENTS IN REPRESENTATIVE FUEL STORAGE RACK Fig. 2 COMPARISON OF CASMO3 AND KENO 5A

- A=15 TEMPERATURE DEPENDENCE ii 8

,,-n

.r

,l,

~

-j j

INTRODUCTION AND

SUMMARY

J 1.0

)

i The objective of this benchmarking study is to verify.

both the NITAWL-KEN 05a,2). methodology with the '27-group SCALE

'I d

0 cross-section library and the CASMO3 code ) for use'in criticality safety calculations of high density spent fuel. storage racks. Both calculational methods are based upon transport theory and have been -

benchmarked against critical experiments that simulate typical spent fuel storage rack designs as realistically as possible.-

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

Results of the benchmark calculations show that - the,

l I

27-group (SCALE) NITAWL-KEN 05a calculations consistently. under-predict the critical eigenvalue by 0.0103 t 0.0018 ok (with a 95%.

probability at a 95% confidence level) for critical experiments")

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

l l

Extensive benchmarking calculations of. critical experi-j ments with CASMO3 have also been reported ),

giving a mean k,,, of f

0 1.0004 i 0.0011 for 37 cases.

With a K-factor of 2.140) for 95%

probability at' a 95% confidence level, and conservatively neglect-

)

ing the small overprediction, the CASMO3 bias then becomes. 0.0000 l

i 0.0024. CASMO3 and MITAWL-KEN 05a intercomparison calculations of

,j infinite arrays of poisoned cell configurations (representative of j

typical spent fuel storage rack designs) show very good agreement, confirming that 0.0000 i 0.0024 'is a reasonable bias and uncertain-ty for CASMO3 calculations.

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

error.

.l 4

A-1 b

d

l 1

The benchmark calculations reported here confirm, that j

either the 27-group - (SCALE) ' NITAWL-XEN05a or ' CASMO3 calculations are acceptable for criticality analysis of high-density spent fuel storage racks.

Where possible', reference calculations for storage l

rack designs should be performed with both code packages to provide j

' independent verification.

CASMO3, however, is not reliable wh'en-large water gaps ( > 2 or 3 inches) are present.

2.0 NITAWL-EEN05a BENCHMARK CALCULATIONE Analysis of a

series of Babcock & Wilcox. critical-l experimente('), including some with absorber panels typical of - a i

poisoned spent fuel rack, is summarized in Table _1, as calculated with NITAWL-KEN 05a using the 27-group SCALE cross-section library and the Nordheim resonance integral treatment in NITAWL.

Dancoff_

l factors for input to NITAWL were calculated with the Oak Ridge SUPERDAN routine (from the SCALE (2) system of codes). The'mean for f

these calculations is 0.9899 1 0.0028 (1 o standard deviation'of

.l the population).

With a one-sided tolerance factor corresponding to 95% probability at a 954 confidence level ('),

the calculational t

bias is + 0.0103 with an uncertainty of the mean of i 0.0018 for the sixteen critical experiments analyzed.

Similar calculational deviations have been reported by-ORNd7) for some 54 critical experiments (mostly clean criticals'

'l i.

'I without strong absorbers), obtaining a mean bias of 0.0100 1 0.0013 (954/95%).

These published results are in good agreement with the j

results obtained in the present analysis and land further credence j

to the validity of the 27-group NITAWL-KENO 5a calculational model j

for use in criticality analysis of high density spent fuel storage racks.

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

gap, with absorber panel reactivity worth, with enrichment or with l

Poison concentration were identified with the 27 group SCALE library, comparable to those previously observeds) with the 123-t group GAM-THERMOS cross-section library.

i A-2 j

1 i

h?'

)

,t Additional benchmarking calculations were also made for j

p a series of French critical experiments") at'4.754 enrichment and ll

'for several of the BNWL criticals. With 4.26% enriched fuel.-

1

^ Analysis'of the French criticals-(Table 2) showed a tendency.to overpredict the reactivity, a result also obtained by'ORNL ').-

The d

calculated. k, values showed a trend toward ' higher values with g

decreasing core size.

In-the absence of a-significant enrichment:

effect. (see section 3 below), this trend and the overprediction is-attributed to a small inadequacy in NITAWL-KEN 05a in calculating

-l neutron leakage from very.small assemblies.

j i

Similar results were observed for the BNWL series.of

.j critical experimentedD, which are also small assemblies (although.

significantly. larger than the French criticals).

In this case I

k, was 0.9959 1 0.0013 (1: o (Table! 2),

the calculated mean g

population, standard deviation).

Because of the small siae o'f.the-l BNWL critical experiments (compared to the B&W criticals usedto determine the KEN 05a. bias) and the absence of any; significant i

enrichment offact, the results also suggest a small ina'dequacy of

[

NITAWL-KEN 05a in treating large neutron leakage from very small' assemblies.

Since the analysis of - high-density spent fuel storage racks _ generally does not entail neutron leakage, the observed inadequacy of NITAWL-KEN 05a: is not significant.

Furthermore, l

omitting results of the French and BNWL critical experiment.

-f analyses from the determination of bias is conservative since any

[

1eakage that might enter into the analysis would tend to result in overprediction of the reactivity.

i i

i I

i A-3 1

t m

S

.y

3.0 CASMO3 BENCHMARK CAIfDIATIONS The CASMO3 Lcode is a multigroup transport theory code utilizing transmission probabilities to accomplish two-dimensional calculations ' of reactivity 'and depletion ' for BWR and PWR. f uel '

assemblies.

As such,. CASMO3 is well-suited to the criticality

(

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

CASMO3 is a modification of the CASMO-2E code and has been extensively benchmarked against both mixed oxide and hot and cold critical. experiments by Studsvik Energiteknik('I.

Reported analyses (53 of 37 critical experiments indicate a mean k,'of 1.0004 g

i 0.0011 (le).

To independently confirm the validity of CASMO3 (and to investigate any effect of enrichment),' a series of calculations were made with CASMO3 and with NITAWL-KEN 05a on identical poisoned storage cells representative of high-density spent fuel storage racks.

Results of these. intercomparison calculations * (shown in Table 3 and in Figure 1) are within the normal statistical variation of KEN 05a calculations and confirm the bias of 0.0000 1 0.0024 (95%/95%) for CASMO3.

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

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

A-4

A second series of CASM03 and KEN 05a intercomparison calculations consisting of-five cases from the BAW critical-experiments were analyzed for the central cell only. The calculat-ed results, Lalso shown ~ in Table 3,

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

This

. lends further credence to the recommended bias for.CASM03.

J 4.0 WORKER ROUTINE The WORKER routine was obtained from ORNL and is intended

~

to interpolate tho' hydrogen scattering matrices for temperature in-

]

ordca to correct for the deficiency noted in NRC Information Notice

'i 91-66 (October 18, 1991). Benchmark calculations were made against.

