Revision as of 01:06, 9 July 2018

Text

Holtec International Report No. HI-2146153, Revision 2, Licensing Report for the Criticality Analysis of the Dresden Unit 2 and 3 SEP for Atrium 10XM Fuel Design
	ML15215A337
Person / Time
Site:	Dresden
Issue date:	07/30/2015
From:	; Holtec
To:	; Office of Nuclear Reactor Regulation
References
	HI-2146153, Rev 2
	Download: ML15215A337 (153)
	v • d • e

ATTACHMENT 3 Holtec International Report No. HI-21 461 53, Revision 2, "Licensing Report for the Criticality Analysis of the Dresden Unit 2 and 3 SFP for ATRIUM 10XM Fuel Design" (Non-Proprietary Version) mnEEm HOLTEC INTERNATIONAL Holtec Center, One Holtec Drive, Marlton, NJ 08053 Telephone (856) 797- 0900 Fax (856) 797 -0909 Licensing Report for the Criticality Analysis of the Dresden Unit 2 and 3 SFP for ATRIUM I OXM Fuel Design -Non Proprietary Version FOR Exelon Holtec Report No: HI-21 461 53 Holtec Project No: 2393 Sponsoring Holtec Division:

HTS Report Class:* SAFETY RELATED Table of Contents 1. INTRODUCTION.........................................................................................

3 2. METHODOLOGY........................................................................................

4 2.1 APPROACH........................................................................................

4 2.2 COMPUTER CODES AN!) CROSS SECTION LiB3RARIES....................................................

4 2.2.1 MCNPS-1.51

........................................................................................

4 2.2.1.1 MCNP5-l1.51 Validation

................................................................................................

4 2.2.2 CASMO-4.............................................................................................

2.3 ANALYSIS METHODS......................

..................................................................

5 2.3.1 Design Basis Fuel Assembly........................................................................

2.3.1.1 Peak Reactivity...........................................................................................................

6 2.3.1.1.1 Peak Reactivity and Fuel Assembly Blurnup.......................................................................

6 2.3.1.1.2 Isotopic Compositions...............................................................................................

7 2.3.1.2 Screening Calculations for the Design Basis Fuel Assembly ........................................................

7 2.3.1.3 Determination of the Design Basis Fuel Assembly Lallice ..........................................................

7 2.3.1.4 Design Basis Model .....................................................................................................

8 2. 3.2 Core Operating Parchneters........................................................................

9 2.3.3 Integral Reactivity Control Devices...............................................................

9 2.3.4 Axial and Planar Enrichment Variations........................................................

10 2.3.5 Fuel Assembly Eccentric Positioning and Fuel Assembly Dc-Channeling

...................

10 2.3.6 Fuel Bundle Orientation in SF1' Rack Cell ......................................................

11 2.3.7 Reactivily Effect of Spent Fuel Pool Water Temperature

......................................

122 2.3.8 Fuel and Storage Rack Man ufacturing Tolerances.............................................

13 2.3.8.1 Fuel Manufacturing Tolerances.......................................................................................

13 2.3.8.2 SFP Storage Rack Manufacturing Tolerances........................................................................

14 2.3.9 Fuel Depletion calculation Uncertainty

..........................................................

S1 2.3.10 Fission Products and Lumped Fission Products Uncertainty..................................

16 2.3.11 Depletion Related Fuel Assembly Geometry Changes..........................................217 2.3.11.1 Fuel Rod Geometry Changes.......................................................................................

17 2.3.11,1.1 Fuel Rod Growth and Cladding Creep ..........................................................................

17 2.3.11.1.2 Fuelkod Crud Buildup...........................................................................................

18 2.3.11.1.3 FuelRod Bow.....................................................................................................

18 2.3.11.2 Fuel Channel lBulging and Bowing................................................................................

18 2.3.12 SFP Storage Rack Interfaces

.....................................................................

19 2.3.13 Maximum keffCalculation for ANormal Conditions

..............................................

20 2.3.14 Fuel Movement, Inspection and Reconstitution Operations

...................................

20 2.3.15 Accident Condition

................................................................................

21 2.3.15.1 Temperature and Water Density Effects ..........................................................................

22 2.3.15.2 Dropped Assembly -Horizontal...................................................................................

22 2.3.15.3 Dropped Assembly -Vertical into an Empty Storage Cell ......................................................

22 2.3.15.4 Missing BORAL Panel .............................................................................................

23 2.3.15.5 Rack movement .....................................................................................................

23 2.3.15.6 Mislocated Fuel Assembly .........................................................................................

23 2.3.15.6.1 Mislocated Fuel Assembly Adjacent to the Storage Rack.....................................................

23 2.3.15.6.2 Mislocated Fuel Assembly in the Corner between 'Two Racks................................................

24 2.3.15.6.3 Mislocated Fuel Assembly in thle Corner between TIhree Racks ..............................................

24 2.3.15.6.4 Mislocated Fuel Assembly in the FPM..........................................................................

25 2.3.16 Reconstituted Fuel Assemblies

...................................................................

26 3. ACCEPTANCE CRITERIA............................................................................

27 4. ASSUMPTIONS..........................................................................................

28 5. INPUT DATA ............................................................................................

29 Project No. 2393 Report No. HI1-2146153 Page 1 Holtec International Proprietary Information 5.1 FUEL ASSEMBLY SPEcwCAFIAION.........................................................................

29 5.2 REACTOR AND SFP OPERATING PARAMETERS.........................................................

30 5.3 STORAGE RACK SPECIFICATION

.........................................................................

30 5.4 MATERIAL COMPOSITIONS................................................................................

30 6. COMPUTER CODES ...................................................................................

31 7. ANALYSIS RESULTS ..................................................................................

32 7.1 DETERMINATION OF THE DESIGN BASIS FUEL ASSEMBLY LA'IfTICEF.................................32 7.2 CORE OPERA'ING PARAMETERS.........................................................................

32 7.3 FUEL ASSEMBLY ECCENTRIIC POSITIONING AN!) FUEL ASSEMBLY DE-CHANNELTNG.............

32 7.4 FUEL BUNDLE ORIENTATION IN TILE SFP RACK CELL .................................................

33 7.5 REACTIVITY EFFECT OF SPENT FUEL POOI. WATER TE.MPERATURE.................................

33 7.6 FUEIL AND STORAGE RACK MANUFACTURING TOLERANCES

........................................

33 7.6.]I Fuel Manufacturing Tolerances..................................................................

33 7. 6.2 SEP Storage Rack Manufacturing Tolerances...................................................

33 7,.6, 3 Fuel Depletion Calculation Uncertainty

........................................................

34 7,6.4 Fission Products' and Lumped Fission Products Uncertainty..................................

34 7.6.5 Depletion Related Fuel Assembly Geometry Changes..........................................

34 7.6.5.1 Fuel Rod Geometry Changes..........................................................................................

34 7.6.5.1.1 Fuel Rod Growth, Cladding Creep and Fuel Rod Crud Buildup ...............................................

34 7.6.5.1.2 Fuel Rod Bow......................................................................................................

34 7.6.5.2 Fuel Channel Bulging and Bowing ...................................................................................

35 7.7 SFP STORAGE RACK INTERFACES

.......................................................................

35 7.8 MAXIMUM CALCULATIONS FOR NORMAL CONDITIONS

.........................................

35 7.9 FUEL MOVEMENT, INSPECTION AND) RFECONSTrII'UION OPERATION.

...............................

35 7.10 ABNORMAL AND ACCIDENT CONDITION S............................................................

35 8. CONCLUSION...........................................................................................

36 9. REFERENCES...........................................................................................

37 Appendix A: CASMO-4 Screening Calculations for Determination of the Design Basis Fuel Assembly......................................................................A-i Appendix B: MCNP5-l .51 Screening Calculations for Determination ofthe Design Basis Fuel Assembly .....................................................................

B-1 Appendix C: MCNP5-1 .51 Design Basis Calculations......................................

C-I Project No. 2393 Report No. H1-2146I 53 Page 2 H-oltec International Proprietary Information

1. INTRODUCTION This report documents the criticality safety evaluation for the storage of B3WR fuel in the Unit 2 and Unit 3 spent fuel pools (SPPs) at the Dresden Station operated by Exelon. The Unit 2 and Unit 3 SFP racks are identical and are designed to accommodate BWR fuel. Currently, the SEP racks credit BORAL for reactivity control. This analysis will include a new fuel .design, ATRIUM I 0XM. This analysis will show that the effective neutron multiplication factor (kerr) in the SFP racks fully loaded with fuel of the highest reactivity, at a temperature corresponding to the highest reactivity, is less than 0.95 with a 95% probability at a 95% confidence level.Reactivity effects of abnormal and accident conditions are also evaluated to assure that under all credible abnormaal and accident conditions, the reactivity will not exceed the regulatory limit.Criticality control in the SEP, as credited in this analysis, relies on the following:

Fixed neutron absorbers o B3ORAL fixed to the SFP rack cell walls* Integrated neutron absorbers o Gadolinium (Gd) in the fuel (peak reactivity isotopic composition).

Criticality control in the SFP, as credited in this analysis, does not rely on the following:

Crediting burnup Project No. 2393 Report No. 11I-2146153 H-oltec International Proprietary Information Page 3

2. METHODOLOGY 2.1 General Approach The analysis is performed consistent with regulatory requirements and guidance.

The calculations are performed using either the worst case bounding approach or the statistical analysis approach with respect to the various calculation parameters.

The approach considered for each parameter is discussed below.2.2 Computer Codes and Cross Section Libraries 2.2.1 MCNP5-I.51 MCNP5-1.51 is a three-dimensional Monte Carlo code developed at the Los Alamnos National Laboratory

[1]. MCNP5-1 .51 calculations use continuous energy cross-section data based on ENDF/B-VII.

MCNP is selected because it has history of successful use in fuel storage criticality analyses and has most of the necessary features (except for fuel depletion analysis) for the analysis to be performed for Dresden Station SFP.The convergence of a Monte Carlo criticality problem is sensitive to the following parameters: (I.) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution.

All M.CNP5 calculations are performed with a minimum of 12,000 histories per cycle, a minimum of 300 skipped cycles before averaging, and a minimum of 300 cycles that are accumulated.

The initial source is specified as uniform over the fueled regions (assemblies).

Convergence is determined by confirming that the source distribution converged using the Shannon entropy [1] and the was confirmed to converge by checking the output file.2.2.1.1 MCNP5-1.51 Validation B~enchmarking of MCNP5-t .51 for criticality calculations is documented in [21. The benchmarking is based on the guidance in [3], and includes calculations for a total of fl critical experiments with fresh U0 2 fuel, fresh MOX fuel, and fuel with simulated actinide composition of spent fuel (HTC experiments

[2]). The results of the benehmarking calculations show few significant trends, and indicate a truncated bias of ' with an uncertainty of +/- (95% probability at a 95%confidence level) for the full set ofall

experiments.

The statistical treatment used to determine those values considered the variance of the population about the mean and used appropriate confidence factors and trend analyses.

Note that the area of applicability for the MCNP5.-1.51 benchmark is presented in Table 2.1(a) and confirms the applicability of benchmarking in [2] to this Dresden analysis.Trend analyses are also performed in [2], and significant trends are determined for various subsets and parameters.

in order to determine the maximum bias that is applicable to the SA positive bias which results in decrease in reactivity is truncated to zero [3].Project No. 2393 Report No. 1-1-2 146153 Page 4 H-oltec International Proprietary Information calculations in this report, the trend equations from [2] are evaluated for the specific parameters of the current analyses.

The subset of all critical experiments with pure water is considered in Table D.3-1 3 of [2] and the tabulated bias and bias uncertainty values for several energy of average lethargy causing fission (EALF) and U3-235 enrichment values are provided in Table 2.1(c).The evaluation of MCNP5-1 .51 bias and bias uncertainty applicable to the current calculations is summarized in Table 2.1t(b) for all experiments and experiments with pure water. As included in Table 2.1(b), the EALF and U-235 enrichment parameters show significant trends for experiments with pure water. The bias and bias uncertainty for each of these independent parameters are calculated using the linear correlation formulas provided in Table 2.1(b) and equations 2-I through 2-6 of [2].Table 2.1(c) provides tabulated bias and bias uncertainty values for several HALF and U-235 enrichment values. The calculated HALF of the rack with pure water is stated in Note 1 of Table 2.1(c). The U-235 enrichment is based on the maximum U-235 enrichment of wt%, and repeated in Note I of Table 2.1 (c). The calculated HALF for the design basis fuel assembly is within two HALF values inl Table 2.1(c). Also, the maximum U-235 enrichment is within two U-235 enrichment values in Table 2.1(c). The bounding bias and bias uncertainty values for these two parameters (HALF and U3-235 enrichment) are selected and compared to the bias and bias uncertainty of the 'all experiments' and 'all with pure water' (as provided in Table 2.1(b)).As can be seen, the set of bias and bias uncertainty of the 'all experiments' is largest, and is used in the maximum k~ff calculations.

