ML20101E996

Rev 0 to Callaway Plant Region 2 Spent Fuel Rack Criticality Analysis

ML20101E996
Person / Time
Site:	Callaway
Issue date:	05/31/1992
From:	UNION ELECTRIC CO.
To:
Shared Package
ML20101E987	List: ML20101E984 ML20101E990 ML20101E996
References
LFNF-92-02, LFNF-92-02-R00, LFNF-92-2, LFNF-92-2-R, NUDOCS 9206240213
Download: ML20101E996 (51)

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LFNF-92-02, REV. O CALIAWAY PLANT REGION 2 SPENT FUEL RACK CRITICALITY ANALYSIS I

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Nuclear Fuel Group Licensing and Fuels Department Union Electric Company St. Louis, MO I

May 1992 I

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STATEMENT OF DISCIAIMER Data, methods, conclusions, and other information contained in this report have been prepared solely for use by Union Electric Company (Union Electric), and nay not be appropriate for uses other than those described herein.

Union Electric therefore makes no claim or warranty whatsoever, express or implied, regarding the accuracy, usefulness, or applicability of information contained in this report.

In particular, UNION ELECiRIC MAKES NO WARRANTY OF MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE, NOR SilALL ANY WARRANTY BE DEEMED TO ARISE TIIROUGH COURSE OF DEALING OR USAGE OF TRADE, with respect to the contents of this document.

In no event shall Union Electric be liable, whether through contract, tort, warranty, or strict or absolute liability, for any damages resulting from the unauthorized use of information contained in this report.

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LFNP-92-02, REV. O CALLAWAY PIANT REGION 2 SPENT FUEL RACK CRITICALI'IY ANALYSIS I.

I May 1992 I

I Prepared By:

NXng S/26 /92 J.

M/ Mc,ose l

Engineer,, Nuclear Fuel r/br

/z Reviewed By:

f P.

G. Justis' Engineer, Nuclear Fuel I07!/2-I Reviewed By:

g[

Are

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j R.' ' J. Irwirf Supe..ising Engineer, Nuclear Fuel Approved By:

rf e-r A.

C.

Passwater Manager, Licensing and Fuels

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I ABSTRACT This report describes Union Electric Company's methodology and techniques for developing the limiting burnup versus enrichment curve for storage of spent fuel in Region 2 of the Callaway spent fuel racks.

Benchmarking of the applicable codes is also presented herein.

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l TADLE OF CONTENTS Section FlLqn i

1.0 Introduction 1

1.1 Purpose 1

1.2 General Methodology 1

1.3 Design Criteria 2

I 1.4 Similarity to Previously-Licensed Methods 2

2.0 Benchmark Calculations 3

2.1 CASMO-3 Benchmarking 3

I 2.2 NITAWL/ KENO-Va Benchmarking 5

2.3 Benchmarking Comparisons 6

3.0 Calculational Approach 22 3.1 General Description 22 3.2 Reactivity Calculations 23 I

3.2.1 OFA/V-5 Fuel 23 3.2.2 Standard Fuel 26 3.3 Equivalent Zero Burnup Enrichment Determination 28 3.4 KENO-Va Calculations 28 3.5 Burnup Curve Development 30 4.0 Accident Conditions 38 4.1 Fuel Element Adjacent to and Outside Rack 39 4.2 Fuel Element Placed in the Wrong Region 40 I

5.0 Summary and Conclusions 41 I

6.0 References 42 I

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r-I LIST OF TABLES U

Pacie 1

CASMO/GRPDQ Benchmark Critical Results 7

2 NITAWL/ KENO-Va Benchmark Critical Results 8

3 Fuel Parameters 32 3

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I LIST OF FIGURES Ficure Paqe 1

Yankee Core I Isotopics - U-235 Atom Percent 9

2 Yankee Core I Isotopics - U-236 Atom Percent 10 3

Yankee Core I Isotopics - U-238 Atom Percent 11 4

Yankee Core I Isotopics - Pu-239 Atom Percent 12 5

Yankee Core I Isotopics - Pu-240 Atom Percent 13 6

Yankee Ctre I Isotopics - Pu-241 Atom Percent 14 7

Yankee Core I Isotopics - Pu-242 Atom Percent 15 8

Yankee Core I Isotopics - Pu-239/U-238 Ratio 16 9

Yankee Core I Isotopics - Pu-239/Pu-240 Ratio 17 10 Yankee Core I Isotopics - Pu-240/Pu-241 Ratio 18 11 Yankee Core I Isotopics - Pu-241/Pd-242 Ratio 19 12 Yankee Core I Isotopics - All Uranium Isotopes 20 13 Yankee Core I Isotopics - All Plutoniunt Isotopes 21 14 Callaway Region 2 Geometry 33 15 Callaway Region 2 - Reactivity vs. Temperature 34 16 Callaway Region 2 - K-infinity vs. Burnup 35 OFAj'V-5 Fuel 17 Callaway Region 2 - K-infinity vs. Burnup 36 I

Standard Fuel 18 Callaway Region 2 - Burnup vs. Enrichment Curve 37 I

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INTRODUCTION I

1.1 PURPOSE The Callaway Region 2 spent fuel rack is an unpoisoned rack, which will be analyzed for storage of Westinghouse 17x17 Standard (STD), Optimized (OFA), and Vantage-5 (V-5) fuel assemblier.