-I CASMO3, based - on the assumption that two independent methods of -

.l

]

analysis would' not exhibit the same error.

Results of these calculations, shown in Table 4,

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

This agreement establishes the validity of the WORKER routine, in conjunction with j

NITAWL-KEN 05a, in calculating reactivities at temperatures between

]

20*C and 120*C.

The deficiency in the NITAWL hydrogen scattering matrix at temperatures above 20 *C does not appear except in the presence.

j of a large water gap where the scattering matrix is important.

{

However, the absolute value of the k= from CASMO3 is not reliable in the presence of a large water gap, although the relative values j

should be accurate.

In the calculations shown in Table 4 and in Figure 2, the absolute reactivity values differ somewhat but the j

trends with temperature are sufficiently in agreement to lend g

credibility to the WORKER routine ~over the temperature range from l

20*C to 120*C.

{

l i

l.

t l

i 5.

t A-5 i

f l

l

.t.

t l

5.0 CLOSE-PACKED ARRAYS i

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

Results of.these analyses, shown 'in Table 5,; suggest a slightly higher bias than that for fuel with normal lattice spacings. 'similar results were obtained by-ORNLU3)

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

This would conservatively encompass all but one of'the cases-i measured.

i i

.i J

?

i l

i 1

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i

)

A-6 k

h i

f i

I

~. -

1

j 6.O' REFERENCES TO APPENDIX A l

1 i

1.

Green, Lucious,. Patrie, Ford, - White,. and ' Wright, "PSR ;

/NITAWL-1. (code _ package) ~ MITAWL Modular Code System For

~!

Generating Coupled, Multigroup ~ Neutron-Gamma Libraries from i

ENDF/B",'ORNL-TM-3706, Oak Ridge National Laboratory, November

'i 1975.

'l 1

1 2..

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

NUREG/CR-0200, 1979.

i A ' Fuel' 3.

A.

Ahlin, M.

Edenius, and H.

Haggblom, "CASMO f

Assembly Burnup' Program", AE-RF-76-4158, Studsvik report.

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

Depletion Code for IMR Analysis", ANS Transactions, Vol. 26,

p. 604, 1977.

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

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

4.

M.N. Baldwin et ' al., " Critical Experiments - Supporting Close

'f Proximity. Water Storage of Power Reactor Fuel",-BAW-1484-7,.

l The Babcock & Wilcox Co., July 1979.

t 5.

M. Edenius and A. Ahlin, "CASM03: New Features, Benchmarking, and Advanced Applications", Nuclear Science and Enaineerina,.

l 100, 342-351, (1988) y

)

6.

M.G.

Natrella, Experimental Statistics, National Bureau of l

Standards, Handbook 91, August 1963.

J 7.

R.W. Westfall.and J. H.' Knight, " SCALE System Cross-section Validation with.. Shipping-cask Critical Experiments", ANS.

Transactions, Vol. 33, p. 368, November 1979 8.

S.E.

Turner and M.K.

Gurley,

" Evaluation. of NITAWL-KENO

)

Benchmark Calculations for High Density Spent Fuel Storage Racks",

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

lj g.7 a

l l

2-l

4.

9.-

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

50, pp 148, September 1980. 2 10.

A.M.

Hathout, et.

al.,

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

11.

S.R.

Bierman, et.

al.,

" Critical Separat5on between Sub-critical Clusters of 4.29 Wt. 4.2% Enriched UO Rods in Water 2

with Fixsd Neutron Poisons",

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

12.

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

13.

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

Department of Energy Office of Civilian Radioactive Waste Management Applications", Section 6, Fuel Consolidation Applications, ORNL/CSD/TM-247 (undated).

9 A-8 s

4

]

Table 1.

EESULTS 'OF 27-GROUP (SCALE) NITAWL-KEN 05a CAICUIATIONS OF B&W CRITICAL EXPERIMENTS 1

i Experiment

' Calculated e

Number k,,,

j I

I-0.9922-i O.0005

{

II 0.9917 i 0.0005 III 0.9931 i 0.0005 i

IX 0.9915 1.0.0006 X

0.9903 1 0.0006 XI 0.9919 i 0.0005, XII 0.9915 1 0.0006-XIII 0.9945

, i 0.0006 i

XIV 0.9902 i 0.0006 t

'i XV 0.9836 i 0.0006

.i XVI 0.9863 i 0.0006 l

t XVII

_0.9875 i 0.0006 t

7 XVIII 0.9880 i 0.0006-XIX 0,9882 i'O.0005

'f XX 0.9885 i 0.0006 i

XXI 0.9862 1 0.0006 i

[

Mean 0.9897 i 0.000703 Bias (95%/954) 0.0103 i 0.0018 i

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

l i

i A-9 l

i i

l i

,9,d Table 2 i

RESULTS OF 27-GROUP'(SCALE) NITAWL-KEN 05a CALCULATIONS I

OF FRENCH and BNWL CRITICAL EXPERIMENTS-French Experiments Separation critical Calculated-Distance, cm Height, cm k,

g

'O 23.8 1.0302 1 0.0008 2.5 24.48 1.0278 1 0.0007 l

5.0 31.47 1.0168 i 0.0007 10.0 64.34 0.9998.1 0.0007 BNWL Experiments calculated Case Erpt. No.

k, g

i No Absorber 004/032 0.9942 i 0.0007 SS Plates (1.05 B) 009 0.9946 i 0.0007 SS Plates (1.62 B) 011 0.9979 i 0.0007 l

SS Plates (1.62 B) 012 0.9968 i 0.0007 SS Plates-013 0.9956 i 0.0007 t

SS Plates 014 0.9967 i 0.0007 Zr Plates 030 0.9955 i O.0007 i

Mean 0.9959 i 0.0013 I

A - 10 i

~-

ipb

[

'I

-i 4:

4 Table 3 RESULTS OF CASMO3 AND NITAWL-KEN 05a

. BENCHMARK'(INTERCOMPARISON) CAICULATIONS i

l i

Enrichmentd3 l

Wt. % U-235 NITAWL-KEN 05a(38 k"CASNO3 l8kl 1

2.5 0.8376't 0.0010 0.8386 0.0010 l

3.0 0.8773 1 0.0010 0.8783

.0.'0010 l

3.5 0.9106lt 0.0010' O.'9097-0.0009 4.0

.O'.9367 i 0.0011

-0.9352 0.0015 i

4.5 0.9563 i 0.0011 0.9565 0.0002 l

f 5.0 0.9744 i 0.0011 0.9746 0.0002

'f Mean 0.0008 i

Expt. No.(3)

{.j XIII 1.1021 i 0.0009 1.1008 0.0013

.I

-i

.XIV 1.0997 i 0.0008 1.1011 0.0014 i

4 1

XV 1.1086 i 0.0008 1.1087 0.0001 l

^

XVII 1.1158 i 0.0007 1.1168 0.0010

'i XIX 1.1215 i 0.0007 1.1237 0.0022 I

Mean 0.0012 j

i i

03 Infinite array of assemblies typical of' high-density spent fuel storage racks.

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

.]

