ML12362A392

Spent Fuel Pool Criticality Analysis with No Credit for Boraflex.

ML12362A392
Person / Time
Site:	Millstone
Issue date:	11/30/2012
From:	Dominion Nuclear Connecticut, Dominion Resources Services
To:	Office of Nuclear Reactor Regulation
References
Download: ML12362A392 (158)

Text

Serial No.12-678 Docket No. 50-336 Attachment 5 Criticality Safety Analysis Report (Non-Proprietary)

DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Nuclef-A'r-I8ys FýueI Millstone Unit 2 Spent Fuel Pool Criticality Analysis with No Credit for Boraflex Nuclear Analysis & Fuel Dominion Resources Services, Inc.

November 2012

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 ii Table of Contents 1 IN T R O D U C T IO N ............................................................................................................... 1 1.1 OBJECTIV E .................................................................................................................................... 1 1.2 REGULATORY CRITERIA ....................................................................................................... 1 1 .3 REFEREN CES ................................................................................................................................ 2 2 CALCULATIONAL TOOLS ...................................................................................... 3 2.1 SCALE 6.0 ....................................................................................................................................... 3 2.2 KENO .............................................................................................................................................. 3 2.3 TRITON ........................................................................................................................................... 3

2.4 REFERENCES

................................................................................................................................ 3 3 M E TH O D O L O G Y .................................................................................................... 4 3.1 CONSIDERATIONS FOR SPENT FUEL POOL CRITICALITY ANALYSIS ......................... 4 3.1.1 Fuel Assembly Selection .......................................................................................................... 4 3.1.2 Depletion Analysis .......................................................................................................................... 5 3.1.3 Conservative Depletion Conditions .......................................................................................... 14 3.1.4 Criticality Code Validation ......................................................................................................... 28 3.1.5 M iscellaneous Items ........................................................................................................................ 29 3.2 SUPPLEM ENTAL ANALY SIS GUID ANCE .......................................................................... 30 3.2.1 Considerations for Evaluation of Accidents ............................................................................... 30 3.2.2 Miscellaneous M odeling Considerations .................................................................................. 30 3.3 REFE REN CES .............................................................................................................................. 31 4 SPENT FUEL STORAGE RACKS & COMPONENTS ......................................... 33 4.1 SPENT FUEL POOL STORAGE CONFIGURATION DESCRIPTION ................................. 33 4.2 INDIVIDUAL STORAGE RACK TYPE DESCRIPTIONS .................................................. 33 4.2.1 Region A and B Storage Racks .................................................................................................. 33 4.2.2 Region C Storage Racks ................................................................................................................. 35 4.2.3 Revised SFP Rack Region Definitions ...................................................................................... 35 4.3 M ISCELLA NEOU S COM PONENTS ....................................................................................... 35 4.3.1 Borated Stainless Steel Poisons Pins .......................................................................................... 35 4.3.2 Control Element Assemblies (CEAs) ...................................... .................................................. 36 4.4 NON-STANDARD STORAGE CONFIGURATIONS ............................................................. 36 4.5 REFE RENCES .............................................................................................................................. 36 5 NEW FUEL STORAGE ANALYSIS .......................................................................... 37 5.1 M ODEL DESCRIPTION ......................................................................................................... 37 5.2 METH OD OLOGY ........................................................................................................................ 39 5.3 ANALY SIS RESULTS ................................................................................................................... 40 5.3.1 Uniform Interspersed M oderator ................................................................................................ 40 5.3.2 Sensitivity Cases ............................................................................................................................. 42 5.3.3 Rack Deformation ........................................................................................................................... 42 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 iii 5.3.4 Misplaced/Dropped Fuel Assembly ........................................................................................... 42 5.3.5 Uncertainties Due to Fuel Assembly Tolerances ........................................................................ 43 5.3.6 Calculation of M aximum Keff .......................................................................................................... 44 5.3.7 Conclusions and Results ........................................................................................................... 46

5.4 REFERENCES

.............................................................................................................................. 46 6 SFP ENRICHMENT, BURNUP & BORON REQUIREMENTS ........................... 47 6.1 ANALYSIS OF ENRICHMENT & BURNUP REQUIREMENTS ......................................... 47 6.1.1 M odel Description and General Calculations ................................................................................. 47 6 .1.2 R egio n 1 A naly sis ........................................................................................................................... 60 6 .1.3 R egio n 2 A naly sis ........................................................................................................................... 65 6 .1.4 R egio n 3 A nalysis ........................................................................................................................... 77 6 .1.5 R egion 4 A n alysis ........................................................................................................................... 88 6 .1.6 R eferen ces ....................................................................................................................................... 95 6.2 ANALYSIS OF SOLUBLE BORON REQUIREMENTS ....................................................... 96 6.2.1 Full Spent Fuel Pool M odel Description .................................................................................... 96 6 .2 .2 Referen ces ..................................................................................................................................... 1 17 7 SUM MARY O F RESULTS ........................................................................................... 118 7.1 ALLOWABLE STORAGE CONFIGURATIONS AND REQUIREMENTS ............................. 118 7 .1.1 Reg io n I ........................................................................................................................................ 118 7 .1 .2 Reg io n 2 ........................................................................................................................................ 1 18 7 .1 .3 Re gio n 3 ........................................................................................................................................ 1 18 7 .1 .4 Reg io n 4 ........................................................................................................................................ 1 18 7.2 SOLUBLE BORON REQUIREM ENTS ..................................................................................... 124 7 .2 .1 N orm al Conditio n s ........................................................................................................................ 124 7.2.2 Accident Conditions ...................................................................................................................... 124 APPENDIX A CRITICALITY CODE VALIDATION .................................................... A-1 APPENDIX B SAMPLE TRITON INPUT FILE ............................................................. B-1 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 iv LIST OF TABLES Table 3.1-1 MP2 Direct Difference Burnup Worth Uncertainty ..................................................... 10 Table 3.1-2 NUREG/CR-6801 Burnup Shapes ............................................................................... 15 Table 3.1-3 Burnup Credit Nodal Burnup Values .......................................................................... 15 Table 3.1-4 MP2 Core Power & Core Flow Data .......................................................................... 19 Table 3.1-5 MP2 Full Power Cycle Average Soluble Boron Concentration .................................. 21 Table 5.3-1 New Fuel Storage Racks Interspersed Moderation Effects Results ............................. 40 Table 5.3-2 Results of Misplaced/Dropped Fuel Assembly Analysis ............................................ 43 Table 5.3-3 Results of New Fuel Storage Rack Tolerance Calculations ......................................... 44 Table 5.3-4 Results of New Fuel Storage Rack KENO Calculations .............................................. 45 Table 6.1-1 MP2 Spent Fuel Storage Rack Specifications ............................................................. 51 Table 6.1-2 Specific Bias and Uncertainty Values .......................................................................... 59 Table 6.1-3 R egion 1 B est Estim ate K,ff ........................................................................................ 60 Table 6.1-4 Region 1 Bias and Uncertainty KENO Cases (0 ppm boron) ....................................... 61 Table 6.1-5 Region 1 Bias and Uncertainty KENO Cases (2000 ppm boron) ................................ 62 Table 6.1-6 Region 1 Bias and U ncertainty .................................................................................... 63 Table 6.1-7 R egion 2 B est Estim ate Keff ......................................................................................... 67 Table 6.1-8 Region 2 Bias and Uncertainty KENO Cases (0 ppm boron, fresh fuel) ..................... 70 Table 6.1-9 Region 2 Bias and Uncertainty KENO Cases (0 ppm boron, depleted fuel) ............... 70 Table 6.1-10 Region 2 Bias and Uncertainty KENO Cases (2000 ppm boron) ................................ 70 Table 6.1-11 Region 2 Bias and Uncertainty (0 ppm boron) ............................................................ 72 Table 6.1-12 Region 2 Bias and Uncertainty Comparison (0 and 2000 ppm boron) ........................ 73 Table 6.1-13 Region 2 Combination Depleted Fuel Cases .............................................................. 75 Table 6.1-14 R egion 2 B urnup C redit ................................................................................................ 75 Table 6.1-15 Region 3 Burnup Credit KENO Cases (3 rodlets) ....................................................... 78 Table 6.1-16 Region 3 Fresh Fuel Bias and Uncertainty (0 ppm boron, 3 rodlets) .......................... 79 Table 6.1-17 Region 3 Depleted Fuel Bias and Uncertainty (0 ppm boron, 3 rodlets) ..................... 81 Table 6.1-18 Region 3 Fresh Fuel Bias and Uncertainty (2000 ppm boron, 3 rodlets) .................... 81 Table 6.1-19 Region 3 Depleted Fuel Bias and Uncertainty (2000 ppm boron, 3 rodlets) ............... 82 Table 6.1-20 Region 3 Bias and Uncertainty (0 ppm soluble boron, 3 rodlets) ................................ 82 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 v Table 6.1-21 Region 3 Burnup Credit (0 ppm soluble boron, 3 rodlets) ......................................... 83 Table 6.1-22 Region 3 Bias and Uncertainty (2000 ppm soluble boron, 3 rodlets) ........................ 83 Table 6.1-23 Region 3 Burnup Credit KENO Cases (CEA) ........................................................... 85 Table 6.1-24 Region 3 Fresh Fuel Bias and Uncertainty (0 ppm boron, CEA) .............................. 85 Table 6.1-25 Region 3 Fresh Fuel Bias and Uncertainty (2000 ppm boron, CEA) .......................... 86 Table 6.1-26 Region 3 Bias and Uncertainty (0 ppm soluble boron, CEA) .................. 86 Table 6.1-27 Region 3 Bias and Uncertainty (2000 ppm soluble boron, CEA) ................................ 87 Table 6.1-28 Region 3 Burnup Credit (0 ppm soluble boron, CEA) ............................................. ....... 87 Table 6.1-29 Region 4 Burnup Credit KENO Cases (No Poison Pins) ......................................... 90 Table 6.1-30 Region 4 Bias and Uncertainty Cases ........................................................................ 90 Table 6.1-31 Region 4 Bias and Uncertainty (0 ppm soluble boron) ............................................... 91 Table 6.1-32 Region 4 Burnup Credit (0 ppm soluble boron) ...................................................... 91 Table 6.1-33 Region 3 Linear Interpolation Enrichment Bias ......................................................... 92 Table 6.1-34 Burnup Credit Requirements (Regions 2 - 4) ....................................................... 94 Table 6.2-1 Keff as a Function of Soluble Boron - Normal Conditions ............................................. 101 Table 6.2-2 KIff for Misload Accidents in Infinite Lattice Region Model (2x2 Assemblies, no soluble bo ron ) ..................... . .................... ............................................................................... 104 Table 6.2-3 Kff for Misload Accidents in Full Pool Model (no soluble boron) ............................ 105 Table 6.2-4 KIff vs Soluble Boron for the Limiting Misload Accident (150'F) .............................. 106 Table 6.2-5 KIff for Four Boundary Misalignment Accidents (no soluble boron) ........................... 107 Table 6.2-6 Assembly Drop/Heavy Load Accident Results (no soluble boron) ................................ 112 Table 6.2-7 Region 1 Optimum Pitch Calculation for Assembly Drop/Crush Accident (no soluble b oron) ........................................................................................................................ 114 Table 6.2-8 Keff vs SFP Temperature (no soluble boron) ................................................................. 114 Table 6.2-9 Simulation for Misload of Nonfuel Components in Empty Region 1 Peripheral Cells.. 115 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A LIST OF FIGURES Figure 3.1-1 MP2 NUREG/CR-7108 Spent Fuel Keff Bias Uncertainty ........................................... 9 Figure 3.1-2 MP2 Direct Difference Burnup Worth Uncertainty ..................................................... 13 Figure 3.1-3 Cumulative Assembly Average Relative Power vs Assembly Burnup ........................ 20 Figure 3.1-4 20 GWD/T TRITON Depletion Convergence - Kff vs Average Depletion Step Size ..... 25 Figure 3.1-5 50 GWD/T TRITON Depletion Convergence - Keff vs Average Depletion Step Size ... 26 Figure 3.1-6 20 GWD/T TRITON Depletion Convergence - Keff vs Decay Time ............................ 27 Figure 4.1-1 MP2 Spent Fuel Pool Existing Region Layout (Not to Scale) ..................................... 34 Figure 5.1-1 KENO3D-Produced Plot of New Fuel Storage Racks (Top Half Removed) ................ 38 Figure 5.3-1 NFSR Kff vs Interspersed Water Density (Volume %)................................................. 41 Figure 6.1-1 M P2 SFP Spent Fuel Pool Layout ............................................................................... 50 Figure 6.1-2 Region I and 2 Rack Cell KENO Model ...................................................................... 52 Figure 6.1-3 Region 3 and 4 Rack Cell KENO Model (Region 3 Fuel With Rodlets Shown) ...... 53 Figure 6.1-4 Region I and 2 Storage Cell KENO Model (Axial Views) ......................................... 54 Figure 6.1-5 TRITON Fuel Assembly Model .................................................................................. 56 Figure 6.1-6 Region I KENO Off-Center Fuel Placement (Inward) ................................................. 64 Figure 6.1-7 Region 2 3-Out-of 4 Storage Pattern .......................................................................... 65 Figure 6.1-8 Region 2 KENO Off-Center Fuel Placement (Inward) ................................................. 69 Figure 6.1-9 Region 2 Enrichm ent Sensitivity ................................................................................. 74 Figure 6.1-10 Region 2 Keff vs Total Burnup Credit Worth ............................................................... 76 Figure 6.1-11 Region 3 Off-Center Fuel Model with Poison Pins ..................................................... 80 Figure 6.1-12 Region 4 3-O ut-of-4 M odel ........................................................................................ 89 Figure 6.1-13 Region 3 Linear Enrichment Interpolation Bias .......................................................... 93 Figure 6.2-1 KENO3D-Plot of Region 1-2 Boundary Misalignment .................................................. 108 Figure 6.2-2 KENO3D-Plot of Second Region 1-2 Boundary Misalignment ..................................... 109 Figure 6.2-3 KENO3D-Plot of Region 2-4 Boundary Misalignment .................................................. 110 Figure 6.2-4 KENO3D-Plot of Region 2-3 Boundary Misalignment .................................................. 111 Figure 6.2-5 KENO3D-Plot of Compacted Region 1 2x2 Rack Module ............................................ 113 Figure 6.2-6 KENO3D-Produced Plot of Region 1 Misload with Unborated Water in Empty Cells.. 116 Figure 7.1-1 Region 2 Burnup Requirements - Assembly Type 2A ................................................... 119 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 Vii Figure 7.1-2 Region 2 Burnup Requirements - Assembly Type 2B .................................................... 120 Figure 7.1-3 Region 3 Burnup Requirements - Assemblies Containing Rodlets ................................ 121 Figure 7.1-4 Region 3 Burnup Requirements -Assemblies Containing CEAs .................................. 122 Figure 7.1-5 R egion 4 B urnup R equirem ents ...................................................................................... 123 Annotation of proprietary information herein corresponds as follows to the specific reasons delineated in the respective affidavits executed by the owners of the information:

1) AREVA information: denoted with A superscript

2) Westinghouse information: denoted with a, c superscripts

3) DNC information: denoted with b, e or d superscripts Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 viii LIST OF ACRONYMS AND ABBREVIATIONS Ltomoustion trngineering Control Element Assembly CFSB Consolidated Fuel Storage Boxes DNC ** iiDinion Nuleai rtonncticut .

EALF energy of the average lethargy of fission EPRI Electric Power Reseahi IsfItut ,,

HTC Haut Taux de Combustion, a series of critical experiments [1 Inner Diameter IH Internatinnil thndhnnk of F1valuated Criticalitv Safe.tv Benchmarkq Fxneriment*

Keff Effective neutron multiplication factor KENO * ' 'criticality analysis computer code employing 7Dimenomf Monte Carlo techniques -. 4

' t .. 2 .

MP2 Millstone Unit 2 MWd/MTlUfI Megawattaays pesjyetricTbn ,U+anium NFSR New Fuel Storage Racks NRC 2: :::::::r Nuclear Regulatory CmmihissIoin .. ._.. ...

OD Outer Diameter ORNL, Oak Ridge National aboratoy '3/4 I  :

ppm parts per million RCA ... * ':*r'adib'clemicaksl

"* *11*

"' '* i" *= * * : * * * ... . '............

. ............> ~

RSS Root Sum Square SCALE computer code .syste jiuedtotperform criticality calculations> 1 [1, SFCC Spent Fuel Cask Crane SFP> ,!*< *Spent Fuel Pool ' 'k"' &'C '

TRITON A module of SCALE used to perform fuel depletion calculations T..... =Techncal' K Specifid'atins ....

wt% Weight percent Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 I 1 INTRODUCTION 1.1 OBJECTIVE This report presents the results of criticality analyses for Millstone Unit 2, which includes analysis for new fuel and spent fuel storage racks. Except where noted specifically in this report, this updated analysis supersedes the existing analysis [Reference 1.1-1].

The primary objectives of this analysis are as follows:

- Eliminate reactivity credit for Boraflex panels in current regions A & B of the spent fuel pool (SFP)

- Confirm the spent fuel pool soluble boron concentration that is credited for reactivity control to meet Keff requirements under normal and accident conditions

- Revise allowed storage patterns for fuel in the spent fuel pool to meet Keff requirements under normal and accident conditions

- Develop assembly burnup versus initial enrichment limitations for spent fuel regions as applicable

- Require use of discharged control element assemblies (CEAs) or borated stainless steel poison rodlets for fuel assemblies stored in revised Region 3

- Eliminate requirement to use spent fuel rack cell blocking devices

- Confirm all analysis objectives are met for the current 4.85 wt% U-235 enrichment limit It is noted that in the display of results and data from the analysis, a conservative enrichment of 5.0 wt%

U-235 is presented. This was chosen to provide a bound for the current enrichment limit and is not intended to represent a revised limit. Unless otherwise noted, enrichment refers to an averaged enrichment derived as noted in the following description. Fuel assemblies used at MP2 may include reduced enrichment fuel rods adjacent to guide thimbles and reduced enrichment axial blanket regions. For simplicity, MP2 SFP criticality calculations are performed using a single enrichment in all fuel rods that is the highest initial planar average U-235 enrichment of the axial regions in the fuel assembly. This averaged enrichment is designated as the initial nominal planar average enrichment.

1.2 REGULATORY CRITERIA The governing regulatory criteria for fresh fuel and spent fuel pool criticality analysis are found in GDC 62 [Reference 1.2-1], 10 CFR 50.68 [Reference 1.2-2] and 10 CFR 70.24 [Reference 1.2-3]. The relevant criteria found in these references are summarized below.

GDC 62: states that "criticalityin the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 2 10 CFR 50.68: part (b)(2) states that "...(k-effective) of thefreshfuel in the freshfuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unboratedwater and must not exceed 0.95, at a 95 percentprobability,95 percent confidence level."

10 CFR 50.68: part (b)(3) states that "If optimum moderation offresh fuel in the freshfuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenousfluid, the k-effective correspondingto this optimum moderation must not exceed 0.98, at a 95 percent probability,95 percent confidence level."

10 CFR 50.68: part (b)(4) states that "If no creditfor soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability,95 percent confidence level, ifflooded with unborated water If credit is taken for soluble boron, the k-effective of the spentfuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability,95 percent confidence level, if flooded with borated water,and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability,95 percent confidence level, ifflooded with unboratedwater."

10 CFR 70.24: instead of specifying a limit on Kff, 10 CFR 70.24 requires controls to be in place to detect and mitigate the consequences of an inadvertent criticality event. However, this requirement is not applicable to Millstone Unit 2, which has committed to comply with 10 CFR 50.68(b).

In addition to criteria found in the sources cited above, the analysis described herein applies guidance from Section 9.1.1 of NUREG-0800 [Reference 1.2-4] and supplemental NRC guidance [Reference 1.2-5] and approaches to perform the detailed calculations. Section 3 describes the application of such supplemental guidance.

1.3 REFERENCES

1.1-1 Letter from J. Alan Price (DNC) to USNRC, "Millstone Power Station, Unit No. 2, Technical Specifications Change Request (TSCR) 2-10-01, Fuel Pool Requirements," November 6, 2001.

(ADAMS Accession No. ML013510295) 1.2-1 Title 10 of the Code of Federal Regulations (10 CFR) 50 Appendix A, General Design Criteria for Nuclear Power Plants Criterion 62, "Prevention of Criticality in Fuel Storage and Handling."

1.2-2 Title 10 of the Code of Federal Regulations (10 CFR) 50.68, "Criticality Accident Requirements."

1.2-3 Title 10 of the Code of Federal Regulations (10 CFR) 70.24, "Criticality Accident Requirements."

1.2-4 USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling," Revision 3, March 2007.

1.2-5 NRC Interim Staff Guidance DSS-ISG-2010-01, Revision 0, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Nuclear Fuel Pools," 9/29/:2011.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 3 2 CALCULATIONAL TOOLS 2.1 SCALE 6.0 The SCALE system [Reference 2.1-1] was developed by Oak Ridge National Laboratory (ORNL) for the Nuclear Regulatory Commission (NRC) to satisfy the need for a standardized method of analysis for evaluation of nuclear fuel facilities and shipping package designs. The SCALE versions that are utilized for this analysis include the control modules CSAS5 and the following functional modules: BONAMI, CENTRM, and KENO V.a. In addition to these codes, SCALE auxiliary modules PMC, WORKER, and CRAWDAD are used for automatic cross section library processing. References to KENO in the report should be interpreted as referring to the KENO V.a module.

2.2 KENO KENO V.a [Reference 2.2-1] (forthwith referred to as KENO) is a multigroup, 3-dimensional Monte Carlo computer program within SCALE that allows a simple description of complex geometries. KENO calculates the Kff of the modeled system and its associated statistical uncertainty.

2.3 TRITON TRITON [Reference 2.3-1] is a SCALE control module that enables depletion calculations to be performed by coordinating iterative calls between cross-section processing codes, KENO-V.a, and the ORIGEN-S point-depletion code. KENO 3D transport calculations are used to calculate weighted burnup-dependent cross sections that are employed to update ORIGEN-S libraries and to provide localized fluxes used for multiple depletion regions. TRITON uses a predictor-corrector approach to perform fuel-assembly burnup calculations. TRITON supports cross-section processing using BONAMI and CENTRM/PMC. TRITON is executed in SCALE using the T5-DEPL sequence, which calls processing modules CRAWDAD, BONAMI, WORKER, CENTRM, PMC, KENO V.a, KMART, COUPLE, ORIGEN-S, and OPUS.

2.4 REFERENCES

2.1-1 "SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," ORNL/TM-2005/39, Version 6, Volumes 1-3, January 2009.

2.2-1 "KENO V.a: An Improved Monte Carlo Criticality Program," ORNL/TM-2005/39, Version 6 Vol.

2, Section F11, January 2009.

2.3-1 "TRITON: A Two-Dimensional Transport and Depletion Module for Characterization of Spent Nuclear Fuel," ORNL/TM-2005/39, Version 6 Vol. 1, Section TI, January 2009.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 4 3 METHODOLOGY NRC issued Reference 3.1-1, in order to reiterate existing guidance, clarify ambiguity in existing guidance, and to identify lessons learned based on recent licensee submittals associated with criticality analysis for spent fuel pools. Reference 3.1-1 contains specific recommendations on developing inputs and performing the required calculations. This analysis adheres to the guidance of Reference 3.1-1, which is denoted herein as Interim Staff Guidance (ISG). Other guidance documents which are referenced within the ISG are also used to determine details of the required calculations. Depletion calculations are performed using TRITON and criticality calculations use KENO.

The analysis determines the limiting Millstone Unit 2 (MP2) fuel assembly design and appropriate depletion conditions for bumup credit calculations to be performed for the MP2 Spent Fuel Pool (SFP).

The ISG and other guidance documents referenced therein are used to determine calculations required.

MP2 plant specific input is developed using the guidance documents, historical plant data, and proposed fuel design. Sensitivity studies have been performed to clarify appropriate TRITON depletion step size, KENO variables important for adequate convergence (neutrons per generation, number of generations, number of generations skipped), and number of decay days for converged and conservative depleted fuel SFP Keff.

3.1 CONSIDERATIONS FOR SPENT FUEL POOL CRITICALITY ANALYSIS This section discusses the specific items contained in the ISG and indicates how each consideration of the guidance was satisfied in the analysis. In certain cases, it is indicated where more detail is provided in the report regarding how the guidance is satisfied. The list of items is found in Section IV, "Technical Guidance" of the ISG.

3.1.1 Fuel Assembly Selection The analysis should include a demonstration that the fuel assembly used in the analysis is appropriate for the specific conditions. Use of a limiting or bounding fuel assembly should consider that the limiting fuel assembly design can change based on the effects of other parameters in the analysis, e.g. depletion parameters, burnup credited, soluble boron present in the SFP, and permanently installed neutron absorbers.

Multiple fuel assembly designs have been used at MP2.

b,e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 5 3.1.2 Depletion Analysis 3.1.2.1 Use of TRITON for Depleted Fuel Isotopic Composition Calculations for MP2 fresh fuel storage and spent fuel pool storage criticality analyses use SCALE version 6.0 with ENDF/B-VII cross section data. Validation of SCALE 6.0 was performed following the recommendations from NUREG/CR-6698 [Reference 3.1-6] and includes use of HTC critical experiments. This is discussed in Section 3.1.4.

