ML17354A015

Enclosure 2 - NET-28091-003-01NP, Rev. 0, Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit

ML17354A015
Person / Time
Site:	Indian Point
Issue date:	11/28/2017
From:	Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation
References
NL-17-144 NET-28091-003-01NP, Rev. 0
Download: ML17354A015 (247)

Text

ENCLOSURE 2 TO NL-17-144 Curtiss-Wright Nuclear Division, NETCO Report NET-28091-003-01, Revision 0 (Non-Proprietary Version)

Entergy Nuclear Operations , Inc.

Indian Point Unit 2 Docket No. 50-24 7

NET- 28091-003-0lNP, Rev. 0 Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit Prepared by:

Curtiss-Wright Nuclear Division, NETCO 44 Shelter Rock Rd.

Danbury, CT 06810 Prepared for:

Entergy Nuclear Operations - Indian Point Energy Center under Contract No. 10502876 Rev: Date: Prepared By: Reviewed By: Approved By:

n 0

~L.-...,

11/28/ J.? ~ *t'~~ ~

y

- ' 7 , - Y:tlitlt- t *Jk< \

\ .Q 1'r, '- ~

This Page Intentionally Left Blank Table of Contents 1 Introduction ................................................................................................. 1 1.1 Background ............................................................................................................................. 1 1.2 Description of the Analysis ..................................................................................................... 2 1.3 Acceptance Criteria ................................................................................................................ 5 2 Methodology ................................................................................................. 6 2.1 Computer Codes ...................................................................................................................... 7 3 Input Data .................................................................................................. 11 3.1 SFP and Storage Rack Specifications .................................................................................. 11 3.2 Fuel Assembly Designs .......................................................................................................... 14 3.3 Fuel Assembly Insert Designs .............................................................................................. 17 3.4 Plant Operation Data ............................................................................................................ 20 4 Validation ................................................................................................... 23 4.1 U02, Structural Materials, and Absorbers Validation ...................................................... 24 4.2 MOX Validation .................................................................................................................... 25 4.3 Critical Experiments Effect on the Final k9s19s ................................................................... 26 5 Depletion Calculations .............................................................................. 28 5.1 Limiting Depletion Parameters - Temperatures ................................................................ 29 5.1.1 Averaged Assembly Radial Peaking Factor ........................................................... 29 5.1.3 Moderator Temperature .......................................................................................... 31 5.1.4 Fuel Temperature ..................................................................................................... 34 5.1.5 Selection of Bounding Model and Temperatures ................................................... 37 5.2 Limiting Depletion Parameters - Burnable Absorbers ..................................................... 42 5.3 Limiting Depletion Parameters - Soluble Boron ............................................................... 43 5.4 Limiting Depletion Parameters - Specific Power ............................................................... 44 5.5 Limiting Depletion Parameters - Control Rod Operation ................................................ 45 5.6 Depletion Analysis Model ..................................................................................................... 51 NET- 28091-0003-01, Revision 0 iii

5. 7 Special Case Depletions ........................................................................................................ 55 5.8 Reduced Power Operation at End of Life and Fission Gases ........................................... 57 5.9 Production of Atom Density Sets ......................................................................................... 58 5.10 Summary of Limiting Depletion Conditions ....................................................................... 60 6 Rack Model ................................................................................................ 62 6.1 SCALE 2x2 Radial Models ................................................................................................... 62 6.2 Axial Model. ........................................................................................................................... 65 6.2.1 Axial Burn up Distribution ....................................................................................... 65 6.3 Dimensional Changes with Irradiation ............................................................................... 72 6.3.1 Clad Creep ................................................................................................................. 72 6.3.2 Grid Growth .............................................................................................................. 79 6.4 Averaged Assembly Peaking Factor Interpolation ............................................................ 82 6.5 Convergence of the 2x2 Infinite Model Calculations ......................................................... 83 6.6 Full Pool Models .................................................................................................................... 84 6.6.1 Sensitivity of the Full Pool Model to Modeling Assumptions ................................ 87 6.6.2 Convergence of the Full Pool Model... ..................................................................... 88 6.7 Summary of Modeling Assumptions ................................................................................... 93 7 Sensitivity Analysis .................................................................................... 94 7.1 Manufacturing Tolerances ................................................................................................... 94 7.2 Burn up Dependent Biases and Uncertainties ..................................................................... 96 7.3 Eccentricity ............................................................................................................................ 99 7.4 Additional Biases and Uncertainties .................................................................................. 102 7.5 Biases and Uncertainties Rack-up ..................................................................................... 104 7.6 Interface Uncertainty Treatment. ...................................................................................... 107 8 Results ....................................................................................................... I 08 8.1 Temperature Effects ........................................................................................................... 108 8.2 Region 1 Fuel Categories 1 and 2....................................................................................... 109 NET-28091-0003-01, Revision 0 iv

8.3 Region 2 Category 4 Batch Grouping Z - Current and Future Fuel ............................. 111 8.3.1 Curve Fit .................................................................................................................. 113 8.3.2 Confirmation Calculations for Category 4 ........................................................... 114 8.4 Determination of Burnup Requirements for Categories 3 and 5 .................................... 115 8.4.1 Cell Category Layout in Region 2.......................................................................... 115 8.4.2 Additional Burnup Requirements for Fuel Categories 3 and 5 .......................... 116 8.4.3 Confirmation of k9s19s for Full Pool (includes Category 3 and 5) ........................ 120 8.5 Alternate Arrangements for Region 1 ............................................................................... 124 8.6 Calculations for Discharged Fuel (IP2 A-X and IP3 A-AA) ........................................... 127 8.7 Cell Blockers ........................................................................................................................ 135 8.8 Region 2 Checkerboard ...................................................................................................... 136 8.9 Burnup Penalty for Hafnium Flux Suppression Inserts .................................................. 136 8.10 Failed Fuel Containers ........................................................................................................ 136 8.11 Fuel Rod Storage Basket .................................................................................................... 138 8.12 Assemblies with Missing Fuel Rods ................................................................................... 139 8.13 Storage of Miscellaneous Materials ................................................................................... 141 8.14 Borated Conditions ............................................................................................................. 141 8.15 Burnup Penalty for High Soluble Boron Conditions ....................................................... 143 9 Normal Operations and Accident Analysis ........................................... 144 9.1 Normal Operations ............................................................................................................. 145 9.2 Misplaced Assembly ............................................................................................................ 146 9.3 Dropped Assembly .............................................................................................................. 149 9.4 Over Temperature .............................................................................................................. 150 9.5 Multiple Misloads ................................................................................................................ 151 9.6 Boron Dilution Accident ..................................................................................................... 152 9.7 Seismic Event ....................................................................................................................... 153 10 Summary .................................................................................................. 154 NET- 28091-0003-01 , Revision 0 V J

10.1 Review of DSS-ISG-2010-01 ............................................................................................... 154 10.2 Fuel Reactivity Categorization .......................................................................................... 158 10.3 Allowable SFP Cells for Each Fuel Category ................................................................... 160 10.4 Fuel and Operating Requirements .................................................................................... 163 References ....................................................................................................... 166 Appendix A: Validation of SCALE 6.1.2 for Criticality Analysis Using Laboratory Critical Experiments ............................................. A-1 A.1. Overview .................................................................................................. A-1 A.2. U02 Laboratory Critical Experiments ................................................. A-1 A.2.1 Introduction ......................................................................................................................... A-1 A.2.2 Definition of the Range of Parameters to Be Validated ................................................... A-2 A.2.3 Selection of the Fresh U02 Critical Benchmark Experiments ........................................ A-2 A.2.4 Computer Analysis of the U02 Benchmark Critical Experiments ............................... A-10 A.2.5 Statistical Analysis of the Fresh U02 Critical Benchmark Results .............................. A-20 A.2.6 Establishing the Bias and the Uncertainty ...................................................................... A-28 A.2.7 Subcritical Margin ............................................................................................................ A-29 A.2.8 Area of Applicability (Benchmark Applicability) .......................................................... A-29 A.2.9 Summary of U02 Laboratory Critical Experiment Analysis ........................................ A-32 A.3. HTC and MOX Critical Experiments ................................................ A-33 A.3.1 HTC Critical Experiments ............................................................................................... A-33 A.3.2 MOX Critical Experiments .............................................................................................. A-39 A.3.3 Bias and Uncertainty from the MOX/HTC Critical Experiments ............................... A-43 A.4. Temperature Dependent Critical Experiments ................................. A-44 A.5. Summary of Validation Using Laboratory Critical Experiments ... A-48 A.6. Appendix References ............................................................................ A-49 Appendix B: Fuel Categorization for Unit 2 Batches A Through X and Unit 3 A through AA ................................................................... B-1 NET- 28091-0003-01, Revision 0 VI

List of Tables Table 2.1: 185 Isotopes Used in the Analysis ..................................................................................... 8 Table 3.1: Region 1 and 2 Storage Rack Dimensions [8, 9) ........................................................... 14 Table 3.2: Fuel Assembly Dimensions [11, 12) ................................................................................ 17 Table 3.3: Control Rod and Hafnium Rod Descriptions [11) ........................................................ 19 Table 3.4: Pyrex and Wet Annular Burnable Absorber Descriptions [11, 12, 15] ...................... 19 Table 3.5: Key Operating Features by Cycle Used in IP2 ............................................................. 21 Table 3.6: Key Operating Features by Cycle Used in IP3 ............................................................. 22 Table 5.1: Moderator Exit Temperature, Texit, versus Peaking Factor for Batch Groups .......... 31 Table 5.2: Moderator Exit Density versus Peaking Factor for Batch Groups ............................. 32 Table 5.3: Enthalpy Node Factor versus Axial Burnup Shape ..................................................... 33 Table 5.4: Moderator Temperature (K) at each Node versus Burnup Profile ............................ 33 Table 5.5: Fuel Temperature (K) at each Node versus Burnup Profile ....................................... 36 Table 5.6: Fit Coefficients for Top Node Moderator Temperature and Density ......................... 40 Table 5.7: Fit Coefficients for 3rd Node Moderator Temperature and Density ........................... 40 Table 5.8: Burnable Absorbers versus Batch Grouping ................................................................ 43 Table 5.9: Soluble Boron versus Batch Grouping .......................................................................... 43 Table 5.10: Assemblies under D-Bank for the First 21 Cycles of IP2 .......................................... .45 Table 5.11: Effect of Modeling the Bite Position rather than Burnable Absorbers .................... 4 7 Table 5.12: Burn up Penalty for Assem. with Burnable Absorbers followed by Bite D-bank ..... 48 Table 5.13: Assemblies with BA Inserts plus under D-Bank in Non-Bite Cycles ........................ 48 Table 5.14: Assemblies under D-Bank for the first 11 Cycles ofIP3 ............................................ 50 Table 5.15: SCALE/TRITON minus CASM0-5 Ak of Depletion at 100 Hours Cooling ............ 53 Table 5.16: SCALE/TRITON minus CASM0-5 Ak of Depletion at 5 Years Cooling ................ 53 Table 5.17: SCALE/TRITON minus CASM0-5 Ak of Depletion at 15 Years Cooling .............. 54 NET- 28091-0003-01 , Revision 0 VII

Table 5.18: Percent Difference in the Ak of Depletion at 100 Hours Cooling .............................. 54 Table 5.19: Percent Difference in the Ak of Depletion at 5 Years Cooling ................................... 54 Table 5.20: Percent Difference in the Ak of Depletion at 15 Years Cooling ................................. 55 Table 5.21: Special Case Depletion Parameters ............................................................................. 55 Table 5.22: Verification of Cooling Time Model in the Interpolation Program .......................... 60 Table 6.1: Axial Burn up Profile vs. Burnup Bin [27) ..................................................................... 67 Table 6.2: Axial Relative Burnups for Blanketed Discharged Fuel.. ............................................ 70 Table 6.3: Axial Relative Burnups for Batch Z Fuel.. .................................................................... 71 Table 6.4: Calculated k versus Number of Nodes Modeled ........................................................... 72 Table 6.5: Full Pool Model Sensitivity Tests ................................................................................... 88 Table 6.6: kerr Changes With Start Source ...................................................................................... 90 Table 7.1: Tolerance Reactivity Effects ........................................................................................... 94 Table 7.2: Eccentricity Results ....................................................................................................... 101 Table 7.3: Total Bias and Uncertainties for Region 1, Categories 1, 2, 3 ................................... 105 Table 7.4: Sample Category 4 and 5 Bias and Uncertainty Rack-up ......................................... 106 Table 7.5: Total Bias and Uncertainty for Fresh Fuel in Region 2 ............................................. 106 Table 8.1: Calculated kerr as a Function of Temperature ............................................................. 109 Table 8.2: Confirmation of Region 1 Requirements for Category 1 and 2 Fuel.. ...................... 110 Table 8.3: Change in kerr with Burn up and number of IFBA Rods ............................................ 110 Table 8.4: Minimum Burnup Requirements (GWd/T) for Category 4 Batch Grouping Z ...... 111 Table 8.5: Curve Fit Coefficients for Category 4 Fuel.. ............................................................... 113 Table 8.6: Calculated kerr Values at each Category 4 Batch Z Burnup Point ............................ 114 Table 8.7: Total Bias and Uncertainty at each Category 4 Batch Z Burnup Point.. ................. 114 Table 8.8: k9s19s for each Category 4 Batch Z Burnup Point ....................................................... 115 Table 8.9: Region 2 Models at Loading Curve (Cat 5 is Cat 4 plus 11 GWd/T) ........................ 121 NET- 28091-0003-01, Revision 0 Vlll

Table 8.10: Eccentric Options for Region 1 .................................................................................. 122 Table 8.11: Maximum Full Pool k9s19s assuming Various Cycle Lengths ................................... 124 Table 8.12: Dependence of kerr on the Region 1 Arrangement .................................................... 126 Table 8.13: Batch A-D Minimum Burnup Requirements (GWd/T) for Category 4 ................ 128 Table 8.14: Batch E-F Minimum Burn up Requirements (GWd/T) for Category 4 ................. 128 Table 8.15: Batch G-L Minimum Burnup Requirements (GWd/T) for Category 4 ................ 129 Table 8.16: Batch M-P Minimum Burnup Requirements (GWd/T) for Category 4 ................ 129 Table 8.17: Batch Q-S Minimum Burnup Requirements (GWd/T) for Category 4 ................. 130 Table 8.18: Batch T-V Minimum Burnup Requirements (GWd/T) for Category 4 ................. 130 Table 8.19: Batch W Minimum Burnup Requirements (GWd/T) for Category 4 .................... 131 Table 8.20: Batch X Minimum Burnup Requirements (GWd/T) for Category 4 ..................... 131 Table 8.21: Batch A-U (IP3) Minimum Burnup Requirements (GWd/T) for Category 4 ....... 132 Table 8.22: Batch V-X (IP3) Minimum Burnup Requirements (GWd/T) for Category 4 ....... 132 Table 8.23: Individual Assembly Analysis for Category 3 ........................................................... 133 Table 8.24: Individual Assembly Analysis for Category 4 ........................................................... 134 Table 8.25: Failed Fuel Container Pin Analysis ........................................................................... 138 Table 8.26: Normal Operations with Boron Dilution ppm (Full Pool Model) ........................... 142 Table 8.27: Burnup Penalty Results at 1200 ppm ........................................................................ 143 Table 9.1: Misplaced 5.0 w/o 64 IFBA Assemblies with 2000 ppm ............................................. 148 Table 10.1: DSS-ISG-2010-01 Checklist ........................................................................................ 154 Table 10.2: Summary of Loading Requirements for Fuel Batch Z ............................................ 159 Table 10.3: Fuel Design Requirements for Batch Z assemblies .................................................. 164 Table 10.4: Fuel Assembly Operating Requirements ................................................................... 165 Table A.l: Selection Review ofOECD/NEA Criticality Benchmarks ........................................ A-3 Table A.2: Critical Experiment Results with SCALE 6.1.2 and ENDF/B-VII ........................ A-11 Table A.3: Summary of Critical Experiments Containing Boron ............................................ A-18 NET- 28091-0003-01, Revision 0 IX

Table A.4: Wilk-Shapiro Test Results Output From DATAPLOT (4) .................................... A-21 Table A.5: Area of Applicability (Benchmark Applicability) ................................................... A-29 Table A.6: HTC Phase 1 Results .................................................................................................. A-34 Table A.7: HTC Phase 2a, Gadolinium Solutions, Results ........................................................ A-35 Table A.8: HTC Phase 2b, Boron Solutions, Results ................................................................. A-36 Table A.9: HTC Phase 3 Results - Water Reflected Assemblies .............................................. A-37 Table A.10: HTC Phase 4 Results - Steel Reflected Assemblies ............................................... A-38 Table A.11: Results of MOX Critical Benchmarks (SCALE 6.1.2, ENDF/B-VII) .................. A-40 Table A.12: LCT-46 with Full Thermal Expansion ................................................................... A-44 Table A.13: LCT-46 with No Thermal Expansion of Solids ...................................................... A-47 Table B.1: Fuel Assembly Reactivity Categorization for Assembly IDs A - X for Unit 2 ......... B-2 Table B.2: Fuel Assembly Reactivity Categorization for Fuel Assembly IDs A - AA for IP3 .. B-5 NET- 28091-0003-01 , Revision 0 X

List of Figures Figure 1.1: Fuel Category Placement in the IP2 SFP (base case) ................................................... 3 Figure 3.1: IP2 SFP Taken From Holtec Drawing #397 [35) ........................................................ 12 Figure 3.2: Small Section of the Region 1 Rack [8] ........................................................................ 13 Figure 3.3: Region 2 Rack Showing Cell Boxes and Resultant Cells [9] ...................................... 13 Figure 5.1: Averaged Assembly Peaking Factors of Assemblies in the IP2 SFP ......................... 30 Figure 5.2: Fuel Temperature Change with Burnup and Relative Power ................................... 35 Figure 5.3: Fuel Temperature (K) at 25 GWd/T vs. Peaking Factor (PF) at the Top Node ....... 36 Figure 5.4: Top Node Moderator Temp (K) vs. Average Assembly Peaking Factor .................. 39 Figure 5.5: Top Node Moderator Density vs. Avg Assembly Peaking Factor ............................ .40 Figure 6.1: Region 1 KENO Model. ................................................................................................. 63 Figure 6.2: Region 2 KENO Model .................................................................................................. 64 Figure 6.3: Comparison of Creep-down for ZIRLO' and Zircaloy-4 [37] ................................ 73 Figure 6.4: Diameter Decrease versus Exposure Time [39) ........................................................... 74 Figure 6.5: Clad Creep Down for Vandellos 2 Nuclear Power Plant [40) .................................... 75 Figure 6.6: Axial Distribution of the Fuel Rod Diameter at 50.5 GWd/T [41] ............................ 76 Figure 6.7: Oxide Layer thickness with Burnup [42) ..................................................................... 77 Figure 6.8: Density of Fuel Pellet as a Function of Pellet Burnup [43) ......................................... 77 Figure 6.9: ZIRLO' Grid Growth [42) ......................................................................................... 80 Figure 6.10: Zircaloy-4 and MS Grid growth versus burnup [44) ................................................ 80 Figure 6.11: Grid Growth of ZIRLO' and Zircaloy-4 versus Elevation [45) ............................ 81 Figure 6.12: Calculated kerr versus Assembly Average Peaking Factor ........................................ 83 Figure 6.13: Full Pool Model ............................................................................................................ 85 Figure 6.14: Model of Module H Showing Control Rods .............................................................. 86 Figure 6.15: Locations of the Start Sources for the Convergence Tests ....................................... 91 NET- 28091-0003-01, Revision 0 XI

Figure 6.16: Change in Average kerr with Progressing Generations ............................................. 92 Figure 7.1: Category 4 Region 2 with 16 Assemblies Eccentrically Placed ................................ 100 Figure 7.2: Eccentric Model for Category 2 with Central Row Shifted Down .......................... 102 Figure 8.1: Calculated kerr as a Function of Category 5 Burnup Using 5.0 w/o Fuel... .............. 118 Figure 8.2: kerr as a Function of Category 5 Burnup Using 4.2 w/o Enriched Fuel ................... 119 Figure 8.3: Calculated kerr as a Function of Category 3 Burnup Using 5.0 w/o Fuel... .............. 120 Figure 8.4: Refueling Arrangement ............................................................................................... 125 Figure 8.5: No Cat 2 Arrangement ................................................................................................ 125 Figure 8.6: Max Cat 1 Arrangement ............................................................................................. 126 Figure 8.7: Example Odd Arrangement ........................................................................................ 126 Figure 8.8: Cell Blocker Region 2 Model ...................................................................................... 135 Figure 8.9: Failed Fuel Container Pin Model ............................................................................... 137 Figure 8.10: Model for the Fuel Rod Storage Basket.. ................................................................. 138 Figure 8.11: kerr versus Number of Missing Fuel Rods ................................................................ 140 Figure 8.12: Model for Assemblies with 36 Missing Fuel Rods [1] ............................................. 140 Figure 9.1: Misplaced Assembly at the Cask Area Corner ......................................................... 147 Figure 9.2: Misplaced Assembly between the Fuel Elevator and the Rack ............................... 148 Figure 9.3: Full Pool Model with Dropped Assembly .................................................................. 150 Figure 10.1: Fuel Category Location Requirements (Base Case) ............................................... 161 Figure 10.2: Refueling Arrangement ............................................................................................. 162 Figure 10.3: Max Cat 1 Arrangement ........................................................................................... 162 Figure 10.4: Example Odd Arrangement ...................................................................................... 162 Figure A.1: Distribution of the Calculated kerr values Around the Mean ................................. A-22 Figure A.2: kerr as a Function of the Energy of the Average Lethargy Causing Fission ......... A-24 Figure A.3: kerr as a Function of the Pin Diameter ..................................................................... A-25 Figure A.4: kerr as a Function of the Lattice Pitch ...................................................................... A-26 NET- 28091-0003-01, Revision 0 XII

Figure A.5: k.rr as a Function of the Fuel Enrichment ............................................................... A-27 Figure A.6: k.rr as a Function of the Soluble Boron Content.. ................................................... A-28 Figure A.7: k 0 rr as a Function of the EALF for the HTC Experiments ..................................... A-39 Figure A.8: Predicted k.rr as a Function of the Plutonium Content.. ........................................ A-42 Figure A.9: Predicted k.rr as a Function of the Am-241 Content .............................................. A-43 Figure A.10: LCT-046 Corrected Calculated k.rr per Case ....................................................... A-45 Figure A.11: LCT-046 Corrected Calculated k 0 rr Versus Temperature ................................... A-46 NET- 28091-0003-01 , Revision 0 Xlll

1 Introduction This report summarizes the 20 17 criticality safety analysis (CSA) for the Indian Point Unit 2 (IP2) spent fuel pool (SFP) taking no credit for absorber panels. The current 200 l CSA of record [36] takes partial credit for Boraflex TM panels which have degraded and continue to degrade. In order to remove the dependence on the Boraflex TM panels thi s new 20 17 CSA credits empty cells, control rods , and leakage along the outer two storage rows of the SFP. In 2015 , a CSA to remove credit for the Boraflex TM panels which used new metal-matrix-composite absorber inserts was previously submitted to the Nuclear Regulatory Conunission (NRC) and was reviewed [l , 2, and 3]. However, the approach taken in this 2017 CSA is expected to result in a more timely resolution of the Boraflex TM degradation issue. Since the Indian Point Energy Center (IPEC) utilizes the Unit 2 (IP2) SFP for temporary storage of Unit 3 (IP3) fuel prior to placement into dry storage casks, this 2017 CSA allows storage in the IP2 SFP of all fuel assemb lies discharged from both IP2 and IP3.

1.1 Background

The IP2 SFP racks currently credit Boraflex TM as the neutron absorber, which is known to degrade over time. Due to this fact, Entergy (the operator of IPEC) will no longer take credit for the Boraflex TM for reactivity hold-down. In early 2015, IPEC submitted a 20 15 criticality analysis which used neutron absorber inserts to replace the negative reactivity of the Boraflex TM [ 1]. During this process, the NRC requested additional infonnation in June 2015 and IPEC iss ued a response in August 2015 [2]. In November 2015 the NRC issued a staff review of this criticality analysis, concluding that "The NRC staff fi nds that the CSA methodology is acceptable for use at IP2" [3]. Subsequent to this 2015 CSA review, scoping studies determined that it would be more timely and less challenging to refueling outages to load additional fuel assemblies into casks for dry cask storage and use empty cells and control rods for criticality control.

NET- 28091-0003-01, Revi sion 0

IP2 and IP3 are both 4-loop Westinghouse power plants that uti lize the l Sx 15 fuel assembly design.

The physical dimension requirements of the fuel of both units are the same, as both units have had all of their fuel assemblies manufactured by Westinghouse. IP3 does not have the capability to load dry storage casks, so fuel from IP3 is moved to the IP2 SFP for temporary storage prior to loading into dry cask storage. Placement of IP3 fuel is currently restricted to Region 1-2 in the IP2 SFP. This 2017 CSA allows the IP3 fuel to be placed anywhere in the IP2 SFP, so long as it meets the reactivity requirements outlined herein.

1.2 Description of the Analysis This 2017 CSA determines the loading criteria for storage of fuel assemblies in the IP2 SFP by taking credit for empty cell locations, control rods , and the periphery (outer two rows) of the SFP. The analysis does not credit any Boraflex TM neutron absorber that might remain in the racks. Taking credit for empty cells and control rods can accommodate the current and future spent fuel inventory.

The analysis defines five reactivity categories for the fuel and defines storage locations for each reactivity category. The categories are numbered from one to five with a Category 1 fuel assembly being the most reactive and a Category 5 fuel assembly being the least reactive. Similarly, each cell in the SFP has also been assigned a category number with a Category 1 cell being able to accept the most reactive fuel whi le a Category 5 cell can only accept the least reactive fuel. For example, a Category 1 cell can accept all categories of fuel while a Category 5 cell can only accept a Category 5 fuel assembly because all other (lower numbered) categories are more reactive. Figure 1.1 below shows the base case arrangement of the fue l categories in the IP2 SFP. Note that the base case arrangement only shows four reactivity categories since it is the most limiting reactivity arrangement. Category 1 fuel , which is missing on Figure 1.1, is only needed for fresh fuel

• or low burned fuel, which should not be present after IP2 shuts down. Category l fuel will be controlled by two rules for placement in Region 1. The

• Throughout this document, fresh fuel is used to describe fuel that has never been in the core.

NET- 28091-0003-01, Revision 0 2

assemb ly's enrichment, bumup, cooling time, and averaged assembly peaking factor* are used to detennine the reactivity category. For permanently di scharged fuel, full use of operating data is used to precisely obtain the change in reactivity with bumup. For fue l that may still be placed in the core, bounding dep letion va lues are used.

I 2 3

• S ' 7 I 9 10 11 U U 14 15 16 17 11 1' 20 Z1 ll lJ 24 25 2fi 27 21 29 30 31 H 0 ~OP ON G

I--+-+-+--+--+--+--+--+--+-, ~--1--1---,l--of--f--+-+-+-+--+--+--+--+--+-+--+-+-+--l--l--,I 0 OM Ol DI(

OJ D

I--+-+-+--!--+--+--+--+--+---< >-t-t-+--+--+--+--+-+-+--+--+--+--+--+---+---+--+---+--t---t....... OH C

I--+-+-+--+--+--+--+--+--+-, ~--1--1-f--f--l--+-+-+-+--+--+--+--+--+-+-+-+--+--1- DG 8

. Of OE Key: CP CP CN HH-+-+-+--+-+--1-+-+-l CH OwaterHole """"'--fl.......--+-+-+-+--+--+-+-+-1-1 CM HH-+-+-+--+-+--+-+--+-l D S0% Water Hole Cl 1--1,--i--+-+-+-t--+--+-+-+-i CL Cl( CK

• category t Fuel CJ HH-+-+-+--+-+--1-+--+-l CJ HH-+-+-+--+-+-+-+-+-l CH D category 2 Fuel QI CG i--1,--+--+-+-+-t--+-+""+-+-i CG 0 category 3 Fuel a HH-+-+-+--t--+--1-+-+-l Cf as HH-+-+-+-++-+-+-+-l C,

• category4Fuel 1--1,--i--+-+-+-t--+--+--+-+-i co F9'=\=*-"p\,={=4=!"'*'~

ca tegory 5 Fuel 8H 1-HH-+-+-+--++-+-+-' BM category s Fuel with a required full length RCCA

[!) Blot ked Cell 8l BJ Bl BJ 8H 8H BG 8f Bf 8( 8{

BO H-t-+-1-+,..-:~rt-H BC Al AJ AH AG j Cask Area Al AO AC AB

_._~--~-~~

l--l,--l--t--t--t--+--t--t-+-r-lt-11-1--t--t--t--r-t--t--t---t-+-'l'-1

........,....~~~............_.._.........._................_.......

Figure 1.1: Fuel Category Placement in the IP2 SFP (base case)

• The averaged assembly peaking factor is the assembly burn up divided by the sum of the cycle burn ups for the cycles the assembly was in the core. T hroughout this document, this averaged assemb ly pealci ng factor is often abbreviated as pealcing factor or PF.

NET- 28091-0003-01, Revision 0 3

For most permanently discharged fuel from Units 2 and 3, a table of the fuel categorization for each assembly ID is provided. This approach is taken since there are multiple groupings of assemblies (with similar operational characteristics) due to changes in core power (temperatures), burnable absorbers, and axial blanket designs. This CSA provides the discharged assembly categories in Appendix B which will be added to the Technical Specifications. Fuel assemblies that may still be placed in the core are categorized by a set of simple equations to determine the fuel reactivity category. The fuel categories are numbered from most reactive fuel (Category 1 fuel) to least reactive fuel (Category 5 fuel) . The storage locations in the SFP that can accommodate each category of fuel are limited by Figure 1.1 and modifications to this arrangement are specified and analyzed in this CSA.

Details of the analysis methodology are provided in the following sections of this report. Section 2 contains a summary of the methodology. Section 3 provides the input data for the rack designs, fuel assembly designs, fuel assembly inserts, and plant operation data . Section 4 describes the computer code validation and determination of code bias and uncertainty. Section 5 describes the depletion analysis and the selection of bounding input parameters. Section 6 describes the basic fuel rack models, while Section 7 contains the sensitivity analysis of manufacturing tolerances and additional biases and uncertainties. Section 8 contains the results of the reactivity calculations. Section 9 contains a summary of normal operations and the accident analysis. Section 10 provides a summary of the analysis including the limits of the analysis. Appendix A contains the detailed results of the validation of SCALE 6.1.2 and its applicability to this analysis, while Appendix B contains the categorization of fuel assemblies previously discharged from IP2 and IP3.

This new CSA for the IP2 SFP follows the most recent methods . The Nuclear Energy Institute (NEI) has been working with the NRC to produce guidance for SFP analysis [4]. This CSA closely follows the NEI guidance. The NEI guidance started with the NRC draft Interim Staff Guidance (ISG) DSS-ISG-2010-01 [5]. All of the requirements set in DSS-ISG-2010-01 are met and are reviewed in Section 10. 1.

Additional guidance was provided by the Kopp Memo [56] for depletion of atom densities, NUREG/CR-NET- 28091-0003-01, Revision 0 4

7109 [22] for the worth of minor actinides and fission products, and NUREG/CR-6998 [26] for the burnup uncertainty.

1.3 Acceptance Criteria The acceptance criteria of the analysis are to ensure compliance with 10CFRS0.68 [6]. Specifically, the analysis demonstrates that:

• the k9s19s of the SFP is less than 1.0 after accounting for all biases and uncertainties when not taking credit for soluble boron (with a 95% probability at a 95% confidence level) [6], and

• the k9519s of the SFP is less than 0.95 after accounting for all biases and uncertainties when taking credit for soluble boron (with a 95% probability at a 95% confidence level) [6].

In addition to meeting the above criteria, an engineering safety margin is provided to cover unanticipated issues. The engineering safety margin used is 1%, so that the k95195 target value is 0.99 for no soluble boron and 0.94 with soluble boron.

NET- 28091-0003-01, Revision 0 5

2 Methodology The CSA performed in this report uses a method that is comprised of the following steps. Each step refers to a section in this report where further information is provided.

1. Review the current IP2 SFP rack design (Section 3.1) .

2. Review the historical and projected fuel designs and inserts for use in IP2 and IP3. Ensure that the analysis covers all of the designs (Sections 3.2 and 3.3) .

3. Review the historical and projected operating cycles of IP2 and IP3 (Section 3.4) .

4. Validate the computer codes for the application (Section 4).

5. Deplete the fuel using a two-dimensional (2-D) lattice representation of the core using bounding depletion values (including bounding burnable absorbers) for sets of fuel assemblies (down to individual assemblies) (Section 5).

6. Develop a radially infinite three-dimensional (3-D) Monte Carlo model of the Region 1 and Region 2 racks using periodic boundary conditions . The axial modeling height is finite, including conservative modeling of the axial bumup distribution (Sections 6.1 through 6.5) .

7. Develop full pool models to take advantage of leakage at the boundaries of the SFP as well as control rods at specific locations. Use this model in checking interfaces between category cells (Section 6.6).

8. Based on the radially infinite 3-D Monte Carlo model , determine the reactivity effects associated with the manufacturing and fuel tolerances (Section 7.1).

9. Determine the bias and uncertainty associated with bumup (Section 7.2).

10. Determine the bias due to eccentric placement of fuel assemblies in the rack cell (Section 7.3).

11. Ascertain through analysis the most limiting SFP temperature by Region (Section 8.1).

12. Use the radially infinite 3-D Monte Carlo model with the combined biases and uncertainties to detennine the minimum bumup as a function of enrichment, averaged assembly peaking factor, NET- 2809 1-0003-01, Revision 0 6

and cooling time for Categories 1, 2, and 4 at the most limiting SFP temperature. This analysis is perfonned with no soluble boron (Sections 8.2 and 8.3).

13. Determine the fuel Category 3 and 5 additional burnup requirement and test the Region and category cell interfaces using a full pool 3-D Monte Carlo model (Section 8.4) .

14. Perform accident analyses (dropped assembly, misplaced assembly, over temperature (boiling SFP water), boron dilution, seismic, and multiple assembly misloads) with the appropriate models (Section 9) .

15. Summarize the resulting loading requirements and the limits of the analysis (Section 10).

2. 1 Computer Codes This analysis uses the t5-depl TRJTON module of SCALE 6.1.2 [7] for the depletion analysis and the CSAS5 module for the criticality analysis. All of the analyses are perfonned using the 238 group ENDF/B-VII.O library (v7-238). The CSAS5 modul e utilizes CENTRM and BONAM! for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of k. rr*. Unless noted, all of the CSAS5 computer runs use a Monte Carlo sampling of at least 8000 generations and 8000 neutrons per generation to achieve a statistical uncertainty in kerr of less than 0.0001.

The t5 -depl sequence of TRJTON utilizes CENTRM and BONAM I for the resonance treatment and then uses KENO V.a for the collapsing of the cross-sections from 238 groups to one group for use in ORJGEN. The input parameter, parm=(addnux=4), is used in the analysis which tracks the maximum number of problem specific collapsed isotopes (388). At the end of the depletion ana lysis, the OPUS module is used to output atom densities for use in the criticality model. In the OPUS input, 185 isotopes are specified, as shown in Table 2.1. The isotopes that are not included have low atom densities (less than l E-12), combined with small cross-sections, in the spent fuel composition. In other words, the eliminated isotopes do not impact the reactivity of the spent fuel and consequently will not impact the criticality

• Throughout this document, ke rr is used as a short hand notation fork-effective.

NET- 2809 1-0003-01, Revision 0 7

analysis. Immediately after shutdown, there is an increase in reactivity in the first few days due to the decay of Xe-135 and Np-239 (poison is being removed and fissile Pu-239 is being added). Rather than follow this change in reactivity and to assure that the peak reactivity occurs at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> , all of the Xe-135 is converted to Cs-135 and all of the Np-239 is converted to Pu-239. As previously mentioned , atom densities less than lE-12 are eliminated.

Table 2.1: 185 Isotopes Used in the Analysis Isotope Isotope Isotope Isotope Isotope Isotope Isotol!e Ag-109 Cm-243 Gd-160 Nd-145 Rb-85 Sm-153 Te-130 Ag-llOm Cm-244 Ge-73 Nd-146 Rb-86 Sm-154 Te-132 Ag-111 Cm-245 Ge-76 Nd-147 Rb-87 Sn-1 15 U-234 Am-241 Cm-246 Ho-165 Nd-148 Rh-103 Sn-116 U-235 Am-242m Cs-133 I-127 Nd-150 Rh-105 Sn-117 U-236 Am-243 Cs-134 I-129 Np-237 Ru-100 Sn-118 U-237 As-75 Cs-135 I-131 Np-238 Ru-101 Sn-119 U-238 Ba-134 Cs-136 I-135 Np-239 Ru-102 Sn-120 Xe-128 Ba-135 Cs-137 In-115 0-16 Ru-103 Sn-122 Xe-129 Ba-136 Dy-160 Kr-82 Pd-104 Ru-104 Sn-123 Xe-130 Ba-137 Dy-161 Kr-83 Pd- 105 Ru-105 Sn-124 Xe-131 Ba-138 Dy-162 Kr-84 Pd-106 Ru-106 Sn-125 Xe-132 Ba-140 Dy-163 Kr-85 Pd- 107 Ru-99 Sn-126 Xe-133 Br-81 Dy-164 Kr-86 Pd-108 Sb-121 Sr-86 Xe-134 Cd-110 Er-166 La-138 Pd-110 Sb-123 Sr-88 Xe-135 Cd-111 Eu-151 La-139 Pm-147 Sb-124 Sr-89 Xe-136 Cd-112 Eu- 152 La-140 Pm-148 Sb-125 Sr-90 Y-89 Cd-113 Eu-153 Mo-100 Pm-148m Se-76 Tb-159 Y-90 Cd-114 Eu-154 Mo-95 Pm- 149 Se-77 Tb-160 Y-91 Cd-115m Eu-155 Mo-96 Pm- 151 Se-80 Tc-99 Zr-91 Cd-116 Eu-156 Mo-97 Pr-141 Se-82 Te-122 Zr-93 Ce-140 Gd-152 Mo-98 Pr-143 Sm-147 Te-124 Zr-95 Ce-141 Gd-154 Mo-99 Pu-23 8 Sm-148 Te-125 Zr-96 Ce-142 Gd-155 Nb-95 Pu-239 Sm-149 Te-126 Ce-143 Gd-156 Nd-142 Pu-240 Sm-150 Te-127m Ce-144 Gd-157 Nd-143 Pu-241 Sm-151 Te-128 Cm-242 Gd-158 Nd-144 Pu-242 Sm-152 Te-129m NET- 28091-0003-01 , Revision 0 8

In addition to using SCALE, a FORTRAN code (INTRPND) is used to interpolate between bumups from the OPUS output and also to decay the isotopic content to the desired cooling time. The INTRPND code, which has been verified and validated [ 1O] , reads an axial burn up profile to get the shape of the bumup axially, so multiple atom density sets can be made quickly. The code was validated by comparing the keff calculated with the code-interpolated number densities to the ketr calculated with number densities directly from SCALE/TRITON, in which no interpolation is used. Furthermore, SCALE/TRITON is used to decay to a given cooling time and similar comparisons were made. All of the differences between the keff values based upon the interpolated isotopics and the SCALE direct isotopics are within the statistical uncertainty of the keff calculations (see Section 5.9).

The INTRPND FORTRAN program is controlled under NETCO's quality assurance program that meets the requirements of 10CFRSO, Appendix B, 10CFR21 , and ASME NQA-1. The program has been audited by NUPIC. NETCO maintains documented procedures and assigned responsibilities to control the engineering activities relative to the acquisition, classification, development, testing, evaluation, modification, use, maintenance, retirement, and user notification of computer software utilized by NETCO for applicatio ns that are safety related or important to safety. Software is controlled under NETCO Standard Operating Procedures (SOPs) in the Standard Operating Procedures Handbook.

Specifically, SOP 2.4, Software Control, provides procedures for software acquisition, software design, Error Notification, Configuration Control, User Documentation, Verification/Validation, Software Testing/Benchmarking and Run Log maintenance. These features are subject to further procedural control as provided for in NETCO Software Control Procedures, SCP-001, Procedure for Classification of NETCO Software Used for Eng ineering Calculations, and SCP-002, Procedure for NETCO Computer Identification and Installed Software In ventory. Additionally, SOP 2.10, Control of Manual and Computerized Calculations, provides procedures for documentation of the accuracy, traceability and verifiability of computerized calculations.

NET- 28091-0003-01, Revision 0 9

Unless otherwise specified, all of the kerr values reported in this document are raw calculated keff values with no adjustment for bias and uncertainty. The final values to be compared to the criticality criteria are the calculated values plus the total bias and uncertainty (notated as "k9s19s").

NET- 28091-0003-01 , Revision 0 10

3 Input Data For the criticality analysis, input data is needed for the SFP and storage racks (Section 3 .1 ), the fuel assemblies (Section 3.2), the fuel assembly inserts (Section 3.3) and the plant operating data (Section 3.4).

3. 1 SFP and Storage Rack Specifications The IP2 SFP is shown in Figure 3.1. It is lined with a 0.25 inch stainless steel plate covering the concrete walls. In this SFP are three Region 1 (flux trap) modules and nine Region 2 modules. The southwest corner (bottom right in Figure 3.1) is an empty area for placement and loading of storage and transport casks. The southeast corner ( top right in Figure 3 .1) contains two large cylinders for containing failed fuel and the new fuel elevator. The placement of the modules is shown on Figure 3 .1.

The Region 1 fuel racks (flux trap design) contain Boraflex TM in sheaths. The Boraflex' is not credited in this analysis. Figure 3.2 shows the general arrangement of the cells in a Region 1 module [8].

Table 3.1 contains the dimensions and tolerances from the manufacturer' s drawing for Region I [8].

Region 2 is an egg-crate design where square storage tubes with Boraflex TM sheaths are joined at the corners via spacer rods creating "resultant" cells between the tubes . Figure 3.3 shows two complete cells in the Region 2 type rack [9]. The cell on the left wou ld be called the "resultant" cell. Notice the fuel in the resultant cell is not bounded by four flat walls but rather by the Boraflex TM sheaths. The dimensions for the Region 2 rack are also shown in Table 3.1 [9].

Two rack positions will be filled with cell blockers. The cell blockers will be made of stainless steel and will not displace more than 50% of the water at the elevations containing fuel rods .

NET- 2809 1-0003-0l , Revision 0 II

'J'

• ~*

Ce:LL OAT (GIOU 11 CI Q.\C S.

l.!f.91 Ct'. LS P1 CH- I&:* 1, :*

~lGICN 2:

1-,1 RAC- 0.750 0 .650 0 .550 - - - - - ~ - - - - -~-----,

5000 15000 25000 35000 45000 55000 65000 Burnup (MWd/MTU)

Figure 5.1: Averaged Assembly Peaking Factors of Assemblies in the IP2 SFP NET- 28091-0003-01, Revision 0 30

Westinghouse Non-Proprietary Class 3 5.1.3 Moderator Temperature The moderator temperature increases as the water rises through the core. For the top node, using the core exit temperature at the averaged assembly radial peaking factor is conservative (the actual node temperature is somewhat less since the temperature becomes the core exit temperature only at the top of the top node). The average core exit temperature (Tcoreexit) is Tcoreexit= T;n + 2 x (Tave - T;n) = 2 x Tave - T;n where Tave and T;n are the average and inlet temperatures for each cycle provided in Tables 3.5 and 3.6.

The enthalpy at T;n and Tcoreexit can be obtained from steam tables at a pressure of 2235 psia (the minimum allowed pressure). For an averaged assembly radial peaking factor of PF, the delta enthalpy is:

6hassembly = PF X (hcoreexit - h;n)

The assemb ly exit enthalpy is then h 0, + ti.hassembly. The temperature corresponding to this assembly exit enthalpy are obtained from steam tables and converted to Kelvin. The exit temperatures for various peaking factors using the Tave and T;n data from Tables 3.5 and 3.6 are shown in Table 5.1 , below.

Table 5.1: Moderator Exit Temperature, T exit, versus Peaking Factor for Batch Groups Texit at 1.4 Tcxit at 1.2 Tcxit at 1.0 Texit at 0.8 Texit at 0.6 Batch Tave (°F) T;u (°F) (K) (K) (K) (K) (K)

A,B,C,D a,c 601.9 596.2 590.3 584 .1 577.6 E thru L 589.9 584.1 578.1 571.9 565.6 M,N, P 598.8 592.5 585.9 579 .1 572 .0 0 , R, S 598.8 592.5 585 .9 579 .1 572.0 T, U, V 603.2 596.8 590.1 582 .9 575 .5 w,x 603.2 596.8 590.1 582 .9 575.5 Z (after X) 603.5 597.7 591.5 585 .0 578.3 All IP3 - 603.5 597.7 591.5 585.0 578.3 The densities corresponding to these exit temperatures are calculated from steam tables and are presented in Table 5.2:

NET- 2809 1-0003-01 , Revision 0 31

Westinghouse Non-Proprietary Class 3 Table 5.2: Moderator Exit Density versus Peaking Factor for Batch Groups density at density at density at density at density at Batch Tave (°F) Tin (°F) 1.4 fa/cc) 1.2 (e:/cc) 1.0 (e:/cc) 0.8 (e:/cc) 0.6 fa/cc)

A,B,C,D a,c 0.6552 0.67 16 0.6874 0.7024 0.7169 E thru L 0.6884 0.7024 0.7158 0.7288 0.7414 M , N,P 0.6645 0.6816 0.6980 0.71 37 0.7287 Q, R, S 0.6645 0.6816 0.6980 0.7 137 0.7287 T, U, V 0.6510 0.6700 0.6879 0.7050 0.72 14 w,x 0.6510 0.6700 0.6879 0.7050 0.72 14 Z (after X) 0.6502 0.6676 0.6842 0.700 1 0.7154 All IP3 - 0.6502 0.6676 0.6842 0.700 1 0.7154 At the second node from the top, the moderator temperature is lower because of the heat added by the top node. The second node enthalpy is hin + NF*PF*(hcore exit - hin) where hin inlet enthalpy hcore exit exit enthalpy (core average)

PF averaged assembly peaking factor NF enthalpy node factor The enthalpy node factor in the equation above depends on how much heat is added by the top node and this depends on the relative axia l power of the top node. Therefore, an axial power shape is needed.

Since this is a depletion calculation, the axial power shape over the life of the assembly is the axial bumup profile. The DOE axial bumup profiles are used for the axial bumup shape [27]. These profiles are a function of the burnup and are discussed in Section 6. To illustrate an example calculation, the profile from the 46+ GWd/T burnup bin is used. The relative burnups at the top five nodes are 1.072, 1.050, 0.992, 0.833, and 0.515 where 0.515 is the top node, 0.833 is the second node from the top , etc.

The enthalpy node factor for the top of the second node is ( 18 - 0.5 15)/ 18 = 0.971 since there are 18 nodes and the heat added by the top node is 0.5 15/ 18 times the total heat added. Similarly, the node factor for the bottom of the second node is (18 - 0.833 - 0.515)/ 18 = 0.925. The average node factor for the second node is the average of these two values or 0.948.

NET- 28 091-0003-01 , Revision 0 32

The average entha lpy node factor for the third node is (36 - 0.992 - 2x0.833 - 2x0.5 l 5) / 36 = 0.898 .

The exit enthalpies ca n be converted to temperature and density using steam tables . Table 5.3 summarizes the node factors for the top five nodes as a function of the axial bumup shape (represented in the table as the bumup bin).

Table 5.3: Enthalpy Node Factor versus Axial Burnup Shape Node 18-22 22-26 26-30 30-38 38-46 46+

Too 1.000 1.000 1.000 1.000 1.000 1.000 2nd 0.96 1 0.956 0.955 0.955 0.949 0.948 3rd 0.9 16 0.913 0.911 0.907 0.898 0.898 4111 0.86 1 0.860 0.858 0.850 0.841 0.841 5111 0.801 0.800 0.798 0.791 0.781 0.78 2 The exit enthalpy is conservatively used for the top node enthalpy so the enthalpy node factor is always 1.00 and there is no bumup shape dependence. In genera l, the node factor increases with lower relative powers at the top because more of the heat is being added at the lower nodes. From the above node factors and using a peaking factor of 1.40, the moderator temperature (K) for Batch Grouping Z (future fuel) as a function of the bumup shape is shown in Table 5.4.

Table 5.4: Moderator Temperature (K) at each Node versus Burnup Profile Node 18-22 22-26 26-30 30-38 38-46 46+

Too 603.5 603 .5 603.5 603 .5 603.5 603 .5 2nd 60 1.9 601.7 60 1.7 601.7 601.4 601.4 3rd 600. 1 600.0 599 .9 599.7 599.3 599.3 4th 597.8 597.8 597.7 597.4 597.0 597.0 5th 595.3 595 .2 595. 1 594.8 594.4 594.5 NET- 2809 1-0003-01 , Revision 0 33

5.1.4 Fuel Temperature As with the moderator temperature, the fuel temperature in the top nodes is a function of the axial profile at the top. The fuel temperature in the top node is the temperature corresponding to a total peaking factor of PF x AF where PF is the averaged assembly radial peaking factor and AF is the axial peaking factor (for example, AF would be 0.515 using the 46+ bumup bin) . Fuel temperatures were calculated by INTERPIN-3 [28] (provided in Ref. [29]) for various total peaking factors and are shown on Figure 5.2.

The fue l temperature starts at a high value and then decreases for a while before starting to increase again .

Since SCALE allows the input of multiple temperatures for a material during depletion as a function ofbumup, six data sets are input to SCALE (at 0, 2, 15, 25, 40, and 70 GWd!T). For a radial PF of 1.4, the relative power to use for the top node is 1.4 x 0.515 = 0. 721. From Figure 5.2, the top node (relative power of0.721) fuel temperatures at 0, 2, 15, 25 , 40, and 70 GWd!T are 837 .5, 812.9, 789.5, 781.6, 786.8, and 8 15.1 K, respectively. At a PF of 1.2, the top node fuel temperatures are 799.2, 778. 1, 758 .1, 751.2, 752.3 , and 774.9 K. At a PF of 1.0, the top node fuel temperatures are 760.9, 743.4, 726.7, 720.8, 717.8, and 734.8 Kand at a PF of 0.8, the top node fuel temperatures are 722.6, 708 .6, 695 .2, 690.4, 683.3, and 694.7 K.

NET- 2809 1-0003-01 , Revision 0 34

INTERPIN-3 Average Fuel Temperature versus Burnup at Various Relative Powers Typical IP-2 Cycle 20 1300 - - - -

1200

___. ~~

1100 '

.---, ~

~ r-6-,

~~

~

E

, ____. 1,---"' ~

~

~ 1000 a.

E i!!-

, It&.,_~ ,i.- t~

th u.. 900

- p-8-j 3--S--~

a :: , IIt-i:r- ia - -i I}-

800

~.  ;

700 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 Burnup (GWD/MTU) 05 -a- 1 0 Figure 5.2: Fuel Temperature Change with Burn up and Relative Power The fuel temperature at the top node as a function of averaged assembly radial peaking factor is linear and can be expressed as:

Top Node Fuel Temperature at O GWd/T = 19l.5 *PF + 569.4 Top Node Fuel Temperature at 2 GWd/T = l 73 .8*PF + 569.5 Top Node Fuel Temperature at 15 GWd/T = 157.2*PF + 569.5 Top Node Fuel Temperature at 25 GWd/T = 152.0*PF + 568. 8 Top Node Fuel Temperature at 40 GWd/T = l 72.5*PF + 545 .3 Top Node Fuel Temperature at 70 GWd/T = 200.7*PF + 534.2 Figure 5.3 demonstrates that the top node fuel temperature at 25 GWd/T is indeed a linear function of the averaged assembly peaking factor (PF) .

NET- 2809 1-0003-01, Revision 0 35

800 ~ - - - - - - - - - - - - - - - - - - - --

780 + - - - - - - - - - - - - - - - - - --=-----

z'-' 760 + - - - - - - - - - - - - - - = c : : : , ' " " " - - - - -

- Linear

~ 740 + - - ---;:F:-:-:it- - - - - - - - - : ; ; , t l ' I E : - - - - - - - -

~ 720 + - - - - - - - - - - - :..lie : : . _ - - - - - - - - -

a.

[ 700 + - - - - - - - ~ " " ' - - - - - - - - - - - - -

~ 680 + - - - - - : ; ; a , - , , , = : . . . - - - - - - - - - - - - - - -

E-,

- 660 +--:::aa=-- - - - - - - - - - - - - - - - - -

<1,1

~ 640 + - - - - - - - - - - - - - - - - - - - - -

620 +---------------------

600 +-----,-----,------..---...........,r------,

0.5 0.7 0 .9 1.1 1.3 1.5 Top Node Peaking Factor Figure 5.3: Fuel Temperature (K) at 25 GWd/T vs. Peaking Factor (PF) at the Top Node As with the moderator temperature, the fuel temperature in the 2 11ct and lower nodes is a function of the axial burnup profile. Table 5.5 provides the fuel temperatures at 25 GWd/T at a radial PF of 1.40 using the range of DOE axial burnup shapes (represented as the burnup bin) .

