ML20246D687

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Rev 1 to Criticality Analysis of Byron & Braidwood Station High Density Fuel Racks
ML20246D687
Person / Time
Site: Byron, Braidwood, 05000000
Issue date: 08/14/1989
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COMMONWEALTH EDISON CO.
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NUDOCS 8908280224
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k f hczCI1/?)C *1 Y CRITICALITY ANALYSIS OF. BYRON AND BRAIDWOOD STATION I

HIGH DENSITY FUEL RACKS REPORT PREPARED FOR COMMONWEALTH EDISON COMPANY REVISION 1 AUGUST 14, 1989 NUCLEAR SAFETY-RELATED f

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1 Table of1 Contents.

1 vp Section.

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. INTRODUCTION ~

12.0-DESIGN BASIS q

.3.0:

SUMMARY

LOF CRITICALITY ANALYSES 3.1:

LNormal Operating Conditions

3. 2 '

Abnormal and Accident Conditions 13 3'.

New Fuel Storage 4.0 REFERENCE FUEL STORAGE CELL

'4.1

' Reference Fuel Assembly.

42-Region'1 Storage Cells 14.3>

. Region 2 Storage Cells'.

5.0-ANALYTICAL METHODOLOGY I

5.1 Analytical Methods and Bias =

5.1.1. Reference 1 Analytical Methods and Bias 5.1.2 Analytical' Methods and Bias for Modified Cases 5.2-Fuel Burnup' Calculations

.5.3.

. Effect of _ Axial Burnup Distribution 5.4 Long-Term Decay 6.0 REGION 1 CRITICALITY ANALYSIS AND TOLERANCE VARIATIONS 6.1.

Nominal Design Case 6.2 Boron Loading Variation

'6.3 Storage Cell Lattice Pitch Variation 6.4, Stainless Steel Thickness Tolerances 6.5" Fuel Enrichment and Density Variation 6.6 Boraflex Width Tolerance Variation 6.7-Axial Cutback of Boraflex n- ^

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' REGION 2 CRITICALITY ANALYSIS AND TOLERANCE VARIATIONS

- 7.1 Nominal' Design Case 7.2 Boron Loading Variation 73 Storage Cell Lattice Pitch Variations 7.4 Stainless Steel. Thickness Tolerance 7.5 Fuel Enrichment and Density Variation 7.6 Boraficx Width Tolerance 8.0 ABNORMAL AND ACCIDENT CONDITIONS 8.1 Eccentric Positioning of Fuel Assembly in Storage Rack 8.2 Temperature and Water Density Effects 8.3 Dropped Fuel Assembly Accident 8.4 Abnormal Location of a Fuel Assembly 8.5 Lateral Rack Movement 9.0 NEW FUEL STORAGE 9.1-Storage in Region 1 Dry 9.2 Storage in Region 2, Flooded 9.3 Storage'in Region 2, Dry 10.0 BORAFLEX SHRINKAGE 11.0

SUMMARY

AND CONCLUSIONS REFERENCES TABLES FIGURES APPENDIX A BENCHMARK CALCULATIONS t_ __ z-

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1.0 INTRODUCTION

This report describes the criticality analysis of high density spent fuel storage racks manufactured by Joseph Oat Corporation (Oat) for the Byron and Braidwood Stations Units 1 and 2.

The plants are owned and operated by Commonwealth Edison Company (CECO). The design, fabrica-tion, nuclear criticality analysis, thermal / hydraulic analysis, struc-tural enalysis, accident analysis, environmental analysis, and cost-benefit appraisal of the high density spent fuel racks are contained in the reports, " Licensing Report on High Density Spent Fuel Racks for Byron Units 1 and 2 (NRL Docket Nos. 50-454 ASS) Revision 2 and Braidwood Units 1 and 2 (NRC Docket Nos. 50-456 A57) Revision 0."

Chapter 4, Nuclear Criticality Analysis, of these licensing reports is I

combined and presented in this report, Sections 2 through 9, in compli-ance with Definition 1.9a and in accordance with Adn.inistrative Controls Section 6.9.1.10 of the Byron /Braidwood Technical specifica-tions.

In addition, a section (10) on Boraflex shrinkage is included to incorporate a later response to the NRC question on this subject.

Section 11 summarizes the results of the criticality analysis. The analysis results indicate that the spacing of the fuel assemblies and the neutron absorbers - Boraficx plus Boral inserts for Region 1 and Boraflex for Region 2 - are adequate to maintain the array in' a sub-critical condition when fully loaded and flooded with nonborated water except for the inadvertent misplacement of a new fuel assembly in Region 2.

For this case, the presence of 300 ppm soluble boron ensures the desired amount of suberiticality is maintained. The maximum l

reactivity is always maintained below 0.95 as required.

l 2.0 DESIGN B ASIS The high density spent fuel storage racks for the Byron /Braidwood Nuclear Power Stations are designed to assure that the neut-on multi-I plication factor (keff) is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity in each of two regions, and flooded with unborated water at temperature corres-l

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ponding to the highest reactivity. :The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations and in mechanical-tolerances, statistically combined, such that the true kerf I' i will-be' equal to or less than 0.95 with 'a 95% = probab111ty at a 955 confidence level.

Applicable codes, standards, and regulations, or ' pertinent sections I

thereof, include the following:

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

o

_ USNRC-Standard Review Plan, NUREG-0800, Section' 9.1.'1, New Fuel Storage, and Section 9.1.2, Spent Fuel. Storage.-

I o

USNRC letter of April 14,.1978,- to all Power Reactor Licensees-- OT Position for Review and Acceptance of Spent l

-Fuel Storage.and Handling Applications, including modifica-tion letter. dated January 18, 1979.

o USNRC Regulatory Guide 1.13, Spent Fuel, Storage Facility Design Basis, Rev. 2 (proposed), December 1981.

o USNRC Regulatory, Guide 3.41, validation of Calculational Methods for Nuclear Criticality Safety-(andrelated ANSI N16. 9-1975 ).

o ANS I /ANS-57. 2-1983, Design Requirements for Light Water Reactor.. Spent Fuel Storage Facilities at Nuclear Power Plants.

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~o ANSI N210-1976.. Design Objectives 'for Light Water Reactor' I

Spent Fuel Storage Facilities at Nuclear Power Plants.

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

To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were mader o

Moderator is pure unborated water at a temperature corres-l I

ponding to the highest reactivity.

o Lattice of storage racks is assumed infinite in all direc-

tions, i.e., no credit is taken for axial or radial neutron leakage (except in the assessment of certain abnormal /

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accident conditions).

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

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The design basis fuel'. assembly is a 17x17 Westinghouse optimized. fuel Lassembly containing UO ' atl a. maximum' initial enrichment of 4.2%' U-235:

2 by weight, corresponding to 48.6 grams U-235 per axial centimeter of-i..

' fuel assembly. Two separate storage regions are provided in the spent fuel storage pool,-with separate criteria defining the: highest antici-'

pated reactivity in. each of the. two regions as follows:

-o-Region 1.is designed to accommodate new fuel with a maximum

' enrichment of 4.2 wt5 U-235, or spent fuel regardless.of the discharge fuel burnup.

o Region-2 is designed to accommodate fuel of various iriitial-enrichments which'hase accumulated minimum 'burnups within an deceptable bound as depicted in Figure 3.1.

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SUMMARY

OF CRITICALITY ANALYSES 3.~ 1 Normal Operating Conditions The criticality analyses results of each of the two separate regions of the spent fuel storage pool are summarized in Tables 31 and: 3 2 for the anticipated normal storage. conditions. The. calculated maximum reactivity. in Region 2 includes a burnup-dependent allowable for uncer-

-tainty in depletion calculations and,. furthermore, provides an addi-tional. margin of more than 25 Ak below the limiting effective multipli-

-cation factor. (keff) of 0.95.

As cooling time increases in long-term storage, decay of Pu-241 results in a significant decrease in reacti-vity, which will provide an increasing suboriticality margin and tends to further compensate for any uncertainty in depletion calculations.

Spacing between the two different rack modules is sufficient to pre-l l

clude adverse nuclear interaction between modules.

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i Region 2 can accommodate fuel of various initial enrichments and dis-charge fuel burnups, provided the combination falls within the accept-able domain illustrated in Figure 3.1.

For convenient reference, the minimum burnup values in Figure 3.1 have been fitted by linear tangents at various values and the results are tabulated below.

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8-14-89 Initial Minimum Initial

' Minimum Enrichment, %

Burnup, MWD /MTU Enrichment, 5 Burnup, MWD /MTU l

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3.00 22,490 1.80 5,230 3.20 25,000 2.00 8,570 3.40 27,510 2.20 11,340 3.60 30,020 2.40 14,390 3.80 32,540 2.60 17,-050 4.00 34,960 2.80 19,700 4.20 37,370 Linear interpolation between the tabulated values will always yield values on or conservatively above the curve of limiting burnups.

These data will be implemented in appropriate administrative procedures to assure verified burnup as specified in draft Regulatory Guide 1.13, Revision 2.

Administrative procedures will also be employed to confirm and assure the presence of soluble poison in the pool water during fuel handling operations, as a further margin of safety and as a precaution in the event of fuel misplacement during fuel handling operations as discussed-in Section 3.2.

32 Abnormal and Accident Conditions Although credit for the soluble poison normally present in the spent fuel pool water is permitted under abnormal or accident conditions,"

most abnormal or accident conditions will not result in exceeding the limiting reactivity (k,pg of 0.95) even in the absence of soluble poison.