(3)

' (3)

Central cell from BAW Critical Experiments A - 11

,w...

,e..

i I

~

Table 4 Intercomparison of WORKER-NITAWL-KEN 05a and CASMO3 Calculations at Various Temperatures Temnerature CASMO3 W-N-KEN 05a(*)

3 4*C 1.2276 1.2345 1 0.0014

.17.5'c 1.2322 1.2328 1 0.0015 25'C 1.2347

'A.2360 1 0.0013 50*C 1.2432 1.2475 1 0'0014 75'c 1.2519 1.2569 i 0.0015' l

120*C 1.2701 1.2746 i 0.0014-I I

I L

• Corrected for bias 1

1 4

l l

l l

l A - 12

U.'

I e -

(he Table 5.

Reactivity.Calculationeifor'Close-Packed Critical.Experimente l

Calc.-

BAW

Pin:

Nodule.

Boron Calculated-i No.

Expt.

Pitch

' Spacing.

Conc.

k,,,

No.

cm.

cm ppm KS01 2500-

. Square' 1.792 1156 0.9891 i'O.0005 1.4097 KSO2 2505 Square 1.792 1068 0.9910 1 0.0005.-

'1.4097~

KS1-2485 Square 1.778 886 0.9845 i 0.0005

.Touching KS2 2491 Square 1.778 746 0.9849 1.0.0005' Touching l

KT1 2452 Triang..

1.86 435 0.9845.i 0.0006L Touching KT1A 2457 Triang.-

1.86 335 0.9865 i 0.0006 Touching l

KT2 2464 Triang.

2.62 361 0.9827 i 0.0006 Touching i

KT3 2472

.Triang.

3.39 121 1.0034 i 0.0006 i

Touching I

A - 13

.A

b.

1.00 t

e.9s

~

-z M

g 0.90 s

t u

CASN KENO-Se 0.85

~

3 0.80 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 FLEL ENRICHT:NT, WTs U-235 i

COPARISCW W KN CALCLA.ATIONS AT VARIOUS ENtIOfENTS IN REPRESENTATIVE FLEL STORA0E RACK 1

A - 14 l

1.28 1.27-.

i

/

1.26.

/

i

' T

--/

l 5

~

6 1.25' 5

- -f i

1.24 5

)Y 1.23 5 l 1.22 0

20 40 60 80 100 120 140 Te mp e r a tur e, Degrees C F ig. 2 COff*ARISON OF CAST 10-3 en d KEN 05.

TEtPERATURE DEPENDENCE I

T 4

A - 15 I

l I

s

6 1

P i

i ATTACHMENT 2 4

)

l i

1

/glm t

ce:

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e

~

i CRITICALITY SAFETY EVALUATION OF THE POOL A'-

i i-i SPENT FUEL STORAGE RACKS IN CRYSTAL RIVER UNIT 3 i

WITH FUEL OF 5.0% ENRICHMENT i

l Prepared for the FLORIDA POWER CORPORATION i

s by

}

i Stanley E. Turner, PhD, PE

.l 1

-i December 1993

.f i

i Holtec Project 3119530514 l

Holtec Report HI-931111 I

i 230 Normandy Circle 2060 Fairfax Ave.

Palm Harbor, FL 34683 Cherry Hill, NJ 08003 4

j s.

i

_f h

f% Ne k4 (a) t / --

L.

.r; TABLE OF CONTENTS

1.0 INTRODUCTION

1 I

2.0

SUMMARY

AND CONCLUSIONS 2

)

l 2.1 Normal Storage Conditions 2

l 2.2 Abnormal / Accident Conditions 3

3.0 CRITICALITY SAFETY ANALYSES.

4 4

I 3.1 Puel Assembly Specifications 4

3.2 Storage Rack Specifications 4

3.3 Manufacturing Tolerances and Uncertainties 4

l 3.4 Calculational Methodology 5

3.4.1 Computer Codes 5

3.4.2 Axial Distribution in Burnup 6

.3.4.3 Checkerboard Configuration.

7

4.0 REFERENCES

8 l

l l

l l

i r

a i

i l

_ _k

f; E]

- e.

I x

J l

l:

l r

List of Tables L

Table 1

SUMMARY

OF CRITICALITY SAFETY ANALYSES 9

POOL A STORAGE RACKS Table 2 FUEL' ASSEMBLY SPECIFICATIONS-10 L

1: -

List of Fiaures

[

F i g. l' ACCEPTABLE BURNUP DOMAIN IN POOL A 11 Fig. 2 Pool A FUEL STORAGE CELL 12 l

l ii

.q

1.0 INTRODUCTION

f The present study is part of.an evaluation of the fuel storage f acilities - at Crystal River. Unit-3 in -order to qualify the facilities for fuel of 5.0% average initial enrichment.

This report addresses the Pool A ' spent fuel pool while a companion report evaluates the_new-fuel vault (HI-931110).

I The CR3 Pool A storage racks are designed to accommodate spent fuel 4

which has' attained a minimum average burnup that is dependent on the average initial enrichment of the fuel assembly.

These racks use a B,C matrix absorber and were previously qualifiedm for fuel-q of 4.5% enrichment burned to 7.0 MWD /KgU and employ a water flux-trap between storage cells as a means of augmenting. reactivity f

control.

In the present study, the previous curve of limiting

]

burnups is extended to encompass 5.0% enriched fuel.

The effect of the ~ axial distribution in burnup has also been considered as i

specified by Regulatory Guide 1.13-(draft, Rev.2).

Calculations were made with both the CASE-3 program and the NITAWL-KENO-Sa code package.

CASE-3 was used for burnup and restart calculations and to define an equivalent enrichment for use in the KENO-Sa calculations.

Both normal'and accident conditions are assessed.

Credit for the soluble poison normally present in the pool water is allowable under accident conditions (double contingency principle).

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

]

" Prevention of Criticality in Fuel Storage and Handling",

the definitive criteria contained in the April 14, 1978 USNRC letter and draft Regulatory Guide 1.13 (Rev. 2) are applicable.

l 1

Calculations were also made to demonstrate the acceptability of a checkerboard loading pattern of fresh un-burned fuel of 5.0%

j enrichment..

o

2.O S'JteRRY AND CONCLUSIONS l

.2.1 Normal Storace Conditions r

The spent fuel storage racks in Pool A use stainless steel boxes to i

define the storage cells with a B C matrix neutron absorber of 4

areal density. A water-gap between the absorber j

0.015 gms B-10/cme panels affords a

flux-trap to ' augment ~ reactivity control.

