NL-19-0796, WCAP-18414-NP, Rev 0, J.M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis (Non-Proprietary Version)

WCAP-18414-NP, Rev 0, J.M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis (Non-Proprietary Version)

ML19275E312
Person / Time
Site:	Farley
Issue date:	09/30/2019
From:	Wenner M Westinghouse
To:	Office of Nuclear Reactor Regulation
Shared Package
ML19275E393	List: NL-19-0796, WCAP-18414-NP, Rev 0, J.M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis (Non-Proprietary Version)
References
NL-19-0796 WCAP-18414-NP, Rev 0
Download: ML19275E312 (155)

Text

SNC to NRC LAR Enclosure NL-19-0796 ENCLOSURE Attachment 6 WCAP-18414-NP J. M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysl~

(Non-proprietary Version)

)

Westinghouse Non-Proprietary Class 3 WCAP-18414-NP September 2019 Revision O J. M. Farley Units 1 and 2 Spent Fuel Pool Criticality' Safety Analysis

• @Westinghouse

Westinghouse Non-Proprietary Class 3 WCAP-18414-NP Revision 0 J. M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety .

Analysis Michael T. Wenner* .

• Core Engineering & Software Development September 2019 Reviewers: Susan M. Nelson* and Andr6w J. Sheaffer*

Core Engineering & Software Development Vefa N. Kucukboyaci*

Fuel Technology and Product Development Sean T. Kinnas*

Safety Analysis Approved: .Stephanie Y. Harsche*, Manager Core Engineering & Software Development

• Electronically approved records are authenticated in the el~ctronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA 0 2019 Westinghouse Electric Company LLC

• All Rights Reserved

ii REVISION HISTORY Revision Description and Impact of the Change Date 0 Original Issue . 09/2019 TRADEMARK NOTICE ZLRLO and Optimized ZIRLO are trademarks Of registered trademarks of Westinghouse El~c Company LLC, its affiliates and/or its subsidiaries in the United States of Ame~ca and may be registered in other countries throughout the world. All rights reseived. Unauthorized ll5e is strictly prohibited. Other names may be trademarks of their respectiv~ OW!}ers.

Boraflex is a trademark or registered trademark of its respective owner. Other µrunes may be trademarks of their respective owners.

WCAP-18414-NP Revision 0

iii TABLE OF CONTENTS LI~T OF TABLES .............................................................................-........................................................... V LIST OF FIGURES ................................................................................................................................... viii LIST OF ACRONYMS, INITIALISMS, AND TRADEMARKS ............................................................... ix INTRODUCTION ......................................... :............................. ,................................................ 1-1 2 OVERVIEW ........ :..................................................... - ............................................... ~................. 2-1 2.1 ACCEPTANCE CRITERIA.'. .......................................................................................... 2-l 2.2 DESIGN APPROACH ....................... :... ;...................................................................... .'.. 2-1 2.3 COtvtPUTER CODES .................................................................................................... .'.2-2 2.3.1 Two-Dimensional Transport Code PARAGON ............................................... 2-2 2.3.2 Scale Code Package ...................................................... :.................................. 2-3 2.3.3 Scale 238 Group Cross-Section Library :***********--***********--********************:.: .....-.. ;2-5 3 FARLEY UNITS 1 & 2 NUCLEAR POWER PLANT ................................................................ 3-1 3.1 REACTOR DESCRIPTION ............................................................................................ 3-1 3.2 FUEL STORAGE DESCRIPTION .......................................................... :...................... 3-3 4 DEPLETION ANALYSIS ........................................................................ :................................... 4-l 4.1 DEPLETION MODELING SIMPLIFICATIONS & ASSUMPTIONS .......................... 4-1 4.2 FUEL DEPLETION PARAMETER SELECTION :................................... :.................... 4-2 4.2.1 Fuel Isotopic Generation ................................................................................. 4-2

, 4.2.2 Reactor Operation Panu:neters ............................ :**********:************** .....: ............ .4-2 4.2.3 Axial Profile Selection ..................................................................................... 4-7 4.2.4 B1,1rn~ble Absorber Usage ................................................................................ 4-9

- 4.2.5 Fuel Assembly Physical Changes with Depletion ...................,. ..................... 4-10 4.3 DESIGN BASIS FUEL SELECTION ........................................................................... 4-10 4.3.1 Fuel Design and Management Modelipg Considerations .............................. 4-10 4.3.2 Reactivity Comparison Methodology ............................................................ 4-11 4.4 -FINAL DEPLETION_PARAMETERS ..............................................................: ..... ,..... 4-13 5 CRITICALITY ANALYSIS ......................................................................................................... 5-1

- 5.1- KENO MODELING APPROACH, SIMPLIFICATIONS & ASSUMPTIONS .............. 5-1 5.1.1 Description of Fuel Assembly and Storage Racks for KEN0 ......................... 5-3 5.1.2 .Impact of Structural Materials on Reactivity ............................. :... _................ 5-5 5.2 KENO MODELING ANALYSrs: ............................................*..... ,................................. 5-6 5.2.1 Array Descriptions and Fuel Categories .......................................................... 5-6 5.2.2 Target kcirCalculation Description ................................................................ 5-1 l 5 .2.3 Bias & Uncertainty Calculations ................................................................... 5-11 5.3 INTERFACE CONDITIONS ............. -........................................................................... 5-26 5.4 NORMAL CONDITIONS '.............................*............... '. ................................................

. 5-27 5.4.1 Type 1 Normal Conditions ...................... :............................... :..................... 5-Tl 5.4.2 Type 2 Normal Conditions ......................................................................... :.. 5-27 WCAP-18414-NP Revision 0

iv 5.4.3 Type 3 Nonna! Conditions .........._..............................:...................... - .... ;.... _5-29 5.4.4 Type 4 Normal Conditions ...................................._....*................................... 5~30 5.4.5 Type 5 Normal Conditions ............................................................ :............... 5-31 5.5 SOLUBLE BORON CREDIT *******************************************************************-******************5-31 5.5.1 Soluble Boron Requirements for Normal Conditions .................................... 5-32 5.5.2 Soluble Boron Requirements for Accident Conditions .: ...................-...... :..... 5-33 5.5.3 Soluble Boron Requirements ................ :*-***************************-~*******-**************5-36 5.6 RODDED OPERATJON ... :...... :............. ,....................................................................... 5-37 6 ANALYSIS RESULTS & CONCLUSION ************************************************************=***-****************6-1 6.1 BURNUP AND IFBA REQUIREMENTS FOR smRAGE ARRAYS .......................... 6-1

- 6.2 ANALYSIS AREA OF APPLICABILITY .............................................................*.........6-9 6.3 SOLUBLE- BORON CREDIT .......................................................................................

6-13 7 . REFERENCES ............................................................................................................................. 7-1 APPENDIX A VALIDATION OF SCALE 6.2.3 ...................................................... :.....*........................ A-_l WCAP-18414-NP Revision 0

V LIST OF TABLES Table2-l Isotopes Used in the Nuclear Criticality Safety Analysis ................................................ 2-3 Table 3-1 Reactor General Specifications ................ ,..................................... , ................................. 3-I Table 3-2 Fuel Design M~hanical Specifications ...................**................... :......................... :........ 3-2 Table 3-3 Non-Mechanical Specifications and Operating History ................................................... 3-3 Table3-4 fuel Storage Rack Specifications ........... :...........................................................*............ 3-6 Table 3-5 Fuel Rod_ Storage Canister ..-.............................................................*............................... 3-7 Table3-6 -Loose Pellet Transport Canister ............................................................................. :.*....... 3-7 T$le4-I Cycle Average Soluble Boron Concentration (ppm) ........................................................ 44 Table4-2 Fuel Design Operating Power .......................................................................................... 4-6 Table 4-3 Burnable Absorber Specifications.................................................................................... 4-9 Table 4-4 Design Basis Comparison Modeled Fuel.Design Parameters ........................................ 4-11 Table4-5 Parameters Used in Depletion Analysis .............................................................*............ 4-1_3 Table4-6 Limiting Axial Burnup and Moderator Temperature Profiles .................... :................... 4-14 Table 5-1 IFBA Criticality Modeling Specifications .................................. '. .................................... 5-2 Table 5-2 Design 'Basis Fuel Assembly Design Modeling Parameters ............................................ 5-4 Table 5-3 Fuel Categories Ranked by Reactivity ............................................................................. 5-8 Table 5-4 Biases & Uncertainties forArray A with STD Fuel... ............................, ....................... 5-19 Table 5-5 Biases & Unce~inties for Array A with OFA Fuel...*******************:*****--*************************5-20 Table,5-6 Biases & Uncertainties for Array B with STD Fuel... ............-.................................... :.... 5-21

• I Table 5-7 Biases & Uncertainties for Array B for OFA Fuel ...........................................-.............. 5-22 Table 5-8 Biases & Uncertainties for Array C with STD Fuel... ............-.......................................-.5-23

.Table 5-9 Biases & Uncertainties for Array C with OFA Fuel... ..:................... :............................. 5-24 Table 5-10 Biases & Uncertainties for Array D with STD Fuel ......................................... .'............ 5-25 Table 5-11 JS.O ******************;***************'**************5-33 Table 5-12 ]a,c.......... ,.......... ., ......................:........... 5-33 Table 5-13 [ - ]"'-c ......................_.................................................... 5-34 Table 5-14 ]a,c ............................... :............................. 5-36 Table 6-1 Fuel Category 2: STD/RFA IFBA Fitting Coefficients .................................................... 6-2 Table 6-2 Fuel Category 2: Example STD/RFA IFBA Requirements(# oflF~A Rods) ................. 6-2 WCAP-18414-NP Revision 0

VI Table 6-3 Fuel Category 2: OFA IFBA Fitting Coefficients ********************************************************-**6-3 Table64 Fuel Category 2: Example OFA IFBARequirements (# of IFBA Rods) ......................... 6-3 Table 6-5 Fuel Category D: STD/RFA Bumup Requirement Coefficients ................. :... -******:********64 Table6-6 Fuel Category D: Example STD/RFA Bumup Requirements (GWd/MTU) ......~ ............ 6-4 Table6-7 Fuel Category 3: STD/RFA Bumup Requirement Coefficients ....................................... 6-5 Table6-8 Fuel Category 3: Example STD/RFA Bumup Requirements (GWd/MTU) .................... 6-5 Table 6-9 Fuel Category 3: OFA Bumup Requirement Coefficients ....................:......... .' ................ 6-6 Table 6-10 Fuel Category 3: Example OFABumup~equirements (GWd/MfU) ............................. 6-6 Table 6-11 Fuel Category 4: STD/RFA Bumup Requirement Coe:fficients ................... ;................... 6-7 Table 6-12 Fuel Category 4: Example STD/RFA Bumup Requirements (GWcL'MTU).:.................. 6-7 Table 6-13 Fuel Category 4: OFA Bumup Requirement Coefficients ............................................... 6-8 Table 6-14 Fuel Category 4: Example OFA Bumup Requirements (GWd/MTU) .......... :**.-***************6-8 Table 6-15 Criticality [ J""' ...................:.. ******:********************_:*****:******6-10 Table A-I Benchmark Values of k.,ff and Respective Uncertainties ................... :........ :: ... :............. A-22

.TableA-2 ]a.c ............. ~ .......... A-24 TableA-3 rc_ ................................................................. _................................. A-28 TableA-4 [

]a.c ********************************************************:***************************************; ................. A-33 TableA-5 [

]11,C ............................. ;*.****.**..***..*.*.***..*****...***.**...*.....*......*...-...*.*.*... A-37 Tabl.eA-6 [

]a." .......................:........................................:........................ ~***:******A-42 Table A [

1a.c .............................................  :..................................................... A-48 TableA-8 [

1a.c........................._. .................................................................... ,.................. A-4~

TableA-9 [

].,

0

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • : ****** A-49 Table A-IO [

]11,C ********************************--*******************************************:********* A-49 TableA-11

].,~ *:*********************************-**************************************************A-5_1 TableA-12 [ ]11,C ......-................................... A-54 WCAP-18414-NP Revision 0

vii TableA-13 [

]a.c ...............................................................................*................*.......*.....*. A-55 TableA-14 [ ]"'c .......... :.............. A-56 TableA-15 [ :r-c ................................................ A-58 TableA-16 [ ]a,c ..............*..................... A-59 TableA-17 [ ]a.c ....... ."............. A-60 TableA-18 Summary of Biases and Bias Uncertainties Detennination .......................................... A-62 WCAP-18414:-NP Revision 0

viii LIST OF FIGURES Figure 3-1 Farley SF];> (Unit I) LayouL ......................................................................... :.................... 3-4 Figure l-2 Farley SFP (Unit 2) Layout ..................................................................................... :........ 3-5 Figure 5-1 KENO Array Rack M~l Planar (x-y) View: Top Left Array A, Top Right Array B, Bottom Left Array C, Bottoni Right Array D (All M~ls: x-y Periodic Boundary Conditions)................, ....... :...................... ,........................... :....... ,................................... 5-3 Figure 5-2 Allowable Storage Arrays******************************:********** ........................ ,................................5-~

Figure 5-3 Currently Used Diunaged Fuel Assembly Configuration (Farley Unit 1)...................... 5-10 r

• Figure 5-4 [ ]a,c ..... ,*.... 5-17 Figure 5-5 [ r,c .............................. 5-26 Figure 5-6 *

• Schematic View of Modeled Failed Fuel Rod Storage Canister.'. ........................... :....... 5.:29 Figure 5-7 [ .. ]a.c..................*....................... 5-30 FigureA-1 . -[ ]a,c.*.................................... A-52 FigureA-2 [ ]a.c.'..........*........................... A-52 FigureA-3 [ ]a,c ..... .' .............. A-53 FigureA-4 [ j11.c ..*...*..***..**...**.**...* :*.* A-53 Figure A-5 [ ]a,c ........................ A-57 WCAP-18414~NP Rev_ision 0

IX LIST OF ACRONYMS, INITIALISMS, AND _TRADEMARKS 1-D One-Dimensional -

D Two-Dimensional 3-D Three-Dimensional AEG Average Energy Group ofNeutrons Causing Fission AoA Area of Applicability at°/o Atom Percent B&U Sum of Biases and Uncertainties B&W Babcock and Wilcox BA Burnable Absorber-BONAM! Bondarenko AMPX Interpolator Boraflex Neutron Absorber Material Comprised of Silicone Polymer and Boron Carbide Powder C.E.A. Commissariat 1r l'Energie Atomique et aux Energies Alternatives Decay time Post-ii:radiation cooling time EALF Energy of Average Lethfill?Y causing Fission En* Enrichment ENDF/B Evaluated Nuclear Data File -

EPRI Electric Power Research Institute FHE Fuel Handling Equipment FOSAR Foreign Object Search and Retrieval FRSC Fuel Rod Storage Canister GT Guide Tube GWd Gigawatt-days -

HTC Haut Tatix de Combustion ID Inner Dimension IFBA Integral Fuel Burnable Absorber ISG Interim Staff Guidance IT Instrumentation Tube kctr Effective neutron multiplication factor LWR Light Water Reactor MTU Metric Ton Uranium

.MWt Megawatts-thermal NPM Non-Parametric Margin NRC U.~. Nuclear Regulatory Commission ORNL Oak Ridge National Lab Optimized ZIRLO Optimized ZIRLo High Performance Cladding Material PNNL Pacific Northwest National Laboratory ppqi parts per million PWR Pressurized Water Reactor RCS Reactor Coolan~ System SFP Spent Fuel Pool SRSC Service de Recherche en Sfirete Criticite, now called Service de Recherche en Neutronique et Sfirete Critioite ss Stainless Steel WCAP-18414-NP Revision 0

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

STD Standard Fuel Assembly TD Percentage of Theoretical Density WABA Wet Annular Burnable Absorber Westinghouse Westinghouse Electric Company LLC wt% Weight Percent

• ZIRLO ZJRLQ'!> High Performance Fuel Cladding Material WCAP-18414-NP Revision 0

1-1 1 INTRODUCTION The-purpose ofthis report is to document the criticaJity safety analysis performed to support the operation of the J.M. Farley Nuclear Power Plant Units I and 2 (hereafter, Farley Units l & 2) spent fuel pools.