The fuel assemblies are stocod in three of four storage locations in the fuel rack array.

The criticality analysis assumes non-borated water in the fuel pool, and utilizes burnup credit to ensure that K is < 0.95.

eff I

Beginning with Callaway Cycle 5, fuel enrichraents are being used which exceed the range of the previous analyses for storage of spent fuel in Region 2 of the spent fuel pool.

The previous burrup versus enrichment curve for OFA/V-5 fuel ranged up to an enrichment of 4.25 w/o, while cycle 5 utilized fuel enrichments up to 4.4 w/o.

The analysis given herein presents the overall methodology for extending the range of applicability to 5.0 w/o enrichment.

1.2 GENERAL METIlODOLOGY Two independent code packages were used for determining criticality cafety.

The overall methodology for determining the enrichment /burnup curve used the CASMO code for cross section generation and the GRPDQ code for reactivity I

determination.

The NITAWL/ KENO-Va code package was utilized for verifying the results of CASMO/GRPDC.

The methodologies for determining criticality safety have been verified by comparison with critical experiment data for configurations that impose a stringent test of the capability of the analytical methodologies.

These experiments are chosen to ensure that the method bias and uncertainty are conservative and, with a high level of confidence, applicable to the Callaway racks.

I 1.3 DESIGN' CRITERIA I

The results of the benchmarking and production runs are used to determine the burnup/ enrichment curve for both OFA/V-5 and STD fuel assemblies.

The burnup curve is dcveloped to ensure that there is a 95 percent probability at a 95 percent confidence level that the effective multiplication j

factor of the Region 2 fuel racks will be < 0.95 as recommended in Reference 1.

1.4 SIlillARITY TO PREVIOUSLY-LICENSED METIIODS L

Union Electric's criticality analysis methodology, as described in this report, is based on methods developed by HOLTEC International, which have been previously accepted by the USNRC. I

I 2.O BENCHMARKING CALCUIATIONS I

Two separate and '" dependent design methods were used to analyze the Callavay Region 2 spent fuel racks.

The first 3

method uses the transport theory CASMO-3 code to generate four group macroscopic cross section data for input into the 4

diffusion theory GRPDQ code.

GRPDQ is then utilized for c31culating k-infinity.

The second method uses the SCALE 4 code system, which includes the NITAWL program to provide cross section data, including self-shielded resonance cross sections, for input into the Monte Carlo theory KENO-Va program.

I controls the use of the codes described above Union Elect.

through firm adherence to procedures governed by Union Electric's Quality Assurance program.

'Ihese procedures address such subjects as preparation of calculations; software validation, verification, installation, and documentation; software development; and control of nuclear analysis activities.

2.1 CASHO-3 BENCHMARKING CASMO-3 is a multigroup two-dimensienal transport theory code used for burnup calculations on BWR and PWR assemblies or simple pin cells.

The nuclear data library contains microscopic cross sections in 40 energy groups covering

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I neutron energies from 0 to 10 Mov.

CASMO-3 generates equivalent transport theory macroscopic cross section data for input into GRPDQ.

GRPDQ is a modified version of PDQ-7 Version 2.

Two sets of critical experiments have been smected for analysis.

The first set is the Babcock & Wilcox (B & W)

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Critical Experiments which consist of low-enriched (2.46 w/o) UO fu 1 pins in a wuter-moderated lattice that 2

simulates close-packed LWR fuel storage configurations.

The critical experiments consist of nino LWR-type fuel assemblies grouped in a 3x3 array, using both spacing ano absorber materials to provide numerous critical configurations.

The second set of experiments are the Battc.lle Northwest Laboratory ( BNWL) Critical Experiments which utihze a higher U-235 enrichment (4.306 w/o) for simulr.cing LWR fuel storage configurations.

A total of 23 e'.periments were analyzed which included various spacings, enrichments, and neutron absorbing materials to adequa%ly demonstrate the accuracy of the methodology and code packages.

14 A summary of the CASMO/GRPDQ results for the 23 critical experiments analyzed is presented in Table 1.

The average lI calculated K for the 23 experiments is 0.9992 with a gff standard deviatlon of the mean of 0.0008 delta k.

