ML20196J724
| ML20196J724 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/08/1988 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20196J720 | List: |
| References | |
| NUDOCS 8803150034 | |
| Download: ML20196J724 (89) | |
Text
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4.25 FUEL ASSEMBLYINITIAL ENRICHMENT,WT.% U 235 l
FIGURE 3.9-1 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2
(
SUMMER - UNIT 1 3/4 9-15 I
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4.25 FUEL ASSEMBLYINITIAL ENRICHMENT,wT.% U 235 l
l i
l l
FIGURE 3.9-2 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 3 SUMMER - UNIT 1 3/4 9-16 l
l AAb4I n s.+8 DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assembly normally containing 264 fuel rods clad with Zircaloy-4, except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel, or by vacancies, may be made if justified by a cycle specific reload analysis.
Each fuel rod shall have a nominal active fuel length of 144 inches.
The initial core loading shall have a maximum enrichment of 3.2 weight percent U-235.
Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of-4-3 weight percent U-235.
CONTROL ROD ASSEMBLIES l
l 5.3.2 The reactor core shall contain 48 full length control rod assemblies.
l The full length control rod assemblies shall contain a nominal 142 inches of l
absorber material.
The nominal values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium.
All control rods shall be i
clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM l
DESIGN PRESSURE AND TEMPERATURE l
5.4.1 The reactor coolant system is designed and shall be maintained:
I a.
In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and ~~
c.
For a temperature of 650*F, except for the pressurizer which is 680'F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9407 2 100 cubic feet at a nominal T,,, of 586.8'F.
5.5' METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
l
_ p _- UNIT 1
_ Amendment No. 27.55 g _
S R
5-6
AH. bmd I i'
p-y. 't. f 9 DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY
- 5. 6.1.1 The spent fuel storage racks consist of 1276 individual cells, each of which accommodates a single fuel assembly.
The cells are grouped into 3 blies with enrichments up to%gt d for storage of freshly discharged fuel assem-l regions.
Region 1 is designa weight percent U-235.
The cells in Region 2 are reserved for accommodating fuel assemblies wi itial enrichments of t S weight percent U-235 and a minimum burnup of 7
MWO/MTV.
Both regions l
"S 1 and 2 are poisoned.
Region 3 cells are capable of accommodating fuel assem-blies with initial enrichments of 4's-3 weight percent U-235 and a minimum burnup 395o 0.1g2,000 MWD /MTV.
The spent fuel h rage racks are designed and shall be maintained with:
A K,ff equivalent to less than or equal to 0.95 when flooded with a.
unborated water, which includes a conservative allowance for uncertainties as described in Section 4.3 of the FSAR.
b.
Nominal center-to-center distance between fuel assemblies of 10.4025" in Region 1, 10.4025" x 10.1875" in Region 2, and 10.116" in Region 3.
5.6.1.2 The new fuel storage racks are designed and shall be maintained with a nominal 21 inch center-to-center distance between new fuel assemblies such that K,ff will not exceed 0.98 when fuel having a maximum enrichment of g weight l
percent U-235 is in place and various densities of unborated water are assumed including aqueous foam moderation.
The K of $0.98 includes the conservative gf allowance for uncertainties described in Section 4.3 of the FSAR.
ORAINAGE 5.6.2 The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 460'3".
SUWER - UNIT 1 5-7 Amendment No.
Attachment II No Significant Hazards Consideration Description of Amendment Reaues_t The proposed amendment would modify Figures 3.9-1 and 3.9-2 contained in Section 3.9.12 "Spent Fuel Assembly Storage," Section 5.3.1, "Fuel Assemblies", and Section 5.6, "Fuel Storage," to reflect the new storage limitations for the Vantage-5 fuel to be utilized in the core during the 5th cycle at the Virgil C. Summer Nuclear Station. The proposed change is necessary to place the required restrictions on the storage of fuel to ensure inadvertent criticality does not occur.
Basis for No Sionificant Hazards Consideration Determination Section 3.9.12 of the Virgil C. Summer Nuclear Station Technical Specifications provides the requirements for initial enrichment and cumulative exposure of spent fuel stored in the different regions of the multi-region spent fuel pool. Region 1 racks are highly poisoned racks which can accept fuel initially enriched to the maximum licensed level with no restrictions on burnup. An analysis of the burnup history of each assembly is required prior to placement in either Region 2 or 3.
A record of this analysis is maintained for the time period that the spent fuel assembly remains in that region of the pool.
Figures 3.9-1 and 3.9-2 graphically depict the enrichment versus burnup requirements for Regions 2 and 3 respectively. The specification is applicable whenever fuel assemblies are stored in the spent fuel pool.
Section 5.3.1 of the Design Features portion of the Technical Specifications describes the physical attributes of the fuel assemblies utilized at the Virgil C. Summer Nuclear Station. Section 5.6 addressed fuel storage and specifically the criticality analyses for the storage racks utilized at the plant.
SCE&G has determined that the proposed changes to the above described Technical Specifications do not involve a significant hazards consideration for the following reasons:
1.
The probability or consequences of an accident previously evaluatad is not significantly increased.
The proposed change does not increase initial fuel enrichment or increase region average discharge burnups for the different regions of the fuel storage racks at the Virgil C. Summer Nuclear Station.
The amendment request is the result of evaluations performed to support the utilization of fuel for the 5th fuel cycle.
These evaluations determined the requirements necessary, and thus the proposed amendment request, to ensure the probability or consequences of previously analyzed accidents were not significantly increased.
l.-
2.
The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis reports is not created.
The proposed amendment does not involve any physical changes to the existing fuel racks currently installed at the plant. The amendment only changes the currently imposed administrative limits to reflect values consistent with the requirement of fuel to be utilized during Cycle 5.
3.
The margins of safety as defined in the bases of the Technical Specifications is not significantly reduced.
The proposed amendment is requested to ensure the design basis for preventing inadvertent criticality in the fuel storage areas is preserved. Therefore, the changes do not involve a significant reduction in the margin of safety.
i 2-
e A % J t TE c
CRITICALITY ANALYSIS OF V. C. SUMMER FUEL RACKS O
M
TABLE OF CONTENTS 1.0 Introduction 1
1.1 Design Description 2
1.2 Design Criteria 2
2.0 Criticality Analytical Method 3
3.0 Criticality Analysis of Region 2 Spent fuel rack 4
3.1 Reactivity Calculations 4
3.2 Postulated Accidents 6
(
3.3 Sensitivity Analysis 6
4.0 Criticality Analysis of Regions 1 and 3 Spent Fuel Racks 7
4.1 Reactivity Equivalencing 7
4.2 Analytical Methods 8
4.3 Reactivity Calculations - Region 1 9
4.4 Reactivity Calculations - Region 3 10
(
4.5 Postulated Accidents 12 4.6 Sensitivity Analysis 12 5.0 Criticality Analysis of Fresh Fuel Racks
?3 l
5.1 Full Density Moderation Analysis 14 l
5.2 Low Density Optimum Moderation Analysis 15 1
6.0 Acceptance Criterion For Criticality 17 Bibliography 44 1
Table of Contents i
g y
---.m
-- -, - -.y
.-n-
m LIST OF TABLES Table
- 1. Benchmark Critical Experiments (5,6]
18 Table
- 2. Fuel Parameters Employed in Criticality Analysis 19 Table
- 3. V. C. Summer Fuel Assembly Minimum Burnup vs initial U8 *
- Enrichment for Storage Spent Fuel Racks 20 j
Table
- 4. Comparison of PHOENIX isotopics Predictions to Yankee Core 5 l
Measurements 21 Table
- 5. Benchmark Critical Experiments PHOENIX Comparison 22 Table
- 6. Data for U Metal and UO Critical Experiments 23 l
l l
l l
l l
l l
List of Tables il
LIST OF ILLUSTRATIONS Figure
- 1. SCE&G Fuel Assembly Minimum Burnup vs. Initial U8 8
- Enrichment for Storage in Region 1 Spent Fuel Racks 25 Figure
- 2. SCE&G Fuel Assembly Minimum Burnup vs. Initial U8 8 '
Enrichment for Storage in Region 3 Spent Fuel Racks 26 Figure
- 3. SCE&G Region 1 Spent Fuel Storage Cell Nominal Dimensions 27 Figure
- 4. SCE&G Region 2 Speni Tual Storage Cell Nominal Dimensions 28 Figure
- 5. SCE&G Region 3 Spent Fuel Storage Cell Nominal Dimensions 29 Figure
- 6. SCE&G Fresh Fuel Storage Cell Nominal Dimensions 30 Figure
- 7. SCE&G Fresh Fuel Rack Layout 31 Figure
- 8. SCE&G Spent Fuel Rack Layout 32 Figure
- 9. SCE&G Region 2 Checkerboard Fuei Assembly Loading Schematic 33 Figure 10. Sensitivity of K.ve to Enrichment in the SCE&G Region 2 Spent Fuel Storage Rack with Two of Four Storage 34 Figure 11. Sensitivity of K.et to Center-to-Center Spacing in the SCE&G Region 2 Spent Fuel Storage Rack with Two of Four Storage 35 Figure 12. Sensitivity of K ve to B 8 ' Loading in the SCE&G Region 2 Spent Fuel Storage Rack with Two of Four Storage 36 Figure 13. Sensitivity of K.., to Enrichment in the SCE&G Region 1 Spent Fuel Storage Racks 37 Figure 14. Sensitivity of K.ve to Center-to-Center Spacing in the SCE&G Region 1 Spent Fuel Storage Racks 38 Figure 15. Sensitivity of K.et to 8 8 ' Loading in the SCE&G Region 1 Spent Fuel Storage Racks 39 Figure 16. Sensitivity of K.ve to Enrichment in the SCE&G Region 3 Spent Fuel Storage Racks 40 i
Figure 17. Sensitivity of K.v to Center-to-Center Spacing in the SCE&G Region 3 Spent Fuel Storage Racks 41 Figure 18. Sensitivity of K.et to Steel Can Thickness in the SCE&G Region 3 Spent Fuel Storage Racks 42 Figure 19. Sensitivity of K.ve to Water Density in the SCE&G Fresh Fuel 43 Storage Racks List of illustrations iil
1.0 INTRODUCTION
The V. C. Summer spent fuel rack (SFR) design described herein employs three arrays of racks, which will be considered as three separate spent fuel racks.
Each of these fuel racks or arrays consists of existing SCE&G fuel racks. The smaller array referred to as Region 2 will be reanalyzed for criticality to show that 5.0 w/o fuel can be stored in the rack in two out of four storage locations.
The larger arrays. Regions 1 and 3, will be reanalyzed to take into consideration the changes in fuel and fission product inventory resulting from depletion in the reactor core up to an enrichment of 5.0 w/o The Regions 1, 2 and 3 spent fuel rack designs are poisoned end non-poisoned stainless steel racks, previ-ously accepted by the NRC, for storage of Westinghouse 17x17 STD fuel where:
1.
Region 1 was designed for storage of freshly discharged fuel assemblies with enrichments up to 4.3 w/o U S 8 8, i
2.
The cells in Region 2 were for accamniodating fuel assemblies with initial enrichments of 4.3 w/o U8 8 8 and a minimum burnup of 20,000 MWD /MTU, l
3.
The Region 3 cell were capable of accommodating fuel assemblies with initial enrichment of 4.3 w/o Unas and a minimum burnup of 42,000 MWDIMTU.
The Region 1 and 3 spent fuel rack reanalysis is based on maintaining Ke v 5 e
0.95 for storage of Westinghouse 17x17 OFA and STD fuel at 5.0 w/o U 8 8 8 with an initial enrichmentiburnup combination in the acceptable area of Figures 1 and 2, respectively and with utilization of every cell permitted for storage j
of the fuel assemblies, l
l The V. C. Summer fresh fuel racks also consist of existing SCE&G fuel racks.
