ML20147E825

Spent Fuel Storage Rack Criticality Analyses for Wolf Creek Generating Station

ML20147E825
Person / Time
Site:	Wolf Creek
Issue date:	12/31/1987
From:	Marsico P PLG, INC. (FORMERLY PICKARD, LOWE & GARRICK, INC.)
To:
Shared Package
ML20147E795	List: ML20147E788 ML20147E810 ML20147E825 ML20147E836
References
PLG-0589, PLG-589, NUDOCS 8803070166
Download: ML20147E825 (62)

Text

.,

4 PLG-0589 -

SPENT FUEL STORAGE RACK CRITICALITY ANALYSIS FOR THE WOLF CREEK GENERATING STATION t

Prepared by >

l PETER J. MARSICO i ,

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Prepared for WOLF CREEK NUCLEAR OPERATING CORPORATION  ;

Wichita, Kansas '

December 1987 ,

I 8803070166 800226 ,

PDR ADOCK 05000482 P DCD Pickard,Lowe andGarrick,Inc.

Engineers AppliedScientists e Management Consultants Newport Beach, CA . Washington, DC l

TABLE OF CONTENTS Section

- Pggg 9.1A.1 THE MAXIMUM DENSITY RACK (MOR) OESIGN CONCEPT 9.1A-1 9.1A.1.1 Introduction 9.1A-1 9.1A.1.2 Design Description 9.1A-2 9.1A.2 CRITICALITY ANALYSIS FOR THE REGION 1 SPENT FUEL STORAGE RACKS 9.1A-3 9.1A.2.1 Analytical Technique 9.1A-3 9.1A.2.2 Evaluation of Criticality Safety for Region 1 9.1A-7 9.1A.2.3 Tolerances and Uncertainties for Region 1 9.1A-8 9.1A.2.4 Accident Analysis 9.1A-10 9.1A.2.5 Analysis Conservatisms

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9.1A-10 9.1A.3 CRITICALITY ANALYSIS FOR THE REGION 2 SPENT FUEL STORAGE RACKS 9.1A-11 9.1A.3.1 Analytical Technique 9.1A-11 9.1A.3.2 Calculational Approach 9.1A-13 9.1A.3.3 Manufacturing and Thernal Considerations 9.1A-13 v.1A.3.4 Design Conservatisms 9.1A-14 9.1A.3.5 Accident Analysis 9.1A-14 9.1A.3.6 Required Burnup as a Function of Initial Enrichment for Region 2 Spent Fuel 9.1A-15 REFERENCES 9.1A-11 e

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LIST OF TABLES LLkLt LLLL 9.1A-1 Criticality Derign Criteria 9.1A-20 9.1A-2 Summary of LEOPA90 Results for Measured Criticals 9.1A-21 9.1A-3 Westinghouse UO2 Zr-4 Clad Cylindrical Core Critical Experiments 9.1A-22 9.1A-4 Battelle Critica15 9.1A-23 9.1A-5 Fuel Assembly Technical Infonnation for Wolf Creek 9.1A-24 9.1A-6 Summary of Perturbations to the Multiplicatfon Factor of the Basic Cell for Region 1 9.1A-25 9.1A-7 Saxton Pu0 2 -002 Critical Experiments 9.1A-26 9.1A-8 ESADA Pu0 2 -UO2 Critical Experiments 9.1A-27 9.1A-9 Summary of Predictions for kert in Criticality Experiments 9.1A-28 9.1 A-10 Summary of Reactivity Perturbations to the Multiplication 9.1A-29 Factor of the Basic Cell 9.1A-11 Region 2 k. as a Function of Initial Enrichment and 9.1A-30 Burnup 9.1 A-12 Region 2 Minimum Burnup as a Function of Enrichment to Obtain 9.1A-31 a k. of 0.915 Prior to the Addition of Tolerances and Uncertainties 111

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LIST OF FIGURES Fioure Elit 9.1A-1 Comparison of Region 1 and Region 2 Fuel Storage Arrangement 9.1A-32 9.1A-2 P0Q Calculational Model for the Wolf Creek Region 1 Spent 9.1A-33 Fuel Racks 9.1A-3 Wolf Creek Region 1 k. versus Fuel Assembly Initial Enrichment 9.1A-34 9.1 A-4 Wolf Creek Region 1 k. versus Pool Termperature 9.1A-35 9.1A-5 Wolf Creek Region 1 k. versus Water Density 9.1A-36 9.1A-6 Wolf Creek Region 1 k. versus Rack Cell Pitch 9.1A-37 9.1A-7 Wolf Creek Region 1 k. versus Stainless Steel Wall 9.1A-38 Thickness 9.1A-8 Net Destruction of U-235 versus Burnup in the Yankee Asymptotic Neutron Spectrum 9.1A-39 9.1A-9 Specific Production of U-235 versus Burnup in the Yankee Asymptotic Neutron Spectrum 9.1A-40 9.1A-10 Net Destruction of U-238 versus Burnup in the Yankee Asymptotic Neutron Spectrum 9.1A-41 9.1A-11 Specific Production of Pu-239 versus Burnup in the '

Yankee Asymptotic Neutron Spectrum 9.1A-42 9.1 A-12 Specific Production of Pu-240 versus Burnup in the l Yankee Asymptotic Neutron Spectrum 9.1A-43 9.1A-13 Specific Production of Pu-241 versus Burnup in'the Yankee Asymptotic Neutron Spectrum 9.1A-44 i

i 9.1 A-14 Specific Production of Pu-242 versus Burnup in the Yankee Asyniptotic Neutron Spectrum 9.1A-45 9.1 A-15 Specific Production of Total Pu and Fissile Pu versus i Burnup in the Yankee Asymptotic Neutron Spectrum 9.1A-46 9.1 A-16 Atom Percent of Total U versus Exposure 9.1A-47 9.1 A-17 Pu-239/U-238 Atom Ratio versus Exposure 9.1A-48 9.t A-18 Atom Percent of Total Pu versus Exposure 9.1A-49 l

l 9.1 A-19 2200 m/s Absorption Cross Sections for 4.50 w/o Fuel at i

45,000 MWO/MTU as a Function of Time After Shutdown 9.1A-50 iv l

s. + . ~ ~ ~ . . . . -

LIST OF FIGURES (continued)

Figure y ,

9.1A-20 P0Q Calculatior.21 Model for the Wolf Creek Region 2 Spent Fuel Racks 9.1A-51 9.1A-21 Wolf Creek of Region 2 k. versus Rack Cell Pitch 9.1A-52 9.1A-22 Wolf Creek of Region 2 k. versus Stainless Steel Wall Thickness 9.1A-53 9.1A-23 Wolf Creek of Region 2 k. versus of Pool Temperature 9.1A-54 9.1A-24 Wolf Creek of Region 2 k. versus of Water Den.sity 9.1A-55 ,

9.1A-25 Wolf Creek Region 2 k. as a Function of Fuel Assembly Burnup for Various Initial Enrichments 9.1A-56 9.1A-26 1631f Creek Minimum Rdquired Fuel Assembly BurnJp as a Function of Initial Enrichment for Storage in Region 2 9.1A-57 L

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APPENDIX 9.lA ' SPENT FUEL STORAGE RACK ANALYSIS 9.1A.1 THE MAXIMUM DENSITY RACK (MDR) DESIGN CONCEPT 9.1A.1.1 Introduction Historically, spent fuel rack designs have been based on conservative assumptions that could be easily accommodated since it was not planned to store large numbers of high exposure spent fuel assemblies on-site.

Previously it was anticipated that only small amounts of high exposure -

fuel assemblies (1/4 to 1/2 of a full core load) would normally be stored in the spent fuel pool at any one time. Additionally, it was anticipated that occasionally (e.g., for inservice inspection of the reactor vessel internals) the entire core would be unloaded and temporarily stored in the spent fuel pool. Therefore, the spent fuel storage rack design was based on the conservative assumption that all fuel rack storage positions would be occupied by fresh unirradiated fuel assemblies of the highest initial enrichment that was foreseen as being useable in that facility.

The penalty in achievable spent fuel storage density associated with this conservative design assumption was relatively small under the circumstances anticipated and easily accommodated by a conservative spent fuel rack design. The potential penalty associated with this conservative design basis is no longer small when long term on-site storage of spent fuel is a necessity. ,

There is no situation where more than one full core load of fresh unirradiated fuel assemblies is to be stored in the spent fuel storage pool. Therefore, it is unnecessary and wasteful to base the entire spent fuel storage rack design on the assumption of fresh unirradiated fuel of the highest initial enrichment.