Since the me asured average of the 23 criticals is 1.0000, the final 1

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I methodology bias to be applied to the CASMO/GRPDQ model is

+0.0008 i 0.0019 delta k, evaluated for a one sided tolerance limit for 95% probability at the 95% confidence level.

The 95%/95% one sided tolerance limit for 23 values is 2.329 To confirm the ability of CASMO-3 to properly perform I

decletion calculations and calculate isotopic inventories, ll another set of benchmark calculations was performed The Union Electric version of CASMO-3 was validated against the 8,

,10 Yankee Rowe Care I isotopic benchmarks These results are shown in Figures 1-13 and clearly show that CASMO-3 correctly performs the depletion /burnup calculations and also calculates the correct isotopic inventor 1ec.

2.2 NITAWL/ KENO-Va BENCRMARKING The same critical experiments used for CASMO/GRPDQ benchmarking were selected for benchmarking of the NITAWL/ KENO-Va methodology.

The 27 group SCALE cross

taction library (derived from the ENDF/B-IV data compilation and collapsed from the 218 group library) was chosen for this analysis since it was developed specifically for criticality safety analysis of more thermal systems.

The Nordheim integral resonance treatment is used to account for

• SCALE is an acronym for fitandardized Computer Analyses for Licensing Evaluation (NUREG/CR-0200) I

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

I the effects of the resonance absorption in U-238.

11ITAWL calculates the self-shielding of neutrons on the basis of a pin cell, accounting for the presence of other fuel in an assembly through the use of a Dancoff factor (evaluated with the ORNL SUPERDAll routine).

15 A summary of the NITAWL/ KENO-Va results for the 23 critical experiments analyzed is presented in Table 2.

The average calculated K f r the 23 experiments is 0.9918 eff with a standard deviation of the mean of 0.0007 delta k.

Since the measured average of the 23 criticals is 1.0000, the final methodology bias to be applied to the MITAWL/

KENO-Va model is +0.0082 1 0.0017 delta k, evaluated at the 95% probability, 95% confidence level (the 95%/95% one sided tolerance liLit for 23 values is 2.329

).

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2.3 BENCHMARK COMPARISC NS The results of the above benchmark calculations are consistent with the published benchmark results of 1

STUDSVIK and ORNL I

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TABLE 1 CASMO / GRPDQ BENCHMARK CRITICAL RESULTS C R'ITIC'AL L ENRICHMENT ABSORBER SOLUBLE 7

EXPERIMENT

' U-235 W/O '

MATERIAL BORON K-EFF' B&Wl 2.46 WATER 0

6.99386 I

B & W ll 2.46 WATER 1037 1.00093 8 & W 111 2.46 WATER 761 1.00364 B & W IX 2.46 WATER 0

1.00597 I

B&WX 2.46 WATER 143 1.00603 B & W XI 2.46 STAINLESS STEEL 514 1.00467 B & W Xil 2.46 STAINLESS STEEL 217 1.00458 B & W Xill 2.46 BORATED AL 15 1.00156 B & W XIV 2.46 BORATED AL 92 0.99783 B & W XV 2.46 BORATED AL 395 0.99284 B & W X'>l 2.46 BORATED AL 121 0.99489 B & W XVil 2.46 BORATED AL 487 0.99617 B & W XVill 2.46 BORATED AL 197 0.99710 B & W XIX 2.46 BORATED AL 634 0.99733 B & W XX 2.46 BORATED AL 320 0.99883 B & W XXI 2.46 BORATED AL 72 0.99945 I

BNWL9 4.306 BORATED SS 0

0.99798 BNWL 11 4.306 BORATED SS 0

0.99823 BNWL 12 4.306 BORATED SS 0

0.99943 BNWL 13 4.306 STAINLESS STEEL 0

0.99665 BNWL 14 4.306 STAINLESS STEEL 0

0.99984 BNWL 29 4.306 ZlRCALOY 0

0.99732 BNWL 32 4.306 WATER 0

0.99608 MEAN = 0.9992 SIGMA = 0.0008 BIAS = 0.0008 +/- 0.001i.

(95%/95%)

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TABLE 2 NITAWL / KENO-Va BENCHMARK CRITICAL RESULT 5 I-i CRITICAL ENRICHMENT -

ABSORBER SOLUBLE EXPERIMENT U-235 W/O MATERIAL BORON K-EFF B&Wl 2.46 WATER 0

0.98808 +/- 0.00308 B & W 11 2.46 WATER 1037 0.99600 +/- 0.00305 f

B & W 111 2.46 WATER 764 0.99540 +/- 0.00260 B & W IX 2.46 WATER 0

0.98842 +/- 0.00300 B&W X 2.46 WATER 143 0.99434 +/- 0.00261 B & W XI 2.46 STAINLESS STEEL 514 0.99024 +/- 0.00279 8 & W Xil 2.46 STAINLESS STEEL 217 0.99297 +/- 0.00265 B & W Xill 2.46 BORATED AL

