ML13199A038

Attachment 5 - Holtec International Report No. HI-2125245, Revision 2, Licensing Report for Quad Cities Criticality Analysis for Inserts - Non Proprietary Version

ML13199A038
Person / Time
Site:	Quad Cities
Issue date:	07/16/2013
From:	Holtec
To:	Exelon Generation Co, Office of Nuclear Reactor Regulation
References
HI-2125245, Rev 2
Download: ML13199A038 (136)

Text

ATTACHMENT 5 Holtec International Report No. HI-2125245, Revision 2 "Licensing Report for Quad Cities Criticality Analysis for Inserts -

Non Proprietary Version"

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HOLTEC Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797- 0900 Fax (856) 797 - 0909 INTERNATIONAL

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Licensing Report for Quad Cities Criticality Analysis for Inserts - Non ProprietaryVersion FOR Exe/on Holtec Report No: HI-2125245 Holtec Project No: 2127 Sponsoring Holtec Division: HTS Report Class: SAFETY RELATED HOLTEC INTERNATIONAL

Summary of Revisions:

Revision 0: Original Issue Revision 1: Supplement I was added to cover a new revision of NETCO-SNAP-IN rack insert.

Revision 2: All Revision 1 revision bars were removed. No other changes were made.

Project No. 2127 Report No. HI-2125245 Page i

Table of Contents

1. INTRODUCTION ............................................................................................................................. 10

2. M ETHODOLOGY ............................................................................................................................ 11 2.1 GENERAL APPROACH ................................................................................................................... 11 2.2 COMPUTER CODES AND CROSS SECTION LIBRARIES.............................................................. 11 2.2 .1 M CNP5 -1.5 1 ......................................................................................................................... 11 2 .2 .1.1 M C N P 5-1.5 1 V alid ation .................................................................................................................................. 11 2 .2 .1 .1.1 ...................................................................... 12 2.2.2 CASMO-4 ...................................................................................................... 13 2.3 ANALYSIS M ETHODS .................................................................................................................... 13 2.3.1 Design Basis FuelAssembly ............................................................................................. 13 2 .3.1.1 P eak R eactiv ity ................................................................................................................................................ 14 2.3.1.1.1 Peak Reactivi and Fuel Assembl Bumup .......................................................................................... 14 2.3.1.1.2 .......................................................... 14 2.3.1.2 15 2.3.1.3 Determination of the Design Basis Fuel Assembly Lattice ......................................................................... 16 2.3.1.4 Optima2 CASMO-4 Model Simplification Effect ....................................................................................... 16 2.3.1.5 C ore O perating Param eters .............................................................................................................................. 18 2.3.1.5.1 R eactor P ow er U prate ................................................................................................................................. 18 2.3.1.5.2 Integral Reactivity Control Devices ........................................................................................................ 19 2.3.1.5.3 Axial and Planar Enrichment Variations ................................................................................................ 19 2.3.1.5.4 Fuel Assembly De-Channeling ............................................................................................................... 19 2 .3.1.6 ......................................................................................... 20 2.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature................................................ 20 2.3.3 Fuel Depletion Calculation Uncertainty........................................................................... 21 2.3.4 Fuel and Storage Rack Manufacturing Tolerances .......................................................... 22 2.3.4.1 F uel M anufacturing Tolerances ....................................................................................................................... 22 2.3.4.2 SFP Storage Rack Manufacturing Tolerances ............................................................................................ 23 2.3.5 Radial Positioning................................................................................................................. 23 2.3.5.1 Fuel Assembly Orientation in the Core ....................................................................................................... 23 2.3.5.2 Fuel R adial Positioning in the R ack ................................................................................................................. 23 2.3.5.3 Inserts Radial Positioning ................................................................................................................................ 25 2.3.5.4 Fuel O rientation in SFP R ack C ell ................................................................................................................... 25 2.3.6 ............................. 25 2.3.6.1 = ............ . ............................................................................................. 26 2.3.6.12 . . . . . . . . . . . . .. ......................................................................... 27 2.3.6 1.2.. ..................................................... ....27 2.3.7 Insert Coupon Measurement Uncertainty....................................................... 27 2.3.8 Maximum keff Calculationfor Normal Conditions............................................................ 27 2.4 M ARGIN EVALUATION .................................................................................................................. 28 2.5 FUEL MOVEMENT, INSPECTION AND RECONSTITUTION OPERATIONS ..................................... 29 2.6 ACCIDENT CONDITION .................................................................................................................. 29 2.6.1 Temperatureand Water Densit&Effects .......................................................................... 30 2.6.2 DroppedAssembly - Horizontal...................................................................................... 30 2.6.3 DroppedAssembly - Vertical into a Storage Cell ........................................................... 30 2.6.4 Storage Cell Distortion .................................................................................................... 31 2.6.5 MisloadedFuel Assembly/Missing Insert......................................................................... 31 2.6.6 MislocatedFuel Assembly ................................................................................................ 32 2.6.6.1 Mislocation of a Fuel Assembly in the Water Gap between the Racks and Pool Wall .............................. 32 2.6.6.2 Mislocation of a Fuel Assembly in the Comer between Two Racks ........................................................... 32 2.6.6.3 Mislocation of a Fuel Assembly between the SFP Rack and the Inspection Platform ............................... 32 2.6.7 Mis-installment of an Insert on Wrong Side of a Cell ....................................................... 33 Project No. 2127 Report No. HI-2125245 Page 1

2.6.8 Insert Mechanical Wear ........................................................................................................ 33 2.6.9 Rack Movement ..................................................................................................................... 33 2.7'"'"'.......................................................33 2.7.1 ....... 35 2.8 SPENT FUEL RACK INTERFACES ............................................................................................... 36 2.9 RECONSTITUTED FUEL ASSEMBLIES ....................................................................................... 36

3. ACCEPTANCE CRITERIA ....................................................................................................... 37 3.1 APPLICABLE CODES, STANDARDS AND GUIDANCE'S .............................................................. 37

4. ASSUM PTIO NS ................................................................................................................................ 38

5. INPUT DATA ..................................................................................................................................... 39 5.1 FUEL ASSEMBLY SPECIFICATION ............................................................................................. 39 5.2 REACTOR PARAMETERS ............................................................................................................... 39 5.3 SPENT FUEL POOL PARAMETERS ............................................................................................. 39 5.4 STORAGE RACK SPECIFICATION ............................................................................................... 40 5.4.1 Material Compositions.......................................................................................................... 40

6. CO M PUTER CO DES ....................................................................................................................... 41

7. ANALYSIS ......................................................................................................................................... 42 7.1 DESIGN BASIS AND UNCERTAINTY EVALUATIONS .................................................................. 42 7.1 .1 ................................................ 42 7.1.2 Determinationof the Design Basis Fuel Assembly Lattice .............................................. 42 7.1.2.1 Fuel Assembly De-Channeling ........................................................................................................................ 42 7.1.3 Optima2 CASMO-4 Model Simplification Effect ............................................................. 42 7.1.4 Core OperatingParameters.............................................................................................. 43 7.1.4 .1 R eactor P ow er U prate ...................................................................................................................................... 43 7.1.5 Water Temperature and Densit, Effect ............................................................................. 43 7.1.6 Depletion Uncertaint ........................................................................................................... 43 7.1.7 Fuel and Rack M anufacturingTolerances........................................................................ 44 7.1.7.1 Fuel Assembly Tolerances ............................................................................................................................... 44 7.1.7.2 S F P R ack To leran ces ....................................................................................................................................... 44 7.1.8 Radial Positioning................................................................................................................. 44 7.1.8.1 Fuel Assembly Radial Positioning in SFP Rack ......................................................................................... 44 7.1.8.2 Fuel Orientation in SFP Rack........................................................................................................................... 44 7.1.9 Fu.1 Rod........I..... . .......................................... ................................................ 45 7 .1.9 .12 ..................................................................................................... . . . 45 7.1.10 .45 7.2 M AXIMUM KErr CALCULATIONS FOR N ORMAL CONDITIONS ................................................... 45 7.3 M ARGIN EVALUATION .................................................................................................................. 45 7.4 ABNORMAL AND ACCIDENT CONDITIONS ................................................................................ 46 7.5 MAXIMUM KE..CALCULATIONS FOR ABNORMAL AND ACCIDENT CONDITIONS ..................... 46 7.6 46 7.7 SPENT FUEL RACK INTERFACES ............................................................................................... 47

8. CONCLUSIO N .................................................................................................................................. 48

9. REFERENCES .................................................................................................................................. 49 Project No. 2127 Report No. HI-2125245 Page 2

Supplement 1: Additional Calculations to Support the Revised NETCO-SNAP-IN Rack Insert D esig n ...................................................................................................... S I-I Project No. 2127 Report No. HI-2125245 Page 3

List of Tables Table Description Page Table 2.1 (a) Summary of the Area of Applicability of the MCNP5-1.51 Benchmark .51 Table 2.1 (b) Analysis of the MCNP5-1.51 calculations 52 Table 2.1 (c) 53 Table 5. 1(a) 54 Table 5.1 (b) 55 Table 5.1 (c) 56 Table 5.1 (d) 57 Table 5.1 (e) 58 Table 5.2(a) Reactor Core and Spent Fuel Pool Parameters 59 Table 5.2(b) Reactor Control Blade Data 60 Table 5.2(c) 61 Table 5.3(a) Fuel Rack Parameters and Dimensions 62 Table 5.3(b) Fuel Rack Insert Parameters and Dimensions 63 Table 5.4(a) Non-Fuel Material Compositions 64 Table 5.4(b) Summary of the Fuel and Fission Product Isotopes Used in Calculations 65 Table 7.1 (a) 66 Table 7.1 (b) 67 Table 7.1 (c) 68 Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122 Table 7.2(a) 69 Lattices Table 7.2(b) Results of the MCNP5-1.51 Calculations for GE14 Lattice Type 5 71 Results of the MCNP5-1.51 Calculations for Design Basis and Table 7.3 72 Simplified Model of SVEA-96 Optima2 Q122 Lattice Type 146 Table 7.4 Results of the MCNP5-1.51 Calculations for Core Operating Parameters 73 Results of the MCNP5-1.51 Calculations for the Effect of Water Table 7.5 74.

Temperature and Density Table 7.6(a) Re-ults nf the MCNP5-l 51 Calculation-, for the Denletion I Jncertaintv 75 Table 7.6(b) 76 Table 7.7 Results of the MCNP5-1.51 Calculations for Fuel Tolerances 77 Table 7.8 Results of the MCNP5-1.51 Calculations for Rack Tolerances 78 Project No. 2127 Report No. HI-2125245 Page 4

Table Description Page Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning in Table 7.9(a) 79 SFP Racks Results of the MCNP5-1.51 Calculations for Fuel Orientation in SFP Table 7.9(b) 80 Racks Table 7.10 81 Table 7.11 Maximum keff Calculation for Normal Conditions in SFP Racks 82 Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate Table 7.12(a) the Effect of Nominal Values Instead of Using Minimum B4C Loading 83 and Minimum Insert Thickness on Reactivity Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate Table 7.12(b) 84 the Effect of the Actual Optima2 Q122 Fuel Assembly Table 7.12(c) Margin Evaluation Summary of the Margin Evaluation 85 Results of the MCNP5-1.51 Calculations for the Abnormal and Table 7.13(a) 86 Accident Conditions on Reactivity of SFP Results of the MCNP5-1.51 Calculations for the Empty Storage Rack Table 7.13(b) 87 Cell without Insert Maximum kff Calculation for Abnormal and Accident Conditions in Table 7.14 88 SFP Racks Table 7.15 89 Results of the MCNP5-1.51 Calculations for Axially Infinite Optima2 Table 7.16 90 Q122 Lattices Table 7.17 Results of the MCNP5-1.51 Calculations for SFR Interface 91 Table 7.18 92 K

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1 Project No. 2127 Report No. HI-2125245 Page 5

Table Di-crintinn Page U

U Table SI-1 Fuel Rack Insert Revised Dimensions SI-5 Table S1-2 Results of the MCNP5 Calculations for Revised Rack Tolerances SI-6 Results of the MCNP5-1.51 Calculations for Revised Fuel Radial Table S1-3 S1-7 Positioning in SFP Racks Results of the MCNP5-1.51 Calculations for Revised Fuel Orientation Table S1-4 SI-8 in SFP Racks Maximum keff Calculation for Normal Conditions in Revised SFP Table S1-5 SI-9 Racks Maximum keff Calculation for the Bounding Accident Condition in SI-10 Table S1-6 Revised SFP Racks Project No. 2127 Report No. HI-2125245 Page 6

List of Figures Figure Description Page 2-D representation of the CASMO-4 models of SVEA-96 Optima2 Q122 lattices (a) Lattice type 146; (b) lattice type 147; (c) lattice Figure 2.1 93 type 148; (d) lattice type 149; (e) lattice type 150; (f) lattice type 151.

