L-2011-032, WCAP-17094-NP, Rev 3, Turkey Point, Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis, Attachment 2 to L-2011-032

WCAP-17094-NP, Rev 3, Turkey Point, Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis, Attachment 2 to L-2011-032

ML110560336
Person / Time
Site:	Turkey Point
Issue date:	02/28/2011
From:	Bishop T, Clarity J, Sanders C Westinghouse
To:	Office of Nuclear Reactor Regulation
References
L-2011-032, LAR 207 WCAP-17094-NP, Rev 3
Download: ML110560336 (187)

Text

Turkey Point Nuclear Plant L-2011-032 License Amendment Request No. 207 Attachment 2 Enclosure Turkey Point Units 3 and 4 LAR NO. 207 FUEL STORAGE CRITICALITY ANALYSIS SUPPLEMENT 1 ATTACHMENT 2 WCAP- 17094-NP, Rev 3, Turkey Point Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis February 2011 This coversheet plus 178 pages

Westinghouse Non-Proprietary Class 3 WCAP-17094-NP Febru ary 2011 Revision 3 Turkey Point Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis

)Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17094-NP Revision 3 Turkey Point Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis Tracy Bishop*

Justin Clarity*

Charlotta Sanders*

U.S. BWR and Criticality February 2011 Reviewer: Vefa Kucukboyaci*

U.S. BWR and Criticality Approved: Ed Mercier*, Manager Westinghouse Core Engineering

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2011 Westinghouse Electric Company LLC All Rights Reserved

ii REVISION HISTORY Revision Description and Impact of the Change Date 0 Not Issued 05/2010 1 Original Issue 05/2010 2 Remove allowances for the current AOR; the analysis documented will 07/2010 completely supersede the current AOR.

Addressed customer comments.

3 Reanalysis to address DSS-ISG-2010-1 02/2011 WCAP- 17094-NP February 2011 Revision 3

iii TABLE OF CONTENTS LIST O F TA BLES ......................................................................... I.............................................................. v LIST O F FIG U RES ................................................................................................................................... viii LIST OF ACRONYMS, INITIALISMS, AND TRADEMARKS .......................................................... ix IN N TRO D U CTION ........................................................................................................................ 1-1 2 O V ERV IEW ................................................................................................................................. 2-1 2.1 A CCEPTAN CE CRITERIA ............................................................................................ 2-1 2.1.1 Spent Fuel Pool Criteria .................................................................................. 2-1 2.1.2 N ew Fuel Storage Rack Criteria ...................................................................... 2-1 2.2 D ESIGN A PPROA CH ..................................................................................................... 2-2 2.3 CO M PUTER CO D ES ...................................................................................................... 2-2 2.4 CO RR ESPON D EN CE TO D SS-ISG -2010-1 .................................................................. 2-4 3 D ESIGN AN D IN PUT D ATA ...................................................................................................... 3-1 3.1 FU EL A SSEM BLY SPEC IFICATION S ...................................................................... 3-I 3.1.1 IFBA Specification .......................................................................................... 3-2 3.2 FUEL A SSEM BLY IN SERTS SPECIFICATION S ......................................................... 3-2 3.3 STO RA G E RA CK IN SERTS SPECIFICATION S .......................................................... 3-4 3.4 FU EL RO D BA SKETS ................................................................................................... 3-4 3.5 STO RA G E RA CK SPEC IFICATION S ........................................................................... 3-4 4 AN A LY TICA L M ETHO D O LO G Y ............................................................................................. 4-1 4.1 BO UN D IN G FUEL A SSEM BLY SELECTION ............................................................. 4-1 4.2 D EPLETION CA LCU LATION S .................................................................................... 4-6 4.2.1 Reactor Param eters .......................................................................................... 4-7 4.2.2 Burnable A bsorber and Insert Usage ............................................................. 4-16 4.2.3 Rodded O peration .......................................................................................... 4-17 4.2.4 Sum m ary of D epletion Param eters ................................................................ 4-18 4.3 CRITICA LITY CA LCULATION S ............................................................................... 4-19 4.3.1 K EN O M odel ................................................................................................. 4-19 4.3.2 U ncertainties .................................................................................................. 4-30 4.3.3 Calculation of Burnup versus Enrichm ent Curves ........................................ 4-34 4.3.4 Soluble Boron Credit ..................................................................................... 4-35 4.3.5 Interfaces ....................................................................................................... 4-35 5 RE SU LTS ..................................................................................................................................... 5-1 5.1 SPEN T FU EL POO L REG ION S I A N D II ...................................................................... 5-1 5.1.1 Fuel Category and Storage A rray Definitions ............................................. 5-1 5.1.2 B ias and U ncertainty Calculations .................................................................. 5-4 5.1.3 M inim um Burnup and IFBA Requirem ents .................................................... 5-8 5.1.4 Curve Fitting Coefficients for Minimum Burnup Requirements ................... 5-14 5.1.5 Confirm atory Criticality Calculations ........................................................... 5-17 5.2 CA SK A REA RACK ..................................................................................................... 5-26 5.3 N EW FUEL STO RAG E RA CK AN A LYSIS ................................................................ 5-27 5.3.1 M odel Description ......................................................................................... 5-27 5.3.2 Rack A nalysis ................................................................................................ 5-29 WCAP-17094-NP February 2011 Revision 3

iv 5.3.3 Sensitivity Analysis for the New Fuel Rack .................................................. 5-32 5.4 FUEL ROD STORAGE BASKET ................................................................................ 5-32 5.5 INTERFACE REQUIREMENTS .................................................................................. 5-35 5.5.1 Allowable Interface Configurations .............................................................. 5-35 5.5.2 Results of Region I - Region I1Interface Calculations ................................. 5-38 5.6 SOLUBLE BORON CREDIT ....................................................................................... 5-41 5.7 NORMAL AND ACCIDENT CONDITIONS ............................................................... 5-41 5.7.1 Evaluation of Normal Conditions in the Spent Fuel Pool ............................. 5-41 5.7.2 A ccident C onditions ...................................................................................... 5-44 6 LIMITATIONS OF ANALYSIS ................................................................................................... 6-1 6.1 FUEL LIMITATIONS .................................................................................................. 6-1 6.2 OPERATIONAL LIMITATIONS ................................................................................ 6-1 6.3 SPENT FUEL POOL LIMITATIONS ............................................................................. 6-2 7 RE FE REN C ES ............................................................................................................................. 7-1 APPENDIX A VALIDATION OF SCALE 5.1 .................................................................................. A-1 WCAP- 17094-NP February 2011 Revision 3

v LIST OF TABLES Table 2-1 Section to Section Correspondence of this Document to DSS-ISG-20 10-1 .................... 2-4 Table 3-I Fuel Assem bly Specifications ..................................................................................... 3-1 Table 3-2 R C C A Specifications ....................................................................................................... 3-2 Table 3-3 Pyrex and WA BA Specifications ..................................................................................... 3-3 Table 3-4 H afnium Insert Specifications .......................................................................................... 3-3 Table 3-5 Metam ic Insert Specifications .......................................................................................... 3-4 Table 3-6 Fuel Rod Basket Specifications ....................................................................................... 3-4 Table 3-7 Fuel Rack Specifications ............................................................................................. 3-5 Table 3-8 M aterial Com positions ................................................................................................ 3-5 Table 4-1 Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (0 ppm)4-2 Table 4-2 Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (500 ppm)

..................................................................................................... .................................. 4 -3 Table 4-3 Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (1600 p p m ) ................................................................................................................................. 4 -4 Table 4-4 Reactivity Comparison of STD Fuel with and without Grids at 39 F ............................. 4-5 0 Table 4-5 Reactivity Comparison of STD Fuel with and without Grids in the Presence of Metamic In serts ............................................................................................................................... 4 -6 Table 4-6 Reactivity Comparison of STD Fuel with and without Grids in the Presence of One E m pty C ell ....................................................................................................................... 4-6 Table 4-7 Calculation of keff for Various 11-3 Enrichment/Bumup Pairings in Array 11-B ............. 4-11 Table 4-8 Exit Moderator Temperature as a Function of PAssembly ................................................. 4-13 Table 4-9 Cycle Average Soluble Boron Concentrations for Pre-EPU Cycles .............................. 4-15 Table 4-10 Cycle Average Soluble Boron Concentrations for Projected EPU Cycles ..................... 4-16 Table 4-Il Parameters Used in Depletion Analysis ......................................................................... 4-18 Table 4-12 Burnable Absorber Modeling For Non-Blanketed Fuel Depletion ................................ 4-18 Table 4-13 [

]a,c ...................................................................................................... 4 -2 3 Table 4-14 Axial Burnup Profiles for Blanketed Fuel ..................................................................... 4-24 Table 4-15 Assemblies with the Most Limiting Axial Burnup Profiles ........................................... 4-25 Table 4-16 Axial Bumup Profiles for Non-Blanketed Fuel ............................................................. 4-27 Table 5-1 Fuel Categories Ranked by Reactivity ............................................................................. 5-1 WCAP-17094-NP February 2011 Revision 3

vi Table 5-2 Unborated Biases and Uncertainties for Region I Fuel Categories .................................. 5-4 Table 5-3 Unborated Biases and Uncertainties for Region II Fuel Categories ................................ 5-5 Table 5-4 Borated Biases and Uncertainties for Region I Fuel Categories ...................................... 5-6 Table 5-5 Borated Biases and Uncertainties for Region II Fuel Categories .................................... 5-7 Table 5-6 Burnup Requirements for Pre-EPU Non-Blanketed Category 1-3 Fuel ........................... 5-8 Table 5-7 Burnup Requirements for Pre-EPU Non-Blanketed Category 1-4 Fuel ........................... 5-8 Table 5-8 Bumup Requirements for Pre-EPU Non-Blanketed Category Il-1 Fuel ......................... 5-8 Table 5-9 Bumup Requirements for Pre-EPU Non-Blanketed Category 11-2 Fuel ......................... 5-9 Table 5-10 Bumup Requirements for Pre-EPU Non-Blanketed Category 11-3 Fuel ......................... 5-9 Table 5-11 Burnup Requirements for Pre-EPU Non-Blanketed Category 11-4 Fuel ......................... 5-9 Table 5-12 Burnup Requirements for Pre-EPU Non-Blanketed Category 11-5 Fuel ......................... 5-9 Table 5-13 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 1-3 Fuel ........ 5-10 Table 5-14 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 1-4 Fuel ........ 5-10 Table 5-15 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-1 Fuel ....... 5-11 Table 5-16 Bumup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-2 Fuel ....... 5-11 Table 5-17 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-3 Fuel ....... 5-12 Table 5-18 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-4 Fuel ....... 5-12 Table 5-19 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category Il-5 Fuel ....... 5-13 Table 5-20 IFBA Requirements for Fuel Category 1-2 .................................................................... 5-13 Table 5-21 Non-Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct) (See Notes 1-4 for use of Table 5-2 1) ......................................................................................................... 5-15 Table 5-22 Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct) (See Notes 1-6 for use of Table 5-22) .......................................................................................................... 5-16 Table 5-23 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 1-3 Fuel ........... 5-17 Table 5-24 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel ........... 5-17 Table 5-25 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 Fuel at 4.3 wt% 235U and 0 IFBA .................................................... 5-18 Table 5-26 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 Fuel at 4.4 wt% 235U and 32 IFBA .................................................. 5-18 Table 5-27 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 Fuel at 4.7 wt% 235U and 64 IFBA .................................................. 5-18 WCAP- 17094-NP February 2011 Revision 3

vii Table 5-28 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 Fuel at 5.0 wt% 235U and 80 IFBA .................................................. 5-19 Table 5-29 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 11-1 Fuel ........... 5-19 Table 5-30 Confirmatory krff Calculations for Pre-EPU Non-Blanketed Category 11-2 Fuel ........... 5-19 Table 5-31 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 11-3 Paired with Fuel C ategory 11-5 Fuel .................................................................................................. 5-20 Table 5-32 Confirmatory k~ff Calculations for Pre-EPU Non-Blanketed Category 11-4 Fuel ........... 5-20 Table 5-33 Confirmatory keff Calculations for Pre-EPU and EPU Blanketed Category 1-3 Fuel ..... 5-20 Table 5-34 Confirmatory k~ff Calculations for Pre-EPU and EPU Blanketed Category 1-4 Fuel ..... 5-21 Table 5-35 Confirmatory keff Calculations for Pre-EPU and EPU Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 at 4.3 wt% 235U and 0 IFBA .................................................... 5-21 Table 5-36 Confirmatory keff Calculations for Pre-EPU and EPU Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 at 4.4 wt% 235U and 32 IFBA .................................................. 5-22 Table 5-37 Confirmatory k~ff Calculations for Pre-EPU and EPU Blanketed Category 1-4 Paired with Fresh Category 1-2 at 4.7 wt% 235U and 64 IFBA .......................................................... 5-22 Table 5-38 Confirmatory keff Calculations for Pre-EPU and EPU Blanketed Category 1-4 Paired with Fresh Category 1-2 at 5.0 wt% 235U and 80 IFBA .......................................................... 5-23 Table 5-39 Confirmatory k~ff Calculations for Pre-EPU and EPU Blanketed Category Il-I Fuel ...5-23 Table 5-40 Confirmatory k~ff Calculations for Pre-EPU and EPU Blanketed Category 11-2 Fuel ...5-24 Table 5-41 Confirmatory keff Calculations for Pre-EPU and EPU Blanketed Category 11-3 Paired with C ategory 1I-5 Fuel .......................................................................................................... 5-24 Table 5-42 Confirmatory keff Calculations for Pre-EPU and EPU Blanketed Category 11-4 Fuel ...5-25 Table 5-43 Results for the New Fuel Storage Racks ........................................................................ 5-30 Table 5-44 keff Values with and Without Fuel Basket in Region II Storage Arrays ......................... 5-34 Table 5-45 Unborated Biases and Uncertainties for Region I and Region II Interfaces .................. 5-38 Table 5-46 Results for the Region I - Region II Interface Calculations .......................................... 5-39 Table 5-47 Results for the Normal Operations with 500 ppm of Soluble Boron ............................. 5-41 Table 5-48 Results of the Accident Calculations ............................................................................. 5-48 WCAP- 17094-NP February 2011 Revision 3

viii LIST OF FIGURES Figure 3-1 Turkey Point Spent Fuel Pool Layout (Unit 4 is the Same Except a Mirror inage) ............... 3-6 Figure 3-2 Top View of the Turkey Point New Fuel Storage Racks ......................................................... 3-7 Figure 3-3 Side View of the Turkey Point New Fuel Storage Racks ........................................................ 3-8 Figure 4-1 Average Relative Power as a Function of Assembly Burnup for Region I Analysis ............... 4-8 Figure 4-2 Average Relative Power as a Function of Assembly Burnup for Region II Analysis .............. 4-9 Figure 4-3 Averaged Assembly Relative Power as a Function of Burnup .............................................. 4-10 Figure 4-4 KENO Model of a Two Insert Case ...................................................................................... 4-19 Figure 4-5 Comparison of the Twelve Node and Twenty-Six Node Axial Approximations to the > 46.2 G W d/M TU B urnup Shape .................................................................................................... 4-26 Figure 4-6 Pellet Density as a Function of Burnup ................................................................................. 4-29 Figure 4-7 Clad Thickness as a Function of Burnup ............................................................................... 4-29 Figure 4-8 Example of a Region I Eccentric Positioning Model ............................................................ 4-33 Figure 4-9 Example of Interfaces between Region II Arrays ................................................................ 4-37 Figure 4-10 Region I and Region II Interface Model ............................................................................. 4-38 Figure 5-1 A llow able Region I Storage A rrays ......................................................................................... 5-2 Figure 5-2 A llowable Region II Storage Arrays ....................................................................................... 5-3 Figure 5-3 KENO Model for the Cask Area Rack .................................................................................. 5-26 Figure 5-4 N ew Fuel Storage Rack M odel ............................................................................................. 5-28 Figure 5-5 keff as a Function of Water Density for 4.5 wt% 235U Fuel in the New Fuel Storage Racks.. 5-31 235 Figure 5-6 klff as a Function of Water Density for 5 wt% U Fuel with 16 IFBA Rods ....................... 5-31 Figure 5-7 Radial View of a Fuel Rod Basket Model ............................................................................. 5-33 Figure 5-8 Allowable Array I-A- Region II Interfaces ............................................................................ 5-35 Figure 5-9 Allowable Array I-B- Region II Interfaces ............................................................................ 5-36 Figure 5-10 Allowable Array I-D- Region II Interfaces .......................................................................... 5-36 Figure 5-11 Interface Restrictions between Region I and Region II ....................................................... 5-40 Figure 5-12 Model of a Misloaded Assembly in Storage Array I1-A ...................................................... 5-45 Figure 5-13 Model for Misloading a Fresh Assembly in the Cask Area Rack Corner Cut ..................... 5-47 WCAP-17094-NP February 2011 Revision 3

ix LIST OF ACRONYMS, INITIALISMS, AND TRADEMARKS 2D Two-Dimensional 3D Three-Dimensional AEG Average Energy Group of Neutrons Causing Fission AOA Area of Applicability B&W The Babcock & Wilcox Company BOL Beginning of Life Ct Cooling Time DR Debris Resistant EALF Energy of Average Lethargy causing Fission ECT Eddy Current Testing En Enrichment ENDF/B Evaluated Nuclear Data File EOL End of Life EPU Extended Power Uprate FA Fuel Assembly FPL Florida Power and Light FRSB Fuel Rod Storage Basket GT Guide Tube GWd Gigawatt days HTC Haut Taux de Combustion HVFD Hafnium Vessel Flux Depression ID Inner Diameter IFBA Integrated Fuel Burnable Absorber IFM Intermediate Flow Mixing ISG Interim Staff Guidance IT Instrumentation Tube Yrff Effective neutron multiplication factor LEU Low Enriched Uranium LV Limiting Value MOX Mixed Oxide Fuel MTU Metric Ton Uranium MWt Mega Watts thermal NFR New Fuel Storage Rack OD Outer Diameter OECD Organization for Economic Co-operation and Development OFA Optimized Fuel Assembly ORNL Oak Ridge National Lab PNL Pacific Northwest Lab ppm parts per million psi pounds per square inch PWR Pressurized Water Reactor RCCA Rod Cluster Control Assemblies SFP Spent Fuel Pool SS Stainless Steel WCAP-17094-NP February 2011 Revision 3

x LIST OF ACRONYMS, INITIALISMS, AND TRADEMARKS (cont.)

