ML20207S106
| ML20207S106 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/31/1986 |
| From: | Boyd W, Fecteau M, Mueller D DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20207R864 | List: |
| References | |
| NUDOCS 8703180534 | |
| Download: ML20207S106 (47) | |
Text
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4-CRITICALITY ANALYSIS OF BEAVER VALLEY 1 FRESH AND SPENT FUEL RACKS August 1986 W. A. Boyd M. W. Facteau D. E. Mueller APPROVED: . _ ,
D. J. Ull,lan'ager CE Core Design B 8703180534 870309 PDR ADOCK 050003?4 P pon .
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TABLE OF CONTENTS Section %
1 Introduction 1 2 Design Description 1 3 Design Criteria 2 4 Criticality Analytical Method 2 5 Criticality Analysis -
SFR Region 1 _Two of Four Storage 3 6 Criticality Analysis - SFR Region 1 - Three 5 of Four Storage 7 Postulated Accidents - SFR Region 1 6 8 Sensitivity Analysis - SFR Region 1 7 9 ' Criticality Analysis - SFR Region 2 - Spent Fuel Storage 7 9.1 Reactivity Equivalencing 7 9.2 Analytical Methods 8 9.3 Reactivity Calculations 9
- SFR Region 2 - Spent Fuel Storage 10 Postulated Accidents - SFR Region 2 11 11 Sensitivity Analysis - SFR Region 2 12 ume-unn $
TABLE OF CONTENTS (Continued)
Section h 12 Criticality Anal: isis - Fresh Fuel Racks 12 12.1 Full De:isity Moderation Analysis 13 12.2 Low Density Optimum Moderation Analysis 15 13 Sensitivity Analysis - Fresh Fuel Racks 16 14 Acceptance Criteria 16 Reftrinces 18 i
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- 1. Introduction The Beaver Valley Unit 1 spent fuel rack (SFR) design described herein employs one array of racks, which will be considered as two separate spent fuel
- racks. Both of these fuel racks consists of existing Duquesne Light fuel racks. The smaller array, referred to as Region 1 will be reanalyzed for criticality to show that 4.5 w/o fuel can be stored in the rack in two out of four storage locations and 4.0 STD w/o fuel in three of four trage -
locations. The larger array, Region 2, will be reanalyzed t i.ake into consideration the changes in fuel and fission product inventory resulting from depletion in the reactor core. The Region l' and 2 spent fuel rack design is a non poisoned stainless steel rack design, previously accepted by the NRC, for enrichments up to 3.3 w/o with no credit taken for the reactivity reduction due to fuel burnup for Westinghouse 17x17 STD fuel.
, The Region 2 spent fuel rack reanalysis is based on maintaining K,ff 0.95 for storage of Westinghouse 17x17 0FA and STD fuel at 4.5 w/o U with an initial enrichment /burnup combination in the acceptable area of Figure 1, and utilization of every cell permitted for storage of the fuel assemblies.
The Beaver Valley Unit 1 fresh fuel racks also consists of existing Duquesne Light fuel racks. These racks will be reanalyzed for criticality to show that 4.5 w/o 0FA and STD fuel can be stored in every storage cell in the rack and maintain k,ff 5 0.95. The fresh fuel rack design is a non poisoned I stainless steel design, previously accepted by the NRC for enrichments up to 3.3 w/o for Westinghouse 17x17 STD fuel.
- 2. Design Description The Region 1 and 2 spent fuel storage cell design is depicted schematically in Figure 2, with nominal dimensions given on the figure. The fresh fuel rack storage cell design is depicted schematically in Figure 3. The fresh fuel rack layout is shown in Figure 4.
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-,,w,-- e yu-n,-v.- -------~7n--- -r--m-,- , - - - --,,-,w,,,,-v- -- . . . - - +---- . - - - - - - - - - - - - - -r -..- - -- e, -.,- m -,-- - +-+--
f- .
- 3. 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,ff) of the fuel assembly array will be less than 0.95 as recommended in ANSI 57.2-1983, ANSI 57.3-1983 and in Reference 1.
- 4. Criticality Analytical Method i
The criticality calculation method and cross-section values are verified by comparison with critical experiment data for assemblies similar to those for which the rejks 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.
The design method which insures the criticality safety of fuel assemblies in the spent fuel storage rack uses the AMPX system of codes (2,3) for cross-section generation and KEN 0 IV(4) for reactivity determination.
i The 227 energy group cross-section library (2) that is the common starting point for all cross-sections used for the benchmarks and the storage rack is generated from ENDF/B-V data. The NITAWL program (3) 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 program (3) which is a one-dimensional S transport theory code. These n
multigroup cross-section sets are then used as input to KEN 0 IV(5) which is a three dimensional Monte Carlo theory program designed for reactivity calculations.
u m e-u nos 2
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 (boroflex, steel, water, etc) that simulate LWR fuel shipping and storage conditions (5) to dry, harder 4 spectrum uranium metal cylinder arrays with various interspersed i materials (0) (Plexiglas and air) that demonstrate the wide range of
- applicability of the method. Table 1 summarizes these experiments.
The average K,ff of the benchmarks is 0.992. The standard deviation of the bias 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.
- 5. Criticality Analysis - Spent fuel Rack Region 1 - Two of Four Storage J
The following assumptions were used to develop the nominal case KENO model for the spent fuel rack Region 1 storage of fresh fuel using two out of four storage locations:
- a. The fuel assembly contains the highest enrichment authorized, is at 4
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 Region 1 racks analyzed herein have t
shown that the W 17X17 0FA fuel assembly yields a larger K,ff than does the W 17X17 Standard fuel assembly when both fuel assemblies 235 enrichment. Thus, only the W 17X17 0FA fuel i have the same U assembly was analyzed for Region 1. (See Table 2 for fuel parameters).
- b. All fuel rods contain uranium dioxide at an enrichment of 4.5 w/o 235 over the infinite length of each rod, 4
U wru-uno5 3 t
.T' -..,m---,m_m-v._,.,3 . , , , , , , , .