2-D representation of the CASMO-4 models of GE14 lattices (a)

Figure 2.2 94 Lattice type 2; (b) lattice type 3; (c) lattice type 4; (d) lattice type 5.

2-D representation of the CASMO-4 models of GE 8x8 lattices (a)

Figure 2.3 95 Lattice 854; (b) lattice 855.

2-D representation of the CASMO-4 models of GE 7x7 lattices (a)

Figure 2.4 96 Lattice type V; (b) lattice type W.

2-D representation of the CASMO-4 models of ATRIUM 9B Figure 2.5 lattices (a) Lattice SPCA9-3.96L-10G6.5; (b) lattice SPCA9-3.96L- 97 S11G6.5; (c) lattice SPCA9-3.96L-I 1G5.5.

2-D representation of CASMO-4 SVEA-96 Optima2 Q122 model Figure 2.6(a) 98 used for depletion analyses.

2-D representation of MCNP5-1.51 SVEA-96 Optima2 Q122 Figure 2.6(b) 99 model equivalent of the CASMO-4 model.

2-D representation of MCNP5-1.51 SVEA-96 Optima2 Q122 Figure 2.6(c) 100 model used in criticality calculations.

2-D representation of the MCNP5-1.51 SFP racks radial positioning for the 2x2 array models (a) Cell centered positioning; (b) every fuel assembly is positioned toward the Figure 2.7 center; (c) every fuel assembly is positioned toward the comer 101 where the insert wings connect; (d) every fuel assembly is positioned away from the comer where the insert wings connect; (e) every fuel assembly is middle between insert and cell walls.

2-D representation of the MCNP5-1.51 SFP racks radial Figure 2.8 positioning for the 8x8 array model Every fuel assembly is 102 positioned toward the center.

2-D representation of the MCNP5-1.51 fuel orientation in SFP rack Figure 2.9 cell for the 2x2 array models (a) The reference case; (b) through (e) 103 the other evaluated cases.

A 2-D representation of the MCNP5-1.51 model of misloaded fuel Figure 2.10(a) 104 assembly/missing insert (eccentric).

A 2-D representation of the MCNP5-1.51 model of misloaded fuel Figure 2.10(b) 105 assembly/missing insert (cell centered).

Project No. 2127 Report No. HI-2125245 Page 7

Figure Description Page A 2-D representation of the MCNP5-1.51 model of mislocated fuel Figure 2.11 (a) 106 assembly in the comer between two racks.

A 2-D representation of the MCNP5-1.51 model of mislocated fuel Figure 2.11 (b) 107 assembly in the comer between two racks (eccentric).

A 2-D representation of the MCNP5-1.51 model of mislocated fuel Figure 2.11 (c) 108 assembly in the comer between two racks (cell centered).

A 2-D representation of the MCNP5-1.51 model of mislocated fuel Figure 2.12(a) 109 assembly adjacent to the platform (eccentric).

A 2-D representation of the MCNP5-1.51 model of mislocated fuel Figure 2.12(b) 110 assembly adjacent to the platform (cell centered).

2-D representation of the MCNP5-1.51 SFP racks radial positioning of the GEl4 fuel assemblies for the 2x2 array models (a) Cell centered positioning with fuel channel; (b) Cell centered Figure 2.13 positioning, without fuel channel; (c) every fuel assembly without 1l1 fuel channel is moved toward the center; (d) every fuel assembly without fuel channel is moved away from the comer where the insert wings connect.

A 2-D representation of the CASMO-4 model of the SVEA-96 Figure 5.1 112 Optima2 fuel lattice 146 in the core.

Figure 5.2(a) Quad Cities Unit 1 SFP. 113 Figure 5.2(b) Quad Cities Unit 2 SFP. 114 Figure 5.3 Insert cross section profile. 115 A 2-D representation of the MCNP5-1.51 model of the SFP rack Figure 5.4 116 cell with insert.

Figure 7.1 117 Figure 7.2 118 Figure 7.3 119 Project No. 2127 Report No. HI-2125245 Page 8

Figure Description Page Figure S I -I Insert cross section profile SI-11 Project No. 2127 Report No. HI-2125245 Page 9

1. INTRODUCTION This report documents the criticality safety evaluation for the storage of spent BWR fuel in the Unit 1 and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by Exelon. The Unit I and Unit 2 SFP racks are identical and are designed to accommodate BWR fuel. Currently, the SFP racks credit BORAFLEX for reactivity control. This new analysis will not credit the BORAFLEX but will instead credit new NETCO-SNAP-INg, rack inserts, which are new to Quad Cities but not new relative to their use for spent fuel pool reactivity control. This analysis will demonstrate that with credit for the inserts the effective neutron multiplication factor (keff) in the SFP racks fully loaded with fuel of the highest anticipated 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 abnormal and accident conditions, the reactivity will not exceed the regulatory limit.

Criticality control in the SFP, as credited in this analysis, relies on the following:

• Fixed neutron absorbers o NETCO-SNAP-IND4 rack inserts in SFP rack cells

• 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:

• Burnup credit

• BORAFLEX.

Project No. 2127 Report No. HI-2125245 Page 10

2. METHODOLOGY 2.1 GeneralApproach The analysis is pe rformed consistent with regulatory requirements and guidance. The calculations are perform ed using either the worst case bounding approach or the statistical analysis approach with respect to th e various calculation parameters. The approach considered for each parameter is discussed below.

2.2 Computer Codes and Cross Section Libraries 2.2.1 MCNP5-1.51 MCNP5-1.51 is a three-dim ensional Monte Carlo code developed at the Los Alam os National Laboratory [1]. MCNP5-1.51 calcula tions use continuous energy cr oss-section data based on ENDF/B-VII. MCNP is selected because it has history of successful use in fuel storage criticality analyses and has m ost of the necessary featu res (except for fuel depletion analysis) for the analysis to be performed for Quad Cities Station SFP.

The convergence of a Monte Carlo criticality problem is sensitive to the following parameters:

(1) 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 MCNP5 calculations are perform ed with a m inimum of 12,000 histories per cycle, a m inimum of 150 sk ipped cycles before averaging, and a minimum of 150 cycles that are accumulated. The initial source is specified as uniform over the fueled regions (assemblies).

2.2.1.1 MCNP5-1.51 Validation Project No. 2127 Report No. HI-2125245 Page 11 Holtec

2.2.1.1.1 Project No. 2127 Report No. HI-2125245 Page 12

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 2.3.3). A validation for CASMO-4 to develop a bias and bias uncertainty is not necessary because the results of the CASMO-4 sensitivity studies are not used as input into the kff calculations. However, the code authors have validated CASMO-4 against MCNP and various critical experiments [5].

The version of the CASMO-4 code used in this application has a built-in limitation in a number of isotopes that may be extracted for specific pins. Therefore, two independent CASMO-4 depletion calculations were perforlned to separately extract the actinides and fission products.

The extracted isotopes were further combined and used in MCNP5-1.51 calculations.

2.3 Analysis Methods 2.3.1 Design Basis Fuel Assembly There are various fuel designs stored in the Quad Cities 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 to determine keff at the 95/95 level. This approach follows the guidance in [2] and [6], and is further described below.

Project No. 2127 Report No. HI-2125245 Page 13

2.3.1.1 Peak Reactivity The BWR fuel designs used at the Quad Cities 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 lattices (including blankets) that can have different axial lengths, uranium loadings (also mixed oxide loading, for MOX fuel), 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 can therefore all have diffentparacity Maxial lattices within a single fuel assembly 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 within 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 of the SVEA-96 Optima2 fuel assembly (as will be seen in Section 7, this is the fuel assembly with the design basis lattice) over a burnup range to determine the burnup at peak reactivity for every lattice. 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.

2.3.1.1.2 Project No. 2127 Report No. HI-2125245 Page 14

2.3.1.2 Project No. 2127 Report No. HI-2125245 Page 15

2.3.1.3 Detenrination of the Design Basis Fuel Assembly Lattice 2.3.1.4 Optima2 CASMO-4 Model Simplification Effect As previously discussed in Section 2.3.1.2, various fuel designs were provided. Of these fuel designs, the SVEA-96 Optima2 designs were specified to be bounding. The Optima2 model in CASMO-4 is described as the SVEA-96 model provided in the CASMO-4 manual [4]. This CASMO-4 internal model is slightly different from the actual fuel assembly geometry.

Therefore, it is important to evaluate and if necessary quantify the reactivity effect of the CASMO-4 model simplifications inherent in the code. The CASMO-4 model geometry of the SVEA-96 Optima2 fuel differs from the SVEA-96 Optima2 fuel as follows:

Project No. 2127 Report No. HI-2125245 Page 16

With respect to the fuel assembly geometry models, the amount of zirconium (and therefore the amount of water) in the CASMO-4 model of the SVEA-96 Optima2 fuel is reasonably similar to that of the actual SVEA-96 Optima2 fuel and therefore these built-in CASMO-4 simplifications are acceptable. However, to evaluate the CASMO-4 model geometry simplification effect on reactivity, an applicable set of code-to-code comparisons is performed. The following cases are evaluated.

For the purpose of showing that the two codes calculate an equivalent reactivity the following comparisons are made:

Project No. 2127 Report No. HI-2125245 Page 17

" Case 2.3.1.4.1 is compared to Case 2.3.1.4.2 at 0 GWD/MTU to show that the two codes calculate similar results with respect to the fuel assembly and storage rack geometry.

• Case 2.3.1.4.1 is compared to Case 2.3.1.4.2 at peak reactivity burnup to quantify the reactivity difference due to the effect of the spent fuel. The two codes use different cross section library versions and calculation sequences. The main calculation sequence difference between the two codes is that CASMO-4 uses a thermal expansion of spent fuel pellet which effects the fuel density [4]. The actual density is conservatively used in MCNP5-1.51. The results are expected to show that the MCNP5-1.51 code is conservative with respect to the CASMO-4 code. Any non-conservative result would be treated as a bias.

• Case 2.3.1.4.3 is compared to Case 2.3.1.4.2 to show the reactivity difference between the simplified MCNP5-1.51 model and the design basis model that is slightly modified to be similar to the CASMO-4 insert orientation. This case is expected to show that the design basis model with respect to the fuel pin pitch (and subsequent sub-bundle pitch) is conservative. This is expected to be conservative because the design basis model fuel compositions are taken from the average fuel pin pitch CASMO-4 calculations and used in the MCNP5-1.51 design basis actual fuel pin locations. Any non-conservative result would be treated as a bias.

Case 2.3.1.4.3 is compared to the result of the actual design basis results (similar to Case 2.3.1.4.3 but with the bounding insert orientation) to show that the design basis model is conservative.

2.3.1.5 Core Operating Parameters As previously discussed, CASMO-4 is used to perform depletion calculations to determine the spent fuel isotopic composition. The operating parameters for spent fuel depletion calculations are discussed in this Section. The 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 axial enrichment distribution and effect of burnable absorbers are discussed in Section 2.3.1.5.3 and Section 2.3.1.5.2, respectively. Sensitivity studies are performed to show the effect of each individual parameter, and to confirm that the selected values are in fact appropriate when combined at their wdnr~t C*.*

2.3.1.5.1 Reactor Power Uprate

ý0 o aetermine tne etiect o0 tne power uprate on tne reactivity oi ruei assemblies in the SFP racks, the following evaluations are performed.

Project No. 2127 Report No. HI-2125245 Page 18

0 2.3.1.5.2 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 Gd. As discussed in Section 2.3.1.1, 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. Note that integrated absorbers do not change the amount of water in the assembly, which is a large part of the effect of non-integral absorbers.

