ML19309H929
| ML19309H929 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/03/1980 |
| From: | Buck P NUCLEAR ENERGY SERVICES, INC. |
| To: | |
| Shared Package | |
| ML18139A244 | List: |
| References | |
| 81A0648, 81A648, NUDOCS 8005200671 | |
| Download: ML19309H929 (28) | |
Text
80052 006 7 l O
oocumenTuo.
81^os*8 Rev. o I
h
, NUCLEAR ENERGY SERVICES. INC. ~
PAGE AF
'~
NUCLEAR DESIGN ANALYSIS REPORT FOR THE NEW FUEL STORAGE RACKS FOR mE SURRY NUCLEAR POWER STATION Prepared Under NES Project No. 5157 for The Virginia Electric Power Company by Nuclear Energy Services, Inc.
i Prom Application Prepared By Date APPROVALS I
TITLE / DEPT.
S I G N, A T U R E l
DATE Dir. General Engineering f
W27((d Project Manager O
9-3-FO Sen. V.P. Eng. Oper.
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FORM e NES 204 2'80
I ggg REVISION LDG MENT NO.
81A0648
. NUCLEAR ENERGY SERVICES. INC.
PAGE O F_ _ 28 "E.
"^ E OESCRIPTION APPROVAL DATE no go, O
e i
I FORM e Nts 20e 2:so
4 DOCUMENT NO.
81A0648 11 NUCLEAR ENERGY SERVICES. INC.
. -[
TABLE OF CONTENTS 1.
SUMMARY
6 2.
INTRODUCTION 8
3.
DESCRIPTION OF NEW FUEL STORAGE RACKS 9
. 4.
CRITICALITY DESIGN CRITERION AND CALCULATIONAL ASSUMPTIONS 10 4.1 Criticality Design Criterion 10 4.2 Calculational Assumptions 10 5.
CRITICALITY CONFIGURATIONS 11 5.1 Normal Configurations 11 5.1.1 Reference Configuration 11 5.1.2 Eccentrically Positioned Fuel 11 5.1.3 Fuel Design Variation 12 5.1.4 Fuel Rack Cell Pitch Variation 12 5.1.5 Low Density Moderator Variation 12 5.1.6 Worst Case Normal Configuration 12 5.2 Abnormal Configuration 12 5.2.1 Fuel Handling Incident 12 5.2.2 High Moderator Density Variation 13 5.2.3 Fuel Drop Incident 13 5.2.4 Seismic Incident 13 5.2.5 Worst Case Abnormal Configuration 13 6.
CRITICALITY CALCULATION METHODS 14 6.1 Method of Analysis 14 2
6 FORMNNES 205 2/80
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TABLE OF CONTENTS (CONT'D) 6.2 Computer Codes 14 6.2.1 HAMMER 14 6.2.2 KENO-IV 14 6.3 Uncertainties and Benchmark Calculations 15 7.
RESULTS OF CRITICALITY CALCULATIONS 19
~
7.1 Reference Configurations,
19 7.2 K
Value for Normal Configurations 19 gf 7.2.1 Moderator Density Variation from 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 Normal Configuration 20 7.3 K,ff for Abnormal 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.
DETAILED PARAMETRIC STUDY VERSUS WATER DENSITY 25 9.
REFERENCES 23 FORM # NES 205 2/80
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op 28 PAGE NUCLEAR ENERGY SERVICES, INC.
LIST OF TABLES 6.1 Fuel Parameters 16 7.1 Results of K,ff Calculations 22 LIST OF FIGURES 6.1 Quarter Storage Location Representation of Infinite Array 17 6.2 Illustration of Single Fuel Pin Model Showing Homogenized Fuel Region 18 7.1 K
vs Water Density 23 df 7.2 K,ff vs Storage Cell Pitch 24 8.1 Detailed Geometric Representation'of Firite Array 26 8.2 K,ff vs Water Density, Finite Array 27 l
l FORM
- NES 205 2/80
nnesmas.arrag2, SIA0648 PAGE 6
op 23 NUCLEAR ENERGY SERVICES. DdC.
- l.
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 configurations considered, the K,ff 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 this report as normal and abnormal.
Normal variations include small changes in 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 variations, dropped or compacted fuel, and cell displacement due to seismic events.
The abnormal variation resulting in the highest increase in the magnitude of K,ff is chosen to represent the worst case abnormal configuration.