Table 5.5: Fuel Temperature (K) at each Node versus Burnup Profile Node 18-22 22-26 26-30 30-38 38-46 46+

Too 723 .0 753 .9 757.3 753.5 78 0.4 781.6 2"d 845 .3 853.5 858.5 878.2 912.0 922 .9 3rd 972.5 941. 8 947.2 992.9 1011. 1002.

4th 1030. 1026. 1029. 1034. 1040. 103 1.

5th 1058. 1060. 1063 . 1046. 1047. 1042.

Note that fo r fuel temperatures , the temperature decreases with lower relative powers at the top which is the opposite of what happens with the moderator temperature.

NET- 28091-0003-01 , Revision 0 36

5.1 .5 Selection of Bounding Model and Temperatures Historically, many criticality analyses used the moderator exit temperature and a single bounding high fuel temperature for all axial nodes. Early analyses non-conservatively ignored radial peaking and used the core exit temperature for the moderator temperature. This non-conservatism is removed by using the averaged assembly radial peaking factor. Using a single set of depletion parameters for all axial nodes is a conservatism that is not needed. The top node is unique since:

• IFBA absorbers are cut short and do not extend into the top node;

• Control rods , if at the bite position, are present in the top node;

• Enrichment is lower for axially blanketed fuel ;

• Fuel temperature is lower due to the sharp decrease in power at the end of the fuel.

For these reasons, the top node is depleted with its own set of depletion parameters. All of the nodes below the top node are depleted with a separate set of temperatures which is conservative for all of the lower nodes . It is possible to use the temperatures in Tables 5.4 and 5.5 and perfonn the depletion analysis using separate depletions for each axial node and burnup bin, but although this approach has been applied in the past [30] , reducing the number of depletion parameters greatly simplifies the analysis.

To simplify the analysis, it is practical to select bounding temperature conditions that cover all burnup bins . Since the fuel and moderator temperature do not change the same amount from burnup bin to burnup bin, a sensitivity study was performed for the fuel and moderator temperatures . It was determined that a change in moderator temperature of 10 K during depletion causes a change in reactivity of 0.0052 t.k while a change in fuel temperature of 100 K causes a change in reactivity of 0.0026 t.k.

Using Tables 5.4 and 5.5 it can be seen that for the top node, the 46+ burnup bin conservatively represents both the moderator and fuel temperatures (the moderator temperature variation is independent of shape but the fuel temperature is highest for the 46+ burnup bin). For the second node, the 46+ burnup bin should also be used because the moderator temperature is a weak function of the shape but the fuel NET- 2809 1-0003-01 , Revision 0 37

temperature is a strong function of the shape. For example, using the 30-38 burnup bin, the moderator temperature at the second node is 0.3 K higher than the 46+ burnup bin but the fuel temperature is 44 .7 K lower. The reactivity effect of the 0.3 K higher moderator temperature is only +0.0002 in keff while the reactivity effect of the 44.7 K lower fuel temperature is -0.0012, so using the 46+ burnup bin for the second node is conservative for all burnup bins.

Since the second node fuel temperature is less than the fuel temperature at the lower nodes, a test is required to determine if the third or second node is more limiting. For the third node, it would appear that using the 38-46 burnup bin is more limiting (the fuel temperature is higher using the 38-46 bin compared to the 46+ bin and the moderator temperature is the same). However, using the 38-46 bin for the top node, the fuel temperature is 1.2 K lower, and for the second node, the fuel temperature is 10.9 K lower, while the 3rd node fuel temperature is 9 K higher. The net effect is that using the 46+ bin is conservative compared to the 38-46 bin when all nodes are considered. However, to check this , a special depletion using the 38-46 shape to obtain fuel and moderator temperatures in the top 5 nodes was perfonned and the result compared to using the 46+ shape for the top and 3'd nodes and then using the 3rd node temperatures for the 2"d, 3'd, 4t\ and 5th nodes (the method that is used for the final selected depletion analysis). The results show that using the 46+ shape to determine fuel and moderator temperatures in this manner is more conservative than using the 38-46 shape by 0.0006 ~k. This difference was obtained by using the burnup shape at 38 GWd/T in the keffca lculation. For additional confirmation, the reactivity effects using three other axial burnup profiles (at bumups of 22, 30, and 45 GWd/T) are 0.0003 ~k, 0.0005 ~k, and 0.0005 ~k, respectively, with the 46+ temperatures always being more conservative.

For the 4th node, using the 3rd node temperatures means that the moderator temperature being used is 2.3 K higher than the 4th node, while the fuel temperature is 29 K lower than the 4th node. The reactivity effect of the 2.3 K higher moderator temperature is +0.0012 ~k while the reactivity effect of the 29 K lower fuel temperature is -0 .0008 ~k. So using the 3'd node temperatures is conservative for the 4th node.

NET- 2809 1-0003-01, Revision 0 38

The same reasoning applies for the 5th and lower nodes. The rationale for this is that the moderator temperature is decreasing faster than the fuel temperature is increasing (in terms of net reactivity), so using the J<d node temperatures is conservative for the 4th and all lower nodes.

For convenience, the moderator temperature at the top node can be conservatively fit with a straight line as illustrated in Figure 5.4 for Batch Grouping Z (future fuel) . The straight line values shown on Figure 5.4 are always the same or conservative with respect to the points.

610 605 z 600 + - - - - - - - - - - - - - - - - :.-,,:.= - - - - -

.._, - Linear Fit

~ 595

= 590

~ + - *- -Points

- - - - - - - :..,.= - - - - - - - - - -

I.

~

c. 585 8

~

E- 580 +---:~"'-- - - - - - - - - - - - - - - -

575 570 + - - - - ~ - - - ~ - - - ~ - - - ~ - - - ~

0.5 0.7 0 .9 1.1 1.3 1.5 Peaking Factor Figure 5.4: Top Node Moderator Temp (K) vs. Average Assembly Peaking Factor Figure 5.5 shows the top node moderator density as a functio n of the averaged assembly peaking factor.

NET- 28091-0003-01, Revision 0 39

0.73 CJ CJ 0.72 oli

'-" 0.7 1

.f' 0.7 C

~ 0.69

~

...0 0.68

....c,:

... 0.67

~ - Linear Fit "O 0.66

~

0

~

"O 0.65 0.64

• Points 0

z 0.63 Q.

0 f,-, 0.5 0.7 0.9 1.1 1.3 1.5 Peaking Factor Figure 5.5: Top Node Moderator Density vs. Avg Assembly Peaking Factor Table 5.6 summarizes the linear fits for the moderator temperature and density for each batch grouping (top node) using the equations:

Exit Temperature (K) =C l + C2 x PF Exit density (glee) = C3 + C4 x PF Table 5.6: Fit Coefficients for Top Node Moderator Temperature and Density Batch Cl C2 C3 C4 A,B,C,D 560.8 29.5 0. 7632 -0.077 1 E thru L 548 .1 30.0 0.78 12 -0.0663 M,N, P 552 .9 33.0 0.7769 -0.0803 O, R, S 552.9 33.0 0.7769 -0.0803 T,U, V 556.6 33 .5 0.7742 -0.0880 w, x 556.6 33 .5 0.7742 -0.0880 Z (after X) 560.5 31.0 0.7643 -0.0815 All IP3 560.5 31.0 0.7643 -0.08 15 Table 5.7 summari zes the linear fits for the 3rd node from the top moderator temperature and density.

Table 5.7: Fit Coefficients for 3rd Node Moderator Temperature and Density (Used for all nodes except the top node)

Batch Cl C2 C3 C4 A,B,C,D 559 .6 27.5 0.7619 -0.0678 NET- 28091-0003-01 , Revision 0 40

E thru L 547.5 27.5 0.7803 -0.0585 M, N, P 552.5 30 .0 0.7755 -0.0705 Q, R, S 552.5 30 .0 0.7755 -0.0705 T, U,V 555.5 31.0 0.7726 -0.0771 W, X 555.5 31.0 0.7726 -0.0771 Z (after X) 559.7 28.5 0.7628 -0.0715 All IP3 559.7 28 .5 0.7628 -0.0715 The fuel temperatures at the 3'd from the top node (used for all nodes except the top node) are:

3rd Node Fuel Temperature at O GW d/T = 351.0*PF + 587 .2 3'd Node Fuel Temperature at 2 GWd/T = 319.5*PF + 584.8 3'd Node Fuel Temperature at 15 GWd/T = 291.5*PF + 580.8 3rd Node Fuel Temperature at 25 GWd/T = 33 l.2*PF + 538 .2 3'd Node Fuel Temperature at 40 GWd/T = 371.0*PF + 514.4 3'd Node Fuel Temperature at 70 GWd/T = 437.8*PF + 493.0 For modeling simplicity, the 3rd node temperatures are used for the 2nd and lower nodes because this is conservative.

For axial blankets, the burnup profile at the top will have relative burnups that are smaller than the DOE profile for full-length fuel. As a result of the smaller relative burnups, the moderator temperature increases slightly but the fuel temperature decrease is more significant. For the same reasons as discussed above regarding using a lower burnup bin, using the DOE profile for the last burnup bin is conservative for axial blanket fuel.

The use of a burnup averaged assembly radial peaking factor assumes that the impact of the temperatures is independent of the power as a function of burnup. In fuel core loading designs, assemblies are depleted with a peaking factor greater than 1.0 during its first cycle (fresh assembl y with burnable absorbers). After the burnable absorber is removed, the assembly is moved and the peaking factor during the second cycle is generally less than 1.0. To show that depletion using the average peaking factor over the life of the assembly is appropriate, a special depletion was performed in which the first 25 GW d/T was depleted at a peaking factor of 1.20 and the second 25 GW d/T was depleted at a peaking factor of0.80. The ke ff for this case at 50 GWd/T (5 .0 w/o fuel) is 0.9371. The keffusing a peaking factor of 1.00 throughout the depletion is 0.9377. This demonstrates that using the average NET- 28091-0003-01, Revision 0 41

peaking factor for the assembly is appropriate and sli ghtl y conservative. As further confirmation, the depletion was repeated using peaking factors of 1.40 and 0.60. The keff for this case is 0.9361, which is even smaller.

5.2 Limiting Depletion Parameters - Burnable Absorbers Burnable absorbers harden the spectrum during depletion, which result in more plutonium production and less U-235 consumption for a given bumup [31 , 32]. The spectrnm hardening comes from the absorption of thermal neutrons by the absorber and displacement of the water in the guide tubes. The effect on reactivity also depends on how long (in terms of GW dff) the burnable absorber remains in the 10 fuel before being removed. Therefore, the burnable absorbers that maximize the B loading and water di splacement should be used. For each batch grouping, the most limiting actual burnable absorbers are used.

IP2 and IP3 have used three types of burnable absorbers: Pyrex, Wet Annular Burnable Absorbers (WABA) and Integral Fuel Burnable Absorbers (IFBA). Table 3.4 provides the dimensions and material details of Pyrex and W ABA inserts. IFBA rods are discussed in Section 3.2.

The Pyrex and W ABA designs consist of rodlets mounted to a base plate which sits on the top of the fuel assembly. The number ofrodlets varies by position in the core to help control power peaking. The most limiting design has 20 rodlets. Table 5.8 provides the worst case burnable absorbers for each batch grouping.

It should be noted that the IFBA and the poison part of the W ABA does not extend to the top of the active fuel. The IFBA starts at least 6 inches from the top (8 inches with 8 inch ax ial blankets) and the poison part of the W ABA starts at least 5 inches fro m the top (6 inches with 6 inch axial blankets and 8 inches with 8 inch axial blankets). For all batch groupings except M - P with IFBA/W ABA, the top node can be depleted with no IFBA and a WABA that has no boron. For M - P, the 8 inch top node is NET- 28091-0003-01 , Revision 0 42

10 depl eted with IFBA that has one quarter B (2 inches of the 8 inches has IFBA) and a W ABA with one 10 half the B (conservatively models the W ABA poison is 4 inches from the top) .

Table 5.8: Burnable Absorbers versus Batch Grouping BA Max BA Max BA Batch Tvoe Loadin2 Burnuo A,B,C,D Pyrex 20 rodl ets 18 .5 E thru F Pyrex 12 rodl ets 12.2 G thru L Pyrex 20 rodl ets 16.7 IFBA 11 6 (I .OX)

M,N, P 28. 1 WABA 20 rodlets IFBA 148 (1.5X)

Q, R, S 26.7 WABA 20 rod lets IFBA 148 (1.5X)

T, U, V 33.8 WABA 20 rodl ets IFBA 148 (1.25X)

W, X 32.6 WABA 20 rod lets Z (after X, and IFBA 148 (1.25X) 33.2 IP3 after U) WABA 20 rodlets IP3 (A-U) Pyrex 20 rod lets 19.4 5.3 Limiting Depletion Parameters - Soluble Boron Soluble boron hardens the neutron spectrum, making the fuel more reactive for a given burnup. It has been shown that performi ng depletion calculations at the burnup averaged soluble boron concentration is acceptable (rather than using a time-dependent soluble boron letdown curve) [57]. Tables 3.5 and 3.6 show the cycle average so luble boron concentration for each cycle. Since nearly every assembly is burned at least two cycles, the soluble boron to use for the depletion analysis is the multi-cycle burnup averaged so luble boron for each assembly. This multi-cycle burnup averaged so luble boron is calculated using the assembly cycle burnups to weight the cycle average ppm. Table 5.9 summarizes the soluble boron used in the depletion analysis for each batch grouping.

Table 5.9: Soluble Boron versus Batch Grouping Boron Used Batch in Depletion A,B,C,D 570 E and F 580 NET- 2809 1-0003-01 , Revision 0 43

G thru L 660 M, N,P 820 Q,R, S 850 T,U, V 880 W, X 880 Z (after X and IP3 after U) 950 IP3 (A-U) 560 5.4 Limiting Depletion Parameters - Specific Power ORNL performed a study of the sensitivity of bumup credit to specific power and determined it is a small effect [33]. For burnup credit using all isotopes, a lower specific power is slightly more reactive.

However, the reactivity effect of moderator temperature and fuel temperature increases with higher specific power. The reactivity effect of higher temperatures is larger than the reactivity effect of a lower specific power. Since the fuel can operate at only one specific power, the specific power used during depletion is determined to match the relative power used for the other depletion parameters. This approach is consistent with DSS-ISG-2010-01 for SFP analysis [5].

The average specific power is the core power divided by the total initial heavy metal mass. Using the stack density (see Section 3.2) of 0.95 multiplied by the U0 2 theoretical density, the initial mass of Uranium metal is then 89.66 metric tons for the Indian Point cores. The specific power (Watts/g U) is then SP= Power x PF x RAB / 89.66 where Power = Total core power (MW) (from Tables 3.5 and 3.6)

PF = Averaged assembly radial peaking factor RAB = Relative Axial Bumup from the DOE shapes For the top node (RAB= 0.515) the specific power ranges from 15.8 to 18.5 W/g U multiplied by the peaking factor for all of the batch groupings (due to changes in the total core power) . Since the specific power has a small effect, the specific power of 16 W/gU multiplied by the averaged assembly radial peaking factor is used for the top node for all fuel batch groupings . The balance of the axial nodes use temperatures developed using the 3rc1 node relative burnup (RAB= 0.992) from the DOE axial burnup profiles . The specific power for the nodes below the top node ranges from 25.6 to 35.6 multiplied by the NET- 28091 -0003-01, Revision 0 44

peaking factor. Again, a simple single specific power value of26 W/gU multiplied by the peaking factor is used for all fuel batch groupings.

5.5 Limiting Depletion Parameters - Control Rod Operation IP2 and IP3 have 193 fuel assemblies in the core. Nine of these assembly locations are under the Control Bank D (less than 5% of the number of assemblies) . Control Bank Dis the only control bank that may be inserted during power operation if the power is greater than 70% of the rated power. IP2 and IP3 have not operated with Control Bank D in the core for any significant bumup except at the "bite" position. The bite position is set as the location where the worth of the lead control bank is 2 pcm per step. The bite position changes from cycle to cycle and during cycle operation but is typically between 207 to 217 steps withdrawn, which corresponds to the rod being inserted 8.7 or less inches into the core.

Operation with Control Bank D at the bite position occurred in IP2 during the first 17 cycles. Cycles 18 and beyond for IP2 and all cycles for IP3 operated with all of the control rods fu lly withdrawn from the core. Table 5.10 shows the fuel assembly IDs that were under control bank D for IP2 for the cycles containing Batches A through X Table 5.10 : Assemblies under D-Bank for the First 21 Cycles of IP2 Core Location Feed B6 BlO F2 F14 HS K2 K14 P6 P lO ofD- Batch Bank Cycle Assembh ID 1 A17 A09 A30 A39 A44 A34 AlO A26 A33 A,B,C 2 BOl B23 B64 B57 B06 B54 B60 B18 B27 D 3 C45 C51 C55 C59 B53 C54 C60 C42 C40 E NET- 28091-0003-01 , Revision 0 45

4 D47 D69 D09 Dl0 D25 D49 D72 D32 D46 F 5 E57 E32 E53 E06 D71 El3 EOl E27 E25 G 6 F06 Fl4 F46 F03 F45 F23 F60 Fl6 F05 H 7 G49 G50 G68 G67 G70 G65 G69 G52 G61 1 8 Jl8 132 122 107 H27 167 130 133 129 K 9 Kl 8 Kl9 K09 Kl2 165 K29 K27 K20 Kl7 L 10 L61 L47 L54 L60 K25 L41 L55 L34 L57 M 11 M54 M63 M69 M60 L09 M65 M62 M53 M46 N 12 P06 P09 P08 P03 MOS P 11 PIO P05 P04 p 13 Q25 Ql7 Q22 Q21 N24 Ql4 Q27 Q26 Ql5 Q 14 R82 R76 R79 R78 R7 1 R75 R77 R72 R 15 S35 S40 S33 S39 S38 S37 S36 S34 s 16 T67 T69 T68 T63 S77 T7 1 T70 T65 T62 T 17 U65 U73 U58 U56 U61 U60 U59 U57 u 18 V32 V51 V46 V52 V44 V30 V49 V31 V 19 W89 W75 W93 W82 W21 W77 W76 W71 W64 w 20 X24 X25 X21 Xl3 W21 X42 X34 X35 X23 X 21 2A64 2A84 2A58 2A55 X03 2A66 2A38 2A70 2A29 2A The reference depletion analyses for all but the future cycles (Batch Grouping Z) are modeled with no control rods inserted. For the assemblies listed in Table 5.10 that are shaded blue or yellow, the burnup requirement for storage is increased by an appropriate burnup penalty. The assembli es marked with green shading on Table 5. 10 did not require a burnup penalty since they did not contain W ABAs and the reactivity effect of WABAs included in the standard depletion analys is is larger than the reactivity effect of the short time that Control Bank D may have been in these assemblies. Note that after Cycle 17, Control Bank D was maintained in an all-out position (no bite).

The assemblies in Table 5. 10 shaded in pink and yellow did not contain burnable absorber inserts during actual core operations but did have the Control Bank D rods in the bite position. A nalysis showed that assemblies depleted with 20 rodlet Pyrex burnable absorbers conservatively bounds assemblies that were operated with Control Bank D at the bite position. No burnup penalty is needed for the pink assemblies since they are modeled with 20 rodlet Pyrex burnable absorbers . The yellow assemblies are depleted with a 20 rodlet WABA. Since W ABA's do not harden the spectrum as much as Pyrex and since W ABA absorber material is not in the top node, a small burnup correction is required. For these NET- 28091-0003-01, Revision 0 46

yellow assemblies , the burnup requirements are increased by 0.5 GWd/T. Table 5.11 shows the cases analyzed to confirm the bumup requirement increase (or shortened to "penalty") for assembl ies where the D-bank was at the bite position and the assembly did not contain a burnable absorber. Note that the L'.k is converted to a L'.GWdfT by use of the sensitivity of ketf to burnup that is discussed in Section 7.

Table 5.11: Effect of Modeling the Bite Position rather than Burnable Absorbers Calculated Calculated kcff A Batch Group, Fuel Burnable kcrr with D- with burnable Ak Burnup Enrichment and Burnup Absorber bank Bite absorber (GWd/T)

G- L, 3.0 w/o, 24.23 GWd/T 20 Pyrex 0.9620 0.9684 -0.0064 -0.90 M- P, 4.2 w/o, 39.28 GWd/T 20WABA 0.9570 0.9554 0.0016 0.30 M-P, 4.6 w/o, 41.31 GWd/T 20WABA 0.9616 0.9608 0.0008 0.20 T - V, 5.0 w/o, 41.94 GWd/T 20 WABA 0.9597 0.9590 0.0007 0.11 The assemblies in the blue shaded portion of Table 5.10 are depleted during the first cycle with a burnable absorber and then contained a control rod inserted to the bite position. For these assemblies the required bumup is increased by 2 GWd/T. The 2 GWdfT penalty is determined by running a separate set of depletions with D-bank at the bite position. The bite depletion is performed as follows :

1. For the top node, the depletion analysis is perfonned with the control rod fully inserted for the entire depletion (this is conservative since this fuel was under D-bank for only one cycle) .

2. For the lower nodes , the burnable absorber is in the fuel until the maximum burnable absorber burnup is reached and then the control rod is placed in the guide tube for 2 GWd/T and then removed. Control rod bumup of 2 GWd/T is considerably more bumup than what has been experienced at IP2.

Calculations of keff using the standard depletions and the bite depletions are performed using the Region 2 three-out-of-four model (see Section 6) . Table 5.12 shows the results of this analysis. The L'.k values shown in the fourth column are converted to delta burnups using the bumup measurement NET- 2809 1-0003-01, Revision 0 47

uncertainty calculations given in Section 7. The maximum delta burnup on Table 5.12 is rounded up to 2 GWd/T which is used for the assemblies in the blue shaded cells on Table 5.10.

Table 5.12 : Burnup Penalty for Assembles with Burnable Absorbers followed by Bite D-bank ABU Batch Group, Fuel Enrichment, and Burnup. k-dbnk k-standard Ak GWd/T A- D, 3.0 w/o, 22.77 GWd/T 0.9721 0.9687 0.0034 0.68 G - L, 3.0 w/o, 24.23 GWd/T 0.9735 0.9684 0.0051 1.02 M-P, 4.2 w/o, 39.28 GWd/T 0.9653 0.9552 0.0101 1.87 M - P, 4.6 w/o, 41.31 GWd/T 0.9708 0.9608 0.0100 1.83 T - V, 4.2 w/o, 36.45 GWd/T 0.9703 0.9621 0.0082 1.64 T - V, 5.0 w/o, 41.94 GWd/T 0.9677 0.9590 0.0087 1.74 The assembly shaded orange (X03 *) represents a separate group. The orange group is for assemblies that contained a burnable absorber insert for their first cycle and then were subsequently placed under D-bank in a later cycle but wi thout "bite" operation . These assemblies may have been burned a short time under D-bank in cycles without bite operation but are modeled as having no D-bank operation. To cover some D-bank operation, 1 GW d/T is added to the burnup requirement of these assemblies. The analysis to support a 1 GWd/T penalty modified the standard depletion analysis by adding 1 and 2 GWd/T ofburnup with control rods to the top node and the lower nodes, respectively, after the burnable absorber is removed. Table 5.13 shows the cases run to confirm this penalty. The results on Table 5.13 are rounded up to arrive at the 1 GWd/T penalty to the burnup requirements.

Table 5.13: Assemblies with BA Inserts plus under D-Bank in Non-Bite Cycles ABU Batch Group, Fuel Enrichment, and Burn up k-dbnk k-standard Ak GWd/T W, 4.6 w/o, 40.07, GWd/T 0.9629 0.9598 0.0031 0.51 X, 4.6 w/o, 40.07 GWd/T 0.9627 0.9597 0.0030 0.49 IP3 (A-U), 3.4 w/o, 29.77 GWd/T 0.9698 0.9666 0.0032 0.54

• X04 was pl aced unde r a D-Bank location in Cyc le 22 and is also inc luded in this group.

NET- 2809 1-0003-01, Revision 0 48

There are two assemblies that are shaded red on Table 5. 10 because they are unique. Assembly R08 was located under D-bank at the bite position for two cycles. This assembly did not contain W ABAs in it. If R08 had been under D-bank for only one cycle then it would also be a " yellow" assembly but it was operated under D-bank for a 2"ct cycle. This assembly has already been casked, and its bumup is

6. 7 GW d/T above the requirements for its assigned category (Category 4 fuel) if it were returned to the SFP . It is concluded that Category 4 is the proper assignment for R08. Assembly U41 is like assembly R08 except that the second cycle of operation for assembly U41 did not have the D-bank at the bite position since bite operation ended with Cycle 17. Thus this assembly would need the yellow bumup penalty of 0 .5 GWd/T plus some additional margin to cover some operational use of D-bank in Cycle 18.

U41 is categorized as Category 5 fuel and exceeds the Category 5 minimum by over 6 GW d/T so the categorization of U4 l is appropriate.

For future cycles (Batch Z) (white cells in Table 5.10) it is not known which assemblies will be under D-bank. To cover power operation with some control rods inserted, the top node for D-bank assemblies is depleted for 1 GWd/T with a control rod and lower nodes are depleted for 2 GWd/T with a control rod.

The rest of the depletion is with a 20 finger W ABA which is never removed plus 148 IFBA ( l .5X). This is the same method that was used in the previous CSA (1]. This approach eliminates the need to check fu ture assemblies for rodded operation under D-bank or when the W ABA was pulled.

IP3 fuel is covered by three depletion batch groupings:

1. Batch Grouping IP3 (A-U) which covers Batches A through U

2. Batch Grouping IP3 (V-X) which covers Batches V , W, and X , and

3. The Z Batch Grouping which covers IP2 and IP3 assemblies beyond X.

NET- 28091-0003-01 , Revision 0 49

The IP3 (A-U) batch grouping is depleted with a 20 rodlet Pyrex burnable absorber which is removed at 20 GWd/f. No control rods are in the base case depleti on. Table 5.1 4 shows the assemblies in IP3 which were under D-bank (assemblies in Batches A through V were not in the IP3 core after Cycle 11 ).

The color coding on Table 5.14 is the same as previously discussed, so there is no burnup penalty for the green shaded assemblies, and the orange shaded assembli es have a burnup requi rement that has been increased by 1 GWd/T.

Table 5.14: Assemblies under D-Bank for the first 11 Cycles ofIP3 Core Feed Location B6 BIO F2 FI4 HS K2 KI4 P6 PIO Batch ofD-Bank Cvcle Assembly ID I A24 A21 AOI A23 A30 Al8 Al7 A06 A02 A,B,C 2 B14 BOS B19 B25 A24 B03 B41 B37 B56 p 3 C45 C64 cso C58 A24 C54 C44 C06 C53 R 4 R32 R53 R71 R25 Pl 1 R06 R43 R47 Rl7 s 5 S04 S3 5 S2 1 S28 R03 S09 Sl3 S l4 S 11 T 6 T74 T41 T70 T46 TSO T61 T76 T64 T65 u 7 U46 U75 U53 U56 TSO U73 U7 1 USO U74 V 8 V38 V37 V40 V35 T53 V34 V36 V33 V39 w 9 W39 W30 WI2 W06 V42 W IS W03 W20 W37 X 10 YI7 Y39 Y28 Y54 V44 Y38 Y40 Y41 Y23 y 11 AA28 AA22 AAS! AA23 U02 AA30 AA24 AA31 AA19 AA NET- 28091-0003-01 , Revision 0 50

Westinghouse Non-Proprietary Class 3 5.6 Depletion Analysis Model The depletion analysis uses the SCALE 6.1.2, t5-depl sequence of TRITON with the ENDF/B-VII.O 238 group cross section library. The model is a simple 2D 15xl5 array of pins centered in an 8.466 x 8.466 inch (assembly pitch in core) square of water (11). The 15x15 array contains 20 guide tubes and a voided instrumentation tube. In the core locations for the incore flux monitoring system, there is a gas-filled tube in which the detectors travel. Rather than model this tube in detail and/or separate out the assemblies that were in instrument locations, the instrument tube of all assemblies are modeled with void inside the tube inner diameter. The 20 guide tubes contain the burnable absorber inserts or control rods as needed for the batch grouping. Some of the fuel rods have ZrB 2 coatings (IFBA rods). This is modeled as a ring next to the pellet outer diameter of 0.001 cm thick of ZrB 2 meeting the 10 B linear density ([ ]a,c mg 10 B/inch except Batch Grouping M- P, which uses [ ]a,c mg 10B/inch) (20).

All of the fuel rods are a single material. Therefore, there is no variation in atom densities across the assembly with depleted fuel. Further, the resonance self-shielding (performed through lattice cell cards in SCALE) is the same for all pins including IFBA pins. A study performed to answer a question from the NRC on the previous CSA showed that ignoring the IFBA in the lattice cell calculation was not significant (2). The depletion problems use 4000 neutrons per generation and 1000 generations. This number of neutrons per generation and number of generations was shown to be adequate by a convergence study detailed in Chapter 4 of reference (2 1].

Small depletion time steps are needed to accurately account for the spectral changes due to Xe and Sm and the initial build in of Pu and other fission products . The initia l time steps (MWd/T) are 150, 350, 500, 500, and 500, followed by steps of 2000 MWd/T until the maximum bumup is achieved. For the Z fuel batch, the depletion contains control rods for the first 2 GWd/T (1 GWd/T in the top node) and then the problem is restarted using the StdCmpMixOOOxx fi le where xx is the material n umber. The fuel temperatures use the "timetable" block to input the bumup dependent temperatures shown below Figure 5.2 and Table 5.7. Due to the restart, the bumups in the timetable block are adjusted for the bumup NET- 28091-0003-01, Revision 0 51

before the restart. For the other fuel groupings, a restart is also used , but to remove the burnable absorber not the control rod. The restart burnup depends on the maximum burnup for any assembly with a burnable absorber insert. This burnup is conservatively rounded up to the next 1 GWd/T .

The depletion block uses the default for the fuel mixture but constant flux option for the burnable absorber materials .

Using the above model inputs, U02 fuel is depleted at fuel enrichments of2.0, 2.2, 2.6, 3.0, 3.4, 3.8, 4.2, 4.6, and 5.0 w/o U-235 at peaking factors of 0.60, 0.80, 1.00, 1.20, and 1.40. Enrichments for axial bl ankets are 2.6, 3.2, 3.4, 3.6, and 4.0 w/o with the same range of peaking factors. The burnup points at which the isotopic data is collected are 0.15, 0.50, 1.0, 1.5, 2.0, and then every 2.0 GW d/T after that.

Although the depletion is carried out with a full set of 388 nuclides, the nuclides used in the SFP model are a reduced set (185 nuclides found on Table 2.1).

In order to confirm the TRITON modeling is adequate, comparisons were made with CASM0-5 benchmarks. The change in kerr as a function of burnup derived from using CASM0-5 and SCALE/TRITON depletion is provided for 198 cases. The CASMO t.k values are published in Reference [20] for specific benchmark conditions.* These include cases with W ABA and IFBA.

Reference [20] is currently under review by the NRC, but the only values used here are the pure CASMO results (not the bias and uncertainty that is under review) .

Tables 5.15, 5.16, and 5.17 provide the difference in the t.k as a function of depletion between CASM0-5 and SCALE/TRITON. Notice that a negative value indicates that SCALE/TRITON conservatively under predicts the reactivity of depletion predicted by CASM0-5. Further, note that the maximum deviation is less than 0.0030. Tables 5. 18 , 5.19, and 5.20, show the percent difference in the

• The CASM0-5 ilk of depletion is not given directly in Reference [20]. Tables C-3 to C-5 of Reference [20]

provide the CASM0-5 ilk of depletion plus the CASMO 5 bias given in Table 10-1 of Reference [20]. For thi s application, the CASM0-5 bias is subtracted from Tables C-3 to C-5 to yield the CASM0-5 ilk of depletion.

NET- 28091-0003-01, Revision 0 52

i:1k of depletion. The max imum percent difference is 1.37%. At these small differences, it is unclear which i:1k of depletion is correct so the util ization of the 5% uncertainty allowed by DSS-ISG-2010-0 1 is appropriate.

In addition, Reference [21] has shown excellent agreement between TRITON/NEWT, which was used for the analysis of chemical assays and the TRITON/KENO approach used in thi s analysis.

Table 5.15: SCALE/TRITON minus CASM0-5 ~k of Depletion at 100 Hours Cooling Case description Case 10 20 30 40 50 60 3.25% enrichment depl eti on 1 -0. 0007 -0.00 12 -0.0015 -0.0025 -0.0022 -0.0020 5.00% enrichment depletion 2 0.0001 0.0002 0.0003 0.0000 0.0004 0.0006 4.25% enrichment depletion 3 0.0004 0.0000 -0 .0002 -0.0005 -0.0008 -0.0005 off-nominal pin depl etion 4 -0.0006 -0.001 2 -0.0015 -0.001 6 -0 .00 19 -0.0022 20 WABA dep leti on 5 0.0002 0.0007 0.0003 -0.0002 -0.000 I -0.0002 I 04 IFBA dep letion 6 0.001 2 0.0011 0.0004 -0.0007 -0.0009 -0.001 8 104 IFBA, 20 WABA dep letion 7 0.0009 0.001 6 0.0008 0.0000 -0 .0001 -0.0008 hi gh boron depletion = 1500 ppm 8 0.0004 -0.0001 -0.0003 -0.0004 -0.0001 -0.0001 branch to hot rack = 338 .7K 9 -0.0002 -0.0003 0.0000 -0.0005 -0.000 I -0.0001 branch to rack boron = 1500 ppm 10 -0.0008 -0.00 15 -0.001 9 -0.0023 -0.0026 -0.0025 high power density depletion 11 0.0000 -0.0007 -0 .0007 -0.0009 -0.0008 -0.0008 Table 5.16: SCALE/TRITON minus CASM0-5 ~k of Depletion at 5 Years Cooling Case description Case 10 20 30 40 50 60 3.25% enrichment depletion I 0.0000 -0.0005 -0.0009 -0.00 14 -0.0008 -0.0004 5.00% enrichment depl etion 2 0.0008 0.00 11 0.0008 0.0008 0.00 10 0.001 3

4. 25% enrichment depletion 3 0.0010 0.0005 0.0003 0.0001 0.0002 0.0003 off- nominal pin depl etion 4 -0.000 1 -0.0003 -0.00 I 0 -0.0011 -0.00 I 0 -0.0009 20 WABA dep letion 5 0.0008 0.0011 0.0009 0.0008 0.0008 0.0010 104 IFBA depletion 6 0.0020 0.0014 0.0009 0.000 1 0.000 1 -0.0005 I 04 IFBA, 20 WABA dep letion 7 0.0020 0.0024 0.0015 0.00 12 0.0006 0.0004 high boron depletion = 1500 ppm 8 0.0007 0.0008 0.0007 0.0003 0.0007 0.0010 branch to hot rack = 33 8. 7K 9 0.0004 0.0006 0.0003 0.0004 0.0009 0.0010 branch to rack boron = 1500 ppm 10 0.0002 -0.0007 -0.00 10 -0.00 17 -0.0015 -0.00 14 high power density depletion 11 0.0006 0.0004 0.0003 0.0004 0.0002 0.0008 NET- 28 091-0003-01 , Revis ion 0 53

Table 5.17: SCALE/TRITON minus CASM0-5 8.k of Depletion at 15 Years Cooling Case description Case 10 20 30 40 50 60 3.25% enrichment dep letion 1 0.0007 -0.0004 -0.00 14 -0.00 16 -0.00 15 -0.0009 5.00% enrichment depletion 2 0.00 11 0.0014 0.0007 0.0008 0.0008 0.0012 4.25% enri chment dep letion 3 0.00 14 0.0010 0.0002 0.0000 0.0000 0.0003 off-nominal pin depletion 4 0.0006 -0.0004 -0 .0006 -0.00 14 -0.00 13 -0.0010 20 W ABA depl etion 5 0.00 14 0.0018 0.0010 0.0005 0.0006 0.0005 104 IFBA depletion 6 0.0025 0.00 19 0.0010 0.0004 -0.0002 -0.0004 104 IFBA, 20 W ABA depletion 7 0.0027 0.0028 0.001 7 0.0010 0.0008 0.0004 high boron depl etion = 1500 ppm 8 0.00 11 0.0009 0.0006 0.000 1 0.0006 0.0006 branc h to hot rack = 338.7K 9 0.0005 0.0004 0.0001 0.0003 0.0006 0.0008 branch to rack boron = 1500 ppm 10 0.0004 -0.0007 -0 .0013 -0.00 16 -0.00 19 -0.00 14 high power density depletion 11 0.00 11 0.0008 0.0001 0.0000 0.0004 0.0005 Table 5.18: Percent Difference in the 8.k of Depletion at 100 Hours Cooling (SCALE/TRITON minus CASM0-5 Llk of Depletion over the Llk of Depletion)

Case description Case 10 20 30 40 50 60 3.25% enrichment deplet ion 1 -0.50 -0.52 -0.48 -0.62 -0.48 -0 .41 5.00% enri chment depletion 2 0.06 0.08 0.10 0.00 0.10 0.12 4.25 % enrichment depletion 3 0.3 1 0.02 -0.06 -0.14 -0.18 -0 .09 off-nominal pin depletion 4 -0.49 -0.54 -0.48 -0.42 -0.40 -0.41 20 WABA depletion 5 0.09 0.31 0.09 -0 .04 -0.02 -0.03 104 IFBA depletion 6 0.71 0.51 0.12 -0 .19 -0.20 -0.36 104 IFBA, 20 WABA depletion 7 0.37 0.65 0.29 -0 .01 -0.01 -0.17 hi gh boron dep letion = 1500 ppm 8 0.36 -0.06 -0.09 -0. 12 -0.02 -0.01 branch to hot rack = 338.7K 9 -0.16 -0 .13 0.00 -0 .13 -0.02 -0.02 branch to rack boron = 1500 ppm 10 -0.79 -0.83 -0.76 -0.72 -0.68 -0.58 hi gh power density dep leti on 11 -0.02 -0 .34 -0.25 -0.25 -0.19 -0.16 Table 5.19: Percent Difference in the 8.k of Depletion at 5 Years Cooling (SCALE/TRITON minus CASM0-5 Llk of Depletion over the Llk of Depletion)

Case description Case 10 20 30 40 50 60 3.25% enrichment depletion 1 -0.01 -0.22 -0.26 -0.32 -0.17 -0.08 5.00% enrichment depletion 2 0.68 0.51 0.28 0.20 0.21 0.25 4.25% enrichment dep letion 3 0.79 0.20 0.10 0.01 0.04 0.05 off-nominal pin depletion 4 -0.09 -0.15 -0.30 -0.25 -0.20 -0.16 20 W ABA depl etion 5 0.38 0.44 0.29 0. 19 0.16 0.18 104 IFBA depl etion 6 1.16 0.61 0.30 0.03 0.01 -0. 10 104 IFBA, 20 WABA dep letion 7 0.78 0.95 0.48 0.30 0.12 0.08 high boron dep letion = 1500 ppm 8 0.56 0.34 0.23 0.06 0.14 0.19 branch to hot rack = 338.7K 9 0.31 0.25 0.10 0.09 0.18 0.18 branch to rack boron = 1500 ppm 10 0.19 -0.40 -0.37 -0.48 -0.37 -0.3 1 high power density depletion 11 0.47 0.16 0.10 0.09 0.04 0. 15 NET- 28091-0003-01 , Revision 0 54

Table 5.20: Percent Difference in the Llk of Depletion at 15 Years Cooling (SCALE(fRITON minus CASM0-5.c.k of Depletion over the t.k of Depletion)

Case description Case 10 20 30 40 50 60 3.25% enrichment depl etion 1 0.49 -0.17 -0. 37 -0.33 -0.28 -0.15 5.00% enrichment depletion 2 0.92 0.62 0.23 0. 18 0.16 0.21 4 .25% enrichment deplet ion 3 1.09 0.40 0.06 -0.0 1 -0.01 0.05 off-nominal pin dep letion 4 0.47 -0.18 -0.17 -0.30 -0.24 -0. 16 20 WABA depl etion 5 0.66 0.69 0.30 0.10 0.1 1 0.08 104 IFBA depletion 6 1.37 0.77 0.28 0.09 -0.03 -0.07 104 IFBA, 20 W ABA depl eti on 7 1.03 1.07 0.52 0.24 0. 17 0.06 high boron depl etion = 1500 ppm 8 0.86 0.37 0.18 0.01 0.11 0.10 branch to hot rack = 338.7K 9 0.38 0.15 0.03 0.06 0. 11 0.1 3 branch to rack boron = 1500 ppm 10 0.38 -0.38 -0.45 -0.42 -0.43 -0.28 high power density depletion 11 0.84 0.32 0.03 -0.01 0.07 0.08

5. 7 Special Case Depletions Due to the limited space in Region 1, assemblies that have been discharged and do not meet the requirements for Region 2 were reviewed to remove conservatisms in the depletion analys is . The largest conservatism is generall y the depletion condition that all assemblies contain the maximum burnable absorbers of the batch grouping. For selected assemblies, analys is is performed using the fu ll available infonnation on the assembly. This section di scusses the change in depletion conditions. Section 8 provides the results of th e analysis for special assemblies. Table 5.21 provides the depletion parameters that are used for special depletion analysis.

Table 5.21: Special Case Depletion Parameters Limiting Fraction of PPM Enrichment Peaking Assembly Fuel Type Theoretical Burnable Absorber (soluble (w/o) Factor ID Density boron)

AI O HIPAR 2.2 1 0.943 None 570 0.92 F44 HIPAR 3.35 0.933 None 540 1.05 L48 LOPAR 3.69 0.944 16WABA 660 0.69 W52 OFA 4.96 0.946 20 W ABA/100 IFBA 880 0.84 XIS OFA 4.95 0.950 20 W ABA/11 6 IFBA 880 0.8 7 U12 (IP3) LOPAR 3.2 1 0.950 12 WABA 560 0.90 V43 (IP3) vs 3.80 0.950 20 W ABA/60 IFBA 650 1.1 2 NET- 28091-0003-01, Revision 0 55

AlO represents 4 other assemblies (A09, A26, A33, and A34) which have the same characteristics but slightly higher burnups. These assemblies were burned for one cycle under the D-bank (therefore, had no burnable absorber). The top node is depleted with a control rod for the entire depletion. The lower nodes were depleted with a control rod for 2 GWd/T. Since the assembly was under D-bank in the bite position, the DOE axial burnup profile between 14 and 18 GWd/T is used.

F44 was in Cycles 4 and 5 and did not contain burnable absorber inserts . With burnup during only two cycles, F44 did not meet the Category 4 fuel requirement by a small amount. When analyzed with its actual burnable absorber (none), it easily made the requirements for Category 4. Of the eight symmetric sisters to F44, six were in the core for three cycles. The remaining sister, F52 has been casked but if it is returned to the SFP, it too meets the Category 4 requirement using this special case analysis.

L48 and its sisters (L37, L38, L39, L44, L51, L52, and L64) spent two cycles on the outside corner of the core. Because of this placement in the core, the burnup after three cycles was too low for Category 4 fuel by about 0. 7 GW d/T if the standard depletion condition for burnable absorbers is used (20 rodlet Pyrex). Since this group of assemblies actually had a 16 rodlet W ABA insert, the analysis of this set of assemblies using the W ABA instead of Pyrex, showed these assemblies meet the Category 4 reactivity requirements.

W52 and X18 are the lowest bumup assemblies of two sets of eight symmetric sister assemblies (W47, WSO, W52, W53, W54, W55, W59, W60 and X09, Xl 1, X12, X14, X16, X18, X44, X45).

Whichever category W52 and X18 qualifies for, then the other seven will also qualify. W52 misses the Category 4 fuel burnup requirement by about 1 GW d/T. X 18 misses the Category 4 requirements by only 0.2 GWd/T. The only benefit in the depletion analysis for these two sets comes from reducing the IFBA rods from 148 to 100 and 116 for the Wand X set respectively. A special depletion is performed for these two sets with the reduced IFBA. In addition to the improved depletion, the actual axial burnup NET- 28091-0003-01, Revision 0 56

profile for these assemblies is used. With these adjustments, these two groups of eight assemblies meet the Category 4 requirements.

Finally, a set of fuel assemblies from IP3 did not meet the Category 4 fuel requirements using the standard depletion analysis . The assemblies Ul2, U2 l, U3 l , and U4 l from IP3 actually contained 12 WABA rather than the 20 Pyrex used in the standard depletion for Batch Group IP3 (A-U) . Modeling these assemblies with the correct burnable absorber inserts allow them to make the Category 4 fuel requirements. V43 and V48 did not initially meet the Category 3 fuel requirement but with special depletions, these two assemblies qualify for Category 3.

5.8 Reduced Power Operation at End of Life and Fission Gases DeHart [34] demonstrated that operating history has a small effect on spent fuel reactivity. However, at the end-of-cycle (EOC), the reactor power may be reduced (for example, a planned coast down) and this can cause a small reactivity change. One of the key absorbing fission products, Sm-149, reaches an equilibrium concentration during power operation that is independent of power. However, its precursor, Pm-149, is directly proportional to power. At a reduced power, there is less Pm-149. Pm-149 decays into Sm-149 with a 2.2 day half-life. Thus, if a reactor reduces power at end of cycle, there would be less Sm-149 in the cooled fuel, which is a positive reactivity effect. Therefore, ignoring low power operation during the last month is non-conservative. To account for this effect, the amount of Pm-149 can be reduced to one half of the full power content for all criticality calculations (which results in a penalty of about 100 pcm). This covers coast downs to 50% power and covers all past operating experience and anticipated future operation at the Indian Point plants.

Furthermore, a significant fraction of once or twice burned assemblies are placed on the core periphery in the last cycle of the assembly's depletion. So a high peaking factor does not reflect a possible very low peaking factor (0.5) at the end of depletion. As discussed above, it is at the end of depletion that the amount of Pm-149 is important. To account for both the "end of depletion" effect (0.5)

NET- 28091-0003-01, Revision 0 57

and the coast down effect (0.5), the amount of Pm-149 is reduced to only 0.5 x 0.5 = 0.25 of the full power isotopic content in all criticality calculations. This results in a reactivity penalty of about 250 pcm in the final kerr calculations.

To conservatively account for fission gases escaping the fuel and migrating to the plenum, all krypton and xenon isotopes are reduced by 32%, all rubidium isotopes are reduced by 44%, and all iodine and bromine isotopes are reduced by 10%. These adjustment factors are justified in the response to an RAI documented in Reference [2].

5.9 Production of Atom Density Sets SCALE TRITON outputs atom densities in two ways, through an output file, StdCmpMixOOOXX where XX is the material number, or OPUS pit files. The StdCmpMixOOOXX supplies the atom densities for the end of the run whereas the OPUS pit files provides atom densities for selected isotopes as a function ofburnup and coo ling time. Rather than rerun SCALE for each desired burnup and cooling time, OPUS pit files are saved. For burnups between the SCALE time steps, the atom densities are linearly interpolated between the time steps. The SCALE time decayed atom densities are on ly valid for the last time step so in order to get the atom densities after cooling, the atom densities are decayed outside of SCALE. The cooling time decay, burnup interpolation , Pm-149 correction, and the fission gas corrections are perfonned using a small FORTRAN code developed by this criticality team called INTRPND.

With isotopics from the depletion calculations recorded every 2 GWdff (see Section 5.6), the isotopics at any particular burnup can be interpolated. Since the burnup delta is small between burnup points, linear interpolation can be used. To validate this approach, the isotopics at 40 GWd/T were interpolated from the OPUS pit files at 38 GWd/T and 42 GWdff at 5.0 w/o enrichment. Using the interpolated isotopics the kerr for the previous CSA Region 2 model was 0.96653 +/- 0.00015 . Taking the isotopics directly from the OPUS plot files at 40 GW d/T, the calculated kerr was 0.96651 +/- 0.000 15. The difference is well within the Monte Carlo statistics . A similar verification was perfonned at 11 GWd/T, NET- 2809 1-0003-01 , Revision 0 58

where an interpolation between 9 and 13 was compared to the direct calculation at 11 GW d/T. The calculated keff using the direct isotopics was 1.1366 +/- 0.0002, while the interpolated case was 1.1362 +/- 0.0002, a difference of only 0.0004, which is within the expected Monte Carlo variation.

Each isotope is decayed using decay constants from the CRC Handbook 85th Edition [58]. Each isotope is decayed into its daughter product which also may be radioactive (the decay is e-I i where 11. is the decay constant) . To ensure that the correct isotopics are obtained, the decay time desired is divided into 10 sub-intervals. The first nine sub-intervals are decayed and the tenth interval is then divided into 100 sub-intervals . The first 99 of these sub-intervals are decayed and the last sub-interval is again divided into 100 sub-intervals. This corrects for the fact that some nuclides are decaying into something else that is also radioactive. If the decay time is not divided into these very fine sub-intervals, the final concentration at the end will not be correct. It was found that this division of the decay time is fine enough such that any finer division resulted in no discernable difference in the concentrations . The correct concentration for all nuclides at the end of the decay time is thus obtained.

To check the cooling time model used in the interpolation program, a special depletion was performed at 5.0 w/o to a bumup of 40 GWd/T. Then SCALE was used to decay the isotopes for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, 1 year, 5 years, and 25 years. The interpolation program was also used to decay the isotopes to the same cooling times . Table 5.22 shows the results of the verification of the cooling time. The differences are within the Monte Carlo statistics (2 sigma of+/- 0.0004) except for the case at 25 years. The calculated keff from the interpolation program at 25 years is conservative as it produces a higher keff-NET- 28091-0003-01 , Revision 0 59

Table 5.22: Verification of Cooling Time Model in the Interpolation Program Interpolation Cooling Time SCALE/ORIGEN k Program k Difference 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1.0023 1.0023 -

1 year 0.9993 0.9996 0.0003 5 years 0.9847 0.9848 0.0001 25 years 0.9449 0.9457 0.0008 5.10 Summary of Limiting Depletion Conditions This section has provided the details on the limiting depletion conditions . The depletion analysis is performed for 11 batch groupings and six sets of special assemblies. For the temperature calculations, several batch groupings were combined, so there are only eight sets of temperatures . For each batch grouping, the depletion analysis is perfonned with five different temperature sets for the fuel and moderator that correspond to the burnup averaged assembly peaking factors of 0.60, 0.80, 1.00, 1.20, and 1.40. For each batch grouping and peaking factor, a depletion analysis is performed for limiting top node conditions and conditions limiting for all nodes below the top node. The depletion analysis is performed over a range of enrichments appropriate for each batch grouping. Atom densities were generated at each 2 GWd/T bumup step. Atom densities for burnups in between these points were determined by interpolation using the small FORTRAN code, INTRPND, described in Section 5.9. The INTRPND code decayed the isotopic concentrations to any cooling time. Finally, the INTRPND code corrected the Pm-149 atom densities for low power operation and corrected the fission gas fractions for their release rates for long-term storage of fuel.

The limiting parameters for each batch grouping are found using:

1. Tables 5.6 and 5.7 for moderator temperature and density,

2. Equations under Figure 5.2 and under Table 5.7 for the fuel temperature, NET- 28091-0003-01 , Revision 0 60

Westinghouse Non-Proprietary Class 3

3. Table 5.8 for the burnable absorber design and bumup at which the burnable absorber is removed (except for Batch Grouping Z for which the burnable absorber is never removed),

and

4. Table 5.9 for the soluble boron concentration .

Specific powers of 16 and 26 W/g multiplied by the peaking factor are used for the top and lower nodes respectively for all groupings . 95% of the U02 theoretical density is used for the stack density except for the special cases given on Table 5.21.

For future fuel (Batch Grouping Z), the fuel assembly is depleted with a contro l rod inserted for 2 GW d/T. Then the assembly is depleted with a 20 rodlet W ABA which is never removed. The initial control rod depletion is to cover future extended part power operation with control rods inserted. The bumup of2 GWd/T burnup requires operation of approximately 1.4 effective fu ll power months . In addition to the WABA, the fuel is modeled as contai ning 148 IFBA pins in all but the top node. The IFBA 10B loading is l .5X [( mg 10B/inch)] a.c to cover future designs (residua l poison from IFBA is not credited in the criticality model).

For all other batch groupings the control rods are not included in the standard depletion. However assemblies under D-bank are identified and the burnup requirements for these assemblies are increased as specified in Section 5.5 .

