ML17229A571

Criticality Safety Analysis for Spent Fuel Storage Rack Using Soluble Boron Credit.

ML17229A571
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Site:	Saint Lucie
Issue date:	10/31/1997
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
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ML17229A570	List: ML17229A569 ML17229A571 ML17309A916 ML17309A917
References
CENPD-387, NUDOCS 9801070057
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CENP D-387, St.Lucie Unit 2 Criticality Safety Analysis for the Spent Fuel Storage Rack Using Soluble Boron Credit.October 1997 PDR ADOCK 05000389 980i070057 97i23i PDRi'BB'ombustion Engineering Nuclear Operations J Ion r IeIr r

CENpD487 Table of Contents Table of Contents.List of Tables.List of Figures.1.0 Introduction 1.1 Design Criteria.1.2 Design Description.

1.3 Analysis Descriptions 2.0 Analysis Methods 2.1 SCALE-PC.................

2.2 The DIT Code.3.0 Spent Fuel Pool and Storage Rack.3.1 Storage Rack Description.....

3.2 Spent Fuel Storage Pattern.4.0 Criticality Safety Analysis.4.1 Analytic Models of the Storage Rack and Module Cells..4.2 K,a Evaluation at Zero Soluble Boron.4.3 K,a Evaluation for Soluble Boron Credit.4.4 Reactivity Equivalencing 4.4.1 Burnup and Decay Time Reactivity Equivalencing.

4.4.2 Gadolinium

Credit Reactivity Equivalencing.

4.4.3 Soluble Boron Credit for Uncertainties in Reactivity Equivalencing.

4.5 Axial Burnup Distribution.

5.0 Postulated

Accidents.

6.0 Soluble Boron Credit Summary.7.0 References.

...'.....3 10 12 15 15 16 17 18..18 19 21.....21..23..25 26 2

CENp0487 Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 Figure 6 Figure 7 Figure 8 Figure 9 Figure 10 Figure 11 Figure 12 Figure 13 Figure 14 Figure 15 Figure 16 Figure 17 List of Figures Spent Fuel Storage Module Installation.

Typical Spent Fuel Storage Rack Module..Typical Spent Fuel Rack Module L-Insert.L-Inserts.

Spent Fuel Storage Module.Fuel Assembly.Spent Fuel Rack Module For Region I Spent Fuel Rack Module For Region II Spent Fuel Loading Pattern For Region I Spent Fuel Loading Pattern For Region II Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region II, 1.3 w/o.Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region II, 1.5 w/o.Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I, 1.4 w/o.Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I, 1.82 w/o Required Fuel Assembly Burnup vs Initial Enrichment, Region I, 2.82 w/o........

K-Infinity At 5.0 w/o With 90%Gad Worth (Through 60000 Mwd/T).............:

K-Infinity At 5.0 w/o With 90%Gad Worth (Through 20000 Mwd/T).........,....

....37....38....39....,40 41 42 43 44 45 46 47 48 49 50 51 52 53 CENpo487 1.0 Introduction This report presents the results of a criticality analysis for the St.Lucie Unit 2 spent fuel storage rack taking credit for assembly burnup, for soluble boron in the spent fuel pool, for gadolinium burnable absorbers and for actinide decay.The methodology employed in this analysis is analogous to that of Reference 1 and employs analysis criteria consistent with those cited in the Safety Evaluation by the Office of Nuclear Reactor Regulation, Reference 2.1.1 Design Criteria The design criteria are consistent with GDC 62, Reference 3, and NRC guidance to all Power Reactor Licensees, Reference 4.Section 2.0 describes the analysis methods including description of the computer codes used to perform the criticality safety analysis.A brief summary of the analysis approach and criteria follows.1.Determine the storage configuration of the spent fuel racks using no soluble boron conditions such that the 95/95 K,ir upper tolerance limit of the system, including applicable biases and uncertainties, is less than unity.2.Next, using the resulting configuration from the previous step, calculate the spent fuel rack effective multiplication factor with the chosen concentration of spent fuel pool soluble boron present.Then calculate the sum of: (a)the latter multiplication factor, (b)the reactivity uncertainty associated with fuel assembly and storage rack tolerances, and (c)the biases and other uncertainties required to determine the final 95/95 confidence level effective multiplication factor and show that at the chosen concentration of soluble boron, the system maintains the overall effective multiplication factor less than or equal to 0.95.3.Use reactivity equivalencing methodologies to determine the minimum fuel assembly burnup for fuel assembly enrichments greater than allowed in step 1, above.As a function of time after discharge and burnup, calculate the reactivity credit due to actinide decay for each fuel assembly.For fuel assemblies containing Gdq03-UO2 rods, evaluate reactivity credit due to the lumped burnable poison.Include these credits in the reactivity equivalencing for each fuel assembly.4.Determine the increase in reactivity caused by postulated accidents and the corresponding additional amount of soluble boron needed to offset these reactivity increases.

1.2 Design Description The 16xl6 ABB CE fuel design characteristics are given in Table 3.The fuel pellet is characterized by the"Value Added" concept, which includes a slightly expanded pellet diameter and higher fuel stack density relative to previous designs.All the spent fuel pool reactivity calculations include the effect of Value Added pellets.N Ik Ik P%ININ

CENPD487 The St.Lucie Unit 2 spent fuel storage racks are described in detail in the Update Final Safety Analysis Report (UFSAR), Reference 5.These storage racks contain no supplemental poison beyond the structural materials and the L-inserts in Region I.Section 3.0 and Figures 1 through 10 provide a description of the storage cells, storage modules and pool configuration.

1.3 Analysis Descriptions Technical Specifications and the UFSAR limit the present utilization of the spent fuel storage rack to fuel assemblies having an initial enrichment of 4.5 w/o U-235 arranged in a checkerboard pattern in Region I and to three out of four positions in Region II.Thus, the primary objective of this analysis of the spent fuel storage rack is to obtain more ef6cient utilization of the available storage capacity consistent with the latest NRC approved methodology, viz., employing credit for soluble boron.In addition, the analyses presented in Section 4.0 and Figures 11-17 demonstrate not only a significant increase in the utilization of available storage cells by taking credit for actinide decay, but also the capability for employing U-235 enrichment levels up to 5.0 w/o in fuel assemblies containing Gadolinia-UO2 rods.Section 5.0 presents the additional boron requirements to protect against several postulated accidents:

fuel assembly drop, loss of spent fuel pool cooling, and fuel assembly misload.Section 6.0 presents the combined soluble boron requirements from this analysis.

E~

CENPD487 2.0 Analysis Methods The analysis methodology used in the evaluation of the storage configuration of the spent fuel storage rack employs: (1)SCALE-PC, a personal computer version of the SCALE-4.3 code package documented in Reference 6, with the updated 44 group ENDF/B-5 neutron cross section library, and (2)the two-dimensional integral transport code DIT, Reference 7, with an ENDF/B-6 neutron cross section library.SCALE-PC is used for both overall storage rack as well as sub-region type K,Q calculations; SCALE-PC modules employed in both the benchmarking analyses and the spent fuel storage rack analyses include CSAS-2, BON-AMI, NITAWL, and KENO-Va.The DIT code is used for simulation of in-reactor fuel assembly depletion and specific types of storage cell calculations.

The following sections describe the application of these codes in more detail.2.1 SCALE-PC Validation of SCALE-PC for purposes of fuel storage rack analyses is based on the analysis of selected critical experiments from two experimental programs.The first is the Babcock 0 Wilcox experiments carried out in support of Close Proximity Storage of Power Reactor Fuel, Reference 8.The second'program is the Pacific Northwest Laboratory Program carried out in support of the design of Fuel Shipping and Storage Configurations; the experiments of current interest to this efFort are documented in Reference 9.Reference 10, as well as several of the relevant thermal experiment evaluations in Reference 11, were found to be useful in updating pertinent experimental data documented in Reference 9.For purposes of code validation, nineteen experimental configurations were selected from the BOW experimental program.These consisted of the following experimental cores: Core X, the seven measured configurations of Core X, Cores XI through XXI, and Core XIIIa.This approach focused on using measured rather than extrapolated configurations to avoid introducing any biases or uncertainties associated with the extrapolation techniques.

This group of experimental configurations employed variable spacing between individual rod clusters in the nominal 3 x 3 array.In addition, the efFects of placing either SS-304 or B/Al plates of different blacknesses in the water channels between rod clusters were measured.Table 1 summarizes the results of these analyses.Similarly, eleven experimental configurations were selected from the PNL experimental program.These included unpoisoned uniform arrays of fuel pins and 2 x 2 arrays of rod clusters with and without interposed SS-304 or B/Al plates of difFerent blacknesses.

Table 2 summarizes the results of these analyses.The approach employed for a determination of the calculational bias is based on Criterion 2 of Reference 12.For a given KENO eigenvalue and uncertainty, the magnitude of K9$/95 is computed by the following equation;by this definition, there is a 95 percent confidence level that in 95 percent of similar analyses the validated calculational model will yield a multiplication factor less than K~5~5.

CENPD487 where: I/2 2 K9//9$KKENo lUCB M95/95(O+%KENO)K~go is the KENO multiplication factor of interest, LQ<B is the mean calculational method bias, M95/95 is the 95/95 multiplier appropriate to the degrees of freedom for the number of validation analyses, a is.the mean calculational method variance deduced from the validation 2 analyses, and a~go is the standard deviation appropriate to the KENO multiplication factor of interest.The equation for the mean calculational methods bias is as follows.~=-g,(1-K)where: K;is the i value of the multiplication factor for the validation lattices of interest, and Mq5;q5 is obtained from the tables in Reference 13.The equation for the mean calculational variance of the relevant validating multiplication factors is as follows.2 0 ave where k'"'s given by the following equation.a, is given by the following equation.2 CENPD48F where G;is the number of generations.

For purposes of this bias evaluation, the datapoints of Tables 1 and 2 are pooled into a single group.With this approach, the mean calculational methods bias, dXa, and the mean calculational variance, (a), calculated by equations given above, are determined to be 0.00259 and (0.00288), respectively.