SCALE 6.1 is considered a state-of-the-art method for calculating depleted fuel :isotopic content and for criticality safety calculations. SCALE 6.1 has been used for the development of bias and uncertainty associated with predicted isotopic compositions in depleted fuel (NUREG/CR-7 108) [Reference 3.1.-7]

and to determine bias and uncertainty related to nuclides that are under-represented in criticality benchmarking (NUREG/CR-7109) [Reference 3.1-8]. Those studies used TRITON in the T-DEPL sequence that uses NEWT rather than KENO to calculate weighted cross sections and provide localized fluxes using ENDF/B-VII cross sections. Section T I of the SCALE documentation [Reference 2.1-1 ]

affirms the use of either NEWT or KENO in the depletion sequence (from the Abstract):

"The more rigoroustreatment of neutron transportavailable within NEWT and KENO, coupled with the accuracy of ORIGEN-S depletion capabilitiesand SCALE resonance self-shielding calculationswithin TRITON-driven lattice physics analyses,provides a rigorousfirst-principlesapproachfor calculation of cross sections and isotopic depletion source termsfor fuel designs."

Pbe Use of Electric Power Research Institute (EPRI) Benchmarks Dominion has performed TRITON calculations to evaluate results of eleven EPRI depletion benchmarks

[Reference 3.1-9, Reference 3.1-10]. The EPRI benchmarks were developed for the purpose of validating depletion reactivity predictions. Sample benchmark results are provided in EPRI Report 1025203

[Reference 3.1-10] that were performed using SCALE 6.1 with ENDF/B-VII cross sections and the T5-DEPL depletion sequence. This analysis evaluated these benchmarks to demonstrate that: 1) burnup worth calculated with the SCALE 6.0 code system is comparable to the results obtained in EPRI Report 1022909 [Reference 3.1-9] and 2) that calculated benchmark burnup worth using SCALE 6.0 compares favorably with measured benchmark burnup worth. The analysis of these benchmarks is for verification only; the numerical results are not used in this analysis for burnup worth uncertainty or in the generation of burnup curves.

For all comparably run depletions (9 sets of cases from 0 to 60 GWd/MTU), Dominion results using SCALE 6.0 differ by no more than 0.74% (average absolute difference < 0.2%) of burnup worth from the Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 6 EPRI SCALE 6.1 results. In addition, these results verify that there is no significant burnup worth bias relative to the EPRI depleted fuel measured benchmark burnup worth. It is concluded that because of the favorable comparison of SCALE 6.0 and 6.1 depletion results, as well as the validation of SCALE 6.1 provided by NUREG/CR-7108 and other ORNL documents, use of SCALE 6.0 with ENDF/B-VII cross sections and the T5-DEPL sequence is appropriate for calculation of depleted MP2 fuel isotopic content.

It is not intended that the EPRI benchmarks be reviewed explicitly as part of this application. Dominion's use of the benchmarks provides additional confirmation that SCALE 6.0 TRITON with KENO provides comparable results to those obtained with SCALE 6.1 TRITON and KENO, as well as good agreement with measured burnup worth benchmarks.

3.1.2.2 Depletion Uncertainty The ISG cites Reference 3.1-2, which contains a statement regarding depletion uncertainty:

"A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculationaluncertainties.In the absence of any other determinationof the depletion uncertainty,an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption."

The ISG further clarifies the appropriate use of depletion uncertainty as defined in Reference 3.1-2:

i. "Depletion uncertainty" as cited in the Kopp memorandum should only be construed as covering the uncertainty in the isotopic number densities generated during the depletion simulations.

ii. The "reactivity decrement" should be the decrement associated with the Keff of a fresh unburned fuel assembly that has no integral burnable neutron absorbers, to the Keff of the fuel assembly with the bumup of interest either with or without residual integral burnable neutron absorbers, whichever results in the larger reactivity decrement.

The MP2 SFP criticality analysis uses decreased fuel reactivity due to fuel burnup (reactivity decrement or burnup worth) in part to demonstrate that SFP Kff requirements are met. Uncertainty in burnup worth in the ISG is discussed by reference to guidance in the Kopp memorandum. However, subsequent to the issuance of the ISG, two NUREGs were released which deal with different aspects of burnup worth uncertainty.

NUREG/CR-7109 [Reference 3.1-8] investigates Keff bias and bias uncertainty associated with criticality validation using critical experiment benchmarking. Reference 3.1-8 concluded that with the inclusion of critical experiments with significant plutonium content, conventional validation is adequate for determining the Keff bias and bias uncertainty related to major actinides (U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). For other burnup worth nuclides, NUREG/CR-7109 recommends applying a Keff bias equal to 1.5% of the worth of minor actinides and fission products. In a pre-meeting with NRC staff, Dominion agreed to include French Haut Taux de Combustion (HTC) critical experiments in code validation Keff comparisons and to apply 1.5% of the minor actinide and Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 7 fission product worth as a K~ff bias. In addition, Dominion agreed to include KIf uncertainty to account for uncertainty in structural material cross sections to be combined by root sum square (RSS) with other uncertainty components. Tables 6.11 and 6.12 of the NUREG provide a value of 0.0008 AK for structural material uncertainty.

NUREG/CR-7108 addresses the question of bias and uncertainty in K~f associated with the difference between predicted and measured fuel isotopic content. During a pre-meeting with NRC staff, Dominion agreed to determine burnup worth uncertainty as a percentage of total bumup worth using data presented in Table 6.3 of the NUREG. The calculated burnup worth uncertainty is applied to all MP2 burnup credit.

calculations as an uncertainty to be combined by RSS with other uncertainty components. Section 3.1.2.3 below documents the calculation of the burnup worth uncertainty allowance.

3.1.2.3 Conservative Assessment of Burnup Worth Uncertainty (NUREG/CR-7108)

Burnup credit for SFP storage requires an assessment of bias and uncertainty related to bumup credit worth (burnup worth). A major element of the overall burnup credit bias and uncertainty is the bias and uncertainty of predicted depleted fuel isotopic composition, which is the subject of NUREG/CR-7108.

During a Dominion/NRC pre-meeting to discuss proposed SFP criticality safety analysis for MP2, a Dominion evaluation of NUREG/CR-7108 was presented. The key observation regarding Keff uncertainty associated with bumup credit isotopic uncertainty is that the NUREG/CR-7108 Kfff bias uncertainty at low burnup is too large and non-physical. Figure 3.1-1 shows the NUREG results from the primary investigation method (Monte Carlo sampling) and results from direct simulation of depleted fuel subjected to radiochemical assay (RCA) measurements. The direct difference method results show a physically realistic trend, correctly approaching zero as burnup worth approaches zero, and shows increasing uncertainty with increasing burnup worth. The Monte Carlo sampling method, which is largely rooted in direct difference radiochemical assay data, exhibits non-physical trends at low burnup.

For these reasons, Dominion used direct difference results to determine an appropriate burnup worth uncertainty. Direct difference data is based on a set of 100 radiochemical RCA samples and is discussed in Section 6.2 of NUREG/CR-7108. Table 6.3 in the NUREG contains KENO modeling results for the 100 samples (sample ID, enrichment, burnup, predicted nuclide Krff, and measured nuclide Ker). For each sample, a detailed SCALE 6.1 ENDF/B-VII TRITON depletion was performed to determine predicted nuclide concentrations for 28 important bumup credit actinides and fission products. Measured RCA results for the 100 samples are listed in NUREG Table C. 1 along with 4 additional precursor nuclides,. Measurements were not all obtained after the same amount of decay time, and were adjusted as described in Section 6.2 of the NUREG to obtain nuclide concentration after 3 days of decay.

Because not all nuclides were measured for every sample, some "measured" values are actually based on surrogate nuclides that were measured. However, all samples had direct measurement of the maj or U and Pu nuclides. Conclusion 7 in Section 8 of NUREG-7108 states that "uncertaintiesin the calculated U-235 andPu-239 concentrationscontributeapproximately 90% to 95% of the K-eff bias uncertainty."

This means that NUREG-7108 Table 6.3 Kff differences represent at least 90% to 95% of potential differences even though total bumup credit worth is significantly understated due to the use of only 28 burnup credit nuclides.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 8 Burnup worth uncertainty (% of total burnup worth) is determined using burnup worth difference (% of total burnup worth) from NUREG-7108 Table 6.3 direct difference data (AK). hi order to convert direct difference AK into relative burnup worth difference (% of total burnup worth), total burnup worth for each sample is needed.

Burnup worth difference (%) = 100 x NUREG Table 6.3 AK / Total burnup worth Total burnup worth (AK) = Fresh Fuel I ff- Depleted fuel Kff (measured isotopes)

Fresh fuel KIff is needed to calculate total burnup worth but is not provided in NUREG-7108. [

Pbe For SFP criticality burnup credit, over-predicting burnup worth (KIff fresh - Keff depleted) is non-conservative. Negative values in Figure 3.1-2 are cases for which [

]bPe Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 9 NUREG/CR-7108 PWR Spent Fuel Keff Bias Uncertainty Monte Carlo Sampling and Direct Difference 0.040 0.030 0.020 0.010 t

S0.000

-0.010

-0.030

-0.040

-0.050 0 10 20 30 40 50 60 Burnup (GWd/MTU)

Figure 3.1-1 MP2 NUREG/CR-7108 Spent Fuel Keff Bias Uncertainty Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 10 Table 3.1-1 MP2 Direct Difference Burnup Worth Uncertainty Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 11 Table 3.1-1 MP2 Direct Difference Burnup Worth Uncertainty (continued) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 12 Table 3.1-1 MP2 Direct Difference Burnup Worth Uncertainty (continued) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 13 b, e Figure 3.1-2 MP2 Direct Difference Burnup Worth Uncertainty Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 14 3.1.2.4 Conservative Treatment of Fuel Assembly Burnup Another element of burnup worth uncertainty is uncertainty associated with the measurement of in-reactor fuel burnup, sometimes called the "reactor record". For MP2, measured burnup uncertainty is composed of calorimetric power measurement uncertainty and assembly relative power measurement uncertainty.

For the MP2 SFP criticality analysis these two contributors are independent and are combined together by RSS, resulting in measured fuel bumup uncertainty of [ ]b,e This value is conservatively set to

[ ]b,e for MP2 measured burnup uncertainty calculations, and is combined together by RSS with other uncertainty components for the calculation of total bias and uncertainty.

3.1.3 Conservative Depletion Conditions A number of the ISG items are associated with modeling choices for parameters that have an impact on the isotopic content of spent fuel. For this analysis, these items are discussed in the context of achieving conservative conditions used in the depletion calculations that establish the condition of the spent fuel.

Several of these considerations are discussed below.

3.1.3.1 Axial Burnup Profile The ISG states "Use of the limiting axial burnup distributionsfrom NUREG/CR.-6801 are acceptablefor existing PWRs, provided they are used in a manner consistent with NUREG/CR-6801, e.g. the profiles are used within the burnup ranges specified."

NUREG/CR-6801 [Reference 3.1-5] evaluated 3169 axial burnup profiles to determine the most reactive representatives in each burnup range. Included in the population are 544 CE 14x14 profiles (Section 3.1) which "may exhibit a smaller end effect on average" (Section 4.2.2) than the other fuel design profiles in the database. This means that it is appropriate and likely conservative to use the NUREG profiles for MP2.

Some MP2 fuel assemblies have reduced enrichment axial blankets. In Cycles 10-14, Gadolinia rods had axial blankets, and fuel in Cycles 15-22 (current cycle) have both Gadolinia cutback regions and axial blankets on all fuel rods. For simplicity, MP2 burnup credit calculations are performed conservatively assuming no axial blankets are present. Section 4.2.3 of NUREG/CR-6801 states "because the axial blankets have significantly lower enrichment than the central region, the end efjfct for assemblies with axial blankets is typically very small or negative... consequently, profilesfrom assemblies with axial blankets were not considered in this or previous bounding profile analyses." Therefore, assuming no axial blankets are present in MP2 fuel is both conservative and consistent with use of the profiles in NUREG/CR-6801. Table 3.1-2 contains the axial burnup shapes that are used for the MP2 SFP criticality burnup credit calculations. Table 3.1-3 shows the same shapes converted to nodal burnup in 10 GWd/T assembly average burnup intervals.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 15 Table 3.1-2 NUREG/CR-6801 Burnup Shapes 3 1 rl eigni ai middle of 6-10 18-22 26-30 38-42 > 46 I node (%) GWDIT GWD/T GWD/T GWDrI GWD/T 2.8% 0.658 0.668 0.619 0.660 0.582 8.3% 1.007 1.034 0.924 0.936 0.920 13.9% 1.091 1.150 1.056 1.045 1.065 19.4% 1.070 1.094 1.097 1.080 1.105 25.0% 1.022 1.053 1.103 1.091 1.113 30.6% 0.989 1.048 1.101 1.093 1.110 36.1% 0.978 1.064 1.103 1.092 1.105 41.7% 0.989 1.095 1.112 1.090 1.100 47.2% 1.031 1.121 1.125 1.089 1.095 52.8% 1.082 1.135 1.136 1.088 1.091 58.3% 1.110 1.140 1.143 1.088 1.088 63.9% 1.121 1.138 1.143 1.086 1.084 69.4% 1.124 1.130 1.136 1.084 1.080 75.0% 1.120 1.106 1.115 1.077 1.072 80.6% 1.101 1.049 1.047 1.057 1.050 86.1% 1.045 0.933 0.882 0.996 0.992 91.7% 0.894 0.669 0.701 0.823 0.833 97.2% 0.569 0.373 0.456 0.525 0.515 Table 3.1-3 Burnup Credit Nodal Burnup Values Ass!emfbly

~Burnup/

(GWD/T) 10 K 20p 30. 40 5 BU1 6.6 13.4 18.6 26.4 29.1 BU2 10.1 20.7 27.7 37.4 46.0 BU3 10.9 23.0 31.7 41.8 53.3 I BU4 10.7 21.9 32.9 43.2 55.3 BU5 10.2 21.1 33.1 43.6 55.7 BU6 9.9 21.0 33.0 43.7 55.5 BU7 9.8 21.3 33.1 43.7 55.3 BU8 9.9 21.9 33.4 43.6 55.0 BU9 10.3 22.4 33.8 43.6 54.8 I BU10 10.8 22.7 34.1 43.5 54.6 BUll 11.1 22.8 34.3 43.5 54.4 BU12 11.2 22.8 34.3 43.4 54.2 BU13 11.2 22.6 34.1 43.4 54.0 BU14 11.2 22.1 33.5 43.1 53.6 BU15 11.0 21.0 31.4 42.3 52.5 BU16 10.5 18.7 26.5 39.8 49.6 BU17 8.9 13.4 21.0 32.9 41.7 BU18 5.7 7.5 13.7 21.0 25.8 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 16 3.1.3.2 Depletion Reactor Parameters The ISG indicates that to be consistent with the guidance in the Kopp memorandum [Reference 3.1-2] for "the spentfuel storage racks loaded with fuel of the maximum permissible reactivity," the depletion simulations should be performed with parameters that maximize the reactivity of the depleted fuel assembly. The analysis should include a demonstration that the reactor parameters used in the depletion analysis are appropriate for the specific conditions. NUREG/CR-6665 [Reference 3.1-3] provides some discussion on the treatment of depletion analysis parameters for PWRs.

Reactor parameters should be chosen to maximize reactivity of the depleted fuel assembly. Bounding values traceable to other licensee documents should be used. Where bounding values are mutually exclusive, the dominant value should be maximized. The following parameter list is developed from NUREG/CR-6665 [Reference 3.1-3] and NUREG/CR-6759 [Reference 3.1-4]. It addresses items 2.b (Reactor Parameters), 2.c (Burnable Absorbers) and 2.d (Rodded Operation) in Section IV of the ISG.

i. Control rods ii. Fuel temperature iii. Moderator temperature iv. Moderator density

v. Soluble boron vi. Operating history vii. Specific power viii. Burnable absorbers ix. Grids Treatment of these items is discussed below in Sections 3.1.3.3 to 3.1.3.6.

3.1.3.3 Fuel Assembly Power, Moderator Temperature and Fuel Temperature Reactor parameters have been chosen to maximize reactivity of the depleted fuel assembly with consideration of guidance in NUREG/CR-6665 [Reference 3.1-3] and NUREG/CR-6759

[Reference 3.1-4] and making use of bounding MP2 fuel assembly power history. Bounding power history means [

]b,e Higher moderator temperature (lower density) and higher fuel temperature during fuel depletion increases SFP Kff. NUREG/CR-6665 guidance on specific power and time-dependent operating history suggests a complex interaction between actinide worth variation with power history and offsetting fission product worth variation with power history. The overall effect is estimated to be small

(- 0.2% AK/K).

Because there are no clearly conservative assumptions with regard to time-dependent power history, and because MP2 fuel assemblies experience a wide range of power histories, [

]b e The approach is based on these principles:

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 17

1) Moderator temperature increases from fuel inlet to fuel outlet based on fuel assembly power and moderator flow rate. Minimizing flow rate and maximizing assembly power maximizes moderator temperature. Table 3.1-4 shows MP2 core power and moderator flow rate data used to select the bounding inlet temperature, core power, and moderator flow rate (highlighted).

2) Fuel temperature is a function of local moderator temperature (a boundary condition for fuel rod heat conduction) and local fuel assembly power. Local in this sense means the fuel power and moderator temperature in the axial node being considered. Minimizing flow rate and maximizing assembly power maximizes fuel temperature.

3) The axial burnup shape represents the average axial power shape for a fuel assembly over its depletion history multiplied by the assembly average burnup.

4) If the [ ]b,e and [ ]b,e are known, then the history-averaged axial power shape can be determined by Pbe

5) Given RPD(z) for some nodal structure and maximum fuel assembly power (MAP), history-averaged local depletion power POW(z) can be determined by

[be

6) Given POW(z) and the core minimum flow rate, maximum history-averaged local moderator temperature (TM(z)) appropriate for fuel depletion can be determined using a simple channel heat balance.

7) Given TM(z) and POW(z), the maximum history-averaged local fuel temperature TF(z) appropriate for fuel depletion can be determined.

]b e To obtain the correct history-averaged TF(z), [

b,e Maximum assembly power used to calculate depletion conditions bounds the depletion history of fuel used at MP2. MP2 assembly average relative power versus assembly burnup for all MP2 fuel assemblies used through Cycle 20 is shown in Figure 3.1-3. The average relative power is [

]b,e The bounding ]b,e for determination of depletion power, moderator temperature, and fuel temperature. Use of this value conservatively assumes that all MP2 fuel assemblies have been operated with [

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 18

]b,e There are no planned changes to MP2 fuel management that would cause b,e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 19 Table 3.1-4 MP2 Core Power & Core Flow Data Cycle Full Power HFP Inlet Vessel Vessel TS IAnimum Flow TS FLODEN Core (WVINth) Temp Avg Temp Avg.Temp (Mlbmlhr) Pvlnimum (kglcm2-hr) Average (F) (F) (K) Core Exit Temp Flow (F)

(Mlbm/hr) 1 2560 549 573.3 573ý9 139.3 135.1 .654 59812 2 2560 549 573.3; 5739. 139-3 135.1 654 598.2

3. 2700 549 573.3 573.9 139.3 135.1 654 600.7 4 2700 549 573.3 573.9 .139.3 135.1 654 600,7 5 2700 549 573.3 573.9 1365 132.4 641 601.6 6 2700 549 573.3 573.9 13.1.8 127.8 619 603.4, 7 2700 549 57313 573.9 131.8 127.8' 619 6034 8 2700 549 573.3 57319 128.0 124.1 601 6048 9 2700 549 *573.3 5739 128.0 124.1 601 604.8 10 2700 549 573.3 573.9 1280 1242 601 604.8 11 2700 549 573 3 573.9Q

- 135223 t73 5375 607,2 12 2700 549 573.3 573.9 135ý5 131.5 637 602,0 13 2700 549 573.3 573.9 1355 1315 637 602.0 14 2700 549 573.3 5739 135.5 131.4 636 602.0 15 2700 549 573.3 573.9 135.5 131A4 636 602.0 16 2700 549 573.3 5739 13525 131.4 636 602.0:

17 2700 549 573.3 57139 135.5 131.4 636. 602.0.

18 2700 545 569.5 571.8 136.3 132.1 640 598,3 19 2700 545 569.5 571.8 136.3 132,1 640 598.3 20 2700 545 569.5 571.8 136,3 132,1 640 598.3 21 2700 545 569.5 571.8 136,3 132.1 64.0 598.3 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 20 b, e Figure 3.1-3 Cumulative Assembly Average Relative Power vs Assembly Burnup Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 21 3.1.3.4 Soluble Boron Depletion with higher soluble boron increases SFP Keff (NUREG/CR-6665). Predicted full power cycle average soluble boron concentration for each MP2 fuel cycle is provided in Table 3.1-5. A soluble boron value of 800 ppm conservatively bounds all MP2 cycle average values and provides margin for variations in future cycle soluble boron.

Table 3.1-5 MP2 Full Power Cycle Average Soluble Boron Concentration Cycle Cycle Average Boron PPM M2C1 396 M2C2 321 M2C3 435 M2C4 486 M2C5 402 M2C6 397 M2C7 437 M2C8 367 M2C9 469 M2C10 501 M2C11 393 M2C12 487 M2C13 619 M2Cl4 696 M2C15 669 M2C16: 661

.M2C17 604

.M2C18 599 M2C19 616 M2C20 625 M2C21 677 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 22 3.1.3.5 Control Rods and Burnable Absorbers Depletion with control rods inserted increases SFP K~ff (NUREG/CR-6759). The bounding maximum MP2 control rod history based on historical operating data is

]be Reduced enrichment rods with various amounts of gadolinia burnable absorber have been used in MP2 fuel since Cycle 10. Three batches of fuel from Cycle 1 contained burnable poison rods in place of fuel rods. No other burnable absorbers have been used at MP2.

NUREG/CR-6760 addresses the reactivity effect of burnable absorbers for SFP burnup credit applications. In the report, various combinations of fuel enrichment and Gadolinia enrichment (as well as the number of Gadolinia rods) are analyzed, including fuel assemblies of both the Siemens as well as Combustion Engineering (CE) designs which are comparable to MP2 fuel. As demonstrated in Figures 16-17 of the report, it is always conservative over the entire postulated bumup range to neglect the presence of Gadolinia in performing depletions for bumup credit applications. Per Figures 16 - 17 of the report, it is extremely conservative to neglect the presence of Gadolinia in performing burnup credit calculations for low burnup fuel (< 18-20 GWd/MTU). It is marginally conservative (approximately 250 -

500 pcm depending on the Gd 20 3 loading) to neglect the presence of Gadolinia for fuel assembly burnup in excess of 18-20 GWd/MTU.

An additional study was performed to assess the impact of Gadolinia on burnup credit calculations specific to MP2 fuel assemblies. TRITON depletions were performed to burnups of 30 GWd/MTU and 60 GWd/MTU, with 6 wt% Gd 20 3 and 2.55 wt% U-235 enrichment in the U0 2-Gd 2O 3 rods, 3.75 wt% U-235 enrichment for the eight fuel rods surrounding the five water holes, with the remaining fuel rods uniformly enriched to 4.25 wt% U-235. The results obtained for MP2 fuel agree with the conclusions presented in NUREG-6760. For MP2 fuel it is conservative to neglect the presence of Gd 20 3 in performing burnup credit calculations.

MP2 used fresh discrete burnable poison rods (BPRs) in Cycle 1 and used depleted BPRs through Cycle

6. These rods, which were made of B 4C-AI 20 3, replaced twelve normal fuel rods in the assembly.

NUREG-6760 (Section 3.3.4) also addresses the impact of B 4C-A120 3 poison rods on burnup credit applications. Figures 35-36 of that report demonstrate that it is conservative to neglect the presence of B 4C-A120 3 burnable poison rods throughout the entire range of burnups studied (0 - 60 GWd/MTU). It should be noted that the configuration of 12 B 4C-A120 3 rods studied in the report is identical to those used at MP2. Nonetheless, two test cases were run to verify that this approximation is definitively conservative. The results of the cases demonstrate the expected result, that it is conservative to neglect the presence of BPRs for MP2 fuel.

Some Cycle 1 MP2 fuel assemblies contained burnable absorber rods in the place of fuel rods. Many later assemblies contained various amounts of gadolinia burnable absorber mixed with U0 2 in some fuel rods.

Depleted fuel reactivity effects of these absorbers were analyzed. In all cases examined, ignoring the Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 23 presence of burnable absorbers is conservative for new and depleted fuel SFP Ken, calculations. No credit is taken in this SFP criticality analysis for the neutron absorption of new or residual burnable absorbers.

3.1.3.6 Spacer Grids A study was performed to determine the effect of depletion with spacer grids on MP2 SFP Kff. TRITON was used to perform single axial node depletions to 50 GWd/MTU. Fuel isotopic: content for fresh and depleted fuel for two initial enrichments (3.0 and 5.0 wt%) was modeled in both types of storage racks with 0 and 2000 ppm soluble boron.

Pbe 3.1.3.7 TRITON Convergence and Decay Time TRITON depletions are required to calculate 18 node depleted fuel isotopic content for each unique fuel enrichment and burnup combination. Performing the TRITON depletions requires:

1) A SCALE / KENO model of the limiting assembly.

2) Conservative depletion conditions.

3) Appropriate TRITON input to ensure convergence of the KENO fluxes and ORIGEN number densities.

4) Decay time that maximizes SFP Keff.

[

b,e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 24

3) There is no statistically significant difference between the results for 4, 5, 6, 7, or 8 days decay. A decay time of 4 to 8 days will provide the maximum KIf.

Decay time results are consistent with NUREG/CR-6665 (Section 4.3), which studied long cooling times at various burnups and concluded "Fuel dischargedfrom a reactor increases in reactivityfor several days due to the decay of short-lived poisons...After this point reactivity decreases continuously with time out to about 100 years".

Figures 3.1-4 and 3.1-5 demonstrate the results of the convergence study. Figure 3.1-6 shows the effect of decay time.

