ML19352A151
| ML19352A151 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/26/1980 |
| From: | Buck P NUCLEAR ENERGY SERVICES, INC. |
| To: | |
| Shared Package | |
| ML19352A149 | List: |
| References | |
| 81A0648, 81A0648-01, 81A648, 81A648-1, NUDOCS 8103110538 | |
| Download: ML19352A151 (28) | |
Text
DOCUMENT NO.
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.ATTACIEE!rt 3 NUCLEAR DESIGN ANALYS5 REPORT FOR THE NEW FUEL STORAGE RACr4 FOR THE SURRY NUCLEAR POWER STATION Prepared Under NES Project No. 5157 for The Virginia E.ectric Power Company by Nuclear Energy Services, Inc.
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DOCUMENT NO.
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t TABLE OF CONTENTS I
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1.
SUMMARY
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8 2.
INTRODUCTION
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9 3.
DESCRIPTION OF NEW FUEL STORAGE RACKS j
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CRITICALITY DESIGN C.RITERION AND CA1.CULATIONAL 10 ASSUMPTIONS 10 l
4.1 Criticality Design Criterion 10 l
4.2 Ca!culational Assumptions 11 5.
CRITICA1.!TY CONFIGURATIONS, 11 5.1 ho mal Configurations 11 5.1.1 Reference Configuration 11
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5.1.2 Eccentrica!!y Positioned Fuel 12 5.1.3 Fuel Design variation 12 5.1.4 Fuel Rack Cell Pitch Variation 12 5.1.5 Low Density Moderator Variation 12 9
5.1.6 Worst Case Normal Configuration 12 5.2 Abnormal Configuration 12 5.2.1 Fuel Handling Incident 13 5.2.2 High Moderator Density Variation 13 5.2.3 Fuel Drop Incident 13 5.2.4 Seismic Incident 5.2.5 Worst Case Abnormal Configuration 13 14~
6.
CRIUCALITY CALCULATION METHODS 6.1 Method of Analysis 14
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DOCUMENT NO.
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TAElE OF CONTENTS (CONT'D) 6.2 Computer Codes 14 6.2.1 HAMMER 14 6.2.2 KENO-IV 14 6.3 Uncertaintns and Benchmark Ca!culations 15
, 7.
RESULTS OF CRil!CALITY CALCULATIONS 19 7.1 Reference Contigurations,
19 7.2 K
Value f r N rmal Configuraticns 19 d
7.2.1 Moderator Density Variation f.'om 0.0 to 1.0 gm/cc of H O 19 2
7.2.2 Fuel Assembly Pitch Variation 19 7.2.3*
Eccentric Fuel Location 19 7.2.4 Worst Case Norma! C: nfiguration 20 7.3 K,ff fer Abnorma! Variations 20 7.3.1 Moderator Density Variation from 0.01 to 1.0 gm/cc of H O 20 2
7.3.2 Fuel Drop Accident 20 7.3.3 Seismic Incident 21 7.3.4. Worst Case Abnormal Configuration 21 7.4 Effects of Calculational Uncertainties 21 8.
DETA! LED PARAMETRIC STUDY VERSUS WATFR DENSITY 25 9.
REFERENCES 28 e
9 EOCV 2 NE4 7M 7/AS
DdCUMENT NO.
31^04S OF G
E NUCLEAR ENERGY SERVICES, iNO.
" LIST OF TABLES 6.1 Fuel Parameters 16
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7.1 Results of K,ff Calculations 22 LIST OF FIGURES 6.1 Quarter Stcrnge Location Representation of Infinite Array 17 6.2 lilusratica of Single Fuel Pin Model Showing Homogenized Fuel Region 18 7.1 K
vs Water Density 23 gf 7.2 K
vs St rage Cell Pitch 24 df 8.1 Detailed Geometric Representation of Finite Array 26 I
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8.2 K,gf vs Water Density, Finitc Array 27 I
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A detailed nuclear analysis has beer. performed for ce new fuel s:crage racks fer the Surry Nuclear Power 5:a:icn. The analysis cemenstra:es that for a!! normal and abnormal configura:icns considered, the Kg, cf the system is less than the criticality criterion of 0.93 fcr 4.1 w/o Tes:inghouse fuel asssemblies stored in the rack.