. , , , , ,, , . . . 3 , ,wo._r ,% y- - ~- _ , . .-. , . . , . , . , _%._,. ,e_ - ,_ _ . _ . . ,,w-_
i
- c. No credit is taken for any U 234 or U236 in the fuel, nor is any ,
credit taken for the buildup of fission product poison material.
- d. The moderator is pure water at a temperature of 68'F. A conservative 3
value of 1.0 gm/cm is used for the density of water.
- e. No credit is taken for any spacer grids or spacer sleeves.
- f. Fuel assemblies are loaded into two of every four cells in a checkerboard pattern in the storage cells as shown in Figure 5.
- g. 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,ff of 0.8889 with a 95 percent probability /95 percent confidence level uncertainty of +0.0054.
The maximum K,ff under normal conditions arises from consideration of mechanical 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 the storage cells has shown that symmetrically placed fuel assemblies yield conservative results in rack K,ff. The manufacturing 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 1 storage racks, the water gap is
, reduced from a nominal value of 2.62" to a minimum of 2.59". Thus, the most conservative, or " worst case", KEN 0 model of the Region 1 storage racks contains a minimum water gap of 2.59" with symmetrically placed fuel assemblies.
l Based on the analysis described above, the following equation is used to
, develop the maximum K,ff for the Beaver Valley Region 1 spent fuel storage racks with two out of four storage:
mn e-unn 4
K,ff = Kworst + Bmethod + [(ks)2 worst + (ks)2 method'l Where:
K worst = w rst case KENO K,ff that includes material tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal B = method bias determined from benchmark critical comparisons method ks worst = 95/95 uncertainty in the worst case KENO K,ff ksmethod = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
K,ff = 0.8902 + 0.0083 + [(0.0061)2 + (0.0018)2)1/2 = 0.9049 Since K,ff is less than 0.95 including uncertainties at a 95/95 probability /
confidence level, the acceptance criteria for criticality is met.
- 6. Criticality Analysis - Spent Fuel Rack - Region 1 - Three of Four Storage The same assumptions that were used to develop the nominal case KEN 0 model for two out of four storage were use to develop the nominal model for three out of four storage except, only Westinghouse 17x17 STD fuel at 4.0 w/o was considered and the model is finite in the axial extent. Figure 6 shows a diagram of fuel assemblies loaded into three of every four storage cells.
The KEN 0 calculations for the nominal case resulted in a k,ff of 0.9348 (model is infinite in axial extent) with a 95 percent probability / 95 percent confidence level uncertainty of +0.0061.
l m "-"" " 5
The maximum K,ff under normal conditions was determined with the same consideration of the mechanical and material tolerances that was used in the two out of four portion of the analysis discussed in Section 5. Based on this discussion the same equation is used to develop the maximum K,ff for the Beaver Valley Region 1 spent fuel storage racks with three out of four storage.
Substituting calculated values in order into the equation in Section 5, the result is:
K,ff = 0.9329 + 0.0083 + [(0.0051)2 + (0.0018)2)1/2 = 0.9466 Since K,ff is less than 0.95 including uncertainties at a 95/95 probability /
confidence level, the acceptance criteria for criticality is met.
- 7. Postulated Accidents - Spent Fuel Rack Region 1 Most accident conditions will not result in an increase in K,ff of the rack. Examples 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 eight 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.
Examples are, not maintaining the proper checker board (2 of 4, or 3 of 4) loading when fuel is placed in the racks, 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 accide'it. 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.
som s-erous s
1
. i The presence of approximately 1000 ppm boron in the pool water will decrease reactivity by about 15 percent AK. Thus, for postulated accidents, should ,
there be a reactivity increase, K,ff would be less than or equal to 0.95 due to the effect of the dissolved boron.
- 8. Sensitivity Analyses - Spent Fuel Rack Region 1 To show the dependence of K,ff on fuel and storage cells parameters as requested by the NRC, the variation of the K,ff with respect to the following parameters was developed using the KENO computer code:
- 1. Fuel enrichment
- 2. Stainless steel thickness
- 3. Center-to center spacing of storage cells.
Results of the sensitivity analysis for the Region 1 storage cells are shown in Figures 7 through 9 for two of four storage and Figures 10 through 12 for three of four storage. All error bars shown on the figures are one sigma uncertainties.
- 9. Criticality Analysis - Spent Fuel Rack Region 2 - Spent Fuel Storage This section develops and describes the analytical techniques and models employed to perform the criticality analyses for storage of spent fuel in Region 2 of the Beaver Valley Unit 1 spent fuel pool.
9.1 Reactivity Eouivalencing Spent fuel storage, in the Region 2 spent fuel storage racks, is achievable by n,eans 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 generate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield 4
the equivalent K,ff when the fuel is stored in the Region 2 racks.
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, -,,,,e , , - - - - - - - - -,,---w---
Figure 1 shows the constant K,ff contour generated for the Beaver Valley Region 2 racks. Note the endpoint at 0 MWD /MTU where the enrichment is 3.10 w/o and at 9,700 MWD /NTU where the enrichment is 4.5 w/o. The interpretation of the endpoint data is as follows: the reactivity of the Region 2 racks containing fuel at 9,700 MWD /MTU burnup which had an initial enrichment of 4.5 w/o is equivalent to the reactivity of the Region 2 racks containing fresh fuel having an initial enrichment of 3.10 w/o. It is important to recognize that the curve in Figure 1 is based on a constant Region 2 rack reactivity and not on a constant fuel assembly reactivity. The data in Figure 1 is also provided as Table 3. Linear interpolation between two data points on this table will yield conservative results.
9.2 Analytical Methods The data points on the reactivity equivalence curve were generated with a transport theory computer code, PHOENIX U) . 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(8) 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 I9) . CINJER is a point-depletion computer code used to determine fission product activities. The fission products were permitted to decay for 30 years after 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 135 , has nearly completely decayed away.
Furthermore, 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 approximated by removing the Xe 135 ,
um e-mus g
l The PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been examined following discharge from a ,
reactor. In addition, an extensive set of benchmark critical experiments has been analyzed with PHOENIX. Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 4. The
. measurements were made on fuel discharged from Yankee Core 5(10) . The data 4 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 81 critical benchmark experiments is summarized in Table 5. Key parameters describing each of the 81 experiments are given in Table 6. These reactivity comparisons again show good agreement between experiment and PHOENIX calculations.
