ML20153G258
| ML20153G258 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/31/1988 |
| From: | Boyd W, Krieg D, Schmidt B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20153G205 | List: |
| References | |
| NUDOCS 8809080093 | |
| Download: ML20153G258 (40) | |
Text
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- Electric Corporation Business Utilt Scz3912 Pmcure Pemsryama 15230 3312 CRITICAUTY ANALYSIS OF SEAVER VALLEY UNIT 2 FUEL RACKS March 1988 l
l W. A. B oyd D. J. Krieg B. W. S:nmidt W. A. Bordogna l - _ 88090g A
. . l TABLE OF CONTENTS 1.0 Introductlen ............................................... I 1.1 Design Description .................................1 1.2 Design Criteria ...................................2 2.0 Criticality Analytical Method ................................. 3 3.0 Criticality Analysis of Region 1 Spent Fuel 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 Criticality Analysis of Fresh Fuel Racks ......................... 12 5.1 Full Density Moderation Analysis ......................13 ,
5.2 Low Density Optimum Moderation Analysis ................. 14 6.0 Acceptance Criterion For Criticality ............................ 15 Bibliography .................................................. 34 ! i i 1 l Table of Contents - l l l l
. - _ _ . -- - - - . - . - - - - ---r,* ,,
i . t 4 LIST OF TABLES l Table 1. Benchmark Critical Experiments (5.6) ................ 16 r 1 Table 2. Fuel Parameters Employed in Criticality Analysis ......... 17 Table 3. Beaver Valley Unit 2 Fuel Assembiy Minimum Burnup vs initial U Enrichment for Region 2 Spent Fuel Mack .......... 18 Table 4. Comparison of PHOENIX isotopics Predictions to Yankee Core 5 Measurements ............................... 19 Table 5. Benchmark Critical Experiments PHOENIX Comparison ...... 20 Table 6. Data for U Metal and UO: Critical Experiments .......... 21 i l 1 l 1 l 1 r 4 l I ' l L i t i , i ! i
- l ;
i ^ i h I 1 I l List of Tablec ll l ) i II
s . . l LIST OF ILLUSTRATIONS Figure 1. Beaver Valley Unit 2 Fuel Assembly Minimum Burnup vs. Initial U Enrichment for Storage in Region 2 Spent Fuel Rocks . 23 Figure 2. Beaver Valley Unit 2 Spent Fuel Storage Cell Nominal Dimensions 24 Figure 3. Beaver Valley Unit 2 Spent Fuel Rack Layout ........... 25 Figure 4. Beaver Valley Unit 2 Fresh Fuel Storage Cell Nominal Dimensions 26 Figure 5. Beaver Valley Unit 2 Fresh Fuel Rack Layout ........... 27
- Figure 6. Beaver Valley Unit 2 Region 1 Three of Four Fuel Assembly Loading Schematic ............................ 28 Figure 7. Sensitivity of Ken to Enrichment in the Beaver Valley Unit 2 Region 1 Spent Fuel Storage Rack with Three of Four Storage 29 Figure 8. Sensitivity of K.n to Center-to-Conter Spacing in the Beaver Valley Unit 2 Region 1 Spent Fuel Storage Rack with Three of Four Storage .... ........................... 30
- Figure 9. Sensitivity of Ken to B 8 ' Loading in the Beaver Valley Unit 2 J Region 1 Spent Fuel Storage Rack with Three of Four Storage 31
- Figure 10. Sensitivity of K.o to Enrichment in the Beaver Valley Unit 2 ;
Region 2 Spent Fuel Storage Rack .................. 32 Figure 11. Sensitivity of K.o to Center-to-Conter Spacing in the Beaver Valley Unit 2 Region 2 Spent Fuel Storage Rack ......... 33 l ! Figure 12. Sensitivity of K.o to B Loading in the Beaver Valley Unit 2 i ! Region 2 Spent Fuel Storage Rack .................. 34 [ ] Figure 13. Sensitivity of K.o to Water Density in the Beaver Valley Unit 2 l j Fresh Fuel Storage Racks ........................ 35 , t l l i i a l 1 i i 1 List of Illustrations lii ; I
)
1.0 INTRODUCTION