I These racks will be reanalyzed for criticality to show that 5.0 w/o OFA and STD fuel can be stored in every storage cell in the rack and maintain K.ve 5 0.95. The fresh fuel rack design is a non-poisoned stainless steel design, pre-viously accepted by the NRC for enrichments un to 4.3 w!o for Westinghouse 17x17 STO fuel.
t Introduction 1
t
1.1 DESIGN DESCRIPTION The Region 1, 2 and 3 spent fuel storage cell design are depicted schematically in Figures 3, 4 and 5, with nominal dimensions given on the figures. The fresh fuel rack storage cell design is depicted schematically in Figure 6.
The fresh and spent fuel rack layout are shown in Figures 7 and 8 respectively.
1.2 DESIGN CRITERIA Criticality of fuel assemblies in a fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between assemblies.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (K it) of the fuel assembly array will be less than 0.95 as recommended in ANSI 57.2-1983, ANSI 57.3-1983 and in Ref-erence 1.
l f
i l
Introduction 2
o 2.0 CRITICALITY ANALYTlCAL METHOD The criticality calculation method and cross-section values are verified by comparison with critical experiment data for assemblies similar to those for which the racks are designed. This benchmarking cata is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which ir.clude strong neutron absorbers, large water gaps and low moderator densities.
The design method which insures the criticality safety of fuel assemblies in the spent fuel storage rack uses the AMPXi 8, 83 system of codes for cross-section generation and KENO IV8
- 3 for reactivity determination.
The 227 energy group cross-section library that is the common starting point for all r:ross-sections used for the benchmarks and the storage rack is generated from ENDFIB-V(
- 3 dat a.
The NITAWL8'3 program includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry.
The Nordheim Integral Treatment is used.
Energy and spatial weighting of cross-sections is performed by the XSDRNPM( 8 2 program which is a one-dimensional Sa transport theory code. These multigroup cross-section sets are then used as input to KENO IV( * ' which is a three dimensional Monte Carlo theory program designed for reactivity calculations.
A set of 33 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and variability. The experiments range from water moderated, oxide fuel arrays separated by various materials (84C, steel, water, etc) that simulate LWR fuel shipphg and storage conditions ( '3 to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials ( '3 (Plexiglas and air) that demonstrate the wide range of applicability of the method. Table 1 summarizes these experiments.
The average K.vs of the benchmarks is 0.992. The standard deviation of the bias value is 0.0008 Ak. The 95/95 one sided tolerance limit f actor for 33 values is 2.19.
Thus, there is a 95 percent probability with a 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0018 Ak.
Criticality Analytical Method 3
4 a
3.0 CRITICALITY ANALYSIS OF REGION 2 SPENT FUEL RACK 3.1 REACTIVITY CALCULATIONS The following assumptions were used to develop the nominal case KENO model for the Region 2 spent fuel rack storage of fresh fuel using two out of four storage locations:
1.
The fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burnable polcon in the fuel rods. Historically, calculations for spent fuel racks similar to the Re-gion 2 racks analyzed herein have shown that the 'N 17x17 OFA fuel as-sembly yields a larger K.vt thr'1 does the W 17x17 Standard fuel assembly when both fuel assemblies have the same U
enrichment.
Thus, only the W 17x17 OFA fuel assembly was analyzed for Region 2.
(See Table 2 for fuel parameters).
2.
All fuel rods contain uranium dioxide at an enrichment of 5.0 w/o U ' ' '
over the infinite length of each rod.
3.
No credit is taken for any U' 8
- or U 8 8
- in the fuel, nor is any credit taken for the buildup of fission product pois,on material.
4.
The moderator is pure water at a temperatu% of 68'F. A conservative value of 1.0 gm/cm is used for the density
.'r water.
8 5.
No credit is taken for any spacer grids or spacer sleeves.
6.
Fuel assemblies are loaded into two of every four cells in a checkerboard pattern in the storage cells as shown in Figure 9 7.
The array is infinite in lateral and axial extent which precludes any neutron leakage from the array.
8.
The minimum poison material loading of 0.0015 grams B
per square centimeter is used throughout the array.
l Criticality Analysis of Region 2 Spent fuel rack 4
l l
(
e O
The KENO calculation for the nominal case resulted in a Kete of 0.8727 with a 95 percent probability /95 percent confidence level uncertainty of 0.0058.
The maximum K.it under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manuf acturing process in addition to asymmetric positioning of fuel assemblies within the storage cells. Studies of asymmetric positioning of fuel assemblies within the storage cells has shown that symmetrically placed fuel assemblies yield con-servative results in rack K.ve. The manuf acturing tolerances are stacked in such a manner to minimize the water gap between adjacent cells, thereby causing an increase in reactivity. The sheet metal tolerances are considered along with construction tolerances related to the cell I.D., and cell center-to-center spacing.
For the Region 2 storage racks, the water gaps are reduced from the nominal values of 1.261" and 1.046" to their minimum values. Thus, the most con-servative, or "worst case". KENO model of the Region 2 storage racks contains minimum water gaps of 1.198" and 0.983" with symmetrically placed fuel as-semblies.
Based on the analysis described above, the following equation is used to de-velop the maximum K et for the V. C. Summer Region 2 spent fuel storage racks with two out of four storage:
]
Kete = K
,o + Bmunee + Bo.,i + (( (ks)8..,n + (ks)8 m.in o where:
worst case KENO K.et that includes material
=
tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal method bias determined from benchmark critical
=
comparisons bias to account for posion partical self-shielding
=
95/95 uncertainty in the worst case KENO K.vt
=
95/95 uncertainty in the method bias
=
Substituting calculated values in the order listed above, the result is:
K.ve = 0.8727 + 0.0083 + 0.0100 + /((0.0065)8 + (0.0018): ) = 0.8977 1
Since K.et is less than 0.95 including uncertainties at a
95/95 probability / confidence level, the acceptance criteria for criticality is met.
Criticality Analysis of Region 2 Spent fuel rack 5
3.2 POSTULATED ACCIDENTS Most accident conditions will not result in an increase in K.tv of the rack. Ex-amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel height of stored assemblies which precludes interaction).
However, accidents can be postulated which would increase reactivity (li.e., or dropping a fuel assembly between the rack and pool wall). For these accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by about 30 percent AK. Thus, for postulated accidents, should there be a reactivity increase, K.es would be less than or equal to 0.95 due to the effect of the dissolved boron.
3.3 SENSITIVITY ANALYSIS To show the dependence of Keef on fuel and storage cells parameters as re-quested by the NRC, the variation of the K.it with respect to the following pa-rameters was developed using the KENO computer code:
1.
Fuel enrichment.
2.
Center-to-center spacing of storage cells.
3.
Poison plate Bio loading.
Results of the sensitivity analysis for the Region 2 storage cells are shown in Figures 10 through 12 for two of four storage.
6 Criticality Analysis of Region 2 Spent fuel rack
o 4.0 CRITICALITY ANALYSIS OF REGIONS 1 AND 3 SPENT FUEL RACKS This section develops and des:ribes the analytical techniques and models em-ployed to perform the criticality analyses for storage of spent fuel in Regions 1 and 3 of the V. C. Summer spent fuel pool.
4.1 REACTIVITY EQUIVALENCING Spent fuel storage, in the Regions 1 and 3 spent fuel storage racks, is achiev-able by means of the concept of reactivity equivalencing. The concept of re-activity equivalencing is predicated upen the reactivity decrease associated with fuel depletion. A series of reactivity calculations are performed to generate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield the equivalent Keve when the fuel is stored in the Regions 1 and 3 racks.
Figures 1 and 2 show the constant Keee contour generated for the V. C. Summer Region 1 and 3 racks respectively. Note in Figure 1 the endpoint at 0 MWD /MTU where the enrichment is 4.25 w/o and at 4.000 MWD /MTU wnere the enrichment is 5.0 w/o. The interpretation of the endpoint data is as fo:iows: the reactivity of the Region 1 racks containing fuel at 4,300 MWDIMTU burnup which had an initial enrichment of 5.0 w/o is ecuivalent te tie reactivity v! the Region 1 racks containing fresh fuel having an initial enrienment of 4.25 w/o. It is important to recognize that the curves in Figures 1 a.id 2 are based on a constant rack reactivity for that region and not on a conslarit fuel assembly reactivity. The data in Figures 1 and 2 is also provided as lible 3.
Linear interpolation be-tween two data points on this table will yield conservative results.
Criticality A>
or he, or.s I and 3 Spent Fuel Racks 7
s' l
4.2 ANALYTICAL METHODS I
The data points on the reactivity equivalence curve were generated with a transport theory computer code, PHOENIX ( ' 8 PHOENIX is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code. A 25 energy group nuclear data library based on a modified version of the British WIMSi * )
library is used with PHOENIX.
A study was done to examine fuel reactivity as a function of time following discharge from the reactor. Fission product decay was accounted for using CINDER ( ' 3 CINDER is a point-depletion computer code used to determine fission product activities. The fission products were permitted to decay for i
30 years af ter discharge. The fuel reactivity was found to reach a maximum l
at approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af ter discharge. At this point in time, the major fission product poison, Xe has nearly completely decayed away. Fur-185 l
thermore, the fuel reactivity was found to decrease continuously from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> l.
to 30 years following discharge. Therefore, the most reactive point in time for f
a fuel assembly af ter discharge from the reactor can be conservatively ap-proximand by removing the Xe8 8 '
i The PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been examined following discharge from a reac-tor. in addition, an extensive set of benchmark critical expe'iments has been l
l analyzed with PHOENIX. Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 4.
The measure-l ments were made on fuel discharged from Yankee Core 58 2 ' 8 The data in Table 4 shows that the agreement between PHOENIX predictions and measured Isotopic compositions is good.
The agreement between reactivities computed with PHOENIX and the results of I
i 81 critical benchmark experiments is summarized in Table 5.
Key parameters describing each of the 81 experiments are given in Table 6.
These reactivity
.I comparisons again show good agreement between experiment and PHOENIX
)
calculations.
An uncertainty associated with the burnup-dependent reactivities computed with PHOENIX is accounted for in the development of the Region 1 and 3 burnup requirements. A bias of 0.01 Ak at 30,000 MWD /MTU is considered to be very conservative since comparison between PHOENIX results and the Yankee Core r
experiments and 81 benchmark experiments indicates closer agreement.
i 6
Criticality Analysis of Regions 1 and 3 Spent Fuel Racks 8
i l
4.3 REACTIVITY CALCULATIONS - REGION 1 1
The nominal and maximum K.te for storage of spent fuel in Region 1 is deter-mined using the methods described in Section 2 for Region 2 in addition to the methods described in Section 4.2.
The actual conditions for tnis determination are defined by the zero burnup intercept point in Figure 1.
The KENO-IV com-puter code is used to calculate the storage rack multiplication f actor with an equivalent fresh fuel enrichment of 4.25 w/o. Combinations of fuel enrichment and discharge burnup yielding the same rack multiplication factor as at the zero burnup intercept are determined with PHOENIX.
The following assumptions were used to develop the nominal case KENO model for the Region 1 storage of spent fuel:
1.
Calculations for the Region 1 racks analyzed herein have shown that the Westinghouse 17x17 OFA fuel assembly yields a larger K et than does the Westinghouse 17x17 standard fuel assembly when both fuel assemblies have the same U8 *
- enrichment. Thus, only the Westinghouse 17x17 OFA fuel assembly was analyzed for Region 1.
2.
The Westinghouse 17x17 OFA spent fuel assembly contains uranium dioxide fuel at an equivalent "fresh fuel" enrichment of 4.25 w/o U8 8 8 3.
The moderator is pure water at a temperature of 6B'F.
A conservative value of 1.0 gm/cm 8 is used for the density of water.
4.
No credit is taken for any spacer grids or spacer sleeves.
5.
The array is infinite in lateral and axial extent which precludes any neutron leakage from the array.
6.
The minimum posion material loading of 0.022 grams 8 5 ' per square cen-timeter is used throughout the array.