The effects on reactivity of credible abnormal and accident conditions are summarized in Table 3 3 Of these abnormal / accident conditions, only one has the potential for a more than negligible positive reactivity effect.

The inadvertent misplacement of a new fuel assembly (either into a Region 2 storage cell or outside and adjacent to a rack module) has the

" Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978, NRC letter (Section 1.2) and implied in the proposed revision (draft) to Regulatory Guide 1.13 (Section 1.L, Appendix A).

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_ potential 'for ' exceeding the limiting reactivity should there. be a' concurrent: and" independent accident condition-resulting 'in?the loss 'of all soluble poison., Administrative procedures to assure the. presence of. soluble poison daring fuel handling operations' will preclude the possibility cf the simultaneous occurrence of these' two independent-J accident conditions. The largest reactivity increase occurs from g

accider; tally placing a'new fuel assembly into a Region 2 storage cell-li',

with all othericells fully loaded. Under this condition, the presence of only.500 ppm soluble boron assures that the infinite multiplication factor would not exceed the design basia reactivity for. Region _2.

With.

the nom'.nal concentration of soluble poison present (2000 ppm boron),

the maximum reactivity, k, is-less than 0.95 even if Region 2 were to be fully loaded withLfresh fuel of 4.2% enrichment.

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New Fuel Storage Region 1 storage racks are designed to safely accommodate new unirradi '

ated fuel of 4.2% enrichment,' when fully flooded with clean unborated water.' Under certain circumstances, it may be desirable to store newJ

- fuel in the dry condition'in Region 1 or to utilize Region 2 for the temporary storage of new fuel, either dry or, fully flooded.

These conditions were analyzed to assure the acceptability of Region 1 in the dry condition and to determine an arrangement in Region 2 that would assure criticality safety in conformance with the requirements of SRP 9.1.1, "New Fuel Storage."

Criticality analyses confirmed that Region 1 does not exhibit a peak in reactivity at low moderator densities (e.g., fog or foam moderation) and that the optimum moderation (highest keff) occurs for the fully flooded condition. This condition is the design basis for Region 1 where the maximum reactivity k, including all uncertainties, is less than 0.943 This corresponds to poison consisting of Boraflex, consi-dering 6" shrinkage at midplane, with double Boral sheet inserts. The maxf aum reactivity k, where Boraflex shrinkage is not considered is k

0.9389 and is 0.9469 in the absence of Boral sheet inserts.

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'In LRegion 2, it was determined that a checkerboard pattern (fuel f{

assemblies aligned diagonally) provided.an acceptable k...in either

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l the; fully.) flooded orj the by -(low. density moderation) cond! tion for n'ew-

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-Tuel' assemblies'of.4.2% enrichment. These calculations indicatec a.

nominal k,of. 0.813'10.014 (l a) wht.n fully flooded with clean unborated water--a value, substantially li.ms than the ~ limiting k,7f of 0.95, even with an additional' allowance-for uncertainties'(maximum k;,.

of. "0.86 at 955 45% tolerance limits).-

Calculations, using Monte Carlo. techniques, did not reveal'a' peak in reactivity at low moderator densities,.and fully flooded condition corresponds to the highest reactivity (optimum moderation). Thus,.the checkerboard patt'ernior new' 4.2% enriched fuel'in Region 2 represents a safe configuration in conformance with SRP 9.1.1. and 9.1.'2.

4.0 REFERENCE FUEL STORA0E CELL 4.1

. Reference Fuel Assembl'y s

The design basis fuel assembly, illustrated.in. Figure 4.1, is a 17x17 array of fuel rods with 25 rods replaced by 24' control rod guide tubes

'and 1 instrument thimble. Table 4.1 summarizes the Westinghouse opti-mized fuel assembly (OFA)- design specifications and the expected range of significant variations.

4.2.

Region 1 Storage Cells The nominal spent fuel storage cells used for the original criticality analyses of Region 1 storage. cells is shown in Figure 4.1.

The rack'is-composed of Boraflex absorber material sandwiched between a 0.060-inch inner stainless steel box and a 0.020-inch outer stainless steel'(SS) coverplate (0.125-inch coverplate for module periphery cell walls).

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The fuel assemblies are centrally located in each storage cell on a j

nominal lattice spacing of 10.320 1 0.050 inches in one direction and 10.420 + 0.050 inches in the other direction. Stainless steel gap channels connect one storage cell box to another in a rigid structure t

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and define an outer. water space between boxes. This outer water space constitutes a flux-trap between the two Boraflex absorber sheets that are essentially opaque (black) to thermal neutrons. The Boraflex absorber has a thickness of 0.075 ;; 0.007 inch and a nominal B-10 areal 2

density of 0.0238 gram per em,

Figure 4.2 shows the nominal cell modified with.Boral inserts.

The 2

Boral has a minimum B-10 loading of.020gm/cm and thickness of.075 +

0.004 inch. The minimum boron loading is combined with the maximum thickness as a " worst case" conservatism. The Boral plates are to be flush with the Borarlex cover plate. However a manufacturing tolerance allowing 0.125 inch clearance is assumed for the analysis.

4.3 Region 2 Storage Cells Region 2 storage cells were initially designed for fuel of 3.2 wt%

U-235 initial enrichment burned to 25,000 MWD /MTU and extended to encompass fuel _of 4.2% initial enrichment burned to 37,370 mwd /MTU.

In this region, the storage cells are composed of a single Boraflex absorber sandwiched between the 0.060-inch stainless steel walls of adjacent storage cells. These cells shown in Figure 4 3, are located on a lattice spacing of 9.03 j; 0.04 inches.

5.0 ANALYTICAL METHODOLOGY 5.1 Analytical Methods And Bias 5.1.1 Reference Analytical Methods and Bias for Nominal Case The CASMO-2E computer code (References 1, 2 and 3), a two-dimensional multigroup transport theory code for fuel assemblies, has been bench-marked (see Appendix A) and is used both as a primary method of analysis and as a means of evaluating small reactivity increments associated with manufacturing tolerances. CASMO-2E benchmarking resulted in a calculational bias of 0.0013 j; 0.0018 (95% /95%). _-__

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'8-14 Y In fuel:ra'ck ' analyses, for independent ' verification, criticality.

analyses of the high density spent fuel storage. racks were also per-formed with the AMPX-KEND computer package (References 4'and 5),'.using the 27-group SCALE" cross-section library (Reference 6)/with'the NITAWL subroutine for U-238 resonance shielding. effects (Nordheim' integral' tre atmer.t ).

Details of: the benchmark calculations.with the 27-group SCALE cross-section 1!brary are also presented in-Appendix A.

These l

benchmark calculations resulted in a bias of 0.0106' + 0.0048 (955 /955).-

In the geometric' model used in KENO, each fuel rod and its cladding L were described explicitly.

For two-dimensional X-Y analysis, a zero

current (white. albedo) boundary condition was applied in the axial direction and, for Region 1, at the centerline through the. outer water space (flux-trap) on all four sides of the cell, effectively creating an infinite array of storage cells. 'In Region 2, the zero current ;

boundary condition was' applied at the center of the Boraflex absorber sheets.between storage cells. The AMPX-KENO Monte Carlo calculations inherently include a statistical. uncertainty due to the random nature'

.of neutron tracking. To minimize the statisticalJuncertainty of the KENO-calculated reactivity, a total of 50,000 neutron histories is normally accumulated for each' calculation, in 100 generations of 500 neutrons each.

CASMO-2E is also used for' burnup calculations, with independent verifi-cation by EPRI-CELL ~ and NULIF calculations.

In tracking long-term

-(30 year) reactivity effect of spent fuel stored in Region 2 of the fuel storage rack, EPRI-CELL calculations indicate a continuous reduc-tion in reactivity with time (after Xe decay) due primarily to Pu-241 decay and Am-241 growth.

A third independent method of criticality analysis, utilizing diffusion / blackness theory, was also used for additional confidence in

' SCALE is an acronym for Standard Computer Analysis for Licensing Evaluation. a

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Revision 1 WT 8-14-89 resultslor'.the primary calculational methods, although no reliance for criticality safety is placed on the reactivity value from the~'

diffusion / blackness theory technique.

This technique, however, is used for auxiliary calculations of small incremental reactivity' effects r

(e.g.,1 axial cutback or mechanical tolerances) that would otherwise be-lost in normal KENO statistical variations, or' would be inconsistent with CASMO-2E geometry limitations.

Cross' sections for the diffusion / blackness theory calculations were derived from CASMO-2E or calculated by the-NULIF computer code (Refer-l?

ence 7),1 supplemented by a blackness theory routine that effectively imposes. a transport. theory boundary condition at-the surface of the Boraflex neutron absorber..Two different spatial diffusion theory codes,' PDQ07 (Reference 8) in two dimensions:and SNEID* in one dimen-H sion, were used to calculate reactivities. The two-dimensional PDQ07.

code was used to describe the actual storage. cell geometry, with NULIF cell-homogenized constants representing each fuel rod and its associ-

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I ated water moderator.' SNEID is one-dimensi.onal model, in cylindrical or slab geometry, used for the calculation of. axial cutback reactivity effects and in the assessment of abnormal occurrences.