2.3.1.5.3 Axial and Planar Enrichment Variations 2.3.1.5.4 Fuel Assembly De-Channeling The SVEA-96 Optima2 fuel assembly (the most reactive fuel assembly, as will be shown in Section 7) cannot be de-channeled for storage in the SFP because of its specific design.

However, GEl4 (the most second reactive fuel assembly, as will be shown in Section 7) may be de-channeled. Studies are performed to evaluate the effect of storage of GEl4 without the Zr channel at various radial positioning in the storage cells. The following cases are evaluated.

" Case 2.3.1.5.4.1: This is the reference for Case 2.3.1.5.4.2 through Case 2.3.1.5.4.4. The MCNP5-1.51 model used herein is a 2x2 array with the cell centered fuel assembly that includes the Zr channel, as shown in Figure 2.13(a).

" Case 2.3.1.5.4.2: The MCNP5-1.51 is a 2x2 array of GEl4 fuel assembly lattice 5 (the most reactive lattice of GEl4, as will be shown in Section 7). The Zr channel is removed, as shown in Figure 2.13(b). The fuel assemblies are cell centered.

• Case 2.3.1.5.4.3: The MCNP5-1.51 is the same as that of Case 2.3.1.5.4.2, except the fuel assemblies are eccentric toward the center, as shown in Figure 2.13(c).

• Case 2.3.1.5.4.4: The MCNP5-1.51 is the same as that of Case 2.3.1.5.4.2, except the fuel assemblies are eccentric away from the corner where the insert wings connect, as shown in Figure 2.13(d).

Project No. 2127 Report No. HI-2125245 Page 19

2.3.1.6 2.3.2 Reactivity Effect of Spent Fuel Pool Water Temperature The Quad Cities Station SFP has a normal pool water temperature operating range below 150 °F.

For the nominal condition, the criticality analyses are to be performed at the most reactive temperature and density [2]. 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 SFP racks which credit inserts, the most reactive SFP water temperature and density is expected to be at 39.2 OF and I g/cc, respectively.

The standard cross section temperature in MCNP5-1.51 is 293.6 K. Cross sections are also available at other temperatures; however, not usually at the desired temperature for SFP criticality analysis. MCNP5-1.51 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(a,13) card.

The S(a,3) 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 S(a,p3) 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,3) 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(a,p3) calculations are performed for both upper and lower bounding S(a,p3) values, if needed.

The studies mentioned above are performed for the following cases for the single cell MCNP5-1.51 SFP model (with periodic boundary conditions through the centerline of the surrounding water 2):

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• Case 2.3.2.1 (reference case): Temperature of 39.2 OF (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(a,3) card corresponds to a temperature of 68.81 OF (293.6 K).

• Case 2.3.2.2: Temperature of 150 OF (338.71 K) and a corresponding density of 0.98026 g/cc are used to determine the reactivity at the high end of the temperature range. The S(a,I3) card corresponds to a temperature of 68.81 OF (293.6 K).

• Case 2.3.2.3: Temperature of 150 OF and a corresponding density of 0.98026 g/cc. The S(a,3) card corresponds to a temperature of 170.33 OF (350 K).

The bounding 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 MCNP5-1.51 models used in the design basis calculation.

2.3.3 Fuel Depletion Calculation Uncertainty To account for the uncertainty of the number densities in the depletion calculations erformed in 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 20 3).

Calculations are performed for the single cell model of design basis fuel assembly.

The uncertainty is determined by the following:

Uncertaintylsotopics = [ (kcalc kcalc-l) + 2 * . (acalc..i 2 + Gcalc..2 2 ) ]

• 0.05 with kcalc-i1 kcalc with spent fuel kcalc-2 = kcalc with fresh fuel ocaic-l = Standard deviation of kcalc-i Ucalc-2 = Standard deviation of kcalc-2 The result of the MCNP5-1.51 calculation for the fuel depletion calculation uncertainty is statistically combined with other uncertainties to determine kff.

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2.3.4 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 [2]. The evaluations use the same MCNP5-1.51 models used in the design basis calculation.

2.3.4.1 Fuel Manufacturing Tolerances The BWR fuel tolerances for Optima2 Q122 fuel (which is the most reactive fuel design evaluated herein) are presented in Table 5.1 (a). Fuel tolerance calculations are performed using the design basis fuel assembly lattice, and 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 MCNP5-1.51 reference case (nominal parameter values) at the 95% probability at a 95% confidence level using the following equation:

delta-kcalc = (kcalc2 - kcalcl) +/-2

• 4j (Cl2 + (2)

The following fuel tolerances are considered in this analysis:

" Fuel enrichment

" Gd loading

" Fuel pellet density (U0 2 and UO 2 +Gd 2O 3 fuel rods)

" Fuel pellet outer diameter (OD)

• Fuel cladding inner diameter (ID)

" Fuel cladding OD

• Fuel pin pitch

" Fuel sub-bundle pitch 3

• Combination of 4 o Water wing canal inner width o Channel outer square width o Channel comer inner radius o Central water canal inner square width

• Combination of 4 o channel wall thickness 3 For fuel sub-bundle pitch uncertainty calculation, the fuel hardware (channel, central water channel and water wings) is fixed. The fuel lattices are moved only.

4 Conservatively, the various tolerances are considered together. The tolerance limits that result in an increase of the amount of water in the core are considered together in one set of uncertainty calculations, and the tolerance limits that result in a decrease of the amount of water in the core are considered together in another set of uncertainty calculations.

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o Water cross wall thickness 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 keff value.

2.3.4.2 SFP Storage Rack Manufacturing Tolerances The SFP rack tolerances are presented in Tables 5.3(a) and 5.3(b). The single cell MCNP5-1.51 model is used to determine the reactivity effect of the tolerance, and the full value of the tolerance is applied for each case. The MCNP5-1.51 tolerance calculation is compared to the MCNP5-1.51 reference case with a 95% probability at a 95% confidence level using the following equation:

delta-kcalc = (kcalc2 - kcaicl) +/-2 * ( + 22)

+] G32 The following SFP rack manufacturing tolerances are considered in this analysis:

" Storage cells:

o Cell ID and cell pitch o Cell wall thickness

" Rack inserts (poison) o Width 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 keff 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; i.e., they are treated as a bias instead of uncertainty, for conservatism and simplification.

2.3.5 Radial Positioning 2.3.5.1 Fuel Assembly Orientation in the Core The fuel assembly orientation in the core with respect to its control blade does not change and therefore the design basis calculations consider the only possible configuration.

2.3.5.2 Fuel Radial Positioning in the Rack The BWR fuel that is loaded in the SFP racks may not rest exactly in the center of the storage cell. Evaluations are performed to determine the most limiting fuel radial location. The following eccentric fuel positioning cases are analyzed:

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" Case 2.3.5.2.1: This is the reference for Case 2.3.5.2.2 through Case 2.3.5.2.5. The MCNP5-1.51 model used herein is a 2x2 array which is the same as the primary single bundle MCNP5-1.51 model used elsewhere in this analysis. In both models the fuel is centered in the rack cell. See Figure 2.7(a).

• Case 2.3.5.2.2: Every fuel assembly is positioned toward the center, for the 2x2 array, as shown in Figure 2.7(b).

" Case 2.3.5.2.3: Every fuel assembly is positioned toward the comer where the insert wings connect, for the 2x2 array, as shown in Figure 2.7(c).

• Case 2.3.5.2.4: Every fuel assembly is positioned away from the comer where the insert wings connect, for the 2x2 array, as shown in Figure 2.7(d).

" Case 2.3.5.2.5: Every fuel assembly is centered between insert and cell walls, for the 2x2 array, as shown in Figure 2.7(e).

• Case 2.3.5.2.6: This is the reference for Case 2.3.5.2.7 through Case 2.3.5.2.10. The MCNP5-1.51 model used herein is an 8x8 array which is the same as the primary single bundle MCNP5-1.51 model used elsewhere in this analysis. In both models the fuel is centered in the rack cell.

• Case 2.3.5.2.7: Every fuel assembly is positioned toward the center, for the 8x8 array, as shown in Figure 2.8.

" Case 2.3.5.2.8: Every fuel assembly is positioned toward the comer where the insert wings connect, for the 8x8 array.

• Case 2.3.5.2.9: Every fuel assembly is positioned away from the corner where the insert wings connect, for the 8x8 array.

• Case 2.3.5.2.10: Every fuel assembly is centered between insert and cell walls, for the 8x8 array.

" Case 2.3.5.2.11: This is the reference for Case 2.3.5.2.12. The MCNP5-1.51 model used herein is a single rack cell where the fuel is centered.

• Case 2.3.5.2.12: The fuel assembly is centered between insert and cell walls, for the single rack cell.

The maximum positive reactivity effect of the MCNP5-1.51 calculations for the fuel radial positioning is added as the bias and the corresponding 95/95 uncertainty is statistically combined with other uncertainties to determine keff.

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Note that the evaluations use the same MCNP5-1.51 models with periodic boundary conditions used in the design basis calculation, except that the array size is larger. The isotopic compositions of the fuel rods are the same as those of the design basis fuel assembly.

2.3.5.3 Inserts Radial Positioning Since the insert width and SFR cell inner diameter are comparable, and each insert is installed into the rack cell such that the insert becomes an integral part of the fuel rack, no uncertainty in the positioning for inserts is evaluated. The water gap between rack wall and insert is not assumed, since it may provide a small flux trap effect. Nevertheless, the orientation of fuel assembly with respect to position of insert is considered in Section 2.3.5.4.

2.3.5.4 Fuel Orientation in SFP Rack Cell As described in Section 5.1, fuel assemblies have various radial fuel enrichments and gadolinium distribution. Also, one comer 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 (more discussion is provided in Section 2.3.1.1.2) and one fuel assembly comer may be more reactive than other comers and therefore 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 with respect to the insert.

The MCNP5-1.51 model of the reference case is the design basis fuel in the 2x2 array, as shown in Figure 2.9(a). The MCNP5.1.51 models of the other four cases are the same as that of the reference case, except with different orientation of fuel assemblies with respect to the inserts.

Figure 2.9(b) through Figure 2.9(e) show the configurations of the fuel assemblies in the SFP cells for the evaluated cases.

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 the same as those of the design basis fuel assembly.

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2.3.6.1 2.3.6.1.1 Project No. 2127 Report No. HI-2125245 Page 26

2.3.6.1.2 -

2.3.6.2 2.3.7 Insert Coupon Measurement Uncertainty There is a measurement uncertainty associated with the B-10 content in the poison test coupons. In this analysis, the minimum B-10 loading and the minimum insert thickness are conservatively used for criticality calculations. Therefore, the coupon measurement uncertainty is not evaluated further in the analysis.

2.3.8 Maximum keff Calculation for Normal Conditions The calculation of the maximum kff 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 insert thickness and its B-10 loading are taken at their worst case values.

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keff is determined by the following equation:

klif = kcai, + uncertainty + bias where uncertainty includes:

and the bias includes Note that each uncertainty is statistically combined with other uncertainties, while biases are added together in order to detennine keff.

The approach used in this analysis takes credit for residual Gd.

2.4 Margin Evaluation The criticality analysis is performed using several conservative assumptions which introduce quantifiable margin into the analysis. Four main conservative assumptions are:

• Minimum insert B 4 C loading

• Minimum insert thickness

" Minimum amount of B-10 in boron

• Bounding lattice throughout the entire length of fuel assembly.

To evaluate this margin, the following cases are evaluated:

• Case 2.4.1: This is the design basis fuel assembly. This is the reference for Case 2.4.2 and Case 2.4.3.

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" Case 2.4.2: This case is the same as Case 2.4.1, except the nominal insert B4 C loading, nominal insert thickness and nominal amount of B-10 in boron are used.

• Case 2.4.3: This case is the same as Case 2.4.1, except the model includes each Optima2 Q122 fuel lattice in the appropriate axial position. However, the top and bottom blankets were conservatively replaced by adjacent fuel lattices. The peak reactivity burnup for each individual Optima2 Q122 lattice under the design basis core operation parameters was determined separately and used in this case (i.e. each lattice is at its individual peak reactivity). Therefore, the model represents a conservative maximum but unrealistic reactivity of the actual Optima2 fuel assembly.

