ML14329A195

NET-300067-01, Rev. 0, Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels

ML14329A195
Person / Time
Site:	Indian Point
Issue date:	11/04/2014
From:	Curtiss-Wright Flow Control Corp, NETCO Products & Services
To:	Entergy Nuclear Operations
References
10351857, NL-14-083 NET-300067-01, Rev. 0
Download: ML14329A195 (202)

Text

ATTACHMENT 2 TO NL-14-083 NETCO Report NET-300067-01 (non-proprietary)

Entergy Nuclear Operations, Inc.

Indian Point Unit 2 Docket No. 50-247

NET- 300067-01, Rev. 0 Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels Prepared by:

NETCO, a business unit of Curtiss-Wright Flow Control Corp.

731 Grant Ave Lake Katrine, New York 12449 Prepared for:

Entergy Nuclear Operations under Contract No. 10351857 Rev: Date: Prepared By: Reviewed By: Approved By:

This Page Intentionally Left Blank Table of Contents 1 Introduction ................................................................................................ 1 1.1 Background .............................................................................................................................. 1 1.2 Description of the Analysis ............................................................................................... 1 1.3 Acceptance Criteria ........................................................................................................... 2 2 M ethodology ................................................................................................ 3 2.1 Computer Codes ............................................................................................................ 4 3 Input Data .................................................................................................... 6 3.1 Storage Rack Specifications .............................................................................................. 6 3.2 Fuel Assembly Designs ...................................................................................................... 8 3.3 Fuel Assembly Insert Designs ........................................................................................... 10 3.4 Absorber Panel Design .................................................................................................... 13 4 Validation .............................................................................................. 16 4.1 Fresh Fuel Validation ...................................................................................................... 16 4.2 Burned Fuel Validation - EPRI Approach ................................................................... 17 4.3 Burned Fuel Validation - Extended ISG-8 Approach .................................................. 18 4.4 Burned Fuel Validation - Most Limiting Approach ..................................................... 20 5 Depletion Calculations .......................................................................... 22 5.1 Limiting Depletion Parameters - Burnable Absorbers ................................................ 26 5.2 Limiting Depletion Parameters - Soluble Boron ......................................................... . 27 5.3 Limiting Depletion Parameters - Temperatures .......................................................... 28 5.4 Limiting Depletion Parameters - Specific Power .......................................................... 30 5.5 Limiting Depletion Parameters - Control Rod Operation ........................................... 31 5.6 Summary of Depletion Assumptions for Fuel < 3.5 wt% U-235 .................................. 32 5.7 Summary of Depletion Assumptions for Fuel > 3.5 wt% U-235 .................................. 34 5.8 Depletion Analysis Details (Time Steps, etc.) ................................................................ 35 5.9 Reduced Power Operation at End of Life ..................................................................... 35 NET- 300067-01 Rev 0 ooi

6 R ack M odel ................................................ e........ . .......

. *................ .. 36 6.1 Region 1 Infinite 2x2 M odel ............................................................................................. 37 6.2 Region 2 Infinite 2x2 Model ............................................................................................. 38 6.3 Axial Burnup Distribution ................................................................................................ 39 6.4 Interpolation of Isotopics and Cooling Time Verification ............................................ 41 6.5 Convergence of Calculations .......................................................................................... 42 6.6 Summary of M odeling Assumptions .............................................................................. 43 7 Sensitivity A nalysis ...................................................................... ........ . 44 7.1 Tolerances .............................................................................................................................. 44 7.2 Calculation of Biases and Uncertainties ........................................................................ 47 8 Results ................................................................................................... . 51 8.1 Region 1.................................................................................................................................. 51 8.1.1 M issing Panel in Region 1................................................................................... 52 8.1.2 Alternate Panel Design in Region 1 ..................................................................... 53 8.2 Region 2 .................................................................................................................................. 53 8.2.1 Curve Fit .................................................................................................................... 54 8.2.2 Confirmation Calculations for Region 2 ........................................................... 55 8.2.3 Use of Control Rods in Region 2 .......................................................................... 56 8.2.4 M issing Panel in Region 2 ................................................................................... 57 8.2.5 Alternate Absorber Panel Design in Region 2 .................................................. 58 8.2.6 Expanded Cooling Time Calculations ................................................................ 58 8.3 Borated Conditions .......................................................................................................... 59 8.4 Depletion Effect of Hafnium Flux Suppression Inserts ................................................ 59 8.5 Axial Reflector ....................................................................................................................... 60 8.6 Volatile Fission Gases ...................................................................................................... 61 8.7 Temperature Effects ........................................................................................................ 61 8.8 Fuel Geometry Changes during Burnup ........................................................................ 62 NET- 300067-01 Rev 0 iv

8.9 Depletion of Fuel < 3.5 wt% with Modern Depletion Assumptions ............................... 62 8.10 Reduced Periphery Requirements & Region 1/Region 2 Interface ............................. 63 8.10.1 Full Pool Model .................................................................................................... 63 8.10.2 Results of Reduced Periphery and Region 1/Region 2 Interface Analysis ........... 70 8.11 Failed Fuel Containers .................................................................................................... 74 8.12 Fuel Rod Storage Basket .................................................................................................. 76 8.13 Assemblies with Missing Fuel Rods ............................................................................... 77 9 Accident Conditions ............................................................................ 79 9.1 Misplaced Assembly ........................................................................................................ 79 9.2 Dropped Assembly .......................................................................................................... 81 9.3 Over Temperature ........................................................................................................... 84 9.4 Multiple Misloads .................................................................................................................. 84 9.5 Boron Dilution Accident ................................................................................................. 85 9.6 Seismic Event ......................................................................................................................... 85 10 Sum m ary ............................................................................................. . 86 10.1 Summary of Allowable Fuel Loading ............................................................................ 86 10.2 Absorber Panel Requirements ....................................................................................... 88 10.3 Fuel Requirements .......................................................................................................... 89 10.4 Reactor Operation Limits ............................................................................................... 90 Reference ................................................................................... 93 Appendix A: Validation of SCALE 6.1.2 for Criticality Analysis of Fresh and Burned Fuel ............................................................................. A..-1 A.1. O bjective .................................................................................................. A-I A.2. M ethod ................................................................................................ A-1 A.3. Com puter Codes Used ............................................................................ A-2 A.4. Analysis .................................................................................................... A-4 A.4.1 Laboratory Critical Experiments ...................................................................................... A-4 A.4.1.1 Introduction ............................................................................................................. A-4 NET- 300067-01 Rev 0

A.4.1.2 Definition of the Range of Parameters to Be Validated ....................................... A-4 A.4.1.3 Selection of the Fresh U0 2 Critical Benchmark Experiments ............................ A-5 A.4.1.4 Computer Analysis of the Fresh U0 2 Benchmark Critical Experiments ......... A-14 A.4.1.5 Statistical Analysis of the Fresh U0 2 Critical Benchmark Results ................... A-22 A.4.1.6 Establishing the Bias and the Uncertainty .......................................................... A-31 A.4.1.7 Subcritical Margin ................................................................................................ A-32 A.4.1.8 Area of Applicability (Benchmark Applicability) .............................................. A-33 A.4.1.9 Summary of Fresh U0 2 Laboratory Critical Experiment Analysis ................. A-35 A.4.1.10 HTC and MOX Critical Experiments ................................................................. A-35 A.4.2 Depletion Reactivity Bias and Uncertainty (EPRI Benchmark Analysis) ................... A-46 A.4.3 Extended ISG-8 Validation ............................................................................................. A-49 A.4.3.1 Selection of Assays ................................................................................................. A-49 A.4.3.2 Analysis of the Chemical Assays .......................................................................... A-55 A.4.3.3 Determination of the Isotopic Depletion Bias By Direct Difference ................. A-69 A.4.3.4 Validation of Isotopic Worth (Extended ISG-8) ................................................. A-76 A.5. Summary of Results ............................................................................. A-77 A.6. Appendix References ............................................................................ A-78 NET- 300067-01 Rev 0 vi

List of Tables Table 3.1: Region I and 2 Storage Rack Dimensions 14,51 ....................................................... 7 Table 3.2: Fuel Assembly Dimensions 124,7] ............................................................................... 10 Table 3.3: Control Rod and Hafnium Rod Descriptions 171 ..................................................... 12 Table 3.4: Pyrex and Wet Annular Burnable Absorber Descriptions 16, 7, 241 ...................... 12 Table 3.5- Absorber Panel Dimensions ...................................................................................... 13 Table 5.1: Key Operating Features by Cycle Used in Indian Point Unit 2 .............................. 24 Table 5.2: Key Operating Features by Cycle Used in Indian Point Unit 3 .............................. 25 Table 5.3: Characteristics of Fuel Inserts ................................................................................. 33 Table 6.1: Axial Burnup Profile vs. Burnup Bin 1201 ............................................................... 40 Table 6.2: Verification of Cooling Time Model in the Interpolation Program ........................ 42 Table 7.1: Tolerance Reactivity Effects ...................................................................................... 44 Table 7.2: Miscellaneous Reactivity Effects .............................................................................. 46 Table 7.3: Rack Up of Biases & Uncertainties in Region 2 for 5 wt% Fuel at 43 GWd/T ........... 49 Table 7.4: Rack Up of Biases and Uncertainties for Region I .................................................. 50 Table 8.1: Calculated k's in Region I .......................................................................................... 52 Table 8.2: Minimum Burnup Requirements (GWd/T) in Region 2 ......................................... 53 Table 8.3: Coefficients for Curve Fit of Minimum Burnup Requirements .............................. 55 Table 8.4: Calculated k Values at Each Burnup Point .............................................................. 55 Table 8.5: Total Bias and Uncertainty at Each Burnup Point ................................................. 56 Table 8.6: k95/95 at Each Burnup Point For Region 2 .................................................................. 56 Table 8.7: Additional Sensitivity Calculations for Region 2 ..................................................... 57 Table 8.8: Hafnium Depletion Results ........................................................................................ 60 Table 8.9: Calculated k as a Function of Temperature ............................................................ 62 Table 8.10: Infinite (Section 6) Versus Finite (Full Pool Model) .............................................. 70 NET- 300067-01 Rev 0 vii

Table 8.11: Region 2 Periphery Tests (No Fuel in Region 1) ................................................... 72 Table 8.12: Region 1 Periphery Tests (4 wt% 28.44/20.44 GWd/T Fuel in Region 2) ............ 73 Table 8.13: Reflector Tests (4 wt% 28.44/20.44 GWd/T Fuel in Region 2) .............................. 74 Table 9.1: Misplaced Fuel Assembly Analysis .......................................................................... 80 Table 9.2: Dropped Fuel Assembly Cases ................................................................................. 83 Table 10.1: Region 2 Minimum Burnup (GWd/T) Requirements(a-c .................. 87 Table 10.2: Summary of Loading Restrictions .................................................................................. 88 Table 10.3: Absorber Panel Requirements .................................................................................. 89 Table 10.4: Fuel Design Requirements ...................................................................................... 89 Table 10.5: Fuel Assembly Operating Requirements for Fuel Enriched > 3.5 wt% .................. 9.1 Table 10.6: Fuel Assembly Operating Requirements for Fuel Enriched <3.5 wt% .............. 92 Table A.3.1: 185 Isotopes Used in the Analysis ............................................................................. A-3 Table A.4.1.1: Selection Review of OECD/NEA Criticality Benchmarks .................................. A-6 Table A.4.1.2: Critical Experiment Results with SCALE 6.1.2 and ENDF/B-VlI ................... A-15 Table A.4.1.3: Summary of Critical Experiments Containing Boron ....................................... A-21 Table A.4.1.4: Wilk-Shapiro Test Results Output From DATAPLOT [41 ............................... A-23 Table A.4.1.5: Bias as Predicted Using the Trend in the Bias as a Function of Pitch ............. A-32 Table A.4.1.6: Bias as Predicted Using the Trend in the Bias as a Function of Enrichment..A-32 Table A.4.1.7: Area of Applicability (Benchmark Applicability) .............................................. A-33 Table A.4.1.8: HTC Phase I Results ............................................................................................ A-36 Table A.4.1.9: HTC Phase 2a, Gadolinium Solutions, Results .................................................. A-37 Table A.4.1.10: HTC Phase 2b, Boron Solutions, Results .......................................................... A-38 Table A.4.1.11: HTC Phase 3 Results - Water Reflected Assemblies ....................................... A-39 Table A.4.1.12: HTC Phase 4 Results - Steel Reflected Assemblies ..... i..,................... ".a ......... A-40 Table A.4.1.13: Results of MOX Critical Benchmarks (SCALE 6.1.2, ENDF/B-VII) ............ A-42 Table A.4.2.1: EPRI Benchmark Results for 100-hour Cooling ............................................... A-47 NET- 300067-01 Rev 0 viio

Table A.4.2.2: EPRI Benchmark Results for 5-year Cooling .................................................... A-47 Table A.4.2.3: EPRI Benchmark Results for 15-year Cooling .................................................. A-48 Table A.4.3.1: Trino Predicted Minus Measured % Deviations (Part 1 of 4) .......................... A-57 Table A.4.3.1: Trino Predicted Minus Measured % Deviations (Part 2 of 4) .......................... A-57 Table A.4.3.1: Trino Predicted Minus Measured % Deviations (Part 3 of 4) .......................... A-58 Table A.4.3.1: Trino Predicted Minus Measured % Deviations (Part 4 of 4) .......................... A-58 Table A.4.3.2: Obrigheim Predicted Minus Measured % Deviations (Part I of 2) ................. A-59 Table A.4.3.3: Turkey Point Predicted Minus Measured % Deviations .................................. A-60 Table A.4.3.4: H. B. Robinson Predicted Minus Measured % Deviations ............................... A-61 Table A.4.3.5: Calvert Cliffs Predicted Minus Measured % Deviations (Part 1 of 2) ............ A-62 Table A.4.3.6: Takahama SF95 and SF96 Predicted Minus Measured % Deviations ............ A-64 Table A.4.3.7: Takaham SF97 Predicted Minus Measured % Deviations ............................... A-65 Table A.4.3.8: TMI Predicted Minus Measured % Deviations ................................................. A-66 Table A.4.3.9: Gdsgen and GKN Predicted Minus Measured % Deviations ........................... A-67 Table A.4.3.10: Vandellos Predicted Minus Measured % Deviations ...................................... A-68 Table A.4.3.11: Performance of All the Chemical Assay Analyses ........................................... A-69 Table A.4.3.12: Direct Difference Results for Each Assay ........................ A-70 NET- 300067-01 Rev 0 ix

List of Figures Figure 3.1: Small Section of the Region 1 Rack 141 .......................................................................... 6 Figure 3.2: Region 2 Rack Showing Cell Boxes and Resultant Cells [51 ......................................... 7 Figure 3.3: L-Shaped Absorber Panel ............................................................................................... 14 Figure 3.4: Example Alternate Panel Design [261 ............................................................................ 14 Figure 6.1: Region 1 KENO Model .................................................................................................... 37 Figure 6.2: Region 2 KENO Model .................................................................................................... 38 Figure 6.3: KENO Model of the Alternate Panel Design .............................. 39 Figure 8.1: Loading Curve vs. Unit 2 Inventory (2013) ................................................................... 54 Figure 8.2: k as a Function of Cooling Time.................................................................................... 58 Figure 8.3: Full Pool M odel ..................................................................................................................... 65 Figure 8.4: Indian Point Unit 2 Spent Fuel Pool Taken From Holtec Drawing #397 121] ............ 66 Figure 8.5: Enlargement of the Top Left Corner of the Pool Model .............................................. 67 Figure 8.6: Enlargement of the Bottom Left Corner of the Pool Model ....................................... 68 Figure 8.7: Enlargement of the Left Side of the Bottom of the Region 1/Region 2 Interface ............ 69 Figure 8.8: Location of the Peripheral Cells with Reduced Requirements ..................................... 71 Figure 8.9: Model for Failed Fuel Container Analysis .................................................................. 75 Figure 8.10: Model for the Fuel Rod Storage Basket ...................................................................... 76 Figure 8.11: k versus Missing Fuel Rods ........................................................................................... 78 Figure 8.12: Model for Assemblies with 36 Missing Fuel Rods .................................................... 78 Figure 9.1: Full Pool Model with Misplaced Assembly .................................................................. 81 Figure 9.2: Full Pool Model with 6 Dropped Assemblies ................................................................ 83 Figure A.4.1.1: Distribution of the Calculated k's Around the Mean .............................................. A-24 Figure A.4.1.2: kff as a Function of the Energy of the Average Lethargy Causing Fission ........... A-27 Figure A.4.1.3: keff as a Function of the Pin Diameter ...................................................................... A-28 NET- 300067-01 Rev 0 X

Figure A.4.1.4: kff as a Function of the Lattice Pitch ....................................................................... A-29 Figure A.4.1.5: k~ff as a Function of the Fuel Enrichment .......................................................... A-30 Figure A.4.1.6: kIf as a Function of the B-10 Areal Density in the Separator Plates ..................... A-31 Figure A.4.1.7: kff as a Function of the Soluble Boron Content ...................................................... A-31 Figure A.4.1.8: kIff as a Function of the Energy of the Average Lethargy Causing Fission for the H TC Experim ents ................................................................................................................................. A-41 Figure A.4.1.9 Predicted k~ff as a Function of the Plutonium Content ............................................. A-44 Figure A.4.1.10 Predicted keff as a Function of the Am-241 Content ............................................... A-45 Figure A.4.3.1: Measured Burnup for Pin H6 in Assembly NJ05YU ......................................... A-50 Figure A.4.3.2: Measured U-235 Content for Pin H6 in Assembly NJ05YU ............................. A-51 Figure A.4.3.3: Measured Pu-239 Content for Pin H6 in Assembly NJ05YU ................................ A-51 Figure A.4.3.4: Measured Burnup for Pin O1 in Assembly NJ070G .............................................. A-52 Figure A.4.3.5: Measured Burnup for Pin 012 in Assembly NJ070G ............................................ A-52 Figure A.4.3.6: Measured Burnup for Pin 013 in Assembly NJ070G ............................................ A-53 Figure A.4.3.7: Measured U-235 Content for Pin Ol in Assembly NJ070G ................................... A-53 Figure A.4.3.8: Measured U-235 Content for Pin 012 in Assembly NJ070G ................................. A-53 Figure A.4.3.9: Measured U-235 Content for Pin 013 in Assembly NJ070G ................................. A-54 Figure A.4.3.10: Measured Pu-239 Content for Pin Ol in Assembly N.1070G ................................ A-54 Figure A.4.3.11: Measured Pu-239 Content for Pin 012 in Assembly NJ070G .............................. A-55 Figure A.4.3.12: Measured Pu-239 Content for Pin 013 in Assembly NJ070G ............................. A-55 Figure A.4.3.13: Direct Difference for the 92 Chemical Assays and the Bounding Uncertainty..A-73 Figure A.4.3.14 Direct Difference for the 92 Chemical Assays and the Bounding Uncertainty As a Statistical Based Uncertainty ............................................................................................................... A-74 Figure A.4.3.15 Direct Difference for the 92 Chemical Assays and the Bounding Uncertainty Including the Corresponding ORNL Results ..................................................................................... A-75 NET- 300067-01 Rev 0 Xi

Figure A.4.3.16 Direct Difference for the 92 Chemical Assays and the Bounding Uncertainty Including All ORNIL Results and ISG-8 Rev. 3 .................................................................................. A-76 NET- 300067-01 Rev 0 xii

Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels I Introduction This criticality safety analysis documents the technical basis and justification for proposed loading criteria for current and future fuel at the Indian Point Unit 2 spent fuel pool. The proposed loading criteria supports placement of the various types of fuel that are used in Unit 2 and Unit 3 reactors into the Unit 2 spent fuel pool. The analysis bounds the fuel from both units.

1.1 Background

Indian Point Nuclear Power Generating Plant Unit 2 spent fuel pool racks currently use Boraflex[M as the neutron absorber, which is known to degrade over time. Due to this fact, Entergy, the operator of the Indian Point Plant, will no longer take credit for the BoraflexiM but rather install new neutron absorber panels into every cell in the spent fuel pool. These absorber panels will cover two adjacent walls of each cell, and arc thin enough to allow for fuel to be inserted and removed from the cells. Once placed in the cells, the absorber panels are not intended to be removed except for repair, replacement, or inspection.

The analysis supports two designs of absorber panels. Design constraints are clearly identified and it is possible for alternate absorber panel designs to be used.

Unit 2 and Unit 3 are both 4 loop Westinghouse power plants that utilize the 15x 15 fuel assembly design. The physical dimension requirements of the fuel from both units are the same. To date both units have had all their fuel assemblies manufactured by Westinghouse.

1.2 Descriptionof the Analysis This criticality analysis determines the loading criteria for fuel assemblies in the Unit 2 spent fuel pool by taking credit for inserted absorber panels. The loading criteria will allow for full core off load, NET- 300067-01 Rev 0 I

while storing all the current and projected fuel for Units 2 and 3. In addition, nearly all fuel at its final discharged burnup will be able to be stored in Region 2 (see Figure 8.4 for region definitions). The analysis does not credit any Boraflex neutron absorber that might remain in the racks.

Further, the analysis supports a simple framework of operation by establishing one loading curve for all fuel. This operating framework is possible by using conservative assumptions which bound all the fuel. Although there is only one loading curve, flexibility is added by two simple burnup corrections (an 8 GWd/T reduction in the burnup requirement for assemblies on the Region 2 periphery and a 2 GWd/T burnup adder for assemblies with Hf inserts). Finally, the analysis allows for future fuel designs by using bounding fuel parameters rather than the traditional nominal plus tolerance approach.

This new criticality safety analysis for the Indian Point Unit 2 pool follows the most recent methods.

This effort has been concurrent with the Nuclear Energy Institute (NEI) working with the NRC to produce guidance for spent fuel pool analysis [27]. The NEI guidance started with the NRC draft Interim Staff Guidance (ISG) DSS-ISG-20 10-1 [I].

1.3 Acceptance Criteria The acceptance criteria of the analysis are to ensure compliance with IOCFR50.68 [2]. Specifically, the analysis demonstrates that:

• the k 95/95 of the pool 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) [2],

• the k 95/95 of the pool 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) [2].

In addition, an engineering safety margin is provided to cover unanticipated issues. The safety margin used is 1%, so that the k 95/95 target value is 0.99 for no soluble boron and 0.94 with soluble boron.

NET- 300067-01 Rev 0 2

2 Methodology The criticality safety analysis performed in this report used 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 historical and projected fuel designs and inserts for use in Units 2 and 3. Assure that the analysis covers all the designs. (See Sections 3.2 and 3.3)

2. Review the historical and projected operating history of Units 2 and 3. (See Section 5)

3. Review the current Unit 2 racks and projected new absorbcr panels. (See Sections 3.1 and 3.4)

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

5. Deplete the fuel using a 2D lattice representation of the core using bounding depletion assumptions. (See Section 5)

6. Develop an infinite 3D Monte Carlo model of the Region I and Region 2 racks using periodic boundary conditions (radially infinite). The axial modeling is finite, including conservative modeling of the axial burnup distribution. (See Section 6)

7. Use the rack and fuel manufacturing tolerances and 3D Monte Carlo model to determine the reactivity associated with the manufacturing uncertainties. (See Section 7)

8. Use the infinite 3D Monte Carlo model with the validation and manufacturing biases and uncertainties to determine the minimum burnup as a function of enrichment and cooling time (loading curve), This analysis is performed with no soluble boron. (See Section 8)

9. Test the Region I/Region 2 interface and the relaxed periphery requirements using a full pool 3D Monte Carlo model. (See Section 8)

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

11. Summarize the resulting loading requirements and the assumptions made in the analysis. (See Section 10)

NET- 300067-01 Rev 0 3

2.1 Computer Codes This analysis uses the t5-depl TRITON module of SCALE 6.1.2 [3] (the most recent version) for the depletion analysis and the CSAS5 module for the criticality analysis. All the analyses are performed using the 238 group ENDF/B-VII library (v7-238) (the most recent library). The CSAS5 module utilizes CENTRM and BONAMI for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of k. All of the CSAS5 computer runs use a Monte Carlo sampling of at least 1500 generations and 6000 neutrons per generation to achieve a statistical uncertainty in k of less than 0.0002.

The t5-depl sequence of TRITON utilizes CENTRM and BONAMI 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 ORIGEN. parn=(addnux=4) is used in the analysis which tracks the maximum number of problem specific collapsed isotopes (388). For more details regarding the depletion model, please see Section A.4.

At the end of the depletion analysis, the OPUS module is used to output atom densities for use in the criticality model. The input for OPUS specifies 185 isotopes that are carried forward to the criticality analysis (see Appendix A.2 for a list of the 185 isotopes). 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.

In addition to using SCALE, a FORTRAN Code was used to interpolate between burnups from the OPUS output and also to decay the isotopic content to the desired cooling time. The FORTRAN code, which has been verified and validated, reads an axial burnup profile to get the shape of the burnup axially, so multiple atom density sets can be made quickly. The code was validated by comparing the k calculated with the code-interpolated number densities to the k calculated with number densities directly Throughout this document, k is used as a short hand notation for k-effective or keff NET- 300067-01 Rev 0 4

from SCALE/ORIGEN-S, in which no interpolation was used. Furthermore, SCALE/ORIGEN-S was used to decay to a given cooling time and similar comparisons were made. All the differences were within the statistical uncertainty of the k calculations (see Section 6.4).

Unless otherwise specified, all of the k values reported in this document are raw calculated k 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 "k95195").

NET- 300067-01 Rev 0 5

3 Input Data For the criticality analysis, input data is needed for the storage racks (Section 3.!), the fuel assemblies (Section 3.2), the fuel assembly inserts (Section 3.3), and the absorber panels (Section 3.4). In addition to this data, plant operating data was used to assure conservative depletion parameters were selected. The plant operating data is found in Section 5.

3.1 Storage Rack Specifications Region 1 uses a flux trap for criticality control, Figure 3.1 shows the general arrangement of the cells in Region 1 [4]. Table 3.1 gives the dimensions and tolerances from the manufacturer's drawing [4] for Region 1.

Figure 3.1: Small Section of the Region 1 Rack 141 Region 2 is a non-flux trap design where receptacle cans with Boraflex sheaths are spaced out creating "resultant" cells between the cans. Figure 3.2 shows two complete cells in the Region 2 type rack [5]. The cell on the left would be called the "resultant" cell. Notice the fuel in the resultant cell is not bounded by four flat walls but rather by the Boraflex sheaths. The dimensions for the Region 2 rack are also shown in Table 3.1 [5].

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Proprietary Information Removed Figure 3.2: Region 2 Rack Showing Cell Boxes and Resultant Cells 151 Table 3.1: Region 1 and 2 Storage Rack Dimensions 14, 51 Value Attribute (inches) Tolerance Region 1 Rack 141 Vertical cell pitch 10.545 Horizontal cell pitch 10.765 Cell ID 8.75 Cell wall thickness 0.075 d0.007 Boraflcx sheathing width 7.70 Minimum Boraflex sheathing thickness 0.035* + 0.003 Boraflex sheathing distance 0.112 from cell wall Region 2 Rack 151 Cell pitch 9.04 Cell ID 8.80 Cell wall thickness 0.075 - 0.007 Boraflex sheathing width 8.00 Maximum Boraflex sheathing thickness 0.035 + 0.003 Boraflex sheathing distance 0.092 from cell wall

• 0.0235" and 0.035" wrappers allowed per drawing The material of the rack is SS304 [4, 5]. The Boraflex is modeled as water. If any Boraflex remained, it would contain B-10, which would decrease reactivity compared to water. EPRI Technical NET- 300067-01 Rev 0 7

Report 103300 [25] concludes the rate of boron carbide loss is proportional to the rate of the silica loss. Therefore, a water displacing material without boron is not credible.

3.2 Fuel Assembly Designs The Indian Point Units 2 and 3 have used a number of fuel designs, but in all cases the fuel design changes had little impact on criticality. All fuel for both units was purchased from Westinghouse.

Westinghouse has used the same fuel clad and pellet dimensions for all its 15X 15 designs. The guide tube dimensions have changed as Indian Point changed from the standard Westinghouse 15xl 5 to Westinghouse OFA 15x 15 fuel. The only difference between Standard (LOPAR) and OFA fuel is that the guide tube outer diameter for OFA fuel is slightly smaller but the thickness of the guide tube is unchanged. The effect on reactivity between the two designs is insignificant (see Table 7.2).

The following nomenclature has been used for Indian Point fuel:

I. HIPAR: This was the initial fuel design and used stainless steel guide tubes and Inconel grids.

2. Standard Fuel (LOPAR): Changed from stainless steel to Zircaloy guide tubes.

3. OFA Fuel: Changed to Zircaloy grids and small change to guide tubes.

4. Vantage + Fuel: Added intermediate flow mixing grids, changed the zirconium alloy to ZIRLOIM, and added axial blankets

5. Performance + Fuel: Added a protective bottom grid and longer bottom end plugs.

6. 15X 15 Upgrade Fuel: Changed the grid design and modified the guide tube dashpot by using a tube-in-tube (more water displacement - less reactive).

Most of the fuel design changes have been related to the grids. The grids are ignored in the criticality modeling, since they displace water in fuel assemblies. PWR fuel designs are under moderated so that there is a negative moderator coefficient of reactivity. Even at cold temperatures the fuel assembly designs are under moderated. With high soluble boron concentrations, it can be non-conservative to NET- 300067-01 Rev 0 8

ignore grids, but for borated cases, there is a large margin to the criticality limits (See Section 8.2.6), so ignoring the grids is still acceptable.

Westinghouse has also changed the pellet theoretical density, chamfering and dishing. This is not identified as a fuel design change. For this analysis, a high stack density is used to cover all the previous changes and to allow for future design changes. The stack density is the smeared density of the U0 2 inside the pellet OD. The chamfering and dishing are not modeled. The stack density is always less than the pellet density.

Finally, the initial fuel for Indian Point did not have axial blankets. Several axial blanket designs have been used at the Indian Point Units. This analysis conservatively does not take credit for the lower enriched axial blankets (see Tables 5.1 and 5.2 for axial blanket types used).

The fuel dimensions and tolerances are taken from References 7 and 24. In general, the dimensions used in the analysis are the nominal dimensions plus or minus the tolerance to maximize the reactivity (see Table 3.2). However, this was not done for all the dimensions. The fuel pellet OD and clad ID are increased by more than the typical tolerance from Reference 7 to allow margin for possible future fuel designs. The guide tube tolerance has been shown to be insignificant to the reactivity [27]. However, a small conservative increase in the guide tube thickness was made in the model.

Having more fuel and more water inside the fuel assembly maximizes the reactivity (see Table 7.2).

Therefore, maximum stack density, maximum pellet OD, minimum clad OD, smaller guide tube OD and larger guide tube ID were used in the analysis. The assembly pitch in the core for Westinghouse 15x 15 plants is 8.466 inches [7]. The most the uniform pin pitch could increase is (8.466 - 15 x 0.563)/15 =

0.0014 inches before assemblies would touch each other in the reactor core. This pin pitch tolerance is analyzed (see Table 7.1) and combined with other manufacturing tolerances. An assumed stack density of 97.5% of the U0 2 theoretical density allows some margin to cover future designs. Dishing and chamfering of the pellets are included in the stack density. A 97.5% stack density means the actual NET- 300067-01 Rev 0 9

chamfered and dished sintered pellet density is nearly 99% of the theoretical density of UO 2. If the fuel design changes, the criticality analysis will be valid as long as the maximum stack density, maximum pellet OD, minimum clad OD and minimum guide tube cross-sectional area are maintained (see Table 10.4).

Table 3.2: Fuel Assembly Dimensions [24,71 Value Value Used in Attribute (inches) Typical Tolerance Analysis Fuel pellet U0 2 stack density 97.5 %TD Maximum expected stack 975 density Fuel pellet OD 0.3659 jax p Dc Fuel clad OD 0.4220 1 ]a.C ]il, Fuel clad ID 0.3734 [ YC ]l, Fuel pin pitch 0.5630 +/- 0.0014 (using assembly pitch) 0.5630 Active fuel length 144 None (has no effect) 144 Standard Fuel Fuel guide tube OD 0.546 Insignificant effect on k [27]

Fuel guide tube ID 0.512 Insignificant effect on k [27] [ 57 OFA Fuel Fuel guide tube OD 0.533 Insignificant effect on k [271 LJ PC Fuel guide tube ID 0.499 Insignificant effect on k [27] ]a The fuel clad and guide tube material is Zirc-4 or Zirlo. For fuel pellets that are coated with ZrB2 (IFBA), the B-10 loading is assumed to be [ mg B ")/inch] " [24] (IX) for IFBA rods assumed in fresh fuel and [ mg B11 /inch] a"(I.5X) for depletion calculations.

3.3 Fuel Assembly Insert Designs The fuel assemblies used in Indian Point Units 2 and 3 have contained a number of different types of inserts in the guide tubes during operation. They are:

1. Pyrex burnable absorbers

2. Wet Annular Burnable Absorbers (WABA)

3. Unclad Hafnium flux suppression assemblies

4. Primary source assemblies

5. Secondary source assemblies NET- 300067-01 Rev 0 10

6. Full Length Control Rods

7. Part Length Control Rods The final criticality calculations assume no inserts in the guide tubes except for the special case of a control rod in the assembly to reduce the reactivity of assemblies that fail to meet the loading requirements. However, depletion calculations are performed with inserts in order to harden the spectrum and maximize the reactivity of burned fuel. The effect of the inserts is maximized by using the highest boron content and the most water displacement. The boron loading for the burnable absorbers has varied, so the maximum boron loading has been used in the analysis. A Pyrex burnable absorber displaces more water and has a higher B- 10 loading than a WABA, so for enrichments where both burnable absorbers are used, the Pyrex burnable absorber is conservatively selected. Pyrex burnable absorbers displace more water than the primary source assemblies. The analysis assumes for fuel < 3.5 wt% (old fuel) that Pyrex burnable absorbers remain in the assembly throughout depletion (see Section 5), so primary sources (which were only in the old fuel) are covered.

Secondary sources displace some water and could have a small effect on the burned fuel reactivity. A WABA rodlet displaces about the same amount of water as a secondary source rodlet but a secondary source assembly contains less rodlets (generally 6 rodlets). Since more water is displaced with the assumed WABA assembly (20 rodlet assembly) than the secondary source assembly, analysis with WABAs would be conservative. Since the WABA is never removed during the depletion, a secondary source that is inserted after the WABA depletion is automatically covered.

Indian Point Units 2 and 3 had 8 part length control rods in Cycle 1. In Cycle I, for both units, the part length rods were in assemblies that were nominally 2.2 wt% U-235 enriched. These assemblies exceed the Region 2 loading curve minimum burnup for their enrichment by over 10 GWd/T. The reactivity effect due to the use of the part length control rods in Cycle I for both units is significantly less than the reactivity effect of the 10 GWd/T excess bumup for these fuel assemblies.

NET- 300067-01 Rev 0 II

Thimble plugs have also been used at the Indian Point Units, but since these do not extend into the active fuel region they can be ignored.

Tables 3.3 and 3.4 provide the input data used in the analysis. Note that the Pyrex burnable absorber used at Indian Point had a range of B-10 loadings. The highest B-10 loading is conservative for the depletion analysis and that value is in Table 3.4. Further, the largest borosilicate glass dimensions were utilized which is also conservative.

Table 3.3: Control Rod and Hafnium Rod Descriptions 17]

Parameter Control Rod Hafnium Number of Rodlets per assembly 20 20 or less' Absorber OD (in) 0.3975 0.3810 Absorber Material Ag-In-Cd I-If (80-15-5 wt%)

Absorber density (g/cc) 10.17 13.3]

• Clad OD (in) 0.4390 0.3810' Clad ID (in) 0.4006 none Clad Material SS 304 none Table 3.4: Pyrex and Wet Annular Burnable Absorber Descriptions [6, 7, 241 Parameter Pyrex WABA Material Inside inner clad Void Water Clad material SS304 Zr Inner Clad ID (in) 0.2235 0.225 Inner Clad OD (in) 0.2365 0.267 Absorber ID (in) 0.2430 0.278 Absorber OD (in) 0.3960 0.318 Outer Clad ID (in) 0.4005 0.329 Outer Clad OD (in) 0.4390 0.381 Absorber Material B 20 3-SiO 2 A120 3-B4C (18.1 wt% B2 0 3 ) (0.00603 gm l1B/cm)

Density = 2.23 gcct 20 rodlets were used in the analysis. To date the maximum rodlets used is 16.

t From the SCALE manual, Ref. [3].

ý This is the OD of the unclad Hafnium.

NET- 300067-01 Rev 0 12

For the special case of crediting a control rod in a fresh assembly in the pool, the Ag-In-Cd content (density) is reduced by 20% to bound manufacturing tolerances and any absorber material loss during operation.

3.4 Absorber Panel Design The spent fuel pool racks are assumed to have L-shaped borated AI/B 4C absorber panels in every cell.

For Region I, the panels are oriented bottom-right, while for Region 2, they are oriented top-left. This resolves any interface issues between Region I and Region 2, since there is a double panel at the interface. There are two analyzed absorber panel designs, which are assumed to have the characteristics shown in Table 3.5. Figures 3.3 and 3.4 illustrate the two designs.

Table 3.5: Absorber Panel Dimensions Attribute Value (inches) Notes Absorber Panel (primary)

Areal Density (g B- I0/cm') 0.015 Minimum Panel width Cell ID - 0.03 Minimum Panel thickness 0.086 Minimum in Region 1 0.096 Maximum in Region 2 Length Covers active fuel length Absorber Panel (alternate)*

Areal Density (g 1-10/cm2) 0.022 (Region 1) Minimum 0.020 (Region 2)

Panel width 7.6 Minimum Panel thickness 0.075 Minimum in Region I 0.094' Maximum in Region 2 Offset from corner 0.64 Length ... Covers active fuel length Minor adjustments to these specific dimensions and areal densities are acceptable provided that the panel is shown to be as effective in absorbing neutrons as the primary design.

t For Region 2, in which maximum water displacement is more reactive, the maximum thickness is increased in the model to 0.101 to account for the water displacement of the stainless steel connector that is attached to the absorber panel.

NET- 300067-01 Rev 0 13

-Cell Width-I

-u U

Figure 3.3: L-Shaped Absorber Panel 7 Offset Panel Width ---- i ...A

... 4 rt*

Figure 3.4: Example Alternate Panel Design* [261 For the primary absorber panel design, the AI/B 4C is L-shaped and continuous and fits snugly inside the rack cell. Therefore, the minimum width of the panel is the cell ID minus the tolerance. Calculations showed that having a minimum panel thickness maximizes the reactivity in Region 1, while a maximum panel thickness maximizes the reactivity in Region 2 (see Table 7.2). For the alternate design, two sheets Minor adjustments to these specific dimensions and areal densities are acceptable provided that the panel is shown to be as effective in absorbing neutrons as the primary design.

NET- 300067-01 Rev 0 14

of AVB 4C are connected at the corner by an L-shaped stainless steel connector (see Figure.3.4). The minimum panel width is 7.6 inches. All of the loading curve calculations were performed with the primary design. For the alternate design, since the panel width is over an inch shorter than the primary design and there is no AI/B 4C in the corner, the minimum areal density had to be increased from 0.0 15 to 0.020" in Region 2 and to 0.0226 in Region 1, so that the loading requirements would remain the same.

Calculations show that any metal (or no metal) used by the alternate design in the corner to connect the panels is acceptable, so long as the connecting metal is no thicker than 0. 1 inches (see Figure 6.3).

No references are given for the absorber panel dimensions, since the dimensions in Table 3.5. are design constraints for the absorber panel to be ordered.