2.2.2 CASMO-4 Fuel depletion analyses during core operation are performed with CASMO-4 Version 2.05.14 (using the 70-group cross-section library), which has been approved by the NRC for reactor analysis (depletion) when providing reactivity data for specific 3D simulator codes. CASMO-4 is a two-dimensional multigroup transport theory code based on the Method of Characteristics and it is developed by Studsvik of Sweden [4]. CASMO-4 is used to perform depletion calculations and to perform various sensitivity studies. The uncertainty on the isotopic composition of the fuel (i.e., the number density) is considered as discussed below (see Section 2.3.9). A validation for CASMO-4 to develop a bias and bias uncertinty is not necessary because the results of the CASMO-4 sensitivity studies are not used as input into the k~r calculations.

However, the code authors have validated CASMO-4 against MCNP and various critical experiments

[5].2.3 Analysis Methods 2.3.1 Design Basis Fuel Assembly There are various fuel designs stored in the Dresden SFP. For the purpose of this analysis, the reactivity of each design is evaluated and the most reactive fuel bundle lattice is determined for use as the design basis fuel assembly (a single lattice (most reactive) along the entire active length) to determine ken- at the 95195 level. This approach follows the guidance in [6] and [7], and is further described below.Project No. 2393 Report No. 1-1I-2146153 Page 5 H-oltec International Proprietary Information 2.3.1.1 Peak Reactivity The BWR fuel designs used at the Dresden Station use Gd as an integral burnable absorber.Initially, the Gd in the fuel assembly holds down the fresh fuel assembly reactivity and then, as core depletion occurs, the Gd begins to burnout until it is essentially fully depleted.

As the Gd depletes the reactivity of the fuel assembly increases until it reaches a peak. This peak reactivity is the fuel assembly's most reactive condition.

Note that most BWR fuel designs are composed of various axial latt ices (including blankets) that can have different axial lengths, uranium loadings, fuel pin arrangements including partial or part-length rods, Gd pin locations and loading, etc. These various lattice components can all effect at what burnup the peak reactivity occurs and the magnitude of the peak reactivity.

The axial lattices within a single fuel assembly can therefore all have different peak reactivity.

Therefore, for each fuel design type, an assessment is made of every lattice to determine the bounding lattice (highest peak reactivity).

These are the screening calculations described in Section 2.3.1.2 and are performed with CASMO-4 only. Note that using the CASMO-4 code is appropriate since all lattices are compared as axially infinite models.Note that for the purposes of this analysis, the term "peak reactivity" is defined as the reactivity of a fuel assembly lattice in the SEP storage rack geometry as determined by MCNP5-1.51 (using CASMO-4 depletion calculation isotopic compositions which include residual Gd). This peak reactivity considers nominal fuel assembly and storage rack dimensions.

For the purpose of determining the design basis fuel assembly and its bounding lattice (see Section 2.3.1.2 and Section 2.3.1.3), the core operating parameters (COP) are varied using four" sets. For all further calculations using the design basis fuel assembly lattice bounding core operating parameters are used (see Section 2.3.2). Note that the fuel assembly orientation in the core with respect to its control blade does not change and therefore the CASMO-4 depletion calculations consider the only possible configuration.

2.3.1.1.1 Peak Reactivity and Fuel Assembly Burnup Typically, a spent fuel assembly is characterized by its assembly average burnup (over all lattices or nodes). In this analysis methodology the fuel assembly average burnup is of no concern and is not credited for reactivity control. Rather, the methodology credits the residual Gd and other depletion isotopic compositions at the fuel assembly peak reactivity (most reactive lattice peak reactivity).

While the peak reactivity occurs at some specific lattice burnup, the peak reactivity lattice burnup varies from lattice to lattice withain a fuel design. Therefore, independent calculations with MCNP5-1 .51 using pin specific compositions (see Section 2.3.1.1.2) are performed for every lattice that is selected as a result of the screening calculations (see Section 2.3.1.2) and all further design basis calculations using MGNP5-1.51.

The MCNPS-1.51 calculations are performed over a burnup range to determine the burnup at peak reactivity for every lattice in the storage rack geometry.

Since each lattice is considered at its peak reactivity (and therefore the lattice or nodal burnup at which that occurs), the fuel assembly average burnup or fuel assembly burnup profile is not applicable because the analysis already considers each lattice at its most reactive composition, independent of the fuel assembly average burnup.Project No. 2393 Report No. 111-2146153 Page 6 Holtec International Proprietary Information 2.3.1.1.2 Isotopic Compositions The BWR fuel design lattices used at Dresden 2 and 3 have complex radial pin compositions.

The radial variation includes enrichment, Gd rod location and loading, part length rods, etc.Furthermore, the fuel assemblies are asymmetric and are designed to a specific control blade orientation.

All fuel compositions are at 0 hours0 days 0 hours 0 weeks 0 months cooling time with the exception of one study to show that this is conservative (see Section 2,3.1.4).

For all calculations in the spent fuel pool racks, the Xe- 135 concentration in the fuel is conservatively set to zero and the Np-239 isotope was considered as Pu-239.2.3.1.2 Screening Calculations for the Design Basis Fuel Assembly The SFP holds various legacy fuel assemblies designs, the current Optima2 design and the future ATRIUM 10OXM design to be qlualified for storage. For many of the legacy fuel designs, it is not necessary to perform calculations because they have a very low lattice average enrichment.

Since it is known that the design basis lattice will have a high lattice average enrichment, a simple assessment of the legacy fuel population is all that is required to determine that they are bounded by the design basis lattice. Therefore, for legacy fuel designs with low latticc enrichments (i.e. less than about fl % U-235), engineering judgment is used to determine that these designs will not need screening calculations since they are well bounded by the more recent fuel designs with much higher lattice average enrichments.

For all of fuel design lattices that require screening calculations, the first step (Step 1) is to perform CASMO-4 calculations to determine the lattices that have the highest peak reactivity in the storage rack geometry (see Appendix A). For Step 1, an arbitrary value of kif > 0.8500 is used to determine the lattices that have the highest peak reactivity in the storage rack geometry.This arbitrary value was selected using engineering judgment.Each of the Step I screening calculations using CASMO-4 includes the in core depletion and restart in SFP rack cell. Note that for the core depletion calculations, four sets of core operating parameters are used and the maximum reactivity over all four is determined (see Section A.2).These four sets of core operating parameters are presented in Table 5 .2.(c) and have been selected to bound the effects of the most important parameters (i.e. void fraction, control blade use and temperatures).

Based on the results of Step 1, the most reactive fuel lattices are identified by selecting the subset of lattices that have a reactivity greater than 0.8500 (see Appendix A). The lattices wvhich meet this criteria are then used for Step 2 calculations as described below.2.3.1.3 Determination of the Design Basis Fuel Assembly Lattice As discussed in Section 2.3.1.2, the Step 1 screening calculations are performed with CASMO-4 for each of the selected lattices.

Based on the results of these screening calculations, the most reactive lattices are determined by comparison to the criteria of kn :> 0.8500. Step 2 calculations are then performed using in-rack MCNP5-1 .51 to determine the peak reactivity for each of the most reactive lattices selected in Step I. See Appendix B.Project No. 2393 Report No. 111-2146153 Page 7 Hloltec International Proprietary Information Step 2 determines the peak reactivity for the most reactive lattices using MCNP5-l.51 calculations in the storage rack geometry.

Note that the peak reactivity of the CASMO-4 depletion calculation model is used only for the screening calculations and is not the peak reactivity as determined by MCNP5-1.51 in rack models. MCNP5-1.51 calculations are performed over a burnup range to independently determine the peak reactivity.

The bounding set of COP determined by Step I in the CASMO-4 screening calculations is confirmed to be consistent with those in Step 2. See Appendix B.The result of the Step 2 calculations are then compared, and the most reactive fuel assembly lattice is determined.

Note that the results of the Step 2 lattice calculations in MCNP5.-1 .51 are useful to show important trends in the reactivity effect of lattice enrichment, Gd rod location, number and loading. These trends are expected to show that the most reactive lattices are those with the highest lattice average enrichment, lowest number of Gd rods and lowest Gd rod loading. The most reactive lattice is then used to construct a new lattice that is much more bounding by increasing the lattice average enrichment to the maximum value (i.e. U wt% U-23 5), decreasing the number of Gd rods to the minimum expected (i.e. II) with the minimum expected Gd loading (i.e. I1%). This new constructed lattice is then used as the design basis fuel assembly lattice and is modeled along the entire active length for all calculations used to determine ker at the 95/95 level.2.3.1.4 Design Basis Model The analysis design basis MCNP5-1 .51 model is a 2x2 array (and larger array sizes as noted below) that considers the formed and fabricated cell design of the storage racks. The storage rack cell wall, poison, and sheathing are all explicitly modeled along the active length of the design basis lattice. The BORAL panels are considered at their minimum thickness and loading.The design basis model explicitly considers the fuel pellet, pellet to cladding gap, cladding, water box and fuel assembly channel (unless otherwise noted below). Various studies are performed with the design basis model to determine the reactivity effect of SFP water, radial position of the fuel assembly within the storage cell, and radial orientation of the fuel in the 2x2 array with respect to the corner of the bundle which was adjacent to the control blade in the core.The reactivity impacts fr'omr these studies are discussed in detail in the sections below. The MCNP5-l.51 model uses periodic boundary conditions radially and 12 inches of water as axial reflectors.

The assembly lattice is considered along the full active length. The storage rack is considered along the full active fuel length only.The design basis model is used for all calculations used to show compliance with the regulatory limit. All calculations with the design basis model are presented in Appendix C. The design basis model differs slightly from the model used to determine the bounding lattice (i.e., the gaseous and volatile isotopes (see Table 5.4(b)) are removed from the spent fuel composition (see Appendix B).Calculations are performed with the design basis model for the four sets of COP to confirm the selection of the bounding set from Appendix B. The design basis MCNP5-1 .51 model is Project No. 2393 Report No. 1-11-2146153 Page 8 Holtec International Proprietary Information presented in Figure 2.2. Note that all calculations are performed at zero hours cooling time.Justification of this cooling time is also presented in Appendix C.The following cases are considered:

Case 2.3.1.4.1:

This is the design basis model. It is a 2x2 array cases MCNP5-1.51 with the fuel assembly centered in the rack cell. The COP used is the "mai" set (see Table 5.2(c)). See Figure 2.2.* Case 2.3.1.4.2:

Same as Case 2.3.1.4.1 except that the COP used are in "nom" set.* Case 2.3.1.4.3:

Same as Case 2.3.1.4.1 except that the COP used are in "max"~ set.* Case 2.3.1.4.4:

Same as Case 2.3.1.4.1 except that the COP used are in "minr" set.* Case 2.3.1.4.5:

Same as Case 2.3.1.4.1 except that the isotopic compositions are at 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months cooling time.The results of these calculations are presented in Table C. 1. The results presented in TFable C.1 also provide the bounding case from Appendix 13 so that a comparison can be made between the two calculations.

2.3.2 Core Operating Parameters As previously discussed, CASMO-4 is used to perform depletion calculations to determine the spent fudel isotopic composition.

The operating parameters for spent fuel depletion calculations are discussed in this Section. The core operating parameters which may have a significant impact on BWR spent fuel isotopic composition are void fraction, control blade history, moderator temperature, fuel temperature, and power density. Other parameters such as the effect of burnable absorbers and axial enrichment distribution are discussed in Section 2.3.3 and Section 2.3,4, respectively.

For the purpose of determining the bounding set of COP for each lattice, four sets of COP are used (see Table 5.2(c)). The bounding set of COP is determined using both CASMO-4 and MCNP5-1 .51 calculations (see Appendix A and Appendix B),. The bounding set of COP for the design basis lattice is used for all design basis lattice calculations (see Appendix C).2.3.3 Integral Reactivity Control Devices The only type of burnable absorber used for the fuel assemblies covered in this analysis is Gd.The use of Gd does not increase the reactivity of the assembly, compared to an assembly lattice where all rods contain fuel and no Gdl. As discussed in Section 2.3.1.1.1, the Gd in the fuel assembly holds down the fresh fuel assembly reactivity and then, as core depletion occurs, the Gd begins to burnout until it is essentially fully depleted.

As the Gd depletes the reactivity of the fuel assembly increases until it reaches a peak. This peak reactivity is the fuel assembly's most reactive condition, which is used for design basis condition.

Project No. 2393 Report No. HIl-2J146153 Page 9 H-oltec International Proprietary Information 2.3.4 Axial and Planar Enrichment Variations All calculations were performed with the design basis fuel assembly lattice pin specific enrichment(s), without any axial variation.

