ML17229A571

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

ML17229A571
Person / Time
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
Download: ML17229A571 (68)

Text

CENP D-387, St. Lucie Unit 2 Criticality Safety Analysis for the Spent Fuel Storage Rack Using Soluble Boron Credit.

October 1997 980i070057 97i23i 05000389 PDR ADOCK PDRi'BB'ombustion Engineering Nuclear Operations rJ Ion IeIr

r CENpD487 Table of Contents Table of Contents.

List of Tables. ...'..... 3 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. 10 3.0 Spent Fuel Pool and Storage Rack .

3.1 Storage Rack Description.....

3.2 Spent Fuel Storage Pattern. 12 4.0 Criticality Safety Analysis. 15 4.1 Analytic Models of the Storage Rack and Module Cells.. 15 4.2 K,a Evaluation at Zero Soluble Boron. 16 4.3 K,a Evaluation for Soluble Boron Credit. 17 4.4 Reactivity Equivalencing 18 4.4.1 Burnup and Decay Time Reactivity Equivalencing. ..18 4.4.2 Gadolinium Credit Reactivity Equivalencing. 19 4.4.3 Soluble Boron Credit for Uncertainties in Reactivity Equivalencing. 21 4.5 Axial Burnup Distribution. .....21 5.0 Postulated Accidents. ..23 6.0 Soluble Boron Credit Summary. ..25 7.0 References. 26 2

CENp0487 List of Figures Figure 1 Spent Fuel Storage Module Installation. ....37 Figure 2 Typical Spent Fuel Storage Rack Module.. ....38 Figure 3 Typical Spent Fuel Rack Module L-Insert. ....39 Figure 4 L-Inserts. ....,40 Figure 5 Spent Fuel Storage Module. 41 Figure 6 Fuel Assembly. 42 Figure 7 Spent Fuel Rack Module For Region I 43 Figure 8 Spent Fuel Rack Module For Region II 44 Figure 9 Spent Fuel Loading Pattern For Region I 45 Figure 10 Spent Fuel Loading Pattern For Region II 46 Figure 11 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region II, 1.3 w/o. 47 Figure 12 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region II, 1.5 w/o. 48 Figure 13 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I, 1.4 w/o. 49 Figure 14 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I, 1.82 w/o 50 Figure 15 Required Fuel Assembly Burnup vs Initial Enrichment, Region I, 2.82 w/o........ 51 Figure 16 K-InfinityAt 5.0 w/o With 90% Gad Worth (Through 60000 Mwd/T).............: 52 Figure 17 K-InfinityAt 5.0 w/o With 90% Gad Worth (Through 20000 Mwd/T).........,.... 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 I/2 2

K9//9$ KKENo lUCB M95/95(O + %KENO) where:

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, 2

a is. the mean calculational method variance deduced from the validation 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,2 is given by the following equation.

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-infinityvs. burnup, K-infinityvs. 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.

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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 IIlk 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) Region II: 1.3 w/o U-235 equivalent,

2) Region II: 1.5 w/o U-235 equivalent,

3) Region I: 4.5 w/o U-235 equivalent,

4) Region I: 4.5 w/o U-235 equivalent with CEA inserted,

5) Region I: 5.0 w/o U-235 equivalent with Gd. poison rods,

6) Region I: 5.0 w/o U-235 equivalent with Gd. poison rods and CEA inserted,

7) Region I: 2.82 w/o U-235 equivalent with CEA inserted,

8) Region I: 1.82 w/o U-235 equivalent, and

9) 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. Ifone 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.

Source Region II 0.97001 2 0.00052 Region I 0.96590 + 0.00054 Regions I + II 0.96731 + 0.00050 A i%IN P%ISIN 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.

SBCToTAi. = SBCg5as + SBCRa + SBCpA where:

SBCToTAi. = total soluble boron credit requirement (ppm).

SBCeses soluble boron credit required for 95/95 K,irto 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 4.4 Reactivity Equivalencing 0 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.

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 Kofthe burned fuel assembly.

Within the first few days after shutdown, a large increase in Kwillresult from decay of the following nuclides:

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 assem bly in the storage cell environment. This effect is favorable since it reduces the pool fL llIk P%NII 18

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 ifall fresh assemblies were unshimmed. Ifburnable 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 Gad loading (w/o) 4 4 8 4 12 4 8 6 16 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 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 Burnup OF-set 4 Gad 4 w/o 4000 8 Gad 4 w/o 3800 12 Gad 4 w/o 3600 8 Gad 6 w/o 3600 16 Gad 6 w/o 3000 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

/%INtN 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-ofthe 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 dX in 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 = 350 ppm SBCRp = 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. Newmyer, W.D., "Westinghouse Spent Fuel Rack Criticality Analysis Methodology,"

WCAP-14416-NP-A, Rev.01, November 1996.