NET- 28091-0003-0 l, Revision 0 61

6 Rack Model This section describes the Keno models used in the analysis . Two-by-two (2x2) storage cell array models are used for the analysis of the three out of four areas in Region 1 (Category 2 cells) and Region 2 (Category 4 cells). A 2x2 model is also used to analyze the checker board arrangements in Region 1 and Region 2 (Category 1 cells). A full pool model is created to confirm the burnup requirements for the Category 3 and Category 5 fuel assemblies, the category cell interfaces, and to perform analyses of the Misplaced Assembly and multiple misload accidents.

6. 1 SCALE 2x2 Radial Models The rack and fuel dimensions are given in Section 3. The nominal dimensions are used in the models with the exceptions mentioned in this section.

The Boraflex TM is modeled as water. As pointed out in Section 3, if any Boraflex TM remains it would still have some 10B so modeling it as water is conservative. The BoraflexTM sheathing is a plate with the outside edge bent down at a 45 degree angle creating a 0.112-inch (Region 1) or 0.092-inch (Region 2) pocket for the Boraflex TM sheet (8,9]. The SCALE model preserves the minimum sheet material for Region 2 (which is less than Region 1) by modeling the sheathing as a squared off box with a width of 7.70 inches (8 ,9]. The same sheathing width is used for both regions . In Section 7 it is shown that using the minimum sheath material is conservative but not significant.

The connecting steel between Region I cells is modeled as an extension of the cell wall rather than a separate piece of steel. The connecting steel is slightly thicker (0.09375 inch) than the cell wall thickness (0.075 inch) [8]. This model was confirmed by accurately modeling the connector steel in a Region 1 checkerboard model of water holes and fresh assemblies of 5.0 w/o enrichment with 64 IFBA rods. The difference calculated is 0.0005 L\k in the conservative direction (the Monte Carlo uncertainty for these nms is only 0.00006 L\k).

NET- 2809 1-0003-01 , Revision 0 62

The Region 1 rack modules are separated by 1.625 inches and the Boraflex TM sheathing on the outside of the module is 0.075 inches thick rather than the nominal 0.0235 inches [8,35]. The normal flux trap is 1.351 inches in the East-West (vertical) direction and 1.571 inches in the North-South (horizontal) direction [8]. The greater separation between cells at the module interfaces and thicker sheathing assures that the infinite model is conservative for the finite racks consisting of three modules. The Region 2 rack modules are separated by at least 1.25 inches [35]. The resultant cells on the outside row of the module have a 0.075 inch wall [9]. Even without the module separation, the infinite model is conservative since, the Region 2 module interfaces possess an additiona l steel plate.

Since Region 2 is an arrangement of cell boxes with resultant cells, a 2x2 model with a periodic boundary condition is required. Since Region l storage patterns include a checkerboard and a 3-out-of-4 arrangement, a 2x2 model is also required for Region 1. Figures 6. 1 and 6.2 are color plots from the KENO models for Region 1 and 2, respectively.

C

-~:,(,;\:C:t'*'\J:"'?~'. 'l°':. \f'!1,'(:' ,*~~~ttf:/li,:,'fi 111111

... **'**.'t

• • • O

' **O**:J***D**O**

• • • • • • • ~)*******

• • • • O*****D****

• • u*********C ** ****O

• • J***O**::J**

• • • • • • • CJ*******

• • • • • o****

• • CJ*********::**

• • • • • • • '.;***n***

• • )*********:;**

)*****:  :

..-~*.

***:_:****,~ ***

)

J**

)*****

)

_, )

ee

• • • • • • o********

Figure 6.1: Region 1 KENO Model NET- 28091-0003-01, Revision 0 63

Figure 6.2: Region 2 KENO Model The 2x2 model s are two cell pitches wide in the x and y directions. In creating the Region 2 model ,

the bottom left is modeled as a complete ce ll box with its Boraflex' sheathing. This requires that the cell wall be split between the top and bottom of the model (likewise for the left and right). When the periodic boundary condition is applied, the two cell wall pieces fit together to precisely match the actual rack dimensions.

NET- 2809 1-0003-01 , Revision 0 64

6.2 Axial Model All of the infinite models discussed in this section are finite axially. Above the active fuel , the models extend the clad 7.3 inches (except for Batches A-F where the plenum is 5.2 inches) creating a plenum composed of a stainless steel spring and void (the spring is 8% of the plenum volume). On top of the plenum region is a 3 1.43 cm of a homogeneous mixture of 50% water /50% steel (to simulate the end fitting) . Below the fuel, the reflector is 50 cm of water. Outside of the reflectors , there is a zero flux boundary condition. For the axial distribution of depletion isotopics, the fuel is modeled as nine discrete axial nodes . For non-blanketed fuel , the top eight nodes are 8 inches each and the bottom node is 80 inches . For assemblies with 6 inch blankets, the top eight nodes are 6 inches and the bottom node is 96 inches. For assemblies with 8 inch blankets, the top node is 6 inches, the second node is 2 inches, the third node is 4 inches, the 4u1, 5th, 6th, 7th, and 3th nodes are 6 inches and the 9th node is 102 inches. For axial burnup models using 13 nodes, the top node is 6 inches, the second node is 2 inches, the third node is 4 inches, the fourth through 12th nodes are 6 inches, and the 13 th node is 78 inches. The bumup distribution that determines the fuel atom densities for each of these nodes is discussed below in Section 6.2.1. Some of the axial blanket designs used annular pellets and these are conservatively modeled using solid pellets (i .e., there is more fuel in the model than reality).

6.2.1 Axial Burnup Distribution To model the axia l variation of a fuel assembly isotopic content, an axial burnup profile is needed .

For this analysis , the limiting profiles from NUREG/CR-6801 [27] are used for the full-length (non-blanketed) fuel. Table 6. 1, below, is reproduced fro m NUREG/CR-6801.

The initial burnup distribution is approximately cos ine shaped shifted down a little due to lower temperatures at the bottom of the core. With increasing burnup, the center reactivity decreases, so the flux moves toward the end and the burnup distribution flattens . Inspection of Table 6.1 for burn up bins 1 through 9 , as expected, shows the top node relative power generally decreases as the bumup decreases.

However, burnup bins 3 and 5 actually have higher relative burnups than bins 2 and 4. This does not NET- 2809 1-0003-01, Revision 0 65

make physical sense. NUREG/CR-6801 is based on end of cycle data collected by the DOE. The shapes used for burnup bins 2 and 4 come from assemblies that experienced some feature that suppressed the burnup at the top of the core such as a transition to axial blankets or perhaps control rod insertion. If the cycle length for these limiting assemblies had been shorter, then these assemblies would probably have had similar relative burnups and would have been counted in the lower bumup bin. It is concluded that the higher relative power seen from going from bin 2 to 3 or from bin 4 to 5 is actually an artifact of the data used to create the database and not due to a phys ical process. To eliminate this artificial increase in relative burnup with decreasing burnup, shapes 3 and 5 are eliminated and shapes 2 and 4 are used to cover the 2/3 and the 4/5 bins, respectively. The top node (97 .22% of the axial height) for bins 6 and 7 is greater than the top node for bin 4, but when the top two or three nodes are averaged, the expected decrease in relative bumup at the top of the core is observed, so bins 6 and 7 are not eliminated.

In summary, only bum up groups 1, 2, 4, 6, 7, 8, and 9 from Table 6.1 are used. Burn up profile groups 10, 11, and 12 are not used because no non-blanketed assembly has a burnup less than 14 GWd/T (an exception is F65 which required special analysi s, see Section 8.6) . The profiles for these low burnup bins were actually selected to match a center peaked flux because the reactivity is not dominated by the top nodes until about 14 GWd/T.

NET- 28091-0003-01 , Revision 0 66

Table 6.1: Axial Burnup Profile vs. Burnup Bin (27)

Burnup group 1' 2' 3 .j 5 6 7 9 10 I J' 1_ t ial Burnu ran es (GWd/MT height C0 ,) 42-46 38 2 34 38 <6 27 0.666 0 660 064 06 I 3 0.944 0.9 6 0924 I 007 13 9 1.0-t I 0-iS I 056 I 135 19.44 I 080 I 104 I 097 I 133

- .00 I 091 I 11 2 I 10 I 09

0. l.()<)3 I. I 06 I IO I I0

.I I 1.092 1. 102 I 103 I IL I 53 4 1 69 1.0 I 090 I 097 I 112 I 11 9 I 047 47.22 1.09-t 1. 125 I 126 1.0 50 1.094 1. 136 I 132 I 060 I 07 I 077 69 44 I L4 I 79 75.00 1.077 I. t:!O 1.073 5 I o-o 1.069 1.057 I 056 I 041 I IOI I 052 6.1 1 0.992 1.010 0.996 0.974 0 . 71 I O-i5 0.996 91.6 0. 33 0. l I 0. 23 0.743 0.7 0.689 0.669 0.894 0.845 97 22 0 447 0 448 0. 73 0 '69 0 - -

The burnup profiles are a step function of burnup, so they are discontinuous at the burnup bin boundaries. To eliminate these discontinuities in a conservative manner, the shape in any bin is conservatively assumed to occur at the maximum burnup in the bin. For any burnup in between these burnup points, the shape is linearly interpolated (the shapes are not changing rapidly between burnup bins) . For example, suppose an assembly has a burnup of39 GWd/T. UsingNUREG/CR-6801 directly, the top node would have a relative burnup of 0.525 . In thi s analysis, however, the top node has a relative burnup of only 0.467 (linearly interpolating between 0.447 at 38 GWd/T and 0.525 at 42 GWd/T). This ensures no discontinuities and all burnup shapes are conservative.

NET- 28091-0003-01 , Revision 0 67

For analysis of full-length fuel , the lower 10 nodes are averaged into one node. This does not affect the calculation of k, since the top half of the assembly dominates the reactivity. In fact, averaging the lower 10 nodes effectively brings the bottom lower burned fuel toward the more reactive top, so the approach is conservative.*

There are only a few assemblies in the IP2 or IP3 SFPs that have less than 18 GW d(T ( 11 Batch A assemblies from IP2 , 30 Batch A assemblies from IP3, IP2 assembly F65 , and IP3 assemblies V43 and V48). If there are any future assemblies with a burnup of less than 18 GWd/T, they must be assigned as Category l fuel , which does not credit burnup. The burnup profile for the 14-18 GWdfT burnup bin on Table 6.1 comes from a unique profile produced by considerable burn up under a control rod. The 15 assemblies having less than 18 GWdfT burnup that were burned under D bank are analyzed with the 14-18 GWdfT burnup profile from NUREG/CR-6801. All of the Batch A assemblies in this group have significant margin since they were all depleted with a 20 rodlet Pyrex burnable absorber, but none of them actually had a burnable absorber. The other assemb lies are analyzed using the 18-22 profile (the 18-22 profile is more reactive than the 10-14 profile) .

For analysis of axial blanket fuel , a conservative burnup shape is obtained by using the smallest relative burnup at each node from all assemblies in the group. The relative burnups are not re-normalized, so the assembly burnup in the analys is is reduced. Since the reactivity is controlled by the top nodes , however, compensating burnup increases in the lower nodes have little effect on k. The ninth node from the top is used for the ninth and all lower nodes. This has been shown to be conservative when compared to using all of the nodes. This ensures a conservative profi le when the reactivity is dominated by either the top or the center (the reactivity is never dominated by the bottom because the bottom nodes always have burnups that are higher than the corresponding top nodes).

• For more discussion on using only the top 8 nodes and an averaged bottom node see the response to the NRC RAI number 16 performed for the 2015 CSA (2].

NET- 2809 1-0003-01 , Revision 0 68

For blanketed assemblies, all ax ial bumup profiles from the plant were reviewed to detennine a limiting axial bumup profile for each of five axia l blanket designs:

6 inch annular* 2.6 w/o (Batch Q, R, S ofIP2) 8 inch annular 3.2 w/o (Batch T, U, V of IP2) 8 inch solid 3.4 w/o (Batch W of IP2) 8 inch solid 3.6 w/o (Batch X of IP2) 8 inch solid 4.0 w/o (Batch 2A+ ofIP2 and Batch GG+ ofIP3)

Assembli es in Batch X are segregated into assembli es that were depleted with no W ABA and those that were depleted with W ABA. The reason for this is that the axial bumup profile for non-W ABA assemblies is more limiting (lower bumup at the top) than for WABA assemblies , because the WABA pushes power toward the ends . In order to not penalize the WABA assemblies with the non-WABA profile, a separate depletion is perfonned by using no W ABA depletion with 148 IFBA. For the WABA group, the most limiting axial profile in the WABA group is used, and kerr is 0.9593. The most limiting axial profile in the non-WABA group is used to analyze the non-W ABA group, and k etr is 0.9532. So, the loading curve for Batch X (based on the W ABA analysis) can be used for all X assemblies. Table 6.2 shows the axial bumup profiles detennined from the plant data.

• The designs which used annular pellets are conservatively modeled using solid pellets (more fuel) .

NET- 28091-0003-01, Revision 0 69

Table 6.2 : Axial Relative Burnups for Blanketed Discharged Fuel X

IP2 Batch : Q, R, S T,U,V w noWABA X

Blanket Length :

6 8 8 8 8 (inches)

Blanket Enrich:

2.6 3.2 3.4 3.6 3.6 (w/o)

Top Node 0.420 0.448 0.471 0 .505 0 .519 2nd Node 0.743 0.598 0.622 0.663 0 .677 3rd Node 0.906 0.76 2 0.772 0.779 0 .805 4th Node 0.989 0.866 0.883 0.884 0.866 5th Node 1.029 0.975 0.986 0 .990 0.987 6th Node 1.048 1.020 1.030 1.033 1.027 7th Node 1.059 1.041 1.050 1.052 1.048 8th Node 1.064 1.051 1.060 1.061 1.057 9th Node 1.068 1.057 1.066 1.067 1.062 For the Batch group Z axial burnup profile, the profiles from Batch 2A assemblies are used. This was the first batch to use 4 .0 w/o blankets . The previous batch used 3.6 w/o blankets. Since the top nodes of the 2A fuel assemblies were surrounded by the 3.6 w/o blankets, the 4.0 w/o blanket would be slightly underburned by the presence of the 3.6 w/o blankets. This "transition" from 3.6 to 4.0 w/o blankets bounds all future 4.0 w/o blanket fuel (designated as Batch Group Z). The Batch 2A assemblies are divided into two groups - assemblies that had a WABA insert plus 148 IFBA and all other assemblies.

Since the WABN 148 IFBA depletion condition is used, only the assemblies that had a WABA insert plus 148 IFBA should be used to find the limiting axial profi le. Results from the Batch X analysis showed that the depletion effect from a reduced amount of burnable absorber is worth more than the effect of the axial shape with reduced absorbers. As further confirmation of this effect, a special depletion was performed for Batch Zin which the fuel was depleted with a W ABA plus 116 IFBA compared to the standard depletion with a W ABA and 148 IFBA. The kerr for the 116 IFBA depletion using the corresponding profile is 0.9569 while the kerr for the 148 IFBA depletion using the 148 IFBA profile is 0.9584. The limiting profile is the minimum relative power in each node for all 2A assemblies that had a WABA insert plus 148 IFBA. The limiting profile fo und at three different bumups is in Table 6.3.

NET- 2809 1-0003-01, Revision 0 70

Table 6.3: Axial Relative Burnups for Batch Z Fuel 48 28.5 21 Burnup:

GWd/T GWd/T GWd/T Top Node 0.562 0.560 0.540 2 0 .727 0.746 0.729 3 0 .815 0.787 0.754 4 0.893 0.826 0.787 5 0 .984 0.950 0.916 6 1.020 1.001 0.976 7 1.039 1.029 1.017 8 1.047 1.042 1.041 9 1.051 1.050 1.061 10 1.055 1.057 1.074 11 1.056 1.052 1.076 12 1.065 1.067 1.095 13 1.061 1.059 1.088 14 1.065 1.061 1.092 15 1.070 1.072 1.102 16 1.069 1.064 1.089 17 1.075 1.071 1.093 18 1.080 1.083 1.095 19 1.081 1.086 1.088 20 1.081 1.085 1.075 21 1.074 1.079 1.054 22 1.047 1.045 1.003 23 0 .976 0.971 0.918 24 0 .875 0 .890 0.848 25 0.713 0 .738 0.718 Bottom Node 0.534 0.545 0 .520 To simplify the analysis, only the top nodes need to be modeled with the last node representing all of the nodes below the last node. The fewer the number of nodes modeled, the more conservative the result because all of the nodes below the last one are modeled with a lower burnup. It was found that using nine nodes for the higher burned fuel is not overly conservative (the k at 48.2 GWd/T using nine nodes is 0.9586 while the k using all 26 nodes is 0.9585). For the analysis at 21 and 28 GWd/T (Z fuel) , 13 nodes were used instead of nine, in order to reduce some of the conservatism. Table 6.4 summarizes the k calculations (the Monte Carlo statistical uncertainty is only+/- 0.00003 t.k).

NET- 28091-0003-01, Revision 0 71

Table 6.4: Calculated k versus Number of Nodes Modeled Burnup 9 nodes 13 nodes 26 nodes (GWd/T) 21.0 (3 of 4 in Region 1) 0.9689 0.9686 0.9686 28.5 (4 of 4 in Region 1) 1.0117 1.0115 1.0114 48.2 (3 of 4 in Region 2) 0.9586 0.9585 0.9585 6.3 Dimensional Changes with Irradiation The fuel assembly dimensions change a small amount with irradiation. This creates a change in reactivity. This change in reactivity is real but too small to include in fuel management analysis.

With irradiation, the fuel pellet densifies and then expands, the clad grows and creeps down to the pellet, and the assembly grid expands. These are changes to the dimensions not the mass. The changes in the dimensions of the fuel pellet (with mass constant) create an insignificant reactivity change . The change in the dimensions of the clad and grid, however, result in more water relative to the fuel. Since the fuel is deliberately designed to be under moderated (to ensure a negative power coefficient), adding more water to the fuel assembly is a positive reactivity effect.

6.3.1 Clad Creep The clad outer diameter initially decreases due to creep caused by the pressure difference between the core pressure (2000+ psi) and the He fill gas (200+ psi) in the rod. At some point the clad contacts the pellet and then the clad outer diameter increases as the pell et swells . When manufactured, per Table 3.2, the difference between the pellet diameter and clad inner diameter is 0.0075 inches or 190.5 microns (l0*6 meters) . If the clad were to creep down to the initial pellet diameter, then this would be a 1.8%

decrease in the clad outer diameter (it is assumed that the clad thickness is constant). The clad creep rate depends on the clad material. However, for all of the clad materials, the pellet densities and then grows back past its initial outer diameter before 190 microns of creep occurs.

NET- 28091-0003-01, Revision 0 72

Figure 6.3 shows clad creep after one cycle at North Anna [37]. As seen from Figure 6.3, Zircaloy-4 creeps down faster than ZIRLO'. The maximum creep down after one cycle is about 90 microns for Zircaloy-4 and 70 microns for ZIRLO'. The creep down data is only for one cycle, since by the end of the second cycle the clad had reached the pellet OD. Sabol, et al, confinns this as fo ll ows:

"Th e profllometry data obtained after one cycle of irradiation has been used to evaluate the differences in the alloys ' in-reactor creep behavior. Fuel-clad contact over most of the rod length occurred during the second cycle, and the profllometry data obtained on the two-cycle rods are controlled by the fuel p ellet swelling rather than the cladding creep. Th erefore, the two-cycle data cannot be used fo r creep analysis". [3 8]

Figure 6.4 provides more data on the creep of Zircaloy-4 [39]. From this plot, the creep down is basically linear with burnup and can reach about -0.8 % of the initial clad outer diameter (about 80 microns which agrees with the creep data from Figure 6.3).

Craepdown (microns) 0 20 40 60 80 100 Zircaloy-4 120 0 so ,oo 1SO 200 2SO 300 Location (cm from bottom of ro(1)

Figure 6.3: Comparison of Creep-down for ZIRLO' and Zircaloy-4 [37)

(One cycle of irradiation at North Anna Unit 1)

NET- 28091-0003-01, Revision 0 73

-1,0

-0,9

• o,e ll P.t e

- 0,7 01 0,64 iii B0

-0,6 02 0,40) A

~

!.. -0,5 63 0,59 ct, dPA:

nelof rods O

<i -04, G) {l

[i) 'l

-0,1 normalized with J0.85

-02

, 200 I

300 400 soo 600 800 1000 1200 Exposur* Tim* (Oaysl Figure 6.4: Diameter Decrease versus Exposure Time [39)

Figure 6.5 shows the clad creep for several alloys for fuel used at the Vandellos 2 Nuclear Power Plant [40]. The initial clad OD for the Vandellos plant is 9.5 nun.

NET- 28091-0003-01 , Revision 0 74

o,,r"-----------------------------------.

-20

,-.. D o* *

'§: -40 *

~ ...

a

..t: A A A

<.)

~ -60 A *

.I...

'O

~ -80 0

ell

• =

'O

'O Segment Type X : SS Cladding Conventional Zr'y-4 0

- -100 0 : WI Low-tin Zr'y-4 6 : \\Z ZIRLO 0 : M.\1 MDA

• : !\fit Low-rui with te.xture control 0

-120 A : ~ t MDA with texture control

• : 121 ZIRLO with texture control

-140 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _....__ _ _ _ _...

0 10 20 30 40 50 60 Segment a\*en,ge bumup (G cLt)

Figure 6.5: Clad Creep Down for Vandellos 2 Nuclear Power Plant [40)

Finally, Figure 6.6 shows the clad creep for a fuel assembly from the Ulchin Unit 2 PWR [41]. This assembly uses a low tin Zircaloy-4 clad. The fuel assembly has a burnup of 50.5 GWd/T. The initial clad OD is 9.5 mm. As can be seen from Figure 6.6, the clad OD with oxide is greater than 9.5 mm in the hi gh reactivity zone. This is due to significant oxide build up and fuel pellet expansion. For this assembly at this burnup it is conservative to ignore creep.

NET- 2809 1-0003-01 , Revision 0 75

9.65------------------------,

---

Average Diameter

- - Diameter(oxide corrected )

9.60 9.55 I-Q)

+-'

9 .50 Q.)

E cu 0 9.45 9.40 0 500 1000 1500 2000 2500 3000 3500 4000 Distance from Bottom End, mm Figure 6.6: Axial Distribution of the Fuel Rod Diameter at 50.5 GWd/T [41]

With bumup, the clad also builds up an oxide coating. This oxide coating displaces water so from a reactivity point of view, it is similar to increasing the clad outside diameter. Figure 6.7 shows a large data base of the oxide layer growth with bumup [42).

After the clad creeps down to the fuel pellet, the clad OD increases due to pellet swelling. Figure 6.8 shows the normal fuel pellet swelling (via pellet density) as a function ofbumup [43).

NET- 28091-0003-01, Revision 0 76

160 Zircaloy-4 140

• ZIRLO OZIRLO - Plant C Olow Tin ZIRLO

• Plant C 120 Cl ZIRLO

• Plant D Clow Tin ZIRLO - Plant D 100 t. Op!tm1zed ZIRLO

• Plant E

¢ Optimized ZIRLO Plants F & G 80 60 40 20 10 20 30 40 so 60 70 80 Bumup (GWd ~ U)

Figure 6.7: Oxide Layer thickness with Burnup [42]

100 98 96 0

f- 94 -- .. .. -- : --

'::!?.

0

.£(/)

92 -- -- ..

C (l) 0 90 -- -- .. --- ,--- ---

88 - - . . .. --

0 20 40 60 80 100 Pe ll et Bu rn up MWd/ kg U Figure 6.8: Density of Fuel Pellet as a Function of Pellet Burn up [43]

NET-28091 -0003-01 , Revision 0 77

With the data presented in these six figures , a creep model has been approximated. Since the reactivity effect is small , the model is not meant to be bounding but to have a slightly conservative mean reactivity effect. The deviation about the mean is an uncertainty that could be statistically combined with the other uncertainties. However, since the entire effect is small, the statistical combination of the uncertainty would be negligible, so it is ignored for this CSA. The modeling is as follows:

1. Per Figure 6.8, the fuel pellet returns to its original density at 30 GWd/T.

2. The pellet density changes are translated to changes in the pellet outer diameter (no axial swelling) . Axial swelling is a planar loss of mass that decreases k.

3. Using the slope of the change in pellet density given on Figure 6.8 ( l % density reduction per 10 GWd/T), the expanded pellet outer diameter is detennined as a function ofburnup.

4. Using Figures 6.4 and 6.5, assume the creep is linear with burnup. The slope for conventional Zircaloy-4 is approximated as 2.5 microns per GWd/T.

5. Using this data, determine the bumup where the clad creeps down to touching the pellet.

6. Use Figure 6.7 to estimate an oxide thickness gain of0.5 micron per GWd/T. This data is used to determine when the clad returns to its original outer diameter.

Using this approach the maximum creep (the point the clad reaches the pellet) for Zircaloy-4 is at 46 GWd/T and is 115 microns. However, at that bumup the oxide layer is 23 microns thick (46 microns diametrical effect) making the net reduction of the clad outer diameter only 69 microns. Upon reaching a burnup of 56 GWd/T, the clad plus the oxide layer is back to its original outer diameter (this agrees with the measurements shown on Figure 6.6) .

The clad creep model used here simplifies the effects in a slightly conservative manner. This CSA assumes the clad outer diameter decreases linearly from O to l 00 microns over the burnup range from NET- 28091-0003-0 l, Revision 0 78

zero to 40 GWdff. As the bumup increases from 40 GWdff to 56 GWdff, the clad outer diameter increases linearly to the point where it is the same as the initial clad outer diameter. Any further increase in the clad outer diameter is conservatively ignored. The creep down as a function of bumup was determined to cover both Zircaloy-4 and ZIRLO' clad fuel. ZIRLO' clad however, creeps down less, so the analysis is conservative for ZIRLO'. ZIRLO' fuel has been used since Cycle 13 (IP2) and Cycle 9 (IP3).

6.3.2 Grid Growth Zircaloy grids grow a small amount with irradiation. Figure 6.9 shows measured grid growth in ZIRLO' and Optimized ZIRLO' [42]. Figure 6.10 shows the grid growth in Zircaloy-4 (and M5 which is not used at Indian Point) [44]. RXA on Figure 6.10 stands for fabricated in stress-relief annealed (SRA) and recrystallized (RXA) conditions. Figure 6.11 shows the grid growth from VC Summer and Wolf Creek measurements [45]. The Wolf Creek assemblies are burned to 50 and 51 GW d/T [46].

ZIRLO' grows less than Zircaloy-4, but they both grow more at higher temperatures than at lower temperatures.

The Inconel grids do not have a growth problem [4 7]. King, et. al. states, "Most Westinghouse fuel designs use Inconel top grids and Inconel bottom grids, and the Inconel grids are designed to maintain a pre-load on the fuel rod until end of life" [48].

NET- 2809 1-0003-01 , Revision 0 79

lJr1o Grid GrOV;ih Data Base 0.80 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - .

+ Plant A

- Plan!B t 060-t---1

• Plant C

.i X GF Opbllllled ZIRLO Data

.c

&40 ---1 , Plant EOp . ed ZIRLO Da a 1 - - - - - - ' - - - ' - - - -

• t

<'5 "O

• c

'- t X

. ' X

(!) ~

t .

• I f X t_ t +

** t

!i ;

+

+ ' X 000 +----,,----,----,---...----,.---,---.,----,--...,..-----i 20000 25000 30000 35000 40000 45000 50000 55000 60000 65000 70000 Bumup ( ,1WD ITU)

Figure 6.9: ZlRLO' Grid Growth r42]

0.9 0

0.8 RXA Zircaloy-4 0 0.7 0

~

0.6 8

~ 0 0

!,, 0.5 0.4 0

0

~

0 t,.

cS 0.3 t

0.2

• j 0.1 MS '

0.0 0 10 20 30 40 50 60 FA burnup (GWd lt U)

Figure 6.10: Zircaloy-4 and MS Grid growth versus burnup [44]

NET- 28091-0003-01 , Revision 0 80

1.2 15

~ZIRLO Grid Envelope

~ Zr-4 Grid Envelope (Assembly 1) 1.0 --tr-Zr-4 Grid Envelope (Assembly 2} 12

- - ZIRLO Fluence > CD

e

-:;- 0 .8 -

• * * * *

• Zircaloy-4 Fluence ...

.,C I\

u 9

. w E

E: 0 .6

.s:.

~

.c 11'1 0

0 C)

• 6 ~

0.4 !C CD

i 3

iL 0 .2 0 .0

• 0 0 1000 2000 3000 4000 Ax ial Elevation (mm)

Figure 6.11: Grid Growth of ZIRLO' and Zircaloy-4 versus Elevation [45)

There are practical limits to grid growth. Excessive or unexpected dimensional changes of guide tubes or spacer grids of a fuel assembly can resu lt in operational issues such as incomplete (control) rod insertion (IRI) [42] or potential fuel assembly interactions and handling concerns due to increases in the fuel assembly envelope resulting from the lateral growth of the grids.

For Indian Point, if the fuel pin pitch expands such that it is uniform for the reactor assembly pitch, closing the inter-assembly gap, then the fuel pin pitch expands only 0.25%. Note that the outer cells of the grid do not have the fuel pin centered between grid straps so there is some additional space even if the grid expanded to uniformly fit the assembly pitch. If the grid straps were touching for two equally expanded assemblies the grid growth would be 0.69% .

NET- 28091-0003-01, Revision 0 81

The grid growth clearly increases with bumup and the increase is more pronounced for higher bumups. A cubic fit has been generated to estimate the grid growth. The fit is roughly drawn on Figures 6.9 and 6.10. It is :

Grid Growth(%)= 4.3E-6*BU3 - 0.000 13*BU 2 + 0.0051 *BU where BU is the bumup in GWd/T.

Although not all measurements of grid growth lie beneath this curve, using this grid growth model as a unifonn pitch expansion is expected to be conservative. The Inconel grids above the active fuel will hold the fuel pins in their original pitch, since the Inconel growth is insignificant (from a material point of view as well as seeing less fluence, since they are outside of the active fuel region). The lower grids expand less. Therefore, in order to get the expansion in the pitch, the fuel rods would have to bow. This analysis conservatively assumes the expanded pitch is uniform over the entire length of the fuel.

For simplicity, a single fit is used to cover Zircaloy-4 and ZIRLO'. However, Zircaloy-4 grows more than ZIRLO'. Zircaloy-4 was used as a grid for only Batches M, N, P, Q, and R in IP2 and Batches T, U , V, Wand X in IP3. All of these assemblies have margin to their assigned reactivity category. In fact, if the grid growth were increased from 0.44% to 1.0% at 50 GWd/T, none of the assigned categories for the Zircaloy-4 grid fuel assemblies would change.

6.4 Averaged Assembly Peaking Factor Interpolation The depletion analysis is perfonned at five discrete averaged assembly peaking factors; 0.6, 0.8, 1.0, 1.2, and 1.4. In order to simplify the dependence on this peaking factor, it is desirable to fit these five points with a single straight line ( one for each burnup , enrichment, cooling time, and batch grouping). It was fo und that interpolating between peaking factors of 0.6 and 1.4 is slightly non-conservative at a PF=l.00. To ensure conservatism, depletions are performed at 0.8 and 1.2 and are linearly interpolated NET- 2809 1-0003-01, Revision 0 82

and extrapolated using these two points. It was found that extrapolation to 0.6 and 1.4 is conservative. A representative graph at 50 GWd/T (Batch Z at 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooling) is shown on Figure 6.12.

0.960

• Points 0.955

- Linear between 0.8 and 1.2 0 .95 0 0 .945 k

0 .940 0.935 0 .930 0.925 0.5 0 .7 0 .9 T

1.1 Peaking Factor 1.3

- 1.5 Figure 6.12: Calculated ken versus Assembly Average Peaking Factor 6.5 Convergence of the 2x2 Infinite Model Calculations The convergence of the 2x2 reflected model keff calculation is generally achieved after only a few hundred generations. However, all of the CSASS computer runs use a Monte Carlo sampling of at least 8000 generations and 8000 neutrons per generation. Convergence could have been a problem in the past, when very few neutron histories were run (300,000 total neutron histories), but due to increasing the number of histories to 64 million, convergence is no longer an issue.

For both the number of generations skipped and the starting source, the SCALE default is used. For the number of generations skipped, the default is 3. However, SCALE calculates the number of generations to skip that gives the minimum uncertainty in the final result. The keff reported for all of the calculations is the keff With the optimum generations skipped. The number of generations skipped spans a wide range but is genera lly between 100 and 200. When running a large number of generations, such as 8000, the input number of generations skipped is not significant to the final results. The default start NET- 28091-0003 -01, Revision 0 83

source is a unifonn source over all of the fissile materials in the model. The number of neutrons per generation is always 8000 or greater and for the four assembly models this sampling is enough to find the most reactive portion of the model.

6.6 Full Pool Models The full pool model is created by taking the 2x2 models for Region 1 and 2 described in Sections 6. 1 and 6.2 and using them as units that are reproduced in arrays. The model has 4 large arrays (see Figure 3.1 for module identification):

1. Region 1 module A (10x8) ,

2. Region 1 modules Band C (combined as 21x9),

3. Region 2 modules D, E-1 , F-1, F-2, G-1 , and G-2 (combined as 24x32), and

4. Region 2 modules E-2, E-3 , and H (combined as l lx32) .

Modules E-2, E-3, and H are 11 cells across (north to south), but since the modeling is using 2x2 units , a new lx2 unit was made and added to the right hand side of the model. This lx2 model correctly removed the Boraflex TM box on the outside of the rack module near the SFP wall but did not add the steel plate that is used to close the resultant cell on the outside of the module.

The full pool model does not model the gap between the rack modules . This means that Module H is placed next to Module C and lowered so that the bottom of Module H is the same as the bottom of Module C. Directly below Module H without any additional space are Modules E-2 and E-3 . To the left of Modules E-2 and E-3 are the other Region 2 modules without any space between modules. This means there is more water on the outside than in the real SFP, because the inside dimensions of the SFP in the model are the actual dimensions of the SFP. The separation in the model between the rack and the SFP liner on the top and left side of Region 1 is the actual separation. Since the separation between the rack modules has been removed in the model, there is more water on the right hand side of the model than is really in the SFP. Similarly, at the bottom of the model there is more water than is actually there. Since NET- 28091-0003-01, Revision 0 84

the 2x2 unit from Section 6 was applied, the exterior of the rack modules genera lly has less steel than the outside wall of a rack module, but it is close. In Section 6.6. 1, analysis of the sensiti vity to the SFP edge is perfo rmed , which shows small sensiti vity.

The separation between the Regi on 1 rack and the top wall is 2.125 inches. The separation between Region 1 and Region 2 racks from the wall on the left is 1.25 inches The SFP has a 0.25 inch stainless steel liner and a concrete wall outside the liner.

Figure 6.13 is a SCALE generated drawing of the full poo l model.

ll.C(M)

O wrn fWHCRUll 2

- l'WHLIUl'll. J

- MlUtlAL"

- l'VIJ[IUAL 7 0 Ml[IUAL I

- Ml[RlR.10 C ) MIOUAL J t

- fWHDUAL 6'

- M I UUnl '3 M IUUAI. 1 10 M l (RIAL 2 1 1 flHlUUAL 2 12

~ Ml(IUtlt. 2 13

(::J mmnAL 2 14 Q M1£11tlAL 2 1!S WHOUAL 2 18 MIUUAl. 217 M lUUAl. 211 MTiltlAI. 312 CJ MlUWll 3 13

- Ml[RIAI 3 14 Ill Ml[IW'll 3 t !S

- Mf(IUnt. Jl6

- MfUUfll. 3 17

- Ml[RlfL 318

- M JUtltll.,10

- Ml(RIAL'lll

- P!AllRIAL41 2

- Mf[MIAL<II J

- PWU UIIAI "11'4

- MJ[RIAL"115

- MllJUf:11'41 6

- Ml[IUtll.<1 17

- Ml(RIAL *H 8

- Ml(RIAL !UI

- MfUUAL!S l 2

- Ml[ MIIW.. !SIJ

- M1CRIAL !il 4

- Mll.RIAL !S l !S MIIM UII. 5 18

- WHUtlAL !Sl7

- MIUUAl. !S 18 O iwHUllnt. 9etl Figure 6.13: Full Pool Model NET- 28091-0003-01 , Revision 0 85

The pink squares in Figure 6.13 are empty cells which can contain up to 50% water displacement with non-fuel components. These cells are modeled with a void fraction of 50% . The water holes (white) are modeled as pure water. However, Section 8.13 provides analysis that shows that up to 50% volume fraction of stainless steel may be in these water holes . Figure 1.1 shows that contro l rods are in specified locations. These contro l rods are in the full pool model but cannot be seen in Figure 6.13 due to poor resolution. Figure 6.14 is a blow up of Module Hand the control rods can be seen.

O vo10""""

- Ml[IIIAL l

- Mrt:RIH..4

- MJ[Rlnl. 7 IW'l1£111nL 8

- MT[RIALIO c:J MlERIAl. 3 1

- PIAll:IUlll.. 66

- PIATtlllnl ll PIAICl'Cll'I 2 10

. ..l'Ht:Mlr... 2 11

~ MT(IUl'l. 2 13

~ Ml(IIIA1 2 1<1 C ) MJUtlfll. 21 5 MIERJnt 2 16 MT(ll: l AL 3 10 MTCIUl'W.. 3 1 1 MrEIUl'll 3 12 MTEll:1,-_ 313

- MTUU nL l l<I

- MICRIAI 3 15

- MT£RIN 3 16

- MT[RIAl.. 3 17

- Mf[Rll'll 3 11

- Mf(ll:lnL<II O

- MTERIAL<lll

- Ml[RIN.._<4 17

- MT[Rll'll.. <II J

- Ml[HUl.'41<1

- MltNlrl.'41 5

- PWOOW'IL '41 1

- MT[ NJAL-41 7

- MT(ll:JAL-411 Ml'l:IUAL '5 10

- MTEftlnL '5 ll

- MT(ftlnL '5 12

- MrERlnL '5 13

- l'IAJfNJN.. '5 H

- Ml [ NIN.. tll '5 l'IAT(ll:fnL 5 Ul

- l'IATEll:Hl. 5 17

- Mf(ll:Hl. 5 11 D "'HENHI. Ne

!;ii;J;!;ii;ii it-:;! ii ni 11:11;!:i;i!;!i g:!\m!..*ii 11:!i;!;J;i!:i! ii:fi:~!{!i !i:!\;i:!;i!:!! !i:rr;m;i!:!! ii:!i;m;ii;ll H:ii;m;i!:!! ii:!m!!;i!:!!

1 I

Figure 6.14: Model of Module H Showing Control Rods NET- 28091-0003-01, Revision 0 86

There could be a concern regarding the content of control rods that have been used in the core. Due to end effects the reactivity is dominated at the top of the active fue l. However, the top third of the control rods is not allowed in the core if the power is above 50% of full power. Any depletion of the control rods is insignificant. However, to confirm that content is not important, a run was performed with the atom densities of the control rod reduced 20%. The calculated keff is within the Monte Carlo uncertainty of the reference case.

6. 6. 1 Sensitivity of the Full Pool Model to Modeling Assumptions A number of model sensitivity cases were performed to determine if the modeling of the outside of the racks and the SFP wall is adequate. Table 6.5 provides the results of these sensitivity cases. The reference cases use the nominal separation from the wall found on the SFP layout drawing for the top and left side of the model [35]. The reference cases are at the limiting enrichment (5.0 w/o) and bumups (21, 27.7, 48.19, and 57 .89) plus asymmetry. For this sensitivity study, the linear dimensional changes are large (tota l elimination and a doubling of the width) . Even with these large changes, the maximum reactivity is only five times the Monte Carlo uncertainty of the cases . The models are slightly more sensitive to the separation from the wall but with moving the racks as close to the wall as possib le, the reactivity effect is at most 0.0005 in k. The model uses the standard regulatory concrete which is a standard mixture in SCALE. An EPRI study of a conservative minimum water concrete is used to find the sensitivity [13]. This very conservative concrete only increased k by 0.0008 for Region 2 and 0.0004 for Region 1. Since all of the extreme changes caused 0.0008 or less change in reactivity, the full pool model is adequate. The maximum k for the analysis is 1% less than the regulatory requirement.

NET- 28091-0003-01, Revision 0 87

Table 6.5: Full Pool Model Sensitivity Tests Case k Sigma '1k Region 1 Reference 0.9687 0.00007 Eliminated the SFP steel liner 0.9689 0.00006 -0.0002 Increase the SFP steel liner from .25 to .5 inches 0.9686 0.00006 0.0001 Decreased rack/liner separation from 2.125 to 0.8225 inches (Top) 0.9691 0.00006 -0 .0004 Increased rack/liner separation from 2.125 to 4.125 inches (Top) 0.9686 0.00006 0.0002 Changed Concrete from reg-concrete to EPRI - minimum water 0.9691 0.00006 -0.0004 Region 2 Reference 0.9584 0.00005 Eliminated the SFP steel liner 0.9585 0.00006 -0.0001 Increase the SFP steel liner from .25 to .5 inches 0.9587 0.00006 -0.0003 Moved rack to meet the SFP steel liner (left side) 0.9589 0.00006 -0.0005 Increased rack/liner separation from 1.25 to 2.98 inches (left) 0.9584 0.00006 0.0000 Changed Concrete from reg-concrete to EPRI - minimum water 0.9592 0.00006 -0.0008 6.6.2 Convergence of the Full Pool Model There is a classic problem for Monte Carlo convergence known as the "k-Effecti ve of the world."

Thi s problem was introduced by Elliot Whitesides in 1971 (50] and more recent papers such as the Brian C. Kiedrowski and Forrest B. Brown paper at ICNC 2011 (51]. Concern has been raised that a large full pool model could face these problems. First, in 1971 when Whitesides raised thi s issue it was common practice to run a few hundred neutrons per generation . The paper by Kiedrowski and Brown used a lot more neutrons per generation but considered solutions with 10,000 neutrons per generation or less as typical. For this CSA, employing computers with more powerful CPUs, 16,000 neutrons are started per generation. This analysis also uses 8,000 generations. The problem posed by Kiedrowski and Brown used a cadmium wrapper to isolate the reactive location. The full-pool-model employed in the current analysis is more neutronically coupled than the model used in the Kiedrowski and Brown problem so at the current number of neutrons per generation, source convergence is not a problem. Kiedrowski and Brown pointed out "suppose the source specification is modified to incorporate only the central sphere. Will this yield more reliable results? The answer is yes, and remarkably so. Even with a batch size of lk, the value ofkeff is always predicted correctly for each of the 100 trials." In the current NET- 2809 1-0003-01, Revision 0 88

analysis, when there is an isolated high reactivity area, the start source is specified at that location, thereby removing the convergence concern.

Many articles are available relating to the topic of source convergence. This issue, however, is not of primary concern for this CSA, as it seeks to determine the kerr, not the flux . In the Kiedrowski and Brown paper, they say, " ' sloshing' of the fission source has an observable impact on k 0 rr.. .. however, the impact on keff itself is small." This sloshing of the fission source increases the uncertainty in the calculated kerr which is included in the bias and uncertainties . The large number of generations used in this CSA assures the mean kerr is correctly predicted within the uncertainty.

To confirm that the full pool model is sufficiently coupled, six different start sources were used to analyze one of the final full-pool cases . The calculated k0 rr is dominated by Region I. For cases including Region I, this CSA uses a start source that covered most of Region I . In order to challenge the convergence issue, the source was started at four boxes as far from the dominate kerr area as possible. A final case was run with the SCALE default source, which is uniform over all fission materials in the problem. Figure 6. 15 shows where the four start sources are located for the model. Table 6.6 shows the calculated kerr values. All of the calculated k0 rr values are within two standard deviations . The cases with the worst start source have a higher reported sigma. It is concluded that the neutrons per generation and the number of generations used for the full models for this CSA is sufficient for convergence.

Figure 6.16 shows how the average kerr behaves with progressing generations. It is clear that the neutron population is moving away from the start source when the start sources are located in the comers of Region 2. All of the cases in this CSA are converged. In some engineering applications, such as checking interface interaction, all that is needed is to determine that the location of the start source is not the most reactive location in the model. When this is the case, then convergence is not required. For example, in this case, it is clear that the three Region 2 start sources are not at the most reactive location in the model.

NET- 28091-0003-0l , Revision 0 89

This analysis uses the ketr produced after a certain number of generations skipped. The number of generations to skip is determined by SCALE, which minimizes the standard deviation . It has been found that this almost always yields the best estimate of k. Table 6.6 shows that the number of generations skipped can vary greatly. Clearly, for this convergence test, if a fixed number of generations to skip were used, the results would be less accurate. Reviewing Figure 6.16 makes it clear that this SCALE feature is correct and needed if the start source is poorly selected. The differences in kerr seen at the right hand side of Figure 6.16 are greater than that seen in Table 6.6 since the variation in the number of generations skipped is not part of the average ketr shown on Figure 6.16.

Table 6.6: kerr Changes With Start Source (Full Pool Model - 8000 Generations, 16000 Neutrons per Generation)

Source Location Calculated k Reported Sigma Generations Skiooed Region I only 0.968708 0.000067 662 Uniform for Problem (Region l and 0.968724 0.000066 372 2)

Region I Bottom Left 0.968704 0.000066 445 Top Right 0.968663 0.000066 525 Bottom Right to Cask Area 0.968809 0.000075 2066 Bottom Far Right Above Cask Area 0.968682 0.00006 8 1282 NET- 28091-0003-01, Revision 0 90

Lftl:NO O vo10 MT[ AIM. 2

- MT[ Rl._. l

- MAT O II .--, '4 O wou:11..._a O M 1u11V... 1 1

- Ml£Rl(I,_ C6

- MlEIW'll. 83

- MTUW1t. 2 10 Ml(HIAl. '212 Q MlUW't , 2 13 t::J MTEIW'lt. 2 1..

O MTUlltl.. 215 MTtRIAl 2 16 M Tt RIAL 3t O MIUtfr.. 3 11 MIUUtl. 3 12 c::J MIOUAI 3 13

- P!Al(Jl:l,W. ) 14

- PV'ITUtl(l,_ ) 15

- Mlf.lW'l. 3 16

- MTERJN._ 3 17

- Ml(M l r.. 318

. ,.HUUi..*nO

- l'W'l'r(R l Jll <11 1

- WIILIW'l.412

- MlUWll.41 3

- MTERIAl..-41'4

- l'WllUW'l. *U 5

- M'JUtlAL ... 16

- IWH(fllAL'417

- IWH£RUll. *118 c:J ,-nEIUnL SIO

-l'W'H[l!IAL 5 11

- IVI IUHl'W.. 1112

- IWIT11UAL ll13

- MT(RIAl. 5 1<1

- 11M ( l!IAL 5 t5 c::J MU::RUW.. 5 16 l'WHl::Hl nl 5 17

- M!EfUN ti 18 Figure 6.15: Locations of the Start Sources for the Convergence Tests NET- 28091-0003-01 , Revision 0 91

0 .97 0 .96 0 .95

..:,: Start Scource Location "C

"S CII

~

0.94

• Region 1 iv

• Uniform u

CII

• Region 1 Left Bottom t,Q

~

CII 0 .93 ------ Top Right er: , Far Right Bottom

• Bottom Right Cask Area 0.92 0 .91 0 .90 T 0 1000 2000 3000 4000 5000 6000 7000 8000 Number of Generations Figure 6.16: Change in Average kcrr with Progressing Generations NET- 28091-0003-01 , Revision 0 92

6. 7 Summary of Modeling Assumptions The following is a summary of the modeling assumptions:

1. Bounding fuel stack density and nominal dimensions for pellet OD, clad OD/ID, and guide tube OD/ID are used.

2. Axial blankets are modeled for five different axial blanket designs . A conservative bumup profile for each blanket design is created by using the smallest relative power at each node from all assemblies having the same blanket design.

3. Grids are ignored (grids displace water between the fuel pins which causes kerr to decrease).

4. No BoraflexTM in the BoraflexTM sheathing and the Boraflex' is replaced with water (if any BoraflexT M remained, it would be less reactive than water).

5. NUREG/CR-6801 bounding axial bumup profiles are used for all full-length fuel. To address discontinuities, the shape in any bin is conservatively assumed to occur at the maximum bumup in the bin. For any bumup between points, the shape is linearly interpolated.

6. Top of fuel assembly models a plenum (length dependent on batch) filled with 8% volume fraction of stainless steel. Above the plenum is a 50/50 mixture of stainless steel and water to simulate the end fitting. Bottom reflector is 50 cm of water.

7. Periodic boundary conditions are used to represent an infinite array.

8. The dimensional changes with bumup are modeled as described in Section 6.3.

a) In reference to clad creep, clad thickness is constant, and creep occurs linearly, in two phases (creep inwards and growth outwards) .

b) In reference to grid growth, pin pitch expands uniformly.

The water temperature in Region 1 of the SFP is modeled at 180 °F which is the SFP design basis maximum temperature. In Region 2, the modeled temperature is 70 °C which is the temperature that maximizes kerr in Region 2.

NET- 28091-0003-01, Revision 0 93

Proprietary Information Removed Westinghouse Non-Proprietary Class 3 7 Sensitivity Analysis This section presents analysis of the sensitivity of the models to the manufacturing tolerances. After the sensitivity is determined, the rack up of the uncertainties and biases is presented.

7.1 Manufacturing Tolerances Calculations are performed to quanti fy the reactivity effect of changes due to manufacturing tolerances. For Region 2, the tolerance calculations are performed at the highest credited bumup conditions (5 w/o at 49.5 GWd/T, PF= l.4) and a low burned condition (4.2 w/o at 30.0 GWd/T, 25 year coo ling, PF=.6). Table 7.1 presents the calculated tolerance reactivities.

Table 7.1: Tolerance Reactivity Effects Re2ion 1 1 2 2 2 Fuel Reactivity Cate2ory 2 3 4 4 5 Arran2ement 3 of 4 4 of 4 3 of 4 3 of 4 4 of 4 Enrichment (w/o) 5 5 4.2 5 5 Burnup (GWd/T) 20.5 28 30 49.5 60 Tolerance Calculated Ak (inches) (no Monte Carlo Uncertainty Adjustment)

Pellet Density +0.35 % 0.0003 0.0004 0.0003 0.0005 0.0005 Pellet OD [ ] a.c 0.0002 0.0001 0.0001 0.0000 0.000 1 Clad ID [ ] a.c 0.0002 0.0002 0.0000 0.0001 0.0002 Clad OD [ ] a.c 0.0008 0.0008 0.0009 0.0009 0.0007 Pin Pitch +0.0014 0.0028 0.0032 0.0016 0.0015 0.0013 Vertical Cell

-- 0.0032 0.0049 0.0016 0.0014 0.0015 Pitch Horizontal

- - 0.0035 0.0050 0.0016 0.0014 0.0015 Cell Pitch Wall Thick - 0.007 0.0037 0.0043 0.0026 0.0025 0.0025 Cell ID 0.0002 0.0002 0.0003 0.0002 0.0002 BoraflexTM Sheathing - 0.003 0.0016 0.0016 0.0010 0.0009 0.0010 Thickness RSS 0.0069 0.0090 0.0040 0.0038 0.0038 NET- 28091-0003-0 1, Revision 0 94

The sign of the tolerance on Table 7.1 shows which direction increases k. No tolerance reactivity effect is calculated for Category 1 fuel , the Category 3 tolerance is applied for Category 1. A checkerboard of Category 1 fuel (see Section 8.1) has a large margin to the criticality safety limit so an approximate tolerance is appropriate.

PWR fuel assemblies are designed to be under moderated at power, so the moderator temperature coefficient is negative to prevent large power excursions. Therefore, increasing water between the fuel rods (and ignoring grids) increases k. This is demonstrated by calculations of the reactivity from varying the pin pitch and the fuel clad outer diameter (shown in Table 7.1) . The grids are conservatively ignored since they displace water around the fuel pins. The fuel pin pitch tolerance (0.0014 inch) used in this analysis is the maximum pin separation possible before the assembly gap becomes zero and all pins in the core are separated by a single enlarged pin pitch.

The fuel enrichment used for determining if the loading requirements are met is the as-built enrichment for each assembly. The uncertainty in the as-built enrichment is +/-0.02% . Note that the uncertainty in the as built enrichment is less than the traditional uncertainty of 0.05% which is based on the nominal enrichment. The reactivity of the fuel enrichment uncertainty is larger at low enrichments.

Calculations show that the reactivity due to enrichment uncertainty for Category 4 fuel is 0.0028 Llk at 2.0 w/o and is 0.0008 L'lk at 5.0 w/o. Fuel Categories 1, 2, and 3 have fixed burnup requirements at an enrichment of 5.0 w/o, so there is no enrichment uncertainty. For Category 4 fuel, the enrichment uncertainty reactivity effect is linearly interpolated using the two points 0.0028 L'lk at 2.0 w/o and 0.0008 L'lk at 5.0 w/o. For Category 5 fuel, the enrichment uncertainty at 5.0 w/o is used because Category 5 fuel has a fixed bumup differential which would have the least margin at 5.0 w/o.