The differences between the reactivity of Cases 2.4.2 and 2.4.3 and the reactivity of reference Case 2.4.1 provide a quantified margin.

Note that the evaluations use the same MCNP5-1.51 models used in the design basis calculation.

The isotopic compositions of the fuel rods of Case 2.4.1 and Case 2.4.2 are the same as those of the design basis fuel assembly.

2.5 Fuel Movement, Inspection and Reconstitution Operations 2.6 Accident Condition The accidents considered are:

" SFP temperature exceeding the normal range

" Dropped assemblies

" Storage cell distortion

" Missing insert

• Misloaded fuel assembly (a fuel assembly in the wrong location within the storage rack)/

Missing an insert

• Mislocated fuel assembly (a fuel assembly in the wrong location outside the storage rack)

• Miss-installment of an insert on wrong sides of a cell

• Insert mechanical wear

" Rack movement Project No. 2127 Report No. HI-2125245 Page 29

Those are briefly discussed in the following sections.

Note that the double contingency principle as stated in [2] 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 keff calculations performed for the accident conditions are done with a 95%

probability at a 95% confidence level.

The accident conditions are considered at the 95/95 level usin the total corrections from the desi 2.6.1 Temperature and Water Density Effects The SFP water temperature accident conditions for consideration are the increase in SFP water temperature above the maximum SFP operating temperature of 150 OF. The decrease in temperature was already considered for the temperature coefficient determination as discussed in Section 2.3.2.

To bound the potential increase in reactivity due to increased SFP temperature, the following case is evaluated:

Case 2.6.1: This case uses a temperature of 255 OF (397.04 K) and a density of 0.84591 g/cc. The S(cx,13) card corresponds to a temperature of 260.33 OF (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 OF.

The evaluation use the same MCNP5-1.51 model used in the design basis calculation.

Note that as discussed in Section 2.3.2, SFP storage racks with strong neutron absorbers, such as inserts, show a higher reactivity at a lower water temperature. The case evaluated above is performed to confirm this statement.

2.6.2 Dropped Assembly - Horizontal For the case 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 more than 12 inches.

Also, the length of the inserts (as indicated in Table 5.3(b)) covers this separation distance. 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.

2.6.3 Dropped Assembly - Vertical into a Storage Cell It is also possible to vertically drop an assembly into a location that might be occupied by another assembly or that might be empty. Such a vertical impact would at most cause a small compression of the stored assembly, if present, or result in a small deformation of the baseplate for an empty cell.

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These deformations could potentially increase reactivity. However, the reactivity increase would be small compared to the reactivity increase created by the 'misloaded fuel assembly/missing insert' accident (discussed in Section 2.6.5) that does not include the insert in one rack cell. The vertical drop is therefore bounded by this misload accident and no separate calculation is performed for this drop accident.

2.6.4 Storage Cell Distortion A storage cell distortion or altered geometry as a result of fuel handling equipment uplift forces is possible. However, the reactivity increase would be small compared to the possible reactivity increase created by the 'misloaded fuel assembly/missing insert" accident that does not include the insert in one rack cell, as discussed in Section 2.6.5. The storage cell distortion is therefore bounded by the 'misloaded fuel assembly/missing insert' accident and no separate calculation is performed for the storage cell distortion accident.

As a result of significant distortion, the storage cell for whatever reason may not be able to contain the insert and also it will be therefore unacceptable for storage of a fuel assembly. This condition is bounded by the 'misloaded fuel assembly/missing insert' accident. However to show that it is acceptable for normal operation and that the empty storage cell decreases the reactivity of the SFR, the model with an empty storage cell, i.e. without a fuel assembly and insert, in the center of a 8x8 array, is evaluated. Two cases with a cell centered and eccentric position of the fuel assemblies are analyzed.

2.6.5 Misloaded Fuel Assembly/Missing Insert The fuel storage racks are qualified for storage of fuel assembly with the highest anticipated reactivity; thus it is not possible to misload a fuel assembly if every cell with a fuel assembly has an insert.

However, there are a few cells in the SFP racks which are exempt from fuel storage. Those locations are blocked or have partial interferences. In a hypothetical scenario, it is assumed that a fuel assembly is misloaded into a cell with a missing insert. To evaluate the effect, the following cases are evaluated:

• Case 2.6.5.1: The MCNP5-1.51 model includes an 8x8 array. One cell near the center of the rack does not have the insert. The misloaded fuel assembly is the design basis fuel assembly.

This fuel assembly is eccentric toward the walls that are not covered by inserts. Other fuel assemblies are also eccentric toward the misloaded fuel assembly. The periodic boundary conditions are used through the centerline of the surrounding water (BORAFLEX replacement). The temperature of the model is set to the minimum (39.2 'F) with its corresponding water density and S(a,13) card. These temperature and density are bounding for the SFP racks. See Figure 2.10(a).

• Case 2.6.5.2: The MCNP5-1.51 model is the same as Case 2.6.5.1, except with all fuel assemblies centered in the rack cells. See Figure 2.10(b).

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2.6.6 Mislocated Fuel Assembly The Quad Cities SFP layout was reviewed to determine the possible worst case locations for a mislocated fuel assembly. Three hypothetical locations where a fuel assembly may be mislocated are:

• In the water gap between the racks and the pool wall

" In the comer between two racks

• Between the SFP rack and the inspection platform.

The three cited scenarios are evaluated, as follows.

2.6.6.1 Mislocation of a Fuel Assembly in the Water Gap between the Racks and Pool Wall A fuel assembly may be mislocated in the water gap between the racks and the pool wall. Due to the neutron leakage to the outside the storage rack area, the effect of this mislocation is bounded by that of 'mislocation of a fuel assembly between the SFP rack and the inspection platform' accident, as discussed in Section 2.6.6.3. No separate calculation is performed for this accident.

2.6.6.2 Mislocation of a Fuel Assembly in the Comer between Two Racks There are some places in the SFP, but outside of the racks, where the mislocated fuel assembly may be in the comer between two racks (thus the mislocated fuel assembly would be adjacent to the fuel assemblies in racks from two sides). To evaluate the effect of the mislocation of a fuel assembly in the comer between two racks, the following cases are evaluated:

• Case 2.6.6.2.1: The MCNP5-1.51 model is three 8x8 arrays of SFP rack cells. The misplaced fuel assembly is in the comer between two racks. The fuel assemblies in the rack are eccentric toward the mislocated fuel assembly. The misplaced fuel assembly is placed as close to the racks as possible. All fuel assemblies in the model are the design basis fuel assembly. Figures 2.11 (a) and 2.11 (b) show the MCNP5-1.51 model used for this analysis.

• Case 2.6.6.2.2: The MCNP5-1.51 model is the same as Case 2.6.6.2.1, except with all fuel assemblies are centered. See Figures 2.11 (a) and 2.11 (c).

• Case 2.6.6.2.3: The MCNP5-1.51 model is the same as Case 2.6.6.2.1, except the temperature of the model is set to the maximum (150 'F).

• Case 2.6.6.2.4: The MCNP5-1.51 model is the same as Case 2.6.6.2.2, except the temperature of the model is set to the maximum (150 'F).

2.6.6.3 Mislocation of a Fuel Assembly between the SFP Rack and the Inspection Platform As discussed in Section 2.5, the fuel handling/inspection/reconstitution platform may have one fuel assembly in it at a time. There is a possibility that a fuel assembly is mislocated between the Project No. 2127 Report No. HI-2125245 Page 32

SFP racks and the fuel assembly in the platform. To evaluate the effect of the mislocation of a fuel assembly between the SFP Rack and the Inspection Platform, the following cases are evaluated:

" Case 2.6.6.3.1: The MCNP5-1.51 model is an 8x8 array of SFP rack cells. The misplaced fuel assembly is adjacent to the SFP rack and the inspection platform. The fuel assembly in the platform is lined up with the mislocated fuel assembly. The fuel assemblies in the rack are eccentric toward the mislocated fuel assembly: The misplaced fuel assembly is placed as close to the rack and fuel assembly in the inspection station as possible. All fuel assemblies in the model are design basis fuel assembly. The side of the fuel in the platform which does not have any fuel has at least 12 inches of water. Figure 2.12(a) shows the MCNP5-1.51 model used for this analysis.

• Case 2.6.6.3.2: The MCNP5-1.51 model is the same as Case 2.6.6.3.1, except with all fuel assemblies are centered. See Figure 2.12(b).

• Case 2.6.6.3.3: The MCNP5-1.51 model is the same as Case 2.6.6.3.1, except the temperature of the model is set to the maximum (150 TF).

• Case 2.6.6.3.4: The MCNP5-1.51 model is the same as Case 2.6.6.3.2, except the temperature of the model is set to the maximum (150 TF).

2.6.7 Mis-installment of an Insert on Wrong Side of a Cell There is a small possibility that an insert is installed on wrong sides of the cell. In this case, there may not be a poison between a fuel assembly placed in that cell and a fuel assembly in an adjacent cell. However, the effect of this mis-installment is bounded by that of 'misloaded fuel assembly/missing insert' accident that does not include the insert in one rack cell, as discussed in Section 2.6.5. No separate calculation is performed for this accident.

2.6.8 Insert Mechanical Wear Handing accidents and other environmental damage may cause scratches and local wear of inserts. The effect of this accident is bounded by that of 'misloaded fuel assembly/missing insert' accident, as discussed in Section 2.6.5.

2.6.9 Rack Movement In the event of seismic activity, there is a hypothetical possibility that the storage rack arrays may move and come closer to each other. Since there is no water gap modeled between cells of a storage rack, the reactivity of the rack movement case is bounded by the reactivity of the design basis calculation.

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0 0

2.8 Spent Fuel Rack Interfaces The spent fuel pool includes a single type of Region 1 spent fuel racks, which are loaded with the neutron absorbing inserts in every storage cell as well as a uniform fuel assembly loading pattern.

Therefore, any possible water gaps and interfaces between the racks are bounded by the infinite array used in the design basis calculations. However, since the neutron absorbing inserts are located in the same corners of rack cells (e.g. south-west), there are two peripheral rows of the cells (correspondingly, north and east periphery of the pool), which are loaded with the fuel assemblies that have one side that is not adjacent to the insert. Furthermore, one fuel assembly in the corner of the spent fuel pool (correspondingly, north-east corner) has two sides that are not adjacent to the insert. Due to the neutron leakage on the periphery of the spent fuel pool the reactivity increase is not expected. Nevertheless, to evaluate the effect of such conditions, the full spent fuel pool model (74x74 array) loaded with the cell centered design basis fuel assemblies and the model where all fuel assemblies are shifted to the fuel assembly in the corner, which is discussed above, were evaluated.

2.9 Reconstituted Fuel Assemblies The SFP contains various reconstituted assemblies which were examined and determined to be relatively old and low reactivity designs. The reconstitution of these fuel assemblies removed fuel rods and replaced them by either fuel rods that are of the same or less initial enrichment and equal or greater Gd loading (with burnup similar to the rod they replaced) or solid stainless steel rods.

The reactivity effect of this reconstitution is not sufficient to make the reconstituted fuel assembly more reactive than the bounding lattice. Therefore, reconstituted assemblies are covered by the design basis Optima2 Q122 lattice 146. Future reconstituted assemblies will replace fuel rods with stainless steel rods.

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3. ACCEPTANCE CRITERIA 3.1 Applicable Codes, Standards and Guidance's 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 Handling."

• Code of Federal Regulations, Title 10, Part 50.68, "Criticality Accident Requirements."

" USNRC Standard Review Plan, NUREG-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.

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4. ASSUMPTIONS The analyses apply a number of assumptions, either for conservatism or to simplify the calculation approach. Important aspects of applying 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. It is conservative to neglect the spacer grids because this spent fuel pool contains no soluble boron, the region around the fuel rods is under-moderated, as confirmed by the fuel tolerances calculations that change the fuel to moderator ratio (Section 7.1.7.1); therefore, neglecting the spacer grid places more water within the calculation model. In addition, the inconel springs within the spacer are a stronger neutron absorber than water. The active fuel region repeats periodically in the vertical direction. Therefore, neutron absorption in upper and lower tie plates, fuel plenums, etc. is neglected.

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, 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. For the inserts, only the worst case bounding material specifications are used (minimum B-10 loading and minimum thickness).