5.1.2-Analytical Methods and Bias.for Modified Cases l

The design method used to demonstrate the criticality safety of fuel in the modified fuel storage racks uses the, AMPX system of computer codes

[4] for cross-section generation' (NITAWL-S and XSDRNPM-S), and the l

KENO-IV (CRC) code [5] for reactivity determination. Tolerance varia-1 tions of the nominal case were then applied.

y An ORNL 227 energy group ENDF/B-V cross-section library [15,16] was i

utilized for the criticality analysis. The NITAWL program computis l

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  • SNEID is one-dimensional diffusion theory routine developed by Black &

Veatch and verified by comparison with PDQ07 one-dimensional calcula tions. -

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x resonance cross-sections and prepares a 227 neutron group P3 CP 88-section data file for.use:in the XSDRNPM one-dimensional S ne utron.

n transport ~ code. 'The Nordheim Integral Treatment [17] was used.

Energy.

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'and spatial seighting of. cross-sections ~ to reduce the' 227 group data to 27 group (SCALE) data is performed b'y the XSDRNPM' program.- Zone weighted cross-sections are used in the fuel regions that are-modeled asdiscretefuelrodsinKENO...Discretemodelikgofstructuralregions such-as ~ the storage cell ' structures also use zone weight cross-sections.

KENO-IV is used.in the evaluation of criticality of the Byron /Braidwood storage racks. KENO-IV is a three-dimensional Monte Carlo theory program designed.for reactivity calculations.

The calculational method and cross-section values which were used in

' the criticality analysis of the fuel storage racks have been verified -

by comparison with data from, criticality experiments for assemblies similar to those for which the racks are designed.. This benchmarking data is sufficiently diverse to. establish that the method bias and

- uncertainty will apply to fuel storage rack conditions which include strong heutron absorbers and low moderator densities.

_ A set of 33 criticality experiments-has been analyzed to demonstrate its applicability to criticality analysis and to establish the method

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bias and variability. The experiments range from water moderated oxide fuel array. separated by various materials -(BORAL, steel and water) that simulate LWR fuel shipping and storage conditions [18], to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials [19] (Plexiglas, paraffin and air) that demonstrate the wide range of applicability of the method.

The results and some descriptive factors about each of the 33 benchmark criticality experiments are given in Table 5.1.

The average Keff OI the benchmarks is 0.9917 which demonstrates that there is a 0.0083 bias associated with the method. The standard deviation of the Keff values is 0.00082 Ak.

The 95 /95 one-side tolerance limit factor for 33 values _ _ _ _ _ _

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There is a 95 percent probability with 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0018 Ak.

5.2 Fuel Burnup Calculations fuel burnup calculations in the hot operating condition were performed primarily with the C ASMD-2E code.

However, to enhance the credibility of the burnup calculations (in lieu of criticality experiments), the CASMD-2E results were independently checked by calculations with the NULIF code (Reference 7) and with EPRI-CELL (Reference 9).

Figure 5.1 compares results of these independent methods of burnup analysis under hot reactor operating conditions. The results agree within 0.008 Ak.

In addition to depletion calculations under hot operating conditions, reactivity comparisons under conditions more representative of fuel to be stored in the racks (cold, xenon-free) are also significant in storage rack criticality analyses. Table 5.2 comparea the cold, xenon-free reactivities calculated by CASMO-2E, NULIF/PDQ07, and EPRI-CELL.

In the cold condition, the CASMO-2E calculations gave a slightly higher reactivity value for the Region 2 fuel storage cell and the good agree-ment generally observed lends credibility to the calculations, particu-larly in view of the known bias and uncertainty in CASMO-2E calcula-tions (Appendix A).

No definitive methods exists for determining the uncertainty in burnup-i dependent reactivity calculations.

All of the codes discussed above have been used to accurately follow reactivity loss rates in operating reactors. CASMD-2E has been extensively benchmarked ( Appendix A; References 1, 2, 3 and 10) against cold, clean, criticality experiments (including plutonium-bearing fuel), Monte Carlo calculations, reactor j

f operations, and heavy-element concentrations in irradiated fuel.

In particular, the analyses (Reference 10) of 11 criticality experiments with plutonium-bearing fuel gave an average keff of 1.002 + 0.011 t

(95% /95%), showing adequate treatment of the plutonium nuclices.

In i

l addition, Johansson (Reference 11) has obtained very good agreement in 1 - - _.

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Revision 1 8-14-89 calculations of close packed, high-plutonium-content, experimental confi gura tio ns.

Since criticality experiment data with spent fuel is not available, it is necessary to assign an uncertainty in reactivity based on other con-siderations, supported by the close agreement between different calcu-lational methods and the general industry experience in predicting reactivity loss rates in operating plants.

Over a considerable portion of the burnup, the reactivity loss rate in PWRs is approximately 0.01 Ak for each 1,000 MWD /MTU, becoming somewhat smaller at the higher burnups.

By conservatively assuming an uncertainty in reactivity" of 0.5 x 10-6 times the burnup in MWD /NTU, a burnup-dependent uncertainty is defined that increases with increasing fuel burnup, as would be reasonably expected. This assumption provides an estimate of the burnup uncertainty that is more conservative and bounds estimates frequently employed in other fuel rack licensing applications (i.e., 5%

of the total reactivity decrement). Table 5 3 summarizes results of the burnup analyses and estimated uncertainties.

These uncertainties are appreciably larger, in general, than would be suggested by the industry experience in predicting reactivity loss rates and boron let-down curves over many cycles in operating plants. The increasing level of conservatism at the higher fuel burnups provides an adequate margin in the uncertainty estimate to accommodate the possible existence of a small positive reactivity increment from the axial distribution in burnup (see Section 5 3).

In addition, although the burnup uncertainty may be either positive or negative, it is treated as an additive term rather than being combined statistically with other uncertainties.

Thus, the allowance for uncertainty in burnup calculations is believed to be a conservative estimate, particularly in view of the substantial reactivity decrease with aged fuel as discussed in Section 5.4.

"Only that portion of the uncertainty due to burnup.

Other uncertainties are accounted for elsewhere. ___-____-___-_ -

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Effect ora Axial. Burnup Distribution' Initially,' fuel loaded into the reactor.will burn with a slightly-skewed cosine power distribution.

As burnup progresses, the burnup:

distribution will tend to flatten' becoming more highly. burned lin the

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-central regions than in the upper and lower ends. This effect may-be clearly..seen in the curves compiled 'in Reference 12. ' At high. burnup, the.more reactive fuel. near the ends of the fuel' assembly-(less than:

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average' burned) occurs in regions of lower reactivity worth'due to neutron leakage. Consequently, it is expected that distributed-burnup fuel assemblies would exhibit a slightly lower reactivity than' that

' calculated for the average burnup.- As burnup progress; ;,l the distribu-l tion, to some extent, tends.to be self-regulating as t. c '.colled by. the axial power distribution, precluding the existence of large regions of significantly reduced burnup.

A number of one-dimensional diffusion theory analyses have. been made based upon calculated and measured. axial burnup distributions. 'These.

-analyses confirm the minor and genera 11y ' negative reactivity effect of-axially distributed burnup. The trends observed,-however,.suggest.the possibility of a'small positive reactivity effect at the high burnup

. values, and the uncertainty' in k, due :to burnup, assigned at : the higher burnups (Section. 5' 2), is adequately conservative to encompass then potential for a small positive reactivity effect of postulated axial burnup ' distributions.

Furthermore, reactivity decreases with time in storage (Section 5.4), and, in addition, there is a large margin.in reactivity (>0.02 Ak) below the limiting k,pp value (0.95) which can accommodate any reasonable reactivity effects that might be larger than expected.

5.4' Long-Term Decay Since the fuel racks in Region 2 are intended to contain spent fuel for long periods of time, calculations were made using EPRI-CELL (which incorporates the CINDER code) to follow the long-term changes in reactivity of spent fuel over a 30 year period.

CINDER tracks the L

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jdecayf and' burnup, dependence cir 'some_179 fission products.: -Early in the

. decay period xenon grows in- (reducing reactivity) and_ subsequently -

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_ decays, with the, reactivity reaching a maximum at: 100-200 hours. The-

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decay of Pu-241 (13 year-half-life') and growth f.Am-241-substantia 11k reduce reactivity during long term storage,'as indicated in Tablei 5.4 The reference design criticality calculations do not take' credit

'for this long-term reduction in reactivity' other 'than to indicate an increasing suboriticality margin in Region 2 of' the spent fuel ' storage.

pool.

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6.'O REDIC,N' 1 CRITICALITY ANALYSIS AND TOLERANCE VARIATIONS 6.1.-

_ Nominal Design Case.

Under normal conditions, with nominal dimensions, the k, values calcu--

lated by the three' methods of analysis are as follows:

Maximum k.

Analytical Method Blas-corrected k_

(955/)55)

CASMO-2E 0.9387 + 0.0018 0.9405 AMPX-KENO-0.9301 7 0.0061 0.9362 Diffusion blackness theory 0.9393 0.9393 The AMPFX-KENO calculations include a one-sided tolerance factor (Reference 13) of 1.799 corresponding.to 955 probability at a 955-confidence limit.

For the nominal design case, the CASMO-2E calcula-tion yields the highest (most conservative) reactivity, and, therefore, the independent verification calculations substantiate CASMO-2E as the prituary calculational method.

6.2' Boron Loading Variation The Borarlex absorber sheets used in Region 1 storage cells are nominally 0.075-inch thick, with a B-10 areal density of 0.0238 2

g /cm.