(SFPs). The report considers past; current, and planned future operating history and fuel design of Farley Units I'& 2.

The main. report details the SFP criticality safety analysis. Appendix A details the validation of the code used for pool eigenvalue C?Jculations.

\

WCAP-18414-NP Revision 0

2-1 2 OVERVIEW The existing SFP storage racks are evaluated for the placement of fuel within the storage arrays described in Section 52.1. Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time ( decay time) for assemblies which have been operated in the reactor. Fuel assemblies which have not been operated in the reactor may take credit for the presence of zirconium diboride (IFBA)

(hereafter referred to as IFBA). While the Farley Units 1 & 2 SFP storage racks may contain Boraflex absorber insel"ts, no credit is taken for the presence of Boraflex absorber. Additionally, credit is taken for the presence of soluble boron in the SFPs.

2.1 ACCEPTANCE CRITERIA This SFP criticality safety analysis ensures that the SFPs operate within the bounds discussed here.

1. 'f1:ie effective neutron multiplication factor (kctf) of all permissible storage arrangements at a

• soluble boron concentration of O.parts per million (ppm) shall be less than 1.0 including a margin for all applicable biases and uncertainties with 95 percent probability at a 95 pe~nt confidence level.

2. The kcJrof all permissible storage arrangements when crediting soluble bo~n shall yield results not exceeding 0.95, including a margin for all applicable biases and uncertainties with 95 percent probability at a 95 percent confidence level.

3. The ketfwhen crediting soluble boron shall not exceed 0.95 under all postul¢ed accident conditions, including a margin for all applicable biases and uncertainties with 95 percent probability at a 95 percent confidence level.

2.2 DESIGN APPROACH For the SFPs; compliance is demonstrated by esta~Iishing the minimum burnup requirements as a function of enrichment and decay time and minimum number of unirradiated IFBA rods as a function of enrichment for storage arrays A, B,.C, and D seen in more detail in Figure 5-2. The fuel storage arrays have been analy~d to determine separate burnup requirements for two major fue1 designs considered, the Standard Fuel Assembly (STD), and the Optimized Fuel Assembly (OFA). Note that the bumup requirements developed for the STD fuel design are applicable to the Robust Fuel Assembly (RF:A) design since their neutronically important characteristics are the same.

A conservative combination of best estimate and bounding values have been selected as input for .

modeling i1;1 this analysis to ensure that fuel represented by the proposed Farley Units 1 & 2 SFP storage Technical Specifications is less reactive than the fuel modeled for this analysis. Therefore, burnup *

. requirements generated here will conservatively bound all fuel to be ~ored in the Farley Units 1 & 2 SFPs.

The acceptability of the storage arrays developed in this analysis is ensured by controlling the assemblies that can be stored in each array. Assemblies are divided into Fuel Categories 1 through 4, and D (assemblies meeting the requirements of the damaged fuel array, Array D), based on assembly reactivity WCAP-18414-NP Revision 0

2-2 determined as a function of assembly average burnup, initial enrichment!, IFBA loading2, and decay time. An assembly's fuel category determines in which storage arrays it may be stored Fuel Category 1 defines the most reactive assemblies, i.e. fresh 5 weight percent (wt°/o) 235 U assemblies without IFBA and Fuel Category 4 defines the least reactive assemblies, i.e., representing l~w reactivity assemblies that can be stored in Array B (see Table 5-3).

• 2.3 COMPUTER CODES The analysis methodology employs the following computer codes and cross.:section libraries: (1) the two dimensional (2-D) transport lattice code PARAGON Version 12.0, as documented in WCAP-16045-P-A; "Qualification of the Two-Dimensional Transport Code PARAGON" (Referel)Ce I) and-its cross-section library based on Evaluated Nuclear Data File Version VI.3 (ENDF/B-VI.3), am;i (2) Scale Version 6.2.3, '

as. documented in ORNL/fM-2005/39, "Scale: A Modular Code System for Performing Standard Computer Analyses for Licensing Evaluation'? (Reference 2), with the ENDF/B-VII 238-group cross-section library. * *

-2.3.1 Two-Dimensional Transport Code PARAGON PARAGON is used in this application to simulate in-reactor fuel assembly depletion to generate isotopics for bumup credit PARAGON is the Westinghouse Electric Company LLC state-of-th~-art 2-D lattice transport code for p~urized water reactor (PWR) applications. It i s ~ ofth~ Westinghouse core design package and provides lattice cell data for three dimensional (3-D) core simulator codes.

~s data includes macroscopic cross-sections, microscopic cross-sections for feedback adjustments, pin factors for pin power reconstructio~ calculations, and discontinuity factors for a 3-D nodal method solution of the diffusion equation. PARAGON uses the collision probability theory within the interface current method to solve-the integral transJ><?rt equation. Throughout the calculation, PARAGON uses the

. l. .

exact. heterogeneous geometry of ¢.e 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 h~mogenization. The 70-group PARAGON cross-section library is based on the ENDF/8-VI.3 basic nuclear data It includes e:xplicit multigroup cross-sections and other nuclear data with.out any lumped fission products or pseudo cross-sections. PARAGON and its 70-group as

• cross-section library are benchmarked, qµalified, and licensed both *a standalone transpoi:t 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. The list of fuel isotopes modeled in PARAGON ~d subsequently modeled in the criticality analysis are given in Table 2-1.

1.

• lnitiaJ enrichment is the enrichment of the cen~ zone region of fuel, excluding axial cirtbacks\blankets and prior to reduction in 235 U content due to fuel depletion. If the fuel assembly contains axial regions of different 235 U enrichment values, such as axial cutbacks or low enriched blankets, the maximum initial enrichment value is to be used.

2 JFBA loading restrictions only apply to fresh fuel being stored as Fuel Category 2.

WCAP-18414-NP Revision 0

2-3 Table24 Isotopes Used in the Nuclear Criticality Safety Analysis a,c Additional qualification of PARAGON for use in spent fuel pool applications has been performed at* .

Westi.nghouse. The Electric Power Research Institute (EPRI) has developed fWR reactivity depletion beJ:ichmarks using a large set of measured flux data (flux maps) in EPRI Report 3002010613, "B~nchmarks for Quantifying Fuel Reactivity Depletion Uncertainty (Reference 3).

A guide for application of the EPRI depletion benchmarks for use in burnup credit calculations is given by way of example in Reference 4, with this methodology repeated.by Westinghouse in "EPRI Depletion Benchmark Calculations Using PARAGON" (Reference 5). Results of this analysis provide*additional confidence in the usage pf PARAGON for SFP reactivity calculations, and provide a sound basis for usage of the 5% decrement approach for depletion uncertainty (See_ Section 5.2.3.1.5), showing that the depletion isotopics generated with P~GON, and input ~to CSAS5 input models in Scale is.

conservative for detennining depletion uncertainty. * .

PARAGON is generically approved for depletion calculations (Reference 1). PARAGON has been chosen for this spent fuel criticality analysis because it has all the attributes needed for burnup credit applications.

There are no Safety Evaluation Report limitations for the use of PARAGON in UOz criti~lity analysis.

2.3.2 Scale Code Package The Scale system was developed for the U.S. Nuclear Regulatory Co~ssion (NRC) to .standardize the method of analysis for evaluation of nuclear fuel facilities and shipping packag~ designs (Reference 2). In WCAP~l8414-NP Revision 0

2-4 this SFP criticality analysis, the Scale code package is used to calculate the reactivity of fissile systems in SFP conditions. Specifically, the Scale package is used to analyze infinite arrays for all storage ~ys in the SFPs, finite rack modules and SFP representations to evaluate interfaces, soluble boron requirements, and postulated accident scenarios to demonstrate that the requirements in Section 2.1 are met.

The Scale package includes the control module Criticality Safety Analysis Sequence with KENO V.a

_(CSASS), which provides reliable and efficient means of performing kctrcalculations for systems that are routinely encountered in engineering practice, especially in the calculation of kerr of 3-D system models.

Updated structurally from prior versions, CSASS implements the modem material and cross section

• processing module XSProc to process material input and provide a temperature resonanctX:Orrected cross section library based on the physical characteristics of the problem being analyzed. XSProc calls several lower level functional modules, some of which perform simple functions that were not called out as separate from CS ASS in *past versions.

XSProc was developed for the Scale 62 release to prepare data for continuous-energy and multigroup calculations. XSProc.

expands

. material input from Standard Composition Library definitions. into atom

. number densities (calling the integrated MixMacros module) and, for multigroup calculations, performs cross section resonance self-shielding,* energy group collapse, and spatial homogeniz.a.tion. XSProc implements capabilities for problem-dependent temperature interpolation, calculation of Dancoff factors (calling the integrated Dancoffmodule), resonance self-shielding using Bondarenko factors with full-rarige intermediate resonance treatment, as well as use of continuous energy resonance self-shielding in the resolved resonance region. XSProc integrates and enhances the capabilities previously implemented independently in BONAMI, CENTRM, PMC, WORKER, ICE, and°XSDRNPM, along with some additional capabilities that were provided by MlPLIB and SCALELIB in prior Scale release. For thi~

work XSProc utilizes the following modules in addition to MixMacros and Dancoff (CENTRM and PMC are called via the CentrmPmc module):

• . BONAMI: The BONAMI module is used to perform Bondarenko calculations for resonance self-shielding. BONAMI obtains problem-independent cross sections and Bondarenko shielding factors from a multigroup (MG) AMPX master library, and it creates a MG AMPX working library of self-shielded, problem-dependent cross sections. Several options may be used to ,

c01:npute the background cross section values \!Sing the narrow resonance or intermediate resonance approximations, with and without Bondarenko iterations. A novel interpolation scheme is used that avoids many of the problems exhibite9 by other interpolation methods for the Bondarenko factors. BONAMI is most commonly used in automated SCALE sequences and is fully integrated within the Scale cross section processing module, XSProc. During_the execution*

of a typical Scale computational sequence using XSProc, Dancoff factors for uniform lattices of square- or triangular-pitched- units are calculat~ automatically for BON AMI by numerical integration over the chord length distribution. Heterogeneous effects are treated using equivalence theory based on an "escape cross section" for arrays of slabs, cylinders, or spheres.

• CENTRM: CENTRM computes continuous-energy neutron spectra for' infinite media, 1-D) systems, 2-D unit cells in a lattice, by solving the Boltzmann transport equation using* a combination ofpointwise and multigroup nuclear data. CENTRM is primarily used to calculate .

problem-specific fluxes on a fine energy mesh to generate self-shielded multigroup cross sections WCAP-18414-NP Revision 0

2-5 for subsequent radiation transport computations. Several calculation options are available, including a slowing-down computation for homogeneous infinite media, 1-D discrete ordinates in slab, spherical, or cylindrical geometries; a simplified two-region solution; and 20 method of characteristics for a unit cell within a square-pitch lattice.

• PMC: PMC ge~rates problem-dependent multigroup cross-sections froin,an existingAMPX multigroup cross-section library, a point wise nuclear data library, and a pointwise neutron flux f~e produced by the CENTRM continuous-energy transport code. In the Scale sequences, PMC is used:primarily to produce self-shielded mnltigroup cross-sections ovei;- a specified energy range such as the resolved resonance energy range of individual nuclides in the system of interest The self-shielded cross-sections are obtained by integrating the point wise nuclear data using the CENTRM problem-specific, continuous-energy flux as a weight function for each spatial zone in the system.

• KENO: The KENO module is a Monte Carlo criticality program used to calculate the kcff' of 3-D models ~ing continuous energy or multigroup cross-sections and is called by CSAS5 once .

XSProc is complete. Flexible geometry features and the availability of various boundary condition prescriptions in KENO allow for accurate and detailed modeling of fuel assemblies in storage racks, either as infinite arrays or in actual SFP models. The version used in this w_ork, KENO V.a, contains a simplified geometry package appropriate for use here. Anisotropic scattering is treated by using discrete scattering angles using Pa Legendre polynomials. KENO uses problem-specific cross~section libraries, processed for resonance self-shielding and for the thermal characteristics of the problem.

\

For this work, the option panp=centrm is used as input, for which the CENTRM/PMC modules are executed to.process shielded multi-group cross sections using continuous energy flux spectra calculated w~th the recommended type of continuous energy transport solver for the designated type of cell. An infinite homoge_neous medium calculation is used for those materials not called out for special processing, uses 2-D Method of Characteristics for a LATIICECELL consisting of cylindrical fuel rods in a square lattice, and uses 1-D discrete Sn transport for all other LATIICECELLs and.MULTIREGION cells.

The criticality sequence of Scale 6.2.3 is validated using fresh UOi critical e!'periments and Haut Taux de Combustion (HTC) critical experiments to form an experiment benchmark suite applicable to fresh and

,spent fuel criticality calculations. See NUREG/CR-6979, "Evaluation of the French Haut_Taux de Combustion (HTC) Critical Experiment Data" (Reference 16) for an overview. of the HTC criticals.

Additional details of the validation are found in Appendix A. The validation shows that Scale 6.2.3 is an accurate tool for calc~~tion of k,,fffor SFP applications. The benchmark calculations use the same computer platform and cross-section libraries that are used for the design basis calculations, The validation considers both fresh U02 and fuel with plutonium designed to have an actinide composition similar to burned fuel.

2.3.3 Scale 238 Group Cross-Section Library The 238-group ENDF/B-VII library included in the Scale package is available for general purpose criticality analyses. The group structure is the same as the 238-group ENDF/B-V and ENDF/B-VI WCAP-18414-NP Revision 0

2-6 libraries in Scale, arid the same weighting spectrum as for the ENDF/B-VI. As with the 238-group ENDF/B-VI Jibrary, the ENDF/B-VII library cannot be used with the NJTAWL-III module for resonance self-shielding calculations in the resolved range.

The 238-group and continuous-energy ENDF/B-Vfl libraries have 417 nocJjdes that include 19 the~al-scattering moderators. The validation of the ENDF/B-Vll 238-group library with the ScaJe Version 6.23 CSASS module is documented mAppendix A.

WCA.P-18414-NP Revision 0

3-1 3 FARLEY UNITS 1 & 2 NUCLEAR POWER PLANT -

This section describes the physical characteristics

. of Farley Units

. I & 2 that are important to SFP criticality safety. Pertinent reactor characteristics and associated fuel design and fuel management history are discussed in Section 3.1. The physical c~stics of the SFPs are discussed in Section 3.2.

3.1 REACTOR DESCRIPTION 1be. Farley Units I & 2 Nuclear Power Plant is a Westinghouse PWR utilizing fuel with . a 17 x 17 lattice.

Farley Units I & 2 have used multiple fuel designs from Westinghouse. All fuel assemblies used at Farley Units I & 2 incorporate a 17 x 17 square array of 264 fuel rods with 24 guide tubes (GT) and I instrument tube (IT). The fuel rod cladding material is Zircaloy cladding and its variants, such as ZIRLO High Performance Fuel Cladding Material. Each fuel rod contains a column of enriched UOi fuel pellets.

The pellets are pressed and sintered, and are dished on both ends.

\

Section 3.1 provides data on the design and operation of Farley Units 1 & 2 as*well as the fuel designs arid fuel management of the plant Table 3-1 provides basic data on the type of reactor and tbe fuel types .

that comprise Farley Units I & 2. The neutronically important mechanical features of the three fuel designs are listed in Table 3-2. *

• Table3-1 Reactor General Specifications Reactor type Westinghouse Historic & current reactor power 1 (MWt) 2652-2775 Fuel lattice- 17 X 17 1 Fuel design 1 \Yestinghouse Standard Fuel Assembly Fuel design 2 Westinghouse Optimized Fuel Assembly

- Fuel design 32 Westinghouse Robust Fuel Assembly Note:

I. Reactor power will be analyzed in this work up to 2831 MWt with current fuel management strategy to address future operation. *

2. The RFA fuel design has not been used at Farley, and there are no current plans to transition to this fuel design. However, it i~ included in this analysis to support potential future use provided the RFA design is operated within the analysis area of applicability.