The KENO calculation for the nominal case resulted in a K.ve of 0.9333 with a 95 percent probability /95 percent confidence level uncertainty of 0.0052.
The maximum K.et under normal conditions was determined with a "worst case" KENO model, in the same manner as for the Reglon 2 storage racks (see Section 3). The uncertainty associated with the reactivity equivalence methodology was included in the development of the burnup requirements. This uncertainty was discussed in Section 4.2.
Based on the analysis described above, the following equation is used to de-velop the maximum K.ve for the storage of spent fuel in the V. C. Summer Re-9 on 1 spent fuel storage racks:
1 Criticality Analysis of Regions 1 and 3 Spent Fuel Racks 9
]
K.it = K. + Bm.ia.a + Bpn + /[(ks) 8..e.i + (ks)8 m.in where:
worst case KENO K tv that includes centered
=
fuel assembly positions, material tolerances, and mechanical tolarance which can result in spacing between assemblies less than nonmlal method bias determined from benchmark
=
critical comparisons bias to account for posion partical
=
self-shielding ks ni 95/95 uncertainty in the worst case KENO
=
K.ve 95/95 uncertainty in the method blas
=
Substituting calculated values in the order listed above, the result is:
(0.0018)8 ) = 0.9496 K.ee = 0.9341 + 0.0083 + 0.0025 + /[(0.0043)8
+
The maximum K.es for Region 1 for this configuration is less than 0.95, including all uncertainties at a 95/95 probability / confidence level. Therefore, the accept-ance criteria for criticality are met for storage of spent fuel at an equivalent "fresh fuel" enrichment of 4.25 w/o U8 8 8 4.4 REACTIVITY CALCULATIONS - REGION 3 The nominal and maximum K.et for storage of spent fuel in Region 3 is deter-mined using the methods described in Section 2 for Region 2 in addition to the methods described in Section 4.2.
The actual conditions for this determination are defined by the zero burnup intercept point in Figure 2.
The KENO-IV com-puter code is used to calculated the storage rack multiplication f actor with an equivalent fresh fuel enrichment of 1.40 w/o. Combinations of fuel enrichment and discharge burnup yielding the same rack multiplication factor as at the zero burnup intercept are determined with PHOENIX.
The following assumptions were used to develop the nominal case KENO model for the Region 3 storage of spent fuel:
1.
Calculations for the Region 3 racks analyzed herein have shown that the Westinghouse 17x17 STD fuel assembly yields a larger K.ve than does the Westinghouse 17x17 OFA fuel assembly when both fuel assemblies have Criticality Analysis of Regions 1 and 3 Spent Fuel Racks 10 1
the same U8 8 8 enrichment. Thus, only the Westinghouse 17x17 STD fuel assembly was analyzed for Region 3.
2.
The Westinghouse 17x17 STD spent fuel assembly contains uranium dioxide fuel at an equivalent "fresh fuel" enrichment of 1.40 w/o U8 8 8 3.
The moderator is pure water at a temperature of 150*F. A conservative value of 0.98 gm/cm Is used for the density of water.
8 4.
No credit is taken for any spacer grids or spacer sleeves.
5.
The array is infinite in lateral and axial extent which precludes any neutron leakage from the array.
The KENO calculation for the nominal case resulted in a K.it of 0.9257 with a 95 percent probability /95 percent confidence level uncertainty of 10.0053.
The maximum Keet unrfer normal conditions was determined with a "worst case" KENO model, in the same n.anner as for the Region 2 storage racks (see Section 3). The uncertainty associated with the reactivity equivalence methodology was included in the development of the burnup requirements. This uncertsinty was discussed in Section 4.2.
Based on the analysis described above, the following eqJation is used to de-velop the maximum K.et for the storage of spent fuel in the V. C. Summer Re-gion 3 spent fuel storage racks:
K.it = K.=n + Sm.m
+ /[(ks) 8
., + (ks)8m.m.a l where:
~~"
worst case KENO K.et that includes centered fuel assembly
=
positions, material tolerances, and mechanical tolerance which car' result in spacing between assemblies less than nonmial method bias determined from benchmark critical comparisons
=
95/95 uncertainty in the worst case KENO K.e#
=
95/95 uncertainty in the method bias
=
Substituting calculated values in the order listed above, the result is:
(0.0018)8 ] = 0.9461 K.iv = 0.9333 + 0.0083 + /[(0.0041)8
+
The maximum K.ve for Region 3 for this configuration is less than 0.95, including all un:ertainties at a 95/95 probability / confidence level. Therefore, the accept-ance criteria for criticality are met for storage of spent fuel at an equivalent "fresh fuel" enrichment of 1.40 w/o U8 8 8 Criticality Analysis of Regions 1 and 3 Spent Fuel Racks 11
4.5 POSTULATED ACCIDENTS Most accident conditions will not result in an increass in K.it of the rack. Ex-amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticall:y is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel height of stored assemblies which precludes Interaction).
However, accidents can be postulated which would increase reactivity (i.e.,
misloading an assembly with a burnup and enrichment combination outside of the acceptable area in Figure 1, or dropping a fuel assembly between the rack and pool wall). For these accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This states that one is not required to assume two unlikely, independent, concurrent eva.its to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a Second unlikely event.
The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by about 30 percent.iK.
Thus, for postulated accidents, should there be a reactivity increase, K ve would be less than or equal to 0.95 due to the effect of the dissolved boron.
4.6 SENSITIVITY ANALYSIS To show the dependence of K.tv on fuel and storage cell parameters as re-quested by the NRC, sensitivity studies were performed in which the poison loading, the fuel enrichment, and the storage celf center-to-center spacing were varied, using tne KENO computer code.
Figures 13 through 18 illustrate the results of the sensitivity studies for spent fuel occupying every cell in the Region 1 and Region 3 fuel racks.
Criticality Analysis of Regions 1 and 3 Spent Fuel Racks 12
5.0 CRITICALITY ANALYSIS OF FRESH FUEi. RACKS This section describes the analytical techniques and models employed to per-form the criticality analysis for storage of fresh fuel in the V. C. Summer fresh fuel racks.
Since the fresh fuel racks are maintained in a dry condition, the criticality analysis will show that the rack Keet is less than 0.95 for the full density and low density optimum moderation conditions. The low density optimum moder-ation scenario is an accident situation in which no credit can be taken for soluble boron. The criticality method and cross-section library are the same as those discussed in Section 2 of this report.
The following assumptions were used to develop the nominal case KENO model for the storage of fresh fuel in the fresh fuel racks under full density and low density optimum moderation conditions:
1.
The fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burnable poison in the l
fuel rods.
2.
All fuel rods contain uranium dioxide at an enrichment of 5.0 w/o U8 *
- over the infinite length of each rod.
3.
No credit is taken for any U8 8
- or Un a e in the fuel, nor is any credit j
taken for the buildup of fission product poison material.
4.
No credit is taken for any spacer grids or spacer sleeves.
Calculations for these racks have shown that the W 17x17 OFA fuel assembly yields a larger K.,e than does the W 17x17 Standard fuel assembly when both fuel assemblies have the same U8 8 ' enrichment in full density water. Thus, i
only the W 17x17 OFA fuel assembly was analyzed (See Table 2 for fuel pa-remeters) in full density water.
l l
Criticality Analysis of Fresh Fuel Racks 13
5.1 FULL DENSITY MODERATION ANALYSIS In the nominal case KENO model for the full density moderation analysis, the moderator is pure water at a temperature of 68'F.
A conservative value of 1.0 gm/cm 8 is used for the density of water. The fuel array is infinite in lateral and axial extent which precludes any neutron leakage from the array.
The KENO calculation for the nominal case resulted in a K.ve of 0.9235 with a 95 percent probability /95 percent confidence level uncertainty of 10.0082.
The maximum K.ev under normal conditions arises from consideration of me-chanicci and material thickness tolerances resulting from the manufacturing process in' addition to asymmetric positioning of fuel assemblies within the storage cells. Studies of asymmetric positioning of fuel assemblies within trie storage cells has shown that symmetrically placed fuel assemblies yield con-servative results in rack K.vt. The manuf acturing tolerances are stacked in such a manner to mmimize the assembly center-to-center spacing and the total vol-ume of steel thereby causing an increase in reactivity. The sheet metal toler-ances are considered along with construction tolerances related to the cell I.D.
and cell center-to-center spacing. For the fresh fuel storage racks, the assembly center-to-center spacing is reduced from a nominal value of 21" to a minimum of 20.94".
Thus, the most conservative, or "worst case" KENO model of the fresh fuel storage racks contains a minimum water gap of 11.72" with sym-metrically placed fuel assemblies.
Based on the entlysis described above, the following equation is used to de-velop the maximum K.,, for the V. C. Summer fresh fuel storage racks:
K.., s i + S m.,a.. + (((k s) 8....i * (ks)8m.in.e K.et =
]
where:
= worst case KENO K.,# that includes material tolerances, and mechanical tolerances which can result in spacings between as,emblies less than nominal method bias determined from benchmark
=
critical comparisons Criticality Analysis of Fresh Fuel Racks 14 l
l
gi95 uncertainty in the worst case KENO
=
~'
95/95 uncertainty in the method bias
=
Substituting calculated values in the order listed above, the result is:
K.ve = 0.9235 + 0.0083 + /((0.0082)8 + (0.0018)8 ) = 0.9402 Since K.ve is less than 0.95 including uncertainties at a 95/95 probability confi-dence level, the acceptance criteria for criticality is rrit.
5.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS in the low density optimum moderation analysis, the fuel array is infinite in only the axial extent which precludes any neutron leakage from the top or bottom of the array.
Calculations have shown that the W.17x17 SYO fuel assembly yields a larger K.se than does the W 17x17 OFA fuel assembly wr.en both assemblies have the same U8 8 8 enrichment at low water densities. Thus, the W 17x17 STD as-sembly was used in the optimum moderation analysis.
Analysis of the V. C. Summer racks has shown that the maximum rack Keve under low density moderation conditions occurs at 0.04 gm/cm water density. The 8
KENO calculation of the V. C. Summer fresh racks at 0.04 gm/cm3 water density resulted in a peak K.e# of 0.8959 with a 95 percent probability and 95 percent confidence level uncertainty of 20.0079. Figure 19 shows the fresh fuel rack reactivity as a function of the water density.
The minimum cell center-to-center spacing, rack modu le spacing and material tolerances have been included in the base case model and result in a storage cell separation distance of 11.86" and a rack module separation distance of 20.94 inches. Studies of asymmetric positioning of fuel assemblies within the storage cells has shown thr.t symmetrically placed fuel assemblies yield con-servative results in rack K.,,
Based on the analysis described above, the following equauon is used to de-velop the maximum K.ev for the V. C. Summer fresh fuel storage racks under low density optimum moderation conditions:
K.ee = K....
+ B...
+ / [ (k s) 8... + (k s) 8
.m.. ]
where:
Criticality Analysis of Fresh Fuel Racks 15
base case KENO K.ft that includes nominal
=
mechanical and material dimension method bias determined from benchmark
=
critical comoarisons k s e...
95/95 uncertainty in the base case KENO K.<f 95/95 uncertainty in the method bias
=
Substituting calculated values in the order listed above, the result is:
K.es = 0.8959 + 0.0083 + (((0.0079)*
(0.0018): ) = 0.9123
+
Since K ef is less than 0.95 including uncertainties at a
95/95 probability / confidence level, the acceptance criteria for criticality is met.
l t
1 l
l l
l Criticality Analysis of Fresh Fuel Racks 16 l
i l
6.0 ACCEPTANCE CRITERION FOR CRITICALITY The neutron multiplication factor in sperst fuel pool and fresh fuel vault shall be less than or equal to 0.95, including all uncertainties, under all conditions.