NET- 300067-01 Rev 0 is

4 Validation The validation of the SCALE 6.1.2, TRITON (t5-depl) and CSAS5 models requires a number of steps. First, since fresh fuel is allowed to be stored in the spent fuel pool, validation for fresh fuel is performed by use of U0 2 critical benchmarks from the OECDINEA handbook [8]. For the burned fuel, two approaches for validation are used: EPRI Benchmarks [9, 10] and an extension of ISG-8, Rev. 3 [11].

The more limiting of the two approaches is used for the validation of the burned fuel. The extension of ISG-8, Rev. 3 (referred to in this report as Extended ISG-8) is the use of all significant isotopes (185 isotopes) rather than 28 isotopes.

The details of the validation arc found in Appendix A. This section describes the method and summarizes the results.

4.1 Fresh Fuel Validation The validation for fresh fuel follows NUREG/CR-6698 [12]. Two hundred thirty six (236) critical experiments were selected from the OECD/NEA handbook that match the conditions of most spent fuel pools but specifically the Indian Point Unit 2 spent fuel pool. These experiments were analyzed with SCALE 6.1.2 using the 238-group ENDF/B-VII cross-section library. The resulting predicted k's were then tested for trends on the key parameters influencing k. Using these trends, the most limiting bias and uncertainty in the area of applicability was determined. Although some of the trends may not be statistically significant, it is conservative to use all of the trends in determining the limiting bias and uncertainty. Table A.4.1.7 is the area of applicability for the validation. The Indian Point spent fuel pool is covered by the area of applicability of the validation. Specifically, I. Enrichment: The benchmarks selected range from 2.35 to 4.74 wt% U-235. The fuel in the spent fuel pool ranges from 2.21 to 5 wt% U-235. The bias decreases with enrichment and the slope is small allowing for a small extrapolation (see Table A.4.1.7).

NET- 300067-01 Rev 0 16

2. Spectrum: The benchmarks cover a wide range of spectrum by varying the pin pitch. The Energy of the Average Lethargy causing Fission (EALF) of the benchmarks ranges from 0.0605 to 0.8485 eV. The calculated EALF in the pool ranges from 0.1 to 0.6 eV.

3. Fuel Pin Pitch: The fuel pin pitch of the benchmarks ranges from 1.075 to 2.54 cm. The Indian Point fuel pin pitch is 1.43 cm.

4. Flux Trap: The benchmarks include spacing between assemblies of 0 to 15.4 cm. The flux trap design for Region I is 3.4 to 4.0 cm.

5. Boron Areal Density: The benchmarks range from 0 to 0.067 g B'°/cm 2. The spent fuel pool will credit absorber panels with areal densities of 0.015 to 0.022 g B' /cm 2.

6. Soluble Boron: The benchmarks have soluble boron concentrations up to 5030 ppm. The maximum ppm used in the analysis is 2000 ppm.

Details on the area of applicability can be found in Appendix A.

The most limiting bias and uncertainty from the validation was due to a trend in the spectrum, EALF.

From this trend, a bias of 0.0029 for EALF up to 0.4 eV and 0.0037 for EALFs from 0.4 to 0.6 eV has been determined. Cases without soluble boron are in the first range of EALF and will use 0.0029 for the bias. Heavily borated cases can have an EALF greater than 0.4 eV and then would use the 0.0037 bias.

The 95/95 uncertainty is 0.0050 for all the analyses.

4.2 Burned Fuel Validation - EPRI Approach EPRI determined the change in k with burnup using measured power distributions from 680 flux maps taken over 44 cycles from 4 different PWRs [9]. With this measured data, EPRI created a set of benchmarks that can be used to validate the ch ange in k with burnup found in other code systems. Since the measured parameter is inferred reactivity, the change in k captures the change in the macroscopic cross-section, which is a function of the isotopic concentrations and cross-sections. The data does not provide any information on individual components of the macroscopic cross-section. The measured NET- 300067-01 Rev 0 17

reactivity effect comes from the change in concentration (and cross-section) of all isotopes. This would include any movement of isotopes during power operations, such as gaseous fission products. The measured reactivity also includes any dimensional changes at power, such as crud buildup or creep down of the clad (but not any change in these conditions between hot full power and cold conditions). The SCALE TSUNAMI module was used by EPRI to account for hot to cold variations by a conservative uncertainty [9].

The EPRI benchmarks were analyzed with SCALE 6.1.2 and the 238 ENDF/B-VilI cross-section library. The depletion analysis for the spent fuel pool analysis followed exactly the same method as used for the EPRI benchmark analysis (same time steps, same method of interpolating the isotopic content, and the same method of accounting for decay). Appendix A shows the results of the analysis. A conservative bias is determined to be 0.003 in k. The uncertainty about the bias is conservatively 0.0064 in k.

The complete validation for burned fuel using this technique is the fresh fuel bias from Section 4.1 (0.0029 or 0.0037 depending on EALF) plus the delta-k of depletion bias (0.003) with two uncertainties, 0.0050 from the fresh fuel critical experiments and the EPRI benchmark uncertainty (0.0064).

Refer to Appendix A and References 9 and 10 for more details.

4.3 Burned Fuel Validation - Extended ISG-8 Approach Spent Fuel Project Office Interim Staff Guidance - 8, Rev. 3 - Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks (ISG-8 Rev. 3), follows the traditional approach for transport and storage casks of limiting the number of isotopes that can be considered in the analysis. Spent fuel pool analysis historically utilizes all isotopes. The Extended ISG-8 approach uses the same techniques as developed for ISG-8 but extends the approach to all isotopes.

NET- 300067-01 Rev 0 18

The ISG-8 approach uses critical experiments to cover the worth of actinide isotopes, a TSUNAMI based estimate of a bias to cover the worth of isotopes not in the critical experiments, and chemical assays to cover the isotopic content.

The critical experiments that most closely match spent nuclear fuel are the HTC critical experiments

[13]. These experiments are Mixed Uranium/Plutonium Oxide (MOX) experiments that were designed to match the Uranium and Plutonium isotopic content of 4.5 wt % U-235 fuel burned to 37.5 GWd/T. These experiments were analyzed for this validation. Since there are fuel assemblies in the spent fuel pool that have less burnup, as well as fuel assemblies that have more burnup, additional critical experiments are needed. MOX experiments from the OECD/NEA handbook were added to cover the higher burned conditions and the fresh U0 2 critical experiments from Section 4.1 were added to cover the lower burned conditions.

The most limiting bias and uncertainty from the three sets of experiments: a) fresh U0 2, b) HTC, and c) MOX is used for the bias and uncertainty for the major actinides (U, Pu, Am-24 1). The most limiting bias and uncertainty comes from the fresh U0 2 experiments, since ENDF/B-VII predicts higher k's for MOX critical experiments than U-235 based systems. Refer to Section A.4.1.10 of Appendix A to see the support for this position. This means that the first component of the bias and uncertainty, the component for the worth of the major actinides, has a bias of 0.0029 or 0.0037 depending on EALF and an uncertainty of 0.0050.

The second component of the validation is to cover the worth of the fission products and minor actinides. ORNL studied the uncertainty in the cross-sections and used TSUNAMI to propagate this uncertainty to a cask system [I 4]. ORNL then concluded that a bias of 1.5% of the worth of the fission products and minor actinides was conservative. This conclusion was then inserted into ISG-8, Rev. 3.

For the Extended ISG-8 approach, the same 1.5% of the worth of the fission products is used, but for all isotopes. This too is supported by the same ORNL report. On page 106 of the ORNL report, the last NET- 300067-01 Rev 0 19

sentence states, "An upper value of 1.5% of the worth is also applicable for SNF isotopic compositions consisting of all nuclides in the SFP configuration, as well as the cask configuration as depicted in Table 7.11 and Table 7.12."

The final component of the ISG-8 approach is a bias and uncertainty to cover the isotopic content.

ISG-8 endorses the direct difference approach to calculate a bias and uncertainty due to the depletion analysis. The Extended ISG-8 approach utilizes the same direct difference approach and produces the same bias and uncertainty. The only difference with the Extended ISG-8 approach is that it allows the calculated bias and uncertainty to apply to all isotopes. The isotopic content bias and uncertainty is dominated by the highest worth isotopes (U-235, Pu-239, Pu-240, Pu-241). The addition of more small worth isotopes would have a small effect on this bias and uncertainty. Without measured data on these isotopes it is impossible to prove the effect is small but due to use of the EPRI benchmarks we can be certain that the effect is not large. Therefore, it is concluded that in order to allow this extension to the ISG-8 approach, the analysis of the EPRI benchmarks should be performed to exercise all the isotopes against some experimental data.

Appendix A provides the details of the analysis of the chemical assays followed by the direct difference calculations. From this analysis, the bias of isotopic content is 0 and the uncertainty is 0.0002 times the burnup in GWd/T.

4.4 Burned Fuel Validation - Most Limiting Approach The validation for this criticality safety analysis uses the most limiting of the EPRI or Extended ISG-8 approach. In both approaches, the bias and uncertainty from the fresh U0 2 validation is used. In the ISG-8 approach this is because the U0 2 critical benchmarks are more limiting than the HTC or MOX critical benchmarks. With the EPRI approach there is a remaining uncertainty of 0.0064 and a bias of 0.003 in k. The Extended ISG-8 approach has a bias of 1.5% of the worth of the minor aetinides and fission products. The worth of the minor actinides is about 10% in k at normal discharge burnups. This NET- 300067-01 Rev 0 20

would create a bias of 0.0015 (0. 1* 1.5%). The 0.003 bias from the EPRI approach would be equaled if the worth of the minor actinides and fission products were 20% in k. This is never met, so the EPRI bias is more limiting. Finally, uncertainty from the chemical assays is 0.0002 times the GWd/T. In order to reach the EPRI uncertainty of 0.0064 a burnup of 32 GWd/T is required. The actual rack up of biases and uncertainty for the Indian Point Unit 2 spent fuel pool (See Appendix A, Section A.5) has shown that for the maximum credited burnup (42.67 GWd/T), the EPRI approach is more limiting than the Extended ISG-8 approach.

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5 Depletion Calculations Prior to performing the depletion analysis, a thorough review of the historical operation of Indian Point Units 2 and 3 was performed. Specifically, the following items were reviewed:

I. What inserts were in assemblies and what were the designs of these inserts.

2. How much burnup was achieved while the inserts were in the assemblies.

3. How were control rods used during the cycles.

4. What was the soluble boron level during the cycles.

5. What was the maximum average power peaking (used for fuel and moderator temperatures).

6. When were axial blankets introduced and what were the designs.

Tables 5.1 and 5.2 summarize the key depletion parameters by cycle for Indian Point Units 2 and 3 respectively [30]. Control rod operating history is discussed in Section 5.5. Ti, is the core inlet temperature and T,, is the average moderator temperature in the active fuel (not the vessel average temperature).

Axial blankets (reduced enrichment at the ends of the fuel) reduce the reactivity compared to assuming the highest enrichment over the entire length of the fuel. Several axial blanket designs have been used in Units 2 and 3: 6 inch or 8 inch long blankets, annular and solid pellets, natural or reduced enriched uranium, same or differing enrichments in the top and bottom blankets. Although it would be possible to credit the reduced reactivity caused by the axial blankets, it was decided not to take the credit.

This approach yields the simplicity of only one loading curve and flexibility for possible future design modifications for the axial blankets such as enrichment or annular hole size changes. Such changes would have no effect on the criticality safety analysis since the fill length uniform enrichment assumption would still be bounding.

The Indian Point fuel management and operating approach changed from annual refueling cycles to two year refueling cycles resulting in higher feed enrichments and heavier use of burnable absorbers. In NET- 300067-01 Rev 0 22

order to not penalize the loading curve for older fuel, the depletion analysis was subdivided into fuel assemblies less than or greater than 3.5 wt% U-235. Fuel assemblies over 3.5 wt% U-235 have never been loaded into the core with Pyrex burnable absorbers. Furthermore, historical fuel assemblies under 3.5 wt% U-235 never had IFBA rods.

With a thorough knowledge of the Indian Point Units 2 and 3 operating history, it was possible to select limiting depletion parameters that bound all fuel.

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Table 5.1: Key Operating Features by Cycle Used in Indian Point Unit 2 Max Peak Feed Cycle Feed Blanket Max BA Assem. Soluble Batch Power Tave Tin Burnup Enrich. Enrich. BA* Max BA Burnup Peaking Boron Cycle ID (MWt) ('F) OF) (GWd/T) (wt%) Blanket (wt%) Type Loading (GWd/T) Factor (m) 1 A,B,C 2758 _ _c 16.4 2.2/2.8/3.3 none Pyrex 20 Rodlets 18.5 1.23 890 2 D 2758 10.7 3.1 none Pyrex 20 Rodlets 13.0 1.22 993 3 E 2758 10.8 3.2 none Pyrex 7 Rodlets 12.0 1.24 948 4 F 27.58 9.8 3.35 none Pyrex 12 Rodlets 12.2 1.25 881 5 G 2758 12.2 3.3 none Pyrex 12 Rodlets 15.0 1.22 908 6 H 2758 13.2 3.2 none - Pyrex 20 Rodlets 16.7 1.27 880 7 J 2758 12.4 3.44 none - Pyrex 16 Rodlets 15.5 1.25 883 8 K 2758 13.8 3.2/3.44 none - WABA 12 Rodlets 17.4 1.27 933 9 L 2758 11.4 3.4/3.7 none - WABA 16 Rodlets 14.6 1.28 881 10 M 2758 3071.4 13.3 3.6/4.2 none - WABA 20 Rodlets 17.3 1.30 1050 11 N 3071.4 18.1 3.75/4.05 none IFBA WABA 116(IX)16 Rodlets ____

24.5 1.35 997 12 P 3071.4 20.7 3.6/4.2/4.6 none IFBA 116(1X) 20 28.1 1.36 1121 WABA Rodlets 13 ~~~IFRA 148(1.5X) 281.8 14 13 Q 3071.4 20.9 4.4/4.8 6" annular 2.6 WABA 12 Rodlets 28.8 1.38 1146 14 R 3071.4 19.0 4.6/4.95 6" annular 2.6 IFBA WABA 148(1.5X) 20 Rodlets 26.7 1.40 1069 15 S 3071.4 22.1 4.8 6" annular 2.6 IFBA 148(1.25X) 29.9 1.40 1172 WABA 16 Rodlets 16 T 3114.4 23.9 1_ 4.6/4.95

_WABA 8" annular 3.2 IFBA 148(1.5X) 20 Rodlets 1 32.4 1.35 1164 17 U 3216 18.7 4.0/4.4 8" annular 3.2 IFBA 148(1.25X) 24.9 1.33 1065 WABA 16 Rodlets 18 V 3216 24.5 4.6/4.95 8" solid 3.2 IFBA 148(.25X) 33.1 1.35 1191 except IFBA WABA 20 Rodlets ...

19 326248 48" solid 3.2 Bot IFBA 148(i.25X) 326 1.31 1244 1 except IFBA 3.4 Top WABA 20 Rodlets 20 X 3216 23.7 4.8/4.95 8" solid 3.2 Bot IFBA 148(1.25X) 32.3 1.36 1230 except IFBA 3.6 Top WABA 20 Rodlets 21 2A 3216 25.6 4.6/4.95 8' solid 3.6 Bot IFBA 148(1.25X) 33.8 1.32 1254

_except IFBA 4.0 Top WABA 20 Rodlets BA is burnable absorber NET- 300067-01 Rev 0 24

Table 5.2: Key Operating Features by Cycle Used in Indian Point Unit 3 Max Peak Feed Cycle Feed Blanket Max BA Assem. Soluble Batch Power TCc Te. Burnup Enrich. Enrich. BA* Max BA Burnup Peaking Boron Cycle ID (MWt) (OF) ('F. (GWd/T) (wt%) Blanket (wt%) Type Loading (GWd/T) Factor (ppm)

I A,B,C 3025 a,c 17.3 2.25/2.8/3.3 none - Pyrex 20 Rodlets 19.4 1.12 899 2 P 3025 11.3 3.1 none - Pyrex 12 Rodlets 10.6 1.26 954 3 R 3025 12.8 3.3 none - Pyrex 12 Rodlets 16.2 1.27 912 4 S 3025 14.1 3.2/3.4 none - Pyrex 20 Rodlets 17.3 1.23 968 5 T 3025 14.3 3.2/3.4 none - WABA 16 Rodlets 17.9 1.26 989 6 U 3025 14.8 3.2/3.6 none - WABA 12 Rodlets 19.5 1.31 1017 7 V 3025 13.4 3.4/3.8 6" solid 0.72 IFBA 60(IX) 18.2 1.35 974 WABA 20 Rodlets 8 W 3025 14.1 3.8/4.2 6" solid 0.74 WABA 12 Rodlets 18.9 1.35 1220 9 X 3025 19.2 4.0/4.4 6" solid 0.74 IFBA 116(1.5X) 25.9 1.35 1169 WABA 20 Rodlets 10 Y 3025 22.4 4.4/'4.6 6"' annular 2.6 IFBA 80(1.5X) 31.1 1.38 1163 1 WABA 20 Rodlets 11 AA 3025 18.7 4.3/4.6 6" annular 2.6 IFBA 80(1.25X) 25.9 1.38 1095 S _ I_ WABA 20 Rodlets 1 133 3025/ 23.0 4.5/4.95 8" annular 3.2 IFBA 100(I.5X) 3. .1 12 1 3067 WABA 20 Rodlets 30.3 1.31 122 13 CC 3067.4 23.7 4.95 8" annular 3.2 IFBA 100(1.5X) 32.1 1.36 1246 I WABA 20 Rodlets.

14 DD 14 DD 3180/

3188 25.2 4.6/4.95 8" annular 3.2 IFBA WABA 148(1.25X) 20 Rodlets 33.1 1.31 1240 15 EE 3188.4 24.9 4.6/4.95 8" annular 3.2 WFBA 148(1.25X) 32.0 1.28 1228

_____ ______ _____ ______WABA 20 Rodlets ____

16 FF 3188.4 24.1 4.6/4.95 8" annular 3.2 Bot IFBA 148(l.25X) 31 6 1.32 1179 3.6 Top WABA 20 Rodlets 1 8" solid 3.6 Bot IFBA 148(1.25X) 332 1.33 1232 17 GG 3188.4 25.0 4.95 except 38TpWB 0Rdes 3. .3 13 eFBA 3.8 Top WABA 20Rodlets 8" solid 3.6 Bot IFBA 148(1.25X) 320 1.30 1258 18 3D 3188.4 24.6 4.6/4.95 except 40TpIWB 0Rdes 3. .0 15 exep 410 Top WABA 20 Rodlets BA is burnable absorber NET- 300067-01 Rev 0 25

5.1 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 burnup [15]. The spectrum hardening comes from the absorption of thermal neutrons by the boron and displacement of the water in the guide tubes. The effect on reactivity also depends on how long (in terms of GWd/T) the burnable absorber remains in the fuel before being removed.

Indian Point Units 2 and 3 have used three types of burnable absorbers: Pyrex, Wet Annular Burnable Absorbers (WABA) and Integral Fuel Burnable Absorbers (IFBA). The Pyrex and WABA designs consist of rodlets mounted to a base plate which sits on the top of the fuel assembly. The number of rodlets varies by position in the core to help control power peaking. The most limiting design has 20 rodlets. The depletion analysis assumes the full 20 rodlets. There have been variations in the boron loading in the burnable absorbers. The highest boron level is most limiting and is assumed in the depletion analysis (see Table 3.4 for the assumed boron level). Pyrex burnable absorbers are more limiting than WABAs since they displace more water. I['BA rods and removable burnable absorbers can be in the assembly at the same time. The number of IFBA rods is varied to aid in power distribution control. The maximum number of IFBA rods (i.e., 148) is assumed. The boron content of the IFBA rods can vary. The maximum boron loading of [ mg B30 per inch]"c (1 .5X) is assumed in the depletion analysis. Most of the IFBA rods have been at a 1.25X loading (see Tables 5.1 and 5.2).

For fuel enriched to 3.5 wt% U-235 or less, Pyrex burnable absorbers are assumed to be in the assembly throughout the entire burnup. For fuel over 3.5 wt% U-235, WABAs are assumed to be in the fuel and never removed. However, control rods are assumed to be in the fuel for 2 GWd/T before the WABAs are inserted to cover future extended part power operation. Since control rods are more limiting than WABAs and since control rods and WABAs cannot be in the assembly at the same time, the analysis is valid for any amount of WABA burnup as long as the control rod exposure is less than 2 GWd/T. This 2 GWd/T is for the assembly burnup but due to the axial burnup distribution, it is only I GWd/T in the NET- 300067-01 Rev 0 26

top node. Therefore, operation for a full cycle with the control rods at the "bite" position is prohibited for fuel enriched to greater than 3.5 wt% U-235. The limitation of 2 GWd/T of assembly bumup under control rods will be checked as part of the reload safety checks.

Finally, gadolinium and erbium have not yet been used as burnable absorbers at Indian Point. These burnable absorbers have a negative impact on reactivity due to incomplete burnout of the absorber [16].

If gadolinium or erbium is used in the future, then this criticality analysis is valid.

5.2 Limiting Depletion Parameters- Soluble Boron Soluble boron hardens the spectrum, making the fuel more reactive for a given burnup. It has been shown that taking the burnup average soluble boron level is acceptable (rather than a time dependent soluble boron) [171. Tables 5.1 and 5.2 show the peak soluble boron concentration. The average cycle soluble boron concentration is less than 70% of the peak values (confirmed for limiting cases) reported on Tables 5.1 and 5.2. Review of the operating history with fuel less than 3.5 wt% U-235 shows that 800 ppm is higher than any cycle averaged ppm, so 800 ppm was selected for the depletion of this fuel. For fuel enriched to more than 3.5 wt0/o U-235, a multi-cycle assembly average bounding value of 1000 ppm soluble boron is selected. Since the reload core designer must limit the soluble boron level in order to maintain a negative moderator temperature coefficient, it is unlikely that 1000 ppm burnup averaged soluble boron will ever be a limiting design feature. However, this parameter will be included in the reload design checks. Early shutdown could cause this average soluble boron to be exceeded. To handle this possibility, a special depletion at 1300 ppm was performed. Fuel with a burnup > 12 GWd/T can be placed anywhere in Region I and fuel with a burnup < 12 GWd/T can be placed on the periphery of Region I or with a control rod inserted and placed anywhere in Region 1.

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5.3 Limiting Depletion Parameters- Temperatures The higher the fuel and moderator temperatures during depletion, the higher the reactivity. As with most parameters assumed during depletion, the time ordering of the effect has a very minor impact on the final reactivity. Therefore, burnup averaging of the fuel and modcrator temperature is appropriate. A limiting fuel and moderator temperature can be found by determiling the maximum burnup averaged peaking factor. Since criticality requires a volume of low enriched fuel that is much greater than a fuel pin, the average peaking factor of interest is the assembly averaged value. This is easy to determine, since it is simply the assembly burnup divided by the sum of the core average burnup for the cycles that the fuel assembly was in the core. Tables 5.1 and 5.2 show the maximum assembly/burnup averaged peaking for any assembly in the core for that cycle. The discharge average peaking factor will be lower since the peaking factor for an assembly is lower in the second and following cycles.

The Indian Point Units 2 and 3 fuel, which was enriched to less than 3.5 wt% U-235 has been reviewed and the burnup averaged assembly peaking factor was always less than 1.35.

The moderator temperature increases as the water rises through the core. This depletion analysis uses a conservative outlet temperature for the entire length of the fuel. The moderator temperature is determined by multiplying the peaking factor times the enthalpy rise across the core. ['his delta enthalpy is added to the inlet enthalpy to determine an outlet enthalpy that is then converted to a temperature using the system pressure. The core inlet temperature and outlet temperature at Indian Point Units 2 and 3 have varied over the cycles, so the highest core enthalpy rise is used with the highest core inlet enthalpy. The calculation accounts for the effects of bypass flow. The resulting moderator temperature with a multi-cycle bumup averaged peaking factor of 1.35 is 628 'F (604.4 'K) and the moderator density is 0.64765 g/ce.

The moderator temperature for fuel where the enrichment is greater than 3.5 wt% U-235 requires projection of future peaking and operating conditions. Currently, the reload safety analysis utilizes a NET- 300067-01 Rev 0 28

limiting assembly average peaking factor of 1.40 at all times during the cycle. Using a multi-cycle assembly burnup average peaking factor of 1.40 and the highest core delta enthalpies from the past cycles, a moderator temperature of 631 TF (605.9 'K) and a moderator density of 0.64264 gm/cc was determined.

This determination accounted for the effects of bypass flow. This approach is conservative, since an assembly cannot operate at the limiting peaking factor throughout its life.

The fuel temperature also needs to be maximized. The same assembly averaged peaking factors were used to burnup average the kw/fi. Figure 5.1 shows the fuel temperature as a function of peaking factor and burnup. These temperatures were generated with INTERPIN-3 [18) and provided in Ref. [32]. The depletion analysis model used a constant fuel temperature for the duration of the burnup. The fuel temperature used in the analysis was conservatively larger than the fuel temperature at any credited time of life (less than 45 GWd/T). The fuel temperature used for fuel under 3.5 wt% U-235 (i.e. fuel with a peaking factor of 1.35 or less) is 1075 'K (1475 'F) and for fuel over 3.5 wt% U-23 5 (i.e. fuel with a peaking factor of 1.40 or less), it is 1100 'K (1520 TF).

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INTERPIN-3 Average Fuel Temperature versus Bumup at Various Relative Powers Typical IP-2 Cycle 20 1300 000 9 5 10 IS 20 25 30 35 4Q 45 50 55 ,0 U5 eurnup (GWDIMNT)

Figure 5.1: Fuel Temperature Change with Burnup and Relative Power 5.4 Limiting Depletion Parameters- Specific Power ORNL has performed a study of the sensitivity of burnup credit to specific power and determined it is a small effect [17]. 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 fucl can operate at only one specific power, a nominal specific power was selected in conjunction with conservatively selected higher temperatures. This approach is consistent with the ISG for spent fuel pool analysis [1].

As with specific power, the power history also has a small effect [17,19]. References 17 and 19 recommend ignoring any down time and assuming that the reactor is operated at full power over the entire depletion. Reference 19 saw a small positive reactivity increase when the last part of the depletion was performed at lower power. Lower specific power at end of cycle is more reactive because less Pm- 149, NET- 300067-01 Rev 0 30

the precursor to Smi-149, is produced. This effect has been specifically accounted for in this analysis by reducing the Pm-149 that could be the result of a power coast down (see Section 5.9).

5.5 Limiting Depletion Parameters- Control Rod Operation The Indian Point Units 2 and 3 have 193 fuel assemblies in the core. Nine of these assemblies are under the Control Bank D (less than 5% of the number of assemblies). Control Bank D is the only control bank allowed in the core at all if the power is greater than 70% of the rated power. Indian Point Units 2 and 3 have not operated with Control Bank D in the core for any significant burnup except at the "bite" position. The bite position is set as the location where the worth of the lead control bank is 2 pem 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.

The top node in the axial model used for the criticality analysis is 8 inches.

For fuel that is less than or equal to 3.5 wt% U-235, the depletion assumption for the top node is that the control bank is in for the entire burnup. Since this fuel has already been discharged, no control rod insertion below the top node is needed in the analysis.

For fuel enriched above 3.5 wt% U-235, the assembly is depleted with a control rod inserted for 2 GWd/T. Due to the low relative power in the top node (about 0.5) (see Table 6.1), the depletion of the top node with a control rod is I GWd/T (all other nodes are burned for 2 GWd/T with the control rod).

Unit 3 never used control rods at the bite position, but Unit 2 positioned its D bank control rods at the bite position through cycle 17. All assemblies that were above 3.5 wt% U-235 enriched that were under Control Bank D for cycles I through 17 were evaluated as a special case in the depletion analysis. The reactivity penalty of the bite position versus no bite position was determined to be approximately I GWd/T additional required burnup. Each of these assemblies exceeded the burnup requirements by more than 4 GWd/T. Thus these assemblies are acceptable to be loaded in Region 2.

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5.6 Summary of Depletion Assumptions for Fuel < 3.5 wt% U-235 Since Pyrex burnable absorbers were not used in fresh fuel for enrichments greater than 3.5 wt%

U-235 and IFBA rods were not used for enrichments less than 3.5 wt% U-235, separate depletion assumptions are used. The last fuel of 3.5 wt% U-235 enrichment or less was loaded in Unit 2 in cycle 9 and Unit 3 in cycle 7. Since it is not anticipated that future cycles will use fuel with enrichments less than 3.5 wt% U-235, it was possible to credit the actual operation of this fuel for the selection of limiting depletion parameters, such as the burnup averaged soluble boron concentration and the maximum assembly and burnup averaged moderator outlet temperature.

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For fuel enriched to less than or equal to 3.5 wt/o U-235 (old fuel), the following depletion assumptions are made:

a. Nominal fuel dimensions at 97.5% theoretical density

b. 800 ppm maximum average boron concentration during depletion (a multi-cycle average)

c. The specific power used during depletion is 39.8 w/g. This is about 8% higher than the nominal specific power (specific power has a very small effect on the final reactivity).

d. Moderator temperature during depletion calculated from a burnup average assembly peaking factor of 1.35 is 628. `F (604.4 'K) and the moderator density is 0.64765 g/cc.

e. Maximum average fuel temperature (or maximum average resonance temperature) during depletion is assumed to be 1075 'K (1475 OF).

f. Assembly is depleted with a 20 rodlet standard BP (Pyrex) and never removed. A Pyrex loading of 18.1 wt% (maximum) is assumed. This depletion will also cover any WABA depletion since WABAs are less limiting than Pyrex. No IFBA is assumed because IFBA was not introduced until later when the utility was ordering fuel greater than 3.5 wt% U-235.

g. A second depletion is performed with a 20 rodlet standard control rod and never removed.

The isotopics from this depletion are used for the top node (top eight inches) in the criticality calculation. This assures that the criticality analysis is conservative even if a control rod were in the bite position for the entire life of the assembly.

h. For Pyrex, WABA and control rod depletion, Table 5.3 provides the dimensions and material used.

Table 5.3: Characteristics of Fuel Inserts Parameter Pyrex Control Rod WABA Absorber ID (inches) 0.243 0 0.278 Absorber OD (inches) 0.396 0.3975 0.318 B203 wt% 18.1 B- 10 mg/cm - _ 6.03 Ag wt% - 80 In wtO/o - 15 -

Cd wt% - 55 Absorber density (jec) 2.23 10.17 NET- 300067-01 Rev 0 33

5.7 Summary of Depletion Assumptions for Fuel > 3.5 wt% U-235 For fuel enriched to more than 3.5 wt% U-235 (modern fuel), the following depletion assumptions are made (hereinafter referred to as modern depletion assumptions). These assumptions are based on a review of the operation history [30] and are selected to cover anticipated future fuel designs and operation.

a. Nominal fuel dimensions at 97.5% theoretical density

b. 1000 ppm maximum average boron concentration during depletion (a multi-cycle average).

c. The specific power used during depletion is 39.8 w/g (specific power has a very small effect on the final reactivity).

d. Moderator temperature during depletion calculated from a burnup average assembly peaking factor of 1.40 is 631 F (605.9 'K) and the moderator density is 0.64264 g/cc.

e. Maximum average fuel temperature (or maximum average resonance temperature) during depletion is assumed to be I 100 'K (1520 TE). Maximum average fuel temperature in the top node is assumed to be 930 'K (1214 'F ), corresponding to a relative power of 1.0.

f. The fuel assembly is depleted with a control rod inserted for 2 GWd/T. Then the assembly is depleted with a 20 rodlet WABA and never removed. The initial control rod depletion is to cover future extended part power operation with control rods inserted. Although generous, the 2 GWd/T burnup with control rods inserted still needs to be checked. This would be violated only if a fuel cycle is operated at less than full power with control rods for a significant amount of time (greater than 1.4 full power months during a 2 year fuel cycle).

g. In addition to the WABA, the fuel is assumed to have 148 IFBA pins in all but the top node.

The IFBA B-I 0 loading is 1.5X [( mg B-10/inch)]i' to cover future designs (residual poison from IFBA is not credited in the criticality model).

h. Similar calculations are performed to confirm that an assembly with an initial WABA depletion that is subsequently placed on the core periphery with a Hafnium flux suppression insert is covered by an adder to the normal loading curve (see Section 8.4).

i. Since IFBA and WABA do not cover the full length of the active fuel, a special depletion for the top node is performed. For fuel enrichments greater than 3.5%, the top node is depleted with a control rod for 1 GWd/T nodal burnup (the axial burnup profile for the top node is always less than 0.5) and then a WABA (no IFBA) at 1/4 of the boron and never removed.

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j. Justification for this is that the practice of positioning the control rod in the bite position has been discontinued. The boron in a WABA never extends more than 6 inches from the top and the IFBA never extends more than 8 inches from the top of the active fuel region [24].

k. A special depletion was performed for 5.0 wt% fuel under a condition of 1300 ppm. This was done to allow burned fuel to be stored in Region I (see Section 8.1).

5.8 Depletion Analysis Details (Time Steps, etc.)

Using the above assumptions, U0 2 fuel was depleted at fuel enrichments of 2.0, 2.5, 3.0, 3.5, 4.0, 4.5, and 5.0 wt% U-235. The burnup points at which the isotopic data is collected are 150, 500, 1000, 1500, 2000, and then every 2000 MWD/MTU after that. Although the depletion is carried out with a full set of 388 nuclides, the isotopics used in the pool model is a reduced set (185 nuclides).

5.9 Reduced Power Operationat End of Life At end of cycle, the reactor power may be reduced (for example, a planned coastdown). 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 is reduced to one half of the full power content for all criticality calculations (which results in a penalty of about 100 pem). This covers coastdowns to 50%

power. This approach covers all past operating experience and anticipated future operation at the Indian Point plants.

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Proprietary Information Removed 6 Rack Model This Section describes the infinite models used in most of the analysis. A full pool model was also created to test the reduced burnup requirements for peripheral assemblies, the Region I/Region 2 interface and to perform analyses of the Dropped Assembly and Misplaced Assembly accidents. The full pool model is described in Section 8.10.

All the infinite models discussed in this Section are finite axially. Section 6.3 describes the axial modeling of the fuiel. Above and below the fuel, the models have 50 cm of water followed by a zero flux boundary condition.

Water at the maximum possible density (1.00 g/cc at 4 UC) is assumed, which results in the largest reactivity for both Regions (see Table 8.9).

The Boraflex sheathing width ranges from 7.70 to 8.00 inches (see Table 3. 1). The minimum value is used in Region 1, while the maximum value is used in Region 2, because calculations demonstrated that this results in the highest reactivity in each respective region (see Table 7.2). The stainless steel plate that covers the Boraflex is 0.112 +/- and 0.092 + inches off the cell walls For Regions 1 and 2 respectively (from Table 3. 1). Since using a smaller separation from the cell wall means that less water is on the inside that is balanced by more water on the outside (and vice versa for a larger distance), this tolerance has no effect on the total mass of water. Therefore, the reactivity effect of this tolerance is insignificant, so the nominal separation is used for all the analyses.

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6.1 Region 1 Infinite 2x2 Model Figure 6.1 is a picture from the KENO model for Region 1.

Figure 6.1: Region 1 KENO Model The absorber panel is shown in green. Inside the Boraflex sheathing is water. No Boraflex is in the model. If any Boraflex material remains, the remaining boron in the material makes the reactivity more negative than water.

The only simplification made is that the connecting steel between cells is modeled as an extension of the cell wall rather than a separate piece of steel. The connecting steel is slightly thicker than the cell wall thickness and modeling it as equal to the cell wall thickness has no effect (as illustrated by the results in Table 7.2).

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6.2 Region 2 Infinite 2x2 Model Figure 6.2 is a picture from the KENO model for Region 2.

Figure 6.2: Region 2 KENO Model All of the features of the rack are modeled explicitly. The bottom left of the model is a complete cell box with its Boraflex sheathing. The model is two cell pitches. This requires cutting the cell wall into two pieces. When the periodic boundary condition is applied, the two cell wall pieces fit together to precisely match the actual rack dimensions.

For the alternate panel design, the panel consists of two absorber panels connected with a stainless steel connector. This is modeled as the two panels connected with a zirconium connector that is 0.10 inch thick (shown in Figure 6.3 below). This conservatively accounts for the water displacement of the connector without adding any significant absorbing material.

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Figure 6.3: KENO Model of the Alternate Panel Design 6.3 Axial Burnup Distribution To model the fuel assembly isotopic content in three dimensions, an axial burnup profile is needed.

For this analysis, the limiting profiles from NUREG/CR-6801 [20] are used. Table 6.1 is taken from NUREG/CR-6801.

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Table 6.1: Axial Burnup Profile vs. Burnup Bin [201 Burtaup gloup 1 2 3 4' 6 7 8 9 10 II 12, Axial -- B-urnu v lGWdnMTU) - - -

height S 406 42.46 38 42 34 38 30-34 26.30 22 26 IS 22 14.18 10-4, 6.10 ý'6 278 0.5,82 0.c66 0.660 0.(.48 0.652 0.610 0630 0668 649 0 633 0.658 0.631 8.33 0920 0Th14 0936 0955 0,967 0.024 0 936 I 0314 I 044 0.989 I 007 1 007 13 8) 1.065 1.048 1.045 I 070 1.074 1.056 I 066 1.150 I 208 1.019 I.09I I 135 19.11 I 105I 081 1,080 I I041 1,103 1 097 I 103 1.094 1215 0857 I 070 I 111 2500 I 113 1.089 1.091 1,112 1.108 1.103 I 108 1.053 1.214 0,776 1022 1098 10.56 1.110 1.00 1093 I 112 1.106 I 101 I 109 1.048 1.208 0.754 0.98M 1069 36 II I 1O5 1.08b 1.091 I 108 1.102 1 103 I 112 1064 I 197 0.785 0.978 1.053 41 69 1 10 1,085 1090 I 1105 1 097 I 112 I 119 1.095 I 189 1 013 0989 1(047 47.22 1.015 1.084 1.084 1.102 1.094 1.125 1.126 1.121 1.188 1,185 1.031 1.050 5780 1.09) 1.081 I 088 L.OAN 1.094 1.136 1.132 1.135 1.192 1.253 1.082 1,060 58 33 1 088 1 085 I 088 1 097 I 095 1.143 I 135 1.140 1.195 1 278 1.110 1 070 61.89 1 084 I 0,6 1 086 I 095 1 09h I 143 I 115 1 138 1 190 1 283 I 121 1 077 609.44 1 080 I 0}86 1 1)84 I 091 10195 1.'36 I 129 1.130 1.156 I 276 1 124 1.079 75.00 1.072 1.083 1.7 1 81 1.098 1.115 1.109 .l9O 1.022 1.251 1,120 1.073 M1.56 I 050 1.06) 1.057 1.056 I 051) 1.047 1.041 1.049 0.756 I 193 I 101 I 053 1

86 II 0.i) 92 1.010 0.996 0.974 0.971 0.882 0.871 0,933 0.614 1.075 1.045 09)96 91.67 0833 0.811 0.823 0.743 0.738 0.701 0689 0.660 0.481 0.863 0.894 0.845 9722 10515 0.512 1) 525 0447 0.162 0456 '1

)48 0373 0284 0.515 10569 0525 In general, the profile becomes less limiting as the burnup increases. However, there are some burnup bins in which the profile becomes more severe as the burnup increases. For example, an assembly having a burnup of 15 GWd/T would have to use a very reactive shape (14-18), but if the burnup were only 13 GWd/T, the shape is much less reactive. To ensure that all burnup bins are conservative, a burnup bin that has this characteristic (the shape is less severe at a lower burnup) is assumed to have the more limiting shape from the higher burnup. So the burnup profile used in the entire range 0-18 GWd/T is the one in the 14-18 GWd/T burnup bin. Furthermore, there are discontinuities at the burnup bin boundaries. To eliminate these discontinuities in a conservative manner, the shape in any bin is assumed to occur at the maximum burnup in the bin and for any burnup in between these bumup points, the shape is linearly interpolated (the shapes are not changing rapidly between burnup bins). For example, suppose an assembly has a burnup of 39 GWd/T. Using NUREG/CR-6801 directly, the top node would have a NET- 300067-01 Rev 0 40

relative burnup of 0.525. In this analysis, however, the top node would have 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.