2.3.5 Fuel Assembly Eccentric Positioning and Fuel Assembly De-Channeling The BWR fulel that is loaded in the SFP racks may not rest exactly in the center of the storage cell, therefore the potential reactivity effect of this eccentric positioning should be evaluated.

The ATRIUM 10OXM fuel assembly (thle most reactive fuel assembly, as will be shown in Section 7) may be de-channeled, therefore the potential reactivity effect of de-channeling should be evaluated.

These two parameters, storage cell eccentric positioning and the fuel assembly de-channeling may occur simultaneously and may impact the reactivity effect of each other.Therefore the two parameters should be evaluated together.

Evaluations are therefore performed to determine the most limiting fuel radial location for fuel with and without a channel.The following cases with the fuel assembly channel present are analyzed:* Case 2.3.5.1: This is the reference for the 2x2 array cases, Case 2.3.5.2 and Case 2.3.5.3.The MCNP5- 1.51 model used herein is a 2x2 array with the fuel assembly centered in the rack cell. This model is the same model as the design basis model. See Figure 2.2.o Case 2.3.5.2: Every fuel assembly is positioned toward the center as shown in Figure 2.3.* Case 2.3.5.3: Every fuel assembly is positioned toward one corner as shown in Figure 2.4.* Case 2.3.5.4: This is the reference for Case 2.3.5.5 and Case 2.3.5.6. The MCNP5-l.51 model used herein is an 8x8 array with the fuel assembly centered in the rack cell. The model is the same as the design basis model but the array size is larger.* Case 2.3.5.5: Every fuel assembly is positioned toward the center as shown in Figure 2.5.* Case 2.3.5.6: Every fuel assembly is positioned toward one corner as shown in Figure 2.6.The following cases with the fuel assembly channel NOT present are analyzed:*Case 2.3.5.7: This is the reference for the 2x2 array cases, Case 2.3.5.8 and Case 2.3.5.9.The MCNP5-1.51 model used herein is a 2x2 array with the fuel assembly centered in the rack cell. This model is the same model as the design basis model except that the fuel channel has been removed.* Case 2.3.5.8: Every fuel assembly is positioned toward the center as shown in Figure 2.7.Project No. 2393 Report No. I--2146153 Page 10 H-oltec International Proprietary Information

Case 2.3.5.9: Every fuel assembly is positioned toward one corner as shown in Figure 2.8.* Case 2.3.5.10:

This is the reference for Case 2.3.5.11 and Case 2.3.5.12.

The MCNP5-1.51 model used herein is an 8x8 array with the fuel assembly centered in the rack cell.The model is thle same as the design basis model but the array size is larger.* Case 2.3.5.11:

Every fuel assembly is positioned toward the center as shown in Figure 2.9.* Case 2.3.5.12:

Every fuel assembly is positioned toward one corner as shown in Figure 2.10.The maximum positive reactivity effect of the MCNP5-1 .51 calculations for the fuel eccentric positioning and de-channeling is added as the bias and the corresponding 95/95 uncertainty is statistically combined with other uncertainties to determine korf.2.3.6 Fuel Bundle Orientation in SFP Rack Cell As described in Section 2.3.1.1.2, fuel asselmblies have various radial fuel enrichments and gadolinium distribution.

Also, one corner of each fuel assembly is adjacent to the control blade during the depletion in the core. As a result, the fuel depletion is not uniform and therefore one fuel assembly corner may be more reactive than other corners and the fuel assembly orientation in the SFP storage cell may have an impact on reactivity.

Five cases are analyzed to assess the fuel assembly orientation variations and to determine the most limiting fuel orientation in SFP rack cell.The MCNP5-1 .51 model of the reference case is the design basis fuel in the 2x2 array, as shown in Figure 2.2. The MCNP5,1.51 models of the other four cases are the same as that of the reference case, except with different orientations.

The following cases are considered:

Case 2.3.6.1: This is the reference for the 2x2 array cases, Case 2.3.6.2 through Case 2.3.6.5. This model is the same model as thle design basis model where the corner of the lattice adjacent to the control blades in the core is oriented towards the north west. See Figure 2.2.*Case 2.3.6.2: The fuel assembly in each cell in the 2x2 array is oriented as shown in Figure 2.11.* Case 2.3.6.3: The fuel assembly in each cell in the 2x2 array is oriented as shown in Figure 2.12.,, Case 2.3.6.4: The fuel assembly in each cell in the 2x2 array is oriented as shown in Figure 2.13.Project No. 2393 Report No. -Il-21461 53 Page 11!Holtec International Proprietary Information

Case 2.3.6.5: The fuel assembly in each cell in the 2x2 array is oriented as shown in Figure 2.14.Note that the evaluations use the same MCNP5-1 .51 models with periodic boundary conditions used in the design basis calculation.

The isotopic compositions of the fuel rods are thle same as those of the design basis fuel assembly.The maximum positive reactivity effect of the MCNP5-l .51 calculations for the fuel bundle orientation is added as the bias and the corresponding 95/95 uncertainty is statistically combined with other uncertainties to determine kcff.2.3.7 Reactivity Effect of Spent Fuel Pool Water Temperature The Dresden Station SFP has a normal pool water temperature operating range below 150 0 F.For the nominal condition, the criticality analyses are to be performed at the most reactive temperature and density [6]. Also, there are temperature-dependent cross section effects in MCNP5-1 .51 that need to be considered.

In general, both density and cross section effects may not have the same reactivity effect for all storage rack scenarios, since configurations with strong neutron absorbers typically show a higher reactivity at lower water temperature, while configurations without such neutron absorbers typically show a higher reactivity at a higher water temperature.

For the SF1P racks which credit neutron absorbers, the most reactive SFP water temperature and density is expected to be at 39.2 "'F and 1 g/cc, respectively.

The standard cross section temperature in MCNP5-I .51 is 293.6 K. Cross sections are also available at other temperatures; however, not usually at the desired temperature for SF1P criticality analysis.

MCNP5-l .51 has the ability to automatically adjust the cross sections to the specified temperature when using the TMP card. Furthermore, MCNP5-1 .51 has the ability to make a molecular energy adjustment for select materials (such as water) by using the S(ct,13) card.The S(c,43) card is provided for certain fixed temperatures which are not always applicable to SFP criticality analysis.

Rather, there are limited temperature options, i.e., 293.6 K and 350 K, etc. Additionally, MCNP5-1.51 does not have the ability to adjust the card for temperatures as it does for the TMP card discussed above. Therefore, additional studies are performed to show the impact of the S(a,f3) card at the two available temperatures.

To determine the water temperature and density which result in the maximum reactivity, MCNP5-1 .51 calculations are run using the bounding values. Additionally, S(o,13) calculations are performed for both upper and lower bounding S&4,3) values, if needed. Additional eases are added to cover the potential increase in temperature beyond normal conditions (i.e. accident condition).

The following cases are considered:

Case 2.3.7.1 (reference case): Temperature of 39.2 0 F (277.15 K) and a density of 1.0 g/cc are used to determine the reactivity at the low end of the temperature range. The S(ct,13) card corresponds to a temperature of 68.81 0 F (293.6 K).Project No. 2393 Report No. 141-2146153 Page 12 H-oltec International Proprietary Information

Case 2.3.7.2: Temperature of. U F K) and a corresponding density of g/cc are used to determine the reactivity at the high end of the temperature range. The S(a,13) card con'esponds to a temperature of 68.81 0 F (293.6 K).*Case 2.3.7.3: Temperature of. U F (K) and a corresponding density glcc. The S(cL,f3) card corresponds to a temperature of 170.33 0 F (350 K).* Case 2.3.7.4: Temperature of 212 0 F (373.15 K) and a corresponding density of 0.95837 g/cc, The S(a,13) card corresponds to a temperature of 170.33 °F (350 K). This is a SEP water temperature accident condition.

Case 2.3.7.5: Temperature of 212 0 F (373.15 K) and a corresponding density of 0,95837 g/cc. The S@4t,3) card corresponds to a temperature of 260.33 0 F (400 K). This is a SEP water temperature accident condition.

Case 2.3.7.6: Temperature of 255 °F (397.04 K) and a corresponding density of 0,84591 g/cc. The card corresponds to a temperature of 260.33 0 F (400 K). In this model, it is assumed that the water modeled includes 10% void. Void is modeled as 10%decrease in density, compared to the density of water at 255 °F. This is a SEP water temperature accident condition.

T'he hounding water temperature and density (the temperature and its corresponding density which result in the maximum reactivity) of the above cases are applied to all further calculations so that the most reactive water temperature and density is considered.

Note that the evaluations use the same MCNP5.-l.51 models used in the design basis calculation.

The pin specific isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.2.3.8 Fuel and Storage Rack Manufacturing Tolerances In order to determine the keff of the SFP at a 95% probability at a 95% confidence level, consideration is given to the effect of the BWR fuel and SFP storage rack manufacturing tolerances on reactivity.

The reactivity effects of significant independent tolerance variations are combined statistically

[6]. The evaluations use the same MCNP5-.1.51 models used in the design basis calculation.

2.3.8.1 Fuel Manufacturing Tolerances The BWR fuel tolerances for ATRIUM 10XM design basis lattice (which is the most reactive fuel design evaluated herein) are presented in Table 5.1(h). Fuel tolerance calculations are performed using the design basis fuel assembly lattice only because the reactivity of the design basis lattice is much greater than lattices from other fuel bundle designs. Therefore, only the tolerances applicable to that lattice are applicable.

Separate CASMO-4 depletion calculations are performed for each fuel tolerance and the full value of the tolerance is applied for each case in both the depletion and in rack calculations.

Pin specific compositions are used. The MCNP5-1 .51 tolerance calculation is compared to the MCNP5-l1.51 reference case (nominal parameter values)at the 95% probability at a 95% confidence level using the following equation: Project No. 2393 Report No. 1-I-2146153 Page 13 Holtec International Proprietary Information delta-kcajc

= (kcalc2 -kcajci) +- 2 * -1(0q2 + a2 2)The following fuel manufacturing tolerances cases are considered in this analysis:* Case 2.3.8.1.1 (reference case): This is the reference for all the other fuel tolerance cases.This MCNP5-l,51 model is the same model as the design basis model. See Figure 2.2.* Case 2.3.8.1.2:

This is the fuel pellet density increase tolerance.

Case 2.3.8.1.3:

This is the fuel pellet diameter increase tolerance.

Case 2.3,8.1.4:

This is the fuel pellet diameter decrease tolerance.

Case 2.3.8.1,5:

This is the minimum cladding thickness tolerance.

In this model, the maximum cladding inner diameter and minimum cladding outer diameter are applied together,* Case 2.3.8.1.6:

This is the increased rod pitch tolerance.

Case 2.3.8.1.7:

This is the decreased rod pitch tolerance.

Case 2.3.8.1.8:

This is the increased channel thickness tolerance.

Case 2.3.8.1.9:

This is the decreased channel thickness tolerance.

o Case 2.3.8.1.10:

This is the increased fuel enrichment tolerance.

All fuel pins have an increase in U-235 enrichment, including the Gd rods, of 0.05 wt% U-235.* Case 2.3.8.1.11

This is the decreased Gd loading tolerance.

The maximum positive reactivity effect of the MCNP5-1 .51 calculations for each tolerance is statistically combined with the other tolerance results, and this result is then statistically combined with other uncertainties when determining the kcff value.2.3.8.2 SFP Storage Rack Manufacturing Tolerances The SEP rack tolerances are presented in Table 5.3. The full value of the tolerance is applied for each case. The MCNP5-1 .51 tolerance calculation is compared to the MCNP5-l1.51 reference case with a 95% probability at a 95% confidence level using the following equation: delta-kca~o

= (kca 1 c 2 -1) +/-- 2

2 + 0y2)The following rack manufacturing tolerances cases are considered in this analysis: Project No. 2393 Report No. 1-1-2 1461]53 Page 14 Iloltec International Proprietary Information

Case 2.3.8.2.1 (reference case): This is the reference for all the other rack tolerance cases.This MCNP5-l.51 model is the same model as the design basis model. See Figure 2.2.* Case 2.3.8.2.2:

This is the increased storage cell inner diameter (ID) tolerance.

Case 2.3.8.2.3:

This is the decreased storage cell inner diameter tolerance.

Case 2.3.8.2.4:

This is the increased wall thickness tolerance.

Note that the tolerance associated with the wall thickness is assumed to be 10% of the wall thickness.

Case 2.3.8,2.5:

This is the decreased wall thickness tolerance.

Note that the tolerance associated with the wall thickness is assumed to be 10% of the wall thickness.

Case 2.3.8.2.6:

This is the increased storage cell pitch tolerance.

.. Case 2.3.8.2.7:

This is the decreased storage cell pitch tolerance.