The analytical methods employed herein conform with ANSI N18.2-1973, "Nu-clear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Hendling System; ANSI 57.2-1983, "Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2; ANSI N16.9-1975, "Validation of Calculational Methods for Nuclear Criticality Safety," NRC Standard Review Plan, Section 9.1.2, "Spent Fuel Storage"; and the NRC guidance, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," ANSI 57.3-1983, "Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants."
l Acceptance Criterion For Criticality 17
Table 1.
Benchmark Critical Experiments (5,6)
Generai E ne Icfmt
$eceratIng Solubte Descrtotton s/o U235 lie f l ec t or Waterial 8-10 pre hgg 1
UO rod lattice 2.46 matar water 0
0.9857 *.00iB 2.
U0 rod lettice 2.46 mater mater 1037 0.9906 7.0018 3.
UO Pod lattice 2.46 water water 764 0.9h 26 T.0015 4
VO rod lattice 2.46
=ater 84C pins 0
0.99ta T.0025 5.
VO rod lattice 2.46
= a t er Bec pins 0
0.9998 T. 0026 l
6.
U0 rod lattice 2.46 water 84C pins 0
0 9955 T.0020 i
7 UO rod lattice 2.46 mater 84C pins 0
0.9989 I.0026 l
8.
VO rod lattice 2.46 safer 84C pins 0
0 9983 7.0025 l
9.
UO rod lattice 2.46 water water 0
0.9931 T.0028 10.
VO rod lattice 2.46 water water 143 0.9928 T.0025 i
l 11 UO rod lattice 2.46 mater stainless steet 514 0.0967 7.0020 12.
UO rod lattice 2.46 safer statntess steel 217 0.9943 7.9019 l
13.
UO rod lattice 2.46 mater borated aluminum 15 0.9892 7.0023 14 UO rod lattice 2.46 satee borated alumtman 92 0 9884 T 0023 l
15.
VO rod lattice 2.46 mater borated aluminum 3ss 0.9fl32 7.0021 l
16.
UO rod lattice 2.46 mater borated stuminum t21 0.9948 T.0024 17.
UO rod lattice 2.46 water borated aluntnum 487 0.9895 7.0020 l
18.
U0 rod lattice 2.46 sater borated aluminum 197 0 9985 7.0022 i
19.
00 rod lattice 2.46 ester borated alu=Inum 634 0.9971 T.0019 20.
UO rod lattice 2.46 mater borated aluminum 320 0.9920 7.0020 fl.
UO rod IsttIce 2.46 mater borated aluutnum 72 0 9939 7.0020 l
l 22.
U tal cylledera 93.2 bare air C
0.9905 7.0020 23.
U metal cylinders 93.2 bare air 0
0.9976 7.0020 24.
U metal cyltnders 93.2 bare ale 0
0.994 7 T.0025 l
25.
U met al cylinders 93.2 bare air 0
0.9928 7,001g 26.
U set al cylineters 93.2 bare air 0
0.9972 7.0026 l
27.
U metal cylinders 93.2 bare air 0
0.9950 T.0027 l
28.
U metal cylinders 93.2 bare plentglass 0
0.9941 7,ooyo 29.
U metal cylinders 93.2 parafftn pleutgfass 0
0.9928 T.0049 30.
U metal cylinders 93.2 bare plealglass 0
0.9968 T.0018 1
31.
U metal cyttnriers 93.2 paraffin plestglass 0
1.0L42 I.0019 l
32.
U petal cylinrters
'l3. 2 paraffin pfestglass 0
0.
33 I.0030 33.
U metal cylinders 93.2 parafftn pleatgtsss 0
0.9919 T,.0032 l
l l
l i
l
-18
Table 2.
Fuel Parameters Employed in Criticality Analysis Parameter W 17x17 OFA W 17x17 STANDARD Number of Fuel Rods per Assembly 264 264 Rod Zirc-4 Clad 0.D.
(i nen) 0 360 0 374 Clad Thickness (Inch) 0.0225 0.0225 Fuel Pellet 0.D. (Inch) 0 3088 0 3225 Fuel Pellet Density
(% of Theoretical) 96 96 i
Fuel Pellet Dishing Factor 0.0 0.0 Rod Pitch (Inch) 0.496 0.496 Number of Zire-4 Guide Tubes 24 24 Guide Tube 0.D.
(i nch) 0.474 0.4841 Guide Tube Thickness (inch) 0.016 0.0188 Number of instrument Tubes I
1 l
Instrument Tube 0.D.
(inch) 0.474 0.4848 l
Instrumen'. Tube Thickness (inch) 0.016 0.0188-t I
Revised cesign data shows the Tube o.D. and thschness to be 0.442 sad 0.010 enches respectively. These changes will have no segnef ecant effect on the results and conclusions of this emelysis.
l t
l 19 -
t
o Table 3.
V. C.
Summer Fuel Assembly Minimum Burnup vs Initial U 2 8 8 Enrichment for Storage Spent Fuel Racks initial U 8 8 8 Assembly Discharge Enrichment Burnup (GWD/MTU)
Region 1 4.25 0
5 00 4.0 Region 3 l
1.4 0
I 17 6.8 2.0 10 9 l
25 18.4 30 24 5 35 31.1 4.0 36.7 45 42.8 50 48.0 J
l l
l l
1 20 l
l
l l
Table 4.
Comparison of PHOENIX isotopics Predictions to Yankee Core 5 l
Measurements Quantity (Atom
% Difference Ratio)
U235/U
-0.67 U236/U
-0.28 U238/U
-0.03 PU239/U
+3.27 PU240/U
+3.63 l
PU241/U
-7.01 PU242/U
-0.20 l
l PU239/U238
+3.24 Mass (PU/U)
+ 1.41 FISS-PU/ TOT-PU
-0.02 21
-e--
,w
,.,,----n--+,r~-
,weng-, -,
..m, -
--+-~v---
.--.e,
-,e,m.--,~.--e e-
-x-r
- m#
4
Table 5.
Benchmark Critical Experiments PHOENIX Comparison Description of Number of PHOENIX K.ve Using Experiment Experiments Experiments Bucklings UOs Al clad 14 0.9947 SS clad 19 0.9944 Borated H2O 7
0.9940 Subtotal 40 0.9944 U-Metal Al clad 41 1.0012 TOTAL 81 0.9978 l
1 l
22
Table 6.
Data for U Metal and UO: Critical Experiments (Part 1 of 2)
Fuel Pellet Clad Clad Lattice Case Cell A/O H20/U Density Olaneter Material 00 7hickness Pitch 8-10 Number Type U-235 Ratto (G/CC)
(CM)
Clad (CM)
(CM)
(CM)
PPM i
Hexa 1.328 3.02 7.53 1.5265 Aluminum 1.6916
.07110 2.2050 0.0 l
2 Hexa 1.328 3.95 7.53 1.5265 Aluminum 1.6916.07110 2.3590 0.0 l
3 Hexa 1.328 4.95 7.53 1.5265 Aluminum f.6916
.07110 2.5120 0.0 l
4 Hexa 1.328 3.92 7.52
.9555 Aluminum 1.1506
.07110 1.5580 0.0 i
5 Hexe f.328 4.89 7.52
.9855 Aluminum 1.1506.07110 1.6520 0.0 6
Hexa 1.328 2.88 10.53
.9728 Aluminum 1.1506
.07110 1.5580 0.0 7
Hexa 1.328 3.58 10.53
.9728 Aluminum 1.1506.07110 1.6520 0.0 8
Hexa 1.328 4.83 10.53
.9728 Aluminum 1.1506
.07110 f.8060 0.0 9
Square 2.734 2.18 10.18
.7620 55-304
.8594
.04085 1.0287 0.0 i
10 Square 2.734 2.92 10.18
.7620
$$-304
.8594
.04085 1.1049 0.0 11 Square 2.734 3.86 10,18
.7620 55-304
.8594
.04085 1.1938 0.0 12 Square 2.734 7.02 10.18
.7620
$$-304
.8594
.04085 1.4554 0.0 i
13 Square 2.734 8.49 10.18
.7620 55-304
.8594
.04085 1.5621 0.0 i
14 Square 2.734 10.38 10.18
.7620 55-304
.859J
.04085 1.6891 0.0 15 Square 2.734 2.50 10.18
.7620
$5-304
.8594
.04095 1.0617 0.0 l
16 Square 2.734 4.51 10.18
.7620 55-304
.8594
.04085 1.2522 0.0 17 Square 3.745 2.50 10.27
.7544
$$-304
.8600.04060 1.0617 0.0 l
18 Square 3.745 4.51 10.37
.7544 55-304
.8600.04060 1.2522 0.0 19 Square 3.745 4.51 10.37
.7544 55-304
.8600
.04060 1.2522 0.0 20 Sqsa rt 3.745 4.51 10.37
.7544 55-304
.8600
.OdO60 1.2522 456.0 21 Square 3.745 4.51 10.37
.7544
$$-304
.8600.04060 1.2522 709.0 22 Square 3.745 4.51 10.37
.7544 SS-304
.8600.04060 1.2522 1260.0 l
23 Square 3.745 4.51 10.37
.7544
$5-304
.8600.04060 f.2522 1334.0 l
24 Square 3.745 4.51 80.37
.7544
$$-304
.8600.04060 1.2522 1477.0 25 Square 4.069 2.55 9.46 1.1278 55-304 f.2090.04060 1.5113 0.0 l
23 Square 4.069 2.55 9.46 1.1278 55-304 1.2090.04060 1.5l13 3392.0 27 Square 4.069 2.f4 9.46 1.1278 SS-304 1.2090.04060 1.4500 0.0 28 Square 2.490
.2.84 10.24 1.0297 Aluminum 1.2060.08130 1.5113 0.0 29 Square 3.037 2.64 9.28 f.1268 S5-304 1.1701
.07163 1.5550 0.0 30 Square 3.037 8.16 9.28 1.1268
$$~304 1.2701
.07163 2.1980 0.0 l
31 Square 4.069 2.59 9.45 1.1268 55-304 f.2701
.07163 1.5550 0.0 32 Square 4.069 3.53 9.45 1.1268 55-304 1.2701
.07163 f.6840 0.0 33 Square 4.069 8.02 9.45 1.1268 SS-304 1.2701
.07163 2.1980 0.0 1
34 Squaro 4.069 9.90 9.45 1.1268 55-304 f.2701
.07163 2.3810 0.0 1
35 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060.08130 1.5113 1677.0 l
36 Hexa 2.096 2.06 10.38 1.5240 Aluminum 1.6916.07112 2.1737 0.0 37 Hexa 2.096 3.09 10.38 1.5240 Aluminum 1.6916
.07112 2.4052 0.0 38 Hexa 2.096 4.12 10.38 f.5240 Aluminu 1.6916.07112 2.6162 0.0 39 Hews 2.096 6.14 10.38 1.5240 Aluminum 1.6916.07112 2.9891 0.0 40 Hexa 2.096 8.20 10.38 1.5240 Aluminum 1.6916.07112 3.3255 0.0 41 Hexa 1.307 1.01 18.90 1.5240 Aluminum 1.6916.07112 2.1742 0.0 42 Hexa 1.307 1.51 18.90 f.5240 Aluminum 1.6916
.07112 2.4054 0.0 l
43 Hexa 1.307 2.02 18.90 1.5240 Aluminum 1.6916
.07112 2.6162 0.0 l
l l
l l
l l
23 l
l l
l 1
1
l' l
t l
1 l
l 1
Table 6.