A margin of error resulting from ' calculational uncertainty is added to ' the numerical results.
The calculation of K,ff values was carried out using the three-dimensional Monte Carlo -)
code KENO-IV.
K,ff 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 K,ff values determined with this simple model may be summarized as follows:
K f the new fuel storage rack dry at 68 F at df nominal dimensions 0.474 K
f the new fuel storage rack including effects df of normal variations and calculational t$ncertainty 0.713 FORM
- NES 205 2/80
DOCUMENT NO.
81A0648 PAGE OF NUCLEAR ENERGY SERVICES, INC.
Final K,ff of the new fuel storage rack including normal variations, calculational uncertainty and the worst case abnormal configuration.
0.973 Because the resulting K,ff, 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 K
to be approximately 0.86.
df These results show clearly that the Surry new fuel storage racks meet the criticality design criterion and are safe under the specifications set forth in the Standard Review' Plan (NUREG-73/087).
1 FORM # NES 205 2/60
- nnr*Hassatran, S I A0641_
PAGE 9
OF 79-MUCWAR ENERGY SEFVICES. WC.
1
- 2. INTRODUCTION
. -..., q The nuclear analysis performed for the Surry Nuclear Power Station is presented in this report in the following order:
Detailed descriptions of the fuel rack and fuel assemblies to be stored within are given in Section 3 including dimensions, tolerances and materials pertinent to the nuclear characteristics of the loaded rack.
The criticality criterion and calculational assumptions made in order to show I
compliance with NRC guidelines are outlined in detail iri Section 4.
Section 5 contains a description of the indvidual criticality cases studied. The presentation in Section 5 is intended to exoand and clarify the scope of the nuclear analysis required for compliance with the NRC guidelines quoted in Spction 4.
The method of analysis and the models used to describe the new fuel storage racks and the fuel assemblies in the various configurations are outlined in
~Section 6. In addition, the computer codes used to carry out the calculations are discussed.
'~
The results of the calculations are presented in Section 7 with their interpreta-tion.
The determination of final K,ff values from calculation results is explained and carried out.
A detailed parametric study versus water density, performed with a more complex geometry, is presented in Section 8.
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81A0648 mr PAGE 9 OF 7R NUCLEAR ENERGY SERVICES, INC.
- 3. DESCRIPTION OF NEW FUEL STORAGE RACKS The new fuel storage f acility at the Surry Nuclear Power Station has a total storage capacity of 126 new fuel assemblies. Each storage location consists of a stainless steel square box 165" tall with 9" I.D. and 1/8" thick walls. These boxes are located in nine parallel rows, with a pitch of 21" between boxes within a row. The pitch between rows is either 21" or 30".
The storage f acility has concrete walls and floor and is normally empty of water.
. The stru;tural supports and bracing which hold the rack together.and provide support during potential seismic 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 absorption cross-section for neutrons so that neglecting them is conserva-tive.
I l
l l
l FORM # NES 205 2/80
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NUCLEAR ENERGY SERVICES. INC.
- 4. CRIBCALITY 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 follows:
"The design of the new fuel storage racks will be such that K,ff will not exceed 0.98 with fuel at the highest anticipated enrichment in place assuming optimum moderation."
This guide is adopted without modification as the criticality design criterion for the Surry new fuel storage racks.
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 rack is assumed to be infinite in lateral extent.
2.
The pitch is assumed to be 21" throughout, whereas in fact some rows are spaced,
at 30".
FORM #NES 205 2/80
DOCUMENT NO.
8AA0648 mer 11 PAGE 1I OF 28 NUCLEAR ENERGY SERVICES. INC.
- 5. CRITICALITY CONFIGURATION To verify the adequacy of the Surry new fuel storage racks for storage of 4.1 w/o fuel, it is necessary to determine multiplication constants corresponding to the different arrangements or configurations possible within the racks. These arrangements or configurations are classified as either normal or abnormal configurations. Normal configurations include the reference configuration, small water density variations, eccentrically, positioned fuel, fuel design variation, fuel rack cell pitch variation and the combination of these effects termed the worst case normal configuration.
Abnormal configurations result from accidents and disturbances not normally encount '
ered. These include fuel handling accidents, large water density variations, fuel drop accident, seismic incident and the worst case abnormal configurations.