6. All models are laterally infinite arrays of the respective configuration, neglecting lateral leakage. The exception is where the model boundaries are water, as specified.

7. 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.

8.

9. The full spent fuel pool model is considered as a 74x74 array of storage cells. The water gaps between the spent fuel racks were conservatively neglected.

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5. INPUT DATA 5.1 Fuel Assembly Specification The SFP racks are designed to accommodate the following fuel assembly types used in the Quad Cities Unit 1 and Unit 2, which are presented in a chronologic order along with the initial maximum planar average enrichment (IMPAE):

The specifications for the most reactive fuel assemblies from the fuel product lines discussed above are presented in Table 5.1. The additional specifications for other fuel design variations are presented in Appendix A.

The fuel assembly MCNP model used for the design basis cacltosis resented in Fi ure 5.4.

The fuel rod, cladding and channel are explicitly modeled.

Axially, the design basis MCNP model considers the bounding lattice along the entire length and uses water reflectors at the top and bottom. The MCNP model for the margin evaluation calculations discussed in Section 2.4 differ from the design basis model in that the active length specifically considers each actual lattice in its actual axial coniuainie l h atcsfo h 12bnl are modeled in the same MCNP .

5.2 Reactor Parameters The reactor core 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 Spent Fuel Pool Parameters The spent fuel pool parameters are provided in Table 5.2(a).

5.4 Storage Rack Specification The storage rack specifications that are used in the criticality analysis are summarized in Tables 5.3(a) and 5.3(b). The Quad Cities Unit I and Unit 2 SFP are shown in Figures 5.2(a) and 5.2(b),

respectively.

The MCNP5-1.51 SFP model consists of a single rack cell with periodic boundary conditions through the centerline of the water (BORAFLEX replacement), thus simulating an infinite array of storage cells. The storage rack cell is modeled the same length as the active fuel and all other storage rack materials are neglected. The neutron absorber is modeled with the worst case bounding values (the minimum B-10 loading and the minimum thickness) provided in Table 5.3(b) and Figure 5.3.

The cell wall thickness of the boundary is different from that of inner walls. The cell wall thickness of the boundary is thicker than the inner wall thickness. The SFP model uses the inner cell wall thickness only, as given in Table 5.3(a), because it decreases the amount of steel in the model, which acts a neutron absorber.

The MCNP5-1.51 SFP rack cell model is shown in Figure 5.4.

5.4.1 Material Compositions The MCNP5-1.51 material specification is provided in Table 5.4(a) for non-fuel materials, and in Table 5.4(b) for fuel materials.

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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. MCNP5-1.51 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.

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7. ANALYSIS 7.1 Design Basis and UncertaintyEvaluations 7.1.1 7.1.2 Determination of the Design Basis Fuel Assembly Lattice As discussed in Section 2.3.1.3, MCNP5-1.51 calculations were performed to determine the design basis lattice. The results for the SVEA-96 Optima2 Q122 fuel assembly are presented in Table 7.2(a) . The results for the GE14 lattice type 5 are presented in Table 7.2(b), along with the bounding result of the SVEA-96 Optima2 Q122. As can be seen, the SVEA-96 Optima2 Q122 lattice type 146 is bounding, and thus it is selected as the design basis lattice. The CASMO-4 model of the SVEA-96 Optima2 bundle Q122 lattice 146 used for depletion calculations is shown in Figure 5.1.

7.1.2.1 Fuel Assembly De-Channeling As discussed in Section 2.3.1.5.4, the reactivity of the second most reactive assembly with no Zr channel at various radial positioning was evaluated. The results are provided in Table 7.2(b) and compared with the reactivity of the design basis lattice (SVEA-96 Optima2 Q122 lattice type 146).

As can be seen, the SVEA-96 Optima2 Q122 lattice type 146 is bounding. Therefore, storage of fuel assemblies without channels is acceptable.

7.1.3 Optima2 CASMO-4 Model Simplification Effect As discussed in Section 2.3.1.4, the effect of CASMO-4 model simplifications on the calculated reactivity of the SVEA-96 Optima2 Q122 lattice 146 was evaluated. The results are provided in Table 7.3. As can be seen, the reactivity of the simplified model is comparable to that of the complete model of SVEA-96 Optima2 Q122 lattice 146 (essentially within the 95/95 uncertainty between the two calculations). Therefore, the results show that the CASMO-4 model simplification Project No. 2127 Report No. HI-2125245 Page 42

does not have a significant impact on the analysis conclusions regarding the determination of the design basis lattice.

7.1.4 Core Operating Parameters As discussed in Section 2.3.1.5, the effects of the core operating parameters on the reactivity were evaluated. The results are provided in Table 7.4. The results show that the two dominant core operating parameters are the control blade insertion and void fraction. The other core operating parameters have an insignificant impact. Therefore, the design basis (bounding) core operating parameters are: control blades inserted, 0% void fraction, maximum fuel and moderator temperature and maximum specific power.

7.1.4.1 Reactor Power Uprate As discussed in Section 2.3.1.5.1, the effect of the MUR on the reactivity was evaluated. The results are provided in Table 7.4. The most important core operating parameters are rodded operation (control blades) and void fraction. Other parameters have relatively negligible effects on reactivity.

As can be seen, the calculations with the increased power density show statistically equivalent results, which confirms the negligible effect of the reactor power uprate on reactivity.

7.1.5 Water Temperature and Density Effect As discussed in Section 2.3.2, the effects of water temperature, and the corresponding water density and temperature adjustments (S(ct,3)) were evaluated for SFP racks. The results of these calculations are presented in Table 7.5.

The results'of the SFP 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 OF at a density of 1 g/cc, and these values are used for all calculations in SFP racks.

7.1.6 Depletion Uncertainty As discussed in Section 2.3.3, the uncertainty of the number densities in the depletion calculations was evaluated. The results of these calculations are presented in Table 7.6(a).

Also, as discussed in Section 2.2.1.1.1, the uncertainty associated with FPs and LFPs was evaluated.

The results of these calculations are presented in Table 7.6(b).

These two uncertainties are statistically combined with other uncertainties to determine k1ff in Table 7.11 and Table 7.14.

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7.1.7 Fuel and Rack Manufacturing Tolerances 7.1.7.1 Fuel Assembly Tolerances As discussed in Section 2.3.4.1, the effect of the BWR fuel tolerances on reactivity was determined. The results of these calculations are presented in Table 7.7. The maximum positive delta-k value for each tolerance is statistically combined.

The maximum statistical combination of fuel assembly tolerances is used to determine kff in Table 7.11 and Table 7.14.

7.1.7.2 SFP Rack Tolerances As discussed in Section 2.3.4.2, the effect of the manufacturing tolerances on reactivity of the SFP racks with inserts was determined. The results of these calculations are presented in Table 7.8. 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 7.11 and Table 7.14.

7.1.8 Radial Positioning 7.1.8.1 Fuel Assembly Radial Positioning in SFP Rack As discussed in Section 2.3.5.2, twelve fuel assembly radial positioning cases in racks were evaluated. The results of these calculations are presented in Table 7.9(a). For each eccentric position case, the result for similar but cell centered case is considered as a reference. The results show that most cases show a negative reactivity effect, however some delta kcaIc values are positive. Therefore, a maximum delta kcai value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Table 7.11 and Table 7.14.

7.1.8.2 Fuel Orientation in SFP Rack As discussed in Section 2.3.5.4, five fuel assembly orientation cases in racks were evaluated. The results of these calculations are presented in Table 7.9(b). The result for the reference case is also included. The results show that all cases are statistically equivalent and the reactivity effect of fuel orientation is negligible. Nevertheless, a maximum positive delta kcalc value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Table 7.11 and Table 7.14.

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7.1.9 Fuel Rod Geometry Change 7.1.9.1 results are presented in lable 7.

The maximum 'kcalc - kcatc,reference' is added as a bias, and the '2

• 4J (acalc 2 + Gcalc,reference2), (95/95 uncertainty) is added as an uncertainty to determine keff in Table 7.11 and Table 7.14.

7.1.9.2 7.1.10 7.2 Maximum keff Calculationsfor Normal Conditions As discussed in Section 2.3.8, the maximum keff for nonnal conditions is calculated. The results are tabulated in Table 7.11. 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.

7.3 Margin Evaluation As discussed in Section 2.4, the margin analyses were performed using the nominal values for poison thickness and loading, as well as the actual lattice configuration of the Optima2 Q122 fuel assembly. The results of calculations are provided in Table 7.12(a) and Table 7.12(b). As can be seen and is expected, the reactivity of design basis is larger. The use of a minimum B-10 loading relative to use of a nominal B-10 loading with tolerance uncertainty provide an additional -1%

reactivity margin to the regulatory limit with a 95% probability at a 95% confidence level.

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The summary of the margin evaluation is presented in Table 7.12(c). The result shows that quantified margin remains in the analysis to offset potential effects not already considered in the model.

7.4 Abnormal andAccident Conditions As discussed in Section 2.6, the effects of empty storage cell, increased temperature, misloaded fuel assembly/missing insert, and mislocated fuel assembly accidents on reactivity were evaluated. The results are provided in Table 7.13(a) and Table 7.13(b).

As can be seen, the increased water temperature will not result in an increase in reactivity.

Both misloaded fuel assembly/missing insert and mislocated fuel accidents may result in an increase in reactivity. For the SFP racks, the effect on reactivity of the missing insert is the limiting case.

Thus, its calculated MCNP5-1.51 kac is used for maximum kff calculations for abnormal and accident conditions, discussed in Section 7.5.

The condition with the empty storage cell without insert in the spent fuel rack shows a lower reactivity than a design basis case, therefore, it is acceptable to have the empty storage cell without insert in the spent fuel pool.

7.5 Maximum keff Calculationsfor Abnormal andAccident Conditions As discussed in Section 2.6, the maximum kfr for abnormal and accident conditions is calculated.

The results are tabulated in Table 7.14. The results show that the maximum kff for abnormal and accident conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95%

confidence level.

7.6 Project No. 2127 Report No. HI-2125245 Page 46

7.7 Spent Fuel Rack Interfaces As discussed in Sections 2.8, the interface between SFRs and pool walls, i.e. effect on reactivity of the peripheral fuel assemblies, that have a side non-adjacent to the insert, was evaluated. The results are provided in Table 7.17. As can be seen, this condition will not result in an increase of SFR reactivity. This result is expected because the infinite array design basis model is an infinite array of storage cells with inserts while the full pool model used for these rack interface calculations includes the rack edge along the pool wall where there is no insert along the water gap edge (i.e. no additional cell with an insert). Therefore, this water gap edge allows for neutron leakage and as the calculations show result in statistically equivalent results.

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8. CONCLUSION The criticality analysis for the storage of BWR assemblies in the Quad Cities SFP racks with NETCO-SNAP-IN inserts has been performed. The results for the normal condition show that keff is M with the storage racks fully loaded with fuel of the highest anticipated reactivity, which is SVEA-96 Optima2 Q122 lattice type 146, at a temperature corresponding to the highest reactivity. The results for the accident condition show that keff is lwith the storage racks f loaded with fuel of the highest anticipated reactivity, which is SVEA-96 Optima2 1 i at a temperature corresponding to the highest reactivity. The maximum calculated reactivity for both normal and accident conditions includes a margin for uncertainty in reactivity calculations with a 95% probability at a 95% confidence level. Reactivity effects of abnormal 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.

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9. REFERENCES

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

[2] 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.

[3] "Nuclear Group Computer Code Benchmark Calculations," Holtec Report HI-2104790 Revision 1.

[4] M. Edenius, K. Ekberg, B.H. Forss~n, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1; and J. Rhodes, K Smith, "CASMO-4 A Fuel Assembly Bumup Program User's Manual," SSP-01/400, 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-94/12, Studsvik of America, Inc., (proprietary).

[6] DSS-ISG-2010-01, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, Revision 0.

[7] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.

[8] HI-2002444, Latest Revision, "Final Safety Analysis Report for the HI-STORM 100 Cask System", USNRC Docket 72-1014.

[9] "Sensitivity Studies to Support Criticality Analysis Methodology," HI-2104598 Rev. 1, October 2010.

[10] "Atlas of Neutron Resonances", S.F. Mughabghab, 5th Edition, National Nuclear Data Center, Brookhaven National Laboratory, Upton, USA.