According to Reference [23], the assembly having the most reactive axial burnup profile of all Indian Point fuel is R-08. This profile was modeled in a 4.5 wt% assembly burned to 39 GWd/T and cooled for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (the approximate loading curve burnup). The k using the R-08 profile was 0.9547. The k using the DOE shape was 0.9632, which shows that the DOE shape is bounding even for R-08. Furthermore, R-08 is actually axially blanketed (top blanket enriched to 2.6 wt%) and the top node was under a control rod for its entire depletion. So the R-08 profile is very unrealistic and not physical for non-blanketed fuel.

In order to determine a more realistic reactivity, the axial blanket and control rod depletion were modeled explicitly and the k decreased to 0.9388. So the DOE shapes in conjunction with non-blanketed fuel are bounding for all the non-blanketed fuel at Indian Point and conservative for axial blanket fuel.

For the analysis, the lower 10 nodes were 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 (but a negligible effect).

6.4 Interpolation of Isotopics and Cooling Time Verification With isotopics from the depletion calculations recorded every 2 GWd/T (see Section 5.8), the isotopics at any particular burMup 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 plot files (OPUS plot files are an output option in the TRITON module of SCALE) at 38 GWd/T and 42 GWd/T at 5.0 wt% U-235 enrichment. The isotopics were input into the Region 2 KENO model and the calculated k was 0.96653 +/-2 0.00015. Another case used isotopics directly from the OPUS plot files at 40 GWd/T. The calculated k was 0.96651 +/- 0.00015. The difference is well within the Monte Carlo statistics. A similar verification was performed at 11 GWd/T, where an NET- 300067-01 Rev 0 41

interpolation between 9 and 13 was compared to the direct calculation at 11 GWd/T. The calculated k 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.

To check the cooling time model used in the interpolation program, a special depletion was performed at 5.0 wt% U-235 to a burnup 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 6.2 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 k from the interpolation program at 25 years is conservative.

Table 6.2: 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 _

I year 0.9993 0.9996 0.0003 5 years 0.9847 0.9848 0.0001 25 years 0.9449 0.9457 0.0008 6.5 Convergence of Calculations The convergence of the 2x2 (or IxI) reflected model k calculation was generally achieved after only a few hundred generations. However, all of the CSAS5 computer runs use a Monte Carlo sampling of at least 1500 generations and 6000 neutrons per generation. Convergence could have been a problem in the past, when very few neutron histories were run (300,000 total neutron histories vs. 9 million today).

The full pool model convergence required knowledge of the most reactive location to be used for a starting source. When the most reactive location was not known, several starting sources were used.

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6.6 Summary of Modeling Assumptions The following is a summary of the modeling assumptions:

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

2. Axial blankets are ignored.

3. Grids are ignored (grids displace water which causes k to decrease).

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

5. NUREG/CR-6801 bounding axial burnup profiles are used for all fuel (bounding for non-blanketed fuel and is conservative for axial blanket fuel).

6. Top and bottom of fuel assembly is 50 cm of water (conservative compared to 50/50 steel/water and other minor end fitting components - see Section 8.5).

7. Isotopics can be linearly interpolated between burnup steps that are 2 GWD/T apart (verified by comparison to non-interpolated isotopics).

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

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Proprietary Information Removed 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 Tolerances Calculations were performed to quantify the reactivity effect of changes due to manufacturing tolerances. For Region 2 the tolerance calculations were performed at the highest credited burnup conditions (5 wt% U-235 and 42.67 GWd/T) and a low burned condition (2.5 wt% U-235 and 12.31 GWd/T'). Table 7.1 presents the tolerance reactivity effects for Region I (fresh 5 wt% U-235 fuel) and Region 2. The tolerance effects were not very sensitive to the enrichment/burnup conditions and the largest values from Table 7.1 are used in the rack up of uncertainties.

Table 7.1: Tolerance Reactivity Effects Tolerance Value (in) Ak Region I Ak Region 2 High Burnup Case Base k = 0.9639 Fuel pin pitch +0.0014 0.0021 Rack cell pitch - horizontal 0.0035" Rack cell pitch - vertical Rack cell ID 0.0002 Rack cell wall thickness +/-0.007 0.0002 Boraflex sheathing thickness +/-0.003 0.0001 Eccentricity 0.0001 Fuel Enrichment +0.05% 0.0030 @ 5.0 Low Burnup Case Base k 0.9717 Base k = 0.9696 Fuel pin pitch +0.0014 0,0022 0.0016 Rack cell pitch - horizontal 0.0039 0.0035" Rack cell pitch - vertical 0.0041 Rack cell ID 0.0020 0.0001 Rack cell wall thickness +0.007 0.0008 0.0002 Boraflex sheathing thickness +/-0.003 0.0004 0.0001 Eccentricity - 0.0001 0.0001 Fuel enrichment +0.05% 0.0022 @ 5.0 wt% 0.0050 @ 2.0 wt%

Both horizontal and vertical pitch were changed at the same time in Region 2 NET- 300067-01 Rev 0 44

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 (shown in Table 7.1), guide tube diameter (Table 7.2), and the fuel clad outer diameter (Table 7.2). For this analysis, the fuel pin clad outer diameter is set to its minimum value. Also, the water displacement of the guide tubes is at the minimum. These parameters have sufficiently small tolerances that bounding values were used. 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 assemblies would touch in the reactor. The impact of increasing the pin pitch is one of the larger reactivity effects from the manufacturing tolerances. Since it causes a change in the spectrum in the assembly, the reactivity effect is about the same in Region I and Region 2.

The fuel enrichment uncertainty used here is the uncertainty associated with ordering fuel. The as-built uncertainty is much smaller. Once the as-built enrichment is known, the as-built enrichment should be used to compare against the loading curve. If the as-built enrichment is not known, then it is acceptable to use the ordered enrichment. The fuel enrichment uncertainty in Region 2 is a function of enrichment and is linearly interpolated using the two points shown in Table 7.1 (the uncertainty changes monotonically between enrichments).

A tighter rack cell pitch increases k of the pool for both regions because the fuel assemblies are closer together. Also, a tighter pitch reduces the moderation prior to the neutrons hitting the absorber panels.

The Region I change in k for reducing the cell pitch is much larger due to the decrease in the flux trap.

However, since the vertical and horizontal tolerances are independent of each other, the net effect for Region I is only mildly larger than Region 2.

NET- 300067-01 Rev 0 45

In calculating the reactivity of increasing the cell ID, the cell pitch is maintained. This means the reactivity effect in Region I is mainly due to decreasing the flux trap. Region 2 is not a flux trap design, so the effect on k is small.

Calculations demonstrated that having maximum water at the edge of the cell maximizes the reactivity in Region 1, while having minimum water at the edge maximizes the reactivity in Region 2 (see Table 7.2). Since bounding values are used for the absorber panel thickness and the Boraflex sheathing width, a minimum panel thickness and minimum Boraflex sheathing width are used in Region 1, while a maximum panel thickness and maximum sheathing width are used in Region 2.

Eccentric positioning of the assemblies in the cell (moving four assemblies together in a repeating infinite array) has very little effect on k. In fact, the effect was smaller than the expected Monte Carlo variation, even after increasing the number of neutrons being followed by a factor of 4. This result is common for racks with absorber panels. This negligible eccentricity is included as an uncertainty.

Calculations confirmed that the highest reactivity occurs with maximum water density (see Section 8.7). Therefore, all analyses were performed at a water density of 1.00 g/cc (4 'C).

Table 7.2: Miscellaneous Reactivity Effects Case (base k =0.9639) Ak from base Notes Max panel thickness in Region I -0.0003 Min thickness is limiting Max sheathing width in Region 1 0.0000 Within statistics Min panel thickness in Region 2 -0.0003 Max thickness is limiting Min sheathing width in Region 2 -0.0002 Max width is limiting Guide and Instrumentation Tube Diameter 0.0000 Within statistics reduced to OFA dimensions Decrease clad ID by [ ],,c inch -0.0001 Increase clad OD by [ ]ac inch -0.0024 All of these results show Decrease pellet density by 1% -0.0001 that the correct bounding Decrease pellet OD by ))a,c inch -0.0002 fuel dimensions are being Decrease GT ID by [ ]a.c inch -0.0002 used Increase GT OD by [ ]*'- inch -0.0002 NET- 300067-01 Rev 0 46

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

7.2 Calculation of Biases and Uncertainties The total bias and uncertainty that needs to be applied to the raw calculated k consists of the following components:

• Bias and uncertainty from the critical experiment validation

• Bias and uncertainty from the depletion reactivity validation

• Fuel tolerance uncertainties for pin pitch and enrichment

" Rack tolerance uncertainties for cell pitch, cell ID, cell wall thickness, and sheathing thickness

" Uncertainty for eccentric positioning

" Uncertainty in the reactor record burnup

" Monte Carlo statistical uncertainty in the k calculations The biases arc added together and the uncertainties are statistically combined. From the validation section, the critical experiment bias is 0.0029 for unborated conditions and 0.0037 for borated conditions.

The critical experiment uncertainty is 0.0050 for both unborated and borated conditions. The depletion reactivity bias is 0.003 using the EPRI benchmarks (see Section A.4.2). Alternatively, using the Extended ISG-8 approach (see Section A.4.3.4) the depletion reactivity bias is 1.5% of the minor actinide and Fission product worth. At 5.0 wt%, 43 GWd/T, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooling in Region 2, a calculation determined that the minor actinide and fission product worth is 0.1168 (1.0677 - 0.9509). So the bias using the extended ISG-8 approach at 43 GWd/T is 0.1168*0.015 = 0.0018. The depletion reactivity uncertainty using the EPRI benchmarks is 0.0064 for all burnups. The depletion reactivity uncertainty using ISG-8 (chemical assays) is 0.0002 x BU from Section 4. So at 43 GWd/T, the chemical assay uncertainty is 0.0086. It turns out that the combination of a smaller bias and a larger uncertainty for the ISG-8 approach NET- 300067-01 Rev 0 47

is less limiting than using the EPRI benchmarks at 43 GWd/T. Since the maximum bumup in the loading curves is only 42.67 GWd/T (5 wt%, 0 years cooling time), the EPRI benchmarks are used for the depletion reactivity bias (0.003) and uncertainty (0.0064).

The uncertainty in the reactor record bumup is assumed to be 5% of the burnup. The effect on reactivity can be calculated by comparing the k calculated for the same enrichment at two different burnups. For example, at 5.0 wt% U-235, the k at 38 GWd/T was 0.9948 while the k at 42 GWd/T was 0.9658. So the Ak due to a 5% burnup uncertainty at 42 GWd/T is (0.9948 - 0.9658)*0.05*42 / (42 - 38) = 0.0152 Ak The same calculation at 3.0 wt%, 18 GWd/T (using calculated k's at 18 and 21 GWd/T) is (0.9943 - 0.9592)*0.05* 18 / (21 - 18) = 0.0105 Ak To simplify the burnup uncertainty calculation for all burnups greater than 18 GWd/T, the following linear relationship is derived from these two points.

Burnup Uncertainty (Ak) = 0.0072 + 0.00019 x BU This relationship was found to be conservative for all burnups greater than 18 GWd/T and overly conservative for burnups less than 18 GWd/T. For burnups less than 18 GWd/T, the maximum slope was found to be 0.005 Ak per GWd/T and so the burnup uncertainty is:

0.005 x .05 x BU = 0.00025 x BU As an example of the total rack up, Table 7.3 summarizes all of the biases and uncertainties for 5.0 wt% U-235 fuel burned to 43 GWd/T in Region 2 under unborated conditions.

NET- 300067-01 Rev 0 48

Table 7.3: Rack Up of Biases & Uncertainties in Region 2 for 5 wt% Fuel at 43 GWd/T Uncertainty Bias Ak Validation (critical experimentsQ 0.0029 0.0050 Depletion reactivity uncertainty' 0.0030 0.0064 Burnup uncertainty - 0.0154 Fuel pin pitch - 0.0021 Fuel enrichment - 0.0030 Rack cell pitch - 0.0035 Rack cell ID - 0.0002 Rack cell wall thickness - 0.0002 Boraflex sheathing thickness - 0.0001 Eccentricity - 0.0001 Monte Carlo statistics - 0.0005 Total Rack Up (Ak = RSS) 0.0059 0.0182 Total Bias and Uncertainty 0.0241 In this example, the bias plus uncertainty is 0.0241. To provide an engineering margin of 0.01, the target k would be 0.99 - 0.0241 = 0.9659. Note that the uncertainty is dominated by the validation, depletion and burnup uncertainty. Small changes in the manufacturing tolerances will have no effect on the statistically combined uncertainties.

Table 7.4 presents the rack up of biases and uncertainties for Region I with fresh fuel. For Region 1, the target k would be 0.99 - 0.0111 = 0.9789. For burned fuel in Region 1, we would also have to add burnup and depletion reactivity biases and uncertainties, which would increase the total bias and uncertainty to 0.0 167 at 12 GWd/T. So the target k For burned fuel in Region 1 would be 0.9733.

For the borated condition, the uncertainties will be slightly different [27] but these differences arc insignificant when compared to the large margin for the borated condition shown in Section 8.

Using the Extended ISG-8 approach, the depletion bias is 0.0018 and the depletion uncertainty is 0.0086. Using these values in the total rack up, the total bias is 0.0047 and the total uncertainty is 0.0191 for a total bias and uncertainty of 0.0238, which shows that using the EPRI depletion bias and uncertainty is more limiting (0.0241) than using the Extended ISG-8 approach.

NET- 300067-01 Rev 0 49

Table 7.4: Rack Up of Biases and Uncertainties for Region 1 Uncertainty Bias Ak Validation (critical experiments) 0.0029 0.0050 Depletion reactivity uncertainty -

Burnup uncertainty - -

Fuel pin pitch - 0.0022 Fuel enrichment - -

Rack cell pitch - horizontal 0.0039 Rack cell pitch - vertical 0.0041 Rack cell ID - 0.0020 Rack cell wall thickness - 0.0008 Boraflex sheathing thickness - 0.0004 Eccentricity - 0.0001 Monte Carlo statistics - 0.0005 Total Rack Up 0.0029 0.0082 Total Bias and Uncertainty 0.0111 If5.0 wt% is assumed in the criticality model, the enrichment uncertainty is zero because the maximum nominal enrichment of the fuel is 4.95 wt%.

NET- 300067-01 Rev 0 50

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

These minimum loading requirements meet the IOCFR50 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 a I% margin in k. Therefore, for this analysis, it has been demonstrated that for all unborated cases k95/ is less than 0.99 and for the borated cases k95 /95 is less than 0.94 after adding biases and uncertainties.

8.1 Region 1 Table 8. 1 below summarizes the Region I results.

Calculations show that fresh 5.0 wt% U-235 fuel with 48 IFBA rods can be stored in Region 1 to meet the criticality requirements (k95/95 < 0.99 after accounting for all biases and uncertainties). The calculated k for this case is 0.9717. After adding Region I biases and uncertainties of 0.0111 from Table 7.4, the k95,95 is 0.9828. For fresh 5.0 wt% U-235 fuel with no IFBA rods, the assemblies can be stored on the periphery of Region I (see Section 8.10) or they can be stored anywhere if they have a control rod inserted.

For burned fuel, no credit is taken for IFBA or any insert in the guide tubes. Calculations show that 5.0 wt% U-235 fuel burned to 12 GWd/T meet the criticality requirements for Region 1. Note that for this particular depletion, a soluble boron concentration of 1300 ppm was used so that once burned fuel that was in a very short cycle can still be stored anywhere in Region I as long as the minimum burnup of 12 GWd/T is met. Burned fuel with less than 12 GWd/T burnup can be stored on the periphery of Region I or stored anywhere in Region I if the fuel assembly has a control rod inserted.

NET- 300067-01 Rev 0

Table 8.1: Calculated k's in Region 1 Bias &

Case k Uncertainty k951 5s Base Case (5.0 wt% U-235 fuel with 48 IFBA) 0.9717 0.0111 0.9828 Burned Fuel (5.0 wt% U-235 burned to 12 GWd/T at 0.9650 0.0167 0.9817 1300 ppm) I I out of 4 cells filled with water (no absorber panel in 0.9144 0.0111 0.9255 cell)

Empty cell filled with air 0.9245 0.0111 0.9356 Empty cell filled with aluminum 0.9255 0.0111 0.9366 Empty cell filled with concrete 0.9222 0.0111 0.9333 Empty cell filled with steel 0.9233 0.0111 0.9344 Empty cell filled with Zircaloy 0.9255 0.0111 0.9366 Missing panel, control rod in fuel assembly (base 0.9729 0.0111 0.9840 case fuel) 5.0 wt% U-235, no IFBA, control rod in fuel 0.7964 0.0111 0.8075 assemblies Base Case with alternate panel design 0.022 areal 0.9716 0.0111 <0.99 density Burned Fuel Case with alternate panel design 0.022 0.9645 0.0167 0.9812 areal density I I _ _ I 8.1.1 Missing Panel in Region 1 A case was also run in which there is an empty cell that has no absorber panel (one empty cell out of four cells in the 2x2 array). The lack of fuel in the cell more than offsets the reactivity due to the lack of an absorber panel in the cell. The cell does not have to be empty, only that there is no fuel. Concrete, steel, air, Zircaloy, aluminum and water were modeled and all of these materials meet the criticality requirements (see Table 8.1). Alternatively, a fuel assembly can be placed in the cell having the missing absorber panel as long as it contains a control rod (a control rod is worth more than an absorber panel, see Table 8. 1). Placing a fuel assembly into a cell that has a missing panel is considered an accident (see Section 9).

To allow for small changes in k due to minor variations in the final design of the alternate panel NET- 300067-01 Rev 0 52

8.1.2 Alternate Panel Design in Region 1 The fresh fuel, 48 IFBA case was run for the alternate panel design at an areal density of 0.022 g 1OB/cm 2 and the calculated k was 0.9716 compared to the base case of 0.9717. Even if there was a slight difference in tolerance effects due to the alternate panel design, there is 0.007 margin to the design goal of 0.99. Since the most reactive fuel was analyzed, the alternate design will result in a k 9 5/9 5 less than 0.99.

8.2 Region 2 The minimum burnup requirements (loading curve) for Region 2 are presented in Table 8.2.

Table 8.2: Minimum Burnup Requirements (GWd/T) in Region 2 Enrichment* Cooling Time (years)

(wt% U-235) 0t 1 2 5 10 15 25' 2.0 3.20 3.10 3.08 3.00 3.00 3.00 3.00 2.5 15.17 14.85 14.66 13.97 13.20 12.84 12.31 3.0 21.28 21.17 20.98 20.66 20.26 19.98 19.65 3.5 27.53 27.10 26.63 25.56 24.29 23.50 22.45 4.0 33.82 33.43 33.05 32.04 30.72 29.70 28.44 4.5 38.98 38.65 37.99 36.49 34.67 33.69 32.60 5.0 42.67 42.14 41.78 40.78 39.72 38.96 37.68 Table 8.2 gives the burmup requirement in GWd/T as a function of initial U-235 enrichment and cooling time. As discussed later, the burnup requirements are adjusted if the assembly had a hafnium insert, has fuel pins removed, or is placed on the periphery. The table can be linearly interpolated to find the required burnup at any enrichment/cooling time combination. No extrapolation is allowed, so fuel at enrichments less than 2.0 wt% U-235 must use the 2.0 wt% U-235 minimum burnup requirements and fuel cooled more than 25 years must use the 25 year cooling value. For illustrative purposes only, the For axially blanketed fuel, the enrichment to be used is the enrichment of the center section between the blanket material t No 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

ý Fuel cooled to more than 25 years must use the 25 year burnup requirements NET- 300067-01 Rev 0 53

loading curve is compared to the 2013 inventory in the Indian Point Unit 2 spent fuel pool as shown in Figure 8.1. This figure shows that only a few assemblies violate the loading curve and would have to be stored in Region I or contain a control rod if in Region 2.

60 .!

50 - - -------

-no cooling

~30 -

- 15 year M 20 Unit 2 inventory 0 ' r-2.0 2.5 3.0 3.5 4.0 4.5 5.0 U-235 Enrichment Figure 8.1: Loading Curve vs. Unit 2 Inventory (2013) 8.2.1 Curve Fit The burnup values in Table 8.2 can be fit to a curve having the following functional form where enr is the initial U-235 enrichment in wt% and CT is the cooling time in years:

(Al + A2enr + A3enr 2 + A4enr 3) exp[-(A5 + A6enr + A7enr 2 + A8enr 3) CT] +A9 + AlOenr

+A I I enr2 + A 12enr3 Two curves were generated - one curve for enrichments from 2.0 to 3.5 and another one for enrichments greater than 3.5 to 5.0. The resulting coefficients are shown in Table 8.3. The coefficients contain an adjustment to ensure that all bumups calculated by the equation are greater than the burnups NET- 300067-01 Rev 0 54

from the table. Using the curve fit results in a maximum penalty of 0.2 GWd/T for low enrichments and 0.4 GWd/T for high enrichments when compared to the tabulated values shown in Table 8.2.

Table 8.3: Coefficients for Curve Fit of Minimum Burnup Requirements Enrichment Enrichment Coefficient 2.0 - 3.5 wt% U-235 >3.5 - 5.0 wt% U-235 Al -226.183 448.616 A2 257.56 -327.34 A3 -95.573 79.9045 A4 11.6921 -6.43034 A5 0.957861 0 A6 -0.57489 0 A7 0.092684 0.013029 A8 0 -0.00203 A9 0.1 -481.261 AI0 -29.7478 343.449 All 22.7422 -78.3549 A12 -3.56276 6.07956 8.2.2 Confirmation Calculationsfor Region 2 To be sure that all burnup/cnrichmentlcooling time combinations given in the loading curve meet the criticality requirements, each loading curve burnup/enrichment/cooling time point was run in the 2x2 KENO model to verify that each point meets the criticality requirements. The calculated k's are shown in Table 8.4.

Table 8.4: Calculated k Values at Each Burnup Point Enrichment Cooling Time (years)

(wt% U-235) 0 1 2 5 10 15 25 2.0 0.9718 0.9710 0.9708 0.9707 0.9700 0.9696 0.9689 2.5 0.9699 0.9698 0.9695 0.9698 0.9704 0.9698 0.9697 3.0 0.9660 0.9657 0.9658 0.9654 0.9652 0.9650 0.9646 3.5 0.9672 0.9671 0.9672 0.9669 0.9672 0.9670 0.9676 4.0 0.9661 0.9659 0.9656 0.9654 0.9657 0.9663 0.9668 4.5 0.9642 0.9641 0.9660 0.9660 0.9656 0.9651 0.9650 5.0 0.9651 0.9646 0.9641 0.9642 0.9632 0.9635 0.9652 NET- 300067-01 Rev 0 55

The total uncertainty is the bias plus a statistical combination of all the uncertainties (see Section 7.2).

These total uncertainties are shown in Table 8.5.

Table 8.5: Total Bias and Uncertainty at Each Burnup Point Enrichment Cooling Time (years)

(wt% U-235) 0 1 2 5 10 15 25 2.0 0.0164 0.0164 0.0164 0.0164 0.0164 0.0164 0.0164 2.5 0.0169 0.0168 0.0168 0.0168 0.0167 0.0167 0.0166 3.0 0.0210 0.0210 0.0210 0.0209 0.0209 0.0209 0.0208 3.5 0.0219 0.0218 0.0217 0.0216 0.0214 0.0213 0.0211 4.0 0.0227 0.0227 0.0226 0.0225 0,0222 0.0221 0.0219 4.5 0.0235 0.0234 0.0233 0.0231 0.0228 0.0226 0.0225 5.0 0.0240 0.0239 0.0238 0.0237 0.0235 0.0234 0.0232 After adding the total uncertainty to the calculated k's, all points are less than 0.99 as shown in Table 8.6. The values in Table 8.6 do not always match the sum of Tables 8.4 and 8.5 due to round off, since each table was developed using more significant digits before rounding for the table.

Table 8.6: k95 /95 at Each Burnup Point For Region 2 Enrichment Cooling Time (years)

(wt% U-235) 0 1 2 5 10 15 25 2.0 0.9882 0.9873 0.9872 0.9870 0.9863 0.9859 0.9852 2.5 0.9867 0.9866 0.9863 0.9866 0.9871 0.9864 0.9863 3.0 0.9871 0.9867 0.9868 0.9864 0.9861 0.9859 0.9854 3.5 0.9890 0.9889 0.9889 0.9885 0.9886 0.9883 0.9887 4.0 0.9889 0.9885 0.9882 0.9879 0.9879 0.9884 0.9887 4.5 0.9876 0.9875 0.9893 0.9890 0.9884 0.9877 0.9875 5.0 0.9891 0.9885 0.9880 0.9879 0.9868 0.9869 0.9884 8.2.3 Use of Control Rods in Region 2 A calculation was performed to show that a control rod inserted into a fresh 5.0 wt% U-235 assembly in Region 2 meets the criticality requirements. The calculated k is 0.9292 (a 2x2 array of fresh assemblies with each assembly containing a control rod). The analysis was performed with fresh assemblies NET- 300067-01 Rev 0 56

containing no IFBA rods. Since the k of the 2x2 calculation is less than the base case for Region 2 (see Table 8.7), this shows that a fresh assembly with a control rod can be stored anywhere in Region 2.

Further, any assembly having any burnup depleted under any condition can be stored anywhere in Region 2 provided it contains a control rod.

8.2.4 Missing Panel in Region 2 As in Region 1, the lack of fuel in a cell more than offsets the reactivity due to the lack of an absorber panel in a cell. A case was run for a 5.0 wt% assembly burned to 42.67 GWd/T with one water hole and no panel in the cell. The cell does not have to be empty, only that there is no fuel assembly. The cell can be filled with a block of concrete, steel, air, Zircaloy, water, or aluminum having the outer dimensions of a fuel assembly. All of these materials meet the criticality requirements (see Table 8.7). Other burnup/enrichment configurations were not run because there is a large margin to the requirements.

Alternatively, a fuel assembly meeting the loading curve can be placed in the cell having the missing panel as long as the fuel assembly contains a control rod. The negative reactivity worth of a control rod is more than that of an absorber panel (see Table 8.7).

Table 8.7: Additional Sensitivity Calculations for Region 2 Bias &

Case k Uncertainty kgs.s Base Case* (5.0 wt% U-235 fuel at 42.67 GWd/T, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) 0.9639 0.0240 0.9879 Empty cell filled with water (no absorber panel in cell) 0.8744 0.0240 0.8984 Empty cell with air (no absorber panel in cell) 0.9195 0.0240 0.9435 Empty cell with aluminum (no absorber panel in cell) 0.9310 0.0240 0.9550 Empty cell with concrete (no absorber panel in cell) 0.9163 0.0240 0.9403 Empty cell with steel (no absorber panel in cell) 0.9205 0.0240 0.9445 Empty cell with zirc (no absorber panel in cell) 0.9333 0.0240 0.9573 Missing panel, control rod in fuel assembly (base case fuel) 0.9567 0.0240 0.9807 No missing panels, fresh 5.0 wt% U-235 no IFBA, control rod in fuel 0.9292 0.0102 0.9394 assembly Alternate panel design, 0.020 areal density (base case fuel) 0.9630 0.0240 0.9870 Low burned fuel (2.5 wt%, 12.31 GWd/T, 25 year, no missing panels) 0.9696 0.0166 0.9862 Alternate panel design for the above low-burned fuel case 0.9683 0.0166 0.9849 The base case identified here was for a depletion in which the WABA was removed after 35 GWd/T.

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8.2.5 Alternate AbsorberPanel Design in Region 2 The base case (5.0 wt% at 42.67 GWd/T) was run for the alternate absorber design at an areal density of 0.020 g BIO/cm 2and results in a k of 0.9630 compared to the base case k of 0.9639. A low burnup case (2.5 wt% at 12.31 GWd/T) was also run with the alternate design. The k was 0.9683 compared to the primary design k of 0.9696. This demonstrates that the loading curves are valid for the alternate design.

8.2.6 Expanded Cooling Time Calculations It is generally accepted that the most reactive time after discharge is immediately after core offload (72 or 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) with no credit for Xe-135 or Np-239 (all of the Np-239 is immediately converted to Pu-239 and all of the Xe-135 is immediately converted to Cs-135). To test this assumption, additional cooling time calculations between 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and I year were performed (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, 8 days, 16 days, 30 days, 60 days, 0.3 year, 0.6 year, and 1 year)..

As shown in Figure 8.2, the k generally decreases with time and the k is largest at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (the k is the same). The statistical uncertainty in these calculations is 0.0002 (2 sigma) so although there appears to be a non-decreasing behavior between 16 and 60 days, it is not significant.

0.9655 0.9650 - -[---- -

0.9645 0.9640 k 0.9635 - -

0.9625-----.----- -

0.9620 .

0 100 200 300 400 Days after Discharge Figure 8.2: k as a Function of Cooling Time NET- 300067-01 Rev 0 58

8.3 Borated Conditions The most limiting acceptance criteria is for the unborated condition so the loading criteria have been set using models that did not contain soluble boron. In order to confirm that the k95/95 is less than 0.95 at a boron content that will be maintained even under a boron dilution accident (Section 9.5) a limited number of additional calculations were performed. The soluble boron concentration of 700 ppm was used for these calculations since this concentration can be easily supported by the boron dilution analysis and it yields significant margin in k. For Region 1, the calculated k at 700 ppm with water at 100 TC is 0.8790 (see Table 8.9). With bias and uncertainty this becomes k95/95 = 0.8900. Due to the difference between 0.89 and the target value of 0.94, no further calculations are warranted. For Region 2, the 5.0 wt% U-235 loading curve points at 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 25 year cooling were run with 700 ppm boron in the pool water (42.67 GWd/T at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 37.68 GWd/T at 25 years). The calculated k at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> was 0.8899, while the k at 25 years was 0.8906. With the bias and uncertainty applied, the k95/95 to be compared to the regulatory limit of 0.95 becomes 0.9139 and 0.9138 for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 25 year cooling times respectively.

The bias and uncertainty used here for both Region I and 2 was 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 would be expected that they would be smaller, since the primary uncertainty is the burnup uncertainty and the reactivity worth of burnup decreases with boron.

Furthermore, ignoring the grids is still conservative at 700 ppm.

8.4 Depletion Effect of Hafnium Flux Suppression Inserts To determine the reactivity effect of having a hafnium rod inserted, a special depletion was performed. For the first 8 GWd/T, the assembly is depleted with a 20 rodlet hafnium insert (full length).

The hafnium insert is then replaced by a fresh 20 rodlet WABA plus 148 IFBA rods and never removed.

Since the hafnium inserts harden the spectrum more than WABAs (Ref. 7, Table 4.7.7), depleting with NET- 300067-01 Rev 0 59

the hafnium first is conservative for the WABA depletion and no estimate of the burnup for changing from WABA to hafnium depletion is required. The burnup requirements for this case were then calculated using the same procedures as were used for the loading curve. Enrichments were 4.0, 4.5, and 5.0 and cooling times were 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 25 years. Table 8.8 shows the burnup required for this special depletion to meet the 0.99 k 95/9 5 target value after accounting for all biases and uncertainties.

Table 8.8: Hafnium Depletion Results Normal Burnup Hafnium Depletion Increased Burnup Enrichment Cooling Normal Burnup Requiremet Required For Hafnium (wt% U-235) Time Requirement Burnup Requirement Depletion Case (GWd/T) (GWd/T) (GWd/T) 4.0 72 hr 33.82 34.89 1.07 4.5 72 hr 38.98 39.69 0.71 5.0 72 hr 42.67 43.98 1.31 4.0 25 yr 28.44 29.13 0.69 4.5 25 yr 32.60 33.20 0.60 5.0 25 yr 37.18 37.82 0.64 For simplicity and to provide margin, a 2 GWd/T increase in the loading curve is required for any assembly having any burnup with a hafnium insert. Please note that if an assembly exceeds the loading curve requirements prior to loading a hafnium insert, no burnup adder is required since additional burnup (where IFBAs are ignored) always reduces reactivity regardless of spectrum.

8.5 Axial Reflector All of the analyses were performed with a water axial reflector (50 cm at top and bottom). Unburned fuel reactivity is dominated by the center and so the results are not sensitive to the axial reflector modeling. Burned fuel is top peaked and so is more sensitive to axial reflector modeling. To test the appropriateness of using 50 cm of water as the axial reflector for burned fuel, several other configurations were analyzed. The bottom reflector has an insignificant effect due to the axial burnup distribution (reactivity is driven by the low burnup in the top nodes), so cases were run to test only the top reflector.

Starting with the 5 wt% U-235, 42.67 GWd/T burnup, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> case, which has a calculated k of 0.9639, NET- 300067-01 Rev 0 60

some modifications to the top reflector were made. With a 50/50 mix of steel and water at the top, the calculated k was 0.9613, so the water reflector k of 0.9639 is more limiting. With 100% steel at the top, k was 0.9729, which demonstrates that pure steel is more limiting. The fuel pin above the active fuel is actually composed of a spring inside a plenum and the rods extend above the active fuel at least 6 inches.

Using the fuel clad OD and the pin pitch, the fuel rod is 44% and water is 56% of the total area. So to simulate the plenum and spring, a mixture of 56% water, 22% steel and 22% void (fission gas) is used for the 6 inches directly above the active fuel. Then 100% steel is used above that. The k for this case was 0.9613, so the current analysis is conservative compared to the realistic case.

8.6 Volatile Fission Gases There may be some leakage of fission gases from the active fuel. To quantify this effect, the 5.0 wt%,

72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> case (42.67 GWd/T) was run in which all of the krypton, xenon, and iodine isotope concentrations were set to zero. The k for this case was 0.9699 compared to the unperturbed k of 0.9639 (see Table 8.7), a difference of 0.0060. If 10% of the fission gases were released [29], the Ak would be 0.0060 x 0.10 = 0.0006. This small amount could be included as a bias in the final calculations.

Alternatively, the top axial reflector could be modeled more realistically, which gives a Ak benefit of 0.0027, which easily covers the 10% fission gas release. Therefore, no bias for volatile fission gases is needed.

8.7 TemperatureEffects Table 8.9 summarizes the base case calculations at 4 different temperatures (4, 20, 70, and 100 'C).

The results demonstrate that the reactivity is largest at 4 TC in both regions under unborated conditions.

The reactivity increases slightly under borated conditions in Region 1 with increasing temperature but there is excess margin for the borated condition.

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Table 8.9: Calculated k as a Function of Temperature Temperature Density Region 1 Region 1 Region 2 Region 2 (00) (g/cc) (0 ppm) (700 ppm) (0 ppm) (700 ppm) 4 1.0000 0.9717 0.8777 0.9639 0.8928 20 0.9982 0.9714 0.8778 0.9634 0.8924 70 0.9778 0.9710 0.8789 0.9553 0.8883 100 0.9584 0.9696 0.8790 0.9492 0.8848 8.8 Fuel Geometry Changes during Burnup There are geometry changes during burnup, such as pellet densification, crud buildup, clad creep down and fuel rod growth. Recent work by the industry has shown that these effects are insignificant [22,271.

8.9 Depletion of Fuel < 3.5 wt% with Modern Depletion Assumptions Historical low enriched fuel used Pyrex burnable absorbers and was burned with the control rods at the bite position. In order to accommodate this, the loading curves were generated using separate depletion assumptions for enrichments below and above 3.5 wt% U-235. However, although not currently anticipated, there may be a need for loading new fuel enriched below 3.5 wt% U-235. A special depletion was performed to cover this possibility. A depletion at 3.5 wt% was performed using the depletion assumptions selected for modern fuel (see Table 10.5). The k for 3.5 wt%/o at 27.53 GWd/T, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooling was 0.9650. The k for 3.5 wt% at 27.53 GWd/T, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> using the older depletion assumptions (Pyrex burnable absorber, lower temperatures, rodded top node, etc.) was 0.9665. So the loading curve for 3.5 wt% or less is valid for either set of depletion assumptions. The reason for this is that although the older depletion assumption is less limiting for fuel below the top node, the older depletion assumes that the top node is rodded its entire life. Modem depletion assumes the top node can be rodded for only a short time (i.e., the bite position is not allowed). The rodded top node balances the less reactive lower nodes so that the net difference between the two depletion assumptions is very small.

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8.10 Reduced PeripheryRequirements & Region 1/Region 2 Interface Since the neutron leakage at the edge of the pool reduces the reactivity worth of the assemblies on the periphery of the pool, it is possible to relax the requirements for these assemblies. In Region 1, the assemblies on the periphery do not need any IFBA rods or buMrup, In Region 2, the assemblies on the periphery, which are 4.0 wt% U-235 or greater, are allowed to have 8 GWd/T less burnup than the Region 2 loading curve. In order to prove that these reductions are acceptable, a full pool model was created. This section describes the full pool model and the results of the periphery assembly analysis.

This section also confirms the interface between Region I and Region 2 does not need any special requirements.

8.10.1 Full Pool Model The full pool model is used to:

a) Verify reduced requirements for the periphery assemblies, b) Verify that the Region 1/Region 2 interface has an insignificant effect on k, and c) Show an acceptable soluble boron concentration for a single dropped assembly or misplaced assembly (see Section 9).

The full pool model was created by taking the 2x2 model for Region 2 and the IxI model for Region I described in Section 6 and using them as units that were reproduced in arrays. The model has 4 large arrays (see Figure 8.4 for module identification):

1. Region I module A (1 0X8),

2. Region I modules B and C (combined as 21X9),

3. Region 2 modules D, E-1, F-I, F-2, G-1, and G-2 (combined as 24X32), and

4. Region 2 modules E-2, E-3, and H.

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Modules E-2, E-3, and H are 11 cells across (north to south) but since the modeling is using 2x2 units, these modules were conservatively modeled as 12 cells across. This conservatively makes the pool larger but the impact is negligible. In the east to west direction these modules have an even number of cells so no modification was necessary. However, module 11 has a cutout for the fuel elevator and failed fuel containers. This cutout was conservatively ignored. The pool has some small extra space between modules. This space is conservatively removed in the full pool model.

The pool dimensions come from Holtec Drawing No. 397 [21 ]. The drawing shows the smallest separation between any module and the wall is 1.25 inches (Region I left wall). This minimum separation is assumed on all sides. The pool has a 0.25 inch stainless steel liner and a concrete wall outside the liner.

Figure 8.3 shows the full pool model. The pink is concrete. The model starts on the left with 19.69 inches (50 cm) of concrete. This is followed by a 0.25 inch stainless steel liner. Next is a 1.25 inch water gap. After this the Region I arrays are added. This then sets the left boundary for the smaller Region 2 array. Then there is a 1.25 inch water gap on the right followed by the 0.25 inch liner and then 19.69 inches (50 cm) of concrete. The large Region 2 array is set to be against the left side of the smaller Region 2 array. This arrangement means the left side of the large array of Region 2 is not directly under the larger Region 1 array. This offset is caused by the separation between Region 2 modules that is ignored in this model. A block of concrete is added to the left of the large Region 2 array. This block does not have the stainless steel liner but is separated from Region 2 by the same 1.25 inches of water.

