ML20117H967

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Rev 1 to NEAD-SR-95/125, Summary of ANO-2 Criticality Safety Analysis for Fuel Enrichments Up to 5.0 Weight % U-235
ML20117H967
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 08/23/1996
From: Laura Smith
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20117H936 List:
References
NEAD-SR-95-125, NEAD-SR-95-125-R01, NEAD-SR-95-125-R1, NUDOCS 9609100099
Download: ML20117H967 (34)
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SUMMARY

OF THE ANO-2 l CRITICALITY SAFETY ANALYSES FOR

! FUEL ENRICHMENTS UP TO 5.0 WT. %  !

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! NEAD-SR-95/125.R1 l EDC FILE QR-204-36 I

AUTHOR L. ALAN SMITH TECHNICAL REVIEW BRIAN J. BOYLE KEITH B. MEGEHEE CENTRAL DESIGN ENGINEERING ENTERGY OPERATIONS, INC.

9609100099 960823 DR ADOCK0500g8 THIS CALCULATIONAL PACKAGE CONFORMS TO THE REQUIREMENTS OF CDE-P-03.01, REV.1

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l TABLE OF CONTENTS TABLE OF CONTENTS 2

1.

SUMMARY

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2. METilODOLOGY 5
3. SPENT FUEL POOL STORAGE RACK: REGION 1 9 3.1 Modeling........ .. .. ..................................................................9 3.2 Storage configurations.. . . . ... . . . . . . . . . . . . . .......................................10 l 3.3 Fuel handling accidents ..... ... .... . . ....... ... . . .. . . . ..................12 l 3.4 Analysis conservatism s......... .. ..... . . . .. ... . .. .... .. ....... ......... . .... .... . . ... .12
4. SPENT FUEL POOL RACK: REGION 2 18 4.1Modeling...............................................................................................18 4.2 Storage con figurat ion s ., . ..... .. .. .. . ...... . . .. .. .... .. . ........ .. .... .... . .. ....... ... .. .. . ... . ... 18 4.3 Fuel handling acc idents ................. .... . .... .. ...... ...... ..... ........ ... . . . . . .. . . . . I 8 4.4 Analysis conservatisms... . . .. . ..... .........................................................19
5. NEW FUEL STORAGE RACKS 23
6. FUEL TRANSFER UPENDER 26 1
7. CONTAINMENT TEMPORARY STORAGE RACK 28
8. REFERENCES 30 I

APPENDIX A: BENCHMARK OF SCALE-4 32 Total number of pages in calculational package: 33 This package is complete, self-contained, and has no external parts.

1 TABLE OF CONTENTS NEAD-SR-95/125.R1 +2 l l

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1.

SUMMARY

The Arkansas Nuclear One Unit 2 (ANO-2) spent fuel racks were originally designed by Westinghouse for the storage of fuel containing up to 4.1 wt. percent U235. Currently, unrestricted storage is allowed in Region 1 of the storage racks since a fixed poison (Boraflex) is included in the rack structure. Two storage configurations are allowed in Region 2 which utilize checkerboarding and burnup credit. Two additional storage areas, the new fuel storage rack and the containment temporary storage rack, also exist at ANO-2.

This report describes calculations used to determine acceptable fuel storage configurations to allow for enrichments up to 5.0 wt. percent at ANO-2. Uncertainties and biases due to method and manufacturing tolerances are addressed in determining the 95/95 upper limit kerr for each rack. The results for all configurations are consistent with the requirements outlined in NUREG-0800 [ Reference 14].

In the Region I calculations, conservative assumptions are made in considering the increase in reactivity due to Boraflex gapping or shrinkage. Additionally, the analysis contains significant conservatisms to compensate for potential long term surface degradation effects. All Boraflex panels are modeled with shrinkage or gapping losses. The shrinkage / gapping is determined based on EPRI reported blackness test data for Westinghouse racks (Reference 12]. To allow storage of 5.0 wt. percent assemblies in the Region 1 racks, cross-hatching (three of four cells contain assemblies) and burnup credit are utilized. All configurations assume no credit for soluble boron and yield 95/95 keff values kss than 0.95.

In the Region 2 calculations, checkerboarding is used to allow storage of undepleted 5.0 wt.

percent fuel assemblies. Two configurations assume burnup credit to allow storage of assemblies with enrichments up to 5.0 wt. %. One configuration allows storage in all cells while the other requires cross-hatching with three of four cells containing assemblies (the fourth cell is a water hole). No soluble boron credit is assumed in the analysis of the Region 2 storage configurations.

In the new storage fuel rack calculations, both possible and optimum moderation conditions were analyzed. The 95/95 kerrvalues were determined to be less than 0.95 when moderated with water at a density of 1.00 gm/cc and less than 0.98 at optimum moderation conditions (0.065 gm/cc). Calculations were also performed to verify 5.0 wt. percent assemblies can be loaded into the fuel transfer upender and stored in the containment temporary storage rack.

With increasing cycle lengths, higher enrichments are needed to meet cycle energy demands.

This criticality analysis demonstrates that fuel assemblies with enrichments up to 5.0 wt. percent U235 can be safely stored in the ANO-2 fuel storage racks. Both nominal and accident conditions are analyzed. The results of these analyses are summarized in Table 1-1.

1.

SUMMARY

NEAD-SR-95/125.R1 + 3

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Table 1-1. Final reactivities for ANO-2 fuel storage areas Storage area description Acceptance criterion Determined 95/95 kegt Spent fuel pool Region I Configuration A(cross-hatch) 0.95 0.938 Configuration B (undepleted) 0.95 0.937 Configuration B (burnup credit) 0.95 0.942 -

Spent fuel pool Region 2 Configuration A (checkerboard) 0.95 0.926 Configuration B (burnup credit / cross-hatch) 0.95 0.944 Configuration C (burnup credit) 0.95 0.9498 New fuel storage rack Possible moderation (1.00 gm/cc) 0.95 0.916 Optimum moderation (0.065 gm/cc) 0.98 0.976 Upender 0.95 0.94997 Containment temporary storage rack 0.95 0.92 _

1.

SUMMARY

NEAD-SR-95/125.R1

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2. METHODOLOGY Computer codes

. All rack criticality calculations are performed with the widely accepted SCALE-4 code package

[ Reference 1]. The SCALE-4 system includes the CSAS25 Control Sequence of the ORNL Criticality Safety Analysis Sequence No. 4 (CSAS4). The functional modules sequentially executed by the CSAS25 control module are the BONAMI-S code, the NITAWL-S code, and the KENO V.a code. The benchmarking of SCALFA is discussed in Appendix A. Unless otherwise noted, the calculated eigenvalue for each case is based on 500,000 neutron histories.

The CASMO-3 two-<limensional integral transport code was used to evaluate the rack and fuel design tolerances and perform reactivity equivalencing calculations [ Reference 8]. This code is widely accepted by the industry for these applications.

i SIMULATE-3 is an advanced 3D nodal simulator which uses the nodal expansion solution j method [ Reference 9]. It has been previously qualified for application to Entergy Operations, i Inc. PWR cores [ Reference 13]. Its use in this study is limited to providing data for the I I

evaluation of burnup distribution and spectral history effects.