A tighter (smaller) rack cell pitch increases ketr of the SFP for both regions because the fuel assemblies are closer together. The Region 1 change in kerr for reducing the cell pitch is much larger due NET- 28091-0003-01 , Revision 0 95

to the decrease in the flux trap. In ca lculating the reactivity effect of decreasing the cell ID, the cell pitch is maintained, so the effect on keff is small.

A conservative minimum width of the Boraflex TM sheathing is used for the model. Calculations are performed where the sheathing width is increased from 7.7 to 8.0 inches in Region 1 and Region 2. The keff decreases by 0.0005 t.k and 0.0004 t.k for Regions 1 and 2, respectively.

Calculations show that the highest reactivity occurs with elevated temperatures (see Section 8. 1).

This is due to the water hole in the 2x2 model holding down reactivity. A higher temperature in the water hole results in a lower water density and the reactivity hold down is reduced. For Region 2, the peak reactivity occurs at a temperature of 70 °C. Above 70 °C, the reactivity begins to decrease because the reduced moderation from lower density water within the fuel array is dominated by the reactivity hold down of the water in the water hole. For Region 1, the reactivity increases with increasing temperature all of the way to boiling. Therefore, for Region 1, the water temperature is modeled at 180 °F, the SFP design basis maximum temperature during normal operation. Boiling conditions are analyzed as an accident where soluble boron credit can be used.

No calculations of tolerances or sensitivities were made with borated water. The borated conditions have excess margin , which covers any differences in sensitivity with borated water.

7.2 Burnup Dependent Biases and Uncertainties Bumup increases the uncertainty in the analysis of k. To account for this there are several biases and uncertainties that are bumup dependent. They are the depletion uncertainty, the minor actinide and fission product bias, the bumup uncertainty, the clad creep bias, and grid growth bias.

The first bumup dependent bias or uncertainty is the depletion uncertainty in the atom densities. This is accounted for by an uncertainty of 5% of the t.k between the zero bumup case and the case at the desired bumup. The 5% has been supported by a number of studies mentioned in Section 4 and is NET- 28091-0003-01 , Revision 0 96

recommended via DSS-ISG-2010-01 [5]. For fuel Categories 2, 3, 4, and 5, the zero burnup kefffor the enrichment of interest is calculated and used with the calculated keff of the burned case to determine the worth of the depletion uncertainty. Note that no bias is applicable. As an example, at 49.5 GWd/T, the delta-k of depletion is 0.3257 ~kin the 3-out-of-4 arrangement of Region 2 (Category 4 cell) . This makes the depletion uncertainty 0.05 x 0.3257 ~k = 0.0163 t.k.

The second burnup dependent bias or uncertainty is the minor actinides and fission product worth bias. This bias, previously mentioned in Section 4, covers the bias and uncertainty due to the lack of criticality data for the minor actinides and fission products . This bias is detennined by calculating kerr in the appropriate model with the 1ninor actinides and fission products removed . The difference in reactivity between the calculations with and without these isotopes is multiplied by 1.5% and included as a bias .

This approach was suggested in NUREG/CR-7109 and conservatively covers the uncertainty [22). As an example, at 49.5 GWd/T for Category 4 fuel , a calculation determined that the actinide and fission product worth is 0.1371 ~k. Therefore the bias at 49.5 GWd/T is 0.0021 ~k.

The third burnup dependent bias or uncertainty is the uncertainty in the declared burnup from the reactor records (shortened to burnup uncertainty). The burnup uncertainty from the reactor records is assumed to be 5% of the burnup [26]. This value is based on comparisons presented in Section 7 .2 of NUREG/CR-6998 [26] of in-core measured burnups that demonstrate that the uncertainty in utility-assigned burnup values is less than 5%. The effect on reactivity is calculated by comparing the kerr calculated for the same case at two different burnups. For example, at 5.0 w/o the keffat 49.5 GWd/T is 0.9560. The keff at 0.9x49 .5=44.55 GW d/T is 0.9813 (10% less burnup ). So the~ due to a 5% burn up uncertainty at 49.5 GWd/T is :

(0 .9813 - 0.9560) I 2 = 0.0127 ~k The bumup uncertainty changes as a function ofbumup and enrichment because the delta-k between two low bumups can be larger than the delta-k between two high bumups. The burnup uncertainty is NET- 28091-0003-01 , Revision 0 97

calculated for each loading curve point by the same procedure (us ing the L'lk between two different burnups with all other parameters staying the same) .

As developed in Section 6.3.1, the clad outer diameter reduction (in microns) due to clad creep is a function ofburnup, starting at Omicrons at O GWd/T, linearly increasing to 100 microns at 40 GWd/T, then linearly decreasing to zero at 58 GWd/T. To determine the reactivity effect at the maximum creep of 100 microns, a special depletion was run with the clad outer diameter reduced by 50 microns (the average reduction between O and 40 GWd/T) . A case was then run at 40 GWd/T with the clad OD reduced by 100 microns using the special depletion for the number densities . When compared to the nominal case at 40 GWd/T, the delta-k at 100 microns is 0.0012 . The clad creep bias (L'lk) is expressed as a function of burnup as fo ll ows:

Clad creep bias (L'lk) = 0.00 12 x BU/40 BU ::S 40

= 0.0012 x (58 - BU)/ 18 BU > 40 As can be seen in Table 7 .1 , the sensitivity to the clad OD is similar for all categories of fuel so the same clad creep bias formulation is used for all categories.

Finally, as discussed in Section 6.3 .2, the grid growth as a percent of the grid cross section is a function of burnup. The tolerance calculation for pin pitch uses 0.0014 inch for the pin pitch tolerance which is 0.25 % of the pitch. The grid growth bias is :

Grid Growth Bias (L'lk) = (0.0000043 x BU 3 - 0.00013 x BU2 + 0.0051 x BU) x 4 x pp where pp = pin pitch tolerance worth from Table 7.1.

NET- 28091-0003-01, Revision 0 98

7.3 Eccentricity Generally the plant intends to place the fuel assembly in the center of the cell. However, it is acceptable to have the assembly in any location within the cell. A study performed for the Millstone 2 license application showed for that plant, the placements were approximately random [49]. For this 2017 CSA, it is assumed that the placement of the assemblies in the cells is random (there is nothing in the cell that would cause the assembly to be preferentially placed in one comer over another) . As was performed for Millstone 2, the number of assemblies that is eccentrically placed in particular quadrants is determined such that the probability of such placement is less than 5% over the lifetime of the plant.

Region 2 contains a 3-out-of-4 set of Category 4 cells surrounded by Category 5 cells in a 4-out-of-4 arrangement on the outer two rows and a 4-out-of-4 arrangement with checkerboarded control rods. To determine an eccentricity bias for the Category 4 cell arrangement, 16 assemblies are placed as close as possible (using the standard pin pitch) to a central assembly. Figure 7.1 shows the placement of the assemblies. The probability of 16 assemblies being randomly placed in the most reactive quadrant is 0.25 16 = 2.3*E-10. There are 504 Category 4 cell locations . It is conservative to estimate that each location could have 100 moves. This makes the probability of getting such an arrangement 2.3*E-10*504*100= 1.2E-5 which is much less than the required 0.05.

NET- 2809 1-0003-01 , Revision 0 99

- ~ - -- - - - - - - - - - - - - - - - -- -- ----,

Figure 7.1: Category 4 Region 2 with 16 Assemblies Eccentrically Placed The model used is an 8x8 model with periodic boundary conditions, where all assemblies except the central 16 eccentric assemblies about a central assembly are centered. This means that there are actually about eight eccentric sets separated by two rows of centered assemblies. The enrichment and bumup used for the fuel is 5.0 w/o and 48 .19 GWd/T. Table 7.2 shows the results of the ana lysis. The reference calculation is the 8x8 model with centered assembli es. The eccentricity bias for Region 2 is negative and therefore conservatively ignored. This is not unexpected since water holes or contro ls rods can break up the effect of eccentricity.

NET- 28091-0003-01, Revision 0 100

Table 7.2: Eccentricity Results Category Calculated k Sigma Ak Reference 4 0.9586 0.00004 16 Eccentric Assemblies 4 0.9585 0.00004 -0.0001 Reference 2 0.9671 0.00007 16 Eccentric Assemblies 2 0.9690 0.00006 0.0019 16 Eccentric Assemblies Shifted down 2 0.9693 0.00007 0.0022 The bumup increment for the Category 5 fuel is determined such that Category 5 fuel is not limiting (see Section 8.4.2) . Therefore, the eccentricity is maximized by moving Category 5 fuel closer to the Category 4 cells. In the final full pool model all of the Category 5 fuel assemblies in Region 2 are moved right in their cells to be as close as possible to the left hand side of Category 4 cells.

Region 1 is more complicated. First, a checkerboard area with Category 1 fuel is designed to have a low keff- The keff is sufficiently low (0.8548) that eccentricity within Category 1 cells is not a concern.

Category 2 fuel eccentricity is analyzed with 16 assemblies placed closest to a central assembly in the same manner as was performed for Category 4 (Region 2). Due to the flux trap, however, it is possible that it is more reactive to move the central assembly from the center of the cell toward one of the four sides. Therefore, analysis was also performed where the central row of assemblies is moved down to be closer to the row of assemblies below it. Figure 7.2 shows this arrangement of fuel assemblies. The results of the analysis for Category 2 are shown on Table 7.2. Category 2 is showing an eccentricity bias and the bias is slightly larger when the central row is moved down. The full pool model incorporates the eccentric placement of the assemblies, so the bias is intrinsic to the analysis. Therefore, it does not need to be added to the final calculated k.

As in Region 2, the bumup penalty for the outer two rows of Region 1 (Category 3 cells) was selected to prevent Category 3 fuel from being more limiting. In the final full pool model, both rows of Category 3 fuel assemblies are moved down toward the Category 2 (or 1) cells to maximize the NET- 28091-0003-01, Revision 0 101

eccentricity/interface effect. Similarly the Category 5 fuel assemblies at the Region 1/Region 2 interface are moved up and to the left to make them as close as possible to Category 2 cells.

1** ** *** ** ** f........ -~ .. .

-- ******~***~***~I --*--*-- - r===============

.::.:::.::.:: I

---

-------- ===============1-

.. .............. ::::::::::::::: i::::::::::::::: ::::::::::::::: ::=::::::=::=::;

............. ::.::.:*:.::.:: .. .. ... .. .. ::.::.:*:.::.:: 1******* .......

'::::::::::::::: i

.. ......... .. 1***************

• • • *** *** *** .. ......... .. 1::.:::::::::.::)

• • • • ***** **** *1  : . : . : : . * . : .*:* * : , : . * . : . : : . : . * *** ***** **** * ******** ****

..,J: :** ******* ** i ::=::=::::::=::

• • • • • • • • • • • • • j

.. 1 ** .........

=::=:::=::=:: ..

• ....*.....*... ***************1 1

=::=:::=::=:: *~::*::*:.:*::*::j

• • • 1 i::::.::*::.:::: !

'-*------***-*-*****+

....... .. **' , --~---~----- --------+--* ********-.-;-~..-;-~~-------

=:::::::

1 1.**............

!::::::::::::::: i' 1

=:::::::::=::

ee

!I i::*::*:.:*::*::

1i:*:*::*:.:*::*::

I

• 1

....------------- _____ 1***************'

• ..**......*.*. .*.*....*.**... !***************- ...*...*....... *1***************i*.

---

~------

---

*-r**********:i

....... ::::=:::::=:::::

.*.............  ::::=::::::::::

• 1

=:::::=::::

..*............  ::::=:::::::::: i*::::=:::::=::::

... .*. **. 1***************

1**

1*************** ..* ... **

....*......**... 1

.... ,***************1

• • • • • • • • • ****1 **** ***** **** **** ***** **** ****

• • ******* ******** 1*.**........*****

• • ********* ** ** ********* ** ** ********* ** * ********** ** 1

• .*.**.*.*.*.*.**.*.**.*.1.* .......

i I

1**************** ***********.**. I***************.!

• * * * * * * * * * * * * * * ' * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 11.

~ :::::::::::::::! t. : . : . : ** * : . :*  :*: : . :*** : *:.::. * !I**

:::::::::::::: r1 I::=:::::::::=:: - ** *** ** ** ,

1::::*::.::*:::: i

i:::::::::::iil I:::*:::*:::*:::

I::*:::::::::*::

t!::::::::::::ii !

*:::::::::*:: j

• • • • ***** *****  !**** ***** **** i**** ***** ****

• • .:*::*::; i:::::::::::::::  :: :::::::::::::: !

~::::~:;--------~~~~~-~~~~~~~--------*-------t************** '

............... i

................ ............... 1***************: 1  :

..;:... .............. ::=:::::::::=::

i::*::*:::*::*::  ::*::*:::*::*::  ::*::*:::*::*:: !

I::::::::::::::: ::::::::::::::: I::::*::.::*:::::

~=.. ............

............. 1 ...............  :::.:::.:::.::: I::::::::::::::::

I

=:::::::::=:: I**:****

• *************** ! , :::*:::*:::*::: *************** 1* * ********* **

• • ********* ****** 1 I** **** ***********1

.. ***************1 ...............

• • .:*::*:: I

!:::::::::::::::  :::::::::::::::  ::*::*:.:*::*::1

-1' **************~---- *************** I I

~:_;;;:::;;;_;;;;L_ f Figure 7.2: Eccentric Model for Category 2 with Central Row Shifted Down 7.4 Additional Biases and Uncertainties The criticality validation for major actinides, absorbers, and structural materials provides a bias and uncertainty due to validation. As provided in Section 4, the validation bias and uncertainty for all calculations except those with very hard spectra such as borated cases or boiling cases are 0.0021 Lik and NET- 28091 -0003-01 , Revision 0 102

0.0087 t.k respectively (for fresh fuel the bias and uncertainty is less as given in Section 4.3). For cases with a hard spectrum (EALF greater than 0.4 eV) the bias and uncertainty are 0.0027 t.k and 0.0112 t.k respectively.

The criticality validation also revealed the need for an additional bias and uncertainty for conditions above room temperature. Due to using water holes to control reactivity, it was found that elevated temperatures have a higher reactivity. For Region 2, the highest reactivity is at a temperature of 70 °C (see Section 8.7). For Region 1, the highest reactivity is at the SFP design basis maximum temperature which is 180 °F (82 °C) . From the validation section, the temperature bias at 70 °C is 0.00043 t.k with an uncertainty of 0.0013 t.k while the temperature bias at 82 °C is 0.00053 t.k with an uncertainty of 0.0013 t.k.

A 2x2 model cannot provide adequate modeling of eccentric placement of fuel assemblies in rack cells. Therefore, a bias is required. This bias is detennined in calculations performed in Section 7.3. It was found that eccentric loading of Category 4 cells (3-out-of-4 of Region 2) does not increase reactivity, so there is no eccentricity bias for Category 4 fuel. The Category 5 fuel bumup penalty, which is derived in Section 8, includes eccentric positioning in the large model, so it is inherently included in the bumup penalty. However, Category 2 (3-out-of-4 in the Region l flux trap design) does have an eccentricity bias of0.0022 t.k (see Section 7.3).

The final additional bias and uncertainty is the reactivity effect due to Monte Carlo statistical uncertainty. The 95/95 Monte Carlo statistical uncertainty in each tolerance calculation is where <:rb and crp are the Monte Carlo standard deviations for the base case and the perturbed case, respectively. The base case calculation was run with 1.024 billion histories to reduce the statistical uncertainty to+/- 0.00002 t.k (1 sigma) for the base case. The perturbed calculations are run for 64 million NET- 28091-0003-01, Revision 0 103

histories for an uncertainty of+/- 0.00008 ~k ( l sigma). This makes the Monte Carlo standard deviations

((2*0.00002) 2 + (2*0.00008) 2 ) 05 = 0.000 16 ~k. In the statistical combination of terms each of these would be squared. Since there are 13 statistical tolerance components the sum of these terms would be 13*(0.00016)2. The fina l step of the statistical combination is taking the square root. The square root of 13 *(0.00016) 2 is 0.0006 ~k. So, 0.0006 ~k is the Monte Carlo statistical uncertainty in the tolerance calculations and is combined in the total rack up of uncertainties.

7.5 Biases and Uncertainties Rack-up Sections 7.1 through 7.3 provides the biases and uncertainties and in the case of the bumup dependent biases and uncertainties, how to calculate them. For Region 1, the fuel categories all have fixed bumup and are valid up to 5.0 w/o enrichment. Therefore, the final bias and uncertainty can be determined.

Table 7.3 provides the total rack up of biases and uncertainties, including the statistical combination of the uncertainties for Region 1 (fuel Categories 1, 2, and 3).

NET- 2809 1-0003-01, Revision 0 104

Table 7.3: Total Bias and Uncertainties for Region 1, Categories 1, 2, 3 Category 1 2 3 Arrangement 2 of 4 3 of 4 4 of 4 Enrichment (w/o) 5 5 5 Burnup (GWd/T) 0 21 28.5 Component Bias Uncertainty Bias Uncertainty Bias Uncertainty Validation (critical 0.0024 0.0035 0.0021 0.0087 0.0021 0.0087 experiments)

Depletion uncertainty - - - 0.0067 - 0.0092 Minor Actinides and Fission

- - 0.0010 0.0013 Products Bias Bumup uncertainty - - - 0.0057 - 0.0081 Manufacturing Tolerances - 0.0090 - 0.0069 - 0.0090 Enrichment Uncertainty - - - - - -

Monte Carlo statistics - 0.0006 - 0.0006 - 0.0006 Eccentricity Bias 0.0 - 0.0022 - o.o* -

Clad creep bias - - 0.0006 - 0.0009 -

Grid growth bias - - 0.0010 - 0.0018 -

Elevated temperature 0.0005 0.0013 0.0005 0.0013 0.0005 0.00 13 Total Rack Up (~k = RSS) 0.0029 0.0098 0.0074 0.0142 0.0066 0.0176 Sum of Bias and 0.0127 0.0217 0.0242 Uncertainties

• Category 3 cells are located on the outer two rows of Region 1 so eccentricity in an infinite model is not relevant. The eccentricity is part of the full pool model and therefore no bias is applied.

For the two categories of fuel in Region 2, the burnup is allowed to change so the bias and uncertainty will change. However, Table 7.4 is provided as an example of the total rack up for Category 4 and 5.

NET- 28091-0003-01, Revision 0 105

Table 7.4: Sample Category 4 and 5 Bias and Uncertainty Rack-up Cate2orv 4 5 Arrangement 3 of 4 4 of 4 Enrichment (w/o) 5 5 Burnup (GWd/T) 49.5 60.5 Component Bias Uncertainty Bias Uncertainty Validation (critical experiments) 0.0021 0.0087 0.002 1 0.0087 Depletion uncertainty - 0.0163 - 0.02 14 Minor Actinides an d Fission Products Bias 0.0021 0.0026 Burnup uncertainty - 0.0124 - 0.0 155 Manufacturi ng Tolerances - 0.0040 - 0.0040 Enrichment Uncertainty - 0.0008 - 0.0008 Monte Carlo statistics - 0.0006 - 0.0006 Eccentricity Bias 0.0 - o.o* -

Clad creep bias 0.0006 - 0.0000 -

Grid growth bias 0.0029 - 0.0041 -

Elevated temperature 0.0004 0.0013 0.0004 0.0013 Total Rack Up (~k = RSS) 0.0081 0.0227 0.0092 0.0282 Sum of Bias and Uncertainties 0.0308 0.0374

• Category 5 cells are located on the outer two rows of Region 2 so eccentricity in an infin ite model is not relevant. The eccentricity is part of the full pool model and therefore no bias is app lied.

Fresh 5.0 w/o fue l with 64 IFBA (and any burned fuel) can also be stored in a checkerboard pattern of assemb li es and water holes in Region 2 and can be stored in the 3-out-of-4 area of Region 2 if it contains a control rod. The rack up of uncertainties that appli es to fresh 5 .0 w/o fuel in Region 2 is shown in Table 7.5.

Table 7.5: Total Bias and Uncertainty for Fresh Fuel in Region 2 Component Bias Uncertainty Validation (critical experiments) 0.0024 0.0035 Manufacturing Tolerances 0.0040 Fuel enrichment -

Monte Carlo statistics 0.0006 E levated temperature 0.0004 0.0013 Total Rack Up (~ k = RSS) 0.0028 0.0055 Sum of Bias and Uncertainties 0.0083 NET- 28091-0003-01, Revision 0 106

None of the above tolerance calculations were performed under borated conditions. The reason is because the borated condition has significant margin. The boron di lution analysis of record shows that a dilution down to 786 ppm is not credible [52]. However, the minimum ppm selected for the borated analysis is 700 ppm. Any small increase in the tolerance uncertainties would be covered by this 86 ppm margin in addition to the large margin from 0.95 reported in Section 8.5 .

7.6 Interface Uncertainty Treatment When analyzing a fu ll pool, the calculated keff will come from the most reactive area of the SFP.

However, when the uncertainty is not the same in all areas, the analysis may not correctly find the most limiting k. In order to address this concern, the bumup of the high burnup regions are adjusted down to account for the difference in the uncerta inty. Tables 7.3 and 7.4 show the uncertainties for the 3-out-of-4 and the 4-out-of-4 areas in Regions 1 and 2. For Category 3 fuel to match the bias and uncertainty of Category 2 fuel, a burnup reduction is required to match the 0.0048 6k difference in bias and uncertainty.

The 0.0047 6k is the sum of0.0025 6k between Category 2 and Category 3 fuel (see bottom of Table 7.3) plus the 0.0022 6k eccentricity effect for Category 2 fuel which will be included in the full poo l model.

To account for this additional uncertainty, the burnup for Category 3 fuel is decreased by 0.8 GWd/T in the full pool calculations (the reactivity due to burnup at 28.5 GWd/T is 0.6% in kefffor every 1 GWd/T burnup). Similarly for analysis of Region 2, the Category 5 fuel bumup must be decreased to match the reactivity difference in the bias and uncertainty between Category 4 and Category 5 fuel. From Table 7.4 this difference is 0.0066 6k. To account for this additional uncertainty, the burnup for Category 5 fuel is decreased by 1.3 GWd/T in the full pool calculations (the reactivity change due to burnup at 60.5 GWdff is 0.51 % in keff for every 1 GWdff bumup).

NET- 2809 1-0003-01, Revision 0 107

8 Results With the biases and uncertainties determined, the minimum loading requirements can be ca lculated.

These minimum loading requirements meet the 10CFR50.68 requirements. Specifically, k 95;95 must be less than 1.0 with no soluble boron credit and less than 0.95 with credit for soluble boron. For this analysis, these limits are met while maintaining about a 1% margin in k. It has been demonstrated that for all unborated cases k 95;95 is less than 0.99 and for the borated cases k 95;95 is less than 0.94 after adding biases and uncertainties .

8. 1 Temperature Effects The criticality analysis must cover the full range of temperatures allowed in the SFP. Rather than perform the criticality ana lysis at a reference temperature and add a bias, the criticality analysis is perfonned at the most limiting temperatures. Table 8.1 summarizes the Region 1 and 2 (3-out-of-4 area) calculations at 12 different temperatures (4, 10, 20, 30, 40, 50, 60, 70, 80, 90, 95 , and 99 °C). From the SCALE validation, there is a temperature bias of 0.0000086 for each °C above 20 °C. For example, the bias at 60 °C is 0.00034, while the bias at 70 °C is 0.00043 , and the appropriate temperature bias is added to the calculated ketr values. The results demonstrate that the bias-corrected reactivity is largest at 70 °C for Region 2 and at 99 °C for Region 1 under unborated conditions. Except for Table 8.1 and the over-temperature accident, all calculated ketr va lues for Region 1 are performed at 180 °F (82 °C), which corresponds to the SFP design basis maximum temperature for the IP2 SFP. In developing the loading curves for Region 2, all of the calculated ketr values are performed at 70 ° C which is the most reactive temperature.

NET- 2809 1-0003-01, Revision 0 108

Table 8.1: Calculated keff as a Function of Temperature Temperatu re Density Region 1 Region 1 Region 2 Region 2 (OC) (glee) ealc. k adj . k ealc. k ad j. k 4 1.0000 0.96386 0.96386 0.95569 0.95569 10 0.9997 0.96388 0.96388 0.95575 0.95575 20 0.9982 0.96372 0.96372 0.95545 0.95545 30 0.9957 0.96508 0.965 17 0.95582 0.95591 40 0.9922 0.96628 0.96646 0.95595 0.95612 50 0.9880 0.96745 0.96771 0.95610 0.95636 60 0.9832 0.968 38 0.96873 0.95606 0.95641 70 0.9778 0.9693 1 0.96974 0.95599 0.95642 80 0.9718 0.97000 0.97052 0.95577 0.95629 90 0.9653 0.97089 0.97149 0.95564 0.95624 95 0.9619 0.97123 0.97187 0.95552 0.95616 99 0.9591 0.97156 0.97224 0.95544 0.95612 8.2 Region 1 Fuel Categories 1 and 2 Using the 2x2 model described in Section 6 with the atom densities developed in Section 5 and with the bias and uncertainties established in Section 7, the loading requirements for Region 1 fuel Categories 1 and 2 are determined. Table 8.2 shows the fuel requirements and the calculated k - for a checkerboard arrangement of fuel in Region 1 (Category 1 cells) and a three out of four arrangement of fuel in Region 1 (Category 2 cells). Category 1 fuel is designed to require 64 IFBA rods in 5.0 w/o fuel assemblies. To provide flexibility in fuel design, the number of IFBA for fresh fuel can be reduced for lower enrichments*. The number of minimum IFBA for fuel less than or equal to 5.0, 4.5 , 4.0, 3.5, and 3.0 w/o is 64, 48 , 32, 16, and 0, respectively. The kefffor these cases are all less than the keffwith 64 IFBA 5.0 w/o. No credit for burnup is taken. For Category 2 fuel , the fuel must be burned at least 21 GW d/T and the maximum enrichment is 5.0 w/o. The Category 2 bumup requirement of 21 GW d/T is based on Batch Z fuel having an 8 inch 4.0 w/o U-235 axial blanket where the relative burnup distribution is from Table 6.3. An axial blanket that is less than 4.0 w/o U-235 or more than 8 inches long is bounded by this analysis, but a shorter or higher enriched blanket is not.

• The IFBA requirement is only for Batch Z.

NET- 2809 1-0003-01, Revision 0 109

Table 8.2: Confirmation of Region 1 Requirements for Category 1 and 2 Fuel Fuel Catee.orv 1 2 Arrane.ement 2-out-of-4 3-out-of-4 Maximum Enrichment (w/o) 5.0 5.0 Minimum Burnup (GWdrf) 0 21 Minimum IFBA Rods 64 -

Calculated k 0.8548 0.9686 Bias and Uncertainty 0.0127 0.0217 k9S/95 0.8675 0.9903*

For burned fuel, no credit is taken for IFBA or any insert in the guide tubes , with the exception of full length RCCAs in designated areas. Calculations show that 5.0 w/o fuel with 64 IFBA rods has lower reactivity at all bumups compared to the BOC equilibrium Xe value. Table 8.3 shows the values of kerrin the core geometry as a function of bumup for fuel with various IFBA loadings. The orange shaded blocks are the bumups where kerr has increased over its initial equilibrium value. The 64 IFBA case is always less than the initial kerr(with equilibrium Xe) of 1.2109.

Table 8.3: Change in k ett with Burn up and number oflFBA Rods (Analysis performed at core conditions for 5.0 w/o fuel)

Number of IFBA Rods + 0 32 64 80 116 Burnup (GWd/T}. Calculated keff at core conditions 0.15 1.2989 1.2526 1.2109 1.1924 1.1 523 0.50 1.2925 1.2489 1.2086 1.1903 1.151 3 1.00 1.2855 1.2441 1.2064 1.1890 1.1 526 1.50 1.2801 1.2417 1.2069 1.1905 1.1 559 2.00 1.2757 1.2403 1.2068 1.1913 1.1 593 3.00 1.2684 1.2360 1.2062 1.1930 1.1 632 4.00 1.252 1 1.2249 1.1 992 1.1874 1.1 6 10 5.00 1.2423 1.2184 1.1963 1.1848 1.1 625 6.00 1.23 19 1.2 11 7 1.1 929 1.1827 1.1 631 8.00 1.2 167 1.2013 1.1 850 1.1782 1.1609 10.00 1.1974 1.1 862 1.1 750 1.1693 1.1563

• The infinite kerr exceeds 0.99 but the actual finite kerr for thi s region is less than 0.99 (0.9881) .

NET- 2809 1-0003 -01, Revision 0 110

Reactivity decreases with increasing category number. Therefore, fuel from any higher numbered category can be placed in any location that allows for a lower numbered category. For example, a fuel assembly categorized as Category 5 can be placed anywhere in the SFP. However, a cell in the SFP that requires Category 5 fuel may not contain a lower category fuel assembly. All of the historical fuel through Batch X of IP2 and Batch AA of IP3 has been categorized.

8.3 Region 2 Category 4 Batch Grouping Z - Current and Future Fuel The minimum burnup requirements (loading curve) for Category 4 fuel for Batch Grouping Z (current design and future fuel assemblies) are presented in Table 8.4. The SFP cells where Category 4 (or above) fuel is required is shown on Figure 1.1 as the green shaded cells in Region 2. The other batch groupings are analyzed separately, and the results are presented in Section 8.6.

Table 8.4: Minimum Burnup Requirements (GWd/T) for Category 4 Batch Grouping Z Enrichment* Cooling Time (years) PF=l.2 (w/o) ot 1 2 5 10 15 25t 4.2 40.27 39.69 38.92 37.23 35.13 33 .75 32.20 4.6 44.27 43.60 42 .83 40.99 38.71 37.25 35.52 5.0 48.19 47 .52 46.61 44.67 42.30 40.71 38.85 Enrichment* Cooling Time (years) PF=0.80 (w/o) ot 1 2 5 10 15 25t 4.2 38.67 38.11 37.48 36.02 34.12 32.96 31.59 4.6 42.60 41.97 41.31 39.70 37.72 36.45 34.90 5.0 46.52 45 .84 45 .16 43 .39 41.14 39.77 38.05 Table 8.4 provides the burnup requirement in GWd/T as a function of initial U-235 enrichment and cooling time for two different peaking factors. For each assembly, the peaking factor is known, and the

• The enrichment to be used is the enrichment of the center section between the blanket material.

t O years cooling is actually 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> . This is the cooling time that maximizes k.

t Fuel coo led to more than 25 years must use the 25 year burnup requirement.

NET- 28091-0003-01 , Revision 0 111

bumup requirement for that assembly can be interpolated between 0.8 0 and 1.20. Over 95% of the fuel inventory has peaking factors between 0.80 and 1.20. Extrapolation above 1.20 and below 0.8 is also acceptable because this has been shown to be conservative (see Section 6.4). After adjusting for an assembly's peaking factor, if an assembly fails the load ing curve, it can be stored anywhere in Region I as long as the bumup is greater than 28.5 GWd/T.

Table 8.4 can be linearly interpolated to find the required bumup at any enrichment/cooling time combination but it is recommended that the curve fit be used instead, as described in Section 8.2. 1. As discussed later, the bumup requirements are adjusted if the assembly contained a hafnium insert or has any fuel pins removed.

Fresh assemblies with at least 64 IFBA rods and an inserted control rod are Category 4. Analysis of an infinite system of Region 2 cells in a 3-out-of-4 arrangement using 5 .0 w/o fresh fuel with 64 IFBA rods and a control rod in every assembly produced a keff of 0.9603 . To be conservative, the control rod atom densities were reduced by 10%. The bias and uncertainty for this case is 0.0083 (see Table 7.5) , so k9s19s is 0.9686, which is well below 0.99. To provide flexibility in fuel design, the number of IFBA for fresh fuel can be reduced for lower enrichments. The number of minimum IFBA for fuel less than or equal to 5.0, 4.5, 4.0, 3.5, and 3.0 w/o is 64, 48 , 32, 16, and 0, respectively. The keff for these cases are all less than the keff with 64 IFBA. As discussed earlier, the reactivity of fuel with 64 IFBA decreases with bumup so a fresh assembly is more limiting than the same assembly having a small amount ofbumup.

NET- 28091-0003-01 , Revision 0 112

8.3.1 Curve Fit The data points of Table 8.4 have been fit with a nine parameter curve having the following form:

Minimum Bumup Requirement where E = U-235 initial enrichment (w/o)

CT = cooling time (years) a1 - a9 = fitting coefficients No extrapolation is allowed, so fuel at enrichments less than 4.2 w/o must use 4.2 for the enrichment, and fuel cooled more than 25 years must use 25 for the cooling time. The curve is purposefully conservative in that the minimum burnup requirement generated from the curve is always equal to or greater than the bumup shown in Table 8.4. The coefficients are shown in Table 8.5.

Several curve fits were attempted but this curve fit matched the data with the least amount of conservatism while being well behaved between the data points of Table 8.4. A spreadsheet for the fit was created to ensure the intermediate points follow the expected behavior. The exponential term in the fit is needed to mimic the physics of radioactive decay.

Table 8.5: Curve Fit Coefficients for Category 4 Fuel (Group Z - Current and Future Fuel)

Coefficient PF= o.so* PF= 1.20 a1 15.1405 -6.26824 az -4.8 11 33 5.29367 a3 0.753855 -0.37154 a4 0.121 252 0.129582 as -0.0150991 -0.0204918 a6 0.00 127009 0.00205596 a1 -16.2293 -0 .13331 as 14.0159 6.9037 a9 -0.687054 0.122068

• Only two peaking factors are needed (0 .8 and 1.2). Calcul ations show that extrapolation below 0.8 and abo ve 1.2 is conservati ve for all other peaking facto rs.

NET- 2809 1-0003-01 , Revision 0 113

8.3.2 Confirmation Calculations for Category 4 To ensure that all bumup/enrichment/cooling time combinations given in the loading curve meet the criticality requirements, each loading curve bumup/enrichment/cooling time point was run in the 2x2 KENO model to verify that each point meets the criticality requirements . The calculated keff values are shown in Table 8.6.

Table 8.6: Calculated k cff Values at each Category 4 Batch Z Burnup Point Enrichment Cooling Time (years) PF=l.20 (w/o) 0 1 2 5 10 15 25 4.2 0.9614 0.9611 0.9613 0.9606 0.9608 0.9608 0.9604 4.6 0.9600 0.9599 0.9597 0.9592 0.9594 0.9593 0.9590 5.0 0.9586 0.9581 0.9585 0.9579 0.9577 0.9576 0.9573 Enrichment Cooling Time (years) PF=0.80 (w/o) 0 1 2 5 10 15 25 4.2 0.96 12 0.961 3 0.9611 0.9605 0.96 11 0.9608 0.9602 4.6 0.9599 0.9600 0.9596 0.9591 0.9590 0.95 88 0.9585 5.0 0.9584 0.9585 0.9578 0.9577 0.9584 0.9579 0.9579 The total uncertainty is the bias plus a statistical combination of all of the uncertainties (see Section 7) . These total uncertainties are shown in Table 8.7.

Table 8.7: Total Bias and Uncertainty at each Category 4 Batch Z Burnup Point Enrichment PF=l.20 w/o 0 1 10 15 25 4.2 0.0278 0.0277 0.0277 0.0274 0.0273 0.0272 4.6 0.0292 0.0292 0.0293 0.0292 0.0291 0.0289 0.0288 5.0 0.0307 0.0308 0.0307 0.0306 0.0306 0.0306 0.0304 Enrichment PF=0.80 w/o 0 1 5 10 15 25 4.2 0.0277 0.0276 0.0276 0.0275 0.0274 0.0273 0.0272 4.6 0.0292 0.0292 0.0293 0.0293 0.0292 0.0290 0.0290 5.0 0.0307 0.0307 0.0308 0.0307 0.0306 0.0306 0.0305 NET- 2809 1-0003-01, Revision 0 114

After adding the total uncertainty to the calculated keff values, all points are less than 0.99 as shown in Table 8.8.

Table 8.8: k9st9s for each Category 4 Batch Z Burnup Point Enrichment Coolin!?: Time (vears* PF=l.20 (w/o) 0 1 2 5 10 15 25 4.2 0.9892* 0.9889 0.9890 0.9882 0.9882 0.988 1 0.9876 4.6 0.9892 0.9892 0.9890 0.988 5 0.988 5 0.9882 0.9877 5.0 0.9893 0.9889 0.9892 0.988 5 0.9883 0.9882 0.9877 Enrichment Coolin!?: Time (vears\ PF=0.80 (w/o) 0 1 2 5 10 15 25 4.2 0.9889 0.9889 0.9887 0.988 0 0.9884 0.9880 0.9875 4.6 0.989 1 0.9892 0.9889 0.9884 0.9882 0.9879 0.9875 5.0 0.989 1 0.989 1 0.9886 0.9883 0.9890 0.9886 0.9885 8.4 Determination of Burnup Requirements for Categories 3 and 5 Cell Categories 3 and 5 utilize the neutron leakage at the edge of the SFP in order to remove the need for one out of four water holes used fo r cell Categori es 2 and 4. Since the edge of the SFP neutron leakage is used, a full pool model is required. Thi s full pool model also allows for calculation of the eccentricity effect and the impact of the interface between cell Ca tegories and Regions.

8.4. 1 Cell Category Layout in Region 2 The concept for Region 2 is to have a three out of four arrangement of fuel with two outer rows of fuel requiring an increased burnup but using the leakage at the edge to reduce the increased burnup requirement. However, IP2 had a set of spare control rods that could be used to reduce the number of water holes such that one less cask would need to be loaded. The control rods in an assembly do not have as large a negative reactivity as a water hole so the control rods needed to be two out of fo ur, rather than

• T he valu es in Tab le 8.8 do not always match the sum of Tabl es 8.6 and 8.7 due to ro und off, since each tabl e was developed using more signifi cant digits before ro unding fo r the tabl e.

NET- 28091-0003-01 , Revision 0 115

one out of four. Rather than make a separate fuel category for the cell area with control rods, this cell area is forced to be the same category as the two outer rows of assemblies (cell Category 5) . Because the checkerboard of control rods reduced kerr more than the two rows on the outside of the SFP, it is possible to reduce the amount of water required in the water holes. This is the reason for the location of the pink cells. Determining the location of the pink cells and the control rods required considerable iteration .

The control rod locations can be water holes since the negative reactivity of a water hole is greater than a Category 5 fuel assembly with a control rod. The control rod may not be placed in or removed while the assembly is in the control rod location. The control rod must be inserted into the assembly while the assembly is in a cell not requiring a control rod, and then the assembly with control rod can be moved into position. Likewise, the control rod may be removed only while the assembly is in a cell not requiring a control rod. It is permissible to remove a control rod at a Category 5 cell as long as all adjacent cells (eight cells) are water holes since this iso lates the assembly, but this is not the expected method.

Category 5 fuel must be placed on both sides of the Region 1/2 interface except in SFP locations J-31 and H-31 (alternate arrangements are allowed as discussed in Section 8.5) . This eliminates the interaction between the two Regions. Placing Category 4 fuel on the Region 1 side of the interface was tried, but it drew the reactivity to the interface with a slight increase in kerr and, therefore, was rejected.

8.4.2 Additional Burnup Requirements for Fuel Categories 3 and 5 The objectives in setting the loading requirements for Categories 3 and 5 are first to make Categories 2 and 4 more limiting than Categories 3 and 5, and second to make them simple. The simplicity is accomplished by making the loading requirements a constant bumup penalty (independent of enrichment, cooling time and peaking factor) to the fuel Category 2 and 4 requirements . The bumup penalty is the smallest fraction of the bumup when the bumup is highest. Therefore, the bumup penalty is determined using the highest bumup requirements in Category 2 and 4 so it is conservative for lower NET- 28091-0003-01 , Revision 0 116

bumups. To find the bumup penalty, the Category 3 and 5 bumups were changed holding the Category 2 and 4 bumups constant. Since the design objective keff for Region 1 is higher than that for Region 2 (due to lower bumup requirements), the Region 2 analysis must be performed with the fuel removed from Region 1, so that the reactivity is dominated by Region 2. The Region 1 analysis contains all of the fuel from both regions. Therefore, the Region 2 analysis is perfom1ed first so that the correct bumups in Region 2 are used while doing the Region 1 analysis.

To determine the additional bumup requirement needed for Category 5 fuel , analysis was performed for Region 2 where the Category 4 bumup is 49.5 GWd/T and the Category 5 bumup is varied. For these cases the enrichment for both categories is 5.0 w/o. For this model, all of the Category 5 fuel is shifted right to maximize the effect of the Category 4/5 interface. Figure 8.1 is a plot of the variation in keff with the Category 5 bumup. As can be seen from Figure 8.1 at lower bumups (such as 54 GWd/T) the Category 5 fuel is determining k. By 59 GW d/T, keff is controlled mainly by the Category 4 fuel and by 61 GWd/T the Category 5 fuel is no longer affecting the SFP k. Based on this information the increased burnup requirement (burnup penalty) for Category 5 fuel is selected as 11 GWd/T.

NET- 28091-0003-01 , Revision 0 117

0 .980 0 .975 0.970 "C

~ 0 .965 u

a 0 .960 0.955 0.950 +

54 55 56 57 58 59 60 61 62 Burnup of Category 5 Fuel (GWd/MTU)

Figure 8.1: Calculated kerr as a Function of Category 5 Burn up Using 5.0 w/o Fuel In order to confirm that using the burnup penalty determined from the highest bumup is conservative, the analysis was also performed using a lower bumup for Category 4 fuel. This model uses 4.2 w/o enriched fuel , 32 GWd/T for the bumup (25 years cooled and a 0.8 peaking factor) . Figure 8.2 shows the results of thi s analysis. As can be seen from Figure 8.2, by a burnup of39 GWd/T the burnup of Category 5 no longer matters. This would imply a penalty of7 GWd/T which means that using the penalty detennined at the higher bumup (11 GWd/T) is indeed conservative.

NET- 28091-0003-01, Revision 0 11 8

1.000 -- --

0 .995 - - -

0 .990 0 .985

.:..: 0 .980

~

... 0 .975

~

iii u 0 .970 0 .965 0 .960 ---

0 .955 0 .950 F 33 T

35 T

37 Burnup (GWd/MTU) 39 41 Figure 8.2: kcrr as a Function of Category 5 Burnup Using 4.2 w/o Enriched Fuel The same process is repeated for Region 1 where the categories of concern are Category 2 and Category 3. However, the burnup requirement for the 3-out-of-4 area of Region 1 (Category 2) is a fixed 2 1 GW d/T independent of enrichment, cooling time, or peaking factor. Figure 8.3 shows the results of the analys is to determine the burnup penalty. As can be seen fro m Figure 8.3 , Category 3 bumup greater than 27 .7 GWd/T has a very small impact on the full pool calculated k. A 7.5 GWd/T additional bumup requirement fo r Category 3 is selected. Thi s makes the Category 3 minimum burnup requirement 28.5 GWd/T.

NET- 28091-0003-01 , Revision 0 11 9

0.984 0.982 0.980

.:ii::

"C 0.978 GI

i"' 0.976 I.I iii 0.974 u

0.972 0.970 0.968 25 25.5 26 26.5 27 27.5 28 28.5 29 Category 3 Burnup (GWd/MTU)

Figure 8.3: Calculated ketr as a Function of Category 3 Burn up Using 5.0 w/o Fuel 8.4.3 Confirmation of k9s19s for Full Pool (includes Category 3 and 5)

Since the total combined uncertainty is higher in Region 2, the design objective kerr for Region 1 is higher than Region 2. This means that a full pool model will produce a kerr value driven by the higher reactivity fuel in Region 1 and not yield any infonnation for Region 2. Before determining the Region 1 kerr, it is desirable to confinn the Region 2 burnup penalty for Category 5. For these cases, the full pool is modeled with water holes in Region 1 so that the reactivity is driven by Region 2 fuel. The burnup for Category 5 fuel is reduced consistent with the difference in the bias and uncertainty between Category 4 and Category 5 fuel. As determined in Section 7 .6, a burnup reduction of 1.3 GW d/T for the Category 5 fuel is required. Calculations were performed for 5.0 w/o fuel at the loading curve (for current fuel) at a peaking factor of 1.2 for zero, two and 25 years of cooling time. The Category 4 burnups are taken from Table 8.4. The Category 5 burnups are 11-1.3 = 9.7 GWd/T higher. Table 8.9 shows calculated k's and the k9s195 ' s for three 5.0 w/o cases at a peaking factor of 1.20. The bias and uncertainty used for Table 8.9 come from Table 8.7 . Table 8.9 shows that the loading criteria of Category 4 plus 11 GWd/T for NET- 2809 1-0003-01 , Revision 0 120

Category 5 meets the k(<0.99) criteria. As expected, there is sli ghtl y more margin for the 2 year cooled case than the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooled case and even more margin for the 25 year cooled case. This is because the burnup requirement decreases with cooling time and so the fixed burnup penalty of 11 GWd/T is a larger fraction of the burnup requirement. It is conceivable that a higher peaking factor mi ght result in less margin. An additional calculation was performed using a peaking fac tor of 1. 3. The highest peaking fac tor fo r fuel in the IP2 SFP meeting the Category 4 burnup requirements is 1.272 . The calculated k95195 for a peaking factor of 1.3 is 0.9897.

Table 8.9: Region 2 Models at Loading Curve (Cat 5 is Cat 4 plus 11 GWd/T)

Category 5 Category 4 Burnup in Peaking Cooling Calculated Bias and Burnup Sigma k 9S/95 Model Factor Time k Uncertainty.

(GWd/T)

(GWd/T)

48. 19 57.89 1.2 72 ho urs 0.9584 0.00005 0.0307 0.9891 46.6 1 56 .3 1 1.2 2 yea rs 0.9578 0.00006 0.0307 0.9885 38.85 48 .55 1.2 25 years 0.956 1 0.00006 0.0304 0.9865 48 .6 1 58 .3 1 1.3 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 0.9588 0.00006 0.0309 0.9897 The model used fo r the Region 2 analysis has the Category 5 fuel eccentrically placed on the right hand side of each cell (the reacti vity effect of moving the Category 5 fuel to the ri ght was calcul ated and is worth 0.0008 ~k) . With this model, which picks up eccentricity as well as any interface effects, the maximum k9s19s is met with more than 1% margin to the regulatory limit.

Upon confirmation of the Region 2 loading requirements, the analysis of Region 1 can proceed by adding Category 2 and 3 fuel in Region 1 of the model. In order to compensate fo r the difference in bias and uncertainty between Category 2 and Category 3, the burnup of the Category 3 fuel is reduced. In Section 7.6, the diffe rence in uncertainties for 7.5 GWd/T is 0.8 GWd/T . The bumup requirement for Category 2 is 2 1 GWd/T and the burnup requirement for Category 3 is 28 .5 GWd/T. Reducing 28.5 GWd/T by 0. 8 means the modeled burnup for Category 3 is 27.7 GWd/T. Consequently, the k9s19s NET- 28091-0003-0 1, Revision 0 121

for the SFP is the calculated keff in the full pool model plus the bias and uncertainty for Category 2 (21 GWd/T) .

The Region 1 eccentricity can be in two forms; in the center of Category 2 cells or at the boundary of Category 2 and 3 cells. The highest keffis when the eccentricity is in the center of the Category 2 cell area of Region 1. Table 8.10 shows the results of the analysis of different eccentric options. The analysis uses the primary arrangement of Region 1 (shown on Figure 1.1) with all fuel at 5 .0 w/o enrichment and the burnups of Categories 2 through 5 fuel are 21 , 27.7, 48.19, and 57.89 GWd/T. For all but the last case on Table 8.10, all of the Category 3 and 5 fuel assemblies are centered (except the case where the eccentric positioning is at the Category 2/3 boundary where only those assemblies with the eccentric grouping are moved in) .

Table 8.10: Eccentric Options for Region 1 Case k Sigma .dk All Region 1 Assemblies centered in cells 0.9665 0.00007 Plus Category 3 and 5 fuel in Region 1 moved 0.9681 0.00007 0.0016 toward the center Also Plus 16 Eccentric at Category 2/3 Interface 0.9686 0.00006 0.0021 16 Eccentric in the middle of Category 2 plus 0.9687 0.00007 0.0022 Category 3 and 5 moved in toward Category 2 From Table 7.3 , the bias and uncertainty is 0.0194 Llk with the eccentricity bias removed. The full pool model contains the eccentricity and interface effects. The final k9s19s for the Figure 1.1 arrangement of Region 1 is 0.9687 + 0.0194 = 0.9881. Thus, for the primary arrangement of Region 1 the target 0.99 is satisfied, leaving more than 1% margin to the 10CFR50.68 limit.

The peaking factors used for Category 2 and Category 3 fuel are 0.9 and 1.2, respectively. These are based on an expected cycle length for the final cycle of 23 .8 GWd/T. There are no assemblies currently in the SFP that are Category 2. Until the final cycle, it is expected that low burned fuel will be returned to the core for more burnup. If the final cycle is 23.8 GWd/T, then the highest peaking factor that matches NET- 28091-0003-01, Revision 0 122

the minimum burnup requirements for Category 2 (21 GWd/T) is 21/23 .8 = 0.88 . Assemblies in that cycle will certainly have higher peaking factors, but they would then exceed the minimum burnup requirement. For example, if an assembly had a peaking factor of 1.4 and the cycle length is 23 .8 GWd/T, its burnup would be 23.8*1.4 = 33 .3 GWd/T which would greatly exceed the Category 2 burnup requirement. The reactivity effect of additional bumup is greater than the effect of higher temperatures during depletion that is caused by a higher peaking factor. For Category 3, the highest peaking factor that matches the minimum burnup requirements for 28 .5 GWD/T and a cycle length of 23.8 GWd/T can be calculated and is 28.5/23 .8 = 1.2. The calculated k using 0.9 and 1.2 peaking factors is conservative for any final cycle burnup greater than 23 .8 GWd/T. The final cycle may be cut short so additional calculations have been done for different cycle lengths. If the final cycle were only 20.3 GW d/T then the peaking factors would be 21/20.3 = 1.03 and 28.5/20.3 = 1.40. The calculation using peaking factors of 1.1 and 1.4 for Category 2 and 3, respectively, resulted in a k9s19s of 0.9895.

With cycle bumups of less than 20.3, no new Category 3 assemblies would be produced since the maximum assembly peaking factor is 1.4 at any point in the cycle (note that this is not bumup averaged).

The most reactive assemblies in the SFP, except for four assemblies , have a bumup of 40 GWd/T. For cycles less than 20.3 GWd/T, the maximum peaking factor of 1.4 is used for Category 2 fuel , and the Category 3 fuel is modeled as 40 GWd/T (5.0 w/o fuel) with a peaking factor of 1.4. The calculated k95195 for this case is 0.9888.

In conclusion, for all final cycle lengths , the calculated k95195 is less than the regulatory limit by more than 1% ink. Although it is not expected that there would be multiple short cycles leaving more reactive fuel in Region 1, a case was analyzed where the peaking factor of 1.4 is used for both Category 2 and 3 fuel using the minimum burnup requirements and 5.0 w/o fuel. This calculated k9s195 is only 0.9904. This exceeds the design objective but still provides 0.96% in k margin to the regulatory limit. These cases are summarized on Table 8.11.

NET- 28091-0003-01 , Revision 0 123

Table 8.11 : Maximum Full Pool k9st9s assuming Various Cycle Lengths Cycle Length Category 2 Category 3 Calculated Bias and Sigma k9S/95 (GWd/f) Peakin11: Factor Peakin11: Factor k Uncertainty

>23.8 0.9 1.2 0.9687 0.00007 0.0194 0.9881 20.3 - 23.8 1.1 1.4 0.9701 0.00006 0.0194 0.9895

<20.3 1.4 1.4 but 40 GWd!f 0.9694 0.00007 0.0194 0.9888 Multiple Short Cycles 1.4 1.4 0.97 10 0.00007 0.0194 0.9904 8.5 Alternate Arrangements for Region 1 A checkerboard arrangement of Category 1 fuel is much less reactive than the base arrangement of fuel in Region 1. The calculated kerr of unburned 5.0 w/o fuel with 64 IFBA rods checkerboarded in Region 1 is 0.8548 . Since this k is so low, it is permitted to replace an area of Region 1 with a checkerboard arrangement of Category 1 cells . In order to not create an increase in k due to interfaces, there are two rules for creating the area for Category 1 storage.