When changing the reflector for studies, the same model setup is maintained, so all the 0.25 inch liners change or all the 1.25 inch water gaps change at once. Figure 8.4 is the pool taken from Holtec Drawing No. 397 [21]. As can be seen when comparing Figures 8.3 and 8.4, the full pool model has extra water below Region I module A and this is consistent with the real pool. The full pool model has extra water above Region 2 module H and this is consistent with the pool drawing.

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Figure 8.3: Full Pool Model NET- 300067-01 Rev 0 65

Figure 8.4: Indian Point Unit 2 Spent Fuel Pool Taken From Holtec Drawing #397 [21]

Figure 8.5 shows an enlargement of the top left comer of the pool. Several points should be noticed.

First, the model has steel extensions on the outside of the module due to replicating the cell. The real modules do not have these. Second, on the wall side of the module, the model has the sheath for the old absorber panels. These are not on the real module outside walls. Thus, there is some excess steel on the outside. A calculation was performed where a 0.1 inch steel plate was placed on the edge of the repeating Region I cells. K.rr increased only 0.0025 so there is confidence that the extra steel on the outside of the model is negligible or slightly conservative. Finally, note that the absorber panels (light blue) are in the bottom right. This means the most reactive periphery is at the top and left side for Region 1.

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LEUMD mwwýt 3 Figure 8.5: Enlargement of the Top Left Corner of the Pool Model Figure 8.6 shows an enlargement of the bottom lcft comner of the full pool model. Note that the resultant cell has the absorber sheath and some of a cell wall on the outside. The actual rack has a plate 0.075 inches thick. The total of the sheath and the partial cell wall is 0.075 inches, so by chance the correct amount of steel is on the outside of the resultant cells. The standard cells in the full pooi model have an absorber sheath on the outside wall that does not exist in the real rack. The calculation with the 0.1 inch steel plate placed on the outside of the model described in the previous paragraph provides NET- 300067-01 Rev 0 67

confidence that this extra sheath in the model is negligible or slightly conservative. Note that the absorber panels (green in this case) for Region 2 are in the top left making the bottom and right sides more sensitive to the periphery.

- Is,, a.

Figure 8.6: Enlargement of the Bottom Left Corner of the Pool Model Figure 8.7 shows an enlargement of the left side of the bottom of the Region I/Region 2 interface.

Note that the absorber panels for the two regions (light blue for Region I and green for Region 2) create an effective flux trap. The distance between the two regions in the full pool model is 0.9 inches. The real separation is 1.375 inches. The smaller separation is conservative as shown by the rack pitch tolerance calculations in Section 7. 1. The real pool has absorber sheaths on the outside of both the Region I and 2 rack modules at the interface. '[he full pool model has the Region I modules consistent with reality but has less steel on the outside of Region 2. Section 7.1 concluded that more water (less steel) between cells NET- 300067-01 Rev 0 68

in Region I is conservative. Therefore, the full pool model with less steel between the racks is acceptable.

Figure 8.7: Enlargement of the Left Side of the Bottom of the Region l/Region 2 Interface Calculations were performed to confirm that the k of the finite model of each region is similar, but lower than the k from the infinite models described in Section 6. Since the reduced periphery burnup criteria (lowered by 8 GWd/T) only applies for enrichments equal to or greater than 4 wt%, the Region 2 analysis of the full pool model is performed at 28.44 GWd/T at 4 wt% U-235 and 25 yr cooling (the lowest burnup) and 42.67 GWdiT at 5 wt% and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooling (the highest burnup). The full pool NET- 300067-01 Rev 0 69

model was run without fuel assemblies in Region 1 for the Region 2 only models. For the Region 1 only models, no fuel was put in the fuel rods of Region 2 (replaced with void). If fuel was in both regions, the k would correspond to the region with the higher k. Table 8.10 shows the k's of the finite and infinite models.

Table 8.10: Infinite (Section 6) Versus Finite (Full Pool Model)

Enrichment Burnup Region (wt% U23_) (GWd/T) Cooling k..f a knitc a 1 5" 0 0 0.97182 0.00005 0.9666 0.0001 2 4 28.44 25 yr 0.96677 0.00011 0.9650 0.0001 2 F 5 42.67 72 hr 0.96395 0.00011 0.9626 0.0001 "The Region I fuel contains 48 1FBA rods As can be seen in Table 8.10, the infinite model is slightly conservative for the pool. The finite model of Region 2 is from 0.13% to 0.17% lower in k. The finite model for Region I is less by more than 0.5% in k. This larger decrease is expected since Region I is much smaller.

8.10.2 Results of Reduced Peripheryand Region 1/Region 2 Interface Analysis This section presents a set of calculations to confirm that the relaxed periphery requirements are acceptable. Figure 8.8 shows the location of the cells where the peripheral relaxation is permitted. Note that due to model simplifications of Module H, the cells near the failed fuel containers are not considered peripheral cells.

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1 3 5 7 noltl 1416 18 20 223 25 27 29 31 4OPM'I 4 gDP Region 1 D.

C DJ DG DE CP CP OlM CW 04 ICH CF 0RegniO 2 cD BN BL BK BJ BG BE 075

• 7 4667 73 AK AF AD AS N PERIPHERAL CELL 40 42 44-46 46 8159 55 67 59 61 636 Figure 8.8: Location of the Peripheral Cells with Reduced Requirements In a large pool model, the calculated k will reflect the k of the most reactive region, yielding little information on less reactive portions of the pool. Since the calculated k for Region 2 is lower than Region 1, the periphery calculations for Region 2 have to be performed without fuel in Region 1. Even though the k is lower in Region 2, Region 2 has less margin, since the uncertainty is higher.

All of the analysis with the full pool model uses at least 3000 generations with 12000 neutrons per generation. Even though this is a very large number of neutrons, care must be taken to assure that the effect of concern is seen, because the model is so large. To do this, the starting source is specified in the region of concern. Sometimes it is not clear where the highest reactivity lies, so several starting source distributions are used in successive runs.

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Table 8.11 presents the results of calculations that were performed to confirm the 8 GWd/T lower burnup requirement for Region 2 periphery assemblies.

Table 8.11: Region 2 Periphery Tests (No Fuel in Region 1)

Bias +

Enrichment* Burnup** Starting Calculated Uncertainty (wt% U135) (GWd/T) Source kj C3 k95195 "-

4 28.44/20.44 Comer 0.9652 0.0001 0.0219 0.9871 4 28.44/20.44 Right 0.9650 0.0001 0.0219 0.9869 4 28.44/20.44 Left 0.9654 0.0001 0.0219 0.9873 5 42.67/34.67 Comer 0.9628 0.0001 0.0240 0.9868 5 42.67/34.67 Right 0.9624 0.0001 0.0240 0.9864 5 42.67/34.67 Left 0.9624 0.0001 0.0240 0.9864 4 wt% cases are cooled 25 years. 5 wt%/o cases are cooled 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

"*The first bumup is the burnup in non-periphery cells. The second burnup is the burnup in periphery cells.

• n*Starting source location: Corner: the cask area corner. Right: right side. Left: Left side.

Bias and uncertainty for the highest burnup taken from Table 8.5

• .... To be compared to licensing limit, less than 1.0. 1%margin to limit is used.

Several observations can be made from the data in Tables 8.10 and 8.11. The periphery cases in

'Fable 8.11 have k's that are nearly equal to the cases without the reduced burnup on the periphery. This means that the local k on the periphery is about equal to the k in the center of the region and therefore, the reduced burnup requirements on the periphery are confirmed. Finally, after adding the bias and uncertainty, the target limiting k95/95 (0.99) is met.

Table 8.12 shows the results of calculations performed with fuel in both regions. These calculations demonstrate that the reactivity is driven by Region 1, since the Region I k is higher. These full pool models use the most reactive Region 2 fuel, which is the 4 wt% fuel with 25 years cooling and 28.44/20.44 GWd/T burnups. Since a constant 8 GWd/T burnup relaxation for the periphery is utilized rather than a fraction of the burnup requirements, the fuel with the lowest non-periphery burnup requirements (i.e., 4 wt%, 25 years cooled) is the most reactive peripheral fuel.

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Table 8.12: Region 1 Periphery Tests (4 wt% 28.44/20.44 GWd/T Fuel in Region 2)

Startinj Calculated Bias +

Source kfr Ca Uncertainty k95195'*

Interface 0.9759 0.0001 0.0111 0.9870 Left 0.9781 0.0001 0.0111 0.9892

Starting source location. Interface: bottom of Region I at the Region 1/2 interface. Left: Left side of Region 1.

'To be compared to licensing limit, less than 1.0. 1%margin to limit is used.

Table 8.12 demonstrates that taking out the IFBAs in the periphery assemblies increases k slightly.

The increased k (0.9781) is slightly larger than the infinite model (0.9718 from Table 8.10). However, there was margin in Region 1, so the final k9 5/95 is still less than the target k95/95 of 0.99. In order to find the most reactive location in a large model, the starting source location can be varied. The two calculations in Table 8.12 only differed in the starting source location. When the source was started at the Region I - 2 interface, the higher reactivity corner was not found.

Table 8.12 also demonstrates that the Region I/Region 2 interface does not have the limiting k. The calculated k, using the starting source at the interface, is lower than the k using the source placed at the left edge of the model, where there is fuel with no IFBA. Clearly, the fully converged calculated k should not depend on the location of the starting source, but placing the starting source at areas of concern provides confirmation that the k of the location of concern was truly evaluated. The k calculation with the starting source at the interface has a lower k since there were generations included in the final calculation of k before the more reactive peripheral assemblies were found. However, there were plenty of neutrons at the interface to assure that k at the interface is within the limits. This result is expected since the absorber panels face each other at the interface creating a flux trap.

Table 8.13 presents the results of calculations performed to test the sensitivity to the pool reflectors.

For some of the runs, no fuel was in the Region I cells, so the reflector sensitivities on Region 2 can be seen. The first observation is that the pool reflectors have no effect on the Region 2 results. This is because leakage has a bigger effect on k than the 8 GWd/T reduction in the burnup requirement. For Region 2, the k is controlled by the center of the region. For Region 1, results show that concrete is a NET- 300067-01 Rev 0 73

better reflector than water (the concrete used is a conservative mixture created by Oak Ridge - named as orconcrete). As the water gap is reduced, k increases. The results are not sensitive to the liner thickness.

Table 8.13: Reflector Tests (4 wt% 28.44/20.44 GWd/T Fuel in Region 2)

Water Gap to Steel Liner Region Starting Calculated Bias +

Cell Wall (cm) Thick. (cm) I Fuel Source* kerr I Uncertainty k9519 .,*

5 5.6 0.635 Yes RI Left 0.9765 0.0001 0.0111 0.9876 12.4 0.635 Yes RI Left 0.9753 0.0001 0.0111 0.9864 3.175 0.635 Yes RI Left 0.9781 0.0001 0.0111 0.9892 5.6 5.0 Yes RI Left 0.9772 0.0001 0.0111 0.9883 5.6 0.001 Yes RI Left 0.9773 0.0001 0.0111 0.9884 3.175 0.635 No Right 0.9650 0.0001 0.0219 0.9869 10.0 0.635 No Right 0.9650 0.0001 0.0219 0.9869 1.0 0.635 No Right 0.9647 0.0001 0.0219 0.9866 3.175 5.0 No Right 0.9649 0.0001 0.0219 0.9868 3.175 0.001 No Right 0.9650 0.0001 0.0219 0.9869 Starting source location. RI Left: Left side of Region 1. Right: Right side of Region 2.

- To be compared to licensing limit, less than 1.0. 1% margin to limit is used.

An additional calculation was performed on the full pool model, where a small layer of steel was added at the outside of the fuel rack. The result was that k increased, which confirms that the extra steel (Boraflex sheathing, and Region 1 cell separator beams) that is at the edge of the full pool model is conservative.

Since all the k 95/9 5 's of the analyses are less than the target limiting k (0.99), the reduced requirements for the periphery are confirmed.

8.11 FailedFuel Containers The southeast comer (please note that on the drawings in this report North points left) of the spent fuel pool contains two 16" circular pipes and are labeled "failed fuel containers." 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 [30]. The neutron sources also have a very small amount of,fissile material. The ANSI/ANS 8.1 standard [33] states that 700 grams of U-235 in any configuration is always subcritical.

However, the failed fuel containers are not fully decoupled from the Module H of Region 2. In order to NET- 300067-01 Rev 0 74

set a conservative limit on the fuel that can be in the failed fuel containers, the 2x2 Region 2 model was modified by (1) removing the absorber panel from the lower right cell and the upper right cell and (2) in each of these cells, placing a regular array of 36 fresh 5.0 wt% fuel pins (6x6). In the two left cells, illustrated in Figure 8.9, are burned 5.0 wt% fuel assemblies at the loading curve burnup. The k for this configuration was 0.9374, which is less reactive than the loading curve. Based on this result, 36 fuel pins could be loaded into each failed fuel container with no criticality concern.

Figure 8.9: Model for Failed Fuel Container Analysis NET- 300067-01 Rev 0 75

8.12 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 wt% fuel rods in Region 2 (see Figure 8.10). The k for this configuration was 0.9195, which is well below the loading curve k. Therefore, the fuel rod storage basket can be stored anywhere in Region I or Region 2.

Figure 8.10: Model for the Fuel Rod Storagc Basket NET- 300067-01 Rev 0 76

8.13 Assemblies with Missing Fuel Rods Usually, when a fuel 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 done, there is no criticality concern since the reconstituted assembly would be less reactive than the original assembly.

However, if a failed fuel rod is removed but not replaced with a stainless steel rod, the reactivity increases because there is more water available. An analysis was done in which one or more fuel rods are removed from an assembly to estimate the effect on k. It was determined that k gradually increases as more fuel rods are removed up to and including 36 missing fuel rods. If more than 36 fuel rods are removed, k begins to decrease (see Figure 8.11). The delta k with 36 missing fuel rods (see Figure 8.12) was 0.0184 (a 2x2 array with all 4 assemblies having 36 missing rods) compared to the base case k of 0.9639 with no missing rods. For simplicity, a burnup adder 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 pool (assembly ID of T67) 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 so the burnup requirement would be 42.67 GWd/T with no cooling time credit. The actual burnup of T67 is 49.81 GWd/T so it exceeds the requirement by more than the 4 GWd/T adder 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 in Region 2. This adder 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 adder is used for both assemblies).

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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 versus Missing Fuel Rods Figure 8.12: Model for Assemblies with 36 Missing Fuel Rods NET- 300067-01 Rev 0 78

9 Accident Conditions The following accidents were analyzed:

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

• A fresh assembly dropped into an empty cell that is missing an absorber panel,

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

• A multiple misload event.

Three more accident conditions were considered, but no analysis was necessary. An assembly dropped horizontally on top of other assemblies was 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 inches 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 that is missing an absorber panel. 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 (See Tables 10.1 and 10.2) are bounded by a fresh assembly dropped into an empty cell that is missing an absorber panel. The third accident condition is an under-temperature accident (water freezing). This was not considered because the analysis 'already assumes water at the maximum possible density (near freezing). The last subsection of this Section describes why a seismic event does not cause a criticality concern.

9.1 Misplaced Assembly For the misplaced fuel assembly accident, it is assumed that a fresh 5.0 wt% U-235 fuel assembly is placed in the pool next to the rack in the most reactive location. There are two locations which could be limiting for a misplaced assembly. One is the comer of the cask loading area. This location is an inside comer so it would interact with the rack on two sides of the misplaced assembly. This corner also allows lower burned peripheral assemblies and the absorber inserts are on the opposite side of the rack cells from NET- 300067-01 Rev 0 79

the misplaced assembly. Figure 9.1 shows the misplaced assembly. The most reactive periphery condition is at 4 wt% (lowest enrichment where a reduced peripheral burnup is allowed) and 25 years cooling. The most reactive misplaced fuel is assumed to be 5 wt% with no IFBA and no burnup. Table 9.1 presents the results of this misplaced assembly accident analysis. As can be seen from Table 9.1, 1200 ppm of soluble boron reduces the final k 9 5 /95 below the target of 0.94. The analysis used a starting source placed near the misplaced assembly.

The second limiting misplaced assembly assumes a fresh fuel assembly is in the fuel elevator and another fresh assembly is wedged in between the rack and the assembly in the fuel elevator. For conservative analysis of this case it is assumed that the misplaced assembly and the assembly in the fuel elevator are both fresh 5 wt% U-235 enriched assemblies with no IFBA rods. Further it is assumed that there is no separation between the two fresh fuel assemblies and no separation from the rack. This misplaced assembly is also shown on Figure 9. 1. Figure 9.1 is the model used for the fuel elevator misplaced assembly. The starting source for the fuel elevator case was concentrated around the fuel elevator. Table 9.1 also contains the results for this misplaced assembly analysis. It is more limiting but the 1200 ppm soluble boron still reduces the final k9 5195 below the target of 0.94.

The minimum soluble boron allowed by the Technical Specifications is much higher than 1200 ppms (currently 2000 ppm), so this accident condition meets the regulatory requirements.

Table 9.1: Misplaced Fuel Assembly Analysis isplaced Assembiy is 5 wt% U-235 with no Burnu or IFBA)

Reg. 2 Enrichment* Reg 2 Burnup** Soluble Calculated Bias +

(wt% U235) (GWd/T) Boron (ppm) kef v Uncertainty k9,s/gj"*

4 28.44/20.44 0 1.04892 0.0002 0.0219 1.0708 4 28.44/20.44 1200 0.89369 0.0001 0.0219 0.9156 4 28.44/20.44 1200 0.90490 0.0001 0.0219 0.9268 4 wt% cases are cooled 25 years.

"The first burnup is the bumup in non-peripheral cells. The second bumup is the burnup in periphery cells.

...To be compared to licensing limit, less than 0.95. 1%margin to limit is used.

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Figure 9.1: Full Pool Model with Misplaced Assembly 9.2 DroppedAssembly For the dropped assembly accident, it is assumed that a fresh 5 wt% U-235 assembly is dropped into a cell where the absorber panel has been removed (e.g.; for inspection). This accident also covers a situation where the dropped assembly severely damages the absorber panel at the same time. It is then further assumed that when the assembly dropped, the grids failed, which allows for full expansion of the pin pitch (this assumption of expansion of the pin pitch goes beyond traditional assumptions but due to NET- 300067-01 Rev 0 81

the available margin, this assumption can be made and 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. The location of the most limiting position for the dropped assembly is not obvious, so several possible locations were analyzed in the same model.

Figure 9.2 shows the full pool model for the dropped assembly analysis with the 6 dropped assemblies that were analyzed. The different potentially limiting dropped assemblies are tested by changing the starting source. The Region 2 fuel is the 4 wt% U-235 fuel and 28.44/20.44 GWd/T burnup at 25 years cooling (the condition with the least margin to the criticality limits, see Table 8.11). Region I contains the most reactive fuel allowed; fresh 5 wt% fuel with 48 IFBA rods, except the periphery Region 1 assemblies have no IFBA rods.

Table 9.2 presents the results of the analysis. The most limiting position for the dropped assembly is the bottom left corner of Region 2. This is most limiting since it puts together 4 low burned assemblies without an absorber panel between two of these assemblies. The right side of Region 2 always has an absorber panel separating the low burned periphery and the interior assemblies. The bottom row of Table 9.2 is highlighted, since it represents a dropped assembly, where the assembly does not come apart and maintains its normal pin pitch. This scenario was considered since at high soluble boron concentration, the larger pitch may not be the most limiting. As seen in Table 9.2, the larger pin pitch is more limiting.

As can be seen in Table 9.2, 1200 ppm (rather than the initial guess of 1500 ppm) is sufficient to meet the limiting criteria of 0.95. The minimum soluble boron allowed by the Technical Specifications (currently 2000 ppm) is much higher than 1200 ppm, so this accident condition also meets the regulatory requirements.

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Table 9.2: Dropped Fuel Assembly Cases Starting Soluble Boron Calculated Bias +

Source (ppm) ka a Uncertainty kgsm*"

Right 1500 0.8773 0.0001 0.0219 0.8992 Comer 1500 0.8737 0.0001 0.0219 0.8956 Center 1500 0.8735 0.0002 0.0219 0.8954 Bottom 1500 0.8817 0.0001 0.0219 0.9036 Comer Bottom 1200 0.9104 0.0001 0.0219 0.9323 Comer Bottom 1200 0.8959'** 0.0002 0.0219 0.9178 Corner Right: right side of Region 2. Comer: cask loading area comer. Center: center of Region 2. Bottom Comer: bottom left comer of Region 2.

• • To be compared to licensing limit, less than 0.95. 1%margin to limit is used.

• '*Same as the case above except no expanded pin pitch Figure 9.2: Full Pool Model with 6 Dropped Assemblies NET- 300067-01 Rev 0 83

9.3 Over Temperature As shown in Section 8.7, raising the temperature lowers k, even up to 100 'C. Under borated conditions, some of the calculated k's go up with temperature, but not enough to overcome the negative reactivity of the boron. The over temperature accident condition meets the regulatory requirements.

9.4 Multiple Misloads A case was run assuming that the entire Region 2 rack was filled with fresh 5.0 wt% fuel having 64 IFBA rods at a 1.25X IFBA loading. With 2000 ppm boron in the pool water (the current Technical Specification requirement at Indian Point), the k was 0.92 10. The total bias and uncertainty of fresh fuel in Region 2 is 0.0095, so the k 9 5/ 95 is 0.9305 which is less than the borated goal of 0.94. Less than 64 IFBA rod fuel has been ordered in the past for Indian Point, but plans for future assemblies use 64 or more IFBA rods. Many years ago Unit 2 had used 48 IFBA rods in 4.4 wt% U-235 fuel assemblies, but never less than 64 IFBA rods at 1.25X for higher enrichments. Unit 3 used a small number of feed assemblies with only 48 IFBA rods in Cycle 15 (current cycle is Cycle 18). A multiple misload of fresh 5.0 wt% fuel with less than 64 IFBA rods is therefore very unlikely. However, analysis was performed to determine how many 5 wt% U-235 enriched assemblies with no IFBA rods would need to be placed together to reach k95195 of 0.95 with 2000 ppm soluble boron. It required 12 of these no IFBA 5 wt%

assemblies to be loaded together in a pool where all the rest of the assemblies were at the loading curve limits to reach the calculated k of 0.9307. Adding bias and uncertainty, the final k95/95 becomes 0.9402.

Another multiple misload of all once burned fuel in Region 2 (5.0 wt% at 18 GWd/T) gives a k of 0.9219 at 2000 ppm. At 18 GWd/T, the total bias and uncertainty is 0.0178, so the k 9 5/9 5 is 0.9397, which meets the criterion of< 0.94 while the regulatory requirement is < 0.95. Normal first cycle burnup is more than 20 GWd/T, so a multiple misload of assemblies with less than 18 GWd/T is very unlikely. In fact, all feed assemblies for all cycles in Indian Point Units 2 and 3 with greater than 4.5 wt% U-235 enrichment have exceeded this burnup in the first cycle. Lower enrichments would need less burnup for the same reactivity.

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For Region 1, if all the fuel was 5 wt% U-235 with no IFBA rods and the pool had 2000 ppm of soluble boron, the calculated k is 0.8022. Since after adding biases and uncertainties this is much less than 0.94, there is no concern about multiple misloads in Region 1.

Two additional features of this criticality analysis make multiple misloads unlikely. First, there is no credit for any checkerboard arrangement. Second, the Region 2 minimum burnup requirements are sufficiently low so that very few burned assemblies would fail to meet the requirements (requiring them to be stored in Region I). Since there is very little burned fuel that needs to be stored in Region 1, there is no reason to store fresh fuel in Region 2. Therefore, a misload of a fresh fuel assembly without a control rod in Region 2 would be easily noticed by its shiny look (compared to all of the other assemblies in Region 2).

At the minimum SFP boron concentration of 2000 ppm, even multiple misloads will bc safely subcritical. It is not credible to assume that a multiple misload and a boron dilution event would take place at the same time. Therefore, the criticality safety requirements are met by use of the double contingency principle.

9.5 Boron Dilution Accident Crediting 700 ppm of soluble boron reduces the calculated k plus biases and uncertainty below 0.94.

The boron dilution analysis of record [28] shows that dilution from the Technical Specification required 2000 ppm to 700 ppm is not credible.

9.6 Seismic Event The absorber panels will be designed to withstand a design basis seismic event. Therefore, the absorber panels will remain in place and the rack configuration will not change significantly following a design basis seismic event. Since the absorber panels remain in place, no special analysis is needed.

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10 Summary This section summarizes the results of the criticality analysis. The section starts with a review of allowable fuel loading for the Indian Point Unit 2 spent fuel pool, This is followed by a review of the assumptions used to justify the allowable loading conditions. The racks are not expected to change, so the assumptions in the analysis are found in Section 3.4, and not repeated here. However, the absorber panels have not yet been ordered, so Section 10.2 reviews the requirements for the absorber panels. The fuel manufacturer may change in the future, so the fuel design requirements are repeated in Section 10.3.

Finally, core operating conditions not precluded by the current Technical Specifications are utilized in this analysis. These core operating condition requirements are listed in Section 10.4.

10.1 Summary of Allowable Fuel Loading Fresh fuel of 5.0 wt% U-235 or lower can be loaded anywhere in Region 1, as long as it contains at least 48 IFBA rods with a minimum IX B130 loading. Burned fuel at a burnup of 12 GWd/T or more can be stored anywhere in Region I. Fresh fuel with less than 48 IFBA rods or burned fuel with less than 12 GWd/T burnup can be stored on the periphery of Region I. The periphery is defined as all the cell locations of the last row on the three sides of Module A that face away from Module B and all the cells on the last row of the face of Module 13 and Module C that is adjacent to the pool wall. Cells in the second row or further in Modules A, B, or C cannot take peripheral credit even if the outer row is empty. See Figure 8.8 for a graphical presentation of the location of the peripheral cells. If any fuel assembly contains a control rod, it can be located anywhere in Region 1, regardless of the number of IFBAs or burnup. A Region 1 cell which does not contain an absorber panel does not affect the loading requirements of any other cell in Region I, so long as the cell that is missing an absorber panel contains a fuel assembly with a control rod or does not contain a fuel assembly.

Region 2 is primarily intended to be for discharged fuel. The only allowable way to load fresh fuel in Region 2 requires a control rod to be inserted in the assembly. The minimum burnup requirements for NET- 300067-01 Rev 0 86

Region 2 are presented in Table 10.1 as a function of fuel enrichment and cooling time. Linear interpolation can be used to determine the minimum burnup requirements at any enrichment or cooling time.

Table 10.1: Region 2 Minimum Burnup (GWd/T) Requirements(B-C'")

I Coolina Time (vears)

Enrichment 0 1 2 5 10 15 25_e_

2.0(d) 3.20 3.10 3.08 3.00 3.00 3.00 3.00 2.5 15.17 14.85 14.66 13.97 13.20 12.84 12.31 3.0 21.28 21.17 20.98 20.66 20.26 19.98 19.65 3.5 27.53 27.10 26.63 25.56 24.29 23.50 22.45 4.0 33.82 33.43 33.05 32.04 30.72 29.70 28.44 4.5 38.98 38.65 37.99 36.49 34.67 33.69 32.60 5.0 42.67 42.14 41.78 40.78 39.72 38.96 37.68 Notes:

(a) Fuel assemblies with initial enrichments _>4.0 wt% that do not meet the burnup requirements may be stored in peripheral cells provided the burnup requirements, reduced by 8 GWd/T, are met. If the fuel assembly contained a Hafnium flux suppressor insert then the burnup requirements may be reduced by 6 GWd/T if stored in a peripheral cell. The peripheral locations are shown on Figure 8.8.

(b) Fuel assemblies that contained a Hafnium flux suppressor insert and the burnup prior to the final cycle does not meet the burnup requirements, then 2 GWd/I' must be added to the burnup requirements.

(c) Linear interpolation between enrichment levels and cooling times to determine minimum burnup requirements is permitted.

(d) Fuel assemblies with initial enrichments less than 2.0 wt% must meet the 2.0 wt% minimum bumup requirement. For axially blanketed fuel, the enrichment to be used is the enrichment of the center section between the blanket material.

(e) Fuel assemblies with cooling times greater than 25 years must meet the 25 year burnup requirement.

(f) Fuel assemblies with any fuel rods removed and not replaced (normally with stainless steel rods) must add 4 GWd/T to the burnup requirements. As for Region I, a Region 2 cell which does not contain an absorber panel does not affect the loading requirements of any other cell in Region 2, so long as the cell which is missing an absorber panel does not contain a fuel assembly or contains a fuel assembly with a control rod.

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Any fuel, including unburned 5 wt% U-235 fuel with no restriction on the number of IFBA rods, may be stored in Region 2 if the assembly contains a control rod.

The above loading requirements have been summarized below in Table 10.2.

Table 10.2: Summary of Loading Restrictions Loading Restriction Fresh fuel assemblies with initial enrichments of 5.0 wt% U-235 or less and 48 or more IFBA rods (@ [ mg B- 10/inch] a.U or greater) may be stored in any location in Region 1.

Burned fuel assemblies with initial enrichments of 5.0 wt% U-235 or less with a burnup Region 1 of 12 GWd/T or more may be stored in any location in Region 1.

Any fuel assembly (fresh or burned) that contains an RCCA may be stored in any location in Region 1 with no restrictions on the number of IFBA rods or burnup.

Any fuel assembly (fresh or burned) may be stored in locations designated as peripheral cells in Region I with no restrictions on the number of IFBA rods or burnup.

Burned fuel assemblies that satisfy the requirements of Table 10.1 may be stored in any location in Region 2.

Any fuel assembly (fresh or burned) that contains an RCCA can be stored anywhere in Region 2 Region 2 with no restrictions on the number of [FBA rods or burnup.

Burned fuel assemblies that satisfy the requirements of Table 10.1 reduced by 8 GWd/T may be stored in locations designated as peripheral cells in Region 2."

For assemblies that had a Hafnium insert and the burnup prior to the Final cycle does not meet the burnup requirements of Table 10.1, then 2 GWd/T must be added to the values given in Table 10.1.

For assemblies on the periphery that had a Hafnium insert and the burnup prior to the final cycle does not meet the burnup requirements of Table 10.1, then 6 GWd/T can be subtracted from the values given in Table 10.1.

10.2 Absorber PanelRequirements To meet the assumptions of this criticality analysis, the absorber panels must satisfy the requirements specified in Table 10.3 below. If the alternate design is chosen, the minimum areal density must be 0.020 g B-10/cm 2 for panels in Region 2 and 0.022 g B-10/cm 2 for panels in Region 1. For the alternate design, the connector can be any material and can be any thickness up to 0.10 inch.

NET- 300067-01 Rev 0 88

Table 10.3: Absorber Panel Requirements Attribute Value (inches) Notes Absorber Panel (primar,)

Areal Density (g B- I 0/cm2) 0.015 Minimum Panel width Cell ID - .03 Minimum 0.086 Minimum in Region 1 0.096 Maximum in Region 2 Length Covers active fuel length Absorber Panel (alternate)"

Areal Density (g 13- 10O/cm 2) 0.020 in Region 2 Minimum 0.022 in Region M Panel width 7.6 Minimum 0.075 Minimum in Region 1 0.094 Maximum in Region 2 Offset from corner 0.64 Length _Covers active fuel length A vendor is not bound by these specific dimensions and areal densities. As long as the panel is shown to be as effective in absorbing neutrons as the primary design, it would be acceptable.

10.3 Fuel Requirements To meet the assumptions of this criticality analysis, the fuel design must meet the design assumptions given on Table 10.4. The guide tube dimensions are flexible as long as the cross-sectional area is at least 0.0243 square inches.

Table 10.4: Fuel Design Requirements Attribute Value (inches) Notes Fuel pellet U0 2 stack density 97.5 %TD Maximum stack density*

Fuel pellet OD [j aj Maximum Fuel clad OD I ],C Minimum Fuel clad ID [ ]3"a Maximum Fuel pin pitch 0.5630 Nominal Guide tube cross sectional area 0.0243 in- Minimum Nominal This density includes the effect of dishing and chamfering.

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10.4 Reactor OperationLimits The depletion parameters were selected to cover anticipated future operation, however, verification is required. Table 10.5 lists the operating assumptions used in the depletion analysis for fuel enriched to greater than 3.5 wt%*. The temperature and soluble boron assumptions are averages over the total burnup (multi-cycle) for a given assembly. These assumptions will be verified as part of the reload design process. The process normally assumes a range of previous cycle burnups. If the plant is shutdown outside of this range or before the reload analysis for the following cycle is completed, the assumptions listed in Table 10.5 must be confirmed for each assembly that is to be placed into Region 2 or a non-peripheral location of Region 1. If an assembly is depleted such that any of the Table 10.5 parameters are not met, then the assembly would have to be stored in Region I or with a control rod inserted until an assembly-specific analysis can be performed and approved. Note that if an assembly meets the burnup requirements (GWd/T) and the operating limits, any additional burnup does not need to meet the operating limits.

Table 10.6 is provided for older fuel enriched to 3.5 wt% or less to verify that the existing fuel meets the assumptions of the criticality analysis. Table 10.6 is not to be used for future fuel.

Iffuel less than or equal to 3.5 wt% is ever used in the future, the same requirements shown in Table 10.5 apply. Table 10.5 should be used for all future fuel regardless of enrichment.

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Table 10.5: Fuel Assembly Operating Requirements for Fuel Enriched > 3.5 wt%*

Parameter Value Notes This value can be demonstrated by a Maximum assembly maximum assembly and burnup averaged average moderator 631 -F (605.9 -K) peaking factor. This corresponds to a outlet temperature maximum burnup averaged peaking factor of during depletion t 1.40 with current thermal design flow conditions.

Maximum WABA 20 rodlet WABA at Design changes that increase water loading 0.00603 g 10B/cm per rodlet displacement are not covered.

Maximum lFBA 148 IFBA rods [ mg Credit for IFBAs used with fresh fuel use a rods and '°B loading I°B/inch]a*c (1.5X) per rod minimum IX loading.

Maximum Operation < 2 GWd/T This control rod inserted burnup covers rods with Control Rods inserted to any depth.

Maximum Burnup This is an average for all cycles in which the Averaged Solublc < 1000 ppm assembly was depleted. For burned Brnassemblies stored in Region 1, this Boron requirement is relaxed to < 1300 ppm.

Average Power To cover reduced power operation at end of During the Last 30 > 50% cycle prior to offload Days of Operation If fuel less than or equal to 3.5 wt% is ever used in the future, the same requirements shown in Table 10.5 apply. Table 10.5 should be used for all future fuel regardless of enrichment.

t If the peaking factor is less than 1.40, the moderator outlet temperature requirement is met.

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Table 10.6: Fuel Assembly Operating Requirements for Fuel Enriched < 3.5 wt%*

Parameter Value Notes This value can be demonstrated by a Maximum assembly maximum assembly and burnup averaged average moderator 628 VF (604.4 °K) peaking factor. This corresponds to a outlet temperature maximum burnup averaged peaking factor during depletion" of 1.35 with current thermal design flow conditions.

Maximum BA 20 rodlet Pyrex at a boron loading No limit on when the Pyrex is removed.

loading of 18.1 wt% B20 3 This control rod inserted bumup is the Maximum Operation assembly average burnup during which with Control Rods < 2 GWd/T control rods were inserted below the top node. The top node can have the rod inserted throughout life.

Maximum Bumup This is an average for all cycles in which Averaged Soluble < 800 ppm the assembly was depleted.

Boron Average Power To cover reduced power operation at end During the Last 30 > 50% of cycle prior to offload Days of Operation of cyclepriortoof__oad This table is provided only to certify that older fuel < 3.5 wt% satisfies the older operating requirements. For future fuel, Table 10.5 should be used regardless of enrichment.

t Ifthe peaking factor is less than 1.35, the moderator outlet temperature requirement is met.

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References

[I] K. Wood, DSS-ISG-2010-1, "Draft Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools." Accession Number ML102220567, Nuclear Regulatory Commission, Rockville, MD, August 2010.

[2] Code of Federal Regulations, Title 10, Part 50, Section 68, "Criticality Accident Requirements."

[3] Scale: A Comprehensive Modeling and Simulation Suitefor 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.

Includes 6.1.2 update dated 2/28/13.

[4] "Rack Construction (SH'T. 2) Region I Storage Racks," Drawing Number 398 Rev5, March 19, 1990, Project No. 81000, P. O. No. 8-24470, Holtec International, Mount Laurel, NJ.

[5] "Rack Construction (SH'T. 1) Region 2 Storage Racks," Drawing Number 400 Rev5, March 19, 1990, Project No. 81000, P. O. No. 8-24470, Holtec International, Mount Laurel, NJ.

[6] G. Radulescu, I. C. Gauld, and G. Hlas, SCALE 5. 1 Predictionsof PWR Spent Nuclear Fuel Isotopic Compositions, ORNL/TM-20 10/44, Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA, March 2010.

[7] Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center, Holtec Report No. HI-2094289. Rev. 6, April 17, 2012.

[8] InternationalHandbook of Evaluated CriticalitySafety Benchmark Experiments, NEA/NSC/DOC(95)3, Volume IV, Nuclear Energy Agency, OECD, Paris, September, 2010.

[9] K. S. Smith, et al., Benchmarksfor Quantifying Fuel Reactivity Depletion Uncertainty, EPRI, Palo Alto, CA, Technical Report Number 1022909 (2011).

[10] D. B. Lancaster, Utilization of the EPRI Depletion BenchmarksJorBurnup Credit Validation, EPRI, Palo Alto, CA, 1025203 (2012).

[11] U.S. Nuclear Regulatory Commission, Spent Fuel Project Office Interim Staff Guidance

- 8, Rev. 3 - Burnup Credit in the CriticalitySafety Analyses of PWR Spent Fuel in TransportandStorage Casks, U.S. Nuclear Regulatory Commission, April, 2012.

[12] J.C. Dean and R.W. Tayloe, Jr., Guidefor Validation of Nuclear CriticalitySafety CalculationalMethodology, NUREG/CR-6698, Nuclear Regulatory Commission, Washington, DC January 2001.

[13] D. E. Mueller, K. R. Elam, and P. B. Fox, Evaluation of the French Haut Taux de Combustion (HTC) CriticalExperiment 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.

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[14] J. M. Scaglione, D. E. Mueller, J.C. Wagner and W. J. Marshall, An Approachfor Validating.Actinide and Fission Product Burnup Credit CriticalitySafety Analyses-Criticality (kef]) Predictions,US Nuclear Regulatory Commission, NUREG/CR-7109, Oak Ridge National Laboratory, Oak Ridge, Tenn. (2012).

[15] J.C. Wagner and C. V. Parks, ParametricStudy of the Effect of Burnable Poison Rods for PWR Burnup Credit, US Nuclear Regulatory Commission, N UREG/CR-6761, Oak Ridge National Laboratory, Oak Ridge, Tenn. (2002).

[16] 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).

[17] C. V. Parks, M. D. DeHart, and J.C. Wagner, Review and Prioritizationof Technical Issues Related to Burnup Credit.forLWR Fuel, US Nuclear Regulatory Commission, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge, Tenn. (2000).

[18] D. Hagrman, INTERPIN-3 User's Manual, SSP-0 1/430, Studsvik Scandpower, Inc.

(2001).

[191 M. D. DeHart, Sensitivity and ParametricEvaluations of Significant Aspects of Burnup Creditfor PWR Spent Fuel Packages, ORNL/TM-1 2973, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, May 1996.