STD Standard Fuel Assembly TD Theoretical Density UP Upgrade USL Upper Safety Limit UT Ultrasonic Testing WABA Wet Annular Burnable Absorber Wt% Weight Percent MetamicTM is a trademark of Metamic, LLC.

PlexiglasTM is a trademark of Atoglas WCAP- 17094-NP February 2011 Revision 3

1-1 1 INTRODUCTION The purpose of this report is to document a criticality analysis performed to support the Extended Power Uprate (EPU) at Turkey Point Units 3 and 4. The EPU has two major impacts on the criticality analysis:

(1) The fuel maximum enrichment Technical Specification is increased from 4.5 wt% 235U to 5.0 wt%

235 U, and (2) the depletion of fuel at the EPU conditions results in the fuel being more reactive at the same burnup than fuel depleted under pre-EPU conditions. This is due to the higher fuel and moderator temperatures that result in a harder neutron spectrum, resulting in more plutonium production.

This report documents the criticality safety evaluation for the storage of pressurized water reactor (PWR) nuclear fuel assemblies in the New Fuel Storage Rack (NFR) and the Spent Fuel Pool (SFP). The SFP consists of the permanent Region I and Region II racks and the removable Cask Area Rack. The criticality analysis contained in this report completely supersedes the criticality analysis in support of Reference 5.

This report has been updated relative to Revision 2 of this document in order to address the draft Interim Staff Guidance (ISG) DSS-ISG-2010-1 (Reference 6.) For ease of review, Table 2-1 includes a summary of the issues raised in the draft ISG and lists the sections of this document that address each issue.

WCAP-17094-NP February 2011 Revision 3

2-1 2 OVERVIEW The existing Region I and II racks are evaluated for the placement of fuel with new allowable storage arrays. Consistent with the current licensing basis, this evaluation credits neutron absorber inserts placed into the Region II racks to partially offset an assumed full loss of the original Boraflex neutron absorber.

Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time.

Additionally, credit is taken for the presence of soluble boron in the spent fuel pool and for the presence of full-length Rod Cluster Control Assemblies (RCCAs) placed in selected fuel assemblies. The presence of Integrated Fuel Burnable Absorber (IFBA) rods is also credited for certain fresh fuel evaluations.

The Cask Area Rack is currently licensed for placement of fresh fuel of up to 4.5 wt% 235U. This analysis evaluates the Cask Area Rack for enrichments up to 5.0 wt% 235U. Similarly, the New Fuel Storage Rack 2 35 is also analyzed for 5.0 wt% U in this analysis.

To clearly distinguish between the inserts placed into rack cells and the control components inserted into fuel assemblies, the term "insert" by itself always refers to the MetamicTM neutron absorber inserts placed into the Region I1 racks. The full-length control components are always referred to as RCCAs. Other burnable absorbers placed into assemblies during depletion are always clearly characterized, e.g., as Pyrex inserts, Wet Annular Burnable Absorber (WABA) inserts, or hafnium inserts.

The relevant fuel assembly and fuel rack specifications are identical between Turkey Point Unit 3 and Unit 4. When assembly or history specific data is used, both units are considered in the compilation of that data. Therefore, all analyses and conclusions presented in this report apply to both units.

2.1 ACCEPTANCE CRITERIA 2.1.1 Spent Fuel Pool Criteria The objective of this analysis is to ensure that all calculations of the effective neutron multiplication factor (keff) performed for each permissible storage arrangement, yield results less than 0.95 when the storage racks are fully loaded with fuel of the highest permissible reactivity, and assuming the pool is flooded with borated water at a temperature corresponding to the highest reactivity. Also, the analysis must demonstrate that keff is less than 0.95 under all postulated accident conditions. Finally, the analysis must demonstrate that keff is less than 1.0 with unborated water in the spent fuel pool. The maximum ken-values must be calculated with a 95% probability at a 95% confidence level (Reference 1) and must include: a margin for statistical uncertainty in the reactivity calculations, the effect of manufacturing tolerances, eccentric positioning, and an allowance for uncertainty in the depletion calculations and the assigned burnup.

2.1.2 New Fuel Storage Rack Criteria The New Fuel Storage Rack analyses must demonstrate:

1. The effective neutron multiplication factor (kerr) of the fresh fuel in the fresh fuel storage racks shall not exceed 0.95, at a 95% probability, 95% confidence level, assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with full density unborated water.

WCAP- 17094-NP February 2011 Revision 3

2-2

2. The New Fuel Storage Rack shall not exceed a keff of 0.98, at a 95% probability, 95% confidence level, when under optimum moderation conditions. In the optimum moderation evaluation the rack is assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid.

The maximum kcf must account for all biases and uncertainties.

2.2 DESIGN APPROACH In order to qualify the existing racks for placement of fuel of higher enrichment and operation under EPU conditions, it is necessary to separate fuel assemblies into categories. Each fuel category is described by a set of minimum burnup requirements which are a function of initial enrichment and post irradiation cooling time. The fuel categories are ranked by reactivity within the appropriate region of the SFP in Table 5-1. The relative placement of assemblies from the fuel categories as well as, the number and location of reactivity suppressing devices necessary to meet the acceptance criteria are then determined.

Acceptable combinations of fuel categories and reactivity suppressing devices are termed storage arrays in this document. Storage arrays allowed by this analysis are presented in Figure 5-1 and Figure 5-2 for Regions I and II of the SFP respectively. It is important to note that each 2x2 array is analyzed with fuel of the maximum allowable reactivity for the category. Therefore, fuel of a lower reactivity (i.e., greater burnup) may also be placed in the array. It is not necessary to use all defined arrays in the configuration of the spent fuel pool.

Additionally, this report documents enrichment and burnable absorber requirements for the NFR.

2.3 COMPUTER CODES The analysis methodology employs the following computer codes: (1) the two-dimensional transport lattice code PARAGON Version 1.2.0, as documented in Reference 4, and its cross-section library based on Evaluated Nuclear Data File (ENDF/B) Version VI.3, and (2) SCALE Version 5.1, as documented in Reference 2, with the 44-group cross-section library based on ENDF/B-V.

PARAGON is used for simulation of in-reactor fuel assembly depletion and SCALE is utilized for reactivity determinations of fuel assemblies in the Turkey Point Units 3 and 4 spent fuel pools.

PARAGON is Westinghouse's state-of-the-art two-dimensional lattice transport code. It is part of Westinghouse's core design package and provides lattice cell data for three-dimensional core simulator codes. This data includes macroscopic cross sections, microscopic cross sections for feedback adjustments, pin factors for pin power reconstruction calculations, and discontinuity factors for three-dimensional nodal method solution of the diffusion equation. PARAGON uses the collision probability theory within the interface current method to solve the integral transport equation. Throughout the whole calculation, PARAGON uses the exact heterogeneous geometry of the assembly and the same energy groups as in the cross-section library to compute the multi-group fluxes for each micro-region location of the assembly. In order to generate the multi-group data, PARAGON goes through four steps of calculations: resonance self-shielding, flux solution, burnup calculation and homogenization. The 70-group PARAGON cross-section library is based on the ENDF/B-VI.3 basic nuclear data. It includes explicit multigroup cross-sections and other nuclear data for 174 isotopes, without any lumped fission products or pseudo cross sections. PARAGON and its 70-group cross-section library are benchmarked, WCAP- 17094-NP February 2011 Revision 3

2-3 qualified, and licensed both as a standalone transport code and as a nuclear data source for a core simulator in a complete nuclear design code system for core design, safety and operational calculations.

PARAGON is generically approved for depletion calculations. The use of PARAGON for spent fuel criticality is chosen since it has improvements relative to the historically used PHOENIX-P (e.g. no lumped fission products) and has all the attributes needed for this work. There are no SER limitations for the use of PARAGON in UO, criticality analysis.

The criticality safety criteria are shown to be met by use of the three dimensional Monte Carlo code, KENO-V.a (Reference 2). Any mention of KENO in this report refers to KENO-V.a. KENO is run using the 44 group cross section library based on ENDF/B-V. Prior to the actual KENO analysis, the 44 group cross sections must be processed for the resonance self shielding and for the thermal characteristics of the problem. The cross section processing and the KENO runs are performed using the CSAS25 sequence, which is part of the SCALE 5.1 code package.

The criticality sequence of SCALE 5.1 is validated using

]IC The details of the validation are found in Appendix A. The validation shows that SCALE 5.1 is an accurate tool for calculation of kff for spent fuel pool applications. The benchmark calculations utilize the same computer platform and cross-section libraries as are used for the design basis calculations. [

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

(1) number of particle histories per generation, (2) the number of generations skipped before averaging, (3) the total number of generations and (4) the initial source distribution. The KENO criticality output contains a great deal of useful information to determine the acceptability of the problem convergence.

This information is used to develop appropriate values for the aforementioned parameters applied in storage rack criticality calculations.

"]a.

WCAP- 17094-NP February 2011 Revision 3

2-4 2.4 CORRESPONDENCE TO DSS-ISG-2010-1 Table 2-1 provides a section to section mapping of the DSS-ISG-2010-1 topics addressed by this analysis.

The topics listed in the "DSS-ISG-20 10-1 Section Number" column refer to the sub-sections of the "Technical Guidance" section of DSS-ISG-2010-1.

Section to Section Correspondence of this Document to DSS-ISG- 2010-1 ll Table 2-1 DSS-ISG-2010-1 DSS-ISG-2010-1 DSS-ISG-2010-1 WCAP-17094-P, Rev. 3 Section Number Section Title Topic Section Number ac February 2011 WCAP- 17094-NP February 2011 Revision 3

3-1 3 DESIGN AND INPUT DATA 3.1 FUEL ASSEMBLY SPECIFICATIONS The design specifications of the fuel assemblies, which are used for this analysis, are given in Table 3-1.

Table 3-1 Fuel Assembly Specifications Parameter Value Assembly type 15xI5 STD or OFA Rod Array size 15xl5 Rod pitch, inch 0.563 ]acc[

Active fuel length, inch 144 Stack density, % TD 97.5 Maximum enrichment, wt% 235 U 5.0 235 Enrichment tolerance (for enrichments less than 5.0 wt% U), wt% +/- 0.05 Total number of Fuel Rods 204 Fuel cladding outer diameter, inch 0.4220 ]a.c p[

Fuel cladding inner diameter, inch 0.3734 +/-[ ]ac Fuel cladding thickness, inch 0.0243 +/-[ ]a,c Pellet diameter, inch 0.3659(21)-[ ]a,c IFBA 10 B loading, mg/inch [ ]axc IFBA Thickness, inch [ ]ac IFBA Length, inch 120 Number of Guide/Instrument tubes 20/1 STD OFA Toler.

Guide/Instrument tube 0D, inch 0.546 0.533 +/-[ ]ac Guide/Instrument tube ID, inch 0.512 0.499 +[ ]a,c Guide/Instrument tube thickness, inch 0.017 0.017 +/-[ ]apc Note:

I.The fuel rod pitch uncertainty is calculated by uniformly expanding the pitch of each unit cell in the lattice such that the assembly would be flush with adjacent assemblies in the core. This value is bounding because there is not a history of assemblies becoming lodged in the core.

2.The nominal pellet diameter is 0.3659 inches. However there have been some fuel rods with pellet diameters outside of the manufacturing tolerances shown. Smaller than nominal diameter pellets are conservatively bounded by this analysis since the smaller pellet diameter results in less fuel thus less reactivity. The effect of the rods with the larger pellet diameter is negligible since there are only 4 rods of this diameter.

February 2011 WCAP- 17094-NP February 2011 Revision 3

3-2 3.1.1 IFBA Specification I

I]a3C 3.2 FUEL ASSEMBLY INSERTS SPECIFICATIONS Table 3-2 shows specifications and tolerances for the full length RCCA used in the analysis. In this report RCCA always refers to full length RCCAs. Part length control elements are not credited in this report.

Table 3-3 provides the specifications for the Pyrex and Wet Annular Burnable Absorber (WABA). These burnable absorbers are only used in the depletion analysis. Although WABA's are only 120 inches they are modeled in all axial nodes (full length). This conservatively increases the impact of the spectral hardening effect by increasing the length of fuel which is starved of water and exposed to additional absorber.

Table 3-2 RCCA Specifications Parameter Value Material Silver-Indium-Cadmium Silver content, wt% 80 ]a.c p[

Indium content, wt% 15[ ]a,c Cadmium content, wt% 5+ ]a.c Poison OD, inch 0.3900 +[ ]a.c Poison length in active fuel region, inch 141.75 Clad inner diameter, inch 0.4005 + ]ac Clad outer diameter, inch 0.4390 ]a pc Clad material SS 3

Poison density, gm/cm 10.17 WCAP- 17094-NP February 2011 Revision 3

3-3 Table 3-3 Pyrex and WABA Specifications Parameter Pyrex WABA Burnable absorber (BA) material Borosilicate Glass B4 C BA inner diameter, inch 0.2440 0.2780 BA outer diameter, inch 0.3890 0.3180 BA clad material, inch SS Zr BA inner clad thickness, inch 0.0065 0.0210 BA inner clad OD, inch 0.2360 0.2670 BA outer clad thickness, inch 0.0188 0.0260 BA outer clad OD, inch 0.4310 0.3810 BA length, inch 144 120 Hafnium Vessel Flux Depression (HVFD) absorbers have been used to reduce the fluence at critical weld locations along the core vessel for the purpose of vessel life extension. Hafnium is used as the absorbing material because of its high neutron absorption and slow depletion characteristics. The hafnium insert is a hafnium rod encased in zircaloy cladding. The hafnium region of the inserts is part-length with the remaining rod length being structural material. The hafnium inserts are designed to be placed in the guide tubes of an assembly similar to discrete burnable absorbers. Table 3-4 below shows the dimensions and physical properties of hafnium inserts.

Table 3-4 Hafnium Insert Specifications Parameter Hafnium Insert Absorber material Hf Absorber outer diameter, inch 0.3700 Clad thickness, inch 0.0310 Clad outer diameter, inch 0.4400 Clad material, inch Zircaloy Absorber density, gm/cm 3 13.225 Length, inches 36 Offset of center of poison down from core midplane, 18 inches WCAP- 17094-NP February 2011 Revision 3

3-4 3.3 STORAGE RACK INSERTS SPECIFICATIONS The physical characteristics for the Metamic neutron absorber inserts are summarized in Table 3-5.

Table 3-5 Metamic Insert Specifications Parameter Value Material AI-B 4C

'0B loading, gm/cm 2 0.016 [c Thickness, inch 0.073 [ ]a,c Width, inch 8.35 [ ]apc Length (in active fuel region), inch [ ]a.c Note that the insert length used in the analysis is shorter than the active fuel length, i.e., it is assumed that the lower 6 inches of the active fuel length are not covered by the insert. The Metamic inserts are modeled as having the minimum length and minimum width allowed by the tolerances, and the nominal thickness.

The uncertainty in thickness is considered in the analysis.

3.4 FUEL ROD BASKETS Fuel rod baskets can be stored in any fuel storage cell in the pool without restriction; this includes fuel storage cells in Region I, Region II, and the Cask Area Rack. These baskets consist of regular arrays of stainless steel tubes. Individual fuel rods are placed in these tubes. The specifications of these fuel rod baskets are given in Table 3-6.

Table 3-6 Fuel Rod Basket Specifications Parameter Value Tube array 8x8 - 4x3 (see Section 5.4)

Number of tubes 52 Tube OD, inch 0.625 Tube thickness, inch 0.035 Tube pitch, inch 0.937 Tube material SS 3.5 STORAGE RACK SPECIFICATIONS The storage cell characteristics that are used in the criticality analysis are summarized in Table 3-7 for the Region 1, Region II, Cask Area Rack, and New Fuel Storage Racks. Note that the poison areal density listed for Regions I and II is not used in the analysis since it is assumed that the Boraflex has completely degraded. The Boraflex material is replaced with water. There is no credible mechanism that would allow WCAP- 17094-NP February 2011 Revision 3

3-5 the ' 0 B to escape without the whole material dissolving so replacing the Boraflex with water is appropriate (i.e., if there is any Boraflex material remaining, it would also contain the neutron absorbing 101).

Table 3-7 Fuel Rack Specifications Value Parameter Region I Region II Cask Area New Fuel Rack Cell ID, inch 8.75 8.80 8.75 9.00

[ ]p °c ]a,c [ ]a,c

[ ]ac Wall thickness, inch 0.075 0.075 0.075

[ ]°.C [ ]pc [ ]a,c Cell pitch, inch 10.60 9.00 10.10 21.0 ))as

[ ]a,c [ a'c ]a,c Poison cavity thickness, inch 0.090 0.064 0.083 -

_[ ]a,c [ ]a,c [ ]a,c Poison thickness, inch 0.078 0.051 0.075

[ ]a~c ]a,c [ ]a,c_[_]ac Sheathing thickness, inch 0.02 0.02 0.0235 -

_[ ]ac ]. [ ]pc Sheathing width, inch 7.50 7.50 7.50 -

__[ ]ac ]a,c [ ]ac Poison areal density, 0.020 0.012 [ ]axc -

l0 B gm/cm 2 (modeled as 0) (modeled as 0)

Figure 3-1 shows the rack layout in the pools. The "1 lx12 NEW REGION 1" rack is called the Cask Area Rack in this report. The minimum separation between rack modules is 1.15 inches.

For this analysis, the pool is modeled as a 2x2 array of four assemblies using a periodic boundary condition, thereby creating an infinite array of 2x2 storage cells. No credit is taken for the Boraflex. All the analysis is performed with the Boraflex material replaced with water. Table 3-8 describes the material composition of all structural and absorber materials used.

Table 3-8 Material Compositions Component Density (gm/cm 3) Material wt%

Rack 7.94 SS304 100.0 Sheathing 7.94 SS304 100.0 B4 C 20.595 Metamic 2.65 Al 79.405 Boral BaC 28.009 (in cask area) Al 71.991 WCAP- 17094-NP February 2011 Revision 3

3-6 r: ..........