Fuelassembly description The base fuel assembly is a Combustion Engineering 16 by 16 assembly as shown in Figure 2-1.

In the rack calculations, nominal values are used for all fuel dimensions and design tolerances are j treated as uncertainties. The fuel parameters are listed in Table 2-1. Axial dimensions were determined using the debris resistant ABB-CE Guardian Grid lower grid design. Consistent with the analysis in Reference 18, the neutron absorption by U234 is credited.

'f lll lll lll Fuel rod A

Guide tube

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fII IIi Iii ili III III Figure 2-1. Combustion Engineering 16 by 16 fuel assembly

2. METHODOLOGY NEAD-SR-95/125.R1 =5

/ l' Table 2-1. ANO-2 base fuel assembly parameters Description Parameter I Fuel enrichment (wt. % Um) up to 5.0 Number of fuel rods (per assembly) 236 ruel rod Zirc-4 clad OD (inch) 0.382 Clad thickness (inch) 0.025 Fuel pellet OD (inch) 0.325 Stack density (gm/cc) 10.25 Maximum U-235 loading per unit length (gm/ inch) 0.614108 Rod pitch (inch) 0.506 Number of Zirc-4 guide tubes (per assembly) 5 Guide tube OD (inch) 0.98 Guide tube thickness (inch) 0.04 Active fuel height (inch) 150.0 j Fuelassembly uncertainties Uncertainties or tolerance factors for the fuel and rack design parameters are evaluated by performing sensitivity studies to determine the reactivity impact of each factor. The rack design j tolerances are discussed in the rack specific portions of this report. The following fuel tolerance factors were analyzed:

1. Fuel enr'chment: In cases considering fuel with enrichments less than 5.0 wt. % or depleted fuel, an increased loading of +0.05 wt. % U 235 is t.ssumed. No uncertainty is assumed for 5.0 wt. % cases since this is the maximum allowed enrichment.
2. Fuel pellet density: The maximum value is assumed in the base calculations; therefore, no tolerance analysis is necessary.
3. Form factor: The maximum form factor increases the fuel loading.
4. Fuel pellet diameter: A larger fuel pellet diameter increases the fuel loading.
5. Fuel clad Inner diameter: A larger fuel clad inner diameter decreases the absorption in the i fuel cladding.
6. Fuel clad outer diameter: A small fuel clad outer diameter reduces parasitic absorption and increases moderation.
7. Guide tube thickness: A smaller guide tube thickness reduces the parasitic absorption and increases moderation.  !

Burnup credit using reactivity equivalencing i For burnup credit, reactivity calculations are performed to generate initiai enrichment-fuel assembly burnup combinations which yield equivalent reactivity when stored in the fuel storage racks. Depletion calculations are performed for various inhial fuel assembly enrichments using CASMO. The fuel assembly is depleted using hot full power operating conditions. A conservative boron history is assumed to enhance the buildup of plutonium and make the assembly more reactive when stored in the spent fuel storage racks. An additional spectral history adjustment factor is also included to account for higher boron levels which may occur if Entergy elects to implement 24 month fuel cycles. The most reactive time after shutdown of the reactor is conservatively approximated by removing Xel35 and 1135,

2. METHODOLOGY NEAD-SR-95/125.R1 + 6

l t Following depletion in core geometry, the assembly is reanalyzed in rack geometry, using l CASMO, to determine reactivity as a function of burnup. An initial enrichment to burnup curve l is then established based on a constant reactivity criteria. Multiple pairs are used to establish a burnup curve which bounds the possible initial enrichment range (up to 5.00 wt. percent) of the r

spent fuel storage rack. This burnup-initial enrichment curve is converted to a burnup-initial U 235 loading curve. When this conversion is performed, reactivity effects due to variations in fuel density and geometry are considered. Potential changes in reactivity due to increases in nominal pellet volume and density while conserving the same U 235 loading were evaluated. The most limiting combination of parameters over the expected range was used to determine the burnup requirement. Each burnup-U 235 loading combination was confirmed to be equivalent to or less reactive than the base configuration.

The effects of the axial burnup distribution on an assembly's reactivity is considered for each i rack geometry. Axial burnup distributions determined from core follow calculations based on SIMULATE-3 [ Reference 9] are modeled in the rack geometry. Assemblies are serted based on average bumup and a typical exposure distribution is determined based on assemblies with similar average burnup. The burnup of the upper and lower axial regions are then reduced by 10% to produce a more severe distribution. The effects of the axial distribution is rack specific and is discussed in the appropriate sections. If the axially dependent exposure distribution is determined to be more reactive than the uniform distribution, a limiting axial burnup distribution adjustment is established based on the reactivity difference between the uniform and axially distributed cases. This adjustment is included when determining 95/95 kerrvalues.

l Two factors are included to account for the uncertainties associated with burnup credit: the depletion history uncertainty and the isotopic calculational uncertainty. The depletion history uncertainty accounts for nodal variations in depletion history such as different moderator densities and inter-assembly burnup tilts. The established uncertainty was confirmed to bound any axially dependent bias. The calculational isotopic uncertainty was determined from previously approved physics methods [ Reference 13]. j I

Uncertainties in fuel and rack materials and geometries l Fuel assembly and rack geometry mechanical tolerances are evaluated using CASMO. The difference between the base case ki nf and the tolerance case i k nf i s the reactivity effect of a tolerance. Negative values are set equal to zero. The Ak values are statistically combined by taking the square root of the sum of the squares of each contributing uncertainty (Reference 8].

Treatment of bias and uncertainty l The KENO calculated eigenvalue for each analyzed storage rack configuration is combined with the appropriate tolerance and method uncertainties using the following equation [ Reference 6].

This 95/95 treatment is the same as that used in previously approved analyses by Entergy

[ Reference 18]. The method bias and method uncertainty are discussed in Appendix A. The 95/95 tolerance factors for each case are determined from Reference 7.

I When considering racks containing Boraflex poison, various Boraflex conGguration assumptions are evaluated. For each set of Boraflex assumptions, several cases, each containing randomly generated panel gap and/or end shrinkage configurations are analyzed. The KENO ker(s of each randomly generated case are combined to obtain an average kerrvalue. The uncertainty due to

2. METHODOLOGY NEAD-SR-95/125.R1 *7

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variations between Boraflex cases is combined with t' -ertainty (atotal) and used to determine the degrees-of-freedom and the 95/95 tok. - (K) to yield the 95/95 kefr value.