Case 2.3.8.2.8:

This is the increased BORAL width tolerance.

Case 2.3.8.2.9:

This is the decreased BORAL width tolerance.

The maximaum positive reactivity effect of the MCNP5- 1.51 calculations for each tolerance is statistically combined with the other tolerance results, and this result is then statistically combined with other uncertainties when determining the ku-r value.The evaluations use the same MCNP5-1 .51 models used in the design basis calculation.

The isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.The poison thickness and loading are used at their minimum values for all calculations; i.e., they are treated as a bias instead of uncertainty, for conservatism and simplification.

2.3.9 Fuel Depletion Calculation Uncertainty To account for the uncertainty of the number densities in the depletion calculations performed in CASMO-4, a 5% depletion uncertainty factor as described in [6] and f 7] is used. Note that an additional uncerztainty factor is used to account for the uncertainty in the cross sections; for fission products see Section 2.3.10.The depletion uncertainty is applied by multiplying it with the reactivity difference (at 95%/95%) between the MCNP5- 1.51 calculation with spent fuel at peak reactivity (includes residual Gd) and a corresponding MCNP5-1.51 calculation with fresh fuel (without Gd 2 0 3).The uncertainty is determined by the following:

Uncertainty Jsotopics

= [ (kcaj.e-2 -kcdle-l)

+ 2 * ."J (o'cale.2 + ]

0.05 Project No. 2393 Report No. H-I-2146153 Page 15 Ho-lotec International Proprietary Information with kcaic-i =- kl with spent fuel k~alo-2 =k~, 0 with firesh fuel Ocalc-1 Standard deviation of k~a 1 0-1= Standard deviation of 2 The following case is considered:

Case 2.3.9.1 (reference case): This is the reference case. This MCNP5-1.51 model is the same model as the design basis model. See Figure 2.2.* Case 2.3.9.2: This is the fresh fuel with no Gd case.The result of the MCNP5-1 .51 calculation for the fuel depletion calculation uncertainty is statistically combined with other uncertainties to determine kerr.2.3.10 Fission Products and Lumped Fission Products Uncertainty Few relevant critical experiments are p~ublicly available for fission products (FP) and minor actinides, and therefore direct validation similar to the actinide validation is not feasible and cannot be directly included in the MCNP5.-1 .51 benchmark bias and bias uncertainty.

The uncertainty in the reactivity worth of FP and minor actinides isotopes is determined based on consideration of uncertainties of cross sections of FPs documented in 1191. The overall uncertainty is derived fr'om the uncertainty associated with each individual isotope's cross section for all FPs and lumped fission products (LFP) and is detenrmined at a 95% probability at a 95% confidence level. Based on the discussion and evaluation presented in [IO0], an uncertainty value of E% is used for both the FPs and LFPs. Note that no statistical approach is used here, i.e., the uncertainty is applied equally to the effect of all FPs (including minor actinides) and LFPs. Also note th~at recent studies [11, 12] indicate that the total cross section uncertainty for 16 prominent fission products is only about 1.5% (one standard deviation) at 95% probability at a 95% confidence level.The uncertainty is applied by multiplying it with the reactivity difference (at 95%/95%) between the MCNP5-1 .51 calculation with all isotopes and a corresponding MCNP5-1 .51 calculation where all FPs and LFPs have been removed. The MCNP-lI.51 model is the same as the design basis model. The uncertainty of the calculations is calculated using the following equation: Uncertainty

= [ (kcaic.-z

-kaic.i) + 2 * (Oci 2 + )] *U with ka- = kcajc with FPs and LFPs included keaIe-2 = kea 1 e with FPs and LFPs removed 0 Ycalc-1 = Standard Deviation of kea 1 e-Uca,)c2 = Standard Deviation of kcaI¢-2 Project No. 2393 Report No. HI1-2146I 53 Page 16 1Holtec International Proprietary Information The following case is considered:

Case 2.3.10.1 (reference case): This is the reference case. This MCNP5-1.51 model is the same model as the design basis model. See Figure 2.2.* Case 2.3.10.2:

This is the spent fuel with FP/LFP removed case.The result of the MCNP5-1 .51 calculation for the FP and LFP calculation uncertainty is statistically combined with other uncertainties to determine kcff.All cases analyzed here have neutron spectra in the thermal energy range and the fission products are predominantly thermal absorbers.

Additionally, fission processes are affected by the resonance integrals of the absorbers.

The fission product cross section uncertainty is evaluated for the thermal neutron energy range and the resonance integral.

The uncertainty is therefore directly applicable to the calculations performed here.2.3.11 Depletion Related Fuel Assembly Geometiy Changes During irradiation the BWR fuel assembly may experience depletion related fuel geometry changes. These changes can be fuel rod growth and cladding creep, crud buildup, fulel rod bow and the fuel channel may bow and bulge. These fuel assembly geometry changes can affect the neutron spectrum during depletion by changing the fuel to moderator ratio. In the spent fuel pool, there are two potential impacts from the depletion related fuel geometry changes: first, the effect during depletion may lead to a different isotopic composition, second, the fuel geometry change itself can also impact reactivity by the change in the fuel to moderator ratio. The effect of these possible fuel geometry changes on the reactivity of the fuel in the SFP are discussed below.Note that since the peak reactivity for the design basis fuel assembly is below fl GWd/mtU (i.e.is about fl GWd/mtU), there is no expected significant reactivity impact associated with any minimal fuel geometry changes which occur below that exposure value.2.3.11.1 Fuel Rod Geometry Changes Possible changes to the fuel rod geometry may occur as a result of fuel rod growth, cladding creep, and crud buildup. These geometry changes have the potential to change the fuel-to-moderator ratio in the geometry, thus potentially increasing reactivity, and are therefore discussed below.2.3.11.1.1 Fuel Rod Growth and Cladding Creep Fuel rod growth and cladding creep is not expected for the design basis lattice at the peak reactivity burnup (i.e. about U GWd/mtU).

Therefore, no additional calculations are performed.

P'roject No. 2393 Report No. 1-1-2146153 Page 17 H-oltec International Proprietary Information 2.3.1 1.1.2 Fuel Rod Crud Buildup Crud buildup on the fuel rod cladding decreases the amount of water around the fuel rods and thus increases the fuel-to-moderator ratio. The amount of crud buildup at peak reactivity is not expected to be significant.

Therefore, no further evaluations are performed.

2.3.11.1.3 Fuel RodlBow Fuel rod bow is a depletion related geometry change that alters the fuel rod pitch. The effect of the fuel rod bow is similar to the fuel rod crud buildup (see Section 2.3.11.1.2).

The reactivity impact ofthis geometry change to the fuel in the SEP is evaluated using the depletion related fuel rod pitch positive tolerance provided in Table 5.1 (h).The following fuel rod bow cases are considered:

Case 2.3.11.1.3.1 (reference case): This is the reference case. This MCNP5-l.51 model is the same model as thle design basis model. See Figure 2.2.* Case 2.3.11 .1.3.2: This is the fuel rod bow case. The isotopic compositions are taken fr'om CASMO4 runs with this geometry change included.

The geometry change is also included in the geometry of the MCNP5-1 .51 model.The results of the MCNP5-1 .51 calculations are used to determine a bias and bias uncertainty.

The bias and bias uncertainty are applied to the design basis results as discussed in Section 2.3.13.The maximum positive reactivity effect of the MCNP5-1 .51! calculations for the fuel rod bow is added as the bias and the corresponding 95/95 uncertainty is statistically combined with other uncertainties to determine kerr.2.3.11.2 Fuel Channel Bulging and Bowing Fuel channel bulging and bowing is a depletion related geometry change that changes the proximity of the channel to the fuel rods. Since the proximity of the channel relative to the fuel rods may change, the temperature and density of the moderator during depletion may change (volume of moderator inside the channel may change). The reactivity effect of fuel channel bulging and bowing is evaluated using the channel outer exposed width tolerance presented in Table 5.1 (h).The following fuel channel bulging and bowing cases are considered:

Case 2.3.11.2.1:

This is the fuel channel bulging and bow case. The isotopic compositions are taken from CASMO4 runs with this geometry change included.

The geometry change is also included in the geometry of the MCNP5-1 .51 model.Project No. 2393 Report No. 1-11-21 46153 Page 18 Hloltec International P~roprietary Information The results of the MCNP5-l.51 calculations are used to determine a bias and bias uncertainty.

The bias and bias uncertainty are applied to the design basis results as discussed in Section 2.3.13.The maximnum positive reactivity effect of the MCNP5-1 .51 calculations for the fuel channel bulging and bowing is added as the bias and the corresponding 95/95 uncertainty is statistically combined with other uncertainties to determine krfc.2.3.12 SEP Storage Rack Interfaces The Dresden SFP storage racks are all the high density egg crate design. BORAL panels are fixed to the outside of all fabricated cells and these fabricated cells are joined to create formed cells. Along the outside of each rack module, BORAL panels are not fixed to the locations where the formed cells reach the edge, thus there is no BORAL panel every other location.

For each rack module, the fabricated cell is placed in each corner of the mnodule so that there is always a BORAL panel beginning and ending each rack module edge. For the location where the formed cell is along the rack module edge there is a steel filler plate welded to cover the hole.The rack design method creates a configuration where there may be no BlORAL between two fuel bundles in adjacent rack mnodules, only the steel filler plates. Therefore, the reactivity effect of this interface condition is evaluated.

The following interface cases are considered:

Case 2.3.12.1.

The MCNP5-1.51 model is a 16x16 array model. The array is the same as the design basis model except that along every 8 columns of cells every other location has both BlORAL panels removed. The two steel sheathings were left in the model to represent the steel plate. Thus, the steel plate thickness considered in the model is thinner than the actual steel plate (see Table 5.3). Note that in this model the gap between racks is not included in the model at all. All fuel is cell centered.

See Figure 2.15.* Case 2.3.12.2:

This is the same as Case 2.3.12.1 except the fuel is eccentric towards the center of the model.For the purpose of the interface calculations, two 1 6x 16 array models that are larger arrays of the design basis model (one cell centered and one with the fuel eccentric towards the center of the model), are used as reference cases. The results of the MCNP5-1 .51 calculations are used to determine a bias and bias uncertainty.

The maximum positive reactivity effect of the MCNP5-1 .51 calculations for the storage rack interface is added as the bias and the corresponding 95/95 uncertainty is statistically combined with other uncertainties to determine ker Project No. 2393 Report No. HI-2146153 Page 19 Holtec International Pr'oprietary Information 2.3.13 Maximum lkfc Calculation for Normal Conditions The calculation of thle maximum kef" of the SFP storage racks fully loaded with design basis fuel assemblies at their maximum reactivity is determined by adding all uncertainties and biases to the calculated reactivity.

Note that the BORAL thickness and its B-10 loading are taken at their worst case values in all design basis cases.koff is determined by the following equation: keff kea 1 e + uncertainty

+ bias where uncertainty includes:* Fuel manufacturing tolerances

SFP storage rack manufacturing tolerances

Fuel eccentricity bias uncertainty

Fuel orientation bias uncertainty

Fuel channel bow bias unceitainty 9 0 Fuel rod bow bias uncertainty Depletion calculation uncertainty FPs and LFPs uncertainty MCNP5- 1.51 bias uncertainty (95% probability at a 95% confidence level)MCNP5-1 .51 calculations statistics (95% probability at a 95% confidence level, 2cr)Interface bias uncertainty and the bias includes* Fuel eccentricity bias* Fuel orientation bias* Fuel channel bow bias* Fuel rod bow bias ,, MCNP5-1.51 bias* Interface bias Note that each uncertainty is statistically combined with other uncertainties, while biases are added together in order to determine ken".The approach used in this analysis takes credit for residual Gd at peak reactivity.

2.3.14 Fuel Movement, Inspection and Reconstitution Operations Fuel movement procedures govern the movement and inspection of the fuel at all times that the fuel is onsite. The new fuel enters the SFP via the fuel prep machine (FPM). The FPM has a single fuel assembly capacity.

There are two FPMs in each SFP, which could be loaded with fuel at the same time. However, the FPMs are greater than U feet apart, which is a low reactivity Project No. 2393 Report No. t-11-2146153 Holtec International Proprietary Info~rmation Page 20 configuration because of the distance between either PPM so no further analysis beyond the normal condition is necessary.

The fuel is then picked up by the refueling platform, which also has a single fuel assembly capacity at any given time, and moved into a storage location in the storage rack. The fuel is always moved above the rack and never moved along the side of the rack. Prom the storage rack, the fuel is picked up by the refueling platform and moved through the refueling slot for transport to the core. The return trip uses the same process in reverse. All of these fuel movement operations involve a single fuel assembly that is never in close enough (i.e., directly adjacent) proximity to any other fuel that the configuration is not bounded by the analysis for normal conditions.