! An uncertainty associatec with the burnup-dependent reactivities computed with
, PHOENIX is accounted for in the development of the maximum Regioh' 2 l multiplication factor. An uncertainty of 0.01 Ak is considered to be very
! conservative since comparison between PHOENIX results and the Yankee Core experiments and 81 benchmark experiments indicates closer agreement.
9.3 Reactivity Calculations - Spent Fuel Rack Region 2 - Spent Fuel The nominal and maximum K,ff for storage of spent fuel in Region 2 is determined using the methods described in Section 4 for Region 1 in addition l
to the methods described in Section 9.2. The actual conditions for this determination are defined by the zero burnup intercept point in Figure 1. The KENO-IV computer code is used to calculated the storage rack multiplication factor with an equivalent fresh fuel enrichment of 3.10 w/o. Combinations of fuel enrichment and discharge burnup yielding the same rack multiplication i factor as at the zero burnup intercept are determined with PHOENIX.
The following assumptions were used to develop the nominal case KENO model for j the Region 2 storage of spent fuel:
1 6JJ1L 6-870228 g
g3
.. ., A .
, , +. g -
s
, a. Historically, calculations for spent fuel racks similar to the Region 2 racks analyzed herein have shown that the Westinghouse
- tl- 17x17 0FA fuel assembly yields a larger K,ff than does the y Westinghouse 17x17 standard fuel assembly when both fuel 235 assemblies have the 'same U enrichment. Thus, only the h Westinghouse 17x17 0FA fuel assembly was analyzed for Region 2.
,~
9,
'i b. The Westinghouse 17x17 0FA spent fuel assembly contains uranium dioxide fuel at an equivalent " fresh fuel" enrichment of 3.10 235 w/o U ,
i
.c. h The moderator is purt. water at a temperature of 68*F. A 3
\ ( conservative value of 1.0 gm/cm is used for the density of b y water.
r; ,
4
- d. No credit is taken for any spacer grids or spacer sleeves.
- e. The array is infinite in lateral and axial extent which precludes any neutron leakage from the array.
l The KEN 0 calculation for the nominal case resulted in a K,ff of 0.9245 with a 95 percent probability /95 percent confidence level uncertainty of + 0.0054.
The maximum K,ff under normal conditions was determined with a " worst case" KENO model, in the same manner as for the Region 1 storage racks (see Section 5). An uncertainty associated with the reactivity equivalence methodology was
- , considered in the development of the maximum K,ff. This uncertainty was
- discussed in Section 9.
Based on the analysis described above, the following equation is used to develop the maximum K,ff for the storage of spent fuel in the Beaver Valley b Region 2 spent fuel storage racks:
's f
l um.-mu. 10 i
1 K,ff = Kworst + 0 method + [(ksworst) + (ksmethod) + (ksre) I l 1
l where: l
=
K worst w rst case KENO K,ff that includes centered fuel assembly position, material tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal B = method bias determined from benchmark critical comparisons method 1
=
ks worst 95/93 uncertainty in the worst case KENO K,ff ksmethod = 95/95 uncertainty in the method bias ks ' = uncertainty in the reactivity equivalence methodology re Substituting calculated values in the order listed above, the result is:
K,ff = 0.9262 + 0.0083 + [(0.0059)2 + (0.0018)2 + (0.01)2 31/2 = 0.9462 The maximum K,ff for Region 2 for this configuration is less than 0.95, including all uncertainties at a 95/95 probability / confidence level.
Therefore, the acceptance criteria for criticality are met for storage of 235 spent fuel at an equivalent " fresh fuel" enrichment of 3.10 w/o U ,
- 10. Postulated Accidents - Region 2 Most accident conditions will not result in an increase in K,ff of the rack. Examples 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 eight inches of water separating it from the active fuel height of stored assemblies which precludes interaction).
t p' q wm .-mm 11
9
~
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 l 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 l condition since not assuming its presence would be a second unlikely event. I The presence of approximately 1000 ppm boron in the pool water will decrease reactivity by about 15 percent AK. 1.1us, for postulated accidents, should !
there be a reactivity increase, K,ff would be less than or equal to 0.95 due ;
to the effect of the dissolved boron. l
- 11. Sensitivity Analysis - Region 2 To show the dependence of K,ff on fuel and storage cell parameters as requested 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 13 through 15 illustrate the results of the sensitivity studies for spent fuel (3.10 w/o U 235 equivalent " fresh fuel" enrichment) occupying every cell in the Region 2 racks.
- 12. Criticality Analysis - Fresh Fuel Racks This section describes the analytical techniques and models employed to perform the criticality analysis for storage of fresh fuel in the Beaver Valley Unit 1 fresh fuel racks.
l l
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12
~
Since the fresh fuel racks are maintained -in a dry condition, the criticality analysis will show that the rack K,ff is less than 0.95 for the full density and low density optimum moderation conditions. The low density optimum moderation scenario is an accident situation in which no credit can be taken for soluable boron. The criticality method and cross-section library are the same as those discussed in Section 4 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:
- a. The fuel assembly contains the highest entichment 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 similar racks have shown that the E 17x17 0FA fuel assembly yields a larger K,ff thandoestheE17x17Standardfuelassemblywhenboth fuel assemblies have the same U 235 enrichment. Thus,onlythey 17x17 0FA fuel assembly was analyzed. (See Table 2 for fuel parameters).
- b. All fuel rods contain uranium dioxide at an enrichment of 4.5 w/o 235 U over the infinite length of each rod.
- c. No credit is taken for any U 234 or U 236 in the fuel, nor is any credit taken for the buildup of fission product poison material,
- d. No credit is taken for any spacer grids or spacer sleeves.
i 12.1 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 3
gm/cm 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.
me-mm 13
The KENO calculation for the nominal case resulted in a K,ff of 0.8959 with a 95 percent probability /95 percent confidence level uncertainty of +0.0079.