2. 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.

3. Code of Federal Regulations, Title 10, Part 50, Appendix A, Criterion 62, "Prevention of Criticality in Fuel Storage and Handling".

4 U.S. Nuclear Regulatory Commission, Standard Review Plan, Section 9.1.2, NUREG-0800, July 1981.

5. "St. Lucie Unit No. 2 Updated Final Safety Analysis Report," Amendment 10, August 7, 1996, Florida Power and Light Company.

6. "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.

7. "The ROCS and DIT Computer Codes for Nuclear Design," CENPD-266-P-A, April 1983, Combustion Engineering, Inc.

8. M.N. Baldwin et al., "Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel; Summary Report," BAW-1484-7, July 1979.

9. 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.

10 S. R. Bierman and E. D. Clayton, "CriticalityExperiments 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.

11 "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-1 COMP-THERM-010, 017 (Rev.0, 8/31/96) 12 W. Marshall, et. al., "Criticality Safety Criteria", TANS Vol. 35, pg. 278, 1980.

13 D.B. Owen, "Factors for One-Sided Tolerance Limits and for Variables Sampling Plans",

SCR-607, Sandia Corp. Monograph, March 1963.

lL IIII PENIS 26

CENP0487 Table 1 Summary of Calculational Results for Cores X Through XXI of the B&W Close Proximity Experiments Core RUN Plate" Spacing "'

No. Type X 2348 0.99610 + 0.00084 none XI 2355 1.00049 + 0.00080 SS-304 1 XI 2359 0.99884 2 0.00077 SS-304 1 XI 2360 1.00315 + 0.00081 SS-304 1 XI 2361 0.99831 + 0.00080 SS-304 1 XI 2362 1.00060 + 0.00078 SS-304 1 XI 2363 0.99957 + 0.00078 SS-304 I XI 2364 1.00246 + 0.00080 SS-304 1 XII 2370 0.99990 + 0.00082 SS-304 2 XIII 2378 0.99754 + 0.00089 B/Al 1 XIIIa 2423 0.99575 + 0.00087 8/Al 1 XIV 2384 0.99465 2 0.00086 B/Al 1 XV 2388 0.99158 + 0.00084 B/Al 1.

XVI 2396 0.99230 + 0.00088 B/Al 2 XVII 2402 0.99478 2 0.00079 B/Al 1 XVIII 2407 0.99440 + 0.00083 B/Al 2 XIX 2411 0.99821 + 0.00081 B/Al 1 XX 2414 0.99498 + 0.00082 B/Al 2 XXI 2420 0.99318 2 0.00094 B/Al 3 (a) - metal separating unit assemblies (b) - spacing between unit assemblies in units of fuel rod pitch AL llwl P%IN1I 27

CENPD487 Table 2 Summary of Calculational Results for Selected Experimental PNL Lattices, Fuel Shipping and Storage Configurations EXPT. COMMENTS No.

043 0.99787 + 0.00106 Uniform rectangular array, no poison 044 1.00104 + 0.00102 CC 045 0.99955 + 0.00101 <c 046 0.99960 + 0.00103 061 0.99792 + 0.00099 2 x 2 array of clusters, no poison 062 0.99628 + 0.00096 064 0.99696 + 0.00103 2 x 2 array of clusters, 0.302 cm SS-304 cross 071 0.99970 + 0.00101 2 x 2 array of clusters, 0.485 cm SS-304 cross 079 0.99463 + 0.00102 2 x 2 array of clusters, cross of 0.3666 g boron/cm 087 0.99423 + 0.00099 2 x 2 array of clusters, cross of 0.1639 g boron/cm 093 0.99787 + 0.00098 2 x 2 array of clusters, cross of 0.1425 g boron/cm N NIk ERNIE 28

CENP0487 Table 3 Fuel Parameters Employed in Criticality Analysis for St. Lucie Unit 2 Spent Fuel Storage Rack Number of Fuel Rods per Assembly 236 UOz Pellet OD (in) 0.3255 Zr-4 Clad Tube OD (in) 0.382 Clad Tube Wall (in) 0.025 Nominal UO2 Stack Density (g/cc) 10.31 Fuel Rod Pitch (in) 0.5065

