ML19332G328

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Criticality Analysis of DC Cook Nuclear Plant Fuel Racks
ML19332G328
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 11/30/1989
From: Boyd W, Fecteau M, Penkrot J
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML17328A232 List:
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NUDOCS 8912210076
Download: ML19332G328 (42)
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+ 1 1 I I 1 l I 1 CRmCAUTY ANALYSIS OF THE 1 DONALD C COOK NUCLEAR PLANT PUEL RACKS j I l lI ll l November 1989. l-I I-W. A. Boyd M. W. Facteau lI J. A. Penkrot W. P. Kovacik J. L Bradfute 'g B. W. Schmidt lg R. C. Cobb W. A. Bordogna lI 'I 'I I

q.y t L. b L-TABLE OF CONTENTS 1.0 Introduction............................................... 1 11.1 Design Description .................................1 1.2 Design Criteria 2 2.0 l Analytical Methods 3 2.1 Criticality Calculation Methodology ......................3 , 2.2 Reactivity Equivalencing Methodology..................... 4 l 3.0 Criticality Analysis of Region 1 Spent Fuel Racks 6 l 3.1-Rsactivity Calculations ..............................6 3.2 Postulated Accidents 8 3.3 Sensitivity Analysis ................................9 4.0 Criticality Analysis of Region 2 Spent Fuel Racks 10 4.1 Reactivity Equivalencing 10 4.2 Peactivity Calculations 10 4.3 Postulated Accidents 13 4.4. Sensitivity' Analysis 13 5.0 Criticality Analysis of Fresh Fuel Racks 14 5.1 Full Density Moderation Analysis 15 5.2 Low Density. Optimum Moderation Analysis 16 6.0 Summary of Criticality Results 18 Bibliography 38 I I Table of Contents 1

q I 1 I 1 I LIST OF TABLES Table

1. Fuel Parameters Employed in Criticality Analysis 19
2. Benchmark Critical Experiments (5,6]

20 Table Table

3. Comparison of PHOENIX lsotopics Predictions to Yankee Core 5 Measurements 21

' Table

4. Benchmark Critical Experiments PHOENIX Comparison 22 Table
5. Data for U Metal and UO Critical Experiments 23 I

g !I !I e ( I l I O lI I List of Tables il .I

r I LIST OF ILLUSTRATIONS Figure

1. Donald C Cook Nuclear Plant Spent Fuel Pool Storage Cell Nominal Dimensions 25 Figure
2. Donald C Cook Nuclear Plant SFP Region 1 Three of Four Fuel I

Assembly Loading Schematic 26 Figure

3. Donald C Cook Nuclear Plant Schematic for SFP Interface Boundary Between Regions 1 and 2 27 i

Figure

4. Donald C Cook Nuclear Plant Fresh Fuel Rack Radial Layout 28 Figure
5. Donald C Cook Nuclear Plant Fresh Fuel Rack Axial Layout 29 I

Figure

6. Sensitivity of K.tv to Enrichment in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading 30 Figure
7. Sensitivity of K.ve to Center-to-Conter Spacing in the Donald C I

Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading 31 Figure

8. Sensitivity of K.ev to B'* Loading in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading 32 Figure
9. Donald C Cook Nuclear Plant SFP Region 2 Fuel Assembly Minimum Burnup vs. Initial U Enrichment Curve 33 Figure 10. Sensitivity of K.ev to Enrichment in the Donald C Cook Nuclear Plant SFP Region 2 Storage Area 34 Figure 11. Sensitivity of K.ev to Center-to-Center Spacing in the Donald C Cook Nuclear Plant SFP Region 2 Storage Area 35 i

Figure 12. Sensitivity of K.ve to B Loading in the Donald C Cook Nuclear Plant SFP Region 2 Storage Area 36 Figure 13. Sensitivity of K.ev to Water Density in the Donald C Cook Nuclear Plant New Fuel Storage Vault 37 List of Illustrations 111

I .I 'I

1.0 INTRODUCTION

{ This report presents the results of the criticality analyses for the storage of Westinghouse 15x15 and 17x17 fuel assemblies in the Donald C Cook Nuclear Plant Spent Fuel Pool (SFP) storage rack and the New Fuel Storage Vault (NFSV). The SFP rack design considered herein is an existing array of Donald C Cook Nuclear Plant SFP poisoned racks, which will be analyzed as two separate spent i fuel arrays or regions. Region 1 will be analyzed for criticality using a three out of four assembly storage arrangement. The Region 1 analysis is presented in Section 3 of this report. Region 2 will be analyzed for criticality with as-sembly storage utilizing all locations. Region 2 will also be analyzed for burnup credit, which takes into consideration the changes in fuel and fission product Inventory resulting from dspletion in the reactor core. The Region 2 criticality . and burnup credit arealysis is presented in Section 4 of this report, Both the Region 1-and 2 analyses are based on maintaining K.et s 0.95 for storage of Westinghouse 15x15 STD and OFA, and 17x17 STD, OFA and VANTAGE 5 fuel. The NFSV rack design considered herein is an existing array of Donald C Cook Nuclear Plant NFSV unpoisoned racks which will be analyzed for criticelity to I show-that Westinghouse 15x15 STD and OFA, and 17x17 STD, O?A and VANTAGE 5 fuel assemblies can be stored using all storage locations. The .g NFSV rcck analysit is based on maintaining K.< $ 0.95 under full wate: density g conditions and $ 0.98 under low water density (optimum morjeration) conditions. The NFSV analysis is presented in Section 5 of this report. I The Westinghouse 15x15 and 17x17 fuel parameters relevant to these analyses are given in Table 1 on page 19. I 1.1 DESIGN DESCRIPTION ~ The Region 1 and 2 spent fuel storage cell design is depicted schematically by Figure 1 on page 25 with nominal dimensions given on the figure. The Region I 1 three out of four storage arrangement is shown in Figure 2 on page 26 and an example of the interf ace boundary between the Region 1 and 2 storage areas is given in Figure 3 on page 27. The total number of SFP locations designated as Region 1 or 2 is left to the utility to determine. The boundary between the two regions can be drawn I Introduction 1

b ~

anywhere within the SFP rocks,- but.the three of four assembly storage ar-(;

r&ngement of the Region 1 area must be carried into the Region 2 area by at least _one row..Therefore, even through Region 2 is analyzed for assembly i k Latorege using all cell locations, some c' ells may need to be left vacant near the T._ Region = 1 to 2 boundary to accomodate the Region 1 pattern carryover by one-row (refer to Figure 3 on page 27), The fresh fuel rock storage rock radial layout is depicted in Figure 4 on page 26 and the axial layout is shown in Figure 5 on page 29, I 1.2 DESIGN CRITERIA 1 ' Criticality of fuel assemblies In a fuel storage rock is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between assemblies. The design basis for preventing criticelity outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level < that the effective neutron multiplication f actor, K.u, of the fuel assembly array _ will be less than 0.95 as recommended in ANSI 57.2-1983, ANSI 57.3-1983 and in Reference 1. The 0.95 K.n limit applies to both the SFP and NFSV under all conditions, except for the NFSV under low water density (optimum moderation) conditions, where the Ken limit is 0.98 as recommended by NUREG-0800. I I i 1 1 ? Introduction 2

t-r L 2.0 ANALYTICAL METHODS 2.1 CRITICALITY CALCULATION METHODOLOGY The criticality calculation method and cross-section values are verified by comparison with critical experiment data for assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps and low moderator densities. The design method which insures the criticality safety of fuel assemblies in the spent fuel storage rack uses the AMPX system of codes for cross-section generation and KENO IV"' for reactivity determination. I The 227 energy group cross-section library that is the common starting point for all cross-sections used for the benchmarks and the storage rack is generated from ENDF/B-V* data. The NITAWL* program includes, in this library, the self-shisided resoriance cross-sections that are appropriate for each particular geometry. The Nordheim Integral Treatment is used. Energy and spatial weighting of cross-sections is performed by the XSDRNPM* program which is [ a one-dimensional Sn transport theory code. These multigroup cross-section sets i are then used as input to KENO IV"' which is a three dimensional Monte Carlo theory program designed for reactivity calculations. A set of 33 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and uncertainty. The experiments range from water moderated, oxide fuel arrays separated by various rnatorials (B4C, steel, water, etc) that simulate LWR .I fuel shipping and storage conditions