In the burnup credit calculation, TRITON depletions are used to calculate depleted fuel isotopic content for a range of initial fuel enrichment at assembly average and nodal burnups indicated in Table 3.1-3. For each assembly average burnup and initial enrichment combination, eighteen TRITON depletions are run with the conservative node-specific depletion conditions described in Sections 3.1.3.1 through 3.1.3.6, using appropriate TRITON input to assure convergence and decay time that maximizes depleted fuel KIf.

Depleted fuel isotopic content for each equal size node is subsequently used in eighteen axial fuel node KENO SFP models.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 25 b, e Figure 3.1-4 20 GWD/T TRITON Depletion Convergence - Keff vs Average Depletion Step Size Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 26 b, e Figure 3.1-5 50 GWD/T TRITON Depletion Convergence - Ken vs Average Depletion Step Size Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 27 20 GWD/T TRITON Depletion Convergence MP2 Region 3 Keno K-effective vs Decay Time 0.9765 0.9760 0.9755 6o,9750 0,974S 0.9740

-Bcst fstima le 0.9735 1 2 4 6 8 10 12 14 16 18 20. 22 DecayTime After Depletion {DaVs)

Figure 3.1-6 20 GWDiT TRITON Depletion Convergence - Ktr vs Decay Time Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 28 3.1.3.8 Rack Model The ISG indicates that the rack model dimensions and materials should be traceable to design documents and neutron absorbers, including any degradation, should be modeled conservatively. All rack dimensions and material characteristics employed in this analysis have been verified using design documents. In all racks containing Boraflex, the Boraflex degradation is conservatively modeled by taking no credit for Boraflex, which is assumed to be replaced by SFP water. [

]b,e Design drawings of the Boraflex wrappers have been reviewed to confirm the existence of vent holes ([ ]a,c diameter) near the top of each wrapper. Gas that may be produced within the wrapper will not displace water since the gas would escape from the vent holes.

3.1.3.9 Interfaces The ISG indicates that analyses which contain more than one storage configuration consider the interface between storage configurations to ensure that the regulatory requirement for KIff to be known with a 95%

probability with 95% confidence level is met. This is addressed in the analysis cases described in Section 6.2.1.3.

3.1.3.10 Normal Conditions The ISG indicates that in addition to the normal static condition where all fuel assemblies are in the approved storage locations, the analysis include appropriate normal conditions, such as fuel movement, fuel inspections and fuel reconstitution. This is addressed in Section 6.2.1.2.

3.1.3.11 Accident Conditions The ISG indicates, citing the Kopp memorandum [Reference 3.1-2] that "the criticality safety analysis should consider all credible incidents and postulatedaccidents." The ISG guidance further states that accidents should be considered with respect to all normal conditions, e.g., fuel inspections and fuel reconstitution and that analysis which credit soluble boron include a boron dilution analysis. Section 6.2.1.3 discusses how these items are addressed.

3.1.4 Criticality Code Validation The ISG indicates that the analysis methods and neutron cross-section data used should be benchmarked by comparison with critical experiments by the organization performing the analysis. The ISG discusses the aspects of criticality code validation under these specific subsection categories: a) Area of Applicability, b) Trend Analysis, c) Statistical Treatment, d) Lumped Fission Products and e) Code-to-Code Comparisons. The validation results are summarized below. Appendix A provides a more detailed description of the validation results.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 29 An independent validation of the SCALE Version 6.0 code system using ENDF/B-VII cross section data was performed. 253 critical experiments were modeled, including cases containing soluble boron, borated poison materials, Ag-In-Cd absorbers, and fuel containing plutonium designed to approximate the composition of depleted fuel. The validation analysis follows the methodology from NUREG/CR-6698

[Reference 3.1-6], and includes area of applicability, trend analysis, and statistical treatment as outlined in the ISG. The result of the calculation is a code bias and uncertainty, valid over the area of applicability (energy of the average lethargy of fission (EALF), U-235 enrichment, Pu content, soluble boron, structural materials, fuel rod diameter, temperature, clad material, fuel rod pitch, and rack cell pitch).

The code bias and uncertainty was calculated using appropriate statistical methods from NUREG/CR-6698. [

]b~e The code bias uncertainty is root sum squared together with the other calculated uncertainties. This is consistent with ANSIIANS 8.27 [Reference 3.1-12]. The code bias is taken as a bias. There is no use of lumped fission product modeling or code-to-code comparisons in the MP2 SFP SCALE 6.0 code validation, so these aspects of validation are not applicable.

3.1.5 Miscellaneous Items The ISG lists several miscellaneous considerations in Section IV.5 which are discussed in this section.

3.1.5.1 Cited Precedents The ISG indicates that cited precedents are justified and used appropriately, including whether any deviations from the precedent are justified. A precedent of previous approval does not itself provide justification for a proposed change, but it may allow use of prior information to facilitate NRC staff review. For the updated MP2 analysis, there are no direct precedents which are cited in this fashion, but the LAR does list some previous submittals for information.

3.1.5.2 Cited References The ISG indicates that cited references are appropriate and used in context. Such references can allow for more efficient NRC staff review, but do have inherent limitations. Use of observations or conclusions in references which are qualified by statements such as "typically" and "very small" should be verified to be used in context and any extrapolation beyond the bounds of the reference should be demonstrated as appropriate. The various citations of references included in this report provide confirmation that they are used in context and within their inherent limitations.

3.1.5.3 Assumptions Numerous assumptions, both implicit and explicit, are involved in performing a complex analysis such as the SFP criticality safety analysis. The ISG indicates that the applicability of an assumption may change with different scenarios in the analysis and that the application should explicitly identify and justify all Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 30 assumptions used. Assumptions are stated and discussion provided to justify their use in applicable sections of this report.

3.2 SUPPLEMENTAL ANALYSIS GUIDANCE In addition to the specific ISG guidance items, additional considerations were employed in the analysis.

This section discusses supplemental items which were derived from discussions with NRC staff at pre-submittal meetings and from other sources. The key source documenting discussions with NRC is the meeting summary for the February 15, 2012 pre-submittal meeting [Reference 3.2-1].

3.2.1 Considerations for Evaluation of Accidents In Reference 3.2-1, several items related to treatment of accident scenarios for the MP2 SFP criticality analysis were noted. These items are listed below along with the section of this report in which each is discussed.

- an analysis of a fuel assembly-misplacement accident scenario, or a probability of occurrence analysis that shows that the accident scenario is not credible with the use of cell blockers This is addressed in Section 6.2.1.3.

- a new analysis of existing accident and event scenarios (e.g., fuel drop, seismic, fuel handling.

etc.) or existing accident analysis should be shown to be bounding This is addressed in Section 6.2.1.3.

- an analysis of a fuel rod misplacement outside the spent fuel pool rack or demonstration that a misplacement is physically impossible.

This is addressed in Section 6.2.1.3.

3.2.2 Miscellaneous Modeling Considerations Certain considerations related to specific modeling details were among the items to be considered, as noted in Reference 3.2-1. These items are listed below along with the section of this report in which each is discussed.

- an analysis showing that there will be no gas entrapment caused by Boraflex, or an analysis of the gas entrapment caused; Boraflex should be modeled appropriately based on these analyses This is addressed in Section 3.1.3.8.

- all of the qualitative data from the TRITON validation study This is addressed in Section 3.1.2.1.

- a study of isotopic modeling including a sensitivity study to demonstrate the method used is conservative This is addressed in Section 3.1.3.

- a discussion on how the borated stainless steel rods are modeled Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 31 This is addressed in Section 4.3.1.

a sensitivity study on the effect of Gadolinium in the spent fuel pool This is addressed in Section 3.1.3.5.

specifically address any rod inserts that are currently in fuel assemblies in the spent fuel pool Credit for CEAs and borated stainless steel rodlets is addressed in Section 4.3.

As part of the analysis for Region 3, credit is taken for the placement of three borated stainless steel pins in fuel assembly guide and instrument thimbles. These borated stainless steel pins were designed specifically for reactivity control.

During the Millstone Unit 2 operating lifetime, numerous CEAs have reached their allowed lifetime fluence limit and are stored in fuel assemblies in the Spent Fuel Pool. The fluence limits were established to prevent control rod failure due to control rod swelling and clad cracking. As part of the criticality analysis credit was taken for the placement of what are called, full length, full strength CEAs in the Spent Fuel Pool Region 3 (Section 4.3.2).

Over the plant lifetime there have been various other non full length, full strength CEA designs.

Millstone Unit 2 Cycle 1 contained 8 part length CEAs (identifier letters A through H, inclusive).

Prior to Cycle 2, the part length CEAs were removed and placed in the spent fuel pool. In addition, eight full length, reduced strength (less than 5 fingers) CEAs were discharged at the end of Cycle 6 to the spent fuel pool (serial numbers 66 through 73, inclusive). Placement of these control rods in Region 3 is not analyzed and does not satisfy the reactivity requirements. Because CEAs by design are strong neutron absorbers, storing a CEA in an assembly in a Region that does not credit CEAs is conservative.

an analysis addressing the effects of any consolidated fuel in the spent fuel pool This is addressed in Section 4.4.

address whether an increase or decrease is the limiting physical tolerance factor in storage cell inside diameter, rack pitch, and cell wall thickness.

This is addressed in the analyses of Section 6.1.

3.3 REFERENCES

3.1-1 NRC Interim Staff Guidance DSS-ISG-2010-01, Revision 0, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Nuclear Fuel Pools," 9/29/2011.

3.1-2 Kopp, L., NRC, memorandum to T. Collins, NRC, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 19, 1998.

(ADAMS Accession No. MLI 1088A013) 3.1-3 NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," February 2002. (ADAMS Accession No. ML003688150)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 32 3.1-4 NUREG/CR-6759, "Parametric Study of the Effect of Control Rods for PWR Burnup Credit,"

February 2002. (ADAMS Accession No. ML020810111) 3.1-5 NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses," March 2003. (ADAMS Accession No. ML031110292) 3.1-6 NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," January 2001. (ADAMS Accession No. ML050250061) 3.1-7 NUREG/CR-7108, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions," April 2012. (ADAMS Accession No. ML12116A124) 3.1-8 NUREG/CR-7109, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality kff Predictions," April 2012. (ADAMS Accession No. ML12116A128) 3.1-9 "Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty." EPRI, Palo Alto, CA:

2011. 1022909 3.1-10 "Utilization of the EPRI Depletion Benchmarks for Burnup Credit Validation." EPRI, Palo Alto, CA: 2012. 1025203 3.1-11 NUREG/CR-6760, "Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit,"

March 2002. (ADAMS Accession No. ML020770436) 3.1-12 Standard ANSI/ANS-8.27-2008, "Burnup Credit for LWR Fuel," American Nuclear Society, August 14, 2008.

3.1-13 Oak Ridge National Laboratory. SCALE Newsletter, Number 42 (Summer/Fall 2010).

3.2-1 "

SUMMARY

OF FEBRUARY 15, 2012, PRE-APPLICATION MEETING WITH DOMINION NUCLEAR CONNECTICUT, INC., TO DISCUSS A PROPOSED MILLSTONE POWER STATION, UNIT NO.2, LICENSE AMENDMENT REQUEST CONCERNING SPENT FUEL POOL CRITICALITY RE-ANALYSIS (TAC NO. ME7943)," Carleen J. Sanders (NRC), March 12, 2012. (ADAMS Accession No. ML120580362)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 33 4 SPENT FUEL STORAGE RACKS & COMPONENTS 4.1 SPENT FUEL POOL STORAGE CONFIGURATION DESCRIPTION This section provides a brief description of the spent fuel storage racks with the objective of defining the characteristics of the hardware that is used in the criticality analyses described in Section 6.

The existing Millstone Unit 2 spent fuel pool and the fuel storage rack types and orientation are illustrated in Figure 4.1-1. The fresh and spent fuel assembly storage area in the pool is divided into three regions:

Region A, Region B and Region C. Regions A and B employ a flux-trap design and a Boraflex panel in each storage cell for reactivity control. Region A and B are primarily employed to store fuel assemblies which have not yet achieved their expected discharge burnup (e.g. fresh fuel assemblies and low burnup fuel assemblies). Region C is employed to store discharged fuel assemblies (both consolidated and intact) which have achieved their expected discharge burnup. Region C does not contain any poison material for reactivity control.

Region A consists of one 8 x 10 module and two 8 x 9 modules with a nominal center-to-center spacing of 9.8 inches. The Region A storage racks can store up to 224 spent fuel assemblies.

Region B consists of two 8 x 10 modules with a nominal center-to-center cell spacing of 9.8 inches. Cell blockers are presently installed in 40 locations. The cell blockers are depicted in Figure 4.1-1 as darkly colored squares. As noted in Section 1.1, one of the objectives of this analysis is to demonstrate that cell blockers are no longer required to maintain fuel assemblies in an acceptable storage configuration.

Region C consists of fourteen modules with a nominal center-to-center spacing equal to 9.0 inches. There are two 8 x 7 modules, three 9 x 7 modules (with one missing 10 storage locations for the fuel elevator function), five 10 x 7 modules, three 11 x7 modules, and one 19 x 9 module. The Region C modules contain a total of 962 storage locations and are capable of storing both consolidated and intact fuel assemblies.

4.2 INDIVIDUAL STORAGE RACK TYPE DESCRIPTIONS Sections 4.2.1 and 4.2.2 describe the individual storage racks in greater detail.

4.2.1 Region A and B Storage Racks Each Region A or B storage cell is centered on a nominal [ ]a,c pitch and consists of an open stainless steel box. The stainless steel box has a nominal inside dimension equal to [ ]a'c and is [ ]a"' thick. Centered in the open stainless steel box is an open poison box which contains the Boraflex material. The inner poison box was constructed with an inner layer of stainless steel, a middle layer of Boraflex, and an outer layer of stainless steel. The inner and outer layers of stainless steel are each [ ]"' thick. The Boraflex layer is [ ]"'C thick and is separated from the stainless steel by a [ la,, void on each side. Overall, the inner box has a nominal inside

• dimension equal to [ ]a"c This analysis takes no reactivity credit for the Boraflex material. The dimensional data employed to model the Regions A and B storage racks is discussed in Section 6.1.1.1.

Millstone Unit 2 Criticality Analysis Report

cj~

0 z0 REGION B Q -T.Y...Y-N**

I Ii 0

0 0

z0 0

4 0

0 Oil 0 e° 0

z C0 I

TQ Cr CD a 0 I

a 0 I

I I tIi I: I I

-1 uuwrx0" A REGION C

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 35 4.2.2 Region C Storage Racks Each Region C storage cell consists of an open stainless steel box with a nominal inside dimension equal to [ ]c and a nominal thickness equal to [ ]a,c The nominal cell-to-cell spacing is

[ ]C Region C racks contain no poison inserts for reactivity control. The dimensional data employed to model the Region C storage racks is discussed in Section 6.1.1.1.

4.2.3 Revised SFP Rack Region Definitions Section 6.1 presents the detailed description for the revised SFP region definitions and the associated allowed storage patterns assumed in this analysis. The resulting region definitions and storage patterns are incorporated into the proposed Technical Specifications changes provided in the accompanying License Amendment Request (LAR). There are no physical changes being made to the existing spent fuel storage racks. The changes involve revised definitions of regions within the physical racks and required storage patterns to meet the objectives of the analysis.

The existing SFP storage rack regions (depicted on Figure 4.1-1) are redefined to create the region definitions shown on Figure 6.1-1 (in Section 6.1.1.1), in the following manner:

Region I is formed from the Western-most module of Region B + the SW module of Region A (these 2 sections of Region 1 are not contiguous)

Region 2 is formed from the 2 remaining modules of Region A + the Eastern-most module of Region B Region 3 is formed from all modules of Region C (except those forming Region 4-below)

Region 4 is formed from the 2 modules of Region C just East of Region A and B 4.3 MISCELLANEOUS COMPONENTS Control Element Assemblies (CEAs) and borated stainless steel pins (poison pins or rodlets) are credited in the MP2 SFP criticality analysis for Region 3. Poison pin modeling includes uncertainty cases for pin diameter and boron loading specified in design documents. CEAs are modeled assuming an extreme degree of depletion has occurred prior to discharge.

4.3.1 Borated Stainless Steel Poisons Pins Borated stainless steel poison pins may be employed in Region 3 to lower the burnup required to store discharged fuel assemblies in this region. The rodlets are modeled in this analysis as cylindrical objects with an outside diameter equal to 0.87 inches. When rodlets are used, three rodlets are required per fuel assembly, with I rodlet in the center guide tube, and the other 2 rodlets stored in diagonally opposite guide tubes. The rodlets are manufactured from borated stainless steel with a nominal boron concentration equal to 2.0 weight percent natural boron. The rodlets are modeled as starting three (3) inches above the bottom of the active fuel height. Fuel assemblies containing rodlets may be stored in any rotational Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 36 orientation. Poison pin modeling includes uncertainty cases for pin diameter and boron loading specified in design documents.

4.3.2 Control Element Assemblies (CEAs)

Region 3 burnup credit with CEAs is similar to the 3 poison pin Region 3 burnup credit development, but substitutes a depleted CEA in place of 3 poison pins. Full length, full strength CEAs are assumed. The CEAs are conservatively assumed to have absorber content [ ]b,e (Section 6.1.4.2.1).

4.4 NON-STANDARD STORAGE CONFIGURATIONS The analysis presented in this report is applicable to storage of fuel assemblies and the associated limitations that are governed by MP2 Technical Specifications. There are non-standard fuel configurations and components, and non-fuel containing components present in the MP2 SFP. These non-standard configurations and components may be stored in fuel assembly locations or in Restricted Locations (see Section 6.1) if they are demonstrated to be non-limiting with respect to the storage patterns that have been analyzed. The same methodology used for the analysis herein that established the Technical Specification requirements will be employed to evaluate non-standard configurations and components. These assessments would employ the requirements for documentation and implementation under the provisions of 10CFR50.59. Non-limiting is understood to mean that the non-standard configuration would meet these key limitations:

- Keff < 1.0 with 0 ppm boron (including biases & uncertainties)

- Keff < 0.95 with 600 ppm boron (including biases & uncertainties)

- Keff < 0.95 with 1400 ppm boron and the most limiting accident event (including biases & uncertainties)

- no change in the burnup requirements for the region in which the non-standard configuration is located Consolidated Fuel Storage Boxes (CFSB) are presently licensed to be stored in Region C. The licensing basis for CFSB storage requirements (burnup versus initial fuel enrichment) is based on a maximum Keff of 0.95, including bias and uncertainty, with no soluble boron [Reference 4.2-1, 4.2-2]. This existing basis is more restrictive than the basis for the revised analysis (maximum Keff < 1.0, including bias and uncertainty, with no soluble boron). Due to the large degree of conservatism associated with boron credit, no change is proposed for the CFSB storage requirements and thus the associated Technical Specifications limitations will remain in effect. The only change is that the region designation for storage of CFSBs (currently Region C) is Region 3 in the proposed TS.

4.5 REFERENCES

4.2-1 Letter from J. Alan Price (DNC) to USNRC, "Millstone Power Station, Unit No. 2, Technical Specifications Change Request (TSCR) 2-10-01, Fuel Pool Requirements," November 6, 2001.

(ADAMS Accession No. ML013510295) 4.2-2 Letter from Richard B. Ennis (NRC) to J. A. Price (DNC), "MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: SPENT FUEL POOL REQUIREMENTS (TAC NO. MB3386)," April 1, 2003. (ADAMS Accession No. ML030910485)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 37 5 NEW FUEL STORAGE ANALYSIS This analysis demonstrates that for all anticipated normal and abnormal configurations of fuel assemblies in the NFSR, the KIffis below the criticality design criteria of 0.95 at a 95 percent probability, 95 percent confidence level when flooded with full density unborated water, and does not exceed 0.98 at a 95 percent probability, 95 percent confidence level assuming optimum moderation conditions.

Key Analysis Assumptions

1) Neutron absorption in minor structural materials is

]b"e All of the stainless steel rack materials were modeled [

b,e The conservative nature of these approximations is demonstrated explicitly with test cases which are described in Section 5.3.

2) All fuel rods contain U0 2 at a maximum enrichment of 5.0 wt% U-235 over the entire length of each rod. No credit is taken for any natural or reduced enrichment axial blankets. This modeling is conservative because it results in more total fissile material in each fuel rod than is actually present, and because it bounds the MP2 initial nominal planar average enrichment limit of 4.85 wt% U-235.

3) The fuel pellets are modeled using a U0 2 density (96% of theoretical density) which bounds the fuel stack density of the evaluated fuel designs, after accounting for dish and chamfer volumes.

This is confirmed to be conservative in Section 5.3.

4) The calculations in this analysis were performed at two temperatures for postulated flooding water: 68°F for the lower bounding value and 212'F for the upper bounding value. Since results are more limiting for the 212'F case, water temperatures less than 68 0F do not need to be considered.

5.1 MODEL DESCRIPTION The new fuel storage racks were modeled using data on plant design drawings. The new fuel storage facility at Millstone Unit 2 has a total storage capacity of 76 fuel assemblies. The new fuel storage vault was modeled as a 32'9" x 16' x 11'4.7" room reflected with 30 cm of concrete on all six faces. As can be observed from Figure 5.1-1, there are eight 2x4 modules and three lx4 modules of storage cells. Each fuel assembly is spaced 20.5 inches center to center within each module. The edge to edge spacing between modules is 31 inches (East-West) and 28 inches (North-South).

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 38 Figure 5.1-1 KENO3D-Produced Plot of New Fuel Storage Racks (Top Half Removed)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 39 5.2 METHODOLOGY Using the model discussed previously, three accident conditions (misplaced/dropped fuel assembly, fully flooded and optimum moderation) were evaluated.

The calculated K.ff is determined for fully flooded and optimum moderation conditions as well as the misplaced/dropped fuel assembly accident. The maximum Keff for each case is determined from the KENO calculated Keff, the calculation bias, and the applicable uncertainties and tolerances (bias uncertainty, calculation uncertainty, fuel tolerances, and [ 1b,e using the following formula:

2 Maximum Keff = Calculated KIff+ biases + Yiuncertainties The effect of fuel assembly manufacturing tolerances on reactivity was evaluated in this analysis by performing sensitivity cases to calculate the tolerance effects. The reference condition for the tolerance cases is the condition with nominal dimensions and tolerances. The Ak associated with a specific fuel manufacturing tolerance is calculated by a comparison of the reference condition to the reactivity from a calculation with the tolerance included. The uncertainty associated with each of the sensitivity calculations is statistically combined and added to the calculated Kff as follows:

Ak= (kcalc - knominal) + 2 / . al All of the uncertainty Ak values are root sum squared (RSS) to determine the final reactivity allowance for fuel assembly manufacturing tolerances. Fuel tolerance analyses were performed [

Pbe In addition, a number of sensitivity studies were performed to verify that the assumptions and modeling simplifications made are conservative. These studies are as follows:

• Evaluation of the effect of spacer grids

" Evaluation of not modeling U-234 and U-236 (only U-235 and U-238 isotopes are considered)

• Evaluation of the sensitivity to the variation of the concrete thickness

• Evaluation of the effects of using the uniform Dancoff factors for the fuel rods near the large water holes, instead of using explicitly calculated values

" Evaluation of the effect of rack materials on criticality conditions Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 40 5.3 ANALYSIS RESULTS The Millstone Unit 2 NFSR was evaluated for both normal and abnormal configurations. This includes the reference configuration, fuel assembly tolerance cases, and water density variations with interspersed moderation. A brief summary of the results of the sensitivity studies is presented.

5.3.1 Uniform Interspersed Moderator The NFSR was modeled at both 293 K and 373 K with a variety of interspersed water densities to simulate varying amounts of water in the normally dry areas adjacent to the fuel rack modules. The interspersed water densities were modeled at various volume percentages between 0 % and 100 %, which covers any postulated density of flooding water (steam, foam, froth, etc.). The results of the interspersed moderation effects analysis is presented in Table 5.3-1 below and plotted in Figure 5.3-1.

Table 5.3-1 New Fuel Storage Racks Interspersed Moderation Effects Results Interspersed Water Density Knff (293 K) Keff (373 K)

(Volume %)

0 0.62827 +/- 0.00015 0.64899 +/- 0.00015 1 0.73258 +/- 0.00017 0.75131 +/- 0.00017 2.5 0.84596 +/- 0.00018 0.86543 +/- 0.00019 4 0.87010 +/- 0.00019 0.89521 +/- 0.00019 5 0.86078 +/- 0.00013 0.88921 +/- 0.00018 6 0.84468 +/- 0.00018 0.87537 +/- 0.00019 7.5 0.82152 +/- 0.00019 0.85291 +/- 0.00018 10 0.79299 +/- 0.00020 0.82513 +/- 0.00021 25 0.67748 +/- 0.00018 0.71204 +/- 0.00018 50 0.69710 +/- 0.00019 0.71734 +/- 0.00018 75 0.81365 +/- 0.00020 0.82779 +/- 0.00019 90 0.87684 +/- 0.00021 0.88885 +/- 0.00020 92.50 0.88625 +/- 0.00020 0.89830 +/- 0.00020 95 0.89614 +/- 0.00021 0.90781 +/- 0.00020 97.50 0.90502 +/- 0.00017 0.91664 +/- 0.00014 100 0.91390 +/- 0.00021 0.92561 +/- 0.00016 As can be observed from the results in Table 5.3-1 above, for both temperatures analyzed, the 100 volume

% moderator condition bounds the optimum hypothetical low density moderation condition (i.e. foam, fog, froth, etc.). Therefore, all of the subsequent cases are performed with 100 volume % moderation at both 293 K and 373 K to ensure that the maximum reactivity condition is captured in the analysis.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 41 109 0.95 C1 ...........

U*

0.9 U.

0.85 0.8 1293K w373K 0.75

[]

B 0 .7 .....................