WCAP-18414-NP Revision 0

3-2 Table3-2 Fuel Design Mechanical Specificatiqns Assembly type STD RFA OFA Rod array siz.e 17 X ]7 17 X 17 17 X 17 Rod pitch, in 0.496

• 0.496 -0.496 Active fuel length; in ' 144 144 144 Total number of fuel rods 264 264 264 Fuel cladding outer dimension (OD), in 0.374 0.374 0.36 Fuel cladding inner dimension (ID), in 0.329 0.329* 0.315 Fuel cladding thickness, in

• 0.0225 0.0225

• 0.0225 Pellet diameter, in 0.3225 0.3225 0.3088 NumbeF of GT/IT 24/1 24/1 24/1 GT/ITOD, in 0.482 0.482 0.474 GT/IT ID, in* 0.450 0.442 0.442 Percent theoreticaJ density, nominal 95.0-96.5

• 95.0-96.5 95.0-96.5 Non-mechanical fuel features which are important to criticality safety 'and how they impact the number of distinct fuel designs are considered in _this analysis. Operational characteristics of every cycle operated at Farley Units I & 2 were reviewed. ATI cycles can be categorized conservatively into one of the following Criticality Fuel Designs. Table 3-3 outlines the key non-mechanical features and fuel man~ement h*istory of each of the fuel designs. ,

WCAP-18414-NP Revision 0

3-3 Table3-3 Non-Mechanical Specifications and Operating History Criticality Fuel 1 2 3 4 5 6 7 Design Assembly type STD STD STD OFA OFA OFA OFA Max. TD 1 96.5  %.5 98.0 96.5 96.5 .96.5 98.0 MaL operating 2652 2652 2831 2775 2652 2775 2831 power,MWt Annular, __ Annular, Axial blanket No No No No No Fully Fully enrichment Blanket Blanket Blanket Blanket Blanket Enriched Enriched Axial blanket length, NIA NIA 6 NIA NIA . NIA 6 in Burnable absorber IFBA/ IFBA/

Pyrex WABA

• IFBA IFBA IFBA (BA)Type WABA

• WABA BAmat~i~I Bz03-Si0i *&c ZrB2 ZrB2/RtC ZrBifB4C ZrBi ZrB2 6.03 1.00X I 6.03 l.OOX/ 6.03 BA maL loading 12.5 wt%, 1.50X 1.50X I.SOX mg 10 B/cm mg 1°B/cm mg1°B/cm IFBA: 132 IFBA: 132 Max BA length, in 144 134 132 132 132 WABA:132 WABA:134 Maximum number of 24 20 200 156 I 8 104/ 12 156 200 rods/ fiQge_rs Note:

I. Percentage of Theoretical Density (TD).

2. The Max. TD for Criticality Fuel Designs 3 and* 7 are chosen to bound potential future operation and are greater than current experienced at Plant Farley as seen in Table 3-2.

Criticality Fuel Design 3 and 7 are selected as the design basis fuel designs for STD/RFA and OF A fuel designs respectively. See Section 4.3.2 for methodology details for determination of the design basis criticality fuel designs.

3.2 FUEL STORAGE- DESCRIP1;10N The physical characteristics of the Farley Units 1 & 2 SFPs. are described in this section. The SFPs are made up of one fuel storage rack design (region). The Farley Units 1 & 2 SFPs each consist oftw_o 6 x 7, nineteen

. 7 x 7, and seven 7 x 8 storage racks. The storage racks are .of flux trap. style with an uncredited Boraflex neutron absorber panel on every side (in the x and y-axis directions) of each storage cell. This results in a flux trap betweeri any two assembly storage locations. A schematic layout of the unit I and unit 2 SFP is given in Figure 3-i and Figure 3-2, respectively. See Section 5.1.1 for modeling details. The specifications for the storage racks are given in Table 3-4.

WCAP-18414-NP Revision 0

3-4 45 f?' )""li-,- - - - - - - - - - * - - - - 1...

5.6'lS- -j6

) '*

t C.B75" 7-x a 7 X 8 7 X 8 7 X '8 7 X 'I 7 XII '! X 8 7 X 7 7 X 7 7 X 7 *7 X 7 7 ,r 'J ., .* 7 27Ft.

I!. 1 x 7 1 x 1 1 x 7 '1. x 7 1 x *1 *1 x ,. _, x 7 *:a. 75*

t---+-----+---+------t---t--------r-""T"'-! Li L,==7=x=7===*=*7=:r.=7===7=x=1=*==*7=x=7===7=ll=7=;;;:!;=6=x=7===6=x::7:::;~-*~2*

4,875' q

.J .-1.111e* Typicd c;ap Benraen Top l.**e-1j

-' 24"

,~

. 1>! Jl&,;kj

• 5/ 8" . .... ~

~~ ntv. OW!HAa -~ *N* .

6 It, 7 2* S4 7 X 7 19 9ll

• 10 7"' a ..J_, ~ * [tGtll!K 2-1 [o~

TOT1,.L H 1407 ,8Pl!Jll' eogi. S'TOJUIG!: - M M . ~

.FARLEY I

. IIUCLEAA i'UllT lmIT. 1.

F'igu.re 3-1 . 'Farley SFP (Unit 1) Layout WCAP-18414-NP Revision 0

3-5 1 1' 7 1 ,c 7 7 x .., 1 1t 7 7 :a: 7 1

• 1. ' ,c 7 27 Ft.,

I I 9.7~ 1 111 7. 7 IC 7 1. ic* 7 7 lC .7 7 :it 7 7 X 7 1. 1t 7 I

L r~

1L_ ~ X? 6 X 7 7 11 l ,- x 7 1 ,c  ? 7 l\ 7 7 it 7 *,e,s*

. _j k -

-+/- ,.1Bs*

a

... typ!~L ot~it -

o;,P Rll!tln)nn 'l'6p s1a*

P'!COOE l*Z

• *:).l~"-+

11N ..  :

• f

,,: I 8

, x,7 s l'Ja!T FURL S'l'O M.CI: IJliW<<lR.l(lf!IT . *t 7 :,c 7 19 l'AllLSY IWCl.,f:'i\l\ PI..UT U!liT 2 jo ~

, ,r. 8 ,.!_  !!!_ ..,*',- .

.. ,TO'?,\L 28 H07 Figure 3:"2 Farley SFP (U~t 2) Layout WCAP-18414-NP Revision 0

3-6

\

Thble3-4 Fuel Storage Rack Spec~tions Value Tolerance Cell pitch, in 10.75 +/-0.06 Cell ID, in 8.9 +/-0.045 Cell wall thickness,-in 0.12 . +/-0.012 BA1 Type Bora.flex NIA BA cavity width, in 8 +/-0.06 BA cavity thickness, in 0.07 NIA BA wrapper thickness, in 0.024 +/-0.003 Note:

I. No credit is taken for the presence of any residual Boraflex. The BA cavity is assumed to be fill~

.with water ofth_e,same composition as the water elsewhere in~ storage racks.

Also present in the SFPs are fuel rod storage canisters (FRSCs) and loose pellet transport canisters (LPTCs). The fuel rod storage canister at Farley Units I & 2 is a rectangular lattice of storage tubes for fuiled fuel rods arranged in an 8 x 8 pattern. Design details of the FRSCs are given in Table 3-5. Section 5.4.3 contains additional modeling details.

The loose pellet ~sport canisters (LPTC~) are stainless steel (SS) canisters designed to st~re up to 5000 loose fuel pellets within individual loose pellet canisters stored within the LPTC:s. Design details are given in Table 3-6. Section 5.4.3 contains additional modeling details .

. (

WCAP-18414-NP Revision 0

3-7 Table3-5 Fuel l;{od Storage Canister 52 rods with fresh Maximum rod loading of FRSC

.. 5 wt°/o 235 U loadfug Length of FRSC, in 155.75 Lattice of storage locations 8x8 Fuel rod storage tubes per row 4, 6, 8, 8, 8, 8, 6, 4 Fuel rod storage ~be OD, in 0.750 Fuel rod storage tube thickness, in 0.120 Fuel rod storage tube material SS-304 (not m9(ieled)

Fuel rod pitch, in 0.937 ..

Table3-6 Loose Pellet Transport Canister Maximum Loading of LPTC

• 5000 pellets at 5 wt%,, no burnup Length of LPTC, in 258 Outer shell material SS-304. _

Outer shell thickness, in 0.375 Loose pellet canister outer dimensions, in 7x5 Loose pellet canister material SS-3_04 Loose pellet canister SS thickness, in 0.078125 WCAP-18414-NP Revision*o

4-1 4 DEPLETION ANALYSIS This section describes the methods usect' to determine the conservative and bounding inputs for the generation of isotopic number densities, which are then used in subsequent Monte Carlo simulations.

[

4.1 DEPLETION. MODELING SIMPLIFICATIONS . & ASSUMYTIONS There are several different combinations of fuel oesigns including"differing mechanical designs, operating conditions, and BA types that need to be considered when performing the analysis. To facilitate the analysis, two bounding design basis fuel asse~bJy types are determined, one for fuel designs with a nominal rod outer diameter of 0.374 inches (STP) and one with a nominal rod outer diameter of 0.360 inches (OFA).

['

]a,c

• Depletion isotopics for use in the Criticality Analysis are generated every 2000 MWd/MTU.*

WCAP-18414-NP

• Revision 0

4-2 4.2 FUEL DEPLETION PARAMETER SELECTION 4.2.1 Fuel Isotopic Generation This section outlines how parameters are selected for use in the fuel depletion calculations to generate isotopic number densities. For the purposes of this analysis, the isotopic number densities generated are differentiated by fuel design, fuel enrichment, burnup, and decay time after discharge.

1a.c Based qn the Farley Units I & 2 fuel management, the fuel has isotopic number densities which are.

calculated at enrichments of 3, 4, and 5 wt%, 235 U and decay times of 0, s*, 10, 15, and 20 years. Fresh fuel modeled in this analysis conservatively excludes 234U and 236U.

4.2.2 Reactor Operation Parameters The reactivity of the depleted fuel in the SFP is determined by the in-reactor depletion conditions. The conditions* experienced in the_ reactor impact the isotopic composition of fuel being discharged to the SFP.

NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnap Credit for LWR Fuel" (Reference 6) provides discussion on the core operation parameters important to SFP criticality.

NEI-12-16, Revision 3, "Guidance for Performing Criti~ity Analyses of Fuel Storage at Light Water Reactor Power Plants" (Reference 14) provid~ practical guidance for criticality safety analysts in line with current recommendations. This section outlines the parameters used in generating the fuel isotopics and why they are appropriate for use in this analysis. The* operating conditions of the fuel selected for modeling are prov!ded in Table 4-5, which provides both the nominal values and the values assumed in the analysis.

4.2.2.1 Soluble Boron Concentration The soluble boron concentration in the reactor during operation impacts the reactivity of fuel being discharged to the SFP. Because boron is a strong thermal neutron absorber, its presence hardens tbe neutron energy spectrum in the core, creating more plutonium.

Based on guidance from Reference 6, "establishment of a bounding value for the maximum average boron per cycle based on boron let-down curves woulµ enable more straightfornrard application of the depletion analyses," a constant cycle average soluble boron concentration (Equation 4-1) which assumes 19.9 at°/o 10 B in place of a soluble boron-letdown curve is considered appropriately conservative. To determine the maximum cycle aver~e soluble boron concentration, fuel management strategies for WCAP-18414-NP Revision 0

4-3 Farley Units I & 2 have been reviewed Table 4-1 provides the cycle average soluble boron concentration i~,

for cycles I through 29 of Unit I and for cycles I through 26 of Unit 2.

[ Equation 4-1

]a,c WCAP-18414-NP Revision 0

4-4 Table4-1 Cycle Average Soluble Boron Concentration (ppm)

Cycle# Unitl Unit2 Cycle I 567.3 547.5 Cy~le2 491.6 482.6 Cycle3 507.8 653.3 Cycle4 4933 703.7 Cycle 5 461.3 685:~

Cycle 6 669.6 806.7 Cycle 7 783.0 769.3 Cycle 8 720.J 724.3 Cycle 9 777.1 851.7 Cycle IO 777.2 694.1 Cycle 11 789.9 856.0 Cyclel2 762.5 807.9 Cycle 13 754.4 722.3 Cyclel4 828.0 684.8 Cycle-15 771.6 783.7 Cycle 16 714.8 767.1 Cycle* 17 755.9 800.7 Cycle 18 785.3 771.1

. Cycle 19 797.0 803.9 Cycle 20 785.3 751.1 Cycle 21 783.9 801.1 Cycle 22 765.4 805.4 Cycle 23 783.2 777.5 Cycle 24 746.0 768.2 Cycle 25 786.0 798.8 Cycle 26 777.9 788.9 Cycle 27 771.3 N/A Cycle 28 787.1 N/A Cycle 2~ 780.6 N/A WCAP-18414-NP Revision 0

4-5 4.2.2.2 FoeJ Temperature The fuel temperatur~ during operation impacts the reactivity of fuel being discharged to the SFP.

Increasing fuel temperature increases resonance absorption in 238 U due to Doppler broadening which leads to increased plutonium production, increasing the reactivity ofthe,discharged fuel. Therefore, utilizing a higher fuel temperature is more conservative.

The temperature input for this analysis is calculated by the F1GHTH code documented in Westinghouse WCAP-9522, "FIGHTH - A Simplified Calculation of Effective Temperatures in PWR Fuel Rods for Use in Nuclear Design" (Reference 7), which determines the fuel temperatures used as input to PARAGON

• for depletion calculations. FIGHTH calculates the steady state radial temperature distribution at each burn up, given the local value of the heat generation rate in the rod, the moderator temperature, and coolant flow rate. The FIGHTH model accounts for radial variations of the heat generation rate, thermal conductivity, themial expansion in the fuel pellet, elastic deflection for the cladding, and ~llet-clad gap

• conductance. The FIG~TH code is used in the development of cross-sections for in-core calculations as part of the standard reload methodology.

As discussed, the important input parameters used by FIGHTH for determining fuel temperature are power level, moderator temperature, and coolant flow rate. [

]a,c Selection of moderator temperature is performed as discussed in Section 4.2.3.2: [

]"-c 4.2.2.3 Operating History and Specific Power The analysis assumes c\mstant full power operation consistent with a bounding assembly average power.

For fission product credit analyses, the conservative direction for specific power '{aries with bumup (see Reference 6). However, assuming a bounding assembly average power (therefore high specific power) ensures high fuel temperatures which is consery~tive throughout life. Interim Staff Guidance (ISG)

DSS-ISG-20 I 0-00 I (Reference 8) 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 consisten! with sensitivity study results reported in Reference 6, the fuel temperature impact on reactivity is greater than the impact from specific power. Guidance in Reference 14 corroborates this assessment This makes the selectiori of a high operating power, and therefore specific power to maximize fuel temperature, appropriate as the subordinate parameter is more conservative than nominally chosen and the 0.002 & bias recommended for a bounding treatment in Reference 6 is unnecessary. For additional conservatism, a 0.002 L\k uncertainty is taken on*operational history.

WCAP-18414;-NP Revision 0

4-6 4.2.2.4 Maximum Average Assembly Power

]""'

Table 4-2 . . Fuel Design Operating Power I ~alue. a,c WCAP~18414-NP

- Revision 0

4-7 4.2.3 Axial Profile Selection This section discusses t}_le selection of bounding axial bumup and moderator temperature profiles. [

].,.,

4.2.3.1 Axial Burnup Profile Selection This section describes the methods used to determine the limiting di~bllre9 axi~ burnup profiles. These pro~les will be used along with the uniform axial burn up profile as .one of the conservative i~put parameters to develop isotopics to ultimately calculate the minimum bumup requirements provided in Section 6.1.