Data for U Metal and UO Critical Experiments (P8rt 2 Of 2) l l
l l
I l
Fuel Pellet Clad Clad tattica Case Call A/O H2D/U Oensity 01ameter Material 00 Thickness Pitch 8-10 Number Type U-235 Ratto (G/CC)
(CM)
Clad (CM)
(CM)
(CM)
PPM i
I i
44 Hexa 1.307 3.01 18.90 1.5240 Aluminum 1.6916
.07112 2.9896 0.0 45 Hexa 1.307 4.02 18.90 1.5240 Aluminum f.6916
.07112 3.3249 0.0 l
46 Hexa 1.160 1.Of 18.90 f.5240 Aluminum 1.6916
.07112 2.1742 0.0 47 Hexa 1.160 t.51 18.90 f.5240 Aluminum f.6916
.07112 2.4054 0.0 48 Hexe f.160 2.02 18.90 f.5240 Aluminum 1.6916
.07112 2.6162 0.0
)
49 Hexa 1.160 3.01 18.90 1.5240 Aluminum 1.6916
.07f12 2.9896 0.0 50 Hexa f.160 4.02 18.90 f.5240 Aluminum f.6916
.07112 3.3249 0.0 51 Hexa 1.040 1.01 18.90 1.5240 Aluminum 1.6916
.07112 2.1742 0.0 52 Hexa 1.040 f.51 18.90 1.5240 Aluminum 1.6916
.07112 2.4054 0.0 53 Hexa 1.040 2.02 18.90 f.5240 Aluminum 1.6916
.07112 2.6162 0.0 54 Here 1.040 3.01 18.90 1.5240 Aluminum 1.6916
.07112 2.9896 0.0 55 Hexa 1.040 4.02 18.90 1.5240 Aluminum 1.6916
.07112 3.3249 0.0 56 Hexa 1.307 1.00 18.90
.9830 Aluminum 1.1506
.07112 1.4412 0.0 57 Hexa 1.307 1.52 18.90
.9830 Aluminum 1.1506
.07112 f.5926 0.0 58 Hexa 1.307 2.02 18.90
.9830 Aluminum 1.1506
.07f12 f.7247 0.0 59 Hexa 1.307 3.02 18.90
.9830 Aluminum 1.1506
.07112 1.9609 0.0 60 Hexa 1.307 4.02 18.90
.9830 Aluminum 1.1506
.07112 2.1742 0.0 St Howa f.160 f.52 18.90
.9830 Aluminum 1.1506
.07112 1.5926 0.0 l
62 Hexa 1.160 2.02 18.90
.9830 Aluminum 1.1506
.07tf2 f.7247 0.0 i
63 Hexa 1.160 3.02 18.90
.9830 Aluminum f.1506
.07f12 1.9609 0.0 64 Hexa 1.160 4.02 18.90
.9830 Aluminum 1.1506
.07112 2.1742 0.0 l
65 Hexa 1.160 t.00 18.90
.9830 Aluminum 1.1506
.07112 I.4412 0.0 l
66 Hexa 1.160 1.52 18.90
.9830 Aluminum 1.1506
.07f12 1.5926 0.0 67 Hexa 1.160 2.02 18.90
.9830 Aluminum f.1506
.07112 f.7247 0.0 68 Hexe
- f. ISO 3.02 18.90
.9830 Aluminum I.1506
.07112 I.9609 0.0 69 Hexa f.160 4.02 18.90
.9830 Aluminum 1.1506
.07112 2.1742 0.0 70 Hexo f.040 1.33 18.90 19.050 Aluminum 2.0574
.07620 2.8687 0.0 71 Hexa f.040 f.58 18.90 19.050 Aluminum 2.0574
.07620 3.0086 0.0 72 Hexa 1.040 1.83 18.90 19.050 Aluminum 2.0574 07620 3.1425 0.0 73 Hexa 1.040 2.33 18.90 19.050 Aluminum 2.0574 07620 3.3942 0.0 74 Hexa 1.040 2.83 18.90 19.050 Aluminum 2.0574
.07620 3.6284 0.0 75 Hexa 1.040 3.83 18.90 19.050 Aluminum 2.0574 07620 4.0566 0.0 76 Hexa 1.310 2.02 18.88 1.5240 Aluminum 1.6916
.07112 2.6160 0.0 77 Hexa 1.310 3.01 18.88 1.5240 Aluminum 1.6916
.07112 2.9900 0.0 78 Hexa 1.159 2.02 18.88 1.5240 Aluminum 1.6916 07f12 2.6160 0.0 79 Hexa 1.159 3.01 18.88 f 5240 Aluminum 1.6916
.07112 2.9900 0.0 80 Hexa 1.312 2.03 18.88
.9830 Aluminum 1.1506
.07f12 I.7250 0.0 86 Hexa 1.312 3.02 18.88
.9830 Aluminum 1.1506
.07112 1.9610 0.0 i
l l
l l
l i
l l
l l
24 l
l
4 i
I i
I i
I l
my i
I H
I I
2 l
l N
I I
O I
I W
t I
3 O
l 1
1 ACCEPTABLE I
I Q.
i I
y L_________
g
_________L_________
x 1
I D
I I
CD I
I I
I W2 s
s O
I I
E I
i l
l I
3 I
Og
_____..___T F_________
1 I
Q l
I I
I
>=
I I
-3 I
I O
l NOT ACCEPTABLE 2
I i
W I
i M
_________L__
_________L_________
w q
l I
I I
I I
I l
k.0 4.5 5.0 U-235 ENRICHMENT (W/0)
Figure 1.
SCE&G Fuel Assembly Minimum Burnup vs. Initial U Enrichment for Storage in Region 1 Spent Fuel Racks 25
o 50 1
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Figure 2.
SCE&G Fuel Assembly Minimum Burnup vs. Initial U8 8 8 Enrichment for Storage in Region 3 Spent Fuel Racks 26
i l
1 8.45020.0625 SORA FL EX 0.0 8 22. 0.0 07 THICK 0.0256 g/ cm 2
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Figure 3.
SCE&G Region 1 Spent Fuel Storage Cell Nominal Dimensions l
27 l
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Figure 4.
SCE&G Region 2 Spent Fuel Storage Cell Nominal Dimensions 28
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Figure 5.
SCE&G Region 3 Spent Fuel Storage Cell Nominal Dimensions 29
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Figure 6.
SCE&G Fresh Fuel Storage Cell Nominal Dimensions 30
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Figure 7.
SCE&G Fresh Fuel Rack Layout l
31
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REGION 3 REGION I e REGION 3 RESERVED
/
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SCE&G Spent Fuel Rack Layout 32
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SCE&G Region 2 Checkerboard Fuel Assembly Loading Schematic l
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Figure 10.
Sensitivity of K.sv to Enrichment in the SCE&G Region 2 Spent Fuel Storage Rack with Two of Four Storage 34
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Figure 11. Sensitivity of K.n to Center-to-Conter Spacing in the SCE&G Region 2 Spent Fuel Storage Rack with Two of Four Storage 35
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l Figure 12. Sensitivity of K.ve to B 5 ' Loading in the SCE&G Region 2 Spent Fuel I
Storage Rack with Two of Four Storage l
36 l
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U-235 ENRICHMENT (W/0) t Figure 13. Sensitivity of K.vf to Enrichment in the SCE&G Region 1 Spent Fuel Storage Racks l
l 37 l
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Figure 14. Sensitivity of K.ve to Center-to-Conter Spacing in the SCE&G Region 1 Spent Fuel Storage Racks 38
l 1
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Figure 15. Sensitivity of K.tv to 8 8 ' Loading in the SCE&G Region 1 Spent Fuel Storage Racks 39 l
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Figure 16. Sensitivity of K.e, to Enrichment in the SCE&G Region 3 Spent Fuel Storage Racks 40
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Figure 17. Sensitivity of K.tv to Center-to-Conter Spacing in the SCE&G Region 3 Spent Fuel Storage Racks l
41 l
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.04
.06
.08
.10
.12
.14
.16
.18 CELLWALLTHICKNESS(INCHES)
Figure 18. Sensitivity of K.,e to Steel Can Thickness in the SCE&G Region 3 l
Spent Fuel Storage Racks l
l 42 i
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.10
.15
.20 WATERDENSITY(G/CC)
Figure 19. Sensitivity of K.ve to Water Density in the SCE&G Fresh Fuel Storage Racks 43
.- ~..
I BIBLIOGRAPHY 1.
Nuclear Regulatory Commission, Letter to All Power Reactor Ll::ensees., from B. K. Grimes OT Position for Review and Acceptance of Spent fuel Storage and Handling Applications.,, Aprll 14 1978.
2.
W.
E.
Ford Ill, CSRL-V:
Processed ENDFIB-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORNL/CSDITM-160, June 1982.
3.
N. M. Greene, AMPY: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Ubraries from ENDFIB, ORNLITM-3706, March 1976.
l 4
L M. Petrie and N. F. Cross, KENO IV--An Improved Monte Carlo Criticality Program, DRNL-4938, November 1975.
5.
M. N. Baldwin, Critical Experiments Support'"g Close Proximity Water Sto: age of Power Reactor Fuel, BAW-1484-7, July 1979.
6.
J.
T.
Thomas, Critical Three-Olmensional Arr.rys of U (93.2) Metal Cy/Ind.tts, Nuclear Science ano Engineering, Volume 52, pages 350-359, i
l 1973.
I 7.
A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Bol/Ing Water Reactors, WCAP-10106, June 1982.
8.
Askew, J. R., Fayers, F. J., and Kemshell, P. B., A General Description of tne lattice Code WIMS, Journal of British Nuclear Energy Society, 5, pp.
564-584, 1966.
9.
England, T.
R., CINDER - A One-Point Depletion and Fission Product Program, WAPD-TM-334, August 1962.
- 10. Meiehan, J.
B., yankee Core Evaluation Program Final Report.
WCAP-3017-6094, January 1971.
l l
Bibliography 44 l
t
9 n%A A JZ SUPPLEMENTARY CRITICALITY ANALYSIS OF V. C. SUMMER FUEL RACKS l
1 l
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TABLE OF CONTENTS 1.0 Introduction 1
1.1 Design Description
.................................2 1.2 Design Criteria 2
2.0 Criticality Analytical Method 3
3.0 Criticality Analysis of Region 3 Spent f uel Racks 4
3.1 Reactivity Calculations 4
3.2 Postulated Accidents
...............................6 3.3 Sensitivity Analysis 6
4.0 Criticality Analysis of Region 2 Spent Fuel Rack 7
4.1 Reactivity Equivalencing 7
4.2 Analytical Methods 8
4.3 Reactivity Calculations
..............................9 4.4 Postulated Accidents 10 4.5 Sensitivity Analysis 11 5.0 Acceptance Criterion For Criticality 12 i
Bibliography 31 l
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l r
Table of Contents i
i l
i LIST OF TABLES Table
- 1. Benchmark Critical Experiments (5,6]
13 Table
- 2. Fuel Parameters Employed in Criticality Analysis 14 Table
- 3. V. C. Summer Fuel Assembly Minimum Burnup vs initial U8 8 8 Enrichment for Region 2 Spent Fuel Rack 15 Table
- 4. Comparison of PHOENIX lsotopics Predictions to Yankee Core 5 Measurements 16 Table
- 5. Benchmark Critical Experiments PHOENIX Comparison 17 l
Table
- 6. Data for U Metal and UO Critical Experiments 18 l
I I
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l f
l List of Tables ii f
1
LIST OF ILLUSTRATIONS Figure
- 1. SCE&G Fuel Assembly Minimum Burnup vs. Initial U8 8
- Enrichment for Storage in Region 2 Spent Fuel Racks 20 Figure
- 2. SCE&G Region 2 Spent Fuel Storage Cell Nominal Dimensions 21 Figure
- 3. SCE&G Region 3 Spent Fuel Storage Cell Nominal Dimensions 22 Figure
- 4. SCE&G Spent Fuel Rack Layout 23 Figure
- 5. SCE&G Region 3 Checkerboard Fuel Assembly Loading Schematic 24 Figure
- 6. Sensitivity of K.re to Enrichment in the SCE&G Region 3 Spent Fuel Storage Rack with Two of Four Storage 25 Figure
- 7. Sensitivity of K.re to Center-to-Center Spacing the SCE&G Region 3 Spent Fuel Storage Rack with Two oi our Storage 26 Figure
- 8. Sensitivity of K.ee to Steel Can Thickness in the SCE&G Region 3 Spent Fuel Storage Rack with Two of Four Storage 27 Figure
- 9. Sensitivity of K.ee to Enrichment in the SCE&G Region 2 Spent Fuel Storage Rack 28 Figure 10. Sensitivity of K.et to Center-to-Center Spacing in the SCE&G Region 2 Spent Fuel Storage Rack 29 Figure 11. Sensitivity of Kere to B 8 ' Loading in the SCE&G Region 2 Spent Fuel Storage Rack 30 4
List of Illustrations ill
l i
i
1.0 INTRODUCTION
The V. C. Summer spent fuel rack (SFR) design described herein employs three arrays of racks, which will be considered as three separate spent fuel racks.
