ML18139A245

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Proposed Change 85 to Tech Specs,Increasing New Fuel Enrichment Limits
ML18139A245
Person / Time
Site: Surry  Dominion icon.png
Issue date: 05/15/1980
From:
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18139A244 List:
References
NUDOCS 8005200663
Download: ML18139A245 (61)
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Text

  • 1

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COMMONWEALTH OF VIRGINIA

)

' ) s. s.

CITY OF RICHMOND

)

Before me, a Notary Public, in and for the City and Common-wealth aforesaid, today personally appeared J. H. Ferguson, who being duly sworn, made oath and said (1) that he is Executive Vice President-Power of the Virginia Electric and Powe1: Company, (2) that he is duly authorized to execute and file the foregoing Amendment in behalf of that Company, and (3) that the statements in the Amendment are true to the best of his kn,owledge and belief.

Given under my hand and notarial seal this \\5~ day of (h_a......,~--' i18_o_..

My Conunis sicin expires --'h....,,.....

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N"otary Public (SEAL)

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e TS 5.3-2

3.

Reload fuel will be similar in design to the initial core.

The enrich-ment of reload fuel will not exceed 4 *. 1 weight percent of U-235.

4.

Burnable poison rods are incorporated in the initial core.

There are 816 poison rods in the form of 12 rod clusters, which are located in vacant control rod assembly guide thimbles.

The burnable poison rods consist of pyrex clad with stainless steel.

5.

There are 48 full-length control rod assemblies in the reactor core.

The full-length control rod assemblies contain a 144-inch length of silver-indium-cadmium alloy clad with stainless steel.*

6.

Surry Unit 1, Cycle 4, Surry Unit 2, Cycle 3, and subsequent cores will meet the following criteria at all times during the operation lifetime.

a.

Hot channel factor limits as specified in Section 3.12 shall be met.

l I

TS 5.4-1 5.4 FUEL STORAGE Applicability Applies to the design of the new and spent fuel storage areas.

Objective To define those,aspects of fuel storage relating to prevention of criticality in fuel storage areas;. to prevention of dilution of the borated water in the reactor; and to prevention of inadvertent draining of water. from the spent fuel storage area.

Specification A.

The reinforced concrete structure and steel superstructure of the Fuel Building_ and spent fuel storage racks are designed to withstand Design Basis Earthquake loadings as Class 1 structures.

The spent fuel pit has a stainless steel liner to ensure against loss of water*.

B.

The new and spent fuel storage racks are designed so that it is im-possible to insert assemblies in other than the prescribed locations.

New.fuel is stored vertically in an array with a distance of 21 inches between assemblies to assure keff ~ 0.98 with fuel of the highest anticipated enrichment in place assuming optimum moderation.* Spent fuel is stored vertically in an array with a distance of 14 inches between

  • E.G., an aqueous foam envelopment.as the result of fire fighting.

I

)

TS 5.4-2 assemblies to assure keff 2, 0.95, even if unborated water were used to fill the spent fuel storage pit.

The enrichment of the fuel stored in the Spent fuel racks shall not exceed 51.7 grams of Uranium

-235 per axial centimeter of fuel assembly.

C.

Whenever there is spent fuel in the spent fuel pit,. the pit shall be filled with borated water at a boron concentration not less than 2,000 ppm to match that used in the reactor cavity and refueling canal during refueling operations.

D.

The only drain which can be connected to the spent fuel storage area is that in the reactor cavity.

The strict step-by-step procedures used during refueling ensure that the gate valve on the fuel transfer tube which connects the spent fuel s~orage area with the reactor cavity is closed before draining of the. cavity commences.

In addition,, the procedures require placing the bolted blank flange on the fuel transfer tube as soon as the reactor cavity is drained.

References FSAR Section 9.5 Fuel Pit Cooling System FSAR Section 9.12 Fuel Handling System

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Project Application

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.TITLE/DEPT.

Dir. General Engineering Project Manager Sen. V.P. Eng. Oper.

Mgr._ Quality Assurance FORM # NES 204 2/80

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NUCLEAR DESIGN ANALYSIS REPORT FOR TiiE NEW FUEL STORAGE RACKS FOR TiiE SURRY NUCLEAR POWER STATION Prepared Under NES Project No. 5157 for The Virginia Electric Power Company by Nuclear Energy Services, Inc.

Prepared By 0.

REV. __

OF-~

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  • REVISION. LOG

_......,.. ::.,.,_NUCLEAR ENERGY ~ERVICES, INC *.

REV.

NO.

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FOAM ~ NES 206 2/80 PAGE

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D E S C R I PT I O N -

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TABLE OF CONTENTS

1.

SUMMARY

2.

INTRODUCTION

3.

DESCRIPTION OF NEW FUEL STORAGE RACKS

4.

CRITICALITY DESIGN CRITERION AND CALCULATIONAL ASSUMPTIONS

5.
6.