  • to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials * (Plexiglas and alr) that demonstrate the wide range of applicability of the method. Table 2 on page 20 summarizes these experiments.

The average K.ve of the benchmarks is 0.992*. The standard deviation of the I bias value is 0.0008 Ak. The 95/95 one sided tolerance limit factor for 33 values is 2.19. Thus, there is a 95 percent probability with a 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0018 Ak. Analytical Methods 3

s /' 2.2 REACTIVITY EQUIVALENCING METHODOLOGY E L_ Spent fuel storage, in the Region 2 spent fuel storage racks, is achievable by means of the concept of reactivity equivalencing. The concept of reactivity [ equivalencing is predicated upon the reactivity decrease associated with fuel L depletion. A series of reactivity calculations are performed to generate a set of enrichment-fuel assembly discherge burnup ordered pairs which all yield the equivalent K.et when the fuel is stored in the Region 2 racks. The data points on the reactivity equivalence curve are generated with a trans-port theory computer code, PHOENIX *. PHOENIX is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code. A 25 energy group nuclear data library based on a modified version of the British WIMS* library is used with PHOENIX. A study was done to examine fuel reactivity as a function of time follo'ning discharge from the reactor. Fission product decay was accounted f ar using CINDER"*. CINDER is a point-depletion computer code used to determire fission 1 product activities. The fission products were permitted to decay for 30 years af ter discharge. The fuel reactivity was found to reach a maximum at approx-imately 100 hours after discharge. At this point in time, the enajor fission I product poison, Xe'", has nearly completely decayed away. Furthermore, the fuel rsectivity was found to decrease continuously from 100 hours to 30 years I following discharge. Therefore, the most reactive point in time for a fuel as-sembly after discharge from the reactor can be conservatively approximated by removing the Xe'". The PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been examined following discharge from a reac-j tor. In addition, an extensive set of benchmark critical experiments has been p analyzed with PHOENIX. Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 3 on page 21. The measurements were made on fuel discharged from Yankee Core 5"". The data in Table 3 on page 21 shows that the agreement between PHOENIX pred-ictions and measured isotopic compositions is good. The agreement between reactivities computed with PHOENIX and the results of 81 critical benchmark experiments is summarl:ed in Table 4 on page 22. Key I parameters describing each of the 81 experiments are given in Table 5 on page

23. These reactivity comparisons again show good agreement between exper-

. iment and PHOENIX calculations. An uncertainty associated with the burnup-dependent reactivities computed with PHOENIX is accounted for in the development of the Region 2 burnup require-m ents. A bias which increases linearly with burnup to 0.01 Ak at 30,000 MWD /MTU is applied to the PHOENIX calculational results. This bias is con-sidered to be very conservative since comparison between PHOENIX results and Analytical Methods 4

,;~.- I r' - the : Yankee Core experiments and 81 benchmark experiments' indicates' closer i greement (see Table 3 on F.upe 21 and Table 4 on page.22). For the Donald n C Cook Nuclear Plant SFP Region ~ 2 analysis, the PHOENIX calculations for the I maximum burnup of 5,550 MWDIMTU include a reactivity bias of 0.0019 Ak. I t I. t. l !I t I II ll-I lI lI I il lI I: g Analytical Methods 5 e e mywm-ws-wv- -e n -s: pew ess w---yweem. ,y gw -. ,we-gy +y-- - - -T+wty-y -'w'w --w v' vp'r I I I I 3.0 CRITICALITY ANALYSIS OF REGION 1 SPENT FUEL RACKS This section develops and describes the analytical assumptions and models employed to perform the criticality analyses for storage of spent fuel in Fiegion 1 of the Donald C Cook Nuclear Plant SFP. h 3.1 REACTIVITY CALCULATIONS The following assumptions were used to develop the nominal case KENO model for the Region 1 SFP rack storage of fresh fuel using three out of four storage locations as shown in Figure 2 on page 26. 1. The Westinghouse 17x17 OFA fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burnable absorber in the fuel rods or any natural enrichment axial I blankets (See Table 1 on page 19 for fuel parameters). Evaluation of the Westinghouse 15x15 and 17x17 fuel assemblies shows that the 17x17 OFA

g.

assembly is the most reactive fuel type when all assemblies have the same ,g enrichment. Therefore, only the Westinghouse 17x17 OFA fuel assembly L was analyzed.- 2. All fuel rods contain uranium dioxide at an enrichment of 4.95 w/o U'" over I the finite 144 inch length of each rod. The fuel pellets are assumed to ( be at 96% of theoretical density, and no credit is taken for dishing or l chamfering. If nominal theoretical density and pellet parameters were used, the resultant enrichment limit would be 5.06 w/o U'". Therefore, the 4.95 w/o U'" enrichment limit can be considered a nominal enrichment limit since the conservative assumptions employed in the pellet modelling bound the standard 0.05 w/o enrichment tolerance. 3. No credit it. taken for any U'" or U'" in the fuel, nor is any credit taken for the build up of fission product poison material. l 4. The moderator is pure water at a temperature of 68'F. A conservative value l of 1.0 gm/cm' is used for the density of water. 5. No credit is taken for any spacer grids or spacer sleeves. I Criticality Analysis of Region 1 Spent Fuel Racks 6 I. w g = my y -,p w e- ,y- - - - - w a w c w wt .--a r -=w