The PPM is not considered to be a long-term storage location for fuel but it is physically possible that a fuel assembly in the PPM. could be approached by another fuel assembly in the refueling platform.

The FPM is only single capacity; therefore, once a fuel assembly is in the P'PM there is no normal operation that would allow the presence of another fuel assembly in close proximity to the PPM. This configuration (i.e., two fuel bundles in or around a PPM) is not considered a normal configuration.

Due to the location of the PPM, only one of the two refueling platforms can ever physically use the PPM at any given time. Furthermore, dimensions for distance fr'om the PPMs to the nearest SFP rack is II inches, which is more than the dimensions of a fuel assembly.2.3.15 Accident Condition The accidents considered are:* SFP temperature exceeding the normal range* Dropped assemblies

Missing BORAL Panel* Rack movement* Mislocated fuel assembly (a fuel assembly in the wrong location outside the storage rack, including the platform area)Those are briefly discussed in the following sections.Note that the double contingency principle as stated in [6] specifies that "two unlikely independent and concurrent incidents or postulated accidents are beyond the scope of the required analysis." This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

The koff calculations performed for the accident conditions are done with a 95%probability at a 95% confidence level.The accident conditions are considered at the 95195 level using the total corrections from the design basis case. Note that the design basis lattice is used for the accident analyses.Project No. 2393 Report NO. H-1-2146153 Page 21 Iloltec International Proprietary Information 2.3.15.1I Temperature and Water Density Effects The SEP water temperature accident conditions for consideration are the increase in SFP water temperature above the maximum SFP operating temperature of[ U F (the decrease in temperature was already considered for the temperature coefficient determination as discussed in Section 2.3.7).The increase in SEP temperature accident cases are discussed in Section 2.3.7 and are bounded by the calculations at reduced temperature.

2.3.15.2 Dropped Assembly -Horizontal For the ease in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a separation distance between the fueled portions of the two assemblies of more than 12 inches. Thus, the horizontally dropped assembly is decoupled from the fuel assemblies in the rack. This accident is also bounded by the mislocated case, where the mislocated assembly is closer to the assembly in the racks. Therefore, the horizontally dropped fuel assembly is not evaluated further in the report.2.3.15.3 Dropped Assembly-Vertical into an Empty Storage Cell It is also physically possible to vertically drop an assembly into a location that might be empty and such a drop may result in deformation of' the rack baseplate.

In that case some part of'the active fuel length may extend beyond the BORAL panel out of the bottom of the rack. This potential configuration is physically similar to the normal condition of insertion and removal of fuel fr'om the storage rack. In thae normal condition of insertion and removal of a fuel assembly from the storage cell, the active fuel in the rack remains well within the length of the BORAL panels, while the part of the moving fuel bundle that is above the length of the B3ORAL panel is physically separated from the fuel in the rack by a sufficient amount of water to preclude neutron coupling.

For the case where the fuel assembly is dropped into an empty cell, the fuel assembly could potentially break through the baseplate.

The design of the rack is such that each storage cell location has a baseplate that is not connected with the adjacent cells. Therefore, this accident condition is physically the same as the normal condition of insertion and removal of fuel in the rack. However, this case is considered to show that there is no reactivity effect associated with this configuration.

The following vertical drop cases are considered:

Case 2.3.15.3.1:

This MCNP5-l.51 model is the same model as the design basis model but the array is 16x16. In the center location, the active length is extended below the active length of the other fuel by the thickness of the baseplate and the distance from the baseplate to the pool floor (see Table 5.3). All fuel is centered in the storage cell. See Figuare 2.16.* Case 2.3.15.3.2:

Same as Case 2.3,15.3.1 but the fuel is eccentric in the storage cell towards the dropped fuel.Project No. 2393 Report No. HI-21461 53 Page 22 IHoltec international Proprietary Inf'ormation 2.3,15.4 Missing BORAL Panel The missing BORAL panel accident is considered to cover the potential that a BORAL panel may have been inadvertently not installed during construction of the rack or that a panel might become dislodged by some other accident force.The following cases are considered:

Case 2.3.15.4.1:

This MCNP5-l.51 model is the same model as the design basis model but the array is 8x8. The cell in the center of the model has I BORAL panel removed.All fuel is centered in the storage cell. See Figure 2.17.* Case 2.3.15.4.2:

This is the same as Case 2.3.15.4.1 but the fuel is eccentric toward the missing BORAL panel.2.3.15.5 Rack movement The racks may move due to seismic activity and the gaps between racks may close. However, the design basis analysis already considers the interface of the racks without any gap, and therefore this condition is already analyzed.2.3.15.6 Mislocated Fuel Assembly The Dresden SFP layout was reviewed to determine the possible worst case locations for a mislocated fuel assembly.

Five hypothetical locations where a fuel assembly may be mislocated are:* Adjacent to the storage rack side where there is no BORAL panel* In the corner between two racks* In the corner between three racks* Between the SEP rack and the FPM a B~etween the two locations on the FPM.The cited scenarios are evaluated, as follows.2.3.15.6.1 Mislocated Fuel Assembly Adjacent to the Storage Rack A fuel assembly may be nilslocated adjacent to the storage rack in one of the alternating locations where there is no BORAL panel. The reactivity effect of this accident is discussed below.The following cases are considered:

Case 2.3.15.6.1.1:

This MCNP5-1.51 model is the same model as the design basis model but the array is 80x80. The mislocated fuel assembly is placed adjacent to the storage rack on one side, aligned vertically with the fuel in the storage rack and in a location that is face adjacent to a location with no BORAL panel. The fuel in the storage rack is cell centered.Project No. 2393 Report No. 1-1-2146153 Page 23 1-oltec International Proprietary Information

Case 2.3.15.6.1.2:

This is the same as Case 2.3,15.6.1.1 but the fuel in the storage rack is eccentrically positioned toward the center of the model.2.3.15.6.2 Mislocated Fuel Assembly in the Corner between Two Racks There are some places in the SFP, but outside of the racks, where the mislocated fuel assembly may be in the corner between two racks (thus thle mislocated fuel assembly would be adjacent to the fuel assemblies in racks from two sides). To evaluate the effect of the mislocated fuel assembly in the corner between two racks, the following cases are evaluated:

Case 2.3.15.6.2.1:

T'his MCNP5-1.51 model is the same model as the design basis model but the array is 80x80 with a corner cut out to model the junction of two racks. The mislocated fuel assembly is in the corner between two racks. The two rack faces where the fuel assembly is mistocated do not have BORAL panels. This configuration is not physically possible because the racks are designed so that the BORAL panels are always in the first location along the outer edge. However, this model is conservative.

The fuel in the storage rack is cell centered.

See Figure 2.18.o Case 2.3.15.6.2.2:

The M.CNP5-1 .51 model is the same as Case 2.3.15.6.2.1, except with all fuel assemblies inl thle storage rack eccentric toward the misplaced fuel assembly.2.3.15.6.3 Mislocated Fuel Assembly in the Corner between Three Racks There is a location in the SEP where the mislocated fuel assembly may be in the corner between three racks (thus the mislocated fuel assembly would be adjacent to the fuel assemblies in racks from thlree sides, although there is a significant gap for the third face). To evaluate the effect of the mislocated fuel assembly in the corner between three racks, the following cases are evaluated:

Case 2.3.15.6.3.t:

This MCNP5-1.51 model is the same model as the design basis model but the array is 80x80 with a corner cut out to model the junction of three racks. The mislocated fuel assembly is in the comer between the three racks. The two rack faces where the fuel assembly is mislocated do not have B3ORAL panels. This configuration is not physically possible because the racks are designed so that the BORAL panels are always in the first location along the outer edge. However, this model is conservative.

The fuel in the storage rack is cell centered.

See Figure 2.19.* Case 2.3.15.6.3.2:

The MCNP5-l .51 model is the same as Case 2.3.15.6.3.1, except with all fuel assemblies in the storage rack eccentric toward the misplaced fuel assembly.* Case 2.3.15.6.3.3:

The MCNP5-1.51 model is the same as Case 2.3.15.6.3.1, except that the gap between the mislocated fuel assembly and the third rack is closed.* Case 2.3.15.6.3.4:

Thle MCNP5-1.51 model is the same as Case 2.3.15.6.3.3, except with all fuel assemblies in the storage rack eccentric toward the misplaced fuel assembly.Project No. 2393 Report No. 1H1-2 146153 Page 24 Holtec International Proprietary Information 2.3.15.6.4 Mislocated Fuel Assemnbly in the FPM The FPM is located adjacent to the SEP storage racks. The FPM has a fuel assembly capacity of two, where the pitch between the two locations on the FPM is specified in Table 5.3. There is a possibility that a fuel assembly could be mislocated between the two FPM locations or between the FPM locations and the storage rack. Note that the pitch is large enough to preclude neutron coupling between PPM locations.

However, for conservatism, the evaluation of this potential mislocated fuel assembly accident condition considers that the distance between the two FPM locations is reduced to about 12 inches and one of them is face adjacent to a missing BORAL panel location.

The gap between the PPM location and the storage rack is 3I inches.The following PPM mislocated fuel assembly accident cases are considered:

Case 2.3.15.6.4.1:

The FPM mislocated MCNP5-l.51 model is a large 80x80 array. The model includes two PPM fuel assemblies.

No FPM structural materials are considered.

The mislocated fuel assembly is placed between the two PPM fuel assemblies with a small gap (position

1) to the closest location.

The fuel is centered in the SFP storage rack cells, See Figure 2.20.* Case 2.3.15.6.4.2:

This is the same as Case 2.3.15.6.4.1 but the fuel is eccentric in the SEP storage rack cells toward the PPM.*, Case 2.3.15.6.4.3:

This is the same as Case 2.3.15.6.4.1 but the mislocated fuel is at a distance (position

2) from the closest PPM location.* Case 2.3.15.6.4.4:

This is the same as Case 2.3.15.6.4.3 but the fuel is eccentric in the SEP storage rack cells towards the mislocated fuel assembly.* Case 2.3.15.6.4.5:

This is the same as Case 2.3.15.6.4.1 but the mislocated fuel is at a distance (position

3) fi'om the closest PPM location.* Case 2.3.15.6.4.6:

This is the same as Case 2.3.15.6.4.5 but the fuel is eccentric in the SEP storage rack cells toward the mislocated fuel assembly.* Case 2.3.15.6.4.7:

This is the same as Case 2.3.15.6.4.1 but the mislocated fuel is at a distance (position

4) from the closest PPM location.* Case 2.3.15.6.4.8:

This is the same as Case 2.3.15.6.4.7 but the fuel is eccentric in the SFP storage rack cells toward the mislocated fuel assembly.* Case 2.3.15.6.4.9:

This is the same as Case 2.3.15.6.4.1 but the mislocated fuel is directly adjacent to the closest PPM location (position 5). See Figure 2.21* Case 2.3.15.6.4.10:

This is the same as Case 2.3.15.6.4.9 but the fuel is eccentric in the SFP storage rack cells toward the mislocated fuel assembly.Project No. 2393 Report No. 1-11-2146153 Page 25 H-oltec International Proprietary Information

Case 2.3.15.6.4.11

This is the saone as Case 2.3.15.6.4.1 but the mislocated fuel is between the SFP rack and the FPM fuel. The mnislocated fuel is directly adjacent to the SFP storage rack location without a BORAL panel (position 6). See Figure 2.22.* Case 2.3.15.6.4.12:

This is the samne as Case 2.3.15.6.4.11 but the fuel is eccentric in the SFP storage rack cells toward the mislocated fuel assembly.* Case 2.3.15.6.4.13:

This is the same as Case 2.3.15.6.4.11 but the mislocated fuel is directly adjacent to the closest FPM location (position 7). See Figure 2.23.* Case 2.3.15.6.4.14:

This is the same as Case 2.3.15.6.4.13 but the fuel is eccentr'ic in the SFP storage rack cells toward the mislocated fuel assembly.2.3.16 Reconstituted Fuel Assemblies The SFP contains various reconstituted assemblies.

The entire population of previously reconstituted fuel has been examined to determine if the reconstitution may have created a more reactive lattice than those which have been evaluated for this analysis.

The evaluation of the population of reconstituted fuel shows that most of the fulel is very old low reactivity legacy fulel and that tlhere has been no reconstituted bundles that may pose a risk of not being bounded by the analysis.

The evaluation also showed that there is a small set of newer Optima2 fuael bundles that have been reconstituted.

However, the enrichment of these bundles is less than fl wt% U-235, and therefore clearly bounded by the analysis.

Therefore, all previously reconstituted fuel is considered hounded by the analysis and no further analysis is required.

