ML20071G840

Nuclear Design Analysis Rept for Surry Nuclear Power Station High Density Fuel Storage Racks

ML20071G840
Person / Time
Site:	Surry
Issue date:	03/14/1980
From:	NUCLEAR ENERGY SERVICES, INC.
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ML18139A244	List: ML20071G840
References
81A0494, 81A494, NUDOCS 8005200676
Download: ML20071G840 (29)

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DOCUMENT NO.

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PAGE 1 np 29

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i NUCLEAR DESIGN ANALYSIS REPORT.

J FOR THE SURRY NUCLEAR POWER STATION HIGH DENSITY FUEL STORAGE RACKS l

Prepared Under Project 3157 for the Virginia Electric Power Company by J, Nuclear Energy Services, Inc.

j .Danbury, Connecticut 06810' i

O Protect Application Prepared By Oa

~ APPROVALS '

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NUCLEAR ENERGY SERVICES. INC. DOCUMENT NO. 81A0494 PAGE 3 op 29 TABLE OF CONTENTS

1.

SUMMARY

5

2. INTRODUCTION 6

3. DESCRIPTION OF SPENT FUEL STORAGE RACKS 7

4. CRITICALITY DESIGN CRITERION AND CALCULATIONAL ASSUMPTIONS 9 4.1 Criticality Design Criterion 9 4.2 Calculational Assumptions 9

5. CRITICALITY CONFIGURATIONS 11 5.1 Normal Configurations 11 5.1.1 Reference Configurations 11 5.1.2 Eccentric Configuration 11 5.1.3 Fuel Design Variation 12 5.1.4 Fuel Rack Cell Pitch Variation 12 5.1.5 Fuel Rack Wall Thickness Variation 12 5.1.6 " Worst Case" Normal Configuration 12 5.2 Abnormal Configuration 12 5.2.1 Single Cell Displacement 12 5.2.2 Fuel Handling Incident 13 5.2.3 Pool Temperature Variation 13 5.2.4 Fuel Drop Incident 13 5.2.5 Seismic Incident 14 5.2.6 " Worst Case" Abnormal Configuration 14

6. CRITICALITY CALCULATIONAL METHODS 19 6.1 Method of Analysis 19 6.2 Benchmark Calculations 19 6.3 Uncertainties 20 6.4 Computer Codes 20 6.4.1 NITAWL 20 6.4.2 KENO IV 20 l

6.4.3 HAMMER , 20 6.4.4 EXTERsi!NATOR 20 l

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81A0494 NUCLEAR ENERGY SERVICES. INC. DOCUMENT NO.

PAGE 4' OF 29

7. RESULTS OF CRITICALITY CALCULATIONS 22 7.1 K,;f Values for Normal Configurations 22 7.1.1 Reference Configuration 22 7.1.2 Eccentric Configuration 22 7.1.3 Fuel Design Variation 22 7.1.4 Fuel Rack Cell Pitch Variation 22 7.1.5 Fuel Rack Cell Wall Thickness Variation 23 7.1.6 Combined Effects of Normal Variations on the Q Reference Configuration K df 23 7.2 K,ff Values for Abnormal Configurations 24 7.2.1 Single Cell Displacement 24 7.2.2 Pool Temperature Variation 24 7.2.3 Seismic Event 24 7 '.4 " Worst Case" Abnormal Configuration 24 7.2.5 Effects of Calculational Uncertainties 25

8. REFERENCES TABLES O

b 5.1 Parameters of 17x17 Westinghouse Fuel Assemblies 18 7.1 Parameters and Results of Criticality Calculations 26 FIGURES 3.1 Fuel Rack 8 5.1.A Reference Configuration 15 5.1.B Eccentric Configuration 16 5.2 Single Cell Displacement - 17 6.1 Quarter Assembly Repeating Array 21 7.1 K,gg vs Fuel Rack Cell Pitch 27 7.2 K,gf vs Pool Water Density 28 r:w .Nas aos sn* /

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. 1 81A0494 NUCLEAR ENERGY SERVICES, INC. DOCUMENT NO.

PAGE 5 op 29

1.