Studies were perfermec cf de effects of varia:i:ns in the physical parame: rs cf the rack and of the fuel assemblies which could affect the nuclear characteristics. Tnese
. variaticas are classified in this. re-cr: as norma! anc abncrmal.
r Nermal variations include sma!I changes in water eenst:y, fuel eccentrica!Iy positiened within a s:: age ce!!, fuel enrichmen: varia-1:n, 5:: rage cell pitch variaden, and de cumula:!ve eff e:: of a!I cf the above, the wers: case normal cen#igu ati:n. Abnorrna!
variadens 1..cluce effec:s cf fuel handling in:Icents, large wa:er density variations, dropped er c = pac:ed fuel, and cell displacemen: cue to seismic events.
The abncrmal variatica resulting in de highes increase in the maEnitude of K,, is e-chosen to represent the worst case abncrma! cenfigura: ion.
A margin cf errer resulting from ca!culational uncer:ainty is acced to the nume-ical results.
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calcula:ica cf K values was carried out unng the three-dimensicnal Monte Carlo
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df code KENO-IV.
K values were first calculated with a very simple geometric mocel with reflecting df boundaries in de x and y directions that e'fectively represented a rcck of in'ini:e lateral extent. The Kg, values de:erminec wid -his simple model may be su nrnarized as ic!!ows:
K f :he new fuel sterage rack cry at 6S F a:
di nominal dimnsions 0.474 K
f the rew fuel s:crage ract. including etfects df of normal ' variations and calculational uncertainty 0.713 s
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DOCUMEtoT WO.
31 AC6' 5 i
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Final K,f f of ce new fuel storage rack including normal variations, calculational uncertainty and the wcrst case abncrmal configuration.
0.973 Because the resulting K,ff, 0.973, is so close to the cri:!cali:y criterien of 0.93, a further study was performed with a more ce: ailed gecme:.-ical mece!' wi:h less inheren: conservatism. The results of the more detailed s:ucy show the maximum K
t be approximately 0.36.
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. These results show clearly that the Surry new fuel s:crage racks meet the criticality design criterien and are safe under the specificati:ns se: for:h in the Stancarc Review-Plan (NUREG-75/037).
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- 2. INTRODUCTION The nuc' ear analysis performed for :ne Surry Nuclear Power Station is presented in this report in the following crder:
Detailed descriptions of the fuel rack and fuel assemblies to ce stored within are given in Section 3 including dimensions, tolerances and ma:e-ials ).-rtinen: to the nuclear characteristics of the loaded rack.
The criticality criterica and calculational assumptions mrde in crder to show compliance with NRC guidelines are outlin e in deta!!in Section 4.
Section 5 centains a description of the individual criticali:y cases studied. The presen:atica in Section 5 is intended to expand and clarify the scope of the nuclear analysis required for compliance with the NRC guidelines c;ucted in Secti n :..
The me:hed of analysis and the models used to describe the new fuel storage racks and the fuel assemblies in the varicus ccnfigurations are cu:11ned in
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Section 6. In addition, the computer codes used to carry out the calculations are discussed.
The results of the ca!culations are presented in Sec:ica 7 with their interpreta-tien.
The deterrnination of final K,ff values frem calculation results is explained and carried ou';.
A detailed parametric study versus water density, perfcrmed with a mere complex geometry, is presented in Section 3.
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f, DOCUMENT NO.
31A0643
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NUCLEAR ENERGY SERVICES. INC.
- 3. DESCRIPTION OF NEW FUEL STORAGE RACKS The new fuel stcrage facility at the Surry Nuclear Power Station has a total storage capacity of 126 new fuel assemblies. Each storage location consists of a stainless stee4 square box 165" ta!! with 9" 1.D. and 1/8" thick walls. These boxes are located in nine parallel rows, v/ith a pitch of 21" between boxes within a row. The pitc'h between rows is either 21" or 30".
The stcrage facility has concrete walls and ficer and is normally empty of water.
. The structural supports and bracing which hold the rack together and provide support during potential s.ismic events will not be considered in this analysis. This omission is' justified because these steel supports are at widely' separated locations and have a fairly large abscrption cross-section fer neutrons so that neglecting them is codserva-tive.