P L: 6. Fuel assemblies are loaded into three of every four cells in a checkerboard r1 pattern in the storage cells as shown in Figure 2 on page 26. L 7. The array is infinite in lateral extent and finite in axial extent which allows neutron leakage from only the axial direction, { 8. The minimum poison material loading of 0.02 grams B'" per square centi-meter is used throughout the array. The KENO calculation for the nominal case resulted in a K.ev of 0.9005 with a 95 percent probability /95 percent confidence level uncertainty of 10.0065. The nominal case result can be compared to the worst case result to determine the relativo impact of applying the worst case assumptions. The nominal case is also used as the center point for the sensitivity analyses discussed in Section 3.3. The maximum K.et under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manufacturing I process in addition to asymmetric positioning of fuel assemblies within the storage cells. Westinghouse internal studies of asymmetric positioning of fuel assemblies within the storage cells have shown that symmetrically placed fuel assemblies yield equal or conservative results in rack K.es. The sheet metal tolerances are considered along with construction tolerances related to the cell I.D., and cell center-to-center spacing. For the Region 1 racks this resulted in a reduction of the nominal center to center spacings to their minimum values. Thus, the " worst case" KENO model of the Region 1 storage racks contains the minimum center to center spacings with symmetrically placed fuel assemblies. Based on the analysis described above, the following equation is used to de-velop the maximum K.tv for the Donald C Cook Nuclear Plant Region 1 spent I fuel storage racks with three out of four storage: K.e# = K orei + Bm.inoa + Be.ri + (( (ks)*. orsi + (ks)'m.inoo 3 where: K orei = worst case KENO K.ve that includes material I tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal Bm.inoo method bias determined from benchmark critical = comparisons Criticality Analysis of Region 1 Spent Fuel Racks 7

e 1 bles-to account for poison particle B,.rt = L swif-shleiding. This standard term accounts for the increased neutron transmission through the poison plate due to the inherent effects of poison particle self-shielding, and has been analytically determined for poison plates similar to those used in this analysis, 95/95 uncertainty in the worst case KENO K.vt k s.or. = 95/95 uncertainty in the method bias ks tn.e = Substituting calculated values in the order listed above, the result is: K.et. = 0.9308 + 0.0083 + 0.0014 + (((0.0046)* + (0.0018)' ) = 0.9454 Since K.et is less than 0.95 including uncertaintles at a 95/95 probability / confidence ' level, the acceptance criteria for criticality is met with fuel enriched to a nominal 4.95 w/o. l 3.2 POSTULATED ACCIDENTS Most accident conditions will not result in an increase in K.et of the rack. Ex-I amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rock (the rack structure-partinent for _ criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active ) .. fuel height of stored assemblies which precludes interaction. However, accidents can be postulated which would increase reactivity (i.e., dropping a fuel assembly between the rack and pool wall). For these accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This . states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by about 0.25 AK. Thus, for postulated accidents, should there be a reactivity increase, K tv would be less than or equal to 0.95 due to the effect of the dissolved boron. Since the Donald C Cook Nuclear Plant.SFP will be maintained at a boron concentration of 2400 ppm, additional margin will exist to the 0.95 limit. Criticality Analysis of Region 1 Spent Fuel Racks 8

l 3.3 SENSITIVITY ANALYSIS To show the dependence of Ken on fuel and storage cells parameters as re-I quested by the NRC"', the variation of the Kere with respect to the following parameters was developed using the KENO computer code: 1. Fuel enrichment, with a 0.50 w/o U'" delta about the nominal case enrichment. 2. Center-to-center spacing of storage cells, with a half inch delta about the nominal case center-to-center spacing. I 3. Poison loading, with a 0.01 gm-B/cm2 delta about the nominal case polson loading. .g Results of the sensitivity analysis for the Region 1 storage cells are shown in Im Figure 6 on par

  • 30 through Figure 8 on page 32 for three of four storage.

I I I I I lI .I I l Criticality Analysis of Region 1 Spent Fuel Rocks 9 i --,___--.,--.,.--.na-,-.-,.w--n..,. .--m.- -n-,---, awe a e

I I I I l 1 4.0 CRITICALITY ANALYSIS OF REGION 2 SPENT FUEL RACKS This section develops and describes the analytical techniques and models em-ployed to perform the criticality analyses for storage of spent fuel in Region 2 of the Donald C Cook Nuclear Plant spent fuel pool. 4.1 REACTIVITY EQUIVALENCING Spent fuel storage, in the Region 2 spent fuel storage racks, is achievable by I means of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with fuel j depletion. A series of reactivity calculations are performed to generate a set

u of enrichment-fuel ascembly discharge burnup ordered pairs which all yltid the equivalent Keet when the fuel is stored in the Region 2 rocks.

Figure 9 on page 33 shows the constant Ke t contour generated for the Donald e C Cook Nuclear Plant Region 2 racks. Note in Figure 9 on page 33 the endpoint at 0 MWDIMTU where the enrichment is 3.95 w!o and at 5,550 MWD /MTU where the enrichment is 4.95 w/o. The interpretation of the endpoint data is es fol-lows: the reactivity of the Region 2 racks containing fuel at 5,550 MWD /MTU burnup which had an initial nominal enrichment of 4.95 w/o is equivalent to the reactivity of the Region 2 racks containing fresh fuel having an initial nominal enrichment of 3.95 w/o. It is important to recognize that the curve in Figure 9 on page 33 is based on a constant rack reactivity for that region and not on a constant fuel assembly reactivity. 4.2 REACTIVITY CALCULATIONS The maximum K.,, for storage of spent fuel in Region 2 is determined using the methods described in Section 2. Figure 9 on page 33 represents combinations of fuel enrichment and discharge burnup yielding the same rack multiplication I f actor (K vi) as the. enrichment of 3.95 w/o U'" at zero burnup. This curve was obtained by first calculating the equivalent reactivity points using PHOENIX and then normalizing the points to the KENO calculation for fresh fuel with a nom-I inal enrichment of 3.95 w/o U'". I Criticality Analysis of Region 2 Spent Fuel Racks 10 I 4

The following assumptions were used to develop the nominal case KENO model for the Region 2 storage of spent fuel: 1. The Westinghouse 17x17 OFA fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burnable absorber in the fuel rods or any natural enrichment axlal i I blankets (See Table 1 on page 19 for fuel parameters). Evaluation.of the Westinghouse 15x15 and 17x17 fuel assemblies shows that the 17x17 OFA assembly is the most reactive fuel type when all assemblies have the same I enrichment. Therefore, only.the Westinghouse 17x17 OFA fuel assembly was analyzed. 2. All fuel rods contain uranium dioxide at an enrichment of 3.95 w/o U"' over the finite 144 inch length of each rod. The fuel pellets are assumed to be at 96% of theoretical density, and no credit is taken for dishing or chamfering. If nominal theoretical density and pellet parameters were used, the resultant enrichment limit would be 4.04 w/o U"'. Theref ore, the 3.95 w/o U'" enrichment limit can be considered a nominal enrichment limit since the conservative assumptions employed in the pellet modelling bound the standard 0.05 w/o enrichment tolerance. 3. No credit is taken for any U"' or U"' in the fuel, nor is any credit taken for the build up of fission product poison material. 4. The moderator is pure water at a temperature of 68'F. A conservative value of 1.0 gm/cm' is used for the density of water. 5. No credit is taken for any spacer gt!ds or spacer sleeves, 6. Fuel assemblies are loaded into three of every four cells in a checkerboard i pattern in the storage cells as shown in Figure 2 on page 26. 7. The array is infinite in lateral extent and finite in axial extent which allows neutron leakage from only the axial direction. 8. The minimum poison material loading of 0.02 grams B" per square centi-meter is used throughout the array. l l The KENO calculation for the nominal case resulted in a Keve of 0.9141 with a lg 95 percent probability /95 percent confidence level uncertainty of 10.0049. The l3 nominal case result can'be compared to the worst case result to determine the relative impact of applying the worst case assumptions. The nominal case is l' also used as the center point for the sensitivity analyses discussed in Section 4.3. The maximum K.tv under normal conditions was determined with a " worst case" KENO model, in the same manner as for the Region 1 storage racks (see Section 3). For the Region 2 racks, the cell center to center spacings are reduced from the nominal value to their minimum value. Thus, the " worst case" KENO model Criticality Analysis of Region 2 Spent Fuel Racks 11 I