The magnitude of M95/9$is deduced from Reference 8 for the total number of pooled data points, 30.The magnitude of K~s~s is given by the following equation for SCALE 4.3 KENO Va analyses employing the 44 group ENDF/B-V neutron cross section library and for analyses where these experiments are a suitable basis for assessing the methods bias and calculational variance.K95/95 K~yo+0.00259+2.22[0.00288+(axago)]A full scale mock-up of the St.Lucie Unit 2 spent fuel storage array, that is six Region I and thirteen Region II storage modules with the nominal two-inch inter-module spacing, was modeled in KENO-Va for basic evaluations of the characteristics of the fuel assembly storage rack.These KENO calculations typically employed one million neutron histories.

DiFerent neutron starting distributions were employed depending upon the type of calculation to ensure conservative multiplication factors were employed in the evaluations.

For some calculations, such as the fuel assembly misload accidents in Region I, a smaller representation of the storage rack was employed which consisted of the whole of Region I plus one row of Region II modules along the interface boundary between the two regions to maintain a more correct representation of the boundary conditions for Region I.A third class of KENO calculations employed individual storage cell representations for both Region I and II type cells, that is module cells with and without L-inserts.

Calculations with these two geometries employed both fresh and burned fuel representations of the ABB-CE 16 x 16 fuel assembly design.These infinite array storage cell models had the disadvantage of a stochastic calculation but the advantage of the highly flexible SCALE-PC geometry capability.

Typical KENO calculations with these cells employed 500,000 neutron histories.

Consequently, the practice of calculating di6erences,in K ir at the one-sigma level had little impact on the quantitative results for cases of importance.

CENP0487 2.2 The DIT Code The DIT (Discrete Integral Transport) code performs a heterogeneous multigroup transport calculation for an explicit representation of a fuel assembly.The neutron transport equations are solved in integral form within each pin cell.The cells retain full heterogeneity throughout the discrete integral transport calculations.

The multigroup spectra are coupled between cells through the use of multigroup interface currents.The angular dependence of the neutron flux is approximated at cell boundaries by a pair of second order Legendre polynomials.

Anisotropic scattering within the cells, together with the anisotropic current coupling between cells, provide an accurate representation of the flux gradients between dissimilar cells.The multigroup cross sections are based on the Evaluated Nuclear Data File Version 6 (ENDF/B-VI).Cross sections have been collapsed into an 89 group structure, which is used in the assembly spectrum calculation.

Following the multigroup spectrum calculation, the region-wise cross sections within each heterogeneous cell are collapsed to a few groups (usually 4 broad groups), for use in the assembly flux calculation.

A B 1 assembly leakage correction is performed to modify the spectrum according to the assembly in-or out-leakage.

Following the flux calculation, a depletion step is performed to generate a set of region-wise isotopic concentrations at the end of a burnup interval.An extensive set of depletion chains is available, containing 33 actinide nuclides in the thorium, uranium arid plutonium chains, 171 fission products, the gadolinium, erbium and boron depletable absorbers, and all structural nuclides.The spectrum-depletion sequence of calculations is repeated over the life of the fuel assembly.Several restart capabilities provide the temperature, density and boron concentration dependencies needed for three dimensional calculations with full thermal-hydraulic feedbacks.

The DIT code and its cross section set have been used in the design of reload cores and extensively benchmarked against operating reactor history and test data.For the purpose of spent fuel pool criticality analysis calculations, the DIT code is used to generate the fuel isotopics as a function of fuel burnup and initial feed enrichment.

These isotopics are input to KENO to generate the K-infinity vs.burnup, K-infinity vs.enrichment, and the burnup vs.enrichment curves.The code is also used to calculate input to the gadolinium burnable absorber reactivity credit and the actinide depletion burnup credit analyses.10 CENPD487 0 3.0 Spent Fuel Pool and Storage Rack The St.Lucie Unit 2 spent fuel storage racks are described in the UFSAR (Reference 5).This section provides a more detailed description of the spent fuel storage rack with the objective of establishing a basis for the analytical model employed in the criticality analysis described in Section 4.0.For purposes of this criticality analysis, the value-added fuel rod parameters of Table 3 were used.3.1 Storage Rack Description The spent fuel storage rack and pool environment are described in Section 9 of the St.Lucie Unit 2 UFSAR.Figure 1, a copy of UFSAR Figure 9.1-5, shows a planar view of the array of nineteen modules within the pool.Figures 2 and 3 (UFSAR Figures 9.1-2 and 9.1-3a)illustrate details of a Region I module with the L-inserts present;note also, the perimeter strip about the top of the modules.Figure 4 (UFSAR Figure 9.1-3b)shows more detail on the L-inserts.

The modules in both regions are positioned in the pool to provide a minimum separation of two inches between adjacent modules.Figure 5 (UFSAR Figure 9.1-4)provides overall dimensions and tolerances on the four basic module types.k A clarification on the details of the L-insert illustrated in Figure 4 is warranted.

As noted in the latter figure, the SS-304 plate stock employed to form the L-insert has a nominal thickness of 0.188 inches.The overall dimensions of the formed L-insert is 8.740+0.000/-0.050 inches, exclusive of the locking tab region illustrated in Figures 3 and 4.Included in the latter overall dimension of the L-insert is the dimension and tolerance of the out-facing elliptical dimples in the two outside faces of the L-insert.Nine dimples are spaced 15 inches apart in the axial interval of the L-insert spanning the active portion of the fuel assembly;the dimples in the two faces of the L-insert are offset axially by one half of the spacing pitch.The height of the dimples above the outside surface of the L-insert is specified as 0.070+0.010 inches.As a consequence, the nominal Region I cell has a 0.070 inch water gap between the module cell wall and the outside surface of the L-insert.To give a better perspective of the vertical dimensioning of the components of a Region I module cell, the L-inserts are approximately 165 inches long and the module cell walls are approximately 178 inches high;the base of the fuel assembly illustrated in Figure 6 (Figure 4.2-6 of the UFSAR)rests on a support plate approximately 5 inches above the base of the module cell.Region I of the storage rack consists of six modules, two 7 x 10 and four 7 x 11 storage modules, containing a total of 448 storage cells;each module cell contains a stainless steel L-insert within the cell to provide a double wall region, with an intervening water region, between each storage location as illustrated in Figure 7 (UFSAR Figure 9.1-5a).Region II of the storage rack consists of 13 modules, one 8 x 10 and twelve 8 x 11 modules, containing a total of 1136 storage cells;each storage cell in this region has a single stainless steel wall separating adjacent storage cells as illustrated in Figure 8 (UFSAR Figure 9.1-5b).IL II lk P%ININ CENPD487 Since the basic module cells are of the same nominal dimensions in each module type in both regions of the storage rack, the nominal Region I storage cell has a smaller internal area due to the presence of the L-insert device.The monolithic structure of each module is formed by welding the 178 inch long juncture between square right angle and slab components formed from 0.135-inch thick plate stock.The resulting cells have internal dimensions of 8.740+0.180/-0.000 inches.Overall dimensions on the monolithic structure, including the 7/16 inch perimeter strip at the top of the module, are shown in Figure 5.The overall dimensions are keyed to the number of cells along a side of the module as 100+I/2, 91+1/2, 73+1/2, and 64+1/2 inches for the 11, 10, 8, and 7 unit cell dimensions.

3.2 Spent Fuel Storage Pattern The spent fuel storage pattern is depicted in Figure 9 for Region I and Figure 10 for Region II as an array of the equivalent uniform enrichment fuel assemblies.

It is noted that for Region I there are 172 water cell locations whereas in Region II there are 52.In Region'II, the four water cells within each module are located in a symmetric pattern relative to the corners of each module.In Region I, the water cell locations are positioned quite differently.

There are eight classes of fuel types that may be stored in the spent fuel storage rack;three classes , of these fuel assembly types employ Control Element Assemblies (CEAs)as supplemental reactivity hold-down devices.The fuel assembly classes are summarized as follows: 1)2)3)4)5)6)7)8)9)Region II: 1.3 w/o U-235 equivalent, Region II: 1.5 w/o U-235 equivalent, Region I: 4.5 w/o U-235 equivalent, Region I: 4.5 w/o U-235 equivalent with CEA inserted, Region I: 5.0 w/o U-235 equivalent with Gd.poison rods, Region I: 5.0 w/o U-235 equivalent with Gd.poison rods and CEA inserted, Region I: 2.82 w/o U-235 equivalent with CEA inserted, Region I: 1.82 w/o U-235 equivalent, and Region I: 1.4 w/o U-235 equivalent All evaluations in this report employ the more reactive value-added fuel rod type, consequentially there is no differentiation between the standard and value-added fuel assembly types.In addition, the more reactive fuel assembly classes listed above were employed in the criticality analysis for the spent fuel rack, i.e.classes 1, 2, 3, 8 and 9..As noted in the above tabulation, only two of the eight fuel classes, viz., the 1.3 and 1.5 w/o equivalent enrichment fuel assembly classes are designated for storage in Region II of the storage rack.The remaining six classes are restricted to Region I.Region II is arranged as follows:~The 1.3 w/o enrichment equivalent fuel assembly locations in Region II are shown as the dark squares in Figure 10 (Class 1).iL MIN PRISSY 12 CENP0487 The gray peripheral squares and the four inboard locations along the interface between Regions I and II are the 1.5 w/o enrichment equivalent locations (Class 2)The white squares located three positions inboard from the corner position of each module are the water cell locations.

The fuel assembly types depicted in Figure 9 for Region I are the 4.5 w/o enrichment equivalent fuel assemblies with and without CEA's, and the 1.82 and 1.4 w/o enrichment equivalent fuel assemblies.

Region I is best described as an annular configuration of fuel assembly types.Progressing from inside to outside: The central 3 x 10 array of dark blue squares adjacent to the boundary between Regions I and II consists of 1.4 w/o fuel assemblies (Class 9).The next ring of medium blue cells consists of 1.82 w/o fuel assemblies (Class 8).The next ring of cells is water cells.The ring beyond, due to lack of an odd number of cell locations, consists of eleven light blue cells with 5 white dots which are 4.5 w/o assemblies without CEAs (Class 3), 14 water cells, and one medium blue 1.82 w/o assembly (Class 8).The latter assembly fills the only inside corner location in this row not occupied by a water cell.The next ring of cells consists of 1.82 w/o assemblies (Class 8)The next ring is all water cells.The next double ring of cells consists of 1.82 w/o assemblies (Class 8)with the exception of the two water cells inboard to the outside corner locations.