1. Each Category 1 cell must be face adjacent with at least three water holes .

2. Each Category 2 cell may not have more than one face adjacent to a Category 1 cell.

Given these constraints, three additional arrangements of fuel in Region 1 were analyzed to confirm the reduction ink with Category 1 checkerboards present. The first additional arrangement represents the expected arrangement in Region 1 prior to loading a new cycle. For this arrangement, fresh fuel is generally in the new fuel vault but some Category 1 fuel is needed to be in the SFP. This arrangement will be called the "Refueling Arrangement." The second additional arrangement covers the case where all of the Category 2 cells are removed but the Category 3 and 5 cells are still used. This arrangement will be called the "No Cat 2" arrangement. The final additional arrangement maximizes the number of Category 1 cells in Region 1. This arrangement is called "Max Cat l."

Figures 8.4 through 8.6 show these arrangements of Region 1. Figure 8.7 shows an arrangement of Region 1 that is not limiting but is useful to illustrate the Category 1 rules. The term "checkerboard" is NET- 2809 1-0003-01, Revision 0 124

used to describe the regular array but the key requirement is the fi rst rule requiri ng at least 3 face adjacent water holes. This requi rement is illustrated in Figure 8.7 showing only one, two , or three Category l cells. The arrangement of other Categories that are not replaced by Category l must not change even with the presence of any number of Category I areas. Although different arrangements of Category 2 cells are possible that would not violate the criticality requirements, the rules to allow these types of rearrangements wou ld be complex and prone to error and therefore are not allowed. In order to fairly compare the arrangement of the base case and the addi tional arrangements , these analyses were perfonned with all of the fuel centered in its cell. Table 8. 12 provides the calculated keff values for each arrangement.

2 3 4 5 6 7 8 9 W ll ll U U ll H V U U W li ll B M ll U V U ~ ~ ll H

G F

E D

C B

Figure 8.4: Refu eling Arrangement Figure 8.5: No Cat 2 Arrangement NET- 28091 -0003 -0 1, Revision 0 125

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 H

G F

E D

C B

Figure 8.6: Max Cat l Arrangement 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31

--- - - ----- - -t= Lj - - - ~

~- ~ 1-------- - -~

H G

• - -- - -~ - ,......,..,

~

F E

D

~

C

- ~

B A X Figure 8.7: Example Odd Arrangement Table 8.12: Dependence of kcff on the Region 1 Arrangement Arrangement ~k from Description k Sigma Number Reference 1 Reference - No Category 1 0.9665 0.00007 2 Refueling Arrangement 0.9654 0.00006 -0.0011 3 No Category 2 Arrangement 0.9651 0.00006 -0.0014 4 Maximum Category 1 Arrangement 0.9584 0.00006 -0.0081 5 Odd Arrangement 0.9655 0.00005 -0.0010 NET- 28091-0003-01, Revision 0 126

8.6 Calculations for Discharged Fuel (IP2 A-X and IP3 A-AA)

As noted in Section 5, depletion conditions vary for each batch. Bounding depletion conditions are used for each batch grouping or individual assembly. Minimum burnup requirements are then determined for each batch grouping or individual assembly.

It is desirable to allow as many assemblies as possible to be stored in Region 2 due to the limited size of Region 1. The minimum burnup requirement of Region 2 is for the 3-out-of-4 area. If an assembly burnup exceeds this minimum requirement, then it is classified as Category 4 fuel. If the burnup exceeds the minimum requirement by 11 GWd/T or more, then it is classified as Category 5 fuel. There are currently a sufficient number of assemblies that meet the Category 5 criteria so that all cells requiring Category 5 can be filled . If the burnup of an assembly is less than the general Category 4 requirement for the batch grouping, its classification is further studied. Many of these cases make the requirements for Category 4 after further analysis (see Table 8.24) . A few remaining assemblies do not have more than 28 .5 GWd/T burnup and further analysis is performed to show that they meet the reactivity requirement for Category 3 (see Table 8.23) .

Tables 8.13 to 8.22 are the Category 4 loading requirements (the 3-out-of-4 area of Region 2) for each batch or batch grouping as a function of initial enrichment, cooling time, and peaking factor.

NET- 2809 1-0003-01, Revision 0 127

Table 8.13: Batch A-D Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.20 (w/o) 10 25 45 2.0 7.90 7.3 1 7.04 2.2 10.56 9.69 9.35 2.6 16.68 15.29 14.64 3.0 22.78 2 1.12 20.32 3.4 27.82 25.90 25 .02 3.8 32.37 30.63 29.69 Enrichment Coolin!! Time (years) PF=0.80 (w/o) 10 25 45 2.0 7. 62 7.09 6.85 2.2 10.18 9.44 9.15 2.6 16.20 14.93 14.37 3.0 22.27 20.74 20.00 3.4 27.25 25.51 24.72 3.8 31.84 30.24 29.34 Table 8.14: Batch E-F Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.20 (w/o) 10 25 45 3.0 21.45 20.02 19.35 3.4 26.48 24.87 24. 16 3.8 31.31 29.66 28.76 Enrichment Coolin!! Time (years) PF=0.80 (w/o) 10 25 45 3.0 21.05 19.76 19. 14 3.4 26.03 24.6 1 23 .96 3.8 30.93 29.36 28 .53 NET- 2809 1-0003-01, Revision 0 128

Table 8.15: Batch G-L Mini mum Burnup Requirements (GWd/T) for Category 4 Enrichment Coolin PF=l.20 (w/o) 10 45 3.0 24.39 22.8 1 22. 11 3.4 29.42 27.43 26.52 3.8 33.64 31.90 31.04 Enrichment Coolin PF=0.80 (w/o) 10 45 3.0 23.97 22.54 21. 86 3.4 28 .91 27. 12 26.26 3.8 33.19 31.56 30.78 Table 8.16: Batch M-P Min imum Burnup Requirements (GWd/T) for Category 4 Enrichment Coolin Time ears PF=l.20 (w/o) 10 15 25 45 3.4 2 9.85 28.92 27.78 26.78 3.8 33.69 32.87 3 1.83 30.92 4.2 39.28 38.59 37.62 36.34 4.6 41 .3 1 40.56 39.66 38.91 5.0 4 5.42 43.96 42.16 41.25 Enrichment ears PF=0.80 (w/o) 10 25 45 3.4 28.92 28.06 27.05 26.17 3.8 33.40 32.64 31.68 30.86 4.2 38.85 38.30 37.17 35.99 4.6 41 .43 40.78 40.01 39.37 5.0 44.58 43 .27 4 1.79 4 1.04 NET- 2809 1-0003-01, Revision 0 129

Table 8.17: Batch Q- S Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Coolin2 Time (years) PF=l.20 (w/o) 10 15 25 3.8 29.75 28 .65 27 .29 4.2 33.63 32.43 30.96 4.6 37.50 36.20 34.58 5.0 4 1.31 39.88 38 .1 4 Enrichment Coolin2 Time (years) PF=0.80 (w/o) 10 15 25 3.8 28 .93 27.94 26.74 4.2 32.83 31.75 30.42 4.6 36.67 35.49 34. 03 5.0 40.4 1 39. 11 37.53 Table 8.18: Batch T-V Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Coolin2 Time (years) PF=l.20 (w/o) 5 10 15 25 3.8 32.40 30. 65 29.44 28 .07 4.2 36.45 34.51 33. 19 3 1.63 4.6 40.41 38.27 36.86 35. 18 5.0 44.24 41.94 40.42 38.64 Enrichment Coolin2 Time (years) PF=0.80 (w/o) 5 10 15 25 3.8 31.41 29.80 28 .8 1 27.58 4.2 35.34 33 .59 32.47 31.1 1 4.6 39.2 1 37.3 6 36.1 1 34. 60 5.0 42.98 40. 87 39.55 37.90 NET- 28091-0003-01 , Revision 0 130

Table 8.19: Batch W Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.2 (w/o) 2 5 10 15 25 3.8 33 .72 32.25 30.43 29.28 27.92 4.2 37.86 36.2 1 34.23 32.89 31.40 4.6 41. 84 40.07 37 .90 36.53 34.87 5.0 45 .75 43.82 4 1.52 40.04 38.27 Enrichment Cooling Time (years) PF=0.80 (w/o) 2 5 10 15 25 3.8 32.49 3 1.23 29 .66 28 .66 27.43 4.2 36.52 35.11 33 .35 32.27 30.88 4.6 40.42 38. 86 36.93 35.71 34.23 5.0 44 .20 42.52 40.50 39.19 37.57 Table 8.20: Batch X Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Coolin!?: Time (years) PF=l.2 (w/o) 2 5 10 15 25 3.8 33.73 32.30 30.48 29.29 27 .96 4.2 37.85 36.21 34.23 32.94 3 1.4 1 4.6 41.8 1 40.08 37.84 36.50 34.8 6 5.0 45 .69 43 .82 41.43 39.94 38.15 Enrichment Cooling Time (years) PF=0.80 (w/o) 2 5 10 15 25 3.8 32.53 31. 25 29.59 28 .63 27.40 4.2 36.53 35 .10 33 .27 32.18 30.85 4.6 40.39 38.81 36.91 35.69 34. 18 5.0 44.16 42.49 40.47 39. 16 37.56 The following loading requirements are for all discharged fuel from IP3 (Batches A thru X).

NET- 28 091-0003-01, Revision 0 131

Table 8.21: Batch A-U (IP3) Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.20 (w/o) 10 25 45 2.2 12.52 11.41 10.92 2.6 18.83 17.25 16.52 3.0 24.71 23.0 1 22.27 3.4 29.79 27.70 26.71 3.8 33.94 32.08 31.21 Enrichment Cooling Time (years) PF=0.80 (w/o) 10 25 45 2.2 12.09 11.10 10.67 2.6 18.28 16.87 16.23 3.0 24.17 22.66 21.97 3.4 29.16 27.28 26.37 3.8 33.39 31.72 30.90 Table 8.22: Batch V-X (IP3) Minimum Burnup Requirements (GWd/T) for Category 4 Enrichment Cooling Time (years) PF=l.20 (w/o) 10 25 45 3.4 28.28 27.24 25.92 3.8 31.90 30.69 29.26 4.2 35.13 33.75 32.19 Enrichment Cooling Time (years) PF=0.80 (w/o) 10 25 45 3.4 27.57 26.56 25.43 3.8 31.04 29.96 28.7 1 4.2 34.1 1 32.95 31.58 The IP3 fuel after Batch U uses the Batch Z depletion parameters with two additional enrichment points at 3.4 and 3.8 w/o. For these lower enrichment points, the fuel is modeled as full length which is conservative for blanketed fuel (Batches V, W, and X have six inch natural urani um blankets while Y and AA have six inch 2.6 w/o blankets). The batches Y and AA (and the higher enriched fuel from Batch X) use the Z loading curve which was based on modeling an 8-inch axial blanket with a blanket enrichment of 4.0 w/o. If the blanket enrichment were the same, modeling a 6-inch blanket as 8 inches wo uld be non-conservative. However, the highest enriched 6-inch blanket is 2.6 w/o. The fo ll owing cases were run to NET- 28 091-0003-0 1, Revision 0 132

show that using the Z loading curve (8 -inch at 4.0 w/o) for the shorter less enriched blanket (6 -inch at 2.6 w/o) is conservative.

4.6 w/o, 42 GWd/T, Z batch: k-eff = 0.97 123 +/- 0.00008 4.6 w/o, 42 GWd/T, 6-inch AA: k-eff= 0.96982 +/- 0.00008 ~k = 0.0014 4.6 w/o, 46 GWd/T, Z batch: k-eff = 0.95008 +/- 0.00008 4.6 w/o, 46 GWd/T, 6-inch AA: k-eff= 0.94884 +/- 0.00008 ~k = 0.00 13 Table 8.23 shows a set of assemblies that failed Category 4 but have a bumup that is less than 28.5 GWd/T (the minimum bumup for Category 3). To show whether they can make Category 3, each assembly was calculated in the infinite 2x2 model of Region 1 containing four fuel assembl ies. The calculated kerrwith 28 .5 GWd/T fuel (5.0 w/o enrichment, PF=l.4) in this model is 1.0133 (note that Category 3 fuel uses leakage from the edge of the SFP to lower its k - this is further explained in Section 8.8) . If the calculated kerr for an assembly is less than 1.0133, then it is classified as Category 3 even though it has a burnup less than 28 .5 GWd/T.

Table 8.23: Individual Assembly Analysis for Category 3 Burnup Cooling Assembly ID Enrich PF Cale. k (MWd/T) (years)

X02 4.802 28357 6.3 1.194 0.9956 XOl 4.802 28460 6.3 1.199 0.9956 F65 3.346 12034 37.7 1.228 0.9911 V43 (IP3) 3.803 14949 26.2 1.11 6 1.0005 V48 (IP3) 3.805 15180 26.2 1.1 33 1.0005 Based on these results, all of these assemblies are classified as Category 3 even though their burnup is less than 28.5 GWd/T.

The assemblies in Table 8.24 fai led Category 4 by a small amount of bumup, so they were analyzed using the depletion conditions for the assembly rather than the depletion conditions for the batch.

Section 5.7 describes the special depletions and the axial bumup profiles used. The atom densities for these assemblies were placed in the Region 2 (3 -out-of-4) 2x2 model (all three assemblies in the model NET- 2809 1-0003-0 1, Revision 0 133

are the sa me). The calculated kerr plus the bias and uncertainty is less than 0.99 in all cases, so they are classified as Category 4 (stored in the 3-out-of-4 area of Region 2).

Table 8.24: Individual Assembly Analysis for Category 4 Assembly ID Enrich Burnup Cooling PF Calc.k Bias & Uncer. k9S/95 AlO 2.21 15038 42.3 0.92 0.9489 0.0166 0.9655 F44 3.35 23017 35.8 1.05 0.9609 0.021 8 0.9827 L48 3.69 29515 25.4 0.69 0.9624 0.0235 0.9859 W52 4.96 40641 6.3 0.84 0.9569 0.0307 0.9876 X18 4.95 42479 4.3 0.87 0.9566 0.0306 0.9872 U12(IP3) 3.21 24800 27.8 0.90 0.9588 0.0223 0.98 11 With the above loading curves, every historical assembly through Batch X of IP2 and Batch AA of IP3 has been categorized as 3, 4, or 5. This categorization is summarized in Appendix B.

Fuel assemblies A l 1, A24, A47, A49, ASO, A51, A54, and ASS in IP2 and assemblies A38, A43 ,

A44, A45, A51, A59, A63, and A64 in IP3 contained part length control rods. These assemblies have an 235 enrichment of 2.25 wt% U-235 (w/o) and burnups greater than 16 GWd/MTU . The Category 4 burnup requirement for these assemblies is less than 11 GWd/T. Although the reactivity effect of control rods is significant [ 14], it is not enough to overcome the 50% excess burnup in these assemblies . These assemblies have insufficient burnup for Category 5 and are dispositioned as Category 4.

NET- 28091-0003-01, Revision 0 134

8. 7 Cell Blockers IP2 will have two cells blocked (no fuel) at locations A22 and BC72 . One of these cells (A22) is at the Region 1 and 2 interface and is not credited. The other cell (BC72) , however, is on the edge of the SFP above the cask loading area. Due to the cell blocker, it is possible to allow a Category 4 cell on both sides of the cell blocker along the cask loading area . Figure 8.8 shows the SCALE model w ith the two extra Category 4 cells. The kerr for this case is 0.95844 +/- 0.00006. This can be compared to the reference kerr of 0.95839 (without the cell blocker and without the two additional Category 4 cells). The difference is within the Monte Ca rl o uncertainty.

L((;(P')

Q wuo MfOIIAL 2

- MIOUAI.. J

- MIUIIAl..-4 c::J ,wnuun1.. a 0 MltlWl.. 3 l

- N'H(RIAL M

- Ml[RIAL.SJ

- MlfHfn.. ,uo

- MIUIIAl411

- Ml{Nl,-_<1 12 a wmM I AL .. 13

- Ml(Nlnl -4 14

- 11Al(RIAI 41 S

- MTUIIJll41 6

- M IOl:ll'tl ,tt 7

- MfCfitlrt."I' MTCIIUL S t O

- MTUUn&.. S Ii

- l'IAT t NIAI Sl 2

- MJUIIN !il l

- l'WH(IUnl.. 11 14 a i.,i1ct11n.. rms Ml(RfAL $ 18 MIOUAL. 5 17

- Ml[RIAl. 1111 Figure 8.8: Cell Blocker Region 2 Model NET- 2809 1-0003-01, Revision 0 135

8. 8 Region 2 Checkerboard Figure 1.1 does not show a Region 2 checkerboard arrangement since it is unexpected that the plant will ever use a Region 2 checkerboard. However, after the removal of a significant number of assemblies, it may be convenient to create a zone of the SFP near the cask loading area where any assembly may be placed (including fresh 5.0 w/o fuel). For this contingency, no interface analysis has been performed, so the Region 2 checkerboard zone must have a row of water holes on all sides. The edge of the SFP can be credited as a row of water holes. Using an infinite 2x2 checkerboard model of Region 2, the calculated keff of 5.0 w/o fuel with 64 IFBA rods is 0.9521. With the bias and uncertainty (0.0083 from Table 7.5) this produces a k9s19s of 0.9604. This easily meets the 0.99 target. All of the fresh and burned fuel from IP2 and IP3 qualify for placement in this Region 2 checkerboard.

8.9 Burnup Penalty for Hafnium Flux Suppression Inserts Hafnium inserts have been used at eight corner locations in the IP3 core starting with Cycle 11 to mitigate concerns over Pressurized Thermal Shock. Hafnium inserts have never been used at IP2 . The burnup penalty for hafnium inserts was determined in the previous CSA to be 1.31 GWd/T (worst case) [1]. To provide more than adequate margin, a penalty of2 GWd/T is added to the bumup requirement for any assembly that had a hafnium insert any time during its life.

8.10 Failed Fuel Containers The southeast comer (please note that on the drawings in this report North points left) of the SFP contains two 16" circular pipes and are labeled "failed fuel containers" on Figure 3 .1. These containers have been used to store pieces of failed fuel rods, neutron sources, and fission chambers . The neutron sources and fission chambers contain too little fissile material to be a concern. The fission chambers have less than 10 mg U-235 each [25]. The neutron sources also have a very small amount of fissile material.

The ANSI/ANS 8.1 standard [53] states that 700 grams ofU-235 in any configuration is always subcritical. However, the failed fuel containers are not fully decoupled from the Module H of Region 2.

NET- 2809 1-0003-01, Revision 0 136

This analysis permits 16 fuel rods in each of the failed fuel containers. Rather than model the actual container, 16 pins are placed close to the fuel modules for each failed fuel container. Since the same SCALE unit is used in two places in the model, a third set of 16 fuel pins is in the model. Figure 8.9 shows the model with the extra pins. The extra pins did not change the k eff of the model. Table 8.25 shows the results of calculations where the number of pins is varied. Although more than 16 fuel pins per failed fuel container wou ld be possible, there are currently pieces from only one fai led fuel pin in the containers now, and greater than 16 rods in each is not credible. The fuel rods are modeled as fresh non-blanketed, no-IFBA, 5.0 w/o rods . The start source was placed near the failed fuel container.

- 1111101111. l

. ,,*'HOU.._*

0 MIOII"'- '

. ..... TfJI IN..t*

c:J -lCAIIII JI

. ... ,0t, ....

. .... TUUfll.ll

- IIITTOUt'll. 410

• *lllllUlUll.411

. ,...10111.4 11

- IWllUlflll.Ul

- IIAlOIIM.414

- llf'll(IIUll. 4 1!1

- IIITT£11111l.*U II

. ,.,!Ull11Ln 7

. .... IUUlll.411 l : J llflllJllll. !U O

- Mlf,IUIII S II

- IWlllllfllt. !1 17

- MTl.lt lll.SU

. ... ,1.11,11.51'

. .... lflllll.. S IS

1. >lCIIIIRISII

- f1A1CIU ,. S 17

- l'W<llltUl !Ut Figure 8.9: Failed Fuel Container Pin Model (Blow up of the top right of the Region 2 model)

NET- 28091-0003-01 , Revision 0 137

Table 8.25: Failed Fuel Container Pin Analysis Pins oer Container Calculated k Sigma 0 0.9584 0.00005 4 0.9586 0.00006 9 0.9585 0.00007 16 0.9585 0.00006 81 0.9635 0.00007 121 0.9793 0.00008 144 0.9959 0.00008

8. 11 Fuel Rod Storage Basket The Indian Point SFP can have movable fuel rod storage baskets that can be used to store fuel rods.

These baskets can fit in a storage cell and they have 52 holes for storing fuel rods. This was modeled as 52 fresh 5.0 w/o fuel rods in the 3-out-of-4 area of Region 2 (see Figure 8. 10). The calculated kefT for this configuration was 0.9584, which is well below the k95195 requirement of 0.99 since the bias and uncertainty for fresh fuel is only 0.0083 t.k (from Table 7.5). Therefore, the fuel rod storage basket is class ified as reactivity Category 4 (refer to Table B. l ).

Figure 8.10: Model for the Fuel Rod Storage Basket NET- 28091-0003-01 , Revision 0 138

8. 12 Assemblies with Missing Fuel Rods Typically, when a fue l assembly has one or more failed fuel rods, the failed fuel rod is removed and replaced with a stainless steel rod having the same outer dimension as a fuel rod. If this is performed, there is no criticality concern since the reconstituted assembly would be less reactive than the original assemb ly. However, if a failed fuel rod is removed but not replaced with a stainless steel rod, the reactivity increases because there is more water ava ilable. An analysis was perfonned for the previous CSA [ 1] in which one or more fuel rods are removed from an assembly to estimate the effect on k. This analysis was not repeated since the approach taken provided a large margin. The model that was used for this analysis contained absorber inserts.

It was detennined that keff gradually increases as more fuel rods are removed up to and including 36 missing fuel rods . If more than 36 fuel rods are removed, keff begins to decrease (see Figure 8. 11 ). The change in keffwith 36 missing fuel rods (see Figure 8. 12) was 0.0 184 ~k (a 2x2 array with all 4 assemblies having 36 missing rods). For simplicity, a bumup penalty of 4 GWd/T would cover this reactivity increase for an assembly with any number of missing rods. There is only one fuel assembly in the SFP (assembly ID ofT67) that has a missing fuel rod . This assembly has only one missing fuel rod .

The initial fuel enrichment for this assembly was 4.952% and the bumup is 49.81 GWd/T and the assembly has cooled more than 10 years. This assembly exceeds the Category 4 loading requirement by over 7 GWd/T, so it exceeds the requirement by more than the 4 GWd/T pena lty for missing fuel rods. If any assembly in the future is re-constituted without replacing fuel rods with stainless steel rods, then 4 GWd/T would have to be added to the loading curve requirement before it could be stored. This penalty covers any number of missing fuel rods , and there is no other loading restriction (two or more fuel assemblies with missing rods could be stored next to each other as long as the 4 GWd/T is added for each assembly).

NET- 2809 1-0003-01, Revision 0 139

k versus Missing Fuel Rods 0.9850 0.9800 0.9750 k

0.9700 0.9650 0.9600 0 10 20 30 40 50 Number of Missing Fuel Rods Figure 8.11: k cff versus Number of Missing Fuel Rods 000000<)00-bbOOOO OOOOODOOOOOOOO.J 000000000000000 000000000000000 oooooco 0000000 o:x)oooo . 0000000 000 .}';; 00000 <4\°'. JOO O~)O >t.. 00000 :,,;4 000 R o. CJ/JO*O*.* * }.0 ...0*.@s§*

v .

• O *~' O 'i>\';O . 00

.'-. 00 OOQ *oo.o . . 0.. oo'"' OOO 00000 >' 0 *.\ 00000 00000 0 0 00 00 00000 0 0 03000 00 0 O O O O 00 00 0 0 0 0 00 00000 0 0 00000 00000 0 0 00000 OCX)OO O 00(:00 ~;:::>QOO O 000 00 000 c oo :JOCJ *YY) i)O:) o oo oo.:) 00:.J ooo :xx ' o J 800 Y)O

• oo-:~oo ooo oo )oo..-~:) * * :f.o,oe:o) o:J***,oo** o o ":oo **-oo 006006000600000 00~006000600600 t1 00r.J()i)0:).)Q ..),.)00*.J.) 08*)000()00000000 000000000000000 0~0000000000000 000000000000000 000000000000000 o<x ,o,x;o o >.)<Y o.) :)')'>oo ~-:, o .. o ~:oo:-::o,)

00:) ,)O _> JO O,Y) ):JO 001 \JO 000

!ooo i.) :..)*) CK). i '..X).:J :.) )O <)oo oo :J ooo

ocy _'i,),) 0 00.>.)C) <Y) .... :JO O 00 ..)0::)

00000 0 0 oo '. __-, :.) C .:) ,_-, 0000~

0:.)

00000 0

J.)  :-, o i..;

0 o GOOOO

,. ; 0.:)

00000 0 0 00000 00060 0 0 00000 oo-* <)O <J oo,- ,) .:) O'.)* --oo o o*s:).:*oo ooo ,** o,) ')c::y*., j:J:) * :J.).J :-*,o,:.) 00 *1 ooo ooo *oo ~oo

• 000 ooo *oo ' oo

• ooo oo . ,O*Y *o ,; :ocJ.. ,,YJ o.:, .oo :*'. o

• o, ,oo-:;oo 000000000600600 000000000000000 000000000000000 000000000000000 Figure 8.12: Model for Assemblies with 36 Missing Fuel Rods [1]

NET-28091-0003-01 , Revision 0 140

8. 13 Storage of Miscellaneous Materials Miscellaneous non-actinide materials (for example, empty or full trash baskets), can be stored in fuel positions of any category. However, there are some special cases where some of the material may be stored in a water hole or 50% water ho le. If the miscell aneous material is any type of steel, Inconel, or absorber material (e.g., absorber coupons, stainless steel coupon trees , control rods, unburned burnable absorbers) it may di splace up to 50% of the water volume at the active fuel zone ( 144 inches) of a water hole or 50% water ho le (there are no restrictions on material above or below the active fuel zone). If the miscellaneous materia l is a very low absorbing material such as a void, zirconium, aluminum, cloth, plastic, concrete, etc., it cannot be placed in a water hole but may be placed in a 50% water ho le so long as the 50% water hole still has 50% water volume in the active fuel zone.

To confirm that a water hole is allowed to contain 50% absorbing material , two cases were run - one with 80% water, 20% sta inless steel and another with 50% water, 50% stainless steel. The difference in keff from the reference case is -0.0100 L'ik and -0.0045 L'ik, respectively. Other materials such as Inconel ,

absorber coupons, unburned burnable absorbers or control rods absorb more neutrons than stainless steel and are covered by this analysis.

Any uranium that is not in a fuel assembly (for examp le a removed or damaged fuel rod) must be stored in the Failed Fuel Containers (see Section 8. 10) or the Fuel Rod Storage Basket (see Section 8. 11).

8. 14 Borated Conditions The most limiting acceptance criteria is for the unborated condition, so the loading criteria have been set using models that do not contain soluble boron. In order to confirm that k9s19s is less than 0.95 at a boron content that is maintained even after a boron di lution accident (Section 9.6), a limited number of additional calculations were performed. The soluble boron concentration of 700 ppm is used for these calculations since this concentration can be easily supported by the boron dilution analysis , and it yields significant margin ink. For Category 2 fuel , at a burnup of2l GWd/f, the calculated keFF With water at NET- 2809 1-0003-0 l , Revision 0 141

180 °F and containing 700 ppm boron is 0.8496. With bias and uncertainty, this becomes k 95195 = 0.87 12.

Due to the difference between 0.87 and the target value of 0.94, no further ca lculations are warranted.

For Category 4 fuel , at the loading curve points for 5.0 w/o fuel at 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (PF=l.20) and the 4.2 w/o fuel at 25 years (PF=0.80) with 700 ppm boron in the SFP water at 70 °C, the calculated keff at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is 0.8665, whil e the kerr at 25 years is 0.8631. With the bias and uncertainty applied, the k 9s19s, to be compared to the regulatory limit of 0.95 , becomes 0.8972 at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 0.8903 at 25 years.

The bias and uncertainty used here for both Region 1 and 2 is from the unborated analysis.

Calculation of borated uncertainties is not needed due to the large margin from the regulatory limit. Even if borated uncertainties were calculated, it is expected that they would be smaller, since the primary uncertainty is the burnup uncertainty and the reactivity worth of burnup decreases with increasing boron concentration due to spectral hardening. Furthermore, ignoring the grids is still conservative at 700 ppm.

In addition to the analysis with the 2x2 model s, a full pool case was run at 700 ppm. The calculated keff is 0.8600. This case had the most reacti ve loading permitted for all five ca tegories and includes eccentricity. Using the highest bias and uncertainty of all regio ns , 0.0374, the k9s19s is 0.8974 which is much less than the target of 0.94. An analysis was also performed for Region 2 only. Table 8.26 provides the results from the full pool models.

Table 8.26: Normal Operations with Boron Dilution ppm (Full Pool Model)

Region Calculated k sigma All Categories at 5.0 w/o, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooled, Peaking Both 0.8 600 0.00008 Factor of 1.2, Loading Curve Burnups Categories 4 and 5 at 5.0 w/o, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooled, Peaking 2 0.8595 0.00005 Factor of 1.2, Loading Curve Burnups NET- 2809 1-0003-01, Revision 0 142

8.15 Burnup Penalty for High Soluble Boron Conditions If a cycle is shut down very early, it is possible that the limiting soluble boron used in the depletion analysis (950 ppm) would not be met. The cycle would have to be shut down extremely early since the 950 ppm would be violated only if the cycle were shut down more than two months early. To cover this unlikely possibility, a special depletion was done at a soluble boron concentration of 1200 ppm throughout the depletion (this value exceeds the highest cycle average ppm at any burnup). If the burnup averaged ppm for any assembly exceeds 950 ppm, burnup penalties of0 .2, 0.3, 0.6, and 0.9 GWd/T would have to be applied to the burnup requirement for Category 2, 3, 4, and 5, respectively. The following table summarizes calculations to show that these burnup penalties are conservative (the Monte Carlo uncertainty is +/- 0.00003) .

Table 8.27: Burnup Penalty Results at 1200 ppm Calculated Case k

Category 2 (3 of 4 in Region 1):

5.0 w/o, 21 GWd/T, 950 ppm 0.9689 5.0 w/o, 21.2 GWd/T, 1200 ppm 0.9686 Category 3 (4 of 4 in Region 1):

5.0 w/o, 28 .5 GWd/T, 950 ppm 1.0117 5.0 w/o, 28.8 GWd/T, 1200 oom 1.0113 Category 4 (3 of 4 in Region 2):

5.0 w/o, 48 .19 GWd/T, 950 oom 0.9586 5.0 w/o, 48 .79 GWd/T, 1200 ppm 0.9583 Category 5 (4 of 4 in Region 2):

5.0 w/o, 59.19 GWd/T, 950 ppm 1.0187 5.0 w/o, 60.09 GWd/T, 1200 ppm 1.0183 NET- 28091-0003-01, Revision 0 143

9 Normal Operations and Accident Analysis The criticality analysis must address all conditions in the SFP that can cause criticality. The normal operations are reviewed in Section 9.1. The accident analysis must assume the worst case conditions from the range of normal operations .

The accident analysis is covered in Sections 9.2 through 9.7 and analyzes the possible upset conditions that increase the reactivity of the SFP. The following accidents are analyzed:

• A fresh assembly misplaced outside of the fuel racks but next to the fuel racks,

• A fresh assembly dropped into an empty cell,

• An over-temperature accident (water boiling in the SFP as a result of a loss of cooling), and

• A multiple assembly misload event.

Two more accident conditions are considered, but no analysis is necessary. An assembly dropped horizontally on top of other assemblies is not specifically analyzed, because the assemblies are de-coupled as a result of the structure above the active fuel. The horizontal assembly would rest more than 20 cm above the top of the active fuel of the assemblies in the rack. This accident would be covered by the more severe accident of a fresh assembly dropped into an empty cell. The second accident condition would be a single misloaded assembly. For example, an assembly that is supposed to have a control rod inserted but does not. All violations of the loading requirements are bounded by a fresh assembly dropped into an empty cell. The last subsection of this section describes why a seismic event does not cause a criticality concern.

NET- 28091-0003-01, Revision 0 144

9. 1 Normal Operations A single isolated assembly at 5.0 w/o and no IFBA will have a k9s19s < 0.99. An assembly is isolated if there is 20 cm of water between assemblies [ 13] (a row of empty cells is 23 cm for Region 2) . The equipment in the SFP can only move one assembly at a time. IP2 does not have any racks in the SFP or refueling canal for temporary storage of fuel. However, there are two locations where it would be possible to place two assemblies within 20 cm of each other outside of the rack: when a fuel assembly is in the new fuel eleva tor and when a fuel assembly is vertical in the upender. Indian Point's procedures will not permit movi ng an assembly outside the rack within 25 cm of fuel in the new fuel elevator or upender. The criticality analysis credits the leakage at the edge of the racks . Therefore, placi ng an assembly within 20 cm of the side of the rack at the active fuel elevations is not permitted. IPEC's procedure wi ll also preclude this.

Moving fuel in and out of rack cell s does not increase kerr since the most reactive portion of the fuel assembly is at the top of the fuel, so moving the bottom of the fuel past the top as it is inserted or removed does not increase k.

Fuel inspection is performed above the rack where the fuel assembly is iso lated. Any reconstitution is performed while the assembly is isolated . Isolation requires a row of water holes on all sides and corners.

The outside of the SFP can be considered as a row of water holes.

The SFP is required by its Technical Specifications to contain at least 2000 ppm of soluble boron .

The SFP water temperature during normal operation ranges fro m above freezing to 180 °F (the SFP design basis maximum temperature).

NET- 28091-0003-01 , Revision 0 145

9.2 Misplaced Assembly For the misplaced fuel assembly accident, a fresh 5.0 w/o fuel assembly with 64 IFBA rods is placed in the SFP next to the rack in the most reactive location. There are two locations which could be limiting for a misplaced assembly; the inside comers of the cask area and near the new fuel elevator when the elevator has a new fuel assembly in it.

For the misplacement of fuel near the new fuel assembly in the elevator the fue l elevator is modeled in the cask loading area of the SFP. The key features of the model are maintained; close to a wall with two rows of Category 5 cells and in a big pool of water. The array structure used in the SCALE model makes it difficult to model at the actual location. The fuel elevator is conservatively modeled as exactly one assembly away from the side of Region 2, so one misplaced assembly can exactly fit between the elevator and the rack.

Figures 9.1 and 9.2 show the location of the misplaced assemblies analyzed. Table 9.1 presents the results of this misplaced assembly accident analysis. As can be seen from Table 9. 1, with the Technical Specification minimum of 2000 ppm of soluble boron, the final k9519s is below the target of 0.94. The analyses used starting sources located near the misplaced assembly.

NET- 2809 1-0003-01, Revision 0 146

L[QHO O vom Ml(AIAl. 2

- MTUUl'll.3

- MTUUnl <I

- rtflf(IWl 7 c:J M!(Ml"'-.8

- MIUIIALIO C ) PW'H[Rlr...JJ

- PIAllNIALSG

- MICMIAl. 113

- Ml[IUAL 2 10 l'IIUUUAL 212 fE:J M1£Rlnt. 21 3 l::3 MTotlRl. 2 1" D PWITLIUAl.21 5 MIUII"- 21 7

- IVH(liUAl.. 218

- Ml[MJAl. 310 Mruunt J tl MICMIAL 3 12 Ml'OUAL 3 13

- Ml[Rlnt JH B I MT(A l nt. 3 15

- MTCMHlL :U8

- J'W'IIUW'll 317

- WIT(M IAI 3 18

-Ml'l"IIMI AL410

- MlUUAL41l

- MlUUl'1L"1 2

- Ml[Mlnt '11 3

• ivuuw.... H

- MTOWl *U'!S

- Mf£MlnL"l8

- Ml(Mlfl.41 7

- MT[IUAl.<11 8 MlOUAL 5 10

- l'WHl:RIAl.. 5 11 Ill MlUllnt. 51 2

- M!CMll'II. 513

- MTCRIRI..Sl<4

- PW'lllfCIAl.. 5 15 Q M!UIIAI !U 8 Ml[AIAl. 5 17

- Ml(AIAL 5 18 Figure 9.1: Misplaced Assembly at the Cask Area Corner NET- 28091-0003-0 1, Revision 0 147

LEGUC)

O vo 10

• IUIUAL 2

- 11Al[fl:IAL3

- IWHCRIAL"I

- MlUUIII 7 Q m1uu111a

- l'IIIITUIIAL 10 CJ 111UUI.IAI 3 1

- JWITUIIAL 6&

- MIUIIAl. 83

- MfutlAl. 2 10 fllAIUUfll. 211 WllUUnl 3 10 WIILIW11.. 3 11 MUR IAi. 312 MJ[ftJAI_ JU

- M I UUAL3l4

- MIUUAL 3 16

- IWll[fl:1111 3 18

- MIUIUll 31.1

- Mf(lltlRl. 318

- M TUIIAL"IIO

- IWllOIIAl"lll

- NfHUtlAl.41 2

- MIUIIAl."1 3

- MltlUAL"II"

- M T(RlnL"IHI

- Ml(lll"-."116

- Mf(MUll <117

- Mf(MIAl."11 8

[:J 1wuow,1..s10

- 1¥11(1WW.. 511

- Ml(Nlnt. Gl2

- MIIIUAL 5 t 3

- Wll(!Unt.5 1<1

- Mlr.Rltll. 5 15 c:J 1wmRIAL !U6 MT(IUnL. 5 17

- Ml[NUt !SIii Figure 9.2: Misplaced Assembly between the Fuel Elevator and the Rack Table 9.1: Misplaced 5.0 w/o 64 IFBA Assemblies with 2000 ppm Location of Mis laced Assembl Calculated k Si ma Elevator 0.7748 0.0001 2 Cask Area 0.7864 0.0001 2 NET- 28091-0003-0 1, Revision 0 148

9.3 Dropped Assembly For the dropped assembly accident, a fresh 5.0 w/o assembly (no IFBA) is dropped into one of the empty cells in Region 2 with the SFP at the Technical Specification minimum of 2000 ppm. Further, the assembly dropped is modeled with the grids failed , which allows for full expansion of the pin pitch into the cell (this maximum expansion of the pin pitch removes any concerns about fuel grid failure after the drop). The pin pitch expansion is modeled as the maximum uniform expansion that would fit in the cell.

Figure 9.3 shows the fu ll pool model for the dropped assembly analysis with the dropped assembly.

Note that Region 1 cells have been removed so that the kerr for Region 2 can be found, which comes with a higher bias and uncertainty than Region 1. The other fuel assemblies are at the loading curve limit for 5.0 w/o, no cooling and a peaking factor of 1.2. The calculated kerr is 0.8700 with the 2000 ppm soluble boron. After adding bias and uncertainty, this would be much lower than the target of 0.94, giving 1% in margin. With this much margin , there is no need to reevaluate the bias and uncertainties for borated conditions.

NET- 28091-0003-01, Revision 0 149

Lf.(i(Nl O wuo IIATCAIAl. 2

- MT{Alnt. 3

- PWHUW'll 4 E:] ...1n:]UAI II

- "'HCfUAI 10 c:] MlllUnL 3 1

- M lli:Wll.66

- MlUWll. U

- MUAlAl.'41 0

- MlHtllV..411

- t'A'IUW'll.41 2

- IIAl[IUlll. <11 3

- Mr£11UnL41"1

- Mf(11Ulll41 6

- Ml(AJAl.<tl ll

- Ml[AIAl.417

- MJ(IUAl. 41 8 MT(JU AL 5 10

- MIDIIAl. 6 11

- Ml(JIIA1&1 2

- Ml(lllA1613

- MIUll"-. 0 14

- 11Al(H l nl 6 15 c:::] MTUUnL 6 16

. ,vouuni i:111 Figure 9.3: Full Pool Model with Dropped Assembly 9.4 Over Temperature In the over-temperature accident case it is assumed the SFP water is boiling and the boiling includes a 20% void fraction . This is consistent with current NEI guidance on over-temperature accidents [4].

Using the highest kerr full pool model , water is modeled with 2000 ppm of soluble boron. The calculated NET- 28091 -0003-01 , Revision 0 150

keffis 0.7464 +/- 0.00007. This case has a large margin to the 0.94 target, so calculation ofa bias and uncertainty for this specific case is not necessary.

9.5 Multiple Misloads The most limiting accident is the multiple misload case. For the Region 1 multiple misload, it is assumed that all cell s are filled with the most limiting assembl ies for IP2 (5.0 w/o enriched fuel with no bumup and 64 IFBA rods) . The calculated keff at 2000 ppm is 0.8196 +/- 0.00008 . Clearly, there is significant margin for the complete misload of assemblies in Region 1.

For the multi-misload analysis for Region 2, all cells that are permitted to contain fuel are modeled as misloaded with once burned 5.0 w/o fuel with a burnup of24 GWd/T, as unburned fuel is easily identifiable and expected to be loaded into Region 1. Nearly all of the once burned fuel assemblies exceed this bumup. For example, if Cycle 23 compl etes its expected burnup, four assemblies will have a burnup of22 GWd/T with all of the rest above 24 GWd/T. Currently (during IP2 Cycle 23), only four assemblies would be in the SFP during refueling with a bumup less than 24 GWd/T. Since the number of assemb lies below 24 GWd/T is so few, misloading them in a reactivity significant way is not credible (note that a single misload is covered by the dropped rod analysis, and if the misloads are not close to each other, the effect on k is the same as a single misload). The calculated keff for this case with 2000 ppm is 0.9124 +/- 0.00006. The bias and uncertainty for this case is 0.0223 Lik (from conservative interpolation using Tab le 7.3, the manufacturing tolerances from Table 7.4, plus the higher validation bias and uncertainty due to the harder spectrum). Adding 0.0223 to 0.9124 results in a k9s19s= 0.9347 which is below the target of 0.94.

The multi-misload analysis did not misload fuel into the water holes or 50% water holes. It also maintained the control rods at the locations required by the Technical Specifications. The water holes and control rod locations in Region 2 are not allowed to change position. With the implementation of this N ET- 28091-0003-01 , Revision 0 151

CSA, the IPEC staff wi ll receive training with emphasis on the fixed positions in Region 2 for the contro l rods and water holes.

Although it is clearly not allowed by the Technical Specifications and the staff will have had specific training to reinforce the requirements of the control rods staying in the specific locations, analysis has been performed where all of the control rods in Region 2 are removed . This case assumes that the fuel is consistent with the most limiting conditions allowed by the Technical Specifications (5.0 w/o fuel with burnups of 48 .19 and 57.89 GW d/T). The calculated ketr with 2000 ppm is 0.7880. Again, this ketr after the addition of an appropriate bias and uncertainty is significantly less than the target 0.94.

The final multi-misload analysis assumes the nonnal loading of Region 2 but all water holes are filled with fresh 5.0 w/o fuel with 64 IFBA. Since the reactivity is dominated by the fresh fuel, it is appropriate to reduce the Category 4 fuel burnup by the difference in the bias and uncertainty. From Table 8. 7, the bias and uncertainty at 48.19 GWd/T is 0.0307 t.k. The Region 2 fresh fuel bias and uncertainty is 0.0083 t.k (from Table 7.5) . Therefore a delta burnup to cover 0.0307 - 0.0083= 0.0224 t.k is taken. This is estimated as 4 GWd/T. The Category 4 fuel modeled is the loading curve 48.19 - 4 = 44.19 GWd/T.

The calculated ketr is 0.9224 +/- 0.00005. The Region 2 bias and uncertainty for Category 1 fuel from Table 7.5 after adjusting for the high EALF is 0.0151 t.k. Thus, the k9s19s is 0.9224 + 0.0151 = 0.9375.

Since this is less than 0.94, this multi-mis load meets the requirement.

9. 6 Boron Dilution Accident Crediting 700 ppm of soluble boron reduces the calculated ketr plus biases and uncertainty to well below 0.94 (see Table 8.26). The boron dilution analysis of record [52] shows that dilution from the Technical Specification required minimum of 2000 ppm to 700 ppm is not credible due to the amount of time and water needed to dilute to this level.

NET- 28091-0003-01, Revision 0 152

9. 7 Seismic Event This CSA is not crediting any absorber plates that could be affected by a seismic event. Further the space between the rack modules is not credited. The space between the rack modu les and the side of the SFP has a very small reactivity effect (see Table 6.5), so the soluble boron would easily cover any possible movement of the rack modules toward the wall. Any random variations in the cell dimensions would have to cover multiple cells to affect kerr, and, again, the soluble boron would easily cover these variations. The dropped assembly that could result from a seismic event is covered in Section 9.3. In summary, this CSA is not sensitive to a seismic event.

NET- 28091-0003-01, Revision 0 153

10 Summary This CSA removes the reliance on BoraflexTM in the IP2 SFP. This is accomplished by use of water holes, control rods, and leakage at the edge of the SFP. Each fuel assembly is categorized by its reactivity and specific locations in the SFP are reserved for each reactivity category. The fuel categorization of historical fuel is accomplished by use of batch groupings with similar characteristics. The categorization takes credit for lower enriched axial blankets, cooling time, and assembly average peaking factor.

This CSA also categorizes IP3 fuel for storage in the IP2 SFP.

Section 10.1 contains a confirmation checklist that the guidance ofDSS-ISG-2010-01 [5] is fo llowed.

Section 10.2 describes the reactivity categorization of fuel assemblies. Section 10.3 identifies the reactivity categorization of each cell. Section 10.4 lists the limitations for fuel assemblies that have not been categorized in Appendix B.

10.1 Review of DSS-ISG-2010-01 Table 10.1 shows the guidance given in DSS-ISG-2010-01 and how this criticality analysis follows that guidance.

Table 10.1: DSS-ISG-2010-01 Checklist Section in this Guidance from DSS-ISG-2010-01 Implementation Report

1. Fuel Assembly Selection All fuel has co me from the same ve ndor with the same clad Section 3.2 Demonstrate all fue l for all outside diameter. Small des ign changes have been conditions insig nifi cant to criticality anal ysis.

2. Depletion Analysis Thi s is fo llowed. Uncertainty fo r the isotopi c content is Section 4 a.i. 5% (Kopp M emo) should considered and impl emented as 5% of the depl etion reactivity only be used to cover uncertainties (i.e., delta-k of depletion) . In addition, a bi as of 1.5 % of the in isotopic concentration worth of the minor actinides and fi ssion products is appli ed to cover their bias and uncertainty.

2. Depletion Analysis No integral bu rnab le absorbers are considered for fresh fuel Section 4 a.i i. Reacti vity decrement should for determining the reacti vity decrement.

not include the integral burnabl e absorbers.

2. Depletion Analysis Bounding values within each batch grouping are used for a ll Sections 5.1 b.i. Bounding values should be parameters. through 5.5 used.

NET-28091-0003-01 , Revision 0 154

Section in this Guidance from DSS-ISG-2010-01 Implementation Report

2. Depletion Analysis The highest power is used which leads to hi g her moderato r b.ii. Use the more limiting and fuel tem peratures, thus in creasing k. To acco un t for Sections 5.4 bounding parameter when a conflict lower power coast down, a Sm-149 correction is made. and 5.8 occurs.

2. Depletion Analysis Bounding values within each batch group ing are used for all Sections 5 .1 b.iii. Non-bounding values are parameters. through 5.5 outside scope ofISG.

2. Dep letion Analysis IP2 and IP3 had standard burnable absorbers, and W ABAs . Sections 5.2

c. i. All removab le burnabl e IP3 also had Hf flu x suppressors. All of these are and 8.9 absorbers must be cons idered. conservatively accounted for.

2. Depletion Analysis The analysis includes the maximum number ofIFBA rods at Section 5.2

c. ii . Limiting integral burnable the hi ghest boron loading for each fuel batch gro uping.

absorbe rs shou ld be used.

2. Depletion Analysis For depletion analysis, the maximum absorber material is Section 5.2

c. iii. Model the burnab le modeled with the max imum water displacement. For the pool abso rbers aooropriately. analysis, all burnabl e absorbers are removed .

2. Depletion Analysis The depletion model correctl y accounts for the increased rate Section 5 c.i v. Consider co mpet ing effects of plutonium production fro m increased fast neutron capture in U-238

2. Dep letion Analysis All hi storical assemblies under D-Bank were identifi ed and Sections 5.5 d.i . Spectrum harden ing from the appropriate treatment is appli ed. For current fuel , it is rodded operation should be assumed that a cont rol rod was fully inserted for 2 GWd/T consi dered. burnup.

2. Depletion Analysis The axial profiles for asse mbli es without ax ial blankets are Section 6.2 d.ii. Effect of control rods on the from NUREG/CR-6801. T hese profiles include rodded cases.

axial burnup profi le shou ld be For blanketed fu el, actual burnup profiles are used. T hese consi dered profil es cover any control rod effects.

3. Criticality Analysis The axial profiles for assemb li es without ax ial blankets are Section 6.2

a. Axial Burnup Profi le from NUREG/CR-6 801 and used in a conservative manner.

i. Use ofNUREG/CR-6 801 is For blanketed fu el, actual burn up profiles are used.

acceptable if done properly

3. Criticality Analysis Site-specific profi les are used . Section 6.2

a. Axial Burnup Profile ii. Site-specific profiles

3. Criticality Analysis For full length fuel , results from uni form and NUREG/CR- Section 6.2

a. Axial Burnup Profi le 6801 shapes were compared at 10 GWd/T and the iii. Uniform profi les NUREG/CR-680 1 shape was more li miting . The lowest burnup of any ful l length fuel is greater than 10 GWd/T. For axially blanketed fu e l, the lowest relative power at each node from all the blanketed fuel burned at Indian Point was used to determine the axial burnup di stributi on for each axial blanket design. Since these relative powers were not renormalized, it covers both the top peaked and center peaked condition.

3. Criticality Analysis The rack dimensions and materials are provided by the Section 3.1

b. Rack Model manu facture r (References 8 and 9).

i. Model inputs shou ld be traceable.

3. Criticality Analysis No credit is take n for absorber panels.

b. Rack Model ii. Effic iency of the neutron absorber sho uld be establi shed .

NET- 28091-0003-01, Revision 0 155

Section in this Guidance from DSS-ISG-2010-01 Implementation Report

3. C riticality Analysis No credi t is taken for absorber panels.

b. Rac k Model iii . Conservati ve degradation should be used.

3. Cri ticality Analysis T he interface ana lysis adj usted the burnu p fo r the d ifference Secti ons 7.6

c. Interfaces - Use the maxi mum in the bi as and uncertainty. and 8.4.3 uncertainties from either s ide.

3. Criticality Analysis There are no temporary storage locations in the IP2 SFP . All Secti on 9.1

d. N ormal Cond itions - All normal operating conditions are covered by the anal ysis.

normal conditi ons suc h as movement of fu el and inspections shou ld be considered.

3. Criti cality Analysis All normal initi al condi tions are considered . For example, in Secti on 9

e. Accid ent Condi tions the mi splaced asse mbl y analysis it is assu med that the fuel

i. Should consider all normal elevato r has a fresh fu el asse mb ly in it when another conditi ons as base conditi ons. asse mbl y is mi splaced next to it.

3. Criticality Analysis Large margins exist for all of the accident conditi o ns with the Secti on 9

e. Acc ident Condi tions excepti on of the multipl e mi sload scenarios . These scenari os, ii. Graded app roac h may be however, still meet the target k95195 (see Sect ion 9.5).

taken when credi ting solub le boron.

4. Criticality Code Validation NUREG/CR-6698 is fo llowed for the validatio n. Append ix A NUREG/CR-6698 endorsed

4. Criti ca lity Code Valid ati on T he HTC critical experiments are included in the analys is. Appendi x A.3

a. Area of Applicability

i. Include the HTC criticals

4. Criticality Code Va lid ati on T he fin al bias and uncertai nty is determi ned by the most Appendix A

a. Area of Applicability limiting of either the M OX and HTC cri ticals or the U02 ii . Use app ropriate cri ticals criticals.

4 . Criti cality Code Validation 328 fresh fu e l criti cal experi ments are used. 11 7 HTC Appendix A

a. Area of Appli cability criti cals are used as well as 63 MOX criticals. G roupings of iii . Sufficient criti cals fo r cri tical sets are ana lyzed to confirm when they shoul d be analysis and app ropri ate groupin g. included in the set as a whole.

4. Crit ica lity Code Validation T he large number of criti cal experiments used and the large Append ix A.2

a. Area of Applicability vari ati on in cri tical configurati ons (geometry and material) iv. Be sure the set is not hi ghl y reduces the concern about being correlated . The analysis corre lated. used 37 diffe rent sets of experiments that were performed in 7 different criti cal fa ciliti es .

4. Cri ticality Code Valid ation T he trend analys is is perfor med on all of the maj or Append ix A. 2.5

b. Trend Analysis parameters. T he trend analysis fo und the best linear tit. No Adequate, appropri ate, not trends were rejected to be conservati ve . The most limiti ng rej ected. bi as and uncertainty for the area of applicability is appli ed assuming either that all trends are real or there are no trends.