Equation for determining 95/95 probability / confidence reactivities CASE k,gg95,95y = gk ,,, + Ak,,,g,g + Ak,o,,,,,,,, + x x a,o,,, + Akadj-spect

  • ONadj-axial where Ag,,, = KENO k,g(case dependent)

Akmeshog = 0.007600 Ntolerances = Overall tolerance uncertainty \

" total " # keno * *Boraflex * " method " " keno * "Boraflex

  • 0*00' A0A "4total f, 4 "4keno *4 *4methad

, Boraflex

((50')- 1) + 2) ((X- 1) + 2) ((21- 1)+ 2)

Note: 500 generations, X Boraflex cases, and 21 critical experiments were analyzed.

r = 95 / 95 tolerance factor forfdegrees-of- freedom l

Ak adj-spect = Spectral history adjustment Akadj-arial = Axial burnup distribution adjustment i

2. METHODOLOGY NEAD-SR-95/125.R1 +8

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3. SPENT FUEL POOL STORAGE RACK: REGION 1 '

l It was shown in a previous analysis that assemblies with enrichmen;s up to 4.1 wt. % U235 may be safely stored in the Region I spent fuel pool storage racks [ Reference 10]. This work shows that assemblies with enrichments up to 5.0 wt. % U235 may be safely stored in these racks.

Conservative assumptions are made to consider the reactivity effects of Boraflex panel gapping and shrinkage. A signi6 cant margin to the acceptance criteria and other conservatisms are 1 i reserved for the potential Ivng term effects of surface degradation. The requirements of NUREG-0800 [ Reference 14] and ANSI /ANS-57.2-1983 [ Reference 15) are met by this analysis.

3.1 Modeling Base models The ANO-2 spent fuel racks were designed by Westinghouse Electric Corporation. The Region 1 rack is an array of cells with stainless steel walls. Stainless steel wrapper plates retain Boraflex sheets on each side of the storage cells. This configuration is illustrated in Figure 3-1. Three-dimensional KENO calculations were performed using a four by four array of rack cells which includes a total of 64 Boraflex panels. The end shrinkage for each Boraflex panel is randomly determined for the upper and lower edges from the distribution shown in Figurc 3-2. The maximum allowed total shrinkage in a single panel is 4.1% of the panel length, the maximum value determined from EPRI testing [ Reference 12].

Boraflex configuration The ANO-2 spent fuel pool Region I racks contain Boraflex panels held in place by a stainless steel wrapper plate. The edges of the wrapper plates are welded to the stainless rack cell. This configuration allows minimal water flow around the Boraflex panels unlike other rack designs which allow a large volume of water to interact with the panels.

In this work, EPRI reported blackness test results of Westinghouse racks were used to determine the impact of Borallex panel shrinkage. The testing was performed on panels which have been exposed to gamma radiation in excess of lx1010 rad, the dose level where shrinkage reaches a saturation condition of 4.1% ofits pre-irradiation length [ Reference 12].

Of the five tested plants, no gaps were detected in the Boraflex panels at two of the plants. At these plants, the shrinkage reduced the length of the Boraflex at the ends of the panels. To l determine the most conservative assumption for the ANO-2 rack design, calculations were performed assuming 1) all panels have gaps,2) all panels have end shrinkage, and 3) 65% of panels have gaps and 45% have end shrinkage.

The gap distribution or end shrinkage in each panel was determined based on random sampling j of gap / shrinkage probability distributions. The location and size of the gaps and/or end shrinkage

3. SPENT FUEL POOL STORAGE RACK: REGION 1 NEAD-SR-95/125.R1 + 9 j r

1 were determined using probability distributions from Reference 12. The most limiting configuration for the ANO-2 rack design was determined to be the case where all panels have end shrinkage. The most limiting end shrinkage distribution from Reference 12 was used and is I

shown in Figure 3-2.

Additional dimensions and characteristics of the Boraflex panels are shown in Table 3-1. The minimum design B IO loading and physical dimensions are assumed. Also, the maximum predicted shrinkage of Boraflex (4.1%) is assumed in the panel width to conservatively account for radiation induced shrinkage. The corresponding increase in density is conservatively not included.

Tolerance analysis The fuel tolerance factors are described in Section 2. The additional storage rack tolerance factors (8-13) are discussed below.

8. Fuel storage area: A larger area increases the neutron moderation around the fuel assembly.
9. Rack wall thickness on the Boraficx side: A thinner wall reduces the absorption in structural material.
10. Rack wail thickness on the fuel assembly side: A thinner wall reduces the absorption in structural material.

I1. Stainless steel wrapper plate thickness: A thinner wrapper plate reduces the absorption in structural material.

12. Minimum separation distance between cell wrapper plates: A reduction in the distance between the cell wrapper plates decreases moderation between rack cells.
13. Lateral assembly position: Off-center positioning of the fuel assembly in the storage cell increases the interaction between assemblies.

The tolerance factor results are shown in Table 3-2. The tolerances were combined using the methodology described in Section 2. Tolerance factors were determined for each storage configuration which are described in detail below.

3.2 Storage configurations Two different fuel storage configurations were evaluated using the previously described approach.

Configurstion A Configuration A utilizes cross-hatching to allow storage of assemblies with enrichments up to 5.00 wt. percent as shown in Figure 3-3. Each fuel assembly must be adjacent to two water holes or located diagonally from four water holes as shown. No credit is assumed to burnable poisons or soluble boron. The 95/95 kerrvalue for this configuration is 0.938 as shown in Table 3-3.

3. SPENT FUEL POOL STORAGE RACK: REGION 1 NEAD-SR-95/1.'t5.R1 + 10

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/ i Configuration B Credit for assembly burnup is taken to allow storage of assemblies with enrichments up to 5.0 wt.

% in an arrangement with fuel assemblies located in all rack cells as shown in Figure 3-3. The 95/95 kerrvalues are shown in Table 3-3 and the exposure-initial U 235 loading per unit length curve generated using reactivity equivalencing is shown in Figure 3-4. Variations in the  ;

assembly reactivity due to changes in fuel density and geometry were considered. If an assembly's fuel loading is less than the analyzed loading, the loading used to determine the assembly's burnup requirement is conservatively considered to be the analyzed loading.

The effects of the axial burnup distribution on assembly reactivity are considered in the l determination of the curve. The axial exposure distribution corresponung to the 0.614108 gm U 235 per inch (5.00 wt. %) statepoint was converted to an equivalent reactivity assembly by determining the equivalent initial enrichment for each axial segment. This assembly was then analyzed in 10 KENO cases containing randomly generated Boraflex configurations as previously discussed. An assembly with the same average burnup, which was uniformly distributed, was also evaluated using this same process. The distributed burnup case was determined to be limiting. The differences between these cases were used to establish an axial ,

burnup distribution adjustment which was applied to lower burnup configurations. This approach )

is conservative because exposure shapes are more uniformly distributed at lower burnups. Other enrichment-burnup combinations were evaluated using uniformly distributed burnup assumptions with reactivity equivalent to the 5.0 wt. % statepoint. These evaluations were performed with the adjustment included to account for the axial burnup shape bias.

1 Interface effects between configurations Cases were evaluated to consider interface effects between various orientations of Configurations l

A and B. In each case the assembly storage constraints of each configuration were ' net. For example, the Configuration A 5.00 wt. % assembly was always adjacent to two water holes or I diagonal to four water holes. The 95/95 kefy values for these cases were less than or equal to the values determined for the limiting configuration shown in Table 3-3.