Calculations were made for fuel burned to 10.5 MWD /KgU for fuel of 5.0% initial enrichment.

t Table i summarizes the criticality safety analysis for 5.0%

enriched fuel at a burnup of 10.5 MWD /KgU.

The maximum k,,,

is 0.9435, including uncertainties, which is within the Regulatory guideline (k,,, of 0.95) and is therefore acceptable.

The previous burnup limit curve W was extended to include 5.0% enriched fuel and the updated curve is shown in Figure 1.

The limiting burnup curve in Figure 1 is well described by a linear fit as shown below, and may be used to calculate the minimum burnup for any initial r

enrichment, E, up to 5.0%.

r

.i i

Acceptable Burnup in MWD /KgU j?

i

= 7.0

• E - 24.5 i

i This fit is the same as previously determined, extended to include 5.0% enriched fuel.

}

r Based upon the calculations reported here (see Table 1 and l

Figure 1), it is concluded that fuel of 5.0% initial enrichment is l

acceptable for storage in Pool A of the Crystal River Unit 3 spent fuel storage facilities, provided the fuel has attained a minimum l

burnup of 10.5 MWD /KgU. Minimum burnup specifications are shown in

, l

?

i f;

.,y-J i

Figure 1-for other enrichments' (from the previous. evaluation)iwith j

assurance that thei maximum. reactivity is within the L regulatory limit.

1

)

Calculations were also made for checkerboard ar,rangements' of fresh

-l 5.0% enriched fuel.

These calculations'show that a checkerboard arrangement.with' empty cells (i.e. filled only with water or.non-maximun f k,,,

of fissile bearing material) is acceptable. with a

}

0.833.

' l 2.2 Abnormal / Accident Conditions 3

The reactivity consequences of abnormal / accident conditions = were

[

considered in the previous analysis") and found to be within.

I{

acceptable-bounds.

However, with the higher enrichment fuel j

(5.0%), the consequence of a mis-placed fuel assembly could differ l

from that previously evaluated.

Calculations with' a mis-placed l

. fuel assembly (fresh assembly of 5.0% enrichment accidentally ;

]

loaded into a Pool A cell) resulted in a maximum. k,,,

of 0.946 i

(including uncertainties) with all other cells filled with fuel of i

the maximum permissible reactivity.

This is within'the Regulatory.

j guideline even without the allowable credit for soluble boron.

I a

I a

I J

)

i 3-l l

6 e'

l I

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j 3.0

-CRITICALITY SAFETYiANALYSES-d

. 3. 1 Fuel Assambly Boecifications:

)

'h The fuel assemblies'.used in^the analyses is th'e Babcock:E Wilcox' 15 x: 15 fuel. ' assembly, the same as that used 'in? th'e ~ previous; q

a'nalyses" 4 Table'2 attached lists the design' specifications for

]

the: fuel.usedlin'the analyses.

..n

- 3.2 Storace Rack Snecifications

}

1 The' storage rack cell design, illustrated in' Figure 2, 'is composed -

of B C absorber material sandwiched' between two 0.060 inch thick 4

stainless steel boxes of 8.9375 inch inside dimension.

The cells 1

.,i

- are arranged on a 10.50. inch.-lattice spacing l with a 1.173 inch

]

water. gap 1between the storage cells.-

The Lstainless ' steel - tabs.

j connecting.the storage boxes have a slightly'negativefreactivityj i

g ef feet and were neglected. in the. calculations. - The B C absorber j;

4 has a thickness of 0.075 ' inches and a B-10 loading of' 0iO15. t- 0.003-

- l

~

-i gms B-10/cnf.

l}

y 3.3 Manufacturina Tolerances and Uncertainties-l q

The small reactivity increments associa' ted.with manufacturing -

>+

d tolerances obtained in the previous evaluatio'n ) - (t' O.0097)" were ;

assumed to remain applicable.

Combined with the uncertainty in l

.I bias (t 0.0024, Appendix A) results'in a total uncertainty of t.

d 0.0100.

Fuel of 5.0% enrichment also requires an-increaselin the'-

l allowance for uncertainty in the depletion calculations. As in'the~

.l original evaluation for 4.5% fuel, the depletion uncertainty was:

j assumed to be 5.0% of the reactivity decrement from beginning-of -

}

life to the burnup of 10.5 mfd /KgU.

This allowance f amount,si to.

0.0035 Ak which is conservatively treated :as an additiive tem j

rather than being statistically combined with the other j

uncertainties.

j O

E 1

i e

O..

j

-Previous: calculations have demonstrated a continuous. reduction in reactivity with storage. time (after Xe decay)'primarily due to..Pu-241 decay and An-241 growth. - No credit is tahan for this reduction l

.in reactivity except: to acknowledge. an additional level of'

.l conservatisnin the calculations.

j 3.4 Calculational Methodoloav l

3.4.1 Computer Codes 3

- The principal'aethods of analysis were CASMO-3"), a two-dimensional I

multigroup transport theory code for. fuel assemblies and NITAWL -

KENO-Sa ), a three dimensional Monte Carlo code package, using'the ll d

27-group SCALE

• cross-section library.

The calculational: methods

'l i

used for the present evaluation are comparable to those used in the j

< original calculations, differing only in that updated versions of, I

the codes were used, ie, CASMO-3 rather than CASMO-2E, and KENO-5a t

rather than KENO-4..Results of these codes are not significantly l

different from.those of the earlier versions, and benchmarking of' the. updated codes resulted in a' bias of 0.0000 i 0.0024 for CASMO-3, l

and 0.0103 i 0.0018 for NITAWL - KENO-5a (95% probability, 95%

confidence leve143).

A summary. of the detailed bench-marking'-

)

analyses is included in Appendix A.

{

CASMO-3 was also used both'for burnup calculations and for restart calculations in the rack geometry.

Since' KENO-Sa cannot perform burnup

analyses, CASMO-3 is used' to define an equivalent enrichment, ie, the U-235 enrichment that-yields the same reactivity in the racks as the burned fuel.

It was found that an j

enrichment of 3.4% yields the same reactivity in the storage racks

{

as 5.0% fuel burned to 10.5 MWD /KgU.

Independent check calculhtion i

-l SCALE is an acronym for Rtandardized Computer Analyses for.

Licensing Evaluation, developed for the=USNRC by the Oak j

Ridge National Laboratory.