I

!!1 i

II I I l

i l

II T H1:-

I II I i il I I I II I 1 I~ hII J

I I I

-+ I I I

. I

-- 4:F-4T:F I- I 1

-T MI I

ý .

E

-- PEDEJAL LOCATION SEE N-T 3- (TYP)4LOAIN UNIT 3 LIFT HOLE LOCATIONS SEE DETAIL ABOVE SPENT FUEL PIT LAYOUT Figure 3-1 Turkey Point Spent Fuel Pool Layout (Unit 4 is the Same Except a Mirror Image)

WCAP- 17094-NP February 2011 Revision 3

3-7 Figure 3-2 and Figure 3-3 show the top and side views of the New Fuel Storage Rack. As can be seen from the top view, the rack is L-shaped with a 10 by 3 array against an 8 by 3 array making for storage of 54 assemblies. The side view is shortened. The side view shows two column-like items, which are funnel type devices used to assure the fuel assemblies are correctly placed in the rack. Between the bottom funnel and top funnel, the rack is open. The funnels are 2 feet 9 inches long. The cell pitch in the rack is 21 +/-[ ]a.C inches and the cell has an ID of 9 inches.

1C Figure 3-2 Top View of the Turkey Point New Fuel Storage Racks WCAP- 17094-NP February 2011 Revision 3

3-8

-5E0f/GQA! (M (DhE)

-305,//.O-I'-o Figure 3-3 Side View of the Turkey Point New Fuel Storage Racks WCAP-17094-NP February 2011 Revision 3

4-1 4 ANALYTICAL METHODOLOGY This section describes the methodology used to determine acceptable storage criteria for the Region I and Region II racks as well as the Cask Area Rack. The results of these calculations are discussed in Section

5. Unless otherwise stated, all calculations assumed nominal characteristics for the fuel and the fuel storage cells. The effect of the manufacturing tolerances is accounted for by combining the reactivity effects associated with manufacturing tolerances (rack, fuel, etc) with other uncertainties as discussed below.

As discussed in Section 2.3, KENO is the criticality code used in the Turkey Point criticality calculations.

KENO is used to determine the absolute reactivity of burned and fresh fuel assemblies, the reactivity sensitivities due to manufacturing tolerances and to perform calculations needed for eccentric fuel positioning, and fuel misloading.

All calculations are performed using an explicit model of the fuel and storage cell geometry. KENO three-dimensional calculations model a 2-by-2 array of cells surrounded by periodic boundary conditions.

The three-dimensional KENO models assume 60 cm of water above and below the active fuel length.

Additional KENO models with more than four cells and different boundary conditions are generated to investigate the effect of eccentric fuel assembly positioning, interfaces between racks, and to analyze accident conditions. These models are discussed in the appropriate sections below.

4.1 BOUNDING FUEL ASSEMBLY SELECTION

[

WCAP- 17094-NP February 2011 Revision 3

4-2

• [

]a,c Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (0 ppm) 2 1Table 4-1 Sa,c WCAP-17094-NP February 2011 Revision 3

4-3 Table 4-1 Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (0 ppm)

(cont.)

a,c

[

Ia~c Table 4-2 Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (500 ppm) I 'C WCAP- 17094-NP February 2011 Revision 3

4-4 1

Table 4-2 Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (500 ppm)

(cont.) la,c

[

Ia~c

-1 Table 4-3 Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (1600 ppm) la,c 7

WCAP- 17094-NP February 2011 Revision 3

4-5 Table 4-3 Comparison of Fuel with Grids Modeled to STD Fuel without Grids Modeled (1600 ppm)

(cont.) ac

]a~c Table 4-4 Reactivity Comparison of STD Fuel with and without Grids at 391F IXC

[

axc WCAP- 17094-NP February 2011 Revision 3

4-6

]a,c Table 4-5 Reactivity Comparison of STD Fuel with and without Grids tsa in the Presence of Metamic I] ac Insser Table 4-6 Reactivity Comparison of STD Fuel with and without Grids in the Presence of One Empty Cell a,c ac 4.2 DEPLETION CALCULATIONS The methodology for depleting fuel assemblies in-reactor to support burnup credit in spent fuel pool criticality safety calculations includes the depletion of two-dimensional (2D) unit assemblies as an infinite array in reactor core geometry with PARAGON (described in Section 2.3) at the bounding reactor core conditions specified in subsection 4.2.1. Once the fuel assembly is depleted to a desired assembly-average burnup, it is allowed to decay to its most reactive state (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />). The assembly-averaged isotopic concentrations from PARAGON are then brought to cold conditions, 687F and 14.7 psi. Credit for subsequent decay of241pu to 2 41Am is taken.

]a,c For the non-blanketed (pre-EPU) assemblies, isotopics are generated at each 2.0 GWd/MTU intervals from 0-72 GWd/MTU for the following initial enrichments (wt% 235U) and cooling times:

1.8, 2.5, 3.0, 3.5, and 4.0 15 yr, 20 yr, 25 yr WCAP- 17094-NP February 2011 Revision 3

4-7 For the blanketed (EPU) assemblies, isotopics are generated at each 2.0 GWd/MTU intervals from 0-72 GWd/MTU for the following initial enrichments (wt% 235U) and cooling times:

2.0, 3.4, 4.0, 4.5, and 5.0 0 yr, 2.5 yr, 5 yr, 10 yr, 15 yr, 20 yr, 25 yr 4.2.1 Reactor Parameters For the purposes of this analysis, there are two criteria that are used to define the depletion conditions used for each set of isotopic concentrations. The two criteria are the region of the pool that will be used to store the assembly and whether or not the fuel contains reduced enrichment axial blankets. Three sets of isotopic concentrations are used in this analysis, a single set is used for non-blanketed fuel, and two sets are used for blanketed fuel. Of the two sets of blanketed fuel isotopic compositions, one set is appropriate for analysis of Region I of the spent fuel pool and another set is appropriate for Region II.

Reference 14 provides guidance on the incorporation of the effects of operating parameters when generating depleted fuel isotopic concentrations. This guidance is utilized for each of the parameters in the following subsections.

4.2.1.1 Moderator Temperature In a PWR, as the moderator temperature increases, the moderator density decreases. Decreased density results in reduced moderation, which results in hardening of the in-core neutron spectrum and increased buildup of the plutonium isotopes.

]ac 4.2.1.1.1 Relative Power Determination II jac WCAP- 17094-NP February 2011 Revision 3

4-8 Blanketed Fuel I

]aC ac Figure 4-1 Average Relative Power as a Function of Assembly Burnup for Region I Analysis I

Ia,c WCAP- 17094-NP February 2011 Revision 3

4-9 I

]a,c a,c Figure 4-2 Average Relative Power as a Function of Assembly Burnup for Region 1I Analysis

]aC Several third cycle assemblies are checked and found to be non-limiting when compared to the limiting two cycle average values. Therefore, the third cycle data is not included in Figure 4-2.

FPL will ensure that all assemblies in the future are bounded by the PA ... mbly values of 1.41 for storage in Region I and 1.28 for storage in Region II.

WCAP- 17094-NP February 2011 Revision 3

4-10 Non-blanketed Fuel I

]a,c

",C Figure 4-3 Averaged Assembly Relative Power as a Function of Burnup 4.2.1.1.2 Exceptions to Non-blanketed PAssembly jac WCAP- 17094-NP February 2011 Revision 3

4-11 I

] as Table 4-7 Calculation of krrfor Various 11-3 Enrichment/Burnup Pairings in Array II-B I a.c a Ic The maximum calculated keff+ 1.6455 across all of the enrichment bumup pairings is 0.93579. The total sum of biases and uncertainties is found in Table 5-3 and is 0.03415. Adding the sum of biases and uncertainties to the raw keff yields 0.96994, which has 0.03006 Akeff margin to the regulatory limit, which accounts for the 0.0028 Akeff penalty.

4.2.1.1.3 Determination of the Bounding Assembly Exit Temperature II

]ac WCAP- 17094-NP February 2011 Revision 3

4-12 ac WCAP-17094-NP February 2011 Revision 3

4-13 Table 4-8 Exit Moderator Temperature as a Function Of PAssembly Ija,c 4.2.1.2 Fuel Temperature Higher fuel temperatures cause Doppler broadening of resonances and lead to increased resonance absorption in 238U nuclei, which results in increased plutonium production, and more reactive fuel at a given burnup than fuel depleted at lower fuel temperatures.

]ac 4.2.1.3 Soluble Boron Concentration Boron is a strong thermal absorber and higher soluble boron concentrations lead to a hardening of the neutron spectrum which causes an increased buildup of plutonium. Bounding soluble boron WCAP- 17094-NP February 2011 Revision 3

4-14 concentrations for depletion are determined based on a review of fuel management studies, both from historical and projected EPU operation.

]ac 4.2.1.4 Specific Power and Operating History The analysis assumes full power operation consistent with the average assembly peaking, PAssembly. This high specific power is non-conservative. The ISG (Reference 6) states:

"It may be physically impossible for the fuel assembly to simultaneously experience two bounding values (i.e., the moderator temperature associated with the "hot channel" fuel assembly and the minimum specific power). In those cases, the application should maximize the dominate parameter and use the nominal value for the subordinate parameter."

As anticipated by the ISG and consistent with Reference 14 sensitivities, the moderator temperature impact on reactivity is greater making the selection of the high specific power appropriate.

The depletion is modeled a single continuous depletion consistent with the recommendations of in Reference 14.

WCAP-17094-NP February 2011 Revision 3

4-15 Table 4-9 Cycle Average Soluble Boron Concentrations for Pre-EPU Cycles Unit 3 Unit 4 Cycle Concentration (ppm) Cycle Concentration (ppm) 1 505 1 505 2 532 2 445 3 422 3 345 4 413 4 451 5 .432 5 206 6 426 6 540 7 453 7 401 8 724 8 401 9 584 9 403 10 691 10 684 11 709 11 706 12 571 12 611 13 593 13 578 14 611 14 672 15 715 15 543 16 668 16 708 17 662 17 692 18 753 18 729 19 711 19 738 20 808 20 768 21 692 21 746 22 793 22 749 23 759 23 709 24 745 24 773 I 25 686 WCAP- 17094-NP February 2011 Revision 3

4-16 Table 4-10 Cycle Average Soluble Boron Concentrations for Projected EPU Cycles Unit 4 Cycle Concentration (ppm) 25"1 866 25 (72 feed) 823 26 838 26 (72 feed) 717 27 892 27 (72 feed) 833 28 883 28 880 28 891 Note:

1. The Cycle 25 value listed corresponds to a projected EPU fuel management study.

4.2.2 Burnable Absorber and Insert Usage 4.2.2.1 Non-Blanketed Fuel For pre-EPU non-blanketed fuel, two different burnable poison insert designs (Pyrex and WABA) were used. These early fuel cycle designs predate the introduction of lFBA rods. While a 15x 15 assembly can accommodate a 20 finger burnable absorber insert, at Turkey Point the maximum number of Pyrex fingers used was 12. To determine the bounding burnable absorber insert for non-blanketed fuel, three sets of PARAGON fuel lattice calculations are performed: [

]F.C The maximum burnup that a non-blanketed fuel assembly accrued in the presence of an insert is 23.2 GWd/MTU, this value will be used for all non-blanketed depletions.

4.2.2.2 Blanketed Fuel For blanketed fuel, [

]ac WCAP- 17094-NP February 2011 Revision 3

4-17 I

Ia~c 4.2.2.3 Hafnium Vessel Flux Depression (HVFD) Absorbers In Turkey Point Units 3 and 4 operations, HVFD absorbers have been inserted in a few highly burned fuel assemblies on the core periphery during the third cycle of operation. These part-length absorber inserts are present only near the mid-plane of the fuel assembly's axial length to reduce the fluence at critical weld locations along the core vessel for the purpose of vessel life extension. HVFDs have not been used since Cycles 23 and 24 at Units 3 and 4, respectively, and they will not be used in EPU fuel. The accumulated burnup under the HVFD is always less than 4 GWdIMTU.

For all assemblies which contained hafnium inserts, the distributed burnup profiles are limiting. The significantly under burned top nodes of the fuel assemblies are most important to the determination of the keff of the problem. The HVFDs push the neutron flux out of the middle of the core and toward the ends of the fuel during fuel depletion. The elevated flux in the top portion of the core causes the nodes that are most important to reactivity to be more depleted than they would have been without HVFDs. The over depletion of the top nodes causes the reactivity of a fuel assembly that contained an HVFD in its third cycle to be bounded by the same fuel assembly that is depleted without an HVFD which also assumed a bounding distributed burnup profile. Note that selection of limiting axial burnup profiles for pre-EPU fuel described in subsection 4.3.1.6.2 included shapes from assemblies with HVFDs.

4.2.3 Rodded Operation Turkey Point does not use control rods as part of their full power operations. Therefore, there is no significant burnup accrued during depletion with RCCAs inserted in the active fuel height, and no need to account for these effects in this analysis.

WCAP- 17094-NP February 2011 Revision 3

4-18 4.2.4 Summary of Depletion Parameters Table 4-11 and Table 4-12 summarize the depletion parameters that are used in the analysis.

Table 4-11 Parameters Used in Depletion Analysis Parameter Non-Blanketed Fuel Value Blanketed Fuel Value Soluble Boron Concentration, ppm [ ]a.c [ ]a.c Core Average Power Level, MWt 2300 2644 Assembly Average Relative Power [ []a,c As described in As described in subsection 4.2.1. 1. 1 subsection 4.2.1.1.1 Moderator Temperature, 'F [ ]ac ]a,c (Bounding Assembly Exit Temperature) As described in As described in subsection 4.2.1.1 subsection 4.2.1.1 Fuel Temperature A function of bumup using the A function of burnup using the Assembly Average Relative Assembly Average Relative Powers above Powers above System Pressure, psi 2250 2250 In-Core Assembly Pitch, Inches 8.465 8.465 Burnable Absorber See Table 4-12 Burnup at which the Absorber is removed See Table 4-12 [ ]ac (GWd/MTU)

Table 4-12 Burnable Absorber Modeling For Non-Blanketed Fuel Depletion Enrichment Burnable Absorber Type Removed 1.8 [ ]ac [ ]a 'c 2.5 -4.0 [ ]ac [ ]pc WCAP- 17094-NP February 2011 Revision 3

4-19 4.3 CRITICALITY CALCULATIONS 4.3.1 KENO Model Using the dimensions and materials described in Section 3, a 2x2 array of the spent fuel rack is modeled in KENO. An illustration of the model for the Region II two insert case is shown in Figure 4-4.

Figure 4-4 KENO Model of a Two Insert Case Modeling Simplifications

1. Calculations are performed assuming an infinite radial array of fuel assemblies or assembly patterns. Specifically, all gaps between adjacent Region 11 rack modules are conservatively ignored, i.e., cells in neighboring Region 11 rack modules are assumed to be separated by a single cell wall only. The actual configuration in the Turkey Point spent fuel pool has a cell wall on each side of the Region II rack-to-rack gap. Region I rack to rack separation was also ignored. Special models were needed for Region I - Region II interfaces, accident analysis, the New Fuel Rack, and eccentricity. These models are described further in the sections.

2. In the KENO models, 60 cm of unborated water is used above and below the active region of the fuel, even when soluble boron is credited in the active fuel region. This is conservative because the end fittings absorb neutrons and if they are to be modeled explicitly, the reactivity would decrease.

3. All of the final two insert calculations are run with the two inserts in the same row (a "parallel" array) because this gives a higher keff than the checkerboard arrangement.

4.3.1.1 Absorber Efficiency of the Metamic Inserts a,c WCAP- 17094-NP February 2011 Revision 3

4-20 I

Ia2c 4.3.1.2 Modeling of RCCAs

[

axc 4.3.1.3 Pool Temperature Bias The normal operating temperature range of the Turkey Point spent fuel pools is 39'F to 150'F. All calculations are performed with a nominal fuel and water temperature of 68'F with a water density of 1.0 gm/cm3 . In order to account for the reactivity effects of operating the pool anywhere in the normal temperature range, calculations are performed with temperatures of 39°F, 50'F, 100'F, and 150'F. For each case the water density is modeled appropriately for the temperature at which the calculation is performed. The difference between the maximum keff of those four cases and the reference case is taken as the temperature bias (including Monte Carlo uncertainties).

Temperatures beyond this nominal range are covered in Section 5.7.2 as accident conditions.

4.3.1.4 Modeling of Burnable Absorber Inserts in the Spent Fuel Pool ac 4.3.1.5 Modeling of Axial Blankets I

a,c WCAP- 17094-NP February 2011 Revision 3

4-21

]a'C 4.3.1.6 Modeling of Axial Burnup Distributions The fuel in this report is either blanketed or non-blanketed. There are no plans to use non-blanketed fuel in the future so the analysis of non-blanketed fuel is specific to the non-blanketed fuel currently in the spent fuel pools. Blanketed fuel is modeled under the most reactive conditions expected. Since depletion under EPU conditions is more limiting than depletion under pre-EPU conditions, the reactivity of blanketed fuel used prior to the EPU is bounded by the assumption that all blanketed fuel is depleted at EPU conditions. The axial burnup profile for the blanketed fuel covering current and future fuel is generated differently than the axial profile for the existing full length fuel whose burnup is already completed.

For each assembly in the model, an axial burnup distribution is needed.

]a,c The next two subsections describe the determination of the bounding axial burnup distributions for both blanketed and non-blanketed fuel.

4.3.1.6.1 Axial Burnup Distribution for the Blanketed Fuel

]a,c WCAP- 17094-NP February 2011 Revision 3

4-22

]ac WCAP- 17094-NP February 2011 Revision 3

4-23

[

]a,c I

I Jl Table 4-13 ia,c

ýý

-91c 2 5 Table 4-14 shows the limiting axial burnup profiles for 8 inch 2.6 wt% 1 U blanketed fuel.

WCAP- 17094-NP February 2011 Revision 3

4-24 a,c WCAP- 17094-NP February 2011 Revision 3

4-25 4.3.1.6.2 Axial Burnup Distribution for the Non-Blanketed Fuel For the pre-EPU non-blanketed fuel, burnup profile data from the Turkey Point reactor records are used to find the most limiting profiles (the pre-EPU non-blanketed loading curves are only used for the existing fuel). The non-blanketed burnup profiles consist of twelve 12 inch nodes.