Region 1- Region 2 Interface Pegion 1-Region 2 racks are not separated from each other by additional water spacing. The Region I cells at the interface contain a Boraflex panel on the cell wall adjacent to the Region 2 cells. Calculations were performed to determine if any limitations exist for the Region 1 storage cells at the Region 2 interface since one Boraflex panel is present between the Region 1 and Region 2 assemblies.

.hese calculations determined that 5.00 wt. % assemblies should be restricted from storage in the first row of Region I cells at the Region 1-Region 2 rack interface. Configuration B assemblies may be placed in the first row. Therefore, no restrictions are placed on the storage of Region I l assemblies which meet the exposure-initial U235 loading requirements of Configuration B.

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3. SPENT FUEL POOL STORAGE RACK: REGION 1 NEAD-SR-95/125.R1 + 11

. l' 3.3 Fuel handling accidents For accident conditions, the double contingency principle applies which states it shall require two unlikely, independent, concurrent events to produce a criticality accident. Therefore, for accidents, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition. The accidents which must be considered are listed below:

  • Fuel assembly dropped on top of the rack: The rack structure is such that an assembly positioned horizontally on top of the rack is approximately 30 inches away from the upper end of the active fuel region of the stored assemblies. This distance precludes interaction between the dropped assembly and the stored fuel.

. Fuel assembly dropped between rack modules or next to the pool wall: This accident cannot occur since the distance between all the rack modules and to the pool walls is less than the width of a fuel assembly.

  • Misplaced fuel assembly: The misplacement of a fuel assembly in a rack configuration was considered. This accident is considered by placing a 5.0 wt. percent assembly from Configuration A in a Configuration B position. The soluble boron concentration was assumed to be 1000 ppm. The 95/95 kerrvalue is 0.80 which is well below 0.95.

3.4 Analysis conservatisms The following conservatisms are included in the analysis of the Region I spent fuel storage racks:

1. Minimum as-<!esigned Boraflex dimensions and the minimum areal B 10 loading are used.

Additionally, maximum poison shrinkage of 4.1% is assumed in the panel width. The boron loading is not increased to account for the densification which occurs as a result of the

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shrinkage. The end shrinkage assumptions are more conservative than gap formation assumptions for the ANO-2 rack geometry.

2. The statistical uncertainty associated with the evaluation of a limited number of Boraflex configurations (10 KENO cases) is considered in the determination of the total uncertainty. ,

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3. No neutron leakage in the x-y directions is assumed. '
4. Water is set to the maximum density of 1.00 gm/cc.
5. The assembly is modeled at its most reactive condition after shutdown when assuming burnup credit. i i
6. U 235 loading-burnup combinations for assemblies less than 5.0 wt. % for Configuration B conservatively assume the axial burnup distribution adjustment of the 5.0 wt. % case. The

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axial exposure shapes are more uniform for assemblies with burnups less than that required ,

for a 5.0 wt. % assembly.

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7. No credit is assumed for burnable poison in the fuel assemblics.
8. Much stainless steel structure is not modeled.
9. Credit for soluble boron is assumed only for accident cases consistent with the double contingency principle.
3. SPENT FUEL POOL STORAGE RACK: REGION 1 NEAD-SR-95/125.R1 12

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Figure 3-1. Radial view of the Region 1 spent fuel pool storage rack End Shrinkap 8 00 06

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Figure 3-2. End panel shrinkage based on blackness test data for

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3. SPENT FUEL POOL STORAGE RACK: REGION 1 NEAD-SR-95/125.R1 = 13

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Table 3-1. Boraflex panel characteristics used in the Region 1 analysis Parameter Design Modeled value specification Minimum 1310 loading 0.026 0.026 (gm Il 10/ cm2 for 0.100 inch thickness)

Ileight (inch) 144.0010.25 143.75a Width (inch) 7.50010.075 7.1206 Thickness (inch) 0.10010.007 0.093 a Changes in the axial dimensions of the Borallex panels due to shrinkage are discussed in the text.

Region 1 Configuration A Assemblies acent to two water holes i

Fuel assembly Water hole Water hole 5.00 wt. % assembly.

4 by 4 array used for KENO analysis j ross-hatch configuration Assembly diagonal to four water holes Region 1 Configuration B 1

N Fuel assembly Water hole l Exposure-initial U-235 loading cun'e l \

4 4 by 4 array used for KENO analysis N 4 cf 4 Configuration i

. Figure 3-3. Region 1 storage configurations

3. SPENT FUEL POOL STORAGE RACK: REGION 1 NEAD-SR-95/125.R1 + 14

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l Table 3-2. Tolerance factor results for Region 1 configurations Configuration l A 11 11 Undepleted assembly Depleted assembly l

Delta k,,/s

'Iolerance number .

l' N/A 0.00311 0.00444 2 N/A N/A N/A 3 0.00118 0.00144 0.00161 4 0.00029 0.00036 0.00041 5 0.00000 0.00002 0.00003 6 0.00263 0.00246 0.00232 7 0.00057 0.00052 0.00049 8 0.00056 0.00092 0.00109 9 0.00000 0.00000 0.00000 10 0.00000 0.00000 0.00000 1I 0.00000 0.00000 0.00000 12 0.00048 0.00043 0.00040 i 13 0.00186 0.00067 0.00000 Isotopic calc. uncert. N/A N/A 0.00495 Derletion hist. uncert. N/A N/A 000425 OveraH tolerance 0.00357 0.00444 0.00849 uncertainty

  • see Section 2 for the description of tolerances I-7 and this section for tolerances 8-13.

Table 3-3. Final reactivities for Region 1 configurations Parameter i Configuration A Configuration B Configuration B Undepleted assembly Depleted assembly Statistical average of the KENO kefr's 0.92154 0.91841 0.91483 Method bias 0.007600 0.007600 0.007600 Tolerances 0.00357 0.00444 0.00849 Spectral history n/ea n/aa 0.0050 Total uncertainty 0.002396 0.002565 0.002605 95/95 factor 2.092 2.176 2.179 Final 95/95 kefr 0.938 0.937 0.942 a Undepleted fuel is considered in these configurations; therefore, no spectral history bias must be included.

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3. SPENT FUEL POOL STORAGE RACK: REGION 1 NEAD-SR-95/125.R1 + 15

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Inmaal Assuubly A=esage U 239 Landing per Duk Imgib(pm U 315 per hm:h)

Curve = 68.008:- 19366 when x > 0399181 Figure 3-4. Region 1 Configuration B exposure vomus initial assembly average U-235 loading curve 1

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3. SPENT FUEL POOL STORAGE RACK: REGION 1 NEAD-SR-95/125.R1 + 16 l

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4. SPENT FUEL POOL RACK: REGION 2 It was shown in a previous analysis that assemblies with enrichments up to 4.1 wt. % U235 may be safely stored in the Region 2 spent fuel storage racks [ Reference 10]. His work shows that assemblies with enrichments up to 5.0 wt. % U235 may be safely stored in these racks. The requirements of NUREG-0800 [ Reference 14] and ANSI /ANS-57.2-1983 [ Reference 15] are met by this analysis.