. j

.l

-l l

sy,

t-w Q: l S:

UM y',

R 3 :=

^

,; a g'.

the f refere'nce ! ; case with.~NITAWL-KENO-Sa. (3 24%1 Jequivalent :

for Q[

.enrichmeht)) gave - ai biasi corrected k;,, of10.9280;i t.0.0010i(without? '

1 g,

uncertainties)~ which confirms the CASN-31 calculation'_.(k,,of.0.9300)'.

~

g

'In the geometric'model.used in the calculations,' each fuel rod and its cladding were described explicitly in both-the 'CASIO-3 and ' KENO-Sa i models'.

Reflecting boundary conditionsf(zero neutron' current) were:

~

' used ini the1 radia11 direction which. has the effect-of creating lanc

-infinite Jarray.: of storage cells in X-Y, directions'.

In. the ~ KENO'-Sa r model,- the. actual' fuel ~ assembly length was : ' used v ini the ~ axial; i

direction, assuming thick (30 cm)-water reflectors' top and bottom.

Monte Carlo '(KENO-Sa) calculations inherently include.aLstatistical-uncertainty due'to the random nature of neutron tracking. To minimize-the statistical' uncertainty. 6f the KENO-calculated reactivity, a.

minimum of ' 500,000 neutron histories in 1000: generations ;of 500:

neutrons each, were: accumulated.in each calculation'.

3.4.2 Axial Distribution in.Burnup Initially, fuel loaded into the reactor will ~ burn with a - slightlyc

~

skewed-cosine power distribution. 'As_burnup' progresses,-theLburnup,

distribution will tend'to flatten, becoming more highly burned in the '

central regions than in the upper and lower. ends'.

At-high burnup, the more reactive fuel near: the ends of the fuel. assembly (Jess than-average burnup) occurs in regions of lower reactivity worth-due to.

neutron leakage. -Consequently, it would be expected that over most of the burnup history, distributed burnup fuel assemblies would' e

exhibit a slightly lower reactivity than that calculated for the average burnup.

As burnup progressen, the.. distribution, to some extent, tends to be self-regulating as controlled by the axial power:

distribution, precluding the existence of large :regi'ons of' significantly reduced burnup.

' L

_a

b Among others, Turner!5)- has provided ' generic analyses of the axial i

burnup effect based upon calculated and measured axial burnup

.i distributions.

These analyses confirm the. minor and ' generally negative reactivity offect of the axially distributed burnups at values less than about 30,000 Mwd /MtU. - Because a burnup of.only 10.5 MWD /KgU is necessary for 5.0% fuel, tho' reactivity consequence of the l

axial distribution in burnup will be slightly negative.

[

e 3.4.3 Checkerboard Configuration i

Fuel up to. 5.0%

initial enrichment may also be stored in a checkerboard pattern, alternating with cells filled with only water or non-fissile material.

For this case, the maximum reactivity, including uncertainties, was calculated to be 0.8331, which is well below the USNRC guideline.

e q

I

rs

_f

(

i i

j f

4.0 REFERENCES

-I 1.

S.E Turner, " Criticality Safety Analysis'of the CrystalT River Unit 3 Pool A Spent Fuel Storage Rack", Southern Science Office i

of Black & Veatch, SS-162.(Undated)-

- i 2.

'A.

Ahlin, - : M.

Edenius',.H. Haggblom, "CASMO

-A Fuel Assembly ?

. j Burnup Program," AE-RF-76-4158, Studsvik report (proprietary).

A. Ahlin and M. Edenius, "CASMO - A Fast-Transport-Thaory Depletion Code for LWR Analysis,"-

ANA._

Transactions, Vol.

26, p. 604, 1977.

5 M. Edenius'et al., "CASMO Benchmark Report,"

Studsvik/ RF-78-6293, Aktiebolaget Atomenergi, March 1978.

i 4

3.

R.M.

Westfall, et.

al.,

"NITAWL-St. Scale Systen Module' for' Performing Resonance Shielding and' Working ' Library Production" in scATR:

A Modular Code syste= for nerformina standardized f

Computer Analyses for Licensina Evaluation, NUREG/CR-0200,.1979.-

L.M. Petrie and N.F. Landers,? KENO 5a. An Improved Monte Carlo i

criticality Program with Supergrouping" in Scale: A Modular code system for nerformina standardized Connuter Analyses for Licensine Evaluation, NUREG/V-0200, 1979.

l R.M.

Westfall at al.,

"ScarR:

A Modular code' system for'

~

oerformina standardized Connuter Analyses for Licensina Evaluation. " NUREG/V-C'200. 1979.

j i

4.

M.G.

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

5.

S. E. Turner, " Uncertainty Analysis - Burnup Distributions", in l

Proceedinas of a Workshop on the use of Burnun Credit in Soent Fuel Transport Casks, Sandia Report SAND-89-0018, October 1989.

t

. i i

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g y

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--,+

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'f t.c r V

('G

'I:

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{

V t

Table 1

-l i

SUMMARY

.OF CRITICALITY SAFETY ANALYSES-l POOL A STORAGE RACKS i

Fuel Enrichment,.wtt U-235

-5.'00 4

I Design Burnup, MWD /KgU' 10.50 Reference Temperature,

'F 68

-l Reference k= (CAS h 3) 0.9300 l

i Calculational Bias, Ak")

'O.0000-

.i Axial'Burnup Distribution Negative Ak Uncertainties (2) 20.0100 i

Total 0.9300 10.0100 Ak' allowance for depletion to.00035 calculations")

~ l Maximum Reactivity (k=).

0.9435 Regulatory Limit ( k,,,

0.95

'i Appendix A l

d)

(2) Section 3.3

{

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r p

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5 TABLE 2 j;

I FUEL ASSEMBLY' SPECIFICATIONS.

g i

Fuel' Rod Data Outside dimension, in.

-0.430-l 0.377 Cladding ID,.in.

i 0.0265 i-Cladding thickness, in.

l Zr-4 l

' Cladding material Pellet diameter,.in.

0.369 r

10.420 t 0.166 UO, density, g/ca Enrichment, wt.% U-235 5.0 t 0.02 r

Fuel Assembly Data i

208 (15x15 array)

Number of fuel rods Fuel rod pitch, in.

~0.568 i

Control rod guide tube Number 530 0.D.,

in.

Thickness, in.

0.016 EE-4' I

Material-a Instrument thimble 1

Number 0.D.,

in.

0.493 Thickness, in.

0.026 Zr-4 Material i

I

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i e !