5.1 NORMAL CONFIGURATIONS 5.1.1 Reference Configuration The reference configuration consists of an infinite array of storage cells having nominal dimensions, each containing a 15x15 Westinghouse fuel assembly of 4.1 w/o enrichment positioned centrally within the cell.
The storage cells are spaced 21.0" on centers and consist of square cans with a 9.0" I.D. and a 1/8" wall thickness.
The new fuel rack and the fuel assemblies are at 68 F.
The reference configuration is shown in Figure 5.1.
5.1.2 Eccentrically Positioned Assemblies It is possible for a fuel assembly not to be positioned centrally within a storage cell because of the clearance allowed between the assembly and the cell wall.
This clearance is nominally 0.2775" on each side of the fuel assembly. The worst eccentric positioning occurs if four adjacent assemblies are displaced within their storage cells as far as possible towards each other.
FORM #NES 205 2/80
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5.1.3 Fuel Design Variation Since 4.1 w/o is the highest enrichment expected to N used at Surry, no calculations have been performed to determine the effects of enrichment changes.
5.1.4 Fuel Rack Cell Pitch Variation Calculations were performed to determine the sensitivity of K,ff to change in pitch, the center-to-center spacing between storage cells. Tne pitch was 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.
5.1.6 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 represents the state of the rack under normal conditions which has the largest K,ff value.
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 or between storage locations within a rack. In the case of Surry new 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.
FORM 4NES 205 2/80
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81A064R my PAGE I1 OF 28 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 value from 0.0 to 1.0 gm/cc. 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 under normal configurations, only densities from 0.01 gm/cc to 1.0 gm/cc will be censidered as abnormal configurations.
5.2..'
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 K,ff.
l l
FORM # NES 205 2/80
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DOCUMENT NO.
8 I A0f;48 PAGE 14 OF 7R NUCLEAR ENERGY SERVICES, INC.
1
- 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 contains information 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 mode' the equivalent of an infinite-array in a horizontal plane. In the vertical direction, nonreflecting boundaries are located below the floor, a concrete slab, and above the top of the storage rack.
)
The 4.1 w/o 15x15 Westinghouse fuel assemblies were modeled using r.e values shown in Table 6.1. Individual fuel pins were represented as concentric cylinders of UO2 ""d zirconium clad (see Figure 6.2). The pellet diameter is assumed expanded to equal the clad inner diameter, thus eliminating the pellet-clad gap.
l 6.2 COMPUTER CODES 6.2.1 HAMMER HAMMER (see Ref. 2) 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 O and light water moderated 2
lattices with good results.
6.2.2 KENO-IV KENO-IV is a 3-D multigroup Monte Carlo code used to determine K,ff (see Ref.
3).
KENO-IV has been benchmarked against critical experiments consisting of typicallight water reactor fuel lattices. Results (see Ref. 5,6) show KENO-IV to be conservative for these configurations.
FORM 4 NES 205 2/80
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81A0648 my PAGE I5 OF 78 NUCLEAR ENERGY SERVICES, INC.
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 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 results 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 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 KENO, fuel pin k, 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.
l FORM i:NES 205 2/80
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my DOCUMENT NO.
81A0648 PAGE 16 OF 28 NUCLEAR ENERGY SERVICES. INC.
l TABLE 6.1 FUEL PARAMETERS Fuel Type 15x15 Westinghouse Fuel Fuel Enrichment 4.1 w/o UO Per Assembly 1122lb 2
Clad I.D.
0.3734 inch Clad O.D.
0.422 inch Clad Material Zircaloy-4 Pitch 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 Guide Tubes, I.D.
0.455
)
Guide Tubes, O.D.
0.512
[
l
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FORM #NES 205 2/80
Document No. 81 A0648 Page 17 of 28 1
1 Moderator Region 304 Stainless Steel
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Guide Tube Fuel Cell Quarter Storage Location Representation of Infinite Array FIGURE 6.1 l
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-. Document No. 81 A0648
, Page 18 of 28, I
ILLUSTRA2 4 SINGLE FUEL PIN MODEL SHOWING HOMOGENIZED FUEL REGION FUEL MIXTURE (UO2) 7 ZIRCALOY CLAD
/
MODERATOR REGION p
FIGURE 6.2
DOCUMENT NO.
SIA0648 PAGE I9 OF 2R NUCLEAR ENERGY SERVICES. INC.