4.1 Modeling Base mod *Is The Region 2 rack is an array of cells with stainless steel walls. This configuration is illustrated in Figure 4-1. The three-dimensional KENO model used in the Region 2 calculations includes a four by four array ofrack cells. The fuel assembly dimensions are given in Section 2.

Tolerance analysis The fuel tolerance factors are described in Section 2. The additional storage rack tolerance factors (8-12) are discussed below.

8. Fuel storage area: A smaller area increases the neutron moderation between rack cells.
9. Rack wall thickness on the outer wall side: A thinner wall reduces the absorption in structural material.
10. Rack wall thickness on the fuel assembly side: A thinner wall reduces the absorption in structural material.

I1. Minimum separation distance between cell walls: .A reduction in the distance between the cell walls decreases moderation between rack cells.

12. Lateral assembly position: Off-center positioning of the fuel assembly in the storage cell increases the interaction between assemblies.

The tolerance factor results for each configuration are shown in Table 4-1. The tolerances were combined using the methodology described in Section 2. Tolerance factors were determined for each storage configuration. Each storage configuration is discussed below.

4.2 Storage configurations Configuration A This configuration utilizes checkerboarding to allow storage of assemblies with fuel enrichments up to 5.0 wt. %. Each fuel assembly must be adjacent to four water holes as shown in Figure 4-

4. SFENT FUEL POOL RACK: REGION 2 NEAD-SR-95/125.R1 + 17

l* $

2. This configuration is modeled using a 4 by 4 array of rack cells with fuel assemblies in every other cell. The 95/95 kerrvalue is 0.926 as shown in Table 4--2.

Configurations B and C Credit for assembly burnup is taken to allow storage of enrichments up to 5.0 wt. % m these configurations. Configuration B requires cross-hatching where each fuel assembly to be adjacent to two water holes or located diagonally from four water holes as shown in Figure 4-2.

Configuration C defines the requirements which must be met for an assembly to be stored in any Region 2 rack location.

Reactivity equivalencing is used to generate an assembly exposure-initial U 235 loading curve which determines if fuel assemblies can be stored in these configurations. Vuriations in the i assembly reactivity due to changes in fuel density and geometry were considered. If an assembly's fuel loading is less than the analyzed loading, the loading used to determine the I assembly's burnup requirement is conservatively considered to be the analyzed loading.

Both uniform and axially distributed burnup distributions were considered. The uniform l distribution was determined to be more limiting. This is due to the fact that the racks do r.o. '

contain fixed poison (Boraflex) and the axial leakage effects offset the burnup distribuGor effects for the range of exposures considered in this analysis. The 95/95 kerrvalue for Confgurations B ,

and C are 0.944 and 0.9498, respectively. The exposure-initial U-235 loading curms tre shown in Figure 4-3.

Interface effects between configurations Cases were evaluated to consider interface effects between various orientations of Coh'.igurations A, B, and C. In each case the assembly storage constraints of each configuration were met. For

{

example, the Configuration A 5.00 wt. % assembly was always adjacent to four water holes. The 95/95 kerrvalues for all cases were equal to or less than the values determined for the limiting l configuration shown in Table 4-2.

4.3 Fuel handling accidents For accident conditions, the double contingency principle applies which states it shall require two unlikely, independent, concurrent events to produce a criticality accident. Therefore, for accidents, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition. The accidents which must be considered are listed below:

e Fuel assembly dropped on top of the rack: The rack structure is such that an assembly positioned horizontally on top of the rack is approximately 30 inches away from the upper end ofine active fuel region of the stored assemblies. This distance precludes interaction between the dropped assembly and the stored fuel.

  • Fuel assembly droppeo between rack modules or next to the pool wall This accident cannot occur since the distance between all the rack modules and to the pool walls is less than the width of a fuel assembly.

. Misplaced fuel assembly: The misplacement of a fuel assembly in a rack configuration was considered. This accides is considered by placing a 5.0 wt. percent assembly from Configuration A in a Configuration C position which requires burnup credit. The soluble

4. SPENT FUEL POOL RACK: REGION 2 NEAD-SR-95/125.R1 + 18

. J l

l boron concentration was assumed to be 1000 ppm. The 95/95 kerrvalue is 0.84 which is well below 0.95.

4.4 Analysis conservatisms The following conservatisms are included in the analysis of the Region 2 spent fuel storage racks:

1. No neutron leakage in the x-y directions is assumed.
2. Water is set to the maximum density of 1.00 gm/cc.
3. The assembly is modeled at its most reactive condition after shutdown when assuming burnup credit.
4. No credit is assumed for burnable poison in the fuel assemblies.
5. Much stainless steel structure is not modeled.
6. Credit for soluble boron is assumed only for accident cases consistent with the double contingency principle.

l n.ir l  :

l 9.80"  ! -

d6.75" as modeled !  ! 7.20" actual h a

. x a 8.580" l0.075" h e- -

l 9 80" l l 1.070" l -

Note: All dimensions are inches. Drawing is not to scale.

/

i.. . ., .., , e Figure 4-1. Radial view of the Region 2 spent fuel pool storage rack l 4. SPENT FUEL POOL RACK: REGION 2 NEAD--SR-95/125.R1 + 19 l

a Region 2 Configuration A Assembly adpcent to four water holes a s a e t

g _

g g Water hole g g 5.00 wt. % assembly NI N e M N g g 4 by 4 array used for KENO analysis ys- g g- N

- w. 13< '

Checkerboard configuration  :

1 Region 2 Configuration B Assemblies a jacent to two water holes {

Water hole y

I 3 j"

[ .

gj "g%

jlf

\ Fuel assembly Water hole Assembly meeting Config. B exposure- initial U-235 loading curve bl A . a -

/

g  %

gg ;g l gj 4 by 4 array used for KENO analysis

- =

we w 'g- m- e gr g ed .

Cross-hatch configuration l

f Assembly diagonal to four water holes I

l Region 2 Configuration C gtp .g ;jgl g g Water hole g g. g g g g g)j gg Fuel assembly fT Assembly meeting Config. C exposure- mitial U-235 loading curve k h l5h M ikf: ($$ $

d r @ n e s se N g g g g j y g g g 4 by 4 array used for KENO analysis l

5 d,' $ Mh % MP k 4 of 4 Configuration Figure 4-2. Region 2 storage configurations

4. SPENT FUEL POOL RACK: REGION 2 NEAD-SR-95/125.R1 + 20

Table 4-1. Tolerance factor results for Region 2 configurations Configuration A B C Tolerance number Delta k,,/s j l' N/A 0.00760 0.01143 2 N/A N/A N/A 3 0.00106 0.00169 0.00188 4 0.00026 0.00044 0.00048 l 5 0.00003 0.00022 0.00030 l 6 0.00163 - 0.00099 0.00072 1 7 0.00035 0.00021 0.00015 I 8 0.00004 0.00016 0.00000 )

9 0.00256 0.00283 0.00286 ]

10 0.00257 0.00282 0.00284  ;

11 0.00000 0.00000 0.00001 )

12 0.00207 0.00060 0.00159 j isotopic calc. uncert. N/A 0.00642 0.00642 -

Depletion hist. uncert. N/A 0.00586 0.00752 Overall toicrance 0.00463 0.01240 0.01586 uncertainty

  • see Section 2 for the description of tolerances I-7 and this section for tolerances 8-12 l

i Table 4-2. Final reactivities for Region 2 configurations Configuration .