In the MDR design concept, the spent fuel pool is divided into two separate and distinct regions which for the purpose of criticality l

?.1A-1 7698W112487

't~

l considerations may be considered as separate pools. . Suitability of this design assumption regarding pool separability is assured through ,

appropriate design restrictions at the boundaries between Region-1 and Region 2. The smaller region, Region 1 of'the pool is designed on the basis of currently accepted conservative criteria which allow for the safe storage of. a' number of fresh unirradiated fuel assemblies (including i-

, a full core loading if that should prove necessary). The larger region of the pool, Region 2, is designed to safely store irradiated fuel

. assemblies which will be discharged from the reactor.in large quantities.

The criteria for Region 2 of the pool are specifically listed and compared to the current criticality criteria in Table 9.1 A-1. The only

. change in criteria is the recognition of actual fuel and fissioa product inventory accompanied by a system for checking fuel prior to moving any fuel assembly from Region 1 to Region 2.

l 9.1A.1.2 Desian Descrintion The difference between the NOR design used in Region 2 and the standard j spent fuel storage rack design used in Region 1 is illustrated in

. Figure g.1A-1. Within Region 2 of the pool, certain positions that had been designated as "water boxes' in Region 1 are used to store spent fuel.

I Region 1 has fuel assemblies stored in two out of four bax positions in a i

checker board pattern. Region 2 has fuel assemblies stored in three out

! of four box positions. During a normal refueling operation, each fuel j assembly is first moved from the core to Region 1. After the refueling j

operation is complete and the suitability of each spent fuel assembly for

] movement into Region 2 is verified the fuel assembly may be moved into i Region 2.

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I l g.1A-2 1698W121087 -

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O Region 1 of the pool is designed to maintain stored fuel of up to 4.50 weight percent of U-235 in a safe, coolable, suberitical (k ,less than 0.95) configuration. Region 1 consists of a minimum of 200 spent fuel storage positions. Region 2 is designed to store fuel which has attained the burnups shown in Figure 9.1A-26 and has an ultimate capacity of 1,140 spent fuel assemblies. Details of the criticality analysis for both Regions 1 and 2 are provided in' Sections 9.1A.2 and 9.1A.3.

9.1A.2 CRITICALITY ANALYSIS-FOR THE REGION 1 SPENT FUEL STORAGE RACKS The initial core of the Wolf Creek plant consists of a Westinghouse fuel design which has been designated as "Standard Fuel Assemblies" or SFA.

The original criticality safety analysis of the MDR spent fuel rack design for Wolf Creek demonstrated that SFA designs with initial enrichments up to 3.50 w/o U-235 could be safely stored in the Wolf Creek spent fuel storage racks.

However, the reload designs for Wolf Creek will utilize SFA enrichments greater than the 3.50 w/o assumed 'or the original criticality safety analysis. Since there were substantial margins to criticality safety limits based on the SFA design at 3.50 w/o U-235, it was possible to increase the limit on initial enrichment for the SFA design to a value of 4.50 w/o U-235.

9.1A.2.1 Analvtical Technioue ,

The analytical techniques described here are the same as those originally used to successfully license spent fuel racks for Wolf Creek.

The LEOPARD (Reference 1) computer program is used to generate macroscopic cross sections for input to four energy group diffusion theory calculations which were performed with the PDQ-7 (Reference 2) 9.1A-3 7698W121087 -

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4 program.- LEOPARO calculates the neutron energy spectrum over the entire

, energy range from thermal up to 10 Mov and determines averaged cross sections over appropriate energy groups. The fundamental methods used in ,

the LEOPARO program are those used in the MUFT (Reference 3) and 50FOCATE (Reference 4) programs which were developed under the Naval Reactor Program and thus are well founded and extensively tested

] techniques. In addition, Westinghouse Electric Corporation, the developers of the original LEOPARD program, demonstrated the accuracy of these methods by extensive analysis of measured critical assemblies consisting of slightly enriched 002, fuel rods (Reference 5).

6 In addition, Pickard, Lowe and Garrick, Inc., (PLG), has made a number of improvements to the LEOPARD program to increase its accuracy for the calculation of reactivities in systems which contain significant amounts of plutonium mixed with UO . PLG has tested the accuracy of these 2

modifications by analyzing a series of U0 and Puo -UO critical 2 2 7 experiments. These benchmarking analyses not only demonstrate the 1

improvements obtained for the analysis of Pu0 2

-002 systems, but also demonstrate that these modifications have not adversely affected the

, - accuracy of the PLG-modified LEOPARD program for calculations of slightly enriched UO systems, l 2 i

l The UO 2 critical experiments chosen for benchmarking include variations in H20/UO2 volume ratios, U-235 enrichments, pellet diameters and l cladding materials. Although the LEOPARD model also accurately calculates the reactivity effects of soluble boron, these experiments

have not been included in the LEOPARD benchmarking criticals since most j of the spent fuel rack analyses do not include soluble boron.

l Neutron leakage was represented by using measured buckling input to

infinite lattice LEOPARD calculations to represent the critical I

assembly. A sunenary of the results is shown in Table 9.1 A-2 for the 27 measured criticals chosen as being directly applicable for benchmarking

! the LEOPARD model for generating group average cross sections for spent i

! fuel rack criticality calculations. The average calculated k,,, is l

9.1A-4 l 7698W112487 l

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e 0.9979 and the standard deviation from this average is 0.0080ak.

Reference 5 raised questions concerning the accuracy of the measured buckling reported fot the experiments number 12 through 19. If these data are excluded, the average calculated k,g, for the remaining 19 experiments is 1.0006 with a standard deviation from this value of 0.0063&k. In all of these experiments there are significant uncertainties in the measured bucklings which are necessary inputs to the LEOPARD analysis.- These uncertainties are the same order of magnitude as the indicated errors in the LEOPARD results, and therefore a more definitive set of experimental data is used to establish the accuracy of the combined LEOPAR0/PDQ-7 model us'ed for the criticality analysis of the spent fuel racks.

The P0Q series of programs have been extensively developed and tested over a period of 20 years, and the current version, PDQ-7 is an accurate and reliable model for calculating the suberitical margin of the proposed spent fuel rack arrangement. This code or a mathematically equivalent method is used by all the U.S. suppliers of light water reactor cores and reload fuel. In addition, this code has received extensive utilization in the U.S. Naval Reactor Program.

As a specific demonstration of the accuracy of the calculational model used for the calculations, the combined LEOPARD /PDQ-7 model has been used to calculate fourteen measured just critical assemblies (References 6 and 7). The criticals are high neutron leakage systems with a large variation in U/H O2 volume ratio and include parameters in i the same range as those applicable to the proposed rack designs.

Experiments including soluble boron are included in this demonstration

! since the ability of PDQ-7 to calculate neutron leakage effects is of I primary interest. The use of soluble boron allows changes in the neutron l 1eakage of the assembly while maintaining a uniform lattice and thus allows a better test of the accuracy of the model. Furthermore, it eliminates the error associated with the measured bucklings which is l

inherent in the LEOPARD benchmarks, thus permitting determinations of the actual calculational uncertainty which must be accounted for in the spent fuel rack criticality analysis.

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9.1A-5 I

7698W112487

These combination LEOPAR0/P00-7 calculations' result in a calculated average k,,, of 0.9928 with a standard deviation about this value of 0.0012&k.. These results, as shown in Table 9.1A-3 demonstrate that the proposed LEOPAR0/P00-7 calculational model can calculate the rur.tivity of the proposed spent fuel rack arrangements with an accuracy of bettar than 0.010ak at the 95 percent confidence level.

In addition to the fourteen critical. assemblies in Table 9.1 A-3, the LEOPAR0/P0Q model was used to calculate the k,pp for seven additional criticalassemblies,twoofwhichincorporatedtt,Instpnlesssteel plates in an array which is geometr'ically similar to a section of the spent fuel racks.

These seven criticals were performed by Batta e Pacific Northwest Laboratories specifically for the purpose of providing benchmark critical experiments in support of spent fuel criticality analysis. They are described in detail in Reference 20. The results of these critical experiments are sumarized ).s Table 9.1 A-4 The overall average k,g, calculated for these seven just critical assemblies was 0.9933, with a standard deviation around this value of 0.0013&k.

Combining the results of benchmarking against both the Westinghouse (Table 9.1A-3) and the Battelle (Table 9.1A-4), critical experiments.

results in a mean calculated k,ff for 21 experiments of 0.9929 with a standard deviation of .0012&k as shown in Table 9.1 A-4 Thus, the final bias to be applied to the combined LEOPAR0/PDQ-7 model is

+.00716k, and tne 2.40a uncertainty to be applied, corresponding to a 95 percent probability at a 95 percent confidence level, is .0029ak.