'EA Guide Tube OD (in) 0.980 CEA Guide Tube ID (in) 0.900 Number of Guide Tubes per Fuel Assembly 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 0.00259 Pool Temperature (50'F to 155'F) 0.00375 Total 0.00634 Tolerances and Uncertainties UO2 Enrichment ( 0.05 w/o) 0.01380 UO2 Stack Density (2%) 0.00381 Cell Wall Thickness ( 0.005 in) 0.00293 Cell ID 2 Pitch 0.01342 Asymmetric FA Position 0.00000 95/95 CEA Worth 0.00545 95/95 Methodology Uncertainty 0.00639 95/95 Calculational Uncertainties 0.00220 Total Uncertainties (statistical) 0.02166 TOTAL 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 0.91497 Calculational and Methodolo Biase Methodology 0.00259 Pool Temperature (50'F to 155'F) 0,00560 Total 0.00819 Tolerances and Uncertainties UOz Enrichment ( 0.05 w/o) 0.01530 UO2 Stack Density (2%) 0.00474 Cell Wall Thickness ( 0.005 in) 0.00225 Cell ID Ec Pitch 0.01675 Asymmetric FA Position 0.00000 95/95 CEA Worth 0.00527 95/95 Methodology Uncertainty 0.00639 95/95 Calculational Uncertainties 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 Region II, 1.3 wlo Years Enrich 0 1 2 3 4 5 6 7 8 9 10 '11 12 13 14 15 16 17 18 19 20 4.5 46697 45865 45071 44314 43592 42906 ~ 42253 41633 41045 40487 39958 39458 389S4 38536 38111 37710 37329 36969 36626 36300 35988 4.4 45943 45060 44219 43419 42657 41932 41244 40590 39970 39381 38824 38295 37794 37320 36871 36446 36043 35662 35299 34956 34629 4.3 44940 44057 43216 42415 41654 40242. 39589 3896S 38381 37824 37296 36795 36321 35872 35447 35044 34662 34300 33956 33630 4.2 43785 42931 42118 41343 40606 39906 39241 38609 38009 37440 36901 36389 35905 35446 35012 34600 34210 33841 33490 33158 32842 4.1 42549 41739 4866 40231 39531 38865 38232 37632 37062 36521 36008 35522 35062 34626 34213 33822 33452 33101 32767 32451 32151 4.0 41280 40517 39790 39096 38436 37809 37212 36645 36107 35597 35'113 34655 34221 33810 33420 33052 32T02 323T1 3205T 31758 31474 3.9 40011 39292 38606 37952 37329 36173 35638 35130 34649 34192 33759 33350 32962 32594 32246 31917 31604 31307 31025 30756 3.8 38758 38075 37423 36802 38211 35113 34604 34122 33664 33230 32819 32429 32061 31711 31381 31067 30770 30488 30219 29964 3.7 37527 36872 36246 35650 35082 34541 34027 33539 33076 32636 32219 31824 31450 31096 30760 30442 30141 29855 29584 29326 29080 3.6 36319 35681 35072 34492 33940 33415 32915 32441 31990 31563 31157 30773 30409 30064 29738 29428 29135 28857 28593 28341 28102 3.5 35126 34497 33898 33327 32T84 32267 31776 31309 30866 30446 30047 29669 29311 28971 28649 28344 28055 27781 27521 27274 27039 3.4 33939 33313 32717 32150 31611 31098 30610 30147 29707 29290 28894 28519 28163 27825 27505 27202 26914 26641 26383 26137 25904 3.3 32747 32121 31525 30958 30419 28958 28519 28102 27706 27331 26975 26638 26318 26014 25727 25454 25195 24950 24717 3.2 31539 30912 30315 29747 29208 28695 28208 27745 27306 26889 26493 26118 25761 25423 25102 24T98 24510 24236 23977 23731 23498 3.1 30309 29681 29084 28516 27977 27465 26978 26516 26077 25660 25264 24889 24532 24194 23873 23569 232S1 23007 22748 22502 22269 3.0 29049 28423 27829 27264 26728 26218 25734 25274 24838 24423 24029 23656 23301 22055 21783 21525 21280 21049 2.9 27755 27138 26551 25993 25463 24959 24481 24027 23595 23186 22797 2242S 22078 21745 21430 21130 20846 20577 20322 20080 19851 2.8 26429 25825 25250 24704 24185 23692 23223 22778 22356 21955 21574 21213 20870 20544 20236 19942 19664 19401 19151 18914 18690 2.7 250T3 2448T 23930 23401 22897 22419 21965 21534 21124 20736 20367 20016 19684 19368 19069 18785 18515 18260 18017 17787 17570 2.6 23692 23130 22596 22088 21605 21146 20710 20297 19904 19531 19177 18841 18522 18219 17932 17660 17401 17156 16923 16703 16493 2.5 22294 21760 21252 20769. 20311 19875 19461 19068 18695 18341 18005 17686 17383 17096 16824 16565 16320 16087 15866 15656 15457 2.4 20884 20381 19903 1S449 19017 18607 18218 17849 17498 17165 16849 16550 16265 15996 15739 15496 15266 15047 14839 14642 14454 2.3 19469 18998 18552 18127 17725 17343 16980 16636 16309 15999 15705 15426 15161 14910 14671 14445 14230 14025 13832 13648 13473 2.2 18051 17613 17198 16804 16431 16077 15742 15424 15122 14S36 14564 14307 14062 13829 13609 13399 13200 13011 12832 12662 12500 2.1 16628 16221 15836 15473 15130 14805 14497 14205 13929 13666 13417 13181 12956 12743 .12540 12347 12164 1199Q 11825 11669 11522 2.0 15187 14811 14456 14123 13809 13512 13232 12967 12716 12478 12251 12036 11832 11638 11452 11276 11109 10950 10800 10659 10526 1.9 13708 13361 13037 12734 12450 12182 11931 11693 11468 11255 11053 10860 10676 10501 10176 10025 9882 9747 9620 9503 1.8 12153 11839 11548 11278 11026 10791 10571 10363 10167 9981 9805 9637 9476 9323 9177 9037 8904 8778 8660 8550 8449 1.7 10467 10194 9944 9715 9503 9306 9123 8952 8790 8637 8492 8354 8221 8094 7973 7857 7746 7641 7544 7453 7372 1.6 8574 8358 8165 799Q l831 7685 7551 7427 7311 7202 7098 6998 6903 6812 6l24 6639 6558 6482 6411 6347 6290 1.5 6370 6241 6129 6033 5949 5876 5811 5753 5700 5651 5605 5561 5518 5477 5437 5398 5360 5324 5292 5263 5240 r