A 11 C KENO keg 0.90937 0.91088 0.91418 Biases Method 0.007600 0.007600 0.007600 Spectral history 0.00000 0.00850 0.00850 Uncertainty KENO 0.00114 0.00088 0.00078 Method 0.001464 0.001464 0.001464 Tolerance 0.00463 0.01240 0.01586 Total 0.00186 0.00171 0.00166 Degreewf-freedom 54 39 34 95/95 factor 2.046 2.133 2.176 (95/95 factor)xTotal 0.00380 0 00364 0.00361 Case 95/95 k,g 0.926 0.944 0.9498

4. SPENT FUEL POOL RACK: REGION 2 NEAD-SR-95/125.R1 + 21

,.' [

a a a m % % v.am i -e u

.e

$19

$$ 0

$2 9 So e -_

Canngmanas c (4 efd esus) 45.0 42 5 --

4o 0 e ns ccc s g 35. n- c-w

  • n9 4- D I' ns y I 25 0 - conAg. B n c.w

<l us- L_ j/

se e f

l19 ,

4 12.5 10 0 - configweden A -

T.S -

so 3.9 -

.. /I hinal Asseme4y Average U-2351mmens per unk Lagth (gm U-215 per buh) 6mm Config. B curve = 239.01 z'- 347.75s' + 243.18x - 41.452 Config C curve =-71435xd+ 1335.80x'-946.44x 2+ 394.52x -47.040 Figure 4-3. Region 2 exposure versus initial assembly average U-235 loading curve

4. SPENT FUEL POOL RACK: REGION 2 NEAD-SR-95/125.R1 + 22

g'. l 4

1 l

i

5. NEW FUEL STORAGE RACKS In a previous analysis, assemblies with enrichments up to 4.1 wt. % U235 show significant i margin to the criticality acceptance criteria [ Reference 11]. His work shows that assemblies with enrichments up to 5.0 wt. % U235 may be safely stored in these racks. The requirements of

) NUREG-0800 [ Reference 14] and ANSI /ANS-57.3-1983 [ Reference 16] are met by this

! analysis.

Tolerance analysis

)

The fuel tolerance factors are described in Section 2. The uncertainty in the lateral position of the ,

fuel assembly in the rack cavity is considered as a tolerance by reducing the assembly pitch to  ;

minimize the separation between assemblies. i i

The tolerance results for the optimum moderation (0.065 gm/cc) and possible moderation (1.00

) gm/cc) cases are shown in Table 5-1. The tolerances were combined using the methodology described in Section 2.

Base model

The following assumptions were used in the development of the base KENO model. The
nominal fuel assembly dimensions are described in Section 2. Scoping calculations to determine the optimum moderator density were performed using 250,000 particles. All other calculations

.' used 500,000 particles.

1. The new fuel rack is filled with unirradiated fuel assemblies containing 5.0 wt. % U235. The 4

UO 2density of the fuel stack is 10.25 gm/ce. The nominal assembly spacing of 26.0" is i assumed.

2. The rack is moderated by pure water at various uniform densities.
3. No burnable poison, control element assembly, or other fixed poison is stored with the fuel i assembly. i
4. The pit walls and the floor are modeled using concrete followed by a reflective boundary condition. The area above the rack is modeled as water moderator followed by a reflective boundary condition.
5. No rack structural material is modeled.

l SCALE cases were run to determine the optimum moderator density. The optimum density was

determined to be 0.065 gm/cc as shown in Figure 5-1. The 95/95 kerrvalues for the 0.065 and I
5. NEW FUEL STORAGE RACKS NEAD-SR-95/125.R1 + 23

)

  • i 1.00 gm/cc cases are shown in Table 5-2. The 95/95 kerris 0.916 for the possible moderation case (1.0 gm/cc) and 0.976 for the optimum moderation case (0.065 gm/cc).

Fuelhandling accidents For accident conditions, the double contingency principle applies which states it shall require two unlikely, independent, concurrent events to produce a criticality accident. Therefore, ' dry' conditions are assumed in the evaluation of the accident where an assembly drops on top of the new fuel rack. The dropped assembly and all assemblies in the rack are assumed to contain 5.0 wt. % fuel. The dropped assembly accident case yields a 95/95 kerrof 0.52 which is well below 0.95.

Analysis conservatisms The following conservatisms are included in the analysis of the new fuel storage racks:

1. No neutron leakage in the x-y directions is assumed.
2. Water is set to the maximum density of 1.00 gm/cc. The optimum moderation condition is also evaluated.
3. None of the rack structure is modeled.
4. No credit is assumed for burnable poison in the fuel assemblies, i

Eeff vs. Moderator Density )

. . , m.

A It

. _? .._ ,

.., l

\ ,

. so 3

, l 5 \ ,/

/

.t \.. -

oss -

\ g

- e 4ENO K.efI(Rai andch

.,, I II IIII Figure 5-1. Determination of optimum moderator dens!ty for the new fuel storage rack j

5. NEW FUEL STORAGE RACKS NEAD-SR-95/125.R1 + 24

,t , j Table 5-1. Tolerance factor results for the new fuel storage rack Moderator density (gui/cc) 1.00 0.065 Tolerance number Delta k,,,/s l' N/A N/A 2 N/A N/A 3 0.00115 0.00036 4 0.00029 0.00010 5 0.00000 0.00053 6 0.00215 0.00074 7 0.00053 0.00012 Lateral assembly position 0.00011 0.00703 Overall tderance 0.00251 0.00710 uncertainty

  • see Section 2 for the description of tolerances I-7 l

Table 5-2. Final reactivities for the new fuel storage rack j Moderator density (gm/ec) 1.00 0.065 KENO keg 0.90113 0.95725 Biases Method 0.007600 0.007600 Uncertainty KENO 0.00113 0.00105 Method 0.001464 0.001464 I

Tolerance 0.00251 0.00710 Total 0.00185 0.00180 Degrees-of-freedom 53 48 95/95 factor 2.051 2.075 (95/95 factor)xTotal 0.00379 0.00374 Case 95/95 ken 0.916 0.976 l

l i

1 i

1 1

J

5. NEW FUEL STORAGE RACKS NEAD-SR-95/125.R1 + 25

A I

6. FUEL TRANSFER UPENDER This work shows that two assemblies with enrichments up to 5.0 wt. % U235 may be safely transponed using the fuel upender. It is shown that kerris less than the 0.95 limit assuming no credit for soluble boron.

Tolerance analysis The fuel tolerance factors are described in Section 2. Worst case lateral assembly position is l

assumed in the base model. The tolerance results are shown in Table 6-1. The tolerances were combined using the methodology described in Section 2.

Base model The following assumptions were used in the development of the base KENO model. The -

nominal fuel assembly dimensions are described in Section 2. Calculations were also performed to model the presence of a single assembly in the upender.