Independent manufacturing tolerance limits are +0.007 inch in 2

thickness and +0.0017 g/cm in B-10 content. This assures that at any

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Revision 1 B-14-89 2

point where the minimum boron concentration (0.0221 gram B-10 /cm ) and minimum Boraflex thickness (0.068 inch) may coincide, the B-10 areal 2

censity will not be less than 0.020 gram /cm.

Differential CASMO-2E calculations indicate that these tolerance limits result in an incre-mental reactivity uncertainty of 10.0021 Ak for boron content and

+0.0047 for Boraflex thickness variations.

h 6.3 Storage Cell Lattice Pitch Variation The design storage cell lattice spacing between fuel assemblies in Region 1 is 10.32 inches in one direction and 10.42 inches in the other direction. A decrease in storage cell lattice spacing may or may not increase reactivity depending upon other dimensional changes that may be associated with the decrease in lattice spacing.

Increasing the water thickness between the fuel and the inner stainless steel box results in a small increase in reactivity. The reactivity effect of the flux-trap water thickness, however, is more significant, and decreasing the flux-trap water thickness increases reactivity.

Both of these effects have been evaluated for independent design tolerances.

The inner stainless steel box dimension, 8.850 + 0.032 inches, defines the inner water thickness between the fuel and the inside box.

For the tolerance limit, the uncertainty in reactivity is 10.0018 Ak as deter-mined by differential CASMO-2E calculations, with k, increasing as the inner stainless steel box dimension (and deri.J tive lattice spacing) increases.

The design flux-trap water thicknesses are 1.160 + 0.040 inches and 1.260 1 0.040 inches, which result in an uncertainty of +0.0038 ak due to the tolerance in flux-trap water thickness, assuming the water thickness is simultaneously reduced on all four sides. Since the manufacturing tolerances on each of the four sides are statistically independent, the actual reactivity uncertainties would be less than

+0.0038 ak, although the more conservative value was been used in tne criticality evaluation.

Revision 1

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S-14-89 6.4 Stainless Steel Thickness Tolerances The nominal stainless steel thickness in Region 1 is 0.060 inch for the inner stainless steel box and 0.020 inch for the Boraflex coverplate (0.125 inch on module boundary).

The maximum positive reactivity effect of the expected stainless steel thickness tolerance variations, statistically combined, was calculated (CASMO-2E) to be 10.0025 Ak.

6.5 Fuel Enrichment and Density Variation The design maximum enrichment is 4.20 1 0.05.wt% U-235.

Calculations of the sensitivity to small enrichment variations by CASMD-2E yielded a coefficient of 0.0047 Ak per 0.10 wt% U-235 at the design enrichment.

For a tolerance on U-235 enrichment of 10.05 in wt%, the uncertainty on k, is +0.0024 Ak.

fuel density increased to maximum Calculations were made with the UO2 value of 97% theoretical density (TD).

For the midrange value (95% TD) used for the reference design calcu'.ations, the uncertainty in reacti-densities expected.

vity is +p.0026 Ak over the range of UO2 6.6 Boraflex Width Tolerance Variation I

I The reference storage cell design for Region 1 (Figure 4.1) uses a l

Borarlex blade width of 7.75 1 0.0625 inches.

A positive increment in reactivity occurs for a decrease in Boraflex absorber width. For i

i reduction in width of the maximum tolerance, 0.0625 inch, the calcu-j i

lated positive reactivity increment is +0.0007 Ak.

J a

J 1

I 6.7 Axial Cutback of Boraflex Boraflex poison material is provided over the entire length of the active fuel length (144 inches) for the Braidwood Region 1 racks.

However for the Byron station Region 1 racks, the axial length of the Boraflex poison material is less than the active fuel length by 284 inches at the top and at the bottom of the storage rack modules. To

! )

1 L

- - - - - 9

p r

"E f! (.

, -(s M*i Revision 1 pW 8-14-89 4-..

F'

$"~ un%Sr' fit d

V

. account 'for the; reactivity effect 'of this axial cutback, one-dimensional :(slab) diffusion theory-calculations were made using flux-

weighted' homogenized. diffusion theory constants edited from CASMO-2E m

calculations of the array of storage-cells, with and without Boraflex present.

In the one-dimensional calculations, an infinite (30-cm) water reflector was used above and below the fuel assembly, with the lengths'. of the' unpoisoned " cutback" regions, top and bottom,' varied in.

a series of parametric calculations.

Results of these calculations showed that the k,77 remains less than the k,. of the reference central storage cell ~ region, until the axial cutback exceeds four inches top and bottom. ' Thus, the actual axial neutron leakage more than compen-sates for the 2'4 inch design cutback, and the reference infinite multi-plication factor, k, remains a conservative overestimate of the true reactivity.

7.0 REGION 2 CRITICALITY ANALYSIS AND TOLERANCE VARIATIONS

'7.1 Nominal Design Case-The principal method of analysis in Region 2 was CASMO-2E code, using the restart ~ option in CASMO to transfer fuel of a specified burnup into the storage rack configuration at a reference temperature of 0 0.

~!

0

- Calculations were made for fuel of several different. initial enrich-ments and, for each enrichment, a limiting k, value established which included an additional factor for uncertainty in the burnup analysis and for the axial burnup distribution. The restart CASMO-2E calcula-tions (cold, clean, rack geometry) were then interpolated to define the burnup value yielding the limiting k, value for each enrichment, as indicated in Table 7.1.

These converged burnup values define the boundary of the acceptable domain shown in Figure 31.

1 At a burnup of 37,000 MWD /MTU, the sensitivity to burnup is calculated

]

to be -0.0079 ok per 1000 MWD /MTU. During long-term storage, the k.

1 i

values of the Region 2 fuel rack will decrease continuously from decay I

J of Fu-241 as indicated in Section 5.4 l

I

..e B

Revision'1-1 -

L vi 3,

g_ta_g9

. Two independent' calculational methods were used tio provide additional

,7

. confidence in the. reference Region 2 criticality analyses.- Fuel-of.

1.5% initial enrichment (approximately equivalent' to the reference. rack design for burned fuel) was analyzed. by AMPX-KENO (27-group ' SCALE 1, n.

cross-section library) and by the CASMO-2E model 'used for the Region 2-rack analysis. Forlthis' case CASMO-2E k. (0.9014)~was within the statistical uncertainty of the bias-corrected value. (0.9043' + 0.0030

'(l o)) 'obtained in the AMPX-KENO calculations.. This agreement confirms the' validity of-the primary CASMO-2E calculations.

The. second. independent method of analysis used the NULIF' code for

~

burnup analysis, and for generating dirfusion theory constants'(cold,-

clean) for' the NULIF-calculated composition at -25,000- MWDMrU with fuel

~

of 3.2% initial, enrichment.- These constants, together with blackness theoryiconstants for the Borarlex absorber,. were then used in a two-dimensional PDQ07 calculation for the storage rack configuration.-

Results of. this calculation - (k. of. 0.9017) compared favorably with the CASMO-2E calculation for the'same conditions (k. of 0.9061) and thus tend to' confirm.the. validity of the primary calculational method.

7.2 Boron Loading Variation

.The Boraflex absorber ' sheets used in the Region 2 storage cells are 2

nominally 0.041 inch thick with B-10 areal density o'r 0.0130 g/cm,

Independent manufacturing limits are +0.007 inches in thickness and 2

+0.0009 g /cm in B-10 content. This assures that at any point where 2

the minimum boron concentration (0.01206 B-10 g/cm ) and the minimum V

Boraflex thickness (0.034 inch) may coincide, the B-10 areal density 2

will not be less than 0.010 g/cm. Differential CASMO-2E calculations indicate that these tolerance limits result in an incremental reacti-5

.vity uncertainty of +0.0028 ok for boron content and +0.0078 ok for Boraflex thickness.

-j 1 i

ts

l

w I

Revision 1 L.. -

33439 7.3 Storage Cell Lattice Pitch Variations l

The value used for the storage cell lattice spacing between fuel assemblles in Region 2 is 9.03 1 0.040 inches, corresponding to an uncertainty in reactivity of 0.0011 Ak.

7.4 Stainless Steel Thickness Tolerance The nominal thickness of the stainless steel box wall;is 0.060 inch with a tolerance limit of 1 005 inch, resulting in an uncertainty in 0

reactivity of 10.0001 Ak.

7.5 Fuel Enrichment and Density Variation Uncertainties in reactivity due to tolerances on fuel enrichment and 00 density in Region 2 are assumed to be the same as those determined 2

for Region 1.

i 7.6 Borarlex Width Tolerance The reference storage cell design for Region 2 (Figure 4.3) uses a Boraflex absorber width of 7.25 1 0.0625 inches.

For a reduction in width of the maximum tolerance, the calculated positive reactivity Increment is 0.0009 ok.

8.0 ABNORMAL AND ACCIDENT CONDITIONS 8.1 Eccentric Positioning of Fuel Assembly in Storage Rack The fuel assembly is normally located in the center of the storage rack cell with bottom fittings and spacers that mechanically limit lateral movement of the fuel assemblies.

Nevertheless, calculations were made with the fuel assemblies moved into the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations indi-cated that the reactivity decreases very slightly in both regions, as - _ _ - - _ - _ _

s q"c q,.