The maximum K,ff under normal conditions arises from consideration of mechanical 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 the storage cells has shown that symmetrically placed fuel assemblies yield conservative results in rack K,ff. The manufacturing 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 fresh fuel storage racks, the water gap is reduced from a nominal value of 11.75" to a minimum of 11.72". Thus, the most conservative, or " worst case", KENO model of the fresh fuel storage racks contains a minimum water gap of 11.72" with symmetrically placed fuel assemblies.
Based on the analysis described above, the following equation is used to develop the maximum K,ff for the Beaver Valley fresh fuel storage racks:
K,ff = Kworst
- 0 method + [(ks)2 ,g73 + (ks)2method Where:
=
K rst w rst case KENO K,ff that includes material tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal B
method
method bias determined from benchmark critical comparisons ks
95/95 uncertainty in the worst case KENO K,ff worst ks method
= 95/95 uncertainty in the method bias som e-nous 14
Substituting calculated values in the order listed above, the result is:
K,ff = 0.9009 + 0.0083 + [(0.0066)2 + (0.0018)21/2 3 = 0.9160 Since K,ff is less than 0.95 including uncertainties at a 95/95 probability /
confidence level, the acceptance criteria for criticality is met.
12.2 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.
Analysis of fresh fuel racks similar to the Beaver Valley Unit 1 racks has shown that the maximum rack K,ff under low density moderation conditions 3
occurs at 0.045 gm/cm water density. The KENO calculation of the Beaver 3
Valley Unit 1 fresh racks at 0.045 gm/cm water density resulted'in a peak K,ff of 0.9194 with a 95 percent probability and 95 percent confidence level uncertainty of +0.0129.
The nominal cell center-to-center spacing, rack module spacing and material tolerances have been included in the base case model and result in a storage cell separation distance of 11.75" and a rack module separation distance of 20.75 inches. Studies of asymme.ric positioning of fuel assemblies within the storage cells has shown that symmetrically placed fuel assemblies yield conservative results in rack K,ff.
Since the Beaver Valley rack tolerances ar'e small (0.06") consideration of minimum cell center-to-center spacing, rack module spacing and material tolerances will have an insignificant effect on the fuel rack k,ff. As a result the maximum k,ff will be 0.9194 with a 95 percent probability and 95 percent confidence level uncertainty of + 0.0129.
Based on the analysis described above, the following equation is used to 1 develop the maximum K,ff for the Beaver Valley fresh fuel storage racks under low density optimum moderation conditions:
sou s-ensse 15
K,ff = Kbase + Beethod + [(ks)2 ,,, + (ks)2 method l where:
=
K base base case KENO K,ff that includes nominal mechanical and material dimensions B
method
method bias determined from benchmark critical comparisons ks
base 95/95 uncertainty in the base case KENO K,ff j ks = 95/95 uncertainty in the method bias method Substituting calculated values in the order listed above, the result is:
K,ff = 0.9194 + 0.0083 + [(0.0129)2 + (0.0018)2)1/2 = 0.9407 Since K,ff is less than 0.95 including uncertainties at a 95/95 probability /
confidence level, the acceptance criteria for criticality is met.
- 13. Sensitivity Analysis - Fresh Fuel Racks To show the dependence of K,ff on fuel and storage parameters as requested 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 16 through 18 illustrate the results of the sensitivity studies for fresh fuel occupying every cell in the fresh fuel racks.
uva- m" 16
- 14. Acceptance Criterion For Criticality The neutros multiplication factor in spent 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, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor i 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," ANSI 57.3-1983, " Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants."
1 * ,
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sem:s-arous 17
References
- 1. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes, April 14, 1978, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications."
- 2. W. E. Ford III, et al . , "CSRL-V: Processed ENDF/B-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies," ORNL/CSD/TM-160 (June 1982).
- 3. N. M. Greene, et al., "AMPX: A Modular Code System for Generatino Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," ORNL/TM-3706 (March 1976).
- 4. L. M. Petrie and N. F. Cross, " KENO IV--An Improved Monte Carlo Criticality Program," 0RNL-4938 (November 1975).
- 5. M. N. Baldwin, et al., " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel," BAW-1484-7, (July 1979).
- 6. J. T. Thomas, " Critical Three-Dimensional Arrays of U (93.2) -- Metal Cylinders," Nuclear Science and Engineering, Volume 52, pages 350-359 (1973).
- 7. A. J. Harris, et al., "A Description of the Nuclear Design and Analysis Programs for Boiling 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. 564-584 (1966).
r l
"""-""" 18
s
- 9. England, T. R., " CINDER - A One-Point Depletion and. Fission Product Program," WAPD-TM-334, August 1962.
i l 10. Melehan, J. B., " Yankee Core Evaluation Program Final Report,"
WCAP-3017-6094, January, 1971.
Y u n a - m iios ig
e
- Table 1
- Benchmark Critical Experiments [5.6]
General Enrichment Separating Soluble J B-10 ppm K Description w/o U235 Reflector Material fp
- 1. UO rod lattice 2.46 water water O O.9857 + .0028
- 2. UO rod lattice 2.46 water water 1037 0.9906 7 .0018
- 3. UO rod lattice 2.46 water water 764 0.9896 7 .0015
- 4. UO rod lattice 2.46 water B4C pins O O.9914 7 .0025
- 5. UO rod lattice 2.46 water B4C pins O O.9891 7. 0026
- 6. UO rod lattice 2.46 water B4C pins O O.9955 7 .0020
- 7. UO rod lattice 2.46 water B4C pins O O.9889 7 .0026
- 8. UO rod lattice 2.46 water B4C pins O O.9983 7 .0025
- 9. UO rod lattice 2.46 water water O O.9931 7 .0028
- 10. UO rod lattice 2.46 water water 143 0.9928 7 .0025
- 11. UO rod lattice 2.46 water stainless steel 514 0.9967 7 .0020 l
- 12. UO rod lattice 2.46 water stainless steel 217 0.9943 7 .0019 l 13. UO rod lattice 2.46 water borated aluminum 15 0.9892 7 .0023
- 14. UO rod lattice 2.46 water borated aluminum 92 0.9884 7 .0023
( 15. UO rod lattice 2.46 water borated aluminum 395 0.9832 7 .0021 i l
- 16. UO rod lattice 2.46 water borated aluminum 121 0.9848 7 .0024 l l
- 17. UO rod lattice 2.46 water borated aluminum 487 0.9895 7 .0020
- 18. UO rod lattice 2.46 water borated aluminum 197 0.9885 7 .0022
- 19. UO rod lattice 2.46 water borated aluminum 634 0.9921 7 .0019 .