Next is a third layer of water cells This is followed by a ring of 4.5 w/o assemblies alternating between those with CEAs (Class 4)and those without CEAs (Class 3).Cells without CEAs (Class 3)have 5 light dots.Lack of an odd number of cells again dictated a mixed array in this ring.The two corner locations are occupied by 4.5 w/o assemblies without CEAs as are alternating storage locations in both directions away from the corner except near the center of the middle-upper module.Here it was necessary to insert two 1.82 w/o assemblies (Class 8)to control the local K.The fourth and outermost ring of water cells follows.An array of fuel assembly enrichments nearly identical to the previous one was employed in the outermost ring of fuel assemblies.

It has 8 more assemblies which merely extend the alternating pattern.The remaining classes of fuel assembly types not addressed in the last paragraph are as follows.A 2.82 w/o equivalent enrichment fuel assembly with a full strength CEA inserted (Class 7)may be interchanged with a 1.82 w/o equivalent enrichment fuel assembly (Class 8)since the former is less reactive than the latter.In a similar vein, a 5.0 w/o U-235 enriched, gadolinium shimmed fuel assembly having a gadolinium loading of the type specified in Section 4.4.2 (Class 5)or the same fuel assembly containing a full strength CEA (Class 6)are less reactive than their 13 1,

CENp0487 unshimmed 4.5 w/o equivalent enriched fuel assembly counterparts (Class 3 and 4, respectively).

CENpo487 4.0 Criticality Safety Analysis 4.1 Analytic Models of the Storage Rack and Module Cells Section 3.1 provided a description of the spent fuel storage rack.Using the data of Section 3.0, analytic models were created in both SCALE-PC and DIT to perform the quantitative evaluations necessary to demonstrate the effective multiplication is: 1)less than unity with zero boron present in the pool, and 2)less than or equal to 0.95 when credit is taken for soluble boron.Applicable biases to be factored into this evaluation are: 1)the methods bias deduced from the validation analyses of pertinent critical experiments (described in Section 2), and 2)any reactivity bias, relative to the reference analysis conditions, associated with operation of the spent fuel storage pool over the temperature range of 50 to 155 F.A second allowance is based on a 95/95 confidence level assessment of tolerances and uncertainties.

Included in the summation of variances are the following:

a)the 95/95 confidence level methods variance, b)the 95/95 confidence level calculational uncertainty, c)tolerance due to enrichment uncertainty, d)tolerance due to UO2 stack density, e)tolerance due to uncertainty in L-insert and module wall thickness, tolerance due to uncertainty in positioning the fuel assembly in the storage cell, g)tolerance due to storage cell ID and pitch, and h)the 95/95 confidence level assessment of calculated CEA worth.Items a)and b)are based on the methods validation analyses.Item h)is based on comparisons of CEA worth measurements and analyses on operating reactors as well as excellent agreement between DIT and KENO-Va calculations of CEA worth.For item c), the uncertainty in enrichment is taken to be&.05 w/o.Assessments of the magnitude of change in multiplication factor per 0.05 change in enrichment may be based on either infinite arrays of a given type of storage cells or a change in the overall storage rack effective multiplication factor for a change of 0.05 w/o in the equivalent enrichment of a given fuel assembly type.The criticality analysis evaluation is done at an enrichment of 4.50 w/o, i.e.at the upper tolerance limit on enrichment, for added conservatism.

Stack density was taken to be the key variable for item d)since this variable is most pertinent to measured parameters in the fuel manufacturing operation.

Pellet stack column weight and stack length are the typical measured parameters for each fuel rod and are carried forward into the as-built fuel evaluations.

A value for the uncertainty in this parameter of 2%was assumed;this magnitude is conservative compared to observed history of variations in this parameter.

iL NIk/%ISIS 15 CENP0487 For item e), the following approach was taken.The tolerances on the thickness of the 0.135 inch thick sheet stock employed for the monolith wall structure and the 0.188 inch thick sheet stock employed for the L-insert were taken to be+0.005 inches based on the manufacturing drawings.For item f), infinite cell arrays of both Region I and II type cells were examined.The fuel assemblies were normally assumed to be centered within each storage cell type for the nominal calculations.

When the fuel assembly was moved diagonally off-center in both Region I and II type cells, the array reactivity decreased.

Consequently, the tolerance for this parameter was taken to be zero.For item g), the tolerance due to storage cell KD and pitch were evaluated in a combined manner using the following approach.Any assessment of tolerances on cell pitch and ID due to tolerances on individual components would appear to be of little relevance due to the method of assembly of the monolithic structures.

It was for this reason that overall dimensional tolerances were imposed on the finished modules.Since the finished modules were found to meet the specified dimensional tolerances, these values were taken to be the more relevant measure of the dimensional variations within a given module.If one takes the nominal overall dimension and subtracts the thickness of the two peripheral strips as well as the thickness of one module box wall, a nominal cell pitch of 8.999 inches is obtained for each cell in each module type.The nominal cell ID is 8.999-0.135 or 8.864 inches.The min/max ID of the module cell would be best approximated by the tolerance on the internal dimension of the 0.135 inch thick angle plates, viz., min/max ID=8.74/8.920 inches.For the Region I cells, the nominal dimensions are determined as follows.The external dimensions of the L-insert are listed as 8.740+0.000/-0.050 inches.Thus, a nominal dimensioned L-insert would have some clearance when inserted into a nominal module cell.The nominal ID of the Region I cell was taken as the average of the min/max ID determined by assuming the L-insert was displaced first to the left and then to the right extremes in the nominal module cell.The nominal Region I cell was evaluated as 8.56 inches.The minimum ID of the Region I storage cell is defined as the minimum overall dimension of the L-insert minus the height of the dimple and thickness of the L-insert wall at their low'er tolerance level.The minimum module cell ID is taken to be the minimum internal dimension of the angle plates or square elements formed from the 0.135 inch thick plate stock, viz., 8.740 inches.4.2 K.fi Evaluation at Zero Soluble Boron The full fuel assembly storage rack was modeled in KENO-Va using the fuel assembly storage pattern of Section 3.2.This KENO calculation served to establish a reference multiplication factor at 50'F with zero soluble boron in the pool water.The effective multiplication factor for the system is evaluated for three assumed initial starting neutron source distributions:

(1)a uniform distribution over Region II, (2)a uniform distribution over Region I, and (3)a uniform distribution over both Regions I and II.The K,ir results are as follows.A i%IN P%ISIN Source Region II Region I Regions I+II 0.97001 2 0.00052 0.96590+0.00054 0.96731+0.00050 16 0

CENP0487 The highest value of K,ir is obtained with a starting neutron distribution in Region II;this value is taken to be the nominal value for the zero boron condition since it is the most conservative interpretation of the data.In reality, the relevant source distribution for a spent fuel storage rack is one that is characteristic to the spontaneous fission distribution and this distribution would favor the more highly burned fuel region of the storage rack.For purposes of evaluating tolerances and uncertainties associated with enrichment and UO2 stack density, perturbations in these parameters were made in the full pool KENO model whereas, for cell wall thicknesses and cell ID and pitch, an infinite array of Region I or II storage cell types was used.The CEA worth allowance is evaluated by calculating the CEA worth in a full pool KENO calculation and multiplying by a conversion factor to derive the 95/95 confidence level uncertainty in CEA worth.Table 4 lists the derived quantities and the margin to unity for the no soluble boron condition.

4.3 K,ii Evaluation for Soluble Boron Credit To determine the amount of soluble boron required to maintain the effective multiplication factor less than or equal to 0.95, a KENO calculation of the full storage rack is employed to establish a nominal reference multiplication factor at 50'F.The calculation of biases and tolerances and uncertainties followed the same procedures as for the no soluble boron condition.

Table 5 lists the derived quantities and the margin to 0.95 for 350 ppm soluble boron.The final soluble boron requirement is the summation of the soluble boron credit requirements determined in steps 2, 3, and 4 of Section 1.These requirements are stated by the following equation.where: SBCToTAi.=SBCg5as+SBCRa+SBCpA SBCToTAi.=total soluble boron credit requirement (ppm).SBCeses soluble boron credit required for 95/95 K,ir to be less than or equal to 0.95 (ppm).SBCRa soluble boron credit requirement required for reactivity equivalencing methodologies (ppm).SBCpA soluble boron credit required for K,ir to be less than or equal to 0.95 under accident conditions (ppm).The total soluble boron credit requirement along with the storage configuration specified in the no soluble boron 95/95 K,ir calculation shows that the fuel rack K,ir will always be less than or equal to 0.95.Furthermore, the no soluble boron 95/95 K,ir storage configuration will ensure that K,ii remains less than 1.0 with no soluble boron in the spent fuel pool.17 CENP0487 0 4.4 Reactivity Equivalencing Reactivity equivalencing is a useful strategy for defining the conditions under which fresh, burned, and shimmed fuel assemblies are interchangeable on an overall reactivity basis;other characteristics of the resulting lattice arrangement may differ.This strategy is used to translate the array of fuel assemblies of difFering enrichments defined for the zero soluble boron condition in a given spent fuel rack into an array of burned fuel assemblies of difFering initial enrichments, decay times, and possible initial burnable poison compositions.

4.4.1 Burnup and Decay Time Reactivity Equivalencing Section 3.2, above, defined the enrichment levels of the various fuel assemblies arranged in the spent fuel storage rack under the zero soluble boron condition.

To establish a reactivity equivalence between, for example, a depleted unshimmed fuel assembly having a uniform initial UO2 enrichment of 4.5 w/o U-235 and the 1.3 w/o U-235 fuel assembly stored in Region II of.the storage rack representation for the zero soluble boron condition, two sets'of data are generated within the environment of a Region II storage cell.First, the K of the Region II storage cell containing the 1.3 w/o U-235 fuel assembly was calculated by KENO under the appropriate coolant temperature and soluble boron levels.Next, the K of the Region II storage cell was calculated by KENO for fuel nuclide compositions appropriate to various fuel assembly burnup levels for fuel assemblies of difFerent initial fresh fuel enrichments and under the same coolant temperature and soluble boron level.The latter burned fuel nuclide compositions were generated by a DIT simulation of an operating reactor.Conservative fissile nuclide compositions versus burnup were obtained by depleting the fuel at the reactor outlet moderator temperature.