4.1 Criticality Design Criterion 4.2 Calculational Assumptions CRITICALITY CONFIGURATIONS 5.1 Normal Configurations 5.1.l Referenc~ Configuration 5.1.2

  • Eccentrically Positioned Fuel 5.1.3 Fuel Design Variation 5.1.l/.
  • Fuel Rack Cell Pitch Variation 5.l.5 Low Density Moderator Variation
  • 5.1.6 Worst Case Normal Configuration 5.2 Abnormal Conf~guration 5.2.1 Fuel Handling Incident 5.2.2 _.High Moderator Density Variation 5.2.3 Fuel Drop Incident

. 5.2.4 Seismic Incident 5.2.5 Worst Case Abnormal Configuratic;m CRITICALITY CALCULATION METHODS 6.1 Method of Analysis FORM; NES 205 2/80 PAGE _

.... 3 __

QF --2g*

6 8

9 10 10 10 11 11 11 11 12 12 12 12 12 12.

13 13 13 13 11.j.

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., 11 e!:1 N~CL~R ENERGY ~RVICES, 1~ ** C DOCUMENT NO.. _..;;;.8=1A...:;;.0;;..;6;...;.4..;;;.8 __ _

PAGE _....:4 __ QF*--=2=8-

7.
8.
9.

6.2 6.3 Computer Codes 6.2.1 HAMMER 6.2.2 KENO-IV TABLE OF CONTENTS (CONT'D)

Uncertainties and Benchmark Calculations RESULTS OF CRITICALITY CALCULATIONS 7.1 Reference Configurations

/

7.2 Keff Value for Normal Configurations 7.2.1 Moderator Density Variation from 0.0 to 1.0 gm/cc of H2o 7.2.2 Fuel Assembly Pitch Variation 7.2.3

  • Eccentric Fuel Location 7.2.4 Worst Case Normal Configuration 7.3 Keff for Abnormal Variations 7.3.1 Moderator Density Variation from 0.01 to 1.0 gm/cc of H20 7.3.2 Fuel Drop Accident 7.3~3 Seismic Incident 7.'3.4 Worst Case Abnormal Configuration 7.4 Effects of Calc1Jlational Uncertainties DETAILED PARAMETRIC STUDY VERSUS WATER DENSITY*

REFERENCES FORM # NES 205 2/80 14 14 14 15 19 19 19 19 19 19 20 20 20 20 21 21 21 25 28

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DOCUMENT NO. --=-8:.:.lA..:..:0:.;:;6...:..48~---

NUCLEAR ENERGY SERVICES, INC.

  • LIST OF TABLES 6,1 Fuel Parameters 7.°l Resu~ts of Keff Calculations LIST OF FIGURES 6.1 Quarter Storage Location Representation of Infinite Array 6.2 Illustration of Single Fuel Pin Model Showing Homogenized Fuel Region 7.1 Keff vs Water Density 7.2 Keff vs Storage Cell Pitch 8.1 Detailed _Geometric Representation of Finite Array 8.2 : Keff vs Water Density, Finite Array FORM # NES 205 2/80 PAGE _..:.5 __

QF __

2_8 __

16 22 17 18 23 24 26 27

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  • ~

ENERGY SERvtCES. lNC..

PAGE 6

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1~

SUMMARY

A detailed nuclear analysis has been performed for the new fuel storage racks for the Surry Nuclear Power Station.

The analysis demonstrates that for all normal and abnormal GOnfigurations considered, the Keff of the system is less than the criticality criterion of 0.98 for 4.1 w/o Westinghouse fuel asssemblies stored in the rack.

Studies were performed of the effects of variations in the physical parameters of the rack and of the fuel assemblies which could affect the nuclear characteristics. These

. variations are classified in thi~ report as norma.t and abnormal.

Normal variations include small changes iri water density, fuel eccentrically positioned within a storage cell, fuel enrichment variation, storage cell pitch variation, and the cumulative effect of all of the above, the worst case normal configuration. Abnormal variations include effects of fuel handling incidents, large water density variation:,

droppe~ or compacted fuel, and cell displacement due to seismic events.

The abnormal variation resulting in the highest increase in the magnitude of Keff is chosen to represent the worst case abnormal configuration.

A margin of error resulting from calculational uncertainty is added to the numerical results.

  • The calc;ulation of Keff values was carried out using the three-dimensional Monte Carlo code KENO-IV *

. Keff values were first calculated with a very simple geometric model_with reflecting boundaries in the x and y directions that effectively represented a rack of infinite lateral extent. The Keff values determined with this simple model may be summarized as follows:

0 Keff of the new fuel' storage* rack dry at 68 F at nominal dimensions Keff of the new fuel storage rack inclu~ing effects of normal variations and calculational uncertaiQty FORM # NES 205 2/80 0.474 0.713 I

111!...li". NUCLEAR ENERGY SERVICES,* INC.

  • _ Final Keff of the new fuel storage rack including normal variations, calculational uncertainty and the worst case abnormal configurati<;m.

DOCUMENT NO. __

8l_A_0_6_~8 ___ _

PAGE ___ 7 _____ QF ___ z_8 __ _

0.973 Because the resulting Keff' 0.973, is so close to the criticality criterion of 0.98, a further study was performed with a more detailed geometrical model with less inherent conservatism. The results of the more detailed study show the maximum Keff to be approximately 0.86.