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- 4. CRITICALITY DESIGN CRITERION AND CALCULATIONAL ASSUMPTIONS 4.1 CRITICALITY DESIGN CRITERION The position of the NRC regarding the criticality of new fuel storage (Ref.1)is as l
follows:
I "The design of the new fuel storage racks will be such that K,ff wn! not exceed 0.98 with fuel at the hig. hest anticipated enrichment in place assuming optimum moderation."
This guide is adopted without modificatien as the criticality design criterien for the Surry new fuel storage racks.
4.2 CALCULATIONAL ASSUMPTIONS The follewing 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.
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- 1.. The rack is assumed to be infinite in lateral extent.
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The pitch is assumed to be 21" throughout, whereas in f act some rows are spaced at 30".
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- 5. CRIDCALITY CONFIGURATION To verify the adegacy of de Surry new fuel s:: rage racks for sterage cf 4.1 w/o fuel, It is necessary to cetermine multiplica:!ca ccastants correspencing :c the cifferen:
arrangements er configura:icas possible within the racks. These arrangements er configura:!ons are classified as either ncima! cr abncemal configurati:ns. Normal configura:!cns incluce the reference configura:icn, sma!1 water censity variations, eccentricc.11y posi:icned fuel, fuei cesign varia:!:n, fue! rack cell pi:ch varia:!:n and the combina:!on of these effects termec de wcrs case n:rmal c nfigura:!cn.
Abnormal configurations result from accidents and dsturbances not ncrmally e ccun: -
ered. These include fuel hand!!ng accicents, large water density varia:!:ns, fue! cr:p accident, seismic inciden: and the wcrst case abn:rmal configurations.
5.1 NOR.Y.AL CONFIGURATIONS 3.1.1 Reference Configura:icn The reference cerifiguratien c:nsists of an infini:e array cf storage cells having nominal cimensicas, each cen:aining a 15x15 Testing 5euse fuel assemb!y cf 4.1 w/o enrichment positioned centrally wi$1n the ce!!.
The s::: age cefis are spaced 21.0" on centers anc consist of sgare cans wit a 9.0" !.D. anc a 1/3" wall thickness.
The new fuel rack and the fuel assemblies are at 68 F.
The reference
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configuration is shewn in Figure 5.1.
3.1.2 Eccentrically Positioned Assemblies It is possible for a fuel assembly no: :: be positiened cen:ra!!y wl:hin a s: rage
. ce!! because of the clearance a!1 owed be: ween the assembly anc the cell wa!!.
This c!earance is nominally 0.2775" en each side of the fuel assembly. The wors:
eccentric positioning occurs if four acjacent assemblies are displaced within their s:crage cells as far as possible towarcs each c her.
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DOCUR1ENT f60.
81 ACfsh?.
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5.1.3 Fuel Design Variation Since 4.1 w/o is the highest enrichment expected to be used at Surry, no calculations have been performed to determine the effects of enrichment changes.
3.1.4 Fuel Rack Cell Pitch Variation Calculations were performed to determine the sensitivity of K,gg to change in pitch, the center-to-center spacing between storage ce!!s. The pitch vras varied 2" above and 2" below the nominal value of 21".
3.1.5 Low Density Moderator Variation The variation of atmospheric humidity in the rack causes a slight variation in moderator (H O) density. The sensitivity of K,ff to the variations in H O 2
2 density over the density range from 0.0 to 0.01 gm/cc was evaluated and is included under normal configurations. The upper limit of 0.01 gm/cc was chosen deliberately high to assure conservatism.
3.1.6 Worst Case Normal Configuration Since any of the above normal configurations can occur simultaneously, it is necessary to evaluate their combined maximum adverse effect.
' The result is the worst case normal configuration.
As the name implies, it l
l represents the state of the rack under normal conditions which has the largest K,gg value.
l 5.2 ABNORMAL CONFIGURATIONS 5.2.1 Fuel Handling Incident in some fuel storage racks it is possible during fuel handling to inadvertently position an assembly beside the loaded rack in a clearance space between racks on between storage locations within a rack. In the case of Surry reew fuel racks however, a steel cover or platform located above the storage cells prevents this incident from occurring. Consequently no calculations have been performed for fuel misplaced in the rack.
DOCUMENT NO.
8!A0663 L
NUCLEAR ENERGY SERVICES. INC.
