ML20097C313

Criticality SE of Fort Calhoun Spent Fuel Storage Racks for Maximum Enrichment Capability

ML20097C313
Person / Time
Site:	Fort Calhoun
Issue date:	12/31/1995
From:	Sarah Turner HOLTEC INTERNATIONAL
To:
Shared Package
ML20097C295	List: LIC-96-0010, Forwards Application for Amend to License DPR-40 & TS, Allowing Increase in Initial Nominal U-235 Enrichment Limit of Fuel Assemblies That May Be Stored in Spent Fuel Pool ML20097C303 ML20097C308 ML20097C313
References
HI-951400, NUDOCS 9602080184
Download: ML20097C313 (43)

Text

.

CRITICALITY SAFETY EVALUATION OF THE FT. CALHOUN SPENT FUEL STORAGE RACKS FOR MAXIMUM ENRICHMENT CAPABILITY Prepared for the OMAHA PUBLIC POWER DISTRICT

, by Stanley E. Turner, PhD, PE i

December 1995 Holtec Project 51085 Holtec Report HI-951400 4

HOLTEC INTERNATIONAL l

230 Normandy Circle 2060 Fairfax Ave.

Palm Harbor, FL 34683 Cherry Hill, NJ 08003 9602080184 960201 PDR ADOCK 05000285 P

PDR j

m a

-[

REVIEW AND CERTIFICATION LOG DOCUMENT NAME:

CRITICALITY SAFETY EVALUATION OF THE FT. CALHOUN SPENT FUEL STORAGE RACKS FOR MAXIMUM ENRICHMENT CAPABILITY HOLTEC DOCUMENT I.D. NO.

HI-951400 HOLTEC PROJECT NUMBER HI-51085 CUSTOMER / CLIENT OMAHA PUBLIC POWER DISTRICT l

l REVISION BIOCK N

1 QUALITY PROJECT ISSUE AUTHOR REVIEWER ASSURANCE MANAGER i

NO.

& DATE

& DATE

& DATE

& DATE S.E. Turner E.REDMOND V. GoPT A S. 6.7'vA N E R

' I m. op %<.

m:.m is./4/W p if W n.-y-%5 e a h*,4.

REV. 2 REV. 3 Must be Project Manager or his Designee.

l NOTE: Signatures and printed names are required in the review block.

This document conforms to the requirements of the design specification and the applicable sections of the governing codes.

(m

=

a

TABLE OF CONTENTS

1.0 INTRODUCTION

and

SUMMARY

1 2.0 CRITICALITY SAFETY ANALYSES 4

2.1 Regio n 1.................................

4 t

1 2.2 Reg ion 2.................................

5 2.3 Peripheral Cells (Region 3) 6 2.4 Accident / Abnormal Conditions 7

2.5. Soluble Boron Considerations 7

3.0 ANALYTICAL BASES 9

3.1 Fuel Assembly Specifications 9

3.2 Storage Rack Specifications 9

3.2.1 Region 1 9

3.2.2 Region 2 10 3.3 Manufacturing Tolerances and Uncertainties 10 3.4 Calculational Methodology 10 3.4.1 Computer Codes

.........................10 3.4.2 Verification Calculations I1 i

4.0 REFERENCES

................................12 i

I

t l

2 l

i 1

List of Tables l

Table 1 Summary of Criticality Safety Analyses 13

)

I Table 2 Evaluation of the Minimum Burnup Requirements in Region 2.....

14 1

Table 3 Evaluation of the Consequences of Fuel Handling Accidents...... 15 i

)

1 Table 4 Design Basis Fuel Assembly Specifications.................

16' List of Figures 1

4 Fig.1 Acceptable Burnup Domain in Region 2................... 17 Fig. 2 Effect of Enrichment on Region 1 Reactivity

'I8 Fig. 3 Reactivity Effect of Soluble Boron (5% Enriched Fuel) 19 j

i Fig. 4 ' Region 1 Cross-Section View........................... 20 i

Fig. 5 Region 2 Cross-Section View........................... 21 j

ii

l

1.0 INTRODUCTION

and

SUMMARY

t 1

i I

The Ft. Calhoun spent fuel storage racks were originally designedW to accommodate fresh fuel of 4.2% enrichment in Region 1 or spent-fuel of 4.2% initial enrichment burned.to 32 MWD /KgU in Region 2, using Boral as the poison material. There was an appreciable margin

~

available below the NRC Regulatory limit and the present study was undertaken to upgrade the capability of the racks to accommodate fuel of higher enrichments. The previous criticality safety.

i evaluation had established that:

Westinghouse fuel results in a higher reactivity than the fuel manufactared by either Combustion Engineering or ANF, 1

i

. - the temperature and void coefficients of reactivity are negative, and

]

I the reactivity effect of eccentric fuel positioning is negative.

In the present evaluation, the analyses were extended to (1) assess the maximum enrichment capability of the racks and (2) to evaluate the additional reactivity control required to enable the racks to safely accommodate fuel with enrichments up to 5%. Result of the analyses established that Region 1 of the racks can safely accommodate fuel of 4.75% enrichment with a maximum

. reactivity within USNRC Guidelines (<0.95 k,y), without any restrictions.

For fuel with enrichments greater than 4.75%, there are three possible restrictions for storage in Region 1, any one of.which will assure the maximum k,y (95% probability, 95% confidence level) will be '

maintained less than the regulatory limit under normal storage conditions. These include (1) the presence of soluble boron in the pool water, (2) checkerboarding of fuel assemblies, and (3) credit for very limited fuel burnup. In Region 2, credit for burnup was included, extending the previous burnup limit curve to encompass fuel up to 5% enrichment. In addition, the criticality safety evaluation included (1) the use of existing spent control rods (CEA's) in Region 2 and (2) the consequence of postulated accident conditions in both storage regions.

i 1

For each region, the reactivity uncertainty associated with manufacturing tolerances and calculational uncertainties were re-evaluated for the higher enrichment fuel and found to essentially the same as that from the previous analysis *.