_ These results show clearly th~t the Surry new fuel storage rac),0..... 6"'"48..._ __ _

PAGE _.L.,J 3..L.---OF 28 111!5 NUCLEAR ENERGY SERVICES, INC.

5.2.2 High Moderator Density Variation Accidents such as fire, pipe break, etc. can result in the presence of foams, steam, water.and other materials containing water in the new fuel storage area.

Under accident conditions it must be assumed the density of water can take any valu1=: from a.a* to 1.0 gm/cc. Therefore, the variation of Keff over the entire range must be evaluated. Since low water densities from 0.0 to 0.01 gm/cc are included under normal configurations, only densities from 0.0 l gm/cc to 1.0 gm/cc will be considered as abnormal configurations.

5.2.3 Fuel Drop Incident A fuel assembly could be dropped during insertion or removal from a storage cell-and compacted* within. A configuration is, therefore, considered in which one storage location contains compacted fuel. For simplicity, this was modeled as a worst case situation in which each location was filled with compacted fuel.

5.2.4 Seismic Incident The effects of a seismic incident are evaluated in terms of pitch variation caused by storage cell displacement.

5.2.5 Worst Case Abnormal Configuration The worst case abnormal configuration is taken to be the single abnormal configuration which results in the most adverse effect on Keff" FORM# NES 205 2/80

DOCUMENT NO. _

_,8.._.I..._A...,,0,..,.6""'4:w.8 __

NUCLEAR ENERGY SERVICES, INC.

PAGE~__,_J~4 ___ QF __ ~z~s.___

6. CRITICALITY* CALCULATION METHODS Calculations in this analysis were performed with KENO..:.Iv using 16 group Hansen Roach cross-sections. The HAMMER code was used as a check for accuracy. This section coi:itains information regarding computer models and codes.

6.1 METHOD OF ANALYSIS It was.stated in Section lJ. that the rack was modeled as an infinite array. This was accomplished by modeling one quarter of a storage cell containing one qu~rter of a

_ fuel assembly and the asso~iated water region surrounding i~ (see Figure 6.1).

Reflecting boundaries on all four sides make this model the equivalent of an infinite*

array in a hor.izontal plane.

In the vertical direction, nonreflecting boundaries are.

located below the floor, a concrete slab, and above the top of the storage rack.

  • The_ lJ..I° w/o 15xl5 Westinghouse fuel assemblies were modeled using the values shown in Table 6.1. Individual fuel pins were represented as concentric cylinders of U07 and zirconium clad (see Figure 6.7). The pellet diameter is assumed expanded to equal the dad inner diameter, thus eliminating the pellet-clad gap.

6.2 COMPUTER CODES 6.2.1 HAMMER

- HAMMER (see Ref. 2) is a multigroup integral transport theory_ code which is used to calculate la~tice cell.cross-sections for diffusion theory codes. This code*

has been extensively benchmarked against o2o and light water moderated lattices with good results.

6.2.2 KENO-IV KENO-IV is a 3-D rnultigroup Monte Carlo cod~ used to determine Keff (see Ref~

3).

KENO-IV has been benchmarked against" critical experiments consisting of typical light water reactor fuel lattices. Results (see "Ref. 5,6) show KENO-IV to

'be conservative for these configurations.

FORM # NES 205 2/80

e lle.5 DOCUMENT NO. --=-8 =1A=0=6....;..48-'----

  • NUCLEAR ENERGY SERVICES, INC.

PAGE J5 OF~-2-8~-

6.3 UNCERTAINTIES AND BENCHMARK CALCULATIONS The uncertainties in Monte Carlo criticality calculations can be divided into two classes:

1.

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

2.

Uncertainty due to bias in the calculational technique.

The first class of uncertainty can be reduced by simply increasing the number of

_ neutrons tracked. For rack c!"iticality calcuations, the number of _neutrons tracked is

  • selected to reduce this error to less than 1 %.

The second class of uncertainty is.accounted for by benchmarking the calculational method against experimental results. In the benchmarking process, the calculational method is used to determine the criticality value for a critical experiment configura-

_tion. The difference between the calculated criticality value and the experimental value is identified as the calculational bias. Once determined, this bias can be applied to-other calculational results obtained for similar configurations to improve the degree of calculational accuracy. If the calculated criticality value found during benchmark-ing is less than the experimental value, then the bias is added to other calculational resi.µts. to ensure a conservative criticality value consistent with experimental results.

Conversely, if the calculational criticality value is greater than the experimental value, it is appropriate to subtract the bias from the other calculated results to improve the accuracy of the criticality determination.

Both HAMMER and KENO-IV have been benchmarked at NES (Ref. 4) and found to be accurate in all cases to better than.:t.l % of the experimental Keff value. Benchmark calculations performed outside NES confirm these findings (see Ref. 5, 6).

Calculations in this analysis were based on KENO-IV.

To check the accuracy of KENO, fuel pin ka) values were determined using both KENO-IV and HAMMER and then compared to assure their agreement to within 1%. * *Thus HAMMER was used solely 'to check accuracy.