'5 0.99905 +/- 0.00334 B & W XIV 2.46 BORATED AL 92 0.98768 +/- 0.00325 B & W XV 2.46 BORATED AL 395 0.98875 +/- 0.00300 E & W XVI 2.46 BORATED AL 121 0.98639 +/- 0.00315 B & W XVI 2.46 BORATED AL 487 0.98939 +/- 0.00241 B & W XVI 2.46 BORATED AL 197 0.98868 +/- 0.00297 8 & W XIX 2.46 BORATED AL 634 0.98905 +/- 0.00264 B & W XX 2.46 BORATED AL 320 0.98877 +/- 0.00289 B & W XXI 2.46 BORATED AL 72 0.98948 +/- 0.00290 BNWL9 4.306 BORATED SS 0-0.99055 +/- 0.00308 BNWL 11 4.306 BORATED SS 0

0.99633 +/- 0.00297 BNWL 12 4.30S BORATED SS 0

0.99622 +/- 0.00328 I

BNWL 13' 4.306 STAINLESS STEEL 0

0.99234 +/- 0.00288 BNWL 14 4.306 STAINLESS STEEL 0

0.99419 +/- 0.00362 BNWL 29 4.306 ZlRCALOY 0

0.99324 +/- 0.00303 BNWL 32 4.306 WATER 0

0.99501 +/- 0.00325 MEAN = 0.9913 SIGMA = 0.0007 B!AS = 0.0082 +/- 0.0017 (95%/95%)

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I 3.0 CALCULATIONAL APPROACH I

3.1 GENERAL DESCRIPTION I

As previously discussed, two separate and independent code packages were utilized in analyzing the Callaway spent fuel racks.

The CASMO/GRPDQ code set was used for developing the limiting spent fuel burnup/ enrichment curve for acceptance into Region 2 of the spent fuel pool.

Due to the design of the Callaway spent fuel racks, which contain a flux trap water cell, an infinite lattice CASMO model cannot be utilized.

The GRPDQ code, a modified version of PDQ-7, was used for determining tne final multiplication factor predictions for the spent fuel racks.

The calculations were 3

performed in four energy groups using macroscopic cross section data generated in CASMO.

The geometry modeled in CASMO took into-account all of the details of the fuel assemblies and the fuel rack storage cells.

The geometry used for the CASMO calculations is a basic cell representing a fuel assembly in a rack cell.

For the CASMO cases to model the water cell, the same basic geometry was used but a reflector was added.

The reference model geometry used for the GRPDQ calculations was a repeating array of four stainless steel boxes, three of which contain fuel assemblies and the fourth which serves as a flux trap.

The specific geometry and nominal dimensions of the reference model are shown in Figure 14. _ _ _ _ _ _ _ _

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The CASMO/GRPDQ calculational approach was to use the reference model to calculate the reactivity of an infinite array of uniform spent fuel racks and to account for any deviations of the actual spent fuel rack array as uncertainties on the calculated reactivity of the basic cell.

Calculational biases, manufacturing tolerances, and uncertainties were evaluated in terms of the reactivity changes to the reference model.

The NITAWL/ KENO-Va code set was used for determining the reference k-infinity for developing the limiting burnup versus enrichlaent curves.

An equivalent zero burnup enrichment, which yields the sane reactivity as a burned assembly, was determined using CASMO/GRPDQ for input into NITAWL/ KENO-Va.

I 3.2 REACTIVITY CAICULATIONS 3.2.1 OFA/V-5 Fuel The V-5 fuel design parameters, Table 3, are identical to OFA fuel for criticality considerations.

Thus, only the V-5 fuel design will be evaluated.

The reference OFA/V-5 fuel assembly cnosen for evaluation was a typical Callaway ascembly with an initial enrichment of 4.0 w/o at a burnup of 33,000 MWD /MTU.

Initial CASMO/GRPDQ calculations were performed to determine the maximum reactivity within the I

I operating temperature range of the spent fuel pool, i.e.

68*F to 248*F.

The reactivity change due to temperature is shown in Figure 15 and shows a peak in reactivity at 90*F; thus the principle calculations were performed at a temperature of 90*F.

I There were a number of tolerances and uncertainties which I

result in reactivity deviations which must be considered in the criticality analysis.

The reactivity effects of all such positive deviations were then combined statistically in accordance with Reference 1 to determine a single reactivity uncertainty which is added to the calculated reference cell I

multiplication factor (including biases) to determine the final spent fuel rack maximum multiplication factor.