[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 (keff) Predictions, NUREG/CR-7109, April 2012.

[13] OECD / NEA Data Bank, Java-based Nuclear Information Software, Janis version 3.3.

Project No. 2127 Report No. HI-2125245 Page 49

[14] EPRI 1003222, "Poolside Examination Results and Assessment, GEl1 BWR Fuel Exposed to 52 to 65 GWd/MTU at the Limerick 1 and 2 Reactors," December 2002.

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Table 2.1 (a)

Summary of the Area of Applicability of the MCNP5-1.51 Benchmark Table Proprietary Project No. 2127 Report No. HI-2125245 Page 51

Table 2.1 (b)

Analysis of the MCNP5-1.51 calculations [3]

Table Proprietary Project No. 2127 Report No. HI-2125245 Page 52

Table 2.1 (c)

Table Proprietary Project No. 2127 Report No. HI-2125245 Page 53

Table 5.1 a)

Table proprietary.

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Table proprietary.

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Table 5.1 (c)

Table proprietary.

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Table 5. 1(d)

Table proprietary.

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Table proprietary.

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Table 5.2(a)

Reactor Core and Spent Fuel Pool Parameters Table proprietary.

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Table 5.2(b)

Reactor Control Blade Data Table proprietary.

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Table 5.2(c)

Table proprietary.

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Table 5.3(a)

Fuel Rack Parameters and Dimensions Table proprietary.

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Table 5.3(b)

Fuel Rack Insert Parameters and Dimensions Table proprietary.

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Table 5.4(a)

Non-Fuel Material Compositions Table proprietary.

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Table 5.4(b)

Summary of the Fuel and Fission Product Isotopes Used in Calculations Table proprietary.

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Table 7.1 (a)

Table proprietary.

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Table 7.1 (b)

Table proprietary.

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Table 7.1 (c)

Table proprietary.

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Table 7.2(a)

Results of the MCNP5-1.51 Calculations for SVEA-96 Optima2 Q122 Lattices Description Burnup Max delta kaIc Uncertainty (GWd/mtU) kccalc (95/95) 15 0.8599 0.0005 16 0.8660 0.0005 17 0.8702 0.0006 Lattice 146 (reference) 18 0.8740 0.0005 0.8740 Reference Reference 19 0.8733 0.0005 20 0.8733 0.0006 21 0.8678 0.0005 15 0.8502 0.0005 16 0.8572 0.0005 17 0.8614 0.0005 Lattice 147 (void) 18 0.8640 0.0006 0.8660 -0.0081 0.0016 19 0.8660 0.0006 20 0.8628 0.0005 21 0.8587 0.0005 15 0.8493 0.0005 16 0.8554 0.0005 17 0.8615 0.0005 Lattice 147 (water) t 18 0.8643 0.0005 0.8643 -0.0097 0.0016 19 0.8641 0.0006 20 0.8614 0.0005 21 0.8567 0.0005 15 0.8476 0.0005 16 0.8538 0.0005 17 0.8591 0.0005 Lattice 148 18 0.8601 0.0005 0.8619 -0.0121 0.0016 19 0.8619 0.0006 20 0.8600 0.0006 21 0.8553 0.0005 Note 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kca1c may occur at an exposure which differs from that shown above.

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Table 7.2(a) Continued Burnup Max delta kcac Uncert.

Description (GWd/mtU) kcaic sigma dlalac (95/95) 15 0.8389 0.0005 16 0.8461 0.0005 17 0.8519 0.0005 Lattice 149 18 0.8533 0.0005 0.8533 -0.0207 0.0016 (void) 19 0.8517 0.0005 20 0.8487 0.0006 21 0.8421 0.0005 15 0.8439 0.0005 16 0.8516 0.0005 17 0.8551 0.0006 Lattice 149 18 0.8548 0.0005 0.8551 -0.0189 0.0016 (water) t 19 0.8524 0.0005 20 0.8476 0.0006 21 0.8400 0.0005 15 0.8464 0.0005 16 0.8543 0.0005 17 0.8583 0.0005 Lattice 150 18 0.8587 0.0005 0.8587 -0.0154 0.0016 19 0.8568 0.0006 20 0.8525 0.0005 21 0.8459 0.0005 15 0.8553 0.0005 16 0.8611 0.0006 17 0.8629 0.0005 Lattice 151 18 0.8622 0.0005 0.8629 -0.0111 0.0016 19 0.8591 0.0005 20 0.8537 0.0005 21 0.8477 0.0005 Note 2: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kaic may occur at an exposure which differs from that shown above.

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Table 7.2(b)

Results of the MCNP5-1.51 Calculations for GE14 Lattice Type 5 Burnup k Uncert.

Description (GWd/mtU) kaic sigma delta kcaic (95/95)

SVEA-96 Optima2 Q122 15.5 0.8991 0.0006 Reference Reference lattice type 146 Single GE14 13 0.8447 0.0005 -0.0543 0.0016 Single GEl4 13.5 0.8481 0.0005 -0.0509 0.0015 Single GE14 14 0.8500 0.0006 -0.0491 0.0016 Single GE14 14.5 0.8521 0.0005 -0.0469 0.0015 Single GEl4 15 0.8517 0.0005 -0.0473 0.0015 Single GE14 15.5 0.8512 0.0005 -0.0479 0.0015 Single GE14 16 0.8505 0.0005 -0.0485 0.0015 Single GEl4 16.5 0.8508 0.0005 -0.0482 0.0015 Single GE14 17 0.8491 0.0005 -0.0500 0.0015 2x2 GE14 - with channel (cell 14.5 0.8517 0.0005 Reference Reference centered) (Case 2.3.1.5.4.1 )

2x2 GE14 - no channel 14.5 0.8473 0.0006 -0.0044 0.0016 (Case 2.3.1.5.4.2) 2x2 GE14-nchannel/eccentric 14.5 0.8345 0.0005 -0.0173 0.0015 center (Case 2.3.1.5.4.3) 2x2 GE14 - no channel / eccentric 14.5 0.8280 0.0005 -0.0238 0.0015 out (Case 2.3.1.5.4.4)

Note 2: The result of the SVEA-96 Optima2 Q122 lattice type 146 is provided as the reference.

Note 3: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kcalc may occur at an exposure which differs from that shown above.

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Table 7.3 Results of the MCNP5-1.51 Calculations for Design Basis and Simplified Model of SVEA-96 Optima2 Q122 Lattice Type 146 Descrption(GWd/mtU)

Description Burnup Code kcalc sigma Simplified model of SVEA-96 Optima2 Q122 lattice 146 15.5 CASMO-4 0.8890 N/A (Case 2.3.1.4.1)

Simplified model of SVEA-96 Optima2 Q122 lattice 146 15.5 MCNP5-1.51 0.8985 0.0006 (Case 2.3.1.4.2)

Model of SVEA-96 Optima2 Q122 lattice 146, similar to design basist 15.5 MCNP5-1.51 0.8968 0.0005 (Case 2.3.1.4.3)

Note 1: These calculations were performed using the design basis core operating parameters as indicated in Table 5.2(c).

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Table 7.4 Results of the MCNP5-1.51 Calculations for Core Operating Parameters Desciptin Power DesityControl Fuel Moderat Void Burnup dlaUcr Description Density Temp. or Temp. Fraction (GWd/ kcalc sigma dla nc(9t.

_(KBlade ) (°F) (%) mtU) kcalc (95/95)

Design basis 23.688 Yes 1176 547 0 15.5 0.8991 0.0006 Ref. Ref.

(reference)

Fuel temperature 23.688 Yes 588 547 0 16 0.8960 0.0006 -0.0030 0.0017 decreasing Moderator temperature 23.688 Yes 1176 528.8 0 15.5 0.8980 0.0005 -0.0011 0.0017 decreasing Voidfraction 23.688 Yes 1176 547 94 22 0.8859 0.0006 -0.0132 0.0017 increasing Un-rodded 23.688 No 1176 547 0 17 0.8789 10.0005 -0.0202 J0.00 17 operation I2

• 24.1617 Yes 1276 547 0 15.5 0.8987 0.0006 -0.0003 0.0017 II 24.1617 Yes 1376 547 0 15.5 0.8998 0.0006 0.0008 0.0017 I

L _ _ _ __ _ _ _ _ _ _

20.1348 Yes 1176 547 0 15.5 0.8984 10.00051 -0.0007 10.0017 Note 1: The burnup calculations for core operating parameters were performed from 14 GWd/mtU to 24 GWd/mtU. For each core operating parameter, only reactivity of the burnup in this range which results in the largest reactivity is reported.

Note 2: The bounding case is bolded.

Note 3: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kc,,

may occur at an exposure which differs from that shown above.

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Table 7.5 Results of the MCNP5-1.51 Calculations for the Effect of Water Temperature and Density Burnup (Gud/mtU Temp. Water Temperature TAusmnUncert.

Water Density Adjustment, al sim detk,, Unr.

Description (GWd/mtU)S(a) kcac sigma delta kC (95/95)

)(g/cc) (OF)

Reference:

lower bound temperature 15.5 39.2 1 68.81 0.8991 0.0006 Reference Ref.

(Case 2.3.2.1)

Upper bound temperature for normal operation, low 15.5 150 0.98026 68.81 0.8950 0.0005 -0.0041 0.0015 S(a,3)

(Case 2.3.2.2)

Upper bound temperature for normal operation, 15.5 150 0.98026 170.33 0.8924 0.0005 -0.0066 0.0015 high S(a,3)

(Case 2.3.2.3)

Note 1: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kcalc may occur at an exposure which differs from that shown above.

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Table 7.6(a)

Results of the MCNP5-1.51 Calculations for the Depletion Uncertainty Depletion Description kcaic sigma Uncertainty (5%)

Design basis 0.8991 0.0006 Reference Fresh fuel, no Gd 1.0262 0.0006 0.0064 Project No. 2127 Report No. HI-2125245 Page 75

T Table proprietary.

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Table 7.7 Results of the MCNP5-1.51 Calculations for Fuel Tolerances Peak Description Reactivity kcalc sigma delta kcaic Max delta kcaic Burnup (95/95) (95/95)

(GWd/mtU)

Design basis (reference) 15.5 0.8991 0.0006 Reference Reference Max fuel enrichment 16 0.9000 0.0005 0.0026 0.0026 Min fuel enrichment 15.5 0.8965 0.0005 -0.0009 Max Gd loading 16 0.8961 0.0006 -0.0013 Min Gd loading 15.5 0.9012 0.0005 0.0038 Max pellet density 16 0.8974 0.0006 0.0000 0.0012 Min pellet density 15.5 0.8986 0.0006 0.0012 Max pellet OD 15.5 0.8989 0.0006 0.0015 0.0015 Min pellet OD 16 0.8985 0.0005 0.0011 Max clad ID 16 0.8984 0.0005 0.0010 0.0010 16 0.8982 0.0005 0.0008 Min clad ID Max clad OD 15.5 0.8971 0.0005 -0.0002 0.0027 Min clad OD 15.5 0.9001 0.0006 0.0027 Max sub-bundle pitch 15 0.9072 0.0005 0.0098 0.0098 16.5 0.8884 0.0006 -0.0089 Min sub-bundle pitch Max pin pitch 15.5 0.9096 0.0006 0.0122 0.0122 Min pin pitch 15.5 0.8888 0.0006 -0.0086 Max combined water wing canal inner width, channel outer square width, channel comer inner radius 15 0.9005 0.0006 0.0031 and central water canal inner square width 0.0031 Min combined water wing canal inner width, channel outer square width, channel comer inner radius 15.5 0.8963 0.0005 -0.0011 and central water canal inner square width Max combination of channel wall thickness and water cross wall 16 0.8981 0.0006 0.0008 thickness 0.0019 Min combination of channel wall thickness and water cross wall 15.5 0.8993 0.0006 0.0019 thickness Statistical combination of fuel tolerances 0.0171 Note 1: The maximum calculation uncertainty (sigma) used to determine the 95/95 delta kcalc may occur at an exposure which differs from that shown above.