t of the Region 2 storage racks contains minimum cell center to center spacings with symmetrically placed fuel assemblies. The uncertainty associated with the L reactivity equivalence methodology was included in the development of the burnup requirements. This uncertainty was discussed in Section 2.2. Based on the analysis described above, the following equation is used to de-velop. the maximum Keet for the storage of spent fuel in the Donald C Cook Nuclear Plant Region 2 spent fuel storage racks: (ks)'m.ines ) Bm.inee + Spri + /( (ks)' or K.ev = Kweest + + where: K or.i - worst case KENO K.ev that includes material = tolerances, and mechanical tolerances which can result in spacings between assemblies less than I nominal Bmeinee = method bias determined from benchmark critical comparisons Spri bias to account for poison particle = I self-shielding. This standard term accounts for the increased neutron transmission through the poison plate due to the inherent effects of poison particle self-shleiding, and has been analytically determined for poison plates similar to those used in this analysis, ' ks. orsi = 95/95 uncertainty in the worst case KENO K.ev ksmeinoe = 95/95 uncertainty in the method bias 5 Substituting calculated values in the order listed above, the result is: K.es = 0.9327 + 0.0083 + 0.0014 + /[(0.0045)* + (0.0018)' ) = 0.9472 The maximum K.tv for Region 2 for this configuration is less than 0.95, including all uncertainties at a 95/95 probability / confidence level. Therefore, the accept-ance criteria for criticality are met for storage of spent fuel at an equivalent fresh fuel nominal enrichment of 3.55 w/o U'". Criticality Analysis of Region 2 Spent Fuel Racks 12 a n s O/

I l 4.3 POSTULATED ACCIDENTS Most accident conditions will not result in an incrasse in K.et of the rack. Ex-amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rock'(the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel height of stored assemblies which precludes interaction). l However, accidents can be postulated which would increase reactivity (i.e., misloading an assembly with a burnup and enrichment combination outside of . the acceptable area in Figure 9 on page 33, or dropping a fuel assembly be-I. tween the rock and pool wall). For these accident conditions, the double con-tingency principle of ANSI N16.1-1975 is applied. This states that one is not i required to assume two unlikely, Independent, concurrent events to ensure pro-I tection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. The presence of approximately 2000 ppm boron in the pool water will decrease t ~ g reactivity by about 0.25 AK. Thus, for postulated accidents, should there be a ,3 reactivity increase, K.ve would be less than or equal to 0.95 due to the effect of the dissolved boron. Since the Donald C Cook Nuclear Plant SFP will be

H

= aa'd * * ' a e a c'" ' 2400 ppm, additionai margin wiii exist su to the 0.95 limit. 4.4 SENSITIVITY ANALYSIS To show the dependence of Kett on fuel and storage cells parameters as re- [ quested by the NRC*, the variation of the K.tv with respect to the following parameters was developed using the PHOENIX computer code: 1. Fuel enrichment, with a 0.50 w/o U* delta about the nominal case l-enrichment. 2. Center-to-center spacing of storage cells, with a half inch delta abo 0t the nominal case center-to-center spacing. 3. ' Poison loading, with a 0.01 gm-B*/cm2 delta about the nominal case poison loading. Results of the sensitivity analysis for the Region 2 storage cells are shown in Figure 10 on page 34 thmugh Figure 12 on page 36 for spent fuel occupying l every cell in the Region 2 fuel racks. 1 l l l Criticality Anslysis of Region 2 Spent Fuel Racks 13

i 5.0 CRITICALITY ANALYSIS OF FRESH FUEL RACKS This section describes the analytical techniques and models employea to per- [ form the criticality analysis for storage of fresh fuel in the Donald C Cook Nuclear Plant New Fuel Storage Vault (NFSV). I Since the fresh fuel racks are maintained in a dry condition, the criticality analysis will show that the rack Ken is less than 0.95 for the full water density condition and less than 0.98 for the low water density (optimum moderation) conditions. The criticality methodology employed in this analysis is discussed In Section 2 of this report. The following assumptions were used to develop the nominal case KENO model for the storage of fresh fuel in the Donald C Cook Nuclear Plant NFSV under full density and low density optimum moderation conditions: 1. The fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burriable polson in the fuel rods or any natural enrichment axial blankets. 'E 2. All fuel rods contain uranium dioxide at an enrichment of 4.55 w/o U'" over 'W the infinite length of each rod. The fuel pellets are assumed to be at 96% of theoretical density, and no credit is taken for dishing or chamfering. If nominal theoretical density and pellet parameters were used, the resultant l enrichment limit would be 4.65 w/o U'" Theref ore, the 4.55 w/o U'" enrichment limit can be considered a nominal enrichment limit since the conservative assumptions employed in the pellet modelling bound the l standard 0.05 w/o enrichment tolerance, l 3. No credit is taken for any U"' or U'" in the fuel, nor is any credit taken for the build up of fission product poison material. 4. No credit is taken for any spacer grids or spacer sleeves. I Criticality Analysis of Fresh Fuel Racks 14 l 1

I I l 5.1 FULL DENSITY MODERATION ANALYSIS in the nominal case KENO model for the full density moderation analysis, the moderator is pure water at a temperature of 68'F. A conssrvative value of 1.0 gm/cm' is used for the density of water. The fuel array is infinite in lateral l and axial extent whleh precludes any neutron leakage from the array. This 2D single cell modelling technique is conservative, however, the overall reactivity effect of neutron leakage from the array under full moderator density conditions is s m all. Calculations for the Donald C Cook Nuclear Plant NFSV array show I total leakage effects to be worth only 0.005 AK Fuel rock calculations have i shown that the Westinghouse 17x17 OFA fuel assemblies are more reactive than i I the other fuel types when all fuel assemblies have the same U'" enrichment. ~ Thus, only the Westinghouse 17x17 OFA fuel assembly was analyzed. The maximum K.et under normal conditions arises from consideration of me-I chenical and material thickness tolerances resulting from the manufacturing process in addition to asymmetric positioning of fuel assemblies within the storage cells. Studies of asymmetric positioning of fuel assemblies within the storage cells has shown that symmetrically placed fuel assemblies yield con-servative results in rack K.et. Since the Donald C Cook Nuclear Plant NFSV rack structure consists of part length angle Irons, all of the structural steel was Thus, the most conservative, or " worst conservatively lef t out of the model. case", KENO model of the fresh fuel storage racks contains no structural steel with symmetrically placed fuel assemblies. Based on the analysis described above, the following equation is used to de-I velop the maximum K et for the Donald C Cook Nuclear Plant New Fuel Storage Vault: K.tv. K ori + Bm.inoe * / ((ks)'.... (ks)'m.inov ) + whert: K orei = worst case KENO K.e# that includes material l tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal Bm.inoe = method bias determined from benchmark critical comparisons k s..,.i = 95/95 uncertainty in the worst case KENO Keet I I l Criticality Analysis of Fresh Fuel Racks 15 v 9>- - -.,, - -p 7.,+w = w vp.-wde yim sr e__,,m t emy--- T='"T- "'---h