SUMMARY

A detailed nuclear analysis has been performed to demonstrate that for all anticipated normal and abnormal configurations of fuel assemblies within the fuel storage racks, the K,ff of the system for 4.1 w/o Westinghouse fuel assemblies is less than the criticality criterion of <0.95. Conservative assumptions about the fuel assemblies and racks have been used in the calculations. The normal configurations considered in the nuclear analysis included the reference configuration (an array of square stainless steel boxes spaced 14.0 inches on centers with centrally positioned fuel), the eccentric positioning of fuel within the storage boxes and the variations permitted in fabrication of the principal fuel rack dimensions. The abnormal configurations included the mislocation of a storage box, box displacement due to a seismic event, and spent fuel pool temperature variations.

The calculations were carried out using the Monte Carlo transport theory code KENO-g IV to evaluate the reference configuration K,ff. Other calculations to determine the V sensitivity of K,ff to the normal and abnormal variations mentioned above were performed using the diffusion theory code EXTERMINATOR-2. The final calculated K,ff for the system including normal and abnormal variations and the effects of calculational uncertainty is 0.938. This value meets the criticality design criterion and is substantially below 1.0. Therefore,it has been concluded that the Surry Nuclear Power Station high density storage racks when loaded with the specified fuel are safe from a criticality standpoint.

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NUCLEAR ENERGY SERVICES, INC. DOCUMENT NO. 81A0494 PAGE 4 OF 29

2. INTRODUCTION The NES design for high density fuel storage racks for Surry consists of a square array of stainless steel boxes (9.12 inches OD with 0.090 inch walls) spaced 14.0 inches on centers. This configuration provides water gaps between the boxes which act as thermal flux traps for neutrons escaping from the fuel assemblies located within the boxes. This flux trap design results in a structurally sound rack which does not depend on additional poisons to achieve a high storage density, A description of the racks !s

/'h given in Section 3.

(d' A detailed nuclear analysis has been performed to demonstrate that, for all antici-pated normal and abnormal configurations of fuel assemblies within the fuel storage racks, the K,ff of the system is substantially below 1.0. Certain conservative assumptions about the fuel assemblies and racks have been used in the calculations.

These are described in Section 4 along with the criticality design criterion for the fuel storage racks. -

The reference configuration which is the basis of the criticality calculations consists m of an array of square stainless steel boxes (9.12 inches OD with a wall thickness of 0.090 inches) spaced 14.0 inches on centers and with fuel assemblies centrally located within the boxes. Variations from this reference configuration were also studied and included the effects of dimensional and spacing variations, fuel enrichment changes, water temperature increases and mislocations of fuel assemblies and boxes. These variations are described in detailin Section 5.

Reference configuration criticality calculations were performea with the transport l thecry Monte Carlo code combination NITAWL/ KENO IV. Sensitivity calculations for normal and abnormal variations on the reference configuration were performed using the diffusion theory code combination HAM'MER/EXTERM'INATOR. Discussion of computer codes can be found in Section 6. The results of the criticality analyses are i presented in Section 7.

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NUCLEAR ENERGY SERVICES. INC. DOCUMENT NO.81A0494 PAGE 7 OF 29 1

3. DESCRIPTION OF SPENT FUEL STORAGE RACKS Each fuel storage rack contains 36 storage locations spaced 14.0 inches on centers in an 6x6 square array (see Figure 3.1). Each storage location consists of a Type 304 stainless steel square box, 9.12 inches in outside dimension with 0.090 inch thick walls except the corner boxes which are 9.56 inches OD with 0.25 inch walls. The spent fuel assembly is located within the stainless steel box.

The square boxes are ~172 inches tall so that the 144 inch active length of each fuel assembly is entirely enclosed by the stainless steel box.

Between boxes is a 4.88 inch wide gap which is filled with water when the rack is located in the spent fuel storage pool. Within this gap are also located certain structural grid members, clips and bracing which locate and space the boxes. This structural material occupies only a small fraction of the water gap at essentially two widely separated elevations.

Each storage rack has structure mounted on the outside which will assure that the center-to-center spacing between cells in adjacent racks is maintained at 14.0 inches d or greater. .

Gua'rd structures are provided at the upper grid of pe-ipheral racks as required to preclude the inadvertent positioning of a fuel assembly too close to a storage rack during fuel handling. The' structure will ensure that the center-to-center distance for such incidents will be in excess of 17 inches.

1 Type 304 stainless steel is used for the square boxes and all of the principal structural grid members, clips and bracing.