The burnup at which the K of the depleted assembly matches the Kof the 1.3 w/o fresh assembly is the minimum required burnup.This process is repeated for each cell type present in the pool.If burnable poison shims are employed in the fuel assembly, this feature must also be factored into the initial composition and nuclide composition as a function ofburnup for this assembly.I-135 and Xe-135 decay Pm-149 decay into Sm-149 Np-239 decay into Pu-239 The no Xenon, peak Samarium and peak Pu-239 condition was used for the determination of the storage rack reactivity without credit for actinide decay.Subsequently, the decay of longer half-life nuclides comes into play, the most important of which is the decay of Pu-241 into Am-241.Pu-241 is a fissile nuclide which contributes several percent of positive reactivity at high burnup.Am-241, on the other hand, is mostly an absorber which has a negative reactivity component.

The half life of Pu-241 being 14.4 years, its decay over the lifetime of the pool storage is important and contributes to a decrease in K of the burned fuel bly in the storage cell environment.

This effect is favorable since it reduces the pool assem fL llIk P%NII 18 Following discharge of a given fuel assembly from the reactor, the decay of fission products and actinide nuclides within the fuel will induce a change in equivalent Kof the burned fuel assembly.Within the first few days after shutdown, a large increase in Kwill result from decay of the following nuclides:

CENP048T reactivity and may permit the transfer of decayed burned fuel assemblies into positions designated for higher burnup in the no soluble boron core representation.

Besides Pu-241, decay of all the actinides and fission products present in the DIT model was accounted for.Credit for actinide decay is used to reduce the minimum burnup required to meet the reactivity requirements.

Actinide decay efFects were calculated over a time interval of up to twenty years for an.initial feed enrichment range of up to 4.5 w/o U-235 in the following manner.First, isotopics in the depleted assembly were calculated by performing a DIT depletion under nominal operating conditions, but at the reactor outlet moderator temperature to maximize the conversion ratio and thus, the reactivity.

At selected time points, depleted isotopics were transferred into another DIT model which represents the geometry of a spent fuel pool cell, at 50'F and 0 PPM boron.This model was then decayed for 20 years, and the reactivity loss with time was translated into a burnup credit using the burnup vs.enrichment curves.Tables 6-1 to 6-4 summarize the derived burnup-enrichment pair data, including credit for actinide decay from 0 to 20 years.These minimum burnup data are tabulated for feed enrichments between 1.5 and 4.5 w/o U-235 and for each burned fuel position type in the spent fuel pool.These results are also plotted in Figures 11-15.4.4.2 Gadolinium Credit Reactivity Equlvalencing The St.Lucie Unit 2 maximum enrichment is currently set at 4.5 w/o.This enrichment is set in part by the spent fuel pool criticality analysis, which assumes a given loading pattern for the fresh fuel assemblies.

The criticality analysis was performed without credit for the burnable absorber reactivity hold-down of the fresh assemblies, that is, it was performed as if all fresh assemblies were unshimmed.

If burnable absorber reactivity hold-down is considered, then the fresh fuel enrichment can be increased until the assembly reactivity matches that of an unshimmed, 4.5 w/o assembly.The enrichment credit was determined for the following five diferent gadolinium burnable absorber (BA)loadings.No of Gad BA Rods 4 8 12 8 16 Gad loading (w/o)4 4 4 6 6 The reactivity gain due to the incr'eased enrichment allowed by the gadolinium credit must be equal to or less than the reactivity hold-down of the gadolinium burnable absorber such that the fresh assembly reactivity never exceeds the fresh unshimmed 4.5 w/o assembly reactivity.

The.reactivity hold-down of the burnable absorber depends on the number and loading of the gadolinium bearing rods, and also of the axial cutback of the burnable absorber.The short neutron difFusion length in cold water magnifies the importance of the cutback.19 CENPD487 The initial reactivity hold-down and depletion effects were first evaluated in a two dimensional geometry, neglecting the effect of the cutback.The initial reactivity hold-down of the lightest Gadolinium loading (4 rods at 4 w/o Gd)is larger than the reactivity gain obtained by increasing the enrichment from 4.5 w/o to 5.0 w/o.The enrichment of the gadolinium assemblies was assumed to be 5.0 w/o, consistent with guidance in the SER contained in Reference 2.The K of the poisoned section of a gadolinium shimmed fuel assembly is lower than that of a uniformly enriched, 5 w/o fuel assembly.As the gadolinium depletes, the assembly Kwill approach that of a comparably burned 5.0 w/o unshimmed assembly and, at a given burnup, will cross the reactivity rundown of a 4.5 w/o unshimmed assembly.For each gadolinium assembly type, a depletion was performed at 5.0 w/o, and the Kvalues compared to those of an unshimmed 4.5 w/o assembly depletion.

The assembly characteristics are summarized in Table 7.The U-235 content of the gadolinium shim rods is reduced during fabrication of these rods to ensure that they are not limiting at any time during the cycle.The reduced gadolinium pin density reflects the uranium displacement by gadolinium.

The fuel assembly depletion calculations were performed with the multigroup transport code DIT.The K'f the various Gadolinium fuel assembly types is plotted in Figure 16.A 5.0 w/o unshimmed assembly is included.To cover possible uncertainties in the gadolinium worth and depletion rate, the gadolinium reactivity hold-down was reduced by 10%for conservatism, leading to conservatively low equivalent burnup.Beyond 20,000 MWD/T, the gadolinium is effectively fully depleted and the reactivity offset is due to the slight difference in assembly average enrichments and gadolinium residual worth.Figure 17 presents the same data at an expanded scale.Because the K of a gadolinium assembly enriched at 5.0 w/o U-235 is always lower than that of a fresh 4.5 w/o unshimmed assembly, the gadolinium assemblies can always be stored in the locations reserved for fresh, 4.5 w/o unshimmed assemblies.

The impact of the gadolinium cutback was determined by performing a three-dimensional KENO calculation of Region I, assigning a 10.5 inch cutback to the 5.0 w/o fresh assembly, and comparing the reactivity to that of a reference case without gadolinium, with a fresh fuel enrichment of 4.5 w/o.At the lowest gadolinium loading (4 rods at 4 w/o Gd), the reactivities of the poisoned case and of the reference case are equal, indicating an exact compensation between the negative worth of the cutback burnable absorber and the positive worth of the increased enrichment.

At the heaviest gadolinium loading (16 rods at 6 w/o Gd), the reactivity of Region I is reduced by 0.0014~below the reference case.The small sensitivity of Region I to reactivity perturbations in the fresh fuel is consistent with earlier findings that the importance of the fresh fuel is small (every other assembly is rodded and the assemblies are separated by a row of water cells).The impact of the gadolinium cutback for depleted assemblies is negligible because of the high burnup requirement imposed by the burnup vs enrichment curves, and by the large conservatism included in the axial burnup distribution effects.AL i%I 20 CENP0487 As shown in Figure 16, the burnup dependence of Kis nearly parallel for all assembly types beyond 20,000 MWD/T.The off-set between the various gadolinium assemblies and a 4.5 w/o unshimmed assembly is: Gad Assembly Burnup OF-set 4000 3800 3600 3600 3000 Burnup OF-set Burnup OF-set@25,000 MWD/T@45,000 MWD/T 4 Gad 4 w/o 3650 4000 8 Gad 4 w/o 3400 3800 12 Gad 4 w/o 3250 3600 8 Gad 6 w/o 3250 3600 16 Gad 6 w/o 2700 3000 The burnup of a Gadolinium fuel assembly enriched at 5.0 w/o U-235 must be reduced by the following amount before applying the burnup equivalencing of Tables 6-1 to 6'-4 for a 4.5 w/o unshimmed assembly: Gad Assembly 4 Gad 4 w/o 8 Gad 4 w/o 12 Gad 4 w/o 8 Gad 6 w/o 16 Gad 6 w/o 4.4.3 Soluble Boron Credit for Uncertainties in Reactivity Equivalencing Soluble boron credit for reactivity equivalencing includes two efFects: (1)an allowance for possible uncertainties in the analytical techniques employed to define the burned composition of the fuel assembly, and (2)an allowance for possible uncertainties in inferring the burnup of a given fuel assembly removed from the core.The allowance for the former uncertainty is taken to be dX=0.005 absolute reactivity per 30,000 MWD/T and, for the latter inferred burnup uncertainty, 5%in burnup.The boron worth for these allowances is evaluated under the highest burnup imposed on a fuel assembly having an initial enrichment of 4.5 w/o U-235.4.5 Axial Burnup Distribution Several reactivity efFects are associated with the axial burnup distribution.

The higher fissile content near the top of the assembly increases its Kand results in a top peaked flux distribution, while the steep flux gradient resulting from the top peaked flux distribution increases the axial leakage and reduces the assembly K,ir.These eFects were evaluated by comparing the results of 2-and 3-dimensional calculations of assembly reactivity at 68'F, 0 PPM boron, and no xenon as a function ofburnup.The 2-dimensional model was depleted at the outlet moderator AL NIk P%lNIN 21 CENPD487 temperature, while the 3-dimensional model was depleted under more realistic axial temperature distributions.

It was found that the conservatism resulting from the 2-dimensional depletion at the outlet temperature outweighs by far any other axial effect.Therefore the assembly reactivities used in the burnup equivalencing are conservatively high, leading to conservatively high burnup values.4L lNIN/%IN tN 22 CENPD487 5.0 Postulated Accidents There is a variety of accidents that can be postulated to occur in connection with operations in the vicinity-of the spent fuel pool.Fuel assembly drop accidents, for example, can usually be shown to not result in any significant increase in reactivity of the spent fuel pool sy'tem.The design of the structures in the spent fuel pool and interfacing systems are such that, in combination with plant administrative controls, they preclude the placement of a fuel assembly into areas not designated as intended storage locations.

At St.Lucie Unit 2, position limit switches on the spent fuel handling machine prohibit placement of fuel outside the region defined by the storage rack modules.The presence of the spent fuel handling machine interlock zones is ensured prior to each fuel handling campaign.Additionally, because the rack modules are free standing without attachment to the floor or walls of the spent fuel pool, no structure exists external to the rack module to support, in the vertical position, any fuel assembly which could be postulated to be placed there.A fuel assembly drop accident resulting in an assembly lying on top of the modules will not result in any significant increase in system K,ir because of the large separation distance between the active volume of the fuel assemblies within specified storage locations and the fuel assembly lying atop the modules.The loss of pool cooling accident has the potential of raising the temperature of the pool coolant to a boiling condition.