c_ ~ i 95/95 uncertainty in the method bias ksmetnee = Substituting calculated values in the order listed above, the result is: ) Keet = 0.9324 + 0.0083 + /[(0.0086)* + (0.0018)' ) = 0.9495 Since K.tv is less than 0.05 including uncertainties at a 95/95 probability confi-dence level, the acceptance criteria for criticality is met. 5.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS in the low density optimum moderation analysis, the fuel array is finite in the radial and axial extent. The nominal model described above is used in KENO except that the concrete walls and floor are explicitly modelled as shown in I Figure 4 on page 28 and Figure 5 on page 29. The Westinghouse 17x17 STD fuel assembly was analyzed in the model (See Table 1 on page 19 for fuel parameters). Calculations have shown that the 17x17 STD fuel assembly is more I reactive than other fuel assemblies under low moderator density conditions. l Analysis of the Donald C Cook Nuclear Plant racks has shown that the maximum rock K.ee under (ow density moderation conditions occurs at 0.045 gm/cm' water density. The KENO calculation of the Donald C Cook Nuclear Plant NFSV at O.045 gm/cm** water density resulted in a peak Keet of 0.8817 with a 95 percent probability and 95 percent confidence level uncertainty of 10.0072. Figure 13 L on page 37 shows the NFSV reactivity as a function of the water density. Based o'1 the analysts described above, the following equation is used to de-velop the maximum K.tv for the Donald C Cook Nuclear Plant fresh fuel storage racks under low water density optimum moderation conditions: K.et = Km...

  • Bm inas * (((ks)'o... + (ks)'m.inoa 3 where:

maximum KENO K et with low density optimum Km... = moderation method bias determined from benchmark critical Bm.ineo = comparisons 95/95 uncertainty in the base case KENO K.vi kso... = I . I I ' ' ' " " " ' " ^ " * ' ' ""'" " '" I

ks mos = 95/95 uncertainty in the method bias l Substituting calculated values in the order listed above, the result is: K.tv = 0.8817 + 0.0083 + /[(0.0072)' + (0.0018)* ] = 0.8974 Since K.tv is less than 0.98 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met, t I I I I I

I I

I' I

I l

I t Criticality Analysis of Fresh Fuel Racks 17

i l I 6.0

SUMMARY

OF CRITICALITY RESULTS The acceptance criteria for criticality requires the effective neutron multipil-cation f actor, Keit, to be less than or equal to 0.95, including uncertalntles, under all conditions for the storage of fuel assemblies in the Spent Fuel Pool (SFP). l For the storage of fuel assemblies in the New Fuel Storage Vault (NFSV), the K.e, must be less than or equal to 0.95, including uncertaintles, under flooded conditions, and less tnan or equal to 0.98, including uncertaintles, under optimum I moderation conditions. This report shows that the acceptance criteria for criticality is met for the Donald C Cook Nuclear Plant Spent Fuel Pool (SFP) and New Fuel Storage Vault (NFSV) for the the storage of Westinghouse 15x15 and 17x17 STD, OFA and VANTAGE 5 fuel assemblies with the following nominal enrichment limits: SFP Region 1 s 4.95 w/o U'" SFP Region 2 s 4.95 w/o U'", with burnup restrictions given by Figure 9 on page 33 NFSV 4.55 w/o U'" The analytical methods employed herein conform with ANSI N18.2-1973, "Nu-I clear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Henaling System; ANSI 57.2-1983, " Design Objectives I for LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2: ANSI N16.9-1975, " Validation of Calculational Metho'ds for Nuclear Criticality Safety"; NRC Standard Review Plan, Section 9.1.2, " Spent Fuel Storage"; and I ANSI 57.3-1983, " Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants." I I I I Summary of Criticality Results 18

l I Toble 1. Fuel Parameters Employed in Crit 6calhy Analysis i Parameter W 15x15 W 17x17 W 17x17. I STD & OFA OFA & VS STD I Number of Fuel Rods per Assembly 204 264 264 I Rod Zlre-4 Clad 0.D. (Inch) 0.422 0 360 0 374 Clad Thickness (Inch) 0.0243 0.0223 0.0225 Fuel Pellet 0.D. (Inch) 0 3659 0 3088 0 3225 ) I \\ i Fuel Pellet Density (% of Theoretical) 96 96 96 l l Fuel Pellet Olshing Factor 0.0 0.0 0.0 Rod Pltch (Inch) 0 563 0.496 0.496 Number of Zlre-4 Guide Tubes 20 24 24 L Guide Tube 0.D. (Inch) 0 533 0.474 0.482 lI Guide Tube Thickness (Inch) 0.017 0.016 0.016 i Number of instrument Tubes 1 1 1 instrument Tube 0.D. (Inch) 0 533 0.474 0.484 lm Instrument Tube Thickness (Inch) 0.017 0.016 0.016 il r !I 19

{ Tow 2. senchmark critical Experiments [s,s) i {- General Ene1chment Separating Soluble Descetption w/o U235 Deflector Mateetal ............................................................. Boron ppm Keff

t. U02 rod lattice 2.46 water water 0

0.9057 */-.0028 1

2. UO2 rod lattice 2.46 water water 1037 0.9906 +/-.0018
9. U02 rod lattice 2.46 water water 764 0.9896 +/-.0015
4. UO2 rod lattice 2.46 water 84C pins 0

0.9914 +/-.0025

5. U02 rod lattice 2.46 water 54C pins 0

0.9001 +/-.0026

8. UO2 rod lattice 2.46

-water 84C pina 0 0.9955 +/..0020

7. UO2 rod lattice 2.46 water B4C pins 0

0.9889 +/-.0027 lI

8. U02 rod lattice 2.46 water 84C pins 0

0.9983 +/-.0025

9. UO2 rod lattice 2.46 water water 0

0.9938 +/-.0028

10. UO2 rod lattice 2.46

-water water 143 0.9928 +/-.0025 1

11. UO2 rod lattice 2.46 water statniess steel 514 0.9947 +/-.0020
12. UO2 rod lattice 2.46 water stainless steel 217 0.9943 +/-.0019
13. UO2 rod lattice 2,46 Water borated aluminum 15 0.9692 */=.0023
14. U02 rod lettice 2.46 water borated aluminum 92 0.9884 +/-.0023
15. f)02 rod lattice 2.46 water borated aluminum 395 0.9832 +/-.0021
16. U02 rod lattice 2.46 water borated aluminum 121 0.9848 +/-.0024
17. U02 rod lattice 2.46 wetse borated aluminum 487 0.9495 +/*.0020
10. UO2 rod lattice 2.46 water borated aluminum 197 0.9885 +/*.0022

(:

19. U02 rod lattice 2.46 water borated aluminum 634 0.9921 */-.0019 -
20. UO2 rod lattice 2.46 water borated aluminum 320 0.9020 +/-.0020
21. UO2 rod lattice 2.46 water borated aluminum 72 0.9939 +/-.0020
22. U metal cylinders 93.2

-bare air 0 0.9905 +/-.0020 r I.'