l
- 20. UO rod lattice 2.46 water borated aluminum 320 0.9920 7 .0020
- 21. UO rod lattice 2.46 water borated aluminum 72 0.9939 7 .0020
- 22. U tal cylinders 93.2 bare air O O.9905 7 .0020 l
- 23. U metal cylinders 93.2 bare air O O.9976 7 .0020
- 24. U metal cylinders 93.2 bare air O O.9947 7 .0025
- 25. O metal cylinders 93.2 bare air O O.9928 7 .0019
- 26. U metal cylinders 93.2 bare air O O.9922 7 .0026
- 27. U metal cylinders 93.2 bare air O O.9950 7 .0027
- 28. U metal cylinders 93.2 bare plexiglass O O.9941 7 .0030
- 29. U metal cylinders 93.2 paraffin plexiglass. O O.9928 7 .0041
- 30. U metal cylinders 93.2 bare plexiglass O O.9968 7 .0018
- 31. U metal cylinders 93.2 paraffin plexiglass 0 1.0042 T .0019
- 32. U metal cylinders 93.2 paraffin plexiglass O O.9963 7 .0030
- 33. U metal cylinders 93.2 paraffin plexiglass O O.9919 7_ .0032 5417L.6-861105
- 1 Table 2 Fuel Parameters Employed in Criticality Analysis Parameter g17X170FA -H17X17 Standard-Number of Fuel Rods 264 264 per Assembly Rod Zirc-4 Clad 0.D. (inch) 0.360 0.374 Clad Thickness (inch) 0.0225 0.0225 Fuel Pellet 0.D. (inch) 0.3088 0.3225 Fuel Pellet Density 96 96
(% of Theoretical)
Fuel Pellet Dishing Factor 1.0 1.0 Rod Pitch (inch) 0.496 0.496 Number of Zirc-4 Guide Tubes 24 24 Guide Tube 0.D. (inch) 0.474 0.482 Guide Tube Thickness (inch) 0.016 0.016
, Number of Instrument Tubes 1 1 Instrument Tube 0.D. (inch) 0.474 0.482 Instrument Tube Thickness 0.016 0.016 (inch)
I mn+""" 21
I Table 3 BEAVER VALLEY FUEL ASSEMBLY MINIMUM BURNUP VS. INITIAL U 235 ENRICHMENT FOR STORAGE IN REGION 2 SPENT FUEL RACKS Initial U 235 Assembly Discharge Enrichment Burnup (GWD/MTU) 3.1 0 3.3 1.6 3.5 3.0 4.4 3.7 3.9 5.8 4.1 7.2 4.3 8.5 4.5 9.7 Note: Linear interpolation yields conservative results.
22
~
Table 4 Comparison of PHOENIX Isotopic Prediction 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 Percent difference is average difference of ten comparisons for each isotope.
)
unes-eenos 23-
Table 5 Benchmark Critical Experiments PHOENIX Comparison Descripti'on of Number of PHOENIX k,ff Using Experiments Experiments Experiment Bucklings UO 2
Al clad 14 .9947 SS clad 19 .9944
' Borated H 2O 7 .9940 Subtotal 40 .9944 i
U-Metal Al clad 41 1.0012 L
TOTAL 81 .9978 f
f m ne-""" 24
Table 6
\
l Data for U Metal and UO2 Critical Experiments Fuel Pellet Clad Clad Lattice
- Casa Cell A/O H20/U Density Diameter Material OD Thickness Pitch B-10 Number Type U-235 Ratio (G/CC) (CM) Clad (CM) (CM) (CM) PPM 1 Hexa 1.328 3.02 7.53 1.5265 Aluminum 1.6916 .07110 2.2050 0.0 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 1.6916 .07110 2.5120 0.0 )
4 4 Hexa 1.328 3.92 7.52 .9855 Aluminum 1.1506 .07110 1.5580 0.0 1 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 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 Square 2.734 2.18 10.18 .7620 SS-304 .8594 .04085 1.0287 0.0 ,
-10 Square 2.734 2.92 10.18 .7620 SS-304 .8594 .04085 1.1049 0.0 11 Square 2.734 3.86 10.18 .7620 SS-304 .8594 .04085 1.1938 0.0
. 12 Square 2.734 7.02 10.18 .7620 SS-304 .8594 .04085 1.4554 0.0 2
13 Square 2.734 8.49 10.18 .7620 SS-304 .8594 .04085 1.5621 0.0 14 . Square 2.734 10.38 10.18 .7620 SS-304 .8594 .04085 1.6891 0.0 i .15 Square 2.734 2.50 10.18 .7620 SS-304 .8594 .04085 1.0617 0.0 16 Square 2.734 4.51 10.18 .7620 SS-304 .8594 .04085 1.2522 0.0 17 Square 3.745 2.50 10.27 .7544 SS-304 .8600 .04060 1.0617 0.0 18 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 0.0 19 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 0.0 20 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 456.0 t 21 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 709.0 22 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1260.0
, 23 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1334.0 L 24 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1477.0 25 Square 4.069 2.55 9.46 1.1278 SS-304 1.2090 .04060 1.5113 0.0 26 Square 4.069 2.55 9.46 1.1278 SS-304 1.2090 _.04060 1.5113 3392.0 27 Square 4.069 2.14 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 1.1268 SS-304 1.1701 .07163 1.5550 0.0 30- Square 3.037 8.16 9.28 1.1268 SS-304 1.2701 .07163 2.1980 0.0 31 Square 4.069 2.59 9.45 1.1268 SS-304- 1.2701 .07163 1.5550 0.0 32 Square 4.069 3.53 9.45 1.1268 SS-304 1.2701 .07163 1.6840 0.0 33 Square 4.069 8.02 9.45 1.1268 SS-304 1.2701 .07163 2.1980 0.0 34- Square 4.069 9.90 9.45 1.1268 SS-304 1.2701 .07163 2.3810 0.0 35 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 1677.0 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 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 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
"""-""" 25 v -- ~ . - - . c. - .. . - - . - - . . _
l Table 6 (continued)
Data for U Metal and UO2 Critical Experiments l Fuel Pellet Clad Clad Lattice Casa Cell A/0 H20/U Density Diameter Material OD Thickness Pitch B-10 Number Type U-235 Ratio (G/CC) (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 l 46 Hexa 1.160 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 l 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 1.6916 .07112 2.6162 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 .07112 2.6162 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 Aluminum 1.1506 .07112 1.5926 0.0 58 Hexa 1.307 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 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 61 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 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 .07112 1.9609 0.0 64 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 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 .07112 1.5926 0.0 67 Hexa 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 1.9609 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 Hexa 1.040 1.58 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 l 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 l 2.9900 0.0 77 Hexa 1.310 3.01 18.88 1.5240 Aluminum 1.6916 .07112 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 .07112 2.9900 0.0 i 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 .07112 1.9610 0.0 l
ses-""" 26
i l
- l .