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 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 '18 19 20 4.5 39660 39059 38484 37935 37410 36910 36433 35980 35550 35141 34753 34386 3403S 33709 33397 33101 32822 32556 32304 32064 31835 4.4 38263 37694 37149 36628 36131 35657 35205 34775 34367 33979 33611 33263 32933 32620 32324 32044 31778 31526 31287 31059 30841 4.3 37043 36486 35954 35445 34959 34495 34054 33634 33235 32856 32496 32155 31833 31527 31237 30963 30704 30457 30223 30000 29787 4,2 35940 35385 34854 34347 33862 33400 32961 32542 32144 31766 31408 31069 30747 30442 30153 29880 29621 29375 29142 28919 28707 4.1 34908 34349 33815 33304 32816 32351 31909 31488 31087 30707 30347 30005 29681 29374 29084 28809 28548 28300 . 28065 27841 27627 4.0 33909 33345 32806 32291 31799 31330 30884 30460 30056 29673 29310 28965 28639 28329 28036 27759 27495 27246 27009 26783 26567 3.9 32915 32347 31805 31287 30793 30323 29874 29448 29043 28658 28293 27947 27619 27309 27014 26735 26471 26220 25981 25755 25538 3.8 31903 31336 30795 30279 29786 29317 28870 28445 28041 27658 27294 26949 26622 26312 26018 25739 25476 25225 24987 24761 24546 3.7 30859 30299 29764 29254 28767 28304 27862 27443 27044 26665 26306 25965 25642 25335 25045 24770 24509 24262 24027 23803 23591 3.6 29777 29228 28705 28206 27730 27277 26845 26435 26045 25675 25324 24991 24674 24375 24091 23822 23567 23325 23095 22877 22669 3.5 28651 28120 27613 27130 26669 26231 25814 2541 T 25040 24682 24342 24020 23714 23424 23150 22889 22642 22408 22186 21975 21774 3.4 27482 26973 26487 26025 25584 25164 24764 24385 24024 23681 23356 23047 22754 22477 22213 21964 21727 21503 21290 21088 20896 3.3 26276 25T92 25331 24891 24472 24074 23695 23335 22993 22668 22359 22066 21788 21524 21275 21038 20813 20600 20398 20206 20024 3.2 25038 245S1 24146 23731 23337 22961 22605 22265 21943 21637 21347 21071 20809 20561 20325 20102 19890 19689 19499 19318 19147 3.1 23776 23347 22938 22549 22179 21827 21493 21175 20873 20587 20315 20056 19811 19578 19357 19148 18949 18761 18583 18414 18253 3.0 22499 22096 21712 21348 21002 20673 20360 20064 19782 19514 19260 19018 18789 18572 18365 18169 17984 17807 17641 17483 17333 2.9'1214 20835 20475 20133 19809 19501 19209 18931 18668 18418 18180 17955 17740, 17537 17344 17160 16987 16822 16666 16518 16378 2.8 19928 195S9 19230 18908 18603 18314 18040 17779 17532 17298 17075 16863 16663 16472 16290 16118 15955 15800 15654 15515 15384 2.7 18644 18304 17981 17676 17388 17114 16855 16610 16376 16155 15945 15746 15556 15376 15205 15043 14889 14743 14604 14473 14349 2.6 17366 17039 16731 16439 16164 15904 15658 15424 15203 14993 14794 14605 14425 14254 14092 13938 13T91 13652 13520 13396 13279 2.5 16088 15773 15477 15197 14934 14685 14449 14227 14016 13S16 13626 13446 13275 13112 12957 12810 12670 12537 12411 12293 12181 2.4 14804 14500 14215 13947 13695 13457 13233 13021 12820 12630 12450 12278 12115 11960 11812 11672 11539 11412 11292 11179 11073 2.3 13500 13210 12938 126S4 12445 12221 12010 1181 0 11622 11443 11273 11112 10959 10813 10674 10542 10416 10297 10184 10078 9978