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4. CRITICALITY DESIGN CRITERION AND CALCULATIONAL ASSUMPTIONS 4.1 CRITICALITY DESIGN CRITERION Determination of a satisfactory value of K,ff for a spent fuel pool requires considera-tion of safety,licensability, and storage capacity requirements. These factors demand aK eff substantially below 1.0 for safety and licensability but high enough to achieve the required storage capacity.

n The published position of NRC on fuel storage criticality is presented in Section 9.1.2 of the NRC Standard Review Plan (Ref.1) which states the following:

" Criticality li1 formation (including the associated assumptions and input parame-ters)in the SAR must show that the center spacing between assemblies results in a subcritical array. A K,ff of less than about 0.95 for this condition is acceptable."

The NRC, in evaluating the design, will " check the degree of subcriticality provided, along with the analysis and the assumptions". In addition, it has been suggested that transport theory calculational methods are more accurate than diffusion theory methods because of the large water gaps present in PWR rack designs.

On the basis of this information, the following criticality design criterion has been established for the Surry Nuclear Power Station high density fuel storage racks: "The multiplication constant (K,ff) shall be less than 0.95 for all normal and abnormal configurations as confirmed by transport theory."

4.2 CALCULATIONAL ASSUMPTIONS

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The following conservative assumptions have been used in the criticality calculations performed to verify the adequacy of the rack design with respect to the criticality design criterion:

1. The pool water has no soluble poison.

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DOCUMENT NO.

lA05 NUCLEAR ENERGY SERVICES. INC.

PAGE 10 op 29

2. The fuei assemblies have no burnable poison.

3. The fuel is fresh and of a specified enrichment higher than that of any fuel available.

4. The rack configuration is infinite in all three dimensions.

5. No credit is taken for structural material other than the stainless steel box.

6. All stainless steel boxes are assumed to be 0.090 inches thick. The minimum allowable thickness for the stainless steel boxes is 0.090 inches O

Q except the corner boxes which have a minimum wall thickness of 0.240 inches.

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NUCLEAR ENERGY SERVICES, INC. DOCUMENT NO. 81A0494 M PAGE II OF 29

5. CRITICALITY CONFIGURATIONS In order to verify the design adequacy of the Surry Nuclear Power Station high density storage rack it is necessary to establish the multiplication constants for the various arrangements or configurations of fuel assemblies and storage cells that are possible within the racks. These arrangements or configurations can be classified as either normal or abnormal configurations. Normal configurations result from the placement of fuel within the storage cell location, and the variation in fuel storage rack

( dimensions permitted in fabrication. Abnormal configurations are typically the result of accidents or malfunctions such as the seismic event, a malfunction of the fuel pool cooling system (abnormal changes in pool water temperature), a dropped fuel assem-bly, etc. The following sections present the normal and abnormal configurations which have been considered in this analysis.

5.1 NORM AL CONFIGURATIONS

. 5.1.1 Reference Configurations The reference configuration consists of an infinite array of storage cells having nominal dimensions each containing a 17x17 Westinghouse fuel assembly of 4.1 w."o enrichment positioned centrally within the cell. The storage cells or boxes are 9.12 inches in outside dimensions, have 0.090 inch walls and are spaced 14.0 inches on centers. The spent fuel pool water temperature is assumed to be 68 F.

This configuration is shown in Figure 5.1.a.

3.1.2 Eccentric Configuration It is possible for a fuel assembly not to be positioned centrally within a storage cell or box because of the clearance allowed between the assembly and the box wall. This clearance is approximately 1/4 inch on each side of the fuel assembly.

If one assembly is displaced 1/4 inch from its nominal centered positioned and if all other assemblies remain centered, the effect on K eff si negligibly small (less l than 0.001). The most unfavorable condition occurs if each of four adjoining assemblies is diagonally offset so as to be as close as possible to the other three.

The effect on K,ff of this condition was determined using the eccentric configuration shown in Figure 5.1.b.

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• . l 81A0494 NUCLEAR ENERGY SERVICES, INC. DOCUMENT NO.

PAGE 12 O F_._79 l

5.1.3 Fuel Design Variation The Surry Nuclear Power Station fuel racks are designed to accommodate both 15x15 and 17x17 fuel designs. Calculations performed by NES show that racks with the 17x17 fuel assemblies were slightly more reactive than the racks with the 15x15 fuel assemblies with equal enrichment. Therefore, NES selected the 17x17 fuel assembly with 4.1 w/o enrichment for the detailed criticality analysis of the Surry fuel storage racks.