Based upon the criticality safety analyses reported here, the following conclusions may i

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p,

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May one: sf the following criteria' ar:eVcsept$i>lAIfoid(etsrminingjhe safes *,

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{

{ st' rags of fuel in:Regi6n 1:c.

_ ~. j-i

,, 2@

! Fuel a semblies with;an enrichmentfof%75%].or?lessf ors

,"s

'r

'y Fsdas^emblies with'eririchmenislup soIS.0*N,5rovidefafminimuin sh~hES

'M s

9hdron concentration"6fe75 ppm or mois:is innintainsd,3er!

w EFuel assemblies with" enrichments uitoj5 0%%bishjhave:-sttained*afmi$Inin m

Ibernup:ofl1000. MWD /MTU (orimbr;eRorf 4

gc f

......,.....y..-.,.

..y W MFuel assemblies with"sufichmen_ts up;to 5.0%l stored.in aithree-ou..

,,7

,,,,.,...A

...,,,,,.,,s...

T

'Y I

ifour chscksiboar'dfpattern withltheffouitlifcelljfd' led lwithSwati@s t-of L

inon-fast biaring thaterikts)R 4

'i dp

}

5Aspf6ne 'oI the foll6 wing cridriaTare. acceptable [forNAlermining the

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fe ; :

istorage of fusiin Regioni26

.,m.

.=

.m M.sFuel. assemblies wikenrichmests up30l5.d'IMhicli[have'ittiinsdial Hm nimsm burnupfwithin ths accept _able domain"of Figure li'or-

-v : Unirradidted;l fuel assemblies wittienrichments up ioM7'A/:with full-Isn$tif a

?CEA' Rods'insefted,7ord

.Ur irradiated [ fuel assemblies with'enrididnents up;td'5.0%istored!in[as V

'two-out-of-four checkerboard.patternTwithP lternate cellsVfilled with) a

< waterL(or: non'-fuel biadng materials).1 -

c 2

To assure criticality safety under e.ll 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 in draft Regulatory Guide 1.13 (Rev.

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

Evaluation of postulated accident conditions in both Region 1 and Region 2 resulted in the following conclusions:

Inn wyWisihimurd concenthafion 6f s$1ublsNron!!rilhAl post wster df5053 phi}[

[Id)reconunehdedSioPassureithath undep]AlijpostdistedicordliUAnshiNei

<.s J.'; f.:........ -- 4 l maximum'kiwill'be mAintainsd le's thhn,,.... :::the Rsgulatoryjimit (0.95%)p s,

... - s s

d...,o...,c,...:'

'+;s...

'''Y LThilsinglelexception7is t'hbLuse of(allichsckssboardl loading?pstterniin3 -

15Eskhhh ![fobhN15diaKminimumkof$5.00NppmMsdlubhNEonY[id[#

[seconuneS5Ad"tsLpr61 sit agiinsi ifdeImisloudhi'g sicidsnid '

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es 3

4 2.0 CRITICALITY SAFETY.' ANALYSES 2.1 ' Region 1 -

-The fuel storage racids in the Ft. Calhoun' spent fuel storage' pool Region 1 use Boral absorber material and a water-gap between cells (flux-trap) to augment reactivity control.. Calculations of.--

Region I with Westinghouse fuel of various enrichments are shown in Figure 2 where the upper ~

' curve represents the maximum reactivity, including bias and uncertainties (at the 95% probability, 95% confidence levelm) and'with creditL for the finite axial length of the active fuel..The maximum k,y begins to excedd the acceptable limit (k,y of 0.95) above an enrichment of 4.75%.

.The presence of axial blankets does not increase reactivity or alter the acceptance criteria,~

provided the assemblies are evaluated for the enrichment of the active zone (without averaging

- to include the blanket enrichment).

' Results of the criticality safety analyses for 4.75% fuel in Region 1 are summarized in Table l'.

These data show a maximum k,y of 0.9498 for fuel of 4.75% enrichment, iricluding bias and uncertainties. Ira addition, the reactivity effects of soluble boron, fuel burnup and checkerboard

- loading patterns.were also evaluated.- For 5.0% enriched fuel, a soluble boron concentration of 75 ppm reduces the maximum k,y to less than 0.95. A three-out-of-four checkerboard loading

- pattern is also acceptable for storage of 5.0 enriched fuel, with a maximum k,y of 0.900.

i Fuel burnup is an alternative way of safely accommodating fuel with enrichments greater *. nan 4.75% in Region 1.

Calculations (with CASMO3, using the re-start option) were made for fuel of 5% initial enrichments. From these calculations, it was determined that a burnup of 850 MWD /MTU would reduce the reactivity to the equivalent of 4.75% enrichment, and therefore acceptable for storage in Region 1. For conservatism, the minimum required burnup was rounded up to 1000 MWD /MTU.

4

)

4

-..~

For protection against a fuel handling accident in Region 1, a minimum 100 ppm soluble boron concentration would be necessary, This concentration would allow unrestricted storage of 5%

fuel in Region I with assurance that the maximum k,a, including uncertainties, will be maintained less than the regulatory limit (0.95) for all conditions. However, Region 2 requires a minimum soluble boron concentration of 150 ppm, which would encompass the Region I requirement.

2.2 Recion 2 For Region 2, the previous criticality analyses were extended to encompass fuel of 5% initial enrichment, including the consequence of a representative axial burnup distribution. Results of these analyses are shown in Figure 1 (and in Table 2) which defines the burnup domain for acceptable storage of spent fuel with initial enrichments up _to 5.0%. Fuel which has attained a burnup within the acceptable domain of Figure 1 may be safely stored in Region 2 with a calculated maximum k,y of 0.935 as indicated in Table 1. Data shown for Region 2 in Table 1 is for fuel of 4.75% initial enrichment burned to 38,900 MWD /MTU. For 5.0% enriched fuel, the corresponding limiting burnup (equivalent reactivity) is 42,300 MWD /MTU. Fuel of other enrichments will have an equivalent maximum reactivity for the corresponding limiting burnup shown in Figure 1. Axial blankets would reduce reactivity in Region 2 (because of the axial burnup distribution) provided the assemblies are evaluated for the enrichment in the central active zone, without averaging in the blanket enrichment.