As a result of this approach to separately benchmark both the cross section and the diffbiion theory calculations against applicable critical assemblies, any differences between transport theory and dif fusion thrscwy calculations are implicitly included in the derived calculational uncertainty f actor and need not be accounted for separately.

9.1A-6 1698W112487

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~9.1A.2.2 ~ Evaluation of Criticality Safety for Recion 1

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The PDO-7 program is used in-the final predictions of the multiplication

. factor of the spent fuel storage pcol. The calculations are performed in four energy groups 'and take'.into acr. aunt all the significant geometric details of thr fuil_ bundles, fuel b.o u s and major structural compenents.

The geometry neo for these calculationc '.s that of a basic cell representing one quarter of the arealo a repeating array of four stainlesssteelboxes. Two of the diagonaly opposing boxes accommodate fuel assemblies Mile the remaining two serve as flux traps. The specific geome";ry and dimensions of this basic cell are shown in u Figure 9.1 A-2,- and[th'e fuel assembly characteristics are; listed in Table 9.1A-5.

.s The calculational. approach is to use the basic cell to chiculan the reactivity of an fofinite array of uniform spent fuel racks'and to account for any deviations of the actual spent fuel rack array f rom this-assumed infinite array as perturbations on the calculated reactivity of s the basic call. The fuel assemblies were issut.cf to be unirradiated .dth, a uniform U-235 enrichment of 4.50 w/ow~ hich is higher than any currently anticipated reload enrichment for the. Wolf Creek core. Most of the calculations were performed at a uniform pool temperature of 68'F, but the reactivity ef fects of poO temper.lture ars also taken into account as a perturbation on the basic cell' calculations.

i ,

The reference basic cell calculation is performed with nominal dimensions on all the stainless steel boxes and results in a k ,of 0.9329.

Tolerances on the geometric array representing the racks are treated as perturbations on this reference basic call calculation.

The calculated variation of the basic cell k, a's a function of the initial enrichment of the fuel assembly is shown in Figure 9.1A-3.

The k, of the basic cell as a function of tempecature is shown in Figure 9.1A-4. Bast.d on this figure, a max %um pot,1 temperature of 160*F t

9.1A-7 7698W112487 ,

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~ a:5 a minimum of 60'F, which correspond to the-design limits for the

!, . pool, result in a maximum perturbation due to temperature effects of ,

\ 4.0023ak.

Based on the results of the benchmarking of the combined LEOPARD /P00-7 analysis model, the bias in the calculated multiplication factor compared to measured just critical arrays is '+.0071ak, and this bias must be added to the calcu'ated basic cell reactivity.

(i Most of the calculations with the basic cell geometry utilized a 36 x 36 two-disansional array of r,;esh points. To test the adequ'acy of this mesh descrirgion, a calculation was run with a 72 x 72 mesh size ano the resultins k ,was .9319. Thus, the perturbation on the basic cell due to mesh spacing effects is .0010ak.

The sensitivity of the spent fuel storage rack multiplication factor to the simultaneous and uniform variation of water density in both the fuel box and water box is illustrated in Figure 9.lA-5. Since the loss of cooling water in the spent fuel pool is considered an accident condition, the double contingency princhle of ANSI N16.1-1975 can be applied. As discussed in Reference 21, this principle allows credit tc be taken for the 7000 ppm of soluble boron normally present in the spent fuel pool.

9.lA.2,3- E le+a'nces and Uncertainties for Region 1 There are a number of tolerances and uncertainties which result in perturbations which must be considered in the criticality analysis. The reactivity effect of all such positive perturbations is then combined statistically in accordance with Reference 21 to determine a single reactivity perturbation which is added to the calculated basic cell multiplication factor (including biases) to determine the final conservttive evaluation of the spent fuel rack maximum possible

! multiplication factor, t

i

/ The basic cell is based upon an inside dimension for the fuel box of i

8.996 inches with a tolerance of t.030 inch. The corresponding minimum 3

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. box pitch is 9.206 inches and the k ,for that case is .9343, as shown in Figure 9.lA-6. Therefore, the perturbation due to box size tolerances-is +.0014ak.

The stainless steel fuel and water boxes are nominally .120 inches thick with a tolerance of 2 004 inches. Assuming a worst case in which all boxes were' at .the minimum thickness of .116 inches, the k, of the basic cell is .9344, as illustrated in Figure 9.lA-7. Therefore, the maximum perturbation on the reactivity of the basic cell due to variations in the stainless steel box thickness is +.0015ak.

With the tuel bundles located in their most reactive positions inside the stainless steel boxes, the k,of the basic cell would be .9375.

Thus, the perturbation on the basic cell reactivity due to positioning uncertainties is +.0046ak.

The nominal density of the pellets contained in the fuel assemblies is 94.5% of theoretical density. Increasing this density by 1%

corresponding to a maximum theoretical density of UO pellets of 95.5%

2 results in a positive reactivity perturbation of .00llak.

As noted above, Reference 21 allows the reactivity perturbations due to tolerances and uncertainties to be combined statistically. When this is done as shown in Table 9.lA-6, the total reactivity perturbation to be added to the biased basic cell reactivity to account for tolerances and uncertainties is +.0059ak. This results in a final conservatively calculated multiplication factor of .9472 for Region 1 of the spent fuel rack.

Realistically, the spent fuel pool coolant contains a concentration of 2000 ppm boron at all times. When the reactivity effect of this minimum boron concentration is included, the actual spent fuel pool multiplication factor is .7043. When all biases, tolerances and uncertainties are included, the multiplication factor is increased to

.7186.

9.lA-9 . ._

7 & O CI.81 ') 16 01

9.lA.2.4 Accident Analysis The fuel racks are designed to prevent any criticality threat caused by a dropped fuel assembly which penetrates and occupies a position other than l

a normal fuel storage location. The only positive effect of such an assembly on the reactivity of the rack would be by virtue of reduction-in axial neutron leakage from the rack. Since no credit is taken for axial leakage, this accident cannot increase the multiplication factor above the reported value.

For an enrichment of 4.50 w/o, the lattice of the fuel assemblies results in an under-roderated configuration. This is shown by the results plotted in Figure 9.lA-5 which demonstrate that a reduction in water density produces a decrease in the multiplication factor. Therefore, any crushing or compaction of the fuel assemblies would tend to reduce the multiplication factor of the spent fuel pool, and the dropping of heavy objects into the fuel pool or deformations from the effects of earthquakes or tornadoes could not produce a criticality accident.

The spent fuel racks include stainless steel standof fs which maintain a spacing of at least 3.0 inches between the racks and any fuel assembly which might be inadvertently located immediately adjacent to a rack.

These standof fs limit the reactivity increase for such an event to less than .0012 ak. In addition, this event is considered to be an abnormal condition for which credit may be taken for the soluble bcron in the pool water which has a negative reactivity worth of about .2286ak.

l Therefore, such an accident scenario would not increase the reactivity of the spent fuel storage rack above the reported maximum.

l 9.lA.2.5 Analysis Conservatisms As discussed previously, the basic cell calculations make the conservative assumption that the fuel pool wate" contains no boric acid when in fact the minimum boron content in the pool water is 2000 ppm. A realistic evaluation of the maximum multiplication factor of the spent fuel racks, even when completely filled with unirradiated fuel, is

.7186.

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m 9.lA.3 CRITICALITY ANALYSIS FOR THE REGION 2 SPENT FUEL STORAGE RACKS

, The following discussion is aoplicable to thesfuel stored in Region 2 of the pool. Since the criticality analyses for Region 2 are subject to potentially larger uncertainties than those applicable to Region 1,.the uncertainties applicable to the Region 2 analyses are independently derived.

9.1A.3.1 Analytical Technioue The analytical methods used for criticality analysis of Region 1 are also incorporated into the criticality analysis of Region 2. The isotopic composition is calculated.as a function of irradiation time, assembly average burnup, and subsequent decay using the LEOPARD (Reference 1) and-CINDER (Reference 8) computer programs. Once the isotopic composition of the fuel assemblies is known, the subsequent criticality calculations for the spent fuel racks in Region 2 are performed in a manner that is analogous to the calculations for Region 1.

The accuracy of the burnup dependent isotopic concentrations calculeted

! with the LEOPARD program is demonstrated in Figuras 9.l A-8 through 9.l A-18. Figures 9.1 A-8 through 9.l A-15 show comparison of LEOPARD calculated data with measured data from a UO fuel assembly irradiated 2

in the Yankee-Rowe reactor while Figures 9.1 A-16 through 9.1 A-18 show corresponding data for a mixed oxide (Pu02-UO2 ) fuel assembly irradiated in the SAXTON reactor.