! The Beaver Valley Unit 2 spent fuel rack (SFR) design described herein employs an array of racks, which will be considered as two separate spent fuel racks. Each of these fuel racks or arrays consists of existing Beaver Valley Unit 2 fuel I racks. This analysis will reanalyze these fuel arrays. The smallest array referred to as Region 1 will be reenalyzed for criticality to show that 4.85 w/o fuel can be stored in the rack in three out of four storage locations. The largest 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 4.85 Wlo. The Regions 1 and 2 spent fuel rack design is a poisoned rack, previously analyzed for storage of 17x17 STD fuel assemblies with enrichments up to 3.6 w/o U8 8 ' utilizing every storage location. i The Region 2 spent fuel rack reanalysis is based on maintaining K.o 5 0.95 for storage of Westinghouse 17x17 STD fuel at 4.85 w/o U's a with an initial enrichment /burnup combination in the acceptable area of Figure 1 with utilization of every cell permitted for storage of the fuel assemblies. I 1.1 DESIGN DESCRIPTION 4 The Region 1 and 2 spent fuel storage cell design is depleted schematically in Figure 2 with nominal dimensions given on the figvru. The spent fuel rack layout is shown in Figure 3. The fresh fuel rack stratage cell design is depicted schematically in Figure 4 and the fresh fuel rack layout is shown in Figure 5. ; I I l l Introduction 1
s . ; 5 i
.v 1.2 DESIGN CRITERIA r Criticality of fuel assemblies in a fuel storage rats .. 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 uncertaintles, there is a 95 percent probability at a 95 percent confidence level a that the effective multiplication factor (Ken) of the fuel assembly array will be 3 less than 0.95 as recommended in ANSI 57.2-1983, ANSI 57.3-1983 and in Ref-
- erence 1.
i o 1 > i l d I l l i i t t e l I ] 1 1 ! I r 4 l ! I l > 6 i ! i 1 . Introduction 2 t
. . i I
f f M 2.0 CRITICALITY ANALYTICAL METHOD ! The criticality calculation method and cross-section values are vserified 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. :
; The design method which Insures the criticality safety of fuel assemblies in the >
spent fuel storage rack uses the AMPXt 8, a> system of codes for cross- ! section generation and KENO IV' 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 ENOF/8-V8 8 8 data. The NITAWLc a s program includes, in this library, the (
self-shielded resonance cross-sections that are appropriate for each particular I
- geometry, The Nordheim Integral Treatment is used. Energy and spatial ,
) weighting of cross-sections is performed by the XSDRNPM' ' 3 program which l is a one-dimensional Sa transport theory code. These multigroup cross-section ; ) sets are then used as input to KENO IV' ' 8 which is a three dimensional Monte f Carlo theory program designed for reactivity calcul!! ions. l i
i A set of 33 critical experiments has been analyzed using the above method to I demonstrate its applicability to criticality analysis and to establish the method [ l bias and verlability. The experiments range from water moderated, oxida fuel arrays separated by various materials (84C, steel, water, etc) that simulate LWR l fuel shipping and storage conditions 8 8 8 to dry, harder spectrum uranium metal ! cylinder arrays with various interspersed materials t (Plexiglas and air) that demonstrate the wide range of applicability of the method. Table 1 surnmarizes , these emperiments. [
. L
- The average K.* of the benchmarks is 0.992. The standard deviation of the bias [
]
value is 0.0008 Ak. The 05/95 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 j 0.0018 Ak.
i Criticality Analytical Method 3 , e i l l' l' I J (
s s L 3.0 CRITICALITY ANALYSIS OF REGION 1 SPENT FUEL RACKS 3.1 REACTIVITY CALCULATIONS t The following assumptions were used to develop the nominal case KENO model for the Region 1 spent fuel rack storage of fresh fuel using three out of four storage locations:
- 1. The W 17x17 STO fuel assembly contains the highest enrichment authorized, ;
i is at its most reactive point in life, and no credit is taken for any burnable
< poison in the fuel rods (See Table 2 for fuel parametersh (
i i
- 2. All fuel rods contain uranium dioxide at an enrichment of 4.85 w/o U8 8 8 i over the infinite length of each rod. (
- 3. No credit is taken for any U8 8
- or U 8 8
- in the fuel, nor is any credit [
taken for the buildup of fission product poison material, j j 4. The moderator is pure water at a temperature of 68'F. A conservativt value of 1.0 gm/cm 8 is used for the density of water. ]
- 5. No credit is taken for any spacer grids or spacer sleeves.