[

]a'C I

Table 4-15 Assemblies with the Most Limiting Axial Burnup Profiles 1 a,c

[I

]ac

1. [ ]a,c WCAP- 17094-NP February 2011 Revision 3

4-26 a,c Figure 4-5 Comparison of the Twelve Node and Twenty-Six Node Axial Approximations to the

> 46.2 GWd/MTU Burnup Shape WCAP-17094-NP February 2011 Revision 3

4-27 The limiting axial profiles for non-blanketed fuel are shown in Table 4-16.

_ Table 4-16 Axial Burnup Profiles for Non-Blanketed Fuel ] a,c WCAP- 17094-NP February 2011 Revision 3

4-28

]a.c 4.3.1.7 Reactivity of Geometry Changes Due to Irradiation

]ac WCAP- 17094-NP February 2011 Revision 3

4-29 9 6 .5 . ..

.. ~ ~I. ~ ... .. ..+.. .....

... c. ...

-I-t-*- -l--- t Ft o 95.0 ,

• 5 - ....... ....

. .... 1 .....

...i- ...... .... ....

...... 4-. 44..

.. .. -F... ..--4. . ... ......

94.5 9 93.0 ...

. 5 10. 1. 2. 2. 3. 3. 4. 4. 5. 5 60 .6 9A.5 Ro BurnupF(GW...TU) 93 .0 U3 If: ,-i,-tfi

.... ~~ ~ ,.~ ~ .... .... *,-,.*

..... ,.,'t*

92.5 7-i -"-'t~

0~~~~~~~~-1 5 0 1 0 25 3 5 475 0 5 0 6 RoT rg uru GDMU Figure 4-6 Pellet Density as a Function of Burnup 0.0246 0.0244 0.0242 W 0.0240 0.0238 4 0.0236 0.0232 0 5 10 15 20 25 30 35 40 45 50 55 60 65 Rod Average Burnup (GWD/MVTU)

Figure 4-7 Clad Thickness as a Function of Burnup WCAP- 17094-NP February 2011 Revision 3

4-30 4.3.2 Uncertainties The following sub-sections contain descriptions of the various sources of uncertainty in the calculated kff of the storage arrays.

4.3.2.1 Manufacturing Tolerances In calculating the final value of keff, the reactivity effect of manufacturing tolerances must be included.

KENO is used to quantify these effects. A bounding fuel density of 97.5% of theoretical density is used in both the depletion calculations and the KENO analysis so no tolerance calculation for density is needed.

Also, the minimum dimensions are used for the Metamic insert length and width so separate tolerance effects for these parameters are not needed. Therefore, the fuel, rack, and insert tolerances that are considered in the analysis are:

1. [

]a.c The reactivity effects of tolerances are established separately for each storage array. The reference condition for each storage array consists of nominal dimensions and properties. Worst case dimensions are used for the Cask Area Rack so separate tolerance calculations for this rack are not performed. For the New Fuel Storage Rack, most of the analysis is performed with the worst case dimensions. For the 4.5 wt% 235U fully flooded case, the worst case dimensions and positioning are separated from the enrichment uncertainty. These uncertainties are then statistically combined with the validation uncertainty.

To determine the Ak associated with a specific manufacturing tolerance, the keff calculated for the reference condition is compared to the kIy from a calculation with an individual tolerance included. The Ak due to a tolerance is then calculated as follows in equation 4.8:

Ak = k, - kR + 1.645ý- + OR 4.8 where:

WCAP- 17094-NP February 2011 Revision 3

4-31 k, =keff with the tolerance included, kR = keff for the reference case, GI = Monte Carlo standard deviation for the case with the tolerance included, GR = Monte Carlo standard deviation of the reference case, and 1.645 = One-sided 95/95 confidence interval factor.

]a.C The manufacturing uncertainty on fuel enrichment is +0.05 wt% 23 5 U. The uncertainty due to enrichment can be obtained from the keff of fuel of different enrichments as follows:

Enrichment uncertainty- = (kftesh + 0.05 w/o - kfresh) + 1.645 1 + R 4.9 Where kfresh + 0.05 w/o is the eigenvalue of the perturbed enrichment and kfresh is the eigenvalue of fresh assembly of the enrichment being considered.

All of the Ak values from the various tolerance effect calculations are statistically combined (square root of the sum of the squares) with the depletion, burnup and validation uncertainties to determine the total uncertainty. All biases are then directly added to the total uncertainty to develop the final reactivity penalty for the array.

2. For blanketed fuel assemblies, this enrichment is the enrichment of the central zone. Analyses are performed to separately account for enrichment uncertainty in the blanketed regions and the effects are found to be insignificant.

WCAP-17094-NP February 2011 Revision 3

4-32 4.3.2.2 Burnup Uncertainty

[

Iac 4.3.2.3 Depletion Uncertainty The depletion uncertainty is taken to be 5% of the delta-k difference between the reactivity at the fresh fuel condition to the reactivity at the burned fuel condition of interest:

Depletion uncertainty = (kfresh - kb.m)

• 0.05 + 1.645 G- OR 4.11 Where kfresh is the eigenvalue of the fresh enrichment with no burnable absorbers and kbumr is the eigenvalue of the burned assembly that has been depleted with the parameters described in subsection 4.2.1.

P.C KENO calculations for each fuel category are performed for each enrichment using unburned fresh fuel to obtain the fresh fuel kef,.

4.3.2.4 Fission Product Uncertainty A common approach to validation of neutron cross sections is through benchmark critical experiments that are designed to closely represent the configurations of the desired criticality application. The validation of fission products, however, is more difficult because few critical experiments are available for the strongest fission products and none are available at all for many of the 'weaker' fission products.

Due to the limited availability of fission product cross section data, a factor of uncertainty is considered in the criticality safety analysis.

II Ia~c WCAP- 17094-NP February 2011 Revision 3

4-33 II Ia~c 4.3.2.5 Eccentric Fuel Assembly Positioning The fuel assemblies are assumed to be nominally located in the center of the storage rack cell; however, it is recognized that an assembly could in fact be located eccentrically within its storage cell.

[

]". A graphical example of the eccentric positioning calculations is presented in Figure 4-8.

a,c Figure 4-8 Example of a Region I Eccentric Positioning Model 4.3.2.6 Other Uncertainties Reference 15 examines the effects of various operating histories (thermal power versus burnup histograms). Reference 14 recommends that a continuous power history be used and that margin (up to 0.002 Akeff) should be included to account for potential variations. FPL rarely operates the Turkey Point units at less than full power, so the 0.002 Akeff margin will be included as an uncertainty.

WCAP- 17094-NP February 2011 Revision 3

4-34 An uncertainty in the predictive capability of SCALE 5.1 and the associated cross section library is considered in the analysis. The uncertainty from the validation of the calculational methodology is discussed in detail in Appendix A.

4.3.3 Calculation of Burnup versus Enrichment Curves All calculations to establish and validate the bumup versus enrichment curves are performed as full three-dimensional criticality calculations considering both an axial burnup distribution and a uniform burnup profile for each assembly in the model. The loading curves and corresponding fitting coefficients are shown in Section 5.

Calculations are run to establish the reactivity of each combination of burnup, enrichment, and cooling time shown in Table 5-6 through Table 5-12 for non-blanketed fuel and Table 5-13 through Table 5-19 for blanketed fuel for all storage arrays. Maximum keff values are calculated using the limiting sets of biases and uncertainties for each storage array presented in Table 5-2 and Table 5-3 by the equation shown below.

keff = k(calc) + Ak(bias) + Ak(uncert) 4.12 where:

k(calc) krff calculated by the KENO model Ak(bias) = sum of biases determined from critical benchmark comparisons (see Appendix A), temperature, and Metamic insert orientation Ak(uncert) = statistical summation of all tolerance and uncertainty components (square root of the sum of the squares)

The calculated kerr values after applying all biases and uncertainties are shown in Table 5-23 through Table 5-42 for all allowable 2x2 arrays. Note that all kerr values are less than 0.99, which is 0.01 Akeff lower than the regulatory limit.

There are several implicit conservatisms that are included in the development of the burnup versus enrichment curves:

I. Each assembly is depleted asc

2. Conservative axial burnup distributions (including the uniform distribution) are used. In reality, only a few assemblies would actually have these burnup distributions. Most assemblies have a nominal axial burnup distribution which is less reactive.

WCAP- 17094-NP February 2011 Revision 3

4-35

3. The loading curves provide the minimum burnup required. In reality, the loading of the pool in each 2x2 array will contain assemblies that exceed these burnup requirements resulting in lower reactivity.

4. The uncertainties on enrichment, depletion, burnup, and fission product worth depend greatly on the initial enrichment of the fuel being considered. This analysis uses the maximum of each of these values when considering all possible enrichments. Because the enrichment uncertainty is most penalizing at low enrichments and the burnup related uncertainties are highest at high enrichments, all assemblies are treated with larger uncertainties than are physical.

The embedded conservatisms will ensure that the actual reactivity of the stored fuel array, under the assumed condition of the loss of all soluble boron in the pool, will always be significantly below 1.0. All burnup versus enrichment curves are therefore acceptable and result in reactivity values below the regulatory limit.

4.3.4 Soluble Boron Credit

].C For all storage arrays, the keff must be less than 0.95 after applying all biases and uncertainties. Boron dilution analysis has previously shown that 500 ppm of soluble boron is acceptable. Calculations are performed to confirm that a soluble boron concentration of 500 ppm in the spent fuel pool ensures that kff does not exceed 0.95 under normal conditions for all fuel storage arrays.

4.3.5 Interfaces The following subsections discuss the methods for the various interface requirements.

4.3.5.1 Within Region Interfaces

]a,c WCAP- 17094-NP February 2011 Revision 3

4-36 WCAP- 17094-NP February 2011 Revision 3

4-37 X h-i II-1 X Il-i 11-1 hI-i Figure 4-9 Example of Interfaces between Region 11 Arrays (Shaded cells contain an insert and X indicates an empty (water-filled) cell.)

Following these rules, every 2x2 array matches an analyzed condition, and therefore no interface-specific analyses are required. Gaps between the same region rack modules are conservatively neglected, i.e., cells located across a rack-to-rack gap are considered the same as cells directly facing each other within a rack.

The arrays wherein Region 1I cells face Region I rack modules require additional analyses and are discussed in subsection 5.5.

No special considerations need be given to cells facing the pool wall or other racks.

4.3.5.2 Region I to Region 1I Interface Ia'C February 2011 WCAP- 17094-NP February 2011 Revision 3

4-38 a'c All allowable Region I- Region II interface configurations must meet a kIr value of less than 0.99 (0.01 Akeff below the regulatory limit), including all biases and uncertainties. The maximum uncertainties and biases calculated for each of the two storage arrays are utilized in the interface calculations. [

The various analyzed interface conditions are described and illustrated in Section 5.5.1. The bias and uncertainty results for these interface configurations are presented in Table 5-45.

WCA P- 17094-NP February 2011 Revision 3

4-39 4.3.5.3 Cask Area Rack Interface The Cask Area Rack has sufficient absorber panels that the maximum rffis much less than the limiting keff in Region I or Region II. Furthermore, the Cask Area Rack has Boral panels on the exterior of the rack so there is no local increase in reactivity at the rack interface. Consequently, there are no interface loading constraints on the Cask Area Rack/Region I or Cask Area Rack/Region II interface.

February 2011 17094-NPP WCAP- 17094-N February 2011 Revision 3

5-1 5 RESULTS 5.1 SPENT FUEL POOL REGIONS I AND II 5.1.1 Fuel Category and Storage Array Definitions Each fuel category is described by a set of minimum burnup requirements which are a function of initial enrichment and post irradiation cooling time. The fuel categories are ranked by reactivity within the appropriate region of the SFP in Table 5-1. The relative placement of assemblies from the fuel categories as well as, the number and location of reactivity suppressing devices necessary to meet the acceptance criteria are shown in Figure 5-1 and Figure 5-2 for Regions I and II of the SFP respectively. It is important to note that each 2x2 array is analyzed with fuel of the maximum allowable reactivity for the category. Therefore, fuel of a lower reactivity (i.e., greater burnup) may also be placed in the array. It is not necessary to use all defined arrays in the configuration of the spent fuel pool.

Table 5-1 Fuel Categories Ranked by Reactivity 1-1 High Reactivity 1-2 Region I 1-3 1-4 Low Reactivity I-1 High Reactivity 11-2 Region I 11-3 11-4 11-5 Low Reactivity Notes:

I .Fuel Category is ranked by decreasing order of reactivity without regard for any reactivity-reducing mechanisms, e.g., Category 1-2 is less reactive than Category I-M,etc. The more reactive fuel categories require compensatory measures to be placed in Regions I and II of the SFP, e.g., use of water filled cells, Metamic inserts, or full length RCCAs.

2. Any higher numbered fuel category can be used in place of a lower numbered fuel category from the same Region.

3.Category I-1 is fresh unburned fuel up to 5.0 wt% 235U enrichment.

4.Category 1-2 is fresh unburned fuel that obeys the IFBA requirements of Table 5-20 or contains an equivalent amount of another burnable absorber.

5.All Categories except I-1 and 1-2 are determined from Table 5-21 and Table 5-22.

WCAP- 17094-NP February 2011 Revision 3

5-2 Array I-A I- X Checkerboard pattern of Category I-1 assemblies and empty (water-filled) cells. X I-1 Array I-B 1-4 1-4 Category 1-4 assembly in every cell. 1-4 1-4 Array I-C Combination of Category 1-2 and 1-4 assemblies. Each Category 1-2 assembly shall contain a full length RCCA. 1-4 1-4 Array I-D 1-3 1-3 Category 1-3 assembly in every cell. One of every four assemblies contains a full length RCCA. 1-3 Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Category I-I is fresh unburned fuel up to 5.0 wt% 235U enrichment.

Category 1-2 is fresh unburned fuel that obeys the IFBA requirements in Table 5-20 or contains an equivalent amount of another burnable absorber.

4. Categories 1-3 and 1-4 are determined from Table 5-21 and Table 5-22.

5. Shaded cells indicate that the fuel assembly contains a full length RCCA.

6. X indicates an empty (water-filled) cell.

7. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

8. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

Figure 5-1 Allowable Region I Storage Arrays WCAP- 17094-NP February 2011 Revision 3

5-3 Array HI-A Category 11-1 assembly in three of every four cells; one of every four cells is empty (water filled); the cell diagonal from the empty cell contains a Metamic insert or full length RCCA.

Array 1I-B f-3 11-5 11j 11-5 Checkerboard pattern of Category 11-3 and 11-5 assemblies with two of 11-5 113 ý- - 11-3 every four cells containing a Metamic insert or full length RCCA.

Array II-C 114 11-4 Category 11-4 assembly in every cell with two of every four cells containing a Metamic insert or full length RCCA.

Array 1I-D 112 11 .

Category 11-2 assembly in every cell with three of every four cells containing a Metamic insert or full length RCCA.

Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Fuel categories are determined from Table 5-21 and Table 5-22.

3. Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.

4. X indicates an empty (water-filled) cell.

5. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

6. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

Figure 5-2 Allowable Region II Storage Arrays WCAP- 17094-NP February 2011 Revision 3

5-4 5.1.2 Bias and Uncertainty Calculations Region I and Region II biases and uncertainties calculated in unborated conditions are tabulated for each fuel category in Table 5-2 and Table 5-3. Borated biases and uncertainties are tabulated in Table 5-4 and Table 5-5. The total bias and uncertainty values given on the last row of each table are used for detennining each fuel category's minimum burnup and soluble boron requirements.

11Table 5-2 Unborated Biases and Uncertainties for Region I Fuel Categories a,c 7

Note:

I. RCCAs are conservatively modeled using the worst dimensions; therefore no tolerance uncertainties are calculated for this configuration.

WCAP- 17094-NP February 2011 Revision 3

5-5 I

11Table 5-3 Unborated Biases and Uncertainties for Region 11 Fuel Categories l1 a,c Note:

I.The Metamic width uncertainty conserved the atom density in the Metamic rather than the boron areal density. This over prediction of the Metamic width uncertainty is conservative.

2.Metamic orientation bias covers all potential permutations of the insert placement within the specified storage cell. This is applicable only to I-A because it is the only asymmetric storage configuration.

WCAP- 17094-NP February 2011 Revision 3

5-6 Table 5-4 Borated Biases and Uncertainties for Region I Fuel Categories I Fuel Category Ic Note:

I . RCCAs are conservatively modeled using the worst dimensions; therefore no tolerance uncertainties are calculated for this configuration.

WCAP- 17094-NP February 2011 Revision 3

5-7 Table 5-5 Borated Biases and Uncertainties for Region II Fuel Categories

-11I Fuel Category a.4,c Note:

lI.The Metamic width uncertainty conserved the atom density in the Metamic rather than the boron areal density. This over prediction of the Metamic width uncertainty is conservative.

2.Metamic orientation bias covers all potential permutations of the insert placement within the specified storage cell. This is applicable only to lI-A because it is the only asymmetric storage configuration.

WCAP- 17094-NP February 2011 Revision 3

5-8 5.1.3 Minimum Burnup and IFBA Requirements The minimum burnup requirements are tabulated in Table 5-6 through Table 5-12 for non-blanketed fuel and Table 5-13 through Table 5-19 for blanketed fuel. The minimum IFBA requirements for category 1-2 fuel are presented in Table 5-20.