5.1.4 Fuel Rack Cell Pitch Variation Calculations were performed to determine the sensitivity of K df t changes in cell pitch (center-to-center spacing). The cell pitch was reduced to 13-15/16 inches and to 13-7/8 inches. The criticality configuration was similar to that of the reference configuration except for the obvious change in center-to-center spacing.

5.1.5 Fuel Rack Cell Wall Thickness Variation Calculations were performed for wall thicknesses of 0.090 and 0.095 inches.

5.1.6 " Worst Case" Normal Configuration O

Q The " worst case" configuration considers the effect of eccentric fuel assembly positioning and minimum average cell pitch (center-to-center spacing) permitted by fabrication.

5.2 ABNORM AL CONFIGURATIONS 5.2.1 Single Cell Disolacement i

Welded clips and shims position the stainless steel cells or boxes centrally within the grid members of the rack structure. If the welds on one of these clips or shims fails, the associated box cannot be displaced. However, calculations were performed to determine the sensitivity of K,ff for the reference configuration to single cell displacement. A cell was arbitrarily displaced 0.25 inches from its proper location as shown in Figure 5.2. In this configuration, the water gap l between the two close boxes is reduced from 4.88 inches to 4.63 inches while the gap on the other side increases to 5.13 inches.

P!mu a NES 208 $/79

NUCLEAR ENERGY SERVICES. INC. DOCUMENT NO.

PAGE 13 OF 29 5.2.2 Fuel Handling incident Structure is provided on the peripheral fuel storage locations which precludes the positioning of a fuel assembly during fuel handling such that the center-to-center spacing between this assembly and the nearest assembly in the rack would be less-than 17 inches.

At this separation and with stainless steel boxes surrounding all but the improperly positioned fuel, the K,ff value will be substantially below the criticality design criterion. Reference 2 verifies this by showing that bare 4.1 w/o,17x17 Westinghouse fuel assemblies spaced 14.2 inches on centers will have a K ,ff value less than 0.95 including variations in configurations and uncertain-ties in calculations. It has been concluded that this type of incident need not be considered fur't her in this analysis.

5.2.3 Pool Temperature Variation Calculations were performed to determine the sensitivity of K gf for the reference configuration to variations in the spent fuel pool temperature. The pool temperature was varied from ~40 F, where water density is a maximum, to

~250 F, the approximate boiling point of water near the bottom of the fuel rack.

Q In addition, the effect of voids in the water was studied.

5.2.4 Fuel Drop Incident The maximum height through which a fuel assembly can be dropped onto the fuel storage racks is limited to 23.5 inches. The dropped fuel assembly will most likely impact the flared tops of the fuel storage rack cells or boxes.

While minor deformation of the flared tops will occur, the close proximity of the upper grid structure to the. Impact point will preclude any significant lateral displacement of the storage cells. Consequently, the change in K df will be negligible. However, it is possible for a dropped fuel assembly to enter a box cleanly and impact directly on the fuel stored in the box. The effect of this type of fuel drop incident was evaluated from a criticality veiwpoint by assuming that the stored assembly would be compressed axially. A calculation based on an axial compression of 2 feet yielded a 0:06 decrease in km of the fuel cell. It has been concluded, therefore,, that this incident would reduce K,ff and need not be considered further in this analysis.

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,- .. l 81A0494 NUCLEAR ENERGY SERVICES. INC. DOCUMENT NO.

PAGE I4 OF 29 5.2.5 Seismic Incident The seismic analyses indicate that the maximum rack structure deflections will be very small (less than 0.120 inches). These deflections have negligible effect on K 'since they do not change the center-to-center spacing between the df storage cells or boxes significantly.

The maximum deflection of the storage cells or boxes due to a seismic event occurs at the middle of the box and is less than 0.050 inches. The effect of box deflections on K df is negligible since the average center-to-center spacing a

between cells or boxes will not change appreciably if the boxes deflect independently in random directions or act together in a single direction.

5.2.6 " Worst Case" Abnormal Configuration The " worst case" abnormal configuration considers the effect of the most adverse abnormal condition in combination with the " worst case" normal config-uration. The results for the " worst case" abnormal configuration are presented in Section 7.2.4.

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81A0 W NUCLEAR ENERGY SERVICES. INC. DOCUMENT NO.