From the GRPDQ runs, the reactivity uncertainties which result in a positive deviation are as follows (in units of delta k):

I TOLERANCE OR UNCERTAINTY REACTIVITY DEVIATION Stainless Steel Box Spacing Decrease 0.00148 I

Stainless Steel Thickness Decrease 0.00167 Fuel Density Increase to 96% of Theoretical 0.00121 Fuel Enrichment Increase of 0.05% w/o 0.00421 I

The final uncertainty to be considered for Region 2 is associated with the calculated reduction in fuel assembly reactivity due to the depletion of heavy metals and the accumulation of fission products as a function of fuel assembly exposure.

The change in reactivity due to burnup 14 is typically 0. 01 delta k / MWD /kgU Calculations of

4

~

reactor lifetimes, i.e. assembly burnups and isotopics, using the same analytical methods as used in this analysis demonstrate an accuracy of better than 5 percent.

Therefore, assuming a 5% uncertainty in the reactivity change due to burnup, and using the base case 4.0 w/o assenbly at 33,000 MWD /MTU, the resulting uncertainty can be conservatively estimated to be 0.0165 delta k.

I The effects of the axial burnup distribution on reactivity must also be evaluated.

As burnup on an assembly increases, the central regions of the assembly become higher burned than the upper and lower ends.

However, at high burnups, the more reactive fuel near the ends of the assembly is in a region of low reactivity due to neutron leakage.

Studies by 16 HOLTEC have provided generic analytical results of the l

axial burnup effect based upon calculated and measured axial burnur distributions.

For the range of burnups considered I

in this analysis, the reactivity effects of axial burnup are minor and generally negative.

As noted above, Reference 1 allows the reactivity deviations due to tolerances and uncertainties to be combined statistically, i.e.

an RMS average.

When this is done, the total reactivity deviation to be added to the reference cell to account for all of the tolerances and uncertainties is 0.0172 delta k.

I The following equation was used to determine the norainal k-infinity for the Region 2 spent fuel racks:

K = 0.95 - B, - [((K)

+ IK ) }

U B

where:

B

= the UE CASMO-3 method bias from the critical experiments

= 0.0008 K

= the reactivity uncertainty due to tolerances g

and uncertainties

= 0.0172 K

= the calculational uncertainty due to the method B = 0.0019 Thus, the resulting k-infinity value is :

K = 0.95 - 0.0008 - [((0.0172)2+ (0.0019)2)

= 0.9319 To provide an additional conservatism, the analysis was performed with an initial k-infinity of 0.9250 for choosing the equivalent zero burnup enrichment for verification with NITAWL/ KENO-Va.

The results of the NITAWL/ KENO-Va runs were then used for the final determination of the k-infinity value used for the limiting burnup/ enrichment curve.

3.2.2 Standard Fuel The same methodology detailed above was utilized for the Standard (STD) fuel assemblies.

While the method bias and uncertainty remain the same, the manufacturing tolerances I I l

vary from the OFA/V-5 case.

From the GRPDQ runs, which are based on a 3.1 w/o STD assembly at a burnup of 23,000 MWD /MTU, the results are as follows:

I TOLERANCE OR UNCERTAINTY

}1EACTIVITY DEVI ATION Stainless Steel Box Spacing Decrease 0.00180 Stainless Steel Thickness Decrease 0.00209 Fuel Density Increase to 96% of Theoretical 0.00241 I

Fuel Enrichment Increase of 0.05 w/o 0.00513 Since the uncertainty due to fuel burnup for the OFA/V-5 case was considered to be conservative, it was applied to the STD fuel analysis, and thus, using the previous formula:

I

[((K)

+ (K } )

K = 0.95 - B B

U I

whern:

B

= the UE CASMO method bias from the critical m

experiments I

= 0.0008 K

= the reactivity deviation due to tolerances and U

I uncertainties g

= 0.0177 K

= the calculational uncertainty due to the method I

g = 0.0019 The resulting k-infinity value is :

K = 0.95 - 0.0008 - [((0.0177)

I

+ (0.0019) )

= 0.9314 As for the OFA/V-S fuel case, to allow for additional conservatism, a k-infinity of 0.9250 was chosen for the analysis.

I I

I 3.3 EQUIVALENT-ZERO BURNUP ENRICHMENT DETERMINATION To utilize the KENO-Va program to verify the CASMO/GRPDQ results, an equivalent zero burnup enrichment must be determined which yields the same reactivity as a burned assembly, since KENO-Va does not perform depletion calculations.

Several values of enrichment were run at zero

-I burnup using CASMO/GRPDQ, and the results interpolated to define the enrichment that yields a k-infinity of 0.925.