Except for the data labeled PLG calculation, the data and curves on l

Figures 9.1 A-8 through 9.1 A-15 and Figures 9.1 A-16 through 9.1 A-18 are taken directly from References 9 and 10, respectively. In a'l cases, the accuracy of the calculations labeled PLG is within the uncertainty in the measured data.

The accuracy of reactivity calculations for irradiated fuel can be I demonstrated in part by the analysis of critical arrays of mixed oxide l

l l 9.1 A-11 7698W112487 i

fuel rods which contain high concentrations of the plutonium isotopes.

Tables 9.lA-7 and 9.lA-8 show results of criticality analyses for the ,

SAXTON (Reference 11) and ESA0A (Reference 12) sets of experiments which cover a wide range of water-to-oxide volume ratios. A summary of these date is shown in Table 9.1A-9. For the mixed oxide criticals, the calculated mean multiplication factor is 0.9969 with a standard deviation about this value of 0.0066ak. Using the 95% probability at 95%

confidence level criterion (one-sided) with 11 data points, implies an uncertainty of 0.0186ak (2.82 o) with a bias to be applied of

+.0031 ak .

The other major uncertainty in the calculations for Region 2 is associated with the calculated reduction in fuel assembly reactivity I associated with the depletion of the heavy metals and the accumulation of fission products as a function of fuel assembly exposure. As an example, I consider a 4.50 w/o (initial enrichment) Wolf Creek fuel assembly at an exposure of 45,000 MWD /MT. The total reactivity loss from the fresh unirradiated case is 0.2576 ak/k, of which approximately 50% can be attributed to the build-up of fission products (ak/k is used to provide a- consistent basis for comparing perturbations at dif ferent values of

[ k,). Calculations of reactor reactivity lifetimes using the same analytical methods as used in this analysis demonstrate an accuracy of l

better than 15%. Therefore, the resulting uncertainty in the calculated fuel assembly k, associated with fuel depletion would be conservatively estimated at 0.0129 ak/k (= .05 x .2576 ok/k). The corresponding uncertainty in the calculated Region 2 multiplication f actor is 0.0116ak on a base case Region 2 k, of 0.8998.

In order to provide further assurance of the conservative nature of these l

calculations, the decay of all fission products following discharge of the fuel assembly was taken into account. This was accomplished with the aid of the CINDER (Reference 8) code which treats a total of 186 nuclides in 84 linear chains. The fission product inventory for each fuel assembly was decayed for thirty years following its removal from the reactor core, and the time point of minimum fission product absorption within that thirty year period was used as the basis for determining the l

1 9.l A-12 ~

7698W121087

3 fission product macroscopic absorption cross sections'for that-particular fuel assembly at that specific exposure. That minimum occurs at approximately 3.5 days into the decay and from then on continues to increase as illustrated in Figure 9.lA-19. Reduction in the fission

. product inventory due to leakage or escape to the plenum has been found to be negligible (Reference 13).

9.1A.3.2 Calculational Approach Reactivity calculations, using the LEOPARD and P0Q-7 models described previously for the Region 1 analysis, were performed for Region 2. The geometry used for these calculations is that of a basic cell representing j a repeating array of four stainless steel boxes, three of which

-accommodate fuel assemblies while the fourth serves as a flux trap. The specific geonetry and dimensions of this basic cell are Snown in l Figure 9.1A-20.

I 9.lA.3.3 Manufacturing and Thermal Considerr.tions l

The calculations. of manuf acturing and thermal considentions were evaluated for depleted fuel stored in the Region 2 configuration. This evaluation assumed a Westinghouse 17 x 17 SFA design, characterized by a 4.50 w/o initial enrichment at 45,000 MWD /MT burnup. This fuel assembly and burnup were selected as being typical of the irradiated fuel being considered.

Specific calculations were performed on the irradiated fuel for variations in all relevant physical dimensions, temperature, water density, pellet density, and fuel assembly position. Table 9.l A-10 provides a sumary of the reactivity perturbations to the Region 2 spent fuel storage racks. Detailed results of these calculations for Region 2 are presented in Figures 9.lA-21 through 9.lA-24 for the effects of cell pitch, stainless steel wall thickness, temperature, and water density variations, respectively.

9.l A-13 7698W121087 - --

. g' 9.1A.3.4 Desian Conservatisms While the MOR concept reduces some of the design conservatisms inherent

.in the earlier spent fuel storage concepts (e.g... assumption of fresh unirradiated fuel), the design and analyses for the MOR, as implemented in Region 2, are still conservative in nature. This is evidenced by the result obtained when Region.2 is assumed to contain fresh unirradiated SFA at 4.50 w/o and 2000 ppm boron as discussed further in the discussion of accident analysis which follows. For these conditions, the maximum multiplication factor, including al'1 applicable biases and uncertainties, is still only .9408.

9.1A.3.5 Accident Analysis The Region 2 fuel racks are designed to prevent a dropped fuel bundle i

f rom penetrating and occupying a position other than a normal fuel storage location.

The only positive reactivity effect of such a 5undle on the multiplication f actor of the rack would be by virtue of a reduction in axial neutron leakage from the rack. As previously described for Region 1, no credit is taken for axial leakage in this analysis, therefore this accident cannot increase the multiplication factor above the reported value.

The positive reactivity effect of a fuel assembly located adjacent to the fully loaded Region 2 spent fuel storage rack is minimized.by 3 inch stainless steel standoffs, and as was the case for Region 1, the negative reactivity ef fect of soluble boron more than compensates for the small positive reactivity effect associated with this abnormal condition.

The other accident that must be addressed is that involving a fuel assembly f rom Region 1 being incorrectly transferred to Region 2 (e.g., a f resh fuel assembly or an insufficiently depleted fuel assembly). This accident was analyzed assuming the extreme case of completely loading s Region 2 with unirradiated 4.50 w/o fuel assemblies that served as the j reference fuel for the Region 1 analysis,

\

f 7698W121087 . . . -

1

7 ,

i This, of course, is considered as an abnormal Region 2 condition and appropriate credit is taken for .1 nominal 2000 ppm of soluble boron present in the spent fuel pool. The resulting Region 2 k ,, including all applicable biases and uncertainties, is then calculated to be 0.9408. Thus, in all cases, the Region 2 spent fuel storage rack design ensures that the multiplication factor is less than the 0.95 limit.

9.lA.3.6 Reouired Burnup as a Function of Initial Enrichment for Region 2 Spent Fuel l

j The maximum combined bias and uncertainty were determined for 4.50 w/o i SFA fuel at a burnup of 45,000 MWO/MT. The biases were added and the uncertainties were combined by summing the squares of the individual I

uncertainties, and taking the square root of the sum. The results of l j these calculations are shown in Table 9.l A-10. The total uncertainty and l bias is 0.0267ak, giving a ak/k value of 0.0297. (The relative l change, ak/k, is used to provide a consistent basis for comparing j uncertainties at different values or k,.) Since a large component of the uncertainty is the uncertainty in the reactivity of the depleted i- fuel, and since this uncertainty increases with exposure, the calculated uncertainty is conservative for all exposures less tuen 45,000 MWD /MT.

Thus, the k ,with uncertainties will be less than 0.95 at the 95%

confidence level if the computed k , satisfies the relationship k ,+ k ,(ak/k) < 0.95 or, using ak/k = 0.0297, if the computed k ,is less than 0.9226.

However, in order to provide additional margin, a target k, of 0.9150 was used. The values of k ,as a function of initial enrichment and burnup given in Table 9.lA-ll are plotted in Figure 9.1A-25. These curves are then interpolated using interpolating polynomials of appropriate order to find the burnup which results in a k ,of 0.9150 for each initial enrichment and the results are presented in Table 9.1 A-12. The curve which fits these points, shown in Figure 9.1A-26, gives the required minimum exposure as a function of i

9.lA-15 7698W121087 . _

initial enrichment.to assure that the value of k ,.in the spent fuel is below 0.95 with a. probability of 95% at the 95% confidence level. ,

~

Because of the well-founded, conservative technique used-for determination of the infinite multiplication factor, there is assurance that this spent fuel rack design will not cause undue risk to the public health and safety resulting from criticality considerations.

f i

i e

i I

i 9.lA-16 7698W121087 - ---

)

.m i

1 REFERENCES i l

l

1. R.F. Barry, "LEOPARD--A Spectrum Dependent Non-Spatial Depletion Code j

~

f or the IBM-7094," WCAP-3269, September 1963.