3 PAGE 18 OF J.9 TABLE 5.1 PARAMETERS OF 17x17 AND 15x15 WESTINGHOUSE FUEL ASSEMBLIES 17x17 15x15 Mass of UO2 in Assembly,Ibs 1154 1122 Number of Fuel Rods 264 204 Number of Guide Tubes 25 21 Clad, ID, inches 0.329 0.3734 Clad, OD, inches 0.374 0.422 Clad Thickness, inches 0.0225 0.0243 Clad Material Zr Zr Spacer Mass,Ibs in Active Fuel Length 12.0 10.5 Spacer Material Inc 718 Inc 718 Number of Spacers 8 7 Pitch Between Fuel Rods, inches 0.496 0.563 Guide, Tube OD, inches 0.482 0.512 Guide, Tube ID, inches 0.450 0.455 O

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DOCUMENT NO. 81A0494 NUCLEAR ENERGY SERVICES. INC.

PAGE 19 OF_ 23

6. CRITICALITY CALCULATIONAL MEWIODS 6.1 METHOD OF ANALYSIS For the reference configuration discussed in Section 5.1.1, the K,ff was determined from a three-dimensional Monte Carlo calculation using NITAWL/ KENO IV with the 123 group XSDRN cross-section set. Check calculations of the reference configura-tion as well as the parametric studies were performed with two-dimensional diffusion theory using HAMMER and EXTERMINATOR. In both the Monte Carlo and diffusion theory methods, an infinite array of fuel assemblies loaded in spent fuel storage d

locations was represented by use of appropriate boundary conditions. An infinite array is used for two reasons: (1) an infinite array has a conservatively higher value of K,ff and (2) the problem can be suitably represented by a repeating portion of the array.

Figure 6.1 shows a representation of one quarter of a storage location with reflecting boundaries on all sides. This duplicates an infinite array of storage locations.

6.2 BENCHM ARK CALCULATIONS

, in order to establish the accuracy of the computer codes used for this analysis, several benchmark calculations have been performed both at NES and elsewhere (Ref. 3,4).

The NITAWL/ KENO IV code combination using the 123 group XSDRN cross-section set n

Q was benchmarked against several recent criticality experiments. Calculated K,ff values for experimental configurations similar to the Surry high density spent fuel storage racks were observed to be ~ 2% higher than the experimental values. No credit will be taken for this conservatism.

Both HAMMER and EXTERMINATOR are used by NES as versions available at Combustion Engineering at Winsdor Locks, Connecticut. This combination has been benchmarked against a cold critical experiment performed at the Lacrosse Boiling Water Reactor with excellent results (Ref. 5). The calculated K,ff differed from the experimental value by only 0.0017.

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DOCUMENT NO. 81A0494

. NUCLEAR ENERGY SERVICES. INC.

PAGE 20 o p_,,,_29 6.3 UNCERTAINTIES The reference configuration K,ff value calculated by KENO IV forms the basis for the final reported K,ff value. To this value we must attach an uncertainty. The errors in Monte Carlo criticality calculations can be divided into two classes:

1. Uncertainty due to the statistical nature of the Monte Carlo methods.

2. Errors due to bias in the calculational technique.

The first class of errors can be reduced by simply increasing the number of neutrons tracked. For rack criticality calculations, the number of neutrons tracked is selected to reduce this error to less than 1%, and in this case 2 0.006. The second class of errors has already been discussed in Section 6.2.

No credit will be taken for the 2% experimental bias. However the statistical error will be conservatively set as 20 01.

6.4 COMPUTER CODES 6.4.1 NITAWL NITAWL performs resonance self-shielding correction and creates a formatted O

(j working library based on the XSDRN cross-section set for use in KENO IV using the Nordheim Integral Method.

6.4.2 KENO IV (Ref. 6)

KENO IV is a 3-D multigroup Monte Carlo code used to determine K,ff.

6.4.3 H AMMER (Ref. 7)

HAMMER is a multigroup integral transport theory code which is used to calculate lattice cell cross-sections for diffusion theory codes. This code has been extensively benchmarked against D 2 0 and light water moderated lattices j with good results.

l 6.4.4 EXTERMINATOR (Ref. 8)

EXTERMINATOR is a 2-D multigroup diffusion theory code used with input from HAMMER to calculate K,ff values.-

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A0494 NUCLEAR ENERGY SERVICES, INC. DOCUMENT NO.

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NUCLEAR ENERGY SERVICES. INC. DOCUMENT NO. 81A0494 PAGE 22 op 29

7. RESULTS OF CRITICALITY CALCULATIONS The following presents the results of calculations for each of the configurations discussed in Section 5 and subsequent contribution to the final rack K,ff.