The following are the I activity values for several enrichments of OFA/V-5 fuel:

B ENRICHMENT K-INFINITY I

1.3 w/o 0.8554 1.5 w/o 0.9027 1.7 w/o 0.9431 1.9 w/o 0.9774 Using the above data, an enrichment of 1.61 w/o was used as input for the initial KENO-Va calculations.

I 3.4 KENO-Va CALCULATIONS I

The results of the CASMO/GRPDQ equivalent zero burnup enrichment runs were utilized as input for the NITAWL/ KENO-Va codes.

The NITAWL program was used to assemble cross sections from the master 27 group SCALE library into the proper format and to evaluate resonance l

I I

shielding, primarily for U-238.

The Dancoff factors required for input into the NITAWu program were calculated using the ORNL SUPERDAN routine.

I Usinc; the methodology delineated previously, the maximum multiplication factor is determined by combining the results of the critical experiments, with the offects of the manufacturing tolerances and uncertainties, evaluated using CASMO/GRPDQ.

Since the individual manufacturing uncertainties (typically i 0.002 delta k) are within the bounds of the uncertainties in the KENO-Va calculations, typically i 0.003 delta k, the manufacturing uncertainties must be obtained from CASMO/GRPDQ.

I An additional uncertainty term to be included is the KENO-Va run str.tistics, which is 0.0010 delta k, evaluated at the one-sided tolerance limit for a 95% probability at the 95%

I-confidence level.

The KENO-Va run utilized 2000 generations, thus the 95%/95% one-sided tolerance limit is 1.703.

Therefore, using the following equatien, the maximuu r;-effective is as follows:

K, - 0.95 - B

- [((K)

+

+ (K ) }

g (g)

B

' I where:

K

= the ma.gimum k-effective m

B

= the UE KENO-Va method bias from the critical m

experiments

= 0.0082 g

- 2e -

I K

= Ge MOWa nn stadsdcs W/M) g = 0.0010 K

= the reactivity deviation due to tolerances and U

uncertainties

= 0.0172 K

= the calculational uncertainty due to the method B = 0.0017 I

Using the above inputs, the maximum allowable calculated k-effective from KENO-Va for an OFA/V-5 assembly is :

K = 0.95 - 0.0082 --

((0.0010)

+ (0,0172)

+ (0.0017) )

= 0.9245 Using the KENO-Va run results and the above k-effective, the maximum enrichment at zero burnup to maintain k $ 0.95 is calculated to be 1.59 w/o.

This results in a maximum k-effective of 0.9480, including biases and uncertainties.

Referring to the equivalent zero burnup enrichment data in Section 3.3, and using the maximum enrichment of 1.59 w/o, the k-infinity for determining the burnup versus enrichment curvas using CASMO/GRIDQ is 0.921.

I 3.5 BURNUP CURVE DEVELOPMENT The limiting burnup versus enrichment curves are determined B

from the CASMO/GRPDQ data.

For OFA/V-5 fuel, the values of k-infinity as a function of initial enrichment and burnup are presented graphically in Figure 16.

The curves are linear and were interpolated to find the burnup which resultc in a k-infinity of 0.921.

For OFA/V-5 fuel, the l

following are the limiting burnups:

l I

g.

OFA/V-5 FUEL INITIAL ENRICHNENT EXPOSURE 3.0 w/o 19,470 MWD /MTU 3.5 w/o 25,072 MWD /MTU 4.0 w/o 30,587 MWD /MTU 4.5 w/o 35,649 MWD /MTU 5.0 W/o 40,747 MWD /MTU I

For standard fuel, the values of k-infinity as a function of initial enrichment and burnup are presented graphica.tly in Figure 17.

The curves are interpolated to find the burnup which results in a k-infinity of 0.921.

For STD fuel, the 2:

following are the limiting burnups:

STANDARD FUEL INITIAL Eb?ICHMENT EXPOSURE 2.1 w/o 3,429 MWD /MTU

- $j 2.6 w/o 15,854 MWD /MTU d

3.1 w/o 21,684 MWD /MTU The limiting burnup data for both OFA/V-5 and STD fuel is presented in graphical form in Figure 18 I

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-____ _ _=_ _ _ _ _ _ _ _ _ _ _ _ _ _

I TABLE 3 I

FUEL PARAMETERS FOR CRITICALITY ANALYSIS I

PARAMETER 17 X 17 OFAN-5 17 X 17 STANDARD I

ASSEMBLY ASSEMBLY Number of Fuel Rods 264 264 per Assembly Cladding O.D. (inch) 0.360 0.374 ClarMing Thickness (inch) 0.0225 0.0225 Fuel Pellet O.D. (inch) 0.3088 0.3225 Fuel Pellet Density 0.9525 0.94867 5

Fuel Pellet Dishing &

0.9925 0.9880 Chamfering Factor I

hod Pitch (inch) 0.496 0.496

'4 umber of Guide Tubes 24 24 Guide Tube O.D. (inch) 0.474 0.482 I

Guido Tube Thickness (inch) 0.016 0.016 Number of Instrument Tubes 1

1 Instrument Tube O.D. (inch) 0.474 0.482

'I Instrument Tube Thickness (inch) 0.016 0.016 I.