FORM # NES 205 2/80

DOCUMENT NO. ____;::8~1A~0:::..:6::...:4.:::.8 ___ _

11es NUCLEAR ENERGY SERVICES, INC.

  • PAGE __-:.1.:::...6 __ QF_....2..... 8'-*-

Fuel Type Fuel Enrichment.*

. U02 per Assembly Clad I.D.

Clad O.D.

Clad Material Pitch Between Rods Active Fuel Length Array Dimensions Guide Tube Material Fuel Rods per Assembly Guide Tubes per Assembly Guide Tubes, I.D.

Guide Tubes, 0.0.

FORM# NES 205 2/80 TABLE 6.1 FUEL PARAMETERS 15xl5 Westinghouse Fuel 4.1 w/o 1122 lb 0.3734 inch 0.422 inch Zircaloy-4 0.563 inch 144.0 inch 15xl5 Zircaloy-4 204 21 0.455 0.512

Moderator Region ---

Document No. *81A0648 Page 17 of 28 304 Stainless Steel I

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Guide Tube Quarter Storage Location Representation of Jnfinite Array FIGURE 6.1

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ILLUSTRATION OF SINGLE FUEL PIN MODEL SHOWING HOMOGENIZED FUEL REGION FIGURE 6.2

-FUEL MIXTURE {U02)

ZIRCALOY CLAD MODERATOR REGION

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.111!.!a *. NUCLEAR ENERGY SERVICES, INC.

DOCUMENT NO. _. _&l_A_0_6_48 ___ _

7. RESULTS OF CRITICAUTY CALCULATIONS Calculations performed with KENO-IV to evaluate Keff for the configuratio*ns described in Section 5 resulted in a final Keff value which is below the design limit of 0.98 impos_ed by the criticality criterion. The final value of Keff = 0.973 allows for variations due to normal and abnormal configurations and the effects of calculational uncertainty.

7.1 REFERENCE CONFIGURATION

. The Keff determined by KENS,-IV using the 16 group Hansen Roa~h cross-section set was 0.47l/. with an uncertainty of.:.0.006 at the 95% confidence level.

7.2 Keff VALUES FOR NORMAL CONFIG URA TIO NS 7.2.1 Moderator Density Variation from 0.0 to 0.1 gm/cc of H2o An increase of water density in the rack from 0.0 to 0.01 gm/cc resulted in a t.Keff of 0.233 (see Fig_ure 7.1 and Table 7.1).

7.2.2 Fuel Assembly Pitch Variation The pitch was varied up and down by 2"; decreasing pitch by 2" caused an increase in Keff of 0.043. The results of pitch variation are shown in Figure 7.2 and Table 7.1. Since the average pitch in the rack is substantially greater than the reference value of 21", no allowance for normal variation in pitch will be made.

7.2.3 Eccentric Fuel Location In the worst case of eccentric location of fuel assemblies,. four adjacent assemblies. 'Yill be located in the corners of th_eir respective cans such that a11

  • four are as close as possible to their three neighbors. In such a case; the pitch between these four neighbors will be reduced by 2 x 0.2775" where 0.2775" is the assembly to can wall clearance.

FORM # NES 205 2/80

NUCLEAR ENERGY SERVICES, INC.

DOCUMENT NO. _8_l_A_0_.64_8 ___ _

PAGE _.,..7.... Q _ __.OF 78 11es This case can conservatively be represented by a configuration in which the average pitch of the whole rack is reduced by 0.555 inches. The average pitch of the rack is much greater than the 21" assigned to the reference case because some gaps are 30".

Therefore the reduction of 0.555" for eccentric can be ignor~d.

7.2.4 Worst Case Normal Configuration The Keff for the worst case normal configuration results from the sum of the

.6K's due to normal variations added to the Keff for the reference co~iguration.

Keff for the worst case I]Ormal configuration is determined as_ follows:

K eff of reference configuration

.6Keff due to moderator density variation

.6Keff due to pitch variation

.6Keff due to eccentric fuel positioning To.tal.6Keff 0.474 0~233 0.00 0.00

  • = 0.233 Adding this value to. the reference Keff gives the value for the worst case normal configuration:

Keff

= 0.474 + 0.233

= 0.707 7.3 Keff FOR ABNORMAL VARIATION 7.3.1 Moderator Density Variation from 0.01 gm/cc to 1.0 gm/cc of H2o Variation of H70 density from 0.1 to 1.0 gm/cc resulted in a.6Keff of 0.260 (see Table 7.1 and Figure 7;1).

7.3.2 Fuel Drop Accident The accidental drop of a fuel assembly resulting in its being compacted in its storage location was modeled by increasing the pell~t O.D. of all fuel contained in the rack by 10%.

Densities were maintained at their reference values for conservatism..6K for this configuration was found to be 0.06.

FORM# NES 205 2/80

11es DOCUMENT NO. __

8_1A_0_6_4_8 ___ _

NUCLEAR ENERGY SERVICES; INC.

PAGE -"""2'""1 __ QF

  • 78.