The data in Figure 1 may be described by an empirical equation as a function of initial

)

enrichments (E,%), up to 5%, as follows:

i

.<.e.

y'.

g iRsgion 2; Minimum Burnap/ MWD /MTU

~

u w

l i35520 d27920f*lEQd600[* FEM.. 426 *.!E'))

gi In addition, it was determined that a checkerboard loading arrangement with a 2-out-of-four 5

pattern in Region 2 was acceptable for fresh unburned fuel up to 5.0% enrichment, with a maximum k,y (including bias and uncertainties) of 0.823.

Soluble baron is necessary in Region 2 as protection against a fuel handling accident (mis-placed assembly - see Section 2.4).

Evaluation of spent control rods (CliA's) was also made for storage in Region 2 of the racks,

~

using the same design and criteria as in the initial analyses. Results of this analysis, listed below, show that fuel up to 4.7% enrichment may be safely stored with the CEA rods installed.

Enrichments above 4.7%, however, would exceed the reactivity limit and would require additional reactivity control.

Reactivity with CEA Rods Enrichment %

Installed 4.2 0.923 4.5 0.939 4.7 0.949 (Interpolated) 4.75 0.951 2.3 Perioheral Cells (Recion 3)

The peripheral cells (high-neutron leakage area in Region 2, called Region 3) were initially qualified for fuel with burnups less than that required for unconditional storage in Region 2.

Extending the enrichment evaluation for these cells resulted in the lower curve shown in Figure 1, with the same maximum reactivity as before. Limiting fuel burnups for the peripheral cells (Region 3) have also been calculated and the results fitted to a polynomial expression over the range from 2% to 5% enrichment, as follows:

JRegion 3 Minimum'Burnup?; MWD /MTUj p.

5237603 (141601*LE)h(4891*, EWi p

._a 6

2.4 Accident / Abnormal Conditions As determined in the original evaluation"', the temperature and void coefficients of reactivity are negative in both regions. Consequently, the maximum reactivity occurs at a water density of 1.0 g/cc and 4 C was therefore used as the design basis temperature. This conclusion was re-affirmed by specific calculations at the higher enrichments for both Region 1 and Region 2.

Other accident conditions were also evaluated in the original analysis. However, the increase in enrichment capability necessitated re-evaluation of the fuel handling accident in which a new-fuel assembly of 5.0% enrichment might be accidentally loaded into, or outside of, a Region 1 or Region 2 cell when all other cells filled with fuel of the maximum permissible reactivity.

Analysis of the postulated fuel handling accidents, listed in Table 3, show that a soluble boron concentration of 425 ppm is required to protect against the most severe credible accident (mis-loading of a 5% enriched assembly when using the checkerboard option). For the more usual storage configurations in Region 2, the most severe accident is the mis-loading of a 5% enriched assembly into a cell with all other cells filled with fuel of the maximum permissible reactivity.

This accidrnt would require a minimum of 180 ppm boron, rounded up to 200 ppm to allow for uncertainty in measuring the boron concentration.

2.5 Soluble Boron Considerations i

Soluble boron is normally maintained in the spent fuel pool water at approximately 2000 ppm.

At this concentration, the normal reactivity of the storage racks is very low (k, of about 0.76) and the soluble boron would be sufficient to compensate fer any credible accident condition.

However, the racks were analyzed under the single failure assumption of the complete loss of all soluble boron. Under other accident conditions, credit for the presence of the soluble boron is permissible under the double contingency principle. Some soluble boron is necessary to protect against adverse consequences of postulated accidents and the minimum concentrations required have been evaluated for credible accident conditions (see Section 2.4).

7

Low concentrations of soluble boron could be used to enable the Region I rack l

accommodate fuel assemblies with up to 5% enrichment. CASMO and KEN 05a calculations of the reactivity effect of soluble boron are shown in Figure 3 for fuel of 5% enrichment a range up to 200 ppm boron, the soluble boron has a reactivity " worth" of approximate 0.0125 Ak per 100 ppm boron. Figure 3 shows that, for fuel of 5.0% enrichment, a solubl boron concentration of about 65 ppm would be required for a k, of 0.95 in Region 1. F conservatism, the required boron concentration was rounded up to 75 ppm.

i A minimum soluble boron concentration of 100 ppm is required in Region I and 18 l

Region 2 to mitigate the consequences of a fuel mis-loading accident (except for the.

checkerboard configuration in Region 2 which requires 425 ppm boron).

A minimum boron concentration of 200 ppm (rounded up from the calculated 180 ppm) would allow unrestricted storage of fuel up to 5.0% in enrichment in both Region 1 and Region 2, subject only to th burnup requirements in Region 2 (as defined in Figure 1), and assuming that a checkerbl loading configuration is not used in Region 2.

If a checkerboard configuration is used in Region 2 for 5.0% unirradiated fuel storage, the calculated minimum boron concentration o ppm should be rounded up to 500 ppm for conservatism and to allow for uncertainty in measurement.

8 i

2

t 3.0 ANALYTICAL BASES 3.1 Fuel Assembiv Specifications The reference fuel assembly used for the analyses is the Westinghouse 14 x:14 fuel assembly.

with 20 rods replaced by control thimbles. the same as used in the original analyses, which had also determined that this assembly gave a higher reactivity than the available alternate designs.-

~

Table 4 lists the design specifications for the fuel used in the analyses.

An axial blanket of UO with an enrichment less than the normal fuel enrichment (e.g.,2%)

2 would result in slightly lower and more conservative reactivities. Fuel enrichments, as used in this report, refer to the enriched fuel zone in the assembly without consideration of any axial.

i blankets that might be present.'

9 3.2 Storage Rack Soecifications 3.2.1 Region 1 The nominal spent fuel storage cell used for the criticality analyses of Region 1 storage cells is shown in Figure 4. The rack is composed of Boral absorber material on the outside of a 8.46-inch I.D., 0.075-inch thick stainless steel box. The fuel assemblies are centrally located in each storage cell on a nominal lattice spacing of 10.363

• 0.080 inches in one direction and 9.821 0.080 inches in the other direction.