4

f l

11 l

,I 10

/

/

I 9-i A,

I B

ACC EPT ABL E B JRNJP

/

i 33 D3NA IN

/

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r s

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t 4

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3' r

/

UIJACCFAIR_ : P " 'Nt P

/

3)ftI 4

~ ~-

l 2

1 3

0 1

3.0 3.5 4.0 4.5 5.0 1

INITIAL FLEL DRIOtENT. Wts U-235 1

Fle. 1 Acceptable Burn up Dome tn tn Pool A i

9 11 -

(

i e

f

I

--4---------

- - 6.687

• 0.063" B4C WATRIX -- ---------l---

2; l

1 4

l 0.075" THK.015 i 0.003 gmB-10/cm e

B&W 15x15 FUEL l

.j O

.i I

3 '"-

DENSITY 10.42 g/cc 1

OO suMBER 208 (is x iS) l 1'

000 Cuo i0 0.377 iN.

i OOO ROD OD 0.430 IN.

l i

00900 PITCH 0.568 IN.

i i

000000 THlWBLES O.060 SS i

O000000 M 0.530IN.~~-*

lNNER & OUTER 90X i

O0000000 70,0 j'78IN.

i 009009OO

"" ? :-

l l

0000000000 00090000000 000000000900 00000^8.937" 80X l.D.^^^^^O0 gt 10.50" P1TCH

!i i

NOT TO SCALA l

l FIG. 2 CR3 P00L A FUEL STORAGE CELL

-APPENDIX A BENCHMARK CAICUIATIONS i

by

\\

)

I Stanley E. Turner, PhD, PE i

HOLTEC INTERNATIONAL I

November, 1993 6

e 4

9

n.

t TABLE OF CONTENTS t

1.0 INTRODUCTION

AND

SUMMARY

A-1 2.0 NITAWL-KEN 05a BENCHMARK CAICUIATIONS A-2 b

3.0 CASMO3 BENCHMARK CAICUIATIONS A-4 A=5 4.0 WORKER ROUTINE t

5.0 CIDSE-PACKED ARRAYS A-6 t

A-7

6.0 REFERENCES

1 i

1 i

f e

i P

i h

i

r:

List of Tables Table 1 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a.

A-9

[

CAICULATIONS OF B&W CRITICAL EXPERIMENTS i

f Table 2 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a

. A-10 CAI4ULATIONS OF' FRENCH and BNWL CRITICAL EXPERIMENTS i

Table 3 RESULTS OF CASNO3 AND HITAWL-KEN 05a....

. A-11 BENCHMARK (INTERCONPARISON) CAICUIATIONS Table 4 Intercomparison of NORKER-NITAWL-KENO 5a... A-12 and CASMO3 Calculations at Various Temperatures Table 5 Reactivity Calculations for close-Packed

. A-13 Critical Experiments List of Flaures I

e Fig. 1 COMPARISON OF CASNO AND KEN 05A CAICUIATIONS.. A-14 AT VARIOUS ENRICHMENTS IN REPRESENTATIVE

-FUEL STORAGE RACK Fig. 2 COMPARISON OF CASNO3 AND KEN 05A A=15 TEMPERATURE DEPENDENCE i

i l

ii 5

k

ppx j

-l su

'1.0

' INTRODUCTION AND

SUMMARY

I E

The ' objective of this benchmarking study. 'is to -verify L

both-.the NITAWL-KEN 05a

>3)' methodology with ' the ~ 27-group. SCALE d

cross-section library and.the CAsM03 code (3) for use in. criticality.

safety calculations of high density. spent fuel storage racks. Both l

calculational methods are based upon transport theory and have been'

{

benchmarked against critical experiments that, simulate typical-spent fuel storage ~ rack designs as realistically as.- possible..

Results of these' benchmark calculations with both methodologies are j

consistent 'with corresponding calculations reported in the literature.

Results of.the benchmark calculations show that.the j

27-group (SCALE) NITAWL-KENO 5a calculations consistently 'under-l predict the. critical eigenvalue by 0.0103 1 0.0018 8k (with a 954 l

probability at a.954 confidence level) for critical experiments (')

l that are as representative as possible of. realistic spent' fuel i

storage rack configurations and poison worths.-

Extensive benchmarking calculations of. critical experi-monts with CASMO3 have also been reported (5), giving ' a mean k,,, of f

1.0004 i 0.0011 for 37' cases.

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

[

probability at a 95% confidence level, and conservatively neglect-ing the small overprediction, the CASM03 bias then becomes 0.0000 l

t 0.0024. CASMO3 and NITAWL-KENO 5a intercomparison calculations of.

i infinite arrays of poisoned cell configurations (representative of typical spent fuel storage rack designs) show very good agreement, confirming that 0.0000 1 0.0024 is a reasonable bias and uncertain-l ty for CASMO3 calculations.

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

error.

M A-1

+.

- - -. + - -

+

rw~

~ -.

-. _ ~ -.

lj.

J s

t The benchmark calculations-reported. here confira.:that-either2 the 27-group (SCALE) NITAWL-KEN 05a. or CASNO3 calculations are acceptable for criticality analysis of high-density spent fuel i

1 storage racks.

Where ~possible, reference calculations for storage =

j

- rack designs should be performed with both code packages to provide independent verification.

CASNO3, however, is not reliable.when j

large water gaps'( > 2 or 3 inches) are present.

l 1

2.0 NITAWL-KEN 05a BENCHMARK CALCUIATIOl($

l Analysis of a series of Babcock & Wilcox critical j

experiments *, including some with absorber panels typical of a.

j poisoned spent fuel rack, is summarized in Table.1, as calculated' with NITAWL-KENO 5a using the 27-group SCALE cross-section library

~

and the Nordheim resonance integral treatment in NITAWL.

Dancoff I

factors for input to NITAWL were -. calculated uith the Oak Ridge

. i SUPERDAN routine (from the SCALE

• system of codes). The mean for f

these calculations is.0.9899 i 0.0028 (1 e standard deviation of the population).- With a one-sided tolerance. factor corresponding to 95% probability'at a' 954 confidence level *r the calculational

. 1 bias is + 0.0103 with an uncertainty of the mean of'i 0.0018 for the sixteen critical experiments analyzed.

l Similar. calculational deviations have been reported by:

. j ORNLM for some 54 critical experiments (mostly clean criticals without strong absorbers), obtaining a mean bias of 0.0100 1 0.0013 l

(95%/95%).

These published results are in good agreement with the j

results obtained in the present analysis and land further credence l

to the validity of the 27-group NITAWL-KEN 05a calculational model for use in criticality analysis of high density spent fuel storage-l racks.

No abnormal deviations in k,,, with. intra-assembly water gap, with absorber panel reactivity worth, with enrichment or with poison concentration were identified with the 27 group SCALE library, comparable to those previously observed

• with the 123-I1 group GAN-THERNOS cross-section library.

l A-2 l

t I

i p.

p.__

q-...-.

w.