7.3.3 Seismic Incident Rack pitch variations due to a seismic event are limited to approximately :t. 0.25 inches. These deflections would likely be in random directions. If, however, we a~sume they combine in the worst case to reduce the average storage cell pitch

~ 0.25 ~nches, it remains clear the effect on Keff is small.

Interpolation from Figure 7.2 shows the t.Keff for a pitch change of 0.25" to be

  • about 0.005 t.K.

7.3.4 Worst Case Abnormal Configuration The worst case abnormal configuration considers the ~Keff of the most adverse*

abnormal configuration in combination with the worst case normal Keff" The most adverse abnormal configuration (large moderator density variation) has a

~Keff of 0.260 which when added to the worst case normal Keff of 0.707 results in the worst case abnormal*Keff of 0.967.

7.4 EFFECTS OF CALCULATIONAL UNCERTAINTIES The statistical uncertainty due to KENO-IV is.:!: 0.006 at the 95% confidence level.

The bias for KENO-IV using 16 groups is negative; in other words, KENO caiculates a Keff higher than the actual Keff of*a critical experiment. This bias is neglected for conservatism.

  • The total effect of all uncertainties is taken as + 0.006. When added to the worst case

. abnormal Keff of 0.967 this re.sults in a final Keff including uncertainties of 0.973.

FORM # NES 205 2/80

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Average Moderator Fuel

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Modeled

. Storage Cell (Water)

Enrichment Keff 0

Configuration Pitch Density (inches)

(gm/cc)

(w/o) z C

Reference 21 10-8 4.1 0.474 0

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Configuration.

n m

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21 4.1 0.514 Variation 10-2 z

21 4.1 0.707 r>

21 10-1 4.1 0.967 21 1.0 4.1 -

0.873 19 10-8 4.1 0.517 oe 10-8

. 0.

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C 3:

m z TABLE 7.1

-0

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z G) 0 RES UL TS OF Keff CALCULATIONS m

N 00 N

0 0\\

~

0 00 "Tl r

0.90 0.80 0.70 0.60 0.50 0.40-;------,-------,-----t------+------t------1------+-----1 Water Density (gms/cc)

Keff vs Water Density FIGURE 7.1 10-0 i*

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Keff vs Storage Cell Pitch FIGURE 7.2 21 22 Document No. 81A064-8 Page 24-of 28 23

DOCUMENT NO. __

8_1A_0_6_4_8 ---

NUCLEAR* ENERGY SERVICES, INC.

PAGE _....;;2:;.;::5 ___ QF_....;;2:;.;::8,___

8. DETAILED PARAMETRIC STUDY VERSUS WATER DENSITY Because the peak Keff 0.973, found in Section 7.4, was so close to the allowed criticality criterion of 0.98, and also because it is possible that a somewhat higher value might exist in the neighborhood of the peak shown in Figure 7.1, a further parametric study was performed with a new, more detailed geometric model for KENO.

Thi.s model, instead of being infinite in lateral extent, represents the north-.south axis

_ of the rack, with the east-wesJ axis remaining infinite in extent (s~e Figure 8.1). This*

representation does ~wo things. First, the actual spacings (pitches) between rows are*

not all 21" but are either 21", 30", or 40", as can be' seen from the figure. Second, since the rack is now finite in the north-south axis, a substantial leakage will occur out the north and south faces of the rack, especially at low water densities. (This model was not used at the start of the work because of the increased complexity and cost.)

The results of a detailed par~etric study of Keff versus water density in the vicinity of 0.1 gms/cc are shown in Figure 8.2. It is seen that there is indeed a peak Keff somewhat higher than the value at.0.1 gm/cc located at about 0.06 gm/cc *. The value of Keff at this point using the more realistic geometric model of. Figure 8.1, is 0.896, whi~h is substantially*below the peak Keff of 0.967 reported for the simpler model (see Figure 7.1) and also substantially below the criticality criterion of 0.98 *

. The final Keff for the more detailed geometric model considering the KENO uncertainty of.:t:. 0.006 is 0.896+ 0.006 = 0.902 A further reduction in the calculated Keff would occur if the_ east-west axis of the pool were modeled instead of being taken as infinite.in extent. Such a calculation was not perform_ed because of* the great complexity and cost of such a large three-dimensional problem _but simple buckling estimates show a further reduction of Keff of about 0.04 would be realized. That is, the final K eff for "the Surr~ racks calculated with a geometry modeled in all three dimensions would be approximately 0.862.

FORM# NES 205 2/80

Document No. 81A0648 Page 26 of 28 DETAILED GEOMETRIC REPRESENTATION OF. FINITE ARRAY REFLECTING BOUNDARIES STORAGE LOCATIONS NORTH FIGURE 8.1 I

I 1DD1 1001 I

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    • Document No. &1A0648 Page 27 of 28 KEFF VERSUS WATER DENSITY, FINITE ARRAY 0.03 0.1.
  • WATER DENSITY, gms/cc FIGURE 8.2 0.3 1.0

---_J DOCUMENl' NO. ---"'8"""'1A:..:.0=6a...;4..;;a.8 __

NUCLEAR ENERGY SERVICES. INC.