- 6. Fuel assemblies are loaded into three of every four cells in a checkerboard pattern in the storage cells as shown in Figure 6.
j 7. The array is infinite in lateral and axial extent which precludes any neutron l
! leakage from the array.
- 8. The minimum poison material loading of 0.0'7 grams B 10 per square cen- ,
timeter is used throughout the array. , The KENO calculation for the nominal case resulted in a K.++ of 0.9139 with a 95 percent probabilityle5 percent confidence level uncertainty of 20.0056. L The maximum K.+, under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manufacturing l . process in addition to asymmetric positioning of fuel assemblies within the , l l Criticality Analysis of Region 1 Spent Fuel Racks 4
- c. .
l
.< j k
storage cells. Studies of asymmetric positioning of fuel assemblies within the l storage cells has shown that symmetrically placed fuel assemblies yield con- l l servative results in rack K o . The sheet metal tolerances are considered along i l with construction tolerances related to the cell I.D., and cell conter-to-center l l spacing. For the Region 1 racks this resulted in a reduction of the nominal 1.104" water gaps to their minimum values. Thus, the "worst case" KENO model j of the Rtgion 1 storage racks contains minimum water gaps of 1.007" with : symmetricitly placed fuel assemblies. f Based on tie analysis described above, the following equation is used to de- ; velop the m3xtrrum K.o for the Beaver Valley Unit 2 Region ' spent fuel storage ; racks with three out of four storage: l K.o a K . o + 8 in..
- 5 , * /( (ks)* n (ks)' in.e ]
where: a worst case KENO K.* that includes material [ tolerances, and mechanical tolerances which can f
. result in spacings between assemblies less than !
nominal j
- method bias determined from benchmark critical comparisons !
a bias to account for posion partical self-shielding
= 95/95 uncertainty in the worst case KENO K.o
= 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
K.o = 0.9267
- 0.0083
- 0.0013 * /((0.0046)8 + (0.0018)8 )=0.9417 Since K.o is less than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met with f
fuel enriched to 4.85 wlo. [ i l I l ! 1 . l i Criticality Analysis of Region 1 Spent Fuel Racks 5 I
i . 1 3.2 POSTULATED ACCIDENTS Most accident conditions will not result in an increase in Keet 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 j fuel height of stored assemblies which precludes interaction). However, accidents can be postuteted 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 l events to snsure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storarje pool water can be l l assumed as a realistic initial condition since not assuming its presence would ! j be a second unlikely event. [ The presence of coproximately 1000 ppm boroa in the pool water will ce: se
- reactivity by about 15 percent .iK. Thus, f or postulated accidents, shoul- e l
be a reactivity increase, K.ee wou.d be less than or equal to 0.95 due to mo ) effect of the dissolved boron. ! i j i l 3.3 SENSITIVITY ANALYSIS , To show the dependence of K.** on fuel and storage cells parameters as re-l quested by the NRC, the variation of the K.et with respect to the following pa- i ! remeters was developed using the KENO computer code: 1 ( t 1. Fuel enrichment. I !
- 2. Center-to-center spacing of storage cells. ;
j 3. Poison loading. i Results of the sensitivity analysis for the Region 3 storage cells are shown in Figures 7 through 9 for three of four storage. 1 l l t I i l j Criticality Analysis of Region 1 Spent Fuel Racks 6 i 1
40 CRITICALITY ANALYSIS OF REGION 2 SPENT FUEL RACK This scetion develops and describes the analytical techniques snd models em-ployed to perform the criticality analyses for storage of spent fuel in Region 2 of the Beaver Valley Unit 2 spsrit fuel pool. 4.1 REACTIVITY EQUIVALENCING Spent fuel eqrage, in the Region 2 spent fuel storage racks, is achievable by means of oc 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 sat of enrichmern fuel assembly discharge burnup ordered pairs which all yield the equivalent K.ve when the fuel is stored in the Region 2 racks. Figure 1 shows the constant K.tv contour generated for the Beaver Valley Unit 2 Region 2 racks. Note in Figure 1 ..ie endpoint at 0 MWD /MTU where the enrichment is 3.6 w/o and at 8,200 MWD /MTU where the enrichment is 4.85 w/o. The interpretation of the endpoint data is as follows: the reactivity of the Rngion 2 racks containing fuel at 8.200 MWD /MTU burnup which had an initial enrichment of 4.85 w/o is equivalent to the reactivity of the Region 2 racks containing fresh fuel having an initial enrichment of 3.6 w/o. It is important to recognize that the curve in Figure 1 is based on a constant rack reactivity for that region 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. l Criticality Analysis of Region 2 bpent Fuel Rack 7
s .