2. W.R. Caldwell, "P0Q-7 Reference Manual," WAPD-TM-678, January 1967.

3. H. Bohl, E. Gelbard and G. Ryan, "MUFT-4-- Fast Neutron Spectrum Code for the IBM-704," WADP-TM-72, July 1957.

4. H. Amster and R. Suarez, "The Calculation of Thermal Constants Averaged Over a Wigner-Wilkins Flux Spectrum: Description of the i

SOFOCATE Code," WAPD-TM-39, January 1957.

5. L.E. Strawbridge and R.F. Barry, "Criticality Calculations for Uniform Water-Moderated Lattices," Nuclear Science and Engineering,

! 23, 58, 1965.

l

6. "Large Closed-Cycle Water Reactor Research and Development Program Progress Report for the Period January 1 - March 31, 1965,"

WCAP-3269-12.

7. "List of Equipment and Apparatus at WREC," Westinghouse Reactor Evaluation Center, February 1967.

8. Electric Power Research Institute, "Fission Product Data for Thermal reactors, Part 1 and Part 2: Data Set for EPRI-CINDER and Users Manual for EPRI-CINDER Code and Data," EPRI NP-356, Final Report (1976).

9. R. J. Nodvik, "Evaluation of Mass Spectometric and Radiochemical Analyses of Yankee Core I and Core II Spent fuel," WCAP-6068 (1965).

10. R. J. Nodvik, "Saxton Core II Fuel Performance Evaluation of Mass Spectometric and Radiochemical Analyses of Irradiated Saxton Plutonium Fuel," WCAP-3385-56, Part II (1970).

9.1A-17 7698W112487

\

l-i

11. W. L. Orr, H. I. Sternberg, P. Deramaix, R. H. Chastain, L. Binder -

and A. J. Impink, "Saxton Plutonium Program, Nuclear Design of the i Saxton Partial Plutonium Core," WCAP-3385-51, December 1965. (Also ,

EURAEC-1490). 1

12. R. D. Leamer, W. L. Orr, R. L. Stover, E. G. Taylor, J. P. Tobin and A. Bukmir, "Pu02-UO2 Fueled Critical Experiments," WCAP-3726-1, July 1967.

~

{

13. R. A. Lorenz, et al., "Fission Product Release f rom Highly Irradiated LWR Fuel," NUREG/CR-0722, February 1980.

i

14. P. W. Davison, et al., "Yankee Critical Experiments Measurements on Lattices of Stainless Steel Clad Slightly Enriched Uranium Dioxide i

~ Fuel Rods in Light Water," YAEC-94, Westinghouse Atomic Power Division (1959).

i

15. V. E. Grob and P. W. Davison, et al., "Multi-Region Reactor Lattice Studies - Results of Critical Experiments in Loose Lattices of UO 2

Rods-in H 0," WCAP-1412, Westinghouse Atomic Power Division (1960).

2 I

16. W. J. Eich and W. P. Kovacik, "Reactivity and Neutron Flux Studies in Multiregion Loaded Core," WCAP-1433, Westinghouse Atomic Power Division (1961).

17. W. J. Eich, Personal Conununication (1963).

18. T. C. Engelder, et al., Measurement and Analysis of Uniform Latticas of Slightly Enriched H0 2M derated by 0 0-H 2O Mixtures,"

2 8AW-1273, the Babcock & Wilcox Company (1963).

l

19. A. L. MacKinney and R. M. Ball, "Reactivity Measurements on Unperturbed, Slightly Enriched Uranium Dioxide lattices," BAW-1199, the Babcock & Wilcox Company (1960).

I 9.1 A-18 I

7698W112487

Y 4

o' _ _

20. Battelle Pacific Northwest Laboratories, "Critical Separation Between Suberitical Clusters of 2.35 Wt% 235-U Enriched UO2 Rods in Water ,

with Fixed Neutron Poisons," PNL-2438.

~

21. "07 Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," U.S. NRC, April 14, 1978, as revised

' January. 18, 1979..

i l

l-i l

i

i

..-l r

TABLE 9.lA 'i. CRITICALITY DESIGN CRITERIA q

I Region 1 Criteria Region 2 Criteria l-i 1. Fresh, unirradiated' l. Actual' irradiated fuel l fuel inventory is and fission product

l. -assumed. inventory is assumed.

2. kef f < 0.95 2. Same

3. Fuel location recorded 3. Controls required for once prior to storage in i each fuel assembly prior Region 1. to transfer from Region 1 to Region 2.,

s e

9.1A-20

taste 9.1 A-2. SupeeARY OF LEOPARD RESULTS FOR SEEASURE3 CRITICAL 5 N Fuel Pellet Clad Clad Lattice Critical Ca se" Re ference Enrichment H2 0/U Denst Disseter Diaeeter Thickness Buckling Number Number Jgca_gy Pitch Calculated (atom 11 Volume ) (cm) (col m2 (cm) (cm) k,gg _ ,

1 16 2. 734 2.18 10.18 0.7620 0.8594 0.04085 1.0287 40.75 1.0015 2 16 2.734 2.SJ 10.18 0.7620 0.8594 0.04085 1.1049 53.23 1.0052 3 16 2.734 3.86 10.18 0.7620 0.8594 0.04085 1.1938 63.26 1.0043 4 17 2.734 7.02 10.18 0.7620 0.8594 0.04085 1.4554 65.64 1.0098 5 17 2.734 8.49 10.18 d.7620 0.8594 0.04085 1.5621 60.07 1.0118 6 17 2./34 10.38 10.18 0.7620 0.8594 0.04085 1.6891 52.92 1.007?

7 18 2.734 2.50 10.18 0.7620 0.8594 0.04085 1.0617 47.5 1.00ra 8 18 2.734 4.51 10.18 0.7620 0.8594 0.04085 1.2522 68.8 0.9987 9 18 3.745 2.50 10.37 0.7544 0.8600 0.0406 1.0617 68.3 1.0010 10 18 3.745 4.51 10.37 0.7544 0.8600 0.0406 1.2522 95.1 1.0025 11 19 3.745 4.51 10.37 0.7544 0.8600 0.0406 1.2522 95.68 1.0009 12 20 4.069 2.55 9.46 1.1278 1.2090 0.0406 1.5113 88.0 0.9889 13 20 4.069 2.14 9.45 1.1278 1.2090 0.0406 1.450 79.0 0.9830 14 21 4.059 2.59 9.45 1.1268 1.2701 0.07163 1.555 69.25 0.9999

? 15 21 4.069 3.53 9.45 1.1268 1.2701 0.07163 1.684 85.52 O.9958 16 21 4.069 8.02 9.45 1.1268 1.2701 0.07163 2.198 92.84 1.0040 ru 17 21 4.069 9.90

>* 9.45 1.1268 1.2701 0.07163 - 2.381 91.79 0.9872 18 .21 3.037 2.64 9.28 1.1268- 1.2701 0.07163 1.555 50.75 0.9946 19 21 3.037 8.16 9.28 1.1268 1.2701 0.07163 2.198 68.81 0.9809 20 13 0.714* l.68 9.52 0.8570 0.9931 0.0592 1.3208 108.8 0.9912 21 13 0.714* 2.17 9.52 0.8570 0.9931 0.0592 1.4224 121.5 1.0029 22 13 0.714* 4.70 9.52 0.8'M 0.9931 0.0592 1.8669 159.6 0.9944 23 13 0.714* 10.76 9.52 0.8570 0.9931 0.0592 2.6416 128.4 1.0008 24 14 0.729* 1.11 9.35 1.2827 1.4427 0.0800 1.7526 89.1 0.9902 25 14 0.729* 3.49 9.35 1.2827 1.4427 0.0800 2.4785 104.72 1.0055 26 14 0.729* 3.49 9.35 1.2827 1.4427 0.0800. 2.4785 79.5 0.9948 27 14 0.729* 1.54 9.35 1.2827 1.4427 0.0800 1.9050 90.0 0.9878

• These are Pupy in Natural UO

• 2 "Cases ! through 19 are with stainless steel clad, Cases 20 through 27 are zircaloy.

7516U111687

• 4

TABLE 9.1A-3. WESTINGHOUSE U02 Zr-4 CLAD CYLIlWRICAL CORE CRITICAL EXPERIMENTS (6,7)

.I i

Material  :

Boron Buckling Radius of Pitch Concentration (for LEOPARD Critical No. Fuel Region. 'keff Experiment In (ppm) CM-2 of Pins (cm) (LEOPARD /P00-7) 1 0.600 0 .008793 489.4 19.021 0.9912 2 0.690 0 .009725 317.0 17.605 0.9941 3 0.848 0 .nnA637- 251.6 19.276 0.9927' 4 0.976 0 .006458 293.0' 23.935 0.9935 5 0.600 306. .007177 659.9 22.088 0.9927.