PAGE__..2_8.__~0F~-28.._~

1.
2.
3.
9. -REFERENCES USNRC Letter to All Reactor Licensees, from Brian K. Grimes, April 14, 1978.

DP-1964, the HAMMER System, J.E. Sutch and H.C. Honeck, January 1967.

ORNL-4938, "KENO-IV -

An Improved Monte Carlo Criticality Program,"

L.M. Petrie, N.F. Cross, November 1975.

4.

NES 81A0260 "Criticality Analysis of the Atcor Vande~burgh Cask," R.J.

  • Weader, Febr*uary 1975.
5.

Bromley, W.D., Olszewski, L.S_. Safety Calculations and Benchmarking of Babcock & Wilcox Designed Close Spaced Fuel Storage Racks, Nuclear Techno-logy, Vol. 41, Mid-December 1978, p. 346.

6.

Bierman, S.R., Durst, B.M., Critical Separation Between Clusters of 4.29 wt%

235u Enriched uo2 Rods in Water with Fixed Neutron Poisons, May 1978, NUREG/GR-0073-RC~

FORM 1 NES 205 2/80 I

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NUCLEAR ENERGY SERVICES, INC.

DOCUMENT NO.

Project Application TITLE/DEPT.

Pro*ect M Vi ualit FOR!','! # NES 204 9/78 NUCLEAR DESIGN ANALYSIS REPORT.-

FOR THE SURRY NUCLEAR POWER STATION HIGH DENSITY FUEL STORAGE RACKS Prepared Under Project.5157 for the Virginia Electric Power Company by Nuclear Energy Servkes, Inc *

.Danbury, Connecticut 06810' Prepared By APPROVALS 800520~(,

81A0494..

REV.

0 PAGE OF* 29 DATE 3-/3-80 3-14--80

NUCLEAR ENERGY SERVICES, INC.

REV.

NO.

DATE FORM # NES 206 9/78 PAGE NO.

REVISION LOG DESCRIPTION e

DOCUMENT NO.

81A0494 PAGE 2

OF 29 ---

APPROVAL

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  • ~I-NUCLEAR ENERGY seRv1ces. 1Nc.

DOCU~ENT NO. 81A __ 0_4 __ 94'------

TABLE OF CONTENTS

1.

SUMMARY

2..

INTRODUCTION

3.

DESCRIPTION OF SPENT FUEL STORAGE RACKS*

4.

CRITICALITY DESIGN CRITERION AND CALCULA TIONAL ASSUMPTIONS

5.
6.

4.1 Criticality Design Criterion 4.2 Calculat_ional Assumpti9ns CRITICALITY CONFIGURATIONS 5.1 Normal Configurations 5.2 5.1.1 5.1.2 5.I.3 5.1.4 5.1.5

~eference Configurations Eccentric Configuration Fuel Design Variation Fuel Rack Cell Pitch Variation Fuel Rack Wall Thickness Variation 5.1.6 "Worst Case" Normal Configuration Abnormal Configuration 5.2.I Single Cell Displacement 5.2.2 Fuel Handling Incident 5.2.3 Pool Temperature Variation 5.2.l/.

Fuel Drop Incident 5.2.5 Seismic Incident 5.2.6 "Worst Case" Abnormal Configuration CRITICALITY CALCULATIONAL METHODS 6.1 Method of Analysis 6.2 Benchmark Calculations 6.3 Uncertainties 6.4, Computer Codes 6.4.1 NITAWL 6.4.2 KENO IV 6.4.3 HAMMER 6.4.4 EXTERMINATOR FORM

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7.

RESULTS OF CRITICALITY CALCULATIONS 22 7.1 Keff 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 Reference Configuration Keff 23 7.2 Keff Values for Abnor.mal Configurations 24 7.2.l Single Cell Displacement 24 7.2.2

_Pool Tempera.ture Variation 24 7.2.3 Seismic Event 24 7.2.4 "Worst Case" Abnormal Configuration 24 7.2.5 Effects of Calculational Uncertainties 25

8.

REFERENCES TABLES 5.1 Parameters of 17xl7 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 Ecce*ntric Configuration 16 5.2 Single Cell Displacement.

17 6.1 Quarter Assembly Repeating Array 21 7.1 Keff vs Fuel Rack Ceil Pitch 27 7.2 Keff vs Pool Water.Density 28 FOAM at NES 205 5/79

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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 Keff of the system for f./..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 arr~y of square stainless steel boxes spaced lf./..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 variatio~s.

The calculations were carried out using the Monte Carlo transport theory code KENO-IV to evaluate the reference configuration Keff" Other calculations to determine the sensitivity of Keff to the normal and abnormal variations mentioned above were performed using the diffusion theory code EXTERMINATOR-2. The final calculated Keff 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 ~he Surry Nuclear Power Station high density storage racks when loaded with the specified fuel are safe from a criticality standpoint.

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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*

cent~rs. 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 is given in Section 3.

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 Keff 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 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 detail in Section 5.