'2.2 12154 11885 11634 11399 11180 109T5 10782 10600'0428 10265 10111 .9964 9824 9691 9564 9443 9328 9219 9116 9019 8930 2.1 10740 10503 10284 10081 9891 9715 9550 9394 9248 9109 8977 8852 8733 8619 8510 8406 8307 8214 8126 8044 7968 2.0, 9219 4

9034 8865 8710 8568 8436 8314 8199 8091 7989 7892 7800 7712 7627 7546 7468 7394 7324 7259 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 Years Enrich 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 4.5 43355 42637 41951 41298 40675 4X$2 39519 38983 38474 37992 37534 37100 36688 36299 35930 35581 35250 3%36 34638 34355 34086 44 42336 41663 41022 40412 39830 39278 38752 38253 37778 37328 36901 36496 36112 35748 35403 35077 34767 34475 34197 33935 33686 4.3 41268 40614 39992 39399 38835 38298 37788 37303 36842 36405 35991 35598 35225 34871 34537 34220 33919 33635 33366 33111 32870 4.2 40163 39512 38891 38300 37738 37202 36694 36210 35751 35315 34902 34509 34138 33786 33452 33136 32837 32554 32285 32030 31789 4.1 39029 38372 37745 37148 36579 36037 35522 35033 34568 34127 33709 33312 32937 32581 32244 31924 31622 31335 31063 30805 30560 4.0 37875 37208 36572 35965 35387 34836 34313 33815 33343 32895 32470 32067 31685 31323 30981 30656 30349 30058 297S1 29518 29268 3.9 36705 36030 35385 34769 34183 33625 33094 32589 32110 31655 31224 30815 30428 30061 29714 29385 29073 28777 28496 28229 27974 3.8 35523 34843 34192 33572 32981 32418 31883 31374 30891 30432 29997 29585 29195 28825 28475 28143 27828 27530 27247 26977 26720 3.7 34333 33652 33001 32381 31790 31227 30692 30183 29700 29241 28807 28394 28004 27634 27283 26951 26636 26338 26054 25784 25527 3.6 33134 32458 31814 31199 30614 30056 29526 29022 28543 28089 27658 27249 26862 26495 26147 25818 25506 25210 24929 24661 24407 3.5 31928 31264 30631 30027 29452 2S905 2S384 27890 27420 26974 26551 26150 25769 25409 25067 24743 24437 24146 23869 23607 23358 3.4 30714 30067 29450 28863 28303 27771 27264 26783 26326 25892 25481 25090 24720 24369 24036 23721 23422 23138 22870 22614 22372 3.3 29492 28867 28271 277Q3 27163 26649 26161 25696 25256 24837 24439 24062 23705 23366 23044 22739 22450 22177 21917 21671 21438 3.2 28261 27660 27088 26544 26026 25534 25066 24621 24199 23798 23417 23056 22713 22079 21787 21509 21247 20998 20762 20539 3.1 27018 26445 25899 25380 24887 24418 23973 23549 23147 22765 22403 22058 21732 21422 21128 2Q585 20334 20097 19&73 19661 3.0 25762 25218 24700 24208 23740 23296 22874 22472 22091 21730 21386 21060 20750 20177 19913 19662 19425 19200 18987 18786 2.9 24492 23977 23487 23022 22581 22161 21762 21383 21024 20682 20358 20050 19758 19480 19217 18968 18731 18507 18294 18093 179Q3 2.8 23204 22718 22257 21819 21403 21008 20276 19938 19616 19311 19022 18747 18486 18238 18003 17781 17569 17369 17180 17001 2.7 21897 21440 21006 20595 20204 19832 19480 19146 18828 18527 18241 17969 17711 17466 17234 17014 16804 16606 16418 16240 16072 2.6 20569 20140 19733 19346 18980 18632 18302 17989 17692 17410 17142 16888 16647 16418 16201 15994 15799 15613 15437 15270 15112 2.5 19217 18815 18433 18072 17729 17404 17096 16804 16527 16264 16015 15778 15553 15340 15137 14945 14762 14589 14424 14268 14120 2.4 17839 17463 17107 16770 16451 16149 15863 15591 15334 15090 14859 14639 14431 14233 14045 13866 13696 13535 13382 13237 13100 2.3 16432 16082 15751 15439 15144 14865 14601 14350 14114 13889 13676 13474 13281 13099 12925 12761 '2604 12455 12314 12180 12054 2.2 14994 14670 14365 14078 13807 13552 13310 13082 12866 12662 12468 12283 12108 11941 11783 11632 11489 11353 11224 11102 10987 2.1 13522 13224 12946 12684 12439 12209 11992 11787 11593 11410 11235 11070 10912 10762 ,10619 10483 10353 10231 10115 10006 9904 2.0 12013 11742 11491 11257 11039 10835 10644 10463 10293 10132 9979 9833 9694 9561 9434 9313 9199 9090 8988 8893 8805 1.9 10462 10220 9998 9793 9603 9427 9262 