<w s-7,,

c_

g-,.,--.--gp e

'1 Additional benchmarking calculations were also made for j

a series of French critical ~experimentam at.4.75% enrichment and for several of the BNWL criticals-with 4.26% enriched fuel.

Analysis of -the French criticals (Table 2) showed - a tendency to.

overpredict the reactivity, a result also obtained by ORNL(*.

The calculated k,,,

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

In the absence of a significant enrichment effect (see Section 3 below), this trend and the overprediction is attributed to a. small inadequacy. in NITAWL-KEN 05a in calculating neutron leakage from very small assemblies.

similar results were observed for the BNWL series of critical experimentstm, which are also small assemblies (although significantly larger than the French criticals).

In this case (Table 2),

the calculated mean k,,, was 0.9959 i 0.0013 (1 o population standard deviation).

Because of the small size of the BNWL critical experiments (compared to the B&W criticals used to determine the KEN 05a bias) and the absence of any significant enrichment effect, the results also suggest a small inadequacy of j

NITAWL-KENO 5a in treating large neutron leakage from very small assemblies.

I Since the analysis of high-density spent fuel storage racks generally does not entail neutron leakage, the observed inadequacy of NITAWL-KEN 05a is not significant.

Furthermore, omitting results of the French and BNWL critical experiment analyses from the determination of bias is conservative since any leakage that might enter into the analysis would tend to result in overprediction of the reactivity.

l i

r f

A-3 t

i t

3.0 CASMO3 BENCHMARK CALCULATIONS The CASMO3' code is a multigroup transport theory code

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

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

CASMO3 is a modification of the CASMO-2E code and has

.j been extensively benchmarked against both mixed oxide'and hot and cold critical experiments by Studsvik Energiteknik s)

Reported t

analyses (53 of 37 critical experiments indicate a mean k,,,'of 1.0004 1 0.0011 (10).

To independently confirm the validity of CASM03 (and to investigate any effect of enrichment),

a series of-calculations were made with CASMO3 and with NITAWL-KEN 05a on identical poisoned storage cells representative of high-density spent fuel storage racks.

Results of these. intercomparison calculations * (shown in Table 3 and in Figure 1) are. within the normal statistical variation of KEN 05a calculations and confirm the bias of 0.0000 1 0.0024 (95%/95%) for CASMO3.

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

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

A-4 a

b

.7

- - _ ~..

~.

=

1 A secondl series of CASMO3 and ' KEN 05a, intercomparison calculations ' consisting _ of five. cases from the' BAW critical 1 experiments were analyzed for the central call only. The calculat-ed results, also. shown in Table -3, indicate a, mean ' difference :

within the 95% confidence limit of the KENO 5a calculations.

This lands further credence to the recommended bias for CASMO3.

5 4.0 WORKER ROUTINE The WORKER routine was obtained from ORNL and.is intended.

I to interpolate the hydrogen scattering matrices for temperature in j

order to correct for the deficiency noted in NRC Information Notice

)

91-66 (October 18,.1991). Benchmark calculations were made against 1

CASMO3, based on the assumption that two independent methods of analysis would not exhibit the same error.

Results of. these

'I calculations, shown in Table 4,. confirm that the trend with temperature obtained by both codes are comparable.

This agreement j

establishes the validity of the WORKER routine,' in conjunction with-j NITAWL-KEN 05a, in calculating reactivities at. temperatures between j

20*C and 120*C.

l The deficiency in the NITAWL hydrogen. scattering matrix f

at temperatures above 20 *C does not appear except in the presence'

)

of a large water gap where the scattering -matrix is important.

i However, the absolute value of'the km from CASMO3 is not reliable in the presence of a large water gap, although the relative. values should be accurate.

In the calculations shown in Table 4'and in:

i Figure 2, the absolute reactivity values differ somewhat but the j

]

trends with temperature are sufficiently in agreement to land credibility to the WORKER routine over the temperature range from

-l 20*C to 120*C.

l'!

j A-5 i

i f

i

=

u:. :

5.O CLOSE-PACKED. ARRAYS The BAW close-packed series of critical experiments"2) intended to simulate consolidated fuel, were analyzed with NITAWL-KEN 05a.

Results of these analyses, shown in Table 5, suggest a w'th normal lattice slightly higher bias than that for fuel i

spacings.

similar results were obtained by ORNLW)

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

This would conservatively encompass all but one of the cases measured.

l-

\\

l l

A-6 l

I

.l 1

l 1

_j

,c 2

a >

h

6.0 REFERENCES

TO APPENDIX A 9

q 1.

Green, Lucious,- Petrie, Ford, E White, and Wright, "PSR-63--

i

/NITAWL-1.(code package) NITAWL - Modular Code System For-Generating Coupled Multigroup Neutron-Gamma Libraries from'

i ENDF/B", ORNL-TM-3706, Oak Ridge National Laboratory, November.

'1975.

l

. 2.

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

i NUREG/CR-0200, 1979.

. 3.~

A.

Ahlin, M.

Edenius, and H..Haggblon,. "CASMO'

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

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

p. 604, 1977.

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

Studsvik/NFA-87/7, Studsvik Energitechnik AB, November.1986 h

4.

M.N. Baldwin et / al., "Criticall Experiments Supporting ' Close ;

i Proximity Water Storage of Power Reactor Fuel",_ BAW-1484-7, 1

The Babcock & Wilcox Co,, July 1979.

5.

M. Edenius and A. Ahlin,."CASMO3: New Features, Benchmarking,

~

and Advanced Applications", Nuclear Science and Enaineerina,'

100, 342-351, (1988)_

6.

M.G. Natrella, Experimental Statistics, National' Bureau, of l

Standards, Handbook 91, August 1963.

7.

R.W.

Westfall and J. H. Knight, " SCALE System Cross-section Validation with Shipping-cask Critical Experiments", Algi,,

Transactions, ~ vol. 33, p. 368, November 1979 l

8.

S.E.

Turner and M.K.

Gurley,- " Evaluation of NITAWL-KENO' Benchmark Calculations for High Density Spent Fuel Storage i

Racks",

Nuclear Science and Enaineerinct,- 80(2):230-2,37, i

February 1982.

i A-7 I

d

,,. l

, _ ~ ~-

-- f'

fN m,-

?

j i

'9.

'J.C. Manaranche, et. al., " Dissolution and Storage Experiment with 4.75% U-235 Enriched U0, Rods", Nuclear Technoloav, Vol..

4-50, pp 148, September 1980.

10.

.A.M.

Hathout, et. al., " Validation ' of Three Cross-section:

' Libraries Used with the SCALE System for criticality ' Analy.

1 sis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981.

j q

11.

.S.R.