5.1.3.1 Pre-EPU Non-Blanketed Fuel Table 5-6 Burnup Requirements for Pre-EPU Non-Blanketed Category 1-3 Fuel 235 Initial Enrichment (wt% U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 y 0.00 6.51 12.07 17.85 22.97 20 y 0.00 6.42 11.85 17.47 22.44 25y 0.00 6.33 11.66 17.16 22.06 Table 5-7 Burnup Requirements for Pre-EPU Non-Blanketed Category 1-4 Fuel 23 Initial Enrichment (wt% "U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 y 0.24 10.71 17.17 22.94 28.42 20 y 0.23 10.50 16.76 22.37 27.73 25y 0.23 10.31 16.41 21.91 27.18 Table 5-8 Burnup Requirements for Pre-EPU Non-Blanketed Category II-I Fuel 23 5 Initial Enrichment (wt% U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 y 0.00 10.64 16.71 22.19 27.48 20 y 0.00 10.39 16.30 21.62 26.78 25y 0.00 10.21 16.05 21.29 26.23 WCAP- 17094-NP February 2011 Revision 3

5-9 Table 5-9 Burnup Requirements for Pre-EPU Non-Blanketed Category 11-2 Fuel 235 Initial Enrichment (wt% U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 y 6.24 19.05 24.45 30.33 37.00 20 y 6.14 18.57 24.18 29.55 35.96 25 y 6.04 18.17 23.94 28.92 35.49 Table 5-10 Burnup Requirements for Pre-EPU Non-Blanketed Category 11-3 Fuel Initial Enrichment (wt% ' 35U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 y 10.05 21.09 27.58 33.52 38.49 20 y 9.82 20.63 26.74 32.68 37.90 25 y 9.60 20.23 26.12 32.12 37.41 Table 5-11 Burnup Requirements for Pre-EPU Non-Blanketed Category 11-4 Fuel 235 Initial Enrichment (wt% U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 y 11.33 23.45 30.19 36.47 41.67 20 y 11.01 22.78 29.29 35.48 40.44 25 y 10.71 22.18 28.54 34.74 39.58 Table 5-12 Burnup Requirements for Pre-EPU Non-Blanketed Category 11-5 Fuel 235 Initial Enrichment (wt% U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 y 15.60 27.52 34.37 39.78 45.78 20 y 15.04 26.59 33.37 38.78 45.07 25 y 14.77 25.86 32.58 38.11 44.67 WCAP- 17094-NP February 2011 Revision 3

5-10 5.1.3.2 Pre-EPU and EPU Blanketed Fuel Table 5-13 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 1-3 Fuel 23 Initial Enrichment (wt% 1U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yr 0.00 19.38 26.21 31.50 36.60 2.5 y 0.00 18.63 25.17 30.31 35.24 5y 0.00 18.02 24.31 29.31 34.11 10 y 0.00 17.09 23.01 27.76 32.37 15 y 0.00 16.46 22.10 26.68 31.16 20y 0.00 16.02 21.48 25.92 30.32 25 y 0.00 15.72 21.05 25.38 29.74 Table 5-14 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 1-4 Fuel 235 Initial Enrichment (wt% U) 2.0 3.4 4.0 4.5 5.0 Cooling Time GWd/MTU 0 yr 5.99 25.40 32.33 37.75 43.01 2.5 y 5.80 24.26 30.98 36.24 41.33 5y 5.64 23.32 29.85 34.98 39.93 My 5.40 21.90 28.13 33.05 37.79 15 y 5.25 20.93 26.93 31.70 36.29 20y 5.15 20.27 26.10 30.75 35.24 25 y 5.08 19.81 25.52 30.09 34.51 WCAP- 17094-NP February 2011 Revision 3

5-11 Table 5-15 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-1 Fuel 23 Initial Enrichment (wt% 1 U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yr 6.10 23.93 30.47 35.58 40.51 2.5 y 5.91 22.92 29.24 34.23 39.01 5y 5.75 22.08 28.22 33.09 37.76 10y 5.51 20.84 26.68 31.35 35.84 15 y 5.35 20.00 25.62 30.13 34.48 20 y 5.23 19.44 24.89 29.28 33.53 25 y 5.16 19.06 24.38 28.68 32.86 Table 5-16 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-2 Fuel 235 Initial Enrichment (wt% U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yr 17.82 32.42 39.27 44.92 50.30 2.5 y 17.28 31.02 37.52 42.95 48.31 5y 16.79 29.87 36.10 41.35 46.63 M0y 15.98 28.13 34.02 38.98 44.04 15 y 15.33 26.94 32.66 37.41 42.20 20y 14.81 26.13 31.76 36.36 40.89 25 y 14.40 25.58 31.17 35.67 39.97 WCAP-17094-NP February 2011 Revision 3

5-12 Table 5-17 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-3 Fuel 235 Initial Enrichment (wt% U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yr 20.08 37.45 44.02 49.44 55.02 2.5 y 19.03 35.80 42.12 47.40 52.77 5y 18.18 34.50 40.57 45.69 50.90 lOy 16.96 32.63 38.26 43.09 48.04 15 y 16.18 31.45 36.71 41.28 46.06 20y 15.68 30.70 35.67 40.02 44.69 25 y 15.35 30.23 34.97 39.15 43.73 Table 5-18 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-4 Fuel 23 Initial Enrichment (wt% 1U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yr 24.31 40.49 47.73 53.66 59.35 2.5 y 23.01 38.68 45.61 51.32 56.90 5y 21.98 37.19 43.86 49.38 54.86 lOy 20.49 34.96 41.23 46.45 51.72 15 y 19.54 33.47 39.44 44.44 49.53 20 y 18.93 32.46 38.23 43.07 47.99 25 y 18.54 31.78 37.41 42.12 46.92 WCAP- 17094-NP February 2011 Revision 3

5-13 Table 5-19 Burnup Requirements for EPU and Pre-EPU Axial Blanketed Category 11-5 Fuel 23 Initial Enrichment (wt% 1U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yr 30.87 46.86 54.06 60.00 65.76 2.5 y 29.15 44.58 51.54 57.33 62.96 5y 27.75 42.74 49.47 55.11 60.61 1Oy 25.67 40.04 46.36 51.73 57.02 15 y 24.29 38.28 44.26 49.38 54.51 20y 23.37 37.12 42.83 47.75 52.75 25 y 22.76 36.37 41.87 46.63 51.53 5.1.3.3 IFBA Requirements for Fuel Category 1-2 Table 5-20 IFBA Requirements for Fuel Category 1-2 23 Enrichment (wt% ;U) Minimum Required Number of IFBA Pins Enr. _<

4.3 0 4.3 < Enr. <4.4 32 4.4 < Enr. <4.7 64 4.7 < Enr. < 5.0 80 WCAP- 17094-NP February 2011 Revision 3

5-14 5.1.4 Curve Fitting Coefficients for Minimum Burnup Requirements For all Fuel Categories except I-I and 1-2, an equation specifying the minimum required burnup as a function of the initial enrichment and post-irradiation cooling time is developed. The uncertainty in the burnup is included in the determination of the minimum burnup requirement and so it is appropriate to use the nominal burnup for comparing to the minimum required burnup determined from the loading curves. The burnup requirements are established as 3rd degree polynomial functions in the form of:

Bu ý (Al + A2*En + A3*En 2 + A4*En 3)* exp [ - (A5 + A6*En + A7*En 2 + A8*En 3 )*Ct]

+A9 + AI0*En +A1 l*En-2 + A12*En3 where:

Bu = Minimum required assembly average burnup (GWd/MTU)

En = Initial 2 3 5 U Enrichment (wt%)

Ct = Post Irradiation Cooling Time (years)

Ai = Coefficients (see Table 5-21 and Table 5-22)

Separate functional relationships are developed for blanketed and non-blanketed fuel assemblies. Pre-EPU blanketed assemblies must use the EPU curves. Note that for blanketed assemblies, the enrichment to be used in the loading curve equation is the enrichment of the axial section between the blanket material (the enrichment of the axial blankets is excluded when determining the assembly enrichment for application of the loading curve).

Since the loading curve is an exponential in cooling time, any cooling time between 0 and 25 years is allowed to be evaluated by the curve for blanketed fuel assemblies and between 15 years and 25 years for non-blanketed fuel assemblies. Fuel assemblies with cooling times greater than 25 years must conservatively use a value of 25 years.

The loading curves are valid for any enrichment between 2.0 and 5.0 for blanketed assemblies and between 1.8 and 4.0 for non-blanketed assemblies.

Coefficients for all loading curves, for both non-blanketed and blanketed assemblies are listed in Table 5-21 and Table 5-22. Required burnup values for selected initial enrichments are determined fron- these coefficients and are listed in Table 5-6 through Table 5-19 for information. Note that some burnups are above the current licensed limits but they are included for completeness.

WCAP- 17094-NP February 2011 Revision 3

5-15 Table 5-21 Non-Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct) (See Notes 1-4 for use of Table 5-21)

Fuel Category Coeff. 1-3 1-4 II-1 11-2 11-3 11-4 11-5 Al 2.04088171 -27.6637884 -11.2686777 20.7284208 29.8862876 -83.5409405 35.5058622 A2 -4.83684164 26.1997193 2.0659501 11.9673275 -37.0771132 94.7973724 -30.1986997 A3 2.59801889 -7.2982252 2.66204924 -14.4072388 16.3986049 -31.9583373 11.0102438 A4 -0.300597247 0.723731768 -0.513334362 2.83623963 -2.1571669 3.55898487 -1.27269125 A5 -0.610041808 0.401332891 -0.0987986108 -1.49118695 1.02330848 0.299948492 1.34723758 A6 0.640497159 -0.418616707 -0.0724198633 1.75361041 -1.21889631 -0.312341996 -1.19871392 A7 -0.219000712 0.144304039 0.106248806 -0.659046438 0.467440882 0.107463895 0.352920811 A8 0.0252870451 -0.0154239536 -0.0197359109 0.080884618 -0.0560129443 -0.0108814287 -0.0325155213 A9 -4.48207836 -5.54507376 -1.34620551 -245.825283 12.1549 39.4975573 -5.2576 A10 -2.12118634 -5.76555416 -10.1728821 243.59979 -22.7755385 -50.5818253 10.1733379 All 2.91619317 6.29118025 8.71968815 -75.7805818 14.3755458 23.3093829 0.369083041 A12 -0.196645176 -0.732079719 -1.14461356 8.10936356 -1.80803352 -2.69466612 0.0443577624 Notes:

1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for bumup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly bumup must exceed the "minimum bumup" (GWd/MTU) given by the curve fit for the assembly "'cooling time" and "initial enrichment." The specific minimum bumup required for each fuel assembly is calculated from the following equation:

Bu = (A1 + A2*En + A3*En2 + A4*En 3)* exp I- (A5 + A6*En + 3Ai*En 2 + As*En 3 )*Ct I

+A9 + Ajo*En + An*En2 + Aiz*En

2. Initial enrichment, En, is the nominal 2 3 5 U enrichment. Any enrichment between 1.8 and 4.0 may be used.

3. Cooling time, Ct, is in years. Any cooling time between 15 years and 25 years may be used. An assembly with a cooling time greater than 25 years must use 25 years.

4. This Table applies only for pre-EPU non-blanketed fuel assemblies. If a non-blanketed assembly is depleted at EPU conditions, none of the burnup accrued at EPU conditions can be credited (i.e., only burnup accrued at pre-EPU conditions may be used as burnup credit)..

WCAP- 17094-NP February 2011 Revision 3

5-16 Table 5-22 Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct) (See Notes 1-6 for use of Table 5-22)

Fuel Category Coeff. 1-3 1-4 II-1 11-2 IH-3 11-4 11-5 Al 5.66439153 -14.7363682 -7.74060457 -7.63345029 24.4656526 8.5452608 26.2860949 A2 -7.22610116 11.0284547 5.13978237 10.7798957 -20.3141124 -4.47257395 -18.0738662 A3 2.98646188 -1.80672781 -0.360186309 -2.81231555 6.53101471 2.09078914 5.8330891 A4 -0.287945644 0.119516492 0.0021681285 0.29284474 -0.581826027 -0.188280562 -0.517434342 A5 -0.558098618 0.0620559676 -0.0304713673 0.0795058096 -0.16567492 0.157548739 -0.0614152031 A6 0.476169245 0.0236575787 0.098844889 -0.0676341983 0.243843226 -0.0593584027 0.134626308 A7 -0.117591963 -0.0088144551 -0.0277584786 0.0335130877 -0.0712130368 0.0154678626 -0.0383060399 A8 0.0095165354 0.0008957348 0.0024057185 -0.0040803875 0.0063998706 -0.0014068318 0.0033419846 A9 -47.1782783 -20.2890089 -21.424984 14.6716317 -41.1150 -0.881964768 -12.1780 A10 33.4270029 14.7485847 16.255208 -10.0312224 43.9149156 9.69128392 23.6179517 All -6.11257501 -1.22889103 -1.77941882 5.62580894 -9.6599923 -0.18740168 -4.10815592 A12 0.490064351 0.0807808548 0.127321203 -0.539361868 0.836931842 0.0123398618 0.363908736 Notes:

1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "cooling time" and "initial enrichment." The specific minimum bumup required for each fuel assembly is calculated from the following equation:

Bu = (A, + Az*En + A3*Enz + A4*En 3)* exp [ - (A5 + 3 2 A 6 *En + A 7*EnZ + As*En )*Ct I

+ A9 + Aio*En + Ai*En + Alz*En3

2. Initial enrichment, En, is the nominal central zone 23SU enrichment. Axial blanket material is not considered when determining enrichment. Any enrichment between 2.0 and 5.0 may be used.

3. Cooling time, Ct, is in years. Any cooling time between 0 years and 25 years may be used. An assembly with a cooling time greater than 25 years must use 25 years.

4. Category I-I is fresh unburned fuel up to 5.0 wt% 235U enrichment.

5. Category 1-2 is fresh unburned fuel that obeys the IFBA requirements in Table 5-20 or contains an equivalent amount of another burnable absorber.

6. This table applies for any blanketed fuel assembly.

WCAP- 17094-NP February 2011 Revision 3

5-17 5.1.5 Confirmatory Criticality Calculations

]LC Table 5-23 through Table 5-42 below show that the maximum keff including all biases and uncertainties remain below 0.99 for both pre-EPU non-blanketed and pre-EPU and EPU blanketed fuel.

5.1.5.1 Pre-EPU Non-Blanketed Fuel Table 5-23 Confirmatory kff Calculations for Pre-EPU Non-Blanketed Category 1-3 Fuel 25 Initial Enrichment (wt% _ U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.95313 0.98908 0.98903 0.98911 0.98842 20 yrs 0.95313 0.98907 0.98895 0.98892 0.98932 25 yrs 0.953 13 0.98875 0.98896 0.98907 0.98884 Table 5-24 Confirmatory krr Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel 23 Initial Enrichment (wt% 1U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.98953 0.98861 0.98850 0.98871 0.98877 20 yrs 0.98971 0.98869 0.98881 0.98833 0.98882 25 yrs 0.98950 0.98872 0.98896 0.98852 0.98909 WCAP-17094-NP February 2011 Revision 3

5-18 Results presented in Table 5-25 through Table 5-28 are the maximum from cases with 1, 2, 3, or 4 Category 1-2 assemblies in the 2x2 Array.

Table 5-25 Confirmatory kIf Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel Paired with 235 Fresh Category 1-2 Fuel at 4.3 wt% U and 0 IFBA 23 Initial Enrichment (wt% SU)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.98877 0.98757 0.98445 0.98345 0.98204 20 yrs 0.98928 0.98697 0.98430 0.98259 0.98175 25 yrs 0.98870 0.98700 0.98423 0.98292 0.98139 Table 5-26 Confirmatory kfrr Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel Paired with 235 Fresh Category 1-2 Fuel at 4.4 wt% U and 32 IFBA 235 Initial Enrichment (wt% U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.98878 0.98730 0.98344 0.98201 0.98090 20 yrs 0.98875 0.98672 0.98319 0.98220 0.98091 25 yrs 0.98877 0.98651 0.98302 0.98188 0.98056 Table 5-27 Confirmatory kff Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel Paired with 23 5 Fresh Category 1-2 Fuel at 4.7 wt% U and 64 IFBA 23 Initial Enrichment (wt% SU)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.98884 0.98707 0.98316 0.98211 0.98113 20 yrs 0.98888 0.98676 0.98337 0.98219 0.98120 25 yrs 0.98892 0.98661 0.98316 0.98210 0.98098 WCAP- 17094-NP February 2011 Revision 3

5-19 Table 5-28 Confirmatory keff Calculations for Pre-EPU Non-Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 Fuel at 5.0 wt% 235U and 80 IFBA 23 Initial Enrichment (wt% -U)

CoolingI 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.98908 0.98705 0.98289 0.98184 0.98049 20 yrs 0.98894 0.98683 0.98295 0.98172 0.98064 25 yrs 0.98873 0.98683 0.98288 0.98181 0.98054 Table 5-29 Confirmatory kerfCalculations for Pre-EPU Non-Blanketed Category 11-1 Fuel 235 Initial Enrichment (wt% U)

CoolingI 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.98827 0.98910 0.98888 0.98916 0.98875 20 yrs 0.98827 0.98943 0.98933 0.98925 0.98892 25 yrs 0.98827 0.98931 0.98905 0.98858 0.98962 Table 5-30 Confirmatory kIffCalculations for Pre-EPU Non-Blanketed Category 11-2 Fuel 235 Initial Enrichment (wt% U)

CoolingI 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.98838 0.98861 0.98847 0.98859 0.98829 20 yrs 0.98781 0.98822 0.98763 0.98846 0.98943 25 yrs 0.98816 0.98858 0.98772 0.98821 0.98756 WCAP- 17094-NP February 2011 Revision 3

5-20 Table 5-31 Confirmatory kIffCalculations for Pre-EPU Non-Blanketed Category 11-3 Paired with Fuel Category 11-5 Fuel 235 Initial Enrichment (wt% U)

Cooling 1.8 2.5 3.0 3.5 4.0 Time GWd/MTU 15 yrs 0.97621 0.98104 0.98302 0.98281 0.98631 20 yrs 0.97633 0.98044 0.98223 0.98344 0.98503 25 yrs 0.97599 0.98086 0.98255 0.98431 0.98442 Table 5-32 Confirmatory kff Calculations for Pre-EPU Non-Blanketed Category 11-4 Fuel 235 Initial Enrichment (wt% U)

Cooling 1.8 2.5 3 3.5 4.0 Time GWd/MTU 15 yrs 0.98812 0.98558 0.98759 0.98793 0.98790 20 yrs 0.98773 0.98749 0.98728 0.98797 0.98767 25 yrs 0.98836 0.98841 0.98788 0.98767 0.98783 5.1.5.2 Pre-EPU and EPU Blanketed Fuel The maximum keff for Category I-1 fuel (fresh 5 wt% 23 WU) is 0.95335.