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Table 7.8 Results of the MCNP5-1.51 Calculations for Rack Tolerances Burnup delta k,.Ic Max delta Description (GWd/mtU) kcalc sigma (95/95) kca (95/95)

Design basis 15.5 0.8991 0.0006 Reference Reference (reference)

Max Max cell cell ID pI 15.5 0.8882 0.0006 -0.0093 N/A Max cell pitch Max wall thickness 15.5 0.9000 0.0005 0.0025 Min wall thickness 15.5 0.8984 0.0005 0.0008 0.0025 Max insert width 15.5 0.8970 0.0005 -0.0005 0.0004 Min insert width 15.5 0.8979 0.0006 0.0004 Statistical combination of rack tolerances 0.0026 Project No. 2127 Report No. HI-2125245 Page 78

Table 7.9(a)

Results of the MCNP5-1.51 Calculations for Fuel Radial Positioning in SFP Racks Burnup Unc.

Description (GWd/mtU) kcaic sigma delta k (95/95) 2x2 (Cas reference

...e ) 15.5 0.8990 0.0005 Reference Ref.

(Case 2.3.5.2.1) 2x2 eccentric center 15.5 0.8937 0.0006 -0.0053 0.0015 (Case 2.3.5.2.2) 2x2 eccentric in 15.5 0.8909 0.0005 -0.0081 0.0013 (Case 2.3.5.2.3) 2x2 eccentric out 15.5 0.8943 0.0005 -0.0047 0.0014 (Case 2.3.5.2.4) 2x2 insert/cell center 15.5 0.8992 0.0005 0.0002 0.0013 (Case 2.3.5.2.5) 8x8 Cas reference 2 e .15.5 0.8981 0.0005 Reference Ref.

(Case 2.3.5.2.6) 8x8 eccentric center 15.5 0.8958 0.0005 -0.0023 0.0014 (Case 2.3.5.2.7)

W eccentric in 15.5 0.8901 0.0006 -0.0080 0.0016 (Case 2.3.5.2.8) 8 eccentric out 15.5 0.8946 0.0005 -0.0035 0.0014 (Case 2.3.5.2.9) 8x8 insert/cell center 15.5 0.8997 0.0005 0.0016 0.0014 (Case 2.3.5.2.10) 1lx 1 reference ICxIseference 15.5 0.8991 0.0006 Reference Ref.

(Case 2.3.5.2.11) l x I insert/cell center 15.5 0.8991 0.0006 0.0000 0.0015 (Case 2.3.5.2.12) 1 1 1 Project No. 2127 Report No. HI-2125245 Page 79

Table 7.9(b)

Results of the MCNP5-1.51 Calculations for Fuel Orientation in SFP Racks Burnup Unc.

Description (GWd/mtU) kcaic sigma delta kc (95/95)

Reference (Shown in 15.5 0.8990 0.0005 Reference Ref.

Figure 2.9(a))

Rotated fuel assembly 15.5 0.8982 0.0005 -0.0008 0.0014 (shown in Figure 2.9(b))

Rotated fuel assembly 15.5 0.8983 0.0005 -0.0007 0.0014 (shown in Figure 2.9(c))

Rotated fuel assembly 15.5 0.8977 0.0005 -0.0013 0.0013 (shown in Figure 2.9(d))

Rotated fuel assembly 15.5 0.8983 0.0005 -0.0007 0.0013 (shown in Figure 2.9(e))

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Table 7.10 Table proprietary.

Page 81 Report No. HI-2125245 Project No. 2127

Table 7.11 Maximum keff Calculation for Normal Conditions in SFP Racks Table proprietary.

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Table 7.12(a)

Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of Nominal Values Instead of Using Minimum B4 C Loading and Minimum Insert Thickness on Reactivity Burnup B-10 Areal Description Density kcale sigma delta kcalc (G d/mtU)_(g/cm 2 )

Description Reference (design basis) 15.5 0.0116 0.8991 0.0006 Reference (Case 2.4.1)

Rack with nominal values for B 4 C 15.5 0.0133 0.8888 0.0005 -0.0103 loading and insert thickness (Case 2.4.2)

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Table 7.12(b)

Margin Evaluation Results of the MCNP5-1.51 Calculations to Evaluate the Effect of the Actual Optima2 Q122 Fuel Assembly Desripion Description (GWd/mtU)

Burnup kcalc sigma Max kcaic delta kcalc Optima2 Q122 Lattice 146 (Design 15.5 0.8991 0.0006 Reference Reference basis)

(Case 2.4. 1) 15 0.8869 0.0005 Optima2 Q122 15.5 0.8873 0.0005 0.8873 Lattice 147 16 0.8868 0.0005 16 0.8834 0.0005 Optima2 Q122 16.5 0.8843 0.0005 0.8843 Lattice 148 17 0.8825 0.0006 14 0.8816 0.0005 Optima2 Q122 14.5 0.8825 0.0005 0.8825 Lattice 149 15 0.8804 0.0006 14 0.8859 0.0006 Optima2 Q122 14.5 0.8863 0.0005 0.8863 Lattice 150 15 0.8856 0.0005 14 0.8857 0.0005 Optima2 Q122 14.5 0.8876 0.0006 0.8876 Lattice 151 15 0.8858 0.0005 Optima2 Q122 Peak Fuel Assemblyt Reactivity 0.8925 0.0006 0.8925 -0.0066 Burnups (Case 2.4.3) (bolded) t The top and bottom natural blankets were conservatively neglected and replaced by adjacent lattice.

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Table 7.12(c)

Margin Evaluation Summary of the Margin Evaluation Description Value Insert Composition Margin, from Table 7.12(a) -0.0103 Actual Optima2 Fuel Assembly Margin, from -0.0066 Table 7.12(b)

Calculated Margin -0.0169 Project No. 2127 Report No. HI-2125245 Page 85

Table 7.13(a)

Results of the MCNP5-1.51 Calculations for the Abnonrmal and Accident Conditions on Reactivity of SFP Table proprietary.

Note 1: The bounding accident is bolded.

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Table 7.13(b)

Results of the MCNP5-1.51 Calculations for the Empty Storage Rack Cell without Insert Description Burnup keale sigma delta kA, Uncertainty (G~dmtU)(95/95)

Design basis 15.5 0.8981 0.0005 Reference Reference (8x8 array)

Empty storage cell (cell 15.5 0.8941 0.0006 -0.0041 0.0016 centered)

Empty storage 15.5 0.8900 0.0005 -0.0081 0.0014 cell (eccentric)

Note 1: The design basis fuel assembly (Optima2 Q122 Lattice Type 146) is used for these calculations.

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Table 7.14 Maximum kIff Calculation for Abnormal and Accident Conditions in SFP Racks Table proprietary.

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Table 7.15 Table proprietary.

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Table 7.16 Results of the MCNP5-1.51 Calculations for Axially Infinite Optima2 Q122 Lattices Burnup kcalc kinf Delta-K Uncertainty Description (GWd/mtU) (reference) (infinite)

Optima2 Q122 Lattice 146 15.5 0.8991 0.9003 0.0013 0.0015 (Design basis)

Optima2 Q122 15.5 0.8873 0.8891 0.0018 0.0015 Lattice 147 Optima2 Q122 16.5 0.8843 0.8851 0.0008 0.0014 Lattice 148 Optima2 Q122 14.5 0.8825 0.8837 0.0011 0.0015 Lattice 149 Optima2 Q122 14.5 0.8863 0.8890 0.0027 0.0014 Lattice 150 Optima2 Q122 14.5 0.8876 0.8886 0.0010 0.0015 Lattice 151 Note: The difference between the MCNP models under the "reference" column and the MCNP models under the "infinite" column is described in Section 5.1.

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Table 7.17 Results of the MCNP5-1.51 Calculations for SFR Interface Description (G d/tU kcale sigma delta kI,,, Uncertainty (G~dmtU)(95/95)

Design basis 15.5 0.8991 0.0006 Reference Reference Full SFP (cell 15.5 0.8983 0.0006 -0.0008 0.0016 centered)

Full SFP (eccentric to 15.5 0.8938 0.0005 -0.0053 0.0015 SFP comer)

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Table 7.18 Table proprietary.

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Figure proprietary.

Figure 2.1 2-D representation of the CASMO-4 models of SVEA-96 Optima2 Q122 lattices (a) Lattice type 146; (b) lattice type 147; (c) lattice type 148; (d) lattice type 149; (e) lattice type 150; (f) lattice type 151.

Project No. 2127 Report No. HI-2125245 Page 93

Figure proprietary.

Figure 2.2 2-D representation of the CASMO-4 models of GE14 lattices (a) Lattice type 2; (b) lattice type 3; (c) lattice type 4; (d) lattice type 5.

Project No. 2127 Report No. HI-2125245 Page 94

Figure proprietary Figure 2.3 2-D representation of the CASMO-4 models of GE 8x8 lattices (a) Lattice 854; (b) lattice 855.

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Figure proprietary.

Figure 2.4 2-D representation of the CASMO-4 models of GE 7x7 lattices (a) Lattice type V; (b) lattice type W.

Project No. 2127 Report No. HI-2125245 Page 96

Figure proprietary.

Figure 2.5 2-D representation of the CASMO-4 models of ATRIUM 9B lattices (a) Lattice SPCA9-3.96L-10G6.5; (b) lattice SPCA9-3.96L-lIG6.5; (c) lattice SPCA9-3.96L- 11G5.5.

Project No. 2127 Report No. HI-2125245 Page 97

Figure proprietary Figure 2.6(a) 2-D representation of CASMO-4 SVEA-96 Optima2 Q122 model in the storage rack geometry.

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Figure proprietary.

Figure 2.6(b) 2-D representation of MCNP5-1.51 SVEA-96 Optima2 Q122 model equivalent of the CASMO-4 model.

Project No. 2127 Report No. HI-2125245 Page 99

Figure proprietary.

Figure 2.6(c) 2-D representation of MCNP5-1.51 SVEA-96 Optima2 Q122 model used in criticality calculations.

Project No. 2127 Report No. HI-2125245 Page 100

Figure proprietary.

Figure 2.7 2-D representation of the MCNP5-1.51 SFP racks radial positioning for the 2x2 array models (a) Cell centered positioning; (b) every fuel assembly is positioned toward the center; (c) every fuel assembly is positioned toward the comer where the insert wings connect; (d) every fuel assembly is positioned away from the comer where the insert wings connect; (e) every fuel assembly is middle between insert and cell walls.

Project No. 2127 Report No. HI-2125245 Page 101

Figure proprietary.

Figure 2.8 2-D representation of the MCNP5-1.51 SFP racks radial positioning for the 8x8 array model Every fuel assembly is positioned toward the center. Note that the insert is located in the bottom left comer of every cell in the model.

Project No. 2127 Report No. HI-2125245 Page 102

Figure proprietary.

Figure 2.9 2-D representation of the MCNP5-1.51 fuel orientation in SFP rack cell for the 2x2 array models (a) The reference case; (b) through (e) the other evaluated cases One sub-lattice from each fuel assembly has a different color to show how fuel assemblies are rotated.

Project No. 2127 Report No. HI-2125245 Page 103

Figure proprietary Figure 2.10(a)

A 2-D representation of the MCNP5-1.51 model of misloaded fuel assembly/missing insert (eccentric).

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Figure proprietary.

Figure 2.10(b)

A 2-D representation of the MCNP5-1.51 model of misloaded fuel assembly/missing insert (cell centered).

Project No. 2127 Report No. HI-2125245 Page 105

Figure proprietary.

Figure 2.11 (a)

A 2-D representation of the MCNP5-1.51 model of mislocated fuel assembly in the comer between two racks.

Project No. 2127 Report No. HI-2125245 Page 106

Figure proprietary.

Figure 2.11 (b)

A 2-D representation of the MCNP5-1.51 model of mislocated fuel assembly in the comer between two racks (eccentric).

Project No. 2127 Report No. HI-2125245 Page 107

Figure proprietary.

Figure 2.11 (c)

A 2-D representation of the MCNP5-1.51 model of mislocated fuel assembly in the comer between two racks (cell centered).

Project No. 2127 Report No. HI-2125245 Page 108

Figure proprietary.

Figure 2.12(a)

A 2-D representation of the MCNP5-1.51 model of mislocated fuel assembly adjacent to the platform (eccentric).

Project No. 2127 Report No. HI-2125245 Page 109

Figure proprietary.