The consequence of this postulated accident on the system K,ir was conservatively estimated by evaluating the dXin an array of both Region I and Region II storage cells containing a 4.5 w/o enriched fresh assembly burned to 50,000 Mwd/t.The magnitude of the change in K was less than 0.0040 and 0.0066 for the zero and 350 ppm soluble boron conditions, respectively, for a temperature change between 155 to 240 degrees Fahrenheit.

To assess the consequence of postulated fuel assembly misload accidents, a variety of scenarios were examined.These scenarios all involved the misplacement of an unshimmed fresh 4.5 w/o enriched fuel assembly without a CEA inserted.This assembly was placed in three possible types of positions:

1)a position designated for a 4.5 w/o fresh fuel assembly containing a CEA, 2)a position designated for the more highly burned fuel (1.3 w/o equivalent enrichment) in Region II, and 3)selected water cell locations.

The latter choice served to quantify the benefit of these isolation cells by maximizing the coupling between subarrays of 1.82 w/o and, or 4.5 w/o fuel assembly locations in Region I.The largest dX observed for postulated misload accidents was 0.1016 for a type 3 misload position.Cases examined for a type 1 misload position resulted in dX values of less than 50%of that for the type 3 misload whereas, for a type 2 misload position, the bK was approximately 75%of that for the type 3 misload position.The soluble boron requirement for a type 3 misload was deduced to be 746 ppm under the assumption that a fresh fuel assembly could be loaded into a water cell.Should physical devices or other means be implemented to preclude the misloading of a fresh 4.5 w/o enriched fuel assembly, or its equivalent, this boron requirement for misload accidents could be reduced by roughly 25%.Since the magnitude of reactivity insertion for the postulated fuel assembly misload accidents is much greater than that resulting from the loss of pool cooling event, the latter event does not Al Nfl/%OSIS 23 CENPD487 require a greater soluble boron level in the pool than the most adverse misload accident.Thus, under the double contingency criterion, an incremental soluble boron allotment of 746 ppm for accidents is sufhcient for the fuel assembly misload and loss of pool cooling events.AL NI 24 CENPD487 6.0 Soluble Boron Credit Summary Spent fuel pool soluble boron is employed in this criticality safety analysis to offset the reactivity allowances for calculational uncertainties in modeling, storage rack fabrication tolerances, and fuel assembly design tolerances, as well as postulated accidents.

The total soluble boron requirement based on the components is designated as SBCToTAL in the equation given in Section 4.3.The components of the latter quantity are summarized as follows.SBC9se5 SBCRp=350 ppm=170 ppm SBCPA=~746 m SBCTpTAi.=1266 ppm This boron requirement is less than the soluble boron concentration required to be present in the St.Lucie Unit 2 spent fuel pool by Technical Specification 5.6.1.AL INES P%NIl 25 0

CENP04&7 References l.2.3.4 5.6.7.8.9.10 11 1 12 13 Newmyer, W.D.,"Westinghouse Spent Fuel Rack Criticality Analysis Methodology," WCAP-14416-NP-A, Rev.01, November 1996.Letter from T.E.Collins, USNRC to T.Greene, WOG,"Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Methodology (TAC NO.M93254)", October 25, 1996.Code of Federal Regulations, Title 10, Part 50, Appendix A, Criterion 62,"Prevention of Criticality in Fuel Storage and Handling".

U.S.Nuclear Regulatory Commission, Standard Review Plan, Section 9.1.2, NUREG-0800, July 1981."St.Lucie Unit No.2 Updated Final Safety Analysis Report," Amendment 10, August 7, 1996, Florida Power and Light Company."SCALE 4.3-Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers," NUREG/CR-200; distributed by the Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee."The ROCS and DIT Computer Codes for Nuclear Design," CENPD-266-P-A, April 1983, Combustion Engineering, Inc.M.N.Baldwin et al.,"Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel;Summary Report," BAW-1484-7, July 1979.S.R.Bierman and E.D.Clayton,"Critical Experiments with Subcritical Clusters of 2.35 wt%'U Enriched UO2 Rods in Water at a Water-to-Fuel Volume Ratio of 1.6," NUREG/CR-1547, PNL-3314, July 1980.S.R.Bierman and E.D.Clayton,"Criticality Experiments with Subcritical Clusters of 2.35 and 4.31 wt%U-Enriched UO2 Rods in Water with Steel Reflecting Walls", Nuclear Technology, Vol.54, pg.131, August 1981."International Handbook of Evaluated Criticality Safety Benchmark Experiments," Nuclear Energy Agency and Organization for Economic Cooperation and Development, NEA/NSC/DOC(95)03IV, LEV-COMP-THERM-001, 2, 3, 4 (Rev.0, 3/31/95);LEU-COMP-THERM-010, 017 (Rev.0, 8/31/96)W.Marshall, et.al.,"Criticality Safety Criteria", TANS Vol.35, pg.278, 1980.D.B.Owen,"Factors for One-Sided Tolerance Limits and for Variables Sampling Plans", SCR-607, Sandia Corp.Monograph, March 1963.lL II II PENIS 26 CENP0487Table 1 Summary of Calculational Results for Cores X Through XXI of the B&W Close Proximity Experiments Core RUN No.X 2348 XI 2355 XI 2359 XI 2360 XI 2361 XI 2362 XI 2363 XI 2364 XII 2370 XIII 2378 XIIIa2423 XIV 2384 XV 2388 XVI 2396 XVII 2402 XVIII 2407 XIX 2411 XX 2414 XXI 2420 0.99610+0.00084 1.00049+0.00080 0.99884 2 0.00077 1.00315+0.00081 0.99831+0.00080 1.00060+0.00078 0.99957+0.00078 1.00246+0.00080 0.99990+0.00082 0.99754+0.00089 0.99575+0.00087 0.99465 2 0.00086 0.99158+0.00084 0.99230+0.00088 0.99478 2 0.00079 0.99440+0.00083 0.99821+0.00081 0.99498+0.00082 0.99318 2 0.00094 Plate" Type none SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 SS-304 B/Al 8/Al B/Al B/Al B/Al B/Al B/Al B/Al B/Al B/Al Spacing"'1 1 1 1 1 I 1 2 1 1 1 1.2 1 2 1 2 3 (a)-metal separating unit assemblies (b)-spacing between unit assemblies in units of fuel rod pitch AL ll wl P%IN1I 27 CENPD487 Table 2 Summary of Calculational Results for Selected Experimental PNL Lattices, Fuel Shipping and Storage Configurations EXPT.No.COMMENTS 043 0.99787+0.00106 044 1.00104+0.00102 045 0.99955+0.00101 046 0.99960+0.00103 061 0.99792+0.00099 062 0.99628+0.00096 064 0.99696+0.00103 071 0.99970+0.00101 079 0.99463+0.00102 087 0.99423+0.00099 093 0.99787+0.00098 Uniform rectangular array, no poison CC<c 2 x 2 array of clusters, no poison 2 x 2 array of clusters, 0.302 cm SS-304 cross 2 x 2 array of clusters, 0.485 cm SS-304 cross 2 x 2 array of clusters, cross of 0.3666 g boron/cm 2 x 2 array of clusters, cross of 0.1639 g boron/cm 2 x 2 array of clusters, cross of 0.1425 g boron/cm N NIk ERNIE 28 CENP0487Table 3 Fuel Parameters Employed in Criticality Analysis for St.Lucie Unit 2 Spent Fuel Storage Rack Number of Fuel Rods per Assembly UOz Pellet OD (in)Zr-4 Clad Tube OD (in)Clad Tube Wall (in)Nominal UO2 Stack Density (g/cc)Fuel Rod Pitch (in)'EA Guide Tube OD (in)CEA Guide Tube ID (in)Number of Guide Tubes per Fuel Assembly 236 0.3255 0.382 0.025 10.31 0.5065 0.980 0.900 4~8L I%IN PINES 29 CENPD487 Table 4 St.Lucie 2 Unit 2 Spent Fuel Rack K,<with No Soluble Boron K-eFective Nominal Reference Value 0.97001 alculational and Methodolo Biases Methodology Pool Temperature (50'F to 155'F)Total 0.00259 0.00375 0.00634 Tolerances and Uncertainties UO2 Enrichment (0.05 w/o)UO2 Stack Density (2%)Cell Wall Thickness (0.005 in)Cell ID 2 Pitch Asymmetric FA Position 95/95 CEA Worth 95/95 Methodology Uncertainty 95/95 Calculational Uncertainties 0.01380 0.00381 0.00293 0.01342 0.00000 0.00545 0.00639 0.00220 Total Uncertainties (statistical)

TOTAL 0.02166 0.99801 lL II 30 CENp0487 Table 5 St.Lucie Unit 2 Spent Fuel Rack K,<with Soluble Boron Credit K-efFective Nominal Reference Value 350 PPM Soluble Boron Calculational and Methodolo Biase 0.91497 Methodology Pool Temperature (50'F to 155'F)Total Tolerances and Uncertainties 0.00259 0,00560 0.00819 UOz Enrichment (0.05 w/o)UO2 Stack Density (2%)Cell Wall Thickness (0.005 in)Cell ID Ec Pitch Asymmetric FA Position 95/95 CEA Worth 95/95 Methodology Uncertainty 95/95 Calculational Uncertainties 0.01530 0.00474 0.00225 0.01675 0.00000 0.00527 0.00639 0.00220 Total Uncertainties (Statistical) 0.02481 Total 0.94797 31