Revision 1 gM$

S-14-89

.n I'

i determine'd.by.'PDQ07 calculations with dif fusion coefficients

  • generated by_NULIF and 'a blackness theory routine. The. highest reactivity there-fore corresponds to the reference; design with the fuel assemblies

, positioned in the'. center of the storage cells.

8.2.
Temperature and Water Density Effects

' The moderator. temperature coefficient of reactivity 'in both regions is

~

0 negative; a moderator temperature of 0 C, with a water density of 1.0 3

g/cm, was assumea for the reference designs, which assures that the true reactisity will always be lower, regardless of. temperature.

Temperature effects on reactivity have been calculated and' the results are shown in Table 8.1.

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

the table. Voids due to boiling will not occur. in the outer (flux-trap)'. water region of Region 1.

With soluble poison present, the temperature coefficients.of reactivity would be expected to differ from those inferred from the data in Table 8.1. ' However, the reactivities.would also be substantially lower at all temperatures with soluble boron present, and the data in Table 8.1.

is pertinent to the higher-reactivity unborated case.

a 8.3 Dropped Fuel Assembly Accident To investigate the possible reactivity effect of a postulated fuel assembly drop accident, calculations were made for unpolsoned J

assemblies separated only by clean unborated water.

Figure 8.1 shows the results of these calculations.

From these data, the reactivity (k ) will be less than 0.95 for any water gap spacing greater than 6 to "This calculational approach was necessary since the reactivity effects are too small to be calculated by KENO, and CASMD-2E geometry is not readily amendable to eccentric positioning of a fuel assembly.

I -___o____1-________

Revision 1

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S-14 -B 9 l

I 7 inches in the absence of any absorber material, other than water, between assemblies.

For a drop on top of the rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum sepa-aation distance of more than 12 inches. Maximum expected deformation under seismic or accident conditions will not reduce the minimum spac-ing between fuel assemblies to less than 12 inches. Consequently, fuel assembly drop accident will not result in an increase in reactivity above that calculated for the infinite nominal design storage rack.

Furthermore, soluble boron in the pool wcter would substantially reduce the reactivity and assure that the true reactivity is always less than the limiting value for any conceivable fuel handling accident.

8.4 Abnormal Location of a Fuel Assembly The abnormal location of a fresh untrradiated fuel assembly of 4.2%

enrichment could, in the absence of soluble poison, result in exceeding ohe design reactility limitation (k. of 0.95).

This could occur if the assembly were to be either positioned outside and adjacent to a storage rack module or loaded into a Region 2 storage cell, with the latter condition producing the larger positive reactivity increment. Soluble poison, however, is normally present in the spent fuel pool water (for which credit is permitted under these conditions) and would maintain the reactivity substantially less than the design limitation.

The large:t reactivity increase occurs for accidentally placing a new j

fuel assembly into a Region 2 storage cell with all other cells fully loaded. Under this condition, the presance of 300 ppm soluble boron j

assures that the infinite multiplication factor would not exceed the design basis reactivity. With the nominal concentration ?f soluble poison present (2000 ppm boron), the maximum reactivity, k, is less than 0.95 even if Region 2 were to be fully loaded with fresh fuel of 4.2% enrichment. Administrative procedures will be used to confirm and assure the continued presence of soluble poison in the spent fuel pool

)

1 I

water during fuel handling operations. l 1

_ _. _ _ _ _ _ _ _. _ _ _ _____ _ _____.____ o

Revision 1 D

8-14-89 8.5 Lateral Rack Movement Lateral motion of the rank modules under seismic conditions could potentially alter the spacing between rack modules. However, girdle bars on the modules prevent closing the spacing to less than 2.0 inches, which is greater than the normal flux-trap water gap in the Region 1 reference design.

Region 2 storage cells do not use flux-traps and the reactivity is insensitive to the spacing between modules.

Furthermore, soluble poison would assure that a reactivity less than the design limitation is maintained under all conditions.

9.0 NEW FUEL STORAGE 9.1 Storage in Region 1, Dry Region 1 is normally designed to accommodate new unirradiated fuel assemblies in the fully flooded condition.

For storage in the dry condition, the racks must also conform to the requirements of SRP 9.1.1 value of 0.98 under optimum low density which specify a limiting keff moderation. Calculations were made, using AMPX-KENO, for several hypothetical low-moderator densities down to 0.05 g /cc simulating tog or foam moderation. These calculations showed a continuously decreas-ing k, as the moderator density decreased, yielding a k of 0.546 1 0.008 (lo) at 10% moderator density. Axial leakage was neglected in these calculations, but would substantially reduce the already low k.

values. These results are consistent with the geneval observation that a low-density optimum-moderation peak in eactivity does not exist in poisoned racks (Reference 14).

9.2 Storage in Region 2, Flooded in a succession of trial-and-error calculations, it was found that a checkerboard stora6e pattern in Region 2 would allow new fuel assemblies of 4.2% enrichment to be safely accommodated without exceed-ing the limiting 0.95 keft value.

In this checkerboard loading pat-f -_. -

Revision 1 B-14-89 e -

tern, the fuel assemblies are located on a diagonal array, as illu-strated below, with alternate storage cells empty of any fuel.

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////

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(empty) r///.

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Monte Carlo calculations (AMPX-KENO) resulted in a k of 0.8133 1 0.0138. With a one sided K-factor (Reference 13) for 955 probability at a 95% confidence level and a ak of ^.009 for uncertainties (Table 3.2 for Region 2), the maximum k. is 0.863, which is substantially less than the 0.95 limiting value. Thus, Region 2 may be safely used for the temporary storage of new fuel assemblies provided the storage configuration is restricted to the checkerboard pattern indicated above.

9.3 Storage in Region 2, Dry As indicated in Section 9.1 above, a peak in reactivity (k.) at low ederator densities is not expected for poisoned rack designs.

AMPX-KENO calculations confirmed the absence of a low-moderator-density peak in Region 2 with 4.2% enriched fuel arranged in the checkerboard pattern.

At 10% moderator density, the calculated k. was 0.552, which would be substantially reduced if axial leakage were to be included.

Thus, Region 2 conforms to tha requirements of SRP 9.1.1 (k. < 0.98 at optimum moderation) for the safe storage of 4.2% enriched fuel, dry, in the checkerboard loading pattern. 1

Revision 1

+

B-14 -8 9 10.0 Boraflex Shrinkage The design of Region 1 poison has been modified to preclude any pos-sible problems from Boraflex shrinkage by adding Boral inserts in the flux-t.*aps (Figure 4.2). The Region 2 design was found to be acceptable as is.

The following is the basis for the confidence in the Byron /Braidwood sgent f uel racks.

Models developed by Northeast Technologies Corporation and recent experiments by BISCO (the manufacturer) indicate that Boraflex shrinks 10 with radiation exposure until an integrated dose of approximately 10 rads has been accumulated. The studies indicate that the Boraflex 13 dimensions then stabilize until approximately 10 rads have accumu-lated at which point the Boraflex begins to powder near the edges.

This data was collected in a combined neutron and gamma field.

Table 10.1 and Figure 10. ) contain results of a calculation of the maximum integrated dose seen by a panel versus time for a typical off load into the Byron spent fuel pool.

This calculation assumes that there was no delay time in transferring the fuel assemblies from the reactor to the spent fuel pit.

10 In this case, the maximum predicted dose, 2.8 x 10 rads, is f ar below the expected dose for powdering to begin and is directly applicable to the Region 2 racks. Therefore, no serious degradation problems should be encountered in this region.

For the upcoming outages, there will be a need to use some Region 2 racks as temporary storage of " hot spent fuel."

" Hot spent fuel" is defined in NUREG 0612. This could result ll in some Boraflex panels receiving a dose of up to 10 rada. A ll Region 1 panel could see a dose of up to 6.6 x 10 rads during its li feti me.

These maximum doses are far below the dose where powdering l

l can be expected. Therefore powdering is of no concern.

l The average Boraflex shrinkage reported in the studies is 2.5%.

How-ever, the maximum shrinkage reported in a sample is 4%. This maximum L.

o'

.., i

. Revision' 1 f,

3 14-89 l

shrinkage was used'to ensure the criticality analyses performed could

~

be. considered bounding.

' The duty cycle-for the Region 1. racks, i.e. many off' loads of newly

' discharged fuel combined with the manufacturing technique used has= led l to.the design modification of these racks. This modification ensures that the criticality criterion of K,77 less.than 0.95 at 95 45 confi-

~

dence level will not be violated even with'an extremely conservative.

assumption of 4% shrinkage.

A further conservatism used in the -

analysis was the assumption that all shrinkage appears as a 6 inch wide.

gap located in all panels at the midplane, thereby maximizing the neutronic width of the gap.

All Region 1. racks have two sheets of Boral with a minimum B-10 loading 2

of 0.020 gm/cm inserted intol the-flux trap between each cell. The analysis. to qualify this design assumed the " worst case" configuration as regards the Boral combined with the Oat uncertainty factors to.

s produce a maximum K,pp of 0.9421 at a 95 M5 confidence level, assuming 6 inch gaps'in the Boraflex of'all panels.

For Region 2, criticality analyses have been performed for various gap configurations. A sufficient number of sensitivity studies were per-formed to allow a plot of gap size versus K,pp, up to 6" gaps in 4 of 4 poison plates. The analyses showed that the racks exhibit a K,pp less than 0.95 for gap conditions up to and including 6" gaps at the mid-plane in 2 of 4 poison plates. The analyses also indicate that if all gaps occur at the midplane with an additional 2" of shrinkage at the top and bottom, a gap of 3 and 3 A",

in 4 of 4 poison plates, will maintain K less than 0.95 at the 95 45 confidence level. Gaps eff located 'at other positions or not aligned result in lower K,fp.