Each of these fuel racks or arrays consists of existing SCE&G fuel racks. This supplemental analysis will reanalyze two of the fuel arrays. The largest array referred to as Region 3 will be. reanalyzed for critica!ity to show that 4.55 w/o fuel can be stored in the rack in two out of four storage locations.
The smallest array, Region 2 will be reanalyzed to take into consideration the changes in fuel and fission product inventory resulting from depletion in the reactor core up to an enrichment of 5.0 w/o. The Regions 1, 2 and 3 spent fuel rach designs are poisoned and non-poisoned stainless steel racks, previously analyzed by Westinghouse for storage of 17x17 STD and OFA fuel to show:
1.
The cells in Region 1 can accommodate fuel assembiles with initial enrichments of 5.0 w/o U 8 8 8 and a minimum burnup of 4,000 MWD /MTU, 2.
Region 2 can store freshiy discharged fuel assemblies with enrichments up to 5.0 w/o U 8 8 8 in a checkerboard pattern, 3.
The Region 3 cells can accommodate fuel assemblies with initial enrichment of 5.0 w/o U 8 8 8 and a minimum burnup of 48,000 MWDIMTU.
The Region 2 spent fuel rack supplemental reanalysis is based on maintaining K.ee s 0.95 for storage of Westinghouse 17x17 OFA and STD fuel at 5.0 w/o U888 with an initial enrichment /burnup combination in the acceptable area of Figure 1 with utilization of every cell permit *ed for storage of the fuel as-semblies.
Introduction 1
1.1 DESIGN DESCRIPTION The Region 2 and 3 spent fuel storage cell design are depicted schematically in Figures 2 and 3 with nominal dimensions giQn on the figures. The spent fuel rack layout is shown in Figure 4.
1.2 DESIGN CRITERIA Criticality of fuel assemblies in a fuel storage rack is prevented by tne design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between assemblies.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (Ken) of the fuel assembly array will be less than 0.95 as recommended in ANSI 57.2-1983 and in Reference 1.
Introduction 2
l 1
,=
2.0 CRITICALITY ANALYTICAL METHOD The criticality calculation method and ;ross-section values are verified by comparison with critical experiment data for assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps and low moderator densities.
l The design method which insures the criticality safety of fuel assemblies in the spent fuel storage rack uses the AMPX( 8, 88 system of codes for cross-section generation and KENO IVt * ) for reactivity determination.
The 227 energy group cross-section library that is the common starting point for all cross-sections used for the benchmarks and the storage rack is generated from ENDF/B-V8 8 3 data. The NITAWL' 8 8 program includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry.
The Nordheim Integral Treatment is used.
Energy and spatial weighting of cross-sections is performed by the XSDRNPMt 8 3 program which is a one-dimensional Sa transport theory code. These multigroup cross-section sets are then used as input to KENO IV( *3 which is a three dimensional Monte Carlo theory program designed for reactivity calcult.tions.
l A set of 33 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and variability. The experiments range from water moderated, oxide fuel arrays separated by various materials (B4C, steel, water, etc) that simulate LWR fuel shipping and storage conditions ( 83 to dry, harder spectrum uranium metal l
cylinder arrays with various interspersed materials'
- 5 (Plexiglas and air) that demonstrate the wide range of applicability of the method. Table 1 summarizes these experiments.
The average K.et of the benchmarks is 0.992. The standard deviation of the bias t
value is 0.0008 Ak.
The 95/95 one sided tolerance limit factor for 33 values is 2.19.
Thus, there is a 95 percent probability with a 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0018 Ak, Criticality Analytical Method 3
l
3.0 CRITICALITY ANALYSIS OF REGION 3 SPENT FUEL RACKS 3.1 REACTIVITY CALCULATIONS The following assumptions were used to develop the nominal case KENO model for the Region 3 spent fuel rack storage of fresh fuel using two out of four storage locations:
1.
The fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burnable poison in the fuel rods. Historically, calculations for spent fuel racks similar to the Re-gion 3 racks analyzed herein have shown that the W 17x17 OFA fuel as-sembly yields a larger K.et than does the W 17x17 Standard fuel assembly when both fuel assemblies have the same U8 8 8 enrichment. Thus, only the W 17x17 OFA fuel assembly was analyzed for Region 3.
(See Table 2 for fuel parameters).
2.
All fuel rods contain uranium dioxide at an enrichment of 4.5 w/o U8 8 8 l
over the infinite length of each rod.
3.
No credit is taken for any U8 8
- or U8 8
- in the fuel, nor is any credit taken for the builduo of fission product poison material.
4.
The moderator is pure water at a temperature of 68'F. A conservative value 8
of 1.0 gm/cm is used f or the density of water.
5.
No credit is taken for any spacer grids or spacer sleeves.
6.
Fuel assemblies are loaded into two of every four cells in a checkerboard pattern in the storage cells as shown in Figure 5.
l l
7.
The array is infinite in lateral and axial extent which precludes any neutron l
leakage from the array.
l The KENO calculation for the nominal case resulted in a K.v. of 0.9325 with a 95 percent probability /95 percent confidence level uncertainty of 0.0069.
I I
Criticality Analysis of Region 3 Spent Fuel Racks 4
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l
The maximum K.ee under normal conditions arises from considerration of me-chanical and materlal thickness tolerances resulting from the manufacturing process in addition to asymmetric positioning of fuel assemblies within the storage cells. Studies of asymmetric positioning of fuel assemblies within the storage cells has shown that symmetrically placed fuel assemblies yleid con-servative results in rack K.ve. The sheet metal tolerances are considered along with construction tolerances related to the cell I.D., and cell center-to-center spacing. For the Region 3 storage racks, the steel thickness is reduced from the nominal value of 0.09" to its minimum value. Thus, the most conservative, or "worst case" KENO model of the Region 3 storage racks contains a steel thickness of 0.085" with symmetrically placed fuel assemblies. This model is also finite in the axial direction to include axial neutron leakage from the array.
Based on the analysis described above, the following equation is used to de-velop the maximum K.ve for the V. C. Summer Region 3 spent fuel storage racks with two out of four storage:
I K.ve = K..< n + Bmnaoa + (( (ks)8..,n + (ks)8 munoe where:
worst case KENO K tv that includes materisl
=
tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal method bias determined from benchmark
=
critical comparisons
- gl95 uncertainty in the worst case KENO 95/95 uncertainty in the method bias
=
Substituting calculated values in the order listed above, the result is:
K.ev = 0.e*J29 + 0.0083 + /[(0.0054)8 + (0.0018)8 ] = 0.9469 Based on the results of the enrichment sensitivty study (Section 3.3), an increase of enrichment from 4.5 to 4.55 w/o will increase the reactivity less than 0.0020 delta k.
Since K.ve is less than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met with fuel enriched to 4.55 w/o.
Criticality Analysis of Region 3 Spent Fuel Racks 5
\\
3.2 POSTULATED ACCIDENTS Most accident conditions will not result in an increase in K.et of' the rack. Ex-amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel height of stored assemblies which precludes interaction).
However, accidents can be postulated which would increase reactivity (i.e., or dropping a fuel assembly between the rack and pool wall). For these accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by about 30 percent.iK.
Thus, for postulated accidents, should there be a reactivity increase, K.ve would be less than or equal to 0.95 due to the effect of the dissolved boron.
3.3 SENSITIVITY ANALYSIS To show the dependence of K es on fuel and storage cells parameters as re-quested by the NRC, the variation of the K.et with respect to the following pa-rameters was developed using the KENO computer code:
1.
Fuel enrichment.
2.
Center-to-center spacing of storage cells.
3.
Cell stainless steel can ti'ickness.
Results of the sensitivity analysis for the Region 3 storage cells are shown in Figures 6 through 8 for two of four storage.
I J
Criticality Analysis of Region 3 Spent Fuel Racks 6
1
4.0 CRITICALITY ANALYSIS OF REGION 2 SPENT FUEL RACK This section develops and describes the analytical techniques and models em-ployed to perform the criticality analyses for storage of spent fue' in Region 2 of the V. C. Summer spent fuel pool.
4.1 REACTIVITY EQUIVALENCING Spent fuel storage, in the Region 2 spent fuel storage racks, is achievable by meant of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with fuel depletion. A series of reactivity calculations are performed to 9*qerate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield the equivalent Keve when the fuel is stored in the Region 2 racks.
Figure 1 shows the constant Ken contour generated for the V. C. Summer Region 2 racks. Note in Figure 1 the endpoint at 0 MWD /MTU where the enrichment is 2.3 w/o cn.1 at 25,700 MWDIMTU where the enrichment is 5.0 wlo. The in-tecpretation of the endpoint data is as follo,ws: the reactivity of the Region 2 racks containing fuel at 25,700 MWD /MTU burnup which had an initial enrichment of 5.0 w/o is equivalent to the reactivity of the Region 2 racks containing fresh fuel having an initial enrichment of 2.3 w/o. It is important to recognize that the curve in Figure 1 is based on a constant reck reactivity for tnat region and not on a constant fuel assembly reactivity. The data in Figure 1 is also pro-vided as Table 3.
Linear interpolation between two data points on this table will yield conservative results.
Criticality Analysis of Region 2 Spent Fuel Rack 7
)
4.2 ANALYTICAL METHODS The data points on the reactivity equivalence curve were generated with a transport theory computer code, PHOENIX ( ')
PHOENIX is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code. A 25 energy group nuclear data library based on a modified version of the British WIMS( ')
library is used with PHOENIX.
A stucy was done to examine fuel reactivity as a function of time following discharge from the reactor.
Fission product decay was accounted for using CINDER ( ')
CINDER is a point-depletion computer code used to determine fission product activities. The fission products were permitted to decay for 30 years af ter discharge. The fuel reactivity was found to reach a maximum at approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after discharge. At this point in time, the major fission product poison, Xe*8' has nearly completely decayed away.
Fur-thermore, the fuel reactivity was found to decrease continuously from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 30 years following discharge. Therefore, the most reactive point in time for a fuel assembly af ter discharge from the reactor can be conservatively ap-proximated by removing the Xe * * *.
The PHOENIX code has been validated by comparisons with experiments wnere isotopic fuel composition has been examined following discharge from a reac-tor. In addition, an extensive set of benchmark critical experiments has been analyzed with PHOENIX. Comrarisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 4.
The mearure-ments were made on fuel diccharged from Yankee Core 58 8 ')
The data in Table 4 shows that the agreement between PHOENIX predictions and measured isotopic compositions is good.
The agreement between reactivities computed with PHOENIX and the rasults of 81 critical benchmark experiments is summarized in Table 5.
Key parameters describing each of the 81 experiments are given in Table 6.
Tnese reactivity comparisons again show good agreement between experiment and PHOENIX calculations.
An uncertainty associated with the burnup-dependent reactivities computed with PHOENIX is accounted for in the development of the Region 2 burnup require-m ents. A bias of 0.01.ik at 30,000 MWD /MTU is considered to be very con-servative since comparison between PHOENIX results and the Yankoe Core experiments and 81 benchmark experiments indicates closer agreement.
Criticality Analysis of Region 2 Spent Fuel Rack 8
4.3 REACTIVITY CALCULATIONS The nominal and maximum K.tr for storage of spent fuel m Region 2 is deter-mined using the methods described in Section 2 for Region 3 in addition to the methods described in Sectic-4.2.