9108 8962 8824 8693 8568 8449 8334 8225 8121 8021 7927 7840 7759 7686 1.8 8866 8653 8461 8285 8125 7977 7840 7712 7592 7479 7371 7268 7168 7073 6982 6894 6811 6733 6660 6594 6537 1.7 7219 7036 6873 6727 6595 6476 6366 6265 6170 60S1 5996 5914 5836 5760 5687 5616 5549 5487 5429 5378 5335 1.6 5514 5360 5225 5107 5002 4908 4824 4747 4675 4608 4544 4483 4424 4367 4311 4257 4206 4158 4114 4076 4046 1.5 3746 3617 3505 3409 3325 3252 3187 3129 3076 3027 2981 2937 2894 2853 2813 2774 2736 2701 2668 2639 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 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 4.5 30444 30005 29587 29189 28809 28446 28099 27768 27452 27150 26863 26589 26330 26085 25855 25640 25442 25260 25097 24953 24830 4.4 29393 28968 28562 28176 27807 27456 27120 26800 26495 26204 25927 25664 25414 25178 24955,24745 24550 24369 24202 240M 23915 4.3 28377 27966 27573 27199 26842 26502 26178 25868 25574 252S3 25027 24773 24533 24305 24089 23886 23694 23515 23348 23193 23050 4.2 27381 26984 26605 26243 25&98 25569 25256 24958 24674 24403 24146 23902 23670 23450 23242 2RNS 22859 22684 22518 22364 22218 4.1 26393 26010 25644 25295 24962 24645 24343 24055 23782 23521 23274 23039 22815 22603 22403 22212 22032 21861 21700 21547 21403 4.0 25404 25035 24683 24346 24026 23720 23429 23152 22889 22638 22400 22174 21959 21755 21562 21378 21204 21039 20882 20734 20593 3.9 24407 24052 23714 23390 23082 22788 22509 22242 21989 21748 21519 21302 21095 20899 20713 20537 20369 20210 20059 19916 19780 3.8 23400 23059 22733 22423 22127 21845 21577 21321 21078 20847 20628 20419 20221 20033 19854 19685 19524 19372 19227 19089 18958 3.7 22379 22051 21739 21441 21158 20888 20631 20387 20155 19934 19724 19524 19334 19154 18983 18821 18667 18521 18383 18251 18127 3.6 21343 21029 20730 20446 20175 19917 19672 19439 19217 19007 18806 18616 18435 18263 18100 17945 17798 17659 17527 17402 17284 3.5 202S4 19993 19707 19436 19178 18933 18700 18479 18268 18068 17878 17697 17525 17362 17207 17059 16919 16787 16662 16543 16431 3.4 19231 18944 18672 18414 18170 17938 17717 17508 17309 17119 16940 16769 16606 16451 16304 16165 16033 15907 15788 15677 15571 3.3 18158 17885 17627 17383 17152 16725 16528 16341 16163 15994 15833 15680 15535 15397 15265 15140 16933'5921 15022 14910 14805 14706 3.2 17076 16817 16574 16344 161 27 15727 15543 15368 15202 15044 14894 14751 14615 14486 14363 14246 14135 14030 13932 13839 3.1 15987 15743 15514 15299 15096 14905 14725 14554 14392 14238 14092 13953 13821 13694 13574 13460 13351 13248 13151 13059 12974 3.0 14893 14665 14451 14251 14063 13887 13720 13563 13414 13273 13139 13012 12890 12774 12664 12558 12458 12363 12273 12189 12111 2.9 13794 13582 13384 13200 13028 12866 12715 12572 12437 12309 12187 12071 11961 11855 11755 11659 11568 11481 11399 11322 11251 2.8 12691 12495 12314 12146 11990 11844 11707 11579 11457 11343 11234 11130 11031 10937 10846 10760 10678 10600 10526 10457 10394 2.7 11581 11402 11238 11087 10947 10817 10696 10582 10475 10374 10278 10186 10099 10016 9936 9859 9786 9717 9652 9591 9536 2.6 10460 10300 10153 10019 9896 9782 9676 9577 9485 9397 9314 9235 9159 9087 9017 8951 8887 8827 8771 8718 8671 2.5 9324 9182 9054 8937 8830 8733 8642 8558 8480 8406 8335 8268 8204 8143 8083 8026 7972 7921 7873 7830 7791 2.4 8164 8041 7930 7831 7741 7660 7585 7515 7451 7390 7332 7277 7223 7172 7123 7076 7030 6988 6948 6913 6882 2.3 6970 6865 6773 6691 6617 6551 6491 6436 6384 6336 6290 6246 6203 6162 6122 6083 6046 6012 5980 5951 5927 2.2 5728 5641 5566 5500 5442 5391 5345 5303 5264 5227 5192 5158 5125 5093 5061 5031 5001 4973 4S47 4925 4906 2.1 4420 4350 4292 4241 4198 4161 4128 4098 4070 4044 4018 3994 3969 3944 3919 3895 3871 3848 3827 3808 3792 2.0 3025 2972 2928 2892 2862 2836 2814 2795 2777 2759 2742 2725 2707 26M 2669 2648 2628 2607 25M 2570 2554 35