4 . Criticality Code Validati on The stati stical approach recomm ended in NUREG/CR-6698 Appendix A. 2.5

c. Stati sti cal Treatment is used. T hus the variance of the popul ation about the mean

i. Use the variance of the is used rather than the va riance of the mean .

populati on about the mean 4 . Cri ticality Code Valid ati on The stati sti cal approach recomm ended in NUREG/CR-6698 Append ix A. 2.5

c. Stati sti cal Treatment is used. T he correct confidence fac tors are used .

ii . U se correct confidence factors.

4. Criticality Code Validation No rmali ty testing is perfo rmed and the app ropriate stati sti cal Appendi x A.2. 5

c. Stati sti cal T reatment treatment is applied.

iii. Consider N ormality NET- 28091-0003-01 , Revision 0 156

Section in this Guidance from DSS-ISG-2010-01 Im lementation Re ort

4. Criticality Code Validation Lumped fission products are not used.

d. Lum ed Fission Products

4. Criticality Code Validation No code-to-code comparisons are used for validati on.

e. Code-to-Code Co mpari sons However, CASM0-5 analysis was used to confirm that the ISG-2010-01 allowed 5% of the delta kerr of depletion is ade uate.

5. Miscellaneous Precedence is not used as a li censing basis.

a. Precedence References used were carefully chosen to be applicable to the

b. References point being made.

c. Assum tions Assum tions are identified.

NET- 28091-0003-01 , Revision 0 157

10.2 Fuel Reactivity Categorization The reactivity category for each fuel assembly must be determined prior to loading in the IP2 SFP.

For IP2 Batches A through X, the reactivity category for each assembly is found on Table B. l and for IP3 Batches A through AA on Table B.2 of Appendix B. Fuel for IP2 and IP3 batches beyond those in Appendix B are labeled " Batch Z".

The equations to determine the categorization for Batch Z are:

B 12 = (-6.26824 + 5.29367*E -0.37 l 54*E2) e - (0.129582 -002049 1s*E + o.00205596*E*El*CT _0 _13 331 +

6.9037*E + 0.122068*E 2 Bo.s ( 15.1405 _ 4 .8 l l 33*E + 0. 753855*E2) e - (0.121252 -0.0 15099 1*E + o.00121009*E*El*CT _16 _2293 +

14.0 159*E - 0.687054*E2 MRB = Bo.s + (PF - 0.8) x (B1.2 - Bo.s) I 0.4 where MRB is the Minimum Required Burnup (GWd/T)

E is the U-235 initial enrichment (w/o)

CT is the cooling time (years)

PF Assembly Burnup I (Sum of Cycle Burn ups)

If an assembly had an inserted hafnium flux suppression insert any time during its life, then 2 GW d/T bumup must be added to the MRB. If an assembly has any number of mi ssing fuel pins that have not been replaced by stainless steel rods, then 4 GW d/T burnup must be added to the MRB. If an assembly was burned with a burnup averaged soluble boron concentration of greater than 950 ppm, then 0.2, 0.3, 0.6, and 0.9 GWd/T must be added to the MRB for fuel Categories 2, 3, 4, and 5 respectively.

If the fuel has a burn up greater than (MRB+ 11) the fuel is Category 5. If the fuel has a bum up greater than MRB but less than the Category 5 requirement the fuel is Category 4. If the fuel has a burnup greater than 28 .5 GWd/T but less than the MRB the fuel is Category 3. If the fuel has a bumup greater than 2 1 GWd/T but less than 28.5 GWd/T the fuel is Category 2. If the fuel has less than NET- 2809 1-0003-01, Revision 0 158

Westinghouse Non-Proprietary Class 3 21 GW d/T or violates any of the requirements of Table 10.4 the fuel is Category 1 or Category 4 if it contains a control rod. For a Batch Z assembly to be stored in the SFP, it must have at least 64, 48 , 32, or 16 IFBA rods for enrichments less than or equal to 5.0, 4.5 , 4.0, and 3.5 w/o, respectively.

If an assembly has a control rod in it, its fuel category increases. If a Category 1 fuel assembly has a control rod in it, it is classified as Category 4 fuel. If a Category 2, 3, or 4 fuel assembly has a control rod in it, the assembly becomes a Category 5 fuel assembly. Control rods that are required in the fuel layout (see Section 10.3) may not be credited to raise the category of that fuel assembly (for example, a Category 2 assembly with an inserted control rod may not be placed in a cell in the control rod area that requires a control rod) .

The above loading requirements have been summarized below in Table 10.2.

Table 10.2: Summary of Loading Requirements for Fuel Batch Z Minimum Burnup Requirement Category 1 Fresh fuel with at least 64, 48, 32, 16, or OIFBA rods for enrichments less than or equal to 5.0, 4.5, 4.0, 3.5, and 3.0 w/o respectively. (@ [ mg 10B/inch] a,c or 2reater). Also, burned fuel with less than 21 GWd/T burnup is Cate2orv 1.

Category 2 Burned fuel assemblies with at least 21 GWd/T burnup, initial enrichment of 5.0 w/o or less.

Category 3 Burned fuel assemblies with at least 28.5 GWd/T burnup, initial enrichment of 5.0 w/o or less.

Category 4 Burned fuel assemblies whose loading requirements are determined from Table 8.4 or the curve fit described in Section 8.3.1. Also, any Category 1 assembly containing a control rod is Category 4.

Category 5 Category 4 burnup requirement plus 11 GWd/T. Also, any Category 2, 3, or 4 assembly containing a control rod is Category 5.

NET- 28091-0003-01 , Revision 0 159

10.3 Allowable SFP Cells for Each Fuel Category The analysis of the IP2 SFP uses a cell dependent reactivity. That reactivity by position is defined in Figure 10.1 by use of cell categories, required contro l rods , and two types of water holes. Each cell in the IP2 SFP has a predetermined minimum fuel category. However, there are a few alternative configurations of the cell categories in the SFP. Figure 10. 1 shows the primary assignment of the cell categories for the SFP. Only Category 5 fuel may be placed in a Category 5 cell. Although Category 5 fuel can be stored anywhere in the SFP, the outer two rows of Region 2, the checkerboard control rod area of Region 2, and the interface between Region 1 and Region 2 must contain only Category 5 fuel or remain empty (alternate arrangements are allowed as discussed in Section 8.5). A Category 4 cell can accept Category 4 or 5 fuel. A Category 3 cell can accept Category 3, 4, or 5 fuel. A Category 2 cell can accept Category 2, 3, 4, or 5 fuel. A Category 1 cell can accept all fuel.

Category 1 cells can replace an area of Region 1 on the Figure 10.1 arrangement (Region 1) using the following two rules :

1. Each Category 1 cell must be face adjacent with at least three water holes.

2. Each Category 2 cell may not have more than one face adjacent to a Category 1 cell.

Figures 10.2 through 10.4 show examples of allowable Category 1 cell locations in Region 1.

It is also permitted to create a checkerboard arrangement of Category 1 cells in Region 2. In order to prevent interaction with other portions of Region 2, the checkerboard area must have a row of water holes on all sides. The water outside of the rack counts as a row of water holes .

Finally, a 3 by 3 block of cells where all of the cell s are water holes except the center cell may be created anywhere in the SFP, and any fuel assembly may be placed in the center position (see Section 9.1).

NET- 2809 1-0003-01, Revision 0 160

1 1

• s , 1 a , 10 11 11 u 14 15 " 11 11 1, 10 u 21 n 24 zs z, l1 1a u 10 31 1.SMl1tltJ101112U1CU 09 ON OM t---t---t--t---11-+---t--+---t-+-il-+-+--+-ll-+-+-+-l-+-+-+-!ll-+-+-+-l-+-+-+-l- l l'-"1---t--t---t--t-t-,--,--,,-r, Ol E *~-+---+--t--+--+-,.-o-o~>-~ M OJ D 11-+---+--+-+--+--+--t--i--,t--l--!

OH C I ~+--+--+--+-+-+--l--1""""1--1-4 1--+--+--+--+--t---+-+--+-!---i rt-1--r-r---r-t-t~-1rt-t-1r-r-r-t-t-1-rt-11 ~-+--+--+-++-1--1--1--1--1-1 0G B ~

A >( OE Key:

'.=;=::=¢:.::¢=;;:=;::;:;:::!;:::;::=¢=;::::;:=;;;;;!;:::!;::=:=¢::;;;=:;:::=;=:;::::!;;::::!;:~:,a;;;;::::~~=:=::=::=::~~

CP CP

--1>---'I >-+--+--+---+--+---+-,.--<'.....--<>-<

CN CN D WaterHole CM CM

"""'1--!f--+-+-+:=+-+-+--+-+-l-H D S0%WaterHole Cl 111111-lt-+-+++--t--+-t---t-l--iH Cl CK CK

• category 1Fuel CJ CJ

....-;t-+-+-++++;--t-t-H CH D category 2 Fuel CH CG CG D categorylfuel CF 111111-lt-+-"1-+--t---t--+-t---t-t--iH CF C£ Cf

• category4Fuel 11---t--+--+---+--+--+--+-+-l--1--+

m co

!!!l!!!!!=='l\,=~~=*'4,=~,=,l,,4,=,\,=I category s Fuel Ill BN I--+-+-+-+--+--+--+--+..+--,.-<

BM BM category S Fuelw~h a required lull lengthRCCA

[!J Blocked Cell Bl Bl BK

&I &I 8H 8G 8(

9G 8(

BO BO

-+--+-+-++---.t-::,-,t-H AM .......,.._.__._.,_........_..__..__.._...., BC Al Alt AJ AH AG / Cask Area AF AE AO AC AB 1-1--+-+-+-+--+--+-+--+--+-+-lt--t--+-+-+-+---+--+-+--+---t--t--i

._._.._.._......................._.__,_- __. ._~ .................................... -~

Figure 10.1: Fuel Category Location Requirements (Base Case)

NET- 28 091-0003-01, Revision 0 161

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 H

G F

E D

C B

Figure 10.2: Refueling Arrangement 1 2 3 4 5 6 7 8 9 10 H

G F

E D

C B

Figure 10.3: Max Cat 1 Arrangement 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 oc: -~-

I

+ .I t H I

- -- - r - - I

~

G F

~ -- i E

D C

B L -~-

I

- - - *- - - ~

• - > - >-- ~

A X Figure 10.4: Example Odd Arrangement NET- 28091 -0003-01 , Revision 0 162

There are restrictions on what can be placed in the water holes or 50% water hol es. The water hole may not have anything in it (except water) in the active fuel area with the fo llowing exceptions:

• If the item is made of stainless steel, Inconel or absorber materia l, it is allowed to displace up to 50% of the water and still be stored in a white cell ( 100% water cell).

• The 50% water holes (pink cells) allow di splacement of the water by any non-fuel material up to 50% volume fraction. This would allow, for example, a component made entirely of zirconium taking up less than 50% of the volume to be stored in a 50% water hole (pink cell) .

• The cell blocker at location BC72 (in the first row of cells above the cask area pit) may not be moved and is required in order to classify the cells on either side of the cell blocker as Category 4 cells .

10.4 Fuel and Operating Requirements The actual fuel and operating conditions are used in the analysis of historical fuel (Batches A through X fo r IP2 and Batches A through AA for IP3 ). Fuel batches after X for IP2 and after AA for IP3 are called Batch Z. This section describes the requirements for Batch Z.

To meet the limitations of this criticality analys is, the fuel design must meet the design requirements given on Table 10.3 .

NET- 28 091-0003 -01, Revision 0 163

Westinghouse Non-Proprietary Class 3 I Table 10.3: Fuel Design Requirements for Batch Z assemblies Attribute Value Notes This includes dishing and Maximum fuel pellet U02 stack density 95 .0%TD chamfering Fuel pellet OD (inches) 0.3659 Nominal Fuel clad OD (inches) 0.4220 Nomina l Fuel clad ID (inches) 0.3734 Nominal Fuel pin pitch (inches) 0.563 Nominal Guide tube OD (inches) 0.533 Nominal Guide tube ID (inches) 0.499 Nominal Maximum enrichment ( wt% m u) 5.0 Maximum blanket enriclunent (wt% m u) 4.0 Minimum blanket length (inches) 8 Requirement for enrichments less than or equal to 4.5, 4.0, Minimum number ofIFBA rods* 64 (IX loading) 3.5, and 3.0 w/o is 48, 32, 16, and O IFBA, respectively Minimum IFBA length (inches - centered)* 128 Design changes that increase 0.00603 g 10B/cm Maximum W ABA loading water displacement are not per rodlet covered.

10 [ mg 10B/inch] "*c ( l .5X)

Maximum IFBA rods and B loadi ng 148 IFBA rods per rod

• These requirements are only for storage of fuel assemblies that have not been in the core.

The depletion parameters are selected to cover anticipated future operation, however, verification is required. Table 10.4 lists the operating requirements from the depletion analysis for Batch Z (recently discharged and future fuel). The temperature and so luble boron assumptions are averages over the total burnup (multi-cycle) for a given assembly. Ifan assemb ly is depleted such that any of the Table 10.4 parameters are not met, then the assembly would have to be classified as Category 1 fuel or classified as Category 4 fuel with a control rod inserted.

NET- 2809 1-0003-01, Revision 0 164

Table 10.4: Fuel Assembly Operating Requirements Parameter Value Notes Maximum Core Inlet 542 .6 Of Temperature Maximum Core 62.4 Of Delta T Maximum Operation This control rod inserted bumup covers rods

.::: 2 GWd/T with Control Rods inserted to any depth.

This is an average for all cycles in which the assembly was depleted. If 950 ppm is Maximum Bumup exceeded, as stated in Section 10.2, a bumup Averaged Soluble .:S 1200 ppm penalty of0.2, 0.3 , 0.6, and 0.9 GWd/T must Boron be added to the MRB for Categories 2, 3, 4, and 5 respectively.

Average Power To cover reduced power operation at end of During the Last 30 > 50%

cycle prior to offload Days of Operation NET- 2809 1-0003-01 , Revision 0 165

References

[1] Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels , NET-300067-01, Rev. 1, Curtiss-Wright Nuclear Division, NETCO, Danbury, CT, February 2015 . (Accession Number: ML15062A200)

[2] " Response to Request for Additional Information Regarding the Indi an Point uclear Generating Unit No. 2 - Spent Fuel Pool Criticality Analysis," Letter L 089, Entergy Nuclear Northeast, Buchanan, NY , August 14, 2015. (Accession Number:

ML15261A527, Non Proprietary

Attachment:

ML15261A528)

[3] "Indian Point Nuclear Generating Unit No. 2 - Staff Review of NETCO Report NET-300067-0 1, "Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels,"" US NRC, Washington, DC, November 23 , 2015 . (Accession Number: ML15292Al61)

[4) Guidance for Performing Criticality Analyses of Fu el Storage at Light- Water Reactor Power Plants, Revision 2, Draft C, NEI 12-16, Nuclear Energy Institute, Washington, DC, August 2017.

[5] K. Wood, "Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 ,

Revision 0, Staff Guidance Regarding the Nuclear Criticality Safety Analysis For Spent Fuel Pools," Access ion Number MLl 10620086, Nuclear Regulatory Commission, Rockvi ll e, MD, October 2011.

[6] Code of Federal Regulations, Title l 0, Part 50, Section 68, " Criticality Accident Requirements."

[7] Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNLff M-2005/39, Vers ion 6. 1, June 20 11. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-785 .

Includes 6.1.2 update dated 2/28/ 13.

[8] "Rack Construction (SH'T. 2) Region l Storage Racks," Drawing Number 398 Rev 5, March 19, 1990, Project No. 81000, P. 0. No. 8-24470, Holtec International , Mount Laurel, NJ.

[9] " Rack Construction (S H'T. I) Region 2 Storage Racks ," Drawing Number 400 Rev 5, March 19, 1990, Project No. 81000, P. 0 . No. 8-24470, Holtec International, Mount Laurel, NJ.

[ 1OJ IN TRPND8: Verification and Validation, CWND, NETCO, Danbury, CT: November 20 17

[ 11] Indian Point Unit 3 Updated Final Safety Analysis Report (UFSAR) , Revision 03, October 3, 2009.

[12) Indian Point Unit 2 Updated Final Safety Analysis Report (UFSAR) , Revision 26, 2016.

NET- 28091 -0003-01 , Revision 0 166

[ 13] Dale Lancaster, Sensitivity Analysis for Spent Fu el Pool Criticality, EPRI, Palo Alto, CA:

2014, Technical Report Number: 3002003073 .

[14] C. E. Sanders and J.C. Wagner, Parametric Study of the Effect of Control Rods for PWR Burnup Credit, NUREG/CR-6759, prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn. , February 2002.

[15] G. Radulescu, I. C. Gauld, and G . Ilas, SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions, ORNL/TM-2010/44, Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA, March 2010.

[ 16] International Handbook of Evaluated Criticality Safety Benchmark Exp eriments, NEA/NSC/DOC(95)3, Nuclear Energy Agency, OECD, Paris, September, 2010.

[ 17] J. J. Lichtenwalter, S. M. Bowman , M. D. DeHart, and C. M . Hopper, Criticality Benchmark Guide for Light- Water-Reactor Fuel in Transportation and Storage Packages, NUREG/CR-6361 (ORNL/TM-13211), Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 , March 1997.

[18] L. W. Newman, et al , Urania-Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810 , Babcock & Wilcox, Utility Power Generation Division, Lynchburg, VA, April 1984.

[19] G. Radulescu, I. C. Gauld, G. Ilas, and J. C. Wagner, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses - Isotopic Composition Predictions, NUREGICR-7108, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC, USA, April 2012.

[20] K. S. Smith, et al., Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty, EPRI, Palo Alto, CA, Technical Report Number 1022909 (201 1).

[21] D. B. Lancaster, Utilization of the EPRJ Depletion Benchmarks for Burnup Credit Validation, EPRI, Palo Alto, CA, 1025203 (2012).

[22] J. M. Scaglione, D. E. Mueller, J.C. Wagner and W. J. Marshall, An Approach for Validating. Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (k,JJ) Predictions , US Nuclear Regulatory Commission, NUREG/CR-7109, Oak Ridge National Laboratory, Oak Ridge, Tenn. (2012).

[23] J.C. Dean and R.W. Tayloe, Jr., Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698 , Nuclear Regulatory Commission, Washington, DC January 2001.

[24] D. E. Mueller, K. R. Elam, and P. B. Fox, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data , NUREG/CR-6979 (ORNL/TM-2007/083),

prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 2008.

NET- 28091-0003-01 , Revision 0 167

(25] Design Input Record, EN-DC-149Rl4 14021 8, Indian Point, February 17, 2014 (26] B. B. Bevard, J.C. Wagner, and C. V. Parks, Review ofInformation for Spent Nuclear Fuel Burnup Confirmation, NUREG/CR-6998, prepared for the US Nuclear Regulatory Commission by Oak Ridge Nationa l Laboratory, Oak Ridge, Tenn., December 2009.

(27] J.C. Wagner, M. D. DeHart, and, C. V. Parks, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, US Nuclear Regulatory Commission, NUREG/CR-6801 , Oak Ridge National Laboratory, Oak Ridge, Tenn. (2003) .

(28] D. Hagrman, INTERPIN-3 User's Manual, SSP-01/430, Studsvik Scandpower, Inc.

(2001).

(29] R. E. Griffith to G. Delfini, "Fuel Temperatures vs Burnup Curves for Indian Point Unit 2," Entergy lnter-Office Correspondence, CE020 13-001 07, August 15, 2013, supported by Entergy Calculation package NEAD-SR-13/023 , EC 46111.

[30] Criticality Safety Evaluation of the North Anna New Fuel Storage Area and Spent Fuel Pool Allowing 5 wt% U-235 Enriched Fuel, Nuclear Engineering and Fuel, Dominion Resources Services, Inc., November 2016. (Accession Number: MLl 7129A452)

(31] J.C. Wagner and C. V. Parks, Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit, US Nuclear Regulatory Commission, NUREG/CR-676 1, Oak Ridge National Laboratory, Oak Ri dge, Tenn. (2002).

(32] C. E. Sanders and J.C. Wagner, Study of the Effect of Integral Burnable Poison Rods for PWR Burnup Credit, US Nuclear Regulatory Commission, NUREG/CR-6760, Oak Ridge National Laboratory, Oak Ridge, Tenn. (2002).

(33] C. V. Parks, M. D. DeHart, and J.C. Wagner, Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fu el, US Nuclear Regulatory Commission, NUREG/CR-6665, Oak Ridge Nationa l Laboratory, Oak Ridge, Tenn. (2000) .

(34] M. D. DeHart, Sensitivity and Parametric Evaluations of Significant Aspects of Burnup Credit for PWR Spent Fuel Packages, ORNL/TM-12973 , Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, May 1996.

(35] "Poo l Layout Spent Fuel Storage Racks," Drawing Number 397 Rev 4, December 8, 1989, Project No. 81000, P. 0. No. 8-24470, Holtec International , Mount Laurel, NJ.

(36] "Criticali ty Analysis for Soluble Boron and Burnup Credit in the Con Edison [the former licensee] Indian Point Unit No. 2 Spent Fuel Storage Racks , NET-173-01, September, 2001. (Accession Number: MLO 12680336)

(37] G.P. Sabol, G. Schoenberger, and M.G. Balfour, "Improved PWR Fuel Claddi ng,"

Materials for Advanced Water Cooled Reactors, Proceedings of a Technical Committee Meeting, Plzen, Czechoslovakia, May 14-17, 1991 , IAEA-TECDOC-665 , IAEA, VIENNA, 1992.

NET- 2809 1-0003-01, Revision 0 168

[38] Sabol, G. P, Comstock, R. J., Weiner, R. A., Larouere, E, and Stanutz, R. N., "In-Reactor Corros ion Performance of ZIRLO and Zircaloy-4," Zirconium in the Nuclear Industry:

Tenth International Symposium, ASTM STP 1245, A. M. Garde and E. R. Bradley, Eds. ,

American Society for Testing and Materials, Philadelphia, 1994, pp. 724-744.

[39] Garzarolli, F. , Manzel, R., Reschke, S., and Tenckhoff, E., "Review of Corrosion and Dimensional Behavior of Zircaloy under Water Reactor Conditions," Zirconium in the Nuclear Industry (Fourth Conference), ASTM STP 681, American Society for Testing and Materials, 1979, pp.91-106.

[40] Y. Irisa, et al, "Segmented Fuel Rod Irradiation Program On Advanced Materials For High Burnup," An International Topical Meeting on Light Water Reactor Fu el Performance, Park City, Utah, Apri l 10-13 , 2000, American Nuclear Society, La Grange Park, Illinois.

[41] C. B . Lee, et al , "Post-irradiation Examination of High Bumup U02 Fuel," Proceedings of the 2004 International Meeting on L WR Fuel Performance, Orlando, Florida, September 19-22, 2004, American Nuclear Society, La Grange Park, Illinois.

[42] David Mitchell , Anand Garde, and Dennis Davis, "Optimized ZlRLO' Fuel Perfom1ance in Westi nghouse PWRs," Proceedings of th e 2010 LWR Fu el Performance Meeting/Top Fuel/ WRFPM, September 26-29, 20 10

• Orlando, Florida, USA, American Nuclear Society, La Grange Park, Illinois.

[43] R . Manzel and C. T. Walker, " High Burnup Fuel Microstructure And lts Effect On Fuel Rod Performance," An International Topical Meeting on Light Water Reactor Fuel Performance, Park City, Utah, Apri l 10-13 , 2000, American Nuclear Society, La Grange Park, Illinois.

[44) Dennis Gottuso, Jean-Noel Canat, Pierre Mollard, "A Family of Upgraded Fuel Assemb lies for PWR," Top Fuel 2006, 2006 International Meeting on LWR Fuel Performance, October 22-26, 2006, Salamanca, Spain, European Nuclear Society.

[45] King, S. J. , Kesterson , R. L., Yueh, K. H. , Comstock, R. J., Herwig, W. M., and Ferguson, S. D., "Impact of Hydrogen on Dimensional Stability of ZIRLO Fuel Assemblies," Zirconium in the Nuclear Industry: Thirteenth International Symposium ,

ASTM STP 1423 , G. D. Moan and P. Rudling, Eds., ASTM International, West Conshohocken, PA, 2002, pp. 471-489.

[46] R. L. Kesterson , S. J. King, and R. J. Comstock, "Impact of Hydrogen on Dimensional Stability of Fuel Assemb li es," An International Topical Meeting on Light Water Reactor Fuel Performance, Park City, Utah, April 10-13 , 2000, American Nuclear Society, La Grange Park, Illinois.

[47] Morize, P., Baicry, J. , and Mardon, J. P., "Effect oflrradiation at 588 Kon Mechanical Properties and Deformation Behavior of Zirconium Alloy Strip," Zirconium in the Nuclear Industry: Seventh International Symposium, ASTM STP 939. R. B. Adamson and L. F. P. Van Swam, Eds., American Society for Testing and Materials, Philadelphia, 1987, pp. 101-119.

NET- 2809 1-0003-01, Revision 0 169 j

[48] Steven J. King, Michael Y. Young, Fabrice M . Guerout, and Nigel J. Fisher, "Fretting-Wear Behavior of Zircaloy-4, OPTIN, and ZIRLO Fuel Rods and Grid Supports Under Various Autoclave and Hydrauli c Loop Endurance Test Conditions," Zirconium in the Nuclear Industry: Fourteenth International Symposium, ASTM STP 1467 , P. Rudling and Bruce Kammenzi nd, Eds., ASTM International, West Conshohocken, PA, 2006, p. 826.

[49] Response To Request For Additional Information Regarding Proposed Technical Specification Change For Spent Fu el Storage (Non-Proprietary), Response to RAI-23 ,

Letter from Dominion Nuclear Connecticut to the USNRC, July 2 1, 2015. (NRC Adams Accession Number: ML15209A729 .)

[50] G.E. Whitesides, "A Difficulty in Computing the k -Effective of the World," Trans. Am.

Nucl. Soc., 14 , pp. 680 (1971 ).

[51] Brian C. Kiedrowski and Forrest B. Brown, "Difficulties Computing kin Non-Unifonn, Multi-Region Systems with Loose, Asymmetric Coupling," Proceedings of the 9th International Conference on Nuclear Criticality Safety (ICNC20 1 l), Edinburgh, Scotland, September 19-22, 2011 .

[52] NET-173-02 . Rev. l , " Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis, September 2001. (Access ion Number: ML012680336)

[53] ANSVANS-8 .1-1998 (R2007), "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," American Nuclear Society, La Grange Park, Illinois.

[ 54] Calculation Notebook, NET-901-02-08, SCALE 6.1.2 Validation for Criticality Analysis

- Amendment Calculations for IP2, Rev.0; CWND, Danbury, CT.

[55] K. Lindquist, et al, Guidelines for Bora.flex Use in Spent-Fuel Storage Racks, EPRI, Palo Alto, CA, Technical Report Number 103300 (1983).

[56] NRC Memorandum from L. Kopp to T. Collins, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 19, 1998. (Access ion Number: MLl 1088A0 13)

[57] J.C. Wagner, "Impact of Soluble Boron Modeling for PWR Burnup Credit Criticality Safety Analyses," Trans. Am. Nucl. Soc. , 89, pp. 120 (2003).

[5 8] D. R. Lide, CRC Handbook of Chemistry and Physics, 85 1h Edition, CRC Press LLC, Boca Raton, FL. (2004).

NET- 28091-0003-01, Revision 0 170

Appendix A: Validation of SCALE 6.1.2 for Criticality Analysis Using Laboratory Critical Experiments A. 1. Overview This appendix detennines the computer code and cross-section library bias and uncertainty in the keff values calculated for the Indian Point Units 2 and 3 spent fuel pools when using SCALE 6.1.2. [ 1] The bias and uncertainties detennined in this Appendix cover the major actinides, absorbers , and structural materials for the spent fuel pool with fresh or burned fuel.

This analysis uses the CSAS5 module of SCALE 6.1.2. [ 1] All of the analyses are performed using the 238 group ENDF/B-VII library (v7-238) . The CSAS5 module executes the CENTRM and BONAM!

programs for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of

k. All of the computer runs use a large Monte Carlo sampling of at least 1500 generations and 6000 neutrons per generation.

This Appendix is divided into three sections: 1) U02 critical experiments, and 2) HTC and MOX critical experiments, and 3) Temperature Dependent critical experiments. After these three sections is a summary section.

A.2. U02 Laboratory Critical Experiments A.2.1 Introduction The validation consists of modeling 328 U02 critical experiments and the determination of the bias and the uncertainty in the calculation of keff for U02 fuel. This validation follows the direction of NUREG/CR-6698, "Guide for Validation of Nuc lear Criticality Safety Calculational Methodology" [2] .

The guide establishes the following steps for performing the validation:

1. Define operation/process to identify the range of parameters to be validated

2. Select critical experiment data

3. Model the experiments

4. Analyze the data

5. Define the area of applicability of the validation and limitations It further defines the steps of "Analyze the data" as :

1. Detennine the Bias and Bias Uncertainty NET- 28091-0003-01, Revision 0 A-1

2. Identify Trends in Data, Including Discussion of Methods for Establishing Bias Trends

3. Test for Norma l or Other Distributions

4. Select the Statistical Method for Treatment of Data

5. Identify and Support Subcritical Margin

6. Calculate the Upper Safety Limit This approach is followed for this validation analysis.

A.2.2 Definition of the Range of Parameters to Be Validated The validation guidance document [2] states:

"Prior to the initiation of the validation activity, the operating conditions and parameters for which the validation is to apply must be identified. Th e fissile isotope, enrichment offissile isotope, fu el density, fu el chemical form, types of neutron moderators and reflectors, range of moderator to fissile isotope, neutron absorbers, and physical configurations are among the parameters to specify. These parameters will come to define the area of applicability for the validation effort . "

Almost all pool applications have common neutronic characteristics and therefore can be validated together. The racks are assumed to be flooded with water at near room temperature and below 100 °C.

The fuel is low enriched uranium dioxide (less than or equal to 5.0 wt% U-235). The fuel is in pellets with a density of greater than 92% of the theoretical density. The only significant neutron moderators are water and the oxygen in the fuel pellet. The neutron absorbers credited are boron (as plates, perhaps rods ,

or in solution) and Ag-In-Cd control rods. The reflectors are water, steel, or concrete. The fuel is in assemblies, but the ana lysis is also valid for disassembled assemblies . The assembly arrangement can vary by design from totally isolated assemblies to a close packed array of assemblies.

A.2.3 Selection of the Fresh U02 Critical Benchmark Experiments The U02 benchmarks that were selected met the following criteria:

• Low enriched (5 wt% U-235 or less) U02 to cover the principle isotopes of concern.

• Fuel in rods to assure that the heterogeneous analysis used in SCALE also is applied in the benchmark analysis.

• Square lattices to assure the lattice features of SCALE used in the rack analysis are also modeled in the critical benchmarks selected.

• Presence of boron as soluble boron, borated steel, boron bearing rods , sheets of aluminum with boron, or BoraflexTM.

• Presence of Ag-In-Cd control rods .

• No emphasis on a feature or material not of importance to the rack analysis .

The OECD/NEA International Handbook of Evaluated Criticality Safety Benchmarks Experiments [3] is now considered as the appropriate reference for criticality safety benchmarks. This handbook has reviewed the available benchmarks and evaluated the uncertainties in the experiments. The appropriate modeling is presented. All of the experiments used in this validation except some experiments for Ag-In-NET- 28091-0003-01, Revision O A-2

Cd control rods were taken from this handbook. Volume IV of the handbook is for low enriched uranium systems. The section of Volume IV of interest to this validation is the "Thermal Compound Systems."

All of the experiments selected are numbered LEU-COMP-THERM-OXX. This validation will refer to the experiments LEU-COMP-THERM-OXX as just XX where any leading zero is not included.

There are more critical experiments in the handbook that meet the requirements for this validation than would be necessary to use. However, most of the applicable available benchmarks were used. There are 95 sets of benchmarks in the 2016 version of the handbook. 25 of these were eliminated since they were for hexagonal arrays . Five more were eliminated due to high enrichments. Seven experiments were not for light water moderated U02 fuel rods . Four experiments were eliminated due to high uncertainties.

Five more were eliminated since they depend on features not in the spent fuel pool. This leaves 49 benchmark sets of which 35 were used for this validation. The 14 unused benchmark sets were reviewed to be sure that there was no feature of the experimental set that was missing in the selected 35 sets .

The international handbook only had two sets of experiments from one laboratory with Ag-In-Cd control rods . In order to fully ascertain if Ag-In-Cd control rods introduce a bias a search of additional criticals was performed. The first compilation searched was NUREG/CR-6361 [5] , which was a report for the validation of SCALE. In that report were Ag-In-Cd critical experiments from two reports, BA W-1810 [6]

and WCAP-3269 (data taken from NUREG/CR-6361). Finally, Lawrence Livermore Laboratory compiled critical experiments from the ANS national meeting. [7] A search of this source did not provide any additional Ag-In-Cd critical experiments.

The selected 35 benchmark sets from the international handbook and the 2 additional sets for Ag-In-Cd rods include critical experiments from eight different critical experiment facilities. The fuel was mainly clad in aluminum but experiments with stainless steel and zirconium cladding were also in the set.

The critical benchmark sets generally contained multiple experiments but not all cases from each critical benchmark set is used . In some sets there are experiments that emphasize features that are out of scope of this validation such as lead or copper reflectors . The 37 selected benchmark sets resulted in 328 experiments that are used for the statistical analysis. 85 experiments used boron (soluble or in absorber plates).

A later section will evaluate the area of applicability provided by this selection of critical benchmarks .

Table A. l provides a summary of all of the low enriched thermal experiments (non-U metal) from the OECD/NEA handbook [3] and why some experiments were not used.

Table A.l: Selection Review of OECD/NEA Criticality Benchmarks (All Experiments Start With LEU-COMP-THERM-)

Benchmark Description Lab Selected?

Number WATER-MODERATED U(2.35)02 FUEL l RODS lN 2.032-C M SQUAR E-PITC HED PNL All 8 ARRAYS WATER-M ODERATED U(4 .31 )0 2 FUEL 2 RODS lN 2.54-CM SQUARE-PITCHED PNL All 5 ARRAYS NET- 28091-0003-01 , Revision 0 A-3

Benchmark Description Lab Selected?

Number WATER-MODE RATED U(2 .35)02 FU EL None. Gd impmity not well RODS IN 1.684-CM SQUARE-PITC HED 3 PNL known . Not benclunark ARRAYS (GA DOLIN IUM WATER quality.

IMPUR ITY)

WATER-MODERATED U(4.31 )02 FUEL None. Gd impu1ity not well RODS IN 1.892-CM SQUARE-PITC HED 4 PNL known. Not benchmark ARRAYS (GADOLIN IUM WATER quality.

IMPUR ITY)

CRITlCAL EX PERIM ENTS WITH LOW-None. No sample SCALE EN RICHED URAN IUM DIOXIDE FUE L 5 PNL decks . Soluble Gd not used in RODS IN WATER CONTA INING pools.

DIS SOL YEO GADOLIN IUM CRIT ICA L ARRAYS OF LOW- EN RICHED U02 FUEL RODS WITH WATER-TO-FUEL 6 JAEA All 18 VO LUME RATIOS RANGING FROM 1.5 TO 3.0 WATER-REF LECTE D 4.738-WT.%-

Only 4 cases used rest are in 7 EN RICHED URAN IUM DIOXIDE FUEL- Ya lduc hexagonal arrays.

ROD ARRAYS CRITI CAL LATT ICES OF U02 FUEL ROD S 8 AN D PERTURBING RODS IN BORATED B&W All 17 WATER WATER-MODERATED RECTANGU LAR CLUSTERS OF U(4.3 1)02 FUEL RODS (2.54- 2 1 cases used. Did not 9 CM PITCH) SEPARATED BY STEEL, PNL include Copper cases sin ce no BORAL, COPP ER, CADMIUM, ALU MINUM , copper in pools.

OR ZrRC ALOY-4 PLATES WATER-MODERATED U(4.3 1)02 FUEL 22 cases used. Did not use 10 RODS REF LECTED BY TWO LEA D, PNL lead cases since no lead in URANIUM , OR STEE L WALLS pools.

CRITlCAL EX PER IM ENTS SUPPORTING CLOSE PROXIM ITY WATER STORAGE OF II B&W All 15 POWER REACTOR FUEL (PART I -

ABSO RB ER RODS)

WATER-MODERATED RECTANGULAR CLUSTE RS OF U(2 .35)02 FUE L None. Gd impurity not well ROD S( l.684-CM PITCH) SE PARATED BY 12 PNL known . Not benclunark STEEL, BORAL, BOROFLEX, quality.

CADM IUM ,OR COPP ER PLATES

/G ADOLIN IUM WATER IMPUR ITY)

WATER-M ODERATED RECTANGU LAR CLUSTERS OF U(4 .3 1)02 FUEL RODS

( 1. 892-CM PITCH) SE PARATED BY STEEL, 5 cases used. Did not use the 13 PNL BORAL, BOROFLEX , CADM IUM, OR 2 cases with copper.

COP PER PLATES , WITH STEE L REFLECTING WALLS WATER-REFLECTE D ARRAYS OF None used. High boron U(4.3 1)02 FUEL RODS ( 1.890-CM AND 14 PNL content uncertainty. Not 1.7 15-CM SQUARE PITCH) IN BORA TED benclunark qua lity.

WATER TH E VVER EX PERIM ENTS: REGU LAR AN D PERTURBED HEXAGONA L 15 KFK I None used due to hex arrays.

LATTICES OF LOW-EN RICHED U02 FU EL RODS IN LI GHT WATER WATER-MODERATE D RECTANGU LAR CLUSTERS OF U(2 .35)02 FUE L RODS 26 cases used. Did not use 16 (2.032-CM PITCH) SEPARATED BY STEEL, PNL the 6 copper cases BORAL, COPPE R, CADMnJM, ALUMIN UM, OR ZrRCA LOY -4 PLATES WATER-M ODERATED U(2 .35)02 FUEL 23 cases used. Did not use 17 RODS REFLECTED BY TWO LEAD , PNL the 6 cases with a lead URAN IUM , OR STEEL WALLS refl ector.

NET- 28091-0003-01 , Revision 0 A -4

Benchmark Description Lab Selected?

Number LIG HT WATER MO DERATED AND 18 REFLECTED LOW ENR ICHED URAN IUM Winfrith None used . Complex system .

DIOXlDE (7 WT.%) ROD LATTICE WATER-MODERATED HEXAGONALLY 19 PITCHED LATT ICES OF U(5%)02 Kurchatov Institute None used due to hex arrays.

STAINLESS STEEL C LAD FUEL RODS WATER-MODERATED HEXAGONALLY PITCHED PARTIALLY FLOODED 20 Kurchatov Institute None used due to hex arrays .

LATT ICES OF U(5%)02 ZIRCONIUM CLAD FUEL RODS , 1.3-CM PITC H HEXAGONALLY PITC HED PART IALLY FLOODED LATT ICES OF U(5%)02 21 ZIRCONIUM CLAD FUEL RODS Kurchatov Institute None used due to hex arrays .

MODERATED BY WATE R WITH BOR IC AC ID UN IFORM WATER-MODERATED 22 HEXAGONALLY PITC HED LATTICES OF Kurchatov Institute None used due to hex arrays .

RODS WITH U( l 0%)02 FUEL PARTIALLY FLOO DED UN IFORM 23 Kurchatov Institute None used due to hex arrays .

LATT ICES OF RODS WITH U( l0%)02 FUEL WATER-MODERATED SQUARE-P ITCHED Did not use either case due to 24 UN IFORM LATT ICES OF RODS WITH Kurchatov Institute IO wt% U-235 enriclunent U( I 0%)02 FUEL WATER-MODERATED HEXAGONAL LY 25 PITCHED LATT ICES OF U(7.5%)02 Kurchatov Institute None used due to hex arrays.

STAINLESS-STEEL-C LAD FUEL RODS WATER-MODERATED U(4.92)02 FUEL RO DS IN 1.29, 1.09, AN D 1.01 CM PITC H 26 IPPE None used due to hex a1rnys.

HEXAGONAL LATT ICES AT DIFFERENT TEM PERATURES WATER-MODERATED AND LEAD-None used due to lead 27 REFLECTED 4.738% ENR IC HED URAN IUM Valduc reflector.

DIOXIDE ROD ARRAYS WATER-MODERATED U(4 .3 1)02 FUEL RO DS IN TRIANGU LAR LATTICES WlT H 28 PNL None used due to hex arrays.

BORON , CADMIUM AN D GADO LIN IUM AS SOLUB LE PO ISONS None used . hf plates cases WATER MODERATED AND WATER without Hf have the same REFLECTED 4.74% ENR ICHED URAN IUM 29 Valduc pitch and pin as benclunark 7 DIOX ID E ROD ARRAYS SURROUNDED BY above. No significant HAFN IUM PLATES additional va lue.

VVER Physics Experiments: Reg ular Hexagonal ( l.27-cm Pitch) Lattices of Low-30 Enriched U(3.5 Wt.% 235U)02 Fuel Rods in Kurchatov Institute None used due to hex a1rnys.

Light Water at Different Core Critica l Dimensions WATER-MODERATED HEXAGONALLY PITCHED PARTIALLY FLOODED 31 Kurchatov Institute None used due to hex arrays.

LATT ICES OF U(5%)02 ZIRCON IUM-C LAD FUEL RODS, 0.8-CM PITC H UN IFORM WATER-MODERATED 32 LATT ICES OF RO DS WITH U{l0%)02 FUEL Kurchatov Institute None used due to hex arrays .

IN RANGE FROM 20°c TO 274°C REFLECTED AND UNREFLECTED 33 ASSEMBLIES OF 2 AN D 3%-ENR ICHED ORNL None used. NotU02 URAN IUM FLUOR IDE IN PARAFFIN FOUR 4.738-WT.%-EN RICHED URAN IUM 6 cases used. Di d not use DlOX lDE ROD ASSEMB LIES CONTA INE D cases with gap less than 2.5 34 IN CADMIUM , BO RATED STAIN LESS Valduc c m due to high uncertainty.

STEE L, OR BORAL SQUARE CAN ISTERS , Did not use Cd plate cases WATER-MODERATED AND-REFLECTED since Cd plates not in poo l.

NET- 28 091-0003 -01 , Revision 0 A -5

Benchmark Description Lab Selected?

Number CR IT ICAL ARRAYS OF LOW-ENR IC HED Used 2 cases. Did not use the U02 FUEL RODS IN WATER WIT H 35 JAEA case with disso lved Gd . (not SOLUBLE GADO LINI UM OR BORON like pool).

PO ISON T HE VVER EX PERCM ENTS: REGULAR AND PERTU RB ED HEXAGONA L 36 KFKI None used due to hex arrays.

LATT ICES OF LOW-ENR IC HED U02 FUEL RODS IN LIG HT WATER - Part 2 WATER-MODERATED AND PARTIALLY CONCRETE-REFLECT ED 4.738- WT.%- None used. No Significant 37 Valduc ENR ICHED URAN IUM DIOXIDE ROD Va lue added .

ARRAYS WATER-MODERATED 4 .738-WT.%-

None used. Used a borated EN RICHED URAN IUM DIOX LD E ROD 38 Valduc concrete reflector (not like ARRAYS NEXT TO A BORA TED pool).

CONC RETE SCREEN INCOM PLETE ARRAYS OF WATER-39 REFLECTED 4 .738-WT.%-ENR IC HED Valduc Used all 17 cases.

URAN IUM DIOXIDE FUEL-ROD ARRAYS FOU R 4 .738-WT.%-EN RIC HED URAN IUM DIO XID E ROD ASSE MBLLES CONTA INED IN BOR ATED STA IN LESS STEEL OR Used 4 cases. Did not use 40 Valduc BORAL SQUARE CAN ISTERS , WATER lead re fl ector cases.

MODERATED AN D REFLECTED BY LEAD OR STEEL STO RAGE ARRAYS OF 3%-EN R!C HED Did not use the 5 cases due to 41 LWR ASSEMBLIES: THE C R ISTO II Cadarache complex geometry.

EX PER IMENT IN Tl-I E EO LE REACTOR WATER-MODERATED RECTANGU LAR C LUSTERS OF U(2.35)02 FUEL RODS

( 1. 684-CM PITC H) SEPARATED BY STEEL, Used 5 cases. Did not use 42 PNL BORAL, BOROFLEX, C ADMIUM , OR copper cases.

COPPER PLATES, W ITH STEE L REFLECTING WALLS CR ITICAL LOADING CONF!GURA TIONS Used only one case. Rest of 43 O F TH E IPEN/MB-0 1 REACTOR WITH A  !PEN cases were not significantly HEAVY SS-304 REFLECTO R different.

C RITICAL LOADING CON FIGURAT IONS Used only one case . Rest of 44 OF Tl-I E IPEN/MB-0 1 REACTOR WIT H U02,  !PEN cases were not significantly STAINL ESS STEE L AND COPPER RODS different.

PLEX IGLAS OR CONCRETE-REFLECTE D None used since not pin 45 U(4. 46)308 W ITH H/U=0.77 AND Rocky Flats geometry.

INTERST IT IAL MODERATION C RITICAL LOAD ING CONF IGURAT IONS OF T HE IPEN/ MB -0 1 REACTO R Used 17 cases that did not 46 IP EN CONSIDERING TEMPERATURE have cooper pins.

VAR IAT ION FROM 14°C TO 85 °C FUEL TRANSPORT FLASK C RITICA L 47 BENCHMARK EXPERCM ENTS WITH LOW- Winfrith None used . 3 complex cases.

EN RICHED URAN IUM DIOX ID E FUEL LIGHT WATER MO DERATED AN D 48 REFLECTED LOW-ENR IC HED (3 WT. % Winfrith A II 5 cases used 235U) URANIUM DIOXID E ROD LATT ICES MARACAS PROGRAMME: POLYTHENE-REFLECTED C RITICAL CONF IGURATIONS None used . Powder rather 49 W ITH LOW-ENR IC HED AND LOW- Valduc than pellets. Not similar to MODERATED URAN IUM DIOX ID E poo ls.

POWDER, U/5)02 149SM SOLUTION TANK IN THE M IDDL E 7 cases used. Did not use OF WATER-MODERATED 4.738-WT. %-

50 Valduc cases with disso lved Sm.

EN RICHED UR AN IUM DIOXI DE ROD This is not typical o f pools.

ARRAYS NET-28091-0003-01 , Revision 0 A-6

Benchmark Description Lab Selected?

Number 9 cases used. Di d not use CR IT ICAL EXPERrMENTS SUPPORTING cases with the borated Al CLOSE PROXrMITY WATER STORAGE OF 51 B&W plates since primaiy source POWER REACTOR FUEL (PART II -

listed a high uncertainty in the ISOLAT ING PLATES) boron content.

URAN IUM DIOX IDE (4.738-WT.%-

ENR ICHED) FUEL ROD ARRAYS 52 Valduc None used due to hex arrays .

MODERATED AND REFLECTED BY GADOLIN IUM NITRATE SOLUT ION VVER PHYSICS EXPE R[MENTS : REGULAR HEXAGONAL ( 1.27 CM PITCH) LA TT ICES OF LOW-ENRICHED U(4.4 WT.% 235U)02 53 Kurchatov Institute None used due to hex arrays .

FUEL RODS IN UGI-IT WATER AT DIFFERENT CORE CR ITICAL DIMENS IONS CR ITICAL LOADING CONF IGURATIONS Used on ly one case. Rest of 54 OF THE IP EN/MB-0 1 REACTOR WITH U02, IPEN cases were not significantly AND U02-Gd203 RODS different.

U GI-I T-WATER MODERATED AN D Neither case used. Comp lex 55 REFLECTED LOW-ENRICHED URAN IUM Winfrith geometry.

(3 wt.% 235U) DIOX IDE ROD LATT ICES CR ITICAL EXPE RIM ENT WITI-I BORAX-V None used. Complex BWR 56 BO ILING WATER REACTOR TYPE FUEL !NL geometry.

ASSEMB LI ES 4.738-WT.%-EN RICHED URAN IUM DIOX IDE FUEL ROD ARRAYS REFLECTED None used. No Significant 57 Valduc BY WATER fN A DRY STORAGE Value added.

CONFIGURAT ION CRITICAL LOAD ING CON FIG URATIONS None used. No Signi ficant 58 OF Tl-I E fPEN/MB-01 REACTOR W!TI-I IPEN Va lue added.

LARGE VO ID IN Tl-IE REFLECTOR 59 Not included in 20 IO Handbook RBMKGRAPI-I ITE REACTOR: UN IFORM CONF IGURAT IONS OF U( I.8, 2.0, or 2.4%

235U)02 FUEL ASSEMBLIES , A D CONF IGURATIONS OF U(2 .0% 235U)02 None used. RBMK - not 60 Kurchatov Institute ASSEMBUES WITI-I EM PTY CHANNELS , typ ical ofLWRs WATER COLUMNS, A D BORON OR THORIUM ABSORBERS, WITI-I OR WITHOUT WATER IN CHANNE LS VVER PHYS ICS EX PERIM ENTS:

HEXAGONAL ( 1.27-CM PITCH) LATTICES OF U(4.4 WT.% 235U)02 FUEL RODS IN LIGHT WATER, PERTURB ED BY BORON, 61 Kurchatov lnstitute None used due to hex arrays .

HAFN IUM , OR DYSPROSIUM ABSORB ER RODS , OR BY WATER GAP WITI-I/WITI-IOUT EM PTY ALUMIN IUM TUBES 2.6%-ENR ICI-IED U02 RODS IN UGI-IT-WATER MOD ERA TOR WITI-I BORA TED None used. No Signi ficant 62 JAEA STAINLESS STEEL PLATE : SINGLE Va lue added.

ARRAYS UGI-I T-WATE R MODERATED AN D REFLECTED LOW-ENR ICHED URAN IUM None used. No Significant 63 Winfrith (3 wt.% 235U) DIOXIDE ROD LATT ICES Va lue added.

WITH DISCRETE POISON-ROD ARRAYS VVER PH YSICS EXPERrMENTS : REGU LAR HEXAGONAL ( 1.27 CM PITCH) LATTICES OF LOW-ENR ICHED U(2 .4 WT. % 235U)02 64 Kurchatov Institute None used due to hex arrays FUEL RODS fN UGI-IT WATER AT DIFFERENT CORE CRITIC AL DIMENS IONS NET- 28091-0003-01 , Revision 0 A-7

Benchmark Description Lab Selected?

Number CR IT ICAL CONFIGURAT IONS OF 2.6%-

ENR IC HED U02 ROD ARRAYS IN LIG HT-None used. No Signi ficant 65 WATER MODERATOR WITH BORA TED JAEA Value added .

STArNLESS STEEL PLATE: COU PLED ARRAYS PLEX IGLAS-REF LECTED, CONCRETE-REFLECTED, OR TH IN STEEL-REFLECTED None used . Not an array of 66 Rocky Flats U(4.46)308 WITH H/U=0.77 AND HEU rods.

DR IVERS CRITICAL LOADrNG CONFIGURAT IONS OF TH E IPEN/ MB-0 1 REACTO R None used since Moly rods 67  !PEN COMPOSED OF FUEL AND are not used in pool.

MOLYBDENUM RODS PLEX IGLAS-REFLECTED, CONCRETE-REFLECTED, OR TH IN STEEL-REFLECTED None used. Not an array of 68 Rocky Flats U(4.48)308 WITH H/U= l.25 OR H/U=2.03 rods.

AND HEU DR IVERS PLEX IGLAS-REFLECTED U(4.48)308 WIT H None used. Not an array of 69 H/U= l.25 OR H/ U=2.03 AND INTERS TITIAL Rocky Flats rods.

MODERATION VVER PHYS ICS EXPERCMENTS: REGULAR HEXAGONAL ( 1.l O-CM PITCH) LATTICES OF LOW-ENRICHED U(6.5 WT.% 235U)02 70 Kurchatov In stitute None used due to hex arrays .

FUEL RODS rN LIG HT WATER AT DLF FE RENT CORE CRITIC AL DIMENS IONS LOW MODERATED 4.738-WT.%-

71 ENR ICHED URAN IUM DIOXIDE FUEL Valduc All 4 cases used.

ROD ARRAYS UNDER-MODERATED 4.738-WT.%-

ENR ICHED URAN IUM DIOXID E FUEL Used 3 cases. Did not use 72 Valduc ROD ARRAYS REFLECTED BYWATER OR Polyethylene refl ector cases .

POLYETHYLENE UNDER-MODERATED 4.738 -WT.%-

ENR ICHED URANIUM DIOXID E FUEL None used. No Significant 73 Valduc ROD ARRAYS REFLECTED BY WATER Value added .