6 0.600 536.4 .006244 807.2 24.429 0.9937 7 0.600 727.7 .005572 950.2 26.504 0.9940 8 0.600 104. .008165 546.3 20.097 0.9919 9 0.600 218. .007599 607.1 21.186 0.9917 10 0.600 330. .007106 669.5 22.248 0.9916 11 0.600 446. .006661 735.3 23.315 0.9909 P 12 0.600 657.1 .005800 895.3 25.727 0.9944 g 13 0.848 104. .007320 321. 0 21.772 . 0.9938 g 14 0.848 218. .006073 420.5 24.919 0.9925 0.9928 Mean 0.0012 Std Notes '

(a) Fuel Region Data Enrichment = 2.719 w/o U-235 (b) Thickness of water reflector is that required to

, Fuel Density = 10.41 g/cm3 attain total radius of 50 cm for model.

Pe11et Radius' = 0.20 in 2 Clad IR = 0.2027 in (c) B (PDQ-7) = .000527 co-2 Clad OR = 0.23415 in Z 4

.I

~

7516UO80185 L___ .. .

^

. _ . . - - - = . - - - - . - . - - --

TABLE 9.1A-4. BATTELLE CRITICALS Length No. of ' Distance Critical' -

Times Assemblies Absorber To Fuel' Separation Case Width Type Thickness Cluster

-keff In Array of Clusters LEOPARD /PDQ 028 20 x 16 3 S.S. .485 cm .645 cm 6.88 cm -0.9922-027 20 x 16 3 S.S. .302 .645 7.43 '0.9919 0028 20 x 18.075 1 None - -

0.9956 015 20 x 17 3 Mone - -

11.94 ca 0.9932-013 20 x 16 3 None - -

8.42 0.9921 022 20 x 15 3 None - -

6.39 0.9933 021 20 x 16 3 None - -

4.46 '0.9946

?

E Statistical Sunnary: .

O Experiments Number Mean k,rf a Battelle 7 .9933 0.0013 8attelle I and Westinghouse 21 .9929 .0012 (Table 9.1A-3) -

Fuel region data: Enrichment = 2.35 w/o, Pellet radius - 0.5588 cm, Clad OR = .635 cm, Wall thickness = .0762 cm Pitch = 2.032 ca .

t 7516U073185

c .

TABLE 9.1A-5. FUEL ASSEMBLY TECHNICAL INFORMATION FOR WOLF CREEK A'ssembly Description -

Rod array 17 x 17 Overall dimensions, in._ 8.426 x 8.426 Fuel Rod Description Fuel rods per assembly '

264 Outside diameter, in. 0.374 Cladding wall thickness, in. 0.0225 Cladding material Zircaloy-4 Pitch, in. 0.496 Active fuel length, in. 144 q Fuel Pellet Description Pellet material UO2 Pellet density, gm/cc. 10.28 l Outside diameter, in. 0.3225

! 002 weight (1b/ft of fuel rod) 0.3643 Control Rod Cell Description

~

Control rod cells per assembly 24 Guide thimble material Zircaloy-4 Guide thimble outside diameter, in. 0.482 Guide thtmble inside diameter, in. 0.450

! Instrument Cell Description

~

Instrument cells per assembly 1 Guide thimble material Zircaloy-4 Guide thimble outside diameter, in. 0.482 Guide thimble inside diameter, in. 0.450 Grids. Ef fective Weight in Core (1b) 4 Inconel 2324 4

9.1A-24 7697W112387

_]

n-

--a.-

I

!- TABLE 9.1A-6.

SUMMARY

OF PERTURBATIONS T0 THE MULTIPLICATION 1

FACTOR OF THE BASIC CELL FOR REGIOR 1 i

i

.-i.

Description 'ak effect k.

Basic cell at 68'F, 4.50 w/o U-235 Pitch = 9.236 inches 0.9329 Calculational Biases

, Increase in temperature f rom 68'F to 160'F +0.0023 LEOPARD /P0Q Model Bias +0.0071 Mesh spacing effect -0.0010 l Total Bias +0.0084 Basic Cell including biases -0.9413

Tolerances and Uncertainties Tolerance on Inside Box Dimension 10.0014 (i.e., minimum pitch)

- Tolerece on SS box thickness to.0015

, Fuel position uncertainty 10.0046

! Maximum fuel pellet density 10.0011 Calculational uncertainty !0.0029

> Total Uncertainty (statistical combination) 10.0059 Maximum reactivity change from biases and 10.0143 uncertainties Maximum k. including Biases and Uncertainties 0.9472 4

9.1A-25 7515U101185 - -- -

, , , ._-.,,..-#-, -+y--w-yw -'FT----- v-_

.q t

j TABLE 9.lA-7. SAXTON Pu02 -UO2 CRITICAL EXPERIMENTS (Reference 11)

Expt. Boron H20/UO2 Pitch (ppm) TVolume) k.ff~ k.ff~-1

.(Incnes) 1 0 1.68 .520' .9912 .0088 2 0 2.17 .560 1.0029 +.0029 3 337 2.17 , .560 1.0084 +.0084 l 4 0 4.70 .735 .9944 .0056 5 0- 10.76 1.040 1.0008 +.0008

1 4

i l

l l

l e

r b

l 9.1A-26 7515UO80185 - -

TABLE 9.1A-8. ESADA Pu02-UO2 CRITICAL EXPERIMENTS l (Reference 12) k Expt. Boron Pu-240 H90/UOp Pitch (ppm) (%) Dolume) (Inches)

. k.ff_ Affd 1 0 8 1.11 .690 .9902 .0098 2 0 8 3.49 .?758 1.0055 +.0055 3 526 8 ,3.49 .9758 .9949 .0051 4 0 24 3.49 .9758 .9948 .0052 5 0 8 1,54 .750 .9878 .0122 i 6 526 8 1.11 .690

.9945 .0055

'l f

1 i

i h

e e

6

• l 1

9.1A-27

• 7515U080185 . -

4

- ----,w-- - , , , - , - - , - - , , - - . - c---, _ - - - - - --,,--nva+-,,,-- ------------,e,n.,, r , - v

l l

TABLE 9.lA-9. SUtNARY OF PREDICTIONS FOR keff IN CRITICALITY EXPERIMENTS -

Experiment Cases hff Saxton Pu0 2-UO2 5 0.9995 + .0068 Esada Pu0 2-UO2 6 0.9946 + .0061 All Pu0 -UO2 2 11 0.9969 + .0066 l -

t I

9.1A-28

~~ '

l 7515V080185

• i If i

i TABLE 9.1A-10.

SUMMARY

OF PERTURBATIONS TO THE MULTIPLICATION FACTOR OF THE,8ASIC CELL FOR REGION 2 Description ak effect k_

Basic rack cell at 68'F, 4.50 w/o U-235, 45,000 MWO/MTU, Pitch = 9.236 inches 0.8998 Calculational Biases Leopard /P0Q model bias +0.0031 j Mesh spacing effect -0.0.007 i

l Most Reactive Temperature over operating range +0.0022 l Most Reactive Water Density +0.0000 t- Total Bias 4.0046 Basic Cell including biases 0.9044 Tolerances and Uncertaiaties (95/95)

Depleted fuel reactivity uncertainties 10.0116 Maximum error due to pitch tolerance 10.0012 Maximum error due to SS thickness tolerance 10.0014 Fuel position uncertainty 10.0000 Maximum error due to pellet density tolerance (t.010) 10.0020 Calculational Uncertainty (2.82o) 10.0186 Total Uncertaintv (statistical combination) _4 .0221 Maximum reactivity change from biases and uncertainties +0.0267 Maximum k., including biases and uncertainties 0.9265 9.1A-29

~

7697W112387

• TABLE 9.1 A-11. ' REGION 2 k. AS A FUNCTION OF INITIAL ENRICHMENT AND BURNUP i

i Initial Burnup Region 2 Enrichment. w/o MWD /MT k.