Elimi-I nating the shrinkags at the ends was found to have a minimal effect on j

K,gg.

]

)

i The Region 2 racks are quite similar to the Quad Cities racks in design, except the Byroneraidwood racks do not use glue to position the Boraflex plates.

Based on the Quad Cities data, the probability of -

.__--_____-_____U

-m,,_.

.--.m

--,-----.-__-,.----m,---.------..----

Revision 1 B-14-89 ec:

gaps this large in all panels at the same level is not a credible event.

Therefore, 'it is concluded that the use of Boraflex in the Region 2 racks poses no hazard. Should degradation of Boraflex be found during the in-service surveillance, an evaluation will be made to ensure that the applicable criticality safety limits are maintained..

11.0 Summary and Conclusions o

The criticality analyses performed on the high density fuel racks for normal, abnormal and accident conditions demonstrate that the design of the fuel storage facilities in the Byron and Braidwood Nuclear Stations meet the acceptance criteria specified in SRP 9.1.1 and 9.1.2 and that the maximum reactivity is always main-tained below 0.95 with the racks fully loaded at a maximum initial enrichment of 4.2 wt % and flooded with unborated water except for the abnormal condition discussed below, o

For Region 1 racks, used for new fuel storage, the maximum reacti-vity including all uncertainties does not exceed 0.943 For Region 2 racks, used for storage of fuel considering a burn up o

enrichments trade-off, the maximum reactivity including all uncer-l tainties does not exceed 0.930.

l l

For the abnormal conditions involving inadvertent misplacement of o

a new fuel assembly (either into a Region 2 storage cell or out-side and adjacent to a rack module) the presence of a minimum of 300 ppm soluble boron ensures that the maximum reactivity is i

always maintained below 0.95 with all other Region 2 cells fully loaded.

. I l

c

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' J.2

.v Revision 1 I

' + =

e.-

8-14-89 REFERENCES l-l 1.

A. Ahlin, M. ' Edenius, H. Haggblom,' "CASMO - Fue1~ Assembly Burnup.

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

'2.

A. Ahlin' and M. Edenius,, "CASMO -L A Fest Transport Theory Depletion -

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

3 M.,Edenius ' et al., "C ASMO Benchmark. Report," Studsvik /RF-78-6293, Aktiebolaget Atomenergi, March 1978.

1 i

4 Green,. Lucious, Petrie,' Ford, White, Wright, "PSR-63 /AMPX-1 (code

' package), AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries: from ENDF /B," ORNL-TM-36-706, Oak Ridge National Laboratory, March 1976..

i i

5.

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

' Criticality Program," ORNL-4938, Oak Ridge National Laboratory, November.1975.

]

I 6.

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

j NUREG /CR-0200,.1979.

l

.1 7.'

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

Generator and' Fuel Depletion Code," BAW-426. The Babcock & Wilcox l

Company, August 1976.

'l 1

8.

W. R. Cadwell,. PDQ07 F eference Manual, WAPD-TM-678, Bettis Atomic Power I

1 Laboratory, January 19 57.

'j z9.

W. J. Eich, " Advanced Recycle Methodology Program, CEM-3," Electric j

Power Research Institute,1976.

I 4. _. _. _ _._____-__ -___ - _

A

i

.c:

Revision 1 l

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10.

E. E. Pilat, " Methods for the Analysis of Boiling Water Reactors (Lattice Physics)," YAEC-1232, Yankee Atomic Electric Co., December 1980.

11.

E. Johansson, " Reactor Physics Calculations on Close-Packed Pressurized Water Reactor Lattices," Nuclear Technology, Vol. 68, pp. 263-2f 8, February 1985, d

12.

H. Richings, Some Notes on PWR (W) Power Distribution Probabilities for LOCA Probabilistic Analyses, NRC Memorandum to P. S. Check, dated July 5, 1977.

1 13 M. G. Natrelia, Experimental Statistics National Bureau of Standards, Handbook 91, August 1963 14 J. M. Cano et al., "Supercriticality Through Optimum Moderation in Nuclear Fuel Storage," Nuclear Tecnrology, Vol. 48, pp. 251-260, May

1980, 15.

W. E. Ford III, et al, "CRSL-V : Processed ENDF /B-V 227 Neutron Group and Pointwise Cross Section Libraries for Criticality Safety, Reactor and Shielding Studies," ORNL /CSD /TM-160, June,1982.

16.

R. Kinsey, Ed. "ENDF /B Summary Documentation," BNL-NCS-17541 (ENDF-201), 3rd Edition, Brookhaven National Laboratory,1979.

17.

T. W. Nordheim, "The Theory of Resonance Absorption," Proceeding of Symposia in Applied Mathematics, Vol. XI, p. 58, Garrett Birkhoff and Eugene Wigner, Eds., AM. Math. Soc.,1961.

18.

M. N. Baldwin et al, " Critical Experiment Supporting Close Proximity Water Storage of Power Reactor Fuel," BAE-1484-7, July,1979.

19.

J. T. Thomas, " Critical Three-Dimensional Arrays of U(93 2) Metal Cylinder 7,

" Nuclear Science and Engineering, Volume 52, pages 350-359, 1973 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ -

Revis ion ' 1 3-14 39 m:.

Table 3.-1

SUMMARY

OF REGION 1 CRITICALITY SAFETY ANALYSES Region 1 Region 1 Nominal Boral Inserts Minimum acceptable burnup 0

0 MWD /MTU

@ 4.2% initial enrienment 0

Temperature assumed for analysis OC 0C Reference k, (nominal) 0.9374 0.9215-Calculational bias 0.0013 0.0083 Uncertainties Bias

+0.0018 0.0018 KENO N /A 10.0041 B-10 concentration (Borarlex)

+0.0021 0.0021 Boraflex thickness 70.0047 0.0047 Borarlex width

+0.0007 10.0007 Inner box dimension 70.0018 0.0018 Water gap thickness

[0.0038 0.0038 SS thickness

+0.0025 10.0025 7.0024 10.0024 0

Fuel enrichment Fuel density

[0.0026 0.0026 Eccentric assembly position negative negative II)

+0,0082 0.0091 Statistical combination Allowance for burnup uncertainty N /A N /A l

Total 0.9387 + 0.0082 0.9298 0.0091 1

Maximum reactivity 0.9469 0.9389 (IISquare root of sum of squares. _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ -

=

if j..,j b "'

,"*h Revision 11'

'k 4 g_ja.gg:

L<

t.

t

. Table 3 2 0

1 L

SUMMARY

OF REGION 2 CRITICALITY SAFETY ANALYSES

_. 7 g/.:.

-1 Region.2-Minimum accept'able;burnup 37,370 MWD /MTU H

  1. 4.2% initial enrichment-p:)

0

- Temperature assumedL for' antlysis 0C

' Reference-'ka (nominal) 0.8999 q-Calculational-bias.

0.0013 Uncertainties-

. Bias

+0,0018 KENO N/A B-10 concentration (Boraflex)

+0.0028-7.0078-0 Boraflex thickness-Boraflex widthz

~[0.0009

~ 1nner box' dimension

+0.0011 Water gap ; thickness N /A

.SS thickness

+0.0001 7.0024 0

Fuel enrichment Fuel density

}[0.0026 Eccentric assembly position

. negative Statistical combination II)

_ + 0.00 93 Allowance for burnup uncertainty 0.0187 1 Total 0.9199 + 0.0093 Maximum reactivity 0.9292 II) Square root of sum of squares..,

$___).__.._-'.-I'.'.

_.u, Revision 1 e.

8-14-89 Table 3.3 REACTIVITY' EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS Accident /Abnonial Conditions Reactivity Effect Temperature increase Negative in both regions Vold (bo11'ing)

Negative in both regions Assembly dropped on top of rack Negligible Lateral rack module movement Negligible Misplacement of a fuel assembly Positive

' it Revision 1 8-14-89 o

Table 4.1 FUEL ASSEMBLY DESIGN SPECIFICATIONS Fuel Rod Data Outside diameter, in.

0 360 Cladding thickness, in.

0.0225 Cladding material Zircaloy-4 Pellet diameter, in.

0 3088 002 pellet density, % TD 95 2,2 3

UO2 stack density, g/cm 10.288 2,0.217 Enrichment, wt% U-235 4.2 2,0.05

. Fuel. Assembly Data Number of fuel rocs 264 (17 x 17 array)

Fuel rod pitch, in.

0.496 Centrol rod guide tube Number 24 Outside diameter, in.

0.474 Thickness, in.

0.016 Material Zircaloy-4 Instrument thimble Number 1

Outside diameter, in.

0.474 Thickness, in.

0.016 Material Zircaloy-4 U-235 loading g / axial em of assembly 48.6 2, 1. 0 l i t

km___________._

i's s

j Revision 1 1+

8-14-89 i

l 1:

TABLE 5.1 L

BENCHMARK EXPERIMENTS Soluble General U-235 Separating Boron Description W /0 Reflector Material ppm Keff 1.

UO2 rod lattice 2.46 Water Water 0

0.9857 + /.0028 L

2.