The actual conditions for this determittation are defined by the zero burnup intercept point in Figure 1.
The KENO-IV com-puter code is used to calculate the storage rack mLitiplication f actor with an equivalent fresh fuel enrichment of 2.3 w/o. Combinatioris of fuel enrichment and discharge burnup yielding the same rack multiplicaticn factor as at the ~2ero burnup intercept are determined with PHOENIX.
The following assumotions were used to develop the nominal case KENO model for the Region 2 storage of spent fuel:
1.
Calculations for the Region 2 racks analyzed herein have shown that the Westinghouse 17x17 OFA fuel assembly yields a larger K.tv than does the Westinghouse 17x17 standard fuel assembly when both fuel assemblies have the same U8 8 ' enrichment. Thus, only the Westinghouse 17x17 OFA fuel assembly was analyzed for Region 2.
2.
The Westinghouse 17x17 OFA spent fuel assembly contains uranium dioxide fuel at an equivalent "fresh fuel" enrichment of 2.3 w/o U 8 * '.
3.
The moderator is pure water at a temperature of 68'F.
A conservative value of 1.0 gm/cm* is used for the density of water.
4.
No credit is taken for any spacer grids or spacer sleeves.
5.
The array is infinite in lateral and axial extent which precludes any neutron leakage from the Stray.
6.
The minimum poison material loading of 0.0015 grams B-10 per square centimeter is used throughout the array.
The KENO calculation for the nominal case resulted in a K.tv of 0.9038 with a 95 percent probability /95 percent confidence level uncertainty of !0.0046.
The maximum K.et under normal conditions was determined with a "worst case" KENO model, in the same manner as for the Region 3 storage racks (see Section 3). For the Region 2 racks, the water gaps are reduced from the nominal values of 1.261" and 1.046" to their minimum values. Thus, the "worst case" KENO model of the Region 2 storage racks contain minimum water gaps of 1.198" and 0.983" with symmetrically placed fuel assemblies. The uncertainty associ-ated with the reactivity equivalence methodology was included in the develop-Criticality Analysis of Region 2 Spent Fuel Rack 9
ment of the burnup requirements.
This uncertainty was discussed in Section 4.2.
Based on the analysis described above, the following equation is used to de-velop the maximum Keit for the storage of spent fuel in the V. C. Summer Re-gion 2 spent fuel storage racks:
K tv =
K..,n + Sm.inoe + Bwi + /((ks) 8
,si + (ks)8 m.inoo 3
where:
= worst :3=a KENO K.ve that includes centered fuel assembly positions, material tolerances, and mechanical tolerance which can result in spacing between assemblies less than nonmial method bias determined from benchmark
=
critical comparisons bias to account for posion partical
=
l self-shielding 95/95 uncertainty in the worst case KENO
=
Kevi 95/95 uncertainty in the method bias
=
Substituting calculated values in the order listed above, the result is:
K.te = 0.9289 + 0.0083 + 0.0045 + /[(0.0052)8 (0.0018)8 ] = 0.9472
+
The maximum K.et for Region 2 for this configuration is less than 0.95, including all uncertainties at a 95/95 probability / confidence, level. Therefore, the accept-ance criteria for criticality are met for storage of spent fuel at an equivalent "fresh fuel" enrichment of 2.3 w/o U8 8 8 4.4 POSTULATED ACCIDENTS Most accident conditions will not result in an increase in K.ve of the rack. Ex-i amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active l
fuel height of stored assemblies which prer.ludes interaction).
I l
Criticality Analysis of Region 2 Spent Fuel Rack 10 1
l
However, accidents can be postulated which would increase reactivity (i.e.,
misloading an assembly with a burnup and enrichment combination outside of the acceptable area in Figure 1. or dropping a fuel assembly between the rack and pool wall). For these accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by about 30 percent AK. Thus, for postulated accidents, should there be a reactivity increase, K.ev would be less than or equal to 0.95 due to the effect of the dissolved boron.
4.5 SENSITIVITY ANALYSIS To show the dependence of K.*
on fuel and storage cell parameters as re-quested by the NRC, sensitivity studies were performed in which the poison loading, the fuel enrichment, and the storage cell center-to-center spacing were varied, using the KENO computer code.
Figures 9 through 11 illustrate the results of the sensitivity studles for spent fuel occupying overy cell in the Region 2 fuel racks.
Criticality Analysis of Region 2 Spent Fuel Rack 11
5.0 ACCEPTANCE CRITERION FOR CRITICALITY The neutron multiplication f actor in spent fuel pool shall be less than or equal to 0.95, including all uncertainties, under all conditions.
The analytical methods employed herein conform with ANSI N18.2-1973, "Nu-clear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Handling System: ANSI 57.2-1983, "Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2:
ANSI N16.9-1975, "Validation of Calculational Methods for Nuclear Criticality Safety," NRC Standard Review Plan, Section 9.1.2, "Spent Fuel Storage"; and the NRC guidance, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications.
Acceptance Criterion For Criticality 12 i
(
Table 1.
Benchmark Critical Experiments (5,6) cenecas enetet-mnt separatin9 soluble Description w/o U235 Reflector matertal B 10 ppm K,gg 1.
UO rod tattice 2.86 water w a t er 0
0.9857 *.0028 2.
VO rod tattIce 2.86 matar m a t ee 1037 0.9906 7.0018 3.
UO rod lattice 2.46 eater watee 764 0.9R96 7.0065 4
UO rod lattice 2.46 mater 84C pins 0
0.9994 7.0025 5.
UO rod lattice 2.46 mater Sec pins 0
0.9R91 T. 0076 6.
UO rod lattice 2.46 mater 94C pins 0
0.9955 7.0020 7.
UO Pod lattice 2.46 unter 94C pins 0
0.9999 7.0076 8.
UO rod lattice 2.46 watee 8*C pins 0
0.9983 7.0075 9.
UO rod lattice 2.46
=a t ee watec 0
0 9938 T.0078 10.
DO Pod tattice 2.86 mater unt er 143 0.992g 7.0075 11 VO rod tattice 2.46 m at er statntess steet 584 0.9967 7.00?0 12.
UO rod lattice 2.86 eater stainless steel 217 0.9943 7.00r9 13.
UO Pod lattice 2.66 w a t er borated aluminum IS O 9992 7.0023 14 UO rod lattice 2.46 sa t ee bor a t ed a l uu t mse 92 0.9994 7.0073 15.
UO rod tattice 2.46 ma t er borat ed s t uminum 295 0.9832 7.0021 16.
UO rod lattice 2.46 enter borat ed aluminum 121 0.gneg 7.o024 17.
UO Pod tattlee 2.84 mater borated aluminum def 0.9995 7.0070 10.
Do rod lattice 2.46 mater bor a t ed s t uminum 197 0 9495 7.0072 19.
DO rod tattIce 2.46 ester borat ed aluminum 634 0.992t T.00tp 20.
VO rod tattlee 2.46 unter borated aluminum 3?0 0.9970 7.0070 21.
UO rod lattice 2.a6 unter bora t ed a luminum 72 0.99 29 I.0020 22.
U tal cyttreees 93.2 bare air 0
0.9905 T.0070 23.
U met al cy t tneers 93.2 bare str 0
0.9916 7.0070 24 U metal cyllneces 93.2 bare ele 0
0.9947 7.0025 i
25.
U metal cyllno*es 93.2 bare air 0
0.9978 7.0019 76.
U metal cylinoers 93.2 bare air 0
0.9972 7.0076 27.
U metal cyllno*rs 93.2 bare ate 0
0.9950 7.0027 78.
U metal cylinders 93.2 bare pleatgtass 0
0.9941 7.0030 29.
U **tal cyttnoers 93.2 paraffin plentglass 0
0.9978 T.004l 30.
U metal cyltnoers 93.2 bare plentglass 0
0.9968 T.0018 31.
U metal cyllneers 93.2 paraffin plealglass o
t 0042 7.0019 32.
U met al cy l i nder s 93.2 parafftn pleatglass 0
0.9963 7.0020 l
33.
U metal cylinders n3.2 paraffin plestglass 0
0.9999 T.0032 l
i 13
c Table 2.
Fuel Parameters Employed in Criticality Analysis Parameter W 17x17 OFA W 17x17 STANDARD Number of Fuel Rods per Assembly 264 264 Rod Zirc-4 Clad 0.D.
(I nch) 0 360 0 374 Clad Thickness (inch) 0.0225 0.0225 Fuel Pellet 0.D. (inch) 0 3088 0 3225 Fuel Pellet Density
(% of Theoretical) 96 96 Fuel Pellet Dishing Factor 0.0 0.0 Rod Pitch (inch) 0.496 0.496 Number of Zirc-4 Guide Tubes 24 24 Guide Tube 0.0.
(Inch) 0.474 0.4848 Guide Tube Thickness (Inch) 0.016 0.0188 Number of instrument Tubes 1
1 Instrument Tube 0.0.
(Inch) 0.474 0.484' instrument Tube Thickness (I nch) 0.016 0.0188 1
in... en.n,..
n.....e
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in
.,v.in. Two. o.o..a in.c a... to e. o,.4 a.ae o.o ie iaca.....
.uve.v.
. m n... a.....
. ci on in. e...ii
.a
....iv
..... i n,..n. i v....
14
l Table 3.
V. C Summer Fuel Assembly Minimum Burnup vs initial U88' Enrichment for Region 2 Spent Fuel Rack initial U 8 8 8 Assembly Discharge l
Enrichment Burnup (GWD/MTU)
Region 2 23 0
2.6 33 30 72 35 12.2 4.0 16.6 45 21.4 50 25 7 Acceptance Criterion For Criticality 15
Table 4.
Comparison of PHOENIX isotopics Predictions to Yankee Core 5 Measurements Quantity (Atom Ratio)
% Difference U235/U
-0.67 U236/U
-0.28 U238/U
-0.03 PU239/U
+ 3.27 PU240/U
+ 3.63 PU241/U
-7.01 PU242/U
-0.20 PU239/U238
+3.24 Mass (PU/U)
+ 1.41 FISS-PU/ TOT-PU
-0.02 l
a i
16 l
+
Table 5.
Benchmark Critical Experiments PHOENIX Comparison Description of Number of PHOENIX X.e, Using Experiment Experiments Experiments Bucklings M931.
Al clad 14 0.9947 SS clad 19 0.9944 Borated H2O 7
0.9940 Subtotal 40 0.9944 U-Metal Al clad 41 1.0012 TOTAL 81 0.9978 l
l l
i 1
17 l
l
Table 6.