Table 5-33 Confirmatory kff Calculations for Pre-EPU and EPU Blanketed Category 1-3 Fuel 2

Initial Enrichment (wt% 11U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yrs 0.98437 0.98933 0.98899 0.98913 0.98901 2.5 yrs 0.98437 0.98880 0.98858 0.98874 0.98847 5 yrs 0.98437 0.98890 0.98863 0.98858 0.98869 10 yrs 0.98437 0.98894 0.98900 0.98938 0.98881 15 yrs 0.98437 0.98870 0.98920 0.98917 0.98920 20 yrs 0.98437 0.98875 0.98889 0.98912 0.98906 25 yrs 0.98437 0.98867 0.98886 0.98920 0.98855 WCAP- 17094-NP February 2011 Revision 3

5-21 Table 5-34 Confirmatory keff Calculations for Pre-EPU and EPU Blanketed Category 1-4 Fuel 23 Initial Enrichment (wt% ,U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yrs 0.98970 0.98834 0.98892 0.98853 0.98893 2.5 yrs 0.98953 0.98829 0.98864 0.98843 0.98842 5 yrs 0.98934 0.98811 0.98836 0.98826 0.98856 10 yrs 0.98930 0.98861 0.98861 0.98865 0.98861 15 yrs 0.98896 0.98869 0.98862 0.98896 0.98865 20 yrs 0.98882. 0.98926 0.98872 0.98878 0.98857 25 yrs 0.98880 0.98926 0.98887 0.98815 0.98827 Results presented in Table 5-35 through Table 5-38 are the maximum from cases with 1, 2, 3, or 4 fresh fuel assemblies in the 2x2 Array.

Table 5-35 Confirmatory kIf Calculations for Pre-EPU and EPU Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 at 4.3 wt% 235 U and 0 IFBA 235 Initial Enrichment (wt% U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yrs 0.98911 0.98806 0.98822 0.98833 0.98834 2.5 yrs 0.98884 0.98787 0.98824 0.98790 0.98791 5 yrs 0.98850 0.98776 0.98832 0.98777 0.98782 10 yrs 0.98858 0.98807 0.98789 0.98788 0.98810 15 yrs 0.98876 0.98820 0.98836 0.98815 0.98798 20 yrs 0.98843 0.98818 0.98796 0.98854 0.98792 25 yrs 0.98803 0.98832 0.98782 0.98773 0.98764 WCAP- 17094-NP February 2011 Revision 3

5-22 Table 5-36 Confirmatory kff Calculations for Pre-EPU and EPU Blanketed Category 1-4 Fuel Paired with Fresh Category 1-2 at 4.4 wt% 235U and 32 IFBA 23 5 Initial Enrichment (wt% U) 2.0 3.4 4.0 4.5 5.0 Cooling Time GWd/MTU 0 yrs 0.98857 0.98774 0.98795 0.98793 0.98837 2.5 yrs 0.98877 0.98749 0.98754 0.98779 0.98752 5 yrs 0.98879 0.98739 0.98771 0.98756 0.98755 10 yrs 0.98828 0.98769 0.98768 0.98796 0.98758 15 yrs 0.98827 0.98774 0.98789 0.98785 0.98781 20 yrs 0.98793 0.98820 0.98803 0.98747 0.98750 25 yrs 0.98771 0.98810 0.98757 0.9876 1 0.98747 Table 5-37 Confirmatory kfr Calculations for Pre-EPU and EPU Blanketed Category 1-4 Paired with Fresh Category 1-2 at 4.7 wt% 235U and 64 IFBA 23 5 Initial Enrichment (wt% U) 2.0 3.4 4.0 4.5 5.0 Cooling I Time GWd/MTU 0 yrs 0.98886 0.98806 0.98800 0.98772 0.98823 2.5 yrs 0.98861 0.98749 0.98770 0.98773 0.98784 5 yrs 0.98858 0.98733 0.98783 0.98771 0.98783 10 yrs 0.98845 0.98793 0.98781 0.98789 0.98796 15 yrs 0.98821 0.98805 0.98791 0.98791 0.98781 20 yrs 0.98804 0.98822 0.98768 0.98776 0.98789 25 yrs 0.98827 0.98817 0.98778 0.98765 0.98761 WCAP- 17094-NP February 2011 Revision 3

5-23 Table 5-38 Confirmatory kff Calculations for Pre-EPU and EPU Blanketed Category 1-4 Paired with 23 5 Fresh Category 1-2 at 5.0 wt% U and 80 IFBA 235 Initial Enrichment (wt% U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yrs 0.98902 0.98808 0.98803 0.98780 0.98774 2.5 yrs 0.98875 0.98731 0.98788 0.98762 0.98747 5 yrs 0.98838 0.98755 0.98786 0.98725 0.98727 10 yrs 0.98831 0.98772 0.98759 0.98732 0.98768 15 yrs 0.98831 0.98806 0.98785 0.98811 0.98778 20 yrs 0.98824 0.98798 0.98778 0.98771 0.98770 25 yrs 0.98815 0.98800 0.98785 0.98770 0.98739 Table 5-39 Confirmatory kfnCalculations for Pre-EPU and EPU Blanketed Category 1I-1 Fuel 23 5 Initial Enrichment (wt% U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yrs 0.98946 0.98901 0.98936 0.98924 0.98935 2.5 yrs 0.98918 0.98898 0.98903 0.98891 0.98904 5 yrs 0.98913 0.98885 0.98905 0.98882 0.98909 10 yrs 0.98932 0.98917 0.98918 0.98930 0.98924 15 yrs 0.98896 0.98953 0.98920 0.98917 0.98914 20 yrs 0.98926 0.98919 0.98904 0.98901 0.98926 25 yrs 0.98893 0.98867 0.98904 0.98884 0.98890 February 2011 WCAP- 17094-NP 17094-NP February 2011 Revision 3

5-24 Table 5-40 Confirmatory kff Calculations for Pre-EPU and EPU Blanketed Category 11-2 Fuel 235 Initial Enrichment (wt% U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yrs 0.98802 0.98804 0.98847 0.98811 0.98876 2.5 yrs 0.98859 0.98875 0.98844 0.98852 0.98832 5 yrs 0.98864 0.98879 0.98857 0.98900 0.98798 10 yrs 0.98829 0.98848 0.98837 0.98865 0.98819 15 yrs 0.98822 0.98836 0.98802 0.98833 0.98875 20 yrs 0.98834 0.98816 0.98953 0.98783 0.98864 25 yrs 0.98911 0.98821 0.98998 0.98686 0.98869 Table 5-41 Confirmatory kIrfCalculations for Pre-EPU and EPU Blanketed Category 11-3 Paired with Category 11-5 Fuel 235 Initial Enrichment (wt% U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yrs 0.98562 0.98267 0.98448 0.98570 0.98586 2.5 yrs 0.98576 0.98272 0.98384 0.98545 0.98575 5 yrs 0.98705 0.98174 0.98366 0.98505 0.98572 10 yrs 0.98911 0.98186 0.98377 0.98539 0.98566 15 yrs 0.98977 0.98268 0.98472 0.98499 0.98628 20 yrs 0.98978 0.98240 0.98383 0.98542 0.98619 25 yrs 0.98933 0.98083 0.98316 0.98478 0.98562 WCAP- 17094-NP February 2011 Revision 3

5-25 Table 5-42 Confirmatory kfrfCalculations for Pre-EPU and EPU Blanketed Category 11-4 Fuel 235 Initial Enrichment (wt% U)

Cooling 2.0 3.4 4.0 4.5 5.0 Time GWd/MTU 0 yrs 0.98770 0.98762 0.98779 0.98827 0.98828 2.5 yrs 0.98784 0.98720 0.98755 0.98740 0.98755 5 yrs 0.98785 0.98858 0.98715 0.98763 0.98755 10 yrs 0.98747 0.98848 0.98725 0.98765 0.98787 15 yrs 0.98748 0.98739 0.98704 0.98816 0.98799 20 yrs 0.98763 0.98822 0.98786 0.98793 0.98772 25 yrs 0.98821 0.98995 0.98823 0.98723 0.98762 February 2011 WCAP- 17094-NP February 2011 Revision 3

5-26 5.2 CASK AREA RACK A picture of the KENO model for the Cask Area Rack is shown in Figure 5-3. For this calculation, all fuel and rack dimensions are set using the worst case tolerances and position. The fuel is fresh 5.0 wt% 235U.

The keff for this model is [ ]1c. The only applicable uncertainties are the bias and uncertainty from the validation (a total adder of [ ]"c). The worst case keff after adding uncertainty is 0.9735, well below the regulatory limit of 1.0.

- TERJ2 00 00 000 00 00 00 ITT 1rýl~

0000000 0000 00000 00000 - ITT- 11 90 00*000000,60 000 000 000 000 000 000 000 000 =I, l~EItc 000000000000000 000000000000000 = ITfERI- 1 00 000000000 00 00 000000'ý000' "ý00 0000 00600 000* 0000 0 '0000 016,00 0000000 0000000 0000000 0000000 1- 1 00 00 000 00 WO 00 00,000 00 00 IVTRI I-M01-000000000000004 000000000000000 000 00 000000000060 000000000000 0 000000000000000 000000000000000 000000000000000 000000000000000 00 0000000 00 0000000000 00 010 00 00 000 00,00 0000060 0000600 0000 00000 0000 0000 00000 0000 00 000000000 0,0 00 000000,000 00 00000000000000-6 000000000,60wWOO 000 000 000 000 000 000 000 000 000000000000000 000000000000000

-0 0 04 0 010 0 0 0 0 0 0 00 000000000,00 0000 00000 00010 0000 00000 0000 0000000 0000000 0000000 0000000 00 00 000 00 00 00 00 000 00 00 1000000000000000 000000,000000000 000000000000000 000000000000000 Figure 5-3 KENO Model for the Cask Area Rack WCAP-17094-NP February 2011 Revision 3

5-27 5.3 NEW FUEL STORAGE RACK ANALYSIS The analysis described in the following subsections supports the safe storage of fuel in the New Fuel Storage Rack provided that the fuel meets one of the following conditions:

2 5 1, Enriched to 4.5 wt% 3 U or less.

21 Enriched to between 4.5 wt% 235U and 5.0 wt% 235U with at least 16 IFBA rods with 1.77 mg

' 0B/inch or higher and the IFBA portion of the rod covers at least the 7 central feet of the fuel or an equivalent amount of another burnable absorber material.

5.3.1 Model Description The analysis is done with SCALE 5.1 (parmn=N1TAWL) and the 44 group ENDF/B-V library. The validation of SCALE is documented in Appendix A. The New Fuel Storage Rack analysis is within the area of applicability shown in Table A-30.

The SCALE model of the rack is shown in Figure 5-4.

February 2011 WCAP- 17094-NP 17094-NP February 2011 Revision 3

5-28 LEGEND VOID MATERIALI MATERIAL2 MATERIAL3

[ MATERIALa

= MATERIAL9 Figure 5-4 New Fuel Storage Rack Model WCAP- 17094-NP February 2011 Revision 3

5-29 The following are comments about the SCALE model:

I. The UO 2 stack density is selected as 97% of the theoretical density. The maximum fresh fuel stack density is 96.5% of the theoretical density. The spent fuel pool analysis used 97.5 % of theoretical density to cover fuel densification during burnup.

2. The water is modeled at the maximum density of 1.00 gm/cm3 .

3. The axial reflector is assumed to be 100% water. Analysis is also performed with the axial reflector modeled as 50% water and 50% steel. In the fully flooded cases, the difference is less than the uncertainty. In the optimum moderation cases, the additional water increased reactivity.

The radial reflector is water at the same density as in the lattice.

4. [

asc 5.3.2 Rack Analysis Analysis showed that 5.0 wt% 235U fuel would not meet the 0.95 acceptance criterion in the fresh fuel rack without additional absorbers. The absorber selected is 16 IFBA rods in the assembly. The minimum number of IFBA rods in any Westinghouse IFBA design is 16 rods. It is assumed that the IFBA rods have a nominal '0B loading of

]*C Analysis further determined the 235 maximum enrichment allowed without absorbers is a nominal 4.5 wt% U.

Table 5-43 shows the results of the final analysis of the New Fuel Storage Racks. The acceptance criteria for dry storage racks are:

1. If the rack is fully flooded by water, keff must be less than 0.95.

2. If the rack is flooded with optimum reduced density water, keff must be less than 0.98.

Each criterion must be met including all biases and uncertainties. The 4.5 wt% 235U fully flooded analysis is done using a nominal base case and the addition of the bias from the validation and a statistical combination of independent uncertainties. The rest of the cases shown in Table 5-43 started with a base case that used [ ]a,c.

As seen from Table 5-43, racks with a large separation between assemblies can have higher reactivity at low water density. The cases with the [ most limiting fuel dimensions and locations ]a Care run changing the moderator density to find the low-density peak. Figure 5-5 and Figure 5-6 show the resulting reactivity as a function of moderation.

WCAP- 17094-NP February 2011 Revision 3

5-30 The analysis for 4.5 wt % 235U assumes an enrichment uncertainty of 0.05 wt% 235U. The analysis of the 5.0 wt% 235U assumes a maximum enrichment of 5.0 wt%. 235 U Although it is anticipated that this will be interpreted as a maximum nominal enrichment of 4.95 wt% 235U, as long as the actual maximum enrichment is 5.0 wt% 235U or less, the safety analysis applies.

235 A statistical combination of uncertainties is used for the 4.5 wt% U fully flooded case.

_ Table 5-43 Results for the New Fuel Storage Racks WCAP-17094-NP February 2011 Revision 3

5-31 a,c 2 35 Figure 5-5 kff as a Function of Water Density for 4.5 wt% U Fuel in the New Fuel Storage Racks a,c 235 Figure 5-6 keff as a Function of Water Density for 5 wt% U Fuel with 16 IFBA Rods WCAP- 17094-NP February 2011 Revision 3

5-32 5.3.3 Sensitivity Analysis for the New Fuel Rack The eccentric positioning of the fuel in the New Fuel Storage Rack cell is evaluated and found to have a small negative effect for the fully flooded case and a small positive impact for the optimum moderation case. In the fully flooded cases the calculated keff values are 0.93373 and 0.93333 for the centered and eccentric fuel, respectively. Since the cell pitch is so large (21 inches), there is little interaction between assemblies in the flooded model, leading to similar results in the centered and eccentric cases. To show this more clearly, a case is run with a single 4.5 wt% 235 U fuel assembly in the middle of a pool of water.

The k~f for this case is [ ] , which means that the interaction of assemblies in the eccentric positioning only increased the keff by [ 1"' in keff.

]rc 5.4 FUEL ROD STORAGE BASKET Storage of a fuel rod basket, instead of a fuel assembly, in one or more Region 1Icells is evaluated. The fuel rod storage basket is an 8x8 array of stainless steel tubes with the three tubes closest to each corner of the array omitted, i.e., the total number of tubes is 8x8 - 4x3 = 52. The KENO model conservatively represents the basket as 64 tubes. The dimensions used in the analysis are listed in Table 3-6. A picture of the KENO model is shown in Figure 5-7. It is conservatively assumed that all the tubes in the basket are filled with fresh 5.0 wt% 235U fuel rods.

February 2011 17094-NPP WCAP- 17094-N February 2011 Revision 3

5-33 The evaluation is performed by demonstrating that substituting fuel rod baskets in place of already analyzed fuel assemblies decreases the reactivity of the storage array. Because the Metamic Inserts are mounted to the top of the assembly it is not possible to place a fuel rod basket in a location requiring an insert. Therefore, the number of storage cells modeled as containing fuel rod baskets is varied from 0 (base case) to the total number of occupied cells, which do not contain a Metamic insert (4 minus the number of empty cells and Metamic inserts). Calculations are performed for Region II configurations which contain varying numbers of inserts. The results of these calculations in Table 5-44 show that replacing an analyzed assembly with a fuel rod basket always results in a reactivity decrease. Because the burnup requirements are higher for Region II than for Region I or the Cask Area Rack, replacing assemblies in those racks with fuel rod storage baskets should cause an even larger decrease in reactivity, and is permissible.

Figure 5-7 Radial View of a Fuel Rod Basket Model WCAP-17094-NP February 2011 Revision 3

5-34 Table 5-44 kIf Values with and Without Fuel Basket in Region If Storage Arrays Number of Baskets in Storage Array Fuel Assembly Description 2x2 Array k~fr +/-c 23 5 1I-A 1.8 wt% U Fresh, 0 yr cooling, 0 0.96860 +/- 0.00015 (One insert and unborated 1 0.92599 0.00017 one empty cell) 2 0.88082 +/- 0.00016 5.0 wt% 235U and 43 GWd/MTU, 0 yr 0 0.94693 +/- 0.00016 cooling, unborated 1.91064+0.00016 2 0.87497 +/- 0.00017 235 II-C 1.8 wt% - U, 7 GWdiMTU, 25 yr 0 0.99338 +/- 0.00013 (Two Inserts) cooling, unborated 1 0.96997 0.00015 2 0.93818 +/-0.00016 5.0 wt% 235U and 61 GWd/MTU, 0 yr 0 0.94425 +/- 0.00013 cooling, unborated 1 0.94274 0.00014 2 0.92529 +/- 0.00016 235 1.8 wt% 'U, 7 GWd/MTU, 25 yr 0 0.68175 +/- 0.00011 cooling, 2300 ppm soluble boron I 0.63250 +/- 0.00012 5.0 wt% 235U and 61 GWd/MTU, 0 yr 0 0.72724 +/- 0.00012 cooling, 2300 ppm soluble boron 1 ~0.67052 +/- 0.00012 II-D 1.8 wt% 23'U, 7 GWd/MTU, 25 yr 0 0.98648 +/-0.00015 (Three Inserts) cooling, unborated 0.94740 0.00015 5.0wt% 235U, 53 GWd/MTU, 0 yr 0 0.94203 +/- 0.00014 cooling unborated 1 0.92577 0.00015 WCAP- 17094-NP February 2011 Revision 3

5-35 5.5 INTERFACE REQUIREMENTS The interface of storage arrays can require restrictions in addition to those imposed by the infinite array analysis. The allowable interface configurations are shown in subsection 5.5.1 and the analysis is presented in subsection 5.5.2.

5.5.1 Allowable Interface Configurations This section illustrates the allowable interface configurations.

Array I-A - Region II Interfaces Storage Array I-A (Region 1) contains checker-boarded fresh fuel of Category 1-1. Figure 5-8 illustrates the allowable Array I-A-Region II interfaces. The shaded cells contain an insert and X indicates an empty (water filled) cell.