2.10 7,000 0.9401 2.10 13,000 0.8891 2.10 19,000 0.8462 2.60 13,000 ~ 0.9469 2.60 19,000 0.9023 j 2.60 25,000 0.8623

.i

! 3.10 19,000 0.9501 i 3.10 25,000 0.9098 3.10 31,000 0.8711 3.80 28,000 0.9479 3.80 39,000 0.8825 l

4.50 33,000 0.9664 4.50 39,000 0.9328 4.50 45,000 0.8998 t  :

i l \

o :

l l t I

l-l 9'.1A-30 7697W112387 l . --

1

• j

! TABLE 9.1 A-12. REGION 2 MINIMUM BURNUP AS A FUNCTION f J0F ENRICHMENT TO OBTAIN A k. OF 0.915 PRIOR TO THE

! ADDITION OF TOLERANCES AND UNCERTAINTIES i

l'

. Initial Minimum

. Enrichment, w/o Burnup, MWD /MT 2.10 9,830 2.60 17,220 3.10 24,200 3.80 33,540 4.50 -

42,210 l

)

I i

l 1 l:

l l

l t

9.1A-31 7697W112387

REGION 1- P.EGION 2 50%' FUEL STORAGE 75% FUEL STORAGE CELL PACKING FRACTIO CELL PACKING FRACTION XEX o X rxo EE!ooEooo oX E;X nX EX EX EX EXoX

! X oXEXiiX E nnC o oooo

! EXEX oXEX -

EX oX oX oX

X EX E X r XE . Co E : J E : E i EX EX E ;X EX JX oX E XoX X EX rX LXE EE E E E E o o EXEXEXoX oX EXEXoX

[ - ruel Secrsge C211

- Water Box Cell FIGURE 9.1A-1. COMPARISON OF REGION 1 AND REGION 2 FUEL STORAGE ARRANGEMENT 9.1A-32

n.

~

F-

~ T l

C L

W  !

h h 9 .236" 2[

M.236M.236'M N

\

3

.24"

.282" water SS N i o \

4. 216 " f uel -

V \

P' 4.216,H ll- M

\ --*i

\

fuel 4.498" water i

[.24"SS

.282" water FIGURE 9.1A-2. 900 CALCULATIONAL MODEL FOR THE WOLF CREEK '

REGION 1 SPENT FUEL RACKS I

9.1A-33 --

aw 7:a -- - . .

'i, ti

^

% .< r,

+3 ,

0.95 -

4

'i 0.93 - -

0.91 - -

0.89 - -

0.87 - -

.t k= --

0.85 - -

0.83 - -

0.81 - -

i 0.79 - -

i l O.77 - -

0.75 --+: : : l: : : H- l : : : : l: : : l l: : : : l: : : :

t l 2.00 2.50 3.00 3.50 4.00 4.50 5.00 i

l' l >

ENRICHMENT ( w/o )

! f!GURE 9.1A-3. WOLF CREEK REGION 1 k VERSip FUEL ASSEMBLY j INITIAL ENRICHMENT 9.1A-34 t w w- -m e.,- , . - - , . , - , - - -,,,.-,p. 1- _

- , 3 3,;

.v i .s - 'e j \l w ,,' .3j'lL -

/

1

- {. . _ -[

i

9' s ;

,  ; - c, x ,

0.937 ..

- - 4

'N.. f ' ,. '-

p/ 4 0.936 - ..

v<3

/N v

- /,

( K} \ ,.

0.935 -- x.-

3 ..

i A

1 7'

)

.- 0.934-- -

y- f koo n;

/ '"

) ,-.

' ^

0.933 - -

( . . .

c i

0.932 -..

0.931 -..

0.930  : : : : t- H : : l: : : : l.: : : l: ::: l: : . : : 1 50 75 100 125 150 175 200 t ,

i o00L TEMPERATURE (o F ). -e FIGURE 9.1A-4. WOLF CREEK REGION 1 k VERSUS PdOL TEMPERATUht ..

r 9.1A-35 '

i . _. .-

-l'

( I

~

pp a .:

g: .  ;, >g-

't). . ~~< \

v ;; ,,

a s +

%"i!k(

gt -

1.050 ..

gE x. - = ==

1.000 -

l oxaux----

_- . , s

'i.s (

0.950 - -]

~

. s

,[]

s . .. s, '

0.'9 0 0 - .

~

' 'Q-4 B __.

0.850- b

~

km 1 0.800 - ..

0,750 - -

(  :

0.700} }

y .

~

0.650  : l  : l  : l  : l  : l  : [: l : l  : l l O.0 0.1 0.20.30.40.50.60.70.80.9 1.0

,. RELATIVE WATER DENSITf FIGURE 9.1A-5. WOLF CREEX REGION 1 k VERSUS WATER DENSITY l'

9.1A-36

~

m, s,p e ----,,--y - - . - -- - y- - - - - . ~ _ .

0.936 0.935 - -

0.934 - -

k co 0.933 - -

1 ..

0.932 - -

0.931 - -

l ..

0.930  :  : l  :  : l  :  : l  :  :

9.176 9.206 9.236 9.266 9.296 CELL PITCH ( in )

FIGURE 9.1A-6. WOLF CREEK REGION 1 k VERSUS RACK CELL PITCH i

9.1A-37

• -qPy - m,w --,-as-,,ww -

---

w-,w -*,,-w- w------e- - - * ----------s-

, ,m---- y-+- --- -- ,si- -- y- y

i 4

0.936 ..

c 0.935 - ..

0.934 - ..

i 0.933~ - -

~

' k=

! 0.932 - ..

! 0.931 - -

=i i

0.930 - ..

~~

0.929  :  :  : l  :  :  : l  :  : : l : : :

O.112 0.116 0.120 0.124 0.128 CELL WALL THICKNESS ( in )

FIGURE 9.1A-7. WOLF CREEK REGION 1 k, VERSUS STAINLESS STEEL WALL THICKNESS 9.1A-38

f .

t i 20

\

l f

/

/

16 ._. .

ll

/

/

14

/

3 i

h@ 12 l

E e

.I

5 j

10 -

t h 8 m _

h D

6 -

4 J LEGEND:

• Inferred from isotopic Data

---

Freehand Fit of Data Previous LEOPARD Unit Cell Calc.

2 E PLG LEOPARD Unit Cell Calc.

0 0 4 8 12 16 20 24 28 l

Burnup (MWD /MTU x 10 3)

FIGURE 9.1A-8. NET DESTRUCTION OF U 235 VERSUS BURNUP IN THE YANKEE ASYMPTOTIC NEUTRON SPECTRUM 9.1A-39

,,,,,_s. . . _ . . - . --- --- - - - - - - - " ' - - - ~ ~ - " ' ' " ~ - ' ~ ~ ~ ~ '

(.

4.0

1. / #

~

3.5 fpr"- ,

s 3.0 N

j a

2.5 g..

.fs j [* '

2 .0 ,

i. ,

.g r

1.5 .  :

• 8 7, .

1.0 LEGEND

f.

• Inferred from itotopic Data

, ==== Freehand Fitof Data Previous LEOPARD Unit Cell Calc.

0.5 .

E PLG LEOPARD Unit CellCalc.

I 1 I I O 28 '

4 8 12 16 20 24 0

Burnup (MWD /MTU x 10 3)

FIGURE 9.1A-9. SPECIFIC PRODUCTION OF U.236 VERSUS BURNUP IN THE YANKEE ASYMPTOTIC NEUTRON SPECTRUM 9.1A-40

22 y I

^

20 18 f

16 I

R 14 g

E m

E e 12

.9 -

U

• 2 .

i 5 10 e i E

Z i R 8 9

3 6

~

LEGEND:

['

>

• Inferred from isotopic Data

~'

---

Freehand Fit of Data

~

Previous LEOPARD Unit Cell Calc.

E PLG LEOPARD Unit Cell Calc.

0 O 4 8 12 16 20 24 28 Burnup (MWD /MTU x 10 3)

FIGURE 9.1A-10. NET DESTRUCTION OF U 238 VERSUS BURNUP IN THE YANKEE ASYMPTOTIC NEUTRON SPECTRUM

, 9.1A-41

- - - - ---.--_,_o- , - , _ _ _ _ _ _ _

9

[

f 8 . .

3 s 6 . ..

W g .

a C

5

.o t

s

1 -

)

E 4 y l  %

2 . ..

N 3

2 I

/

f/, LEGEND:

• Inferred from isotopic Data -

! ---- Freehand Fit of Data 1 '

- Previous LEOPARD Unit Cell Calc.

E PLG LEOPARD Unit Cell Calc.

i I I I I I I O

l 0 4 8 12 16 20 24 28 Burnup (MWD /MTU x 10 3)

FIGURE 9.IA-11. SPECIFIC PRODUCTION OF PU.239 VERSUS BURNUP l IN THE YANKEE ASYMPTOTIC NEUTRON SPECTRUM i

9.1A-42 l .

2.0

/

1.8 - LEGEND; l>

• Inferred from isotopic Data

---

Freehand Fit of Data

,/ [

l j

Previous LEOPARD Unit Cell Calc.

1.6 g /

PLG LEOPARD Unit Cell Calc. /

/

/

1.4 E

E

(

1.2 E

c 0

t 1.0 e bw -

z o 0.8 7

l  ?