~
4.2 ANALYTICAL METHODS Tne 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 modiflod version of the British WIMS( ') library is d 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 ( ' ). 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 hour., af ter discharge. At this point in time, the major fission product poison, Xe 888 , has nearly completely decayed away. Fur-thermore, the fuel reactivity was found to decrease continuously from 100 hours to 30 years following discharge. Therefore, the most reactive point in time for a fuel arsembly after discharge from the reactor can be conservatively ap-proximated by removing the Xe * * * . The PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been examined following discharge from a rear-tor. 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 measure-ments were made on fuel discharged from Yankee Core 58 5
- 3 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 81 critical benchmark experiments is summarized in Table 5. Key parameters describing each of the 81 experiments are given in Table 6. These reactivity t omparisons again show good agreement between experiment and PHOENIX calculations. l An uncertainty associated with the burnup-dependent reactivities computed with PHOENIX is accounted for in the development of the Region 2 burnup require-ments. A blac of 0.003 Ak at 8,200 MWDIMTU is considered to be very con-servative since comparison between PHOENIX results and the Yankee Core experiments and 81 benchmark experiments indicates"closer agreement. l l l l i Criticality Analysis of Region 2 Spent Fuel Rack 8 l
& O 4.3 REACTIVITY CALCULATIO' ; The nominal and maximum K.et for storage of spent fuel in Region 2 is deter-mined using the methods described in Section 2 for Region 1 in addition to the methods described in Section 4.2. The actual conditions for this determint.tlon 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 3.6 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 2 storage of spent fuel:
- 1. The Westinghouse 17x17 SYD fuel assembly was analyzed for R ..on 2.
- 2. The Westinghouse 17x17 STD spent fuel assembly contains uranium dioxide fuel at an equivalent "fresh fuel" enrichment of 3.6 w/o U8 8 8
- 3. 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.
- 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 poison material loading of 0.017 grams B-10 per square cen-timeter is used throughout the array.
The KENO calculation for the nominal case resulted in a K.vt of 0.9246 with a 95 percent probability /95 per:ent confidence level uncertainty of t0.0046. The maximum K.e, under normal conditions was determined with a "worst case" KENO model, in the same manner as for the Region 1 storage racks (see Section 3). For the Region 2 racks, the water gaps are reduced from the nominal value of 1.106" to their minimum value. Thus, the "worst case" KENO model of the Region 2 storage racks contains minimum water gaps of 1.007" with symmet-rically placed fuel assemblies. The uncertainty associated with the reactivity l equivalence methodology was included in the development of the burnup re-quirements. This t ncertainty was discussed in Section 4.2. Based on the analysis described above, the following equation is used to de- l velop the maximum K ve for the storage of spent fuel in the Beaver Valley Unit 2 Region 2 spent fuel storage racks: Criticality Analysis of Region 2 Spent Fuel Rack 9
K.if = K..,.i + Bm.m a + B..,i + / [ (k s) 8..,.i + (ks)8m.inoe ] where:
= worst case KENO K.et ' that includes centered fuel assembly positions, material tolerances, and mechanical tolerance which can result in spacing between assemblies less than nominal 0*** * = method bles determined from benchmark critical comparisons
"" = bias to account for poison partical self-shielding s..ni
= 95/95 uncertainty In the worst case KENO K.#,
= 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
K.vf = 0.9336 + 0.0083 + 0.0018 + /((0.0046)8 + (0.0018)8 ] = 0.9486 The maximum K.ve 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 3.6 w/o U8 8 8 4.4 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 (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. Criticality Analysis of Region 2 Spent Fuel Rack 10 ; l
' ~ ~
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.tv would be less than or equal to 0.95 due to the effect of the dissolved boron. 4.5 SENSITIVITY ANALYSIS To show the depenci-'nce of K.ve on fuel and storage cell parameters as re-quested by the NRC, sensitivity studies were performed in which the poison
'oading, the fuel enrichment, and the storage cell center-to-center spacing were varied, using the PHOENIX computer code.