- 7. RESULTS OF CRITICALITY CALCULATIONS Calculations performed with KENO-IV to evaluate K fr the configurations df described in Section 5 resulted in a final K value which is below the design limit of df 0.98 imposed by the criticality criterion. The final value of Keff = 0.973 allow: for variations due to normal and abnormal configurations and the effects of calculational uncertainty.
7.1 REFERENCE CONFIGURATION The K determined by KENO-IV using the 16 group Hansen Roach cross-section set df was 0.474 with an uncertainty of 1 006 at the 95% confidence level.
0 t
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 AK f 0.233 (see Figure 7.1 and Table 7.1).
df 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 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 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.
FORM #NES 205 2/80
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81A0648 mr PAGE 70 OF 7R 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 to the reference case because some gaps are 30".
Therefore the reduction of 0.555" for eccentric can be ignored.
1 7.2.4 Worst Case Normal Configuration The K,ff for the worst case normal configuration results from the sum o2 the AK's due to normal variations added to the rg for the reference configuration.
K,ff for the worst case normal configuration is determined as follows:
K,ff of reference configuration 0.474 AK due to moderator density variation 0.233 df j
A K,ff due to pitch variation 0.00 AK due to eccentric fuel positioning 0.00 df Total AK
= 0.233 df Adding this value to the reference K,ff gives the value for the worst case normal configuration:
K,ff
= 0.474 + 0.233
= 0.707 7.3 K
FOR ABNORMAL VARIATION df 7.3.1 Moderator Density Variation from 0.01 gm/cc to 1.0 gm/cc of H _O 2
Variation of H O density from 0.1 to 1.0 gm/cc resulted in a AK of 0.260 (see 2
eff 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 j
storage location was modeled by increasing the pellet O.D. of all fuel contained l
in the rack by 10E Densities were maintained at their reference values for conservatism. AK for this configuration was found to be 0.06.
l l
)
FORM #NES 205 2/80
my DOCUMENT NO.
81A0648 PAGE 21 OF M NUCLEAR ENERGY SERVICES, INC.
7.3.3 Seismic Incident Rack pitch variations due to a seismic event are limited to appro...aately 2 25 0
inches. These deflections would likely be in random directions. If, 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 f r a pitch change of 0.25" to be df 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 0.260 which when added :o the worst case normal K,ff of 0.707 results in the worst case abnormal K,ff of 0.967.
7.4 EFFECTS OF CALCULATIONAL UNCERTAINTIES The statistical uncertainty due to KENO-IV is 1 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 conservatism.
The total effect of all uncertainties is taken as 2 006. When added to the worst case 0
abnormal K f 0.967 this results in a final K,ff including uncertainties of 0.973.
df FORM #NES 205 2/80
d
~
f..
g a
Ua e
Average Moderator Fuel 8
Modeled Storage Cell (Water)
Enrichment K,ff l
Configuration Pitch Density (inches)
(gm/cc)
(w/o) z C
Refercnce 21 10-8 4.1 0.474 h
Configuration 21 10-6 4,g 0.475 5
21 10
'41 0.486 Moderator g
21 10-N 41 0.473 21 Density g
21 10-3 41 0.514
{
m Variation
-2 b
21 10 4,1 0.707 21 10-I 4.1 0.967 21 1.0 4.1 0.873
-8 19 10 4.1 0.517 o
-8 o
23 10 4.1 0.446
@E Z
TABLE 7.1
-i
.o RESULTS OF K CALCULATIONS m
eff U
3; m
a
0.90 0.80 O.70 q g K cif 0.60 0.50 9
O 9
s mo Y
w 3 0.40 u o P.,
?+
Z 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-I 10-0 wm Water Density (gms/cc)
{
K vs Water Density eff FIGURE 7.1
Document No. 81A0648 Page 24 of 28 0.550 -
0.525 4
0.500 -
eff 0.475 0.450 I
I 19 20 21 22 23 Average Pitch (Inches)
K,gg vs Storage Cell Pitch FIGURE 7.2
DOCUMENT NO.
81A0648 PAGE 25, _3F 28 NUCLEAR ENERGY SERVICES. INC.
- 8. DETAILED PARAMETRIC STUDY VERSUS WATER DENSITY Because the peak K,ff 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.