As discussed in NUREG/CR-6801, "Reco~ndations for Addressing Axial Bum~p in PWR Bumup Credit Analyses" (Reference 9), as fuel is operated in the reactor, the axial center of each assembly genera~ more power than the ends. This leads to the bumup of each assembly varying along its length.

Because the axial center of each assembly generates most of the power, the burn up .in the axial center of the assembly is greater than the assembly average. At the same time, the ends of the assembly are less burned than the assembly average. When the burnup difference between the axial center and end of an assembly is large enough, reactivity becomes driven by the end of the assembly rather than the axial center, as the unde! depletion of the ends (the end-effect) overcomes the reactivity lo~ due to neutron leakage.

As driven by the end-effect, the following methodology was used to ensure that the_ appropriate axial bumup profiles were selected for this analysis. Fuel management calculations containing readily available data from 25 cycles of operation were utilized to develop a database of axial burnup profiles specific to Farley Un.its 1 & 2. [

WCAP-18414-NP Revision 0

4-8

[

]a,c, For the reasons discussed abov~ it is typical for fuel modeled assuming a uniform axial burnup profile to be more reactive ~ly i~ life than fuel modeled with a distributed profile. To address this, isotopics were created for fuel assuming-both a tmiform profil~ and distributed profile (for the design basis fuel). These isotopics were used during the Monte Carlo calculations to determine the minimum burnup requirements to ensure the limiting profile (uniform vs design basis distn"buted) has been used.

4.2.3.2 Axial Moderator Temperature Profile Selection This section describes the methods used to determine the limiting axial moderator temperature profiles.

These profiles will be used together with axial. distributed and uniform burnup profiles to calculate the isotopics used in generating the burnup requirements provided in Section 6.1. .

as Selecting an appropriate moderator temperature profile is important it impacts the moderator density and therefore the neutron spectrum during depletion as djsc~ in Reference 6. An appropriate moderator temperature ensures the impact of moderator density on the neutron spectral effects is bounded, conservatively biasing the isotopic inventory of the fuel.

]ll,C WCAP-18414-NP Revision 0

4-9 4.2.4 Burnable Absorber Usage Burnable absorber usage at Farley Units I & 2 has been considered for this analysis and conservative assumptions have been used to bound the effects ofBAs on fuel isotopics. The BAs that have been evaluated include both discrete and integral BAs. The BA rod parameters are shown in Table 4-3.

.. Table4-3 Burnable Absorber Specifications Parameteri Pyrex WABA IFBA BA material Bz()3-SiO:z Ah(n-B,iC ZrB2 BA type Discrete Discrete Integral B4CTO, % NIA 70 NIA Boric Oxide Content, wt°/o 12.5 NIA NIA 1 2.35 (STD/RFA) I

°8 abundance or loading 19.9 at% 19.9 at°/o 2.25 (OfA) mg/in BA thickness, in 0.073 0.02 0.00022 BAJD, in 0.1900 0.2780 NIA BAOD,in 0.3360 0.3180 NIA BA clad material Stainless Steel Zirc-4 NIA

_BA inner clad 9D, in 0.1810 Q.2670 NIA BA inner clad thickness, in 0.0070 0.0210 NIA BA outer clad OD, in 0.3810 . 0.3810 NIA BA outer clad thickness, in 0.0185 0.0260 NIA BA length, in I See Table 3-3 See Table 3~3 NIA Max. BA exposure~ MWd/MTU 3 40000 40000 NIA Notes:

1.) Additional BA information is contained in Table 3-3 for each Criticality Fuel Design considered.

The maximum BA length and loading are modeled for each Criticality Fuel Design.

2.) Coating on the fuel pellet IFBA depletion input captures the desired absorber per unit length with 0.2 mils coating thickness. Specific criticality analysis input is given in Table 5-1.

3.) Pyrex and WABA exposure is conservatively modeled to 40000 MWd/MTU. IFBAresidual 10B is removed from the fuel for spent fuel pool criticality calculations.

Criticality Fuel Designs 2, 4, and 5 contain WA13A. The WABA length and number ofWABA rodlets present will impact the final assembly reactivity, as they are directly related to the amount and location of absorber presenfwithin the assembly. For past operation with these designs, a minimum of a 5" cutback is observed, however less conservative 6" cutbacks are utilized in the depletion calculations. The maximum number of WABA rodlets used for past operating cycles falling within these designs is 20 yet 24 rodlets were analyzed. The additional rodlets ensures any reactivity impact lost from the additional inch of WCAP-184 I 4-NP

. Revision 0

4-10 WABA during operation is accounted for. For conservatism, discrete BAs (Pyrex and WABA) were not removed from the core until 40 GWd/MTU of operation.

4.2.5 Fuel Assembly Physic~.I C~nges with Depletion Reference 14 discusses fuel assembly" physical changes with depletion, and specifically calls out the need .

to address the potential reactivity impact ~m fuel rod changes (clad creep, fuel densification/swelling) arid material dependent grid growth. Appendix B of Reference 14 is based on Westinghouse methodology and indicates that holistically the impact of fuel rod changes with depletion are conservative. Fuel assembly grid growth impact has been shoym to be negligible during depletion and is conservatively addressed for the in-pool storage reactivity impact in Section. 5.2.3.1.11.

43 DESIGN -BASIS FUEL SELECTION 4.3.1 .Fuel Design and Management Modeling Considerations

• To develop conservative storage requirements Jor the Farley Units 1 &*2 SFPs, the different unique fuel designs (criticality fuel designs) and the conditions in which those designs were operated or are planned to be operated were considered. All the criticality fuel designs considered are discussed in Section 3.1. [

. ]11,c WCAP-18414-NP Revision 0

4-11 Table4-4 Desi~ Basis Comparison Modeled Fuel Design Parameters Fuel Design Parameter STD/RFA OFA )

Rod pitch, in 0.4% 0.4%

Active fuel length, in 144 144 Total number of fuel rods* 264 264 Pellet OD, in *0.3225 0.3088 Clad OD, in 0.374 0.360 Clad ID, in 0.329 0.315 Number of GT/IT- 24/1 24/1 f

GT/ITOD, in 0.482 - 0.474 GT/IT ID, in 0.442 0.442 -

Notes:

I. Blanket/Cutback, BA type and loading information is given in Table 3-2

2. RFA fuel model GT ID is used. Section 6.2 discusses the applicability of the RFA fuel design.

[ '

]11,C At Farley Units 1 & 2, the STD/RFA and OFA fuel designs have-used the Pyrex, WABA and IFBA BAs.

As discussed in Nl)REG/CR-6761, "Parametric Study o_f the Effect of Burn_able Poison Rods for the PWR Burnup*Credit" (Reference 11 ), the presence of discre~e burnable absorbers such as Pyrex and. WABA

• displace water and absorb thermal neutrons, thereby hardening the neutron spectrur:n and creating more plutonium isotopes. Therefore, Pyrex and WABA must be modeled in the depletion analysis. As discussed in NURE<J/CR-6760, "Study of the Effect oflntegral Bu~able Absorbers on PWR Burnup Credit

(Reference 10), the presence of integral absorbers during depletion hardens the neutron spectrum, resulting in lower 235 U depletion and higher production of plutoniu~ isotopes. As a result, the IFBA integral absorber must also be modeled in the depletion analysis for determination of the bounding criticality fuel design.

4.3.2 Reactivity Comparison Methodology The final bounding assembly design is the determination of a limiting combination of fuel type and conservative depletion input parameters, denoted as a criticality fuel design. The design basis co*nservatively _covers past, current, and expected future spent fuel operation for Farley Units 1 & 2. This WCAP-18414-NP Revision 0

4-12 section outlines the methodology used to detennine a bounding assembly design, including the selected criticality fuel 'type. For this analysis,-a bounding fuel type is determined for the STD/RFA and OFA fuel designs. The potentially limiting a.'Cial burnup profiles identified using the methodo1ogy approach described in Section 4.2.3.1 are implemented together with the *limiting depletion parameters for each Criticality Fuel Design.

Reactivity comparisons w~re performed across all burnup bins at 3, 4! and 5 wt°/o mu _wi~ all potentially limiting axial burnup profiles and the limiting modern.tor temperature profiles-for each criticality fuel.

design with STD/RFA and OFA fuel ~blies. Comparisons were performed for both Array B an.d Array C (See Section s.2: I for a description of storage arrays). Based on the reactivity comparison Criticality Fuel Design I was chosen as the limiting STD/RFA criticality fuel design, and-Criticality Fuel Design 7 was chosen as the limiting OFA criticality fuel design. Note that when necessary, reacti'(ity .

comparisons focused on a range of reactivity of interest to spent fuel pool criticality to ensure that the limiting design was appropriate. The two largest differences were 0.00058 ~k at 30 GWd/MTU and 0.00031 & at 35 GWd/MTU for Array B comparisons of Criticality Fuel Design 7 with 3 wt°/o 235 U fuel with two different potentially limiting axial ~umup profiles. -

Criticality Fuel Design I was selec,ted as limiting over Criticality Fuel Design 2 and 3 for SID (RFA) fuel. It is the combination of Criticality Fuel Design 1 input (including the burnable absorber usage during operation) thadeads to the:overa1'1 bounding n ~ despite the lower fuel percent of theoretical density (96.5 vs 98.0) when comp,ared with Criticality_Fuel Design 3. See Section 6.2 for the analysis area of appli~ility. *. * '

WCAP-18414-NP Revision 0

4-13 4.4 FINAL DEPLETION PARAMETERS This section outlines the parameters used in the final depletion calculations. The depletion parameters discussed in this section are:

• Core Operation Parameters

• Fuel Assembly Dimensions

• Axial Bumup and Moderator Temperature Profiles The ~I isotopics used in the reactivity calculatio~ were generated based on the data presented in Table 4-4, Table 4-5, and Table 4-6.

Table 4-5 Parameters Used in Depletion Analysis Parameter Nominal Values Depletion Analysis Maximum cycl~ average soluble boron ]a,c 461.3 - 856.0 [

concentration, ppm lRated thermal power', MWt 2652 - 2775 [ ]""'

Average assembly po~er, MWt 16.90- 18.04 2 [ ]&,C Soluble boron 1

°13 atom percent, % 19.9 [ ]a,c Minimum core lpading, kg U 72443 (STD), 66417 (OFA) [ ]&,C System pressure, psia 2250 [ ]a,c Core outlet moderator temperature, °F 618.9 (max) [ f'c Core inlet moderator temperature, °F 541.1 (max)_ [ ]&,C Minimum RCS flow rate (thermal design ]a,c 86000 [

flow), gpm/coolant pump Fuel designs STD/RFA, OFA [ ]a,c Fuel assembly cutback/blanket region See Table 3-3 [ ]a,c

[

Blanket type See Table 3-3 ]a,c TD ,- 94.0-%.5 [ ]a,c BA Pyrex,,WABA, IFBA [ ]a,c Max BA lengths~ in 144 (STD), 134 (STD/OFA), [ ]a,c 132 (SID/OFA)

Notes:

1. The current rated thermal power is 2775 MWt, with a proposed uprate power of 2831 MWt.

2. This number is calculated by dividing the rated thermal power by the number offuel assemblies.

3. [ '

]..c

4. See Section 6.2 for overall applicability offuel designs covered by the analysis area of applicability.

WCAP-18414-NP Revision 0

4-14 Table4-6 Limiting Axial Burn op and Moderator Tern peratore Profiles a,c WCAP-18414-NP Revision 0

5-1 5 CRTIJCALITY ANALYSIS This section describes the reactivity calculations and evaluations perfonned in developing the bumup requirements for fuel storage in the Farley Units I & 2 SFPs. The section also confinns continued safe SFP operation during both normal and accid~t conditions.

5.1 KENO MODELING APPROACH, SIMPLIFICATIONS & ASSUMPTIONS As discussed in Section 2.32, KENO is the criticality code used to support this analysis. KENO is used to determine the absolute reactivity of burned and fresh fuel assemblies loaded in storage arrays.

Additionally, KENO is used to detennine the reactivity sensitivity of these storage arrays to effects sµch as manufacturing tolerances, fuel depletion, eccentric positioning., and the allowable temperature range of the SFPs. KENO is also used to model accident scenarios and confirm there is sufficient soluble boron to

-meet the requirements of Section 2.1.

The methods used to model the fuel in normal and accident scenarios are discussed in the following

.sections. [

].a,c .

WCAP-18414-NP Revision 0

5-2 Table 5-1 IFBA Criticality Modeling Specifications Parameter *I 1.00X I 1.25X I I.SOX a.c

. [

]a.<:

• Acceptable storage arrays are described in Section 5.2.1. Figure 5-1 shows a planar view (x-y) of each storage array as model~c;l (periodic boundary conditions applied in the x-y directions).

WCAP-18414-NP Revision 0

5-3 Figure 5-1 KENO Array Rack Model Planar (x-y) View: Top Left Array A, Top Right Array B, Bottom Left Array C, Bottom Right Array D (All Models: x-y Periodic Boundary Conditions) 5.1.1 Description of Fuel Assembly and Storage Racks for KENO This section outlines the dimensions and tolerances of the design basis fuel assembly designs and the fuel storage racks. These dimensions and tolerances are the input basis for the KENO models used to determine the bumup requirements for each fuel storage array and to confirm the safe operation of the SFPs under normal and accident conditions.

5.1.1.1 Fuel Assembly Dimensions and Tolerances This section provides the dimensions and tolerances for the design basis fuel assembly designs. Table 5-2 provides this data for STD/RFAand OFA fuel as modeled. Selection of these fuel designs is discussed in Section 4.3 . As identified in Section 4.3.2, Criticality Fuel Design I was selected as limiting over Criticality Fuel Design 2 and 3 for STD (RFA) fuel. It is the combination of Criticality Fuel Design I input (including the burnable absorber usage during operation) that leads to the overall bounding nature, despite the lower fuel percent of theoretical density (96.5 vs 98.0) when compared with Criticality Fuel Design 3. See Section 6.2 for the analysis area of applicability.

WCAP-18414-NP Revision 0

5-4 Table5-2 Design Basis Fuel ~ b l y Design Modeling Parameters STD (RFA) Fuel ~ b l y Parameter Value Tolerance Rod array siz.e 17 X 17 NIA Rod pitch, in 0.4% [ r,c Active ~el length, in 144 NIA Nominal fuel theoretical density, % TD  %.5 NIA Maximum pellet enrichment, wt°1 235U 5 [ ]a,c Total number of fuel rods 264 NIA

/

Fuel cladding OD,' in 0.374 [ 1a..c Fuel cladding ID, in 0.329 [ ]a,c Pellet diameter, in 03225 [ . ]"'c Number of 9T/IT .2411 NIA GT/IT OD, in 0.482 [ ]"."'

GT/ITID, in 0.442 [ ]a,c OFA Fuel Assembly ..

Parameter: Value 1'.olerance Rod array st.re 17 X 17 NIA Rod pitch, in 0.4% [ ]a,c Active fuel length, in 144 NIA Nominal fuel theoretical density, % TD 98.0 NIA Maximum pellet enrichment, wt°/o 235 U 5 [ ]a,c Total number of fuel rods 264 NIA

• Fuel cladding OD, in 0.360 [ ]a,c

(

Fuel cladding_ ID, in 0.3I°5 [ . ]a,c Pellet diameter, in 0.3088 [ ]a,c Number ofGTIIT 24/1 NIA GT/IT OD, in 0.474 [ ]a,c GT/IT ID, in 0.442 [ ]a,c WCAP-18414-NP Revision 0

5-5 Notes:

1. [ ]a.c

2. The maximum pellet enrichment tolerance is used for all enrichments evaluated as identified in Section 523.1.2.

5.1.1.2 Fuel Storage Cell Rack Dimensions and Tolerances The storage racks used at Farley Units 1 & 2 SFPs are described in Section 3.2. The fuel storage cell characteristics, as t ~ are modeled in the criticality analysis, are shown in Section 3.2. Dimensions

including tolerances are given in Table 3-4: Tolerance models were created and the reactivity impacts were acc~mnted for in the form of uncertainties added to the final reactivity cakulation as shown in Table 5-4 through Table 5-10.