U02 rod lattice 2.46 Water Water 1037 0.9906 ' + /.0018 1

3 UO2 rod lattice 2.46 Water Water

'764

-0.9896 + /.0015

'4 UO2 rod lattice 2.46 Water B4C pins 0

0.9914 + /.0025 5.

Uo2 rod lattice 2.46 Water B4C pins 0

0.9891 + /.0026 6.

UO2 rod lattice 2.46 Water B4C pins 0

0.9955 + /.0020 7.

UO2 rod lattice 2.46 Water B4C pins 0

0.9889 + /.0026 '

8.

UO2 rod-lattice 2.46 Water B4C pins 0

0.9983 ' + /.0025 9.

UO2 rod lattice 2.46 Water Water 0

0.9931 + /.0028 10.

UO2 rod lattice 2.46 Water Water 143 0.9928 + /.0025 11.

UO2 rod lattice 2.46 Water SS 514 0.9967 + /.0020 12.

UO2 rod lattice 2.46 Water SS 217 0.9943 + /.0019 13 UO2 rod lattice 2.46 Water Borated Al 15 0.9892 + /.0023 14 UO2 rod lattice 2.46 Water Borated A1 92 0.9884 + /.0023 15.

U02 rod lattice 2.46 Water Borated Al 395 0.9832 + /.0021 16.

UO2 rod lattice 2.46 Water Borated Al 121 0.9848 + /.0024 17.. UO2 rod lattice 2.46 Water Borated A1 487 0.9895 + /.0020 18.

UO2 rod lattice 2.46 Water Borated Al 197 0.9885 + /.0022

'19.

U02 rod lattice 2.46 Water Borated A1 634 0.9921 + /.0019 20.

UO2 rod lattice 2.46 Water Borated Al 320 0.9920 + /.0020

21. 'UO2 rod lattice 2.46 Water Borated A1 72 0.9939 + /.0020 22.

U metal cylinders 93 2 Bare Air 0

0.9905 + /.0020 23 U metal cylinders 93 2 Bare Air 0

0.9976 + /.0020 24 U metal cylinders 93.2 Bare Air 0

0.9947 + /.0025 25.

U metal cylinders 93.2 Bare Air 0

0.9928 + /.0019 26.

U metal' cylinders 93.2 Bare Air 0

0.9922 + /.0026 27.

U metal cylinders 93.2 Bare Air 0

0.9950 + /.0027 28.

U metal cylinders 93.2 Bare Plexiglas 0

0.9941 + /.0030 29.

U metal cylinders 93 2 Paraffin Plexiglas 0

0.9928 + /.0041 30.

U metal cylinders 93 2 Bare Plexiglas 0

0.9968 + /.0018

31. U metal cylinders 93 2 Paraffin Plexiglas 0

1.0042 + /.0019 32.

U metal cylinders 93 2 Paraffin Plexiglas 0

0.9963 + /.0030 33 U metal cylinders 93.2 Paraffin Plexiglas 0

0.9919 + /.0032 - _ __ ___

F Revision 1 c*-

314 g9 Table 5.2 COMPARISON OF COLD, CLEAN REACTIVITIES CALCULATED AT 25,000 MWD /MTU BURNUP AND 3 2% ENRICHMENT O

k,(Xe-free, O C)

Calculational Method Fuel Assembly In Region 2 Cell CASMO-2 E-1.1206 0.9061 NULIF /PDQ07 1.1294 0.9017 EPRI-CELL 1.1201(l)

II)EPRI-CELL k, at maximum value during long-term (30-year) storage.

Table 5.3 ESTIMATED UNCERTAINTIES IN REACTIVITY DUE TO FUEL DEPLETION EFFECTS Design 0.5 x 10-6 Initial Burnup Times Reactivit{I)

Enrichment MWD /MTU Burnup, Ak Loss, ak 1.8 5,230 0.0026 0.0475 2.5 15,720 0.0079 0.1575 3.2 25,000 0.0125 0.2337 37 31,280 0.0156 0.2757 4.2 37,370 0.0187 0.3107 II) Total reactivity decrease, calculated for the cold, Xe-free condition in the fuel storage rack, from the beginning-of-life to the design burnup.

l l _ _ _ _ - _ _ _

Revision 1

+ :' -

B-14 -89 Table 5.4 LONG-TERM CHANGES IN REACTIVITY IN STORAGE RACK (XENON-FREE)

Ak from Shutdown (Xenon-free)

Storage 3 2%E 4.2%E Time, years 625,000 MWD /MTU 637,000 MWD /MTU 0.5

-0.0046

-0.0057 1.0

-0.0080

-0.0103 10.0

-0.0406

-0.0529 20.0

-0.0588

-0.0756 I

30.0

-0.0692

-0.0886 l

\\

.j l

)

1 l

1 I i

Revision 1

+

8-14-89 Table 7.1 FUEL BURNUP VALUES FOR REQUIRED REACTIVITIES (k.)

WITH FUEL OF VARIOUS INITIAL ENRICHMEffrS Fuel j

' Initial Reference Uncertainty Design Burnup,

Enrichment k.

In Burnup, Ak Limit k.

MWDAiTU 1.58 0.9186 0

0.9186 0

1.8 0.9186 0.'0026 0.9160 5,230 2.5 0.9186 0.0079 0.9107 15,720 32 0.9186 0.0125 0.9061 25,000 3.7 0.9186 0.0156 0.9030 31,280 4.2 0.9186 0.0187 0.8999 37,370 l I

___]

Revision 1

.o 8-14-89'

. + - + >

Table 8.1 EFFECT OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF STORAGE RACK Incremental Reactivity Change, Ak Case Region 1 Region 2 0

0C Reference Reference 0

20 C

-0.0022

-0.0047 50 C

-0.0084

-0.0081 80 C

-0.0165

-0.01 21 0

120 C

-0.0298

-0.0178 0

120 C + 20% void

-0.0953

-0.0520 l'

l l

l a

1 i

l 1 :

i

_..____.._______________J

Revision 1 8-14-89 Table 10.1 Integrated Dose at the Center Line Between Two Spent Fuel Assemblies" in h'ater at the ByronMraldwood Stations Time After Reactor Shutdown Dose (Days).

( Ra d-Carbon )

0 1-6.3 x 108 2

9.2 x 10 8 3

1.2 x 109' 4

1.4 x 10 9

5 1.6 x 10 6

1.8 x 109 9

'I 2.0 x 10 9

14 3.0 x 10 9 30 4.7 x 10 9

60 6.8 x 10 90 8.2 x 109 9

120 9.2 x 1010 150 1.0 x 1010 180 1.1 x 10 10 240 1.2 x 10 10 365 1 3 x 10 10 730 1.4 x 1010 1,095 1.5 x 1010 1,460 1.6 x 1010 1,825 1.7 x 10 10 l

2,555 1.8 x 1010 3,650 1.9 x 10 10 5,475 2.2 x 1010 7,300 2 3 x 10 10 9,125 2.4 x 10 10 10,950 2.6 x 1010 12,770 2.7 x 10 10 14,600 2.8 x 10 i

  • The assemblies have been in an operating reactor for a period of three years.

l._

l Revision 2 8-14-89 l

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1.5 2.0 2.5 3.0 3.5 4.0 4.5 INITIAL ENRICHMENT,WTX U-232i Fig. 3.1 ACCEPTABLE BURNUP DOMAIN IN REGION 2 OF THE SPENT FUEL STORAGE RACKS. i l

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fuel assemblies.

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APPENDIX A BENCHMARK CALCULATIONS i

A-1

9'*:

t Revision 1

' 4_ t <:

8-14 l 11.

INTRODUCTION AND SUMMARN 1

JThe objective 'of this benchmarking study is to verify both the AMPX (NITAWL)-KENO (Refs.1 and 2) methodology -with the 27 group SCALE cross-l section library -(Refs. 3 and 4) and the CASMO-2E code- (Refs. 5, 6, 7, and

8) for use in. criticality calculations of high density spent fuel storage racks.

Both calculational methods are based on 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 cor-l

=

responding' calculations reported in the literature and with the require-l' ments ;of Regulatory Ouide 3.41,* Rev.1, May 1977.

l 1.

Results of these benchmark calculations show that the 27 group (SCALE)

AMPX-KENO calculations consistently underpredict the critical eigenvalue by 0.010610.0048 Ak (with a 955 probability at a 955 confidence level) for critical experiments selected to be representative of realistic spent fuel storage. rack configurations and poison worths.

Similar calculations by Westinghouse suggest a bias of 0.012 1 0.0023, and the results of ORNL

' analyses of 54 relatively " clean" critical experiments.show a bias of -

0.0100 i 0.0013.-

Similar calculations with CASMO-2E for clean critical experiments resulted in a bias of 0.0013 i 0.0018 (955 /955). C ASMO-2E and AMPX-KENO intercomparison calculations of infinite arrays of poisoned cell con-figurations show very good agreement and suggest that a' bias of 0.0013 i 0.0018 is the reasonably expected bias and uncertainty for CASMO-2E calculations.

The benchmark calculations reported here indicate that either the 27-group (SCALE) AMPX-KENO or CASMO-2E calculations are acceptable for criticality analysis of high density spent fuel storage racks. The preferred methodology, however, is to perform independent calculations

  • Validation of Calculational Methods for Nuclear Criticality Safety.

(See also ANSI N16.9-1975.)