CENPD487 0 Table 7 Summary of Gadolinium Fuel Assembly Types Value Added Fuel Design No. Gad Pins Gad Loading Enrich Enrich Density Assembly

('v/o Gd203) Non Gad pins Gad Pins Gad Pins Average (g/cc) Enrich 5.0 5.00 5.0 10.1895 5.0 4.5 10.1895 12 5.0 10.1895 4.97 5.0 4.0 ,10.1238 4.97 16 5.0 4.0 10.1238 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 a5 )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 n

TI II I

180 II O II II

+ O II I))

E lh U 0>

Om C2 r ~D ALCS r+ r m> m TTT C 2m th I t)A m

r~

rO I

zr I21) 6 121 I'LC'5 21 12 C 552 r.

0g ~ A 2

~ 4 3 D

CENPD487 Figure 2 Typical Spent Fuel Storage Rack lLtiodute L INSERT L. INSERT LOCKING HOLE oorr FUEL ASSEMBLY SUPPORT PLATE SLOT FLOW PASSAGES FLORIOA POY/ER II.LIGHT COMPANY ST. LUCIE PLAIIT UNIT 2 TYPICAL SPENT FUEL STORAGE RACK MOOULE 38 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 FIGURE 9.1 3b 40

H~ )I 7al0 fLKLSTO11 Qf. INXWLE i%VIEW AllfUEL STOIIAQE INXN U DETAIL Z 73+

DETAIL Y F10 fLKL STOIIhQS MODULE 4sll fUfL STQIIAQf. ICXNLK

~

~P m

C) l/l m D DETAILW DETAIL V DETAIL LI VIEW 8-8 n

c gr-178 Yi'./i O BOTTOM VIEW C>

I C)

C) i ELEVATION NOTE: ALL OIMENSIONS ARE IN INCIIES

CENP0487 Figure 6 Fuel Assembly ALIGNMENTPOST UPPER END FITTING

~ ~ ~

)

~

6.371 I

'4I I4 j

SPACER GRID CEA ~ %4 I.'

GUIDE TUBE 0040404044400000 ASSEMBl. Y TOP VIEW FUEL ROD 158,1" 136.r OVERALL ACTIVE FUEL LENGTH 040404 440 444 04 REGION G AND 136.7 BEYOND ACTIVE FUEL LENGTH REGION A-F 888888 88 8888 300 8

oooo 0088 HALI 0088 II 80 o

0 040 ~ 444 ~ 44044 ~ 0 ~

BOTTOM VIEW 4.732 REQ. Q 4.703 3.413 REIL 4 BEG.F 3.112 AEG. Q LOWER END FITTING AMENDMENT NO. 8 (9/93)

FLORIDA POWER 8( LIGHT COMPANY ST. LUCIE PLANT UNIT 2 FUEL ASSEMBLY FIGURC 4.2-6 42

CENPDQ87 Figure 7 Spent Fuel Rack Module For Region I FUEL ASSEMBLY CELL BLOCKING DEVICE "L" INSERT FLORIDA POWER 8 "LIGHT COMPANY ST. LUCIE PLAHT UHIT 2 SPENT FUEL RACK MOOULE FOR REGION I FIGURE 9.1.5a 43