5.1.2 Impact of Structural Materials on R~ctivity Over the years, different fuel types have: been developed to meet the needs of the utiliti~s. Differences between the fuel types include changes in rod pitch, fuel rod dimensions such as pellet and cladding dimensions, and structural components such as grid material and volumes.

Each of the fuel types which have been or are planned to be operated at the plant neaj to be considered.

For Farley Units l & 2, the determination of the design basis fuel assembly 'd.esigns for the analysis has been performed as outlined in Section 4.3. The structural ma~rials of each fuel type do not need to be considered in the determination of the bounding fuel assembly design as discussed in Reference 14 in regards-to grid material where 50 ppm is added to soluble boron requirements as recommended to neglect modeling grids 1*

• 5.1.2.1 Composition of Structural ~aterials Various zirconium-based materials and SS have traditionally been used ~ structural materials for fuel assembly designs. [

]11,C 5.1.2.2 Top and Bottom No7.Zles

]a,c 1 Reference 14 indicates this 50 ppm is also sufficient to offset the change in- reactivity effect of tolerances under borated conditions (if modeling only unborated conditions for bias and uncertainty caloulations).

WCAP-18414-NP Revision 0

5-6 5.1.2.3 Grids and Sleeves A generic study determined the impact of grids and sleeves being present in an assembly both during core operation and storage in tbe.SFP shows that the impact of grids and sleeves is negligible. The study was based on the [

]a.c The study incorporated a variety of depletion parameters and several different [ ]a.c Additionally, Reference 14 indicates for depletion that 50 ppm additional soluble boron in determination of normal and accident condition soluble boron requirements is acceptable if grids are not explicitly modeled in the SFP criticality analysis. An additional 50 ppm was used.

5.2 KENO MODELING ANALYSIS KENO models generated for STD/RFA and OFA bounding criticality fuel designs were evaluated for different storage arrays. The enrichment, assembly average burnup, and decay.time input were varied to determine appropriate storage limits based on resulting reactivity. Reactivity margin is added to the KENO reactivity calculations for the generation of burnup1requirements as discussed in Section 5.2.2 to account for manufacturing deviations.

5.2.1 Array Descriptions and Fuel Categories

)

Assembly storage is controlled through the storage arrays defined in this section. For storage arrays A, B and C, separate requirements are detem.uned by fuel type (STD/RFA as one type, OFA as the second type). Each storage array contains assemblies which are defined by a fuel category as given in Table 5-3.*

A fuel category is a ranking of assemblies by the maximum .allowable reactivity 9f the individual assembly in a storage cell within each storage location. A lower fuel category is more reactive than a

• higher fuel category. Unique storage locations were determined for which to assign fuel categories using the desired storage patterns for use in the Farley Units 1 and 2 SFPs. Figure 5-i shows the allowable storage arrays, including the fuel categories discu5!ied in this section.

Reactivity as discussed in this paragraph pertains to the maximum allowable reactivity in_ a storage cell.

Fuel Category 1 locations can contain the highest rea~tivity fuel assemblies, up to 5 wt°/o mu as~mblies with no bumup, IFBA, or decay time credit required. Fuel Category 2 locations can contain fresh fuel

~emblies with up to 5 wt°/o mu but are subject to IFBA requir~ments (thereby reducing the reactivity

• compared to Fuel Category l fuel assemblies). Additionally, Fuel Category 2 assemblies must have .

accumulated at least l O GWd/MTU of bumup once they have been exposed to .ensure peak reactivity is considered for IFBA burnout, with the exception of a Fuel Category 3 or 4 assembly*( discu5;5ed in this

.section). Fuel Category 3 locations are storage cells defined within Storage Array C, a.ii "all-cell" storage cell. Fuel Category 3 locations can contain fuel assemblies ,af up to 5 wt°/o mu and have minimum bumup and/or decay time requirements which determine acceptability for storage in Array C.

To determine that Fuel Category 3 assemblies are less reactive than Fuel Category 2 assemblies (which require IFBA credit), a comparison of the burnup requirements of Fuel Category 4 assemblies-that are stored with the Fuel Category 2 assemblies (in Array B) is neces~ary._ If the three Fuel Category 4 assemblies are individually Jess reactive than Fuel Category 3 assemblies, then Fuel Category 2 assemblies will be allowed a-higher reactivity than Fuel Category 3 assemblies because the overall WCAP-18414-NP Revision 0

5-7 storage array reactivities are the same (iso-reactive). Additionally, the IO GWd/MTU required for (5 wr'/o 235 U fuel assemblies) Fuel Category 2 which have bumup is significantly I~ than the bumup requirements for 5 wr'/o 2350 Fuel Category 3 assemblies.

As can be seen by comparing the storage requirements ofFu,el Category 4 assemblies and the storage requirements of Fuel Category 3 assemblies, the. Fuel Category 3 assemblies require less bumup for the

_same enrichment, indicating the individual assemblies are allowed to be more reactive.,

An array can.only be populated by assemblies of the fuel category defined in the array definition or a lower reactivity fuel category ( e.g., Fuel Category 3 assemblies can be stored in locations for Fuel Categories I, 2, or 3, but cannot be stored in Fuel Category 4 locations). !flower reactivity fuel category requirements are met for an assembly, they need not meet the requirements of the fuel category cell for which they are stored). This is a' unique*_occurrence because Fuel Category 2 requirements for exposed fuel were generated for the bounding case (I~ GWd/MTU for 5 wt% mu assemblies) and applied for all burned fuel stored in Fuel Category 2 locations.

_In addition to theSe defined fuel categories, Array D contains I I fu~l storage locations generically evaluated for storage within the Farley SFPs with STD fuel. This allows assemblies which meet the damaged fuel array storage requirements to be stored anywhere in the SFPs for which an array.of this* size is met Array D assemblies were given the. Fuel Category D label, indicating- these storage_ cells are for damaged fuel. An example _of Array D, which is currently employed at the Farley Unit 1 SFP, is shown in Figure 5-3. including the assembly ID used for storage. All I I damaged assembli~ meet the requirements of Array 0.

WCAP-18414-NP Revision 0

5-8 Table5-3 Fuel Categories Ranked by Reactivity Fuel Category I High Reactivity Fuel Category 2 Fuel Category 3 Fuel Categoi:y 4 Lqw Reactivity Not!!S:

l. Assembly storage is controlled through the storage arrays defined in Figure 5-2.

2. Fuel Categories are ranked in order of decreasing reactivity, e.g., Fuel Category 2 is less reactive thari Fuel Category 1, etc.

3. Each storage cell.in an array can only be populated with assemblies of the fuel category defined in the array definition or a lower reactivity fuel category.

4. F1:1el Category I contains fuel with an initial maximum enrichment up to 5 wt% 235 U. Neither burnup nor IFBA is required.

5.. Fuel Category 2 contains fuel with an initial maximum enrichment up to 5 wt'% mu. Storage of fresh fuel is .

determined from the mini111um IFBA equation and coefficients provided in Table 6-1 for STD/RF A fuel and Table 6-3 for OFA fuel. Fuel Category 2 fuel which has been operated in ~ereactor requires at least JO.O GWd/MTU ofburnup with the exception ofa Fuel Category 3 or 4 assembly.

6. Fuel Categories 3 and 4 are determined from the minimum burnup equation and coefficients provided in Table 6-7 and Table 6-11 for STD/RFA fuel, and in Table 6-9 and Table 6-13 for OFA fuel, respectively.

Example btirnup requirements at several initial enrichments and decay times are provided for Fuel Categories 3 and 4 in Table 6-8 and Table 6-12 for STD/REA fuel, and Table 6-10 and Table 6-14 for OFA fuel, respectively. *

7. Example IFBA requirements at several initial enrichments for IFBA thicknesses of I .OX, 1.25X. and l .50X are provided in Table 6-2 for STD/RF A fuel and in Table 6-4 for OFA fuel.

WCAP-18414-NP Revision 0

5-9 ArrayA v 1 X Two Category 1 assemblies with two empty storage locations. The C~tegory 1 fuel assemblies must only be face adjacent to an empty storage location.

X 1 ArrayB 4 4 One Category 2 assembly with three Category 4 ~blies.

4 2 ArrayC 3 3 Four Category 3 assemblies.

. J. 3 X X x X X *.*X ArrayD X D X* D D X Eleven Category D assemblies arranged in an array of four assemblies by three assemblies. One storage cell along the four storage cell wide side of the outside of the Array m~ remain empty. The storage array must have at X D D D D X least one row of empty*ceJls between it and any other

. (Array A, B, C, .or D). A row of empty cells array . are not needed on any section of the configuration face adjacent X D D D D X to the SFP wall.

X X X X X. X Notes:

1. Any storage array location designated for a fuel assembly.may be replaced with non-fissile material or an empty (water-filled) cell.

2.- Empty locations designated with. an X must remain completely empty.

3. Storage )'.equirements are determined for different fuel types (RFA/STD and OFA) for fuel category_ 1 through 3. Only RFA/STD fuel is evaluated for storage as Category D.

Figure 5-2 Allowable Storage Arrays WCAP-18414-NP Revision 0

5-10 F31 Empty F30 F06 F18

• F1?- F19 F02

• , F15 F20

• F05 F32 F~re.5-3 Currently Used Damaged Fuel Assembly Configuration (Farley Unit 1)

WCAP-18414-NP Revision 0

5-11 5.2.2 Target ketr Calculation Description As discussed in Section 2.1, this analysis provides burnup requ~ents such that the Farley Units 1 & 2 SFPs remain subcritical in unborated conditions. To ensure that the burnup requirements generated are appropriate, a target keirvalue is created for each array at differen~ enrichments (maximum fresh,-

3, 4 and 5 wt%, 235 U). The target kewvalue accounts for the reactivity effect of applicable biases and uncertainties and includes administrative margin to ensure safety as shown in Equation 5-1. -

Target~= Acceptance Criterion - Admin Margin - 'f.(Biases & Uncertainties) Equation 5-1 where, Acceptance Criterion = the maximum allowable keirfor a storage array (see-Section 2.1)

A'.dmiri Margin = the administrative margin (0.005 &) taken to provide additional certainty of safe operation L(Biases & Uncertainties) = the amount of reactivity that accounts for biases and uncertainties in the reactivity' calculation for each storage array The sum of biases are simply additive while the sum of uncertainties are statistically added as the root sum sq~ of the individual reactivity uncertainties.

5.23 Bias & Unce.rj"ainty Calculations Reactivity biases are known variations between the real 8:nd analyzed syste-m and their reactivity impact is added directly to ~he calculated k.,ff. Examples include the SFP temperature and code validation biases.

Uncertainties account for allowable variations within the real model whether they are physical (manufacturing tolerances), analytical (depletion uncertainty and validation bias uncertainty), or measurement related (bumup measurement uncertainty). Biases have a greater impact due to their direct addition to the total sum of bias and uncertai,nty._Uncertainties are statistically added 8$ the root sum square of the in~ividual reactivity-uncertainties.

]a,c ,

5.2.3.1 Bias & Uncertainty Descriptions including Manufacturing Tolerances Reactivity biases and uncertainties as a result of manufacturing tolerances and other SFP characteristics

_are ,discussed in this s~ction and the following subsections. KENO is used to quantify reactivity effects.

WCAP-18414-NP Revision 0

5-12

[

]a,c WCAP-18414-NP Revision 0

5-13 5.2.3.1.1 Cladding Tolerance Reactivity Uncertainty

[

]a.<:

5.2.3.1.2 Initial Feel Enrichment Reactivity Uncertainty

)S.C 5.2.3.1.3 Guide Tube and Instrument Tube Reactivity Uncertainty

]a,c 5.2.3.1.4 Bnrnup Measurement Uncertainty WCAP~18414-NP

. Revision 0

5-14 5.2.3.1.5 Depletion Uncertainty The depletion uncertainty takes into account the potential reactivity misprediction of the depletion code.

[

]l,C .

5.2.3.1.6 Operational Uncertainty

]a,c 5.2.3.1.7 Flux Trap Gap Reactivity Uncertainty The fl~ trap gap tolerance worth is not explicitly calculated; however the rack cell pitch tolerance cases explicitly include the effect of cell pitch and flux trap gap tolerance since the rack pitch would change the flux gap width and vice versa The rack pitch tolerance was chose~ as it fa the larger change.

5.2.3.1.8 Borated Sheath Width Reactivity Uncertainty Reference 14 indicates that for flux-trap rack designs, the uncertainty due to the manufacturing tolerance on the sheathing width is small but cannot generically be declared negligible. Despite not creqiting the J

borated in~rt, sheath width tolerance reactivity uncertainty is determined.

WCAP-18414-NP

- Revision 0

5-15 5.2J.1.9 Borated Insert Cavity Width Uncertainty Jo addition to the ~rated sheath width reactivity uncertainty, the borated insert cavity width uncertainty is determined 5.2.3.1.10 Other Uncel'tainties

• An uncertainty in the predictive capabi!jty of Scale 6.2.3 and the associated cross-section library is considered in1he analysis. The ~rtainty from the validation of the calculational methodology is discussed in detail in Appendix A.

5.2.3.1.11 Assembly Envelope Expansion Bias The assembly envelope expansion bias is comprised of r

]a,c WCAP-18414-NP Revision 0

5-16

[

]a,c 5.2.3.1.12 fission Product and .Minor Actinide Worth Bias A_cornmon approach to the_validation of cross-sections*is by benchmarking critical experiments that are oesigned to closely represent the configurations 9f the desired criticality application. The validation of fission praj~~ however, is moi;e difficult because few critical experiments are available. Due to the limited availability of fission product benchmark data, a factor of uncertainty was considered in the criticality safety analysis.

NUREG/CR-7109, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (kdr) Predictions" (Reference 13) presents findings that show fo_r minor

• actini9e and fission product nuclides for which* adequate critical e:q,eriment data aie not available, calculations of kotr uncertainty due to nuclear data uncertainties can be used to establish a bounding bias value which was approximately 1.5 perc:ent of the worth of the minor actinides and fission products.

[

.* I

. \

]a,c WCAP-18414-NP Revision 0

5-17 5.2.3.1.13 Eccentric Fuel Assembly Positioning Bias The fuel assemblies are assumed to be nominally located in the center of the storage rack ceJI; however, it an is recognized that assembly could in fact be located eccentrically within its storage cell. Reference 14 indicates that assembly eccentric positioning should be considered in racks without absorber panels.

Racks in this analysis contain two absorber panels between each storage location, however, they are not credited in this analysis; so an eccentric positioning bias is determi~

• To quantify the reactivity effects ,9[ eccentrically located fuel within a fuel storage cell, [

JO.C 8,C

• I Figure 5-4 WCAP-18414-NP Revision 0

5-18 5.2.3.1.14 SFP Temperature Bias The Farley Units 1 & 2 SFPs do not have a nominal temperature; instead it operates within an allowable range. [

5.2.3.1.15 Borated and Unborated Biases and Uncertainties Technical Specifications require each SFP to have kc£r,to be< 0.95 under borated conditions accounting for all applicable biases and uncertainties. [

]a,c 5.2.3.2 Storage Array Biases & Uncertainties Results Tables 5-4 thi:ough 5-10 give the calculated biases and uncertainties for Array A, B, and C, and D for STD and OF~ fuel as well as the total sum of biases and uncertainties and administrative margin for determination of the listed Target kc£r values. Not~ that for Array B, ,the initial enrichments. shown at the top of Table 5-6 for STD fuel and Table 5-7 for OFA fuel coqespo_nd to Fuel Category 4. [

]a,c Fuel Category 2 assemblies are fresh fuel assemblies up to 5 wt°/o, with fresh IFBA requirements discussed in Section 6.1.