Stainless steel channels connect one storage cell box to another in a rigid structure and define a water flux trap between the two (thermal-neutron opaque)

Boral absorber panels. The 7.25-inch wide Boral absorber has a nominal thickness of 0.075

• 2 0.004 inch and a nominal B-10 areal density of 0.0151 0.0011 g/cm,

9

c.

3.2.2 Region 2 In Region 2, the storage cells are composed of a single Boral absorber panel between the stainless steel walls of adjacent storage cells. These cells, shown in Figure 5, are located on a lattice spacing.of 8.652

• 0.040 inches. The Boral absorber has a thickness of 0.075 0.004 inch and a nominal B-10 areal density of 0.0151

• 0.0011 g/cm (minimum of 0.014 g/cm ),

2 2

3.3 Manufacturine Tolerances and Uncertainties The small reactivity increments' associated with manufacturing tolerances developed in the previous evaluation were verified to assure that the higher 5.0% enrichment would result in a significantly different tolerance uncertainty than thit from the previous evaluation. Results, shown in Table I are consistent with the previous antaysis.

3.4 Calculational Methodolony 3.4.1 Computer Codes The principal method of analysis was the CASMO-3* code, a two-dimensional multi-group transport code for assemblies and the NITAWL-KENOW code package, a three dimensional Monte Carlo code package, using the 27-group SCALE cross-section library. Supplementary analyses for independent verification were performed with the 218-group cross-section library in KEN 05a and with the MCNP code * (a continuous energy Monte Carlo code developed by the Los Alamos National Laboratory).

Benchmarking of the codes, summarized in Appendix A, resulted in the following bias values (at the 95% probability, 95% confidence level *):

CASMO-3 0.0000

• 0.0024 NITAWL-KENO 5a 27-Group Library 0.0103 0.0018 218-Group Library 0.0128

• 0.0020 MCNP 0.0032 0.0020 10

in the geometric model used in the calculations. each fuel rod and its cladding were described t

explicitly.

Reflecting boundary conditions (zero neution current) were used in the radial i

]

direction which has the effect of creating an infinite array of storage cells in X-Y directions. In the KENO-Sa and McNP models, the actual fuel assembly length was used in the' axial direction, assuming thick (30 cm) water reflectors top and bottom. Since Monte Carlo (KENO-Sa and MCNP) calculations inherently include a statistical uncertainty due to the random nature of neutron tracking, a minimum of I x 10 neutron histories were accumulated in each calculation.

i 6

CASMO3 was used for depletion analyses and in the evaluation of the small reactivity effects of.

manufacturing tolerances. As in the previous analysis, an uncertainty in depletion calculations equal to 5% of the reactivity decrement from the beginning oflife to the burnup ofintemt was

' assumed. CASMO3 was also used to determine equivalent enrichments corresponding to.10

' zones of the axial burnup distribution. Three-dimensional KEN 05a calculations, with the 10 axial zones, enabled the reactivity effect of the distribution in burnup to be determined.

3.4.2 Verification Calculations Independent verification calculations 'were made with both the 27-group and the-218-group SCALE cross-section libraries in KEN 05a, and with MCNP and CASMO3 for selected cases.

These results are shown below (maximum k,,, including bias and uncertainties):

CASE CASMO3 MCNP 27-Grouc 218-Group Region I with 4.75% Fuel 0.9498 0.9486 0.9493 0.9483 Region I with 5.0% Fuel 0.9586 0.9573 0.9589 0.9579 Region 2 with 1.6513% Fuel

• 0.9438 0.9437 0.9450 0.9454

• Equivalent to 4.75% Enriched Fuel @ 37.500 MWD /MTU i

The good agreement for these various cases tend to confirm the validity of the analytical results reported here.

i 11

4.0 REFERENCES

1.

Licensinc Report for Soent Fuel Storace Caoacity Expansion. Holtec Report HI-92828 2

Omaha Public Power District, Fort Calhoun Station, Rev. 5, November 1992.-

2.

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

3.

-A Ahlin, M. Edenius, H. Haggblom, "CASMO - A Fuel Assembly Burnup Program,"

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

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

"CASMO-3 A Fuel Assembly Burnup Program, Users Manual", Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986 M. Edenius and A. Ahlin, "CASMO-3: 14ew Features, Benchmarking, and Advanced Applications", Nuclear Science and Engineering, 100, 342-351,(1988) 4.

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 oerforming Standardized Comouter Analyses for Licensing Evaluation., NUREG/CR-0200, I

1979.

L.M. Petrie and N.F. Landers," KENO 5a. An Improved Monte Carlo Criticality Program with Supergrouping" in Scale: A Modular Code System for oerformine Standardized Computer Analvses for Licensine Evaluation. NUREG/V-0200,.1979.

R.M. Westfall et al., " SCALE: A Modular Code System for oerformine Standardized Gmouter Analvses for Licensine Evaluation." NUREG/V-0200.1979 5.

J.F. Briesmeister. Ed., "MCNP - A General Monte Carlo N-Particle Transoort Code.

Version 4A.", Los Alamos National Laboratory. LA-12625-M (1993).

12

Table 1 1

Summary of Criticality Safety Analyses l

Region 1 Region 2 Design Basis 4.75% enrichment 4.75 % enrichment at 38,900 MWD /MTU l

Temperature for analysis 4C 4*C l;

1 L

Reference k. (CASMO-3) 0.9392 0.9016 Uncertainties In Bias

• 0.0024 0.0024 B-10 loading-0.0031 0.0032 Boral width 0.0008 0.0006 Inner box dimension

• 0.0009 0.0011 Water gap thickness

• 0.0093 NA SS thickness 0.0004

• 0.0002 Fuel enrichment")

• 0.0018

• 0.0018 m

• 0.0022

• 0.0022 Fuel density Eccentric position Negative Negative Statistical combination.

2 0.0106

• 0.0051 of uncertaintiesm Burnup Uncertainty NA 0.0153 Axial Burnup Distribution NA

+ 0.0130 Total 0.9392' O.0106 0.9299 0.0051 Maximum Reactivity (k.)