Array II-A Array II-B Array II-C Figure 5-8 Allowable Array I-A- Region II Interfaces WCAP- 17094-NP February 2011.

Revision 3

5-36 Array 1-B-Region II Interfaces Region I Array I-B contains four burned assemblies with no inserts (Fuel Category 1-4). Figure 5-9 illustrates the allowable Array I-B-Region II interfaces. The shaded cells contain an insert and X indicates an empty (water filled) cell. Note that any number of Category 1-4 fuel assemblies can be substituted with a fresh Category 1-2 fuel with an RCCA and 1FBA rods. There is no restriction on the placement of Category 1-2 fuel at the interface.

Array I-B Array I-B Array I-B 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 11-1 x 11-1 x 11-1 A 11-1 11-5 11-3 11-5 11-3 Array 11-A Array II-B Array II-C Array II-D Figure 5-9 Allowable Array I-B- Region 11 Interfaces Array I-D - Region 11 Interfaces Region I Array I-D consists of fuel assemblies stored in a 2x2 array with one of the fuel assemblies containing a full length RCCA (Fuel Category 1-3). Figure 5-10 illustrates the allowable Array I-D-Region II interfaces. The shaded cells either contain an insert (11-1) or a RCCA (1-3) and X indicates an empty (water filled) cell.

Array I-D Arra I-D Array I-D 1-3 1-3 1-3 1-3 1-3 1-3 1-3 1-3 1-3 1-3 1-3 1-3 1-3 1- 1-3] -

1 I1-3 1-3 1-3 1-3 1-3 II-I X II-1 X

n1-1 1- 11-5 11-3 11-5 11-3 Array II-A Array 1I-B Array II-C Figure 5-10 Allowable Array I-D- Region II Interfaces WCAP- 17094-NP February 2011 Revision 3

5-37 Notes:

1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Fuel categories are determined from Table 5-21 and Table 5-22.

3. Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.

4. X indicates an empty (water-filled) cell.

5. For Array I-D, the RCCA shall be in the row adjacent to the Region II rack.

6. For Array II-A, the empty cell shall be in the row adjacent to the Region I rack.

7. Array I-A shall not interface with Array 1I-D.

8. If no fuel is stored adjacent to Region II in Region I, then the interface restrictions are not applicable.

9. Figure 5-8 through Figure 5-10 are applicable only to the Region I - Region II interface. There are no restrictions for the interfaces with the Cask Area Rack.

10. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

WCAP-17094-NP February 2011 Revision 3

5-38 5.5.2 Results of Region I - Region II Interface Calculations This section discusses the results of the interface analysis for Region I-Region II. The allowable interface configurations meet a keff of less than 0.99 (0.01 Akcff lower than the regulatory requirement). This is demonstrated by analysis of the model described in Section 4.3.5. As previously stated, the biases and uncertainties for the interface can either be taken from the maximum of the individual storage arrays of both sides of the interface (Table 5-2 and Table 5-3) or can be recalculated for the interface configuration.

Biases and uncertainties are calculated for interfaces of arrays I-B - II-A, I-B - II-C, I-B-II-D, l-D - hIA and I-D - 1I-D. Table 5-45 below shows the total biases and uncertainties for these interfaces.

Table 5-46 shows the keff values (including biases and uncertainties) calculated for the allowable interface configurations shown in Figure 5-8 through Figure 5-10.

Table 5-45 Unborated Biases and Uncertainties for Region I and Region 11 Interfaces .,c WCAP- 17094-NP February 2011 Revision 3

5-39 a,c Table 5-46 Results for the Region I - Region II Interface Calculations Max kIf Region I Array Region 11 Array (Including Biases and Uncertainties)

II-A 0.98804 I-A II-B 0.98267 lI-C 0.98872 II-A 0.98460 1I-B 0.98266 I-B lI-C 0.98749 1I-D 0.98603 II-A 0.98455 Il-B 0.98508 I-D lI-C 0.98953 II-D 0.98448 WCAP- 17094-NP February 2011 Revision 3

5-40 Each Region I - Region 11 interface was explicitly analyzed to ensure that all safety criteria are met.

Figure 5-11 summarizes the interface restrictions that result from this analysis.

DEFINITION ILLUSTRATION Region I Rack Array 1I-A, as defined in 1-4 1-4 1-4 1-4 Figure 5-2, when placed on the interface with 1-4 1-4 1-4 1-4 Region I shall have the empty cell in the row adjacent to the Ar-1 X IX-1 X Region I rack.

• I1 '9 I-1 Array 11-A l- egion I Rack Region I Rack Region I Rack Arrays 1I-B, 1I-C and lI- 1-4 I-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 D, as defined in Figure 5-2, when placed on the 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 1-4 interface with Region I shall have an insert in every cell in the row adjacent to the Region I rack. 11-5 11-3 11-5 11-3 11-4 11-4 11-4 11-4 11-2 ay1I-Array I-B Array I-C Array 11-D Notes:

I. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.

2. Fuel categories are determined from Table 5-21 and Table 5-22.

3. Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.

4. X indicates an empty (water-filled) cell.

5. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.

Region I Array I-B is depicted as the example; however, any Region I array is allowed provided that

a. For Array I-D, the RCCA shall be in the row adjacent to the Region II rack, and

b. Array I-A shall not interface with Array II-D.

6. If no fuel is stored adjacent to Region II in Region I, then the interface restrictions are not applicable.

7. Figure 5-11 is applicable only to the Region I - Region 1Iinterface. There are no restrictions for the interfaces with the cask area rack.

8. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.

Figure 5-11 Interface Restrictions between Region I and Region II WCAP-17094-NP February 2011 Revision 3

5-41 5.6 SOLUBLE BORON CREDIT A minimum soluble boron concentration of 500 ppm must be maintained in the spent fuel pool to ensure that keff is less than 0.95, under all normal conditions. Table 5-47 presents the maximum keff values for normal conditions including biases and uncertainties. It is demonstrated that 500 ppm is sufficient to comply with the regulatory limit with at least 0.01 Akeff margin.

Table 5-47 Results for the Normal Operations with 500 ppm of Soluble Boron Max kfr Storage Array (Including Biases and Uncertainties)

I-A 0.87300 I-B/I-C 0.89009 I-D 0.89223 lI-A 0.92082 II-B 0.92180 II-C 0.92923 11-D 0.93184 5.7 NORMAL AND ACCIDENT CONDITIONS 5.7.1 Evaluation of Normal Conditions in the Spent Fuel Pool While the primary purpose of the spent fuel pool is storage of new and spent fuel assemblies, many other activities occur in the pool. The criticality safety analysis is intended to cover the following SFP activities in addition to normal fuel handling and storage activities.

1. Ultrasonic Testing (UT) of fuel assembly to determine leaking rods:

The rack cell location to perform UT must be selected based on allowed SFP configurations for the bounding fuel design. This configuration involves raising and lowering the assembly within the 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 interaction between assemblies. This configuration is bounded by the analysis.

2. Fuel assembly raised on pedestal or placed in new fuel elevator during fuel inspection (failed and healthy fuel inspections):

WCAP-17094-NP February 2011 Revision 3

5-42 No special consideration needs to be made for the new fuel elevator; a single isolated assembly is bounded by the criticality safety analysis. A single assembly will be less reactive than the analyzed storage array I-A.

The fuel assembly must meet allowed SFP configurations for the rack location. The presence of the pedestal will raise the assembly up higher than the surrounding assemblies, this is not unlike when an assembly is inserted or removed from a cell. Axially offsetting one assembly from the others will reduce the reactivity of the system by reducing the neutronic interaction between assemblies. This configuration is bounded by the criticality safety analysis.

3. Reconstitution of fuel assembly:

The removal of a fuel rod will change the moderator to fuel ratio of the assembly and could potentially cause a small increase in reactivity. However, the fuel assembly undergoing reconstitution will be kept at least one cell pitch from other fuel in the pool while the rod is removed. This will serve to isolate the fuel assembly and a single isolated assembly is bounded by the criticality safety analysis.

The lattice configuration with a stainless steel replacement rod is less reactive than the original lattice.

4. Fuel rod inspection (ECT, visuals, etc):

This configuration involves a fuel assembly being placed on a pedestal, a fuel rod being removed from the assembly, inspected, then returned to the assembly or placed in damaged rod storage.

The fuel assembly must meet allowed SFP configurations for the rack location. The presence of the pedestal will raise the assembly up higher than the surrounding assemblies, this is not unlike when an assembly is inserted or removed from a cell. Axially offsetting one assembly from the others will reduce the reactivity of the system by reducing the neutronic interaction between assemblies. The removal of a fuel rod will change the moderator to fuel ratio of the assembly and could potentially cause a small increase in reactivity. However, the fuel assembly undergoing inspection will be kept at least one cell pitch from other fuel in the pool while the rod is removed.

This will serve to isolate the fuel assembly and a single isolated assembly is bounded by the criticality safety analysis.

A single fuel rod is bounded by this analysis.

5. Fuel Rod Storage Basket (FRSB) raised on pedestal during fuel inspection to facilitate damaged fuel rod insertion:

The rack cell in which the FRSB is located must meet the requirements of the criticality analysis.

The raising of the fuel rod storage basket will reduce neutronic interaction with surrounding fuel and the resulting reactivity will be the same or lower than the reactivity of the configurations analyzed. The FRSB is analyzed in Section 5.4.

WCAP- 17094-NP February 2011 Revision 3

5-43

6. Storage of damaged fuel rods, fuel rod inserted in FRSB or fuel assembly guide tubes:

The FRSB is evaluated for fresh 5.0 wt% fuel, therefore fuel with any nominal initial enrichment less than 5.0 wt% and any amount of burnup may be stored in the FRSB.

Fuel rods stored in fuel assembly guide tubes tend to lower reactivity by displacing moderator.

This activity is bounded by the criticality safety analysis.

7. Fuel assembly inspection:

During fuel inspection, the fuel assembly will be raised and lowered in the same rack cell or in another SFP location (e.g., transfer canal, cask lay down area).

The fuel assembly must meet allowed SFP configurations for the rack or other location. This activity is bounded by the criticality safety analysis.

8. UT fuel assembly cleaning:

This configuration is analyzed by the UT vendor. No special consideration is given here.

9. Bottom nozzle inspections:

The fuel assembly is inspected during normal handling evolutions in the refueling cavity, transfer canal, or cask lay down area.

A single isolated assembly is bounded by the criticality safety analysis. A single assembly will be less reactive than the analyzed storage array I-A.

10. Fuel assembly debris removal:

Debris removal is performed away from other fuel assemblies (e.g., transfer canal cask laydown area)

A single isolated assembly is bounded by the criticality safety analysis. A single assembly will be less reactive than the analyzed storage array I-A.

11. Top Nozzle Separation visual inspection:

The fuel assembly is raised four feet and a camera is used to inspect guide tube sleeves.

The fuel assembly must meet allowed SFP configurations for the rack location. This activity is bounded by the criticality safety analysis.

12. Debris/Trash Storage Baskets:

WCAP- 17094-NP February 2011 Revision 3

5-44 A debris/trash storage basket may not be stored in a cell that is required to be empty by the criticality analysis (e.g. empty cells in configurations I-A and II-A). In each configuration, the debris/trash storage basket will replace a fuel assembly; the debris/trash storage basket contents listed have no significant fissile material and will therefore reduce the reactivity of the configuration relative to the analysis. This activity is bounded by the criticality safety analysis because the storage basket must be placed in a location intended for a fuel assembly.

13. Physically damaged SFP cells (with pre-existing fuel assembly or with no fuel assembly). Cells that are bent or damaged during fuel handling:

There are two possible usages for damaged cells. The cell may be used for storage of an assembly as intended or credited as an empty cell. Storage of fuel in the damaged cell is permissible if the damage is not in the active fuel region. Credit for an empty cell is permissible since the amount of water and steel is maintained relative to the analysis contained herein.

14. Metamic inserts or RCCAs stored in empty rack cell or stored between racks and SFP wall:

The Metamic inserts or RCCAs stored in empty rack cells or stored between racks and the SFP wall may not be credited as part of a configuration. The addition of a neutron absorber relative to the analyzed configuration is acceptable for usage as an empty cell.

15. A single fuel assembly in transit (e.g., refueling canal, fuel upender, or fuel elevator):

No special consideration needs to be made for the refueling canal, fuel upender, or fuel elevator; a single isolated assembly is bounded by the criticality safety analysis. A single assembly will be less reactive than the analyzed storage Array I-A.

5.7.2 Accident Conditions The following reactivity increasing accidents are considered in this analysis:

eMisloaded fresh fuel assembly into incorrect storage rack location elnadvertent removal of an absorber insert eSpent fuel pool temperature greater than normal operating range (150 'F) eLoss of water gap between Region I and Region II due to seismic event eDropped fresh fuel assembly

• Misplaced Fuel Assembly 5.7.2.1 Misloading of an Assembly into the Storage Racks The misloaded fresh fuel accident scenario is analyzed by placing a 5.0 wt% 2 35U fresh fuel assembly into the water-filled cell required by Array II-A. This is expected to be the bounding condition since the fresh assembly is being surrounded by the most reactive fuel allowed in Region II. A 6x6 model is utilized with periodic boundary conditions containing one misloaded fresh fuel assembly. Figure 5-12 illustrates Array 11-A with one misloaded assembly. A misload into one of the empty cells in Array I-A is also analyzed and found to be less limiting. This accident requires 1683 ppm of boron to maintain kYfr less WCAP- 17094-NP February 2011 Revision 3

5-45 than 0.95, including biases and uncertainties. The results of these calculations are presented in Table 5-48.

Figure 5-12 Model of a Misloaded Assembly in Storage Array 11-A 5.7.2.2 Inadvertent Removal of an Absorber The removal of an absorber insert from an already analyzed array is bounded by the misload because the incorrectly placed assembly will be more heavily burned than the analyzed misload case, therefore this accident is covered.

5.7.2.3 Temperature Greater than Normal Operating Range The spent fuel pool is to be operated at less than 150 OF. However, under accident conditions this temperature could be higher. Due to the large volume of water in the spent fuel pool, boiling off the pool water before remnediation is not credible; therefore the lowest density of the water is the water density at boiling and atmospheric pressure, 0.96 gm/cm3 . Calculations are run with voiding and 1683 ppm of soluble boron. Although not credible, an additional case with a moderator density of 0.75 gmn/cm 3 is performed. The results for these calculations are presented in Table 5-48.

5.7.2.4 Loss of Region I to Region 11 Gap A seismic event could reduce the spacing between rack modules in the spent fuel pool. This accident scenario is analyzed by not modeling any water gap between a representative Region I and Region 11 interface model described in subsection 4.3.5.2. Additionally, the fuel assemblies are eccentrically positioned toward the interface. Results shown in Table 5-48 demonstrate a large margin to the regulatory limit, therefore the remaining interfaces are not analyzed.

WCAP- 17094-NP February 2011 Revision 3

5-46 5.7.2.5 Dropped Fuel Assembly During placement of the fuel assemblies in the racks, it is possible to drop the fuel assembly from the fuel handling machine. The dropped assembly could land horizontally on top of the other fuel assemblies in the rack. In this case, there is significant separation between the dropped fuel assembly and the rest of the fuel assemblies due to the top nozzle, fuel rod plenum, fuel rod end plug, and the separation between the fuel rod and the top nozzle. It is clear that the misloaded fresh fuel assembly described above is far more limiting than a single assembly lying horizontally on top of other assemblies in the rack. It is also possible that a fuel assembly could be dropped in its location with such force that the resultant fuel assembly deforms the support structure such that more of the fuel assembly is below the absorbers. The removal of an absorber insert represents 100% of the assembly below the absorber and this is seen in subsection 5.7.2.2 to be non-limiting.

5.7.2.6 Misplaced Fuel Assembly It is possible to misplace a fuel assembly in a location not intended for fuel. Any assembly placed outside of the racks is surrounded by water on at least two sides. The misloaded fresh assembly discussed above is surrounded by fuel on all four sides. The additional neutron leakage of the two sides not facing fuel ensures that this condition is bounded by the misload event.

There are two conditions with the Cask Area Rack that must be considered. First, the Cask Area Rack has a corner where there is no storage cell box. It is possible, though very unlikely, that a fresh fuel assembly could be placed in this corner such that there is only one panel of Boral separating this misplaced fuel assembly from the fuel assemblies in the Cask Area Rack. The infinite 2x2 model of the Cask Area Rack is modified to model this scenario. See Figure 5-13 below.

WCAP- 17094-NP February 2011 Revision 3

5-47 0 0 0 C C) 0 0 0 0 0 0 G 0 () 0 000000000000000 00 00 Go* 00 00 0000000 0000000 0 C)0 0 Goo 0 0 0 ý*16'660 0000 00600 0000 0 0 0 0 (:) 0 Q 0 0 0 0 00 000000000 00 0000000 *000000 04DOO 000 *00'900 00100 060064000 00 00000000000000100000000oO00000 00 0000000OG 00!000 COO (DOO Goo

• 606, 00060 0,00016o e000000 0000ooo!00 000000000 00 (DO 00 000 00 001i 0000 00000 0000 000000000000000 0000000 0000000 00 00 000 00ý00 000000000000000 0000180000000000 000000000000000 *20000000M50.90

'000000000000000 Goo 0 *Goo* 0 00".0 00 00 004D 00 00 00 00 000 00 00 0000000 0000000 0000000 0000000 0000 OaGeo 6800; 0000 000%.p. 60"160 00 000000000 001ý 100 00000"001 "GO 000000000000000 000000000000000 000 Goo GOO COO 04DO 000 Goo 000 4bo OoO000004,86 00 000000000 00 0000 00000 0000 0000 00000 0000 0000000 6009000 0000000 0000000 s o Goo (DO 00 00 06 Goo 00 00 00000(Z)OO0000400 000000000000000 0*00100000000000 j

ý)

Figure 5-13 Model for Misloading a Fresh Assembly in the Cask Area Rack Corner Cut The second condition concerns the fact that one side of the Cask Area Rack does not contain any Boral absorber because it is designed to face the pool wall. While it is considered extremely unlikely that the cask rack could be mis-positioned, if the entire rack is rotated 180 degrees, then the side with no Boral will be facing fuel assemblies in Region II. The infinite cask area model is modified to remove the Boral absorber on two sides. This very conservatively represents the rack facing fuel assemblies in Region II.