Reference configuration criticality calculations were performed with the transport theory Monte Carlo code combination NITAWL/KENO IV. Sensitivity calculations for normal and abnormal variations on the reference configuration were performed using

~he diffusion theory code combination HAM.MER/EXTERMINATOR.

Discussion of computer codes can be found in Section 6. The results of the criticality analyses_ are presented in Section 7.

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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 or greater.

Guard structures are provided at the upper grid of peripheral 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.

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

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  • JI-NUCLEAR ENERGY SERVICES, 1Nc.

IJ. CRITICALITY DESIGN CRITERION AND CALCULA TIONAL ASSUMPTIONS

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

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 information (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 Keff 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 fuef storage racks: "The multiplication constant (Keff) shall be less than 0.95 for all normal and abnormal configurations as confirmed by transport theory."

4.2 CALCULATIONAL ASSUMPTIONS 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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2.

The fuel 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 except the corner boxes which have a minimum wall thickness of 0.24-0 inches.

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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 NORMAL CONFIGURATIONS 5.1.1 Reference Configurations The reference configuration consists of an infinite array of storage cells having nominal dimensions each containing a l 7xl 7 Westinghouse fuel assembly of l.J.. l.

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 11.J..O inches on centers. The spent fuel pool water temperature is assumed to be 68°F.

This configuration is shown in Figure 5.1.a.

5.1.2 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/ l.J. inch on each side of the fuel assembly.

If orie assembly is displaced 1/4 inch from its nominal centered positioned and if all other assemblies r~main centered, the effect on Keff is negligibly small (less 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 Keff of this condition was determined *using the eccentric configuration shown in Figure 5.1.b.

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5.1.3 Fuel Design Variation The Surry Nuclear Power Station fuel racks are desi~ned to accommodate both 15xl5 and 17xl7 fuel designs. Calculations performed by NES show that racks with the 17xl7 fuel assemblies were slightly more reactive than the racks with the 15xl5 fuel assemblies with equal enrichment. Therefore, NES selected the 17xl7 fuel assembly with 4.1 w/o enrichment for the detailed criticality analysis of the Surry fuel storage racks.

5.1.l/.

Fuel Rack Cell Pitch Variation Calculations were perform~d to determine the sensitivity of Keff to 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-c~nter 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 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 ABNORMAL CONFIGURATIONS 5.2.1 Single Cell Displacement Welded clips and shims position the stainless steel cells or boxes centrally within the gr id members of the rack structure.

  • If the welds on one of these clip~ or shims fails, the associated box cannot be displaced. However, calculations were performed to determine the sensitivity of Keff 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 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.

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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 arid 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 Keff 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 Keff 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 further in this analysis.

5.2.3 Pool Temperature Variation Calculatiot1s were performed to determine the sensitivity of Keff 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

-i5o°F, the approximate boiling point of water near the bottom of the fuel rack.

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 pre~lude any significant lateral displacement of the storage cells *. Consequently, the change in Keff 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 o_f 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 k= of the fuel cell. It has been concluded, therefore,. that this incident would reduce Keff and need not be considered further in this analysis~

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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 Keff

  • since they do not change the center-to-center spacing between the*

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 Keff. is negligible since the average center-to-center spacing 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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1 TABLE 5.1 PARAMETERS OF 17xl7 AND 15xl5 WESTINGHOUSE FUEL ASSEMBLIES 17xl7 l5xl5 Mass of ~o2 in Assembly, lbs 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, lbs 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 O.f./.82 0.512 Guide, Tube ID, inches 0.450 0.455 FORM # NES 205 5/79

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6. CRITICALITY CALCULA TIONAL METHODS 6.1 METHOD OF ANALYSIS For the reference configuration discussed in Section 5.1.l, the Keff 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 a~ 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 spen-t fuel storage 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 Keff 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 BENCHMARK 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 was benchi:narked against several recent criticali_ty experiments.

Calculated Keff val_ues 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 resµlts (Ref. 5). -The calculated Keff differed from the experimental value by only 0.0017.

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6.3 UNCERTAINTIES The reference configuration Keff yalue calculated by KENO IV forms the basis for the nnal reported Keff 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 calcula tional 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 l %, and in this case +/-. 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 +/-_0.0 l.

6.4 COMPUTER CODES 6.4.l

  • NITAWL NIT AWL performs resonance self-shielding correction and c.reates a formatted working library base~ 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 Keff' 6.4.J HAMMER (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 o2o and light water moderated lattices with good results.

6.4.4 EX TERM INA TOR (Ref. 8)

EXTERMINATOR is a: 2-D multigroup diffusion theory code used with input from HAMMER to calculate Keff values.

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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 Keff" 7.1 Keff VALUES FOR NORMAL CONFIGURATIONS 7.1.1 Reference Configuration The Keff value for the reference configuration described in Section 5.1.1 was calculated to be 0.914 using NIT AWL/KENO IV.

7.1.2 Eccentric Configuration The Keff 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 refere!"lce configuration value by tiK = 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 17xl7 Westinghouse design to have a Keff value 0.007 higher that the 15xl5 Westinghouse design of the.same enrichment. Variation of Keff with fuel enrichment was calculated using both KENO IV and EXTERMIN-ATO~ *. Results show that the fuel rack Keff 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 Keff will be conservative with respect to enrichment and fuel design.