0.9498 0.935 i;

")

For fuel tolerances, uncertainties in Region 2 assumed l

to be the same as those for Region 1.

m Square root of sum of squares.

i' 13 l

l

Table 2 Evaluation of the Minimum Burnup Requirements in Region 2 Initial Calculated Depletion Axial Burnup Limiting Enrichment k,,

Uncert. Ak Dist. Ak Burnup 2.0%

0.9232 0.0025 0.0 5,270")

2.5%

0.9197 0.0060 0.0 12,1000) 3.0%

0.9169 0.0088 0.0 18,310")

3.5%

0.9148 0.0109 0.0 24,240")

4.0%

0.9130 0.0127 0.0 29,810")

4.2%

0.9124 0.0133 0.0038 32,0000) 4.5%

0.9074 0.0143 0.0082 35,600 4.75 %

0.9016 0.0153 0.0130 38,890 5.0%

0.8958 0.0161 0.0180 42,290 m From initial analysis 14 i

Table 3 Evaluation of the Consequences of Fuel Handling Accidents Soluble Boron Case Maximum k,,

Required Region 1 5% E Assembly positioned 0.962 100 outside and adjacent to Region 1 5% E Assembly mis-loaded into 0.950 None an otherwise filled Region 1 Rack l

5% E Assembly misplaced 0.918 None within a Region 1 Checkerboard Region 2 5% E Assembly positioned 0.944 None outside and adjacent to Region 2 5% E Assembly mis-loaded into 0.980 180 an otherwise filled Region 2 Rack 5% E Assembly misplaced 1.000 425 within a Region 2 Checkerboard 15

Table 4 Design Basis Fuel Assembly Specifications FUEL ROD DATA Westinghouse Fuel Outside diameter, in.

0.440.

Cladding inside diameter, in.

0.384 Cladding material Zr-4 Pellet density, % T.D.

95 Stack density, g UOj, ec (* 0.20) 10.29 Pellet diameter, in.

0.376 Enrichment, wt % U-235 (t0.05) 4.75 - 5.0 ASSEMBLY DATA Fuel rod array 14x14 Number of fuel rods 176 Fuel rod pitch, in.

0.580 Number of control rod guide and 5

instrument thimbles Thimble O.D., in. (nominal) 1.115 Thimble I.D., in. (nominal) 1.035 16

.45000

)

40000

~

35000 2

- ACCEPTa8LE BURNUP DOMAIN m

30000 N

1 2

32 25000 f

da z

m 20000 m/

i m

S m

e

+

D' 15000

/ /

"""*'F 10000

/f BURNllP DO WAIN

'l

[

(Requires toglon 1 !foroge)

I 5000

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O' 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 INITIAL ENRICHMENT, wtz U-235 Fig. 1 LIMITING BURNUP CRITERIA FOR ACCEPTABLE STORAGE IN REGION 2 Notes:

(1)

Any fuel assembly (s 4.7% average U-235 enrichment), mechanically coupled with a full length CEA, may be located anywhere in Region 2.

(2)

Peripheral cells are those adjacent to the spent fuel pool wall or the cask laydown area.

17

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l l

1 l

l l

l 0.960 l

0.955 1

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0.950

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0.925

~

O.920

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0.915 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 i

Enrichment, wt z U-235 Fig. 2 Effect of Enrichment on Region 1 Reactivity 18

i 0.960 0.958 -(,

1 s

0.956 y

b 0.954 i

A T

.o O.952

=

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.O.950 J

m g

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c

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2 0.944 o

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j 0.942 s

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CA 5M03 5

h 3 0.936 N

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'c i

0.934 u

0.932 0

25 50 75 100 125 15$' '17$' 'dO$' 'd25 ppm Soluble Boron i

Fig. 3 Reactivity Effect of Soluble Boron (5s Enriched Fuel) 19

9.821" i 0.08" 3

(1.00" WG) 1 i

1 CELL l.D.

1 y

I'

8. 4 6"

• 0.0 4"

. _ - _..i.. _.. _ t _.

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a 1

TABLE OF CONTENTS

1.0 INTRODUCTION

AND

SUMMARY

A 2.0 NITAWL-KEN 05a BENCHMARK CALCULATIONS....

A-2 3.0 CASMO3 BENCHMARK CALCULATIONS A-4 4.0 MCNP BENCHMARK CALCULATIONS.............

A-5 5.0 INTERCOMPARISON CALCULATIONS A-5 5.1 Enrichment Effect A-5 5.2 Temperature Effect 5.3 Effect of Water-Gap Size......................

A-6 6.0 CLOSE-PACKED ARRAYS A-7

7.0 REFERENCES

A-8 i

1.0 INTRODUCTION

AND

SUMMARY

l 1

The objective of this benchmarking study is to verify the NITAWL-KENO 5a"J) methodology m and MCNPW for use in criticality safety (WORKER-NITAWL-KEN 05A), the CASMO3 code calculations of high density spent fuel storage racks. These 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 literature.

Three different cross-section libraries have been benchmarked for use with KEN 05A. Results -

of these calculations show that NITAWL-KENO 5a calculations consistently underpredict the critical eigenvalue for all three of the cross-section libraries. These libraries and their related calculational bias (for 95% probability at a 95% confidence level)* determined from critical experiments

• are the_following:

27 group Library (27GROUPNDF4) 0.0103 0.0018 27 group with fission products (27BURNUPLIB) 0.0095

• 0.0016

=

218 group library (218GROUPNDF4) 0.0128 0.0020 For CASMO-3, extensive benchmarking calculations of critical experiments have also been reportedm in the literature, giving a mean k,r,of 1.0004

• 0.0011 for 37 cases. With a K-factor of 2.14W for 95% probability at a 95% confidence level, and conservatively neglecting the small overprediction, the CASMO3 bias then becomes 0.0000 0.0024. CASMO3 and NITAWL-KENO 5a intercomparison calculations of infinite arrays of poisoned cell configurations

-(representative of typical spent fuel storage rack designs) show very good agreement, confirming that 0.0000 i 0.0024 is a reasonable bias and uncertainty 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 error.