5.2.2 Hirh Mode ator Density Variation Accidents such rs 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 value from 0.0 to 1.0 gm/ce. Therefore, the variation of K,ff over the entire range must be evaluated. Since low water densities from 0.0 to 0.01 gm/cc are included urider normal configurations, only densities from 0.01 gm/cc to 1.0 gm/cc will be considered as abnormal configurations.
5.2.3 Fuel Drop Incident
~
A fuel asser:.bly 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 inciden3 The effects of a seismic incident are evaluated in terms of pitch variation caused by storage cell displacement.
I 5.2.5 Worst Case Abnormal Configuration The worst case abrmrmal configuration is taken to be the single abnormal configuration which results in the most adverse eff ect on Kgf.
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DbCUMENT NO.
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NUCt. EAR ENERGY SERVICES, INC.
- 6. CRITICALITY CALCULATION METHODS Calculations in this analysis were performed with KENO-IV using 16 group Hansen This Roach cross-sections. The HAMMER code was used as a check for accuracy.
section contains iniormation regarding computer models and codes.
6.1 METHOD OF ANALYSIS it was stated in Section 4 that the rack was modeled as an infinite array. This was accomplished by modeling one quarter of a storage cell containing one quarter of a
, fuel assembly and the asso,ciated water region surrounding it (see Figure 6.1).
Reflecting boundaries on all four sides make this model the equivalent of an infi:.ite-array in a horizontal plane. In the vertical direction, nonreflecting boundaries are located below the floor, a concrete slab, ind above the top of the storage rack.
The 4.1 w/o 15x15 Westinghouse fuel assemb!!es were modeled u:ing the values shown in Table 6.1. Individual fuel pins were represented as concentric cylindert of UO2 #Ud zirconium clad (see Figure 6.2). The pellet diameter is assumed expancec to ee,ual the clad inner diameter, thus eliminating the pellet-clad gap.
6.2 COMPUTER CODES 6.2.1 HAMMER HAMMER (see Ref. 2) is a multir,roup integral transport theory code which is r
used to calculate lattice cell cross-sections for diffusion tneory codes. This code has been extensively benchmarked against D O and light water mocerated 2
lattices with good results.
6.2.2 KENO-IV KENO-IV is a 3-D multigroup Monte Carlo code used to determine Kd (see Ref.
3).
KENO-IV has been benchmarked against critical experiments consisting of typicallight water reactor fue! lattices. Results (see Ref. 5,6) show KENO-IV to be conservative for these configurations.
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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 criticality calcuations, the number of neutrons tracked is selected to reduce this error to less than IE i
The second c!:ss of uncertainty is, accounted for by benchmarking the calculational metod against experimental results. In the benchmarking process, the calculational metod is used to determine the criticality value for a critical experiment configura-
. tion. The difference between the calculated criticality value and the experimental i
i value is identified as the calculational bias. Once determined, this bias can be applied j
to other calculational results obtained for similar configurations to improve the degree l
of calculational accuracy. If the calculated criticality value found during benchmark-I i
Ing is less than the experimental value, then the bias is added to other calculational t
results to ensure a conservative criticality value consistent with experimental results.
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Conversely, if the calculational criticality value is greater than the experimental l
. value, it is appropriate to subtract the bias from the other calculated results to improve the accuracy of the criticality determination.
)i Both HAMMER and KENO-IV have been benchmarked at NES (Ref. 4) and found to be
- accurate in a'l cases to better than 11% of the experimental K,ff 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 w
KENO, fuel pin k, values were determined using both Mr.NO-IV and HAMMER and I
then compared to assure their agreement to within 1%.
Yhus HAMMER was used solely to check accuracy.
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DOCUMEMT NO.
81 A043
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NUCLEAR ENERGY SERVICES, INC.
TABLE 6.1 t
FUEL PARAMETERS I
l Fuel Tyoe 15x15 Westinghouse Fuel l
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Fuel Enrichment 4.1 w/o UO Per Assembly 1122lb,
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0.3734 inch 0.422 inch Clad O.D.
l Clad Material Zircaloy-4 l
Pi ch Between Rods 0.563 inch Active Fuel Length 144.0 inch Array Dimensions 15x15 Guide Tube Material Zircaloy-4 Fuel Rods per Assembly 204 Guide Tubes per Assembly 21 0.455 Guide Tubes, l.D.