CENPD487 Table 6-1 St.Lucie Unit 2 Spent Fuel Rack Tabulation of Burnup vs Initial Enrichment and Decay Time Years Region II, 1.3 wlo Enrich 4.5 4.4 4.3 4.2 4.1 4.0 3.9 3.8 3.7 3.6 3.5 3.4 3.3 3.2 3.1 3.0 2.9 2.8 2.7 2.6 2.5 2.4 2.3 2.2 2.1 2.0 1.9 1.8 1.7 1.6 1.5 r 0 46697 45943 44940 43785 42549 41280 40011 38758 37527 36319 35126 33939 32747 31539 30309 29049 27755 26429 250T3 23692 22294 20884 19469 18051 16628 15187 13708 12153 10467 8574 6370 1 45865 45060 44057 42931 41739 40517 39292 38075 36872 35681 34497 33313 32121 30912 29681 28423 27138 25825 2448T 23130 21760 20381 18998 17613 16221 14811 13361 11839 10194 8358 6241 2 45071 44219 43216 42118 4866 39790 38606 37423 36246 35072 33898 32717 31525 30315 29084 27829 26551 25250 23930 22596 21252 19903 18552 17198 15836 14456 13037 11548 9944 8165 6129 3 44314 43419 42415 41343 40231 39096 37952 36802 35650 34492 33327 32150 30958 29747 28516 27264 25993 24704 23401 22088 20769.1S449 18127 16804 15473 14123 12734 11278 9715 799Q 6033 4 43592 42657 41654 40606 39531 38436 37329 38211 35082 33940 32T84 31611 30419 29208 27977 26728 25463 24185 22897 21605 20311 19017 17725 16431 15130 13809 12450 11026 9503 l831 5949 39906 38865 37809 34541 33415 32267 31098 39241 38232 37212 36173 35113 34027 32915 31776 30610 28695 27465 26218 24959 23692 22419 21146 19875 18607 17343 16077 14805 13512 12182 10791 9306 7685 5876 28208 26978 25734 24481 23223 21965 20710 19461 18218 16980 15742 14497 13232 11931 10571 9123 7551 5811 5 6 42906~42253 41932 41244 40242.7 41633 40590 39589 38609 37632 36645 35638 34604 33539 32441 31309 30147 28958 27745 26516 25274 24027 22778 21534 20297 19068 17849 16636 15424 14205 12967 11693 10363 8952 7427 5753 8 41045 39970 3896S 38009 37062 36107 35130 34122 33076 31990 30866 29707 28519 27306 26077 24838 23595 22356 21124 19904 18695 17498 16309 15122 13929 12716 11468 10167 8790 7311 5700 9 40487 39381 38381 37440 36521 35597 34649 33664 32636 31563 30446 29290 28102 26889 25660 24423 23186 21955 20736 19531 18341 17165 15999 14S36 13666 12478 11255 9981 8637 7202 5651 10 39958 38824 37824 36901 36008 35'113 34192 33230 32219 31157 30047 28894 27706 26493 25264 24029 22797 21574 20367 19177 18005 16849 15705 14564 13417 12251 11053 9805 8492 7098 5605'11 39458 38295 37296 36389 35522 34655 33759 32819 31824 30773 29669 28519 27331 26118 24889 23656 2242S 21213 20016 18841 17686 16550 15426 14307 13181 12036 10860 9637 8354 6998 5561 12 389S4 37794 36795 35905 35062 34221 33350 32429 31450 30409 29311 28163 26975 25761 24532 23301 22078 20870 19684 18522 17383 16265 15161 14062 12956 11832 10676 9476 8221 6903 5518 13 38536 37320 36321 35446 34626 33810 32962 32061 31096 30064 28971 27825 26638 25423 24194 21745 20544 19368 18219 17096 15996 14910 13829 12743 11638 10501 9323 8094 6812 5477 14 38111 36871 35872 35012 34213 33420 32594 31711 30760 29738 28649 27505 26318 25102 23873 21430 20236 19069 17932 16824 15739 14671 13609.12540 11452 9177 7973 6l24 5437 15 37710 36446 35447 34600 33822 33052 32246 31381 30442 29428 28344 27202 26014 24T98 23569 21130 19942 18785 17660 16565 15496 14445 13399 12347 11276 10176 9037 7857 6639 5398 16 37329 36043 35044 34210 33452 32T02 31917 31067 30141 29135 28055 26914 25727 24510 232S1 22055 20846 19664 18515 17401 16320 15266 14230 13200 12164 11109 10025 8904 7746 6558 5360 17 36969 35662 34662 33841 33101 323T1 31604 30770 29855 28857 27781 26641 25454 24236 23007 21783 20577 19401 18260 17156 16087 15047 14025 13011 1199Q 10950 9882 8778 7641 6482 5324 18 36626 35299 34300 33490 32767 3205T 31307 30488 29584 28593 27521 26383 25195 23977 22748 21525 20322 19151 18017 16923 15866 14839 13832 12832 11825 10800 9747 8660 7544 6411 5292 19 36300 34956 33956 33158 32451 31758 31025 30219 29326 28341 27274 26137 24950 23731 22502 21280 20080 18914 17787 16703 15656 14642 13648 12662 11669 10659 9620 8550 7453 6347 5263 20 35988 34629 33630 32842 32151 31474 30756 29964 29080 28102 27039 25904 24717 23498 22269 21049 19851 18690 17570 16493 15457 14454 13473 12500 11522 10526 9503 8449 7372 6290 5240 32 CENPD487 Table 6-2 St.Lucie Unit 2 Spent Fuel Rack Tabulation of Burnup vs tnitial Enrichment and Decay Time Region ll 1.5 wlo Years Enrich 0 1 4.5 39660 39059 4.4 38263 37694 4.3 37043 36486 4,2 35940 35385 4.1 34908 34349 4.0 33909 33345 3.9 32915 32347 3.8 31903 31336 3.7 30859 30299 3.6 29777 29228 3.5 28651 28120 3.4 27482 26973 3.3 26276 25T92 3.2 25038 245S1 3.1 23776 23347 3.0 22499 22096 2.9'1214 20835 2.8 19928 195S9 2.7 18644 18304 2.6 17366 17039 2.5 16088 15773 2.4 14804 14500 2.3 13500 13210'2.2 12154 11885 2.1 10740 10503 2.0, 9219 9034 4 2 3 4 38484 37935 37410 37149 36628 36131 35954 35445 34959 34854 34347 33862 33815 33304 32816 32806 32291 31799 31805 31287 30793 30795 30279 29786 29764 29254 28767 28705 28206 27730 27613 27130 26669 26487 26025 25584 25331 24891 24472 24146 23731 23337 22938 22549 22179 21712 21348 21002 20475 20133 19809 19230 18908 18603 17981 17676 17388 16731 16439 16164 15477 15197 14934 14215 13947 13695 12938 126S4 12445 11634 11399 11180 10284 10081 9891 8865 8710 8568 5 36910 35657 34495 33400 32351 31330 30323 29317 28304 27277 26231 25164 24074 22961 21827 20673 19501 18314 17114 15904 14685 13457 12221 109T5 9715 8436 6 7 8 9 36433 35980 35550 35141 35205 34775 34367 33979 34054 33634 33235 32856 32961 32542 32144 31766 31909 31488 31087 30707 30884 30460 30056 29673 29874 29448 29043 28658 28870 28445 28041 27658 27862 27443 27044 26665 26845 26435 26045 25675 25814 2541 T 25040 24682 24764 24385 24024 23681 23695 23335 22993 22668 22605 22265 21943 21637 21493 21175 20873 20587 20360 20064 19782 19514 19209 18931 18668 18418 18040 17779 17532 17298 16855 16610 16376 16155 15658 15424 15203 14993 14449 14227 14016 13S16 13233 13021 12820 12630 12010 1181 0 11622 11443 10782 10600'0428 10265 9550 9394 9248 9109 8314 8199 8091 7989 10 11 34753 34386 33611 33263 32496 32155 31408 31069 30347 30005 29310 28965 28293 27947 27294 26949 26306 25965 25324 24991 24342 24020 23356 23047 22359 22066 21347 21071 20315 20056 19260 19018 18180 17955 17075 16863 15945 15746 14794 14605 13626 13446 12450 12278 11273 11112 10111.9964 8977 8852 7892 7800 12 3403S 32933 31833 30747 29681 28639 27619 26622 25642 24674 23714 22754 21788 20809 19811 18789 17740, 16663 15556 14425 13275 12115 10959 9824 8733 7712 13 33709 32620 31527 30442 29374 28329 27309 26312 25335 24375 23424 22477 21524 20561 19578 18572 17537 16472 15376 14254 13112 11960 10813 9691 8619 7627 14 15 33397 33101 32324 32044 31237 30963 30153 29880 29084 28809 28036 27759 27014 26735 26018 25739 25045 24770 24091 23822 23150 22889 22213 21964 21275 21038 20325 20102 19357 19148 18365 18169 17344 17160 16290 16118 15205 15043 14092 13938 12957 12810 11812 11672 10674 10542 9564 9443 8510 8406 7546 7468 16 17'18 32822 32556 32304 31778 31526 31287 30704 30457 30223 29621 29375 29142 28548 28300.28065 27495 27246 27009 26471 26220 25981 25476 25225 24987 24509 24262 24027 23567 23325 23095 22642 22408 22186 21727 21503 21290 20813 20600 20398 19890 19689 19499 18949 18761 18583 17984 17807 17641 16987 16822 16666 15955 15800 15654 14889 14743 14604 13T91 13652 13520 12670 12537 12411 11539 11412 11292 10416 10297 10184 9328 9219 9116 8307 8214 8126 7394 7324 7259 19 20 32064 31835 31059 30841 30000 29787 28919 28707 27841 27627 26783 26567 25755 25538 24761 24546 23803 23591 22877 22669 21975 21774 21088 20896 20206 20024 19318 19147 18414 18253 17483 17333 16518 16378 15515 15384 14473 14349 13396 13279 12293 12181 11179 11073 10078 9978 9019 8930 8044 7968 7200 7146 33 CENPD487 Table 6-3 St.Lucie Unit 2 Spent Fuel Rack Tabulation of Burnup vs Initial Enrichment and Decay Time Region I,'1.4 wlo Enrich 4.5 44 4.3 4.2 4.1 4.0 3.9 3.8 3.7 3.6 3.5 3.4 3.3 3.2 3.1 3.0 2.9 2.8 2.7 2.6 2.5 2.4 2.3 2.2 2.1 2.0 1.9 1.8 1.7 1.6 1.5 Years 0 43355 42336 41268 40163 39029 37875 36705 35523 34333 33134 31928 30714 29492 28261 27018 25762 24492 23204 21897 20569 19217 17839 16432 14994 13522 12013 10462 8866 7219 5514 3746 1 42637 41663 40614 39512 38372 37208 36030 34843 33652 32458 31264 30067 28867 27660 26445 25218 23977 22718 21440 20140 18815 17463 16082 14670 13224 11742 10220 8653 7036 5360 3617 2 41951 41022 39992 38891 37745 36572 35385 34192 33001 31814 30631 29450 28271 27088 25899 24700 23487 22257 21006 19733 18433 17107 15751 14365 12946 11491 9998 8461 6873 5225 3505 3 41298 40412 39399 38300 37148 35965 34769 33572 32381 31199 30027 28863 277Q3 26544 25380 24208 23022 21819 20595 19346 18072 16770 15439 14078 12684 11257 9793 8285 6727 5107 3409 4 40675 39830 38835 37738 36579 35387 34183 32981 31790 30614 29452 28303 27163 26026 24887 23740 22581 21403 20204 18980 17729 16451 15144 13807 12439 11039 9603 8125 6595 5002 3325 5 4X$2 39278 38298 37202 36037 34836 33625 32418 31227 30056 2S905 27771 26649 25534 24418 23296 22161 21008 19832 18632 17404 16149 14865 13552 12209 10835 9427 7977 6476 4908 3252 6 39519 38752 37788 36694 35522 34313 33094 31883 30692 29526 2S384 27264 26161 25066 23973 22874 21762 19480 18302 17096 15863 14601 13310 11992 10644 9262 7840 6366 4824 3187 7 38983 38253 37303 36210 35033 33815 32589 31374 30183 29022 27890 26783 25696 24621 23549 22472 21383 20276 19146 17989 16804 15591 14350 13082 11787 10463 9108 7712 6265 4747 3129 8 9 10 38474 37992 37534 37778 37328 36901 36842 36405 35991 35751 35315 34902 34568 34127 33709 33343 32895 32470 32110 31655 31224 30891 30432 29997 29700 29241 28807 28543 28089 27658 27420 26974 26551 26326 25892 25481 25256 24837 24439 24199 23798 23417 23147 22765 22403 22091 21730 21386 21024 20682 20358 19938 19616 19311 18828 18527 18241 17692 17410 17142 16527 16264 16015 15334 15090 14859 14114 13889 13676 12866 12662 12468 11593 11410 11235 10293 10132 9979 8962 8824 8693 7592 7479 7371 6170 60S1 5996 4675 4608 4544 3076 3027 2981 11 37100 36496 35598 34509 33312 32067 30815 29585 28394 27249 26150 25090 24062 23056 22058 21060 20050 19022 17969 16888 15778 14639 13474 12283 11070 9833 8568 7268 5914 4483 2937 12 36688 36112 35225 34138 32937 31685 30428 29195 28004 26862 25769 24720 23705 22713 21732 20750 19758 18747 17711 16647 15553 14431 13281 12108 10912 9694 8449 7168 5836 4424 2894 13 36299 35748 34871 33786 32581 31323 30061 28825 27634 26495 25409 24369 23366 21422 19480 18486 17466 16418 15340 14233 13099 11941 10762 9561 8334 7073 5760 4367 2853 14 35930 35403 34537 33452 32244 30981 29714 28475 27283 26147 25067 24036 23044 22079 21128 20177 19217 18238 17234 16201 15137 14045 12925 11783 ,10619 9434 8225 6982 5687 4311 2813 15 35581 35077 34220 33136 31924 30656 29385 28143 26951 25818 24743 23721 22739 21787 19913 18968 18003 17014 15994 14945 13866 12761 11632 10483 9313 8121 6894 5616 4257 2774 16 35250 34767 33919 32837 31622 30349 29073 27828 26636 25506 24437 23422 22450 21509 2Q585 19662 18731 17781 16804 15799 14762 13696'2604 11489 10353 9199 8021 6811 5549 4206 2736 17 3%36 34475 33635 32554 31335 30058 28777 27530 26338 25210 24146 23138 22177 21247 20334 19425 18507 17569 16606 15613 14589 13535 12455 11353 10231 9090 7927 6733 5487 4158 2701 18 34638 34197 33366 32285 31063 297S1 28496 27247 26054 24929 23869 22870 21917 20998 20097 19200 18294 17369 16418 15437 14424 13382 12314 11224 10115 8988 7840 6660 5429 4114 2668 19 34355 33935 33111 32030 30805 29518 28229 26977 25784 24661 23607 22614 21671 20762 19&73 18987 18093 17180 16240 15270 14268 13237 12180 11102 10006 8893 7759 6594 5378 4076 2639 20 34086 33686 32870 31789 30560 29268 27974 26720 25527 24407 23358 22372 21438 20539 19661 18786 179Q3 17001 16072 15112 14120 13100 12054 10987 9904 8805 7686 6537 5335 4046 2616 CENPD487 Table 6Q St.Lucie Unit 2 Spent Fuel Rack.Tabulation of Burnup vs initial Enrichment and Decay Time Region I, 1.82 w/o Years Enrich 0 4.5 30444 4.4 29393 4.3 28377 4.2 27381 4.1 26393 4.0 25404 3.9 24407 3.8 23400 3.7 22379 3.6 21343 3.5 202S4 3.4 19231 3.3 18158 3.2 17076 3.1 15987 3.0 14893 2.9 13794 2.8 12691 2.7 11581 2.6 10460 2.5 9324 2.4 8164 2.3 6970 2.2 5728 2.1 4420 2.0 3025 1 2 30005 29587 28968 28562 27966 27573 26984 26605 26010 25644 25035 24683 24052 23714 23059 22733 22051 21739 21029 20730 19993 19707 18944 18672 17885 17627 16817 16574 15743 15514 14665 14451 13582 13384 12495 12314 11402 11238 10300 10153 9182 9054 8041 7930 6865 6773 5641 5566 4350 4292 2972 2928 3 4 29189 28809 28176 27807 27199 26842 26243 25&98 25295 24962 24346 24026 23390 23082 22423 22127 21441 21158 20446 20175 19436 19178 18414 18170 17383 17152 16344 161 27 15299 15096 14251 14063 13200 13028 12146 11990 11087 10947 10019 9896 8937 8830 7831 7741 6691 6617 5500 5442 4241 4198 2892 2862 5 28446 27456 26502 25569 24645 23720 22788 21845 20888 19917 18933 17938 16933'5921 14905 13887 12866 11844 10817 9782 8733 7660 6551 5391 4161 2836 6 7 8 28099 27768 27452 27120 26800 26495 26178 25868 25574 25256 24958 24674 24343 24055 23782 23429 23152 22889 22509 22242 21989 21577 21321 21078 20631 20387 20155 19672 19439 19217 18700 18479 18268 17717 17508 17309 16725 16528 16341 15727 15543 15368 14725 14554 14392 13720 13563 13414 12715 12572 12437 11707 11579 11457 10696 10582 10475 9676 9577 9485 8642 8558 8480 7585 7515 7451 6491 6436 6384 5345 5303 5264 4128 4098 4070 2814 2795 2777 9 27150 26204 252S3 24403 23521 22638 21748 20847 19934 19007 18068 17119 16163 15202 14238 13273 12309 11343 10374 9397 8406 7390 6336 5227 4044 2759 10 26863 25927 25027 24146 23274 22400 21519 20628 19724 18806 17878 16940 15994 15044 14092 13139 12187 11234 10278 9314 8335 7332 6290 5192 4018 2742 11 12 26589 26330 25664 25414 24773 24533 23902 23670 23039 22815 22174 21959 21302 21095 20419 20221 19524 19334 18616 18435 17697 17525 16769 16606 15833 15680 14894 14751 13953 13821 13012 12890 12071 11961 11130 11031 10186 10099 9235 9159 8268 8204 7277 7223 6246 6203 5158 5125 3994 3969 2725 2707 13 26085 25178 24305 23450 22603 21755 20899 20033 19154 18263 17362 16451 15535 14615 13694 12774 11855 10937 10016 9087 8143 7172 6162 5093 3944 26M 14 15 25855 25640 24955,24745 24089 23886 23242 2RNS 22403 22212 21562 21378 20713 20537 19854 19685 18983 18821 18100 17945 17207 17059 16304 16165 15397 15265 14486 14363 13574 13460 12664 12558 11755 11659 10846 10760 9936 9859 9017 8951 8083 8026 7123 7076 6122 6083 5061 5031 3919 3895 2669 2648 16 17 18 25442 25260 25097 24550 24369 24202 23694 23515 23348 22859 22684 22518 22032 21861 21700 21204 21039 20882 20369 20210 20059 19524 19372 19227 18667 18521 18383 17798 17659 17527 16919 16787 16662 16033 15907 15788 15140 15022 14910 14246 14135 14030 13351 13248 13151 12458 12363 12273 11568 11481 11399 10678 10600 10526 9786 9717 9652 8887 8827 8771 7972 7921 7873 7030 6988 6948 6046 6012 5980 5001 4973 4S47 3871 3848 3827 2628 2607 25M 19 20 24953 24830 240M 23915 23193 23050 22364 22218 21547 21403 20734 20593 19916 19780 19089 18958 18251 18127 17402 17284 16543 16431 15677 15571 14805 14706 13932 13839 13059 12974 12189 12111 11322 11251 10457 10394 9591 9536 8718 8671 7830 7791 6913 6882 5951 5927 4925 4906 3808 3792 2570 2554 35 CENPD487 0 Table 7 Summary of Gadolinium Fuel Assembly Types Value Added Fuel Design No.Gad Pins Gad Loading Enrich Enrich ('v/o Gd203)Non Gad pins Gad Pins 5.0 Density Gad Pins (g/cc)Assembly Average Enrich 5.00 12 16 5.0 5.0 5.0 5.0 5.0 4.5 4.0 4.0 10.1895 10.1895 10.1895 ,10.1238 10.1238 4.97 4.97 4.93 AL IN Ik P%INIS 36 I 20)l)9<NO)t 1 NOOV)t)A)VST Vi)NSTAILTO W)TN 9)ICI NVVSIIIS OHIIN Ttl)AS SH(h'IN l~ll VVN~4 a 5)NCVI S el ICiu 50$STOAACt~)4 TOTAL CILLS 1)l VSA4Lt Ct LLS A(C<(W ll 095~965 LNCNI 5 If))thi)5>STOIIACI ll)C TOTAL CILL)45)VSA4)t Cttl'I E lh 0>Om C2 r+m>C 2m th I m r~rO r.0 g 2 TI I O+O r D r~m TTT I t)A zr C~A~4 3 D 180 n II II II II II U I21)552 121 6 I'LC'5 21 I))ALCS 12 CENPD487 Figure 2 Typical Spent Fuel Storage Rack lLtiodute L INSERT L.INSERT LOCKING HOLE oorr FUEL ASSEMBLY SUPPORT PLATE SLOT FLOW PASSAGES 38 FLORIOA POY/ER II.LIGHT COMPANY ST.LUCIE PLAIIT UNIT 2 TYPICAL SPENT FUEL STORAGE RACK MOOULE FIGURE 9.l 2 CENPD487 Figure 3 Typical Spent Fuel Rack Module L-Insert FlORIDA POITER 8 LIGHT COMPANY ST.LUCIE PLANT UNIT 2 TYPICAL SPEiVT FUEL RACK h>>ODVLE L INSERT FIGURE 9, I 3;)