A-1

~

MCNP (a Los Alamos continuous energy Monte-Carlo code) has also been benchmarked against critical experiments, giving a calculational bias of 0.0032 1 0.0020. Several cross-section libraries are available for MCNP and for the analyses reported here the library identified as "50.C" was used (for a few nuclides, the recommended library was identified as "55.C" or "56.C"). Calculations at Los Alamos have demonstrated that MCNP can readily handle complex geometries with consistent and accurate results. In general, however, MCNP tends to yield slightly lower reactivity values relative of KEN 05A.

The benchmark calculations reported here confirm that either the NITAWL-KENO 5a, MCNP or CASMO-3 calculations are acceptable for criticality analysis of high-density spent fuel storage racks, provided the appropriate calculational bias is used. Where possible, reference calculations for storage rack designs should be confirmed with an attemative analysis using either different -

cross-section libraries or different methods of analysis (or both), to provide independent verification. It should be noted, however, that CASMO-3 is not reliable when large water gaps

( > 2 or 3 inches) are present.

2.0 NITAWL-KENO 5a BENCHMARK CALCULATIONS Analysis of a series of Babcock & Wilcox critical experiments *, including some with absorber panels typical of a poisoned spent fuel rack, is summarized in Table 1, as calculated with NITAWL-KENO 5a using the three available SCALE cross-section libraries and the Nordheim resonance integral treatment in NITAWL. Dancoff factors for input to NITAWL were calculated with the Oak Ridge SUPERDAN routine (from the SCALEm system of codes). The mean for these calculations for each of the four cross-section sets are shown in Table 1 together with the standard deviation of the mean). The calculational bias, shown above and in Table 1, were calculated with a one-sided tolerance factor (2.523) corresponding to 95% probability at a 95%

confidence level? for the sixteen critical experiments analyzed.

A-2

_. _. _ _ _ _ _ _ y e'

I l

i Similar calculational deviations have been reported by ORNL* for some 54 critical experiments (mostly clean criticals without strong absorbers), obtaining a mean bias of 0.0100

• 0.0013 (95 %/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 NITAWL-KENO 5a calculational model for use in criticality analysis of high density spent fuel storage racks. No abnormal deviations in 1 rr4 with intra-assembly water gap, with absorber panel reactivity worth, with enrichment or with poison concentration were identified with either the 27 group or the 218-group SCALE library or with MCNP. CASMO was found to be unreliable for the larger water-gaps.

Additional benchmarking calculations were viso made for a series of French critical experiments

• at 4.75% enrichment and for several of the IsNWL 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 rr values showed a trend toward higher values with

]

e decreasing core size. In the absence of a significant enrichment effect (see Section 5.2 below),

this trend and the overprediction is attributed to a small inadequacy in NITAWL-KENO 5a in calculating neutron leakage from very small assemblies.

Similar results were observed for the BNWL series of critical experimentsuo, which are also small assemblies (although significantly larger than the French criticals). In this case (Table 2),

the calculated mean k,,, was 0.9959 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 NITAWL-KENO 5a 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-KENO 5a 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 overpred-iction of the reactivity.

A-3

6 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 radial array of storage cells,

. neglecting leakage effects.

CASMO3 has been extensively benchmarked against both mixed oxide and hot and cold critical experiments by Studsvik Energiteknikm. Reported analyses of 37 critical experiments indicate a mean k r, of 1.0004 0.0011 (10). To independently confirm the validity of CASMO3 (and o

to investigate any effect of enrichment), a series of intercomparison calculations were made (see Section 5) with CASMO3, NITAWL-KENO 5a and MCNP 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) show very good agreement and confirm the bias of 0.0000 i 0.0024 (95%/95%) for CASMO3.

A second series of CASMO3, MCNP, and KEN 05a intercomparison calculations consisting of five cases from the BAW critical experiments were analyzed for the central cell only. The calculated results, also shown in Table 3, indicate a mean difference within the 95% confidence f

i limit of the KENO 5a calculations. This lends further credence to the recommended bias for CASMO3.

Intercomparison between analytical methods is a technique endorsed by Reg. Guide 3.41,

" Validation of Calculational Methods for Nuclear Criticality Safety" A-4

4.0 MCNP BENCIIMARK CALCULATIONS 1

MCNP (Monte Carlo N-Particle) is a continuous energy Monte Carlo code with very flexible geometry capabilities. For these benchmark calculations, the recommended "50.C" library was

- used except for two nuclides whose cross-section libraries have been corrected - iron (55.C) and zirconium (56.C).

MCNP was benchmarked against the same set of critical experiments as KEN 05A and the comparison is shown in Table 1. These benchmark calculations gave a bias of 0.0032 10.0020.

Independent calculations confirmed that there is no significant dependence on enrichment or temperature (see Section 5).

5.0 INTERCOMPARISON CALCULATIONS 5.1 Enrichment Effect Calculations were made with CASMO-3, MCNP, and KEN 05A (27-Group) for various fuel enrichments in a representative high-density (poisoned) spent fuel storage rack cells. Results of these calculations, shown in Table 3 and illustrated in Figure 1, show very good agreement for the three independent methods of analysis. Smce three 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 enrichments involved in power reactor fuel.

A-5

,a 5.2 Temocrature Effect The WORKER routine was obtained from ORNL and is intended to interpolate the hydrogen scattering. matrices for temperature in order to correct for the deficiency noted in NRC Information Notice 91-66 (October 18,1991). Benchmark calculations were made against MCNP and CASMO3, based on the assumption that independent methods of analysis would not exhibit the same error. Results of these calculations, shown in Table 4 and in Figure 2, confirm that the slope with temperature obtained by CASMO3 and KEN 05a are essereially the same. This agreement establishes the validity of the WORKER routine, in conjunction with NITAWL-KENO 5a, in. calculating reactivities at temperatures between 20*C and 120*C. MCNP is comparable but does not incorporate the Doppler effect which accounts for the slight difference in slope.