Guide Tubes, O.D.
0.512 l
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Document No. S I A0643 Page 17 of 28 Moderator Region 304 Stainless
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,fage 18 of 28 ILLUSTRATION OF SINGLE FUEL PIN MODEL SHOWI!3G HOMOGENIZED FUEL REGION y UEL MIXTURE (UO2)
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DOCUMENT NO.
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- 7. RESULTS OF CRITICALITY CALCULATIONS Calculations performed with KENO-IV to evaluate K,ff for the configuratio'ns described in Section 3 resulted in a final K value which is below the design limit of eff 0.98 imposed by the criticality criterion. The final value of Keff = 0.973 allows for variations due to nvrmal and abnormal configurations and the effects of calculational uncertainty.
7.1 REFERENCE CONFIGURATION The K determined by KEN,0-IV using the 16 group Hansen Roach cross-section set gf 1 006 at te 9596 confidence level.
was 0.474 with an uncertainty of 0
7.2 K,ff VALUES FOR NORMAL CONFIGURATIONS 7.2.1 Moderator Density Variation from 0.0 to 0.1 gm/cc of H O 2
An increase of water density in the rack from 0.0 to 0.01 gm/cc resulted in a 4K f G.233 (see Figure 7.1 ano Table 7.1).
d 7.2.2 Fuel Assembly Pitch Variation The pitch was varied up and down by 2"; decreasing pitch by 2" caused an increase in K,ff 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 variatlan in pitch will be made.
7.2.3 Eccentric Fuel Location In the worst case of eccentric location of fuel assemblies, four adjacent assemblies will be located in the corners of their respective cans such that all 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.
DOCUMEMT NO.
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NUCLEAR ENERGY SERVICES, INC.
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 so the reference case because some gaps are 30".
Therefore the reduction of 0.555" for eccentric can be ignored.
7.2.4 Worst Case Normal Configuration The K for the worst case normal configuration results from the sum of the df AK's due to normal variations added to the K f r the reference configuration.
df K,ff for the worst case normal configuration is determined as follows:
K,ff of reference codguration 0.474 AK due t m derator density variation 0.233 eff AK cue to pitch variation 0.00 df A K,ff due to eccentric fuel positioning 0.00 Total AK
= 0.233 df Adding this value to the reference K,ff gives ce value for de worst case normal configuration:
it,ff
= 0.474 + 0.233
= 0.707 7.3 K
FOR ABNORMAL VARIATION di 7.3.1' Moderator Density "ariation from 0.01 gm/cc to 1.0 gm/cc of H O 2_
of 0.260 (see Variation of H O density from 0.1 to 1.0 gm/cc resulted in a. AKeff 2
Table 7.1 and Figure 7.1).
7.3.2 Fuel Droo Accident The accidental-drop of a fuel assembly resulting in its being compacted in its storage location was modeled by increasing the pellet O.D. of aR fuel contained in the rack by 10E Densities were maintained at their reference values for conservatism. 6K for this configuration was found to be 0.06.
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81A0643 "e*em
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7.3.3 Seismic Incident Rack pitch variations due to a seismic event are limited to approximately 3 23 0
inches. These deflections would likely be in random directions. II, however, we assume they combine in the worst case to reduce the average storage cell pitch 0.25 inches, it remains clear the effect on K,ff is small.
Interpolation from Figure 7.2 shows the AK,ff for a pitch change of 0.25" to be about 0.005 AK.
7.3.4 Worst Case Abnormal Configuration The worst case abnormal configuration considers the AK,ff of the most adverse-abnormal configuration in combination with the worst case normal K,ff. De most adverse abnormal configuration (large moderator density variation) has a A K,ff of f)160 which when added to the worst case normal K f 0.707 results di in the worst case abnormal K,ff 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 calculates a K,ff higher than the actual K,ff of'a critical experiment. This bias is neglected for cor.servatism.
The total effect of all uncertainties is taken as 1 006. When added to the worst case 0
abnormal K f 0.967 this results in a final K including uncertainties of 0.973.