W ITH HETEROGENEITIES MIRTE PROGRAM FOUR 4.738- WT.%-

ENR ICHED URAN IUM-D IOXIDE FUEL-None used. 2 cases without 74 ROD ARRAYS rN WATER SEPARATED BY Valduc Ti screen cou ld be used.

A CROSS-SHAPED SCREEN OF T ITAN IUM (5 MM AND 10 MM THIC K)

VVER PHYS ICS EXPER IMENTS:

HEXAGONAL ( 1.1 0 CM PITC H) LATT ICES OF LOW-ENRIC HED U(6.5 WT.% 235U)02 75 Kurchatov Institute None used due to hex arrays.

FUEL RODS rN LIGHT WATER, PERTURB ED BY BORON ABSORBER RODS AND WATER HOLES LIGHT WATER MODERATED AND REFLECTED LOW ENR IC HED URAN IUM None used . No Significant 76 Win frith (3 WT.% 235U) DIOX IDE ROD LATTICES Value added.

WITH EX-CORE DETECTOR FEA TVRE On ly one case used . Rest of CR IT ICAL LOADrNG CON FIG URAT IONS cases same materials with 77  !PEN OF THE IPEN/MB-01 REACTO R sma ll modification of arrays.

Not sufficientlv independent.

WATER-MODERATED SQUARE-P ITC HED U(6.90)02 FUEL ROD LA TT ICES WITH 0.52 None used . No Significant 78 Sandia FUEL-TO-WATER VOLUME RAT IO Value added.

(0.855 CM PITCH)

WATER-MODERATED U(4.31)02 FUEL 79 ROD LATT ICES CONTA INING RHOD IUM Sandia None used due to hex arrays .

FO ILS NET- 28091-0003 -01, Revision 0 A -8

Benchmark Description Lab Selected?

Number WATER-MODERATED SQUARE-P ITC HED None used. No add iti onal 80 U(6.90)02 FUEL RO D LATT ICES WITH 0.67 Sandia Signi ficant Va lue added.

FUEL TO WATER VOLUME RATI O PWR TYPE U02 FUEL RODS WITH ENR ICHMENTS OF 3.5 AND 6.6 WT.%

81 Single case not use. Unusual WITH BURNABLE ABSORBER ("OTTO ANEX case.

HAHN" NUCLEAR SHIP PROGRAM, SECOND CORE)

CR ITICAL LOADING CONF IGURAT IONS Used only one case. Rest of OF TH E IPEN/MB -0 1 REACTOR WITH LOW 82  !PEN cases were not signi ficantly ENRICHED FUEL AND BURNABLE different.

POISON RODS CR ITICAL LOADING CONF IGURAT IONS Used on ly one case. Rest of 83 OF THE IPEN/MB-0 1 REACTOR WITH A IPEN cases were not significantly BIG CENTRAL VOID different.

CR ITICAL LOADING CONF IGURAT IONS 84 OF TH E IPEN/MB-0 1 REACTOR WITH A IPEN Used the single case.

CENTRAL CRUC IFORM ROD VVER PHYS ICS EX PER[MENTS: REGULAR HEXAGONAL ( 1.27 CM PITCH) LA TT ICES OF LOW-ENR ICHED U(6.5 WT.% 235U)02 85 Kurchatov Institute None used due to hex arrays.

FUEL RODS IN LIGHT WATER AT DIFFERENT CORE CR ITICAL DIMENS IONS VVER PHYS ICS EXPER IMENTS:

HEXAGONAL LATT ICES ( 1.275 CM PITC H) 86 OF LOW ENR ICHED U(3.6, 4.4 WT.% NR I None used due to hex arrays.

235U)02 FUEL ASSE MBLIES IN LI GHT WATE R WITH H3B03 VVER PHYS ICS EXPER IMENTS:

HEXAGONAL LATT ICES ( 1.22-CM PITC H)

OF LOW-ENR ICHED U(3.6, 4.4 WT.%

87 NR I None used due to hex arrays.

U235)02 FUEL ASSE MBLIES IN LI GHT WATER WITH VAR IABLE FUEL-ASSEMBLY PITCH CR ITICAL LOADING CONF IGURAT IONS OF TH E [PEN/MB-0 1 REACTOR WITH 88 IPEN Used all 35 .

HEAVY REFLECTORS COMPOSE D OF CARBON STEEL AND NICKEL CRITICAL LOADING CONF IGURAT IONS Used only one case. Rest of OF TH E IPEN/MB-0 I REACTOR WITH U02 89 IPEN cases were not significantly AN D BORATED STAINLESS STEEL different.

PLATES CR ITICAL LOADING CONF IGURATIONS Used only one case. Rest of 90 OF TH E IPEN/MB-0 I REACTOR WITH U02 IPEN cases were not significantly AND STA IN LESS STEEL RODS different.

CR ITICAL LOAD ING CON FIGURATIONS Used on ly one case. Rest of 91 OF TH E IP EN/MB -0 1 REACTOR WITH U02 , IPEN cases were not significantly STA IN LESS STEE L AND GD203 RODS di fferent.

CR IT ICAL LOADING CONF IGURAT IONS 92 OF THE IPEN/MB-0 I REACTOR WITH IPEN Used all 6.

SOLUB LE BO RON DEUTERfUM CR ITICA L ASSEMBLY WITl-1 1.2% ENR ICHE D URANIUM VAR YING Not used since cases use D20 93 PNC COOLANT VO ID FRACT ION AN D rather than H20 LATT ICE PITCH VVER PHYS ICS EXPER[MENTS: REGULAR HEXAGONAL ( 1.10 CM PITC H) TWO-REGION LATT ICES OF LOW-ENR ICHED 94 Kurchatov Institute None used due to hex arrays.

U(6.5 AND 4.4 WT. % 235U)02 FUEL RODS IN LIGHT WATER AT DI FFERENT CORE CR ITICAL DI MENS IONS 95 Not included in the 20 16 Handbook NET- 28091-0003-01 , Revision 0 A -9

Benchmark Description Lab Selected?

Number PARTI ALLY -R EFLECTED WATER-Used all 19 (Even though MODERATED SQUARE-PITC HED high enrichment, adds an 96 U(6 .90)02 FUE L ROD LATI!CE S WITH 0.67 Sandia independent lab and EALF FU EL TO WATER VOLUM E RATIO coverage)

(0 .800 CM PITC H)

TITANIUM AND/OR ALUMTN UM ROD-REPLACEMENT EXP ERIMENTS IN FULLY- None used. High enrichment REFLECTE D WATER-MODERATED and Ti in many cases . No 97 Sandia SQUAR E-PITCHED U(6 .90)0 2 FUEL ROD additional Significant Value LATTIC ES WITH 0.67 FUEL TOW ATER added .

VOLUME RAT IO (0.800 C M PITC H)

A.2.4 Computer Analysis of the U02 Benchmark Critical Experiments SCALE input decks exist on the OECD/NEA handbook [3] disc for many of the critical experiments. In general, these input decks were used with minor modifications. None of the decks (except LCT-96) were for SCALE 6.1.2 or the ENDF/B-VII library. The number of neutrons per generation and the number of generations were, in general, too low. All of the decks were modified to 6000 neutrons per generation and 1500 generations or more. This was sufficient to make the Monte Carlo uncertainty to be 0.0002 or about one tenth the experimental uncertainty. The input decks matched the isotopic content given in the handbook but this was confirmed. The geometric modeling in the decks also matched the descriptions in the handbook but this too was confirmed. In short, although there was considerable help by starting with the input fi les given in the handbook, the ownership of the files was taken, as required by NUREG/CR-6698 [2] and as stated in section 2.3:

For sp ecific critical exp eriments, the facility or site may choose to use input files generated elsewhere to expedite the validation process. Th e site has the responsibility for ensuring that input files and the options selected are appropriate for use. Regardless of th e source of the input file, the site must have reviewed the description of each critical exp eriment and determined that the representation of the experiment, including simplifying assumptions and options, are consistent with the intended use. In other words, the site must assume ownership of th e input file.

For LCT-8 the input decks were actually 2D models. As part of the International Handbook independent review of LCT-08 eva luation, Virginia Dean, perfonned 3D analysis, found a 0.002 bias, and declared it not "significant." (Appendix D of the evaluation in the International Handbook.) 0.002 is not insignificant when the bias from all of the critical experiments given in Reference l is only 0.0024. The LCT-08 evaluation provided all of the detail for a 3D analysis. It was chosen to reanalyze LCT-08 with a 3D model. With the 3D modeling the average calculated kerr of the LCT-08 cases is 0.9978 (the 2D model given on the International Handbook files yielded an average kerr of 0.9970) .

BAW-1810 [6] reports 23 critical cores. This analysis uses 9 of these cores. Since these cores are not part of the international handbook, the cases were limited to those related to Ag-In-Cd cases . All of the Ag-In-Cd cores were selected as well as the cores that were the closest match where water holes replaced the Ag-In-Cd rods. The B&W faci lities were used for 3 evaluations in the International Handbook, LCT-08 , LCT- 11, and LCT-51 . All three of these sets are used in the validation. Some BAW- 18 10 cases are NET- 28091-0003-01, Revision 0 A-10

used in NUREG/CR-6361 [5]. The input decks for the cases started with the NUREG/CR-6361 decks but changed the Al alloy clad to match the Al alloy atom densities reported in Table 18 of LCT -08 . The input ignores the bottom and top grids. The original source, BAW-1810, was carefully reviewed to be sure that there was good agreement with the International Handbook and NUREG/CR-6361 input decks .

Cases 1-2, 3-4, 5-6, 5A-6A and 8-9 are pairings where the only change was interchanging 16 Ag-In-Cd rods for water holes. The maximum difference between any two cases is 0.00075. The mean difference is 0.00002 where the uncertainty in each case is 0.00008. It is clear that there is no significant difference in the ability to predict keff when there are Ag-In-Cd control rods present.

BAW-1810 does not assign an experimental uncertainty. The mean uncertainties assigned to LCT-08, LCT-11, and LCT-51 are 0.0012, 0.00251, and 0.00207 respectively. An uncertainty of 0.0025 (the largest of the three) was assigned.

The WCAP-3269 input decks are directly from NUREG/CR-6361 with modifications for the newer cross section library, small changes in SCALE input format caused by a newer version of the code, and more neutrons per generation and generations. The uncertainty assigned to these cases is 0.004 . This uncertainty estimate is one of the largest for the entire set of experiments. The average uncertainty of all of the experiments is 0.0019.

Table A.2 shows the results of the analysis of the 328 critical experiments, along with parameters that are used to check for trends in the results. The spectral index, the Energy of the Average Lethargy of the neutrons causing Fission (EALF) is a calculated value from the SCALE output.

Table A.2: Critical Experiment Results with SCALE 6.1.2 and ENDF/B-VII Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. keff ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (~ k)

LCT-1 1 2.350 1.270 2.032 0.0964 0.003 0.9981 2 2.350 1.270 2.032 0.0957 0.003 0.9977 3 2.350 1.270 2.032 0.0950 0.003 0.9970 4 2.350 1.270 2.032 0.0955 0.00 3 0.9976 5 2.350 1.270 2.032 0.0942 0.003 0.9956 6 2.350 1.270 2.032 0.0952 0.003 0.9978 7 2.350 1.270 2.032 0.0934 0.0031 0.9974 8 2.350 1.270 2.032 0.0945 0.003 0.9964 LCT-2 1 4.310 1.415 2.540 0.11 32 0.002 0.99 71 2 4 .310 1.415 2.540 0.1129 0.002 0.9987 3 4 .310 1.415 2.540 0.1129 0.002 0.9984 4 4.310 1.415 2.540 0.1119 0.00 18 0.9979 5 4.310 1.415 2.540 0.1103 0.001 9 0.9962 LCT-6 1 2.596 1.417 1.849 0.23 66 0.002 0.9977 2 2.596 1.417 1.849 0.2432 0.002 0.9987 3 2.596 1.41 7 1.849 0.2495 0.002 0.9987 4 2.596 1.41 7 1.956 0.181 8 0.002 0.9984 5 2.596 1.41 7 1.956 0.1871 0.002 0.9986 6 2.596 1.41 7 1.956 0.1927 0.002 0.9983 7 2.596 1.41 7 1.956 0.1977 0.002 0.9989 NET- 28091-0003-01, Revision 0 A- 11

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. k,rr ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (Ll k) 8 2.596 1.41 7 1.956 0.2028 0.002 0.9986 9 2.596 1.417 2.150 0.1359 0.002 0.9988 10 2.596 1.417 2.150 0.1394 0.002 0.9988 11 2.596 1.417 2.150 0.1427 0.002 0.9985 12 2.596 1.41 7 2.150 0.1462 0.002 0.9982 13 2.596 1.41 7 2. 150 0.1497 0.002 0.9981 14 2.596 1.41 7 2.293 0.1147 0.002 0.9988 15 2.596 1.417 2.293 0.11 74 0.002 0.9983 16 2.596 1.417 2.293 0.1200 0.002 0.9991 17 2.596 1.417 2.293 0.1228 0.002 0.9987 18 2.596 1.41 7 2.293 0.1254 0.002 0.9985 LCT-7 1 4 .738 0.940 1.260 0.2411 0.0014 0.9959 2 4.738 0.940 1.600 0.1090 0.0008 0.9980 3 4.738 0.940 2.100 0.0708 0.0007 0.9976 4 4.738 0.940 2.520 0.0605 0.0008 0.9983 LCT-8 1 2.459 1.206 1.636 0.2845 0.0012 0.9976 2 2.459 1.206 1.636 0.2502 0.001 2 0.9984 3 2.459 1.206 1.636 0.2502 0.0012 0.9990 4 2.459 1.206 1.636 0.2506 0.0012 0.9980 5 2.459 1.206 1.636 0.2506 0.0012 0.9976 6 2.459 1.206 1.636 0.2502 0.0012 0.9977 7 2.459 1.206 1.636 0.2502 0.001 2 0.9971 8 2.459 1.206 1.636 0.2486 0.0012 0.9960 9 2.459 1.206 1.636 0.2479 0.0012 0.9963 10 2.459 1.206 1.636 0.2534 0.0012 0.9978 11 2.459 1.206 1.636 0.2586 0.0012 0.9985 12 2.459 1.206 1.636 0.2524 0.0012 0.9985 13 2.459 1.206 1.636 0.2523 0.0012 0.9985 14 2.459 1.206 1.636 0.2547 0.0012 0.9982 15 2.459 1.206 1.636 0.2546 0.001 2 0.9980 16 2.459 1.206 1.636 0.2315 0.0012 0.9981 17 2.459 1.206 1.636 0.2017 0.001 2 0.9974 LCT-9 1 4.310 1.415 2.540 0.11 27 0.0021 0.9980 2 4.310 1.4 15 2.540 0.11 22 0.0021 0.9986 3 4.310 1.415 2.540 0.1125 0.0021 0.99 79 4 4.310 1.415 2.540 0. 11 21 0.0021 0.9981 5 4.310 1.415 2.540 0.1136 0.0021 0.9993 6 4.310 1.415 2.540 0.1127 0.0021 0.9985 7 4.310 1.415 2.540 0.11 37 0.0021 0.9994 8 4.310 1.415 2.540 0.1130 0.0021 0.998 1 9 4.310 1.415 2.540 0.11 35 0.0021 0.9986 16 4.310 1.415 2.540 0.11 35 0.0021 0.9987 17 4.310 1.415 2.540 0.1127 0.0021 0.9991 18 4.310 1.415 2.540 0.1138 0.0021 0.9977 19 4.310 1.415 2.540 0.1129 0.0021 0.9986 20 4.3 10 1.415 2.540 0.11 37 0.0021 0.9982 21 4.310 1.415 2.540 0.11 29 0.0021 0.9988 22 4.310 1.415 2.540 0.1138 0.0021 0.9984 23 4 .3 10 1.415 2.540 0.1130 0.0021 0.9994 NET- 28091-0003 -01 , Revision 0 A-1 2

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kerr ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (Li k) 24 4.310 1.415 2.540 0.1122 0.0021 0.9979 25 4.310 1.415 2.540 0.1120 0.0021 0.9983 26 4.310 1.415 2.540 0.1121 0.0021 0.9987 27 4.310 1.415 2.540 0.1119 0.0021 0.9985 LCT- 10 5 4.310 1.415 2.540 0.3547 0.0021 1.0000 6 4.310 1.415 2.540 0.26 15 0.0021 1.0003 7 4.310 1.415 2.540 0.2092 0.0021 1.0006 8 4.310 1.415 2.540 0.1844 0.0021 0.9979 9 4.3 10 1.415 2.540 0.1221 0.0021 1.0007 10 4 .310 1.415 2.540 0.1183 0.0021 1.001 3 11 4.310 1.415 2.540 0.1154 0.0021 1.0006 12 4 .310 1.415 2.540 0.1122 0.0021 1.0000 13 4.310 1.415 2.540 0.1105 0.0021 0.9968 14 4.310 1.415 1.892 0.307 1 0.0028 1.0014 15 4.310 1.415 1.892 0.29 50 0.0028 1.0018 16 4.310 1.415 1. 892 0.2853 0.0028 1.0021 17 4.310 1.415 1.892 0.2787 0.0028 1.0021 18 4.310 1.415 1.892 0.2749 0.0028 1.0010 19 4.310 1.415 1.892 0.2677 0.0028 1.0008 24 4.310 1.415 1.892 0.5990 0.0028 0.9994 25 4.310 1.41 5 1.892 0.5536 0.0028 1.001 0 26 4.310 1.415 1.892 0.5122 0.0028 1.001 0 27 4.310 1.41 5 1.892 0.4780 0.0028 1.001 7 28 4.3 10 1.415 1.892 0.4485 0.0028 1.0017 29 4.310 1.415 1.892 0.4232 0.0028 1.0016 30 4 .3 10 1.415 1.892 0.3679 0.0028 0.9996 LCT-1 1 1 2.459 1.206 1.636 0.1685 0.0018 0.9968 2 2.459 1.206 1.636 0.2450 0.0032 0.9967 3 2.459 1.206 1.636 0.1920 0.0032 0.99 71 4 2.459 1.206 1.636 0.1927 0.0032 0.99 72 5 2.459 1.206 1.636 0.1935 0.0032 0.9970 6 2.459 1.206 1.636 0.1951 0.0032 0.9970 7 2.459 1.206 1.636 0.1959 0.0032 0.9967 8 2.459 1.206 1.636 0.1972 0.0032 0.9974 9 2.459 1.206 1.636 0.1984 0.0032 0.9975 10 2.459 1.206 1.636 0.1866 0.0017 0.9945 11 2.459 1.206 1.636 0.1628 0.0017 0.9940 12 2.459 1.206 1.636 0.1670 0.0017 0.9950 13 2.459 1.206 1.636 0.1475 0.001 7 0.9943 14 2.459 1.206 1.636 0.1508 0.001 7 0.9946 15 2.459 1.206 1.636 0.1387 0.0018 0.9959 LCT-13 1 4.3 10 1.415 1.892 0.2862 0.0018 1.0005 2 4.310 1.415 1.892 0.2939 0.0018 1.0004 3 4 .310 1.415 1.892 0.2974 0.0018 1.0003 4 4.310 1.415 1.892 0.2969 0.0018 1.0007 5 4 .310 1.415 1.892 0.2961 0.0032 1.0003 LCT-16 l 2.350 1.270 2.032 0.0957 0.0031 0.9973 2 2.350 1.270 2.032 0.0954 0.0031 0.9962 3 2.350 1.270 2.032 0.0954 0.0031 0.9967 NET- 28091-0003-01 , Revision 0 A-13

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kerr ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) ,~ k) 4 2.350 1.270 2.032 0.095 6 0.0031 0.9960 5 2.350 1.270 2.032 0.0952 0.0031 0.9970 6 2.350 1.270 2.032 0.0961 0.0031 0.9971 7 2.350 1.270 2.032 0.0959 0.0031 0.9973 8 2.350 1.270 2.032 0.0969 0.0031 0.9972 9 2.350 1.270 2.032 0.0961 0.0031 0.9977 10 2.350 1.270 2.032 0.0970 0.0031 0.9971 11 2.350 l .270 2.032 0.0962 0.0031 0.9978 12 2.350 1.270 2.032 0.0974 0.003 1 0.9972 13 2.350 1.270 2.032 0.0965 0.0031 0.9979 14 2.350 1.270 2.032 0.0975 0.0031 0.9974 21 2.350 1.270 2.032 0.0971 0.0031 0.9977 22 2.350 1.270 2.032 0.0968 0.003 1 0.9974 23 2.350 1.270 2.032 0.0963 0.0031 0.9977 24 2.350 1.270 2.032 0.0967 0.0031 0.9970 25 2.350 1.270 2.032 0.0963 0.0031 0.9972 26 2.350 1.270 2.032 0.0969 0.0031 0.9976 27 2.350 1.270 2.032 0.0963 0.0031 0.9979 28 2.350 1.270 2.032 0.0951 0.0031 0.9972 29 2.350 1.270 2.032 0.0950 0.0031 0.9969 30 2.350 1.270 2.032 0.0949 0.0031 0.9965 31 2.350 1.270 2.032 0.0950 0.0031 0.9979 32 2.350 1.270 2.032 0.0949 0.0031 0.9972 LCT- 17 4 2.350 1.270 2.032 0.201 7 0.0031 0.9983 5 2.350 1.270 2.032 0.1779 0.0031 0.9994 6 2.350 1.270 2.032 0. 1685 0.0031 0.9989 7 2.350 1.270 2.032 0.1597 0.0031 0.9994 8 2.350 1.270 2.032 0.1333 0.0031 0.9972 9 2.350 1.270 2.032 0.1092 0.0031 0.9973 10 2.350 1.270 2.032 0.0998 0.0031 0.9973 11 2.350 1.270 2.032 0.0979 0.0031 0.9979 12 2.350 1.270 2.032 0.0968 0.0031 0.9977 13 2.350 l.270 2.032 0.0953 0.0031 0.99 76 14 2.350 1.270 2.032 0.0946 0.0031 0.9985 15 2.350 1.270 1.684 0.1777 0.0028 0.996 1 16 2.350 1.270 1.684 0.1711 0.0028 0.9983 17 . 2.350 1.270 1.684 0.1665 0.0028 0.9987 18 2.350 1.270 1.684 0.1648 0.0028 0.9974 19 2.350 1.270 1.684 0.1622 0.0028 0.9978 20 2.350 1.270 1.684 0.1607 0.0028 0.9971 21 2.350 1.270 1.684 0.1592 0.0028 0.9966 22 2.350 1.270 1.684 0.1584 0.0028 0.9959 26 2.350 1.270 1.684 0.3741 0.0028 0.9958 27 2.350 1.270 1.684 0.3203 0.0028 0.9972 28 2.350 l.270 1.684 0.2806 0.0028 0.9974 29 2.350 1.270 1.684 0.2505 0.0028 0.9984 LCT-34 4 4.738 0.940 1.600 0.1367 0.0039 1.0003 5 4.738 0.940 1.600 0. 1330 0.0039 0.9999 6 4.738 0.940 1.600 0.1298 0.0039 1.0017 NET- 2809 1-0003-01 , Revision 0 A-14

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. keff ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) ,~ k) 7 4.738 0.940 1.600 0.1279 0.0039 1.0002 8 4 .738 0.940 1.600 0.1258 0.0039 0.9992 15 4 .738 0.940 1.600 0.1348 0.0043 0.9947 LCT-35 1 2.596 1.417 1.956 0.2086 0.0018 0.9983 2 2.596 1.417 1.956 0.2126 0.0019 0.9976 LCT-39 1 4.738 0.940 1.260 0.22 18 0.0014 0.9953 2 4.738 0.940 1.260 0.2119 0.0014 0.9969 3 4.738 0.940 1.260 0.1923 0.0014 0.9965 4 4.738 0.940 1.260 0.1836 0.0014 0.9961 5 4 .738 0.940 1.260 0.1393 0.0009 0.9978 6 4.738 0.940 1.260 0.1452 0.0009 0.9977 7 4.738 0.940 1.260 0.2132 0.0012 0.9962 8 4.738 0.940 1.260 0.2031 0.0012 0.9963 9 4.738 0.940 1.260 0.1976 0.0012 0.9969 10 4.738 0.940 1.260 0.1732 0.0012 0.9970 11 4.738 0.940 1.260 0.22 18 0.0013 0.9953 12 4.738 0.940 1.260 0.2 166 0.0013 0.9951 13 4.738 0.940 1.260 0.2146 0.0013 0.9951 14 4.738 0.940 1.260 0.2 124 0.0013 0.9954 15 4.738 0.940 1.260 0.2 112 0.0013 0.9959 16 4 .738 0.940 1.260 0.2 104 0.0013 0.9967 17 4 .738 0.940 1.260 0.2099 0.0013 0.9960 LCT-40 1 4 .738 0.940 1.600 0.1427 0.0039 0.9966 5 4.738 0.940 1.600 0.1377 0.0042 0.9951 9 4 .738 0.940 1.600 0.1470 0.0046 0.9993 10 4.738 0.940 1.600 0.1419 0.0046 0.9931 LCT-42 1 2.350 1.270 1.684 0.1690 0.0016 0.9971 2 2.350 1.270 1.684 0.1753 0.0016 0.9968 3 2.350 1.270 1.684 0.1819 0.0016 0.9981 4 2.350 1.270 1.684 0.1804 0.0017 0.9980 5 2.350 1.270 1.684 0.1775 0.0033 0.9981 LCT-43 2 4 .349 0.980 1.500 0.1553 0.0010 1.0007 LCT-44 1 4.349 0.980 1.500 0.1474 0.0010 0.9993 LCT-46 1 4.349 0.981 1.500 0.1488 0.00044 0.9991 2 4.349 0.981 1.500 0.1525 0.00044 0.9989 3 4.349 0.981 1.500 0.1542 0.00044 0.9988 4 4.349 0.98 1 1.500 0.1556 0.00044 0.9989 5 4.349 0.981 1.500 0.1573 0.00044 0.9986 6 4.349 0.981 1.500 0.1595 0.00044 0.9987 7 4 .349 0.981 1.500 0.1479 0.00044 0.9991 8 4 .349 0.981 1.500 0.1550 0.00044 0.9988 9 4 .349 0.981 1.500 0.1594 0.00044 0.9987 10 4 .349 0.98 1 1.500 0.1621 0.00044 0.9988 11 4.349 0.98 1 1.500 0.1672 0.00044 0.9988 12 4.349 0.98 1 1.500 0.1539 0.00044 0.9986 13 4.349 0.981 1.500 0.1570 0.00044 0.9986 14 4.349 0.981 1.500 0.1596 0.00044 0.9986 15 4.349 0.981 1.500 0.1618 0.00044 0.9984 16 4.349 0.981 1.500 0.1655 0.00044 0.9983 NET- 28091-0003-01 , Revision 0 A-15

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. k.rr ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (~ k) 17 4.349 0.981 1.500 0.1724 0.00044 0.9983 LCT-48 1 3.005 1.094 1.320 0.6771 0.0025 0.9990 2 3.005 1.094 1.320 0.6508 0.0025 0.9983 3 3.005 1.094 1.320 0.6824 0.0025 0.9984 4 3.005 1.094 1.320 0.6838 0.0025 0.9988 5 3.005 1.094 1.320 0.6736 0.0025 0.9983 LCT-50 1 4.738 0.940 1.300 0.1998 0.0010 0.9983 2 4.738 0.940 1.300 0.1907 0.0010 0.9978 3 4 .738 0.940 1.300 0.2075 0.0010 0.9978 4 4.738 0.940 1.300 0.1977 0.0010 0.9972 5 4.738 0.940 1.300 0.2230 0.0010 0.9983 6 4.738 0.940 1.300 0.2 141 0.0010 0.9991 7 4.738 0.940 1.300 0.2095 0.0010 0.9992 LCT-51 1 ClO 2.459 1.206 1.636 0.1472 0.0020 0.9965 2 cl la 2.459 1.206 1.636 0.1968 0.0024 0.9972 3 cl 1b 2.459 1.206 1.636 0.1964 0.0024 0.9972 4 cl le 2.459 1.206 1.636 0.1979 0.0024 0.9975 5 cl ld 2.459 1.206 1.636 0.1989 0.0024 0.9970 6 cl le 2.459 1.206 1.636 0.1998 0.0024 0.9972 7 cl If 2.459 1.206 1.636 0.2000 0.0024 0.9973 8 cl lg 2.459 1.206 1.636 0.20 11 0.0024 0.9971 9 cl2 2.459 1.206 1.636 0.1669 0.0019 0.9969 LCT-54 I 4 .349 0.980 1.500 0.1508 0.0005 0.9996 LCT-71 1 4.738 0.949 1.100 0.7592 0.00076 0.9955 2 4.738 0.949 1.100 0.6972 0.00076 0.9954 3 4.738 0.949 1.100 0.6610 0.00076 0.9948 4 4 .738 0.949 1.075 0.8485 0.0008 0.9951 LCT-72 l 4.738 0.949 1.600 0.111 7 0.0012 0.9990 2 4 .738 0.949 1.600 0.1077 0.0012 0.9985 3 4 .738 0.949 1.600 0.1099 0.0012 0.9988 LCT-77 3 4.349 0.980 1.500 0.1621 0.0010 1.0006 LCT-82 3 4.349 0.980 1.500 0.1497 0.0010 1.0005 LCT-83 1 4.349 0.980 1.500 0.1516 0.0010 1.0001 LCT-84 I 4.349 0.980 1.500 0.1541 0.0010 1.0008 LCT-88 1 4.349 0.98 1 1.500 0.1543 0.00044 0.9993 2 4.349 0.981 1.500 0.1556 0.00044 0.9992 3 4.349 0.981 1.500 0.1561 0.00044 0.9992 4 4.349 0.98 1 1.500 0.1560 0.00044 0.9997 5 4 .349 0.981 1.500 0.1560 0.00044 0.9996 6 4.349 0.981 1.500 0.1560 0.00044 0.9998 7 4.349 0.98 1 1.500 0.1559 0.00044 0.9999 8 4.349 0.981 1.500 0.1560 0.00044 0.9994 9 4.349 0.981 1.500 0.1560 0.00044 0.9994 10 4.349 0.981 1.500 0. 1561 0.00044 0.9989 11 4.349 0.981 1.500 0.1561 0.00044 0.9986 12 4.349 0.981 1.500 0.1563 0.00044 0.9980 13 4.349 0.981 1.500 0.1565 0.00044 0.9975 14 4.349 0.981 1.500 0.1564 0.00044 0.9971 15 4.349 0.981 1.500 0.1566 0.00044 0.9967 NET- 28091 -0003-01 , Revision 0 A -16

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kerr ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (d k) 16 4 .349 0.981 1.500 0. 1566 0.00044 0.9963 17 4.349 0.981 1.500 0.1567 0.00044 0.9957 18 4 .349 0.981 1.500 0. 1568 0.00044 0.9954 19 4 .349 0.981 1.500 0.1560 0.00044 0.9994 20 4 .349 0.981 1.500 0.1567 0.00044 0.9990 21 4.349 0.981 1.500 0.1571 0.00044 0.9992 22 4.349 0.981 1.500 0.1573 0.00044 0.9991 23 4.349 0.981 1.500 0.1575 0.00044 0.9991 24 4.349 0.981 1.500 0.1576 0.00044 0.9991 25 4.349 0.981 1.500 0.1578 0.00044 0.9991 26 4.349 0.981 1.500 0.1578 0.00044 0.9992 27 4.349 0.981 1.500 0.1579 0.00044 0.9992 28 4 .349 0.981 1.500 0.15 80 0.00044 0.9991 29 4 .349 0.981 1.500 0.1582 0.00044 0.9992 30 4 .349 0.981 1.500 0.1582 0.00044 0.9995 31 4 .349 0.981 1.500 0.1583 0.00044 0.9995 32 4.349 0.981 1.500 0. 1584 0.00044 0.9995 33 4.349 0.9 81 1.500 0.1585 0.00044 0.9994 34 4.349 0.981 1.500 0.1584 0.00044 0.9996 35 4.349 0.981 1.500 0. 1584 0.00044 0.9996 LCT-89 1 4.349 0.980 1.500 0. 1530 0.0010 1.0000 LCT-90 1 4.349 0.980 1.500 0. 1459 0.0010 0.9994 LCT-91 4 4.349 0.980 1.500 0.1508 0.00 10 0.9999 LCT-92 1 4.349 0.981 1.500 0.1543 0.00044 0.9996 2 4 .349 0.981 1.500 0.1545 0.00044 0.9994 3 4 .349 0.981 1.500 0.1545 0.00044 0.9996 4 4.349 0.981 1.500 0.1549 0.00044 0.9994 5 4.349 0.981 1.500 0.1555 0.00046 0.9988 6 4.349 0.981 1.500 0. 1559 0.00055 0.9994 LCT-96 1 6.903 0.635 0.800 0.5690 0.00095 0.9973 2 6.903 0.635 0.800 0.5674 0.00095 0.99 72 3 6.903 0.635 0.800 0.4191 0.00095 0.9993 4 6.903 0.635 0.800 0.5704 0.00095 0.9971 5 6.903 0.635 0.800 0.561 7 0.00095 0.9971 6 6.903 0.635 0.800 0.5492 0.00095 0.9968 7 6.903 0.635 0.800 0.5304 0.00095 0.9965 8 6.903 0.635 0.800 0.5068 0.00095 0.9966 9 6.903 0.635 0.800 0.4929 0.00095 0.9965 10 6.903 0.635 0.800 0.4929 0.00095 0.9963 11 6.903 0.635 0.800 0.4630 0.00095 0.9974 12 6.903 0.6 35 0.800 0.43 17 0.00095 0.9975 13 6.903 0.635 0.800 0.4032 0.00095 0.9978 14 6.903 0.635 0.800 0.3800 0.00095 0.99 77 15 6.903 0.635 0.800 0.3604 0.00095 0.9979 16 6.903 0.635 0.800 0.4320 0.00095 0.9978 17 6.903 0.635 0.800 0.3756 0.00095 0.9984 18 6.903 0.635 0.800 0.33 17 0.00095 0.9986 19 6.903 0.635 0.800 0.2997 0.00095 0.9989 BAW-1810 1 2.460 1.206 1.636 0.2477 0.00250 0.9990 NET- 2809 1-0003-01, Revision 0 A-17

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kerr ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (d k) 2 2.460 1.206 1.636 0.2470 0.00250 0.99 82 3 2.460 1.206 1.636 0.2465 0.00250 0.9983 4 2.460 1.206 1.636 0.2452 0.00250 0.9991 5 2.460 1.206 1.636 0.2461 0.00250 0.9982 5A 2.460 1.206 1.636 0.2457 0.00250 0.998 1 6 2.460 1.206 1.636 0.2453 0.00250 0.9981 6A 2.460 1.206 1.636 0.2450 0.00250 0.9982 8 2.460 1.206 1.636 0.2453 0.00250 0.9982 9 2.460 1.206 1.636 0.244 8 0.00250 0.998 1 WCAP-3269 2.7 2.720 1.189 1.524 0.2599 0.00400 0.9988 5.7 5.700 0.993 1.422 0.3006 0.00400 0.9978 3.7-12 3.700 0.860 1.105 0.4309 0.00400 0.9985 3.7-24 3.700 0.860 1.105 0.42 88 0.00400 0.9979 3.7-48 3.700 0.860 1.105 0.455 8 0.00400 0.9969 Since boron credit is used it is important validate boron with critical experiments. Table A.3 shows the boron information on boron-containing benchmarks, along with the calculated k.

Table A.3: Summary of Critical Experiments Containing Boron Benchmark Case No. Soluble Separator Plate No. of kerr 10 ID Boron B Areal Density Boron (ppm) (gm/cm 2) Rods LCT-8 1 1511 0.9976 2 1334 0.9984 3 1337 0.9990 4 11 83 36 0.9980 5 11 81 36 0.9976 6 1034 72 0.9977 7 1031 72 0.997 1 8 794 144 0.9960 9 779 144 0.9963 10 1245 72 0.9978 11 1384 0.998 5 12 1348 0.998 5 13 1348 0.9985 14 1363 0.9982 15 1362 0.9980 16 1158 0.998 1 17 921 0.9974 LCT-9 5 0.004549 0.9993 6 0.004549 0.9985 LCT-9 7 0.006904 0.9994 8 0.006904 0.998 1 NET- 2809 1-0003-01, Revision 0 A-1 8

Benchmark Case No. Soluble Separator Plate No. of kerr 10 ID Boron B Areal Density Boron (ppm) (gm/cm 2) Rods 9 0.066946 0.9986 LCT-11 2 1037 0.9967 3 769 0.997 1 4 764 0.9972 5 762 0.9970 6 753 0.9970 7 739 0.9967 8 721 0.9974 9 702 0.9975 10 84 0.9945 11 64 0.9940 12 64 0.9950 13 34 0.9943 14 34 0.9946 LCT-13 2 0.004549 1.0004 3 0.030173 1.0003 4 0.056950 1.0007 LCT-16 8 0.004549 0.9972 9 0.004549 0.9977 10 0.006904 0.997 1 11 0.006904 0.9978 12 0.066946 0.9972 13 0.066946 0.9979 14 0.066946 0.9974 LCT-34 4 0.002521 1.0003 5 0.002521 0.9999 6 0.00252 1 1.0017 7 0.00252 1 1.0002 8 0.00252 1 0.9992 15 0.04601 1 0.9947 LCT-35 1 70 0.9983 2 147.7 0.9976 LCT-40 1 0.002521 0.9966 5 0.046011 0.9951 9 0.04601 1 0.9993 10 0.046011 0.9931 LCT-42 2 0.004549 0.9968 3 0.030 173 0.998 1 4 0.056950 0.9980 LCT-50 3 822 0.9978 4 822 0.9972 5 5030 0.9983 6 5030 0.999 1 7 5030 0.9992 LCT-51 1 ClO 143 0.9965 NET- 2809 1-0003 -01 , Revision 0 A-19

Benchmark Case No. Soluble Separator Plate No. of kerr 10 ID Boron B Areal Density Boron (p pm) (gm/cm 2) Rods 2 cl la 510 0.9972 3 cl lb 514 0.9972 4 cl le 501 0.9975 5 cl ld 493 0.9970 6 cl le 474 0.9972 7 cl lf 462 0.9973 8 cl lg 432 0.9971 9 cl2 217 0.9969 LCT-77 3 4 1.0006 LCT-82 3 6 1.0005 LCT-92 1 0.1 0.9996 2 6 0.9994 3 11 0.9996 4 22 0.9994 5 43 0.9988 6 95 0.9994 BAW-1 810 1 1337 0.9990 2 1250 0.9982 3 1239 0.9983 4 1171 0.9991 5 1208 0.9982 SA 1191 0.9981 6 1156 0.9981 6A 1136 0.9982 8 11 71 0.9982 9 1131 0.9981 A.2.5 Statistical Analysis of the Fresh U02 Critical Benchmark Results The statistical treatment used follows the guidance provided in NUREG/CR-6698 [2]. The NUREG approach weights the calculated kerr values by the experimental uncertainty. This approach means the hi gher quality experiments (i.e.: lower uncertainty - see Table A.2) affect the results more than the low quality (i.e.: higher uncertainty) experiments . The uncertainty weighting is used for the analysis of the set of experiments as a whole, as well as for the analysis for trends .

Before seeking trends the 328 critical benchmarks set are reviewed as a whole. The unweighted mean kerr of the 328 samples is 0.9981 with a standard deviation of0.0015 . The weighted mean is 0.9985 and the weighted standard deviation is 0.0015. The average uncertainty of the experiments (interpreted as one sigma) is 0.0019. Since the total one sigma standard deviation is only 0.0015, this suggests that the experimental uncertainty dominates the uncertainty and there is little to be gained with improved methods. Unless stated otherwise all of the results presented will come from the weighted analysis. The bias of the set as a whole is 0.0015. The uncertainty is the standard deviation multiplied by the single-sided lower tolerance factor (taken as 2.065 from Reference 2 for more than 50 samples) so it is 0.0031.

NET- 28091-0003-01, Revision 0 A-20

As recommended by NUREG/CR-6698, the results of the validation are checked for normality. The National Institute of Standards and Technology (NIST) has made publicly avai lable a statistical package, DATAPLOT [4). The 328 critical experiments were tested with the Wilk-Shapiro normality test and were found to adhere to a normal distribution at the 90% level. The test results are shown in Table A.4. A histogram plot of the data is shown on Figure A. l.

Table A.4: Wilk-Shapiro Test Results Output From DATAPLOT [4]

NET- 2809 1-0003-0 1, Revision 0 A-21

Calculated keff Distribution Versus a Normal Distribution 60 so C

iii 40

.l:

-a

~

8"' 30

~

b.

Ill

.a E

~ 20 10 0

Figure A.l: Distribution of the Calculated kcrr values Around the Mean If a feature of a subset of the critical experiments creates a statistically subset, this feature needs to be corrected before combining all of the critical results. There are 89 experiments that have boron in them.

The average keff of the boron containing cases is 0.9978 which is very close to the average of all cases (0.9981 ). Similarly, there are 17 cases that used pure Cadmium absorbers. The mean of these cases is 0.9982. There are 51 cases that use Ag-In-Cd control rods with a mean of 0.9987 . Since the standard deviation of the set as a whole is 0.0015 (unweighted) it is clear these features are not skewing the results.

The next step in the analysis is to look for trends in the data . The math will always find a trend but only the real or statistically significant trends are of interest. Section 3.2.2 of the DOE/RW technical report in support of validation for burnup credit [8] describes an appropriate trend test. In this test, the null hypothesis is that the slope of the trend is zero (no trend) and it tests to determine if there is confidence that the calculated slope is a more accurate representation than a zero slope. The equations for this test are presented here.

NET- 28091-0003-01, Revision 0 A-22

Let the regression fit be of the form:

k= a+ b X Let x-bar be the average value ofx for then cases and define:

sxx = ~)x; -x)2 i=l,n and define:

then the test statistic is:

(n -2)

• Sxx T=\bl*

This test statistic is then compared to the Student's t-distribution at the desired confidence level and n-2 degrees of freedom. In the past it was assumed that unless there is a high confidence level (95%) that the slope was non-zero, the analysis would assume a zero slope (no trend) on the given parameter. Since the analysis will include consideration of the data as non-trended, it is more conservative to assume there is also a trend. Inverting the statistical test to requiring a high confidence that the slope is zero will result in all cases having a trend. At this time, although a test on the confidence of the trend is performed, the analysis assumes all calculated trends are real.

For this work the weighted k.,ffvalues are used to determine the fit to a straight line. Refer to NUREG/CR-6698 [2] equations 10 through 13.

NUREG/CR-6698 [2] describes the appropriate tolerance band for criticality validation. This work simply applies the equations (equations 23 to 30) given in the NUREG. Note that the tolerance band is found using the weighted experimental data. The width of the tolerance band is the uncertainty.

In the final analysis, the calculated keff of the system must be less than the minimum of k(x) minus the uncertainty minus the administrative safety margin. The uncertainty in k.,ff from other independent uncertainties, such as the manufacturing tolerances, burnup, and depletion uncertainties can be statistically combined with the uncertainty in the criticality validation. The rest of this section will evaluate the trends in keff as a function of trending parameters using the methods described above.

Historically, an Upper Subcriticality Limit (USL) was assigned from the criticality validation analysis.

This is not done here, since the other uncertainties (e.g., manufacturing tolerances of the rack, depletion uncertainty, etc.) are not known at this time.

NET- 28091-0003-01, Revision 0 A-23

Neutron spectrum Trends in the calculated kerr of the benchmarks were sought as a function of the neutron spectrum. Since a large number of things can affect the spectrum, a single index calculated by SCALE is used. This index is the Energy (eV) of the Average Lethargy caus ing Fission (EALF) . Figure A.2 shows the distribution of kerr values around the mean k, which is shown as the red line. Visual inspection of the graph and the statistical analysis of the results of the statistical analysis suggest that there is a statistically significant trend on neutron spectrum. Using NUREG/CR-6698 [2] equations 10 through 13 and the data from Table A.2 , the predicted mean kerr as a function of EALF is:

k(EALF) = 0.999406 - 0.00459

• EALF The units for EALF are eV . The bias at 0.4 eV is 0.0024. The uncertainty at 0.4 eV is 0.0030 . The bias at 0.65 eV is 0.0036. The uncertainty at 0.65 is 0.0034 .

1.003 ------ - - -----

1 .002 1.001

li::

1.000

i: 0 .999 QI

-o 0 .998 QI 1.., 0 .997

- - - - ~ - - - - -- - - -

~

u 0 .996 0 .995 0 .994 0.993 0 .992 ,---- T 0 0 .1 0 .2 0 .3 0.4 0 .5 0 .6 0 .7 0 .8 0 .9 Energy of the Average Lethargy Causing Fission (eV)

Figure A.2: k eff as a Function of the Energy of the Average Lethargy Causing Fission NET- 28091-0003-01 , Revision 0 A-24

Geometry tests Two trend tests were performed to determine if lattice/geometric parameters are adequately treated by SCALE 6.1.2. The first parameter is the fuel pin diameter. A small, statistically insignificant trend was found when the critical experiment analysis results were correlated to the fuel pin diameter. The second lattice parameter tested is the lattice pitch. A statistically significant trend on lattice pitch was found . The trend on pitch or pin diameter could be caused by the spectral trend found in the previous subsection.

Using NUREG/CR-6698 [2] equations 10 through 13 and the data from Table A.2, the predicted mean kerr as a function of pin diameter is:

k(Pin Diameter) = 0.997805 + ( 7.43E-04)*Pin Diameter where the pin diameter is in cm. The predicted mean kerr as a function of pitch is:

k(Pitch) = 0.997098 + ( 9.646E-04)*Pitch where lattice pitch is in cm.

The tolerance band widths are 3.4E-03 and 3.SE-03 for the pin diameter and pitch respectively. Figures A.3 and A.4 graphically present kerr as a function of the pin diameter and the lattice pitch.

1.003 1.002 1 .001 1.000

i: 0 .999 o/

lI

0.998 41 3...

~

0 .997 0 .996 0 .995 0 .994 i 0 .993 0 .992 0.8 0 .9 1 1.1 1.2 1.3 1.4 1.5 Pin Diameter (cm)

Figure A.3: kerr as a Function of the Pin Diameter NET- 28091-0003-01 , Revision 0 A-25

1.003 1.002 1.001 1.000

=

QI 0 .999

=-::

"C 0 .998 QI Ill

i 0 .997 V

iv u 0 .996 0 .995 0 .994 0 .993 -+-- - - -

0 .992 1 1.2 1.4 1.6 1.8 2 2.2 2.4 2 .6 Pitch (cm)

Figure A.4: keff as a Function of the Lattice Pitch Enrichment The fuel to be stored in the racks ranges in enrichment from 1.6 wt% 235 U to 5 wt% mu. It was determined that there is no statistically significant trend on enrichment. Although not statistically significant, the trend in the mean keff is:

k(Enrichment) 0.998824 - ( 6.4E-05)*Enrichment where Enrichment is wt% mu.

The tolerance band width is 3.4E-03 . Figure A.5 graphically presents the results .

NET- 28091-0003-01 , Revision 0 A-26

1.003 - - - - - - - -- -- - - - - -

1.002 1.001 1

~

41 0.999 *

"O 41 I'll

i 0 .998 0 .997 V

7ii u 0 .996 0.995 0 .994 0 .993 0 .992 2 2.5 3 3.5 4 4.5 5 5.5 6 6.5 7 Enrichment (wt% U-235)

Figure A.5: keff as a Function of the Fuel Enrichment Boron Content A trend test was performed to determine whether the calculated kerr values of the benchmark experiments contain a statistically significant trend as a function of the soluble boron ppm. No statistically significant trend was found . However, it is conservatively assumed that the trends are real. The follo wing equation is the best fit of the data for kerr versus soluble boron ppm. Figure A.6 shows the results of the analyses .

The uncertainty around the mean values given in the following equations is 0.0029 at O ppm and 0.0035 at 2000 ppm.

k(ppm soluble boron) 0.99856 + ( 1.56E-07)*ppm NET- 28091-0003-01 , Revision 0 A-27

0.9995 ------ -

0.9990 0 .9985 **

i:

~

~ 0.9980 "O

~

i 0.9975

• ia u

0 .9970 0.9965 0 .9960 0 1000 2000 3000 4000 5000 6000 Soluble Boron (ppm)

Figure A.6: kcff as a Function of the Soluble Boron Content A.2.6 Establishing the Bias and the Uncertainty To make the incorporation of the bias and bias uncertainty in the criticality analysis conservative, the most limiting bias and bias uncertainty from the trends in the range of interest is used. At the lattice pitch for Westinghouse 15xl5 fuel (1.43 cm) the bias and uncertainty are 0.0015 and 0.0029 respectively. At the Westinghouse 15xl5 fuel pin diameter (1.072 cm) the bias and uncertainty are 0.0014 and 0.0029 respectively. Thus the bias as a function of pitch is more limiting.

The bias as a function of enrichment is greatest at 5 wt% and is 0.0015 . The uncertainty over the range of enrichments is 0.0034. The maximum bias and uncertainty as a function of soluble boron ppm occurs at 2000 ppm and is 0.0018 and 0.0035 respectively.

The spectrum as measured by the EALF in the pool with no soluble boron is genera lly between 0.2 and 0.4 eV. The bias increases as the spectrum hardens and the bias at 0.4 eV is 0.0024. This is the most limiting bias. For heavily borated cases the EALF can get almost as high as 0.65 eV. At 0.65 eV the bias is 0.0036 . For the criticality analysis a bias of 0.0024 is used for all EALF less than 0.4 (limiting cases for no boron credit) and 0.0036 for EALF values between 0.4 and 0.65 eV (heavily borated cases). The maximum uncertainty for any trends is 0.0035 which comes from the soluble boron analysis.

In order to make the analysis simple 0.0035 is selected for the uncertainty in the bias.

NET- 2809 1-0003-01 , Revision 0 A-28

The uncertainty of the set as a who le is 0.0031 . The uncertainty for the trended analysis is generally less since taking advantage of the trend reduces the difference between the experimental value and the predicted value.

A.2. 7 Subcritical Margin In the USA, the NRC has established subcritical margins for rack analysis . The subcritical margin for borated spent fuel pools, casks, and fully flooded dry storage racks is O when the analysis is perfonned with unborated water. This is actually saying the subcritical margin is contained in the uncredited soluble boron. To make sure there is sufficient so luble boron, analysis is also performed with soluble boron and a subcritical margin of 5% in kerr is required. For dry storage racks analyzed with optimum moderation, the subcritical margin is 2% and 5% with full moderation. In the analysis of 328 critical experiments, which generously cover the range of expected conditions, the lowest calculated kerr was 0.9931 . This supports the position that the subcritical margin is more than sufficient.

A.2.BArea of Applicability (Benchmark Applicability)

The critical benchmarks selected cover all commercial light water reactor fuel storage racks or casks. To summarize the range of the benchmark applicability (or area of applicability), Table A.5 is provided below.

Table A.5: Area of Applicability (Benchmark Applicability)

Parameter Range Comments Fissionable Material/Physical U02 Form Enrichment (wt% U-235) 2.35 to 6.903 Some extrapolation of the bias to lower enrichments may be needed .

Enrichments less than 2.35 are rarely limiting and generally only used in 1st cores. The bias is becoming smaller at low enrichments. Using the maximum bias and uncertainty for all of the trends easily covers the small extrapolation needed. The maximum enrichment of 5% is within the range of experiments.

NET- 28091-0003-01 , Revision 0 A-29

Parameter Range Comments Spectrum Expected range in applications:

- EALF (eV) 0.0605 to 0.1 to 0.6 0.8485 The experiments easily cover the entire expected range of limiting conditions.

Lattice Characteristics Type Square Hex lattices have been excluded Pin Pitch (cm) 1.075 to 2.54 Pin pitch of 1.43 cm is within the range.

Assembly Spacing in Racks This covers all spacing. Neutron transport through larger than 15 .4 cm Distance between Assemblies 0 to 15.4 has a small effect on k. Note that the (cm) spacing is assumed to be filled with full density water. If the water density is less this separation effectively increases. Therefore, optimum moderation cases of wide spaced racks are covered.

Absorbers Ag-In-Cd control rods Contained in No significant difference in bias 51 critical between Ag-In-Cd critical experiments experiments and those that did not contain the control rods.

Absorbers Soluble Boron 0 to 5030 ppm All designs are within this range.

Concentration Absorbers Cd bearing experiments showed no dependence on the number of rods.

Cd (component of Ag-In- Cd Absorber Credit for these rods is acceptable.

Cd rods) panels NET- 28091-0003-01 , Revision 0 A-30

Parameter Range Comments Reflector Experiments included water Reflectors Most rack analysis will assume an and steel adequately infinite system. Full pool model covered reflectors are adequately covered.