This model, instead of being infinite in lateral extent, represents the north. south axis of the rack, with the east-wes,t axis remaining infinite in extent (see Figure 8.1). This representation does two 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 parametric study of K versus water density in the vicinity df of 0.1 gms/cc are shown in Figure 8.2.
It is seen that there is indeed a peak K,ff somewhat higher than the value at 0.1 gm/cc located at about 0.06 gm/cc. The value of K,ff at this point using the more realistic geometric model of Figure 8.1, is 0.896, which is substantially below the peak K,ff of 0.967 reported for the simpler model (see Figure 7.1) and also substantially below the criticality criterion of 0.08.
The final K f r the more detailed geometric model considering the KENO df 2 006 is uncertainty of 0
0.896 + 0.006 = 0.902 A further reduction in the calculated K w uld ccur if the east-west axis of the df pool were modeled instead 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,ff of about 0.04 would be realized. That is, the final K f r the Surry racks calculated df with a geometry modeled in all three dimensions would be approximately 0.862.
FORM 4 NES 205 2/80
Document No. 21A0648 Page 26 of 28 DETAILED GEOMETRIC REPRESENTATION OF FINITE ARRAY I
I REFLECTING BOUNDARIES D
a
.=.:.
, ll. lll STORAGE LOCATIONS
.==.
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1 M'
m'
= =.
n
=
. =e I
NORTH FIGURE 8.1 1
Document No. 81A0648 Page 27 of 28 l
K VERSUS WATER DENSITY, FINITE ARRAY EFF 1.0 --
.9 -
.8 -
.7 -
.6 -y l
i I
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i l I i l
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i i I l t 0.01 0.03 0.1 0,3 1,o WATER DENSITY, gms/cc I
FIGURE 8.2 9
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81A0648
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PAGE 78 OF 23 NUCLEAR ENERGY SERVICES. INO.
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- 9. REFERENCES 1.
USNRC Letter to All Reactor Licensees, from Brian K. Grimes, April 14,1978.
2.
DP-1064, the HAMMER System,3.E. Sutch and H.C. Honeck, January 1967.
3.
ORNL-4938, " KENO-IV - An Improved Monte Carlo Criticdity Program,"
L.M. Petrie, N.F. Cross, November 1975.
4.
NES 81A0260 " Criticality Analysis of the Atcor Vandenburgh Cask," R.J.
Weader, February 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 UO Rods in Water with Fixed Neutron Poisons, May 1978, 2
NUREG/GR-0073-RC.
n..
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l l
FORM NES205 2/80
8005200(o76 NUCLEAR ENERGY SERVICES, INC.
~
PAGE 1
np 29 l
l NUCLEAR DESIGN ANALYSIS REPORT.
FOR THE SURRY NUCLEAR POWER STATION HIGH DENSITY FUEL STORAGE RACKS Prepared Under Project 3157 for the Virgin'.a Electric Power Company by Nuc '. ear Energy Services, Inc.
.Danbury, Connecticut 06810' Project Acclication
' Prepared BY Da
~ APP ROV A LS TITLE / DEPT.
SIGN ATU R E DATE AJ-fd
/
Profect Manager Us2H( A/w S~l0~SO Vice President. Encineerino
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NO Ouality Assurance Manager I
8-M-8 0 A #h M R K l _ I l.p Y.
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- NES 204 9/78
DOCWENT NO.
MW NUCLEAR ENERGY SERVICES, INC.
"^o" REVISION LOG "E.
PA[E DESCRIPTION APPROVAL DATE n0 g
1 DORM
- NES 208 9/78
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 6.4.3 HAMMER 20 6.4.4 EXTERsilNATOR 20 PCMM 8 NES 20$ S/79
81A0494 NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
PAGE 4
OF 29 7.
RESULTS OF CRITICALITY CALCULATIONS 22 7.1 K,ff 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 K,ff 23 7.2 K,ff Values for Abnormal Configurations 24 7.2.1 Single Cell Displacement 24 7.2.2 Pool Temperature Variation 24 1
7.2.3 Seismic Event 24 j
7.2.4
" Worst Case" Abnormal Configuration 24 7.2.5 Effects of Calculational Uncertainties 25 8.
REFERENCES TABLES 3.1 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,ff vs Fuel Rack Cell Pitch 27 7.2 K,ff vs Pool Water Density 28 I
l OOmu a NES 20S S/79
I 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-IV to evaluate the reference configuration K,ff. Other calculations to determine the 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.