1. Both upender assembly positions contain unirradiated fuel assemblies with 5.0 wt. % U235 fuel. The UO 2density of the fuel stack is 10.25 gm/cc.
2. Water is set to the maximum density of 1.00 gm/cc.
3. No burnable poison, control element assembly, or other fixed poison is stored with the fuel assembly.
4. Three cases were examined. Case I assumed the two assemblies are completely ot1tside the canal (infinite moderator), Case 2 assumed the assemblies are centered in the transfer canal, and Case 3 assumed half of the assembly is in the canal and halfis outside the canal. Case 1 yielded the highest 95/95 kerrvalue.  ;

l The final 95/95 cigenvalue is 0.94997 as shown in Table 6-2 which is based on 1,500,000 neutron histories. This value is less than the 0.95 limit. The 95/95 eigenvalue for a single assembly in infinite moderator is 0.92. The minimum allowed separation distance between the loaded upender and a third assembly adjacent to the upender was determined to be 12 inches (assembly edge to assembly edge). At this separation distance, the final 95/95 eigenvalue is less than 0.95 assuming 0 ppm soluble boron.

Fuelhandling accidents For accident conditions, the double contingency principle applies which states it shall require two unlikely, independent, concurrent events to produce a criticality accident. Therefore, credit for - I soluble boron can be assumed in the evaluation of the accident where an assembly drops next to

6. FUEL TRANSFER UPENDER NEAD-SR-95/125.R1
  • 26

J '* l ,

I i

i the upender. The dropped assembly and the two assemblies in the upender are assumed to contain 5.0 wt. % fuel. The dropped assembly accident case yields a 95/95 kefrof 0.87 when assuming credit for 1000 ppm soluble boron. The soluble boron concentration during refueling operations is maintained at a much higher level. j Analysis conservatisms The following conservatisms are included in the analysis of the fuel transfer upender:

1. The base model assumes the fuel assemblies are in the most reactive g, ometry (smallest possible separation distance).
2. Water is set to the maximum density of 1.00 gm/ce.
3. The stainless steel upender structure is not modeled.
4. No credit is assumed for burnable poison in the fuel assemblies. l
5. No credit is assumed for soluble boron in the water.

Table 6-1. Tolerance factor results for the fuel transfer upender Tolerance number Delta k,,,r's 1' N/A 2 N/A j 3 0.00114 4 0.00029 5 0.0001I 6 0.00203 7 0.00049 Lateral assembly position" N/A Overall tolerance uncertainty 0.00240

  • scc Section 2 for the description of tolerances I-7

" The base model assumes worst case positioning.

I i

Table 6-2. Final reactivity for the fuel transfer upender  !

KENO ken 0.93640 Biases Method 0.007600 Uncertainty KENO 0.000'3 Method 0.001464 Tolerance 0.00240 Total 0.00159 Degrees-of-freedom 29 95/95 factor 2.232 (95/95 factor)xTotal 0.00356 Case 95/95 ken 0.94997

6. FUEL TRANSFER UPENDER NEAD-SR-95/125.R1 + 27

,r . ]

7. CONTAINMENT TEMPORARY STORAGE RACK This work shows that assemblies with enrichments up to 5.0 wt. % U235 may be safely stored in the containment temporary storage racks. It is shown that ke rris less than 0.95 for both normal and accident conditions.

i Tolerance analysis The fuel tolerance factors are described in Section 2. The uncertainty in the lateral position of the i fuel assembly in the rack cavity is considered as a tolerance by reducing the assembly pitch.  !

l The tolerance results are shown in Table 7-1. The tolerances were combined using the methodology described in Section 2.

Base model The following assumptions were used in the development of the base KENO model. The ,

nominal fuel assembly dimensions are described in Section 2. I

1. All four positions in the rack contain unirradiated fuel assemblies with 5.0 wt. % U235 fuel.

The UO2density of the fuel stack is 10.25 gm'ce The nominal value of 17.812" is assumed for spacing between assemblies.

2. Water is set to the maximum density of 1.00 gm/cc.

1

3. No burnable poison, control element assembly, or other fixed poison is stored with the fuel i assembly.
4. The wall supporting the storage rack is modeled as concrete followed by a reflective boundary condition. The upper, lower, and other sides of the rack are modeled as water followed by reflective boundary conditions.
5. No rack structural material is modeled.

The final 95/95 eigenvalue is 0.92 as shown in Table 7-2. This value is less than the 0.95 limit. l FuelhandIIng accidents For accident conditions, the double contingency principle applies which states it shall require two unlikely, independent, concurrent events to produce a criticality accident. Therefore, credit for soluble boron can be assumed in the evaluation of the accident where an assembly drops next to the containment temporary storage rack. The dropped assembly and all assemblies in the rack are assumed to contain 5.0 wt. % fuel. The dropped assembly accident case yields a 95/95ek rrof 0.83 when assuming credit for 1000 ppm soluble boron. The soluble boron concentration during refueling operations is maintained at a much higher level.

7. CONTAINMENT TEMPORARY STORAGE RACK NEAD-SR-95/125.R1 + 28 j

J '. A Analysis conservatisms The following conservatisms are included in the analysis of the containment temporary fuel storage racks:

1. No neutron leakage in the x-y directions is assumed.
2. Water is set to the maximum density of 1.00 gm/cc.
3. None of the stainless steel rack structure is modeled.
4. No credit is assumed for burnable poison in the fuel assemblies.
5. Credit for soluble boron is assumed only for accident cases consistent with the double contingency principle.

Table 7-1. Tolerance factor results for the containment temporary storage rack Tolerance number Delta k,js

!" N/A 2 N/A 3 0.00114 4 0.00029 5 0.00000 6 0.00214 7 0.00053 Lateral assembly position 0.00087 i Overall tolerance uncertainty 0.00265

  • see Section 2 for the description of tolerances I-7 Table 7-2. Final reactivities for the containment temporary storage rack KENO k., 0.90537 Biases Method 0.007600 lincertainty KENO 0.00!!8 Method 0.001464 Tolerance 0.00265 l Total 0.00188 Degrees-of-freedom 57 9535 factor 2.034 (95/95 factoexTotal 0.0382 Case 95/9_5 k., 0.92 J

l

7. CONTAINMENT TEMPORARY STORAGE RACK NEAD-SR-95/125.R1
  • 29

j '. i

8. REFERENCES
1. " SCALE 4 - A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation",CCC-545,NUREG/CR-0200 REV.4 (ORNL/NUREG/CSD-2/R4)

Vols. I, II, and Ill, Oak Ridge National Laboratory, February 1990.

2. M. N. Baldwin et al.," Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel", BAW-1484-7, Babcock and Wilcox Company, July 1979.
3. S. R. Bierman, E. D. Clayton, and B. M. Durst, " Critical Separation Between Suberitical Clusters of 4.29 wt. %235 U Enriched UO2Rods in Water with Fixed Neutron Poisons",

NUREG/CR-0073, Battelle Pacific Northwest Laboratories, May 1978.

l 4. S. R. Bierman, E. D. Clayton, and B. .M. Durst, " Critical Separation Between Suberitical Clusters of 2.35 wt. % 235 U Enriched UO2Rods in Water with Fixed Neutron Poisons",

PNie-2438, Battelle Pacific Northwest Laboratories, October 1977.