[20] J.C. Wagner, M. D. DeHart, and, C. V. Parks, RecommendationsforAddressing Axial Burnup in PWR Burnup CreditAnalyses, US Nuclear Regulatory Commission, NUREG/CR-6801, Oak Ridge National Laboratory, Oak Ridge, Tenn. (2003).

[21] "Pool 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.

[22] "NEI 12-16, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Plants," Slides from NEI/NRC on meeting September 24, 2013, Adams Accession Number, ML13264A008.

[23] NET-300042-01, Rev. 1, "Interim Reactivity Analysis of the Indian Point Unit 2 Spent Fuel Pool," September, 2012.

[24] Indian Point Unit 2 Updated Final Safety Analysis Report (UFSAR), Revision 24, 2013.

[25] K. Lindquist, et al., Guidelinesfor Boraflex Use in Spent-FuelStorage Racks, EPRI, Palo Alto, CA, Technical Report Number 103300 (1993).

[26] Letter from Tom Duberville of Holtec International to Joe DeFrancesco of Entergy-Indian Point, "I-Ioltec International DREAM Insert; basic specification data," September 19,2013.

[27] Guidancefor PerformingCriticalityAnalyses of Fuel Storage at Light- Water Reactor Power Plants, Revision 1, NEI 12-16, Nuclear Energy Institute, Washington, DC, April 2014.

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[28] NET-173-02. Rev. 1, "Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis, September 2001.

[29] Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July, 2000.

[30] Design Input Record, EN-DC-149RI4 140218, Indian Point, February 17, 2014

[31] [NOT USED]

[32'1 R. E. Griffith to G. Delfini, "Fuel Temperatures vs Burnup Curves for Indian Point Unit 2," Entergy Inter-Office Correspondence, CEO2013-00107, August 15, 2013, supported by Entergy Calculation package NEAD-SR-2013/023, EC 4611.

[33] ANSI/ANS-8.1-1998 (R2007), "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," American Nuclear Society, La Grange Park, Illinois.

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Appendix A: Validation of SCALE 6.1.2 for Criticality Analysis of Fresh and Burned Fuel A. 1. Objective This appendix determines the computer code and cross-section library bias and uncertainty in the k's calculated for the Indian Point Units 2 and 3 spent fuel pools when using SCALE 6.1.2 [ I] and the 238 group ENDF/B-VII cross-section library. The bias and uncertainties determined in this Appendix covers both fresh and burned fuel.

A.2. Method The bias and uncertainty is determined as a bias and uncertainty in the initial condition of the fuel and then an additional bias and uncertainty due to the change in reactivity with burnup. This Appendix is divided into three sections: I) Laboratory critical experiments (Fresh U0 2, HTC criticals, and MOX criticals), 2) EPRI benchmarks, and 3) Chemical Assays.

The initial condition bias and uncertainties come from the fresh U0 2 criticals. With the burned fuel, two approaches will be used to determine the bias and uncertainty in the depletion reactivity: 1) Extended ISG-8, and 2) EPRI benchmark based.

In the Extended ISG-8 approach, a bias and uncertainty due to the change in isotopic content is determined and an additional bias and uncertainty due to the cross-sections of isotopes not contained in the initial fresh fuel criticals is included. The HTC and MOX criticals are analyzed to determine if the major actinides introduce a bias and uncertainty not found in the fresh U0 2 experiments. Then a bias of 1.5% of the fission product and minor actinide worth is added. This bias and uncertainty (zero uncertainty is used) is based on the NRC approved method for the transport analysis given in ISG-8, Rev. 3 [9]. ISG-8, Rev. 3, however, is restricted to 28 isotopes and this analysis "extends" this to all significant (185) isotopes (see Table A.3. 1). The last sentence of Section 7 ofNUREG/CR-7109 [32]

states, "An upper value of 1.5% of the worth is also applicable for SNF isotopic compositions consisting of all nuclides in the SFP configuration." The 1.5% bias on the worth of fission products and minor actinides is applied to all of the non-major actinides in the 185 isotopes. The analysis of the EPRI benchmarks [10,1 ] provide confidence that the isotopes not included in the chemical assays do not have a significant error that would overwhelm the 1.5% bias.

The EPRI benchmarks provide a direct measure of the change in reactivity with burnup. They can be used directly for the desired validation. In order to assure conservatism, the more limiting bias and uncertainty from either the Extended ISG-8 approach or the EPRI benchmark approach will be used.

Note that the EPRI benchmarks are also used as part of the Extended ISG-8 approach. Although the EPRI benchmarks do not provide information for individual isotopes, all the isotopes used in the Extended ISG-8 are in the analysis of the EPRI benchmarks. The reactivity of the extension isotopes is NET- 300067-01 Rev 0 A-1,

generally based on only cross-section measurement data. The EPRI benchmarks provide some assurance that there is not a gross transcription error in the cross-section data.

A.3. Computer Codes Used This analysis uses the CSAS5 module of SCALE 6.1.2 [1] for the criticality analysis and the t5-depl module for the depletion analysis. All the analyses are performed using the 238 group ENDF/B-VII library (v7-238). The CSAS5 module executes the CENTRM and BONAMI programs for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of k. All the computer runs use a large Monte Carlo sampling of at least 1500 generations and 6000 neutrons per generation.

The t5-depl implements CENTRM and BONAMI 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 ORIGEN.

parm=(addnux=4) is used in the analysis which utilizes the maximum number of problem specific collapsed isotopes (388). At the end of the depletion analysis, the OPUS module is used to output atom densities (for the EPRI benchmark work) or grams per MTU for the chemical assay work. The input for OPUS could specify any set of isotopes. For this validation, the OPUS input is set for 185 isotopes that are carried forward to the criticality analysis.

Although it may seem that there was a significant reduction in isotopes from the 388 isotopes with the addnux=4 set, many of the isotopes are for structural materials rather than fission products. All of the eliminated isotopes have low atom densities in spent fuel and do not have a large cross section to compensate for the low atom density. Of the 388 isotopes only 185 isotopes have any significant impact on k. The 185 isotopes are listed in Table A.3. 1. Since for the short cooling times the peak reactivity is desired, the Xe-135 is directly converted to Cs-135 and the Np-239 is directly converted to Pu-239. 1-135 and Ru-I 05 have sufficiently short half-lives that they decay away to daughters which arc followed before 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of cooling. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, only 181 isotopes are followed. At low burnups some additional isotopes are eliminated if their atom densities are less than lE-12.

Since SCALE 6.1 does not output isotopic data for burnups after cooling for less than the maximum specified burnup, a small Fortran program was used to decay the isotopic content to the desired cooling time. The Fortran program was only used in support of the EPRI benchmark analysis by decaying the isotopes for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, 5 years and 15 years. The computer code was confirmed by comparison to direct decay performed by SCALE. The Fortran program could be avoided for the validation, but was used to match the method used in the pool criticality analysis, where multiple cooling times and burnups make direct calculation with SCALE very time consuming.

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Table A.3.1: 185 Isotopes Used in the Analysis Ag-109 Crn-243 Gd-160 Nd-145 Rb-85 Sm-153 Te-130 Ag-i 1Oim Cm-244 Ge-73 Nd-146 Rb-86 Sm-154 Te-132 Ag-1 Il Cm-245 Ge-76 Nd-147 Rb-87 Sn- 115 U-234 Am-241 Cm-246 Ho-165 Nd-148 Rh-103 Sn-1 16 U-235 Am-242m Cs-133 1-127 Nd-150 Rh-105 Sn-117 U-236 Am-243 Cs-134 1-129 Np-237 Ru-100 Sn-118 U-237 As-75 Cs-135 1-131 Np-238 Ru-101 Sn-1 19 U-238 Ba-134 Cs-136 1-135 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-I10 Er-166 La-138 Pd-110 Sb-123 Sr-88 Xe-135 Cd-I I I Eu-151 La-139 Pm-147 Sb-124 Sr-89 Xe-136 Cd-i 12 Eu-152 La-140 Pm-148 Sb-125 Sr-90 Y-89 Cd-113 Eu-153 Mo-100 Pm-148m Se-76 Th-159 Y-90 Cd-1 14 Eu-154 Mo-95 Pm-149 Se-77 Tb-160 Y-91 Cd-I 1i5m 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-238 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 Sn-I150 Te-127mn Ce-144 Gd-157 Nd-143 Pu-241 Sm-151 Te-128 Cm-242 Gd-158 Nd-144 Pu-242 Sm-152 Te-I29m NET- 300067-01 Rev 0 A-3

A.4. Analysis The analysis is broken into three major sections: Laboratory Critical Experiments, EPRI Benchmarks, and Extended ISG-8 Chemical Assay Analysis.

A.4.1 LaboratoryCriticalExperiments A. 4.1.1 Introduction The validation consists of modeling 236 fresh U0 2 critical experiments and the determination of the bias and the uncertainty in the calculation of k for fresh fuel. This validation follows the direction of NUREG/CR-6698, "Guide for Validation of Nuclear 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 arca of applicability of the validation and limitations It further defines the steps of "Analyze the data" as:

1. Determine the Bias and Bias Uncertainty

2. Identify Trends in Data, Including Discussion of Methods for Establishing Bias Trends

3. Test for Normal 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 will be followed for this validation analysis.

A.4.1.2 Definition of the Range of Parametersto Be Validated The validation guidance document [2] states:

"Priorto the initiation of the validation activity, the operatingconditions and parametersfor which the validation is to apply must be identified The fissile isotope, enrichment offissile isotope,fuel density,fuel chemicalform, types ofneutron moderators and reflectors, range of moderator tofissile isotope, neutron absorbers,andphysical configurationsare among the parametersto specify. These parameterswill come to define the area of applicabilityfor 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 94% 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, NET- 300067-01 Rev 0 A-4

or in solution) and Ag-In-Cd control rods. The reflectors are water, steel, or concrete. The fuel is in assemblies, but the analysis 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.4.1.3 Selection of the Fresh U0 2 CriticalBenchmark 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 ofconcern.

" 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 soluble boron, borated steel, boron bearing rods, sheets of aluminum with boron, Boraflex, or Ag-In-Cd.

" No emphasis on a feature or material not of importance to the rack analysis.

The OECD/NEA InternationalHandbook of Evaluated CriticalitySafety Benchmarlck 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 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-'FHERM-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 85 sets of benchmarks in the September 2010 version of the handbook. 22 ofthese sets were eliminated, since they were for hexagonal arrays. 4 more sets were eliminated due to enrichments of 7 wt% U-235 or higher. 10 experimental sets were not for water moderated fuel rods. 4 experimental sets were eliminated due to high uncertainties. This leaves 45 benchmark sets of which 31 sets 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 3 1 sets.

The selected 3 1 benchmark sets include critical experiments from six 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 the scope of this validation, such as lead or copper reflectors. The 31 selected benchmark sets resulted in 236 experiments that are used for the statistical analysis. Since boron absorption is important to criticality of the spent fuel pool, it is important that of these 236 experiments, 41 experiments used soluble boron and 28 experiments used boron containing absorber plates.

A later section will evaluate the area of applicability provided by this selection of critical benchmarks.

NET- 300067-01 Rev 0 A-5

Table A.4. 1.1 provides a summary of all the low enriched thermal experiments (non-U metal) from the OECD/NEA handbook [3] and why some experiments were not used.

Table A.4.1.1: Selection Review of OECD/NEA Criticality Benchmarks (All Experiments Start With LEU-COMP-THERM-)

Benchmark Description Lab Selected Number-MODERATEDU_2.35)O__FUEL WATER-MODERATED U(2.35)0 2 FUEL RODS IN 2.032-CM SQUARE-PITCHED PNL All 8 ARRAYS WATER-MODERATED U(4.31)0 2 FUEL 2 RODS IN 2.54-CM SQUARE-PITCHED PNL All 5 ARRAYS WATER-MODERATED U(2.35)O 2 FUEL RODS3NLwell IN 1.684-CM SQUARE-PITCHED None. Gd impurity known. Notnot ARRAYS (GADOLINIUM WATER benchmark quality.

IMPURITY)

WATER-MODERATED U(4.31)0 2 FUEL RODS IN 1.892-CM SQUARE-PITCHED PNL well known. Not ARRAYS (GADOLINIUM WATER benchmark quality.

IMPURITY)

CRITICAL EXPERIMENTS WITH LOW-ENRICHED DIOXIDE FUEL URANIUMCNAIIGPNL None. No sample SCALE 5 OS NWAE decks. Soluble Gd not RODS IN WATER CONTAINING used in pools.

DISSOLVED GADOLINIUM CRITICAL ARRAYS OF LOW-ENRICHED 6 U0 2 FUEL RODS WITH WATER-TO-FUEL JAEA All 18 VOLUME RATIOS RANGING FROM 1.5 TO 3.0 WATER-REFLECTED 4.738-WT.%-

7 ENRICHED URANIUM DIOXIDE FUEL- Valduc in hexagonal arrays.

ROD ARRAYS CRITICAL LATTICES OF U0 2 FUEL RODS 8 AND PERTURBING RODS IN BORATED B&W All 17 WATER WATER-MODERATED RECTANGULAR CLUSTERS OF U(4.31)0 2 FUEL RODS 21 cases used. Did not 9 (2.54-CM PITCH) SEPARATED BY STEEL, PNL include Copper cases since BORAL, COPPER, CADMIUM, no copper in pools.

ALUMINUM, OR ZIRCALOY-4 PLATES WATER-MODERATED U(4.31)02 FUEL 22 cases used. Did not use 10 RODS REFLECTED BY TWO LEAD, PNL lead cases since no lead in URANIUM, OR STEEL WALLS pools.

NET- 300067-01 Rev 0 A-6

Benchmark Description Lab Selected Number CRITICAL EXPERIMENTS SUPPORTING CLOSE PROXIMITY WATER STORAGE OF POWER REACTOR FUEL (PART I -

ABSORBER RODS)

WATER-MODERATED RECTANGULAR CLUSTERS OF U(2.35)0 2 FUEL 12 RODS(1.684-CM PITCH) SEPARATED BY PNL well known. Not STEEL, BORAL, BOROFLEX, benchmark quality.

CADMIUM,OR COPPER PLATES (GADOLINIUM WATER IMPURITY)

WATER-MODERATED RECTANGULAR CLUSTERS OF U(4.31)0 2 FUEL RODS 13 (1.892-CM PITCH) SEPARATED BY PNL 5 cases used. Did not use STEEL, BORAL, BOROFLEX, CADMIUM, the 2 cases with copper.

OR COPPER PLATES, WITH STEEL REFLECTING WALLS WATER-REFLECTED ARRAYS OF 14 U41)2FEROS1.9-MADPNL U(4.31)0 2 FUEL RODS (1.890-CM AND None used.

content High boron uncertainty. Not 1.715-CM SQUARE PITCH) IN BORATED 14 benchmark quality.

WATER THE WER EXPERIMENTS: REGULAR 15 AND PERTURBED HEXAGONAL KFKI None used due to hex LATTICES OF LOW-ENRICHED U0 2 FUEL arrays.

RODS IN LIGHT WATER WATER-MODERATED RECTANGULAR CLUSTERS OF U(2.35)0 2 FUEL RODS 16 (2.032-CM PITCH) SEPARATED BY PNL 26 caued Dnse STEEL, BORAL, COPPER, CADMIUM, ALUMINUM, OR ZIRCALOY-4 PLATES WATER-MODERATED U(2.35)0 2 FUEL 23 cases used. Did not use 17 RODS REFLECTED BY TWO LEAD, PNL the 6 cases with a lead URANIUM, OR STEEL WALLS reflector.

LIGHT WATER MODERATED AND None used. Only I case 18 REFLECTED LOW ENRICHED URANIUM Winfrith with no SCALE sample DIOXIDE (7 WT.%) ROD LATTICE deck. Complex system.

WATER-MODERATED HEXAGONALLY 19 PITCHED LATTICES OF U(5%)0 2 Kurchatov Institute None yse STAINLESS STEEL CLAD FUEL RODS WATER-MODERATED HEXAGONALLY 20 PITCHED PARTIALLY FLOODED Kurchatov Institute None used due to hex LATTICES OF U(5%)0 2 ZIRCONIUM arrays.

CLAD FUEL RODS, 1.3-CM PITCH NET- 300067-01 Rev 0 A-7

Benchmark Description Lab Selected Number HEXAGONALLY PITCHED PARTIALLY FLOODED LATTICES OF U(5%)02 None used due to hex 21 ZIRCONIUM CLAD FUEL RODS Kurchatov Institute arrays.

MODERATED BY WATER WITH BORIC ACID UNIFORM WATER-MODERATED 22 HEXAGONALLY PITCHED LATTICES OF Kurchatov Institute arrays.

RODS WITH U(10%)0 2 FUEL PARTIALLY FLOODED UNIFORM 23 LATTICES OF RODS WITH U(10%)0 2 Kurchatov Institute None used due to hex arrays.

FUEL WATER-MODERATED SQUARE-PITCHED Did not use either case due 24 UNIFORM LATTICES OF RODS WITH Kurchatov Institute to 10 wt% U-235 U(10%)0 2 FUEL enrichment WATER-MODERATED HEXAGONALLY 25 PITCHED LATTICES OF U(7.5%)0 2 Kurchatov Institute None used due to hex arrays.

arrays.

STAINLESS-STEEL-CLAD FUEL RODS WATER-MODERATED U(4.92)0 2 FUEL RODS IN 1.29, 1.09, AND 1.01 CM None used due to hex PITCH HEXAGONAL LATTICES AT arrays.

DIFFERENT TEMPERATURES WATER-MODERATED AND LEAD- None used duetor to lead REFLECTED 4.738% ENRICHED Valduc rneu e 27 reflector.

URANIUM DIOXIDE ROD ARRAYS WATER-MODERATED U(4.31)0 2 FUEL 28 RODS IN TRIANGULAR LATTICES WITH PNL None used due to hex BORON, CADMIUM AND GADOLINIUM arrays.

AS SOLUBLE POISONS None used. No SCALE WATER MODERATED AND WATER sample decks. hf plates 29 REFLECTED 4.74% ENRICHED URANIUM cases without hrhave the DIOXIDE ROD ARRAYS SURROUNDED same pitch and pin as BY HAFNIUM PLATES benchmark 7 above. No significant additional value VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.27-CM PITCH) LATTICES 30 OF LOW-ENRICHED U(3.5 WT.% Kurchatov Institute None used due to hex 235U)0 2 FUEL RODS IN LIGHT WATER arrays.

AT DIFFERENT CORE CRITICAL DIMENSIONS WATER-MODERATED HEXAGONALLY 31 PITCHED PARTIALLY FLOODED Kurchatov Institute None used due to hex LATTICES OF U(5%)0 2 ZIRCONIUM- arrays.

CLAD FUEL RODS, 0.8-CM PITCH NET- 300067-01 Rev 0 A-8

Benchmark Description Lab Selected Number __________

UNIFORM WATER-MODERATED None used due to hex 32 LATTICES OF RODS WITH U(10%)0 2 Kurchatov Institute arrays.

FUEL IN RANGE FROM 20"C TO 274"C REFLECTED AND UNREFLECTED 33 ASSEMBLIES OF 2 AND 3%-ENRICHED ORNL Nonc used. Not U0 2 URANIUM FLUORIDE IN PARAFFIN FOUR 4.738-WT.%-ENRICHED 6 cases used. Did not use URANIUM DIOXIDE ROD ASSEMBLIES cases with gap less than 34 CONTAINED IN CADMIUM, BORATED Valduc 2.5 cm due to high STAINLESS STEEL, OR BORAL SQUARE uncertainty. Did not use CANISTERS, WATER-MODERATED AND Cd plate cases since Cd

-REFLECTED plates not in pool.

CRITICAL ARRAYS OF LOW-ENRICHED U0 2 FUEL RODS IN WATER WITH Used 2 cases. Did not use 35 SOLUBLE GADOLINIUM OR BORON JAEA the case with dissolved AOLIN ON Gd. (not like pool).

POISON THE VVER EXPERIMENTS: REGULAR 36 AND PERTURBED HEXAGONAL None used due to hex LATTICES OF LOW-ENRICHED U0 2 FUEL arrays.

RODS IN LIGHT WATER - Part 2 WATER-MODERATED AND PARTIALLY CONCRETE-REFLECTED 4.738-WT.%- VadNone used. No SCALE ENRICHED URANIUM DIOXIDE ROD sample decks.

ARRAYS WATER-MODERATED 4.738-WT.%- None used. No SCALE 38 ENRICHED URANIUM DIOXIDE ROD Valduc sample decks. Used a ARRAYS NEXT TO A BORATED borated concrete reflector CONCRETE SCREEN (not like pool).

INCOMPLETE ARRAYS OF WATER-39 REFLECTED 4.738-WT.%-ENRICHED Valduc Used all 17 cases.

URANIUM DIOXIDE FUEL-ROD ARRAYS FOUR 4.738-WT.%-ENRICHED URANIUM DIOXIDE ROD ASSEMBLIES 40 CONTAINED IN BORATED STAINLESS V Used 4 cases. Did not use STEEL OR BORAL SQUARE CANISTERS, lead reflector cases.

WATER MODERATED AND REFLECTED BY LEAD OR STEEL STORAGE ARRAYS OF 3%-ENRICHED Did not use the 5 cases due 41 LWR ASSEMBLIES: THE CRISTO II Cadarache to complex geometry and EXPERIMENT IN THE EOLE REACTOR no SCALE sample deck.

NET- 300067-01 Rev 0 A-9

Benchmark Description Lab Selected Number____________

WATER-MODERATED RECTANGULAR CLUSTERS OF U(2.35)0 2 FUEL RODS 42 (1.684-CM PITCH) SEPARATED BY PNL Used 5 cases. Did not use STEEL, BORAL, BOROFLEX, CADMIUM, copper cases.

OR COPPER PLATES, WITH STEEL REFLECTING WALLS CRITICAL LOADING CONFIGURATIONS Used only one case. Rest 43 OF THE IPEN/MB-O1 REACTOR WITH A IPEN of cases were not HEAVY SS-304 REFLECTOR significantly different.

CRITICAL LOADING CONFIGURATIONS OF THE IPEN/MB-01 REACTOR WITH Used only one case. Rest 44 IPEN of cases were not UO 2 , STAINLESS STEEL AND COPPER significantly different.

RODS__________________

PLEXIGLAS 45OR CONCRETE-REFLECTED None used since not pin 45 U(4.46) 30 8 WITH H/U=0.77 AND Rocky Flats uedsnetpi INTERSTITIAL MODERATION geomet_.

46 Not included in 2010 Handbook FUEL TRANSPORT FLASK CRITICAL BENCHMARK EXPERIMENTS WITH None used. 3 complex 47 LOW-ENRICHED URANIUM DIOXIDE Winfrith cases. No SCALE sample decks.

FUEL LIGHT WATER MODERATED AND 48 REFLECTED LOW-ENRICHED (3 WT.% Winfrith All 5 cases used 235U) URANIUM DIOXIDE ROD LATTICES MARACAS PROGRAMME: POLYTHENE-REFLECTED CRITICAL CONFIGURATIONS None used. Powder rather 49 WITH LOW-ENRICHED AND LOW- Valduc than pellets. Not similar to MODERATED URANIUM DIOXIDE pools.

POWDER, U(5)0 2.

149SM SOLUTION TANK IN THE MIDDLE OF WATER-MODERATED 7 cases used. Did not use 50 Valduc cases with dissolved Sm.

4.738-WT.%-ENRICHED URANIUM This is not typical of pools.

________ DIOXIDE ROD ARRAYS _________

CRITICAL EXPERIMENTS SUPPORTING 9 cases used. Did not use CLOSE PROXIMITY WATER STORAGE OF cases with the borated Al 51 POWER REACTOR FUEL (PART - OF B&W plates since primary source listed a high uncertainty in ISOLATING PLATES) the boron content.

URANIUM DIOXIDE (4.738-WT.%-

52 ENRICHED) FUEL ROD ARRAYS Valduc None used due to hex MODERATED AND REFLECTED BY arrays.

GADOLINIUM NITRATE SOLUTION NET- 300067-01 Rev 0 A-10

Benchmark Description Lab Selected Number____________

VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.27 CM PITCH) LATTICES 53 OF LOW-ENRICHED U(4.4 WT.% Kurchatov Institute None used due to hex 235U)0 2 FUEL RODS IN LIGHT WATER arrays.

AT DIFFERENT CORE CRITICAL DIMENSIONS CRITICAL LOADING CONFIGURATIONS Used only one case. Rest 54 OF THE IPEN/MB-01 REACTOR WITH IPEN of cases were not U0 2, AND U0 2-Gd 2O 3 RODS significantly different.

LIGHT-WATER MODERATED AND Neither case used.

55 REFLECTED LOW-ENRICHED URANIUM Winfrith Complex geometry no (3 wt.% 235U) DIOXIDE ROD LATTICES KENO-V.a sample deck CRITICAL EXPERIMENT WITH BORAX-V None used. No sample 56 BOILING WATER REACTOR TYPE FUEL INL SCALE decks. Complex ASSEMBLIES BWR geometry.

4.738-WT.%-ENRICHED URANIUM 57 DIOXIDE FUEL ROD ARRAYS REFLECTED Valduc None used. No sample BY WATER IN A DRY STORAGE SCAIF. decks.

CONFIGURATION CRITICAL LOADING CONFIGURATIONS 58 OF THE IPEN/MB-01 REACTOR WITH IPEN NonE u SCALE edNsa decks.

LARGE VOID IN THE REFLECTOR 59 Not included in 2010 Handbook RBMK GRAPHITE REACTOR: UNIFORM CONFIGURATIONS OF U(1.8, 2.0, or 2.4% 235U)0 2 FUEL ASSEMBLIES, AND CONFIGURATIONS OF U(2.0% 235U)O 2 None used. RBMK - not ASSEMBLIES WITH EMPTY CHANNELS, typical of LWRs WATER COLUMNS, AND BORON OR THORIUM ABSORBERS, WITH OR WITHOUT WATER IN CHANNELS WER PHYSICS EXPERIMENTS:

HEXAGONAL (1.27-CM PITCH) LATTICES OF U(4.4 WT.% 235U)0 2 FUEL RODS IN 61 LIGHT WATER, PERTURBED BY BORON, Kurchatov Institute None used due to hex HAFNIUM, OR DYSPROSIUM ABSORBER arrays.

RODS, OR BY WATER GAP WITH/WITHOUT EMPTY ALUMINIUM TUBES 2.6%-ENRICHED U0 2 RODS IN LIGHT-62 WATER MODERATOR WITH BORATED JAEAsapedc. None used. No SCALE STAINLESS STEEL PLATE: SINGLE sample decks.

ARRAYS NET- 300067-01 Rev 0 A-1II

Benchmark Number Description Lab Selected LIGHT-WATER MODERATED AND REFLECTED LOW-ENRICHED URANIUM None used. No SCALE (3 wt.% 235U) DIOXIDE ROD LATTICES sample decks.

WITH DISCRETE POISON-ROD ARRAYS 64 Not included in 2010 Handbook CRITICAL CONFIGURATIONS OF 2.6%-

ENRICHED U0 2 ROD ARRAYS IN LIGHT-65 WATER MODERATOR WITH BORATED JAEA None u edNsC STAINLESS STEEL PLATE: COUPLED ARRAYS PLEXIGLAS-REFLECTED, CONCRETE-66 REFLECTED, OR THIN STEEL-REFLECTED Rocky Flats None used. Not an array U(4.46)3 0 8 WITH H/U=0.77 AND HEU of rods.

DRIVERS 67 Not included in 2010 Handbook PLEXIGLAS-REFLECTED, CONCRETE-68 REFLECTED, OR THIN STEEL-REFLECTED Rocky Flats None used. Not an array U(4.48) 30 8 WITH H/U=1.25 OR of rods.

H/U=2.03 AND HEU DRIVERS PLEXIGLAS-REFLECTED U(4.48) 30 8 None used. Not an array 69 WITH H/U=1.25 OR H/U=2.03 AND Rocky Flats of rods.

INTERSTITIAL MODERATION VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.10-CM PITCH) LATTICES 70 OF LOW-ENRICHED U(6.5 WT.% Kurchatov Institute None used due to hex 235U)0 2 FUEL RODS IN LIGHT WATER arrays.

AT DIFFERENT CORE CRITICAL DIMENSIONS LOW MODERATED 4.738-WT.%-

71 ENRICHED URANIUM DIOXIDE FUEL Valduc All 4 cases used.

ROD ARRAYS UNDER-MODERATED 4.738-WT.%-

ENRICHED URANIUM DIOXIDE FUEL Used 3 cases. Did not use 72 ROD ARRAYS REFLECTED BY WATER OR Valduc Polyethylene reflector cases.

POLYETHYLENE UNDER-MODERATED 4.738-WT.%-

ENRICHED URANIUM DIOXIDE FUEL None used. No SCALE ROD ARRAYS REFLECTED BY WATER sample decks.

WITH HETEROGENEITIES 74 Not included in 2010 Handbook NET- 300067-01 Rev 0 A-12

Benchmark Number Description Lab Selected VVER PHYSICS EXPERIMENTS:

HEXAGONAL (1.10 CM PITCH) LATTICES 75 OF LOW-ENRICHED U(6.5 WT.% Kurchatov Institute None used due to hex 235U)O 2 FUEL RODS IN LIGHT WATER, arrays.

PERTURBED BY BORON ABSORBER RODS AND WATER HOLES LIGHT WATER MODERATED AND REFLECTED LOW ENRICHED URANIUM None used. No KENO Va (3 WT.% 235U) DIOXIDE ROD LATTICES sample decks.

WITH EX-CORE DETECTOR FEATURE Only one case used. Rest CRITICAL LOADING CONFIGURATIONS of cases same materials 77 CRITIA ING CONFIGURATON IPEN with small modification of OF THE IPEN/MB-01 REACTOR arrays. Not sufficiently independent.

78 Not included in 2010 Handbook WATER-MODERATED U(4.31)0 2 FUEL 79 ROD LATTICES CONTAINING RHODIUM SNL None ys FOILS arrays.

80 Not included in 2010 Handbook PWR TYPE UO2 FUEL RODS WITH ENRICHMENTS OF 3.5 AND 6.6 WT.% Single case not use. No 81 WITH BURNABLE ABSORBER ("OTTO AN EX sample SCALE deck.

HAHN" NUCLEAR SHIP PROGRAM, Unusual case.

SECOND CORE)

CRITICAL LOADING CONFIGURATIONS OF THE IPEN/MB-01 REACTOR WITH Used only one case. Rest LOW ENRICHED FUEL AND BURNABLE IPEN of cases were not significantly different.

POISON RODS CRITICAL LOADING CONFIGURATIONS Used only one case. Rest 83 OF THE IPEN/MB-01 REACTOR WITH A IPEN of cases were not BIG CENTRAL VOID significantly different.

CRITICAL LOADING CONFIGURATIONS 84 OF THE IPEN/MB-01 REACTOR WITH A IPEN Used the single case..

CENTRAL CRUCIFORM ROD VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.27 CM PITCH) LATTICES 85 OF LOW-ENRICHED U(6.5 WT.% Kurchatov Institute None used due to hex 235U)0 2 FUEL RODS IN LIGHT WATER arrays..

AT DIFFERENT CORE CRITICAL DIMENSIONS NET- 300067-01 Rev 0 A-13

Benchmark Nchmber Description Lab Selected VVER PHYSICS EXPERIMENTS:

HEXAGONAL LATTICES (1.275 CM 86 PITCH) OF LOW ENRICHED U(3.6, 4.4 NRI None yse t WT.% 235U)0 2 FUEL ASSEMBLIES IN arrays..

LIGHT WATER WITH H3BO3 VVER PHYSICS EXPERIMENTS:

HEXAGONAL LATTICES (1.22-CM PITCH) 87 OF LOW-ENRICHED U(3.6, 4.4 WT.% NRI None used due to hex U235)0 2 FUEL ASSEMBLIES IN LIGHT arrays..

WATER WITH VARIABLE FUEL-ASSEMBLY PITCH 88 Not included in 2010 Handbook CRITICAL LOADING CONFIGURATIONS OF THE IPEN/MB-01 REACTOR WITH Used only one case. Rest 89 OFIPEN of cases were not UO 2 AND BORATED STAINLESS STEEL significantly different.

_________ ~PLATES__ _ _ _ _ _ _ _ _ ___ _______

CRITICAL LOADING CONFIGURATIONS Used only one case. Rest 90 OF THE IPEN/MB-01 REACTOR WITH IPEN of cases were not U0 2 AND STAINLESS STEEL RODS significantly different.

CRITICAL LOADING CONFIGURATIONS OF THE IPEN/MB-01 REACTOR WITH Used only one case. Rest 91 IPEN of cases were not UO 2 , STAINLESS STEEL AND GD203 significantly different.

RODS__ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _

92 Not included in 2010 Handbook DEUTERIUM CRITICAL ASSEMBLY WITH 93 1.2% ENRICHED URANIUM VARYING PNC Not used since cases use COOLANT VOID FRACTION AND D 20 rather than 1420 LATTICE PITCH VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.10 CM PITCH) TWO-REGION LATTICES OF LOW-ENRICHED Kurchatov Institute None used due to hex U(6.5 AND 4.4 WT.% 235U)0 2 FUEL arrays.

RODS IN LIGHT WATER AT DIFFERENT

_ __ CORE CRITICAL DIMENSIONS A.4.1.4 Computer Analysis of the Fresh U0 2 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 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 the decks were modified to 6000 neutrons per generation and 1500 generations.

This was sufficient to make the Monte Carlo uncertainty to be 0.0002 or about one tenth the experimental NET- 300067-01 Rev 0 A-14

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 files 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 specific criticalexperiments, the facility or site may choose to use inputfiles generated elsewhere to expedite the validationprocess. The site has the responsibility.forensuring that inputfiles and the options selected are appropriatefor use. Regardless of the source of the input file, the site must have reviewed the description of each critical experiment and determined that the representationof the experiment, including simplifying assumptionsand options, are consistent with the intended use. In other words, the site must assume ownership of the input file.

Table A.4.1.2 shows the results of the analysis of the 236 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.4.1.2: Critical Experiment Results with SCALE 6.1.2 and ENDF/B-VII Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kerr ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (delta k)

LCT- 1 I 2.350 1.270 2.032 0.0964 0.003 0.9979 2 2.350 1.270 2.032 0.0957 0.003 0.9975 3 2.350 1.270 2.032 0.0950 0.003 0.9968 4 2.350 1.270 2.032 0.0955 0.003 0.9974 5 2.350 1.270 2.032 0.0942 0.003 0.9954 6 2.350 1.270 2.032 0.0952 0.0027 0.9976 7 2.350 1.270 2.032 0.0934 0.0031 0.9972 8 2.350 1.270 2.032 0.0945 0.003 0.9962 LCT-2 1 4.310 1.415 2.540 0.1132 0.002 0.9971 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 0.0018 0.9979 5 4.310 1.415 2.540 0.1103 0.0019 0.9962 LCT-6 1 2.596 1.417 1.849 0.2366 0.002 0.9977 2 2.596 1.417 1.849 0.2432 0.002 0.9987 3 2.596 1.417 1.849 0.2495 0.002 0.9987 4 2.596 1.417 1.956 0.1818 0.002 0.9984 5 2.596 1.417 1.956 0.1871 0.002 0.9986 6 2.596 1.417 1.956 0.1927 0.002 0.9983 7 2.596 1.417 1.956 0.1977 0.002 0.9989 8 2.596 1.417 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 1.417 2.150 0.1462 0.002 0.9982 13 2.596 1.417 2.150 0.1497 0.002 0.9981 14 2.596 1.417 2.293 0.1147 0.002 0.9988 NET- 3 00067-01 Rev 0 A-15

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kff ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty

.,_ _ 235) (cm) (delta k) 15 2.596 1.417 2.293 0.1174 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 1.417 2.293 0.1254 0.002 0.9985 LCT-7 I 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 0.2792 0.0012 0.9969 2 2.459 1.206 1.636 0.2467 0.0012 0.9971 3 2.459 1.206 1.636 0.2465 0.0012 0.9978 4 2.459 1.206 1.636 0.2465 0.0012 0.9970 5 2.459 1.206 1.636 0.2468 0.0012 0.9967 6 2.459 1.206 1.636 0.2461 0.0012 0.9977 7 2.459 1.206 1.636 0.2459 0.0012 0.9964 8 2.459 1.206 1.636 0.2440 0.0012 0.9963 9 2.459 1.206 1.636 0.2437 0.0012 0.9962 10 2.459 1.206 1.636 0.2498 0.0012 0.9969 11 2.459 1.206 1,636 0.2549 0.0012 0.9980 12 2.459 1.206 1.636 0.2489 0.0012 0.9970 13 2.459 1.206 1.636 0.2489 0.0012 0.9975 14 2.459 1.206 1.636 0.2510 0.0012 0.9970 15 2.459 1.206 1,636 0.2509 0.0012 0.9967 16 2.459 1.206 1.636 0.2278 0.0012 0.9972 17 2.459 1.206 1.636 0.1991 0.0012 0.9971 LCT-9 1 4.310 1.415 2.540 0.1127 0.0021 0.9980 2 4.310 1.415 2,540 0.1122 0.0021 0.9986 3 4.310 1.415 2.540 0.1125 0.0021 0.9979 4 4.310 1.415 2.540 0.1121 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 0.1137 0.0021 0.9994 8 4.310 1.415 2.540 0.1130 0.0021 0.9981 9 4.310 1.415 2.540 0.1135 0.0021 0.9986 16 4.310 1.415 2.540 0.1135 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 4.310 1.415 2.540 0.1137 0.0021 0.9982 21 4.310 1.415 2.540 0.1129 0.0021 0.9988 22 4.310 1.415 2.540 0.1138 0.0021 0.9984 23 4.310 1.415 2.540 0.1130 0.0021 0.9994 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 0.2615 0.0021 1.0003 7 4.310 1.415 2.540 0.2092 0.0021 1.0006 NET- 300067-01 Rev 0 A-16

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kerf ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (delta k) 8 4.310 1.415 2.540 0.1844 0.0021 0.9979 9 4.310 1.415 2.540 0.1221 0.0021 1.0007 10 4.310 1.415 2.540 0.1183 0.0021 1.0013 11 4.310 1.415 2.540 0.1154 0.0021 1.0006 Q2 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 0.3071 0.0028 1.0014 15 4.310 1.415 1.892 0.2950 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 1.415 1.892 0.5536 0.0028 1.0010 26 4.310 1.415 1.892 0.5122 0.0028 1.0010 27 4.310 1.415 1.892 0.4780 0.0028 1.0017 28 4.310 1.415 1.892 0.4485 0.0028 1.0017 29 4.310 1.415 1.892 0.4232 0.0028 1.0016 30 4.310 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 0.9971 4 2.459 1.206 1.636 0.1927 0.0032 0.9972 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 0.0017 0.9943 14 2.459 1.206 1.636 0.1508 0.0017 0.9946 15 2.459 1.206 1.636 0.1387 0.0018 0.9959 LCT- 13 1 4.310 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 1 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 4 2.350 1.270 2.032 0.0956 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 NET- 300067-01 Rev 0 A-17

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. keff ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm_ (delta k)

I1 2.350 1.270 2.032 0.0962 0.0031 0.9978 12 2.350 1.270 2.032 0.0974 0.0031 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 0.0031 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 0.2017 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 0.003 t 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 1.270 2.032 0.0953 0.0031 0.9976 14 2.350 1.270 2.032 0.0946 0.0031 0.9985 15 2.350 1.270 1.684 0.1777 0.0028 0.9961 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 1.270 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 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 I 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 I 4.738 0.940 1.260 0.2218 0.0014 0.9953 2 4.738 0.940 1.260 0.2119 0.0014 0.9969 NET- 300067-01 Rev 0 A-18

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kfr

[D No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (delta k) 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 0.2218 0.0013 0.9953 12 4.738 0.940 1.260 0.2166 0.0013 0.9951 13 4.738 0.940 1.260 0.2146 0.0013 0.9951 14 4.738 0.940 1.260 0.2124 0.0013 0.9954 15 4.738 0.940 1.260 0.2112 0.0013 0.9959 16 4.738 0.940 1.260 0.2104 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-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 0.2141 0.0010 0.9991 7 4.738 0.940 1.300 0.2095 0.0010 0.9992 LCT-51 I CIO 2.459 1.206 1.636 0.1472 0.0020 0.9965 2 c2c la 2.459 1.206 1.636 0.1968 0.0024 0.9972 3 c I Ib 2.459 1.206 1.636 0.1964 0.0024 0.9972 4 c I Ic 2.459 1.206 1.636 0.1979 0.0024 0.9975 5 c I Id 2.459 1.206 1.636 0.1989 0.0024 0.9970 6cIIe 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_1Ig 2.459 1.206 1.636 0.2011 0.0024 0.9971 9 cc12 2.459 1.206 1.636 0.1669 0.0019 0.9969 LCT-54 1 4.349 0.980 1.500 0.1508 0.0005 0.9996 LCT-71 1 4.738 0.949 I.100 0.7592 0.00076 0.9955 N~~~l1'0 3006.07592A-1 NET- 300067-01 Rev 0 A-19

Benchmark Case Enrichment Fuel Pin Fuel Pin EALF Meas. kff ID No. (wt% U- Diameter Pitch (cm) (eV) Uncertainty 235) (cm) (delta k) 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 1 4.738 0.949 1.600 0.1117 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 1 4.349 0.980 1.500 0.1541 0.0010 1.0008 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 0.0010 0.9999 Since boron is important to criticality analysis of racks, two methods are used to parameterize the boron:

the B-1 0 areal density for plates and the soluble boron ppm. Table A.4.1.3 shows the boron information on boron-containing benchmarks, along with the calculated k.