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~

I..

l FIGURE 14 CALLAWAY REGION 2 GEOMETRY I

I I

AN'

\\

I N xx xxx h sN b '

Ns 9+

I M*

~h b

I k 9.236 9.236d I

K A

4.216-FUEL -

\\

.282-WATER

\\

.28 -WA I-x

~

^

4.216-FUEL -

J x

4.216-FUELd ll h

,y 4.4 9 8-WATER d.282-WATER W 8 m.

g)

. I

- l FIGURE 15 CALLAWAY REGION 2 l

REACTIVITY VS. TEMPERATURE g

0. 9 o.

1 I

I O.902 -

I

' I g

0.90,-

-b I

E i*

g 0.900-E I

s B

- I 0.898-i i

i i

i i

i r

0 50 100 150 200 250 300 TEMPERATURE (DEGREES F)

-u-

.l l

noune 1e CALLAWAY REGION 2 K-INFINITY VS. BURNUP - OFA/V-5 FUEL l

o.n -

' 4.0 W/0 k

.\\

5.0 W/0.

O.96-I 0.95-a4 -

I 3.0 W/0 3.5 W/O l

0.93-e I

E E 0.92-T

{

0.91-f

\\

ue.

l l

u9-I

\\

\\

us.

I 0.571-i

',y.

/

35-ABOVE THE CURVE IS ACCEPTABLE k/

I

"~

s I

sN 25-Q I

E

/

a.

/

/

2 20 4-y

' STANDARD FUEL- --

/'

I h

[

/

~

)<_

/

/

BELOW THE CURVE IS NOT ACCEPTABLE

/

10-

/-

I.

I

}:

I 0-i ri i

i i

i.3 2.0 2'

u a.5 4.0 4.5

0 ENRICHMENT, W/O U-235 I I

I 4.0 ACCIDENT CONDITIONS I

To ensure the safety of the spent fuel racks, an evaluation of the reactivity consequences of abnormal / accident conditions must be performed.

These conditions are as follows19:

(1) Fuel assembly positioned eccentrically in the cells I

(2) Fuel element located outside and adjacent to the rack (3) Fuel element dropped on top of the rack (4) Fuel element placed in the wrong region Items 1 & 3 will not result in an increase in k-infinity.

Studies of asymmetric positioning performed by Westinghouse 12 for similar rack configurations have shown that

- I symmetrically positioned fuel assemblies yield equal or conservative results.

For the case of the fuel assembly dropped on the racks, the dropped assembly is separated from the active fuel height of the assemblies in the rack by more than 21 inches of water.

The distance from the bottom of the rack to the top of the lead-in guides is 169.05", and the distance from the bottom of the bottom nozzle to the top of the active fuel is 147.499", resulting in a separation of

>21".

Since 30 cm of water (~12") is considered infinite reflection, the separation precludes neutron interaction between the assemblies.

I Items 2 & 4 are postulated conditions which would result in an increase in reactivity.

For these conditions, the double contingency principle of ANSI N16.1-1975 can be applied;

I lg l g I

I-thus the presence of borated water can be assumed as a realistic initial condition.

The typical boron concentration of the Callaway spent fuel pool is 2000 ppm boron.

The above accident conditiGns were thus analyzed using a soluble boron concentration of 2000 ppm.

4.1 FUEL ELEMENT ADJACENT TO AND OLTSIDE RACK l

I-This case assumes that an assembly is accidentally placed outside of, but adjacent to, the fuel storagr racks.

As stated above, this accident condition allows for analysis with the presence of soluble boron.

The assemblies in the

-I.

rack were assumed to be the typical 4.0 w/o V-5 assembly in a3 out of 4 condition, at a burnup of 30,587 MWD /MTU.

The assembly outside of the racks was assumed to be a fresh fuel assembly of the maximum reactivity, which was datermined from the Callaway Region 1 analysis as a 3.85 w/o V-5 assembly with no Integral Fuel Burnable Absorbers (IFBAs).

(NOTE - enrichments above 3.85 w/o are permitted in the Callaway spent fuel pool provided the assembly has the proper number of IFBA rods to assure a k-infinity equivalent to the 3.85 w/o assembly).