A-2

['

Revision 1 8-14-89

+

s

[

with both code packages and to utilize the higher, more conservative l

value for the reference design infinite multiplication factor.

l l

2.

AMPX (NITAWL)-KENO BENCHMARK CALCULATIONS Analysis of a series of Babcock & Wilcox (B&W) critical experiments (Ref. 9), which include some with absorber sheets typical of a poisoned spent fuel rack, is summarized in Table 1, as calculated with AMPX-KENO using the 27 group SCALE cross-section library and the Nordheim resonance integral treatment in NITAWL. The mean for these calculations is 0.9894 1 0.0019, conservatively assuming the larger standard deviation calcu-lated from the K,7f values. With a one-sided tolerance factor (K =

2.502), corresponding to 95% probability at a 955 confidence level (Ref.10), the calculational bias is +0.0106 with an uncertainty of 0.0048.

Similar calculational deviations reported by Westinghouse (Ref.11) are also shown in Table 1 and suggest a bias of 0.012 1 0.0023 (955/95%).

In addition, ORNL (Ref.12) has analyzed some 54 critical experiments using the same methodology, obtaining a mean bias of 0.0100 1 0.0013 (955 /95% ).

These published results are in good agreement with the results obtained in the present analysis and lend further credence to the validity of the 27 group AMPX-KENO calculational model for use in criti-rality analysis of high density spent fuel storage racks. Variance analysis of the data in Table 1 suggests the possibility that an unknown factor may be causing a slightly larger variance than might be expected from the Monte Carlo statistics alone.

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

No trends in k,pg with intra-assembly water gap, with absorber sheet reactf.vity worth, or with soluble poison concentration were identified.'

'Significantly large trends in k,77 with water gap and with absorber reactivity worth have been reported (Ref.16) for AMPX-KENO calculations with the 123-group GAM-THERMOS library.

A-3

N '"*

Revision 1 4.

i' B-14 -8 9 y

3 CASi40-2E' BENCHMARK CALCULATIONS

[x i.'

131.. GENERAL' g

lj

.. The. CASMO-2E code :is a. multigroup transport theory code util'izing trans-n

mission probabilities to acconiplish two-dimensional calculations of j:

reactivity. and~ depletion for BWR and PWR fuel assemblies, s such.

e,

' CASMO-2E is well-suited to the criticality analysis of spenc fuel-storage racks, since: general practice is to treat the racks as an infinite medium of storage cells, neglecting leakage effects.

CASMO-2E is closely analogous to the EPHI-CPM' code (Ref.13) and has been extensively benchmarked against hot and cold critical experiments by Studsvilc EnergiteknikJ(Refs. 5, 6, 7, and 8).

Reported analyses of 26 critical' experiments ' indicate a mean k,7f of 1.00010.0037. (1 o).

Yankee Atomic -(Ref.14) has' also reported results of extensive benchmark calou-:

lations with CASMO-2E..Their analysis of 54 Strawbridge and Barry. criti-

~

cal exper.w;s (Ref.15) using the reported buckling indicates a mean.or' O.9987 i = 0.0009 (10), or a bias of' O.0013

0.0018 (with 955 probability.

at a 955 sonfidence. level).- Calculations were repeated for seven of. the Strawbridge and Barry 4xperiments selected at random, yielding a mean -

.k,7f of.0.9987:

0.0021 :(1 o), thereby confirming that the cross-section-library and analytical methodology being used for the present calcula-

~

tions are the same'as those used in the Yankee analyses. Th 1s, the expected bias for CA3MO-2E in the analysis of." clean" critical.experi-ments is 0.0013 1 0.0018 (955 6 55).

3.2 BENCMMARK CALCULATIONS CASMO-2E benchmark calculations have also been made for the B&W series of critical experiments with absorber sheets, simulating high density spent fuel storage racks. However, CASMO-2E, as an assembly code, cannot A-4

W!s

" [

Revision.1-

. s Ln 3-14-89 dir=ctly represent an entire core configuration" without introducing l..

uncertainty due to reflector constants and the appropriateness of their l-spectral weighting.

For.. this reason, the poisoned cell configurations of the central assembly, as calculated by CASMD-2E, were benchmarked against l

corresponding calculations with the 27-group (SCALE) AMPX-KENO code packa ge.

Results of this compactson are shown in Table 2.

Since the differences are well within the normal KENO statistical variation, these calculations confirm the validity of CASMO-2E calculations for the typi-cal high density poisoned spent fuel rack configurations. The differ-ences shown in Table 2 are also consistent with a bias of 0.0013 1 0.0018, determined in Section 3.1 as the expected bias and uncertainty of CASMD-2E calculations.

l l

l~

" Yankee has attempted such calculations (Ref.14) using CASMO-2E-generated constants in a two-dimensional, four group PDQ model, obtaining a mean kepp of 1.005 for 11 poisoned cases and 1.009 for 5 unpoisoned cases.

Thus, Yankee benchmark calculations suggest that CASMO-2E tends to slightly overpredict reactivity.

A-5

Revj sion 1 8-14 -B 9 i

Table 1 RESULTS OF 27-OROUP (SCALE) AMPX-KENO CALCULATIONS OF B&W CRITICAL EXPERIMENTS l

l Westinghouse l

Experiment Calculated Calculated-me as.

Number k

o k

gf7 eff I

0.9889 0.0049

-0.008 II 1.0040 10.0037

-0.012 III 0.9985 0.0046

-0.008 II)

IX 0.9924 10.0046

-0.016 X

0.9907 10.0039

-0.008 XI 0.9989 10.0044

+0.002 XII 0.9932 0.0046

-0.013 XIII 0.9890 0.0054

-0.007 XIV 0.9830 0.0038

-0.013 XV 0.9852 10.0044

-0.016 XVI 0.9875

+0.0042

-0.015 l

XVII 0.9811 0.0041

-0.015 XVIII 0.9784 0.0050

-0.015 XIX 0.9888 to.0033

-0.016 XX 0.9922 10.0048

-0.011 XXI 0.9783 10.0039

-0.017 Mean 0.9894 0.0011(2)

-0.0120 t 0.0010 Bias 0.0106 0.0019(3) o,o3po 1 o,oogo Bias (95% /95%)

0.0106 0.0048 0.0120 0.0023 1

Maximum Bias 0.0154 0.0143 Il) Experiments IV through VII used B C pin absorbers and were not considered q

j i

representative of poisoned storage racks.

I2) Calculated from individual standard deviations.

(3) Calculated from Reff values and used as references, j

)

l' i

A-6 l

(_

  • ' (..-

Revision 1 3-14-89 Table 2 RESULTS OF CASMO-2E BENCHMARK (INTERCOMPARISON) CALCULATIONS 1

k (l)

B&W Experiment No.II)

AMPX-KEN 0(2)

CASMO-2E Ak XIX 1.1203 1 0.0032 1.1193 0.0010 XVII 1.1149 0.0039 1.1129 0.0020 XV 1.1059 i 0.0038 1.1052 0.0007 Interpolated (3) 1.1024 0.0042 1.1011 0.0013

.XIV 1.0983 t 0.0041 1.0979 0.0004 XII 1.0992 t 0.0034 1.0979 0.0013 Mean 1 0.0038 0.0011 Uncertainty 0.0006 BWR fuel rack 0.9212 1 0.0027 0.9218

-0.006 5I) Infinite array of central assemblies of 9-assembly B&W critical configura-tion (Ref. 9).

(2)

53) k, from AMPX-KENO corrected for bias of 0.0106 ok.

Interpolated from Fig. 28 of Ref. 9 for soluble boron concentration at critical condition.

A-7

o-

+3 Revision 1 o.

  • 8-14-89 REFERENCES TO APPENDIX A 1.

Green, Lucious, Petrie, Ford, White, Wright, "PSR-63 /AMPX-1 (code packaSe), AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF /B," ORNL-TM-3706, Oak RidSe National Laboratory, March 1976.

2.

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

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

4 W. E. Ford, III et al., "A 218-Neutron Group Master Cross-section Library for Criticality Safety Studies," ORNL /TM-4,1976.

5.

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

6.

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

7.

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

8.

"CASMO-2E Nuclear Fuel Assembly Analysis, Application Users Manual,"

Rev. A, Control Data Corporation,1982.

9.

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

10.

M. G. Natrella, Experimental Statistics, National Bureau of Standards.

Handbook 91, August 1963 11.

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

12.

R. M. Westf all and J. R. Knight, " Scale System Cross-section Validation with Shipping-cast Critical Experiments," ANS Transactions, Vol. 33,

p. 368, November 1979.

13 "The EPRI-CPM Data Library," ARMP Computer Code Manuals, Part II, Chapter 4, CCM3, Electric Power Research Institute, November 1975.

14 E. E. Pilat, " Methods for the Analysis of Boiling Water Reactors (Lattice Physics)," YAEC-1232, Yankee Atomic Electric Co., December 1980.

A-8

y.

~ h.., )

Revision 1 o

8-14-89 f.

15.

L. E. Strawbridge and R. F. Barry, " Criticality Calculations for Uniform, l-Water-moderated Lattices," Nujiear Science and Engineering, Vol. 23,

p. 58, September 1965.

l l

16. -

S. E. Truner and M. K. Gurley, " Evaluation of AMPX-KENO Belichmark Calculations for High Density Spent Fuel Storage Racks," Nuclear Science and Engineering, 80(2): 230-237, February 1982.

1 A-9

_ - - _ _ - _

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