All future reconstituted bundles will have to be evaluated to determine if they are bounded by the analysis.Project No. 2393 Report No. 1-1I-2146153 Hloltec International Proprietary Information Page 26

3. ACCEPTANCE CRITERIA Codes, standard, and regulations or pertinent sections thereof that are applicable to these analyses include the following:

Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62,"Prevention of Criticality in Fuel Storage and H-andling."* Code of Federal Regulations, Title 10, Part 50.68, "Criticality Accident Requirements."* USNRC Standard Review Plan, NURIEG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Revision 3 -March 2007.* L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.Collins, August 19, 1998.* ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors (withdrawn in 2004).* USNRC, NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, January 2001.* DSS-ISG-2010-01, Revision 0, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools.Project No. 2393 Report No. 1-1-2146153 H-oltec International Proprietaty¢ Information Page 27

4. ASSUMPTIONS The analyses apply a number of assumptions, either for conservatism or to simplify the calculation approach.

important aspects ofapplying those assumptions are as follows: 1. Bounding or sufficiently conservative inputs and assumptions are used essentially throughout the entire analyses, and as necessary studies are presented to show that the selected inputs and parameters are in fact conservative or bounding.2. Neutron absorption in minor structural members of the fuel assembly is neglected, e.g., spacer grids are replaced by water.3. The neutron absorber length in the rack is more than the active region of the fuel, but it is modeled to be the same length.4. The fuel density is assumed to be equal to the pellet density for the design basis calculations, and is conservatively modeled as a solid right cylinder over the entire active length, neglecting dishing and chamfering.

This is acceptable since the amount of fuel modeled is more than the actual amount.5. All models are laterally infinite arrays of the respective configuration, neglecting lateral leakage. The exception is where the model boundaries are water, as specified.

6. All fuel cladding materials are modeled as pure zirconium, while the actual fuel cladding consists of one of several zirconium alloys. This is acceptable since the model neglects the trace elements in the alloy which provide additional neutron absorption.

7. T/he SEP storage rack cell ID and cell wall thickness tolerances are assumed values presented in Table 5.3.Project No. 2393 Report No./--I1-2146153 H-oltec International Proprietary Information Page 28

5. INPUT DATA 5.1 Fuel Assembly Specification The SFP racks are designed to accommodate various fuel assembly types used in Dresden Unit 2 and Unit 3. A subset of these fuel designs are presented here for information purposes (the much older fuel designs are not shown): The specifications for the above fuel assemblies designs are presented in Table 5.1. Note that the fuel assembly tolerance information is provided for the bounding fuel design only. As it can be seen in Section 7.1, the reactivity difference between the reactivity of the bounding lattice from the most reactive fuel design and the next most reactive design is large enough to preclude tolerance calculations for both designs.Additional Snecification of the ATRIUM I 0XM 2 Note: Thifs is the expected actual IMPAE; the design basis lattice uses 4.95 wt% U-235.Project No. 2393 Report No. H-1-21 46153 Holtec international Proprietary Information Page 29 5.2 Reactor and SFP Operating Parameters The reactor core and SFP operating parameters are provided in Table 5.2(a). The reactor control blade data are provided in Table 5.2(b). The reactor control parameters used in CASMO-4 screening and design basis calculations are provided in Table 5.2(c).5.3 Storage Rack Speciiication The spent fuel pool rack parameters are provided in Table 5.3. The rack cells are constructed by fixing BORAL panels to the outside of a fabricated steel cell box with sheathing.

The fabricated cells are then joined to create formed cells. On the exterior of every rack module, the location of the formed cells along the exterior without BORAL is closed with a filler plate. Thus, beginning at the corner of each module, the first location has BORAL and then every other location does not have BORAL.The SEP layout is shown in Figure 5.1.5.4 Material Compositions The MCNP5-1 .51 material specification is provided in Table 5.4(a) for non-fuel materials, and Table 5.4(b) specifies isotopes followed in the fuel pellet.Project No. 2393 Report No. 1H1-21t46153 Hioltec International Proprietary Information Page 30

6. COMPUTER CODES The following computer codes were used in this analysis.* MCNP5-1 .51 [1] is a three-dimensional continuous energy Monte Carlo code developed at Los Alamos National Laboratory.

This code offers the capability of performing full three dimensional calculations for the loaded storage racks. MCNP5-l1.51 was run on the PCs at Holtec.* CASMO-4 [4] is a two-dimensional multigroup transport theory code developed by Studsvik.

CASMO-4 is used to perform the depletion calculation for the pin-specific approach, and for various studies. CASMO-4 was run on the PCs at Holtec.Project No. 2393 Report No. HI-2146153 1-Jooltec International Proprietary Information Page 31

7. ANALYSIS RESULTS 7.1 Determination of the Design Basis Fuel Assembly Lattice As discussed in Section 2.3.1I, a complete evaluation of the legacy fuel bundles, current fuel bundle designs and future fuel bundle designs (i.e. the ATRIUM I0XM design) has been performed.

Based on the method described in Section 2.3.1, and the discussion presented in Appendix A, CASMO-4 screening calculations were performed for all Optirna2 lattices, all ATRIUM 10OXM lattices, three ATRIUM 9B lattices and one GEl 4 lattice. The results of the screening calculations determined a subset of lattices with an in-rack CASMO-4 reactivity greater than 0.8500. The subset of most reactive lattices has been further evaluated using MCNP5-1 .51 to determine the bounding lattice. This evaluation is documented in Appendix B.The results presented in Appendix B show that the most reactive ATRIUM 10OXM lattice is, as expected, the lattice with the combination of the highest lattice average enrichment, least number of Gd rods, and lowest Gd rod loading. This lattice is shown to be the ATRIUM 10OXM lattice~(see Figure 7.1). As discussed in Section 2.3.1.3, this lattice was then used to construct a lattice with the maiumpssible lattice average enrichment ofin wt%UO 2 , a lower number of Gd rods and the Gd loading was left at nitue ) This constructed lattice was then labeled the ATRIUM 10OXM Lattice fl(see Figure 7.2). An alternate version has also been constructed

~jljnqjaet lattice with two alternate Gd rod locations, ATRIUM 10OXM Lattice (see Figure 7.3) .Calculations were then performed and document in Appendix B to compare the v ofrthese lattices.

As can be seen in Appendix B TFable B. 1, the ATRIUM 10XM lattice, has an statistically equivalent reactivity to the ATRIUM I 0XM lattice (the onlyiffrn cebtente two lattices is the location of two Gd rods). The ATRIUM 10OXM lattice was selected as the design basis lattice for simplicity and is used for all design basis calculations to show compliance with the regulatory limit.7.2 Core Operating Parameters As discussed in Section 2.3.2, the effects of the core operating parameters on the reactivity were evaluated both during the design basis lattice screening calculations in Appendix A and Appendix B, as well as in the final design basis models calculations presented in Appendix C, Table C.1. As can be seen from the results in Appendix C, Table C. 1 the bounding COP for the design basis lattice is the "min" set (see Table 5.2(c)). Therefore, all design basis calculations use the "min" set of COP. Since the bounding configuration is determined for the various design basis calculations, there is no bias and bias uncertainty associated with COP.7.3 Fuel Assembly Eccentric Positioning and Fuel Assembly De-Channeling As discussed in Section 2.3.5, the reactivity effect of the fuel assembly position in the storage cell and the reactivity effect of the channel have been evaluated.

The results of these calculations are presented in Appendix C, TFable C.2. The result show that the bounding fuel Project No. 2393 Report No. I-JI-2146153 Page 32 1-oltec International Proprietary Information assembly position is cell centered and the bounding condition is channeled fuel. Therefore, all design basis calculations consider the fuel cell centered and with a channel with the exception of specific cases that are otherwise noted. Since the bounding configuration is determined for the various design basis calculations, there is no bias and bias uncertainty associated with fuel assembly eccentric positioning and fuel assembly de-channeling (i.e. the value is zero as presented in Table 7.1 and 7.2).7.4 Fuel Bundle Orientation in the SFP Rack Cell As discussed in Section 2.3.6, the reactivity effect of the fuel assembly orientation (i.e.orientation of the in core control blade corner) has been evaluated.

The results of these calculations are presented in Appendix C, Table C.3. The results of these calculations show that Case 2.3.6.2 has a small bias and bias uncertainty.

This small bias and bias uncertainty are therefore considered in the determination of (see Table 7.1 and 7.2).7.5 Reactivit'y Effect of Spent Fuel Pool Waler Temperature As discussed in Section 2.3.7, the effects of water temperature, and the corresponding water density and temperature adjustments (S(cL,f3))

were evaluated for SFP racks. The results of these calculations are presented in Appendix C, Table C.4.The results of the SEP temperature and density calculations show that as expected (for poisoned racks) the most reactive water temperature and density for the SFP racks is a temperature of 39.2 °F at a density of I g/cc, and these values are used for all calculations in SFP racks with the exception of specific accident conditions.

7.6 Fuel and Storage Rack Manufacturing Tolerances 7.6.1 Fuel Manufacturing Tolerances As discussed in Section 2.3.8.1, the effect of the BWR fuel tolerances on reactivity was determined.

The results of these calculations are presented in Appendix C, Table C.5. The maximum positive delta-k value for each tolerance is statistically combined.The maximum statistical combination of fuel assembly tolerances is used to determine k~fr in Table 7.1 and Table 7.2.7.6.2 SFP Storage Rack Manufacturing Tolerances As discussed in Section 2.3.8.2, the effect of the manufacturing tolerances on reactivity of the SFP racks was determined.

The results of these calculations are presented in Appendix C, Table C.6. The maximum positive delta-k value for each tolerance is statistically combined.The maximum statistical combination of the SFP rack tolerances is used to determine keff in Table 7.1 and Table 7.2.Project No. 2393 Report No. HI-2146153 Page 33 H-oltec International Proprietary Information 7.6.3 Fuel Depletion Calculation Uncertainty As discussed in Section 2.3.9, the uncertainty of the number densities in the depletion calculations was evaluated.

The results of these calculations are presented in Appendix C, Table C.7. As can be seen in Appendix C, Table C.7, thle depletion uncertainty is calculated as 5% of the reactivity difference between the design basis case and a calculation with fresh fuel and no Gd.The depletion uncertainty is included in the statistical combination of uncertainties used to determine keff in Table 7.1 and Table 7.2.7.6.4 Fission Products and Lumped Fission Products Uncertainty As discussed in Section 2.3.10, the uncertainty of the FP and LFP in the depletion calculations was evaluated.

The results of these calculations are presented in Appendix C, T!able C.8. As can be seen in Appendix C, Table C.8, the FP and LIP uncertainty is calculated as 1l% of the reactivity difference between the design basis case and a calculation with no PP or LFP.The FP and LFP uncertainty is included in the statistical combination of uncertainties used to determine kdyr in Table 7.1] and Table 7.2.7.6.5 Depletion Related Fuel Assembly Geometry Changes As discussed in Section 2.3.1 ], the reactivity effect of depletion related fuel assembly geometry changes has been evaluated.

These evaluations are discussed further below.7.6.5.1 Fuel Rod Geometry Changes As discussed in Section 2.3.1 I .1, the reactivity effect of fuel rod geornetly changes is evaluated.

These evaluations consider fuel rod growth and cladding creep, fuel rod crud buildup and fuel rod bow and are discussed below. As previously discussed, the fuel assembly is not expected to undergo significant depletion related geometry changes at peak reactivity (i.e. about l GWd/m~tU).

However, specific effects are evaluated as discussed below.7.6.5.1.1 Fuel Rod Growth, Cladding Creep and Fuel Rod Crud Buildup As discussed in Section 2.3.11.1.1 and Section 2.3.11.1.2, the effect of the fuel rod growth, cladding creep and fuel rod crud buildup on reactivity was not evaluated due to the low burnup at peak reactivity.

7.6.5.1.2 Fuel Rod Bow As discussed in Section 2.3.11.1.3, the reactivity effect of the fuel rod bow was evaluated by calculation.

The fuel rod bow calculation results are presented in Appendix C, Table C.9. The Project No. 2393 Report No. l-1-2 146153 P'age 34 H-oltec International Proprietary Information results presented in Appendix C, Table C.9 show a small bias and bias uncertainty.

This bias and bias uncertainty are considered in the determine of kenf as presented in Table 7.1 and 7.2.7.6.5.2 Fuel Channel Bulging and Bowing As discussed in Section 2.3.11.2, the reactivity effect of fuel channel bulging and bowing was evaluated by calculation.

The fuel channel bow calculation results are presented in Appendix C, Table C.9. The results presented in Appendix C, Table C.9 show a small bias and bias uncertainty.

This bias and bias uncertainty are considered in the determine of kerr as presented in Table 7.1 and 7.2.7.7 SFP Storage Rack Interfaces As discussed in Section 2.3.12, the reactivity effect of the SFP storage rack interfaces, specifically the interface of one storage rack module with another storage rack model has been evaluated.