, 0.4 .

1 1

l 0.2 g,

0 0 4 8 12 16 20 24 28 Burnup (MWD /MTU x 10 3) 1 FIGURE 9.1A-12. SPECIFIC PRODUCTION OF PU 240 VERSUS BURNUPIN THE YANKEE ASYMPTOTIC NEUTRON SPECTRUM 9.1A-43 en-

Y .

1 10 i i i i -

, LEGEND:

• 1.4 -

• Inferred from isotopic Data ##

---

Freehand Fit of Data Previous LEOPARD Unit Cell Cale.

1.2 -

E PLG LEOPARD Unit Cell Calc.

sw E

j/

F y 1.0 6

i .

c

.9 C /

s 0.8 /

1 I

$ 0.6 5

4 .

I 0.4 I

0.2 0

0 4 8 12 16 20 24 28 Burnup (MWD /MTU x 10 3)

FIGURE 9.IA-13. SPECIFIC PRODUCTION OF PU 241 VERSUS BURNUPIN THE YANKEE ASYMPTOTIC NEUTRON SPECTRUM 9.1A-44

., ,,.,-,_.-.,n..._, -.- -,m,--.--,-. _,

-~ ,

0.28 0.24 LEGEND:

/

• Infstred from isotopic Data .

---

Freehand Fit of Data Previous LEOPARD Unit Cell Calc. N 0.20 '

E PLG LEOPARD Unit Cell Calc.

3>

l E 0.16

$c

.9 t

= 0.12 E

st z

0.08 E .

l , 0.04 g l

n

)

0 4 8 12 16 20 24 28 Burnup (MWD /MTU x 10 3)

FIGURE 9.1A-14. SPECIFIC PRODUCTION OF PU 242 VERSUS BURNUP IN THE YANKEE ASYMPTOTIC NEUTRON 5PECTRUM 9.lA-45 e

. .- d I I I I I I

~

LEGEND:

12 - O Total Pu (Pu-239 + Pu-240 + Pu-241 + Pu-242)

• Fissile Pu (Pu-239 + Pu-241)

--- Freehand Fit of Data F 10 -

Previous LEOPARD 1.Init Cell Calc.

7 s

E PLG LEOPARD Unit CellCalc.

Total Pu

/ -e -M o d p h h o 6 oj _

Fissile Pu L g g. 2 1

4 2

f" 0

0 2 4 6 8 10 12 14 16 18 20 22- 24 26 Burnup (MWD /MTU x 10-3)

FIGURE 9.IA-15. SPECIFIC PRODUCTION OFTOTAt.PU AND FISSILE PU VERSUS BURNUP IN TifE YANKEE ASYMPTOTIC NEUTRON SPECTRUM .

100 LEGEND:

E Calculations l A Saxton Data 10 i 3 3

c 1.0 E

e '

g 7  ;- P U235 8

0.1 , .

e

, , i i

i i 1

i i =

i 7Y' U236 l

M

/

/

.01 /

l 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 MWD /KgM FIGURE 9.1 A-16. ATOM PERCENT OF TOTAL U VERSUS EXPOSURE 9.1A-47

_ . ~ . - - . . . - -

1lfj ll'l s

nt a

i oaD 2

. t 2

l a n D uc tox N l E aCSa G

E

% 1 0 2

L Ee

% 8 1

E R

U

~

6 1

S O

P X

E

_ S U

4 S 1

R E

M V N 2/

1 D g

K O

I T

A R

W M M O 0 T 1 A 8

3 2-U

/

8 9 3

2 u

1

% 6

~P 7

1 A

l 4 9 E

R U

G I

F 2

0 7 6 5 4 0 0 ~

0 0 os-O {<c om3gm=k

.~~r 1

l l , Illll

7 100 6%'**3gsggEPu239:

I

- g_ _ Pu2e 10 w

_s Pu241

_ T. -

s o cw -

g w r 5 j r-l 4 /

o f E 1.0 '

E E __

^ Pute2 o JO 4 n x

l

/

[ /

i g r 0.1 ,e

/ ,

X i av b i f l

~

i  ;  ;

LEGEND:

i E Calculations i

e Sarton Data i

0.01 8 10 12 14 16 18 20 22 24 26 28 MWD /Kg M FIGURE 9.lA.18. ATOM PERCENT OF TOTAL Pu VERSUS EXPOSURE l 9.IA-49

u

-u o

75 ..

70 - -

(

1 .. )

65 -..

C "

O ". .

'i 60 - -

%m E ..

j 55 -- -

50 -..

45 -..

40 l l l l 1E-3 1E-2 1E-1 1 10 Time after Shutdown (years)

FIGURE 9.lA-19. 2200 M/S ABSORPTION CROSS SECTION FOR 4.5 W/0 FUEL AT 45,000 MWD /MTV AS A FUNCTION OF TIME AFTER SHU!DOWN 9.1A-50 . _.

)..s, ,

i t

t

'. h h -

3 S[ Sk U O h'if _

q

\ '

9 36"

.  !. . y. .

N9.236M.236'M i

T N N j 4.216" fuel -

.282". water %

Y \ \

.24" ss

.282" water h N 4.~ 216 " f uel -

U \

W ,,6.-H lil+-

+ .

fuel 4.498" water

.24" ss .

.282" water FIGURE 9.1A-20. P00 CALCULATIONAL MODEL FOR THE WOLF CREEK REGION 2 SPENT FUEL RACKS 9.lA-51

g  :-4 :

0.903 ..

c ..

0.902 -..

0.901 -..

l 0.900 - ..

! koo ..

~

0.899 {

0.898 h ..

l 0.897 - ..

~~

0.896  : l l  : l :  : l  : :

9.176 9.206 9.236 9.266 9.296 CELL PITCH (in )

FIGURE 9.1A-21. WOLF CREEK REGION 2 k. VERSUS RACK CELL PITCH 9.1A-52

?

1 0.903 ..-

- 0.902 - -

I a ..

l 0.901 - ..

0.900 {~-

k*  ::

0.899 - -

j ..

0.898 {

~ ..

0.897 - ..

0.896  :  :  : l  :  :  : l  :  : : l  :  : :

0.112 0.116 0.120 0.124 0.128 CELL WALL THICKNESS (in )

l FIGURE 9.1A-22. WOLF CREEK REGION 2 k. VERSUS STAINLESS STEEL WALL THICKNESS 9.1A-53

.}

l e r ~

0.903 I

70.902 - -

L .. }

0.901 - -

g ..

0.900 - -

t O.899 - -

0.898  : : : : l: ::: l: : : : l: : : : l: : : : .l: : : :

50 75 100 125 150 175 200 POOL TEMPERATURE ( F )

FIGURE 9.1A-23. WOLF CREEK REGION 2 k , VERSUS POOL TEMPERATURE 9.1A-54

1 L.

3 1.000 ..

' O.950 -..

0.900 -

)

0.850 - ..

k,ec 0.800.

1 ..

l

0.750 - ..

i ..

0.700 .

0.650 - ..

0.600 l l  : l  : l  : l  : l  : !: l  : l : l  :

0.00.10.20.30.40.50.60.70.80.91.0 RELATIVE WATER DENSIT(

FIGURE 9.1A-24. WOLF CREEK REGION 2 k. VERSUS WATER DENSITY 9.1A-55

i 1 f

0.990 4.50 w/o 0.965 - -

-0.940 2 -

.( -0.915 - ---------

0.890 2 -

3.80 w/o

~~

3.10 w/o 0.865 - -

2.aow/o 2.to w/o 0.-840 ....e,,,,i,,,,e,,,,i,,,,i,,,,i,,,,i,,,,e,,,,i,,,,

.. i ..

l BURNUP ( MWD /KgU) l

' FIGURE 9.1A-25. WOLF CREEK 2 k. AS A FUNCTION OF FUEL ASSEMBLY BURNUP FOR VARIOUS INITIAL ENRICHMENTS 9.1A-56

T i

i 50 . ....

40 2 ..

Acceptable-n 3

."~

g x 30 2 N ..

o ..

y ..

2 v

g ..

g ..

E

--)

20 ii..

m ..

Unacceptable 10 2..

O" ::::::::: 1:::::::::l:::::::::l:::::::::

1.0 2.0 3.0 4.0 5.0 ENRICHMENT (w/o)

FIGURE 9.1A-26. WOLF CREEK MINIMUM REQUIRED FUEL ASSEMBLY BURNUP AS A FUNCTION OF INITIAL ENRICHMENT FOR STORAGE IN REGION 2 9.1A-57 h ,