7.1.4 Fuel Rack Cell Pit~h Variation The Keff variation for fuel rack cell pitch values ranging down to 13-7/8 ihches 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 14.0.:t. 0.062 inches. The change in Keff for a 0.062 inch reduction in the average pitch is.6.K = 0.002.

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7.l.5 Fuel Rack Cell Wall Thickness Variation Fuel. rack cell wall thickness will be controlled to 0.090 +_~*~°o5o inches.

The change~ Keff due to variation in cell wall thickness was calculated to be ll.K =

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 Keff will always be conservative relative to this parameter.

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

Minimum Average Pitch Eccentric Positioning of Four Assemblies ll.Keff

+0.002

+0.006 Statistical combination of the above normal variations yields:

~(0.002)2 + (0.006)2

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

Worst case normal Keff FORM # NES 205 5/79

= 0.9 ll/. + 0.007

= 0.921

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7.2 Keff _VALUES FOR ABNORMAL CONFIGURATIONS.

7.2.1 Single Cell Displacement Calculations wer:e performed in which a single fuel cell was arbitrarily displaced 0.25 inches towards an adjacent cell. The resultant tiK was 0.001.

7.2.2 Pool Temperature Variation The Keff 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 Keff value, reached at apptoxim.ately 250°F, is 0.007 greater *than the Keff 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 Keff by 0.002. If the SSE is assumed to occur with the pool temperature at l 70°F (the maximum temperature during a full core off load) the increase in Keff due to the combined seismic and temperature effect is 0.006.

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

FORM # NES 205 5/79

1.

Worst Case Normal Configuration (per Section 7.1.6)

2.

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

3.

Resulting* Keff

~ff 0.921' 0.007 0.928

e

-~I.NUCLEAR ENERGY SERVICES, INC; e

DOCUMENT NO. _BI_A_0_4_94 ____ _

PAGE ---2.5 OF~.2__

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 Keff for the Surry high density spent fuel storage racks is *o.938. This conservative result meets the criticality design criterion set forth in Section 4 and clearly shows that the racks are safe from a criticality standpoint.

FORM # NES 205 5/79

"11 0

0

~

Q, z

m 1/1

~

TABLE 7.1 PARAMETERS AND RESULTS OF CRITICALITY CALCULATIONS

,0 Water Box Wall Enrichment Spacing Temp Density Thickness Configuration (w/o)

(inches)

J.oF)

(gm/cc)

(inches)

Reference Configurations 4.10 14.0 68 0.998 0.090 Maximum Water Density, 39°F 4.10 14.0 39 1.000 0.090 90°F 4.10 14.0 90 0.995 0.090 150°F 4.10 14.0 150 0.980 0_.090 212°F 4.10 14.0 212 0.958 0.090 250°F 4.10 14.0 250 0.941 0.090 250°F, voided 4.10 14.0 250 0.925 0.090 Close Spacing 4.10 13-15/ 16.

68 0.998 0.090 13-7/8 68 0.998 0.090 Eccentric Position 4.10 14.0 68 0.998 0.090 Displaced Can (base can) 4.10 14.0 68 0.998 0.090 Displaced Can (can moved U,11) 4.10 0.090 Kg/f

/!JKeff 0.914 3.50-4 1.25-3 3~60-3 5.92-3 6.16-3 3.86-3 1.75-3 3.68-3 5.69-3 1.42-3 1.73-3

>i.

~--

j

)>

G) m N °'

0 "Tl N

\\0 z

C 0

~

]J m

z m

u G)

-< ~-

0 m

z p*

.ge n

C:

3:

m z

-i.

z p 00 I-'

~

~

\\0

~

0.004 I I

t l I 0.003 J I

' t 0.002 *-,,,l.

0.001 0

Figure 7.1 13.90 PITCH, IN.

13.95

~cument No.: 81A0494 Page 27 of 29 14.00

..6.Keff vs Pitch.For 17 x 17 Westinghouse Fuei,.

l/-.1 w/o, 0.090" Stainless Steel Boxe~, 6&°F

.

  • 006

.005

-004,

.003

.002 Figure 7.2 Docum.ent No. : 81A0494 Page 28 of 29 1.0

.98

-.96

  • 3 Water Density, gm/cm

.94,.

40 T

OF Water emperatur.e,

.92 Keff vs Water Density fo_~ 17 x 17 4.1 w/o Westinghouse Fuel,_0.090" S~ainless Steel Boxes, 14.0" Spacing

/.

.~

~ *.~...

  • 1

~ NUCLEAR ENERGY SE~CES. 1Nc.

    • OCUMENT NO. _ 8_1_A_o4_9_4 ___

PAGE _2=9 ___ QF....,....=2_9_..,.

8.. REFERENCES

1.

USNRC Standard Review Plan: "Spent Fuel Storage," Section 9.1.2 (February 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 u235 Enriched uo2 Rods in Water with Fixed Neutron Poisons," May 1978.

5.

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

I

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-~4, 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.

FORM # NES 205 5/79

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