The calculation results are presented in Appendix C, Table C.10. The results presented in Appendix C, Table C.10 show a bias and bias uncertainty.

This bias and bias uncertainty are considered in the determine of kerr as presented in Table 7.1 and 7.2.7.8 Maximum k,,ff Calculations for Normnal (Conditions As discussed in Section 2.3.13, the maximum keff for normaal conditions is calculated.

The results are tabulated in Table 7.1. The results show that the maximum keff for the normal conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95% confidence level.7.9 Fuel Movement, Inspection and Reconstitution Operation.

As discussed in Section 2.3.14, the fuel movement, inspection and reconstitution operations are normal conditions that are bounded by the analysis.

No further evaluations are required.7.10 Abnormal and Accident Conditions As discussed in Sections 2.3.15, the effects of various accident conditions has been evaluated.

The results of these calculations are presented in Appendix C, Table C.4 (increased SEP temperature only) and Appendix C, Table C. 11 (all other accidents).

The maximum reactivity accident has been determined to beThe calculated results of this accident are used, along with all applicable biases and uncertainties, to show compliance with the regulatory limit in Table 7.2. As it can be seen in Table 7.2, the maximum calculated reactivity is less than 0.95 at a 95% probability and at a 95%confidence level.Project No. 2393 Report No. HI-21 46153 Page 35 H-oltec International Proprietary Information

8. CONCLUSION The criticality analysis for the storage of BWR assemblies in the Dresden SFP racks with BORAL has been performed.

The results for the normal condition show that keff is 1 with the strg ak ul oddwith fuel of the highest anticipated reactivity, which is the strae acsful oaedat a temperature corsodn othe highest reactiviy Terslsfor the boudn acietcondition, i.e. theshow that ke is with of the highest anticipated reactivity, which is 1 ,at a temperature corresponding to the highest reactivity.

The maximum calculated reactivity for both normal and accident conditions include a margin for uncertainty in reactivity calculations with a 95% probability at a 95% confidence level.Reactivity effects of abnolrmal and accident conditions have been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95.Project No. 2393 Report No. H1-2 146153 H~oltec International Proprietary information Page 36

9. REFERENCES

[1] "MCNP -A General Monte Carlo N-Particle Transport Code, Version 5," Los Alamnos National Laboratory, LA-UR-03-1987, April 24, 2003 (Revised 2/1/2008).

[2] "Nuclear Group Computer Code Benchmark Calculations," H-oltec Report 1H1-2104790 Revision 1.[3] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001I.[4] M. Edenius, K. Ekberg, B.H. Forss~n, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," StudsviklSOA-95/1; and J. Rhodes, K Smith,"CASMO-4 A Fuel Assembly Burnup Program User's Manual," SSP-0l/400, Revision 5, Studsvik of America, Inc, and Studsvik Core Analysis AB (proprietary).

[5] D. Knott, "CASMO-4 Benchmark Against Critical Experiments," SOA-94/13, Studsvik of America, Inc., (proprietary);

and D. Knott, "CASMO-4 Benchmark Against MCNP," SOA-94/l12, Studsvik of America, Inc., (proprietary).

[6] L.i. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.Collins, August 19, 1998.[7] DSS-ISG-201 0-01, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, Revision 0.[8] HI1-2002444, Latest Revision, "Final Safety Analysis Report for the HI-STORM 100 Cask System", USNRC Docket 72-10 14.[9] "Atlas of Neutron Resonances", S.F. Mughabghab, 5th Edition, National Nuclear Data Center, Brookhaven National Laboratory, Upton, USA.[10] "Sensitivity Studies to Support Criticality Analysis Methodology," HI1-2104598 Rev. 1, October 2010.[11] "Spent Nuclear Fuel Burnup Credit Analysis Validation", ORNL Presentation to NRC, September 21, 2010.[12] An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (k~ff) Predictions, NUREG/CR-71 09, April 2012.Project No. 2393 Report No.111-2146153 Page 37 H-oltec International Proprietary Information Table 2.1 (a)Summary of the Area of Applicability of the MCNP5-1 .51 Benchmark Validated by Validation Extrapol Parameter Analysis Bench mark Gps ation.........3-235, U3-238, Fuel Pu-239, Pu-240, assemblies U0 n D ul Pu-241, Pu-242, nn /Am-241 _______Initial fuel Up to

wt% U-235, < 5 wt% U3-235, " enrichments

___________

1.5 to 20 wt% Pu none N/A Fuel density g/cc 9.2 to 10.7 g/cc none N/A Burnp rage <I G~/mtU0 and 37.5 Bunprne<lG dmUGWd/mtU none N/A Moderator material H 2 0 1-120 none N/A.............

B-SS, BORAL, '...Neutron B-10 (rack insert) B~oraflex, Cadmium none NiA poison Gd (residual) or Gadoliniunm

___IetsialSteel Steel or Lead none N/A material Fuel cladding Zr a~lloy Zr alloy none .. N/A Peridic oundty wter Reflective or Reflector Peioi 'onay ae periodic boundary, none N/A water reflectors Lattice type Square Square, triangle none N/A Neutron Thermal spectrum Thermal spectrum none N/A energy(eV) IIIIII, none N,/A SThe set of benchimarked experiments include the experiments with Gd 2 0 3 rods and gadolinium dissolved in water. However, it's acceptable because the isotope composition and distribution (Gd 2 O 3 rods) is similar.Project No. 2393 Report No.1-11-2146153 Holtec International Proprietary hnformation Page 38 Table 2.1 (b)Analysis of the MCNP5-l.51 calculations

[2]Note 1: The single sided lower tolerance factor forE[ samples was conservatively used.Project No. 2393 Report No. HI-2146153 H-oltec International Proprietary" Informaation Page 39 Table 2. l(c)Bias and Bias Uncertainty as a Function of Independent Parameter for SEP Racks Filled with Pure Water [21 r T I I V Independent Parameter:

EALF Calculated keff Bias Bias Uncertainty Independent Parameter:

U-235 Enrichment Calculated kctf Bias Bias Uncertainty Note 1: For U-235 enrichment ofin wt% (maximum fuel enrichment used in the analysis which has the largest bias uncertainty) and BALE of I(larger than the maximum EALF determined in the analysis), the bolded numbers show the bounding bias and bias uncertainty values.Note 2: The positive biases (which mean decrease in reactivity) are truncated to zero [31].Project No. 2393 Report No. HI-2146153 Holtec International Proprietary Information Page 40 S f'm Ug II zz~t r-- ÷ /zzII Project No. 2393 Report No. 1-1-21 46153 1-oltec International Proprietary Information Page 41 Project No. 2393 Report No. H-I-21461 53 Hloltec International Proprietary Information Page 42 Project No. 2393 Report No. HI1-2146153 H-oltec International P~roprietary Information Page 43 Project No. 2393 Report No. H-1-21 46153 H-oltec International Proprietary Information Page 44 Table 5.1 (e)I!Project No. 2393 Report No. 1-]1-21 46153 Holtec International Proprietary Information Page 45 Table 5 1(Ct Project No. 2393 Report No. 1-11-2146153 H-oltec International Proprietary Information Page 46 Table 5.1 (g)Ku ZZi 6191 UI Project No. 2393 Report No. 111-2146153 1-oltec international Proprietary Information Page 47 4---,'----,--,---El 4-1--t mm I-__ U--HE Ui II IU El Project No. 2393 Report No. H-1-2146153 H-oltec International Proprietary Information Page 48 Table 5.2(a)Reactor Core and Spent Fuel Pool Parameters Description (Unit)...

Value Licensed thermal power (MWth) -F Power density (W/gU) Maximum fuel pin temperature (K)___l Moderator temperature range (0 F)Moderator saturation temperature (0 F) ......Design basis core average void fraction (%) 1_____________

Maximum bundle core exit void fr'action

(%)Spent Maximum temperature (0 F)2 Project No. 2393 Report No. HJ-2146153 Holtec International Proprietary Information Page 49 Table 5.2(b)Reactor Control Blade Data Description (Unit)Noia au initial equipment m Project No. 2393 Report No. HI-2146153 H-oltec ]nternational Proprietary lnform~ation Page 50 Table 5.2(c)Reactor Core Parameters used for CASMO-4 Screening and Design Basis Calculations It is assumed that the minimum power density is 15% less than the nominal value.tt it is assumed that the minimum fuel temperature is half of the maximum value. Also, the nominal fuel temperature is the average of the maximum and minimum values.!i The nominal moderator temperature is the average of the maximum and minimum values.Project No. 2393 Report No. 1-t1-2 146153 Holtec International Proprietary Information Page 51 Table 5.3 SFP Storage Rack Parameters and Dimensions Description (Unit) Nominal Value [ Tolerance SFP Racks-__n _ Ij n U I m m U-U-m+4--BO zIv RAL P: m Fuel Prep Machine IF tThese are assumed values.TlThis is the design value. The value used in the interface model (see Section inches.tt This representation of the fuel prep machine (FPM) is a simplification.

physically separate FMPs in the SFP each with a capacity of one assembly.2.3.12) is -There are two Project No. 2393 Report No. 11i-2146153 H-oltec International Proprietary Information Page 52 Table 5.4(a)Non-Fuel Material Compositions Element MCNP ZAID [l] "weight Fraction Steel (density g/cc) [8Ijt 24050.70c I__________

Cr24052.70c__________

Cr '- 24053.70c_______ 24054. 70c Mn 25055.70c 26054.70e 26056.70e Fe26057.70e

______ 26058.70c 28058.70c 28060.70c Ni ...28061l.70e 28062.70c_____28064,70c

_______________

Zr (density__6.55 g/cc)J[8]j" 40090.70c 0.50706120 40091 .70c 0.11180900 Zr 40092.70c 0.17278100 40094.70e 0.1'7891100

______ 40096.70c

0.0 2943790

_________Pure water (density=

1.0 g/ce)[8]1 1001.70c 0.11188600 1002.70c 0.00002572 8016.70c 0.88579510

______ 8017.70c 0.00229319 BORAL (density =i g/c&)B 5010.70c________ 5011,70c C 6000.70e __Al 13027,70c

__chemical element.Project No. 2393 was expanded to represent the full list of natural isotopes for each Report No. 1-11-2146153 Page 53 H-oltec International Proprietary Information Table 5.4(b)Summary of the Fuel and Fission Product Isotopes Used in Calculations ASO MCNP5 ZAID CMO MCNP5 ZAID Isotope Isotope U-234 92234.70c Xe-1 31 t 54131.70c U-235 92235.70c s-3 55133.70c U-236 92236,70c 55134.70c U-238 92238.70c Cs'135 55135.70c U-239 92239.70c C-3t 55137.70c Np-237 93237.70c Nd- 143 60143.70c Np-239 added to Pu-239 Nd-145 60145 .70c Pu-238 94238.70c Pro-147 61147.70c Pu-239 94239.70c Pio-148 61148.70c...Pu-240 94i240.70c Pro-149 61 149.70c Pu-241 .....94241 .70c Sm-147 62147.70c Pu-242 94242.70c Sm-I149 62149,70c Amn-241 95241.70c Sm-150 62150.70c Amn-242m ' 95242.70c.

Sm-I 51 62151.70c Am-243 95243 .70c Sm-I 152 621 52.70c Cmi-242 96242.70c Eu-153 63153.70c Cmn-243 96243.70c Eu- 154 63154.70c Cm-244 96244.70c Eu-155 63155.70c Cm-245 96245.70c Gd-152 64152.70c Cm-246 96246.70c Gd-154 .......64154.70c Kr-83t 36083.70c , Gd-155 64155.70c Rh-103 45103.70c Gd- 157 64157.70c Rh-1O5 45105.70c Gd-160 64160.70c Ag-109 47109.70c 0-16 8016.7Cc 1-135t 53135.70c Gd-158 64158.7Cc Gd-156 64156.70c LFP 1/LFP2 tNt:These isotopes are removed for all design basis applications because they are either gaseous or volatile nuclides.Project No. 2393 Report No. 1-1-2146153 H-oltec International Proprietary Information Page 54 Table 7.1 Maximum ken Calculation for Normal Conditions in SFP Racks Parameter Value Uncertaint~iest Fuel tolerance uncertainty, from Table C.5 -Rack tolerance uncertainty, fr'om Table C.6 -Fuel eccentricity and de-channeling bias uncertainty, from Table C.21 Fuel orientation bias uncertainty, from Table C.31 Fuel channel bow bias uncertainty, from Table C.9 ________Fuel rod bow bias uncertainty, from Table C.9 -Depletion uncertainty, from Table C.7 -FP/LFP uncertainty, from Table C.8 -MCNP5-1 .51 code bias uncertainty (95%/95%), from Table 2. 1(b)1 MCNP5-1 .51 calculations statistics (95%/95%, 2ar), from Table C.l 1_____1 __Interface bias uncertainty, from Table C. 10 -Statistical combination of uncertainties-Biases Fuel eccentricity and dc-channeling bias, fr'om TFable C,21 Fuel orientation bias, fr'om Table C.3-Fuel channel bow bias, from Table C.9-Fuel rod bow bias, from TFable C.9-MCNP5-1.51 code bias, from Table 2.1(b)-Interface bias, from Table C. 10-D eterm ination of keff __ _ _ __ _Calculated MCNP5-1 .51 k 4 a 1 e, from Table C.l -Maximum kcrff _____Regulatory Limit 0.9500 Margin to Limit________

tTeprovided value is the 95%/95% delta 1 uncertainty.