7.1 K,ff VALUES FOR NORMAL CONFIGURATIONS 7.1.1 Reference Configuration m

The Keff value for the reference configuration described in Section 5.1.1 was (v) calculated to be 0.914 using NITAWL/ KENO IV.

7.1.2 Eccentric Configuration The K,ff value for the eccentric configuration described in Section 5.1.2 (four assemblies displaced diagonally toward each other the maximum amount allowed by clearance) was determined to increase over the reference configuration value by aK = 0.006.

7.1.3 Fuel Design Variation There are two fuel designs used at the Surry facility which can be placed in the high density fuel storage racks (see Section 5). Calculations have been performed which show the 17x17 Westinghouse design to have a K,ff value 0.007 higher that the 15x15 Westinghouse design of the.same enrichment. Variation of K,ff with fuel enrichment was calculated using both KENO IV and EXTERMIN-ATOR. Results show that the fuel rack K,ff increases 0.056 per weight percent increase in fuel enrichment. Since the fuel enrichment used in this analysis (4.1 w/o) is higher than any fuel available at the Surry facility, the reference configuration value of K,ff will be conservative with respect to enrichment and fuel design.

7.1.4 Fuel Rack Cell Pitch Variation The Keff variation for fuel rack cell pitch values ranging down to 13-7/8 inches are shown in Figure 7.1.The cell pitch of interest, of course, is the minimum value that can occur'in fabrication. The mechanical design of the fuel rack is such that the average pitch (center-to-center distance) between the cells or boxes in one rack is 14.0f, 0.062 inches. The change in K,ff for a 0.062 inch reduction in the average pitch is AK = 0.002.

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NUCLEAR ENERGY SERVICES, INC. DOCUMENT NO. 8.1A0494 PAGE 71 OF 79 7.1.5 Fuel Rack Cell Wall Thickness Variation Fuel rack cell wall thickness will be controlled to 0.090 +0 ,0.000 inches. The change in K,ff due to variation in cell wall thickness was calculated to be AK =

0.003 for a 0.005 inch increase in cell wall thickness. Since 0.090 inch is the minimum allowed value of cell wall thickness, the reference configuration value of K eff will always be conservative relative to this parameter.

7.1.6 Combined Effects of Normal Variations on the Reference Configuration K di To establish the maximum variation of the reference configuration K,ff due to normal variations, we statistically add the individual positive components. For this case the positive components are those of:

AK

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ef f Minimum Average Pitch +0.002 Eccentric Positioning of Four Assemblies + 0.006 Ch V Statistical combination of the above normal variations yields:

AK,ff =k(0.002)2 + (0.006)2

= 0.007 The " worst case" normal configuration is defined as the reference configuration K,ff value plus the variation in K,ff due to the combined effect of all adverse normal variations.

Worst case normal K,ff = 0.914 + 0.007

= 0.921

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I DOCUMENT NO. 81A0494 UCLEAR ENERGY SERVICES, INC.

PAGE J4 OF 29 7.2 K,ff VALUES FOR ABNORMAL CONFIGURATIONS 7.2.1 Single Cell Displacement Calculations were performed in which a single fuel cell was arbitrarily displaced 0.25 inches towards an adjacent cell. The resultant 4K was 0.001.

7.2.2 Pool Temperature Variation The K ,ff as a function of pool water temperature and water density is presented in Figure 7.4 for pool temperatures up to 250 F. The maximum K,ff value,

(] reached at approximately 250 F, is 0.007 greater than the K,ff value for the reference configuration.

7.2.3 Seismic Event The maximum deflection of a storage cell or box in the active fuel region is less than 0.050 inches for the Safe Shutdown Earthquake (SSE). As stated in Section 5.2.5, cell or box deflections will not result in significant reductions in the average cell pitch. For conservatism, however, it will be assumed that the SSE reduces the average pitch by the cell deflection, 0.050 inches. A reduction in cell pitch of 0.050 inches will increase K,ff by 0.002. If the SSE is assumed to occur with the pool temperature at 170 F (the maximum temperature during a p) full core off load) the increase in K,ff due to the combined seismic and temperature effect is 0.006.

7.2.4 " Worst Case" Abnormal Configuration The " worst case" abnormal configuration combines the change in K,ff due to the occurrence of the most adverse abnormal condition (increase in pool tempera-ture) with the K,ff value associated with the worst case normal configuration.