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 p p, N

a~

~ 0 Class 3 or 5 Class 7 or 8 Class 9 EmPty (0) Class 3 or 5 Class 4 or 6 I

Black and White Pattern Using Class Numbers 3/5 4/6 3/5 3/5 4/6 3/5 4/6 3/5 4f6 3/5 4/6 3/5 7/8 7/8 3/5 4/6 3/5 4f6 4f6 3/5 4/6 3/5 4/6 3/5 4/6 3/5 4f6 3/5 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 4/6 3/5 4/6 3/5 3/5 4/6 3/S 4/6 3/S 4/6 3/5 4/6 3/S 7/8 7/8 4/6 3/5 4/6 3/5 4/6 3/5 4/6 3/S 4/6 3/5 4/6 3/5 0 3/5 4/6 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 4/6 0 4/6 3/5 3/S 0 7/8 7/8 7/8 7/8 7/8 7/8 7/8'/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 0 3/5 0 3/5 4/6 0 7/8 7/8 7/8 7/8 7/8 7/8 .7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 0 7/8 0 4f6 0 4/6 3/5 3/5 0 7/8 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 7/8 7/8 0 3/5 0 3/5 4/6 0 7/8 7/8 0 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 0 7/8 7/8 0 4/6 0 4f6 3/5 3/5 0 7/8 7/8 0 7/8 0 3/5 0 3/5 0 3/5 0 3/5 0 3/5 0 3/5 0 3/5 0 7/8 7/8 0 7/8 7/8 0 3/5 0 3/5 4/6 0 7/8 7/8 0 7/8 0 0 0 0 0 0 0 0 0 0 0 0 0 0 7/8 0 7/8 7/8 0 4/6 0 4/6 3/5 3/S 0 7/8 7/8 0 7/8 3/5 0 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 7/8 0 3/5 7/8 0 7/8 7/8 0 3/5 0 3/5 4/6 0 7/8 0 7/8 0 0 7/8 9 9 9 9 9 9 9 9 9 9 7/8 0 0 7/8 0 7/8 7/8 0 4/6 0 4/6 3/5 3/S 0 7/8 0 7/8 3/5 0 7/8 9 9 9 9 9 9 9 9 9 0 3/S 7/8 0 7/8 7/8 0 3/5 0 3/S 4/6 0 7/8 7/8 0 7/8 0 0 7/8 9 9 9 9 9 9 9 9 9 0 0 7/8 0 7/8 7/8 0 4/6 0 4/6 Key Class Limits

0) Empty

3) Region 1 4.5 w/o U-235 equivalent

4) Region 1 4.5 w/o U-235 equivalent with CEA inserted

5) Region 1 5.0 w/o V-235 equivalent with Gd. poison rods

6) Region 1 5.0 w/o U-235 equivalent with GtL poison rods and CEA inserted

7) Region 1 2.82 w/o U-235 equivalent with CEA inserted ,

8) Region 1 1.82 w/o U-235 cquivalcnt

9) Region 1 1.4 w/o U.235 equivalent IA I%IN r+Irrr

I'l

~ ~

'I

CENPD487 Figure 10 Spent Fuel Loading Pattern For Region II Y. a S

8 S

a 8 Yj.. .Y, Y M MmiMMBlIMS5NHSL'ml Qgimglmggggjlngg BQQQQ~QQQQggggQigKN Class 1 Class 2 Empty (0)

Key Class Limits

0) Empty l) Region 2: l.3 w/o U-235 equivalent

2) Region 2: l.S w/o U-235 equivalent fL ENID P%ISIN 46

CENp0487 Figure 11 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region ll, 1.3 w/o 50000 0 years 5 years D

40000 I- 10 years Acceptable Bumup 15 years 20 years g 30000 K

~ 20000

~ 1OOOO

'.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 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region II, 1.5 w/o 40000 0 years 5 years 10 years 15 years D 20 years 30000 Acceptable Bu rnup C

L 20000 10000 IL 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/0)

CENPD487 Figure 13 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I,1.4 w/o 0 years 40000 years I-Acceptable Bu lTlup 10 years Q

15 years 20 years

> 30000 E

Gl

~ 20000 10000 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 initiai U-235 Enrichment (w/o) iL IN II

. PRIS%

CENPD487 Figure 14 Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I, 1.82 wlo 0 years 5 years I- 10 years c 25000 Acceptable Bumup 20 years 20000 I

LQ C)

Eg 15000 Cl E

10000 I

LL 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/0)

AL MIN PL&IN 50

CENPD487 Figure 15 Required Fuel Assembly Burnup vs Initial Enrichment Region I, 2.82w/0 15000 0 years D

I-10000 Acceptabte Burnup > -484.92E-E"3+ 4504.5'E'2- 6086 7783.1 C

5ooo 1.5 2.0 2.5 3.0 3.5 4,0 4.5 5.0 Initial U-235 Enrichment (wlo}

iL MIN PRISING 5I

t C

CENPD487 Figure 16 K-infinityat 5.0 wlo with 90% Gad Worth 1.3 hrough 60000 NlWD/T) 1.25 1.2 1.15

~0 Gad 4.5w/o

~4 Gad-4 C

1.05 ~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 40000 5QOQQ 6OOQQ Assembly Bun1up (MWD/MTU) 4 SIN P'LlÃll 52

CENPD487 Figure 17 K-infinityat 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

'yai~

Proposed License Amendment Md't.

Lucie Unit 2 Spent Fuel Pool Dilution Analysis, PSL-ENG-SENS-97-068, Revision 0: FPL Nuclear Engineering, November, 1997.