The deficiency in the NITAWL hydrogen scattering matrix at temperatures above 20

• C does not appear except in the presence 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 should be accurate. In the calculations shown in Table 4 and in Figure 2, the absolute CASMO-3 reactivity values differ somewhat from the other calculations, but the trends with temperature are sufficiently in agreement to lend credibility to the WORKER routine over the temperature range from 20 C to 120*C.

5.3 Effect of Water-Gao Size Calculations were made for a fuel assembly suspended in water for the reactivity with various water-gap spacings between assemblies. Results, shown in Table 5, indicate good agreement between MCNP and KEN 05A. CASMO-3, however, showed significant deviations for 4 inch or larger water-gaps, depending upon the number of mesh intervals specified in the CASMO-3 calculations. For this reason, CASMO-3 is not considered reliable for water-gap spacing greater

. than about 2 to 3 inches, although differential calculations for small design variations (e.g.,

tolerances) would be acceptable.

A-6

l l

6.0 CLOSE-PACKED ARRAYS The BAW close-packed series of critical experiments *) intended to simulate consolidated fuel, l

were analyzed with NITAWL-KENO 5a (27-Group). Because there are so few cases rosilable I

for analysis, results of these analyses, shown in Table 6, suggest the possibility of a slightly larger bias than that for fuel with normal lattice spacings. Similar results were obtained by i

ORNL(").

Therefore, the maximum bias for close-packed lattices may be taken as 0.0155,

)

including uncertainty, which would very conservatively encompass all but one of the cases measured.

)

i j

i i

i A-7

I.0 '

' REFERENCES TO APPENDIX A-7 1.

Green, Lucious, Petrie, Ford, White, and Wright, "PSR-63/NITAWL-1 (code package)

NITAWL Modular Code System For Generating Coupled Multigroup Neutron-Gamma Libraries frorn ENDF/B", ORNL-TM-3706, Oak Ridge National Laboratory, Novem'oer.

1975.

2.

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

3.

A. Ahlin, M.'-Edenius, and H. Haggblom, "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 Bumup Program, Users Manual", Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986.

4.

J.F. Briesmeister, Ed., "MCNP - A General Monte Carlo N-Particle Transport Code, Version'4A.", Los Alamos National Laboratory, LA-12625-M (1993).

5.

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

6.

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

7.

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

Applications", Nuclear Science and Engineerine, 100, 342-351, (1988).

8.

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.

A-8 i

P References (Continued) 9.

J.C. Manaranche, et. al., " Dissolution and Storage Experiment with 4.75% U-235 Enriched UO Rods", Nuclear Technolocv, Vol. 50, pp 148, September 1980.

2 4

10.

A.M. Hathout, et. al., " Validation of Three Cross-section Libraries Used with the SCALE System for Criticality Analysis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981.

11.

S.R. Bierman, et. al., " Critical Separation between Sub-critical Clusters of 4.29 Wt. %

i 2"U Enriched UO Rods in Water with Fixed Neutron Poisons", Battelle Pacific 2

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

12.

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

]

(1981).

i 13.

R.M. Westfall, et al., " Assessment of Criticality Computational Software for the U.S.

'l Department of Energy Office of Civilian Radioactive Waste Management Applications",

Section 6, Fuel Consolidation Applications, ORNL/CSD/TM-247 (undated).

4 A-9

Table 1 Results of NITAWL-KENO 5a and MCNP Benchmark Calculations k,, 2 l a Expt Number -

27-Group

27. group Burn 218-group MCNP I

0.9922 2 0.0006 0.9919 0.0006 0.9886 : 0.0006 0.9918

• 0.0010 II 0.9917 e 0.0005 0.9924 2 0.0005 0.98 % 0.0005 0.9995

• 0.0010 Ill 0.9931 0.0005 0.9923 0.0005 0.9907 0.0005 0.9988

• 0.0010 IX 0.9915 : 0.0006 0.9912 2 0.0006 0.9884 e 0.0006 0.9881 0.0009 X

0.9903 2 0.0006 0.9915 2 0.0006 0.9871 2 0.0006 0.9951

• 0.0009 XI 0.9919 0.0005 0.9917 2 0.0005 0.9904 e 0.0005 0.9991 i 0.0009 XII 0.9915 0.0006 0.9929 i 0.0006 0.9898

• 0.0006

- 0.9883

• 0.0009 XIII 0.9945 0.0006 0.9951 0.0006 0.9904 0.0006 1.0004 0.0009 XIV 0.9902 = 0.0006 0.9910 0.0006 0.9885

• 0.0006 0.9974 0.0010 XV 0.9836 0.0006 0.9858

• 0.0006 0.9800

• 0.0006 0.9925 $ 0.0010 XVI 0.9863 z 0.0006 0.9P55 0.0006 0.9817 0.0006 0.9916 2 0.0010 XVII 0.9875 : 0.0006 0.9885 0.0005 0.9853 2 0.0006 0.9958 2 0.0009 XVIII 0.9880 z 0.0006 0.9892 0.0006 0.9857 e 0.0006 0.9985 e 0.0010 XIX 0.9882 2 0.0005

(..% 99 0.0005 0.9858 2 0.0006 0.9939 0.0009 XX 0.9885 2 0.0006 0.9887 2 0.0006 0.9859 2 0.3006 0.9981 z 0.0010 XXI 0.9862 e 0.0006 0.9909 2 0.0006 0.9878 2 0.0006 1.0006 i 0.0009 Mean"'

O.9897 2 0.0007 0.9905 0.0006 0.9872

• 0.0008 0.9968 : 0.0008 Blas*

0.0103

• 0.0018 0.0095 i 0.0016 0.0128 0.0020 0.0032

• 0.0020

• Standard Deviation of the Mean, calculated from the individual k,y values.

• Bias at the 95% Probability / 95% Confidence level.