NET- 300067-01 Rev 0 A-20

Table A.4.1.3: Summary of Critical Experiments Containing Boron Benchmark Case No. Soluble Separator Plate No. of kf ID Boron B-10 Areal Density Boron (ppm) (gm/cm 2 ) Rods LCT-8 1 1511 0.9969 2 1334 0.9971 3 1337 0.9978 4 1183 36 0.9970 5 1181 36 0.9967 6 1034 72 0.9977 7 1031 72 0.9964 8 794 144 0.9963 9 779 144 0.9962 10 1245 72 0.9969 11 1384 0.9980 12 1348 0.9970 13 1348 0.9975 14 1363 0.9970 15 1362 0.9967 16 1158 0.9972 17 921 0.9971 LCT-9 5 0.004549 0.9993 6 0.004549 0.9985 LCT-9 7 0.006904 0.9994 8 0.006904 0.9981 9 0.066946 0.9986 LCT-11 2 1037 0.9967 3 769 0.9971 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 0.9971 11 0.006904 J 0.9978 12 0.066946 0.9972 NET- 300067-01 Rev 0 A-21

Benchmark Case No. Soluble Separator Plate No. of kff ID Boron B-10 Areal Density Boron (ppm) (gm/cm 2) Rods 13 0.066946 0.9979 14 0.066946 0.9974 LCT-34 4 0.002521 1.0003 5 0.002521 0.9999 6 0.002521 1.0017 7 0.002521 1.0002 8 0.002521 0.9992 15 0.0460 t1 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 0.046011 0.9993 10 0.046011 0.9931 LCT-42 2 0.004549 0.9968 3 0.030173 0.9981 4 0.056950 0.9980 LCT-50 3 822 0.9978 4 822 0.9972 5 5030 0.9983 6 5030 0.9991 7 5030 0.9992 LCT-51 I CIO 143 0.9965 2 clla 510 0.9972 3 cl lb 514 0.9972 4cllc 501 0.9975 5 cl ld 493 0.9970 6 c.1 le 474 0.9972 7clIf 462 0.9973 8cllg 432 0.9971 9 c12 217 0.9969 LCT-77 3 4 1.0006 LCT-82 3 6 1.0005 A.4.1.5 StatisticalAnalysis of the Fresh U0 2 CriticalBenchmark Results The statistical treatment used follows the guidance provided in NUREG/CR-6698 [2]. The NUREG approach weights the calculated k's by the experimental uncertainty. This approach means the higher quality experiments (i.e.: lower uncertainty - see Table A.4.1.2) 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.

NET- 300067-01 Rev 0 A-22

Before seeking trends, the 236 critical benchmarks are reviewed as a whole. The unweighted mean k of the 236 samples is 0.9979 with a standard deviation of 0.0016. The weighted mean is 0.9978 and the weighted standard deviation is 0.0024. These results show that the weighting has a negligible effect on the mean, but does increase the standard deviation. This increase in the standard deviation may be dominated by differences in the experimental uncertainty, which ranges from 0.0005 to 0.0046. Further, the average uncertainty of the experiments (interpreted as one sigma) is 0.0022. Since the total one sigma standard deviation is only 0.0024, this suggests that the experimental uncertainty dominates the uncertainty and there is little to be gained with improved methods. Unless stated otherwise, all the results presented will come from the weighted analysis. The bias of the set as a whole is 0.0022. The uncertainty is the standard deviation multiplied by the single-sided lower tolerance factor (taken as 2.065 from Table 2.1 of Reference 2), so it is 0.0049.

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 available a statistical package, DATAPLOT [4]. The 236 critical experiments were tested with the Wilk-Shapiro normality test and were found to adhere to a normal distribution at only the 99% level. The test results are shown in Table A.4.1.4. Since the Wilk-Shapiro test shows normality only at 99%, a histogram plot of the data was made. This plot suggests that a normal distribution assumption is valid. This plot is Figure A.4. 1.1.

Notice that the calculated k's are a little closer to the mean than expected in a normal distribution. This means assuming a normal distribution is conservative for this data.

Table A.4.1.4: Wilk-Shapiro Test Results Output From DATAPLOT [41 WILK-SHAPIRO TEST k WILK-SHAPIRO TEST FOR NORMALITY

1. STATISTICS:

NUMBER OF OBSERVATIONS = 236 LOCATION PARAMETER = 0.9979254 SCALE PARAMETER = 0.1639522E-02 WILK-SHAPIRO TEST STATISTIC VALUE = 0.9844111

2. CRITICAL VALUES:

P-VALUE = 0.10903651E-0l

3. CONCLUSIONS:

AT THE 90% LEVEL, WE REJECT THE NORMALITY ASSUMPTION.

AT THE 95% LEVEL, WE REJECT THE NORMALITY ASSUMPTION.

AT THE 97.5% LEVEL, WE REJECT THE NORMALITY ASSUMPTION.

AT THE 99% LEVEL, WE ACCEPT THE NORMALITY ASSUMPTION.

NET- 300067-01 Rev 0 A-23

Calculated keff Distribution Versus a Normal Distribution 45 40 35 30 25 20 S15 10 5

0 1 1 1 K-eff by 0.0005S bins Figure A.4.1.1: Distribution of the Calculated k's Around the Mean One possible source of the limited normality could be due to a difference in results with some subsets, but this is not the case. There are 76 experiments that have boron in them. The average k of the boron containing cases is 0.9975, which is very close to the average of all cases (0.9979) (Note: the Monte Carlo uncertainty is 0.0002). Similarly, there are 15 cases that used Cadmium absorbers. The mean of these cases is 0.9981. Since the standard deviation of the set as a whole is 0.0016 (unweighted), it is clear these features are not causing the limited normality.

Numerous sources [5, 6, 7] suggest that for the large sample size used here (236 experiments), normality testing is not important. For example, the guide to the PROPHET statistical program [6] states:

NET- 300067-01 Rev 0 A-24 I

"With large sample sizes, most normal-theory-basedtests like the t test are robust to non-normality, and if the non-normality is not apparentin the normalprobabilityplol for a large data sample, it probably won't have a serious effect on the results of a normal-theory-basedtest. "

In the textbook, Statistics for Social Science by R. Mark Sirkin [7], it states:

"Law of large numbers. A law that states that if the size of the sample, n, is sufficiently large (no less than 30; preferably no less than 50), then the central limit theory will apply even if the population is not normally distributedalong variablex ...

If: Then:

n>= 100 It is always safe to relcx the normality assumption 50<=n<1O0 It is almost always safe 30<=n<50 It is probably safe. "

The analysis in this validation assumes that the techniques used here are sufficiently robust for the limited normality data. However, a non-parametric check has been performed. The 236 cases were ranked by increasing k. 95% of the cases are above the I I"' case. The k of the I I"'case is 0.995 1. Since the average of the experiments is 0.9979, a standard deviation of (.9979-.995 1)/1.96 = 0.00 14 can be inferred.

This inferred standard deviation is less than the standard deviation of 0.0024 (or the unweighted value of 0.00 16) predicted assuming a normal distribution. Again, this is expected from a visual inspection of Figure A.4. 1. 1, since the predicted distribution shows a significantly larger number of data points in the center compared to a normal distribution. In conclusion, although the data does not meet the normality tests, it will be treated as normally distributed, yielding a conservative bias uncertainty.

The next step in the analysis is to look for trends in the data. Section 3.2.2 of the DOE/RW technical report in support of validation for bumup 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.

Let the regression fit be of the form:

k =a+ b x Let x-bar be the average value of x for the n cases and define:

ad= In and define:

NET- 300067-01 Rev 0 A-25

SSR= Z(k, -a-b*x,)2 then the test statistic is:

TbI ý (n-2)*S.,,

SSI?

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's 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. The width of the tolerance band is a small function of the trending parameter. For this analysis, the width of the tolerance band is calculated at the maximum and minimum value of the trending parameter and the maximum of these two widths is taken as the uncertainty. The change in the width as a function of trending parameter is so small that any change in reported uncertainty does not depend on which tolerance band width is used.

In the final analysis, the calculated k of the system must be less than the minimum of k(x) minus the uncertainty minus the administrative safety margin. The uncertainty in k from other independent uncertainties, such as the manufacturing tolerances, bumnup, and depletion uncertainties can be statistically combined with the uncertainty in the criticality validation. The rest of this section will evaluate the trends in k as a function of trending parameters using the methods described above.

Historically, an Upper Suberiticality 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.

Neutron spectrum Trends in the calculated k 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 causing Fission (EALF). Figure A.4.1.2 shows the distribution of k's 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.4.1.2, the predicted mean k as a function of EALF is:

k(EALF) = 0.99865 - 0.00386

• EALF NET- 300067-01 Rev 0 A-26

The units for EALF are eV. The uncertainty about the trend is calculated using the second term of NUREG/CR-6698 [2] equation 23 and is 0.00431 in k.

1.003 1.002 4 1.o001***

1.000

• 0.999 0.998

• 0.997 0.996

• 0.995 0.994 0.993 0.992 I 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.4.1.2: ktrf as a Function of the Energy of the Average Lethargy Causing Fission 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 significant 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 [2J equations 10 through 13 and the data from Table A.4.1.2, the predicted mean k as a function of pin diameter is:

k(Pin Diameter) = 0.99592 + ( 1.662E-03)*Pin Diameter where the pin diameter is in cm. The predicted mean k as a function of pitch is:

k(Pitch) = 0.99547 + ( 1.389E-03)*Pitch NET- 300067-01 Rev 0 A-27

where lattice pitch is in cm.

The tolerance band widths, using the second term of NUREG/CR-6698 [2] equation 23, are 4.343E-03 and 4.270E-03 for the pin diameter and pitch, respectively. Figures A.4.1.3 and A.4.1.4 graphically present kff as a function of the pin diameter and the lattice pitch.

1.003 1.002 1.001 0.999 0.994 0.993 0.992 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5 Pin Diameter (cm)

Figure A.4.1.3: ker as a Function of the Pin Diameter NET- 300067-01 Rev 0 A-28

1.003 1.002 1.001 1.000 0 .999 0.998 0.997 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 Pkch (cm)

Figure A.4.1.4: kerr as a Function of the Lattice Pitch Enrichment The fuel to be stored in the racks ranges in enrichment from 1.6 wt%

• 5 Uto 5 wt/o Z'U. It was determined that there exists a statistically significant trend on enrichment. Using NUREG/CR-6698 [21 equations 10 through 13 and the data from Table A.4. 1.2, the trend in the mean k is:

k(Enrichment) = 0.99686 + ( 2.42E-44)*Enrichment where Enrichment is wt%0U 23 The tolerance band width is 4.35E-03. Figure A.4.1.5 graphically presents the results.

NET- 300067-01 Rev 0 A-29

1.003 1.002 1.001 I

S0.999

• 0.998 2 0.997 Y-0.996 S0.9975 0.994 0.993 0.992 2 2.5 3 3.5 4 4.5 5 Enrichment (wt% U-235)

Figure A.4.1.5: kiff as a Function of the Fuel Enrichment Boron Content A fit of the calculated k's as a function of the B-I10 areal density in the absorber plates or the soluble boron ppm was performed using NUREG/CR-6698 [2] equations 10 through 13 and the data from Table A.4.1.2. Both fits failed the statistically significance test compared to a zero slope. However, to be conservative, both the zero slope and the calculated fit are used for determining the limiting k as a function of boron content.

The following equations are the best fit of the data for k versus soluble boron and areal density. Figures A.4.1.6 and A.4.1.7 show the results of the analyses. The uncertainty around the mean values given in the following equations is 0.00440 and 0.00497.

k(ppm soluble boron) = 0.99777 + ( 5.24E-08)*ppm k(B-10 Areal Density) = 0.99777 + ( 5.24E-08)*(B-10 Areal Density) where the B-10 areal density is in gm B-10/cm2 .

NET- 300067-01 Rev 0 A-30

1.0020 1.000o'*

0 .9980 o 909960 0.9940 0.050 0.060 0.070 0.020 0.030 0.040 0.99200.000 0.010 ISm B-10/cm'Cm)

Areal S-10 Density Separator Plates of the B-i 0 Areal Density in the Figure A.4.1.6: keras a Function 0.9995 0.9990 0.9985 9,

0.9980 0,9975

• 0.9965 0.9960 6000o 4000 5000 2000 3000 0 1000 Baron (Ppm)

-.. .Soluble of the Soluble Boron Content Figure A.4.1.7: kitras a Function A.4.1.6 Establishingthe Bias andthe Uncertainty analysis conservative, the make the incorporation of the bias and bias uncertainty in the criticality is used. The lattice pitch To bias uncertainty from the trends in the range of interest trend on most limiting bias and on Table A.4.1.5 along with the bias, as predicted with the for Wcstinghouse fuel is tabulated lattice pitch. A-31 NET- 300067-01 Rev 0

Table A.4.1.5: Bias as Predicted Using the Trend in the Bias as a Function of Pitch Fuel Type Pitch (cm) Bias 14x14 1.412 0.0026 15x15 1.430 0.0025 16x16 1.232 0.0028 17x17 1.260 0.0028 The bias as a function of pin diameter decreases with increasing pin diameter and for the smallest pin diameter (Westinghouse 17X 17 OFA fuel, 0.360 inches) is 0.0026. The bias for the 15x 15 pin diameter (0.422 inches) is 0.0023. Thus the bias as a function of pitch is always more limiting.

The bias as a function of enrichment is given in Table A.4.1.6. As can be seen by comparing Table A.4.1.5 and A.4.1.6, the lattice pitch bias is more limiting at enrichments consistent with reloads.

However, for low enriched fuel, the limiting bias is a result of the enrichment trend.

Table A.4.1.6: Bias as Predicted Using the Trend in the Bias as a Function of Enrichment Enrichment (wt% U-235) Bias 5 0.0019 4 0.0022 3 0.0024 2 0.0027 1.6 0.0028 The biases as a function of soluble boron or areal density decrease with increasing boron and are not the most limiting biases at zero boron.

The spectrum, as measured by the EALF in the pool with no soluble boron, is generally between 0.2 and 0.4 eV. The bias increases as the spectrum hardens and the bias at 0.4 eV is 0.0029. This is the most limiting bias. For heavily borated cases, the EALF can get almost as high as 0.6 eV. At 0.6 cV the bias is 0.0037. For this criticality analysis, a bias of 0.0029 will be used for all EALF less than 0.4 (limiting cases for no boron credit) and 0.0037 for EALF values between 0.4 and 0.6 eV (heavily borated cases). The maximum uncertainty for any trend is 0.0050, which comes from the areal density analysis.

This uncertainty is the largest, since it is the smallest sample (only 28 cases), but in order to make the analysis conservative, 0.0050 is selected for the uncertainty in the bias.

The uncertainty of the set as a whole is 0.0049. The uncertainty for the trended analysis is generally less, (not so for the areal density, due to the small sample size) since taking advantage of the trend reduces the difference between the experimental value and the predicted value.

A.4.1.7 SubcriticalMargin NET- 300067-01 Rev 0 A-32

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 0 when the analysis is performed with unborated water. This is actually saying the subcritical margin is contained in the uncredited soluble boron. To make sure there is sufficient soluble boron, analysis is also performed with soluble boron and a subcritical margin of 5% in k is required. For dry storage racks analyzed with optimum moderation, the subcritical margin is 2% and 5% with full moderation. In the analysis of 236 critical experiments, which generously cover the range of expected conditions, the lowest calculated k was 0.9931. This supports the position that the subcritical margin is more than sufficient.

A.4.1.8 Area of Applicability (BenchmarkApplicability)

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.4.1.7 is provided below.

Table A.4.1.7: Area of Applicability (Benchmark Applicability)

Parameter Range Comments Fissionable Material/Physical U0 2 The fuel material is the same as in the Form benchmark experiments Enrichment (wt% U-235) 2.35 to 4.74 The first core enrichments require extrapolation of the bias to lower enrichments. The predicted difference in bias from the lowest measured data to the lowest enrichment manufactured is only 0.01% in k. An error in the slope used for extrapolation would only produce errors in the hundredths of a percent.

The extrapolated bias is less than the limiting bias, Therefore extrapolation of the bias to lower enrichments is justi fied.

An extrapolation from 4.74 to 5 wt%

will also be needed but in this direction the bias is decreasing so the data is adequate for this extrapolation.

Spectrum Expected range in applications:

EALF (eV) 0.0605 to 0.1 to 0.6 eV 0.8485 The experiments cover the entire expected range of limiting conditions.

For intended applications, dry conditions are never limiting.

NET- 300067-01 Rev 0 A-33

Parameter Range Comments Lattice Characteristics Type Square Hex lattices have been excluded.

Pin Pitch (cm) 1.075 to 2.54 Expected range of 1.2 to 1.6 cm.

The expected range of all fuel types, including both PWR and BWR fuel is covered.

Assembly Spacing in Racks This covers all spacing. Neutron Distance between Assemblies 0 to 15.4 transport through larger than 15.4 cm (cm) has a small effect on k. Note that the 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 Spent fuel storage designs are within Boron Areal Density (B-10 0 to 0.067 this range.

g/em 2)

Absorbers All designs are within this range.

Soluble Boron Concentration 0 to 5030 ppm Absorbers Cd bearing experiments showed no Cd (for Ag-In-Cd rods) Cd Absorber dependence on the number of rods.

anels Credit for these rods is acceptable.

Reflector Most rack analysis will assume Experiments included water Reflectors optimum reflection or in the case of and steel adequately casks, steel walls. Both of these covered assumptions are adequately covered.

Temperature Room The criticality calculations are Temperature performed with the fuel at low temperatures. No significant Doppler feedback.

Moderating material Water The moderator in all benchmark experiments is water, therefore water as a moderating material is covered NET- 300067-01 Rev 0 A-34

A. 4.1.9 Summary of Fresh U0 2 Laboratory CriticalExperiment Analysis This validation follows the guidance of NUREG/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.

236 U0 2 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 k is small, The results of the criticality analysis were tested for trends against 6 different parameters important to reactivity. It was conservatively assumed that any trend found was significant. Using the trends, the most limiting bias and bias uncertainty is determined to be 0.0029 for the bias for EALF up to 0.4 eV and 0.0037 for EALF's in the range of 0.4 and 0.6 eV and the uncertainty is 0.0050 for all analyses.

While this validation is intended to cover all LWR fuel racks, the specific area of applicability is found in Table A.4.1.7.

A.4.1.10 HTC andMOX CriticalExperiments Burned fuel contains a low concentration of plutonium (about I 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.4.1.10.1 HTC Critical Experiments All 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 em and different quantities of pins were used to change the critical height. An 18 th 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 k of the Phase I cases was 0.99910.

Phase 2 [14] consisted of 20 cases where gadolinium 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 k of the gadolinium cases was 0.99815 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 k of these 26 cases was 0.99890.

NET- 300067-01 Rev 0 A-35

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 analysis, since lead reflectors are not common in spent fuel pools. The 33 steel reflector experiments were included.

The average k of these cases was 0.99858.

References 13 through 16 provided all the details for the analysis. The modeling was straight forward.

The references gave a simple model and a detailed model. The model created for this work followed the detailed 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.

Tables A.4.1.8 through A.4.1.12 present the results of the analysis. A statistical analysis of the 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 k is 0.99878 and the uncertainty is 0.00590. This makes the bias 0.00122. Since all 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 (performed consistent with NUREG/CR-6698) on EALF yielded the following function:

k(EALF) = 0.999538 - 0.00548

• EALF The units for EALF are eV. The uncertainty about the trending k is 0.0057 in k. Figure A.4.1.8 shows the results of the HTC analysis.

Table A.4.1.8: HTC Phase I Results Case No. kff Monte EALF Pitch Carlo (eV) (cm)

Sigma

. 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- 300067-01 Rev 0 A-36

Table A.4.1.9: HTC Phase 2a, Gadolinium Solutions, Results Case No. kIr Monte EALF Pitch Gadolinium Carlo (eV) (cm) Concentration Sigma (g/)

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 0.00018 0.26756 1.3 0.1005 5 0.99784 0.00018 0.26333 1.3 0.1005 6 0.99683 0.00018 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 I1 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 0.10631 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- 300067-01 Rev 0 A-37

Table A.4.1.10: HTC Phase 2b, Boron Solutions, Results Case No. kef Monte EALF Pitch Boron Carlo (eV) (cm) Concentration Sigma (g/)

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- 300067-01 Rev 0 A-38

Table A.4.1.11: HTC Phase 3 Results - Water Reflected Assemblies (1.6 cm pin pitch)

Case No. kff Monte EALF Absorber Assembly Carlo (eV) Box Separation Sigma Material (er) 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 0.00018 0.12391 Borated SS 3 5 0.99699 0.00018 0.13503 Borated SS l 6 0.99987 0.00019 0.12974 Boral 0 7 0.99614 0.00019 0.12866 Cd 2 8 1.00381 0.00018 0.13904 Cd 0 9 0.99646 0.00017 0.13345 Cd I 10 0.99672 0.00018 0.12952 Cd 1.5 11 0.99571 0.00019 0.13726 Cd 0.5 12 0.99901 0.00017 0.11277 none 18 13 0.99915 0.00018 0.11167 none 14.5 14 0.99934 0.00018 0.11183 none 11 15 0.99910 0.00019 0.11093 none 10 16 0.99961 0.00019 0.11030 none 9 17 0.99930 0.00018 0.10842 none 8 18 0.99980 0.00017 0.10656 nonc 6 19 1.00016 0.00018 0.10421 none 4 20 1.00044 0.00018 0.10206 none 4 21 0.99976 0.00018 0.10470 none 2 22 1.00047 0.00019 0.10714 none 1 23 0.99893 0.00018 0.11506 none 0 24 0.99949 0.00020 0.15073 none 0 25 0.99996 0.00018 0.12672 none 4 26 0.99937 0.00020 0.11550 none 10 NET- 300067-01 Rev 0 A-39

Table A.4.1.12: HTC Phase 4 Results - Steel Reflected Assemblies (1.6 cm pin pitch)

Case No. lkfe 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 Boral 0 0.0 13 0.99653 0.00018 0.13386 Boral 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- 300067-01 Rev 0 A-40

1.006 1.004, 1.002 +

S0.998

• 0.996 0.994 0.992 0.99 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.4.1.8: k~f as a Function of the Energy of the Average Lethargy Causing Fission for the HTC Experiments A.4.1.10.2 MOX Critical Experiments The selection of the MOX critical experiments was limited to the low enriched MOX lattice critical experiments. All 63 of the low enriched MOX pin critical experiments documented in the OECD handbook [17] were utilized. The actual input decks were initiated from available decks found in NUREG/CR-6102 [18] and OECD [17]. The decks were modified to update to the new cross-section library and changes in the SCALE input format.

Table A.4. 1.13 presents the results of the 63 selected MOX critical experiments. The Reference column has the evaluation number from the OECD benchmark handbook [17]. For example, OECD-7 refers to the OECD case MIX-COMP-THERM-07.

Trends were investigated as a function of EALF, plutonium content, and the Am-24 I/U-238 ratio. As the spectrum hardens (higher EALF), there is a small trend to higher k. With more plutonium content, k increases. This is seen in Figure A.4.1.9. This means that the more limiting bias comes from the zero plutonium cases (fresh U0 2). This is confirmed by the average calculated k of the HTC critical experiments being higher than the average calculated k for the fresh U0 2 cases. The average uncertainty NET- 300067-01 Rev 0 A-41

weighted k of the fresh U0 2 criticals is 0.9978. The average uncertainty weighted k for the HTC criticals is 0.9988. The average uncertainty weighted k of the MOX experiments is 0.9984.

The change in k with cooling time is dominated by the reactivity of the decay of Pu-241 to Am-24 1. By plotting k versus the Am-241/U-238 ratio, it is possible to determine if the bias should be changed for cooling. Figure A.4.1.10 shows that with increasing Am-241 content, the calculated k of the critical experiments increases. This observation shows that the zero cooling time bias conservatively covers the cooling time.

Table A.4.1.13: Results of MOX Critical Benchmarks (SCALE 6.1.2, ENDF/B-VII)

Case ID Reference kff sigma EALF Pu Pu Am241/U238

__________cV ASL 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 7.55 E-05 105array OECD-7 0,9960 0.00025 0.1377 2.00 16 7.55E-05 105bI 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 I143arra OECD-7 0.9980 0.00026 0.1166 2.00 16 8.13E-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 OECJ)-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 k 1mct009 OECD-9 0.9994 0.00024 0.5169 1.50 8 1.06E-05 k2mct009f OECD-9 0.9941 0.00027 0.2943 1.50 8 9.77E-06 k3mctOO9 OECD-9 0.9934 0.00024 0.1528 1.50 8 8.96E-06 K4mctOO9 OECD-9 0.9921 0.00024 0.1155 1.50 8 8.96E-06 K5mctO09 OECD-9 0.9925 0.00021 0.0947 1.50 8 8.96E-06 K6mctOO9 OECD-9 0.9937 0.00024 0.0905 1.50 8 9.77E-06 omct6l OECD-6 0.9954 0.00026 0.3570 2.00 8 2.24E-05 omct62 OECD-6 0.9990 0.00029 0.1885 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 mct8c I 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 NET- 300067-01 Rev 0 A-42

Case lID Reference lff sigma EALF Pu Pu Am241IU238 Case ED_____

Reference sigma(eV) wt% 240%

mct8cal OECD-8 0.9967 0.00025 0.1375 2.00 24 8.59E-05 mct8cb I 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 mctcb2 OECD-8 0.9937 0.00024 0.1385 2.00 24 8.59E-05 mctcb4 OECD-8 0.9942 0.00026 0.1378 2.00 24 8.59E-05 mixo251k OECD-5 1.0011 0.00032 0.3732 4.00 18 1.59E-04 mixo252k OECD-5 0.9985 0.00027 0.2476 4.00 18 1.59E-04 mixo253k OECD-5 1.0044 0.00027 0.1712 4.00 18 1.59E-04 mixo254k OECD-5 1.0004 0.00029 0.1425 4.00 18 1.59E-04 mixo255k OECD-5 1.0034 0.00028 0.1058 4.00 18 1.59E-04 mixo256k OECD-5 1.0023 0.00024 0.0917 4.00 18 1.59E-04 mixo257k OECD-5 1.0036 0.00024 0.0875 4.00 18 1.59E-04 saxtn 104 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 0.1820 6.60 8.6 8.43 E-05 saxtn792 OECD-3 0.99951 0.00031 0.1505 6.60 8.6 8.43E-05 Saxton52 OECD-3 0.99977 0.00028 0.8517 6.60 8.6 8.43E-05 Saxton56 OECD-3 1.00018 0.0003 0.5177 6.60 8.6 8.43E-05 tcal OECD-4 0.99572 0.00027 0.1418 3.01 22 1.04E-04 tca 10 OECD-4 0.9988 0.00024 0.0792 3.01 22 9.31E-05 tcal I OECD-4 0.99886 0.00023 0.0788 3.01 22 2.06E-04 tca2 OECD-4 0.9964 0.0003 0.1409 3.01 22 1.99E-04 tca3 OECD-4 0.99665 0.00028 0.1403 3.01 22 2.96E-04 tca4 OECD-4 0.99644 0.00026 0.1172 3.01 22 9.88E-05 tca5 OECD-4 0.9974 0.00027 0.1167 3.01 22 2.02E-04 tca6 OECD-4 0.99848 0.00025 0.1156 3.01 22 3.90E-04 t=a7 OECD-4 0.99753 0.00025 0.0917 3.01 22 8.88E-05 tca8 OECD-4 0.99801 0.00025 0.0913 3.01 22 2.03E-04 tca9 OECD-4 0.99864 0.00025 0.0909 3.01 22 3.02E-04 NET- 300067-01 Rev 0 A-43

1 IK00 4 1 1 I

0-992 0.99 1.0 2.0 4.0* 7.0 8.0 Pu WM 4 ff cfi n of thePltnu uet F'gire A.4.1 .9 Prcdict 8.0 NFET 300067.01 Rev 0 A-44 A-44

1.0060 1.0040 1.0020 4.

0O9940 0.99204 3 35 ~ 4 4 0 .0 . E3 .~

04 2 .0E Q42.SE-04 .OE_04 5

09~o .0E÷(Jo -0E-0 5 1.0E434 1 .5E .. -E0

.E0 .E0 Ratio OfAmn-24 to U283'E0 ANA-2 4 1 Conten Fig~~.4. 10 redcted k. as a Funct 0 1,Of the NET. 300067-01 Rev 0 A-45

A.4.2 Depletion Reactivity Bias and Uncertainty(EPRI Benchmark Analysis)

Measured data from power plants can be used to validate the delta-k of depletion,. EPRI has used power distribution measurements to infer the change in reactivity due to burnup using 680 flux maps taken over 44 cycles from 4 different PWRs [10]. The reactivity change is then captured as a series of benchmarks that can be analyzed using the methods selected for the criticality analysis. The analysis for the benchmarks for utilization in criticality analysis is demonstrated for SCALE 6.1 and the 238-group ENDF/B-VII library in Reference 11, but this work uses a slight modification which requires reanalysis of the benchmarks.

Due to the axial burnup profiles, this effort requires the generation of atom density sets at a large number of different burnups. Reference 11, which demonstrated how to use the EPRI benchmarks, required only 6 burnups (10, 20, 30, 40, 50, and 60 GWd/T). In Reference 11, each burnup atom density set was generated by a separate burn from the initial fresh fuel conditions. This is consistent with the design of TRITON. Reference 11 restarted TRITON to produce the two cooling times needed after the initial 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> cooling time. The depletion performed for this work matched that used in Reference I1, however, the isotopes used in the final criticality analysis were reduced from 388 to 185 isotopes and the decay was performed outside of SCALE. The reduction in isotopes is needed for this analysis because the axial burnup profiles multiply the number of isotopes needed for processing by 9. With this multiplier, the processing time becomes large, making the reduction in isotopes significant to the total effort.

For this work, the same steps were followed as in Reference 11, so the reader is referred to Reference I I for more details regarding this process. Tables A.4.2. 1, A.4.2.2, and A.4.2.3 show the results of the benchmark analysis using the methods selected for this analysis.

NET- 300067-01 Rev 0 A-46

Table A.4.2.1: EPRI Benchmark Results for 10O-hour Cooling

.i. uenricnmenE depiction -0.0008 -0.0017 -0.0024 -0,0037 -0.0040 -040044 5.00% enrichment

-0.0001 -0.0003 -0.0005 -0.0012 -0.0014 -0.0018 depletion 4.25% enrichment depletion 0.0002 -0.0004 -0.0010 -0.0018 -0.0026 -0.0029 off-nominal pin depiction

-0.0008 -0.0016 -0.0023 -0,0029 -0.0037 -0.0046 20 WABA depletion 0.0000 0.0003 -0.0005 -0.0014 -0.0018 -0.0025 104 IFHA depletion 0.0010 0.0007 -0.0004 -0.0020 -0.0027 -0.0042 104 IFBA, 20 WASA depletion 0.0007 0.0011 0.0000 -0.0013 -0.0018 -0.0032 high boron depletion =

1500 ppm

-0.0003 -0.0006 -0.0011 -0.0017 -0.0018 -0.0024 branch to hot rack 338.7K -0.0004 -0.0007 -0.0008 -0.0017 -0.0019 -0.0025 branch to rack boron =

1500 ppm -0.0009 -0.0019 -0.0027 -0.0036 -0.0044 -0.0049 high power density

-0.0002 -0.0012 -0.0016 -0.0022 -0.0026 -00032 depletion Table A.4.2.2: EPRI Benchmark Results for 5-year Cooling 3.25% enrichment -0.0010 -0.0017 -0.0026 -0.0026 -0.0028 depiction -0.0002 5.00% enrichment depletion 0.0006 0.0006 0.0000 -0.0005 -0.0008 -0.0011 4.25% enrichment 0.0008 0.0000 -0.0005 -0.0012 -0.0016 -0.0021 depletion off-nominal pin depletion -0.0003 -0.0008 -0.0018 -0.0023 -0.0028 -0.0033 20 WABA depietion 0.0006 0.0006 0.0001 -0.0005 -0.0010 -0.0014 104 IFBA depletion 0.0019 0.0009 0.0001 -0.0011 -0.0017 -0.0029 104 IFBA, 20 WABA depletion 0.0018 0.0019 0.0007 -0.0001 -0.0012 -0.0020 high boron depletion =

1500 ppm 0.0005 0.0003 -0.0001 -0.0010 -0.0011 -0.0014 branch to hot rack 338.7K 0,0002 0.0001 -0.0005 -0.0009 -0.0009 -0.0014 branch to rack boron =

1500 ppm 0.0000 -0.0012 -0.0018 -0.0029 -0.0033 -0.0038 high power density depiction 0.0004 -0.0001 -0.0005 -0,0009 -0.0016 -0.0016 NET- 300067-01 Rev 0 A-47

Table A.4.2.3: EPRI Benchmark Results for 15-year Cooling 3.25% enrichment 0.0005 -0.0009 -00022 -0.0028 -0.0033 -0.0033 depletion 5.00% enrichment depletion 0.0009 0.0009 -0.0001 -0.0005 -0.0010 -0.0012 4.deenrichment 0.0012 0,0005 -0.0006 -040013 -0.0018 -0.0021 depletion off-nominal pin 0.0004 -0.0009 -0.0014 -0.0026 -0.0031 -0.0034 depletion 20 WABA depletion 0.0012 0.0013 0.0002 -0.0008 -0.0012 -0.0019 104 IFBA depletion 0.0023 0.0014 0.0001 -0.0009 -0.0019 -0.0028 104 IFBA, 20 WABA 00025 0.0024 0.0009 -0.0002 -0.0009 -0.0020 depiction h borondepleton=

1gh 0.0009 0.0004 -0.0002 -0.0012 -0.0012 -0.0018 1500 ppm branch to hot rack = 0.0003 -0.0001 -0.0007 -0M0010 -0U0012 -0U0016 338.7K branch torackboron= 0.0002 -0.0012 -0.0021 -0.0028 -0.0037 -0.0038 1500 ppm hlghpowerdensity 0.0009 0.0003 -0.0007 -0.0013 -0.0014 -0.0019 depletion The maximum bias from all the cooling times is 0.0025. This bias is from Case 7 at 10 GWd/T burnup.

However, a number of the 1 lattice conditions given on Tables A.4.2.1 to A.4.2.3 can occur simultaneously. To find the highest bias, one starts with the bias from Case 3 and adds the delta biases between Case 3 and the condition of interest. The most limiting condition is 5 years cooling and 30 GWd/T. The values for this case will be used to show how a rack up of biases is performed. The Case 3 bias at 30 GWd/T and 5 years cooling is -0.0005 (see Table A.4.2.2). Since the fuel can be a higher or lower enrichment, the difference between the bias for Case 2 and Case 3 (+0.0005) is added. That is the base -0.0005 plus +0.0005 resulting in a corrected bias of zero. Now we correct for the pin diameter.

Case 4 uses a smaller pin diameter than Case 3. The 15x 15 fuel pin diameter is larger than the 17X 17 fuel by about the same amount as Case 4 is smaller than standard 17X 17 fuel. The difference in the bias between Case 4 and Case 3 is added to further correct the bias. That is -0.0005 - (-0.0018) or a +0.0013 correction, making the new corrected bias +0.0013. Note that it is assumed here that the effect is due to the pin diameter rather than the spectrum. 15X 15 and 17X 17 fuel have about the same spectrum, so this correction may not be necessary if it is actually due to the spectral difference rather than the pin diameter.

Case 7 covers both IFBAs and WABAs. The bias is more positive by 0.0007 -(-0.0005) = 0.0012.

Adding this effect, the corrected bias is now 0.0025. Finally, the depletion at a higher ppm shows a more positive bias (-0.0001- (-0.0005)). Adding the effect of a higher ppm depletion raises the total corrected bias to 0.0029. (Note that the depletions are performed at less than 1500 ppm, so this effect is over estimated.) The higher power and the higher rack temperature did not change the bias from the reference, Case 3 and adding absorbers to the pool made the bias less positive. The final selected bias is rounded up to 0.003.

A similar rackup of biases at 20 GWd/t produces a higher bias, but that is due to the depletion of the burnable absorbers. The depletion reactivity, as defined by the EPRI benchmarks, includes the change in reactivity from the burnable absorbers. From reviewing Tables A.4.2. i through A.4.2.3, it is clear that NFT- 300067-01 Rev 0 A-48

there is an over-prediction of the depletion rate of the burnable absorbers. Although burnable absorbers are included in the depletion analysis in order to harden the spectrum, the final application does not credit any absorption from burnable absorbers. Table 6-1 of Reference 11 shows that at 20 GWd/T the residual boron worth is greater than 0.008, which means the assumption to ignore the residual boron is much larger than the small positive biases from Cases 5, 6, and 7. Ignoring the positive biases for Cases 5, 6, and 7 at 20 GWdf I and below makes the 30 GWd/T, 15 years cooled case most limiting.

The depletion uncertainty is the uncertainty in the EPRI benchmarks which are reported in Table C-I of Reference 10. The benchmark uncertainty is a function of the specific power. With radial peaking, the specific power of this analysis is between the Case 10 and 11 values. The depletion uncertainty of 0.0064 from the highest power, EPRI Case 11. is used for this analysis.

A.4.3 Extended ISG-8 Validation The validation of the analysis only needs validation of the initial condition (Fresh U0 2 critical experiments) and validation of the depletion reactivity (EPRI benchmarks). However, an alternate approach (the Extended ISG-8 approach) can be taken, where the isotopic content is validated (validation of the depletion calculations), followed by validation of the reactivity worth of the isotopes (MOX and HTC critical benchmarks used for the most important isotopes and conservative use of TSUNAMI analysis for the rest of the isotopes). This validation uses the most limiting of the two approaches.