5. S. E. Turner and M. K. Gurley, " Evaluation of AMPX-KENO Benchmark Calculations for High-Density Spent Fuel Storage Racks", Nuclear Science and Engineering,80:230-237 (1932).
6. W. Marshall et. al., " Criticality Safety Criteria", Transactions ofthe American Nuclear Society,35:278-279 (1980).
7. R. E. Oden et. al., Tables for Normal Tolerance Limits, Sampling Plans, and Screening, Marcel Dekker, Inc., New York, New York,1980.
8. "CASMO-3: A Fuel Assembly Burnup Program, Version 4.7", Studsvik/NFA-89/3.R2, March 1992.

1 l

9. " SIMULATE-3: Advanced Three-Dimensional Two-Group Reactor Analysis Code",  ;

Studsvik/SOA-92/01, April 1992. l 1

10. Letter from John F. Stolz and Robert A. Clark (NRC) to John M. Griffin (AP&L), l L " Amendments Nos. 76 and 43 to Facility Operating Licenses Nos. DPR-51 and NPF-6 for

! Arkansas Nuclear One, Units Nos. I and 2 (ANO-l & 2)", Docket Nos. 50-313 & 50-368, April 15,1983.

I h

1

8. REFERENCES NEAD-SR-95/125.R1
  • 30
j. t i
11. Letter from Robert S. Lee (NRC) to John M. Griffin (AP&L)," Issuance of Amendment No.

71 to Facility Operating License NPF-6, Arkansas Nuclear One, Unit No. 2", Docket No.

50-368, April 16,1986.

12. K. Lindquist et. al., "Boraflex Test Results and Evaluation", EPRI TR-101986, Project 2813-04, February 1993.
13. " Qualification of Reactor Physics Methods for the Pressurized Water Reactors of the Entergy System", TAC Nos. M88565, M88566, and M88567, Document ENEAD-01-A.R0, Entergy Operations, Inc., September 1995.
14. "U.S. Nuclear Regulatory Commission Standard Review Plan", NUREG-0800.
15. "American National Standard Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants", ANSI /ANS-57.2.1983, October 1983.
16. "American National Standard Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants", ANSl/ANS-57.3-1983, January 1983.
17. Brian K. Grimes,"U.S. NRC Letter to All Power Reactor Licensees", Docket No. 50-289, April 14,1978.
18. Letter from Chandu P. Patel (NRC) to Ross P. Barkhurst (EOI), " Issuance of Amendment No.108 to Facility Operating License NPF-38, Waterford Steam Electric Station, Unit 3 (TAC NO. M91460)", Docket No. 50-382, June 14,1995.

l l

8. REFERENCES NEAD-SR-95/125.R1 + 31 i

s '. :d i

APPENDIX A: BENCHMARK OF SCALE-4 A benchmarking of the SCALE-4 system of codes against critical experiments is performed in order to qualify and validate its use for performing fuel storage rack criticality safety analyses at Entergy Operations, Inc. plants.

The benchmarked SCALE-4 codes includes the CSAS25 Control Sequence of the ORNL Criticality Safety Analysis Sequence No. 4 (CSAS4). The functional modules sequentially executed by the CSAS25 control module are the BONAMI-S code, the NITAWL-S code, and the KENO V.a code [ Reference 1].

A total of twenty-one critical experiments are used to benchmark these codes. These criticals are selected from a list of seventy-five critical experiments conducted by Babcock and Wilcox and Pacific Northwest Laboratory [ References 2,3, and 4]. The twenty-one criticals are chosen because their fuel characteristics, lattice geometry's water gap spacing and materials are l reasonably representative of those found in Entergy Operations, Inc. fuel storage rack arrays.

i An evaluation of the 218 group cross section library benchmarking results identifies one significant bias in the SCALE calculated keffs. The bias is an observed trend toward over-prediction of reactivity and increasing Boron loading in the Boral plates for four of the B&W cores, as also reported for the 123 group library in Reference 5. These four cases are corrected for the Boron loading bias before further data reduction is performed. For conservatism, this credit is not taken in analyzing fuel storage racks.

Statistical analysis of the twenty-one calculated keffs, using Criterion 2 [ Reference 6], gives a mean kefrof 0.992402 with a Monte Carlo uncertainty of10.000987, a method bias of 0.007600, and a method uncertainty of10.001464. Table A-1 provides a summary of the benchmark statistics.

APPENDIX A: BENCHMARK OF SCALE-4 NEAD-SR-95/125.R1 + 32

\

s' og? i

e.

Table A-1. Summary of SCALE-4 benchmark )1 Ca'e Case name Enrichment SCALE Standard Corrected Number of Number of number (wt.% kefy deviation kefr generations neutrons per U235) generation 1 BWil . 2.46 0.99I41 0.00082 0.99I41 500 1000 f B Will 2 2.46 0.98996 0.00086 0.98996 500 1000 3 BWXI 2.46 0.98837 0.00091 0.98837 500 1000 4 BWXIll - 2.46 0.99761 0.00095 0.99292 500 1000 )

5 BWXIV 2.46 0.99522 0.00096 0.99209 500 1000 l 6 B W X Vil 2.46 0.99233 0.00088 0.99364 500 1000 l 7 BWXIX 2.46 0.98988 0.00089 0.99181 500 1000 i 8 PNL5 2.35 0.99329 0.00099 0.99329 500 1000 l 9 PNL26 2.35 0.99340 000095 0.99340 $00 1000 t 10 PNL28 2.35 0.99384 0.00093 0.99384 500 1000  !

II PNL32 2.35 0.99392 0.00092 0.99392 500 1000 ,

12 PNL33 2.35 0 99486 0.00098 0.99486 500 1000 l 13 PNL38 2.35 0.99276 0.00102 0.99276 500 1000 l 14 PNL39 2.35 0.99487 0.00096 0.99487 500 1000 , l 15 PNL8 4.29 0.99075 0.001IO 0.99075 500 1000  !

16 PNL9 4.29 0.99254 0.00111 0.99254 500 1000  !

17 PNL10 4.29 0.99491 0.00106 0.99491 500- 1000 {

18 PNL11 4.29 0.98996 0.00112 0.98996 500 1000  ;

19 PNL12 4.29 0.99137 0.00104 0.99137 500 . 1000 l 20 PNL13 4.29 0.99243 0.00110 0.99243 500 1000 ,

21 PNL32 4.29 0.99169 j 0.00109 0.99169 500 1000

  • Avg. Corrected Keff = 0.992402 l Monte Carlo Uncertainty = 0.000987

{

Method Blas = 0.007600 i Method Uncertainty = 0.001464

  • l i

I APPENDIX A: BENCHMARK OF SCALE-4 NEAD-SR-95/125.R1

  • 33 ,

l

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