Figures 10 through 12 illustrate the results of the sensitivity studies for spent fuel occupying every cell in the Region 2 fuel racks. l L L l l i l l 1 Criticality Analysis of Region 2 Spent Fuel Rack 11 l 1
i . 5.0 CRITICALITY ANALYSIS OF FRESH FUEL RACKS This section describes the analytical techniques and models employed to per-form the criticality analysis for storage of fresh fuel in the Beaver Valley Unit 2 fresh fuel racks. Since the fresh fuel racks are maintained in a dry condition, the criticality adalysis will show that the rack K.et 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 criticJllty 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 fuel rods.
- 2. All fuel rods contain uranium dioxide at an enrichment of 4.85 w/o U 8 8 8 over the infinite length of each rod.
- 3. No credit is taken for any U8 8
- or U 8 8
- In the fuel, nor is any credit taken for the buildup of fission product poison material.
- 4. No credit is taken for any spacer grids or spacer sleeves.
The . 57x17 STD fuel assembly was analyzed. (See Table 2 for fuel parameters) l I I Criticality Analysis of Fresh Fuel Raevs 12 j J
4 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 is used for the density of water. The fuel array is infinite in lateral 8 and axial extent which precludes any neutron leakage from the array. The maximum K.es under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the nanufacturing 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.tv . The manufacturing tolerances are stacked in such a manner to minimize the assembly center-to-center spacing and the total voi-ume of steel thereby causing an increase in reactivity. The sheet metal toler-an es are considered along with construction tolerances related to the cell 1.D. and cell center-to-center spacing. For the fresh fuel starage racks, the assembly center-to-center spacing is reduced from a nominal voue of 21" to a minimum of 20.9375". 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. Dased on the analysis described above, the following equation is used to de-velop the maximum Keef for the Beaver Valley Unit 2 fresh fuel storage racks: Kete = K ..<et + 0m. .e + / [ (k s) * ..< u + (k s) 8 m.m.a ) where:
= worst case KENO K tv that includes material tolerances, and mechanical tolerances which can result in spacings between assemblies less than j nominal 1
= method bias determined from benchmark critical comparisons
= 95/95 uncertainty in the worst case KENO K.ve I
= 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
K.tv = 0.9099 + 0.0083 + /((0.0080)8 + (0.0018)8 ) = 0.9264 l l Since K.ve is less than 0.95 including uncertainties at a 95/95 probability confi- ; dence level, the acceptance criteria for criticality is met. Criticality Analysis of Fresh Fuel Racks 13 l l 1
5.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS In the low density optimum moderation analysis, the fuel array is finite in the radial and axial extent. Analysis of the Beaver Valley Unit 2 racks has shown that the maximum rack K.ve under low density moderation conditions occurs at 0.076 gm/cm water 8 density. The KENO calculation of the Beaver Valley Unit 2 fresh racks at 0.076 gm/cm 8 water density resulted in a peak K.tv of 0.9233 with a 95 percent probability and SS percent confidence level uncertainty of 10.0080. Figure 13 shows the fresh fuel rack reactivity as a function e' the water density. The minimum cell center-to-center spacing, rack module spacing and material tolerances have been inciudad in the base case mode! and result in a storage cell separation distance of 11.72" and a rack module separation distance of 20.9375 inches. Based on the analysis described above, the following equation is used to de-velop the maximum K.tv for the Beaver Valley Unit 2 fresh fuel storage racks under low density optimum moderation conditions: K.ve = Km... + Bm.inee + (((ks) 8o... + (ks) 8m.inoe ] where:
= base case KENO K.tv that includes nominal mechanical and material dimension
= method bias determined from benchmark critical comparisons
= 95/95 uncertainty in the base case KENO K.ve
= 95/96 uncertainty in the methoC bias Substituting calculated values in the order listed above, the result is:
Keve = 0.9233 + 0.0083 + /((0.0080)8 + (0.0018) 8 ] = 0.9398 Since K.ev is less than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance critcria for criticality is met. Criticality Analysis of Fresh Fuel Racks 14 a
6.0 ACCEPTANCE CRITERION FOR CRITICALITY The neutron multiplication f actor 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, "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," ANSI 57.3-1983, "Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants." r
'lJ l
Acceptance Criterion For Crit'cality 15 l
,s Table 1. Benchmark Critical Experiments (5,6)
Cener al Enctchment Separating Soluble Descelotton s/o U235_ Reflector matertal Boron opa E gg
- 1. UO rod lattice 2.46 water setec 0 0.9952 * .0020
- 2. U0 rod lattice 2.46 matee watee 1037 0.9906 7 .0008
- 3. UO rod tattice 2.46 mater water 764 0.9896 7 ,00:5 4 U0 rod lattice 2.46 watec 84C pins 0 0.9914 7 .0025
- 5. VO cod lattice 2.46 saler 84C pins 0 0.9999 T. 0026
- 6. U0 rod fattIce 2.46 er.#pr 84C pins 0 0.9955 7 .0020
- 7. UO rod lattice 2.46 unter 54C pins 0 0.9889 7 .0026
- 8. U0 rod tattIce 2.46 mater 64C pins 0 0 9993 T .0025
- 9. UO rod lattice 2.46 water watec 0 0 9931 T .0029
- 10. VO rod lattice 2.46 sater water 143 0.9923 T .0025 II. U0 rod lat t ice 2.46 vater stainless steel 514 0.9967 7 .0020
- 12. UO rod lattice 2.46 mater stainless steel 217 0.9943 7 ,colg
- 13. U0 rod lattice 2.46 water borated aluntnum 85 0.9992 7 .0023 14 U0 rod tattIce 2.46 mater bor a t ed s t un t mae 92 0.9684 7 .0023
- 15. UO rod lattice 2.46 mater bocated aluminum 395 0.9832 7 .0021
- 16. UO rod lattice 2.46 mater borated aluminum 125 0.9448 T .0024
- 17. UO red lattice 2.46 u4ter bocated stoutnum 487 0.9895 7 ,0o7o
- 10. 00 rod lattice 2.46 matve borated alumtrue 197 0.ggg5 7 .0022
- 19. 00 rod lattice 2.46 estee berated aluminum 634 0.992 T cogg
- 20. UO rod lattice 2.46 sater borated aluminum 320 0.9920 7 .0020
- 28. U0 rod lattice 2.46 mater bocated aluminum 72 0.9939 7 .0020
- 22. U tal cylinders 93.2 b4*e ate 0 0.9905 7 .0070
- 23. U metal cylinders 93.2 bare air 0 0.9976 Y .0020 24 U metal cylinders 93.2 bare ate 0 0.9947 7 .0025
- 25. U metal cylinders 93.2 bare ate 0 0.9929 7 .0019 i
- 23. U metal cyltrwiefs 93.2 bare age 0 0.9922 7 .0026 l
- 27. U met al cyt tnders 93.2 b4re ate 0 0.9950 T .0027
- 26. U metal cylindees 93.2 bare plealglass 0 0.9941 T ,co3o
- 29. U met al cylinders 93.2 parafffn plealglass 0 0.9928 T .0041
- 30. U metal cylindars 93.2 bare plestglass 0 0.9960 T .0010
- 31. U metal cylinders 91.2 paeaffin plestglass 0 1.0042 7 .0019
- 32. U met al Cyl trwters 93.2 paraffin pleatglass 0 0.9963 7 ,co3o J3. U metal cyltnders 93.2 paraffin Olestglass 0 0.9919 7 .0032 l
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l l l l Table 2. Fuel Parameters Employed in Criticality Analysis Parameter W 17x17 STANDARD Number of Fuel Rods per Assembly 264 Rod Zirc-4 Clad 0.0. (inch) 0 374 Clad Thickness (Inch) 0.0225 Fuel Pollet 0.D. (Inch) 0 3225 Fuel Pellet Density (% of Theoretical) 96 Fuel Pellet Dishing Factor 0.0 Rod Pitch (i nch)- 0.496 Number of Zirc-4 Guide Tubes 24 Guide Tube 0.0. (Inch) 0.482 Guide Tube Thickness (inch) 0.016 Number of Instrument Tubes 1 Instrument Tube 0.0. (inch) 0.482 Instrument Tube Thickness (Inch) 0.016 17
l Table 3. Beaver Valley Unit 2 Fuel Assembly Minimum Burnup vs initial U8 8 8 Enrichment for Region 2 Spent Fuel Rack initial U 8 8 ' Assembly Discharge Enrichment Burnup (GWD/MTU) Region 2 36 0 4.0 2.6 4.4 53 4.85 8.2 18 l I
i .