Pcmu a NES 206 S/79
NUCLEAR ENERGY SERVICES. INC.
DOCUMENT NO.
81A0494 PAGE 6
op 29
- 2. INTRODUCTION The NES design for high density fuel storage racks for Surry consists of a square array of stainless st' eel 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 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 K,ff of the syst<:m is substantially below 1.0.
Certain conservative assumptions about the kd 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 i
variations are described in detailin 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 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 presented in Section 7.
Pomu e Nes 20s sne
I NUCLEAR ENERGY SERVICES,INC.
DOCWENT NO.NM PAGE 7
OF 29 i
- 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.
Gua'rd 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 all of the principal structural grid members, clips and bracing.
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DOCUMENT NO.
PAGE 9
OF_ 39
- 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 a K,ff 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 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 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:
l 1.
The pool water has no soluble poison.
FCatu o NES 206 5/F9
/
I l
I NUCLEAR ENERGY SERVICES,INC.
DOCUMENT NO. 81A0494 PAGE 10 op 29 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 0.240 inches.
.e
I NUCLEAR ENERGY SERVICES,INC.
DOCUMENT NO.
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.
3.1 NORMAL CONFIGURATIONS 3.1.1 Reference Configurations The reference configuration consists of an infinite array of storage cells having nominal dimensions each containing a 17xi7 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 Keff s negligibly small (less i
)
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.
FORM e NES 206 5/79
I NUCLEAR ENERGY SERVICES,INC.
0W DOCUMENT NO.
PAGE I2 O F.,_ _79 5.1.3 Fuel Design Variation Tha 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 tne 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,ff 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-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 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 Displacement 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. Hewaver, 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 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.
FORM eNfS 208 5/79
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' her in this analysis.
t 5.2.3 Pool Temperature Variation Calculations were performed to determine the sensitivity of K,ff 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.
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 2?.5 inches. The dropped fuel assembly will most i
likely impact the flared tops c the fuel storage rack cells or boxes.
While minor deformation of the flared tops will occur, the close proximity of the i
upper grid structure to the. impact point will preclude any significant lateral displacement of the storage cells. Consequently, the change in K,ff will be negligible. However, it is possible for a dropped fuel assembly to enter a box l
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 i
the stored assembly would be compressed axially. A calcula ion 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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I NUCLEAR ENERGY SERVICES,INC.
81A0494 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 Kd{ 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 K,ff 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.
e Pemu a NES 206 S/79
e i
NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
PAGE 15 op '29 I
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NUCLEAR ENERGY SERVICES, INC.
PAGE 16 op.,_29 q-o a
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comu ANES 206 5/79
I NUCLEAR ENERGY SERVICES,INC.
^
DOCU?.1ENT NO.
PAGE 17 OF _.29 a
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POmu a NES 205 S/79
I NUCLEAR ENERGY SERVICES,INC.
81A0494 DOCUMENT NO.
PAGE I8 OF_ 2. 9 TABLE 5.1 PARAMETERS OF 17x17 AND 15x15 WESTINGHOUSE FUEL ASSEMBLIES 17x17 15x15 Mass of UO in Assembly,Ibs 1154 1122 2
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 Materal 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 l
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PORM
- PdES 206 S/F9
I NUCLEAR ENERGY SERVICES,INC.
DOCUMENT NO.
81A0494 PAGE I9 OF_ 29
- 6. CRITICALITY CALCULAT'ONAL METHODS 6.1 METHOD OF ANALYSIS For the reference configuration discussed in Section 5.1.1, the K., was determined g
from a three-dimensional Monte Carlo calculation using N;TAWL/ 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 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 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.
1 FCMM e NES 206 5/79
NUCLEAR ENERGY SERVICES. INC.
DOCUMENT NO.
81A0494 PAGE 20 op 79 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 er.lculational 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 2 01.
0 6.4 COMPUTER CODES 6.4.1 NITAWL NITAWL performs resonance self-shielding correction and creates a formatted working library based on the XSDRN cross-section set for use in KENO IV using the Nordheim Integral Method.
i 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 HAMMER (Ref. 7) l 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 O and light water moderated lattices 2
with good results.
6.4.4 EXTERMINATOR (Ref. 8)
EXTIRMINATOR is 12-D multigroup diffusion theory code used with input from HAMMER to calculate K,ff values.