CEl4PD487 Figure 4 L-Inserts 8.740 188 8.74o SECTION A-A I I I I 4 DETAIL Z 46'WEI.DED CELL BLOCK ING DEVICE"L" INSERT IVIODIFIED"L" INSERT FLORIDA POWER 8, LIGHT COIIPANY ST.LUCIE PLANT UNIT 2 L INSERTS 40 FIGURE 9.1 3b H~)I 7al0 fLKL STO11 Qf.INXWLE All f UEL STOIIAQE INXN U i%VIE W DETAIL Z 73+DETAIL Y F10 f LKL STOIIhQS MODULE 4sll f Uf L STQIIAQf.ICXNLK l/l D n g c r-O C>I C)C)i~~P m C)m 178 Yi'./i VIEW 8-8 DETAIL W DETAIL V DETAIL LI BOTTOM VIEW ELEVATION NOTE: ALL OIMENSIONS ARE IN INCIIES CENP0487 Figure 6 Fuel Assembly ALIGNMENT POST 6.371 SPACER GRID I j UPPER END FITTING CEA~~~)~'4II4~%4 I.'0040404044400000 GUIDE TUBE ASSEMBl.Y TOP VIEW FUEL ROD 158,1" OVERALL 040404 440 04 444 136.r ACTIVE FUEL LENGTH REGION G AND BEYOND 136.7 ACTIVE FUEL LENGTH REGION A-F 30 0 oooo 0088 II o 80 0 888888 88 8888 8 0 088 HALI 040~444~44044~0~BOTTOM VIEW 3.413 REIL 4 3.112 AEG.Q 4.732 REQ.Q 4.703 BEG.F LOWER END FITTING AMENDMENT NO.8 (9/93)FLORIDA POWER 8(LIGHT COMPANY ST.LUCIE PLANT-UNIT 2 42 FUEL ASSEMBLY FIGURC 4.2-6 CENPDQ87 Figure 7 Spent Fuel Rack Module For Region I FUEL ASSEMBLY CELL BLOCKING DEVICE"L" INSERT 43 FLORIDA POWER 8"LIGHT COMPANY ST.LUCIE PLAHT UHIT 2 SPENT FUEL RACK MOOULE FOR REGION I FIGURE 9.1.5a