0.65 0.6 I 0% 10% 20% 30% 40% 50% 60% 70%,6 80% 90% 100%

Interspersed Water Density (Volume %)

Figure 5.3-1 NFSR Keff vs Interspersed Water Density (Volume %)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 42 5.3.2 Sensitivity Cases Small amounts of U-234 and U-236 typically present in the fuel are neglected in this analysis. Three different test cases were run at both 293 K and 373 K with U-234 and U-236 concentrations of 0.05, 0.1, and 0.25 wt% and 100 volume % interspersed moderator to evaluate the reactivity effect of neglecting U-234 and U-236 in the NFSR models. All test cases confirm that it is conservative to neglect U-234 and U-236 in performing NFSR calculations.

The effect of modeling the NFSR with different concrete wall thicknesses was evaluated at both temperatures (293 K and 373 K) by performing calculations with wall thicknesses of 15 cm and 45 cm.

The results demonstrate that the reactivity effect of the NFSR concrete wall thickness is statistically insignificant. The assumed reference NFSR concrete thickness of 30 cm is therefore acceptable for this analysis.

The effect of assuming a uniform lattice of identical fuel rods in the resonance calculation (default for SCALE) was also investigated. Location specific Dancoff factors, which are used in the resonance Pe calculation, [

Results of this study [

]b,e in the vicinity of the guide and instrument thimbles.

To verify that the effect of not modeling the stainless steel rack materials is conservative, cases were run with stainless steel structure surrounding each of the 76 fuel assemblies. Cases were run at both temperatures with full flooding. The results confirm that addition of stainless steel structure reduced Keff.

The effect of spacer grids was also evaluated with KENO sensitivity cases. Cases were performed at 0 %,

4 %,and 100 % interspersed moderation at both 293 K and 373 K with Zircaloy grids added. Results Pe confirmed that it is [

5.3.3 Rack Deformation Rack deformation is not considered to be a credible accident scenario and is therefore not analyzed. As can be observed from Table 5.3-1, the Keff value is below 0.65 for dry conditions; therefore rack deformation of any kind could not cause the reactivity of the NFSR to exceed the regulatory limit except under flooded conditions, where the double contingency principle of ANS-8. l/N16.1-1975 [Reference 5.3-1] precludes analyzing simultaneous occurrences of multiple accident conditions.

5.3.4 Misplaced/Dropped Fuel Assembly It is possible that during movement of fuel assemblies in the new fuel storage area that a new fuel assembly may be inadvertently dropped into the storage area. It may either be dropped on top of an existing fuel assembly (either end on end axially or possibly lying horizontally on the rack) or in an area adjacent to the racks next to an existing new fuel assembly. The cases documented in this analysis were performed so as to maximize the severity of this accident by placing the misplaced fuel assembly face adjacent to an existing new fuel assembly.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 43 For the fuel assembly drop accident, a scenario exists where the fuel assembly can be compacted within the cell due to the weight of the assembly being dropped on top of it. The reactivity effect of this accident is calculated by increasing the fuel theoretical density to 120% and increasing the fuel rod diameter by 10% to simulate the compacting of the fuel. The active fuel height of the crushed fuel assembly was reduced by three feet to simulate the effect of crushing the fuel assembly. These values were selected somewhat arbitrarily and were selected to conservatively bound realistic amounts of compaction and densification. To bound both of the potential fuel assembly drop scenarios (fuel assembly dropped directly on top of another assembly creating a fuel region which is essentially twice the height of a single fuel assembly and dropping an assembly horizontally where it is lying across the top of the racks), the fuel assembly crush scenarios were run with mirror reflective boundary conditions on the upper and lower axial boundaries.

For the fuel assembly misplacement accident, a new fuel assembly is placed directly in contact with another new fuel assembly in a normally unoccupied area of the NFSR. Since the rack materials are not being modeled in this analysis, the two new fuel assemblies are placed in direct contact with each other in an area which would otherwise be occupied by rack material.

Results for both the dropped and misloaded fuel assembly are presented in Table 5.3-2.

Table 5.3-2 Results of Misplaced/Dropped Fuel Assembly Analysis Accident Scenario Keff (273 K) Keff (373K)

Single Misloaded Assembly 0.62968 +/-0.00016 0.65017 +/- 0.00015 Dry Conditions Single Crushed Fuel Assembly 0.63147 +/- 0.00015 0.65317 +/- 0.00015 Dry Conditions Nominal NFSR Configuration 0.62827 +/- 0.00015 0.64899 +/- 0.00015 5.3.5 Uncertainties Due to Fuel Assembly Tolerances As mentioned previously, the calculation of the final Kff includes an added uncertainty which accounts for the effects on reactivity of fuel assembly tolerances. The evaluations include tolerances in fuel assembly dimensions, as well as guide tube tolerances, enrichment uncertainty, and fuel rod pitch uncertainty. The methodology requires that the selected tolerance values bound the manufacturing tolerance for the parameter of interest. To obtain the reported RSS calculated result, the maximum positive value of Aka, for each case was used in the calculation. Cases were performed at flooded conditions at both temperatures (293 K and 373 K). The results are provided in Table 5.3-3 below.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 44 Table 5.3-3 Results of New Fuel Storage Rack Tolerance Calculations b, e 5.3.6 Calculation of Maximum Keff The maximum Keff values, based on the formula presented in Section 5.2, was calculated using the previously presented results. The results presented below in Table 5.3-4 show the maximum Keff values for both temperatures analyzed at 100 volume % moderator and confirm that Keff values are below the KIff limit of 0.98 at optimum moderation and the KIf limit of 0.95 when fully flooded with un-borated water at a 95% probability and at a 95% confidence level.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 45 Table 5.3-4 Results of New Fuel Storage Rack KENO Calculations Parameter Value Temperature (Kelvin) 293 373 Moderator Density (% Volume) 100% 100%

Uncertainties b, e KENO Code Bias Uncertainty KENO Calculation Statistics (95%/95%, 2a)

Calculated Tolerances Statistical Combination of all Uncertainties and Tolerances Calculated KENO Keff KENO Code Bias Maximum Kff 0.932291 0.942658 Regulatory Limit 0"957 0.95"

__________________________________________ I I______

1. Note that since the optimum moderation environment was found to be with full density water, the lower value (0.95) was used as the regulatory limit in Table 5.3-4. Both the optimum moderation and the full density water flooding criteria were met in this analysis.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 46 5.3.7 Conclusions and Results The acceptance criteria for criticality analyses of a NFSR requires that KIff be less than or equal to 0.95, including bias and uncertainties, when flooded with full density unborated water at a 95 percent probability, 95 percent confidence level and does not exceed 0.98 at a 95 percent probability, 95 percent confidence level assuming optimum moderation conditions.

This calculation demonstrates that the acceptance criteria are met for the Millstone Unit 2 NFSR for the storage of 76 fuel assemblies with a maximum initial planar average enrichment up to 5.0 wt% U-235.

The maximum Keff for the Millstone Unit 2 NFSR was calculated to be 0.942658 including all statistical biases and uncertainties.

5.4 REFERENCES

5.3-1 ANS-8.1-1975 (N16.1), "Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors."

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 47 6 SFP ENRICHMENT, BURNUP & BORON REQUIREMENTS This section discusses the KENO models and their usage in the analyses that determine enrichment, burnup and boron requirements for the SFP regions. Separate discussions are provided for details of the model and the bias and uncertainty calculations performed for use of the model.

6.1 ANALYSIS OF ENRICHMENT & BURNUP REQUIREMENTS Section 6.1 provides the detailed discussion of the analysis to determine the enrichment and burnup requirements that apply to each of the separate SFP regions.

Key Analysis Assumptions

1) Not all sensitivity and bias cases are run for all conditions in each SFP region. Items demonstrated for each rack type as nearly insignificant and/or reasonably insensitive to different conditions are assumed to have the same sensitivity as determined for a similar region. This is justified because these items have only a minor effect on total bias and uncertainty. Justification of adequacy for bias and uncertainty calculation data is provided in each section. Comparisons to data trends in a similar SFP region may be used; some cases assume a conservative value (e.g.,

fraction of total burnup worth represented by minor actinides and fission products).

2) Due to complexity and computer resource requirements, [

lb,e Development of the burnup credit requirements involves linear or cubic spline interpolation of available data. Justification is provided for the adequacy of cubic spline interpolation in Section 6.5. Conservative bias correction is performed to accommodate the use of linear interpolation between available enrichment points in Section 6.1.5.2.

3) Restricted storage locations are also referred to in this analysis as empty cells. Unless noted, empty cells are assumed to contain only SFP water inside the normal rack storage cell structure.

Fuel assemblies analyzed for storage are assumed to be standard fuel configurations. Analysis of non-standard fuel configurations and components and non-fuel containing components is discussed in Section 4.4.

6.1.1 Model Description and General Calculations 6.1.1.1 Model Description The MP2 SFP has two types of storage racks. A total of 384 storage locations are flux trap poison racks containing Boraflex panels. Boraflex is enclosed between thin stainless steel wrappers welded to comer supports that form a Boraflex box located inside each storage cell. The Boraflex boxes are removable.

The existing flux trap storage racks are subdivided into two regions (Region 1 and Region 2) that have different administrative requirements.

An additional 962 storage locations are in non-poisoned racks. The non-poisoned racks are subdivided into two regions (Region 3 and Region 4) that have different administrative requirements. Figure 6.1-1 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 48 shows the orientation of each region in the SFP. The allowed storage patterns for each region are as follows:

Region 1: 2-Out-of-4 checkerboard with two fuel assemblies and two empty cells.

• Region 1 Region 2: 3-Out-of-4 checkerboard with two higher reactivity fuel assemblies, one lower reactivity fuel assembly, and one empty cell

• Region 2 2A (Higher reactivity) 2B (Lower reactivity)

Region 3: 4-Out-of-4 fuel assemblies with either 3 borated stainless steel pins (rodlets) or a CEA 3 3 Region 3 Region 4: 3-Out-of-4 fuel assemblies with one empty cell

• Region 4 The existing arrangement of the MP2 spent fuel pool includes 40 blocking devices installed in selected storage locations in Region B (the dark squares on Figure 4.1-1). The analysis described herein no longer assumes the use of these blocking devices.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 49 Fuel rack dimensions used for this analysis are listed in Table 6.1-1. SFP storage rack models include these axial regions (described from bottom to top):

1) Concrete floor

2) Stainless steel SFP liner

3) Storage rack structure and water below the fuel assembly lower tie plate (assumed to be

]bPe I

4) Lower tie plate modeled as [

b,e

5) Fuel rod end plugs

6) Fuel rods

7) Fuel rod upper plenum

8) Upper tie plate region modeled as [ b,e

9) Storage cell walls

10) Water above storage rack Figure 6.1-2 is a visualization of the KENO model for a Boraflex rack model (single storage cell with fuel, top half removed). Figure 6.1-3 is a visualization of the KENO model non-poisoned rack model (single storage cell with fuel, top half removed, with 3 borated stainless steel rodlets). Figure 6.1-4 is a cutaway view of the Boraflex model showing detail near the bottom and top of the fuel with various materials removed for clarity.

Millstone Unit 2 Criticality Analysis Report

6Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 50 1~ M M M 2 M*2* 3 313131313131313131 31313 4444444 4 3 3 3 3 3 3 333 a a 0 M M9A22M M 4 3313333333 3 3 I

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_3 3 3 3 3 33 3 31 3 3 3 3 3 3 31 3 3 313 213 a n MI a 333333333O 12A 313 3 3 3 3 3 *3 9"3 3I 3 i a 0 U 333333333 3 3133 3 13 33f 3 3* 1.3333 33 11 U B a " HHHH"Ol33 1 3- 34 3-i- 3 1 i 3333 2 3131313131 WTF I Woin i Mf27 Fbicn 2 LUT.

Ž2 Fbgion 3 E M4 A Igion A [ Fuet Elwaw Re*tricId Locafion*

MN~ 2AM2 A Restricted Location may contain non-standard f"e configurations or components, or is empty Figure 6.1-1 MP2 SFP Spent Fuel Pool Layout Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 51 Table 6.1-1 MP2 Spent Fuel Storage Rack Specifications Region Design Feature, Boraflex Racks Value 1,2 Storage Cell Pitch (cm) [a,c 1,2 Storage Cell lID (cm) [ ]a"c 1,2 SS cell wall thickness (cm) ]a'c 1,2 SS cell wall height (cm) [ ]a' 1,2 Boraflex cover sheet ID (cm) [ ]a,c 1,2 Boraflex cover sheet length (cm) [ ]a,c 1,2 Gap between Boraflex cover sheet and Boraflex (cm) [ 1c 1,2 Boraflex thickness (cm) I ]ac 1,2 Boraflex width (cm) [ ]ac 1,2 Poison box ID (cm) [ ]ac 1,2 SS Boraflex cover sheet thickness (cm) I]'c 1,2 SS angle bracket edge width (cm) [ ]a,c 1,2 SS angle bracket thickness (cm) [ ]a,c 1,2 SS angle bracket length (cm) [ ]ac 1,2 Storage cell height (cm) [ ]ac 1,2 Storage cell height at bottom of fuel assembly (cm) [ ]a,c 1,2 SFP SS liner thickness (cm) 1]c

]

1,2 Concrete floor thickness (arbitrary, cm) [ ]aC Region Design Feature, Non-poison racks Value 3,4 Cell Pitch (cm) [a Cell ID (cm) ]a,c 3,4 3,4 SS cell wall thickness (cm) [ ]ac 3,4 Storage cell height (cm) [ a,c 3,4 Storage cell height at bottom of fuel assembly (cm) [ 12C 3,4 Borated SS rodlet diameter (cm) 2.2098 [d Borated SS rodlet length (cm) 387.35 d 3,4 3,4 Borated SS rodlet boron content (wt%) 2.0 [

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 52 Figure 6.1-2 Region 1 and 2 Rack Cell KENO Model Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 53 Figure 6.1-3 Region 3 and 4 Rack Cell KENO Model (Region 3 Fuel With Rodlets Shown)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 54 Figure 6.1-4 Region 1 and 2 Storage Cell KENO Model (Axral Views)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 55 6.1.1.2 Calculation of Enrichment and Burnup Requirements with No Soluble Boron A general outline of bumup credit methodology is presented in the introductory Section 3 of NUREG/CR-7109 [Reference 6.1-1]. The generalized K~ff equation reduces to this:

K~ff (predicted) + Bias + Uncertainty < Klimit For the calculation of enrichment and burnup requirements (burnup credit), Klimit is 1.0. Keff bias and uncertainty terms are composed of multiple components:

1) Bias components considered P~e

2) Uncertainty components considered b,e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 56 Figure 6.1-5 TRITON Fuel Assembly Model Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 57 Enrichment and burnup requirements are determined for each region using regional SFP rack models.

Confirmation of the absence of region-to-region interface effects is performed using a full pool model. To determine the regional enrichment and burnup requirements, the following general process is used:

1) Perform TRITON depletions to produce candidate sets of 18 axial node fuel nuclide compositions. Composition data is calculated for [

Pbe

2) Run KENO SFP Keff cases with 0 ppm soluble boron. [

b,e

3) Run KENO bias and uncertainty cases with fresh and depleted fuel at 0 ppm soluble boron.

Additional cases are run with 2000 ppm soluble boron as needed to determine appropriate bias and uncertainty for use in determination of soluble boron required for K,.ff < 0.95.

4) Run KENO cases to determine burnup worth for use in calculating bias and uncertainty values related to bumup worth (burnup measurement uncertainty, burnup worth uncertainty, and minor actinide and fission product worth bias).

5) Select limiting bias and uncertainty results and combine them to determine the limiting best-estimate KENO Kff, (the target Keff) which is a function of bumup.

6) Target Keff = 1 - bias - uncertainty be Although evaluations are performed with up to 5.0 wt% U-235 initial nominal planar average enrichment, the Technical Specification limit of 4.85 wt% is not being changed. Calculated burnup requirements at 5.0 wt% are provided for use in preparing plots and tables for administration of the bumup credit requirements.

For uncertainty calculations using KENO results, sensitivity or uncertainty values are calculated as follows:

Sensitivity (AK) = (K2-K1) + 2 x RSS (K1SD, K2SD)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 58 where KI is the base case KENO K~ff, K2 is the sensitivity case Keff, RSS is the root sum square, K1SD is the KENO Keff standard deviation for the base case, and K2SD is the KENO K ff standard deviation for the sensitivity case. Sensitivity values treated as bias are calculated as K2-KI. Significance is defined as K2 - K1 > K1SD. All sensitivity cases are treated as uncertainty except for temperature change, grid effects, and boundary conditions (rodlet orientation in Region 3) which are treated as bias.

6.1.1.3 Calculation of Bias and Uncertainty - Specific Bias and Uncertainty Values Table 6.1-2 lists specific values associated with each bias (B) or uncertainty (U) component. Fuel in fuel rods in the MP2 KENO SFP models is modeled as a solid right circular cylinder. Fuel density uncertainty is composed of variations in fuel pellet density and variations in dish and chamfer volume. A value of +/-1.6% bounds the RSS effect of pellet density manufacturing tolerance and dish and chamfer volume uncertainty.

Fuel assemblies used at MP2 typically feature reduced enrichment fuel rods adjacent to guide thimbles in the central zone (a non axial blanket region comprising roughly 90% of the fuel stack) to control local power peaking. For simplicity, MP2 SFP criticality calculations are performed using a single enrichment in all fuel rods equal to the simple average enrichment of all fuel rods in the central zone. This averaged enrichment is designated as the initial nominal planar average enrichment. The impact of this simplifying assumption on SFP Keff was investigated using TRITON fuel depletions and KENO models of muti-enrichment and single-enrichment fuel assembly designs. For Regions 1 and 2, the single enrichment model produced either the highest SFP K~ff or a KIff within the uncertainty of the calculation when compared to the more exact multi-enrichment model. For the Region 3 and 4 racks, a few statistically significant differences of [ ]b,e were calculated such that the single pin simplification could be slightly non-conservative. A bias of [ ]b, is applied to Region 3 and 4 bias and uncertainty calculations to bound this effect.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 59 Table 6.1-2 Specific Bias and Uncertainty Values

]bA

[ ]a,c

[ ]ac

[ ]a,c

[ ]a,c

[ ]A

]A

[ []A J

[]A

]A Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 60 6.1.2 Region 1 Analysis 6.1.2.1 Calculation of Bias, Uncertainty and Burnup Credit - Region 1 SFP rack Kff results for Region 1 are shown in Tables 6.1-3 through 6.1-5. Bias and uncertainty items with significant reactivity impact are shaded. Limiting cases are designated with bold print. Table 6.1-6 is the calculation of total bias and uncertainty. Bias items are added together, uncertainty items are combined by RSS. Bias and uncertainty for 2000 ppm is provided for use in boron credit calculations for Keff < 0.95 with soluble boron. Figure 6.1-6 is a 2-D X-Y plot of a Region 1 off-center fuel placement model. The results for Region 1 demonstrate the following:

1) Dominion administrative margin of [ ]b and NRC administrative margin of I ]b has been explicitly included in the total bias and uncertainty calculation.

2) With poison boxes installed, fresh fuel with 4.85 wt% U-235 may be stored in a 2 out of 4 checkerboard. The maximum best estimate Keff for this configuration is 0.9086 with 5.0 wt%

fuel, which is well below the maximum allowable best estimate Kn [ lb" Additional margin for Region 1 may be obtained by modeling 4.85 wt% fuel.

3) There are no burnup requirements for Region 1.

4) Poison boxes are worth roughly 3% AK, which is much less than the available Region 1 Kff margin [ ]b,e If poison boxes are removed from Region 1 cells in which fuel will be stored, the uncertainty case for off-center fuel must be re-evaluated and total bias and uncertainty re-calculated.

5) With 2000 ppm soluble boron, total uncertainty and bias is [

be

6) Zircaloy grids have a Pbe Table 6.1-3 Region 1 Best Estimate Keg Initial KENO KENO Enrichment Poison Burnup K- K-effective (w/o U235) Box (GWd/MTU) effective Deviation 5.00 YES 0 0.90859 0.00021 5.00 NO 0 0.93407 0.00022 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 61 Table 6.1-4 Region 1 Bias and Uncertainty KENO Cases (0 ppm boron) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 62 Table 6.1-5 Region 1 Bias and Uncertainty KENO Cases (2000 ppm boron) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 63 Table 6.1-6 Region 1 Bias and Uncertainty b,e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 64 Figure 6.1-6 Region 1 KENO Off-Center Fuel Placement (Inward)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 65 6.1.3 Region 2 Analysis 6.1.3.1 Model Description Region 2 fuel storage racks are identical to Region 1 fuel storage racks. Region 2 uses a repeating 3-Out-of-4 storage pattern shown in Figure 6.1-7. In the Figure, the designation 2A stands for higher reactivity fuel (typical of fuel that has been used in one operating cycle) and 2B stands for lower reactivity fuel (typical of fuel that has been used in two operating cycles).

EMPTY 2A 2A 2B Figure 6.1-7 Region 2 3-Out-of 4 Storage Pattern 6.1.3.2 Calculation of Bias, Uncertainty and Burnup Credit - Region 2 Region 2 has the most complex analysis of the four regions because two interdependent burnup credit curves are developed for Region 2, one for the 2A fuel and one for the 2B fuel. The curves are interdependent because both types of fuel contribute to the Keff of the 2x2 region.

Development of the two curves is both iterative and strategic. The process is strategic because the goal is to maintain adequate storage to manage the population of fuel projected to be present during future re-fueling campaigns. The process is iterative, because an assumption must be made about one curve to develop the other due to the computer time resources required.

The general process used for determination of burnup requirements is as follows:

1) Develop the 2x2 Region 2 SFP rack model with fresh fuel.

2) Estimate the fresh fuel enrichments for 2A and 2B fuel based on target Kff calculated for Region 1 (same rack design).

3) Run bias and uncertainty cases using the KENO 2x2 fresh fuel model.

[

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 66 bPe SFP rack KENO Region 2 best-estimate Keff results are shown in Table 6.1-7. These results are the primary Keff values used for burnup curve enrichment and burnup requirement determination. Two fuel enrichment points (initial nominal planar average) [

]b,e such that each point on the Region 2 burnup credit curve may be obtained by linear interpolation on enrichment or by linear interpolation on burnup (5.0 wt%).

Also indicated in the table is the total burnup worth. Burnup worth is used in Table 6.1-11 to provide uncertainty and bias values related to burnup worth (minor actinide and fission product worth uncertainty, burnup worth uncertainty, and assembly burnup uncertainty). The fraction of total burnup worth represented by minor actinides and fission products was not directly calculated for Region 2, but is derived from the Region 3 calculation using similar depleted fuel compositions.

Total burnup worth includes the effect of all TRITON depleted fuel actinides and fission products after five days of decay following irradiation. Major actinide worth includes only the effect of U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241. As detailed for Region 3, in the burnup range 10 to 50 GWd/MTU, the maximum minor actinide and fission product share of total burnup worth is [ ]b,e and occurs at [ ]b,e The conservative [ 1b,e value is used in the Region 2 depleted fuel bias and uncertainty calculation for all burnups for convenience.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 67 Table 6.1-7 Region 2 Best Estimate Keff b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 68 Burnup worth in Table 6.1-7 is conservatively calculated for use in the total bias and uncertainty calculation as Burnup worth = (K1-K2) + 2 x RSS (KSDI, KSD2) where KI is the 0 burnup KENO Kff, K2 is the depleted fuel Kff, RSS is the root sum square, KSD1 is the KENO Keff standard deviation for the fresh fuel case, and KSD2 is the KEN() K~ff standard deviation for the depleted fuel case. Burnup worth is a strong function of burnup, but a weak function of initial fuel enrichment. In addition, burnup worth of cases with depleted 2A and 2B fuel in combination (bottom portion of Table 6.1-7) is significantly less than the sum of the individual burnup worth cases (top portion of Table 6.1-7). This confirms that burnup related bias and uncertainty components may be conservatively calculated using the sum of the burnup worth from the single burnup curve cases to approximate the total burnup worth that would be present in combination cases with depleted 2A and 2B fuel.

Region 2 bias and uncertainty results are shown in Table 6.1-8 for fresh fuel with no soluble boron.

Table 6.1-9 contains analogous cases run at high burnup credit conditions. Several of the less significant uncertainty cases were not re-run for Region 2 because the fuel and rack design are the same as Region 1.

Figure 6.1-8 shows a 2-D (X-Y) plot of a KENO off-center fuel model.

bke Table 6.1-9 depleted fuel biases and uncertainties are slightly larger than corresponding fresh fuel values.

A fuel density increase case was not run for Region 2 depleted fuel, however, a Region 3 depleted fuel density increase case indicates a slightly smaller density sensitivity for depleted fuel. The dominant terms in the uncertainty calculation for depleted fuel are [

]b,e Other values make small contributions to the RSS total uncertainty. [

Pbe Table 6.1-10 contains bias and uncertainty results for fresh and depleted fuel with 2000 ppm soluble boron. All significant cases from the 0 ppm evaluation are included for fresh fuel and the three most significant 2000 ppm fresh fuel cases are provided for depleted fuel. Taken together, the 2000 ppm boron results indicate very little change in total uncertainty versus soluble boron, but a significant increase in bias with increased boron driven by the water temperature sensitivity. Total burnup worth with 2000 ppm soluble boron is significantly less than at 0 ppm (0.102 AK in Table 6.1-10 versus 0.136 AK using data in Table 6.1-7). Essentially all of the difference in total bias and uncertainty for 2000 ppm versus 0 ppm boron is attributable to temperature bias, and essentially all of the increase with burnup is attributable to the burnup worth bias and uncertainty items.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 69 Figure 6.1-8 Region 2 KENO Off-Center Fuel Placement (Inward)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 70 k b,e Table 6.1-8 Region 2 Bias and Uncertainty KENO Cases (0 ppm boron, fresh fuel) -

Table 6.1-9 Region 2 Bias and Uncertainty KENO Cases (0 ppm boron, depleted fuel) b, e Table 6.1-10 Region 2 Bias and Uncertainty KENO Cases (2000 ppm boron) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 71 Table 6.1-11 is the calculation of total bias and uncertainty versus burnup for Region 2 with 0 ppm soluble boron. For each burnup in Table 6.1-11, bumup worth related uncertainty and bias components are calculated as a function of burnup worth (Burnup worth, Burnup, 1.5% minor actinides and fission products). Enrichment uncertainty is a function of enrichment. Other bias and uncertainty components used the appropriate fresh or depleted fuel value or the more conservative of the two for intermediate burnup cases. A comparison of Region 2 bias and uncertainty for 0 and 2000 ppm soluble boron is in Table 6.1-12.