WCAP-18414-NP Revision 0

5-19 TableS-4 Biases & Uncertainties for Array A with STD Fuel a,c WCAP-18414-NP Revision 0

5-20

,___T:_a_b_le_S-_5_*_ _B_iases

__&_u_>>_ce_na_in_ti_*es_ror_Arra

__Y_A_w_i_tb_O_F_'.A_F_uel L._-

_ _.... 11,-C J

/

WCAP-18414-NP Revision 0

5-21 J

Table5-6 .. Biases & Uncertainties for Array B with. STD Fuel a,c

, I WCAP-18414-NP Revision 0

5-22 ThbleS-7 Biases & Uncertainties for Array B for OFA Fuel a,c WCAP-18414-NP Revision 0

5-23 Table 5-8 Biases & Uncertainti~ for Array C with ~ Fuel a,c WCAP-18414-NP Revision 0

5-24 Table5-9 Biases & Uncertainties for Array C with OFA Fuel a,c

\

r WCAP-18414-NP Revision 0

5-25 Table 5-10 Biases & Uncertainties for Array D with STD Fuel a,c WCAP-18414-NP

• Revi_sion 0

. 5-26 5.3 INTERFACE CONDITIONS Interfaces are the locations where there is a change in either the storage racks or the storage requirements of the fuel in question. In this analysis, only intra-region interfaces are evaluated since all racks are of the same design and no pool region interfaces are present In addition to the intra-region interfaces, Array D assemblies are required to have a row of empty. storage cells ( or the pool wall) face. adjacent to all sides of

• the storage array.

The only interface conditions that need to be addressed in this analysis are those.between different fuel storage arrays. [

]11,C a,c Figure 5-5 la,c Additionally, Array D contains 11 fuel storage locations generically evaluated for storage within the Farley SFPs. One-storage cell along the four storage cell wide side of the outside ofthe-Artay must remain empty i.e. water-filled. The storage array must have at least one row of empty cells b6tween it and any other array (Array A, B, C, or D). A row of empty cells is not needed on any section of the configuration face adjacent to the SFP wall.

WCAP-18414-NP Revision 0

5-27 5.4 NORMAL CONDmONS This section discusses nonna1 conditions within the SFPs which are in addition to the steady-state storage of fresh and spent fuel assemblies. During nonnal operation, the SFPs have a soluble boron concentration of great_er than 2000 ppm and a moderator temperature :S 185°F. Beyond the storage of fuel assemblies, there are five major types of normal conditions covered in this analysis. These five conditi9ns are explained in subsections 5.4.1 through 5.4.5.

• 5.4.1 Type 1 Normal Conditions Type I conditions involve the placement of components io the guide tubes and/or instrument tube of intact fuel ass;emblies while normally stored in the storage racks. This also includes removal and reinsertion of these comp_onents into the fuel when stored in the rack positions using specifically designed tooling. Examples include control rods, neutron sources, guide tube probing, fuel assembly guide tube length measurement, and ultrasonic test equipment being placed in a testing location on top of a spent fuel storage rack.

• The Type 1 normal conditions typically include the insertion of components into fuel assemblies for storage in the SFPs (e.g., depleted Pyrex). The SFPs as _single systems aie over moderated. A single fuel assembly however, is significantly undennoderated, and reducing the interstitial hydrogen to uranium ratio lowers the system ke11 as seen by the fact. that all rod pitch uncertainty cases show that a reduction in rod pitch reduces reactivity. Additionally, calculations have been performed which show-that [

]11,c Any components designed to -be inserted into-~ asse~bly may be stored in a fu.el assembly in the SFPs.

5.4.2 Type 2 Normal Conditions Type 2 conditions involve evolutions or transitional fuel assembly actio~ where the fuel assembly is removed from its normal storage rack location for a specific procedure and reinserted after the completion of the procedure. Examples of Type 2 conditi()llS include fuel assembly visual inspection, reconstitution, cleaning and sipping. During the Type 2 assembly evolutions only one fuel assembly will be manipulated at a time and all manipulations will oeccur outside the storage cell and not within one assembly pitch of other assemblies. pescriptions of each of these items are provided, along with the evalµation of the impact on this criticality safety analysis.

• One c;ell pitch of separatiqn is defined as one rack.pitch distance away fi-om all sides of the assembly (including both face a*djacent and corner adjacent cells). Outside of a storage cell b o t h ~ to fuel which has been removed to a location outside of the storage rack as well as-to the areas of an assembly exposed above the rack due to partial insertion.

Fuel assembly cleaning is defined as placing cleaning equipment adjacent to a single assembly and either jetting water from or into a nozzle. The cleaning equipment will displace water adjacent to tpe assembly and can use demineralized (unborated) water to clean assemblies. The demineralized water used in this process is not confined to a.particular volume but would be readily dispersed into the bulk water of the SFP. In all cases, only one fuel assembly will be manipulated at a time and all manipulation*s will occur WCAP- I 8414-NP Revision 0

5-28 outside the storage cell and not within one assembly pitch of other assemblies. The large delta between the Technical Specification required boron concentration and the boron concentration credited in this analysis and the relatively small volume of demineralized water used for this operation guarantees that the addition of unborated water does not constitute a significant dilution event_

Fuel assembly inspection is defined as placing non-destructive examination equipment against at least one face* of an assembly. Periscopes and underwater . . cameras can be ' placed against all four faces of the assembly sim~ltaneously and will displace water. In all cases, only one fuel assembly will be manipulated at a till}e and:a11 manipulations will occur outside the storage cell and not within one assembly pitch of

. other assemblies ..

Fuel assembly reconstitution involves rod movement from and/or to an assembly. In most cases, damaged rods will be replaced with SS rods, but natural uranium rods may also be used If a rod is replaced with either SS or a rod made_ofnatural uranium the reactivity of the available fissile material of the single assembly will be decreased while capable moderation remains the same resulting in a reduction in reactivity. If the fuel rod in question is either replaced by a fuel rod from another assembly or the fuel rod

. is to be removed without replacement, adjustments must be made. to the bumup storage requirement of the assembly being reconstituted.

. I

]a,c Fuel assembly sipping is defined as placing one fuel assembly in the sipping equipment. The fuel assembly is separated from all other stored fuel by at least one assembly pitch via the equipment design. While the sipping equipment C8J1. be placed within one assem.bly pitch adjacent to a storage rack l~aded with fuel, the fuel assembly loaded into the sipping equipment must be more than one assembly pitch removed, from the fuel located in the storage racks. During this operation, demineralized water may be introduced to the sipping container, exposing 1;he assembly(s) to an unborated environment WCAP-18414-NP Revision 0

5-29 Fuel assembly ~leaning, inspection, reconstitution, and sipping are bounded by this criticaJity analysis.

[ ,,

5.43 ~3 Normal Conditions Type 3 conditions. involve insertion of compo~nts that are not intact fuel assemblies, into the fuel storage a

rack cells. For Farley Units 1 & 2 SFPs, these include a loose pellet canister as well as failed fuel rod storage canister. Additionally, any components that-do not contain fissile materials can replace a fuel .

assembly of any fuel category in one of the approved storage configurations described in Section 5.2.1.

The-fuel rod storage canister at Farley Units I & 2 is a rectangular lattice of storage tubes for failed fuel rods arranged in an 8 x 8 pattern. Not all rows contain 8 tubes as can be seen from the inodeled schematic in Figure 5-6. Table 3-5 contains pertinent design information.

8,C Figure>6 Schematic View of Modeled Failed Fuel Rod Storage Canister

]8,C.

WCAP-18414-NP Revision 0

5-30

[

]11,C The loose pellet transport canisters (LPTCs) are SS canisters designed to store upto 5000 loose fuel pellets. Design details are given in Table 3-6. Modeling of the LPTCs [

]a,.c a,c

. Figure 5-7 ]a,<

]a,c 5.4.4 fype 4 Normal Conditions Type 4 normal conditions include temporary installation of non-fissile* components *on the rack periphel)'

facing the pool wall. Analyses of the storage arrays contained within this criticality analysi? assume an infinite array*of storage cells. This assumption bounds the installation of any non-fissile components on the ~riphel)' of racks.

WCAP-18414-NP Revision 0

5-31 5.4.5 Type 5 Normal Conditions lype 5 conditions involve miscellaneous ~onditions that do not fit into the first four.normal condition types. Examples include usage of fuel handling tools for their intended purpose, miscellaneous debris under the storage racks, and damaged storage cells.

-A damaged storage ~I I is defined as a cell where the cell liner is out of tolerance or the entry channel has been damage~L These cells should not be used to store fuel assemblies, but they may be used to store items that 'need to be stored as a fuel assembly (i.e.; non-fissile material or-a fuel rod canister, etc.).

Insertion of handling tools into the top of fuel assemblies or other components occurs frequently in the SFP environment. The insertion of handling tools into. the top of an assembly is bounded by the storage of inserts in fuel assl?mblies and therefore, from a criticality perspective, all fuel handling tools are acceptable for their intended purpose.

Performance ofForeign Object Search and Retrieval (FOSAR) from fuel assemblit:5 and/or _storage cells must meet the following guidelines.

l. If a FOSAR is done on a storage cell, any fuel assembly residing in- the storage cell must be removed before the action takes place.

I ,

2. For FOSAR done on a fuel assembly, if the operations do not occur in the active fuel region and do not require tooling to reside in the active fuel region, the FOSAR does not impact criticality and the assembly can remain in its storage cell.

3. If the FOSAR requires tooling to be present in the active fuel region,-then the fuel assemb_LJ:must be separated from other fuel assemblies by at least one assembly pitch.

The Farley SFPs have pumpifiltration systems which sit on top of the fuel racks. These systems displace water above the assembly, ~hich is conservative in unborated_conditions. Borated models for accident -

analysis use unborated water above and below the active fuel befo~ a reflective boundary is applied. In -

addition, the separation from the_fuel rods due to the top nozzle and the fuel rod end plug is sufficient to prevent signifi~t neutron interaction. Therefore, there is no restrictiC!n on.the location of the filtration systems.

5.5 .SOLUBLE BORON-CREDIT Section 2.1 contains kotr-requirements under bqth the assumption that the pool is flooded with pure water and that the pool contains soluble boron. This section outlines the calculations that were performed to demonstrate the 59luble boron concentration necessary to meet the soluble boron requirements in Section

2. L In reporting the soluble boron requirements, the atomic percent (at%) of 10B in boron is conservatively assumed to be 19.4 to. bound the potential variation in the isotopic concentration of boron within the SFP~.

WCAP-18414-NP Revision 0

5-32

-5.5.1 Soluble Boron Requirements for Normal Conditions Soluble boron credit for Nonna] Operating Conditions is evaluated for Farley Units 1 & 2. Additional pertinent details for modeling each storage array conservatively for the Normal Operating Conditions soluble boron determination are as follows. While additional models were evaluated with all fresh fuel, cases with the highest bumup are conservative as expected due to the reduction in boron ~orth with bumup due to spectral hardening. As a result, the following models are conservativ~ for each storage array: ,

• Array A: Fuel assemblies were modeled as fresh 5 wt°/o mu fuel.

• Array B: Fuel is modeled as three 5 wt°/o initial enrichment fuel-at 48,GWd/MTU and one fresh 5 wt% 235 U fuel assembly.

• Array C: Fuel is modeled as 5 wto/o initial enrichment at 34 GWd/MTU.

• Array D: Fuel is.modeled as 3 wt% mu initial *enrichment at 4 GWd/MTU To d~termine the maximum soluble boron concentration for normal conditions to meet a 95/95 kctr of<

0.95 including biases and uncertainties, where 95/95 kctr is defined as 95/95 keff = KENO keff + 2aketf + B&U + Adm.margin, Equation 5 where:

• KENO keff = The simulated i,tCCident condition ,kq,-

aketf = The simulated accident condition~ Morte Car!o simulati~n standard deviation B&U = The total bias and uncertainty term 1 Adm. Margin = Administrative margin.

The ~nirnum soluble boron concentration*to.tnaintain kctr< 0.95 for the limiting nonnal condition including biases, uncertainties, and administn;ltive margin is 320 ppm, conservatively rounded up frqm value determined from linear interpolation (plus rounding) of Array B is 270 ppm. Results are given in Table 5-11 for STD/RFA fuef and in Table 5-12 for OFA-fuel for all storage array. '

1 Biases and unce_rtainties are taken from the nominal storage c~ndition for all storage arrays.

WCAP-18414-NP Revision 0

5-33 Table 5-11 -[ a.c Table 5-12 [ ]""' a,c

\!

5.5.2 . Soluble Boron Requirements for Accident Conditions In addition to maintaining ~ not to exceed 0.95 during normal operations, soluble boron is used to offset the potential reactivity insertion events in the SFPs. the.following accidents are considered in this analysis:

• Assembly misload

• SFP temperature greate~ than normal operatin'g range(> 185°F)

• Dropped & misplaced fresh fuel assembly

• Seismic event WCAP- I 8414-NP Revision 0

5-34 5.5.2.1 Assembly Misload This section addresses the potential for an l,lSSeffibly or assemblies to be placed in a storage cell location, which is not allowed by the burnup requirements in Section 6.1, in addition to an assembly misloaded between the SPP storage rack and concrete wall. This analysis addresses both the misload of a single

.assembly into an unacceptable storage location and multiple assemblies being misloaded in series into unacceptable storage locations.

5.5.2.1~1 Single Assembly Misload

[

]a,c 5.5.2.1.2 Multiple Assembly Misload A multiple assembly misload is a postulated accident where assemblies are misloaded in series due to a common_cause. [

]8,C Table5-13 ]a,c a,c WCAP-18414-NP :

Revision 0

5.5.2.2 Spent Fuel Temperature Outside the Normal ~perating Range The J.M. Farley Units 1 and 2 SFPs are to be operated at less than 185°F. However, under accident conditions this temperatute could be higher. [

]O.C 5.5.2.3 Dropped & Misplaced Fresh 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. [

]a.c 5.5.2.4 Seismic Event In the event of an earthquake or similar seismic event, the SFPs storage racks can shift position. This can cause the rack modules to slide together eliminating the space between modules and between modules and the SFP wall. [

_]S.C WCAP-18414-NP Revision 0

5-36 5.5.3 Soluble Boron Requirements Table 5-14 data indiqites-that [

)11.C Table 5-14 [ ]""'

r-1L-~~~~~~~~~~~-'--~~~~~~-'--~~~---'~-'--'-~~~~~~~~--11---,~c I

WCAP-18414-NP Revision 0

5-37 5.6 RODDED OPERATION While standard operation is perfonned unrodded, it is allowable to operate at hot full power with rods inserted to the power dependent insertion limits. Operating with control rods inserted into the core impacts the assemblies in the rodded locations. The i~on of a control rod into an assembly during operation has several effects.

The reactiy~ty of an assembly experiencing rodded operation can increase relative to an assembly which does not experience rodded operation. The loss of moderator as water displaced in the GTs when a control rod is insei:J:ed into the assembly will harden the neutron spectrum increasing plutonium production. The control rod will also preferentially absorb thennal neutrons, further hardening the neutron spectrum. In addition to the spectral hardening, the control rod will lower the power in the area of the .

~mbly where it is inserted This will lower the burnup accumulated in the top of the assembly, increas.ing the .end'effecl These effects can all increase the reactivity of an assembly, making it possible for an assembly operated with rods inserted to be more reactive than an assembly of the same assembly average burnup which experienced unrodded operation.

While the,se items can increase reactivity, there are competing effects which reduce assembly reactivity due to rod insertion. When a control rod is inserted into an assembly, the power in that assembly will be reduced. This will reduce both the fuel and moderator temperatures. The reduction in fuel temperature will decrease Doppler broadening leading to less neutron c.apture by 238U, thus lowering plutonium production. The reduction in moderator temperature will increase moderator density, increasing neutron moderation and therefore softening the neutron spectrum.