The validation of the isotopic content has three major steps. The first step is to determine what chemical assays are appropriate. The second step is the analysis of the chemical assays using the depletion methods that will be used for the final criticality analysis of the application. The third step is to convert the results of the prediction of the isotopic content of the chemical assays to a delta-k bias and uncertainty for the final application. This third step is often called the "direct difference" approach. [19]

A.4.3.1 Selection of Assays ORNtL reviewed the chemical assays available and made selections in ORNL/TM-2010/44 [20] and then in NUREG/CR-7108 [19]. This validation generally follows the ORNL selections. ORNL eliminated several sets of chemical assays. Yankee Rowe assays were eliminated due to lack of information on the control rod locations. Many in the industry consider Yankee Rowe assays high quality assays [21] but this validation follows the ORNL recommendation. Mihama samples were eliminated with the statement, "Future use of this set of data was not recommended because of the unexplained variation in the results, which may indicate problems related to the radiochemical analysis of samples or the incomplete documentation of operating data." Soluble boron data is not available for the Mihama samples but previous studies [22] only eliminated one sample. This analysis follows the ORNL recommendation to not use Mihama data. The OECD/NEA has a web site, http://www.oecd-nea.org/sfcompo/, which summarizes chemical assay data. At the time of the writing of this validation (11/1/2013), the only PWR data on the sfcompo website not covered by ORNL are two chemical assays from the Genkai- I reactor in Japan. These two assays do not include fission products and there is no information provided as to which fuel rod in the assembly was assayed. The Genkai-1 assay data is not used in this validation.

NET- 300067-01 Rev 0 A-49

ORNL/TM-2010/44 analyzed 118 assays but NUREG/CR-7108 did not use 18 of these. They were:

I. 3 G6sgen fuel samples that are proprietary.

2. 12 Obrigheim assays taken from rods on the periphery. Obrigheim had used some MOX assemblies and the location of the MOX assemblies was not known, so the peripheral pin could have been next to MOX fuel.

3. 3 Takahama assays were excluded due to their close proximity to the edge of the active core.

The flux gradient is strong in this region and this effect is not relevant to the criticality analysis this effort supports, Like NUREG/CR-7108., this validation does not use these 18 assays.

In addition to the assays selected for NUREG/CR-7108, this validation will use 5 chemical assays performed on fuel from the Spanish reactor, Vandell6s I1. These samples have been analyzed by ORNL.

The ORNL report is NUREG/CR-7013 [23]. The reason these assays were not included in NUREG/CR-7108 was because the data was not available in time to be included in the analysis. These samples include fission products and are at a higher burnup and enrichment compared to the rest of the assays used in this validation. Note that there are actually six chemical assays in the Vandell6s 11 set. However, the sample from rod WZtR 160 does not include accurate uranium data [23]. When using the direct difference approach, it is inappropriate to use samples missing high worth isotopes that vary with irradiation, such as U-235 in this case.

This validation does not use the 11 TMI chemical assays performed on Assembly NJ05YU. These assays were done by ANL on a single pin. The 8 other chemical assays from TMI were performed by GE on three rods from Assembly NJ070G. These assays do not have the problems found with the NJ05YU assembly. Figures A.4.3. l, A.4.3.2, and A.4.3.3 show the change in data as the assays go up the same pin axially. Except for the grids, which are shown as red lines, there is nothing in the assembly that should produce non-smooth results, but the figures show strikingly irregular behavior. The data from TMI Assembly NJ05YU will not be used due to flaws in the measured data.

Measured Burnup Up Pin H6 in Assembly NJOSYU 60 58 56 54 E 48 46 4) 0 100 150 2001 250 300 350 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.1h Measured Burnup for Pin H6 in Assembly NJOSYU NET- 300067-01 Rev 0 A-50

Measured U-235 Up Pin H6 in Assembly NJ05YU 9.thJE-03 8.50E03 8.0011-03 7.50E-03 L.OOt-03 1ý 6.50E-0.1 6.00iE03 5.50E-03 5.OOt-03 150 200 250 350 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.2: Measured U-235 Content for Pin H6 in Assembly NJO5YU Measured Pu-239 Up Pin H6 in Assembly NJ05YU 5.60E-03 S5.0t-03 5.40F-03 5,320E03 5-ODF-03 4.90E-03 4.80 E-03 0 50 100 150 200 250 300 3S0 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.3: Measured Pu-239 Content for Pin H6 in Assembly NJ05YU The 8 chemical assays performed on the TMI assembly NJ070G were also reviewed for obvious errors in the experimental data. Figures A.4.3.4 through A.4.3.12 show the change in burnup, the change in U-235 content and the change in Pu-239 content going up assembly NJ070G for Pins 01, 012, and 013. An NET- 300067-01 Rev 0 A-51

inspection of the figures shows that the trend in burnup and U-235 content is basically the same for the three pins. However, the Pu-239 content of pin 01 does not follow the trend observed in the other two pins. The burnup at 197 cm from the bottom is significantly higher than at 39 cm (about 15% more), but the Pu-239 content of pin 01 only increases 2.6%. This is not expected and does not follow the increase observed in the other pins (about 9%). It is concluded from review of the data that there is a significant error in the Pu-239 content and this assay will not be used. (TMI Assembly NJ070G pin 01 height 197.1 sample will not be used.)

Measured Burnup in Pin 01 of Assembly NJ070G 30 29 28 27 26 25 E 24 23I 22 0 50 1W0 150 200 250 300 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.4: Measured Burnup for Pin Ol in Assembly NJ070G Measured Burnup in Pin 012 of Assembly NJ070G 30 29

• 28.

i 27 E26 .

• 25 24 23 22 0 50 100 150 200 250 300 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.5: Measured Burnup for Pin 012 in Assembly NJ070G NET- 300067-01 Rev 0 A-52

Measured Burnup in Pin 013 of Assembly NJ070G 30 29 28 27

- 26 25 24 23 22 50 100 150 200 250 300 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.6: Measured Burnup for Pin 013 in Assembly NJ070G Measured U-235 in Pin 01 of Assembly NJ070G 0.025

" 0.023 0.021

-0.019 N

0.015 1 1 0 50 100 150 200 250 300 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.7: Measured U-235 Content for Pin O1 in Assembly NJ070G Measured U-235 in Pin 012 of Assembly NJ070G 0.025 0.023

_ 0.021 0S.019 en

- 0.017 0.015 0 50 100 150 200 250 300 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.8: Measured U-235 Content for Pin 012 in Assembly NJ070G NET- 300067-01 Rev 0 A-53

Measured U-235 in Pin 013 of Assembly NJ070G 0.025 oo23 i -

0.023

• 0.017 0.015 I 0 50 100 150 200 250 300 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.9: Measured U-235 Content for Pin 013 in Assembly NJ070G Measured Pu-239 in Pin 01 of Assembly NJ070G 0.0062

, 0.006 0-0058 0.0056 0.0054

' 0.0052

. 0.005 0.0048 0 50 100 1w0 20 250 300 Height From Bottom Of Active Fuel (cm)

Figure A.4.3.10: Measured Pu-239 Content for Pin O1 in Assembly NJ070G Measured Pu-239 in Pin 012 of Assembly NJ070G 0.0064 0.0062 S 0.006 1

• 0.0058 S0.0056 m 0.0054 0.0052

. 0.005 0.0048 0 50 100 150 200 250 300 Height From Bottom Of Active Fuel (cm)

NET- 300067-01 Rev 0 A-54

Figure A.4.3.11: Measured Pu-239 Content for Pin 012 in Assembly NJ070G Measured Pu-239 in Pin 013 of Assembly NJ070G 0.006

' 0.0056 0.0054 W 0.0052 0.005 0.0048 -

0 50 100 150 200 250 300 Height From Bottom Of Active Fuel (cmi Figure A.4.3.12: Measured Pu-239 Content for Pin 013 in Assembly NJ070G Sample N-9C-J for H. B. Robinson is eliminated since it is taken from under an Inconel grid. The strong thermal absorption of Inconel makes the 2D methods used to produce the correct assembly average reactivity inadequate for this sample.

The total set used for this analysis consists of 92 chemical assays from 10 different plants. Of these 92 samples, about 25 have significant fission products.

A.4.3.2 Analysis of the ChemicalAssays In general, the information for chemical assay analysis is taken from ORNL/TM-2010/44 [20] and NUREG/CR-7013 [23]. The computer decks for Reference 20 reports are on line at htLp:Ilscale.ornl.gov/spent fuel isotopic char input.shtml. The computer decks for Reference 23 are in the report. The moderator temperatures, soluble boron concentrations, power densities, and fuel temperatures as a function of time used for this validation were obtained from the ORNL references

[20,23].

For this analysis, KENO-V.a is used for the collapsing of the cross-sections to create the one-group cross-sections for depletion. The ORNL effort used NEWT. In general, the input needed for these two modeling approaches is similar; however, each input deck used in this analysis is different than that produced by ORNL due to the KENO/NEWT change. One other change from the ORNL decks is related to the burnup step sizes. Historically, the industry uses a small step (0.15 GWd/T) for getting equilibrium Xe in the model. The 0.15 GWd/T step is used for this analysis and production runs. Steps used are:

0.15, 0.35, .5, .5, .5, 1, 1, 1, 1, 1, 1, 1, 1, and 2 GWd/T for all burnups thereafter. This step size will be used for all chemical assays to match the production runs. Some variations in these steps for the chemical assays have been taken to match the down times.

The following sections discuss each chemical assay set.

NET- 300067-01 Rev 0 A-55

A.4.3.2.1 TRINO VERCELLESE The original references for Trino have been reviewed to check the ORNL modeling. Reference 24 gives the assembly drawing reproduced in ORNL reports. References 25 and 26 provide details on the one cycle and two cycle irradiation experiments, respectively. Unfortunately, information is missing relating to the assembly design and pitch.

First, the reports state that there is an empty location in the assembly center for the instrumentation tube.

No information is given about the instrumentation tube and the drawings suggest it is removed during operation. This is a Westinghouse designed plant and US Westinghouse plants have a tube in the center of the assembly for the instrumentation to traverse. ORNL modeled the hole as empty (no tube) and that modeling is followed in this analysis.

Second, the assembly drawing provides little detail on the can around the assemblies. The ORNL model includes a 0.6 mm stainless steel can inside the standard fuel pin pitch. There is insufficient information to support the positioning of the can, so for the modeling in this analysis, the can is placed outside the standard pin pitch. The ORNL model sets up the separate pitch for the cruciform fuel followers. For this analysis, the cruciform fuel pins are accurately modeled, but the pin pitch is maintained to be the same as the pin pitch for the non-cruciform pins. This creates a full empty pin cell at the ends of the cruciform pins rather than the partial empty cell created by the more accurate ORNL model. Both models have the same fuel, clad and water volumes. Since the sampled pins are not next to the cruciform fuel followers, this model simplification is appropriate. There is no information on a can surrounding the cruciform fuel followers. ORNL's model contains a can and models the can around the fuel assembly to follow the irregular boundary. In the model used for this analysis, the can is a simple square around the entire model. The can modeling is not accurate for either ORNL or this effort. The can is perforated. The amount of perforation is not given. Figure 3 of Reference 25 shows a picture of the assembly. From this picture it is assumed that 50% of the steel is removed. ORNL assumed the can was not perforated. This analysis assumes 50% steel and 50% water. The water has soluble boron of varying concentrations, and rather than create a can mixture with 50% water, the thickness of the steel is reduced from 0.6 mm to 0.3 mm.

Third, the assembly separation is not given. Tables in the references give the assembly "side of square cross-section" as 20 cm. This was assumed by ORNL as the assembly pitch. The assembly drawing may suggest an assembly pitch of 20.1 cm. This analysis uses the same 20 cm selected by ORNL.

The ORNL model includes an empty cell with water at the corner opposite from the main cruciform fuel followers. This is incorrect. That corner should have a fuel follower pin. That error is corrected for this analysis.

ORNI. did not use all the chemical assay data available from Trino. The pins from the corner were not included. The references report these pins as "perturbed spectrum." However, these corner pins are at the corner away from the cruciform followers and may actually not have much of a "perturbed spectrum."

Since the data on assembly separation is weak, these pins are also ignored for this work.

There are 31 chemical assays that are used from the Trino data. They come from three assemblies that were in the Trino core for one cycle and one assembly that was in the core two cycles.

ORNL used reflective boundary conditions. The assemblies are only symmetric on the diagonal, so neither reflective nor periodic boundary conditions match the actual assembly. This work uses periodic boundary conditions, since the cruciform fuel rod assemblies are one row thick. A reflective boundary NET- 300067-01 Rev 0 A-56

condition makes them effectively 2 rows thick. The periodic boundary condition is not precise, since the cruciform rods are on different halves of the assembly, but this error is less significant than doubling up the thickness.

One assembly, 509-104, was on the edge of the core. ORNL modeled this assembly with the reflector.

ORNL points out in Reference 20 that since the sampled pin is II rows from the periphery, "the core periphery has negligible effects on the measured rod." For this analysis the reflector is not modeled.

Further, the cut out section of this assembly actually has a stainless steel filler. No details for this filler were found in the references but ORNL has assumed it is rods. For this work, the filler is assumed to be the standard cruciform fuel rods, since the sampled rod is closer to a standard cruciform fuel insert. In fact, the pin location has been adjusted in this model to preserve the distance from the cruciform fuel inserts.

The ORNL fuel densities were confirmed using the original references. The footnotes in Reference 20, Table 49 are confusing, but the final values for the densities are the same as independently confirmed.

The fuel density in the cruciform rods seems low, because no pellet diameter was known, so it was assumed that the pellet extended to the clad ID (no gap). The clad ID was an assumed value based on the control rod clad thickness (see Reference 20).

Note that the ORNL deck for assembly 509-032, pin El 1, Plane 7, has an error in the fuel temperatures.

The impact of that error is expected to be small, but was corrected for this analysis.

Table A.4.3.1 shows the percent deviation between the predicted and measured content of the measured isotopes for all the Trino samples.

Table A.4.3.1: Trino Predicted Minus Measured % Deviations (Part 1 of 4) 104- 032- 032- 032- 032- 032- 032- 032-MIIP7 EIIPI EIlP4 EIIP7 EIIP9 H9P4 H9P7 H9P9 Enrichment 3.897 3.13 3.13 3.13 3.13 3.13 3.13 3.13 Burnup 12.04 7.24 15.38 15.9 11.53 16.56 17.45 12.37 U-234 15.20 -6.64 34.67 28.95 27.29 U-235 1,15 2.73 3.20 4.80 1.28 0.98 -0.44 3.84 U-236 U-238 0.07 0.03 0.02 0.00 -0.02 0.03 0.02 1.26 Pu-238 Pu-239 -1.94 -3.27 1.44 2.45 0.41 1.74 0.71 -1.30 Pu-240 -6.00 -7.60 -1.14 0.85 -1.42 -1.06 -0.23 -2.45 Pu-241 -10.33 -17.91 -3.46 -0.65 -7.97 -6.27 -2.39 -8.93 Pu-242 -17.07 -36.69 -5.98 -9.70 -8.67 -9.24 0.79 Table A.4.3.1: Trino Predicted Minus Measured % Deviations (Part 2 of 4) e049-J8P 049-J8P4 049-J8P7 049-J8P9 049-L5P1 049-L5P4 049-L5P9 069-ElIPI Enrichment 2.719 2.719 2.719 2.719 2.719 2.719 2.719 3.13 Burnup 8.71 14.77 15.49 11.13 14.16 14.49 10.18 12.86 U-234 33.76 29.39 38.05 23.15 23.40 U-235 1.76 5.42 3.34 2.03 -0.29 0.83 1.87 0.57 U-236 1 -5.52 NET- 300067-01 Rev 0 A-57

U-238 0.00 0.00 0.07 0.01 0.02 0.01 0.03 -0.11 Pu-238 -6.75 Pu-239 3.84 4.38 1.23 1.96 0.71 2.52 0.87 6.03 Pu-240 5.45 -0.85 -2.98 -1.74 0.70 0.16 -1.98 7.03 Pu-241 2.17 -5.46 -6.34 -5.90 -3.36 -3.63 -9.11 4.90 Pu-242 3.33 -13.54 -12.06 -10.26 -5.22 -3.81 -10.72 2.72 Table A.4.3.1: Trino Predicted Minus Measured % Deviations (Part 3 of 4) 069- 069- 069- 069- 069- 069- 069- 069-EIlP2 EIIP4 EIIP5 EIIP7 EI1P8 EIIP9 E5P4 E5P7 Enrichment 3.13 3.13 3.13 3.13 3.13 3.13 3.13 3.13 Burnup 20.6 23.72 24.52 24.3 23.41 19.25 23.87 24.68 U-234 U-235 2.84 5.00 3.40 2.87 3.80 2.21 1.12 2.91 U-236 -5.46 -6.32 -4.61 -5.73 -6.35 -7.26 -3.54 -2.31 U-238 -0.09 -0.12 -0.10 -0.13 -0.09 -0.12 -0.10 -0.05 Pu-238 -11.01 -11.16 -12.51 -16.12 -29.21 -18.37 -18.46 -12.34 Pu-239 6.20 9.14 7.11 4.94 5.96 2.11 8.20 6.64 Pu-240 3.77 3.66 2.73 1.01 3.34 1.50 4.19 4.15 Pu-241 -0.94 2.01 6.57 0.51 -3.07 -2.88 0.91 1.72 Pu-242 -4.87 -6.51 2.89 -8.20 -7.37 -10.18 -4.16 -3.99 Table A.4.3.1: Trino Predicted Minus Measured % Deviations (Part 4 of 4) 069- 069- 069- 069- 069- 069- 069-E5P9 J9P4 J9P7 LI1P4 L1IP7 L5P4 L5P7 Enrichment 3.13 3.13 3.13 3.13 3.13 3.13 3.13 Burnup 19.21 24.85 25.26 23.93 24.36 24.33 24.31 U-234 U-235 1.32 3.09 2.59 1.56 3.37 -1.05 3.11 U-236 -8.57 -4.51 -4.81 -10.22 -0.39 -1.08 -3.63 U-238 -0.05 -0.01 -0.10 -0.03 -0.02 -0.10 0.09 Pu-238 -9.46 -14.17 -27.13 -6.07 -15.41 -6.07 -13.50 Pu-239 8.15 7.26 6.57 6.05 6.24 6.63 6.57 Pu-240 4.15 4.63 1.89 1.97 1.16 5.38 1.47 Pu-241 6.79 0.94 2.29 1.45 2.37 2.39 1.11 Pu-242 1.26 -6.08 -6.86 -5.67 -7.91 -0.45 -3.78 The Trino analysis shows an over-prediction of the Pu-239 and U-235 content. This implies that predicted k's after depletion are conservative. This is the same direction as seen by the EPRI benchmarks, but larger in magnitude. The direct difference section (A.4.3.3) that follows will show the impact on k (see Table A.4.3.12).

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A.4.3.2.2 OBRIGHEIM For the analysis of the Obrigheim chemical assays, the ORNL decks were checked and then used with the following modifications:

1. The geometric model was changed from NEWT to KENO V.a modeling. (This is a change only in input format. There was no change in input parameters, such as materials or dimensions.)

2. The time steps were changed to match the smaller time steps used in the first cycle and the number of libraries in the following cycles was changed to be less than 2 GWd/T between libraries. (Burnup in GWd/T divided by 2, truncated, then plus 1).

Table A.4.3.2 shows the percent deviation between the predicted and measured content of the measured isotopes for all the Obrigheim samples.

Table A.4.3.2: Obrigheim Predicted Minus Measured % Deviations (Part I of 2)

BE124- BE124- BE124- BE124- BE124- BE124- BE124- BE124-E3PI E3P2 E3P3 E3P4 E3P5 G7PI G7P2 G7P3 Enrichment 3 3 3 3 3 3 3 3 Burnup 20.18 29.35 36.26 30.92 22.86 17.13 25.83 31.32 U-235 -1.12 -5.16 -5.70 1.29 -3.93 -2.83 -7.77 3.98 U-236 7.87 -1.31 -2.78 -2.18 -10.39 -5.96 -10.23 -4.27 U-238 -0.11 0.71 -0.09 -0.04 0.00 -0.01 -0.01 -0.07 Pu-239 -9.55 -16.57 -13.63 -6.14 -14.02 -10.42 -8.80 -22.89 Pu-240 4.50 6.18 8.73 5.19 3.73 4.97 8.73 8.98 Pu-241 3.07 4.29 3.50 -0.55 3.07 3.27 4.22 0.68 Pu-242 9.98 6.21 8.25 2.54 2.30 2.49 5.42 4.70 Am-241 9.79 4.53 2.49 -3.47 5.09 0.75 4.29 -6.29 Cm-244 -22.44 -5.48 -4 1.00 -30.68 -16.61 -21.17 -34.55 Table A.4.3.2: Obrigheim Predicted Minus Measured % Deviations (Part 2 of 2)

Sample ID BE124- BE124- BE168 BE170 BE171 BE172 BE176 G7P4 G7P5 B68 E7 EI BEI72 BEI76 Enrichment 3 3 3.13 3.13 3.13 3.13 3.13 Burnup 27.71 25.81 29.35 27.01 28.74 27.89 28.78 U-235 -6.64 2.24 -1.81 -1.74 -1.74 -0.23 -0.73 U-236 -12.42 -3.35 0.91 0.48 1.03 1.44 0.89 U-238 0.00 0.00 -0.03 -0.03 -0.04 -0.08 -0.05 Pu-239 -16.17 -1.68 -18.23 -7.15 -7.57 -3.59 -10.48 Pu-240 6.73 5.48 3.06 4.30 4.28 5.01 4.14 Pu-241 1.69 -0.44 1.84 2.46 3.89 4.07 3.05 Pu-242 2.34 0.40 -1.61 -1.36 0.02 1.45 -1.28 Am-241 -3.76 -7.73 -14.75 -12.30 -9.80 -12.13 -13.35 Cm-244 -35.48 -34.67 -43.81 -33.44 -43.49 -31.23 NET- 300067-01 Rev 0 A-59

A.4.3.2.3 TURKEY POINT UNIT 3 Although full details of the operating history of the two assemblies is available through Turkey Point, rather than use potentially proprietary data, only the data documented by ORNL in Reference 20 is used.

The ORNL input decks are used to initiate the analysis and only the same two changes noted for Obrigheim are made.

Table A.4.3.3 shows the percent deviation between the predicted and measured content of the measured isotopes for all the Turkey Point samples.

Table A.4.3.3: Turkey Point Predicted Minus Measured % Deviations Sample ID DOI-GI0 DOI-G9 DOI-H9 D04-GIO D04-G9 Enrichment 2.556 2.556 2.556 2.556 2.556 Burnup 30.51 30.72 31.56 31.31 31.26 U-234 4.86 4.32 9.71 3.40 17.24 U-235 2.56 -1.87 -2.79 -1.61 1.47 U-236 2.84 3.09 6.25 3.82 6.51 UI-238 -0.17 -0.16 -0.14 -0.17 -0.18 Pu-238 -5.79 -4.31 -5.34 -1.58 -2.61 Pu-239 4.24 4.90 2.16 5.74 2.51 Pu-240 2.48 3.71 4.38 5.70 3.26 Pu-241 1.30 2.28 0.44 2.93 -1.45 Pu-242 -2.24 3.15 1.66 3.03 -0.37 A.4.3.2.4 H. B. ROBINSON UNIT 2 The data from Reference 20 is uscd for the modeling. The ORNL decks were used to initiate the analysis.

"Tihegeometric modeling was converted from NEWT to KENO-V.a format and the time steps were changed to the same step sizes used for all the analyses.

Notes:

1. Reference 20, Table 63 has the fuel temperatures for N-9C-D under the column for N-9C-J and vice versa. The input decks are correct.

2. Sample N-9C-J was taken from under the Inconel grid [27]. Inconel is a strong absorber resulting in a significant reduction of the thermal flux. This clearly challenges the 2D methods. This sample is eliminated.

Table A.4.3.4 shows the percent deviation between the predicted and measured content of the measured isotopes for all the H1. B. Robinson samples.

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Table A.4.3.4: H. B. Robinson Predicted Minus Measured % Deviations Sample ID B05-N9BN B05-N9BS B05-N9CD Enrichment 2.561 2.561 2.561 Burnup 23.81 16.02 31.66 U-234 1.71 1.35 11.31 U-235 0.84 0.92 3.40 U-236 -2.78 -2.85 -0.66 U-238 -0.59 0.18 -0.72 Np-237 -13.05 -11.77 -0.95 Pu-238 -19.19 -17.65 -15.36 Pu-239 2.71 2.76 5.31 Pu-240 1.75 2.71 3.18 Pu-241 -2.67 -2.88 1.46 Pu-242 -4.87 -2.76 -3.01 Tc-99 10.93 13.80 13.24 Nd- 143 6.77 4.87 10.03 Nd-145 3.49 3.16 6.42 A.4.3.2.5 CALVERT CLIFFS UNIT I The modeling of the Calvert Cliff chemical assays starts with the ORNL input decks and Reference 20.

Once again, the NEWT model was converted to a KENO V.a model. The large guide tubes, which are not adjacent to the sample, were homogenized with the water in the cell. This allowed for each unit to be the same area (the pin pitch squared). The homogenization preserved the mass of the water and the tube.

Thus, for assemblies Dl 01 and D047, the guide tubes in the model are homogenized and for assembly BT03, the central guide tube is modeled with the actual geometry and the other guide tubes are modeled as homogenized.

The soluble boron letdown data was copied from the ORNL decks as was the fuel temperature change with cycles.

Notes:

I. Table 71 in Refercnce 20 does not come from the referenced source but was instead found to correspond to Table 3.3 in Reference 34.

2. Reference 20, Table 74 values for cumulative burnup (except final burnup) have an unknown source. Attempting to reproduce these values using Table 71 results in the sample burnups as given in References 22 and 28. The specific powers in the ORNL input decks do not match Table

74. For MKP109, the total burnup in the ORNI., input decks is too low by from about 0.5% to 1%. The specific powers from Refercnce 22 with a small adjustment for historical round-off are used in this analysis for MKP 109.

3. Reference 20 and the ORNL input decks remove the BPs at the end of cycle 1. CE BPs are not removable. BT03 is a test assembly, so it is possible to remove the BPs but this was not done.

The ORNL decks for BT03 used two separate runs due to the BP mistake. This effort uses a single input deck. The soluble boron and temperature adjustment table from the second ORNL decks were used, but the cycle I time of 811 was added to each entry. Reference 20 states that the specific power should be calculated by taking the values in Reference 20, Table 74 and divide NET- 300067-01 Rev 0 A-61

them by the duration days. This was not followed by the ORNL input decks and the ORNL decks have some burnup errors. For this analysis, the specific power was calculated by the Fable 74 burnups divided by the duration days. The cumulative burnups on Table 74 match the cumulative burnups in References 22 and 28.

Table A.4.3.5 shows the percent deviation between the predicted and measured content of the measured isotopes for all the Calvert Cliffs samples.

Table A.4.3.5: Calvert Cliffs Predicted Minus Measured % Deviations (Part I of 2)

Sample ID DIO0- DI01- D101-MLA098-BB MLA098-JJ MLA098-P Enrichment 2.72 2.72 2.72 Burnup 26.62 18.68 33.17 U-234 13.32 11.31 4.86 U-235 0.98 1.01 6.13 U-236 -2.61 -3.16 -2.84 U-238 -1.77 -1.22 -1.04 Np-237 -9.50 -0.80 4.20 Pu-238 -14.31 -27.88 -8.63 Pu-239 3.92 2.16 9.28 Pu-240 2.90 0.58 4.02 Pu-241 -3.10 -5.72 3.07 Pu-242 -6.99 -8.29 -8.15 Am-241 -1.31 -3.73 1.91 Tc-99 6.13 2.16 6.37 NET- 300067-01 Rev 0 A-62

Table A.4.3.5: Calvert Cliffs Predicted Minus Measured % Deviations (Part 2 of 2)

Sample ID BT03- BT03- BT03- D047- D047- 1D047-NBDI07-GG NBDI07-MM NBDI07-Q MKPI09-CC MKPI09-LL MKPI09-P Enrichment 2.453 2.453 2.453 3.038 3.038 3.038 Burnup 37.27 31.4 46.46 37.12 27.35 44.34 U-234 -17.60 -29.80 21.35 -2.75 -1.40 1.39 U-235 -7.62 -3.31 -0.26 -0.87 -1.16 2.31 U-236 -0.84 1.11 0.10 2.42 2.05 2.08 U-238 -1.28 -0.77 -0.26 -0.38 -0.67 -0.19 Np-237 8.89 12.34 14.18 11.73 5.45 4.07 Pu-238 -13.18 -13.10 -14.31 -8.50 -8.71 -7.80 Pu-239 -0.70 -0.71 3.36 5.01 3.62 7.82 Pu-240 1.25 0.31 1.88 3.66 2.47 4.52 Pu-241 -5.76 -6.70 -1.21 0.36 0.85 2.52 Pu-242 -7.32 -7.53 -8.71 -2.61 -1.53 -5.00 Am-241 -13.38 -5.82 -59.56 -4.01 -0.20 -1.63 Tc-99 32.08 32.59 30.62 8.07 6.03 12.83 Rh- 103 -29.50 Cs- 133 1.42 0.96 2.27 Nd-143 -0.84 1.47 1.19 2.88 Nd- 145 -2.71 -1.63 -0.73 -1.94 Sm- 147 3.18 4.71 6.63 -3.73 Sm- 149 -31.28 Sm-150 2.55 7.38 7.19 3.57 Sm-151 -8.24 -1.01 12.33 -3.64 Sm- 152 2.51 7.67 7.76 -1.43 Eu-153 -6.67 3.23 7.91 0.43 Eu-155 -11.29 5.53 5.13 3.29 Gd- 155 20.62 7.68 -30.59 2.93 A.4.3.2.6 TAKAHAMA UNIT 3 There were 16 chemical assays taken from three pins (two from assembly NT3G23 and one from NT3G24). For each pin, a sample was taken so close to the top of the fuel that 2D analysis does not correctly pick up the spectrum. ORNL eliminated these samples in Reference 19 and they are also eliminated for this effort. ORNL in Reference 19 referred to the eliminated samples as "near the extreme ends of the fuel rods (< 20 cm)." This is misleading (but correct). The 20 cm includes the top plenum. A more useful description is that they are < 6 cm from the top of the active fuel (actually, 5.4, 2.9 and 1.6 cm from the top of the active fitel). One of the sampled pins is one of the corner pins of the assembly.

This pin is not typical of the average of the pins in the assembly. For example, the average pin pitch used for resonance treatment is not correct. Further, that pin is adjacent to pins with a different burnup and may have a different enrichment. There is no information available on the adjacent assemblies.

Normally, this pin would be rejected, but since there are only a limited number of chemical assays that include fission products, this pin is used in this analysis.

NET- 300067-01 Rev 0 A-63

For the corner pin, the assembly pitch is very important. Reference 20 gives the assembly pitch as 21.4 cm. Takahama 3 is a MHI 3 loop plant. MHI was a Westinghouse licensee and it is expected that MHI did not change the core design. The Westinghouse 3 and 4 loop plant assembly pitch is 21.5036 cm (see Turkey Point and 1-B Robinson). This slightly larger pitch is used in the analysis. The ORNL decks used 21.42 cm for the assembly pitch. No reference for this value was found. The pin pitch is given in Reference 20 as 1.259 cm but 1.26 cm is used in the ORNL input decks. This analysis uses 1.26 cm for the pin pitch, which is consistent with the ORNL input decks.

The pin pitch used for the resonance self-shielding for the corner pin and side pins is adjusted to include the assembly gap. The effective pin pitch is defined as the pin pitch derived by averaging the pitch to the four closest pins.

The fuel temperature data is not provided in the references. The ORNL decks use an approximate value of 900' K for all samples. Clearly, there should be a burnup dependent value, but none was used by ORNL and this analysis also uses 900' K for all samples.

The depletion analysis for the ORNL decks used the power option for all depleting material. It is not clear whether the flux option should be used for the Gd pins, but since the fission power dominates over the absorption power, the inputs for this analysis uses the power option as ORNL did.

The specific powers in the ORNL decks did not agree with Reference 20, Table 77, but the differences, though small, can be as much as 1%. 'fhis analysis uses the specific powers from Reference 20, Table 77 for all but the last cycle. For the last cycle, the specific power is forced to match the sample bumup. The adjustments needed for matching the burnup were small (less than 0.1%).

Tables A.4.3.6 and A.4.3.7 show the percent deviation between the predicted and measured content of the measured isotopes for all the Takahama samples.

Table A.4.3.6: Takahama SF95 and SF96 Predicted Minus Measured % Deviations Sample ID NT3G23- NT3G23- NT3G23- NT3G23- NT3G23- NT3G23- NT3G23- NT3G23-SF95-2 SF95-3 SF95-4 SF95-5 SF96-2 SF96-3 SF96-4 .SF96-5 Enrichment 4.11 4.11 4.11 4.11 2.63 2.63 2.63 2.63 Burnup 24.46 35.68 37.01 30.45 17.43 29.69 30.41 25.42 U-234 -0.19 22.08 21.11 -7.33 -2.88 -3.15 -3.37 -3.01 U-235 2.52 2.63 3.29 1.70 5.06 7.21 8.12 5.68 U-236 -2.45 -0.46 -0.56 -1.54 -7.01 -3.96 -4.25 -4.33 U-238 -0.10 -0.10 -0.14 -0.07 0.02 0.06 0.07 0.04 Np-237 25.00 35.07 32.25 29.45 Pu-238 -11.83 -2.97 -2.86 -5.69 -20.48 -18.63 -23.36 -16.08 Pu-239 7.10 8.28 7.44 7.88 4.53 4.73 3.82 4.57 Pu-240 3.12 7.23 6.32 6.42 0.03 1.09 0.44 2.74 Pu-241 -0.55 -1.48 1.22 0.84 -4.51 -1.87 -3.01 -1.38 Pu-242 -5.52 -5.64 -4.43 -1.57 -12.50 -10.78 -12.21 -7.51 Am-241 20.41 21.08 43.10 17.78 29.44 23.99 14.44 32.25 Am-243 8.33 9.53 12.03 10.16 -3.11 0.54 -3.79 2.02 Cm-242 -41.01 -55.63 -77.36 -18.98 -35.76 -30.33 -32.83 -26.56 Cm-244 -8.77 -0.02 -0.38 3.02 -25.73 -18.50 -24.14 -14.32 Nd-143 -1.32 -1.47 -0.62 -0.79 -6.47 -4.64 -3.86 -5.70 Nd-145 -0.79 -1.28 -0.74 -0.39 -5.73 -4.65 -4.05 -4.96 NET- 300067-01 Rev 0 A-64

Table A.4.3.7: Takaham SF97 Predicted Minus Measured % Deviations NT3G24- NT3G24- NT3G24- NT3G24- NT3G24-Sample ID SF97-2 SF97-3 SF97-4 SF97-5 SF97-6 Enrichment 4.11 4.11 4.11 4.11 4.11 Burnup 30.48 42.1 47.07 47.26 40.85 U-234 9.34 6.86 6.42 6.68 8.02 U-235 0.77 1.64 2.12 0.71 2.12 U-236 -0.57 -0.16 -0.31 -0.19 -0.82 U-238 -0.09 -0.06 -0.03 0.03 -0.08 Np-237 0.23 1.32 -0.58 -3.66 -2.45 Pu-238 -9.27 -12.07 -14.70 -17.30 -10.93 Pu-239 5.03 5.48 5.05 2.89 6.44 Pu-240 7.06 9.07 8.05 6.95 7.72 Pu-241 -1.58 -1.38 -1.50 -3.53 0.27 Pu-242 -2.59 -5.22 -5.92 -5.89 -3.82 Am-241 3.30 2.75 1.00 -1.27 4.64 Am-243 7.45 5.35 3.01 0.88 6.95 Cm-242 1.83 7.96 12.22 16.17 8.96 Cm-244 -4.44 -6.79 -9.95 -14.50 -3.27 Nd-143 1.83 1.83 2.37 1.56 1.98 Nd-145 0.64 -0.05 0.08 -0.01 -0.11 Sm- 147 4.16 3.61 2.80 3.07 3.71 Sm-149 -11.85 -8.09 0.13 0.13 -7.83 Sm- 150 2.79 1.75 1.22 0.48 2.88 Sm- 151 -2.52 -3.01 -4.73 -8.09 -0.86 Sm-152 4.34 3.21 2.67 2.73 3.00 A.4.3.2.7 TMI UNIT I As described in Section A.4.3. 1, all the samples analyzed by ANL are not included due to some problem in the chemical assay process that produced noticeable errors. For assembly NJ070G, data was used for 7 of the 8 samples. The sample 0 1S2 showed errors noted in Section A.4.3.1.

This analysis utilized the input decks from ORNL, while making the model changes noted previously for the other chemical assays from the other reactors. The following errors were found in the ORNL decks and were corrected for this analysis:

1. The BPs should be depleted with a constant flux. The ORNL decks used power rather than the constant flux option. That is corrected for this validation.

2. For the pin 01 samples, an adjacent assembly was burned for one cycle. In the ORNL burnup analysis for the adjacent assembly, the BPs were not depleted at all. This was changed to a constant flux depletion.

3. For the 01 samples, the 4.75 wt% adjacent assembly had Gd pins. These pins were modeled by ORNL as having 4.75 wt% U-235. The actual wt% was 4.19 wt% [29]. This was corrected.

NET- 300067-01 Rev 0 A-65

4. Although not an error, the Gd pins were modeled by ORNL as a single cylinder. Due to the onion ring depletion nature of Gd, for this analysis the Gd pins are modeled as three concentric rings each of one third the area of the fuel pin.

Table A.4.3.8 shows the percent deviation between the predicted and measured content of the measured isotopes for all the TMI samples.

Table A.4.3.8: TMI Predicted Minus Measured % Deviations NJ070G- NJ070G- NJ070G- NJ070G- NJ070G-012S4 012S5 012S6 013S7 013S8 Enrichment 4.657 4.657 4.657 4.657 4.657 Burnup 23.54 26.26 24.09 23.21 26.1 U-234 0.19 2.38 1.12 -1.33 0.86 U-235 3.23 3.58 2.09 4.97 3.72 U-236 -4.20 -3.32 -4.99 -4.86 -2.94 U-238 -0.06 -0.06 0.05 -0.03 -0.06 Np-237 -9.23 -4.56 -11.22 -7.96 -6.24 Pu-238 -20.17 -22.21 -34.36 -27.98 -24.13 Pu-239 4.39 2.88 -2.22 3.08 4.32 Pu-240 0.78 -1.45 -4.39 -2.83 -0.11 Pu-241 -3.22 -5.76 -10.63 -7.72 -5.88 Pu-242 -13.17 -14.50 -18.50 -21.85 -15.04 Am-241 -6.67 -2.94 11.85 -5.36 -2.22 Amn-243 -1.50 -6.66 12.69 -12.06 -3.23 Cm-242 -30.63 -38.81 -18.09 -36.01 -47.1 I Cm-244 -15.68 -23.28 -6.97 -30.97 -19.38 Nd-143 2.45 3.06 1.68 0.72 3.04 Nd- 145 1.31 2.00 1.27 -0.67 1.76 Sm- 147 -2.16 1.49 -0.65 -0.23 2.06 Sm-149 6.30 6.26 -0.91 6.76 6.34 Sm-150 -0.20 -0.52 -1.18 -2.29 -0.15 Sm- 151 0.31 -2.31 -8.55 0.64 -2.92 Sm- 152 3.08 3.60 5.22 1.38 3.92 Eu-151 -6.32 0.66 -14.44 3.73 0.73 Eu- 153 -2.18 -1.37 -3.28 -4.10 -1.92 Gd- 55 -18.63 -14.28 -31.95 -15.90 -16.30 A.4.3.2.8 GOSGEN: ARIANE PROGRAM The ORNL input decks from Reference 20 were correct. The only modifications needed were related to changing from NEWT to KENO V.a. The ORNI, analysis used separate input decks and collected data via files to put together the final analysis. For this work, the input decks were stacked as a single run.