Data for U Metal and UO Critical Experiments (Part 1 of 2)
Fuel Pellet Clad Clad Lattice Case Call A/O H20/U Density Diameter Mateetal DO 7hteknese Pitch 8-10 Numeer 7ype U-235 Ratio (G/CC)
(CM)
Clad (CM)
(CM)
(CM)
PPM i
Hexa 1.328 3.02 7.53 1.5265 aluminum f.6916
.07110 2.2050 0.0 2
Hexa 1.328 3.95 7.53 1.5265 Aluminum 1.6916.07110 2.3590 0.0 3
Hexa 1.328 4.95 7.53 1.5265 Aluminum 1.6916
.07110 2.5120 0.0 l
4 Hexa 1.328 3.92 7.52
.9855 aluminum 1.1506
.07110 1.5580 0.0 l
5 Hexa 1.328 4.89 7.52
.9855 Aluminum 1.1506
.07110 1.6520 0.0 6
Hexa 1.328 2.88 10.53
.9728 Aluminum 1.1506
.07110 1.5580 0.0 t
7 Hexa 1.328 3.58 10.53
.9728 Aluminum 1.1506
.07110 1.6520 0.0 8
Hexa 1.328 4.83 10.53
.9728 Aluminum 1.1506
.07110 1.8060 0.0 9
Souare 2.734 2.18 10.18
.7620
$5-304
.8594 94085 1.0287 0.0 to Square 2.734 2.92 10.18 7620 55-304
.8594
.04085 1.1049 0.0 11 Souare 2.734 3.86 10.18
.7620
$$-304
.8594
.04085 1.1938 0.0 12 Souare 2.734 7.02 10.18
.7620 55-304
.8594
.04085 1.4554 0.0 13 Souare 2.734 8.49 10.18
.7620 53-304
.8594
.04085 1.5621 0.0 14 Souare 2.734 10.38 10.18
.7620 55-304
.8594
.04085 1.6891 0.0 15 Souare 2.734 2.50 10.18
.7620 55-304
.8594
.04085 1.0617 0.0 16 Square 2.734 4.51 10.18
.7620 55-304
.8594
.04085 1.2522 0.0 17 Souare 3.745 2.50 10.27
.7544
$$-304
.8600
.04060 1.0617 0.0 18 Souare 3.745 4.51 10.37
.7544
$$-304
.8600
.04060 1.2522 0.0 l
19 Souare 3.745 4.51 10.37
.7544
$$-304
.8600
.04060 1.2522 0.0 20 Souare 3.745 4.51 10.37
.7544 55 304
.8600
.04060 1.2522 456.0 21 Souare 3.745 4.51 10.37
.7544 55-304
.8600
.04060 1.2522 7 09.0 22 Souare 3.745 4.51 10.37
.7544
$5-304
.8600
.04060 1.2522 1260.0 23 Souare 3.745 4.51 10.37
.7544
$$-304
.8600
.04060 1 2522 1334.0 24 Souare 3.745 4.51 10.37
.7544 55-304
.8600
.04060 1.2522 1477.0 25 Souare 4.069 2.55 9.46 1.1278 55-304 1.2090.04060 1.5113 0.0 26 Souare 4.069 2.55 9.46 1.1278
$$.304 1.2090
.04060 1.5113 3392.0 27 Square 4.069 2.14 9.46 1.1278
$$-304 1.2090
.04060 1.4500 0.0 2B Souare 2.490
.2.84 10.24 1.0297 Aluminum 1.2060.08130 1.5113 0.0 29 Square 3.037 2.64 9.28 1.1268 55-304 1.1701
.07163 1.5550 0.0 30 Square 3.037 8.16 9.28 1.1268
$$-304 1.2701
.07162 2.1980 0.0 31 Square 4.069 2.59 9.45 1.1268
$5-304 1.2701
.07163 f.5550 0.0 32 Square 4.069 3.53 9.45 1.1268 55-304 1.2701
.07163 f.6840 0.0 33 Square 4.069 8.02 9.45 1.1268 55-304 1.2701
.07163 2.1980 0.0 34 Souare 4.069 9.90 9.45 1.1268 55-304 1.2701
.07163 2.3810 0.0 35 Souare 2.490 2.84 10.24 1.0297 Aluminum 1.2060
.08130 1.5113 1877.0 36 Hexa 2.096 2.06 10.38 1.5240 Aluminum 1.6916
.07112 2.1737 0.0 37 Hexa 2.096 3.09 10.38 1.5240 Aluminum 1.6916
.07112 2.4052 0.0 38 Hexa 2.096 4.12 10.38 1.5240 Aluminum 1.6916
.07112 2.6162 0.0 39 Hexa 2.096 6.14 10.38 1.5240 Aluminum 1.6916
.07112 2.9891 0.0 40 Hews 2.096 8.20 10.38 1.5240 Aluminum 1.6916
.07112 3.3255 0.0 4i Hexa 1.307 1.01 18.90 1.5240 Aluminum 1.6916
.07112 2.1742 0.C 42 Hexa 1.307 1.51 18.90 1.5240 Aluminum 1.6916
.07112 2.4054 0.0 43 Hexa 1.307 2.02 18.90 1.5240 Aluminum 1.6916
.07112 2.6162 0.0 18
?
Table 6.
Data for U Metal and UO: Critical Experiments (Part 2 Of 2)
Fuel Pellet Clad Clad Lattice case Cell A/O H2D/U Donalty Diameter Matertal 00 Th t ek nes s Pitch 8-10 Number Type U-235 Ratio (G/TC)
(CM)
Clad (CM)
(CM)
(CM)
PPM 44 Hexa 1.307 3.01 18.90 1.5240 Aluminum 1.6916
.07112 2.9896 0.0 45 Hexa 1.307 4.02 18.90 1.5240 Aluminum 1.6916
.07112 3.3249 0.0 46 Hexa 1.160 1.01 18.90 1.5240 Aluminum 1.6916
.07112 2.1742 0.0 47 Hexa 1.160 1.51 18.90 1.5240 Aluminum 1.6916
.07112 2.4054 0.0 48 Hexa 1.160 2.02 18.90 1.5240 Aluminum f.6916
.07112 2.6f62 0.0 49 Hexa 1.160 3.01 18.90 1.5240 Aluminum 1.6916
.07112 2.9896 0.0 50 Hexa 1.160 4.02 18.90 1.5240 Aluminum 1.6916
.07112 3.3249 0.0 51 Hexa 1.040 1.01 18.90 1.5240 Aluminum 1.6916
.07112 2.1742 0.0 52 Hexa 1.040 1.51 18.90 1.5240 Aluminum 1.6916
.07112 2.4054 0.0 53 Hexa 1.040 2.02 18.90 1.5240 Aluminum 1.6916
.07tf2 2.6f62 0.0 54 Hexa 1.040 3.01 18.90 1.5240 Aluminum 1.6916
.07112 2.9896 0.0 55 Hexa 1.040 4.02 18.90 1.5240 Aluminum 1.6916
.07112 3.3249 0.0 56 Hexa 1.307 1.00 18.90
.9830 Aluminum 1.1506
,07112 1.4412 0.0 57 Hexa 1.307 1.52 18.90
.9830 A l um f itum 1.1506
.07112 f.5926 0.0 58 Hexa 1.307 2.02 18.90
.9830 Aluminum 1.1506
.07112 f.7247 0.0 59 Hexa 1.307 3.02 18.90
.9830 Aluminum 1.1506
.07tt2 f.9609 0.0 60 Hexa 1.307 4.02 10.90
.9830 Aluminum 1.1506
.07112 2.1742 0.0 61 Hexa 1.160 1.52 18.90
.9930 Aluminum 1.1506
.07112 f.5926 0.0 62 Hexa 1.160 2.02 18.90
.9830 Aluminum 1.1506
.07112 1.7247 0.0 63 Hexa 1.160 3.02 18.90
.9830 Aluminum 1.1506
.07t12 1.9609 0.0 64 Hexa 1.160 4.02 18.90
.9830 Aluminum 1.1506
.07tf2 2.I742 0.0 65 Hexa 1.160 1.00 18.90
.9830 Aluminum 1.1506
.07112 1.4412 0.0 66 Hexa 1.160 1.52 18.90
.9830 Aluminum 1.1506.07f12 f.5926 0.0 67 Heus 1.160 2.02 18.90
.9830 Aluminum 1.1506
.07112 1.7247 0.0 68 Hexa 1.160 3.02 18.90
.9830 Aluminum 1.1506
.07112 f.9609 0.0 69 Hexa 1.160 4.02 18.90
.9830 Aluminum 1.1506
.07112 2.1742 0.0 70 Hexa 1.040 1.33 18.90 19.050 Aluminum 2.0574
.07620 2.8687 0.0 71 Heus 1.040 1.58 18.90 19.050 AlumJnum 2.0574
.07620 3.0086 0.0 72 Hexa 1.040 1.83 18.90 19.050 Aluminum 2.0574
.07620 3.1425 0.0 73 Hexa 1.040 2.33 18.90 19.050 Aluminum 2.0574
.07620 3.3942 0.0 74 Hexa 1.040 2.83 18.90 19.050 Aluminum 2.0574
.07620 3.6284 0.0 75 Hexa 1.040 3.83 18.90 19.050 Aluminum 2.0574
.07620 4.0566 0.0 76 Hexa 1.310 2.02 18.88 1.5240 Aluminum 1.6916 07112 2.6160 0.0 77 Hexa 1.310 3.01 18.88 1.5240 Aluminum 1.6916
.07tt2 2.9900 0.0 78 Hexa 1.159 2.02 18.88 1.5240 Aluminum 1.6916
.07112 2.6160 0.0 79 Hexa 1.159 3.01 18.88 1.5240 Aluminum 1.6916
.07f12 2.9900 0.0 80 Hexa 1.312 2.03 18.88
.9830 Aluminum 1.1506
.07112 1.7250 0.0 81 Hexa 1.312 3.02 18.88
.9830 Aluminum 1.1506
.07tf2 1.9610 0.0 19
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Figure 1.
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l
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BOR AFLEX- 0.0322.0 07 THICK O.00 20 g /em 2 j
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(PITCH)
( N OT TO SCALE)
Figure 2.
SCE&G Region 2 Spent Fuel Storage Call Nominal Dimensions 21
dk y
00000000000000000 00000000000000000 t
t 00000000000000000 R
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(INSIDE BOX) b 9.030 y
10.I16 = 0.032
]
( PIT C H )
l (NOT TO SCALE)
Figure 3.
SCE&G Region 3 Spent Fuel Storage Cell Nominal Dimensions 22
O t
e W
12 0 hE T}$
'ni REk 6
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N RE610N 3 REGION I,
REGION 3 RESERVED
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SCE1G Spent Fuel Rack Layout l
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Figure 5.
SCE&G Region 3 Checkerboard Fuel Assembly Loading Schematic 24 i
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Figure 6.
Sensitivity of K.et to Enrichment in the SCE&G Region 3 Spent Fuel Storage Rack with Two of Four Storage 25
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Figure 7.
Sensitivity of Keet to Center-to-Center Spacing in the SCE&G Region 3 Spent Fuel Storage Rack with Two of Four Storage 26
+
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Figure 8.
Sensitivity of %, to Steel Can Thickness in the SCE&G Region 3 Spent Fuel Storage Rack with Two of Four Storage 27
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Figure 3.
Sensitivity of K.ve to Enrichment in the SCE&G Region 2 Spent Fuel Stc, rage Rack t
28
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Figure 11. Sensitivity of K.it to B 8 ' Loading in the SCE&G Region 2 Spent Fuel Storage Rack.
30
c 7
1 BIBLIOGRAPHY 1.
Nuclear Regulatory Commission, Letter to All Power Reactor Licensees,, from B. K. Grimes OT Postuon for Review and Acceptance of Spent Fuel Storage and Handling Applications.,, April 14 1978.
2.
W.
E.
Ford \\ \\ \\, CSRL-V:
Processed ENDFIB-V 227-Neutron-Group and Pointw!se Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORNL/CSDITM-160, June 1982.
3.
N.
M.
Greene AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDFIB, ORNLITM-3706, March 1976.
4.
L M. Petrie and N. F. Cross, KENO IV--An Improved Monte Carlo Criticality Program ORNL-4938, November 1975.
5.
M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, B AW-1484-7, July 1979.
l 6.
J. T. Thomas, Critical Three-Dimensional Arrays of U(93.2) Metal Cylinders, j
Nuclear Science and Engineering, Volume 52, pages 350-359,1973.
l 7.
A. J. Harris, A Description of the Nuclear Design and Analysis Programs for l
Bolling Water Reactors, WCAP-10106, June 1982.
8.
Askew, J. R., Fayers, F. J.,
and Kemshell, P. B., A General Description of the Lattice Code WIMS, Journal of British Nuclear Energy Society, 5, pp.
l 564-584, 1966.
9.
England, T.
R., CINDER - A One-Point Depletion and Fission Product l
Program, WAPD-TM-334, August 1962.
i
\\
- 10. Melehan, J.
B., yankee Core Evaluation Program Final
- Report, WCAP-3017-6094, January 1971.
l l
Bibliography 31 l