23. U metal cylinders 93.2 bare air 0

0.9975 +/-.0020

24. U metal cylinders 93.2 bare air 0

0.9947 +/-.0025 25.' U metal cylinders 93.2 bare air 0 0.9928 +/-.0019

26. U metal cyttnders 93.2 bare air 0

0.9922 +/.0026

27. U metal cylinders 93.2 bare ate 0

0.9950 +/-,0027

28. U metal cylinders 93.2 bare plexiglass 0

0.9949 +/ .0030 1

29. U metal cylinders 93.2 paraffin plexiglass 0

0.9928 +/-.0041

30. U metal cyll'Wers 93.2 bare plexiglass 0

0.9968 +/..0018 [

31. U metal cyttnders 93.2 paraffin plextglass 0

1.0042 +/-.0019

32. U metal cylinders 93.2 paraffin plexiglass 0

0.9963 +/-.0030

33. U metal cyttnders 93.2 paraffin pleutglass 0

0.9919 +/-.0032 4 I I s I 20

I I Table 3. Comparison of PHOENIX isotopics Predictions to Yankee Core 5 Measurements 1 I 1 Quantity (Atom Ratio) % Difference U235/U -0.67 i U236/U -0.28 U238/U -0.03 PU239/U +3.27 PU240/U +3.63 PU241/U -7.01 PU242/U -0.20 ) PU239/U238 + 3.24 Mass (PU/U) 41.41 I FISS-PU/TOTfU -0.02 l I g t 9 i 21

. ~,. ) I Table 4. Benchmark Critical Experiments PHOENIX Comparison i Description of Number of PHOENIX K.n Using Experiment Experiments Experiments Bucklings UO: ] Al clad 14 0.9947 SS clad 19 0.9944 Borated H 0 7 0.9940 Subtetal 40 0.9944 1 U-Metal l Al clad 41 1.0012 i l TOTAL 81 0.9978 ) i I I I I l l l lI 22

l Table 5. Data for U Metal and 00: Crhical Experiments (Part 1 of 2) I I I Fuel Pellet Clad Clad Lattice ' Case Cell A/D H20/U Density Otameter Motorial 00 7htekness P1 ten Boron Nuntper 7ype U-235 Ratio (G/CC) (CM) Clad (CM) (CM) (CM) PPM t Hexa 1.328 3.02 7.53 1.5265 Aluminum 1.6916 .071 0 2.2050 0.0 2 Hexa 1.328 3.95 7.53 1.5265 Aluminum 1.6916.07110 2.3590 0.0 3 Hexa 1.328 4.95 7.53 1.5265 Aluminum 1.6916.07110 2'.5120 0.0 4 Hexa 1.328 3.92 7.52 .9855 Aluminum 1.1506.07110 1.5580 0.0 5 Hexa 1.328 4.89 7.52 .9855 Aluminum 1.1506.07110 1.6520 0.0 I 6 Hexa 1.328 2.88 10.53 .9728 Aluminum 1.1506 .0711) 1.5580 0.0 7 Hexa 1.328 3.58 10.53 .9728 Aluminum 1.1506.07110 1.6520 0.0 8 Hexa 1.328 4.83 10.53 .9728 Aluminum 1.1506.07110 1.8060 0.0 9 Square 2.734 2.18 10.18 7620 55-304 .8594 .04085 1.0287 0.0 10 Square 2.734 2.92 10.18 7620 55-304 .8594 .04085 1.1049 0.0 11 Square 2.734 3.86 10.18 .7620 55-304 .8594 .04085 1.1938 0.0 12 Square 2.734 7.02 10.18 .7620 $$ 304 .8594 .04085 1.4554 0.0 13 Square 2.734 8.49 10.18 .7620 $5-304 .8594 .04085 1.5621 0.0 to Square 2.734 10.38 10.18 .7620 $5-304 .8504 .04085 1.6891 0.0 15 Square 2.734 2.50 10.18 7620 $$ 304 .8594 .04085 1.0617 0.0 14 Square 2.734 4.51 10.18 .7620 55 304 .8594 .04045 1.2522 0.0

1 17 Square 3.745 2.50 10.27

.7544 $$ 304 .8600 .04060 1.0617 0.0 18 Square 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 0.0 19 Square 3.745 4.51 10.37 .7544 55-304 .8600.04060 1.2522 0.0 20 Square 3.745 4.51 10.37 .7544 55-304 .8600.04060 1.2522 456.0 [I 21 Square 3.745 4.51 10.37 .7544 55-304 .8600..04060 1.2522 709.0 22 Square 3.745 4.51 10.37 7544 55-304 .8600.04060 1.2522 1260.0 23 Square 3.745 4.51 10.37 .7544 $5-304 .8600.04060 1.2522 1334.0 24 Square 3.745 4.51 10.37 .7544 $$-304 .8600 .04060 1.2522 1477.0 25 Square 4.069 2.55 9.46 1.1278 $5 304 1.2090.04060 1.5113 0.0 26 Square 4.069 2.55 9.46 1.1278 $$-304 1.2090.04060 1.5113 3392.0 I 27 Square 4.069 2.14 9.46 1.1278 55-304 1.2090.04060 1.4500 0.0 28 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060.08130 1.5113 0.0 29 Square 3.037 2.64 9.28 1.1268 $$-304 1.1701 .07163 1.5550 0.0 30 Square 3,037 8.16 9.28 1.1268 55-304 1.2701 .07163 2.1980 0.0 31 Square 4.049 2.59 9.45 1.1268 55-304 1.2701 .07163 1.5550 0.0 I 32 Square 4.069 3.53 9.45 1.1268 55-304 1.2701 .07163 1.6840 0.0 33 Square 4.069 8.02 9.45 1.1268 55-304 1.2701 .07163 2.1980 0.0 34 Square 4.069 9.90 9.45 1.1268 55-304 1.2701 .07163 2.3810 0.0 35 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060.08130 1.5113 1677.0 36 Hexa 2.096 2.06 10.38 1.5240 Aluminum 1.6916 .07112 2.1737 0.0 l 37 Hexa 2.096 3.03 10.38 1.5240 Aluminum 1.6916 .07112 2.4052 0.0 34 Hexa 2.096 4.12 10.36 1.5240 Aluminum 1.6916.07112 2.6162 0.0 39 Hexa 2.096 6.14 10.38 1.5240 Aluminum 1.6916.07112 2.9891 0.0 40 Hexa 2.096 8.20 10.38 1.5240 Aluminum 1.6914.07112 3.3255 0.0 41 Hexa 1.307 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 I 42 Hexa 1.307 1.51 18.90 1.5240 Aluminum 1.6916.07112 2.4054 0.0 43 Hexa 1.307 2.02 18.90 1.5240 Aluminum 1.6916.07112 2.6162 0.0 I I l I