Figure 2.12(b)

A 2-D representation of the MCNP5-1.51 model of mislocated fuel assembly adjacent to the platform (cell centered).

Project No. 2127 Report No. HI-2125245 Page 11I0

Figure proprietary.

Figure 2.13 2-D representation of the MCNP5-1.51 SFP racks radial positioning of the GE14 fuel assemblies for the 2x2 array models (a) Cell centered positioning with fuel channel; (b) Cell centered positioning without fuel channel; (c) every fuel assembly without fuel channel is moved toward the center; (d) every fuel assembly without fuel channel is moved away from the comer where the insert wings connect.

Project No. 2127 Report No. HI-2125245 Page I1I1

Figure proprietary.

Figure 5.1 A 2-D representation of the CASMO-4 model of the SVEA-96 Optima2 fuel lattice 146 in the core.

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Figure proprietary.

Figure 5.2(a)

Quad Cities Unit 1 SFP.

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Figure proprietary.

Figure 5.2(b)

Quad Cities Unit 2 SFP.

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Figure proprietary.

Figure 5.3 Insert cross section profile.

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Figure proprietary.

Figure 5.4 A 2-D representation of the MCNP5-1.51 model of the SFP rack cell with insert.

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Figure proprietary.

Project No. 2127 Report No. HI-2125245 Page 117

Figure Proprietary Figure 7.2 Project No. 2127 Report No. HI-2125245 Page 118

Figure Proprietary Figure 7.3 Project No. 2127 Report No. HI-2125245 Page 119

Appendix A Appendix proprietary.

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APPendix B Appendix proprietary.

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Appendix C Appendix proprietary Project No. 2127 Report No. HI-2125245 Page C- I

Supplement 1 Additional Calculations to Support the Revised NETCO-SNAP-IN Rack Insert Design (11 pages including this page)

Project No. 2127 Report No. HI-2125245 Page Sl-1

SI.1 Introduction This Supplement documents the criticality safety evaluation for the storage of spent BWR fuel in the Unit I and Unit 2 spent fuel pools (SFPs) at Quad Cities Station operated by E xelon. The purpose of this analysis is to justify that the specified changes in the NETCO-SNAP-IN rack insert design [S 1.1] are accep table and bounded by th e current analysis, presented in the m ain part of the report.

S1.2 Methodology See Section 2 of the main report and as otherwise discussed below.

S1.3 Acceptance Criteria See Section 3 of the main report.

S1.4 Assumptions See Section 4 of the main report and as otherwise discussed below.

S1.5 Input Data See Section 5 of the main report. The revised dimensions of the NETCO-SNAP-IN rack insert are presented in Table SI-1 and Figure SI-1.

S1.6 Computer Codes See Section 6 of the main report.

S1.7 Analysis The comparison of the revised insert parameters presented in Table S l-1 with the previous insert design in Table 5.3(b) shows that changes are minor and therefore a si gnificant impact on the conclusions made in the m ain part of the rep ort is no t expected. Nevertheless, to verify the negligible or minor impact of the revised insert design on results presente d in the main part of the report additional calculations are presented in this Supplem ent. The additional calculations presented in this Supplement are similar to those in report for the following cases:

" SFP rack tolerances

" Fuel assembly radial positioning in the SFP rack

" Fuel orientation in the SFP rack These cases are selected because the NETCO-SNAP-IN rack insert design change m ay impact the reactivity in the rac k. All o ther calculations from the main report are not affected by the NETCO-SNAP-INO rack insert design change and the re sults of the unaffected calculations are Project No. 2127 Report No. HI-2125245 Page S 1-2

used in this Supple ment where applicable. Th is approach is considered for both norm al and accident conditions.

S 1.7.1 SFP Rack Tolerances As discussed in Section S 1.7, the effect of the manufacturing tolerances on reactivity of the SFP racks with revised inserts was determ ined. The results of these calcula tions are presented in Table S 1-2. The maximum positive delta-k value for each tolerance is statistically combined.

The maximum statistical combination of the SFP rack tolerances is used to determ ine klfe in Table S1-5 and Table S 1-6.

SI.7.2 Fuel Assembly Radial Positioning in the SFP Rack As discussed in Section S1.7, twelve fuel asse mbly radial positioning cases in th e racks were evaluated. The resu Its of these calculations are pres ented in Ta ble S 1-3. For each eccentric position case, the result for similar but cell centered case is considered as a reference. The results show that most cases show a negative re activity effect, however so me delta kcalc values are positive. Therefore, a maximum delta kcalc value is a pplied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Table S 1-5 and Table S 1-6.

S1.7.3 Fuel Orientation in the SFP Rack As discussed in Section S1.7, five fuel assem bly orientation cases in racks were evaluated. The results of these calculations are presented in Table S 1-4. The result for the reference case is also included. The results show that all cases are statis tically equivalent and the reactivity effect of fuel orientation is negligible. Nevertheless, a maximum positive delta k cic value is applied as a bias and the correspondent 95/95 uncertainty is statistically combined with other uncertainties in Table S1-5 and Table S1-6.

S1.7.4 Maximum kff Calculations for Normal Conditions The calculations of the maxi mum kefr for normal conditions are described in Section 2.3.8 of the main part of the report. The results for the revised NETCO-SNAP-IN rack insert design and the results from the main part of the report are tabulated in Table S 1-5. The results show that the maximum k~ff for the normal conditions in the SFP rack s is less than 0.95 at a 95% probability and at a 95 % confidence level for the revised NETCO-SNAP-IN rack insert design and are bounded by the results from the main part of the report.

Project No. 2127 Report No. HI-2125245 Page S 1-3

SI1.7.5 Maximum k~f Calculations for Abnormal and Accident Conditions The calculations of the maximum klff for accident conditions are described in Section 2.6 of the main part of the report. The bounding accident case from the main report is recalculated using the revised NETCO-SNAP-IN' rack insert design. The results for the revised NETCO-SNAP-IN rack insert design and the results from the main part of the report are tabulated in Table S 1-6. The results show that the maximum keff for abnormal and accident conditions in the SFP racks is less than 0.95 at a 95% probability and at a 95% confidence level for the revised NETCO-SNAP-IN rack insert design and are bounded by the results from the main part of the report.

S1.8 References

[S 1.1] Transmittal of Design Information NF 1100434, Revision 1, "Quad Cities SFP Rack Insert Design Information", dated 09/11/2012.

S1.9 Conclusions The criticality analy sis for the storage of BW R assemblies in the Quad Cities SFP racks with revised NETCO-SNAP-INO inserts has been performed. The results show that keff is M with the stora racks full lo aded with fuel of th e highest anticipated reactivity, which is SVEA-96 Optima2 , at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for u ncertainty in reactiv ity calculations with a 95 %

probability at a 95% confidence level. Reactivity effects of abnor mal 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.

The results show that the speci fied changes in the insert de s Prp nropntnlilp nnil liniinApti liv the current analysis. Dresented in the main Dart of the reDort. I Therefore, any insert width dim ension between the value used in the main report including the specified m anufacturing tolerances and the value eva luated in th is Supplement is acceptable.

Project No. 2127 Report No. HI-2125245 Page S1-4

Table Si-i Fuel Rack Insert Revised Dimensions [S 1.1]

Table proprietary.

t For the details of the insert dimensions, see Figure S I-I.

" See Table 5.3(b)

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Table S1-2 Results of the MCNP5 Calculations for Revised Rack Tolerances Revised Reference Burnup delta k Max delta Max delta Description (GWd/mtU) Filename kcalc sigma (95/95) kcalc kcalelt (95/95) (95/95)

Design basis 15.5 opl46-rt2Oll55r 0.8963 0.0005 Reference Reference Reference (reference)

Max cell Max cell ID It 15.5 op146-rt2O2l55r 0.8856 0.0005 -0.0091 0.0000 0.0000 Max cell pitch Max wall thickness 15.5 op146-rt203155r 0.8964 Min wall thickness 15.5 op146-rt204155r 0.0005 0.8958 10.0006 0.0011 0.00111 Max insert width 15.5 op146-rt206155r 0.8964 0.0005 0.0016 0.0030 0.0004 Min insert width 15.5 op146-rt207155r 0.8978 0.0005 0.0030 Statistical combination of rack tolerances 0.0035 0.0026 t See Table 7.8 Project No. 2127 Report No. HI-2125245 Page S1-6

Table S 1-3 Results of the MCNP5-1.51 Calculations for Revised Fuel Radial Positioning in SFP Racks Revised Revised Reference Reference Burnup sigma ded Unc. t Unc.

Description Filename kcalc (GWd/nitU) delta kai, (95/95) delta kal, (95/95) 2x2 (Cas reference e 15.5 2x2dbrot0I55r 0.8954 0.0005 Ref. Ref. Ref.

(Case 2.3.5.2. 1) Ref.

2x2 eccentric center 15.5 2x2ecnt155r 0.8926 0.0006 -0.0028 0.0015 -0.0053 0.0015 (Case 2.3.5.2.2) 2x2 (cseeccentric in 3 ) 15.5 2x2einl55r 0.8900 0.0005 -0.0054 0.0015 -0.0081 (Case 2.3.5.2.3) 0.0013 2 eccentric out 15.5 2x2eoutl55r 0.8940 0.0005 -0.0014 0.0015 -0.0047 0.0014 (Case 2.3.5.2.4) 2x2 insert/cell center 15.5 2x2icntl55r 0.8956 0.0006 0.0001 0.0016 0.0002 0.0013 (Case 2.3.5.2.5) 8 reference 15.5 8x8dbc155r 0.8966 0.0005 Ref. Ref. Ref. Ref.

(Case 2.3.5.2.6) 8x8 eccentric center 15.5 8x8ecntl55r 0.8934 0.0006 -0.0032 0.0015 -0.0023 0.0014 (Case 2.3.5.2.7) 8x8 (ce eccentric 2 in 8 15.5 8x8ein155r 0.8895 0.0006 -0.0071 0.0015 -0.0080 (Case 2.3.5.2.8) 0.0016 8 eccentric out 15.5 8x8eout155r 0.8931 0.0006 -0.0035 0.0016 -0.0035 0.0014 (Case 2.3.5.2.9) 8x8 insert/cell center 15.5 8x8icntl55r 0.8975 0.0005 0.0009 0.0014 0.0016 0.0014 (Case 2.3.5.2.10) lxI reference 15.5 op146- 0.8963 0.0005 Ref. Ref. Ref. Ref.

(Case 2.3.5.2.11) dbcl55r lxi insert/cell center 15.5 lxlicntl55r 0.8967 0.0006 0.0004 0.0015 0.0000 0.0015 (Case 2.3.5.2.12)

ý See Table 7.9(a)

Project No. 2127 Report No. HI-2125245 Page S 1-7

Table S 1-4 Results of the MCNP5-1.51 Calculations for Revised Fuel Orientation in SFP Racks Burnup Revised Revised Reference Reference Description (GWd/mtU) Filename kcalc sigma delta kaic delta kcalt Unc.U 95/95) (95/95)

Reference (Shown in 15.5 2x2dbrot0l55r 0.8954 0.0005 Ref. Ref. Ref. Ref.

Figure 2.9(a))

Rotated fuel assembly 15.5 2x2dbrotl 155r 0.8958 0.0005 0.0004 0.0014 -0.0008 0.0014 (shown in Figure 2.9(b))

Rotated fuel assembly 15.5 2x2dbrot2l55r 0.8965 0.0005 0.0011 0.0015 -0.0007 0.0014 (shown in Figure 2.9(c))

Rotated fuel assembly 15.5 2x2dbrot3l55r 0.8971 0.0005 0.0016 0.0014 -0.0013 0.0013 (shown in Figure 2.9(d))

Rotated fuel assembly 15.5 2x2dbrot4l55r 0.8978 0.0006 0.0024 0.0016 -0.0007 0.0013 (shown in Figure 2.9(e))

t See Table 7.9(b)

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Table S 1-5 Maximum klff Calculation for Normal Conditions in Revised SFP Racks Table proprietary.

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Table S1-6 Maximum kff Calculation for the Bounding Accident Condition in Revised SFP Racks Table proprietary.

Project No. 2127 Report No. HI-2125245 Page SI-10

Figure proprietary.

Figure S1 -1 Insert cross section profile [S 1.1]

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