Enrichment uncertainty is interpolated from Table 6.1-7 and 6.1-8 data using cubic spline interpolation.

Enrichment sensitivity Figure 6.1-9 demonstrates the adequacy of the interpolation technique. Burnup uncertainty is equal to the difference between the burnup worth at the indicated burnup and the cubic spline interpolation of burnup worth versus burnup evaluated at 96.5% of the indicated burnup. Minor actinide and fission product worth is conservatively calculated as [ 1 b,e of the total burnup worth.

Using the target KIff values from Table 6.1-11, enrichment limits are determined by linear interpolation versus enrichment. [

]b,e Table 6.1-13 shows the results of four combination depleted fuel cases. For each combination case a double linear interpolation of the four bounding cases was performed using enrichment and burnup combinations allowed by the 2A and 2B burnup credit curves. The resulting interpolated Kff was compared to the target Keff. One of the combined burnup cases had an interpolated Keff greater than the target Keff.

[

1b,, which was sufficient for all cases to meet the actual target Keff. The combined burnup case with the least Keff margin to the target Keff is for 5.0 wt%

fuel (16.12 GWd/MTU for Type 2A and 34.1 GWd/MTU for Type 2B). This 5.0 wt% case bounds the allowable initial nominal planar average enrichment for MP2 (4.85 wt% U-235) and provides assurance that at the maximum allowable enrichment, the Region 2A and 2B burnup credit curves are sufficiently conservative. Figure 6.1-10 illustrates these results, and confirms that the target Keff is [

]b,e Table 6.1-14 contains the resulting Region 2 burnup credit values.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 72 Table 6.1-11 Region 2 Bias and Uncertainty (0 ppm boron) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 73 Table 6.1-12 Region 2 Bias and Uncertainty Comparison (0 and 2000 ppm boron) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 74 Region. 2 Enrichment Sensitivity Spline Interpolation Confirmation 0.07 0 06i

.*0 .0 6 .. . ..........................................

0.04 0 .03 ..................................................

0OM .............. ...

0 .0 1 ..... ...

1.5 2.0 2,5 3.0 3.5 4.0 CS 5.0 Initial Fuel Enrichment (w/o U-235)

Figure 6.1-9 Region 2 Enrichment Sensitivity Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 75 b, e Table 6.1-13 Region 2 Combination Depleted Fuel Cases Table 6.1-14 Region 2 Burnup Credit Region 2 Region 2 Region 2 2A Burnup Region 2 2B Burnup 2A Enrichment (GWd/MTU) 2B Enrichment (GWd/MTU) 2.95 0 1.95 0 4.22 10 2.64 :10 5.00 16.12 3.61 20 4.60 .30 5.00 34.1 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 76 b, e Figure 6.1-10 Region 2 Keff vs Total Burnup Credit Worth Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 77 6.1.4 Region 3 Analysis 6.1.4.1 Calculation of Bias, Uncertainty and Burnup Credit - Region 3 with Rodlets Two independent burnup curves are developed for Region 3. The first is for use with three borated stainless steel poison pins (rodlets) inserted and is detailed in this section. The second is for use with a CEA inserted.

SFP rack KENO Kff results for the Region 3 rodlet cases are shown in Table 6.1-15. Also indicated in the table are the total bumup worth, fraction of total burnup worth represented by the minor actinides and fission products, and enrichment sensitivity. These worth and sensitivity values are used in Table 6.1-20 to provide uncertainty values related to enrichment or burnup worth.

Region 3 bias and uncertainty results are shown in Table 6.1-16 for fresh fuel with no soluble boron.

Table 6.1-17 contains analogous cases run at near maximum burnup credit conditions (5.0 wt% fuel with 50 GWd/MTU bumup). Items with significant reactivity impact are shaded. Limiting cases are designated with bold print. Table 6.1-17 depleted fuel biases and uncertainties are bounded by the corresponding fresh fuel value. A plot (2-D X-Y) of the off-center fuel placement models are shown in Figure 6.1-11. Because the diagonally installed rodlets can be placed in the same or opposite orientations in neighboring assemblies, two different models are used. The Kff difference associated with reflective versus periodic poison pin diagonal placement is taken as a bias in Tables 6.1-20 and 6.1-22.

Table 6.1-18 and Table 6.1-19 contain bias and uncertainty results for fresh and depleted fuel, respectively, with 2000 ppm soluble boron. Table 6.1-20 is the calculation of total bias and uncertainty versus burnup for Region 3 (3 poison pins) with 0 ppm soluble boron. A bias term [

]b e Administrative Keff margin of [ ]b is included in these values

]b Using the target K~ff values from Table 6.1-20, burnup credit values shown in Table 6.1-21 were determined by linear interpolation on enrichment.

Total bias and uncertainty is calculated in Table 6.1-22 with 2000 ppm soluble boron. Enrichment sensitivity and bumup worth are assumed to be the same as for 0 ppm boron. Based on the enrichment sensitivity in Table 6.1-16 and Table 6.1-18, increased boron slightly reduces enrichment sensitivity, making the assumption of no change conservative. Burnup worth is reduced roughly [ ]b with 2000 ppm soluble boron, making the assumption of no change in burnup worth versus soluble boron in Table 6.1-22 conservative. The burnup worth and enrichment conservatism is made for convenience. Given these conservative assumptions, total bias and uncertainty with 2000 ppm boron would be slightly smaller than for 0 ppm except for a substantial increase in [ Pb,e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 78 Table 6.1-15 Region 3 Burnup Credit KENO Cases (3 rodlets) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 79 Table 6.1-16 Region 3 Fresh Fuel Bias and Uncertainty (0 ppm boron, 3 rodlets) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 80 1i 000000000 100000 t ZOO

ýC00 00ý000 000ý00 00o(~ 000000 000 0000, 0,0 00 oooooooooaoo0oC)C /0oonoooo ooo 000000000000 M)0$,0-7-C0100OO 000C0000000 10000000000000 100000lap000 o 0100000000co ,-I0000ýo OOOOD 0 0 0 000 0000001" 0000000 600,000 000~ 00 1000 0~00000 000

-0 C000000C0000000 00 00 0 00 0 r?[5000a000 00(D000000 C0O0@00 0 0100010 0 00 1>00,0000 00 00 cc0ý00000 00(

O1@ 000@

oooooooT 10000ý

- 00 OG 1oO0C,00010000a0o 000&00 0ý010(0(Do00oa 00!00C0,00000000 0000000000000 00C (:00000000 V000o0Oooo0oo0 oooo0 COCCooooo~oolo Sg ::o 00000o,ý--00000000000 :000000000CH,-0 0 C0100:

-0 00 01(o'00000000 D0000,000000000 061opoDocccoloo 0,00000000!000,00 200c00000000o IIL Figure 6.1-11 Region 3 Off-Center Fuel Model with Poison Pins Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 81 Table 6.1-17 Region 3 Depleted Fuel Bias and Uncertainty (0 ppm boron, 3 rodlets) b, e Table 6.1-18 Region 3 Fresh Fuel Bias and Uncertainty (2000 ppm boron, 3 rodlets) bye Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 82 Table 6.1-19 Region 3 Depleted Fuel Bias and Uncertainty (2000 ppm boron, 3 rodlets) b, e Table 6.1-20 Region 3 Bias and Uncertainty (0 ppm soluble boron, 3 rodlets) be Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 83 Table 6.1-21 Region 3 Burnup Credit (0 ppm soluble boron, 3 rodlets)

__ _ eU, Burnup (GWd/MTU) Enrich Limit (w/o U-235) 0 1.620 10 2.213 20 2.724 30 3.315 40 4.231 50 4.982 Table 6.1-22 Region 3 Bias and Uncertainty (2000 ppm soluble boron, 3 rodlets) b, Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 84 6.1.4.2 Calculation of Bias, Uncertainty and Burnup Credit - Region 3 with CEAs 6.1.4.2.1 Conservative Modeling of Discharged CEAs Region 3 burnup credit with CEA credit is similar to the 3 poison pin Region 3 bumup credit development, but substitutes a depleted full length CEA in place of 3 poison pins. For this analysis, CEAs are conservatively assumed to have absorber content depleted [ ]b,e This amount of depletion reduces CEA worth by over [

b,e Based on industry CEA failure studies and CEA fluence limits intended to preclude CEA failure, B4 C depletion of [ Id in any part of the CEA B4C stack is expected. TRITON modeling performed for MP2 confirmed that depletion effects in Silver-Indium-Cadmium portions of the CEA are not significant. KENO results for Region 3 burnup credit calculations indicate [ ]b,e conservatism in the CEA burnup credit calculation using [ ]b,e The assumption of [ 1b,, is acceptable because it bounds any expected depletion in new or irradiated CEAs.

6.1.4.2.2 Region 3 SFP Rack Model Results with Discharged CEAs SFP rack KIff results for the Region 3 CEA cases are shown in Table 6.1-23. These results are the primary K*ff values used for burnup curve enrichment and burnup determination. [

]b,e In the process of developing the burnup credit curve, extrapolation of the burnup credit curve versus enrichment produced a higher burnup requirement than burnup interpolation at 5.0 wt%. Rather than running additional cases to justify the lower result, the higher extrapolated value was used. Table 6.1-23 also indicates the total burnup worth and enrichment sensitivity.

Region 3 CEA bias and uncertainty results are shown in Table 6.1-24 for fresh fuel with no soluble boron.

Cases from the Region 3 rodlet analysis with significant Keff changes were run for the CEA cases. No cases with higher enrichment depleted fuel were run due to the similarity of these results to fresh fuel results for the rodlet analysis. Analogous cases for 2000 ppm soluble boron are shown in Table 6.1-25.

Table 6.1-26 is the calculation of total bias and uncertainty versus burnup for Region 3 (CEA) with 0 ppm soluble boron. Uncertainty results not directly calculated for Region 3 with CEAs were retained from the Region 3 rodlet analysis. A bias term [ ]b,e to accommodate use of a single assembly average enrichment rather than modeling the assembly with multiple enrichment rods is included. Because the only change to the Region 3 rodlet model for the CEA credit calculation is the exchange of rodlets and depleted CEAs, minor actinide and fission product worth was not calculated directly. Instead, the largest fraction from the Region 3 rodlet cases [ .]b,e was conservatively used for all burnups. Results of the CEA Target Keff calculation are very similar to the Region 3 rodlet calculation.

Table 6.1-27 Contains the calculation of total bias and uncertainty versus burnup for Region 3 (CEA) with 2000 ppm soluble boron. As with the 3 rodlet results, bias and uncertainty is nearly the same as for 0 ppm Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 85 except for the bias due to temperature. Using the target Kef values from Table 6.1-27, burnup credit values for 0 - 30 GWd/MTU shown in Table 6.1-28 were determined by linear interpolation on enrichment. The 40 GWd/MTU value was determined by bumup curve extrapolation as discussed.

Table 6.1-23 Region 3 Burnup Credit KENO Cases (CEA) b, e Table 6.1-24 Region 3 Fresh Fuel Bias and Uncertainty (0 ppm boron, CEA) b e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 86 Table 6.1-25 Region 3 Fresh Fuel Bias and Uncertainty (2000 ppm boron, CEA) b, e 7

Table 6.1-26 Region 3 Bias and Uncertainty (0 ppm soluble boron, CEA) b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 87 Table 6.1-27 Region 3 Bias and Uncertainty (2000 ppm soluble boron, CEA) b, e Table 6.1-28 Region 3 Burnup Credit (0 ppm soluble boron, CEA) b, e Burnup (GWd/MTU) Enrich limit 0 2.257 10 3.043 20 3.845 30 4.610 35.1 5.000 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 88 6.1.5 Region 4 Analysis 6.1.5.1 Calculation of Bias, Uncertainty and Burnup Credit - Region 4 The Region 4 storage rack design is identical to Region 3. In Region 4, no rodlets or CEAs are credited.

Instead, a 3-Out-of-4 repeating storage pattern is used to control Keff so that less burnup credit is required for fuel storage. Figure 6.1-12 is a cutaway view (top half removed) of the KENO Region 4 model.

SFP rack KENO KIff results for the Region 4 cases are shown in Table 6.1-29. These results are the primary Keff values used for burnup curve enrichment and burnup requirement determination. Also indicated in the table are the total burnup worth and enrichment sensitivity.

Region 4 bias and uncertainty results are shown in Table 6.1-30. Major bias and uncertainty items from the Region 3 analysis are provided for fresh fuel with no soluble boron. Depleted fuel (5.0 wt%, 40 GWd/MTU) [

]b,e insensitive to burnup as indicated in the Region 3 results. A 2000 ppm temperature sensitivity case and a 2000 ppm grid case are also provided to confirm the boron sensitivity seen in Region 3 results. Items with significant reactivity impact are shaded in Table 6.1-30. Limiting cases are designated with bold print. Bias and uncertainty results are similar to those for Region 3.

Table 6.1-31 contains the calculation of total bias and uncertainty versus burnup for Region 4 with 0 ppm soluble boron. Because the only change to the Region 3 model is the removal of poison pins and one empty location in each 2x2, minor actinide and fission product worth was not calculated directly. Instead, the largest fraction from the Region 3 poison pin cases [ Pb,e was conservatively used for all burnups for convenience. Total bias and uncertainty for Region 4 is smaller than Region 3 (rodlets) by approximately the amount of the rodlet orientation bias (-0.003 AK).

Using the target K~ff values from Table 6.1-3 1, enrichment limits shown in Table 6.1-32 were determined.

Administrative Kerr margin of [ ]b is included in these values [

]b Region 4 storage racks are the same as those analyzed in Region 3.

Results in Table 6.1-30 confirm that the trends for Region 4 with fuel burnup and soluble boron are analogous to those in Region 3. Because of this similarity, the same Region 3 (rodlets) total bias and uncertainty increase for boron credit calculations is applicable to Region 4.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 89 Figure 6.1-12 Region 4 3-Out-of-4 Model Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 90 Table 6.1-29 Region 4 Burnup Credit KENO Cases (No Poison Pins) b, e Table 6.1-30 Region 4 Bias and Uncertainty Cases ), e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 91 Table 6.1-31 Region 4 Bias and Uncertainty (0 ppm soluble boron) e Table 6.1-32 Region 4 Burnup Credit (0 ppm soluble boron) b, e Burnup (GWd/MTU) Enrichment limit (w/o) 0 1.739 10 2.506 20 3.134 30 3.854 40 4.958 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 92 6.1.5.2 Adjustment of Burnup Credit for Potential Linear Interpolation Bias Linear interpolation is used to determine the fuel enrichment at which the target Keff is met for Region 2, 3, and 4 burnup credit curves. The goal of each interpolation is to determine the enrichment that produces a best-estimate Keff equal to the target Kff [ 1b e Review of the burnup credit results for Regions 2, 3, and 4 shows that Region 3 with rodlets has the largest range of enrichment (roughly 1.6 wt% to 5.0 wt%) and the highest enrichment sensitivity (AK/wt%). Because Keff is somewhat non-linear versus enrichment, using linear interpolation may introduce a small bias, depending on how far apart the enrichment values are and how far the target Keff is from the nearest KENO case Kffr.

Table 6.1-33 contains KENO Keff versus initial enrichment data for Region 3 with 0 ppm soluble boron.

Also calculated in Table 6.1-33 is the maximum interpolation error resulting from using linear interpolation with known data points that are [ 1b,e Interpolation error is the difference between calculated enrichment using linear interpolation to a target Keff midway between the data values and the calculated enrichment obtained by cubic spline interpolation when the midpoint enrichment Keff value is known. Figure 6.1-13 shows that the interpolation error is a function of enrichment bounded by the following equation:

Bias (wt%) [ ]b,e Interpolation gap refers to the difference between the linear interpolation enrichment and the nearest data point being used for the interpolation.

Correction of Region 2, 3, and 4 burnup credit tables is shown in Table 6.1-34.

Table 6.1-33 Region 3 Linear Interpolation Enrichment Bias b,e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 93 b, e Figure 6.1-13 Region 3 Linear Enrichment Interpolation Bias Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 94 Table 6.1-34 Burnup Credit Requirements (Regions 2 - 4)

Linear Interpolation Corrected Enrichment Interpolation Interpolation Enrichment Region (w/o) Gap (w/o) Bias (w/o) (w/o) Burnup Comments 2 (2A) 2.95 0.00 0 2.95 0 2 (2A) 4.22 0.22 0.01 4.21 14)

Point obtained by burnup credit curve 2 (2A) 5.00 0.00 0 5.00 16.27 extrapolation**

2 (213) 1.95 0 0 1.95 0 2 (213) 2.64 0.14 0.01 2.63 10) 2 (213) 3.61 0.11 0.01 3.61 20 2 (213) 4.60 0.10 0.00 4.59 30 Point obtained by burnup credit curve 2 (213) 5 0 0 5.00 34.16 extrapolation**

3 (rodlets) 1.62 0.09 0.01 1.61 0 3 (rodlets) 2.21 0.21 0.02 2.19 10 3 (rodlets) 2.72 0.22 0.02 2.71 20 3 (rodlets) 3.32 0.18 0.01 3.30 30 3 (rodlets) 4.23 0.23 0.01 4.22 4) 3 (rodlets) 4.98 0.02 0.00 4.98 50 3 (CEA) 2.26 0.24 0.02 2.24 0 3 (CEA) 3.04 0.04 0.00 3.04 10 3 (CEA) 3.84 0.16 0.01 3.84 20 3 (CEA) 4.61 0.11 0.01 4.60 30 Point obtained by burnup credit curve 3 (CEA) 5.00 0.00 0.00 5.00 35.:14 extrapolation**

4 1.74 0.04 0.00 1.74 0 4 2.51 0.01 0.00 2.51 10 4 3.13 0.13 0.01 3.12 20 4 3.85 0.15 0.01 3.85 301 4 4.96 0.04 0.00 4.96 40

• • As discussed in the Region 2 and Region 3 (CEA) sections, the 5.0 wt% endpoint of the burnup credit curve could not be obtained by enrichment interpolation because no Keff data for enrichment > 5.0 wt% is calculated.

Instead, interpolation of two 5.0 wt% cases at two burnup values that bound the endpoint were used for linear interpolation. The burnup requirement obtained by linear interpolation on burnup is less than the burnup value obtained by extrapolating last two known burnup credit curve points to 5.0 wt%. Rather than running additional cases to justify the bumup interpolation, the more conservative extrapolated values are used for convenience.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 95 6.1.6 References 6.1-1 NUREG/CR-7109, "An Approach for Validating Actinide and Fission Product Bumup Credit Criticality Safety Analyses-Criticality (kef) Predictions," April 2012. (ADAMS Accession No. ML12116A128)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 96 6.2 ANALYSIS OF SOLUBLE BORON REQUIREMENTS Section 6.2 provides the detailed discussion of the analysis to determine the minimum SFP boron requirements under normal and postulated accident conditions.

Key Analysis Assumptions

1) The SCALE Standard Composition Library (SCL) boron was used in this calculation. The B-10 atom % for the SCL boron is 19.9 at%. The required soluble boron concentrations calculated herein are conservative for these percentages in excess of 19.9 at%. For B-10 atom percentages lower than 19.9 a proportionally higher amount of soluble boron is required.

2) In the determination of the nominal soluble boron requirement, no soluble boron is credited as being present in the empty guide tubes, in the water/steel mixture used to approximate the fuel assembly nozzle, or in the w~ter modeled in the Boraflex space. These approximations are acceptable because they conservatively bound the actual content of these regions.

3) The nominal pool temperature was assumed to be 68°F for this analysis. This is a reasonable estimate for a lower bound pool temperature. In addition, it is shown in this analysis that higher SFP temperatures are bounding for reactivity. Temperatures lower than 68°F, particularly if soluble boron is present, are not limiting.

4) For the accident cases which use fresh fuel equivalent enrichments, an enrichment of 1.624 wt%

U-235 is used for the Region 3 fuel (with poison rodlets) instead of 1.61 wt% which is the maximum fresh fuel enrichment allowed by the Region 3 (with poison rodlets) bumup curve.

This value conservatively bounds the actual enrichment limit.

6.2.1 Full Spent Fuel Pool Model Description KENO was used to model the entire SFP at the zero boron condition to verify that the multiplication factor does not exceed 1.0 with no soluble boron present, including all biases and uncertainties. The full pool model was subsequently used to analyze the accident scenarios. In the full pool model, each of the individual 19 racks are in direct contact with each other (i.e., assuming no gaps between modules).

The SFP was modeled as being surrounded by a concrete wall, with 50 cm of concrete beneath the pool and 15 feet of water above the fuel assembly racks. These two values are somewhat arbitrary. Test cases were documented with the full pool model with 10 cm of concrete below the pool and 100 cm of concrete below the pool. The results were shown to be insensitive to the thickness of the concrete below the pool (less than 0.001 AK). Similarly, the height of the pool water above the racks was demonstrated with test cases to be inconsequential.

The fresh fuel enrichments used for the full pool model are as follows:

Region 1: two 5.0 wt% U-235 assemblies and two empty locations Region 2: two 2.95 wt% U-235 assemblies and one 1.95 wt% U-235 with one empty location Region 3: four 1.61 wt% U-235 assemblies with three diagonally placed poison rodlets Region 4: three 1.74 wt% U-235 assemblies and one empty location Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 97 The calculated KENO Keff for the SFP is 0.96747 +/- 0.00017 (without biases and uncertainties) with no soluble boron present. Applying the maximum individual region bias and uncertainty value from Table 6.1-20 for 0 GWd/MTU (0.0320 Ak from Region 3 with three poison rodlets and 0 ppm soluble boron) to the full pool model Keff produces a final full pool Keff value of [ 1 b,e which is less than the design basis limit of 1.0 with zero soluble boron. The bias and uncertainty used includes [ ]b,e administrative margin.

The KENO full pool model demonstrates that there is no evidence of any increase in overall pool reactivity due to region boundaries. Again it must be noted that the racks in the pool, unlike in the model, have two to three inches of separation between them. Hence, the boundary interactions in the actual pool are even less than those from this full pool model. Furthermore, the full pool Keif is 0.002 AK less than the most reactive individual region. Therefore, it can safely be concluded that there are no detrimental boundary effects with respect to pool reactivity because the overall pool KIff is less than the maximum regional Keff value.

6.2.1.1 Boron Requirements for Normal Conditions (Keff < 0.95)

To determine the amount of soluble boron required to maintain Keff less than or equal to 0.95, with all applicable biases and uncertainties, cases were performed using both the whole core model, as well as the individual region models at 0, 200, 400, 600, 800, 1000, 1200, and 1400 ppm of soluble boron. For each of the individual regions which require burnup credit (every region except Region 1), the most conservative fuel assembly was used, which was selected to minimize the reactivity worth of soluble boron, and therefore maximize the calculated amount of boron required. This corresponds to the maximum enrichment and maximum allowable burnup. This is verified by performing a series of test cases which measure the boron worth of an assembly in a Region 3 rack with a fresh 5.0 wt% enriched assembly and a 5.0 wt% enriched assembly depleted to 50 GWd/MTU. The results confirm that the worth of the soluble boron was approximately 25% less for the depleted fuel assembly, and therefore it is conservative to use the highest allowable burnup for the determination of the nominal boron concentration.

For Region 2, two 5.0 wt% assemblies depleted to 20 GWd/MTU and one 5.0 wt% fuel assembly depleted to 40 GWd/MTU were used. For Region 3, 5.0 wt% enriched assemblies depleted to 50 GWd/MTU with three poison pin rodlets were used. Also, for Region 3, 5.0 wt% assemblies depleted to 40 GWd/MTU with a CEA inserted were used. For Region 4, a 5.0 wt% enriched assembly depleted to 40 GWd/MTU in a 3/4 configuration was used.

The amount of soluble boron required to maintain Keff less than 0.95 under nominal conditions is determined using the results from Table 6.2-1 below. The table provides the Keff of the respective region at various soluble boron concentrations. The column in the table labeled AKeff is the reactivity change from the reference Ker value of the unborated spent fuel pool / storage region. The change in Keff is conservatively calculated by adding the one-sigma standard deviation to the borated case and subtracting the one-sigma standard deviation of the reference unborated value.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 98 The target value for calculating the soluble boron required to maintain Keff less than 0.95 is [ ]b,e This value was derived using a value for total bias and uncertainty of 0.0502 AK. This value is acceptable because it bounds the maximum region bias and uncertainty calculated in Section 6.1 [

]b~e The target Keff[ b,e is considerably conservative for the following reasons:

" The maximum bias and uncertainty used (0.0502 AK) bounds the maximum regional bias and uncertainty of any region. This value is being conservatively used for all regions of the MP2 SFP.

" Table 6.1-22 bias and uncertainty calculations include SFP temperature effects up to 210°F (as a bias) and the nominal boron requirement only needs to cover up to 150TE

" Table 6.1-22 bias and uncertainty calculations include soluble boron effects up to 2000 ppm, which is higher than needed for this calculation. This results in a larger temperature bias than would really be present for boron concentrations encountered in this analysis.

For Regions 2 and 3 (with a CEA), the target Keff was reduced by the difference between the modeled Keff and the allowable Keff from the respective burnup curve. The other regions have 0 ppm values which are very close to those on the respective burnup curves, and therefore this adjustment is not required.