Sample G4 is only 7.42 cm from the bottom of the active fuel. This sample is the closest to the end of the active fuel of any samples used. The Takahama sample that was 5.4 cm from the end was not used due to 3D effects. It is not clear how much the end of the fuel affects this sample. Even though this sample is near the end of the fuel, the 2D analysis of this sample is included in the validation.

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The measured values for all the chemical assays were taken from NUREG/CR-7108 [19]. An error in Table C.1 of NUREG/CR-7108 was found for the U3 sample. The error is in the cooling time used for some of the isotopes in the sample. Table C.I was to adjust the isotopic content to the cooling time, but several of the isotopes had zero cooling time for Sample U3 in Table C. 1. Reference 20, Table 42 identified isotopes which were at zero cooling time by using italics. These isotopes were not adjusted when put into NUREG/CR-7108, Table C. I. Confirmation of this was not straightforward, since Reference 30 did not have the decay time correct. The primary reference is Reference 31 and by using that reference, it was possible to confirm the cooling time error.

Table A.4.3.9 shows the percent deviation between the predicted and measured content of the measured isotopes for all the G6sgen samples.

Table A.4.3.9: Giosgen and GKN Predicted Minus Measured % Deviations Gosgen G6sgen GOsgen GKN Sample ID 1240-GUI 1701-GU3 1701-GU4 419-MiI Enrichment 3.5 4.1 4.1 3.8 Burnup 60.7 53.2 31.1 54 U-234 14.72 29.63 31.17 17.21 U-235 7.94 1.88 3.18 6.43 U-236 1.22 3.46 1.74 -0.40 U-238 -0.36 2.39 1.64 -0.34 Np-237 -9.56 -37.1 I1 18.38 Pu-238 -6.40 -4.07 -4.10 -12.32 Pu-239 7.15 6.02 7.62. 9.72 Pu-240 3.81 8.06 5.20 6.09 Pu-241 3.93 3.29 2.74 2.50 Pu-242 -6.46 1.57 0.36 -6.77 Am-241 10.99 17.01 1.72 24.37 Am-243 9.94 18.18 17.03 21.71 Cm-242 -1.82 -20.81 Cm-244 -1.10 0.04 -12.84 -9.31 Mo-95 1.61 1.80 0.97 10.01 Tc-99 5.73 11.17 24.93 -1.97 Ru-101 7.33 4.62 0.30 22.38 Rh- 103 16.37 24.23 2.03 18.74 Ag-I09 54.32 11.31 24.55 Cs-133 5.75 6.97 4.52 6.06 Nd- 143 6.99 6.28 0.10 3.96 Nd-145 0.73 3.72 0.58 -0.73 Smr147 0.94 11.09 11.26 1.84 Sm- 149 -6.21 15.44 3.51 0.90 Sm-150 4.56 9.89 9.67 1.05 Sm- 151 0.40 4.37 6.09 0.61 Sm- 152 -4.32 10.34 11.58 -0.73 Eu- 153 2.14 5.24 8.60 1.03 Eu- 155 -2.37 -4.49 2.09 -22.29 Gd- 155 3.96 13.01 -39.88 -3.39 NET- 300067-01 Rev 0 A-67

A.4.3.2.9 GKN II The only modifications needed to the ORNL input deck [20] were for the model changes to KENO V.a.

Table A.4.3.9 above shows the percent deviation between the predicted and measured content of the measured isotopes for the GKN 11 sample.

A.4.3.2.10 VANDELUOS II The starting point for this analysis is Reference 23. A sample set of input decks is provided in the appendix of Reference 23 for one of the chemical assays.

As noted in Section A.4.3.1, the data for sample 160-800 is not used since there was a problem with the uranium measurements [23]. It is inappropriate to do direct difference analysis missing the U-235 isotope that has such a high reactivity worth. Furthermore, fuel rod 160-800 was next to the core baffle for one cycle and data on the core baffle was not provided. Finally, the soft spectrum due to being next to the core baffle for a cycle makes the pin atypical and therefore not appropriate for this analysis.

With sample 160-800 not being used, the 5 remaining samples come from two pins that were initially loaded in Assembly EC45, burned for four cycles and then transferred to Assembly EF05 for the final cycle.

Table A.4.3.10 shows the percent deviation between the predicted and measured content of the measured isotopes for the Vandellos samples.

Table A.4.3.10: Vandellos Predicted Minus Measured % Deviations Sample ID Van-58-88 Van-58-148 Van-58-260 Van-58-700 Van-165 Enrichment 4.4982 4.4982 4.4982 4.4982 4.4982 Burnup 42.5 54.8 64.6 77 78.3 U-234 5.02 -4.50 18.02 14.23 17.18 U-235 -1.04 -11.33 5.95 6.37 10.20 U-236 7.45 -1.21 3.42 6.62 5.01 Np-237 -6.20 -18.40 -10.27 -2.56 -19.52 Pu-238 -2.49 12.58 -8.86 -7.01 -14.80 Pu-239 3.75 7.30 6.36 7.53 -1.58 Pu-240 5.37 8.27 5.49 5.11 0.04 Pu-241 3.64 10.53 4.31 7.86 -2.18 Pu-242 58.96 10.34 88.33 92.24 92.65 Am-241 24.97 24.97 20.23 24.60 15.05 Am-243 29.08 13.37 15.46 14.16 -12.94 Cm-244 49.50 47.68 39.70 45.68 33.07 Tc-99 5.55 Cs- 133 1.07 3.33 -6.92 6.38 10.47 Nd- 143 -1.45 1.25 0.40 1.63 7.71 Nd- 145 -0.83 0.04 -2.75 -3.99 1,41 Sm-147 1.96 2.38 9.70 6.32 -2.87 Sm- 149 -4.86 2.49 8.29 16.98 17.75 Sm-150 4.99 3.50 5.85 5.96 8.94 Sm-151 -3.55 6.15 5.82 0.70 4.95 NET- 300067-01 Rev 0 A-68

Sm- 152 2.31 0.60 -0.14 -1.05 -6.82 Eu-153 -7.77 -8.81 -13.04 -10.76 -13.69 Eu-155 5.78 -0.14 0.25 9.82 10.63 Gd-155 20.58 12.34 18.50 12.47 27.97 A.4.3.3 Determination of the Isotopic Depletion Bias By Direct Difference The reactivity worth of the error in the isotopic content is determined by finding the difference in the predicted k of the application (the spent fuel pool) between using the measured isotopic content and using the predicted isotopic content. The depletion analysis of the chemical assays was described in the previous section. The isotopic content was output by using the SCALE OPUS module.

Before going into the details of the direct difference calculation of the bias and uncertainty in k due to depletion analysis, the integrated results of the chemical assays should be reviewed. Table A.4.3. I1 shows the results of the chemical assay analysis for all the isotopes. The two isotopes that have the biggest effect on k are UJ-235 and Pu-239 (highlighted in yellow in Table A.4.3.1 1). The isotopic content of both of these isotopes is over predicted. This implies that the depletion analysis is conservative. The over prediction of some of the absorbing isotopes compensates for the over prediction of U-235 and Pu-239, but the net effect, as will be seen shortly, is that the depletion analysis is conservative and no bias will be needed. However, the standard deviation for the chemical assays is large. This implies that the uncertainty will be significant in the computation of k 95,95 .

Table A.4.3.11: Performance of All the Chemical Assay Analyses Number of Average Maximum Standard Average Isotope Assays Deviation Deviation Deviation M/C U-234 56 9.12 38.05 11.4 0.9088 U-235 92 1.48 10.20 3.3 0.9852 U-236 77 -1.72 7.87 3.8 1.0172 U-238 87 -0.07 2.39 0.4 1.0007 Np-237 36 0.23 35.07 13.0 0.9977 Pu-238 77 -12.64 12.58 8.0 1.1264 Pu-239 92 4.51 9.72 2.8 0.9549 Pu-240 92 2.39 9.07 3.1 0.9761 Pu-241 92 -0.64 10.53 4.8 1.0064 Pu-242 91 -2.14 92.65 19.0 1.0214 Am-241 50 2.27 43.10 29.1 0.9773 Am-243 39 0.74 29.13 13.6 0.9926 Cm-242 51 -28.20 16.17 17.7 1.2820 Cm-244 58 -1.81 49.50 17.3 1.0181 Mo-95 4 3.60 10.01 4.3 0.9640 Tc-99 17 12.96 32.59 10.7 0.8704 Ru-101 4 8.66 22.38 9.6 0.9134 Rh- 103 5 6.37 24.23 8.5 0.9363 Ag- 109 3 30.06 54.32 22.0 0.6994 Cs-133 12 3.52 10.47 4.3 0.9648 Nd-143 36 1.51 10.03 3.5 0.9849 Nd-145 36 -0.25 6.42 2.5 1.0025 NET- 300067-01 Rev 0 A-69

Sm-147 25 2.74 11.26 4.1 0.9726 Sm-149 22 1.37 17.75 8.5 0.9863 Sm-150 25 3.14 9.89 3.7 0.9686 Sm- 151 25 -0.40 12.33 5.0 1.0040 Sni-152 25 2.66 11.58 4.1 0.9734 Eu- 151 7 -3.82 3.73 6.3 1.0382 Eu-153 20 -2.34 8.60 6.3 1.0234 Eu-155 13 0.15 10.63 9.3 0.9985 Gd-155 20 -4.31 27.97 21.4 1.0431 For the calculation of the direct differences in reactivity, a model of the application is needed. The model selected is that of the Indian Point Region 2 with an absorber plate in every cell. The predicted g/gU were converted to atom densities for KENO calculations by multiplying by the gU in 97% TD pellets times 0.6022 barn atoms per mole and dividing by the atomic weight of the isotope. For isotopes that were assayed, the measured g/gU were similarly converted to atom densities. If an isotope was not assayed, then the "measured" atom density was determined as the predicted atom density multiplied by the average ratio of the measured divided by calculated (M!C) g/gU (given in Table A.4.3.1 I). Since the cooling times were not the same for all the chemical assays, they were all adjusted to zero and 15 years cooling time.

For the determination of the direct difference bias and uncertainty, 368 input decks (92 for each: predicted zero cooling, measured zero cooling, predicted 15 years cooling and measured 15 years cooling) for the Indian Point Region 2 were generated. Table A.4.3.12 shows the direct differences for each assay. The results for this analysis are plotted on Figure A.4.3.13. Also on Figure A.4.3.13 is a bounding line. There is one sample out of the 92 samples that exceeds the bounding curve, so this curve meets the 95/95 criteria. The bounding line is Isotopic Uncertainty in k = 0.0002

• Burnup in GWd/T Notice that the bias is negative, meaning the predictions over-predict k. This negative bias is ignored and only the uncertainty is used.

Table A.4.3.12: Direct Difference Results for Each Assay Measured Measured Measured ID Enrichment Burnup Minus Minus Minus (wt% U-235) (GWd/T) Predicted k Predicted k Predicted k ORNL No Cooling 15 yr Cooling BE124-E3PI 3.00 20.t8 -0.0004 -0.0028 -0.0011 BE124-E3P2 3.00 29.35 0.0027 -0.0011 0.0006 BE124-E3P3 3.00 36.26 -0.0065 -0.0102 -0.0085 BE124-E33P4 3.00 30.92 -0.0058 -0.0090 -0.0094 BE124-E3P5 3.00 22.86 0.0067 0.0030 0.0035 BE124-G7P1 3.00 17.13 0.0045 0.0019 0.0020 BE124-G7P2 3.00 25.83 0.0058 -0.0001 0.0016 BE124-G7P3 3.00 31.32 -0.0148 -0.0193 -0.0196 BE124-G7P4 3.00 27.71 0.0058 0.0005 0.0005 BF124-G7P5 3.00 25.81 -0.0047 -0.0093 -0.0102 BE168-- 3.13 29.35 0.0092 0.0011 -0.0002 BE 170-- 3.13 27.01 0.0070 -0.0002 -0.0016 NET- 300067-01 Rev 0 A-70

Measured Measured Measured Enrichment Burnup Minus Minus Minus ID (wt% U-235) (GWd/T) Predicted k Predicted k Predicted k ORNL No Cooling 15 yr Cooling BE171-- 3.13 28.74 0.0077 -0.0011 -0.0012 BE 172-- 3.13 27.89 0.0013 -0.0040 -0.0043 BE176-- 3.13 28.78 0.0064 -0.0017 -0.0028 104-M 11 P7 3.90 12.04 0.0000 -0.0007 -0.0015 032-El IPI 3.13 7.24 -0.0036 -0.0039 -0.0046 032-E I IP4 3.13 15.38 -0.0035 -0.0054 -0.0064 032-E 11 P7 3.13 15.90 -0.0068 -0.0085 -0.0101 032-E 1 P9 3.13 11.53 -0.0004 -0.0011 -0.0021 032-H9P4 3.13 16.56 0.0010 -0.0016 -0.0025 032-H9P7 3.13 17.45 0.0027 0.0016 0.0014 032-H9P9 3.13 12.37 -0.0022 -0.0029 -0.0043 049-J8P 1 2.72 8.71 -0.0042 -0.0046 -0.0044 049-J8P4 2.72 14.77 -0.0111 -0.0120 -0.0139 049-J8P7 2.72 15.49 -0.0032 -0.0052 -0.0070 049-J8P9 2.72 11.13 -0.0038 -0.0040 -0.0053 049-L5P 1 2.72 14.16 0.0027 0.0016 0.0010 049-L5P4 2.72 14.49 -0.0001 -0.0023 -0.0029 049-L5P9 2.72 10.18 -0.0029 -0.0029 -0.0042 069-El IPI 3.13 12.86 -0.0031 -0.0040 -0.0036 069-E I 1P2 3.13 20.60 -0.0080 -0.0082 -0.0094 069-E I IP4 3.13 23.72 -0.0149 -0.0162 -0.0181 069-EI IP5 3.13 24.52 -0.0110 -0.0161 -0.0154 069-E I I P7 3.13 24.30 -0.0067 -0.0108 -0.0121 069-E I IP8 3.13 23.41 -0.0083 -0.0098 -0.0122 069-EI 1IP9 3.1.3 19.25 -0.0027 -0.0041 -0.0054 069-E5P4 3.13 23.87 -0.0070 -0.0095 -0.0104 069-E5P7 3.13 24.68 -0.0083 -0.0106 -0.0111 069-E5P9 3.13 19.21 -0.0076 -0.0096 -0.0094 069-J9P4 3.13 24.85 -0.0103 -0.0105 -0.0114 069-J9P7 3.13 25.26 -0.0079 -0.01 10 -0.0117 069-L 11 P4 3.13 23.93 -0.0058 -0.0083 -0.0090 069-L 113P7 3.13 24.36 -0.0082 -0.0112 -0.0115 069-L5P4 3.13 24.33 -0.0014 -0.0046 -0.0048 069-L5P7 3.13 24.31 0.0000 -0.0130 -0.0141 D0I-GI0 2.56 30.51 -0.0026 -0.0073 -0.0084 DOI-G9 2.56 30.72 0.0027 -0.0039 -0.0043 DO 1-119 2.56 31.56 0.0088 0.0025 0.0024 D04-G 10 2.56 31.31 0.0007 -0.0052 -0.0057 D04-G9 2.56 31.26 0.0050 -0.0013 -0.0030 D101-MLA098-BB 2.72 26.62 -0.0019 -0.0055 -0.0073 D101-MLAO98-JJ 2.72 18.68 -0.0008 -0.0028 -0.0046 DIOI-MLA098-P 2.72 33.17 -0.0194 -0.0219 -0.0232 BT03-NBD107-GG 2.45 37.27 0.0151 0.0045 0.0015 BT03-NBD 107-MM 2.45 31.40 0.0166 0.0094 0.0068 BT03-NBDI07-Q 2.45 46.46 0.0055 -0.0093 -0.0105 D047-MKPI09-CC 3.04 37.12 -0.0033 -0.0041 -0.0045 D047-MKP 109-LL 3.04 27.35 -0,0010 -0.0042 -0.0037 D047-MKPI09-P 3.04 44.34 -0.0135 -0.0142 -0.0150 NET- 300067-01 Rev 0 A-71

Measured Measured Measured Enrichment Burnup Minus Minus Minus Sample ID (wt% U-235) (GWd/T) Predicted k Predicted k Predicted k ORNL No Cooling 15 yr Cooling B05-N9BN 2.56 23.81 0.0006 -0.0020 -0.0039 B05-N9BS 2.56 16.02 0.0005 -0.0015 -0.0021 B05-N9CD 2.56 31.66 -0.0059 -0.0094 -0.0105 NT3G23-SF95-2 4.11 24.46 -0.0109 -0.0085 -0.0095 NT3G23-SF95-3 4.11 35.68 -0.0134 -0.0085 -0.0115 NT3G23-SF95-4 4.11 37.01 -0.0142 -0.0095 -0.0108 NT3G23-SF95-5 4.11 30.45 -0.0110 -0.0074 -0.0088 NT3G23-SF96-2 2.63 17.43 -0.0054 -0.0078 -0.0100 NT3G23-SF96-3 2.63 29.69 0.0090 -0.0093 -0.0125 NT3G23-SF96-4 2.63 30.41 0.0092 -0.0087 -0.0123 NT3G23-SF96-5 2.63 25.42 0.0037 -0.0077 -0.0101 NT3G24-SF97-2 4.11 30.48 -0.0056 -0.0025 -0.0038 NT3G24-SF97-3 4.11 42.10 -0.0068 -0.0041 -0.0062 NT3G24-SF97-4 4.11 47.07 -0.0072 -0.0036 -0.0068 NT3G24-SF97-5 4.11 47.26 -0.0014 0.0015 -0.0024 NT3G24-SF97-6 4.11 40.85 -0.0110 -0.0066 -0.0088 NJ070G-0 12S4 4.66 23.54 -0.0069 -0.0070 -0.0082 NJ070G-0 12S5 4.66 26.26 -0.0072 -0.0058 -0.0082 NJ070G-012S6 4.66 24.09 -0.0001 -0.001 I -0.0036 NJ070G-013S7 4.66 23.21 -0.0068 -0.0093 -0.0118 NJ070G-013S8 4.66 26.10 -0.0089 -0.0069 -0.0099 NJ070G-0 IS3 4.66 26.84 -0.0073 -0.0065 -0.0087 NJ070G-O I S 1 4.66 25.53 -0.0088 -0.0069 -0.0087 1240-GUI 3.50 60.70 -0.0171 -0.0143 -0.0160 1701-GU3 4.10 53.20 0.0148 0.0023 -0.0004 1701 -GU4 4.10 31.10 -0.0081 -0.0094 -0.0096 419-M 11 3.80 54.00 -0.0168 -0.0123 -0.0193 Van-58-88 4.50 42.50 Not Done 0.0026 0.0040 Van-58-148 4.50 54.80 Not Done 0.0016 0.0058 Van-58-260 4.50 64.60 Not Done -0.0062 -0.0079 Van-58-700 4.50 77.00 Not Done -0.0074 -0.0051 Van-165 4.50 78.30 Not Done 0.0124 0.0118 NET- 300067-01 Rev 0 A-72

0.0160 0.0120 #

0.0080 V 0.0040 50.0000 zero cooling

-0.0040 15 yrcooling

-o.0080 --U-depletion uncertainty

-0.0120 LI.

-0.0160 4 +/-

-0.0200

-0.0240 0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 Burnup (GWO/MTU)

Figure A.4.3.13: Direct Difference for the 92 Chemical Assays and the Bounding Uncertainty In addition to the traditional graphical approach to determining a limiting uncertainty, a statistical uncertainty was calculated using the equations from NUREG/CR- 6698 [2]. The laboratory uncertainty for the chemical assays is isotope dependent, so an estimate of the reactivity uncertainty for each point would be difficult, so a constant uncertainty was used. Figure A.4.3.14 shows the direct differences with the selected uncertainty and the statistical analysis uncertainty. Note that the equations used for both the bias and the uncertainty do not take into account that the bias and uncertainty is known to be zero at zero burnup. This is a small effect on the bias but would have a large effect on the uncertainty. The non-statistical uncertainty uses this information to establish the intercept. Proper statistical analysis would result in an uncertainty curve that would be more similar to the non-statistical approach taken.

The statistical analysis determines an uncertainty from the mean bias. However, since credit for a negative bias is conservatively ignored, the uncertainty is the difference between zero and the statistical limiting curve found on Figure A.4.3.14. The EPRI benchmark based uncertainty is 0.0064. For all uncertainties greater than 0.0064, the engineering based uncertainty is larger than the statistically based uncertainty. Since this validation uses the most limiting of the EPRI or the Extended ISG-8 approaches, the engineering based limiting curve is always more conservative than the statistically based curve at the burnups where it is used.

The results of this analysis differ from that reported by ORNL in NUREG/CR-7108 [19]. However, ORNL provided their direct difference results in Table 6.3. Figure A.4.3.15 shows the ORNL results for the same assays (note that there are no ORNL results for the 5 Vandellos assays). The rack design for the ORNL work did not have absorber panels and the rack design for this work has absorber panels. Some of NET- 300067-01 Rev 0 A-73

the difference could be due to this but the points that differ the most above the limiting curve all come from cases where significant model improvements have been made in this analysis. There are three points that stand out. Two of these points come from Calvert Cliffs assembly BT03, where ORNL had incorrectly removed the burnable absorber rods. The third point is from GSsgen sample U3 where ORNL used an improper cooling time.

0.0160 0.0120 0.0080

• zero cooling tW 0.0040 0.0000 A 15 yr cooling V~-0.0040

-*-Bounding Isotopic

-0.0080 Content Uncertainty

-0.0120 -Isotopic Content Bias

-0.0160

- Statistically Based

-0.0200 *A Uncertainty

-0.0240 0.00 10.00 20.00 30.00 40. 50.00 60.00 70.00 80.00 Bumup (GWD/MTU)

Figure A.4.3.14 Direct Difference for the 92 Chemical Assays and the Bounding Uncertainty As a Statistical Based Uncertainty NET- 300067-01 Rev 0 A-74

0.0200 0.0160 XX X 0.0120

• ' 0.0080 0.0040

• zero cooling 0.0000 A 15yr cooling

-0.0040 il

-0.0080 -- depletion uncertainty

• -0.0120 ORNL 7108

-0.0160 A

-0.0200 I

-0.0240 0.00 20.00 40.00 60.00 80.00 Burnup (GWD/MTU)

Figure A.4.3.15 Direct Difference for the 92 Chemical Assays and the Bounding Uncertainty Including the Corresponding ORNL Results One final note on the chemical assay bias and uncertainty. Removing the bad TMI data would not affect the engineering approach to the uncertainty (and may not affect the statistical results). Figure A.4.3.16 is provided, which shows the eliminated data. Note that all the TMI points are well below the limiting curve. The HB Robinson sample that came from under the Inconel grid is the only point not utilized that is above the depletion uncertainty curve, but this sample unquestionably should be excluded, unless 3D analysis were to be used. Figure A.4.3.16 also shows the uncertainty selected for ISG 8, Rev. 3 [9].

NET- 300067-01 Rev 0 A-75

0.0400 0.0300 0.0200 i

'*0.0100 S0.0000

• zero cooling 9L

.; -0.o0 oo00_ A 15 yr cooling V 0-0depletion uncertainty Iooo ,ORNL 7108 N Bad ANIL TMI Experiments

-0.0400 0 Bad TMI NJO70G Sample

-0.0500 l HB Robinson Under Grid

-0.0600 -ISG 8 Rev 3 0.00 20.00 40.00 60.00 80.00 Burnup (GWD/MTU)

Figure A.4.3.16 Direct Difference for the 92 Chemical Assays and the Bounding Uncertainty Including All ORNL Results and ISG-8 Rev. 3 A.4.3.4 Validation of Isotopic Worth (Extended ISG-8)

Unlike the EPRI depletion reactivity benchmarks where the validation of the isotopic content and worth are done simultaneously, the Extended ISG-8 approach has separate biases and uncertainties for the isotopic content and the isotopic worth. The NRC commissioned ORNI. to determine an appropriate method for validation of the isotopic worth of the isotopes not covered by the Laboratory critical experiments. NUREG/CR-7109 [32] found that it is conservative to add a bias of 1.5% of the worth of the minor actinides and fission products to cover the bias and uncertainty in the isotopic worth. This validation follows that recommendation.

NET- 300067-01 Rev 0 A-76

A.5. Summary of Results SCALE 6.1.2, using the 238 group ENDF/B-VII cross-sections, has been validated for analysis of fresh and burned fuel. More specifically, the CSAS5 module for calculation of k's and the tS-depl module with parm=(addnux=4) for calculating depleted isotopic concentrations has been validated. For fresh fuel, the bias is 0.0029 for EALF up to 0.4 eV and 0.0037 for EALF's in the range from 0.4 to 0.6 eV and the uncertainty is 0.0050 for all analyses.

For burned fuel, the bias and uncertainty is taken from the more limiting of the EPRI method or the Extended ISG-8 method. The EPRI method depletion bias is 0.003. The depletion uncertainty is 0.0064 at all burnups. For the EPRI method, both bias and uncertainty are not burnup dependent. For the Extended ISG-8 method, there is a bias for the worth of minor actinides and fission products of 1.5%

of this worth. [here is no uncertainty on this term, since the bias is actually an estimate of the uncertainty [9, 32]. There is no bias for the chemical assay portion, since it is negative. The uncertainty from the chemical assays is 0.0002

• Burnup in GWd/T. For both the EPRI method and the Extended ISG-8 method, the fresh fuel bias and uncertainty in the previous paragraph is used. (The Extended ISG-8 method requires the most limiting of bias and uncertainty for the major actinides at all burnups, which happens to be the fresh fuel condition. The EPRI method requires the use of the fresh fuel bias and uncertainty.)

The worth of minor actinides and fission products at 40 GWd/T is about 11% in k. This means the bias for the minor actinides and fission products is about 0.0016 at 40 GWd/T. Thus, at 40 GWd/T, the EPRI method bias of 0.003 is larger than the Extended ISG-8 method bias. At 40 GWd/T, the EPRI method uncertainty is 0.0064 and the Extended ISG-8 method uncertainty is 0.0080. The uncertainties can be statistically combined (not so for biases), so although the uncertainty at 40 GWd/T is larger for the Extended ISG-8 method, the EPRI method is more limiting. After the statistical combination of uncertainties, it was found that the EPRI bias and uncertainty is more limiting through 43 GWd/T.

NET- 300067-01 Rev 0 A-77

A.6. Appendix References

[P] 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. Tayloc, Jr., Guidefor Validation of Nuclear CriticalitySafety CalculationalMethodology, NUREG/CR-6698, Nuclear Regulatory Commission, Washington, DC January 2001.

[3] InternationalHandbook of Evaluated CriticalitySafety Benchmark Experiments, NEA/NSC/DOC(95)3, Volume IV, Nuclear Energy Agency, OECD, Paris, September, 2010.

[4] DATAPLOT 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] Kleinbaum, Kupper, and Muller, Applied Regression Analysis and Other Multivariable Methods, Second Edition, page 48, PWS-KENT Publishing Company, Boston, MA 1988.

[6] "PROPI lET StatGuide: Examining normality test results,"

http://www.basic.norghwestem.edu/statguidefiles/n-dist exam res.html, located on 6/8/09.

[7] R. Mark Sirkin, Statistics for the Social Sciences, Third Edition, 2005, page 245, Sage Publications, Thousand Oaks, CA.

[8] M. Rahimi, E. Fuentes, and D. Lancaster, Isotopic and Criticality Validationfor PWR Actinide-Only Burnup Credit, DOE/RW-0497, U. S. Department of Energy, Office of Civilian Radioactive Waste Management, Washington, DC, May,1997.

[9] U.S. Nuclear Regulatory Commission, Spent Fuel Project Office Interim Staff Guidance - 8, Rev. 3 - Burnup Credit in the CriticalitySafety Analyses of PWR Spent Fuel in Transportand Storage Casks, U.S. Nuclear Regulatory Commission, April, 2012.

[10] K. S. Smith, et a]., Renchmarksfor QuantifyingFuel Reactivity Depletion Uncertainty, EPRI, Palo Alto, CA, Technical Report Number 1022909 (2011)

[11] D. B. Lancaster, Utilization of the EPRI Depletion Benchmarksfor Burnup Credit Validation, EPRI, Palo Alto, CA, 1025203 (2012).

[12] D. E. Mueller, K. R. Elam, and P. B. Fox, Evaluation of the French Haut Taux de Combustion (HTc) CriticalExperimentData, 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. Fernex, "Programme HTC Phase I : Rdseaux de crayons dans l'eau pure (Water-moderated and reflected simple arrays) R66valuation des exp6riences," DSU/SEC/T/2005-33/D.R., Institut de Radioprotection et de Sfiretd Nucldaire, 2008.

NET- 300067-01 Rev 0 A-78

[14] F. Fernex, ProgrammeHTC- Phase 2: R&seaux simples en eau empoisonn~e (bore et gadolinium) (Rejlected simple arraysmoderatedby poisoned water with gadolinium or boron) Rj~valuationdes expiriences, DSU/SEC/T/2005-38/D.R., Institut de Radioprotection et de Sfiret6 Nucldaire, 2008.

[15] F. Fernex, Programme HTC -- Phase 3 ."Configurations "stockage en piscine" (Pool storage)

Rdivaluation des experiences, DSU/SEC/T/2005-37/D.R., Institut de Radioprotection et de Sfiretd Nucl6aire, 2008.

[16] F. Fern ex, ProgrammeHTC - Phase 4 : Configurations "chdteamc de transport' (Shipping cask) - Rývaluation des expiriences, DSU/SEC/T/2005-36/D.R., Institut de Radioprotection et de SOretd Nucldaire, 2008.

[1 7] InternationalHandbook of EvaluatedCriticalitySajfty Benchmark Experiments, NEA/NSC/DOC(95)3, Volume VI, Nuclear Energy Agency, OECD, Paris, September, 2010.

[18] M. D. Del lart and S. M. Bowman, Validation of the SCALE BroadStructure 44-Group ENDF/B-V Cross-Section Libraryfor Use in CriticalitySafety Analyses, NUREG/CR-6102 (ORNL/TM- 12460), Oak Ridge National Laboratory, Oak Ridge, TN, September 1994.

[19] G. Radulescu, I. C. Gauld, G. llas, and J. C. Wagner, An Approachfor Validating Actinide and Fission Product Burnup Credit CriticalitySafety Analyses - Isotopic Composition Predictions,NUREG/CR-7108, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC, USA, April 2012.

[20] G. Radulescu, i. C. Gauld, and G. [las, SCALE 5.1 Predictionsof PWR Spent Nuclear Fuel Isotopic Compositions, ORNL/TM-2010/44, Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA, March 2010.

[21] Alan H. Wells, Burnup Credit - Contributionto the Analysis of Yankee Rowe Radiochemical Assays, EPRI Technical Report Number 1022910, Electric Power Research Institute, Palo Alto, CA, USA, October 2011.

[22] Meraj Rahimi, Elmilio Fuentes, and Dale Lancaster, Isotopic and CriticalityValidationfor PWR Actinide-Only Burnup Credit, DOE/RW-0497, Office of Civilian Radioactive Waste Management, U. S. Department of Energy, Washington, DC, USA, May 1997.

[23] G. [las, and 1. C. Gauld, Analysis ofExperimental Datafor High-Burnup PWR Spent Fuel

!sotopic Validation- Vandell6s H Reactor, NUREG/CR-7013, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC, USA, January 2011.

[24] S. Guardini and G. Guzzi, BENCHMARK, Reference Data on Post IrradiationAnalysis of Light Water Reactor Fuel Samples - A review of 12 years experience, results obtainedand their characterization,EUR 7879 EN, Commission of the European Communities, Joint Research Centre ISPRA Establishment, ECSC-EAEC, Brussels-Luxembourg, 1982.

NET- 300067-01 Rev 0 A-79

[25] A. M. Bresesti, et at., Post-IrradiationAnalysis of Trino Vercellese Reactor Fuel Elements, EUR 4909 e, Commission of the European Communities, Joint Research Centre ISPRA Establishment, 1972.

[26] P. Barbero, et al., Post irradiationexamination of the frel dischargedfrom the Trino Vercellese reactorafter the 2"'n irradiationcycle, EUR 5605 e, Commission of the European Communitics, Joint Research Centre ISPRA Establishment, ECSC, EEC, EAEC, Luxembourg, 1977.

[27] J. 0. Barrer, CharacterizationoJ'LWR Spent Fuel MCC-Approved Testing Material- A T74-101, PNL-5109, Rev. 1, Pacific Northwest Laboratory, 1985.

[28] 0. W. Hermann, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the Scale System for PWR Spent Fuel Isotopic CompositionAnalyvses, ORN L/TM- 12667, Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA, March 1995.

[29] TMI-I Cycle 10 Fuel Rod Failures, Volume 1: Root Cause FailureEvaluations, EPRI, Palo Alto, CA; GPU Nuclear, Parsippany, NJ; and Duke Power Company, Charlotte, NC: 1998.

EPRI Report TR-108784-V 1.

[30] G. Hlas, I. C. Gauld, and B. D. Murphy, Analysis of Experimental Data fbr High Burnup PWR Spent Fuel Isotopic Validation--ARIANE and REBUS Programs(U0 2 Fuel), NUREG/CR-6969 (ORNL/ITM-2008/072), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tennessee (2010).

[31] ARIANI InternationalProgramme-FinalReport, ORNL/SUB/97-XSV750-1, Oak Ridge National Laboratory, Oak Ridge, Tenn.., May 2003.

[32] J. M. Scaglione, D. E. Mueller, J.C. Wagner and W. J. Marshall, An Approachfor Validating.

Actinide and Fission Product Burnup Credit CriticalitySafety Analyses-Criticaliry(kefj)

Predictions,US Nuclear Regulatory Commission, NUREEG/CR-7109, Oak Ridge National Laboratory, Oak Ridge, Tenn. (2012).

NET- 300067-01 Rev 0 A-80

ATTACHMENT 3 TO NL-14-083 Affidavits in Support of Request to Withhold Information Entergy Nuclear Operations, Inc.

Indian Point Unit 2 Docket No. 50-247

Westinghouse Electric Company Engineering, Equipment and Major Projects 1000 Westinghouse Drive, Building 3 Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8560 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 Proj letter: NF-IP-14-37, Rev. I CAW-14-4059 November 6, 2014 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

NET-300067-01, Rev. 0, "Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels" (Proprietary)

The proprietary information on pages 10, 24-26, 34, 46, 88, 89 & 91 for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-14-4059 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Entergy Nuclear Operations, Inc.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse Affidavit should reference CAW-14-4059, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

Very truly yours, James A. Gresham, Manager Regulatory Compliance Enclosures

CAW- 14-4059 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

I, James A. Gresham, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

JJames A. Gresham, Manager Regulatory Compliance

2 CAW-14-4059 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC (Westinghouse),

and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) 1am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-.14-4059 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

4 CAW-14-4059 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.3 90, it is to be received in confidence by the Commission.

(v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in pages 10, 24-26, 34, 46, 88, 89 & 91 ofNET-300067-01, Rev. 0, "Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels" (Proprietary), for submittal to the Commission, being transmitted by Entergy Nuclear Operations, Inc. letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the Indian Point Unit 2 spent fuel pool criticality analysis, and may be used only for that purpose.

(a) This information is part of that which will enable Westinghouse to:

(i) Assist customers in obtaining licensing changes.

5 CAW-14-4059 (ii) Assist customers in analyzing the spent fuel pool and absorber panels to ensure criticality does not occur.

(b) Further this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of the information to its customers for the purpose of assisting in obtaining license changes.

(ii) Westinghouse can sell support and defense of spent fuel pool criticality analyses.

(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar criticality analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information Would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith is the proprietary version of a document furnished to the NRC associated with the Indian Point Unit 2 spent fuel pool criticality analysis, and may be used only for that purpose In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

HnItAr~ C~AntAr C)nA Hnlt~~ flri~~ Mnrltr~ri NJ.I flRfl~

Telephone (856) 797-0900 HOLTEC INTERNATIONAL Fax (856) 797-0909 SENT BY ELECTRONIC MAIL ONLY Document ID: 2441-003 Nov. 7, 2014 Mr. Roger Waters Regulatory Assurance Indian Point Energy Center 450 Broadway GSB Second Floor Buchanan, NY 10511-0249

Subject:

Affidavit Related to "Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels"

Dear Mr. Waters:

Holtec is pleased to approve the release, from a proprietary perspective, the following information to the United States Nuclear Regulatory Commission (USNRC):

Attachment I- NET-300067-01 Rev. 0, "Criticality Safety Analysis of the Indian Point Unit 2 Spent Fuel Pool with Credit for Inserted Neutron Absorber Panels" (Proprietary)

Attachment 2 - AFFIDAVIT PURSUANT TO 10 CFR 2.390 We require that you include this letter along with the affidavit (Attachment 2) pursuant to IOCFR2.390 when submitting Attachment I to the USNRC.

Holtec is looking forward to working with Indian Point to complete this project. If you have any questions or concerns please do not hesitate to reach out to Haizhen Pan (h.pan@holtec.com) or Jordan Landis (j.landis@holtec.com).

Sincerely, Haizhen Pan Program Manager Holtec International CC: Mr. Jordan Landis (j.landis@holtec.com)

Mr. Rick Trotta (r.trotta@holtec.com)

Document ID: 2441-003 HO Lr E c Confidential - Holtec International Proprietary Information 2441-001 INT"ERNATIONAL

U.S. Nuclear Regulatory Commission Document ID 2441-003 Non-Proprietary Attachment 2 AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Haizhen Pan, being duly sworn, depose and state as follows:

(1) I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.

(2) The information sought to be withheld is information provided in Attachment 1 to Holtec letter 2441-003. This Attachment contains Holtec Proprietary information denoted by the shaded areas.

(3) In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC See. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(1) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information",

and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

I of 5

U.S. Nuclear Regulatory Commission Document ID 2441-003 Non-Proprietary Attachment 2 AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a, 4.b and 4.e above.

(5) The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as 2 of 5

U.S. Nuclear Regulatory Commission Document ID 2441-003 Non-Proprietary Attachment 2 AFFIDAVIT PURSUANT TO 10 CFR 2.390 proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Holtec International is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec's competitor to copy our technology and offer it for sale in competition with our company, causing us financial injury.

3 of 5

U.S. Nuclear Regulatory Commission Document ID 2441-003 Non-Proprietary Attachment 2 AFFIDAVIT PURSUANT TO 10 CFR 2.390 (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

4 of 5

U.S. Nuclear Regulatory Commission Document ID 2441-003 Non-Proprietary Attachment 2 AFFIDAVIT PURSUANT TO 10 CFR 2.390 STATE OF NEW JERSEY )

) ss:

COUNTY OF BURLINGTON )

Haizhen Pan, being duly sworn, deposes and says:

That she has read the foregoing affidavit and the matters stated therein are true and correct to the best of her knowledge, information, and belief.

Executed at Marlton, New Jersey, this 7th day of November, 2014.

Haizhen Pan Program Manager Holtec International Subscribed and sworn before me this 7 -th day of ANovember ,2014.

1 Il1 ~4 AUMY 0. NIEBLE NOTWAY PUUIC STATE OF NEW JERSEY 10

• 2444047 MY COMMISSION EXPIRES APRIL 14. 2019 5 of 5