K

--eff

1. Worst Case Normal Configuration 0.921 (per Section 7.1.6)

2. Most Adverse Abnormal Configuration 0.007 (pool temperature increase per Section 7.2.2)

3. Resulting K,ff 0.928

, P2feu o pees 206 5/F9 L

DOCUMENT NO. 81A0494 mr I,NUCt. EAR ENERGY SERVICES, INC. PAGE__25 OF _29 7.2.5 Effects of Calculational Uncertainity As discussed in Section 6, a value of 0.01 will be added to the result of 0.928 obtained thus far to account for any statistical fluctuations in the KENO IV result. The final resulting K,ff for the Surry high density spent fuel storage racks is 'O.938. This conservative resuit meets the criticality design criterion set forth in Section 4 and clearly shows that the racks are safe from a criticality standpoint.

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S TABLE 7.1 b t

4 PARAMETERS AND RESULTS OF CRITICALITY CALCULATIONS h

as Water Box Wall K Q Enrichment Spacing Temp Density Thickness' Ng O Configuration '

(w/o) (inches) [F] (gm/cc) (inches) A K ,gg $

s Reference Configurations 4.10 14.0 68 0.998 0.090 0.914 Q

Maximum Water Density,39 F 'us 4.10 14.0 39 1.000 0.090 3.50-4 -

90 F b

4.10 14.0 90 0.995 0.090 1.25-3 150 F 4.10 14.0 150 0.980 0.090 3.60-3 212 F 4.10 14.0 212 0.958 0.090 5.92-3 250 F 4.10 14.0 250 0.941 0.090 6.16-3 250 F, voided 4.10 14.0 250 0.925 0.090 3.86-3 Close Spacing 4.10 13-15/16 68 0.998 0.090 1.75-3 13-7/8 68 0.998 0.090 3.68-3 8 o

Eccentric Position 4,10 14.0 68 0.998 0.090 5.69-3 [

Displaced Can (base can) 4.10 14.0 68 0.998 0.090 1.42-3 ,

> 2 Displaced Can (can moved K") 4.10 0.090 1.73-3 0 .O 8

0 m

• '

• s',' 6 ,.. ,,' Docum:nt No.: 81A0494 s Page 27 of 29 0.004

• 0.003 - .

0.002 -- g A K,ff O

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0.001 -

4

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Figure 7.1 f

' A K ,ff vs Pitch For 17 x 17 Westinghouse Fuel, 4.1 w/o,0.090" Stainless Steel Boxes,68 E e

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• Documnt No. : 81A0494 i

i Page 28 of 29 4 ,

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Water Density, gm/c[n

! I 1 40 Water Temperature, F 250-Figure 7.2 K,ff vs Water Density for 17 x 17 4.1 w/o Westinghouse Fuel,.0.090" Stainless Steel Boxes, 14.0" Spacing

0 NUCLEAR ENERGY SERVICES. INC. DOCUMENT NO.

8. REFERENCES

1. USNRC Standard Review Plan: " Spent Fuel Storage," Section 9.1.2 (February f 1975).

2. Spier, E.M., et al: " Transactions of the American Nuclear Society," p. 306, Westinghouse Spent Fuel Storage Rack Calculational Techniques, November 1975.

'3. Bromely, W.D., Olszewski, J.S.: " Safety Calculations and Benchmarking of Babcock & Wilcox Designed Close Spaced Fuel Storage Racks," Nuclear Technology, Vol. 41, Mid December 1978.

4. Bierman, S.R., Durst, B.M.: NUREG/ER-0073-RC, " Critical Separation Between Clusters of 4.29 w/o U 235 Enriched UO Rods in Water with Fixed Neutron 2

Poisons," May 1978.

5. Weader, R.J.: " Criticality Analysis of the Atcor Vandenburgh Cask," Nuclear Energy Services, Inc., NES 81A0260, May 1978.

6. Petrie, L.M., Cross, N.F.: " KENO IV - An Improved Monte Carlo Criticality Program," ORNL - 4938, November 1975.

7. Sutich, J.E., Honeck, H.C.: "The HAMMER System," DP-1064, January 1967.

8. Fowler, T.B., et al.: " EXTERMINATOR-2," ORNL-4078, April 1967.

9. VEPCO Specification POM-14, July 2,1976.

10. Dr. P. Buck: " Nuclear Design Analysis Report for the Beaver Valley Power Station Unit 1 High Density Fuel Storage Racks," NES 81A0441, March 1976.

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