l . .
.t 12
- , , -,= , ., -t 10
)
. . . . . . . . . . /!
s kB TE l M STN E l /:
/ l b
./ ., ,
g W6 , ,
5 .
h i 3
4
/
/ .
2
, NOT ACEPTELE FDR STNAGE i
. . . . . . 1 ,
FIGURE 1 .
BEAVER VALLEY FUEL ASSEMBLY MINIMUM BURNUP VS. INITIAL U-235 ENRICHMENT FOR STORAGE IN REGION 2 SPENT FUEL RACKS som
27
FIGURE 2 BEAVER VALLEY REGION 1 & 2 SPENT FUEL STORAGE CELL NOMINAL DIMENSIONS
\
, 2.62" .
c 8.94"
~
! ! 9
- ~
- 1 1 1 1 4 >
CELL CENTER TO CENTER ( 12.06" )
eem.- o i. 28
.e- . .
l l
FIGURE 3 BEAVER VALLEY FRESH FUEL STORAGE CELL NOMINAL DIMENSIONS
~ ~ ~
, . . , , \
h .
, 11.75" . '
c 8.94" :
1 o ,
4 CELL CENTER TO CENTER ( 21.0" ) >
um- o
29 I
_j
FIGURE 4 BEAVER VALLEY FRESH FUEL RACK LAYOUT
=
m...es.mos 3.13t'3NQ3_ L "m aa.o;C i
h lI
'A i i .
.i l
Tsi' i g L -- c-- - c.----J-d--l ;b 2 i i i l-E-i::
-g D-[-E
- . :i .c? 9gg-
.!"l, I EE :M_ Efh ,d,i f ;
g-M i ;
I l
+ M-E,4 ;,
/
o i j
W '
i!
- * ~, * *
$lIl m ,- --
! . 7.%. , qgJ*a 4 s a
i i l [ 7ji l7:
% l 1 s- i i ,
__. . . .: d~ a. d a i
i i i
i a eJ4 l
. . , .g i 2 i
i _. . . .
i ,
~
zd l l l k
I I I I n- __. . . . a.
I J_ _L .
l _
l
-@-E-h g ..fi
.i ,
l -Q-g-EtB E t,g,a I
________2___
-a-amm a f.: i w._;. a .
a, n . : ,
! i ! I i 2,. .s =
.. .=.=...== d ,-lc' , .
so n ..
30
_-__________-_______.___ _=____=__________ . - - _ _ _ - _ _
.-n i .
FIGURE 5 BEAVER VALLEY REGION 1 CHECKERBOARD FUEL ASSEMBLY LOADING SCHEMATIC
)
1 l
E g.
D um.-no i. 31
e D
e FIGURE 6 BEAVER VALLEY REGION ONE 3 0F 4 FUEL ASSEMBLY LOADING SCHEMATIC l
um - o.ie 32
- +
4
.92 *
...i . . - - s .. . ...i.... . 4. .. +.. .f .
..i ...
..} . . ....i.... . .i...
..<.... ...+... ...y .
4
.91 '
.. .f .. ...... ..L 4... . . . . . . .
. . . . . . . -
- r
- 4) !
.9
....s..... . 4. ....i.... .. . . . .e.. - ..d. . -~4.--i.
.. .. .. i. - . . i .... . . . . . .
- . =-; *
.99 6 0 b ... .
. . a_ - . . . ....... .
. .. 4 . ...f... . . . .
Q . . . . . . . . . .
x .
.88 .
.+.. -- -
....... ........ ... . . . .... . 7..
.57 :. :. :
g NOTE ENTER-TO-CENTER SPACIfG = 12.06 (IPOES) l---i. .
...,.... ...a.. ..-4.... .,.
.2S INCKS l STAINLESS STEEi THICKNESS o . .
i
.06 : :
.. 4 . . ..
... f..p.
. . . . . . .. . . . . . . . . . . . . . . . . . . . . .. . . . .p. .
4.4 4.5 4.6 4.7 4.8 4.9 S. 5.1 5.2
.s3.8 3.9 4. 4.1 4.2 4.3 U-235 Deti0404i (W/0) r i
1 t
I 1
i FIGURE 7
! SENSITIVITY OF K-EFF TO ENRICHMENT IN THE BEAVER VALLEY REGION 1 SPENT FUEL STORAGE RACKS WITH TWO OF FOUR STORAGE I
I i
i I
5031L:6-880810 f
s.
~
.92
. . , . --i y .......j p ..
9
= .i
.91 i
k -4....... ......, 4 ,
y .. 4
.9
(>
B .
i
.89 (pi 4 >
. . . .. ...m-- .
NOTE ENRIC N T = 4.5 W/0 l 4
.88 STAlrLESS STEEL THICKESS = .25 INCES l !
..... .... , --4 , ., ... .. .