Therefore, for Region 2, the target is calculated as follows:

Target Keff= [ 1 b,e = 0.87057 where [ ]b,, is the 0 ppm value from Table 6.2-1 and [ 1 b,, is the maximum allowable Keff from Table 6.1-11.

For Region 3 (with CEA), the target is calculated as follows:

Target Kef = [ ]b,e = 0.86078 where [ 1b,, is the 0 ppm value from Table 6.2-1 and [ ]b,e is the maximum allowable Keff from Table 6.1-28.

The boron values in Table 6.2-1 were interpolated to obtain the minimum SFP boron requirement for each region. As can be observed in the table, the most limiting region (one with the maximum amount of soluble boron required to maintain Keff less than 0.95, including all biases and uncertainties) is Region 2.

The absence of region interface effects has been previously verified by comparing the full pool model Keff to the maximum region model Keff. Therefore, the Region 2 boron requirement is sufficient for the full SFP as well as each individual region.

To verify that this boron concentration is adequate, the full pool model was run with fresh fuel equivalents in each region. The Kff of the full pool, using fresh fuel equivalent enrichments and the Region 3 boron concentration (453 ppm) is 0.87122 +/- 0.00015. The limiting boron value from Region 2 (540 ppm) will Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 99 cause the Keff value to be much lower. These results confirm that the normal configuration soluble boron requirement is less than 600 ppm, the value assumed in the SFP dilution analysis.

6.2.1.2 Additional Normal Conditions Scenarios exist during the normal operation of the SFP for which the exact conditions are not explicitly modeled in this analysis. These will generally be related to the movement of single fuel assemblies into or out of the SFP. These conditions, which are all bounded by the analyses performed in this calculation, are discussed below.

Loading and Removal of Assemblies During the fuel movement process to either fill or empty the SFP, scenarios will arise in which a single fuel assembly is axially offset from the other fuel assemblies. For example, in the process of lowering a fuel assembly into a rack or removing a fuel assembly from a rack, the fuel assembly being moved is elevated above the rest of the assemblies, which is not directly analyzed in this calculation for the normal storage analysis. The configuration of a single axially offset fuel assembly will reduce the reactivity of the SFP by reducing the neutronic coupling between neighboring fuel assemblies. The overall effect is to increasingly isolate the fuel assembly the more axially offset it is in the configuration. This configuration, therefore, is bounded by the corresponding analysis performed in this calculation where all of the fuel assemblies are aligned axially.

Ultrasonic Testing of a Fuel Assembly Ultrasonic Testing (UT) involves raising and lowering an assembly within its rack cell while the inspection is being performed. This is similar to when the assembly is being loaded or unloaded from the rack. Axially offsetting one assembly from the others will reduce the reactivity of the system by reducing the neutronic coupling between neighboring assemblies. Therefore, this configuration is bounded by this analysis.

Fuel Rod Inspection/Reconstitution This particular scenario involves a fuel assembly being placed in a certain location outside of its rack cell, a single fuel rod being removed from the assembly, that rod being inspected, and either placed back into the assembly, replaced by a non-fuel replacement rod, or replaced by another fuel rod. The removal of a fuel rod from an assembly will change the moderator to fuel ratio, which could potentially cause a slight increase in reactivity. The assembly, however, will be kept isolated from the other fuel assemblies in the SFP by a distance in excess of one rack length, effectively isolating the assembly which is bounded by the normal criticality safety analysis documented in this calculation.

Should the removed fuel rod be replaced by another fuel rod, the assembly average bumup must be recomputed and the fuel assembly must be requalified according to the revised burnup, which will be bounded by the criticality safety analysis.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis -Attachment 5 100 If the rod is replaced by a stainless steel rod or another non-fissionable material rod, the reduction in fuel content will be less reactive and is bounded by this analysis.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 101 Table 6.2-1 Keff as a Function of Soluble Boron - Normal Conditions Region Type Boron (ppm) Keffr A~ff 0 0.96683 +/- 0.00015 0 200 0.92851 +/- 0.00015 -0.03802 400 0.89433 +/- 0.00016 -0.07219 Full Pool 600 0.86343 +/- 0.00016 -0.10309 800 0.83464 +/- 0.00014 -0.1319 1000 0.80895 +/- 0.00015 -0.15758 1200 0.79044 +/- 0.00018 -0.17606 1400 0.77408 +/- 0.00016 -0.19244 Required Boron (ppm) 350 0 0.94057 +/- 0.00015 0 Region 2 200 0.91263 +/- 0.00015 -0.02764 two 5.0 wt% assemblies 400 0.88735 + 0.00014 -0.05293 depleted to 20 GWd/MTU 600 0.86396 +/- 0.00016 -0.0763 and 800 0.84288 +/- 0.00014 -0.0974 one 5.0 wt% assembly 1000 0.82322 +/- 0.00015 -0.11705 depleted to 40 GWD/MTU 1200 0.80460 +/- 0.00015 -0.13567 1400 0.78759 +/- 0.00015 -0.15268 Required Boron (ppm) 540 0 0.96237 +/- 0.00014 0 Region 3 200 0.93313 +/- 0.00014 -0.02896 4 out of 4 400 0.90642 +/- 0.00014 -0.05567 5.0 wt% enriched assembly 600 0.88219 +/- 0.00013 -0.07991 depleted to 50 GWdIMTU 800 0.85939 +/- 0.00013 -0.10271 with 3 poison rodlets 1000 0.83878 +/- 0.00014 -0.12331 1200 0.81903 +/- 0.00013 -0.14307 1400 0.80100 +/- 0.00013 -0.1611 Required Boron (ppm) 453 0 0.93158 +/- 0.00018 0 Region 3 200 0.90263 +/- 0.00017 -0.0286 4 out of 4 400 0.87687 +/- 0.00017 -0.05436 5.0 wt% enriched assembly 600 0.85278 +/- 0.00016 -0.07846 depleted to 50 GWdIMTU 800 0.83057 +/- 0.00017 -0.10066 with one CEA 1000 0.80999 +/- 0.00015 -0.12126 1200 0.79113 +/- 0.00017 -0.1401 1400 0.77319 +/- 0.00015 -0.15806 Required Boron (ppm) 534 0 0.96537 +/- 0.00014 0 Region 4 200 0.93347 +/- 0.00014 -0.03162 3 out of 4 400 0.90615 +/- 0.00016 -0.05892 5.0 wt% enriched assembly 600 0.88102 +/- 0.00014 -0.08407 depleted to 40 GWd/MTU 800 0.85842 +/- 0.00015 -0.10666 1000 0.83755 +/- 0.00014 -0.12754 1200 0.81805 +/- 0.00013 -0.14705 1400 0.80026 +/- 0.00014 -0.16483 Required Boron (ppm) 448 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 102 6.2.1.3 Boron Requirements for Accident Scenarios The soluble boron concentration (ppm) required to maintain K~ff less than or equal to 0.95 under accident conditions is calculated by initially determining the limiting credible accident scenario in the MP2 SFP (the accident which produces the largest increase in pool reactivity). The accident scenario which produces the largest increase in SFP Keff is utilized to determine the required soluble boron concentration necessary to mitigate this and all less severe accident scenarios.

The accident scenarios were largely derived from the existing FSAR (Section 9.8.2.1.2) accidents. Due to the modeling of the various rack modules in direct contact with each other, with no spacing in between, the seismic/lateral rack movement accident was removed from consideration, since the spacing between racks modules cannot be reduced. In addition, in response to LER #2012-002-00 submitted for Indian Point Unit 2 [Reference 6.2-1], the evaluation of incorrect boundary alignment between regions which require empty cells on the boundary was included for evaluation.

The dropping of a fuel transfer cask into the SFP was not analyzed. Due to the single failure proof design of the Spent Fuel Cask Crane (SFCC) at MP2, this accident is not considered to be credible. Note that this is the current licensing basis at MP2 per Section 9.8.2.1.2 which states that load drop events are "not crediblefor loads lifted by the SFCC when handled and rigged in accordancewith the singlefailure criteria".

The placement of an assembly outside of the spent fuel pool rack is bounded by the single fresh fuel assembly misload. In addition, the evaluation of the misplacement of a 5.0 wt% U-235 fresh fuel assembly between Region 3 and the new fuel elevator, with a fresh 5.0 wt% U-235 fuel assembly in the new fuel elevator, places two 5.0 wt% enriched fuel assemblies in direct contact with each other with no rack material in between. Placing a fuel assembly outside of the rack geometry will result in far less neutronic coupling than would be present in either the fresh assembly misload or the fresh assembly next to the new fuel elevator. This scenario is therefore bounded by these two accident analyses.

Based on the assessment of potentially limiting incidents and other considerations noted above, the following credible accident scenarios were analyzed:

" Misplacement or dropping of a single 5.0 wt% enriched fresh fuel assembly into a Region 1, 2, 3, or 4 storage location.

" Temperature reaches boiling conditions which increases reactivity in Regions 3 and 4 of the SEP.

" Dropping of a fuel assembly on top of a fuel rack which comes to rest in a horizontal or vertical position.

" Misplacement of a 5.0 wt% U-235 fresh fuel assembly between Region 3 and the new fuel elevator, with a fresh 5.0 wt% U-235 fuel assembly in the new fuel elevator.

" Incorrect fuel assembly placement in which the boundaries between the different regions of the MP2 SFP are out of alignment in such a way which maximizes the reactivity.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 103 Fuel Assembly Misload Accident The fuel assembly misload accident involves the inadvertent placement of a fresh fuel assembly in an area in which it is not permitted to be placed. The most severe event involves the placement of a fresh assembly in a region which is required to be left empty. A number of misload scenarios were explored to ensure that the most limiting scenario was found for the soluble boron requirement. The misloads were analyzed bothin the full pool model as well as in the infinite lattice regional model. These are described by region below.

Misload in Region 1 The misloads for Region I in the regional model consist of placing a fresh 5.0 wt% assembly into one of the empty cells in the 2-Out-of-4 configuration, thus becoming essentially a 3-Out-of-4 configuration for that rack cell. In the full pool, the fresh assembly can be loaded either in the interior of the region or on one of the edges of the region which face regions 2 and 3. These scenarios were analyzed and are presented in Table 6.2-2 and Table 6.2-3.

Misload in Region 2 Region 2 presents a number of possible misload scenarios. The misload scenarios analyzed are a fresh 5.0 wt% assembly placed in the empty location of the 3-Out-of-4 configuration, replacing the lower reactivity assembly with a fresh 5.0 wt% assembly and replacing one of the higher reactivity assemblies with a fresh 5.0 wt% assembly. For the full pool analysis, a number of boundary scenarios exist since Region 2 is in contact with each of the other three Regions. Therefore misloads are analyzed on the boundary of Regions 1, 3, and 4 to determine the most severe scenario. These scenarios were analyzed and are presented in Table 6.2-2 and Table 6.2-3.

Misload in Region 3 The misload scenarios analyzed for Region 3 are a fresh 5.0 wt% assembly placed in a location normally occupied by a low reactivity assembly with 3 poison pins. For the full pool analysis, a number of boundary scenarios exist since Region 3 is in contact with each of the other three Regions. It is also possible to mistakenly place or drop a fuel assembly in the area between Region 3 and the fresh fuel elevator, with a fresh fuel assembly on the elevator. These scenarios were analyzed and are presented in Table 6.2-2 and Table 6.2-3.

Misload in Region 4 The misload scenarios analyzed for Region 4 are a fresh 5.0 wt% assembly placed in a location normally occupied by a low reactivity assembly with no poison pins, and the more severe scenario where a fresh 5.0 wt% assembly is placed in a location which should be empty. For the full pool analysis, a number of boundary scenarios again exist since Region 4 is in contact with two other regions. These misloads are analyzed to determine the most severe scenario. These scenarios were analyzed and are presented in Table 6.2-2 and 6.2-3.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 104 Table 6.2-2 Keff for Misload Accidents in Infinite Lattice Region Model (2x2 Assemblies, no soluble boron)

Region Keff Region I - 5.0 wt% in 2-Out-of-4 0.90859 +/- 0.00021 configuration Region 1 Misload 1.12213 +/- 0.00021 Region 2 Out-of-4 0.96865 +/- 0.00019 configuration Base Case Region 2 Misload- fresh assembly placed in medium reactivity location Region 2 Misload- fresh 1.06688 +/- 0.00021 assembly placed in low reactivity location Region 2 Misload- fresh 1.13413 +/- 0 assembly placed in empty location Region 3 Out-of-4 Base Case 0.96968 +/- 0.00015 Region 3 - Misload 1.13062 +/- 0.00023 Region 4 Out-of-4 Base Case 0.97001 +/- 0.00019 Region 4 Misload - fresh assembly placed in low reactivity assembly location Region 4 Misload - fresh 1.19383 +/- 0.00017 assembly placed in empty location Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 105 Table 6.2-3 Keff for Misload Accidents in Full Pool Model (no soluble boron)

Region Keff Region 1 Misload - fresh assembly placed in interior 1.06586 +/- 0.00019 empty location Region 1 Misload- fresh assembly placed in Region 2 1.05672 +/- 0.00020 boundary empty location Region 2 Misload - fresh assembly placed in interior 1.06263 +/- 0.00021 empty location Region 2 Misload - fresh assembly placed in Region 1 1.05410 +/- 0.00020 boundary empty location Region 2 Misload - fresh assembly placed in low reactivity 1.04327 +/- 0.00022 assembly location Region 3 Misload - fresh assembly placed in low reactivity 1.06576 +/- 0.00019 assembly location Region 3 Misload - fresh assembly placed on Region 1 1.06585 +/- 0.00021 boundary Region 3 Misload - fresh assembly placed on Region 2 1.06575 +/- 0.00020 boundary Region 3 Misload - fresh assembly placed on Region 3 1.06590 +/- 0.00019 boundary Region 3 Misload - fresh assembly placed in between 1.06551 +/- 0.00021 Region 3 and fresh assembly on new fuel elevator Region 4 Misload - fresh assembly placed in interior 1.10020 +/- 0.00025 empty location Region 4 Misload - fresh assembly placed in low reactivity 1.01335 +/- 0.00020 assembly location Region 4 Misload - fresh assembly placed in empty 1.09133 +/- 0.00019 location on Region 2 boundary Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 106 These misloads are analyzed to determine the most severe scenario. These scenarios were analyzed and are presented in Table 6.2-2 and 6.2-3. Note that Table 6.2-2 uses 2x2 region models (infinite lattice) to select the most limiting misload region. The Keff values in Table 6.2-2 greatly overstate the full pool effect of a single misload.

As can be observed from Table 6.2-2 and Table 6.2-3, the most limiting assembly misload, in both the infinite lattice as well as the full pool model, is the placement of a fresh assembly in a normally empty location in Region 4. The interior empty location produces the largest reactivity increase in the full pool model. To calculate the amount of boron needed to mitigate this misload, the Keff of a 6x20 region at 150 0F, which represents Region 4, was calculated with 1100, 1200, 1300 and 1400 ppm of soluble boron, assuming a fresh 5.0 wt% assembly loaded into an empty location in Region 4. Periodic x-y boundary conditions were used to ensure no net neutron leakage effects are credited. The other assemblies are 5.0 wt% depleted to 40 GWd/MTU, which is the maximum enrichment/burnup allowed for fuel stored in Region 4. This represents the configuration which produces the minimum boron worth encountered in Region 4, which will produce the maximum amount of soluble boron needed to offset the effects of the accident. The results of these cases are provided in Table 6.2-4 below.

Table 6.2-4 Keff vs Soluble Boron for the Limiting Misload Accident (150'F)

Boron (ppm) Keff 1100 ppm 0.91768 +/- 0.00024 1200 ppm 0.90565 +/- 0.00026 1300 ppm 0.89514 +/- 0.00026 1400 ppm 0.88385 +/- 0.00023 Using the same approach as was used previously, the amount of soluble boron required to lower Kff to the target value [ Ib,e with a misloaded fresh fuel assembly in an empty location in Region 4 was calculated to be 1256 ppm. Therefore, the soluble boron required to mitigate the worst misload scenario is 1256 ppm - 540 ppm = 716 ppm.

The approximation used to model the water/steel mixture for the top and bottom nozzles of the fuel assembly [ ]b affects the overall anmount of soluble boron present in the nozzles for the misload cases. To evaluate this, a misload test case was run with 1300 ppm and no soluble boron present in the top and bottom nozzles [

]b This results in a KIff of 0.89500 +/- 0.00023, compared with 0.89514 +/- 0.00026 from Table 6.2-4 above. Therefore, not taking credit for the 1300 ppm of soluble boron in the nozzles still produces a Keff less than 0.898.

These results confirm that the current MP2 SFP minimum soluble boron requirement of 1400 ppm continues to meet the regulatory criteria for maintain sufficient KIff margin when taking credit for soluble boron in the SFP.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 107 Boundary Misali2nment A scenario exists whereby the row placement of assemblies in a particular region could be inadvertently reversed. This would result in two consecutive rows of assemblies with some fuel in direct contact with four neighboring assemblies, without the required empty cells in between. This scenario is analyzed for all of the regions where empty cells are employed and this accident could be realized.

This potential accident scenario was analyzed, in part, due to the event addressed in LER #2012-002-00

[Reference 6.2-1], in which 11 fresh fuel assembles were incorrectly loaded in such a way that the fresh fuel assemblies were in direct contact with depleted fuel assemblies without the required empty locations in between, which is required by that Unit's Technical Specifications. All of the boundary regions for the MP2 SFP are analyzed for this potential event.

The first boundary misalignment analyzed is for the Region 1-2 boundary, where the fresh assemblies in Region 1 are out of alignment, resulting in the fresh assemblies being in direct contact with the neighboring assemblies from Region 2 as shown in Figure 6.2-1 below. In all figures of the full pool model in this section, they are depicted with the top of the pool removed for clarity.

The second boundary misalignment analyzed is also for the Region 1-2 boundary, where the Region 2 assemblies are out of alignment, resulting in the fresh assemblies being in direct contact with the neighboring assemblies from Region 2 as shown in Figure 6.2-2 below.

The third boundary misalignment analyzed is for the Region 2-4 boundary, where the Region 4 rows are reversed, causing the rows from Regions 2 and 4 to be in direct contact in a 4-Out-of-4 configuration without the empty cells in between as shown in Figure 6.2-3 below.

The fourth boundary misalignment analyzed is for the Region 2-3 boundary, where the Region 2 rows are reversed, causing the rows from Regions 2 and 3 as well as Region I to be in direct contact in a 4/4 configuration without the empty cells in between as shown in Figure 6.2-4 below.

The results of the four boundary misalignment scenarios are provided in Table 6.2-5 below and confirm that this potential accident is well bounded by the fuel assembly misload. No further analysis is required.

The soluble boron required to mitigate the effects of the fuel assembly misload accident cover the four worst boundary misalignment accidents.

Table 6.2-5 KIfr for Four Boundary Misalignment Accidents (no soluble boron)

Boundary Misalignment Keff 1-2 0.97614 +/- 0.00016 2 nd 1-2 Boundary 1.02942 +/- 0.00017 2-4 1.01135 +/- 0.00016 2-3 1.01697 +/- 0.00017 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 108 Region 1-2 boundary misalignment Figure 6.2-1 KENO3D-Plot of Region 1-2 Boundary Misalignment Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 109 Region 1-2 boundary misalignment Figure 6.2-2 KENO3D-Plot of Second Region 1-2 Boundary Misalignment Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 110 Region 2-4 Boundary Misalignment Figure 6.2-3 KENM3D-Plot of Region 2-4 Boundary Misalignment Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 ill Figure 6.2-4 KENO3D-Plot of Region 2-3 Boundary Misalignment Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 112 Fuel Assembly Drop The fuel assembly drop/crush accident involves the dropping of a fuel assembly into the pool on top of another assembly, landing either horizontally or vertically on top of the fuel rack. It has already been demonstrated that dropping an assembly into the region between the fuel elevator and Region 3 is bounded by the Region 4 misload accident. The heavy load/drop accident is evaluated for each different rack type in each different storage region using a 2x2 regional representation. To simulate a crushing/compacting of the storage region, the inner diameter of the storage cell rack was reduced by one cm for the Region 3/4 rack type and was reduced to the point where the angle brackets were touching the outer stainless steel cell in the Region 1/2 rack type. All four of the racks were treated as being compacted so as to minimize the intra-assembly spacing. The thickness of the stainless steel rack was maintained for both rack types. Figure 6.2-5 provides a visual representation of a Region 1 compacted rack module. The potential also exists for the compacting of the assembly, which is modeled by increasing the fuel theoretical density to 100%. To bound both of the potential fuel assembly drop scenarios (fuel assembly dropped directly on top of another assembly creating a fuel region which is essentially twice the height of a single fuel assembly and dropping an assembly horizontally where it is lying across the top of the racks), the fuel assembly crush scenarios were run with mirror reflective boundary conditions on the upper and lower axial boundaries (+z and -z).

The results for each of the four regions is provided in Table 6.2-6 below and demonstrate that the fuel assembly drop/heavy load accident are bounded by the most severe fuel assembly misload in Region 4.

Table 6.2-6 Assembly Drop/Heavy Load Accident Results (no soluble boron)

Region Keff 1 1.09607 +/- 0.00019 2 1.04882 +/- 0.00020 3 1.00600 +/- 0.00012 4 1.08116 +/- 0.00017 The Region 1 heavy load accident, demonstrated to be the most severe from the results in Table 6.2-6 was evaluated to determine the optimum rod pitch, which could potentially be obtained during a deformation caused by a drop/crush accident. The crush cases for Region 1 were run for a variety of rod pitch values in order to determine the maximum Kff value which can be obtained by this deformation. The search was performed for an infinite array of crushed Region I storage modules, with infinite boundary conditions to cover both (horizontal and vertical assembly landing) drop scenarios. This is an extremely conservative representation of this accident. The largest rod pitch value (1.5232 cm) is the value which produces the largest Kff value. Table 6.2-7 below provides the results and demonstrates that the accident is bounded by the reactivity of the Region 4 assembly misload accident from Table 6.2-2.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 113 Figure 6.2-5 KENO3D-Plot of Compacted Region 1 W Rack Module Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 114 Table 6.2-7 Region 1 Optimum Pitch Calculation for Assembly Drop/Crush Accident (no soluble boron)

Pitch (cm) Keff 1.4782 1.10108 +/- 0.00019 1.4822 1.10594 +/- 0.00018 1.4862 1.11030 +/- 0.00019 1.4922 1.11660 +/- 0.00018 1.5122 1.13619 +/- 0.00017 1.5222 1.14587 +/- 0.00019 1.5232 1.14658 +/- 0.00018 Spent Fuel Pool Temperature Greater than 150 Degrees Fahrenheit The temperature deviation of the spent fuel pool is analyzed for temperatures in excess of 150T. The temperature bias for pool temperatures up to 150°F is covered in the temperature bias calculated for the bias and uncertainty used in the generation of burnup curves. Region 3 was shown to have the largest positive moderator temperature coefficient up to 150'F. Therefore, an infinite array of Region 3 assemblies (5.0 wt% depleted to 50 GWd/MTU) was analyzed along with the full pool model.

Temperatures from 150'F through boiling (212'F) were analyzed. In addition, an approximation was made to analyze the saturation temperature at the bottom of the SFP, estimated to be about 250'F for a pressure of about two atmospheres. The temperature was applied homogeneously throughout all of the water in the model. The results are shown in Table 6.2-8 below and demonstrate that this accident is well bounded by the misloaded assembly accident analyzed previously.

Table 6.2-8 Keff vs SFP Temperature (no soluble boron)

Temperature (TF) Region 3 Keff Full Pool Keff 150 0.96377 +/- 0.00013 0.96970 +/- 0.00015 160 0.96383 +/- 0.00014 0.96992 +/- 0.00016 175 0.96392 +/- 0.00013 0.96965 +/- 0.00016 200 0.96458 +/- 0.00013 0.97078 +/- 0.00015 212 0.96493 +/- 0.00014 0.97116 +/- 0.00016 Tsat (250) 0.96551 +/- 0.00015 0.97172 +/- 0.00015 To address possible voiding conditions which could occur as a result of pool overheating, water densities of 0.80 g/cm3 and 0.73 g/cm 3 with a bulk water temperature of 212TF were analyzed for Region 3. A full Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 115 pool model at 212'F and 0.80 g/cm 3 was also run. All three of these conditions resulted in a decrease in reactivity from the nominal 212°F condition.

Demonstration of Misload with Nonfuel Related Components Loaded in Region 1 Peripheral Cells A final set of four test cases was performed to demonstrate that the conclusions of this calculation remain valid even if a nonfuel item is loaded into the empty peripheral cells in Region 1. This is demonstrated by performing full core misload cases in Region 4 (demonstrated above to be the limiting region for this accident) as well as on the edge of Region 1 (shown in Figure 6.2-6 below) with a soluble boron concentration of 1400 ppm. The empty cells (shown with arrows on Figure 6.2-6) are filled both with unborated water as well as void. These two materials bound any nonfuel item which would potentially be placed in the empty cells. The water displacement is satisfied by the void. The elimination of only the boron but the inclusion of the moderation is satisfied by the unborated water cases. Any credible object which might be placed in the empty cells would be made of either some sort of modest resonance absorber (stainless steel, Inconel, zirconium, etc.) or a completely neutronically inert material.

It is conceivable that the 1256 ppm minimum soluble boron requirement could be affected by the placement of nonfuel objects in these cells. An additional test case was performed with a misload in Region 1 and void in the Region 1 peripheral empty cells. The result of this case is also presented in Table 6.2-9. Nominal cases with 1256 ppm and with water filled and void filled empty cells are also provided in Table 6.2-9.