Both of the conditions described above for the Cask Area Rack have been evaluated and it was noted that these conditions are bounded by the misload event.

When a fuel assembly is positioned in the upender, it possible to bring another fuel assembly in close approach to the upender. The condition of having two fully enriched fuel assemblies in direct contact with one another is bounded by the misload into the storage array I-A. While the assemblies are slightly closer together in the upender event, the I-A misload accident surrounds the misloaded assembly with fresh fully enriched fuel on all sides. The additional leakage associated with having no neighboring assemblies ensures that this accident is bounded.

WCAP- 17094-NP February 2011 Revision 3

5-48 5.7.2.7 Summary of Accident Results All of the accident calculations were performed with 1683 ppm of soluble boron. A summary of the results of all calculations are presented in Table 5-48.

Table 5-48 Results of the Accident Calculations Max kef Accident Description (Including Biases and Uncertainties)

Misload into 11-A 0.93981 Misload into I-A 0.82654 Loss of Region I - Region II gap 0.81340 Temperature Above Normal Operations (density 0.82441

= 0.96 gm/cm 3)

Temperature Above Normal Operations (density 0.82785

= 0.75 gm/cm 3 )

WCAP-17094-NP February 2011 Revision 3

6-1 6 LIMITATIONS OF ANALYSIS The following inputs to this analysis will be checked in order to ensure compliance with the criticality safety design basis.

6.1 FUEL LIMITATIONS The following limitations apply to the as manufactured fuel characteristics:

I. This analysis is applicable to the Westinghouse 15x1 5 STD fuel design. Additionally, the fuel designs in Section 4.1 have been shown to be bounded. Any other fuel designs will be checked against the main fuel design characteristics.

2. The initial stack density shall be less than 96.5% of the theoretical density of uranium dioxide.

The stack density may be calculated as the ratio of mass of uranium dioxide contained in the fuel rod of interest to the mass of a 144 inch tall cylinder of uranium dioxide of full theoretical density.

3. All EPU fuel shall contain annular blankets that are at least 8 inches in length and enriched to 2.6 wt% 235U or less. However, in the special case of 6 inch natural uranium solid and annular blankets, it is shown that the 8 inch enriched blanket depleted under EPU conditions is conservative (see subsection 4.3.1.5).

4. All future fuel assemblies shall be limited to a maximum of 148 IFBA rods if containing a WABA with 16 fingers or a maximum of 100 IFBA rods if a 20 fingered WABA insert is used.

6.2 OPERATIONAL LIMITATIONS I. Turkey Point shall not take credit for any burnup accrued on non-blanketed fuel under EPU conditions. All non-blanketed fuel is assumed to be characterized by the non-blanketed fuel currently in the Turkey Point spent fuel pools and exposed to pre-EPU conditions. Turkey Point has no plans to use non-blanketed fuel once the EPU is implemented.

2. The burnup averaged soluble boron concentration for all future fuel assemblies shall be less than 1000 ppm.

3. The PAssembly values for all future fuel assemblies shall be less than 1.41 for Region I storage and 1.28 for Region II storage.

4. All future fuel assemblies shall not accrue more than 30 GWd/MTU while exposed to a WABA insert.

5. Fuel assemblies that do not meet operational limits and assumptions will be specifically evaluated and classified following the same methodology used in this report.

WCAP-17094-NP February 2011 Revision 3

6-2 6.3 SPENT FUEL POOL LIMITATIONS I. Each Metamic insert shall have an areal density greater than or equal to 0.015 1

°B gm/cm 2.

2. Each Cask Area Rack Boral panel shall have an areal density greater than or equal to 0.0204 '0B 2

gm/cm

3. The center to center spacing of Region I shall be greater than or equal 10.48 inches, the center to center spacing of Region II shall be greater than or equal to 8.97 inches, and the center to center spacing of the Cask Area Rack shall be greater than or equal to 10.06 for each cell.

WCAP- 17094-NP February 2011 Revision 3

7-1 7 REFERENCES I. Code of Federal Regulations, Title 10, Part 50, Section 68, "Criticality Accident Requirements."

2. "SCALE: A Modular Code System for Performing Standard Computer Analyses for Licensing Evaluation," ORNL/TM-2005/39, Version 5.1, Vols. I-Il, November 2006. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-732.

3. L. 1. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," Nuclear Regulatory Commission, Rockville, MD, August 1998.

4. M. Ouisloumen, H. Huria, et al, "Qualification of the Two-Dimensional Transport Code PARAGON," WCAP-16045-P-A, Westinghouse Electric Company, Pittsburgh, PA, August 2004.

5. FPL License Amendment Request No. 178, Boraflex Remedy, L-2007- 112.

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

7. J. C. Wagner, M. D. DeHart, and C. V. Parks, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses", NUREG/CR-6801 (ORNL/TM-2001/273), U.S. Nuclear Regulatory Commission, Washington, DC, March 2003.

8. A. H. Wells, et al., "Criticality Effect of Neutron Streaming Between Boron Carbide Granules in Boral for a Spent Fuel Shipping Cask," Trans. Am. Nucl. Soc., 54, 205 (1987).

9. S. E. Turner, "Reactivity Effects of Streaming Between Discrete Boron Carbide Particles in Neutron Absorber Panels for Storage or Transport of Spent Nuclear Fuel," Nucl. Sci. Eng., 15 1, 344 (2005).

10. S. E. Turner, T. G. Haynes III, "Analytical and Experimental Investigation of Recent Challenges to Neutron Attenuation and the Relationship with Criticality Analyses," Rad. Prot., 169, 195 (2010).

11. W.B. Henderson, "FIGHTH -A Simplified Calculation of Effective Temperatures in PWR Fuel Rods for Use in Nuclear Design", WCAP-9522, Westinghouse Electric Company, Pittsburgh, PA, May 1979.

12. J. C. Wagner, "Impact of Soluble Boron Modeling for PWR Burnup Credit Criticality Safety Analyses," Trans. Am. Nucl. Soc. 89, 120-122 (2003).

13. J. C. Wagner, et al., "NRC/ORNL Methods Presentation on Spent Nuclear Fuel Burnup Credit Analysis Validation," NRC HQ Rockville, MD, September 2010.

WCAP-17094-NP February 2011 Revision 3

7-2

14. C. V. Parks et al., "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel", NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge, TN, February 2000.

15. M. D. DeHart, "Sensitivity and Parametric Evaluations of Significant Aspects of Burnup Credit for PWR Spent Fuel Packages", ORNL/TM-12973, Oak Ridge National Laboratory, Oak Ridge, TN, May 1996.

February 2011 WCAP- 17094-NP 17094-NP February 2011 Revision 3

A-I APPENDIX A VALIDATION OF SCALE 5.1 A.1 INTRODUCTION The purpose of this appendix is to summarize the validation of the SCALE Version 5.1 code system (Reference A. 1) to be used in the Turkey Point Units 3 and 4 criticality safety analysis of fresh and spent fuel storage.

]aC The methodology validated in this Appendix is based on the use of the CSAS25, NITAWL-II, BONAMI, and KENO V.a modules of the SCALE code system. Any mention of KENO in this report refers to KENO V.a.

This validation is designed to cover fresh and spent fuel storage for Turkey Point Units 3 and 4. It also covers the criticality analysis for any movement of fuel from the spent fuel pool to the core or cask loading and other normal operations in the spent fuel pool. The validation is adequate to cover all present and anticipated (non-mixed-oxide) LWR fuel designs at Turkey Point. The area of applicability is discussed in more detail in subsection A.5.3.5.

A.2 CALCULATIONAL METHOD The analysis methodology employs SCALE Version 5.1, as documented in Reference A. 1, using a 44

-group Evaluated Nuclear Data File Version 5 (ENDF/B-V) neutron cross section library.

The SCALE system is developed for the NRC to satisfy the need for a standardized method of analysis for evaluation of nuclear fuel facilities and shipping package designs. The SCALE version that is utilized here is a code system that runs on LINUX clusters and includes the control module CSAS25 and the following functional modules: BONAMI, NITAWL-II, and KENO V.a. All references to KENO in the text to follow should be interpreted as referring to the KENO V.a module.

Standard material compositions are employed in the SCALE analyses consistent with the material descriptions in References A.3 through A.8.

All calculations are performed on systems with the following hardware and software characteristics:

• SCALE Version 5.1

• SUSE Linux 9.0 and 10.0 Correct installation and operation of the SCALE code system is verified by performing test cases on the platforms described above.

WCAP- 17094-NP February 2011 Revision 3

A-2 A.3 VALIDATION METHOD a,c WCAP- 17094-NP February 2011 Revision 3

A-3

]a,c A description of the critical experiments modeled is contained in Section A.4. The results and area of applicability are contained in Section A.5.

A.3.1 Determination of Bias and Bias Uncertainty

]ac WCAP-17094-NP February 2011 Revision 3

A-4 I

Iac A.3.2 Test for Normal Distribution

[

Iac WCAP- 17094-NP February 2011 Revision 3

A-5

]ac A.3.3 Identify Trends in the Data Ias WCAP- 17094-NP February 2011 Revision 3

A-6 ac WCAP- 17094-NP February 2011 Revision 3

A-7 aIc WCAP-17094-NP February 2011 Revision 3

A-8

]a,c A.4 CRITICAL EXPERIMENT DESCRIPTION ac WCAP- 17094-NP February 2011 Revision 3

A-9 A.4.1 [

a,c WCAP- 17094-NP February 2011 Revision 3

A-10

[

A.4.2 a,c

]a,c WCAP- 17094-NP February 2011 Revision 3

A-Il

]8C A .4 .3 a,c A.4.4 [a~c ala, la,c WCAP- 17094-NP February 2011 Revision 3

A-12 a]c A.4.5 [

ac

]a,c WCAP-17094-NP February 2011 Revision 3

A-13 a~c A.4.6 [

a,c

]a WCAP- 17094-NP February 2011 Revision 3

A-14 II

] a~c A.4.7 [

Ia,c II

]ac WCAP- 17094-NP February 2011 Revision 3

A-15 A.4.8 [

Ia,c I

]ao WCAP- 17094-NP February 2011 Revision 3

A-16 I

.]Lc A.4.9 [ ISaC

[

]ac a,c WCAP- 17094-NP February 2011 Revision 3

A- 17 a,c ja,c A.4.10 [

]ac WCAP- 17094-NP February 2011 Revision 3

A-18 aW 4

WCAP- 17094-NP February 2011 Revision 3

A-19 a.c A.5 RESULTS This section presents the results of the validation analysis. This includes the raw calculational results, the calculation of the bias and bias uncertainty, the detailed statistical trending results, and the definition of the AOA.

A.5.1 Raw Calculational Results Table A-I shows the raw calculational results for each of the critical experiments considered in this validation.

WCAP- 17094-NP February 2011 Revision 3

A-20 I

A-20 Raw Calculational Results ac ITable A-1 I

WCAP- 17094-NP February 2011 Revision 3

A-21 Table A-1 Raw Calculational Results (Cont.)

a,c WCAP- 17094-NP February 2011 Revision 3

A-22 Table A-1 Raw Calculational Results (cont.) Iac WCAP- 17094-NP February 2011 Revision 3

A-23 ITable A-I Raw Calculational Results I(cont) Sa c WCAP- 17094-NP February 2011 Revision 3

A-24 Table A-1 Raw Calculational Results (Cont.) a,c WCAP- 17094-NP February 2011 Revision 3

A-25 A-25 Table A-I Raw Calculational Results I(Cont.)

Ia,_c WCAP- 17094-NP February 2011 Revision 3

A-26 T able A-1 Raw Calculational Results

__ (cont.) a,c February 2011 WCAP- 17094-NP February 2011 Revision 3

A-27 Raw Calculational Results ITableA-1 I a,c WCAP- 17094-NP February 2011 Revision 3

A-28 Table A-1 Raw Calculational Results (Cont.)

a,c WCAP- 17094-NP February 2011 Revision 3

A-29 Table A-I Raw Calculational Results (cont.)

WCAP- 17094-NP February 2011 Revision 3

A-30 A-30 Raw Calculational Results (cont.)A-1 Table I ac WCAP- 17094-NP February 2011 Revision 3

A-3 1 Table A-1 Raw Calculational Results (Cont.) I a, c WCAP- 17094-NP February 2011 Revision 3

A-32 Table A-1 Raw Calculational Results (cont.)

Iax_

February 2011 WCAP- 17094-NP February 2011 Revision 3

A-33 T able A-I Raw Calculational Results S](cont.)

WCAP- 17094-NP February 2011 Revision 3

A-34 A.5.2 Determination of Bias and Bias Uncertainty and Normality Check This validation suite is intended to be used for fresh and spent fuel storage for LWRs. There are several situations which occur frequently with respect to LWR fuel storage a[

]a*c a,c WCAP-17094-NP February 2011 Revision 3

A-35 a,c a,c A.5.3 Trending Analysis The regression fits and goodness of fit tests described in subsection A.3.3 are applied to each subset of data for each of the following parameters:

a [

Iac WCAP- I 7094-NP February 2011 Revision 3

A-36 0

Ia~c A.5.3.1 axc II

]a~c ac WCAP-17094-NP February 2011 Revision 3

A-37 ac a[c WCAP-17094-NP February 2011 Revision 3

A-38

_a,c a,c

]a~c WCAP- 17094-NP February 2011 Revision 3

A-39 II

]a,c a,c a,c WCAP-17094-NP February 2011 Revision 3

A-40

]aac a,c a,c WCAP-1 7094-NP February 2011 Revision 3

A-41 a,c F

ac Sa,c WCAP-17094-NP February 2011 Revision 3

A-42 a,c

]a,c a,c WCAP- 17094-NP February 2011 Revision 3

A-43 a,c

[

]a.c ac WCAP- 17094-NP February 2011 Revision 3

A-44

]a,c l',C A.5.3.2 II

]a,c WCAP- 17094-NP February 2011 Revision 3

A-45 a,c a,c a~c WCAP- 17094-NP February 2011 Revision 3

A-46 II Ia~c A-46 WCAP-17094-NP February 2011 Revision 3

A-47 1a~c axc a,c WCAP-17094-NP February 2011 Revision 3

A-48 a,c m

[

-Ae a,c WCAP- 17094-NP February 2011 Revision 3

A-49 a,c a,c WCAP-1 7094-NP February 2011 Revision 3

A-50 a,c a,c A.5.3.3 aC

]ac WCAP-17094-NP February 2011 Revision 3

A-5 1

[

]a,c ac ac WCAP- 17094-NP February 2011 Revision 3

A-52 ac a,c WCAP- 17094-NP February 2011 Revision 3

A-53

[

ac a,c WCAP-17094-NP February 2011 Revision 3

A-54 a,c

[

a~c a,c WCAP- 17094-NP February 2011 Revision 3

A-55 a,c ac WCAP- 17094-NP February 2011 Revision 3

A-56 a,c a,c a ,c WCAP- I 7094-NP Februarv 2011 Revision 3

A-57 I]I a,c a,c WCAP-17094-NP February 2011 Revision 3

A-58 ac a]c WCAP- 17094-NP February 2011 Revision 3

A-59 a,c A.5.3.4 Summary of Trend Results II

]ac WCAP-17094-NP February 2011 Revision 3

A-60 lC

))a~c

[

A.5.3.5 Area of Applicability Definition The area of applicability (AOA) of this benchmark is defined by the range of parameters in the validation suite. Table A-30 summarizes the Areas of Applicability. a,c WCAP-1 7094-NP February 2011 Revision 3

A-61 a,c WCAP- 17094-NP February 2011 Revision 3

A-62 A.5.3.6 Summary of the Bias and Bias Uncertainty Table A-31 summarizes the results of the validation for Westinghouse 15x 15 fuel.

__a,c WCAP- 17094-NP February 2011 Revision 3

A-63 A.6 REFERENCES A. 1 "SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," ORNL/TM-2005/39, Version 5.1, Vols I-III, November 2006. Available from Radiation Safety Infonnation Computational Center at Oak Ridge National Laboratory as CCC-732.

A.2 [

Iasc A.3 IaWc A.4

]a.c A.5

]a.c A.6 Ia~c A.7 I

]a.c A.8 Ia~c A.9 Sac WCAP-17094-NP February 2011 Revision 3

Turkey Point Nuclear Plant L-2011-032 License Amendment Request No. 207 Attachment 3 Enclosure Turkey Point Units 3 and 4 LAR NO. 207 FUEL STORAGE CRITICALITY ANALYSIS SUPPLEMENT 1 ATTACHMENT 3 Westinghouse Affidavit for Attachment 4 February 4, 2011 This coversheet plus 7 pages

Westinghouse Electric Company

(

HhWestinghouse Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 Proj letter: NF-FP-1 1-25 CAW-11-3100 February 4, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

WCAP- 17094-P, Revision 3, "Turkey Point Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-1 1-3100 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations:

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Florida Power & Light.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-1 1-3100, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Very truly yours, J A Gresha m , Manager Regulatory Compliance Enclosures

CAW-11-3100 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

(J. A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 4th day of February 2011 NoayPublic COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member. Pennsylvania Association of Notaries

2 CAW- 11-3 100 (1) 1 am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-11-3100 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

4 CAW-11-3100 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390; it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in WCAP-17094-P, Revision 3, "Turkey Point Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis" (Proprietary) dated February 2011, for submittal to the Commission, being transmitted by Florida Power &

Light letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse. is that associated with Turkey Point Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis, and may be used only for that purpose.

This information is part of that which will enable Westinghouse to:

(a) Demonstrate the sub-criticality of the spent fuel pool.

5 CAW-11-3100 (b) Address the draft Interim Staff Guidance (ISG) DSS-ISG-2010-1, "Draft Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools," August 2010.

Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of similar information to its customers for purpose of demonstrating the sub-criticality of the spent fuel pool.

(b) Westinghouse can sell support and defense of spent fuel pool criticality analysis.

(c) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar calculations and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

K

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license,

.permit, order, or regulation subject to the requirements of 10 CFR 2.3 90 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.