s N Table 5. Data for U Metal and UO: Critical Experiments (P8rt 2 of 2) L I I Fuel Pellet 01ad Clad Lattice case Cell A/O H20/U Donetty Otameter Matertal 00 7htekness Pitch Boron Nunter Type U-239 Ratio (G/CC) (CN) Clad (CN) (CN) (CM) PPM 44 Hexa 1.307 3.01 18.90 1.5240 Aluminum 1.4916.07112 2.9896 0.0 I 46 Hexa 1.307 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 46 Hexa 1.140 1.01 18.90 1.5240 Aluminum 1.6916.07112 2.1742 0.0 47 Hexa 1.160 1.51 18.90 1.5240 Aluminum 1.6916.07112 2.4064 0.0 48 Hexa 1.140 2.02 18.90 1.5240 Aluminum 1.4916.07112 2.6162 0.0 49 Hexa 1.160 3.01 18.90 1.5240 Aluminum 1.6916.07112 2.9896 0.0 1 50 Hexa 1.140 4.02 18.90 1.5240 Aluminum 1.6816.07112 3.3249 0.0 51 Hexa 1.040 1.01 18.90 1.5240 Aluminum 1.6916.07112 2.1742 0.0 92 Hexa 1.040 1.51 18.90 1.5240 Aluminum 1.6816 .07112 2.4054 0.0 53 Hexa 1.040 2.02 18.90 1.5240 Aluminum 1.6816 .07112 2.4162 0.0 54 Hexa 1.040 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 l SS Hexa 1.040 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 54 Hexa 1.307 1.00 18.90 .9830 Aluminum 1.1506.07112 1.4412 0.0 57 Hexa 1.307 1.52 18.90 .9830 Aluminum 1.1506.07112 1.5926 0.0 58 Hexa 1.307 2.02 18.90 .8830 Aluminum 1.1506.01112 1.7247 0.0 SS Hexa 1.307 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609' O.O i l SO Hexa 1.307 4.02 18.90 .9830 Aluminum 1.1506.07112 2.1742 0.0 61 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506.07112 1.5926 0.0 62 Hexa

1. ISO 2.02 18.90

.9830 Aluminum 1.1506.07112 1.7247 0.0 63 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506.07112 1.9609 0.0 64 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506.07112 2.1742 0.0 1 65 Hexa 1.160 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 66 Hexa 1.160 1.5? 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 67 Hexa 1.140 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 48 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 69 Hinxa 1.140 4.02 18.90 .9830 Aluminum 1.1506.07112 2.1742 0.0 1 70 Hexa 1.040 1.33 18.90 19.050 Aluminum 2.0574 .07620 2.8607 0.0 71 Hexa 1.040 1.58 18.90 19.050 Aluminum 2.0574 .07620 3.0086 0.0 72 Hexa 1.040 1.83 18.90 19.050 Aluminum 2.0574 .07620 3.1425 0.0 73 Hexa 1.040 2.33 18.90 19.050 Aluminum 2.0574 .07620 3.3942 0.0 74 Hexa 1.040 2.83 18.90 19.050 Aluminum 2.0574 .07620 3.6284 0.0 1 75 Hexa 1.040 3.83 18.90 19.050 Aluminum 2.0574 .07620 4.0566 0.0 76 Hexa 1.310 2.02 18.88 1.5240 Aluminum 1.6916.07112 2.6160 0.0 77 Hexa 1.310 3.01 18.08 1.5240 Aluminum 1.6916.07112 2.9900 0.0 78 Hexa 1.159 2.02 18.80 1.5240 Aluminum 1.6916.07112 2.6160 0.0 79 Hexa 1.159 3.01 18.88 1.5240 Aluminum 1.6916.07112 2.9900 0.0 80 Hexa 1.312 2.03 18.88 .9830 Aluminum 1.1506.07112 1.7250 0.0 St Hexa 1.312 3.02 18.88 .9830 Aluminum 1.1506.07112 1.9610 0.0 24

l l~ i SEE DETAIL A d V I r 8.884* = I E o.196* --e e-- I _ l.224' 5 0 ~ l = CELL CENTER TO CENTER (10,5")

== l .oa stainless steel i h A l l t l l 111 l l l l l11111111111111111111111 h " l.olAL .071* D,C AL DORAL p.olAL g I lIll!ittillvilililitil!Illilittliti! F .o75" Stainicas Steel I i INSIDE OF CELL l l DETAIL A Figure ' 1. Dona 6d C Cook Nucleer Plant Spent Fuel Pool Storage Cell Nominal Dimensions l. l 2s j

?> L F L I. t-I l I - i l 1 Empty Cell Fuel Cell 1 l I Figure 2. Donald C Cook Nuclear Plant SFP Region 1 Three of Four Fuel As-sembly Loading Schematic I.

I I l Region 1 to 2 Boundary -I I lll lfllIIIIIl I l III IIIIIIII i l I i son ! ) j 1 i l Reglon 2 ll l ll l ll l l I 1 Empty Cell !g Fuel Cell !u ~ il } j Figure 3. Donald C Cook Nuclear Plant Schematic for SFP Interface Boundary Between Regions 1 and 2 27

i I I I i I ....a..... -..,,,,,..g - -..*u.:. a --._. . t u.... -,7 i mg_________;p l % g...;' "-- ..... o &y.

===- wr %4, l .! p' !+,! . d, ~d ' n'E W . l,a v, a 2 ^ N r g .q _j d inla I a a.,.[ m.e (r::7 i: i i

i i :. ;

2 - ic \\ ':M, 1 MI J,Qg ' I I l .w... ,? I e r Ip / -* D - - p;,m i I i a eM =- ] e. g l ..l..l....l._...l _..l..l._.l _ l l.gf!;;!_!gl-2 'y i t -i i-i t i= t i ! q y y 1. I - a .,1 .+- i- ,.{ - fi,-----fff l- -] Qy W,g I j **$ QL g4%,, 'O' i I i;i iti I ! e/ / ',/ / [..,.Zr ' !IID l),,. ..e 1 _. .,s..s O. G. l l, t t It i .. =n IiI y; I l I a ,--*--.-.-.-.-.-.~y.p,., -.! la 1 j ! 1 lJi-y I l l d !il l l x! i

  • ~#

'I.'y*M 4-.I -.-.l -.l -.l -.l -. L.~.. w@- y _.-_777_ _7g __'g-t h-..;................. d ' i' b 1 .m n - cc....s. ;.... ' _ _ _ ___ _T,O m ...c t ( -j u ,f .q O v........ e i

I
I I

a Figure 4. Donald C Cook Nuclear Plant Fresh Fuel Rack Radial Layout 28 l e. + e w +-i.--, e w, -e.- ---e

- ~ -_ --. ..... ~..-.-..-.. - - I I I I e I a , * ' * ', '. + I J *...'. i 3 l 89 l + 5 ,/ E2 /T EE EE g E 7 - {. j + _a I ^ aw. i g n 3 i <::06 ,00, <20i < d., <d i \\ l 1,,,, i t i I I l I i4*'. g.g1

  • l mt F.

l l out6tNE or FUEL Al&Y ig*!g gy m h ;3 h % E k d.0 0.s.4. J 0 s. t i W O LE \\,.g,j j ( . 0 6. ,, % m m w w = <=- w p@ip;ggegs ~st g m l.h',: YVI b c',cc 1 %'.., t n 6i .e s I / ? 6 i I i 'I

mas, b t s OS T AI L. E, stE DETAlb'O 4 00 m t.00 s.t t A'.'S.1 g

s 668.00 L4 I h I . Figure 5. Donald C Cook Nuclear Plant Fresh Fuel Rock Axial Layout 2. .g

5. t L F L .93 i i i i I I I I l l I l ] J.. _ _....._L.. . __...l...v.....l... 'j i l I l I I I i n I I I I I .92 i i a i l i l i l I I I I .___I......._L..__ _._-.l.___... _ _I_ _ _ _ - 1 I I I I I i l i i i l I .91 i I l i I I I I l I I I I b i i i i x I I I I I I I I I .90 1 I I I I I I I .....l_.....____l..... ___I..._- 1 I I I I I I I .69 I I I I u I I I I I .____I_... .___I I ..I....- g--.. 3-... 1 I I I I I I I '0k.0 4.5 5.0 5.5 6.0 VM ENRICHMENT (W/0) l I BORAL HELD AT.02 GM B10/CM2 CENTER TO CENTER HELD AT 10.5" Figure 6. SensitMty of K.ev to Enrichment in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Leading. 30