.n[1.2 11.4 11.6 11.8 12. 12.2 12.4 12.s 12.s MNTER-TO-ENTER SPACING (INCIES) l FIGURE 8 SENSITIVITY OF K-EFF TO CENTER-TO-CENTER SPACING IN THE BEAVER VALLEY REGION 1 SPENT FUEL STORAGE RACKS WITH TWO OF FOUR STORAGE sem - as 34
. s
.92 i-- ; e- ,- ...
.....4........ --4
.91
.9 .
- -4 . . . . . . . . . -4 i - - - - t
.59
E :
.. ..I ,-- -i i ,
i o
.58 .
--a : . .
i NOTE EPRICFt1ENT = 4.50 WO l
.s7 CENTER-TO-CENTER SPACDC = 12.06 INCHES l i !
.2 .25 .3 .35 .4
- ".1 15 STAINLESS STEEL THICK!ESS (INCES)
FIGURE 9 SENSITIVITY OF K-EFF TO STEEL CAN THICKNESS IN THE BEAVER VALLEY REGION 1 SPENT FUEL STORAGE RACKS WITH TWO 0F FOUR STORAGE ums-no'* 35
e e
.M
- j : : :
......p..... ..)...... 4..- m.4, .m.
. . . ' > .. + . . . .
..g..... ...g..... 4-
... p .. ....q....-
- g i.
.97 .
! ! ! 4.
- 4. 4. . . . . . . . .. . . .
....... ....;..- ....:... . . . . . . . . .e ..
4....
- i 1
- : - :.
.96 l
l
- 1 :
.4.... .....
1 .. ...:. .. .....: - - 4..
. J.-,.
.........Tw.
6
~
? *
- * . (>
.95 : :* : : *
...4.... ....4 -- s. . ..4,.. ...4.... ..4..-. . . . . . , . .
4....
..4...
g
..4..... ~ g I '
.94 .
.l qp....'..... ...!..- .7.... . . . . . ....'....
[W.93 :*
M :.
.~.4.m. .-4.-
. . .6,~--
.. >.... ... ). ....+. . q
. . . .,m- .,_..g m ..--t.-. . .y-..-4...
, , , . g g
.92 .
- . 6
- 4. ....4.... .. 4.. ... .e -...4. . . . . . - . . . . . -
.4... ..4... ....a.... . . . .
l l : * :
l
.91 . : .
... . . . . . ,r ... .. ... ... .
- 7. . . . , .- ;-
g NOTE CENTER-TO-ENTER SPACING = 12.06 IINCHESI I
- 9 :
- t. i..
STAlblESS STEEL .THICKNESS
= .25 INCHES ., . . . .. . ..4'.
. . . . . . . . . . 4.
>- -.> ....+-- . . . : :
.se 4
.. ... --i.
...4,- --5,.. . . . > . . 4..4. ..
4.
.4-- --4,....
4.6 4.1 4.s 4.4 4.5
'85.4 s.5 s.s 3.7 s.s 3.9 4. 4.2 U-235 ENHQ4ENT (W/0)
FIGURE 10 SENSITIVITY OF K-EFF TO ENRICHMENT IN THE BEAVER VALLEY REGION 1 SPENT FUEL STORAGE RACKS WITH THREE OF FOUR STORAGE sosu-sees 36 e d - a
.98 .
- )
- : : l
... 4 -- - ..j - - -..i.......;.............;........
....2- - 4...... ...l ....
i 1
.97 : i i : . : ...... y .. .. ,._.
- 4
- 3. . .
....9...... . . . . . . . . . .gg ..t_ .
.
- qg
.96 .
l ,
I
== : - - , .-
i q.. --.y__ : 1
- e : :
- :- . - I
^
- N ,
- - - - - f.- -
.. .. 4_ _ .
I-
.94 r
- :- - -i .
- g- .
I hs : .
.93 * :
i ! .
.~
i . ; _. .... ......+.
.92 :
i -
. ....,-- -4.....
...... ._ ,__ -., i
!. !4I i .
.91 . -
. . , ..-- -,__ ,. -y- -
- r I
I I NOTE ENRID+1ENT = 4.0 LI/O I
,, i !.
STAlft.ESS STEEL THICKNESS = .2S INCES I...
. i -. .. . ~ .
......t...... , :.
12.2 12.4 12.6 12.8 11.6 11.8 12.
'Y1.2 11.4
~ CENTER-T(H: ENTER SPACING (INCES)
FIGURE 11 .
SENSITIVITY OF K-EFF TO CENTER-TO-CENTER SPACING IN THE BEAVER VALLEY RETION 1 SPENT FUEL STORAGE RACKS WITH THREE OF FOUR STORAGE S03:LS-060st0 37
- e e
e
.96 :
-+ .. ....q. 4 4 **********6
') : : : :
4
.97 : : :
.96 . . .
! A
- s. i.
I I : :.
.96 ,
t .
......4._ .-l......... -4 4...... 4 g
.94 .
g _ ,.
x ,
. : sim.=
6-
> -+ .
4 4,. ; ~~.6
. :. 1 >
- I2 .
1 i' .......4--
J. - -
- 4. - . & .
.91 :
' 2..... ..+. .-.:. .. .
NOTE EFRICNT = 4.00 WO l
- 9
- ENTER-TO-CENTER SPEDG = 12.06 ItOES l :
.. ... .. .! -- .. . 4. .. ;
! ~ ..
.59 .
-, 4 - -.i i y
. I
- 88
.1 15 .2 .25 .3 .35 .4 STAINLESS STEEL THICKESS (INCHES)
FIGURE 12 SENSITIVITY OF K-EFF TO STEEL CAN THICKNESS IN THE BEAVER VALLEY REGION 1 SPENT FUEL STORAGE RACKS WITH THREE OF FOUR STORAGE-
.em .. i.
38
m 6
-a 6 l% = rR ?
a .4 -
t 1
.97 .
..!- .......... . . .l...
' 4 :
j i. .
i .
.N .
, .... .t
. . w..
. 4.. ..... .... . . . .....
1 (p .
.95 :
i i :.
- 3
- .m.. ....t.. m..
l I.
~ . .
t : :
? : ?
.94 .
?
.93 . . .