CENPD487 Figure 8 Spent Fuel Rack Module For Region II FUEL ASSEMBLY CELL BLOCKING OEVICE FLORIDA POWER 8'IGHT COMPANY ST.LUCIE PLANT UNIT 2 SPENT FUEL RACK MODULE FOR REGION II FIGURE 9.1-5b CENPD487 Figure 9 Spent Fuel Loading Pattern For Region I Color Coded Pattern pp, N~a~0 Class 3 or 5 Class 4 or 6 Class 7 or 8 Class 9" EmPty (0)Class 3 or 5 I Black and White Pattern Using Class Numbers 3/5 3/5 3/5 4/6 3/5 0 0 0 3/5 4/6 3/5 4/6 0 0 3/S 0 7/8 4/6 0 7/8 3/5 0 7/8 4/6 0 7/8 7/8 3/5 4/6 0 0 3/5 4/6 0 0 7/8 7/8 7/8 7/8 0 0 0 7/8 3/5 4/6 0 0 3/S 4/6 0 0 7/8 7/8 7/8 7/8 0 0 7/8 7/8 3/5 4f6 3/5 4/6 0 0 0 0 3/S 4/6 3/5 4/6 0 0 0 0 7/8 7/8'/8 7/8 7/8.7/8 7/8 7/8 0 0 0 0 7/8 7/8 7/8 7/8 3/5 7/8 7/8 0 0 0 3/S 7/8 7/8 0 0 0 7/8 7/8 7/8 7/8 7/8 7/8 0 0 0 7/8 7/8 7/8 3/5 4/6 3/5 4f6 0 0 0 4/6 3/5 4/6 0 0 0 7/8 7/8 7/8 7/8 7/8 7/8 0 0 0 7/8 7/8 7/8 3/5 4f6 3/5 4/6 0 0 0 4/6 3/5 4/6 0 0 0 7/8 7/8 7/8 7/8 7/8 7/8 0 0 0 7/8 7/8 7/8 3/5 4/6 3/5 0 0 0 3/S 4/6 3/5 0 0 0 7/8 7/8 7/8 7/8 0 7/8 0 7/8 7/8 0 7/8 7/8 4/6 3/5 0 0 4/6 3/5 0 4/6 0 3/5 0 4f6 0 3/5 0 4/6 4f6 3/5 0 4/6 0 3/5 0 4/6 0 3/5 0 4/6 0 3/5 0 4f6 3/5 3/5 3/5 3/5 0 7/8 4/6 0 7/8 3/S 0 7/8 4/6 0 7/8 3/S 0 7/8 4/6 0 7/8 7/8 7/8 7/8 7/8 0 7/8 0 7/8 0 7/8 0 7/8 0 7/8 0 7/8 0 3/5 0 0 3/5 0 0 0 3/5 0 0 0 0 3/5 0 3/5 0 0 0 0 7/8 7/8 7/8 7/8 7/8 9 9 9 7/8 9 9 9 7/8 9 9 9 0 3/5 0 0 0 0 7/8 7/8 7/8 9 9 9 9 9 9 9 9 9 3/5 7/8 9 0 3/5 0 0 0 0 7/8 7/8 7/8 9 9 9 9 9 9 9 9 9 3/5 7/8 7/8 0 7/8 7/8 0 0 7/8 0 3/5 7/8 0 0 7/8 0 3/S 7/8 0 0 7/8 0 7/8 7/8 0 7/8 7/8 0 7/8 7/8 0 7/8 7/8 0 7/8 7/8 0 7/8 7/8 0 3/5 0 4/6 0 3/5 0 4/6 0 3/5 0 4/6 0 3/5 0 4/6 0 3/5 0 4/6 0 3/S 0 4/6 IA I%IN r+Irrr Class 0)3)4)5)6)7)8)9)Limits Empty Region 1 Region 1 Region 1 Region 1 Region 1 Region 1 Region 1 Key 4.5 w/o U-235 equivalent 4.5 w/o U-235 equivalent with CEA inserted 5.0 w/o V-235 equivalent with Gd.poison rods 5.0 w/o U-235 equivalent with GtL poison rods and CEA inserted 2.82 w/o U-235 equivalent with CEA inserted, 1.82 w/o U-235 cquivalcnt 1.4 w/o U.235 equivalent

'I'l~~'I CENPD487 Figure 10 Spent Fuel Loading Pattern For Region II Y.a S 8 a Yj..M MmiMMBlIMS5NHSL'ml S 8.Y, Y Qgimglmggggjlngg BQQQQ~QQQQggggQigKN Class 1 Class 2 Empty (0)Class 0)l)2)Key Limits Empty Region 2: l.3 w/o U-235 equivalent Region 2: l.S w/o U-235 equivalent fL ENID P%ISIN 46 CENp0487 Figure 11 50000 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region ll, 1.3 w/o 0 years-40000 D I-g 30000 K~20000~1OOOO Acceptable Bumup 5 years 10 years 15 years 20 years'.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/0)1L All P%RNIS I-CENPD487 Figure 12 40000 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region II, 1.5 w/o 0 years 5 years D 30000 C L 20000 10000 IL Acceptable Bu rnup 10 years 15 years 20 years 1.5 2.0 2.5 3.0 3.5 Initial U-235 Enrichment (w/0)4.0 4.5 5.0 CENPD487 Figure 13 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I,1.4 w/o 40000 I-Q>30000 E Gl~20000 10000 Acceptable Bu lTlup 0 years years 10 years 15 years 20 years 1.5 2.0 2.5 3.0 3.5 initiai U-235 Enrichment (w/o)4.0 4.5 5.0 iL IN II.PRIS%

CENPD487 Figure 14 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I, 1.82 wlo 0 years I-c 25000 Acceptable Bumup 5 years 10 years 20 years LQ I C)Eg Cl E I LL 20000 15000 10000 1.5 2.0 2.5 3.0 3.5 Initial U-235 Enrichment (w/0)4.0 4.5 5.0 AL MIN PL&IN 50 CENPD487 Figure 15 15000 Required Fuel Assembly Burnup vs Initial Enrichment Region I, 2.82w/0 0 years D I-10000 C 5ooo E"3+4504.5'E'2-6086 Acceptabte Burnup>-484.92E-7783.1 1.5 2.0 2.5 3.0 3.5 Initial U-235 Enrichment (wlo}4,0 4.5 5.0 iL MIN PRISING 5I CENPD487 t C Figure 16 1.3 1.25 K-infinity at 5.0 wlo with 90%Gad Worth hrough 60000 NlWD/T)1.2 1.15 C 1.05~0 Gad 4.5w/o~4 Gad-4~8 Gad.4~12 Gad-4~8 Gad.6~16 Gad-6~0 Gad 5.0wlo 0.95 0.9 0.85 0.8 0 10000 20000 30000 Assembly Bun1up (MWD/MTU)40000 5QOQQ 6OOQQ 4 SIN P'LlÃll 52

CENPD487 Figure 17 K-infinity at 5.0 w/o with 90%Gad Wolth ThriiI h':20000;IN@fDlT.':-.";.'-:-':

0 Gad 4.5w lo~4 Gad.4~BGad 4~12 Gad.4~B Gad.6~16 Gad 6~0 Gad 5.0w la 2000 4000 eaao eaoo 1oaaa 12000 14ooo 1eooo 1eooo 2oooo Assembly Burnup (MWD/MTU)53 C'

St.Lucie Unit 2 Docket No.50-389 Proposed License Amendment'yai~Md't.Lucie Unit 2 Spent Fuel Pool Dilution Analysis, PSL-ENG-SENS-97-068, Revision 0: FPL Nuclear Engineering, November, 1997.