4 -. ? FL .96 f 1 l l I I I r--- ,--- --i--- r-- r--- B 1 I I I I i .95 i o I l i l i I I T-~~' 1 - -' ~ ~~~'l--~ ~- r-- ~~~T---~7-- l l l 1 1 I '94 6 I I I I I I I I l l 1 I T---'- 7--- -~~1--- i - - T - - " i l i l I i 1 I i i 1 i i l i l l i I I i 1 1 I i 1 i i i i 1 .92 I I I I I I L. g i I I I i 1 g i i 1 1 1 1 I l i I I I ,93 I I i 1 1 I 1. _J. _ _I _ ____L__ L__...J 1 I i l 1 I I I I I n i I I ,9g I I I I l I __.1 ...__J.. _I ___L__ ___L ..._J I l l I I I I I I I e i I ,gg I i i l 1 i _1 J.. . _ _I _.L ..L_... J__ I i l i l I i i i e i i ,gg I I I I a l 1 1 ..-.J _.l... _ _ t. _ _ ___L ..J__ I I I I i i I 'gy i i 9.00 9.50 10.0 10.5 11.0 11.5 12.0 I CENTER-TO-CENTER SPACING (INCHES) I BORAL HELD AT.02 GM B10/CM2 ENRICHMENT HELD AT 4.95 W/O I I Figure 7. Sensitivity of Keve to Center-to-Conter Spacing in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading I 31 I _____h_m

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.03 .04 .00 .01 10 2 POSION LOAD GW B /CW ) I CENTER TO CENTER HELD AT 10.5" ENRICHMENT HELD AT 4.95 W/O ~' l II Figure S. Sensitivity of K.et to B Loading in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading lI

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J___ .L-- ...L I i i i i I I I i i i i I t I I i 1 h.9 I 0-235 ENRICHWENT (t/0).7 4.1 4.3 4.5 4 4.9 5.1 I Figure 9. Donald C Cook Nuclear Plant SFP Region 2 Fuel Assembly Minimum Burnup vs. Initial U* Enrichment Curve 33

I I 94 I i i i i i i l 1 1 I I I 1 1 1 I I i '*7--- / .._p...- T--- 1--- -- i--- r--' l l l l l/ l i i i i i 1 1 1 I I I I -__I i 1 I i 1 p.-- T--- -~7--- l-- r--' 7-- I I I I I I I I I I i 1 .92 j-, g i g l I i / I I I _ _ _ l_ _ _..__j__. _ _ J, _ __8 ___[__..__j___ l l I I l 6g i l I l i 1 I l l / l l l l -._L-... _ _.!. _ J... _ _ _ l_ _ _ L....l_ I I I I i I. l 1 l l l .90 I /1 1 1 I I I l / I l l l 1 -__L_ J.__. J. _ _ _ _ i_ _ _ ___L-_.. ...l l i I I I I l3 I I I I i 1 ,g ,gg 1 i i i t l [I I I I I I I i 1 l l I I ._L__.. _.L _. __J _ _ _ l_ _ _ 1.__....1__ 1 I I I I I I I I I I I t t f 1 1 l lI 4.0 4. 4.4 4.6 U-235 ENRICHWENT (W/0)2 3.4 3.6 3.6 BORAL HELD AT.02 GM B10/CM2 CENTER TO CENTER HELD AT 10.5" l l 1 'I Figure 10. Sensitivity of K.ee to Enrichment in the Donald C Cook Nuclear Plant SFP Region 2 Storage Area 34 1

I I I .96 i i i i i i I I I I I I i . I J __p__ .q._..__i_. __p _ _.i... _ _i_ _. I I I I I I i e i e i e i i ,93 I I i i i I I ___I__. -_J...l. 1__ __J..l_. I I I I I I I i I I I I I I I 94 I I T I i i i i l i I I I i 1 ._i_. r-- 7 "' ** '---l-**- "* - r - --7-'-- l--- i I I i l I i I I .93 I I I i l i 1 __l.. L .. -J l_ _. L. .J...l. I i l l l l l 1 b I I I \\ l i I i j I I I I I I I I I I I I I I I _7_ _q... _7 .q.. 1 I I I I I I .91 i i i i i i g i 'l l 1 1 I I I I W l_ _ 4. .l ..__I_. 4. .l ..l. I l l l l l .90 I\\ l I I I I I i i l 1 1 I I 1 i i l i i l l 1 1 1 1 I I 89 lW i i i i i iN i .qhL__i. I I l I i l I I . _i ..p. .i _ ...-i. __p_ t l 1 l i I I I I 1 t t f f i i l 'g 9.80 10.0 10.2 10.4 10.6 10.8 11.0 11.2 t CENTER-TO-CENTERSPACING(INCHES) 'g I BORAL HELD AT.02 GM B10/CM2 ENRICHMENT HELD AT 3.95 W/O I 1 Figure 11. Sensitivity of K.ve to Center-to-Conter Spacing in the 9onald C Cook . I 36 4

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  • 80.01

.C2 .03 .04 .05 .06 .07 .08 Hg0 DENSITY (G/CC) l Figure 13. Sensitivity of K.ve to Water Density in the Donald C Cook Nuclear Plant New Fuel Storage Vault g. e-ec-- m m m-

I I BIBLIOGRAPHY 1. Nuclear Regulatory Commission, Letter to All Power Reactor Ucensees, from B. K. Grimes OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications., Aptl\\ 14, 1978, ii 2. W. E. Ford \\\\l, CSRL-V: Processed ENDFIB-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Crlticality Safety, Reactor and Shielding Studies, ORNLICSDITM-160, June 1982. F 3. N. M. Greene, AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDFIB, ORNLITM-3706, March 1976. 4. L M. Petrie and N. F. Cross, KENO 'IV--An Improved Monte Carlo Criticality Program, ORNL-4938, November 1975. 5. M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage l of Power Reactor fuel, BAW-1494-7, July 1979. 6. J. T, Thomas, Critical Three-Dimensional Arrays of U(93.2) Metal Cy/Inders, Nuclear Science and Engineering, Volume 52, pages 350-359,1973. l 7. D. E.

Mueller, W.

A.

Boyd, and M.

W. Fecteau (Westinghouse NFD), Qualification of KENO ' Calculations with ENDFIB-V Cross Sections, American Nuclear Society Transactions, Volume 56, pages 321-323, June 1986. 8. A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Bol/Ing Water Reactors, WCAP-10106, June 1982. 9. Askew, J. R., Fayers, F. J., and Kemshell, P. B., A General Description of the Lattice Code WIMS, Journal of British Nuclear Energy Society, 5, pp. 564-584, 1966.

10. England. T.

R., CINDER A One-Point Depletion and Fission Product Program, WAPD-TM-334, August 1962.

11. Melehan, J.

B., Yankee Core Evaluation Program Final Report. WCAP-3017-6094, January 1971. I Bihiiogranny 38 j

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