These cases were run at 150 OF with the same fresh fuel enrichments and other parameters as discussed previously for the misload cases. The results are presented in Table 6.2-9 below and demonstrate that the results of this calculation are not affected by the placement of nonfuel items in the empty peripheral cells of Region 1.

Table 6.2-9 Simulation for Misload of Nonfuel Components in Empty Region 1 Peripheral Cells Misloaded Region Empty Cell Filler Material Full Pool Keff 1 (1400 ppm boron) Void 0.87256 +/- 0.00018 1 (1400 ppm boron) Water 0.87081 +/- 0.00017 4 (1400 ppm boron) Void 0.74909 +/- 0.00016 4 (1400 ppm boron) Water 0.74698 +/- 0.00018 1(1256 ppm boron) Void 0.88795 +/- 0.00016 Nominal / No Misload Void 0.76186 +/- 0.00016 (1256 ppm boron)

Nominal / No Misload Water 0.76087 +/- 0.00019 (1256 ppm boron)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 116 Figure 6.2-6 KENO3D-Produced Plot of Region 1 Misload with Unborated Water in Empty Cells Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 117 6.2.1.4 Boron Dilution Analysis A soluble boron dilution analysis of potential scenarios which could dilute the boron concentration in the SFP demonstrates that sufficient time is available to detect and mitigate a boron dilution prior to reaching the minimum soluble boron required under normal conditions, thus not exceeding the 0.95 KIff design basis (including biases and uncertainties). The existing SFP boron dilution analysis [Reference 6.2-2]

was approved by the NRC in Reference 6.2-3, establishing the current SFP minimum TS 3.9.17 soluble boron limit of 1720 ppm. The existing analysis also requires a minimum of 600 ppm under normal operating conditions in the SFP to assure that the 0.95 Kff design basis (including biases and uncertainties) is met. For accident conditions, the existing analysis credited up to 1400 ppm soluble boron in the criticality analysis.

The potential dilution sources described in the existing analysis are not credible threats to the SFP soluble boron concentration from 1720 ppm to 600 ppm due to either volume or flow rate considerations. The large volume of water required to dilute the SFP, the TS controls on SFP boron concentration, observation during plant operator rounds as well as engineered alarms, would effectively detect a dilution event prior to KIff reaching 0.95. All of these considerations and mitigation measures remain unchanged for the analysis herein.

As discussed in Sections 6.2.1.1 and 6.2.1.3 of this report, the 600 ppm and 1400 ppm values for normal and accident conditions, respectively, are retained for this analysis. The existing dilution volumes, flow rates and event mitigation response times remain unchanged for this analysis. The Reference 6.2-2 boron dilution analysis conclusions remain applicable and unchanged.

6.2.2 References 6.2-1 Letter from John A. Ventosa (Entergy) to USNRC, "Licensee Event Report #2012-002-00, Technical Specification (TS) Prohibited Condition Caused by New Fuel Assemblies Stored in a Configuration Prohibited by the TS," Indian Point Unit No. 2, Docket No. 50-247, DPR-26, April 13, 2012.

6.2-2 Letter from J. Alan Price (DNC) to USNRC, "Millstone Power Station, Unit No. 2, Technical Specifications Change Request (TSCR) 2-10-01, Fuel Pool Requirements," November 6, 2001.

(ADAMS Accession No. ML013510295) 6.2-3 Letter from Richard B. Ennis (NRC) to J. A. Price (DNC), "MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: SPENT FUEL POOL REQUIREMENTS (TAC NO. MB3386)," April 1, 2003. (ADAMS Accession No. ML030910485)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 118 7

SUMMARY

OF RESULTS This section presents the results of the Millstone Unit 2 spent fuel pool criticality analysis, performed with reactivity credit for burnup and with credit for soluble boron, but with no credit for Boraflex panels.

7.1 ALLOWABLE STORAGE CONFIGURATIONS AND REQUIREMENTS Included are descriptions of allowable storage configurations and minimum burnup requirements (as applicable) for each of the four spent fuel regions identified in the analysis.

7.1.1 Region 1 The Region 1 "2-Out-of-4" storage configuration (Figure 6.1-1) may be used to store new fuel with initial nominal planar average enrichment up to 4.85 wt% U-235 or depleted fuel. Region 1 has no minimum bumup requirements.

7.1.2 Region 2 The Region 2 "3-Out-of-4" storage configuration (Figure 6.1-1) may be used to store two types of fuel assemblies, denoted Type 2A and 2B. Assembly Type 2A has higher reactivity and must meet the minimum burnup requirements of Figure 7.1-1. Assembly Type 2B has lower reactivity and must meet the minimum burnup requirements of Figure 7.1-2.

7.1.3 Region 3 The Region 3 "4-Out-of-4" storage configuration (Figure 6.1-1) may be used to store fuel assemblies which contain two different types of poison inserts: borated stainless steel rodlets or CEAs. Assemblies containing rodlets must meet the minimum burnup requirements of Figure 7.1-3. Assemblies containing CEAs must meet the minimum burnup requirements of Figure 7.1-4.

7.1.4 Region 4 The Region 4 "3-Out-of-4" storage configuration (Figure 6.1-1) may be used to store fuel assemblies which meet the minimum burnup requirements of Figure 7.1-5.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 119

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Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 120

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Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 121 30 _ _--_- -~- -_-_-_ -------

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Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 122 40

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Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 123 40 77 CL --[,ACCEPTA-BLE C I C E

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Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 124 7.2 SOLUBLE BORON REQUIREMENTS 7.2.1 Normal Conditions The total soluble boron concentration required to maintain the Keff less than or equal to 0.95 with 95%

probability at 95% confidence level during normal conditions is 540 ppm. Since this value is less limiting than the value of 600 ppm currently documented in MPS-2 FSAR Section 9.8.2.1.2.c and TS Basis 3/4.9.17, the existing limitation of 600 ppm is retained.

7.2.2 Accident Conditions The total soluble boron concentration required to maintain the Keff less than or equal to 0.95 with 95%

probability at 95% confidence level during the most limiting SFP accident scenario is 1256 ppm. Since this value is less limiting than the value of 1400 ppm currently documented MPS-2 FSAR Section 9.8.2.1.2.c and TS Basis 3/4.9.17, the existing limitation of 1400 ppm is retained.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-1 APPENDIX A CRITICALITY CODE VALIDATION Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-2 This appendix describes validation for the use of the SCALE 6.0 system for criticality safety analysis with the 238 group ENDF/B-VII cross-section library distributed with SCALE. A bias and bias uncertainty for SCALE 6.0 is determined for use in the criticality analyses. The SCALE versions that are utilized for this analysis include the control modules CSAS5 and the following functional modules: BONAMI, CENTRM, and KENO V.a. In addition to these codes, SCALE auxiliary modules PMC, WORKER, and CRAWDAD are used for automatic cross section library processing. This analysis applies the methodology described by NUREG/CR-6698 [Reference A-I] to validate the use of SCALE 6.0 with the 238-group ENDF/B-VII neutron cross section library.

A.1 VALIDATION RANGE OF PARAMETERS This validation is performed as part of the analysis for the Millstone Unit 2 SFP and NFSR. The pool is assumed to be flooded with water at near room temperatures and below 100TC. 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 moderator is water and the oxygen in the fuel pellet. The neutron absorbers credited are boron (as poison rods 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 other configurations within the area of applicability. The assembly arrangement can vary from totally isolated assemblies to a close packed array of assemblies. A separate area of applicability is defined for the NFSR which excludes the use of critical experiments modeling depleted fuel or soluble boron.

A.2 CRITICAL EXPERIMENT DATA The validation of SCALE 6.0 for purposes of fuel storage rack analyses is based on the analysis of selected critical experiments from two experimental programs. The first program is the French Haut Taux de Combustion critical experiment data (referred to as HTC).The second program is the OECD/NEA International Handbook of Evaluated Criticality Safety Benchmarks Experiments (referred to as IH).

The HTC critical experiments consist of 4 phases of experiments, consisting of rectangular pitch lattice arrays of zirconium-clad fuel rods in water. The fuel rods were composed of a mixture of low-enriched U0 2 and PuO 2 designed to simulate light water reactor fuel with initial 4.5% U0 2 depleted to 37,500 MWD/MTU, but without fission products. Each experiment consisted of a different combination of number of rods, lattice pitch, assembly pitch, reflector spacing, and boron concentration. Only cases within the area of applicability of this analysis were selected for this validation. Experiments using lead reflectors, soluble gadolinium, cadmium poison plates, etc., were excluded. A total of 66 HTC experiments were utilized.

The OECD/NEA International Handbook of Evaluated Criticality Safety Benchmarks Experiments is the second source of critical experiments used for this validation. Volume IV of the handbook is for low enriched uranium systems. The section of Volume IV of interest to this validation is the "Thermal Compound Systems." All of the experiments selected are numbered, LEU-COMP-THERM-OXX.

There are 85 sets of benchmarks in the September 2010 version of the handbook. 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 area of applicability of this Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-3 validation, such as lead reflectors, cadmium absorbers, or hexagonal arrays. The 25 selected benchmark sets resulted in 187 experiments that are used for the statistical analysis. Forty-one experiments used soluble boron.

A.3 EXPERIMENTAL DATA MODELING Table A-1 lists the results of the critical experiments modeled using SCALE 6.0. Note that experimental Keff values and uncertainties for the HTC experiments are removed from the table as they are proprietary information bound by a Non-Disclosure Agreement between Dominion and UT-Battelle. The removed data are denoted by 'PI' in Table A-1.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-4 Table A-i: Critical Experiment Results with SCALE 6.0 Pitch Pin Soluble SoulBrn Boron EALF U-235 Pu Experiment Oep kcaic cac ELU25 Pu ith in Concentration knexp Fexp kcalc (eV) (wt%) (wt%) (cm) (cm) (ppm)

(ppm)

HTC-PI-CO1 Pi PI 0.99924 0.00020 6.91199E-02 1.57 1.10 PI Pi 0 HTC-P1-C02 PI PI 0.99915 0.00019 6.62346E-02 1.57 1.10 PI Pi 0 HTC-P1-C03 PI PI 0.99929 0.00019 6.61296E-02 1.57 1.10 PI PI 0 HTC-P1-C04 PI PI 1.00021 0.00024 8.44981E-02 1.57 1.10 PI PI 0 HTC-P1-C05 PI PI 1.00028 0.00023 8.23172E-02 1.57 1.10 PI PI 0 HTC-PI-C06 PI PI 0.99974 0.00022 8.17074E-02 1.57 1.10 PI PI 0 HTC-P1-C07 PI P1 0.99992 0.00024 1.01762E-01 1.57 1.10 PI PI 0 HTC-P1-C08 PI PI 0.99958 0.00024 1.00194E-01 1.57 1.10 PI PI 0 HTC-P1-C09 PI PI 0.99917 0.00024 9.92774E-02 1.57 1.10 PI PI 0 HTC-P1-C1o PI P1 1.00017 0.00025 1.39681E-01 1.57 1.10 PI PI 0 HTC-P1-C(11 P1 PI 0.99882 0.00023 1.35008E-01 1.57 1.10 P1 PI 0 HTC-P1-C12 PI PI 0.99864 0.00023 1.33062E-01 1.57 1.10 P1 PI 0 HTC-P1-C13 PI P1 0.99834 0.00027 2.54188E-01 1.57 1.10 PI PI 0 HTC-P1-C14 PI PI 0.99813 0.00024 2.32169E-01 1.57 1.10 Pi PI 0 HTC-P1-C15 P1 PI 0.99766 0.00025 2.28552E-01 1.57 1.10 PI PI 0 HTC-P1-016 PI PI 1.00008 0.00023 1.00989E-01 1.57 1.10 Pi PI 0 HTC-P1-(17 PI PI 0.99937 0.00021 9.89056E-02 1.57 1.10 PI PI 0 HTC-P1-C18 P1 P1 0.99696 0.00024 1.01103E-01 1.57 1.10 PI PI 0 HTC-P2-BOR-CO1 PI P1 0.99878 0.00024 2.45110E-01 1.57 1.10 Pi PI 100 HTC-P2-BOR-C02 PI P1 0.99783 0.00026 2.42638E-01 1.57 1.10 PI PI 106 HTC-P2-BOR-C03 PI P1 0.99790 0.00024 2.52969E-01 1.57 1.10 PI PI 205 HTC-P2-BOR-C04 Pi Pi 0.99880 0.00024 2.61183E-0i 1.57 1.10 Pi Pi 299 HTC-P2-BOR-C05 P1 P1 0.99855 0.00022 2.72138E-01 1.57 1.10 Pi PI 400 HTC-P2-BOR-C06 P1 P1 0.99823 0.00023 2.68813E-01 1.57 1.10 PI PI 399 HTC-P2-BOR-C07 PI P1 0.99934 0.00027 2.77584E-01 1.57 1.10 PI PI 486 HTC-P2-BOR-C08 PI PI 0.99847 0.00022 2.84657E-01 1.57 1.10 PI PI 587 HTC-P2-BOR-C09 Pi PI 0.99930 0.00022 1.65208E-01 1.57 1.10 Pi Pi 595 HTC-P2-BOR-C10 PI PI 0.99789 0.00022 1.60011E-01 1.57 1.10 PI PI 499 HTC-P2-BOR-C11 PI P1 0.99959 0.00023 1.55491E-01 1.57 1.10 PI PI 393 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-5 Table A-i: Critical Experiment Results with SCALE 6.0 Soluble Boron EALF U-235 Pu Pitch Pin Concentron Experiment kexp Yexp kcalc Ocalc (eV) (wt%) (wt%) (cm) (cm) Concentration (ppm)

HTC-P2-BOR-C12 PI PI 0.99963 0.00021 1.49222E-01 1.57 1.10 PI Pi 295 HTC-P2-BOR-C13 PI PI 0.99893 0.00024 1.44503E-01 1.57 1.10 PI PI 200 HTC-P2-BOR-C14 Pi PI 1.00255 0.00026 1.39080E-01 1.57 1.10 P1 PI 89 HTC-P2-BOR-C15 Pi PI 1.00337 0.00024 1.02584E-01 1.57 1.10 PI PI 90 HTC-P2-BOR-C16 PI Pi 1.00162 0.00024 1.06601E-01 1.57 1.10 PI Pi 194 HTC-P2-BOR-C17 P1 PI 1.00309 0.00021 1.09778E-01 1.57 1.10 PI Pi 286 HTC-P2-BOR-C18 PI Pi 0.99343 0.00020 1.15191E-01 1.57 1.10 PI Pi 415 HTC-P2-BOR-C19 PI Pi 1.00041 0.00023 1.04132E-01 1.57 1.10 PI Pi 100 HTC-P2-BOR-C20 P1 PI 0.99279 0.00020 8.92436E-02 1.57 1.10 Pi PI 220 HTC-P2-BOR-C21 PI PI 0.99689 0.00026 8.57123E-02 1.57 1.10 PI Pi 110 HTC-P3-C12 PI Pi 0.99965 0.00024 1.12085E-01 1.57 1.10 Pi Pi 0 HTC-P3-C13 P1 PI 0.99956 0.00028 1.10964E-01 1.57 1.10 Pi Pi 0 HTC-P3-C14 Pi PI 0.99990 0.00023 1.11089E-01 1.57 1.10 PI PI 0 HTC-P3-C15 PI P1 0.99938 0.00017 1.10305E-01 1.57 1.10 PI PI 0 HTC-P3-C16 Pi PI 0.99949 0.00027 1.09776E-01 1.57 1.10 PI P1 0 HTC-P3-C17 P1 PI1 0.99991 0.00023 1.07865E-01 1.57 1.10 PI PI 0 HTC-P3-C18 P1 PI 0.99955 0.00023 1.05957E-01 1.57 1.10 PI PI 0 HTC-P3-C19 Pi PI 1.00008 0.00022 1.03631E-01 1.57 1.10 PI PI 0 HTC-P3-C20 PI PI 0.99967 0.00023 1.01581E-01 1.57 1.10 PI PI 0 HTC-P3-C21 Pi PI 1.00014 0.00022 1.04065E-01 1.57 1.10 PI . PI 0 HTC-P3-C22 Pi PI 1.00056 0.00023 1.06508E-01 1.57 1.10 P1 PI 0 HTC-P3-C23 PI PI 0.99996 0.00023 1.14074E-01 1.57 1.10 PI PI 0 HTC-P3-C24 Pi PI 0.99981 0.00025 1.49740E-01 1.57 1.10 PI Pi 0 HTC-P3-C25 Pi PI 0.99952 0.00023 1.26086E-01 1.57 1.10 PI Pi 0 HTC-P3-C26 Pi PI 0.99922 0.00026 1.14808E-01 1.57 1.10 PI Pi 0 HTC-P4-ST-C22 PI Pi 1.00070 0.00023 1.72443E-01 1.57 1.10 PI PI 0 HTC-P4-ST-C23 P3 Pi 1.00096 0.00024 1.65006E-01 1.57 1.10 PI Pi 0 HTC-P4-ST-C24 PI PI 0.99964 0.00024 1.57252E-01 1.57 1.10 PI Pi 0 HTC-P4-ST-C25 -PI Pl 0.99960 0.00021 1.55742E-01 1.57 1.10 PI Pi 0 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-6 Table A-i: Critical Experiment Results with SCALE 6.0 U-235 Pu Pitch Pin Soluble SoulBrn Boron Experiment aex kcaic Oci EALF AF U3 u Pth Pn Concentration kexp Cexp Ocalc (eV) (wt%) (wt%) (cm) (cm) (ppm)

(ppm)

HTC-P4-ST-C26 Pi Pf 0.99967 0.00017 1.54343E-01 1.57 1.10 PI PI 0 HTC-P4-ST-C27 Pi PI 0.99917 0.00024 1.53327E-01 1.57 1.10 PI PI 0 HTC-P4-ST-C28 Pi PI 0.99909 0.00025 1.52264E-01 1.57 1.10 PI PI 0 HTC-P4-ST-C29 PI PI 0.99914 0.00023 1.43086E-01 1.57 1.10 PI PI 0 HTC-P4-ST-C30 PI PI 0.99993 0.00023 1.33505E-01 1.57 1.10 PI PI 0 HTC-P4-ST-C31 Pi Pi 0.99925 0.00021 1.27756E-01 1.57 1.10 Pi Pi 0 HTC-P4-ST-C32 PI PI 0.99990 0.00028 1.24432E-01 1.57 1.10 Pi Pi 0 HTC-P4-ST-C33 PI PI 0.99961 0.00022 1.22402E-01 1.57 1.10 Pi PI 0 LEU-COMP-THERM-001-001 1.00000 0.00300 0.99793 0.00026 9.60223E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-001-002 1.00000 0.00300 0.99780 0.00027 9.54523E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-001-003 1.00000 0.00300 0.99701 0.00026 9.44835E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-001-004 1.00000 0.00300 0.99742 0.00026 9.52462E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-001-005 1.00000 0.00300 0.99564 0.00026 9.39287E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-001-006 1.00000 0.00300 0.99783 0.00024 9.47277E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-001-007 1.00000 0.00310 0.99754 0.00024 9.30922E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-001-008 1.00000 0.00300 0.99636 0.00025 9.41497E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-002-001 1.00000 0.00200 0.99742 0.00019 1.12915E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-002-002 1.00000 0.00200 0.99941 0.00030 1.12849E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-002-003 1.00000 0.00200 0.99817 0.00032 1.12769E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-002-004 1.00000 0.00180 0.99781 0.00030 1.11734E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-002-005 1.00000 0.00190 0.99633 0.00026 1.10072E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-006-001 1.00000 0.00200 0.99765 0.00028 2.35080E-01 2.60 0.00 1.849 1.417 0 LEU-COMP-THERM-006-002 1.00000 0.00200 0.99809 0.00027 2.41958E-01 2.60 0.00 1.849 1.417 0 LEU-COMP-THERM-006-003 1.00000 0.00200 0.99848 0.00029 2.48401E-01 2.60 0.00 1.849 1.417 0 LEU-COMP-THERM-006-004 1.00000 0.00200 0.99828 0.00029 1.81216E-01 2.60 0.00 1.956 1.417 0 LEU-COMP-THERM-006-005 1.00000 0.00200 0.99788 0.00027 1.86581E-01 2.60 0.00 1.956 1.417 0 LEU-COMP-THERM-006-006 1.00000 0.00200 0.99922 0.00030 1.91345E-01 2.60 0.00 1.956 1.417 0 LEU-COMP-THERM-006-007 1.00000 0.00200 0.99880 0.00019 1.96262E-01 2.60 0.00 1.956 1.417 0 LEU-COMP-THERM-006-008 1.00000 0.00200 0.99903 0.00030 2.01764E-01 2.60 0.00 1.956 1.417 0 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-7 Table A-i: Critical Experiment Results with SCALE 6.0 Soluble Boron EALF U-235 Pu Pitch Pin Concentron Experiment kexp Cexp kcaic ycalc (eV) (wt%) (wt%) (cm) (cm) Concentration (ppm)

LEU-COMP-THERM-006-009 1.00000 0.00200 0.99868 0.00027 1.35151E-01 2.60 0.00 2.150 1.417 0 LEU-COMP-THERM-006-010 1.00000 0.00200 0.99826 0.00026 1.38813E-01 2.60 0.00 2.150 1.417 0 LEU-COMP-THERM-006-011 1.00000 0.00200 0.99854 0.00028 1.42056E-01 2.60 0.00 2.150 1.417 0 LEU-COMP-THERM-006-012 1.00000 0.00200 0.99787 0.00028 1.45591E-01 2.60 0.00 2.150 1.417 0 LEU-COMP-THERM-006-013 1.00000 0.00200 0.99860 0.00018 1.48648E-01 2.60 0.00 2.150 1.417 0 LEU-COMP-THERM-006-014 1.00000 0.00200 0.99909 0.00028 1.14215E-01 2.60 0.00 2.293 1.417 0 LEU-COMP-THERM-006-015 1.00000 0.00200 0.99913 0.00024 1.17080E-01 2.60 0.00 2.293 1.417 0 LEU-COMP-THERM-006-016 1.00000 0.00200 0.99909 0.00026 1.19603E-01 2.60 0.00 2.293 1.417 0 LEU-COMP-THERM-006-017 1.00000 0.00200 0.99902 0.00030 1.22255E-01 2.60 0.00 2.293 1.417 0 LEU-COMP-THERM-006-018 1.00000 0.00200 0.99871 0.00029 1.24857E-01 2.60 0.00 2.293 1.417 0 LEU-COMP-THERM-007-001 1.00000 0.00140 0.99576 0.00029 2.40635E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-007-002 1.00000 0.00080 0.99861 0.00028 1.08853E-01 4.74 0.00 1.600 0.940 0 LEU-COMP-THERM-007-003 1.00000 0.00070 0.99763 0.00028 7.07226E-02 4.74 0.00 2.100 0.940 0 LEU-COMP-THERM-007-004 1.00000 0.00080 0.99803 0.00024 6.04989E-02 4.74 0.00 2.520 0.940 0 LEU-COMP-THERM-008-001 1.00070 0.00120 0.99718 0.00020 2.77959E-01 2.46 0.00 1.636 1.206 1511 LEU-COMP-THERM-008-002 1.00070 0.00120 0.99758 0.00022 2.45199E-01 2.46 0.00 1.636 1.206 1336 LEU-COMP-THERM-008-003 1.00070 0.00120 0.99814 0.00021 2.45049E-01 2.46 0.00 1.636 1.206 1336 LEU-COMP-THERM-008-004 1.00070 0.00120 0.99762 0.00020 2.45757E-01 2.46 0.00 1.636 1.206 1182 LEU-COMP-THERM-008-005 1.00070 0.00120 0.99692 0.00025 2.45390E-01 2.46 0.00 1.636 1.206 1182 LEU-COMP-THERM-008-006 1.00070 0.00120 0.99733 0.00019 2.44533E-01 2.46 0.00 1.636 1.206 1033 LEU-COMP-THERM-008-007 1.00070 0.00120 0.99723 0.00023 2.44519E-01 2.46 0.00 1.636 1.206 1033 LEU-COMP-THERM-008-008 1.00070 0.00120 0.99672 0.00022 2.42606E-01 2.46 0.00 1.636 1.206 794 LEU-COMP-THERM-008-009 1.00070 0.00120 0.99709 0.00021 2.41889E-01 2.46 0.00 1.636 1.206 779 LEU-COMP-THERM-008-010 1.00070 0.00120 0.99728 0.00022 2.48087E-01 2.46 0.00 1.636 1.206 1245 LEU-COMP-THERM-008-011 1.00070 0.00120 0.99772 0.00020 2.53435E-01 2.46 0.00 1.636 1.206 1384 LEU-COMP-THERM-008-012 1.00070 0.00120 0.99772 0.00019 2.46975E-01 2.46 0.00 1.636 1.206 1348 LEU-COMP-THERM-008-013 1.00070 0.00120 0.99764 0.00015 2.47396E-01 2.46 0.00 1.636 1.206 1348 LEU-COMP-THERM-008-014 1.00070 0.00120 0.99737 0.00022 2.49245E-01 2.46 0.00 1.636 1.206 1363 LEU-COMP-THERM-008-015 1.00070 0.00120 0.99690 0.00023 2.49589E-01 2.46 0.00 1.636 1.206 1363 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-8 Experiment kexp O'exp kcalc II Table A-i: Critical Experiment Results with SCALE 6.0 caic c c eV)

EALF (eV)

U-235 (wt%)

Pu Pu (wt%)

Pch Pitch (cm)

Pin (cm)(pm SoulBrn Concentration (ppm)