In addition to impacting the neutron spectrum, rodded operation can also affect_ the axial burnup profile of assemblies. Operation with a control'rod inserted in an assembly will shift power down, under-depleting

• the top of the assembly while the control rod is present Once the control rod has been withdrawn from the assembly, power preferentially moves to the under-depleted top of the assembly, and over time the axial burnup profile developed will return to a profile :typical of unrodded operation. Therefore, time-in-life before final discharge of an assembly is an important faGtor in the impact of rodded operation on assemb~y reactivity.

NUREG/CR-6759, "Parametric Study of the Effect of Control Rods for PWR Burnup Credit" I

(Reference 15) defines a significant amount of control rod insertion as more than 20 cm into the core.

Farley Units 1 and 2 have not operated at full power with control rods inserted a significant length into the core. Therefore, there is no significant burnup accrued during depletion with rods inserted in the active fuel height, and no need to account for these effects in burnup li~ts contained within this analysis.

Any assemblies incurring significant rodded operation going forward must not credit the rodded burnup.

While typical operation for Farley Units 1 & 2 is performed unrodded, there is potential to operate at reduced power levels with rods inserted. Short term reduced power operation may be the result of plant equipment issues or economic considerations and has occurred at Farley Units I & 2. Any impact from short term operation at reduced power levels with rods inserted will be negligible.

WCAP-18414-NP Revision 0

6-1 6 - - ANALYSiS RESULTS & CONCLUSION

_ This section documents the final storage re~mlts of the Farley Units I & 2 Spent Fuel Pool criticality safety analysis. Included in this section are the bumup requirements for the fuel storage arrays documented in this analysis. This section also contains the Area of Appli~ility of this analysis. The Area of Applicability (AoA) of the criticality code validation suite is discussed in Appendix A.

6.1 BURNUP AND IFBA REQUIREMENTS FOR STORAGE ARRAYS

• Assembly storage is controlled through the storage arrays defined in Section 52.1. An army:can only be populated by ~mblies of the fuel category defined in the array definition or a lower reactivity fuel category (see Table 5-3). Fuel Category I does not require burnup or fresh.IFBA for storage. Fuel Category 2 assembly storage requirements require that they either must b.tve not been operated in the reactor and the IFBA loading must exceed the "minimum JFBA" (# rods per assembly) given by the IFBA requirements coefficients or have at least 10.0 GWd/MTU of exposure C9Vering the peak reactivity of IFBA bearing assemblies with 5 wt°/o 235 U enrichment (No IFBA requirements are needed beyond 10.0 GWd/MTU). Fresh IFBA requirements coefficients as well as sample 1FBA require)J'}ents are given in Table 6-1 and Table 6-2 for STD/R,FAfuel and in Table 6-3 and Table 6-4 for OFA fuel. Fuel categories D, 3, and 4 are defined by assembly average bumup, initial enrichment*; and decay time with burnup requirement coefficients and sample evaluated bumup !~its given iri Table 6-5 through Table 6-14.

This analysis has provided burnup requirements at discrete decay times, measured in years. Howeve~, it is acceptable to interpolate between these decay times to determine bumup requirements at alternate decay times. Using linear interpolation between two already analyzed decay times will give a conservative bumup requirement for the decay time in question. Linear interpolation based on actual decay-time should be performed between cal~ulated values of minimum bumup associated with tabulated decay times'._

greater and less than the actual decay time. No extrapolation beyon~ 20 years is permitted. This is acceptable because isotopic decay is an exponential function which means assembly reactivity will decay faster than the calculations using .linear interpolation would predict Initial enrichm~nt is the o:iaximum nominal enrichment of the fuel,_prior to reduction in 235 U content due to fuel depletion. If the fuel assembly contains axial regions of differe~t wu_enrichment values, such,as axial cutbacks, the maximum initial enrichment val-ue is to be used.

WCAP-18414-NP Revision 0

6-2 Table 6-1 Fuel Category 2: STD/RFA IFBAFitting Coefficients Fitting Coefficients

,IFBA Thickness Al A2 A3 '

1.00X 5.2750 83325 -79.9546 1.25X 3.7476 10.8046 -72.0974 I.SOX L8593. 19.8050 -81.5075 Notes:

I. For a fuel assembly to meet the requirements the assembly mu,st either:

a Not have been operated in ~e reactor and the IFBA loading must exceed the "minimum IFBA" (# rods per assembly) given by the curve fit for the assembly "initial enrichment,"

or,

b. Have at least I 0.0 GWd/MTU of exposure.

2. The specific minimum IFBA required for each fuel assembly is calculated from the following equation:

1. of lFBA Rods= Al

• En2 + A2 *En+ A3

3. Initial enrichmen~ En, is the maxirtmm radial average 235 U enrichment. Any enrichment greater than 3.2 wto/o 235 U and less than or equal to 5 wt% 235 U may be used. The number ofIFBA i:ods required must be rounded up to the next whole number. Below 3.2 wt% 235 U, IFBA is not

. required.

Table 6-2 Fuel Category*2: Example STD/RFA IFBA Requirements(# of IFBA Rods)

- Average Initial Enrichment, wt% 235U IFBA Thickness 3.2 3.8 '-

4.2 4;6 5 l.OOX 1 -28 49 70 - 94 1.25X I 24 40 57 76 I.SOX 1. 21 35 49 64*

Note: . . .

I. The values provided in this table are provided as an example. The requirements must be calculated using the coefficients in Table 6-1.

WCAP-18414-NP Revision 0

6-3 Table6-3 Fuel Ca~ory 2: OFA IFBA Firiing Coefficients Fitting Coefficients IFBA Thickness Al A2 A3 1.00X 62658 0.8890 -65.4949

. 1.2SX 3.9144 '9.3963 -68.9414

• 1.SOX 1.5898 21.84:36

.. -84.9630 Notes:

• I. For a fuel assembly to meet the requirements the assembly must either:

a Not have been operated in the reactor _and the _IFBA loading must exceed the "minimum IFBA" (# rods per assembly) given by the curve fit for the assembly "initial enrichment,"

or,

b. Have at least 10.0 GWd/MfU ofexposure.

2. The specific minimum IFBA required for each fuel assembly is cal~ulataj from the following equation:

1. ofIFBARods ::AJ *Err+ A2 *En+ A3

• J. Initial enrichment, En, is the maximum radial average ~ 5Uenrichment Any enrichment greater than 32 wt°/o 235 U and less than or equal to 5 wt°/o 235 µ may be used The*number ofIFBA r0s required must be rounded up to the next whole number. Belo~ 3.2 wt°/o 235 U, IFBA is not

. required.

Table6-4 Fuel Category"2: Example OFAIFBAReqtiirements (# ofIFBARods)

Average Initial Enrichment,*wt°/o 235U IFBA Thickness 3.2 3.8 4.2 4.6 5

• • 1.oox 2 29 49 72' 96 1.25X 2 24 40 58 76 1.50X 2

• 21 35 50 64 Note: -*

1. The values provided in this table are pr<:Jvided as an example. The requirements must be calculated using the coefficients in Table 6-3.

WCAP-18414-NP Revision 0

6-4 Table6-5 Fuel Ca~ory D: ~/RFA Burnup Requirement Coefficients Coefficients Decay Time (yr.) A1 A2 Al, At 0 0 0 9.6344 -24.5678 5 0 0 9.45~ -24.1047 10 0 0 9.3343 -23".8025 1.5 0 0 9.2508 -23.. 58% -

20 0 0 9.1965 -23.4510 Notes:

I. 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 the assembly burnup must exceed the "minimum bumup"

. (GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment" If the computed minimum bumup value is negative, zero shall be used. The specific minimum bumup

• required for each fuel assembly is calculated from the following equation:

BU= A1

• En3 + A2 ~ En2 + A3

• En + Ai [GWd/MTU]

• 2. Initial enrichment, En, is the maximum mu enrichment Any enrichment between 2.55 wt°/o m.u and 3 wt% mu may be used. Below 2.55 wt°/o mu, burnup credit is not required.

3. An assembly with a decay time greater than 20 *years must use the 20-ye'ar ( or less decay time) limits. '

Table~ Fuel Category D: Example STD/RFA Burnap Requirements (GWd/MTU)

Average Initial Enrichment, wt% 235U . - I Decay Time (yr.) 2.55 3 0 0 4.336 5 0 4.254 10 0 4.200 15 0 4.163 20 0 4.138 Note:

1. This table is included as an example, the purnup requirements will be calculated using the coefficients provided.

WCAP-18414-NP Revision 0

6-5 Table6-7 Fuel Category 3: STD/RFA Bomop Requirement Coefficients Coefficients Decay Time (yr.) . A1 Az AJ ~

0 0.2251 -2.5199 21.4065 -36.6115 5 03002 -3.4376 24.0978 -38.9002 10 0.1856 -2.3309 202704 -34.6503 15 0.0892 -1.3905 17.0683 -31.1550 20 0.0388 -0.9253 15.5082 -29.4500 Notes:

1. All relevant uncertainties are explicitly included in the criticality analysis. For i n ~ , no additional allowance for bumup uncertainty* or enrichment uncertainty is required. For a fuel assembly to meet the requirements the assembly .burnup must exceed the "minimum burnup" . _

(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichme11t." If the computed minimum burnup value is 11egative, zero shall be used. The specific minimum burnup required for each fuel assembly is calculated from the following equation: ..

BU= A1

• En3 + A2

• En 2 + A3

• pn + fu [GWd/MTU]

2. lnitial enrichment, En, fa the maximum mu enrichment Any enrichment between 2.15 wt°/o mu and 5 wt%, mu may be use4 Below 2.15 wt°/o mu, burnup credit is not required.

3. An assembly with a decay time greater than 20 years must use the 20-year (or less decay time) limits.

/

Table 6-8

• Fuel Category 3: Eumple STD/RFA Bornop Requirements (GWd/MTU)

• Radial Average Initial Enrichment, wt% 235U Decay. . .

Time (yr.) . 2.15 3 I 4 5 0 0 11.007 23.103 35.561 5 0 10.560 21.702 33.174 10 :O 10.194 21.015 31.629 15 0 9.944 20.579 3Q.574 20 0 9.795 20.261 29.809 Note:

1. This table is included as an example, the burnup requirements will be calculated using the coefficients provided. .. '

WCAP-18414-NP Revision 0

6-6

'(able 6-9 Fuel Category 3: OFA Bornup Requirement Coefficients r, Coefficients Decay Time (yr.) A1 Az & ~

0 0.1692 -1.8852 18.5219 * -32.7830 5 0.0191 *-0.4154 13.4482 -27.1777 10 -0.0705 0.4300 10.5987 -24.0722 15 -0.1420 1.1146 '8.2825 -21.5440 20 -0.1959 1.6375 6.5093 -19.6130 Notes:

I. 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 the assembly burnup must exceed the "minimum burnup" *

(GWd/MTU) given by the curve fit for the assembly "decay time"' and "initial enrichment." If the computed minimum bumup value is negative, zero shall be used. The specific minimum bumup required for each fuel assembly is calculated from the following equation:

BU = A1

• En3 + A2

• En2 + A3 **En+ ~ [GWd/MTU]

2 . Initial enrichment, En, is the maximum mu enrichment Any enrichment between 2.15 wt°/o mu

. and 5 wt°/o 235 U may be used. Below 2.15 .wt°/o 235 U, b~up credit is not required.

3. An asse~bly with a decay time greater than 20 years must use the 20-year (or less decay time) limits.

Table 6-10 Fuel Category 3: Example OFABurnnp Requirements~ (GWcl/MTU)

Radial Average Initial Enrichment, wt% 235 U Decay Time (yr.) 2.15 3 4 __5 0 0 10.384

• 21.970 33.847 5 0 9.944 21.191 32.066 10 0 9.690 20.691 30.859 1S 0 9.501 ..

20.332 29.984 20 0 9.363 20.087 29.384 Note: I

1. This table is included as an example, the bumup requirements will be calculated using the coefficients provided.

WCAP-18414-NP Revision 0

6-7 Table6-ll Fuel Category 4: STD/RFA Burnup Requirement Coefficients .

Coefficients Decay Tune (yr.) A1 Az ,AJ At 0 --0.6112 4.6655 6.7127

-21.8911 5 --0.3326 2.0713 12,8468 -26.1880 10 --0.1305 0.0505 183242 -30.7080 15 0.1360 .-2.6856 26.5239 -38.3300 20 0.2321 -3.7177 29.5977 -41.1200 Notes:

I. All relevant uncertaintfes are explicitly included in the criticality analysis. For instanGe, no additional allowance for bumup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements the assembly burnup must exceed the "minimum bumup" *

(GWd/MI1J) given by the curve fit for the assembly "decay time" and "initial enrichment." If the computed minimum burnup value is negative, zero shall be used. 1be specific minimum bumup required for each fuel as.5e111bly is calculated from the following equation:

BU= A1

• En3 + A2

• En2 + A3

• En + At [GWd/MTU]

2. Jnitial enrichment, En, is the maximum ~35U enrichment Any enrichment between 1.7 w1°/o mu and 5 wt% 235 U may be used. Below 1.7 wt% mu, burnup credit is not required.

3. An assem~Iy with a decay time we-mer than 20 years must use the 20-year (or Jess decay time)

• limits.

Table 6-12 Fuel Category 4: Example STD/RFABornup Requirements (GWd/MTU)

Radial Average Initial Enri~hment, wt% 2350 Decay Time*(yr.) 1.7 3 4 5 I ,

0 0 .23.734 . 40.477 51.916 5 0 21.836 37.008 48.240 10 0 21.183 34.918 45.812 15 0 20.735 33.344 44.140

. 20 0 20.480 32.371 42.890 Note:

1. This table is included as an, example, the bumup requirements will be calculated using the coefficients provided.

WCAP-18414-NP Revision 0

"6-8 Table6-13 F!Jel Category 4: OFA Burn up Requirement Coefficients Coefficients Decay Tune (yr.) A, Ai *& ~

0 0.4957 -6.0715 372851 -49.1282 5 0.7476 -8.7581 45.3241 -56.5172 10 0.9041 -10.4334 50.3246 -61.0800 15 1.0799 -12.2326 55.7508 ~6.1820 20 1.2541 -13.9154 60.5977 - -70.5720 Notes:

l.

• All relevant uncertainti~ are explicitly included in the criticality analysis. For instance, no

• additional ~lowance for burnup uncertainty or.enrichment uncertainty is required. For a fuel assembly t(? meet the requi~ments the assembly bumup must exceed the "minimum bumup" (G Wd/MI1J) given by the curve fit for the assembly "decay time" and "initial enric~t." If the computed minimum bumup value is negative, zero shall be used. The specific minimum burnup required for each fuel assembly is calculated from the following equation:

BU = A1

• En3 + A2

• En2 + AJ

• En + Ai [GWd/MTU]

2 .. Initial enrichmen~ En, is the maximum 235 U enrichment Any enrichment between 1.75 wt°/o 235 U and 5 wt°/o 235 0 may be used. Below 1.75 wt°/o 235 U, bumtip credit is not required.

3. An assembly with a decay time greater than 20 years must use the 20-year ( or less decay time) limits.

Table 6-14 Fuel Category 4: Example OFABurnup Requirements (GWd/MTU)

Radial Average Initial Enrichment:, wt°/o 235U Decay Time (yr.) 1.75 3* 4 5 0 0 21 .401 34.600 47.456 5 0 20.806 32.495 44.606 10 0 20.401 31.027 42.668 15 0 20.138 30.017 41.691 20 0 19.845 29.273 41.277 Note:

I. This table is included as an exam.ple, the bumup requirements will be calculated using the coefficients provided.

• WCAP-184l+NP Revision 0

'6-9 6.2 ANALYSIS AREA OF APPLICABILITY This section details the area of applicability of the analysis concerning assembly characteristics and associated fuel management, including a suminary of the data which needs to be confirmea to assure that the results presented here remain valid. Additionally, restrictions are given for other nonnal SFP conditions. Farley Units I & 2 have operated with the STD and OFA fuel designs. [

]11,C WCAP-18414-NP Revision 0

o-10 IL~~----T:..:....a....:..bl-e6---15_-_ _Cri_*_ticali..:__*ty_l--:--------]a,c-----:--,1 ~,

\

\ ...