.. l Table 4. Comparison of PHOENIX lsotopics Predictions to Yankee Core !i 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 Mase(PU/U) + 1.41 FISS-?UITOT-PU -0.02 19
e . l l 1 4 I Table 5. Benchmark Critical Experiments PHOENIX Comparison Description of Number of PHOENIX K.n Using Experiment Experiments Experiments Bucklings UO: Al clad 14 0.9947 SS clad 19 0.9944 Borated H 0 7 0.9940 Subtotal 40 0.9944 U-M etal Al clad 41 1.0012 TOTAL 81 0.9978 1 20
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Table 6. Data for U Metal and UO: Critical Experiments (P8rt 2 Of 2) Fuel Pellet Clad Clad Lattice Case Cell A/O H20/U Density Diameter Matental 00 7 hick nes s Pitch Boron Number 7ype U-235 Ratto (0/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 Hexe 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- 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 18.00 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 51 Hexa 1.040 1.01 18.90 1.5240 Aluminum f.6916 . 2.1742 0.0 52 Hexa 1.040 1.51 18.90 1.5240 Aluminum 1.6916 .'07112 07112 2.4004 0.0 53 Hexa 1.040 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.616; O.0 54 Hexa 1.040 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 4 55 Hexa 1.040 4.02 18.90 1.5240 Aluminum t.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 Hoxa 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 Hera 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 65 0.0 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 67 Hexa 0.0 1.160 2.02 18.90 .9830 Aluminum 1.1566 .07112 1.7247 0.0 68 Hera 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 69 Hexa 1.160 t 9609 0.0 4.02 18.90 .9830 Aluminu3 1.1506 .07112 2.1742 0.0 70 Hexa 1.040 1.33 18.90 19.050 Alumirva 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 i3.90 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 10.90 19.050 Aluminum 2.0574 75 .07620 3.6284 0.0 l 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 77 Hexa 1.310 3.01 18.88 2.6160 0.0 1.5240 Aluminum 1.6916 .07112 2.990J 0.0 78 Hexa 1.159 2.02 18.88 1.5240 Aluminum 1.6916 79 Hexa 3.01
.07t12 2.6 t40 0.0 1.159 18.88 1.5240 Aluminum 1.6916 .07112 2.99CO O.0 80 Hexa 1.312 2.03 18.88 .9830 Aluminum 1.1506 .07112 )
81 Hexa 1.312 3.02 18.88 1.7250 0.0
.9830 Aluminum 1.1506 .07tt2 1.9410 0.0 l l
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Figure 7. SensitMty of Ken to Enrichment in the Beaver Valley Unit 2 Region 1 l Spent Fusi Storage Rack with Three of Four Storage i I 29 l
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, POIS0NLOAD(GMB10/CM) 1 4 Figure 9. Sensitivity of K.et to B1' Loading in the Beaver Valloy Unit 2 Region 1 Spent Fuel Storage Rack with Three of Four Storage 31
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Figure 10. Sensitivity of K.ve to Enrichment in the Beaver Valley Unit 2 Region 2 Spent Fuel Storage Rack 32
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l l l 1 j Figure 13. Sensitivity of Les to Water Density in the Beaver Valley Unit 2 Fresh Fuel Storege Racks ] j 35
I . BIBLIOGRAPHY
- 1. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes OT Position for Review and K:eptance of Spent fuel Storage and Handling Applications, April 14, 1978.
- 2. W. E. Ford lll, 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, AMPX: A Modular Code System for Generatir>g Coupled Multigroup Neutron-Gamma Librarles 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 BAW-1484-7, July 1979.
l 6. J. T. Thomas, Critical Three-Dimensional Arrays of U(93.2) Metal Cylinders, t Nuclear Science and Engineering, Volume 52, pages 350-359,1973,.
- 7. A. J. Harris, A Descrip*lon of the Nuclear Design and Analysis Programs for 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.
) 564-584, 1966.
- 9. England, T. R., CINDER - A One-Point Depletion and Fission Product Program. WAPD-TM-334, August 1962.
l 10. Melenan, J. B., Yankee Core Evaluation Program Final Report, j' WCAP-3017-6094, Janu y 1971. I 1 1 1 t bibliography 36 1 I i i -}}