Temperature Room This temperature range covers all Temperature normal operating temperatures . Over to 358 K temperature accident conditions have significant margin due to ppm boron.

Moderating material water The moderator in all benchmark experiments are water, therefore water as a moderating material is covered NET- 28091 -0003-01 , Revision 0 A-3 1

A.2.9 Summary of U02 Laboratory Critical Experiment Analysis This validation follows the guidance ofNUREG/CR-6698. Key aspects of the guidance are the selection of experiments, analysis of the experiments, statistical treatment, determination of the bias and the bias uncertainty, and finally identification of the area of applicability.

328 U02 critical experiments have been selected that cover the range of conditions for rack analysis . The experiments have been analyzed using SCALE 6.1.2 and the ENDF/B-VII 238 group cross sections and the resulting bias in keff is very small. The results of the criticality analysis were tested for trends against 5 different parameters important to reactivity. It was conservatively assumed that the any trend found was significant. Using the trends, the most limiting bias and bias uncertainty is determined to be 0.0024 for the bias for EALF up to 0.4 eV and 0.0036 for EALF's in the range of 0.4 and 0.65 eV and the uncertainty is 0.0035 for all analysis.

The area of applicability is found in Table A.5.

NET- 2809 1-0003-01, Revision 0 A-32

A.3. HTC and MOX Critical Experiments Burned fuel contains a low concentration of plutonium (about 1 wt%), as well as the uranium and thus is actually Mixed Oxide (MOX) fuel. Most classical MOX experiments have plutonium concentrations at least twice as high as that contained in burned fuel. A series of experiments were performed in France and purchased by the US for domestic use, which model the uranium and plutonium concentration, which matches 4.5 wt% U-235 fuel burned to 37.5 GWd/T [12]. This fuel has 1.1 wt% plutonium and 1.57 wt%

U-235. Both the HTC critical experiments and a large series of classical MOX experiments were analyzed.

A.3. 1 HTC Critical Experiments All of the HTC critical experiments used the same fuel pins . The criticality of these experiments was controlled by adjusting the critical water height. The fuel pins were used in 156 critical arrangements.

117 of these were relevant to spent fuel pool analysis. The experiments were performed in four phases.

Phase 1 [ 13] consists of 17 cases where the pin pitch was varied from 1.3 cm to 2.3 cm and different quantities of pins were used to change the critical height. An 18 111 case was done where the array was moved to the edge of the tank, so the boundary was the steel tank followed by void. This condition is not typical of a spent fuel pool , so this case was not analyzed. The average keff of the Phase 1 cases was 0.99910.

Phase 2 [ 14] consisted of 20 cases where gadolini um of various concentrations was dissolved in the water (Phase 2a) and 21 cases where boron was dissolved in the water (Phase 2b). These experiments also varied the pitch ( 1.3 to 1.9 cm) and the number of pins. The average keff of the gadolinium cases was 0.998 15 and the average for the boron cases was 0.99897.

Phase 3 [ 15] consists of 26 experiments where the pins were arranged as 4 "assemblies." Each assembly used a 1.6 cm pin pitch. The assembly separation was varied, as well as the number of pins in each assembly. Finally, eleven cases boxed the assemblies with an absorber (borated steel, boral, or cadmium).

The average keff of these 26 cases was 0.99890.

Finally, Phase 4 [ 16] consisted of redoing the same type of experiments as Phase 3, except with reflector screens. The 38 experiments which used the lead reflector screen were not included in this ana lysis, since lead reflectors are not common in spent fuel pools. The 33 steel reflector experiments were included.

The average keff of these cases was 0.99858.

References 13 through 16 provided all of the detai ls for the analysis. The modeling was straight forward.

The references gave a simple model and a detailed model. The model created for this work fo ll owed the detai led model , except that the top grid outside of the array and the basket supports were not modeled.

Both of these assumptions were part of the simplified model and have a negligible impact on k. The model used actually exceeded the detailed model, since the spring above the fuel was modeled by homogenizing it with the void.

NET- 28091-0003-01, Revision O A-33

Tables A.6 through A.10 present the results of the analysis. A statistical analysis of the HTC set as a whole was performed consistent with the method provided in NUREG/CR-6698 , where the experimental uncertainties were taken from References 13 through 16. The mean uncertainty weighted keff is 0.99878 and the uncertainty is 0.00590. This makes the bias 0.00122. Since all of the pins are the same, trend analysis on the pin diameter and enrichment are not possible. The pin pitch changes are made to adjust the spectrum, so the only trend analysis performed is on the spectrum (EALF) . The trend analysis on the HTC set (performed consistent with NUREG/CR-6698) on EALF yielded the following function:

k(EALF) = 0.999541 - 0.00548

• EALF The units for EALF are eV. The uncertainty about the trending keff is 0.0076 ink. Figure A.7 shows the results of the HTC analysis.

Table A.6: HTC Phase 1 Results Case No. kerr Monte EALF Pitch Carlo (eV) (cm)

Sigma 1 0.99913 0.00015 0.069486 2.3 2 0.99893 0.00016 0.066544 2.3 3 0.99892 0.00016 0.066412 2.3 4 0.99974 0.00017 0.084957 1.9 5 0.99983 0.00017 0.082795 1.9 6 0.99946 0.00020 0.082123 1.9 7 0.99977 0.00019 0.102248 1.7 8 0.99962 0.00018 0.100654 1.7 9 0.99903 0.00019 0.099687 1.7 10 0.99991 0.00019 0.140669 1.5 11 0.99898 0.00020 0.135753 1.5 12 0.99906 0.00019 0.133996 1.5 13 0.99813 0.00021 0.256212 1.3 14 0.99776 0.00019 0.234183 1.3 15 0.99812 0.00022 0.230564 1.3 16 0.99952 0.00020 0.101408 1.7 17 0.99882 0.00019 0.099384 1.7 NET- 28091-0003-01 , Revision 0 A-34

Table A.7: HTC Phase 2a, Gadolinium Solutions, Results Case No. k.rr Monte EALF Pitch Gadolinium Carlo (eV) (cm) Concentration Sigma (Q:/1) 1 0.99784 0.00020 0.25279 1.3 0.0520 2 0.99792 0.00021 0.24946 1.3 0.0520 3 0.99777 0.00019 0.27074 1.3 0.1005 4 0.99771 0.0001 8 0.26756 1.3 0.1005 5 0.99784 0.0001 8 0.26333 1.3 0.1005 6 0.99683 0.0001 8 0.28513 1.3 0.1505 7 0.99684 0.00019 0.27847 1.3 0.1505 8 0.99623 0.00016 0.29552 1.3 0.1997 9 0.99608 0.00018 0.29253 1.3 0.1997 10 0.99689 0.00017 0.16982 1.5 0.1997 11 0.99766 0.00019 0.16252 1.5 0.1495 12 0.99771 0.00018 0.16101 1.5 0.1495 13 0.99868 0.00017 0.15392 1.5 0.1000 14 0.99861 0.00018 0.15223 1.5 0.1000 15 0.99983 0.00020 0.14727 1.5 0.0492 16 0.99976 0.00019 0.14432 1.5 0.0492 17 1.00053 0.00018 0.1063 1 1.7 0.0492 18 1.00070 0.00017 0.08783 1.9 0.0492 19 0.99707 0.00016 0.11369 1.7 0.1010 20 1.00050 0.00019 0.10648 1.7 0.0492 NET- 28091-0003-01, Revision 0 A-35

Table A.8: HTC Phase 2b, Boron Solutions, Results Case No. kerr Monte EALF Pitch Boron Carlo (eV) (cm) Concentration Si2ma {l?:/1) 1 0.99835 0.00020 0.24780 1.3 0.100 2 0.99760 0.00020 0.24450 1.3 0.106 3 0.99816 0.00020 0.25528 1.3 0.205 4 0.99904 0.00020 0.26400 1.3 0.299 5 0.99886 0.00019 0.27475 1.3 0.400 6 0.99852 0.00019 0.27125 1.3 0.399 7 0.99933 0.00018 0.27977 1.3 0.486 8 0.99894 0.00019 0.28781 1.3 0.587 9 0.99952 0.00016 0.16627 1.5 0.595 10 0.99811 0.00019 0.16087 1.5 0.499 11 0.99990 0.00017 0.15663 1.5 0.393 12 0.99987 0.00018 0.15007 1.5 0.295 13 0.99887 0.00018 0.14559 1.5 0.200 14 1.00192 0.00018 0.14024 1.5 0.089 15 1.00338 0.00018 0.10325 1.7 0.090 16 1.00202 0.00017 0.10717 1.7 0.194 17 1.00313 0.00017 0.11049 1.7 0.286 18 0.99367 0.00017 0.11577 1.7 0.415 19 1.00021 0.00021 0.10473 1.7 0.100 20 0.99251 0.00017 0.08965 1.9 0.220 21 0.99642 0.00017 0.08611 1.9 0.110 NET- 28091-0003-01, Revision 0 A-36

Table A.9: HTC Phase 3 Results - Water Reflected Assemblies *

(1.6 cm pin pitch)

Case No. kerr Monte EALF Absorber Assembly Carlo (eV) Box Separation Si2ma Material (cm) 1 0.99774 0.00022 0.12377 Borated SS 3.5 2 0.99986 0.00019 0.14095 Borated SS 0 3 0.99710 0.00019 0.12939 Borated SS 2 4 0.99715 0.0001 8 0.12391 Borated SS 3 5 0.99699 0.00018 0.13503 Borated SS 1 6 0.99987 0.00019 0.12974 Bora) 0 7 0.99614 0.00019 0.12866 Cd 2 8 1.00381 0.0001 8 0.13904 Cd 0 9 0.99646 0.00017 0.13345 Cd 1 10 0.99672 0.0001 8 0.12952 Cd 1.5 11 0.99571 0.00019 0.13726 Cd 0.5 12 0.99901 0.00017 0.11 277 none 18 13 0.999 15 0.0001 8 0.11167 none 14.5 14 0.99934 0.0001 8 0.111 83 none 11 15 0.99910 0.00019 0.11093 none 10 16 0.99961 0.00019 0.11030 none 9 17 0.99930 0.0001 8 0.10842 none 8 18 0.99980 0.00017 0.10656 none 6 19 1.00016 0.0001 8 0.10421 none 4 20 1.00044 0.0001 8 0.10206 none 4 21 0.99976 0.0001 8 0.10470 none 2 22 1.00047 0.00019 0.10714 none 1 23 0.99893 0.0001 8 0.11506 none 0 24 0.99949 0.00020 0.15073 none 0 25 0.99996 0.0001 8 0.12672 none 4 26 0.99937 0.00020 0.11550 none 10 NET- 28091-0003-01, Revision 0 A-37

Table A.10: HTC Phase 4 Results - Steel Reflected Assemblies (1.6 cm pin pitch)

Case No. kcrr Monte EALF Absorber Assembly Separation Carlo (eV) Box Separation From Reflector Sigma Material (cm) (cm) 1 1.00157 0.00019 0.15363 Borated SS 0 0.0 2 0.99845 0.00018 0.15069 Borated SS 0.5 0.0 3 0.99797 0.00018 0.14674 Borated SS 1 0.0 4 0.99826 0.00018 0.14227 Borated SS 1.5 0.0 5 0.99839 0.00019 0.13923 Borated SS 2 0.0 6 0.99712 0.00018 0.13820 Borated SS 2 0.5 7 0.99634 0.00018 0.13705 Borated SS 2 1.0 8 0.99650 0.00018 0.13598 Borated SS 2 1.5 9 0.99658 0.00018 0.13518 Borated SS 2 2.0 10 0.99834 0.00018 0.13430 Borated SS 3 0.0 11 0.99821 0.00018 0.13234 Borated SS 3.5 0.0 12 1.00095 0.00018 0.13558 Bora! 0 0.0 13 0.99653 0.00018 0.13386 Bora! 0.5 0.0 14 1.00431 0.00017 0.14979 Cd 0 0.0 15 0.99818 0.00020 0.14323 Cd 1 0.0 16 0.99769 0.00017 0.13683 Cd 2 0.0 17 0.99615 0.00018 0.13568 Cd 2 0.5 18 0.99536 0.00019 0.13423 Cd 2 1.0 19 0.99513 0.00018 0.13315 Cd 2 1.5 20 0.99465 0.00018 0.13235 Cd 2 2.0 21 0.99869 0.00018 0.13390 Cd 2.5 0.0 22 1.00060 0.00018 0.17427 none 0 0.0 23 1.00057 0.00018 0.16641 none 1 0.0 24 0.99973 0.00018 0.15852 none 2 0.0 25 0.99935 0.00018 0.15709 none 2 0.5 26 0.99946 0.00018 0.15559 none 2 1.0 27 0.99939 0.00018 0.15431 none 2 1.5 28 0.99937 0.00019 0.15351 none 2 2.0 29 0.99941 0.00019 0.14426 none 4 0.0 30 0.99964 0.00018 0.13456 none 6 0.0 31 0.99953 0.00018 0.12886 none 8 0.0 32 0.99947 0.00017 0.12537 none 10 0.0 33 0 .99940 0.00018 0.12333 none 12 0.0 NET- 28091-0003-01, Revision 0 A-3 8

1.006 1.004 - - ----.-

• 1.002 ---.

1

',:J 41

~ 0.998 + - - - - - - - - -

• *:I u

0.996 + - - - - - - - - - - - #

I 0.994 + - - - - - - -

0 .992 + - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

0 .99 + - - - - - , - - - - - - - , - - - - - - r - - - - - , - - - - - . , - - - - - - , - - - - - - - ,

0 .000000 0.050000 0 .100000 0 .150000 0 .200000 0.250000 0 .300000 0 .350000 Energy of the Average Lethargy of Fission (EALF) (ev)

Figure A.7: k eff as a Function of the EALF for the HTC Experiments A.3.2 MOX Critical Experiments The selection of the MOX critical experiments was limited to the low enriched MOX lattice cri ti cal experiments. All 63 of the low enriched MOX pin critical experiments documented in the OECD handbook [1 7] were utilized. The actual input decks were initiated from available decks found in NUREG/CR-6102 [18] and the International Handbook. [1 7] The decks were modified to update to the new cross-section library and changes in the SCALE input format.

Table A. 11 presents the results of the 63 selected MOX critical experiments. The Reference column has the evaluation number from the International Handbook. [17] For example, OECD-7 refers to the OECD International Handbook case MIX-COMP-THERM-07 .

Trends were investigated as a function of EALF, plutonium content, and the Am-241/U-238 ratio. As the spectrum hardens (higher EALF), there is a small trend to higher k. With more plutonium content, keff increases . This is seen in Figure A.8.

NET- 28 091 -0003 -01 , Revision 0 A-39

The change in kerr with cooling time is dominated by the reactivity of the decay of Pu-241 to Am-241 . By plotting kerr versus the Am-241 /U-238 ratio , it is possible to determine if the bias should be changed for cooling. Figure A.9 shows that with increasing Am-241 content, the calculated kerr of the critical experiments increases . This observation shows that the zero cooling time bias conservatively covers the cooling time.

Table A.11: Results of MOX Critical Benchmarks (SCALE 6.1.2, ENDF/B-VII)

EALF Pu Pu Case ID Reference k eff sigma Am241/U238 (eV) wt% 240%

093array OECD-7 1.0009 0.00025 0.1903 2.00 16 6.82E-05 105al.in OECD-7 0.9942 0.00027 0.1369 2.00 16 7.55E-05 105array OECD-7 0.9960 0.00025 0.1377 2.00 16 7.55E-05 105bl OECD-7 0.9914 0.00026 0.1379 2.00 16 7.55E-05 105b2 OECD-7 0.9921 0.00024 0.1377 2.00 16 7.55E-05 105b3 OECD-7 0.9933 0.00025 0.1373 2.00 16 7.55E-05 105b4 OECD-7 0.9940 0.00026 0.1371 2.00 16 7.55E-05 l 143arra OECD-7 0.9980 0.00026 0.1166 2.00 16 8. l 3E-05 132array OECD-7 0.9971 0.00022 0.0953 2.00 16 8.13E-05 1386arra OECD-7 0.9942 0.00023 0.0906 2.00 16 6.97E-05 epri70b OECD-2 0.9992 0.00025 0.7209 2.00 7.8 7.29E-05 epri70un OECD-2 0.9974 0.00027 0.5409 2.00 7.8 7.29E-05 epri87b OECD-2 1.0019 0.00022 0.2710 2.00 7.8 7.29E-05 epri87un OECD-2 0.9981 0.00032 0.1852 2.00 7.8 7.29E-05 epri99b OECD-2 1.0012 0.00024 0.1772 2.00 7.8 7.29E-05 epri99un OECD-2 1.0007 0.00027 0.1333 2.00 7.8 7.29E-05 klmct009 OECD-9 0.9994 0.00024 0.5169 1.50 8 l .06E-05 k2mct009f OECD-9 0.9941 0.00027 0.2943 1.50 8 9.77E-06 k3mct009 OECD-9 0.9934 0.00024 0.1528 1.50 8 8. 96E-06 K4mct009 OECD-9 0.9921 0.00024 0.1155 1.50 8 8.96E-06 K5mct009 OECD-9 0.9925 0.00021 0.0947 1.50 8 8.96E-06 K6mct009 OECD-9 0.9937 0.00024 0.0905 1.50 8 9.77E-06 omct61 OECD-6 0.9954 0.00026 0.3570 2.00 8 2.24E-05 omct62 OECD-6 0.9990 0.00029 0.1 885 2.00 8 2.24E-05 omct63 OECD-6 0.9943 0.00027 0.1374 2.00 8 2,24E-05 omct64 OECD-6 0.9982 0.00025 0.1167 2.00 8 2.24E-05 omct65 OECD-6 0.9994 0.00025 0.0956 2.00 8 2.24E-05 omct66 OECD-6 0.9956 0.00024 0.0907 2.00 8 2.24E-05 mct8cl OECD-8 0.9978 0.00029 0.3776 2.00 24 7.93E-05 mct8c2 OECD-8 0.9977 0.00028 0.1922 2.00 24 7.27E-05 mct8c3 OECD-8 0.9967 0.00024 0.1383 2.00 24 8.59E-05 mct8c4 OECD-8 1.0006 0.00027 0.1170 2.00 24 9.88E-05 mct8c5 OECD-8 1.0000 0.00026 0.0955 2.00 24 9.56E-05 mct8c6 OECD-8 0.9992 0.00023 0.0905 2.00 24 7.27E-05 mct8cal OECD-8 0.9967 0.00025 0.1375 2.00 24 8. 59E-05 mct8cbl OECD-8 0.9931 0.00024 0.1387 2.00 24 8.59E-05 mct8cb3 OECD-8 0.9941 0.00025 0.1381 2.00 24 8.59E-05 NET- 28091-0003-01, Revision 0 A-40

EALF Pu Pu Case ID Reference kerr sigma Am241/U238 (eV) wt 0/o 240%

meteb2 OECD-8 0.9937 0.00024 0.1385 2.00 24 8.59E-05 meteb4 OECD-8 0.9942 0.00026 0.1378 2.00 24 8.59E-05 mixo25lk OECD-5 1.0011 0.00032 0.3732 4.00 18 l.59E-04 mixo252k OECD-5 0.9985 0.00027 0.2476 4.00 18 l.59E-04 mixo253k OECD-5 1.0044 0.00027 0.1712 4.00 18 l.59E-04 mixo254k OECD-5 1.0004 0.00029 0.1425 4.00 18 l .59E-04 mixo255k OECD-5 1.0034 0.00028 0.1058 4.00 18 l.59E-04 mixo256k OECD-5 1.0023 0.00024 0.0917 4.00 18 l .59E-04 mixo257k OECD-5 1.0036 0.00024 0.0875 4.00 18 l .59E-04 saxtnl04 OECD-3 1.00044 0.00027 0.0987 6.60 8.6 8.43E-05 saxtn56b OECD-3 0.99962 0.00028 0.6133 6.60 8.6 8.43E-05 saxtn735 OECD-3 0.99999 0.00031 0.1 820 6.60 8.6 8.43E-05 saxtn792 OECD-3 0.99951 0.00031 0.1505 6.60 8.6 8.43E-05 Saxton52 OECD-3 0.99977 0.00028 0.8 517 6.60 8.6 8.43E-05 Saxton56 OECD-3 1.00018 0.0003 0.5177 6.60 8.6 8.43E-05 teal OECD-4 0.99572 0.00027 0.1418 3.01 22 l.04E-04 tealO OECD-4 0.9988 0.00024 0.0792 3.01 22 9.3 lE-05 teal 1 OECD-4 0.99886 0.00023 0.0788 3.01 22 2.06E-04 tea2 OECD-4 0.9964 0.0003 0.1409 3.01 22 l .99E-04 tea3 OECD-4 0.99665 0.00028 0.1403 3.01 22 2.96E-04 tea4 OECD-4 0.99644 0.00026 0.1172 3.01 22 9.88E-05 tea5 OECD-4 0.9974 0.00027 0.1167 3.01 22 2.02E-04 tea6 OECD-4 0.99848 0.00025 0.1156 3.01 22 3.90E-04 tea7 OECD-4 0.99753 0.00025 0.0917 3.01 22 8.88E-05 tea8 OECD-4 0.99801 0.00025 0.0913 3.01 22 2.03E-04 tea9 OECD-4 0.99864 0.00025 0.0909 3.01 22 3.02E-04 NET- 28091-0003-01 , Revision 0 A-41

1.0060 ------ -----

1.0040 1.0020 1.0000

~

.x 0.9980 0 .9960 I --.-- * ------

0 .9940 0 .9920

---.11-

• t 0.9900 0 .0 1.0 2.0 3 .0 4 .0 5.0 6.0 7.0 8 .0 Pu wt%

Figure A.8: Predicted kerr as a Function of the Plutonium Content NET- 28091-0003-01 , Revision 0 A-42

1.0060 1.0040 1.0020 t * **

1.0000 i * *

i:

QI 0.9980 ---

~

0.9960

~

• 0.9940 0.9900 + - - - ~ - - ~ - - - . - - - - ~ - - ~ - - - , - - - - - r - - - ~ - - - - , , - - -

0.0E+OO 5.0E-05 1.0E-04 1.SE-04 2.0E -04 2.SE-04 3.0E-04 3.SE -04 4.0E -04 4.SE-04 5.0E -04 Ratio of Am-241 to U-238 Figure A.9: Predicted kerr as a Function of the Am-241 Content A.3.3 Bias and Uncertainty from the MOX/HTC Critical Experiments The bias and uncertainty of burned fuel depends on the amount of plutonium in the burned fuel. As shown on Figure A.9 the bias decreases with plutonium content. However, the uncertainty increases with plutonium content. In order to determine appropriate biases and uncertainties the HTC and MOX critical benchmarks are combined. The MOX experiments with plutonium content above 2 wt% were useful for confirmation that the bias decreases with plutonium content, but the maximum plutonium content in spent nuclear fuel about 1.5 wt% plutonium, so using experiments above 2 wt% plutonium needlessly increases the uncertainty. The bias and uncertainty in the bias for the HTC/MOX (2 wt% PU or less) set is controlled by the EALF trend. For EALF's less than 0.4 eV the maximum bias and uncertainty are 0.0021 and 0.0087 respectively. For EALF's from 0.4 to 0.65 eV the maximum bias and uncertainty are 0.0027 and 0.0112 respectively.

NET- 28091-0003-01 , Revision 0 A-43

A.4. Temperature Dependent Critical Experiments Since the criticality analysis for spent fuel pools must consider the full range of temperatures allowed for the pool the LCT-46 set of critical experiments are needed to assure the correct bias and uncertainty is used for conditions where the pool is at its highest temperatures . The suite of critical experiments other than LCT-46 contains a range of fuel to moderator ratios that should adequately cover the impact of the density change in the water as the pool temperature rises but no other experiments test the Doppler broadening of the cross sections or the change in the thennal scattering.

LCT-46 consists of 22 experiments but the last 5 experiments contain copper rods. Since copper is not normally in spent fuel pools only the first 17 experiments are analyzed here.

Section 3 of LCT-046 specifies the critical benchmark and the SCALE models used follow that specification. The specification has a couple of minor ambiguities related to the thermal expansion given as Table 29 of LCT-046. For this analysis all of the expansion factors from Table 46 were applied to all of the x-y dimensions. That means that the same SS component expansion factor was applied to pitch and the inner and outer diameter of the clad. This is consistent with the MCNP samples given in the Appendix of LCT-046. For the axial expansion only the fuel was expanded. As with the MCNP sample input the same expansion factor was used for the radius and the axial direction.

Table A.12 shows the corrected SCALE 6.1.2 ENDF/B-VII results for the 17 critical experiments.

Corrected results in this case means they were divided by the kerr of the benchmark which was not quite 1.0.

Table A.12 : LCT-46 with Full Thermal Expansion Case Temperature (K) Corrected SCALE k SCALE sii!ma 1 297.05 0.999082 0.000065 2 310.41 0.998902 0.000071 3 315.43 0.998817 0.000067 4 319.96 0.998908 0.000073 5 324.93 0.998629 0.000067 6 332.53 0.998746 0.000067 7 287.22 0.999148 0.000067 8 315.91 0.998819 0.000066 9 330.27 0.998696 0.000068 10 337.44 0.998804 0.000065 11 351.99 0.998829 0.000068 12 303.6 0.998649 0.000065 13 312.95 0.998641 0.000069 14 321. 16 0.998556 0.000067 15 328.24 0.998401 0.000068 16 338.26 0.998318 0.000067 17 358.3 1 0.99825 6 0.000065 NET- 28091-0003-01, Revision 0 A-44

Figure A.10 plots the results of the analysis as a function of case. As can be seen from this plot there does seem to be a trend with temperature. Figure A.11 is the data plotted against temperature with the least squares linear fit. The fit is statistically significant. The slope is -8 .6E-6 deltak/°C. The uncertainty around the fit is 0.0013 . The bias is determined by multiplying the change from room temperature in °C by 8.6E-6. The uncertainty of 0.00 13 is an independent uncertainty that can be statistically combined with the other uncertainties.

k versus Case (three sets with increasing temperature in each set) 0.999200 X

0 .999100 X

0 .999000 0 .998900

~ 0.998800

-i, QI

.!! 0.998700 X

I u X X a o .998600 X

X 0 .998500 0 .998400 -x -

0 .998300 x Rectangu lar Set Rou nded Set 4 Gd Rods Set X 0 .998200 T ---,-- ---, T ,-- T -,-----,

0 2 4 6 8 10 12 14 16 18 Case Number Figure A.10: LCT-046 Corrected Calculated k eff per Case NET- 28091-0003-01 , Revision 0 A-45

0.999200 t 0.999 100 0.999000 - -

...., 0.998900

,r.

.1!:,

u 0.998800 a... o.998700

~

~ 0.998600 -j----

§ 0.998500

• 0.998400 0.998300 *

• 0.998200 +

290 300 310 320 330 340 350 360 370 Temperature (K}

Figure A.11: LCT-046 Corrected Calculated keffVersus Temperature It is common practice not to thermally expand the solids when doing analysis of elevated temperatures in criticality analysis. Table A.13 shows the results of the analysis repeated where the temperatures of the materials were increased and the density of the water decreased but no thermal expansion of the solids (fuel , pitch, etc.). As can be seen in Table A.13 the difference between just expanding the water (lowering the density) and full thennal expansion is similar to the Monte Carlo uncertainty. The maximum difference is 0.00027 which is less than 4 times the Monte Carlo one sigma uncertainty of one of the two calculations used in the difference.

NET- 28091-0003-01, Revision 0 A-46

Table A.13: LCT-46 with No Thermal Expansion of Solids Difference in kert From Case Temperature (K) Corrected SCALE k SCALE si!!.ma Full Thermal Expansion 1 297.05 0.999114 0.000069 -0.00003 2 310.41 0.998899 0.000068 0.00000 3 315 .43 0.998736 0.000067 0.00008 4 319.96 0.998666 0.000074 0.00024 5 324.93 0.998491 0.000069 0.00014 6 332.53 0.998 743 0.000069 0.00000 7 287.22 0.999302 0.000067 -0.00015 8 315 .91 0.9988 74 0.0000 70 -0.00006 9 330.27 0.99859 7 0.000068 0.00010 10 337.44 0.998564 0.000067 0.00024 11 351.99 0.998558 0.000070 0.00027 12 303.6 0.998633 0.000070 0.00002 13 312.95 0.998651 0.000070 -0.00001 14 32 1.1 6 0.998469 0.000067 0.00009 15 328.24 0.998420 0.000067 -0.00002 16 338.26 0.998500 0.000067 -0.00018 17 358.3 1 0.998042 0.000067 0.00021 Since the hi gher temperatures have a harder spectrum, the effect of the higher temperatures could have already been captured in the trend on spectrum (EALF). This was tested by using the slope of the change in keff with EALF from the full set of critical experiments . Using the EALF biased ks the linear fit was reanalyzed. The maximum bias (0 to 100 C) changed from 0.00086 to 0.00071. The spectrum is a small amount of the temperature effect and is therefore ignored for the final conclusion.

The analysis of the only set of thermal critical experiments in the International Handbook that uses elevated temperatures below boiling has shown a small increase in the bias with temperature. The bias is determined by multiplying the change from room temperature in °C by 8.6E-6. The uncertainty of 0.0013 is an independent uncertainty that can be statistically combined with the other uncertainties.

NET- 2809 1-0003-01, Revision 0 A-47

A.5. Summary of Validation Using Laboratory Critical Experiments Nuclear fuel starts as U02 and as it bums it becomes a mixture of U02 and Pu 0 2. SCALE 6.1.2 with the ENDF/B-VII.O cross sections calculates a slightly higher keff as the Pu02 content increases. The correct bias and uncertainty should be a function of the plutonium weight percent but this would be overly complicated for a small effect. The bias for the initial condition from U02 critical experiments would be conservative for spent nuclear fuel. However, the uncertainty from the U02 only set is smaller than the uncertainty from the MOX set. To conservatively cover the all of the conditions of the fuel the final 95/95 keff is calculated twice, once using the U02 critica l experiments bias and uncertainty and once using the MOX/HTC bias and certainty. The higher final 95/95 keff is used for comparison to the keff criteria.

The two bias and uncertainty sets are:

1. Based on the U0 2 experiments: For EALF ' s less than 0.4 eV the bias is 0.0024. For EALF ' s between 0.4 ev and 0.65 eV the bias is 0.0036 . The uncertainty for the entire range of EALF is 0.0035.

2. Based on the MOX/HTC experiments: For EALF 's less than 0.4 eY the bias is 0.0021. For EALF ' s between 0.4 ev and 0.65 eV the bias is 0.0027. The uncertainty for the range of EALF 0 to 0.4 eV is 0.0087 . The uncertainty for the range ofEALF 0.4 to 0.65 eV is 0.0112 For all burned fuel the MOX/HTC bias and uncertainty actually determine the 95/95 k.

For most cases in the pool analysis, the most dense water conditions are most limiting. However, if higher temperature cases are more lirniting,then a temperature bias of 8.6E-6 multiplied by the change from room temperature in °C is applied. In addition the uncertainty in this bias, 0.0013 , needs to be included in the uncertainty rack up.

NET- 28091-0003-01, Revision 0 A-48

A. 6. Appendix References

[ 1] Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, June 2011. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-785 .

[2] J.C. Dean and R.W. Tayloe, Jr., Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698 , Nuclear Regulatory Commission, Washington, DC January 2001.

[3] International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)3, Volume IV, Nuclear Energy Agency, OECD, Paris, September, 2016.

[4] DAT APLOT is statistical software supported by the National Institute of Standards and Technology. It can be down loaded at: http://www.itl.nist.gov/div898/software/dataplot/

[5] J. J. Lichtenwalter, S. M. Bowman, M. D. DeHart, and C. M. Hopper, Criticality Benchmark Guide for Light-Water-Reactor Fuel in Tran sportation and Storage Packages , NUREG/CR-6361 (ORNL/TM-13211), Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, March 1997.

[6] L. W. Newman, et al, Urania-Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BA W-1810, Babcock & Wilcox, Utility Power Generation Division, Lynchburg, VA, April 1984.

[7] Brian L. Koponen and Viktor E. Hampel, Nuclear Criticality Safety Experiments, Calculations, and Analyses-1958 to 1982, UCRL-53369, Lawrence Livermore Laboratory, University of California, Livermore, California, October 21 ,1982.

[8] M. Rahimi , E. Fuentes, and D. Lancaster, Isotopic and Criticality Validation/or PWR Actinide-OnlyBurnup Credit, DOE/RW-0497, U.S. Department of Energy, Office of Civilian Radioactive Waste Management, Washington, DC, May,1997 .

[9] [NOT USED]

[10] [NOT USED]

[11] [NOT USED]

[12] D. E. Muel ler, K. R. Elam, and P. B. Fox, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data , NUREG/CR-6979 (ORNL/TM-2007/083),

prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 2008.

[13] F. Femex, "Programme HTC - Phase 1 : Reseaux de crayons dans l'eau pure (Water-moderated and reflected simple arrays) Reevaluation des experiences," DSU/SEC/T/2005-33/D.R. , Institut de Radioprotection et de Surete Nucleaire, 2008.

NET- 28091-0003-01 , Revision 0 A-49

[14] F. Fernex , Programme HTC - Phase 2: Reseaux simples en eau empoisonnee (bore et gadolinium) (Reflected simple arrays moderated by poisoned water with gadolinium or boron) Reevaluation des experiences, DSU/SEC/T/2005-38/D.R. , lnstitut de Radioprotection et de Surete Nucleaire, 2008 .

[ 15] F. Fernex, Programme HTC - Phase 3 : Configurations "stockage en piscine" (Pool storage)

Reevaluation des exp eriences , DSU/SEC!T/2005-37/D.R., lnstitut de Radioprotection et de Surete Nucleaire, 2008 .

[16] F. Fernex, Programme HTC - Phase 4: Configurations "chateaux de transport " (Shipping cask) - Reevaluation des experiences , DSU/SEC!T/2005-36/D.R., Institut de Radioprotection et de Surete Nucleaire, 2008.

[17] International Handbook ofEvaluated Criticality Safety Benchmark Exp eriments, NEA/NSC/D0C(95)3, Volume VI, Nuclear Energy Agency, OECD, Paris, September, 2010.

[18] M. D. DeHart and S. M. Bowman, Validation of the SCALE Broad Structure 44-Group ENDFIB-V Cross-Section Library for Use in Criticality Safety Analyses , NUREG/CR-6102 (ORNL!TM-12460), Oak Ridge National Laboratory, Oak Ridge, TN, September 1994.

NET- 28091-0003-01 , Revision 0 A-50

Appendix B: Fuel Categorization for Unit 2 Batches A Through X and Unit 3 A through AA All of the early discharged fuel has been categorized. Nearly all of the fuel is either Category 4 or Category 5. The table has been color coded to quickly identify the Category. Category 3 is yellow, Category 4 is green, and Category 5 is blue. A range of assembly IDs that have the same Category are grouped together to reduce the length of the table.

All but two assemblies for historical fuel at Unit 3 have been categorized as Category 4 even though about half of them could have been Category 5. Since the spent fuel pool at Unit 2 is used only temporarily for Unit 3 fuel while it is being casked, the lower reactivity Category is not needed.

NET- 2809 1-0003-01, Revision 0 B-1

Table B.1: Fuel Assembly Reactivity Categorization for Assembly IDs A through X for Unit 2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category A01-A65 4 E43-ESS 4 K01-Kl3 4 E56 3 K14-Kl5 5 801-807 4 E57-E60 4 K16-K57 4 808-813 5 K58 5 814-823 4 FOl 3 K59-K68 4 B24-B26 5 F02-F20 4 827-B64 4 F21 3 L01-L07 4 F22-F30 4 L08-Ll0 5 C01-C04 4 F31-F34 5 Lll-L63 4 COS-C06 5 F35 4 L64 3 C07-C12 4 F36 3 L65-L68 4 C13 5 F37-F39 4 (14 4 F40 3 M01-M04 4 C15-C18 5 F41-F49 4 MOS 5 C19-C28 4 FSO 3 M06-M08 4 C29 5 F51-F60 4 M09 5 C30-C64 4 F61 3 M10-M12 4 F62-F64 4 M13-M14 5 D01-D25 4 F65 3 M15-M20 4 D26 5 F66 4 M21 5 D27-D60 4 F67-F68 5 M22-M23 4 D61-D68 5 M24 5 D69-D72 4 GOl-GOS 4 M25-M27 4 G06 5 M28 5 E01-E14 4 G07-G37 4 M29-M30 4 ElS 3 G38 5 M31 5 E16-E19 5 G39-G72 4 M32-M34 4 E20 4 M35 5 E21-E24 5 H01-H38 4 M36-M37 4 E25-E27 4 H39-H51 5 M38-M44 5 E28-E31 5 H52-H54 4 M45 3 E32-E33 4 HSS 5 M46 4 E34-E35 5 H56 4 M47-M48 5 E36-E40 4 M49-MSO 4 E41-E42 5 J01-J68 4 M51-M52 5 NET- 2809 1-0003-01 , Revision 0 B-2

Table B.1: F uel Assembly Reactivity Categorization for Assembly IDs A through X for Unit 2 (Continued)

Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category M53-M54 4 Q71-Q73 4 T42-T43 4 MSS-MSG 5 Q74-Q76 5 T44-T46 5 M57 4 Q77 4 T47 4 M58-M59 5 Q78 5 T48 5 MGO 4 Q79-Q80 4 T49-T51 4 M61 3 T52-T53 5 M62-M63 4 R01-R07 5 T54 4 M64 3 ROB 4 TSS 5 MGS 4 R09-R38 5 T56-T72 4 M66 5 R39 4 T73-T80 5 M67 3 R40-R43 5 M68 5 R44-RSO 4 M69-M71 4 R51-R69 5 U01-U04 5 M72 5 R70 4 uos 4 R71-R72 5 U06-U13 5 N01-N08 4 R73-R74 4 U14 4 N09-N12 5 R75-R79 5 U15-U16 5 N13-N14 4 R80-R81 4 U17-U21 4 N15-N16 5 R82 5 U22 5 N17-N23 4 R83-R85 4 U23 4 N24-N32 5 U24-U49 5 N33-N47 4 S01-S44 5 USO 4 N48 5 S45 4 USl 5 N49-N80 4 S46-S47 5 U52 4 S48 4 U53-U61 5 P01-P02 4 S49-S61 5 U62-U64 4 P03 3 S62 4 UGS 5 P04-P47 4 S63-S65 5 U66-U68 4 P48 5 S66 4 U69-U73 5 P49-P60 4 S67-S77 5 P61-P72 5 V01-V16 5 V17-V29 4 QOl-QGS 5 T01-T32 5 V30-V35 5 QGG 4 T33-T34 4 V36 4 Q67-Q68 5 T35-T36 5 V37-V38 5 Q69 4 T37 3 V39 4 Q70 5 T38-T41 5 V40-V41 5 NET- 28091-0003-0 1, Revision 0 B-3

Table B.1: Fuel Assembly Reactivity Categorization for Assembly IDs A through X for Unit 2 (Continued)

Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category V42-V43 4 W21 5 X01-X02 3 V44-V49 5 W22 4 X03 -X04 5 vso 4 W23 5 XOS-X37 4 V51-V54 5 W24 4 X38 5 VSS-V57 4 W25 5 X39-X49 4 V58-V61 5 W26 4 XSO-XSl 5 V62 4 W27 5 X52-X53 4 V63 5 W28-W34 4 X54-XSS 5 V64-V65 4 W35 5 X56-X58 4 V66-V67 5 W36-W38 4 X59-X60 5 V68 4 W39 5 X61-X62 4 V69-V77 5 W40 4 X63 5 V78-V79 4 W41-W43 5 X64-X65 4 V80-V81 5 W44-W45 4 X66 5 V82 4 W46 5 X67 4 V83 5 W47 4 X68-X69 5 V84 4 W48-W49 5 X70-X73 4 V85 5 wso 4 X74 5 V86 4 WSl 5 X75 4 V87-V88 5 W52-WSS 4 X76 5 V89 4 W56-W58 5 X77 4 V90-V91 5 W59-W60 4 X78 5 V92 4 W61 5 X79 4 W62 4 X80-X93 s WOl-WlO 4 W63-W67 5 X94-X95 4 Wll 5 W68 4 X96 5 W 12-W15 4 W69-W71 5 W16 5 W72 4 FRss* 4 W17 4 W73-W83 5 W18-W19 5 W84 4 W20 4 W85-W93 5

• FRSB is the fuel rod storage basket NET- 28091-0003-01 , Revision 0 B-4

Table B.2: F uel Assembly Reactivity Categorization for Fuel Assembly IDs A through AA for Unit 3 Indian Point Unit 3 Fuel Assembly ID I Category I Assembly ID I Category I Assembly ID I Category V43 I 3 I V48 I 3 I I All Other Indian Point 3 fuel (Batches A through AA) are Category 4 NET- 28091-0003-01, Revision 0 B-5

ENCLOSURE 3 TO NL-17-144 Indian Point Unit 2 NEI 12-16 Draft Revision 2c Checklist Entergy Nuclear Operations, Inc.

Indian Point Unit 2 Docket No. 50-247

Criticality Analysis Checklist - Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation 1.0 Introduction and Overview Purpose of submittal YES Section 1. Remove credit for Boraflex' Changes requested YES Section 1 and 10 Summary of physical changes YES Section 1. Boraflex' loss Summary of Tech Spec changes YES Section 10 Summary of analytical scope YES Section 1.2 2.0 Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance YES Requirements documents referenced YES Section 1.3 Guidance documents referenced YES Section 1.2 and 10.1 Acceptance criteria described YES Section 1.3 3.0 Reactor and Fuel Design Description Describe reactor operating parameters YES Section 3.4 Describe all fuel in pool YES Section 3.2 Geometric dimensions (Nominal and YES Section 3.2 Tolerance)

Schematic of guide tube patterns YES Figure 6.1 Material compositions YES Section 3 Describe future fuel to be covered YES Section 3.2 Geometric dimensions (nomina l and YES Section 3.2 to lerance)

Schematic of guide tube patterns YES Section 3.2 Material compositions YES Figu re 6.1 Describe all fuel inserts YES Section 3.3.

Geometric dimensio ns (nominal and YES Section 3.3 Tolerances are not used tolerance) since Depletion An alysis uses nominal dimensions .

Schematic (axia l/cross section) NO Standard Westinghouse Designs Material compositions YES Section 3.3 Describe non-standard fuel YES Section 8.10-8.12 Geometric dimensio ns YES Section 8.11 Describe non-fuel items in fuel cells NO Limits given in Section 8.13 Nominal and tolerance dimensions NO Limits given in Section 8.13 4 .0 Spent Fuel Pool/Storage Rack Description New fuel vault and Storage rack description N/A No change in current license needed Nominal and tolerance dimensions N/A Schematic (axia l/cro ss section) N/A Material compositions N/A Spent fuel pool, Storage rack description YES Section 3.1 Nominal and tolerance dimensions YES Section 3.1, Tab le 3.1 Page 1 of 7

Criticality Analysis Checklist - Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes/ Explanation Schematic (axial/cross section) YES/NO Cross sections in Figures 3.2 and 3.3. No axial details given since there is no variation in the over relevant axial heights Material compositions YES Section 3.1 (SCALE elemental brea kdown of SS 304 is used but not specified)

Other Reactivity Control Devices {Inserts) N/A There are no rack inserts . Control rods inserted in the fuel assembli es are credited and covered with the fuel inserts.

Nom inal and tolerance dimensions N/A Schematic (axial/cross section) N/A Material compositions N/A 5.0 Overview of the Method of Analysis New fuel rack analys is description N/A Storage geometries N/A Bounding assembly design(s) N/A Integral absorber cred it N/A Accident analysis N/A Spent fuel storage rack analysis description YES Section 2.0 Storage geometries YES Figure 1.1 and Section 8.5 Bounding assembly design(s) YES Batch Groupings are used. Introduced in Section 5. Fuel designs given in Section 3.2 So luble boron credit YES Soluble boron credit is taken by inference with the criteria se lected in Section 1.3 .

Boron dilution analysis Section 9.6 which references the YES current approved analysis .

Burnup credit YES Section 2.0 Decay/cooling time credit YES Section 10.2 Integral absorber cred it YES Section 10.2 Other credit YES Figure 1.1 shows the credited control rods in specific locations Fixed neutron absorbers N/A Not taking credit for any fixed neutron absorbers Aging management program N/A Accident analysis YES Section 9 Temperature increase YES Section 9.4 Assembly drop YES Section 9.3 Single Assembly misload YES Section 9.3 Multiple misload YES Section 9.5 Page 2 of 7

Criticality Analysis Checklist - Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Boron dilution YES Section 9.6 which references the current approved analysis.

Other YES Section 9.2 (Misplaced Assembly),

Section 9.7 (Seism ic)

Fuel out of rack analysis (Normal Operations) Yes Section 9.1 Handling YES Section 9.1 Movement YES Section 9.1 Inspection YES Section 9.1 6.0 Computer Codes, Cross Sections, and Validation Overview Code/Modules Used for Calculation of kett YES Section 2.1 Cross section library YES Section 2.1 Description of nuclides used YES Table 2.1 Convergence checks YES Section 6.5 and Section 6.6.2 Code/Module Used for Depletion Calculation YES Section 2.1 and Section 5.6 Cross section library YES Section 2.1 and Section 5.6 Description of nuclides used YES Section 2.1 Convergence checks YES Section 5.6 Validation of Code and Library YES Section 4 and Appendix A Major Actinides and Structural Materials YES Sections 4.1 and 4.2 Minor Actinides and Fiss ion Products YES 1.5% bias (NUREG/CR-7109) -Section 4 Absorbers Credited YES Sections 4.1 and A.2.3 7.0 Criticality Safety Analysis of the New Fuel Rack Not part of License Application Rack model N/A Boundary conditions N/A Source distribution N/A Geometry restrictions N/A Limiting fuel design N/A Fuel density N/A Burnable Poisons N/A Fuel dimensions N/A Axial blankets N/A Limiting rack model N/A Storage vault dimensions and materials N/A Temperature N/A Multiple regions/configurations N/A Flooded N/A Low density moderator N/A Eccentric fuel placement N/A Tolerances N/A Fuel geometry N/A Fuel pin pitch N/A Page 3 of 7

Criticality Analysis Checklist - Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Fuel pellet OD N/A Fuel clad OD N/A Fuel content N/A Enrichment N/A Density N/A Integral Absorber N/A Rack geometry N/A Rack pitch N/A Cell wall thickness N/A Storage vault dimensions/materials N/A Code uncertainty N/A Biases N/A Temperature N/A Code bias N/A Moderator Conditions N/A Fully flooded and optimum density N/A 8.0 Depletion Analysis for Spent Fuel Section 5 Depletion Model Considerations Time step verification YES Section 5.6 Convergence verification YES Section 5.6 Simplifications YES Section 5.6 Non-uniform enrichments YES Axial Blankets, Sections 5.6 and 6.2 Post depletion nucl ide adj ustments YES Section 5.9 Cooling time YES Section 5.9 Depletion Parameters YES Sections 5.1-5.5 Burnable Absorbers YES Section 5.2 Integral absorbers Yes Section 5.2 Soluble Boron YES Section 5.3 Fuel and Moderator Temperature YES Section 5.1 Specific power YES Section 5.4 Control rod insertion YES Section 5.5 Atypical Cycle Operating History YES Section 5.3 utilized full details of cycle lengths (some short cycles) for determining average ppm . Section 5.1 utilized most limiting temperatures in cycles where power changed (IP2 cycle 10 and IP3 cycle 12) 9.0 Criticality Safety Analysis of Spent Fuel Pool Storage Racks Rack model YES Boundary conditions YES Section 6.1 Source distribution Yes Section 6.5, Section 6.6.2 Geometry restrictions YES Section 10.3 Page 4 of 7

Criticality Analysis Checklist - Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Design Basis Fuel Description NO Multiple batch groupings as well as assemb ly specific analyses are utilized.

The secti ons given be low for this to pic are where the batch va lu es are identified .

Fue l density YES Section 3.2 Burnable Poisons YES Section 5.2 Fuel assembly inserts YES Section 5.2 Fuel dimensions YES Section 3.2 Axial blankets YES Section 3.2 Confi gurations conside red Borated YES Section 8.14 Un borated YES All but Section 8.14 Multiple rack designs YES Sections 8.2, 8.3, and 8.4 Alternate storage geometry YES Section 8.5 Reactivity Control Devices Fuel Assembly Inserts YES Control Rod Credit, Sections 6.6 and 8.3 Storage Cell Inserts N/A No used .

Storage Cell Blocking Devices YES Section 8.7 Axial burnup shapes Uniform/Distributed YES Section 6.2 .1 Nodalization YES Section 6.2 .

Blankets mode led YES Section 6.2 \J Tolerances/Uncertainties Fuel geometry Fuel rod pin pitch YES Section 7.1 Fuel pellet OD YES Section 7.1 Cladding OD YES Section 7.1 Axial fuel pos ition NO Insignificant reactiv ity since the rack has no axial variation (Applies to racks crediting absorbers that are not full length which doe s not apply to Indian Point.)

Fuel content En rich ment YES Section 7.1 Density YES Section 7.1 Assembly insert dimens ions and materials NO Depletion uses nominal dimensions, Control rod densities are reduced a bounding 20% (Section 6.6)

Rack geometry Flux trap size (width) NO Reduction of the cell pitch reduced the flux trap width Rack cel l pitch YES Section 7.1 Rack wall thickness YES Section 7.1 Page 5 of 7

Criticality Analysis Checklist - Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes/ Explanation Neutron absorber dimensions N/A No credit taken for rack absorbers Rack insert dimensions and materials N/A No rack inserts used Code val idation uncerta inty YES Section 7.5 Criticality case uncerta inty (statistical) YES Section 7.5 Depletion uncertainty YES Section 7.2 Burnup uncertainty YES Section 7.2 Biases Design basis fuel design NO Used most limiting fuel in each batch grouping Fuel geometry Clad creep YES Section 7.2 Grid growth (pin pitch) YES Section 7.2 Minimum grid volume NO Conservatively ignored grids .

Minor actinides and fission product worth YES Section 7.2 Code bias YES Section 7.5 Temperature NO Analysis was performed at most limiting temperature . See Section 8.1 for determining most limiting temperature .

Eccentric fuel placement YES Section 7.4 . Include in full pool analysis rather than a bias. See Section 8.4.3.

lncore thimble depletion effect NO Included in the analysis rather than a bias. See Section 5.6.

NRC administrative margin NO Rather than specify a bias for the NRC administrative margin, the k95 ; 59 is calculated showing at least 1% margin .

Calculated k95 ; 59 in Sections 8.3.2 and 8.4.3 Modeling simplifications Identified and described YES Section 6.2 10.0 Interface Analysis Interface configurations analyzed Between dissimilar racks Section 6.6 Between storage configurations within a rack Section 6.6 Interface restrictions NO Categorization of cell rather than an interface restriction .

11.0 Normal Conditions Fuel handling equipment NO Fuel handling equipment can only handle one assembly at a time and therefo re do not pose a criticality concern . Fuel handling operations are in Section 9.1.

Administrative controls YES Section 9.1 Fuel inspection equipment or processes YES Section 9.1 Page 6 of 7

Criticality Analysis Checklist - Indian Point 2 Storage Rack and Spent Fuel Pool Storage Racks Proposed License Amendment Request Subject Included Notes / Explanation Fuel reconstitution YES Section 9.1 12.0 Accident Analysis Boron dilution YES Section 9.6 which references the current approved analysis .

Norm al conditions YES Section 9 .1.

Accident conditions YES Section 9.6 which references the current approved analysis .

Single assembly mislead YES Section 9 .3 Fuel assembly misplacement YES Section 9 .2 Neutron absorber insert mislead YES Section 9.5 addresse s withdrawa l of required control rods .

Multiple fuel mislead YES Section 9.5 Dropped assembly YES Section 9.3 Temperature YES Section 9.4 Seismic event or other natural phenomena YES Section 9.7 13.0 Analysis Results and Conclusions Summary of results YES SectionlO Burnup curve(s) YES Section 10.2 Intermediate decay time treatment YES Section 10.2 New administrative co ntrols YES Section 9.1 Techn ical Specification markup covered Technical Specification markups YES in an attachment separate from the CSA report .

14.0 References Appendix A Computer Code Validation :

Code validation methodology and bases YES Appendix A New Fuel YES Section A.2 Dep leted Fuel YES Section A.3 MOX crit ical YES Section A.3.2 HTC critical YES Section A.3.1 High temperature crit icals YES Section A.4 Convergence NO Convergence of the Critical Experiments is cove red by the same discussion of convergence for all the analysis. See Section 6.5 Trends YES Section A.2.5 Bias and uncertainty YES Section A.2.6 Range of applicability YES Section A.2.8 Analysis of area of appl icability coverage YES Section A.2 .8 Page 7 of 7