LEU-COMP-THERM-008-016 1.00070 0.00120 0.99799 0.00022 2.27234E-01 2.46 0.00 1.636 1.206 1158 LEU-COMP-THERM-008-017 1.00070 0.00120 0.99700 10.00021 1.98152E-01 2.46 0.00 1.636 1.206 921 LEU-COMP-THERM-009-001 1.00000 0.00210 0.99833 0.00033 1.12619E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-009-002 1.00000 0.00210 0.99781 0.00028 1.11946E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-009-003 1.00000 0.00210 0.99801 0.00027 1.12284E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-009-004 1.00000 0.00210 0.99819 0.00027 1.12066E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-009-024 1.00000 0.00210 0.99821 0.00028 1.11967E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-009-025 1.00000 0.00210 0.99836 0.00029 1.11765E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-009-026 1.00000 0.00210 0.99859 0.00028 1.11910E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-009-027 1.00000 0.00210 0.99822 0.00028 1.11733E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-005 1.00000 0.00210 0.99952 0.00027 3.47759E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-006 1.00000 0.00210 1.00005 0.00029 2.56671E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-007 1.00000 0.00210 1.00033 0.00031 2.05823E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-008 1.00000 0.00210 0.99723 0.00027 1.81943E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-009 1.00000 0.00210 1.00081 0.00029 1.21866E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-010 1.00000 0.00210 1.00038 0.00026 1.17949E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-011 1.00000 0.00210 1.00086 0.00028 1.15244E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-012 1.00000 0.00210 0.99917 0.00026 1.12098E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-013 1.00000 0.00210 0.99708 0.00027 1.10376E-01 4.31 0.00 2.540 1.415 0 LEU-COMP-THERM-010-014 1.00000 0.00280 1.00084 0.00028 3.06441E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-015 1.00000 0.00280 1.00144 0.00027 2.94126E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-016 1.00000 0.00280 1.00215 0.00028 2.84458E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-017 1.00000 0.00280 1.00127 0.00029 2.78589E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-018 1.00000 0.00280 1.00156 0.00030 2.73578E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-019 1.00000 0.00280 1.00070 0.00029 2.66767E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-024 1.00000 0.00280 0.99880 0.00028 5.90544E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-025 1.00000 0.00280 1.00014 0.00030 5.44794E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-026 1.00000 0.00280 1.00068 0.00026 5.05596E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-027 1.00000 0.00280 1.00104 0.00031 4.72194E-01 4.31 0.00 1.892 1.415 0 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-9 Table A-i: Critical Experiment Results with SCALE 6.0 EALF U-235 Pu Pitch Pin SoulBrn Experiment Ien kcaic cac ELU23 Pu ith in Concentration nkexp Yexp Ocalc (eV) (wt%) (wt%) (cm) (cm) (ppm)

(ppm)

LEU-COMP-THERM-010-028 1.00000 0.00280 1.00159 0.00029 4.41850E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-029 1.00000 0.00280 1.00154 0.00030 4.17687E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-010-030 1.00000 0.00280 0.99867 0.00029 3.64185E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-011-001 1.00000 0.00180 0.99697 0.00024 1.67660E-01 2.46 0.00 1.636 1.206 0 LEU-COMP-THERM-01 1-002 1.00000 0.00320 0.99627 0.00023 2.43586E-01 2.46 0.00 1.636 1.206 1037 LEU-COMP-THERM-01 1-003 1.00000 0.00320 0.99672 0.00021 1.91545E-01 2.46 0.00 1.636 1.206 769 LEU-COMP-THERM-011-004 1.00000 0.00320 0.99739 0.00026 1.91672E-01 2.46 0.00 1.636 1.206 764 LEU-COMP-THERM-011-005 1.00000 0.00320 0.99671 0.00023 1.92646E-01 2.46 0.00 1.636 1.206 762 LEU-COMP-THERM-011-006 1.00000 0.00320 0.99676 0.00022 1.93682E-01 2.46 0.00 1.636 1.206 753 LEU-COMP-THERM-011-007 1.00000 0.00320 0.99747 0.00024 1.94914E-01 2.46 0.00 1.636 1.206 739 LEU-COMP-THERM-011-008 1.00000 0.00320 0.99679 0.00022 1.96179E-01 2.46 0.00 1.636 1.206 721 LEU-COMP-THERM-01 1-009 1.00000 0.00320 0.99689 0.00023 1.97085E-01 2.46 0.00 1.636 1.206 702 LEU-COMP-THERM-011-010 1.00000 0.00170 0.99416 0.00026 1.85372E-01 2.46 0.00 1.636 1.206 0 LEU-COMP-THERM-011-011 1.00000 0.00170 0.99420 0.00025 1.62230E-01 2.46 0.00 1.636 1.206 0 LEU-COMP-THERM-011-012 1.00000 0.00170 0.99390 0.00026 1.66445E-01 2.46 0.00 1.636 1.206 0 LEU-COMP-THERM-011-013 1.00000 0.00170 0.99454 0.00026 1.46584E-01 2.46 0.00 1.636 1.206 0 LEU-COMP-THERM-011-014 1.00000 0.00170 0.99494 0.00028 1.49835E-01 2.46 0.00 1.636 1.206 0 LEU-COMP-THERM-011-015 1.00000 0.00180 0.99526 0.00025 1.38204E-01 2.46 0.00 1.636 1.206 0 LEU-COMP-THERM-013-001 11.00000 0.00180 0.99995 0.00030 2.83620E-01 4.31 0.00 1.892 1.415 0 LEU-COMP-THERM-016-001 1.00000 0.00310 0.99725 0.00026 9.50629E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-002 11.00000 0.00310 0.99584 0.00026 9.48495E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-003 1.00000 0.00310 0.99742 0.00024 9.47196E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-01 6-004 1.00000 0.00310 0.99635 0.00030 9.48572E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-005 1.00000 0.00310 0.99664 0.00028 9.45.121E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-006 1.00000 0.00310 0.99695 0.00025 9.55380E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-007 1.00000 0.00310 0.99698 0.00027 9.52749E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-028 1.00000 0.00310 0.99769 0.00023 9.42677E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-029 1.00000 0.00310 0.99661 0.00023 9.41519E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-030 1.00000 0.00310 0.99669 0.00027 9.42659E-02 2.35 0.00 2.032 1.270 0 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-10 Table A-i: Critical Experiment Results with SCALE 6.0 Experiment Gex kcaic cac ELU-3 EALF U-235 Pu Pu ith Pitch Piin

.n SoulBrn Concentration kexp Yexp Ocalc (eV) (wt%) (wt%) (cm) (cm) (ppm)

(ppm)

LEU-COMP-THERM-016-031 1.00000 0.00310 0.99784 0.00025 9.43161E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-016-032 1.00000 0.00310 0.99699 0.00025 9.41851E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-004 1.00000 0.00310 0.99785 0.00018 1.97878E-01 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-005 1.00000 0.00310 0.99908 0.00031 1.74945E-01 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-006 1.00000 0.00310 0.99964 0.00025 1.65171E-01 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-007 1.00000 0.00310 0.99896 0.00024 1.57544E-01 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-008 1.00000 0.00310 0.99728 0.00027 1.31567E-01 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-009 1.00000 0.00310 0.99696 0.00023 1.08423E-01 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-010 1.00000 0.00310 0.99800 0.00026 9.92831E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-011 1.00000 0.00310 0.99789 0.00025 9.75449E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-012 1.00000 0.00310 0.99766 0.00027 9.63018E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-013 1.00000 0.00310 0.99748 0.00027 9.49642E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-014 1.00000 0.00310 0.99832 0.00026 9.42375E-02 2.35 0.00 2.032 1.270 0 LEU-COMP-THERM-017-015 1.00000 0.00280 0.99742 0.00027 1.76297E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-016 1.00000 0.00280 0.99741 0.00017 1.70459E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-017 1.00000 0.00280 0.99909 0.00025 1.65640E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-018 1.00000 0.00280 0.99724 0.00027 1.64000E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-019 1.00000 0.00280 0.99722 0.00029 1.61494E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-020 1.00000 0.00280 0.99644 0.00026 1.60038E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-021 1.00000 0.00280 0.99690 0.00029 1.58663E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-022 1.00000 0.00280 0.99560 0.00025 1.57547E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-026 1.00000 0.00280 0.99500 0.00028 ' 3.65166E-0i 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-027 1.00000 0.00280 0.99705 0.00030 3.14364E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-028 1.00000 0.00280 0.99794 0.00027 2.74751E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-017-029 1.00000 0.00280 0.99808 0.00030 2.46321E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-035-001 1.00000 0.00180 0.99810 0.00029 2.07286E-01 2.60 0.00 1.956 1.417 70 LEU-COMP-THERM-035-002 1.00000 0.00190 0.99708 0.00028 2.1111OE-01 2.60 0.00 1.956 1.417 148 LEU-COMP-THERM-039-001 1.00000 0.00140 0.99528 0.00029 2.21580E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-002 1.00000 0.00140 0.99676 0.00030 2.11227E-01 4.74 0.00 1.260 0.940 0 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-I11 Table A-i: Critical Experiment Results with SCALE 6.0 Pin Soluble Baron Concentron EALF U-235 Pu Pitch Experiment kexp Texp i kcalc calc (eV) (wt%) (wt%) (cm) (cm) Concentration (ppm)

LEU-COMP-THERM-039-003 1.00000 0.00140 0.99636 0.00028 1.91986E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-004 1.00000 0.00140 0.99545 0.00028 1.83413E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-005 1.00000 0.00090 0.99795 0.00022 1.39070E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-006 1.00000 0.00090 0.99789 0.00031 1.45245E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-007 1.00000 0.00120 0.99637 0.00030 2.12411E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-008 1.00000 0.00120 0.99584 0.00029 2.02607E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-009 1.00000 0.00120 0.99672 0.00030 1.97023E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-010 1.00000 0.00120 0.99733 0.00028 1.72665E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-011 1.00000 0.00130 0.99498 0.00032 2.21378E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-012 1.00000 0.00130 0.99558 0.00029 2.15945E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-013 1.00000 0.00130 0.99534 0.00028 2.13975E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-014 1.00000 0.00130 0.99569 0.00031 2.11982E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-015 1.00000 0.00130 0.99581 0.00029 2.10932E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-016 1.00000 0.00130 0.99633 0.00033 2.09911E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-039-017 1.00000 0.00130 0.99651 0.00031 2.09562E-01 4.74 0.00 1.260 0.940 0 LEU-COMP-THERM-042-001 1.00000 0.00160 0.99717 0.00026 1.68005E-01 2.35 0.00 1.684 1.270 0 LEU-COMP-THERM-048-001 1.00000 0.00250 0.99782 0.00030 6.74014E-01 3.00 0.00 1.320 1.094 0 LEU-COMP-THERM-048-002 1.00000 0.00250 0.99838 0.00027 6.46665E-01 3.00 0.00 1.320 1.094 0 LEU-COMP-THERM-048-003 1.00000 0.00250 0.99766 0.00025 6.77065E-01 3.00 0.00 1.320 1.094 0 LEU-COMP-THERM-048-004 1.00000 0.00250 0.99833 0.00026 6.78792E-01 3.00 0.00 1.320 1.094 0 LEU-COMP-THERM-048-005 1.00000 0.00250 0.99765 0.00027 6.69080E-01 3.00 0.00 1.320 1.094 0 LEU-COMP-THERM-050-001 1.00040 0.00100 0.99800 0.00026 1.99223E-01 4.74 0.00 1.300 0.940 0 LEU-COMP-THERM-050-002 1.00040 0.00100 0.99760 0.00030 1.90624E-01 4.74 0.00 1.300 0.940 0 LEU-COMP-THERM-050-003 1.00040 0.00100 0.99737 0.00028 2.07226E-01 4.74 0.00 1.300 0.940 822 LEU-COMP-THERM-050-004 1.00040 0.00100 0.99714 0.00028 1.97634E-01 4.74 0.00 1.300 0.940 822 LEU-COMP-THERM-050-005 1.00040 0.00100 0.99868 0.00029 2.21841E-01 4.74 0.00 1.300 0.940 5030 LEU-COMP-THERM-050-006 1.00040 0.00100 0.99898 0.00027 2.13406E-01 4.74 0.00 1.300 0.940 5030 LEU-COMP-THERM-050-007 1.00040 0.00100 0.99917 0.00028 2.09380E-01 4.74 0.00 1.300 0.940 5030 LEU-COMP-THERM-051-001 1.00000 0.00200 0.99698 0.00027 1.46810E-01 2.46 0.00 1.636 1.206 143 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-12 Table A-I: Critical Experiment Results with SCALE 6.0 U-235 Pu Pitch Pin Soluble Boron SoulBrn Experiment aci EALF Oex kcaic AF U3 u Pth Pn Concentration nkexp Texp kcalc (eV) (wt%) (wt%) (cm) (cm) (ppm)

(ppm)

LEU-COMP-THERM-051-002 1.00000 0.00240 0.99893 0.00026 1.95349E-01 2.46 0.00 1.636 1.206 510 LEU-COMP-THERM-051-003 1.00000 0.00240 0.99825 0.00018 1.95140E-01 2.46 0.00 1.636 1.206 514 LEU-COMP-THERM-051-004 1.00000 0.00240 0.99804 0.00028 1.96828E-01 2.46 0.00 1.636 1.206 501 LEU-COMP-THERM-051-005 1.00000 0.00240 0.99839 0.00024 1.97383E-01 2.46 0.00 1.636 1.206 493 LEU-COMP-THERM-051-006 1.00000 0.00240 0.99766 0.00027 1.98857E-01 2.46 0.00 1.636 1.206 474 LEU-COMP-THERM-051-007 1.00000 0.00240 0.99822 0.00026 1.99085E-01 2.46 0.00 1.636 1.206 462 LEU-COMP-THERM-051-008 1.00000 0.00240 0.99796 0.00026 1.99996E-01 2.46 0.00 1.636 1.206 432 LEU-COMP-THERM-051-009 1.00000 0.00190 0.99776 0.00030 1.65995E-01 2.46 0.00 1.636 1.206 217 LEU-COMP-THERM-071-001 1.00000 0.00076 0.99431 0.00027 7.55299E-01 4.74 0.00 1.100 0.949 0 LEU-COMP-THERM-071-002 1.00000 0.00076 0.99452 0.00029 6.91495E-01 4.74 0.00 1.100 0.949 0 LEU-COMP-THERM-071-003 1.00000 0.00076 0.99433 0.00028 6.56265E-01 4.74 0.00 1.100 0.949 0 LEU-COMP-THERM-071-004 1.00000 0.00080 0.99381 0.00031 8.43154E-01 4.74 0.00 1.075 0.949 0 LEU-COMP-THERM-072-001 1.00000 0.00120 0.99845 0.00028 1.10056E-01 4.74 0.00 1.600 0.949 0 LEU-COMP-THERM-072-002 1.00000 0.00120 0.99734 0.00027 1.06185E-01 4.74 0.00 1.600 0.949 0 LEU-COMP-THERM-072-003 1.00000 0.00120 0.99799 0.00030 1.08328E-01 4.74 0.00 1.600 0.949 0 LEU-COMP-THERM-077-003 1.00040 0.00100 1.00087 0.00027 1.61776E-01 4.35 0.00 1.500 0.980 0 LEU-COMP-THERM-082-003 1.00060 0.00100 1.00109 0.00026 1.49387E-01 4.35 0.00 1.500 0.980 0 LEU-COMP-THERM-083-001 1.00070 0.00100 1.00056 0.00032 1.51210E-01 4.35 0.00 1.500 0.980 0 LEU-COMP-THERM-084-001 1.00040 0.00100 1.00012 0.00026 1.54145E-01 4.35 0.00 1.500 0.980 0 LEU-COMP-THERM-089-001 1.00030 0.00100 1.00030 0.00028 1.52861E-01 4.35 0.00 1.500 0.980 0 LEU-COMP-THERM-090-00i 1.00050 0.00100 11.00019000026 45795E-01 T 435 1 000 1. 00 0.980 1 0 Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-13 A.4 EXPERIMENTAL DATA ANALYSIS A.4.1 Data Trends The next step in the analysis is to determine if there are any trends in the data. Only real or statistically significant trends are of importance. The null hypothesis is that the slope of the trend is zero (no trend) and is tested to determine if there is 95% confidence that the calculated slope is a more accurate representation than a zero slope. The equations from NUREG/CR-6698 are applied to the results in Table 1 to calculate the fitting coefficients. A test statistic is calculated and compared to Student's t-distribution with 95% confidence and n-2 degrees of freedom. If the absolute value of the test statistic is greater than the t-distribution, this indicates a statistically significant trend exists. Results are shown in Tables A-2 and Table A-3.

If a trend in a parameter is found, NUREG/CR-6698 describes the use of a one-sided tolerance band for criticality validation. The width of the tolerance band is the bias uncertainty. This analysis applies the equations given in NUREG/CR-6698.

Table A-2: Trend Analysis Results - SFP Data Set b, e Table A-3: Trend Analysis Results - NFSR data set b, e Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-14 A.4.2 Data Statistical Distribution Tests The statistical treatment used follows the guidance provided in NUREG/CR-6698. The NUREG approach weights the calculated Keff values by the experimental uncertainty. This approach means the high quality experiments affect the results more than the low quality experiments. The uncertainty weighting is used for the analysis of the set of experiments as a whole as well as for analysis for trends.

As recommended by NUREG/CR-6698, the results of the validation are checked for normality. The critical experiments were tested with the D'Agostino D' test of Normality, which is appropriate for the sample size (253 data points for the SFP data set). The Keff values from Table 1 were found to adhere to a normal distribution at the 95% confidence level. The D' test was also performed for the NFSR data set of 146 critical experiments. The results also support the assumption of normality.

A.4.3 Statistical Data Treatment Method The results of the trending analysis above showed statistically significant trends in 5 relevant trending parameters: EALF, U-235 enrichment, Plutonium content, lattice pitch, and fuel rod diameter. Results are shown in Tables A-4 and Table A-5.

Table A-4 One-Sided Tolerance Band Results for Statistically Significant Trended Parameters (SFP Data Set) b, e Table A-5 One-Sided Tolerance Band Results for Statistically Significant Trended Parameters b, e (NFSR Data Set)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A- 15 A.4.5 Subcritical Margin The subcritical margin for borated spent fuel pools, casks, and fully flooded dry storage racks is 0 when the analysis is done with unborated water. This is 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 253 critical experiments, which cover the range of expected conditions, the lowest calculated Keff was 0.99279. This supports that the subcritical margin is more than sufficient.

A.4.4 Determine Bias and Bias Uncertainty The bias and bias uncertainty is determined for SFP and NFSR analyses based on the statistical analysis.

The data is normally distributed. Since trends for physical parameters were shown to exist, a one-sided tolerance band is utilized. The code bias is calculated by (1 - kfit) where kfi is most limiting for the parameters analyzed. The code uncertainty is the one-sided tolerance bandwidth associated with this kfit using the equations from NUREG/CR-6698.

For SFP analysis, the limiting tolerance band is for EALF The bias and uncertainty for SFP analysis are based on the EALF of the system analyzed. The limiting EALF for the SFP uncertainty cases is 0.65 eV. Interpolating the EALF kfi values from Table 4 for 0.65 eV, results in a kfit (code bias) of [ ]be with a bandwidth (bias uncertainty) of [ Pb,e For NFSR analysis, the bias and uncertainty is also based on the EALF of the system analyzed. The EALF for the limiting fully flooded NFSR conditions is 0.24 eV. Interpolating the EALF kfit values from Table A-5 for 0.24 eV, results in a kfit (code bias) of [ ]b " A bandwidth of [ b,e is conservatively used as the bias uncertainty based on the maximum bandwidth of all trends for NFSR analysis, although it is not required to do so.

A.5 DEFINE AREA OF APPLICABILITY & LIMITATIONS The critical benchmarks were selected to cover the materials, geometry, and energy spectrum applicable to the Millstone 2 analysis. Little to no extrapolation for key parameters important to criticality is needed.

Table A-6 is provided below to summarize the range of the benchmark applicability (or area of applicability) for this analysis Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-16 Table A-6 Physical Parameters for Area of Applicabillity

~Parameter~ ~ Range Comments~

Fissionable Material/Physical Form U0 2 PuO 2 only for the spent fuel pool.

U0 2+PuO 2 Enrichment (wt% U-235) 1.57 to 4.74 2.35% minimum U-235 enrichment for new fuel storage area.

Enrichment (wt% Pu) 1.10 Experiments using plutonium are limited to the HTC critical experiments, which all use the same fuel rods. The fuel rods were manufactured to approximate the uranium and plutonium isotopic concentration of PWR fuel with 37.5 GWd/MTU burnup (without fission products).

Spectrum The experiments cover the expected range of EALF (eV) 0.0605 to conditions.

0.8468 Lattice Characteristics Hex lattices have been excluded. The range Type Square/ covers fuel types used in this analysis with the Pin Pitch (cm) Rectangular exception of disassembled fuel e.g. the 1.075 to 2.54 consolidated rod basket.

Assembly Spacing in Racks Note that the spacing is assumed to be filled Distance between Assemblies (cm) 0 to 18 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 All configurations are within this range.

Soluble Boron Concentration 0 to 5030 ppm Reflector All configurations are covered.

Experiments included water steel, Reflectors and concrete. adequately covered Temperature Room The criticality calculations are done with the Temperature fuel at the low temperatures. No significant Doppler feedback.

Moderating material water The moderator in all benchmark experiments is air water, therefore water as a moderating material oxygen in fuel is covered. Experiments typically have a portion of the fuel above the critical water level, so air or vacuum conditions are also applicable.

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 A-17

• Parameter Range Comments Cladding & Structural materials Al, SS, Cladding material for most IH experiments is Zircaloy, aluminum or stainless steel which covers the Concrete use of stainless pins and structural materials.

Cladding material for HTC experiments is zirconium which covers normal fuel rods.

A.6 REFERENCES A-1 NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," January 2001. (ADAMS Accession No. ML050250061)

Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 B-1 APPENDIX B SAMPLE TRITON INPUT FILE Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 B-2 During discussions at pre-submittal meetings, NRC staff requested detailed information regarding use of the TRITON computer code which is employed in this analysis. This appendix provides a sample input file for a case representative of the MP2 fuel to satisfy this request.

TRITON Input File - 14x14 Representative Fuel (with grids), 50 GWd/MTU

=t5-depl parm=(addnux=3)

Sample 14x14, dimensions approximate, 800 ppm, node 17, zr grid, no control rod, 50 GWD/T v7-238 Read Comp

'Node 17 5.0 w/o 50 GWD/MTU' U02 1 0.946 907 92235 5.0 92238 95.0 END

'Clad' ZIRC4 2 1.0 598 END

' Mod 800 ppm with max size zr grid smeared around fuel pins' H20 3 den=0.667 0.96 598 END WTPTBOR 3 0.667 15000 100768e-6 598 END ZIRC4 3 0.04 598 END

'GT' ZIRC4 4 1.0 598 END

'GT Mod' H20 5 den=0.667 1 598 END WTPTBOR 5 0.667 1 5000 100 800e-6 598 END H20 6 den=0.667 1 598 END WTPTBOR 6 0.667 1 5000 100 800e-6 598 END END COMP READ CELLDATA LATTICECELL SQUAREPITCH PITCH=1.4732 3 FUELD=0.96 1 GAPD=0.98 0 CLADD=1.12 2 end END CELLDATA READ DEPLETION 1 end END DEPLETION READ BURNDATA power=32.4 burn=10 down=0 nlib=l end power=32.4 burn=40 down=0 nlib=l end power=32.4 burn=50 down=0 nlib=l end power=32.4 burn=50 down=0 nlib=1 end power=32.4 burn=50 down=0 nlib=l end power=32.4 burn=1086 nlib=16 down=5 end END BU RNDATA READ MODEL Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 B-3 read parm gen=2000 npg=2000 run=yes plt=no htm=no far=no end parm READ GEOM UNIT 1

' Fuel pin' CYLINDER 110.48 19.29 0.

CYLINDER 0 10.49 19.29 0.

CYLINDER 2 1 0.56 19.29 0.

CUBOID 3 1 0.7365 -0.7365 0.7365 -0.7365 19.29 0.

UNIT 2

'Guide thimble' CYLINDER 5 1 1.31 19.29 0.

CYLINDER 4 1 1.42 19.29 0.

CUBOID 5 1 1.473 -1.473 1.473 -1.473 19.29 0.

UNIT 3

'2x2 fuel pin array' ARRAY 1 -0.7365 -0.7365 0 GLOBAL UNIT 4

'Fuel assembly surrounded by water in-core pitch' array 2 0.089 0.089 0 cuboid 6 1 20.8 0 20.8 0 19.29 0.

END GEOM READ ARRAY ARA=1 NUX=2 NUY=2 nuz=1 typ=cuboidal FILL 11 11 END FILL ARA=2 NUX=7 NUY=7 nuz=1 typ=cuboidal FILL 3333333 3233323 3333333 3332333 3333333 3233323 3333333 END FILL end array read bnds xfc=mirror yfc=mirror zfc=periodic end bnds Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 - Docket No. 50-336 - Proposed LAR-Criticality Analysis - Attachment 5 B-4 end data end model end

=shell cp stdcmp_mixOO01 /naf/ncd/ScaleWork/MP2_TRITON/compositions/sample-triton end Millstone Unit 2 Criticality Analysis Report

Serial No.12-678 Docket No. 50-336 Attachment 6 Affidavit of AREVA NP, Inc.

DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. (AREVA NP) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in Attachment 4 to a letter from J. Alan Price (Dominion) to U.S. NRC entitled "Dominion Nuclear Connecticut, Inc.,

Millstone Power Station Unit 2, License Amendment Request Regarding Proposed Technical Specifications Changes for Spent Fuel Storage," Serial No.12-678, and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information. The proprietary information is identified by its enclosure within pairs of brackets ("[ ]").

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be

withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this , j*

day of __________ ___2012.

Kathleen Ann Bennett NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31/15 Reg. # 110864 I KATHLEEN ANN BENNETT Notary Public Commonwealth of Virginia E

110864 My Commission Expires Aug 31, 20151!

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