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Average Moderator Fuel 3
Modeled Storage Cell (Water)
Enrichment K,gg t; \\
Configuration Pitch Decisity (inclies)
(gni/cc)
(w/o)
C Reference 21 10-8 4.1 0.474 O
Configuration 9
m 21 10-6 4,g 0.475 O
5 21 10 4.1 0.486
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4.1 0.473 s
Densit y Q
21 10-)
4.1 0.514
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21 10 4.1 0.707 9
21 10-I 4.1 0.967 21 1.0 4.1 0.873 19 10' 4.1 0.j l 7 g
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m TABLE 7.1 N
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0.90 0.80 0.70 Keff 0.60 0.5b y
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Water Density (gms/cc)
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K vs Water Density cgg FIGURE 7.1
Document No. 31A0608 Page 24 of 23 0.550 -
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'O.525 0
0.500 4
Keff 0.475 0.450 3
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Average Pitch (Inches) i
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K,ff vs Storage Cell Pitch I
' FIGURE 7.2 =
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q,ar DOCUMENT NO.
ME NUCLEAR ENERGY SERVICES. INC.
- 3. DETAII.ED PARAMETRIC STUDY VERSUS TATER DENSITY Because the peak K,ff 0.973, found in Secticn 7.4, was so close to the a!!cwed criticality r,riterion of 0.98, and also because it is possible that a somewhat Figher value might exist in the neighborhood of the peak shown in Figure 7.1, a further parametric study was performed with a new, m:re cetailed gecmetric model for KENO.
This model, instead of being infinite in later: exten:, represen:s the ncrth-south axis
, of the rack, with the east-wes,t axis remaining infinite in exten: (see Figure 3.1). This representatica does two things. First, the actual spacings (pitches) between rows are no: all 21" but are either 21", 30", or cG", as can ber seen from the figure. Second, since the rack is now finite in the north-scu h axis, a subs:an:ia! leakage will occur ou the north and south faces of the rack, especia!!y a: low water densities. (This mede!
was not used a: the star: cf the werk because of the increased complexity and cost.)
The resu!:s of a cetailed parameric stucy of K versus water density in the vicini:y df cf 0.1 gms/cc are shown in Figure S.2.
1: is seen :na: there is indeed a peak Kgf somewhat higher than the value at 0.1 gm/cc located at about 0.06 gm/cc. The value of K at this point using the more realistic gecmetric model of Figure 3.1, is 0.396, di which is substan:ially below the peak K,ff of 0.967 reper:ed for the simpler model (see Figure 7.1) and also substantia!!y below :he criticalig criterien of 0.98.
I The final K f r the more detailed geometric model considering the KENO df uncertainty cf 3 006 is 0
0.396 + 0.006 = 0.902 A further reductien in the calculated K would occur if the east-west axis <:f the gf pool were meceled incead of being taken as infinite in extent. Such a calculation was not performed because of -the great complexity and cost of such a large three-dimensional problem but simple buckling estimates show a further reduction of K of gg about 0.04 would be realized. That is, the final K for 'the Surry racks calculated g
with a geometry modeled in all three cimensions would be approximately 0.S62.
- CW.t :NES 205 2/80
Document No.1,l AC MC Page 26 of 2S DETAILED GEOMETRIC' REPRESENTATION OF FINITE ARRAY I
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Document No. SI AC643 Page 27 of 28 i
K VERSUS WATER DENSITY, FINITE ARRAY EFF I
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WATER DENSITY, gms/cc u.
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- 9. REFERENCES l
- 1
(
l.
USNRC Lener to All Recctor Licensees, from Brian K. Grimes, April 14,1978.
J 2.
DP-1064, the HAMMER System,3.E. Sutch and H.C. Honeck, January 1967.
i h
3.
ORNL-4938, " KENO-IV - An Improved Monte Carlo Criticality Prog am "
L.M. Petrie, N.F. Cross, November 1975.
4.
NES 81A0260 " Criticality Analysis of the Atcor Vandenburgh Cask," R.J.
i Weader, February 1975.
5.
- Bromley, W.D.,
Olszewr,ki, L.S. Safety Calculations and Benchmarking of Babcock & Wilcox Designed Close Soaced Fuel Storage Racks, Nuclear Techno-logy, Vol. 41, Mid-December 1978, p. 346.
6.
Bierman, S.R., Durst, B.M., Critica! Secaration Between Clusters of 4.29 wt%
235U Enriched UO Rods in Water with Fixed Neutron Poisons, May 197R.
2 NUREG/GR-0073-RC.
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