Note I : The negative biases were conservatively truncated.

Project No. 2393 Report No. HIl-21461 53 Holtec International Proprietary Information Page 55 Table 7.2 Maximum kerr Calculation for Abnormal and Accident Conditions in SFP Racks Parameter

[ Value Uncertaintiest Fuel tolerance uncertainty, from Table C.5 -Rack tolerance uncertainty, from Table C.6 -Fuel eccentricity and de-channeling bias uncertainty, from Table C.21 Fuel orientation bias uncertainty, from Table C.31 Fuel channel bow bias uncertainty, from Table C.91 Fuel rod bow bias uncertainty, from Table C.9 -Depletion uncertainty, from Table C.7 -FP/LFP uncertainty, from Table C.8 -MCNP5-1 .51 code bias uncertainty (95%/95%), from Table 2.1(b)1 MCNP5-l .5] calculations statistics (95%1o95%, 2or), from Table C.I 1 Interface bias uncertainty, fr'om Table C. 10 -Statistical combination of uncertainties1 Biases Fuel eccentricity and de-channeling bias, from Table C.21 Fuel orientation bias, from Table C.3 -Fuel channel bow bias, from Table C.9 -Fuel rod bow bias, from Table C.9 -MCNP5- 1.51 code bias, from Table 2.1(b)-Interface bias, from Table C. 10-Determination of k~1 y Calculated MCNP5-1.51 from Table C.1 1 ______Maximum keffr Regulatory Limit 0.9500 Margin to Limit-SThe provided value is the 95%/95% delta uncertainty.

Note 1 : The negative biases were conservatively truncated.

Project No. 2393 Report No. 1-1-2146153 Iloltec International Proprietary Information Page 56 Figure 2.1 A representation of the Design Basis CASMO-4 Model with the Design Basis Lattice.This figure is proprietary.

Project No. 2393 Report No. 111-2146153 H-oltec International Proprietary Information Page 57 Figure 2.2 A 2-D Representation of the MCNP5-1 .51 Design Basis Model with the Design Basis Lattice, Case 2.3.1.4.1 This figure is proprietary.

Project No. 2393 Report No. 1-1-2146153 1-oltec International Proprietary Information Page 58 Figure 2.3 A 2-D Representation of the 2x2 Channeled Fuel Eccentric Positioning MCNP5-1 .51 Model, Case 2.3.5.2 This figure is proprietary.

Project No. 2393 Report No. HI-2146 153 H-oltec International Proprietary Infornation Page 59 Figure 2.4 A 2-D Representation of the 2x2 Channeled Fuel Eccentric Positioning MCNP5.-1 .51 Model, Case 2.3.5.3 This figure is proprietary, Project No. 2393 Report No. H11-2146153 1-oltec International Proprietary Information Page 60 Figure 2.5 A 2-D Representation of the 8x8 Channeled Fuel Eccentric Positioning MCNP5-I1.51 Model, Case 2.3.5.5.This figure is proprietary.

Project No. 2393 Report No. 111-2146153 H-oltec International Proprietary Information Page 61 Figure 2.6 A 2-D Representation of the 8x8 Channeled Fuel Eccentric Positioning MCNP5-1 .51 Model, Case 2.3.5.6.This figure is proprietary.

Project No. 2393 Report No. 1-1-2146153 H-oltec International Information Page 62 Figure 2.7 A 2-D Representation of the 2x2 De-channeled Fuel Eccentric Positioning MCNP5-! .51 Model, Case 2.3.5.8.This figure is proprietary.

Project No. 2393 Report No. HI1-2146153 H-oltec International Proprietary Information Page 63 Figure 2.8 A 2-D Representation of the 2x2 De-channeled Fuel Eccentric Positioning MCNPS-1 .51 Model, Case 2.3.5.9.This figure is proprietary.

Project No. 2393 Report No. 111-2146153 Holtec International Proprietary Information 1Page 64 Figure 2.9 A 2-D Representation of the 8x8 De-channeled Fuel Eccentric Positioning MCNP5-1 .51 Model, Case 2.3.5.11 This figure is p~roprietary.

Project No. 2393 Report No. 1H1-2146153 1-oltec International Proprietary Information Page 65 Figuare 2.1I0 A 2-D Representation of the 8x8 De-channeled Fuel Eccentric Positioning MCNP5-1 .51 Model, Case 2.3.5.12 This figure is proprietary.

Project No. 2393 Report No. HI-2146 153 1-oltec International Proprietary Information Page 66 Figure 2.1]A 2-D Representation of the 4x4 Fuel Orientation MCNP5-1 .51 Model, Case 2.3.6.2 This figure is proprietary.

Project No. 2393 Report No. 11t1-21 461 53 Floltec International Proprietary Information Page 67 Figure 2.12 A 2-D Representation of the 4x4 Fuel Orientation MCNP5-1 .51 Model, Case 2.3.6.3 This figure is proprietary.

Project No. 2393 Report No. 111-2146153 H-oltec International Proprietary Information Page 68 Figure 2.13 A 2-Dl Representation of the 4x4 Fuel Orientation MCNP5-1.5 1 Model, Case 2.3.6.4 This figure is proprietary.

Project No. 2393 Report No. 111-2146153 Hioltec International Proprietary Information Page 69 Figure 2.14 A 2-D Representation of the 4x4 Fuel Orientation MCNP5-1 .51 Model, Case 2.3.6.5 This figure is proprietaly,.

Project No. 2393 Report No. H-I-2146153 H-oltec International Proprietary Information Page 70 Figure 2.15 A Partial 2-D Representation of the MCNPS-1.51 Interface Model, Case 2.3.12.1 This figure is proprietary.

Project No. 2393 Report No. 111-2146153 H-oltec International Proprietary Information Page 71 Figure 2.16 A partial 2-D Representation of the ]6x 16 Vertical Fuel Drop Accident MCNP5-1.51 Model, Case 2.3.15.3.1 This figure is proprietary.

Project No. 2393 Report No. 1-1-2146153 1-oltec International Proprietary Information Page 72 Figure 2.17 A partial 2-D Representation of the 8x8 Missing BORAL Panel Accident MCNP5-l1.51 Model, Case 2.3.15.4.2 This figure is proprietary.

Project No. 2393 Report No. 1-1-2146153 1-oltec International Proprietary Information Page 73 Figure 2.18 A partial 2-D Representation of the 80x80 Mislocated in a Corner of Two Racks Accident MCNP5-1.51 Model, Case 2.3.15.6.2.1 This figure is proprietary.

Project No. 2393 Report No. 1-I-2146153 JHoltec International Proprietary Information Page 74 Figure 2.19 A partial 2-D Representation of the 80x80 Mislocated in a Corner of Three Racks Accident MCNP5-1 .51 Model, Case 2.3,15.6,3.1 This figure is proprietary.

Project No. 2393 Report No. 1-1I-2146153 H-oltec international Proprietary Information Page 75 Figure 2.20 A partial 2D representation of the SFP Platform Mislocated Fuel Assembly Accident MCNP5-1.51 Model, Position 1 (Case 2.3.15.6.4.1)

This figure is proprietary.

Project No. 2393 Report No. HI1-2146153 H-oltec International Proprietary Information Page 76 Figure 2.21 A partial 2D representation of the SFP Platform Mislocated Fuel Assembly Accident MCNP5-.1.51 Model, Position 5 (Case 2.3.15.6.4.9)

This figure is proprietary.

Project No. 2393 Report No. HI-2146]53 1-oltec International Proprietary Information P age 77 Figure 2.22 A partial 2D representation of the SFP Platform Mislocated Fuel Assembly Accident MCNP5-1.51 Model, Position 6 (Case 2.3.15.6.4.11)

This figure is proprietary.

Project No. 2393 Report No. HI1-2146153 1Holtec International I~roprietary Information Page 78 Figure 2.23 A partial 2D representation of the SEP Platform Mislocated Fuel Assembly Accident MCNP5-1.51 Model, Position 7 (Case 2.3.15.6.4.13)

This figure is proprietary.

Project No. 2393 Report No. 1-11-2146153 H-oitec International Proprietary Information Page 79 Figure 5.1 Layout of the SFP I::i UNIT 3 Project No. 2393 Report No. 1-1-2146153

Holtec International Proprietary Information Page 80 Figure 7.]This figure is proprietary.

Project No. 2393 Report No. 1-11-2146153 Holtec International Proprietary Information Page 81 Figure 7.2 This figure is proprietary.

Project No. 2393 Report No. 1-11-2146153 1-oltec International Proprietary Information Page 82 Figure 7.3 This figure is proprietary.

Project No. 2393 Report No. HI-2146153 H-oltec International Piroprietary Information Page 83 Appendix A CASMO-4 Screening Calculations for Determination of the Design Basis Fuel Assembly (Number of Pages 43)Project No. 2393 Report No. 1-11-2146153 H-oltec International Proprietary Information Page A-I A. 1 Introduction The purpose of Appendix A is to present the results of the Step I CASMO-4 screening calculations (see Section 2.3.1.2 in the main report).A.2 Methodology The CASMO-4 screening calculations are performed using CASMO-.4 depletion calculations and in-rack restart kin calculations for four sets of core operating parameters (COP) (minimum COP, minimum COP with control blades inserted, nominal COP and maximum COP), see Table 5.2(e)in the main report. The screening calculations are performed in order to determine the peak reactivity for every Optima2, every ATRIUM 10XM lattice, a GEl4 lattice and three ATRIUM 9B lattices.

The other legacy fuel lattices (i.e. , etc.) all have an average enrichment less than

wt% U-235. Engineering judgment is used to screen these lattices fr'om further consideration because their reactivity will be bounded by the other fuel designs with average enrichments greater than fl wt% U-235. All lattices with natural uranium are neglected because of their low reactivity.

The screening calculations determaine the peak reactivity for each of the four sets of COP for each lattice. Using the maximum overall value fi'om the four sets of COP for each lattice, the results are further screened to select the subset of most reactive lattices (and the two most reactive fuel designs).

For- the purpose of determining the most reactive subset of lattices, the lattices with an in-rack kinf- of 0.8500 or greater are selected for further analysis in the main report (see Section 2.3.1.3 in the main report).A.3 Assumptions No assumptions are made specifically for the screening calculations that are different than those listed in Section 4 of the main report..A.4 Acceptance Criteria In order to screen out low reactivity lattices from unnecessary additional calculations, the entire set of lattices are screened for in-rack reactivity kinf values of 0.8500 or more. The criteria of kinf > 0.8500 is chosen based on the overall range of reactivity seen in the results presented in this Appendix.A.5 input Data All input data has been specified in Section 5 of the main report.A.6 Results The results of the CASMO-4 screening calculations are presented in Table A. 1 for the ATRIUM IOXM design, Table A.2 for the Optima2 design, Table A.3 for the ATRIUM 9B design and Table A.4 for the GEI4 design. The results presented in Table A.I through A.4 are screened for lattices with an in-rack peak reactivity greater than 0.8500. The results of this screening are presented in Table A.5.Project No. 2393 Report No.1-1I-2146153 Page A-2 H-oltec International Proprietary Information A.7 Conclusion Based on the results presented in Table A.5, the most reactive lattices from the ATRIUM 10OXM and Optima2 fuel designs are selected because they meet the acceptance criteria of an in-rack restart peak reactivity greater than 0.8500. These lattices are considered for additional calculations as described in Section 2.3.1.3 in the main report.Project No. 2393 Report No. 1H1-2146153 1-oltec International Proprietary Information Page A-3 Table A. I Results of the CASMO-4 in-rack k 1~f Screening Calculations for the ATRIUM 10OXM Fuel Design (1 of 12)neak Burnup (gwd))cinf Bumup kinf I Burnup kinf flumup kinf-max' (~wd) -minr" Bounding COP U m III U ml mm I -] i- j II-I III I-I~-A i-m --i [ H~ -m -m II -l m m ,m~ m -u-i Il m m m, _mm -a, m m....... n mI --m- -m m "m -_ _ ..m -l llm -I _, -m -m mL m m m m -l m [ -Note: the peak reactivity values are bolded. Any lattice that meets the criteria of peak reactivity