=j l : :
.. . . . . . . - . . . . . . . . [.....=q p ... ,.
.92 Li : : :
g . .
.91 . . . 3 .
- : : p.m.
& i go . .. - - .
.....p... .... . . . . . . . . , . . - . . _ .. - . . , . . . . . , . .. --9. m.
. .-9..m. .
.9 l : : : :
- '.g . . . ? .
.. :-p ..4....4,... g. ..6.....
...y.... ..y... ...q..... ...q..... . .
.89 . .
....y..".." . . . . > . . ..4..
..q.....
..g__
. --6..*.. t.
..!.*. F. .4,-....
4,- .-.*4. ...6... .
) .
.as
.__ i NOTE CENTER-TO-CENTER SPACING = 12.06 IItOESI l i STAltLESS STEEL THICKtESS,_= .25 INCFES- l, . . i.
. !. ...,.. . . , . . .....,.. 4__ 4.
....4.....
. 4. .. . :. . .
.87 . .
,.. .. , .+ . . ,.. .
. 4.. .. ...4... 4.. . .
- b.S 2.6 2.7 2.8 2.9 3. 3.1 3.2 3.3 3.4 3.5 3.6 3.7 U-235 D5tlOMNT (W/0)
FIGURE 13 SENSITIVITY OF K-EFF TO ENRICHMENT IN THE BEAVER val.l.EY REGION 2 SPENT FUEL STORAGE RACKS so m .. oei. 3g
---_i__.___* __
e l .97 :
2 g -9<.
- g. g.
. . . . 9-- -.9 9 l : : : :.
t l
.96 ,
- q, :
- f. p.. .j. .. .
.......d..... .......j.... .f. .. 5 *
. .. 4 a e :
.95 :*
- ********* t********
......y-. 9'.. 9....- 9
-f 1
l l* :
y : :
.93
-g ..
x ! . !.
.92 .
- . ~. - - --:
+.... .
NOTE DRICHfNT = 5.1 W/0 l
.91 STAltLESS STEEL THICKNESS = .2S INCES
....2.....
- - -+ .. ..--.i i
- i i
- i i
.9 : :
-- - - 4..w -. . . ,
.....4.......
- i. o
.s9 _ .
4..-
- l
- 12. 12.2 12.4 12.6 12.8
'Y1.2 11.4 11.6 11.8 ENTER-TO-CENTER SPACING (INOES)
FIGURE 14 SENSITIVITY OF K-EFF TO CENTER-TO-CENTER SPACING IN THE BEAVER VALLEY REGION 2 SPENT FUEL STORAGE RACKS som.. ** 40 s . ---i-___...____u._____...__ _ _ _ _ _ _ .
- e e
o
.96 ,
5
- g. . 9 9
- p. ..
y
.95 f... . 4 y. . . _
.94 y
.y... * . q. .- . 9-
.93
. . .! - . -4...... i.. 3 ,......i , ,
t:-
E .92 -
x
--2-. -
9... g. .m...q -g
.......f . . . . . . . , 7 i
j --
.91 .
-q g.....___ 9 q.
..t__
.9
.e9 NOTE EWICNT = 5.10 WO I i
. i .
! . CENTER-TO-CENER SPFCING = 12.06 INCES l
.; .. u . . ,
.ns.1 .15 .2 .25 .3 .35 .4 STAINLESS STEEL THICKfESS (INDES)
FIGURE 15 I
SENSITIVITY OF K-EFF TO STEEL CAN THICKNESS IN THE BEAVER VALLEY REGION 2 SPENT FUEL STORAGE P.ACKS so m s. i.
41
n
- e s
e
(,
.93
...; ... . i.... , . . . . , . .. . , . ....! ... ....; _- -
- 1
.92
..I i. . . ......
.91 :
?
...h. ........
...p. .,
.....t.***- **t**** -**t.'-***' -******* --*h,*** * * * - > - * * -*****t****
...f.....
.9 .
E i i i l : :_ ,
. ,.... .. ....,.. .,.._ ....,-. . _ , . - .~,.-. ~ , . .~.,~.t.-
....,= . . , . .
.89 :
NOTE CENTER-TO-CENTER ! PACING = 21.0 IINCHES) l
- I : I .
I I I e I I I STAIN.ESS STEEL THICXPESS = .125 INCTS l t : : : : : i :
.86 q l . .
,-... . -- . 4.- .-.-. --; - _ J -., - . , - .
....,. ...,. ,. 4.- .
- b.8 3.9 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5. 5.1 S.2 U-235 DetlOMNT (W/0)
FIGURE 16 SENSITIVITY OF K-EFF TO ENRICHMENT IN THE BEAVER VALLEY FRESH FUEL STORAGE RACKS some ooe 42
t ,
n
.92
., . 4 4.
.91
--+ _.......9
-9. . ..
.9 4> ; 4>
X t l :
.89 J .
,, NOTE ENRIOfENT = 4.5 WO l .
SmINLISS STE. NICKNESS = .125 INCHES I I i 19 20 21 22 23 MNTER-TO-CENTER SPACIMI (if0ES)
FIGURE 17 SENSITIVITY OF K-EFF TO CENTER-TO-CENTER SPACING IN THE BEAVER VALLEY FRESH FUEL STORAGE RACKS 1
l 5031L1-440ste g
P-A .
n
.93 .
i l
...q..
. _ . 9.
.92
. - , _ . = - . .
.91
. .. . ...j . . .I. .. ,.......
- - - - -.-..--t-.-
I. .
in.
i D .9 u
4 > ,
qp
. .y .
--a__ 4. .
.89 ,
i i --
NOTE ENRIOfENT = 4.5 W/O I
.se . ,
CENTER-TO-CENTER SPACING = 21.0 ICES l 8 i i
. 1. . a ~ -.
, ~.a-.--- -~~..-t-
--f__
.15 .2 .25
'UO. .05 .1 STAINESS STEEL THlWIESS (INOES)
\
FIGURE 18 SENSITIVITY OF K-EFF TO STEEL CAN THICKNESS IN THE BEAVER VALLEY FRESH FUEL STORAGE RACKS som . oete y
.O
_