The results of this accident show that with a boron

,I concentration of 2000 ppm, the k-infinity of the spent fuel I

racks is 0.7723, including biases and uncertainties.

l l,I

l-

=

4.2 FUEL ELEMENT PLACED IN WRONG REGION This case assumes that an assembly is accidentally placed in the wrong region of the spent fuel racks.

The most limiting case for this scenario is placing a fresh fuel assembly of the maximum reactivity in Fegion 2.

As in the previous accident case, a 3.85 w/o V-5 assembly was used, and is placed in the middle of an 8 X 8 array of 4.0 V-5 fuel at a burnup of 30,587 MWD /MTU.

The results of th2 s accident show that with a boron concentration of luGO ppm, the K-infinity of the spent fuel racks is 0.8027, including biases and uncertainties.

I I

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I I

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I I

5.0

SUMMARY

AND CONCLUSIONS I

The NITAWL/ KENO-Va code package was utilized to determine the final k-infinity value of 0.921 used for developing the limiting burnup versus enrichment curves.

Using KENO-Va, the maximum calculated k-effective for the spent fuel racks is 0.9480, including biases and uncertainties, ensuring that the final k-effective of the Callaway Region 2 spent fuel racks is < 0.95.

The burnup versus enrichment curves for both OFA/V-5 and Standard fuel were developed using data from the CASMO/GRPDQ codes, and include the methodology bias and the manufacturing tolerancas and uncertainties.

I Results of the critical experiment benchmarks and the comparisons to measured isotopic results, demonstrate that Union Electric's methods for performing criticality analyses are both appropriate and valid.

The results are consistent I

with previous analyses performed for the Callaway spent fuel racks.

Furthermore, Union Electric's criticality analysis methods are similar to methods previously accepted by the NRC.

Therefore, in view of the demonstrated validity of the methods described herein, Union Electric concludes that the criticality analysis for the Callaway Region 2 is acceptable.

LI I._

i I

6.O REFERENCES 1.

OT Position for Review and Acceptance of Spent Fuel Storage I

and Handling Applications, Nuclear Regulatory Commission Letter to All-Power Reactor Licensees, from B.

K.

Grimes, April 14, 1978 2.

Licensing Report for Reracking Indian Point 2 Spent Fuel Pool, HI-89327, HOLTEC International, June 1989 3.

CASMO-3, A Fuel Assembly Burnup Program, User's Manual, Version 4.4, Studsvik/NFA-89/3 4.

GRPDQ User's Manual, Version 1.01, Revision 3, January 1988 5.

SCALE 4, A Modular Code System for Performing Standardized I

Computer Analyses for Licensing Evaluation, Oak Ridge National Laboratory I

6.

BAW-1484-7, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, M.

N.

Baldwin, et al, July 1979 7.

NUREG/CR-0073, Critical Separation Between Subcritical Clusters of 4.29 wt% U235 Enriched UO2 Rods in Water with Fixed Neutron Poisons, S.

R.

Bierman, et al, August 1979 sI 8.

WCAP-6071, Experimentally Determined Burnup and Spent Fuel Composition of Yankee Core I, July 1965 9.

WCAP-6068, Evaluation of Mass Spectrometric and Radiochemical Analysis of Yankee Core I Spent Fuel, March 1966 I

10.

WCAP-6086, Supplementary Report On Evaluation of Mass Spectrometric and Radiochem ~T1 Analysis of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium through Curium, August 1969 11.

NFDC 91-016, CASMO Benchmark Against Yankee Rowe Core I I

Isotopics 12.

Criticality Analysis of Callaway Spent Fuel Racks with IFBA Fuel, November 1989, M.

W.

Fecteau, et al 13.

National Bureau of Standards Handbook 91, Experimental Statistics 14.

NFDC 91-027, Callaway Region 2 Spent Fuel Rack Analysis -

CASMO/GRPDQ Methodology 15.

NFDC 92-002, Callaway Region 2 Spent Fuel Rack Analysis -

KENO-Va Methodology I

I 16.

S.

E. Turner, "An !incertainty Analysis - Axial Burnup Distribution Effccts, Proceedinas of a Workshon on the Uc2 of Burnun Credit in Spent Fuel Transport Casks, Sandia I, ~-

Report SAND 89-0018, October 1989 r-17.

M. Edenius and A. Ahlin, "CASMO-3: New Features, Benchmarking, and Advanced Applications", Nuclear Science and Encineerina, 100, 342-351, (1988),

12.

E W. Westfall and J. H. Knight, " SCALE System Cross-Section validation with Shipping Cask Critical Experiments", b)){E Transactions, Vol. 33, p. 368, November 1979 19.

Callaway Plant Final Safoty Analysis Report I

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- 43 1.

. - -