Bierman,

'e t. -

al.,

" Critical Separation between Sub-235 critical clusters of 4.29 Wt. 4 U Enriched.UO,' Rods in Water a

with Fixed Neutron Poisons",

Battelle Pacific Northwest J

Laboratories, NUREG/CR/0073, May 1978- (with August 1979 errata).

12.

.G.S. Hoovler, etLal., " Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations ofiSpent Fuel Pins", BAW-1645-4, Babcock & Wilcox company'(1981).

13.-

R.M. Westf all, et al., " Assessment of Criticality Computation.

]

al Software for the U.S.

Department' of Energy Office of Civilian Radioactive Wasta Management Applications", SectionT 6, Fuel Consolidation Applications, ORNL/CSD/TM-247 (undated).

i 1

i

.g a

1 A-8 b

c 3

~:

, n s;

. Table 1-

- RESULTS 'OF 27-GROUP (SCALE) NITAWL-KENO 5a CAIrULATIONS' OF B&W CRITICAL EXPERIMENTS

. Experiment

. Calculated.

o

. Number-k,,,

1 a

I 0.9922.

t 0.0006 II 0.9917 i'O.0005' III 0.9931.

i 0.0005

.IX 0.9915 1 0.0006 X

0.9903 1 0.0006 XI 0.9919 i 0.0005 XII 0.9915 i 0.0006 XIII 0.9945-

, i 0.0006 XIV 0.9902 i 0.0006 XV 0.9836 i 0.0006 XVI 0.9863 1 0.0006 i

XVII 0.9875 i 0.0006 XVIII 0.9880 i 0.0006 XIX 0.9882 1 0.0005

-l XX 0.9885 i 0.0006 XXI 0.9862 1 0.0006 Mean 0.9897 1 0.00070)

Bias (95%/95%)

0.0103 i 0.0018 "I

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

A_9 s

4 s

i s4-Table 2 RESULTS OF 27-GROUP (SCALE) NITAWL-KEN 05a CAI4ULATIONS OF FRENCH and BNWL CRITICAL EXPERIMENTS French Experiments separation Critical Calculated Distance, cm Height, cm k,,,

0 23.8 l'.0302 i 0.0008 2.5 24.48 1.0278 i 0.0007 5.0 31.47 1.0168 i 0.0007 10.0 64.34 0.9998 i 0.0007 BNWL Experiments calculated Case Expt. No.

k,,,

No Absorber 004/032 0.9942 i 0.0007 SS Plates (1.05 B) 009 0.9946 i 0.0007 SS Plates (1.62 B) 011 0.9979 i 0.0007 SS Plates (1.62 B) 012 0.9968 i 0.0007 SS Plates 013 0.9956 f.0.0007 SS Plates 014 0.9967 i 0.0007 Zr Plates 030

.0.9955 i 0.0007 f

Mean 0.9959 i 0.0013 A - 10 h

e

.c.

Table 3 RESULTS OF CASMO3-AND NITAWL-XEN05a l

' BENCHMARK.(INTERCOMPARISON) CALCUIATIONS j

. i Enrichmentd8 Nt. % U-235.

NITAWL-KENO 5a )

k"CASMO3 l8kl ta 2.5 0.8376 i 0.0010 0.8386~

0.0010 i

3.0 0.8773'i 0.0010 0.8783 0.0010 3.5 0.9106 i 0.0010 0.9097 0.0009 4.0 0.9367 1 0.0011 0.9352 0.0015 j

4.5 0.9563 i 0.0011 0.9565 0.0002 5.0 0.9744 i'0.0011 0,9746 0.0002 i

Nean 0.0008 i

Expt. No.(38 i

i XIII 1.1021 1 0.0009 1.1008 0.0013 XIV 1.0997 i 0.0008 1.1011 0.0014 XV 1.1086 i 0.0008 1.1087 0.0001 XVII 1.1158 i 0.0007 1.1168 0.0010' i

XIX 1.1215 i 0.0007 1.1237 0.0022 4

Nean

-0.0012-(

l Infinite array of assemblies typical of high-density spent fuel

~

"8 storage racks.

~

.l

. tr>

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

(38 Central Cell from BAW Critical Experiments A - 11 1

e

je l,

~*

1 Table 4 i

i Intercomparison of WORKER-NITAWL-KEN 05a and CASMo3 Calculations at various Temperatures l

Igmoerature CASMO3 W-N-KEN 05m(*)

4*C 1.2a76 1.2345 i 0.0014 f

17.5'C.

1.2322 1.2328 i 0.0015 t

25*C 1.2347 1.2360 1 0.0013 50*C 1.2432 1.2475 1 0.0014-75'C 1.2519 1.2569 i 0.0015 120*C 1.2701 1.2746 i 0.0014

• Corrected for bias t

i o

i r

i i

r A - 12 f

i a

l 1

P 9,-

i Table 5 t

Reactivity Calculations-for Close-Packed Critical Experiments i

Calc.

BAW

Pin Module Boron' Calculated No.

Expt.

Pitch Spacing.

k,,,

Conc.

No.

cm en ppa i

-KS01 2500-Square 1.792 1156 0.9891't 0.0005' 1.4097 l

i KS02 2505 Square 1.792 1068 0.9910 1.0.0005 1.4097 KS1 2485 Square 1.778 886 0.9845 i 0.0005

[

Touching KS2 2491 Square 1.778 746 0.9849 i 0.0005 i

Touching l

KT1 2452 Triang.

1.86 435

.0.9845 1.0.0006 Touching j

i KT1A 2457-Triang.

1.86 335 0.9865 1 0.0006 l

Touching KT2 2464 Triang.

2.62 361 0.9827 1 0.0006 Touching

-j i

KT3 2472 Triang.

3.39 121 1.0034 i 0.0006-Touching I

l l

)

I i

A - 13 5

4 i

E i

1.00 W

C 2

0.95

.I g

w 3

z f

a S 0.90 H

n u

i CAST 1)

KEN} 5.

0.85 3

I

~

~

0.80 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 Ft.EL ENRIOfENT, WTs U-235 I

Fts. 1 I.

CorPARISON OF CASMO APO KEN 06. CALCU.ATIONS AT VARIOUS ENtIOtENTS IN REPRESENTATIVE RKl. STORAGE RACK i

e e

A - 14 k

I 9

e i

J

6-i 1.28

+

1.27 i

/

i t

p,0

'-i 1.26_

ry j

/

3, s

/

• " 1.25 i

'Y /

~

1.24 e

)-

1.23 3!

1.22' i

O 20 40 60 80 100 120 140 Te mp e r a tu r e, Degrees C I

Fto. 2 cot 1'ARISON OF CASf10-3 end KEN 05.

TEtPERATURE DEPENDENCE i

P f

h A - 15 i

i 8

ATTACHMENT 3 1

4 I