DORM
- NES 206 5/F9
Om DOCUMENT NO.
NUCLEAR ENERGY SERVICES, INC.
PAGE 21 op 29
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- NE5 206 $/79
NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO. 81A0494 PAGE 22 29 o p,,
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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 K,ff.
7.1 K,ff VALUES FOR NORM AL CONFIGURATIONS 7.1.1 Reference Configuration The K,ff value for the reference configuration described in Section 5.1.1 was 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 clagonally 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 value 0.007 df 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 increates 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.t.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 + 0.062 inches. The change in K,ff for a 0.062 inch reduction in the average pitch is AK = 0.002.
PORM
- NES 208 $/79
NUCLEAR ENERGY SERVICES. INC.
DOCUMENT NO. _81 A0494 PAGE_ 21 OF 79 7.1.5 Fuel Rack Cell Wall Thickness Variation 0.090 +0'.00O Fuel rack cell wall thickness will be controlled to inches. The 0
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 will always be conservative relative to this parameter.
eff 7.1.6 Combined Effects of Normal Variations on the Reference Configuration K df 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=ff Minimum Average Pitch
+0.002 Eccentric Positioning of Four Assemblies
+ 0.006 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
'Je to the combined effect of all adverse normal variations.
Worst case normal K,ff
= 0.914 + 0.007
= 0.921 FCRM eNES 206 5/79
81A0494 NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
PAGE -._24 OF 29 7.2 K
VALUES FOR ABNORMAL CONFIGURATIONS eff i
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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 AK was 0.001.
7.2.2 Pool Temperature Variation
'l he 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 re!.erence cor. figuration.
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 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 ncrmal configuration.
kff 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 l
rew.nes nos sm
I NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO. 81A0494 PAGE 75 OFJ9 l
l 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 1
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 result 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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TABLE 7.1 5
PARAMETERS AND RESULTS OF CRITICALITY CALCULATIONS h
3)
E Water Box Wall.
K Enrichment Spacing Temp Density Thickness Ng Q
Configuration (w/o)
(inches)
(_ F) gm/cc)
(inches)
AKeff M
$s Reference Configurations 4.10 14.0 68 0.998 0.090 0.914 Maximum Water Density,39 F 4.10 14.0 39 1.000 0.090 3.50-4 b
90 F 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 8o Eccentric Position 4.10 14.0 68 0.998 0.090 5.69-3 y
Displaced Can (base can) 4.10 14.0 68 0.998 0.090 1.42-3 2
Displaced Can (can moved %")
4.10 0.090 1.73-3 M
9 W
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E.
m a
~
Documsnt No.: 81A0494 Pago 27 of 29 0.004-,2 0.003 -
0.002 - -
f I
eff i
0.001 -
0 n
j I
T 13 9 13.95 14.00 PITCH, IN.
Figure 7.1 r
A K,ff vs Pitch For 17 x 17 Westing, house Fuel, 4.1 w/o,0.090" Stainless Steel Boxes,68 F l
Docusant No.: 81A0494 Pzge 28 of 29
/
.006 -
.l
.005 -
1 SQgg 004 -
.003 -
s
/
.002 ~
.001 -
C' i
i I
i 1.0
.98
'.96
.94 '
.92 Water Density, gm/cm l
-l 40 Water Temperature, F 250-l Figure 7.2 K,ff vs Water Density fop 7 x 17 4.1 w/o Westinghouse Fuel,0.090" Stainless Steel Boxes, l
14.0" Spacing
81A0 m NUCLEAR ENERGY SERVICES. INC.
DOCUMENT NO.
PAGE 29 OF 29
- 8. REFERENCES 1.
USNRC Standard Review Plan: " Spent Fuel Storage," Section 9.1.2 (February 1975).
2.
Spier, E.M., et at: " Transactions of the American Nuclear Society," p. 306, Westinghouse Spent Fuel Storage Rack Calculational Techniques, November 1975.
3.
- Bromely, W.D., Olszewski, 3.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 235 Clusters of 4.29 w/o U Enriched UO Rods in Water with Fixed Neutron 2
Poisons," May 1978.
5.
Weader, R.J.:
" Criticality Analysis of the Atcor Vandenburgh Cask," Nuclear 1
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, 3.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 81 A0441, March 1976.
l l
FORM ebeES 206 5/79