A-10 m

e e

D

.o o

Table 2 Results of 27-Group (SCALE) NITAWL-KENO 5a Calculations of French and BNWL Critical Experiments i

1 French Experiment Separation Critical Calculated Distance, em Height, cm k,,

0 23.8 1.0302 z 0.0008 2.5 24.48 1.0278

• 0.0007 5.0 31.47 1.0168 0.0007 10.0 64.34 0.9998

• 0.0007 l

BNWL Experiments Calculated i

l Case Expt. No.

k,,

No Absorber 004/032 0.9942

• 0.0007 l

SS Plates (1.05 B) 009 0.9946

• 0.0007 SS Plates (1.62 B) 011 0.9979

• 0.0007 SS Plates (1.62 B) 012 0.9968

• 0.0007

[-

SS Plates 013 0.9956 0.0007 i'

SS Plates 014 0.9967 0.0007 Zr Plates 030 0.9955

• 0.0007

]

Mean 0.9959 0.0013 1

l i

i A-11 1

I l

e Table 3 i

Results of CASMO3, MCNP, AND NITAWL-KENO 5a Benchmark (Intercomparison) Calculaitions L. (Bias Corrected)

Enrichment

• Wt. % U-235 NITAWL-KENO 5a*

CASMO3 MCNP*

2.5 0.8371 0.0010 0.8386 0.8350 0.0010 3.0 0.8776

• 0.0010 0.8783 0.8730 i 0.0011 3.5 0.9082 2 0.0010 0.9097 0.9075 *0.0010 4.0 0.9370 2 0.0011 0.9352 0.9326 0.0011 4.5 0.9561 2 0.0011 0.9565 0.9557

• 0.0011 5.0 0.9747

• 0.0011 0.9746 0.9717 0.0012 Expt. No.*

XIII 1.1021

• 0.0009 1.1008 1.1030 0.0010 I

XIV 1.0997 3 0.0008 1.1011 1.1032 2 0.0008 XV 1.1086 2 0.0008 1.1087 1.1104 2 0.0009 XVII 1.1158 3 0.0007 1.1168 1.1177

• 0.0008 XIX 1.1215 0.0007 1.1237 1.1235 3 0.0008 Infinite array of assemblies typical of high-density spent fuel storage racks.

m

k. from NITAWL-KENO 5a corrected for bias (+0.0103).

MCNP calculation corrected for bias (+0.0032)

Central cell of BAW Critical Experiments (Ref. 6).

A-12 9

m

.O Table 4 Intercomparison of Calculations"' at Various Temperatures with MCNP, KEN 05A and CASMO3 MCNP j

Temperature (S(a 8) at 300 *K)

CASMO3 W-N-KENO 5a )c l

4* C 1.2320 2 0.0008 1.2276 1.2345

• 0.0014 i

1 17.5'C 1.2342 0.0008 1.2322 1.2328 0.0015 25'C 1.2346 0.0008 1.2347 1.2360 e 0.0013 j

50*C 1.2431 0.00180) 1.2432 1.2475

• 0.0014 i

75'C 1.2540 0.000'/')

1.2519 1.2569 2 0.0015' 120*C 1.2779 0.0010 1.2701 1.2746 2 0.0014

")' Bias Corrected

0) WORKER-NITAWL-KEN 05a Code Package.

0) Interpolated between S(a,B) at 27 'C and S(a,B) at 127 *C; Doppler effect not included.

1 A - 13 9

a i

Table 5 Intercomparison of Calculations"' at Various Water Gaps with MCNP, KENOSA and CASMO3 Interassembly Water Gap, in.

MCNP CASMO3*

W-N-KEN 05a*

1.0 1.4755

• 0.0007 1.467 - 1.468 1.4655

• 0.0010 2.0 '

!.3244

• 0.0008 1.321 - 1.326 1.3169 0.0012 4.0 1.0791

• 0.0010 1.088 - 1.102 1.0769

• 0.0014 9

6.0 0.9871

• 0.0011 0.983 - 1.008 0.9893

• 0.0015 8.0 0.9591 2 0.0011 0.916 - 0.986 0.% 22

• 0.0015 12.0 0.9469s 0.0013 0.844 - 0.979 0.9476

• 0.0016 m Bias Corrected

• Range due to different mesh intervals used.

m WORKER-NITAWL-KEN 05a Code Package.

4 A-14

.4

a. m..

m_.

I

?

a Table 6 KEN 05A (27-Group) Calculations for Close-Packed Critical Experiments Calc.

BAW Pin Module Boron Calculated i

No.

Expt.

Pitch Spacing Conc.

k,,

No.

cm em ppm i

KS01 2500 Square 1.792 1156 0.9891 2 0.0005 1.4097 KS02 2505 Square -

1.792 1068 0.9910 2 0.0005 1.4097 KS1 2485 Square 1.778 886

.0.9845 2 0.0005 Touching KS2-2491 Square 0.778 746 0.9849 e 0.0005 Touching KTl 2452 Triang.

1.86 435 0.9845 e 0.0006 Touching KTIA 2457 Triang.

1.86 335 0.9865 2 0.0006 i

Touching KT2 2464 Triang.

2.62 361 0.9827 0.0006 j

Touching i

KT3 2472 Triang.

3.39 121 1.0034 2 0.0006 i

Touching 1

A-15

= _.

e o

1.00 b

/

.J O.95

~

f

/

/

}

~

z

/, '

@ 0.90 b

~

/

KENO -5a MCNP

'~~

/

CASMO

--~~-

O.85

/

k

0. 80.,,,, ' d ' ' $, d ' ' j, f ' ' 4, g..

2.0 2.

FUEL ENRICHMENT, WTs U-235 Fig. 1 Intercomparison of CASMO-3, MCNP, and KENO-So for Fuel of Various Enrichments

i 4

n.

- y 1.28 -

n C;

i i

I

~

i i

1.27 e

n li'/

}

i i

1.26'

/

/

n l

-/,'

4 s

i

/-

p i

/

C 1.25 b

~f.f i

.x

~

1.24_

b

.f.

1.23 b

c 1.22 0

20 40 60 80 100 120 140 Temperature, Degrees C Fig. 2 COMPARISON OF TEMPERATURE DEPENDANCE FOR CASMO-3, MCNP, AND KENO-5A i

I

-