WCAP-18414~

.. Revision 0

6-11 n,c WCAP-18414-NP Revision 0

6-12 8.C Additional restrictions for fuel storage are given here. One assembly pitch is defined as one cell in any direction, tncluding both face adjacent and corner adjacent cells.

• Fuel assembly evolutions (fuel cleaning, inspection, reconstitution, and sipping) must 'occur with at least one assembly pitch of water between the assembly in question and other assemblies. It is also acceptable t.o perform these actions above the top of the storage racks.

• Fuel assemblies stored with or1e or more rods missing, leaving a water hole, need to be stored as fresh fuel.

Fuel assemblies which have had fuel rods replaced with SS, natural uranium, or zirconium alloy rods may be stored as normal (by initial enrichment and bumup ). .

• Reconstituted fuel which contains fuel rods from other fuel assemblies wili be controlled as follows:

1. The fuel assembly enrichment will be assumed to be the higher of the in5;erted rod or reconstituted fuel assembly's initial enrichment;*and

2. The fuel assemb]y*bumup will be assumed to be the lower of the reconstituted rod Qr reconstiµrted fuel assembly's bumup.

• In all cases, only one fuel assembly will be manipulated at a fune and all manipulations will occur outside the storage cell and not within one assembly pitch of other assemblies.

• An inspection *can occur within the storage racks without-restriction if it does not invo_lve unborated water and nothing occurs within the assembly envelope or below the top of the active fuel. .

• Any storage cells considered damaged (outside of their allowable tolerances) cannot be used to store fuel assemblies without further evaluation. These ~maged cells may be used to store as non-fuel assembly components such failed fuel baskets in a storage array.

WCAP-18414-NP Revision 0

6-13 6.3 _SOLUBLE BORON CREDIT Soluble boron is credited in the Farley Units 1 & 2 SFPs to keep kc<<< 0.95 under all normal and credible

. accident scenarios. Under normal conditions, this requires less than 320 ppm of soluble boron: Under accident conditions the most limiting accident is the multiple misload accident requiring 17 l O ppm of soluble boron.

WCAP-18414-NP Revision 0

7-1 7 REFERENCES

1. Westinghouse Document WCAP-16045-P-A, Rev. O; "'Qualification of the Two-Dimensional Transport Code PARAGON," August 2004.

2. "SCALE Code System, ORNUfM-2005/39, Version 6.2.3, March 2018.

3. EPRJ Report 3002010613, "Benchmarks for Quantifying Fuel Reactivity Depletion .

1 Uncertainty-Revision l," Adams Number ML18088B397,_ Electric Power Research Institute, 2017. *

/

4. EPRI Report 3~010614, "Utilization of the EPRI Depletion Benchmarks for Burnup Credit Validation-Revision 1," Ad81l).5 Number ML18088B395, Electric Power Research Institute, 2018.

5. V. Kucukboyaci, "EPRJ Depletion Benchmark Calculations Using PARAGON," ANS NCSD, October 2013:

6. NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," U.S. Nuclear Regulatory Commission, February 2000.

7. Westinghouse Document WCAP-9522, "FlGHTH -A Simplified Calculation of Effective Temperatures in PWR Fuel Rods for Use in Nuclear Design," May 1979:

8. . ~- Wood, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools," USS-ISG-2010-001, Accession Number ML102220567, Nuclear Regulatory Commission, Rockville, MD, August'2010.

9. NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup*Credit

• Analyses," Oak Ridge National Laboratory, March 2003.

I 0. NUREG/CR-6760, "Study of the Effect of Integral Burnable Absorbers on PWR Burn up Credit,"

Oak Ridge National Laboratory, March 2002.

11. NUREG/CR-6761, "Parametric Study of the Effect of Burnable Poison Rods for the PWR Bornup Credit," Oak Ridge National Laboratory, March 2002.

12. L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at

. Light-Water Reactor Power Plants," Nucl~ Regulatory Commission, Rockville, MD, August 1998.

13. NUREG/CR-7109, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (kotr) Predtctlons," Oak Ridge National Laboratory, April 2012.

WCAP-18414-NP Revision 0

7-2

14. NEI-p-16, Revision 3, "Guidance for Performing Criticality Analyses of Fuel Storage at Light W~r Reactor Power Plants," March 2018.

15. C.E. Sanders, et al., "Parametric Study of the Effect of Control Rods for PWR Burnup Credit,"

NUREG/CR-6759, Oak Ridge National Laboratory, Oak Ridge, TN, February 2002.

16. NUREG/CR-6979, "Evaluation of the French Haut Taux de Combustion (HTC) Critical

• Expe~ment Data," U.S. Nuclear Regulatory Commission, Septem1?et" 2008.

17. of ML19189Al 11, ~'Final Safety Evaluation by the Office Nuclear Reactor Re~ation Topical Report 3002010613, 'Benchmarks for Qualifying Fuel Reactivity Depletion Uncertainty-Revision l' and Topical Report 3002010614, 'Utilization of the EPRI Depletion

• Benchmarks for Burnup Credit Validation-Revision l,"' U.S: Nuclear Regulatory Commission,

' ' I 2019. .

WCAP-18414-NP

• Revision 0

. A-1 APPENDIXA VALIDATION OF SCALE 6.2.3 Al INTRODUCTION This validation suite is intended to be used for fresh and spent fuel storage in the Farley Units 1 & 2 Spent Fuel Pool Criticality Safety Analysis. [

In order to validate the Scale Version 6.2.3 code system with the 238-group ENDF/B-VII library (referred to hereafter as Scale) for the Farley Units I & 2 Spent Fuel Pool criticality safety analysis, guidance from the NRC publication "Guide for Validation of Nuclear Criticality Safety Calculational Methodology" (Reference Al) was used and, as recommended in Reference Al, the "International Handbook of Evaluated Criti.cality Safety Benchmark Experiments" (Reference A2), has been used as the primary source of critical benchmarks for the validation effort. References A3 through A 7 were also used as fu:iditional sources of critical benchmarks.

Section 3 in each of the "International Handbook of Evaluated Criticality Safety Benchmark Experiments". (Reference A2) individual evaluations provides benchmark material compositions as

• tmmber densities which were reviewed and used for modeling experiments.

• Per Reference Al, the following are important parameters when defining the_area of applicability of a benchmark suite: fissile isotope, enrichment of the fissile isotope, fuel.density, fuel.chemical form, type of neutron moderators and reflectors, range of moderator to fissile isotope, neutron absorbers, and physical configurations. Therefore, these were the parameters considered when choosing which critical experiments to include in this validation suite.

This validation suite is designed to cover fresh and spent fuel storage for Farley Units 1 & 2. It 'also covers the criticality analysis of all normal operations and postulated accidents 'in the SFPs and fresh fuel storage. The validation is adequate to cover all present and anticipated (non-mixed-oxide) light water

~ctor (LWR) fuel designs at Farley.

WCAP-18414-NP Revision 0

,J A-2 A.2 METHOD DISCUSSION The validation methods recommended in Reference Al are the basis of this validation of Scale for nuclear criticality safety calculations. The code methodology bias and the uncertainty associated with t~e bias will be used in combination with other biases and uncertainties, as well as additional subcritical margin to ensure 1he regulatory requirements are met St?tlstical analysis is performed to determine whether trends exisrin the bias for three subsets of experiments; fresh fuel with strong absorbers, fresh fuel without strong absorbers, and fresh and burnt fuel with strong absorbers. No critical experiments containing Gadolinia, IFBA or Erbia were used because they will not be credited in the Farley Criticality Safety Analysis either as fresh or residual absorbers.

[

111,C According to NUREG/CR-6979, "Evaluation of the French Haut Taux de Combustion (HTC) Critical

~periment Data" (Reference A4), the HTC experiments are a series of experiments performed with mixed oxide rods designed to have U and Pu isotopic compositions equal to that of. UOi PWR fuel with initial enrichment.of 4.50 ,Wt°/o 235U at 37,500 MWd/MTU bumup. No fission products are included in the compositions. The HTC experiments are included to ensure the validation suite covers spent fuel as well.

Normality testing for.the data subsets is performed as outlined in References Al and A8 using the Shapiro-Wilk test'for data sets with a sample size of 50 or less and the D' Agostino normality test for the data sets with a sample size of more than 50. For the. cases which fail the normality tests, the non-

.parametric statistical treatment recommended-in Reference Al is _used.

A.2.1 Tesffor Normality (Goodness-of-Fit Test)

As stated in Re(erence Al, the statistical evaluation performed must be appropriate foi: the distribution of the data. A goodness-of-fit I

test is a procedure designed to examine whether a sample has come from ' a postulated distribution. Among the methods for testing goodness-of-fit, some are superior to others in their sensitivity to different types of departures from the hypothesized distribution. Some of the tests are quite general in that they can apply to just about any distribution, while other tests are more specific, such as tests that apply only to the nonnal distribution. [

]a,c A.2.1.1 Shapiro-Wilk Test for Normality References Al and A8 discuss the Shapiro-Wuk test for normal.ity (W-test). The W-test is applicable when neither the population mean (µ) nor t):ie population standard deviation ( o) is specified. The W-test is considered an omnibus test for normality because of its superiority to other procedures over a wide range of problems and conditions that depend on an-assumption of normality. The W-t~tis superior to the chi-square test (used by USLSTATS from Scale package) in many situations. This analysis thus uses the W-WCAP-18414-NP Revision 0

A-3 test as reco~ended in Reference AS for sample sizes between 3 and 50, the range over which Table T-6b provides the critical value Wq (n).

The null and alternative hypotheses are:

Ho: The sample.comes from anormal_distnbutjon.

• H!: The underlying distribution is not normal.

The W-test statistic is: __

82 W=-- Equation A-1 (n-t).52 where, n is the number of experiments in the group, s1 is the dataset variance, and

• Equation A-2 where, k = n/2 if n is even or (n-1)/2 ifn is odd a, = i coefficients obtained from Table T-6a ofNUREG- ~ 475, "'Applying Statistics" (Reference A8) associated with sample size n -

{Y(t), Y(z), ... , Y(n)} is the-normalized ke!f of each ~xperimeni arranged in ascending order The null hypothesis Ho oJ normality is rejected at the a level of signi:fiQaJ1ce ,if the calculated value of Wis less than the critical value Wq (n) obtained from Table T-6b of Reference A8. Note that in-this table, the quantile q=a.

. A.2.1.2 D' Agostino*Test for Normality Reference A8 discusses the D' Agostino test for normality (D' test). Like the W-test, the D' test is also applicable when neither µ nor cr is specified. Like the W-test, the D' test is also considered .an omnibus test for normality because of its superiority to other procedures o'ver a wide range of problems and conditions that depend on an assumption of normality. The D' test complements the W-test, which is used for samples no _larger than 50, and can be used for any sample size greater than 50.

The null and alternative hypotheses for D' test are:

Ho: The sample comes from a normal distribution.

H1: The underlying distribution is not nonnaL WCAP-18414-NP Revision 0

A-4 The test statistic is:

D' = T . Equation A-3

,/s2 (n-1) where, n is the number of ~periments s2 is the dataset variance T -~ L.,f=l

~n (*

L-(n+l))

Y(i) Equation A-4

{y(i), y c2 ), *** , y (n)} is the normalized ~ of each experiment arranged in ascendi~g order

[

]"-c The D' test involves a comparison of the calculated D' value with tw<? quintiles from Table T-14 of Reference AS. The test is two-sided and requires two critical values that pound a noncritical region. For each combination of n and a, the critical values are found in Table T-14 under the row that corresponds ton and the columns for 'lan(n) and qi. an(n). If the calculated D' is not between these two values, the null hypothesis is rejected.

If the null hypothesis is rejected~ a non-parametric treatment may be applied.'lf the null hypothesis is not

.rejected, tJ:ien a technique _such as a one-sided tolerance limit described in Ref~rence Al can 1?e used to determine the appropriate bias and bi~ uncertainty.

• A.2,2 Determination of Bias and :1Jias'Uncertainty The statistical analysis presented in Section 2.4 of Reference Al is followed for all datasets that passed the appropriate test for Normality. This approach involves determining a weighted mean that incorporates the uncertainties from poth the measurement (crcoq,) and the calculation method (crco1c). The benchmark experiments ~hosen from References A2, A6, and A7, use the experimental uncertainties presented in .

References A2,'A6 and A 7, respectively. Experimental uncertainty is not presented for the experiments contained in NUREG/CR-6361, "Criticality Benchmark Guide for Light-Water-Reactor-F~el in Transportatiqn and Storage Packages" (Reference A3), so the average value of experimental uncertainties of similar experiments documented in Reference A2 is used. This is consistent with the recommendation in Reference Al that engineeringjudgment be used to approximate typical experimental uncertainties rather than assume no experimental uncertainty.

If the critical-experiment being modeled is at a state other than critical (i.e., k-:f:, 1.0) then~ adjustment is made to the calcul!ited vah,ie of k,,ff. This adjustment is done by non:nalizing the calculated eigenvalue to the experimental value. This normalization assumes that the inherent bias in the calculation is not affected by the normalization, which js valid for small differences in k.,ff. To normalize 'koir, the calculated keff Ckcaic) is divided by the kcff evaluated in the expe~ent Ckcxi,):

WCAP-18414-NP Revision 0

A-5 I, - kca1c Equation A..:s

"-nonnal - --

kv:p The normalized kcir(lc..ormai)values are used in the subsequent detennination of the bias and bias uncertainty, therefore all subsequent instances of kcJr sho~ld be taken to mean the normalized kett-value.

)be Monte Carlo calculational uncertainty (uca1c) and experimental uncertainties (uexp) are root-sum-sq~ to create a combined uncertainty (o"t) for each experiment:

Equation A-6 A ~eighted.~ kc1r (keff) is ~Jculated by using the 'weighting factor IfCJ[. The use of this factor reduces the "weight" of the data with high un~nty. Within a set of.data, the "i"'" member of that set is shown with a subscript "i." Henceforth, unless otherwise specified, the ~mbined uncertainty CJr for an.

"i11,,, k..1r is shown as a;. The weighted equation varia~les for the single-sided lower tolerance limit are as follows:

• Variance about the mean:

Equation A- 7 Ayerage total uncertainty:

-2 n

(}" =-1 Equation A-8 L~

The weighted mean kc1r value:

~quation A-9 The square root of the pooled variance:

Sp= ..Jsz + 7i2 Equation A-10 .

where,.

s1 = variance about the mean n = number of critical experiments used in the validation ff = average total uncertainty

- WCAP-18414-NP

• Revision 0

A-6 Bias is determined by the relation:

Bias= ketr LO if~~11<LO Equation A-11 0.0 if keff"2LO Reference Al states that when a relationship between a calculated kcfrand an independent variable cannot

• . be determined (no trend exists), a one-sided lower tolerance limit should be used. This method provides a

• single lower limit above which a defined fraction of the~ population ofkcir is expected to lie, with a prescribed confidence and within the area of applicability. Use of this method requires the experimental results to have a normal statistical distnbution: Lower tolerance limits, at a minimum., should be calcul~ with a 95% confidence .that 95% of the data lies above Kr,. The equation for the one-sided lower tolerance band from Reference Al is:

K_L = keff - USp Equation A-12 Or, if keff ~ l, KL= 1- USp 'Equation A-13 Where, Sp is the pooled variance, U is the one sided lower tolerance fact~r (found in Table T-11 _b of Reference A~ where n is the number of experiments contained in the data set).

USP is then taken as the uncertainty to the untrended bias (~trended bias uncertainty).

A.2.3 Identify Trends in the Data Trends.are determined using regression fits to the calculated results. Based on a visual inspection of the data plots, it is determined that a linear fit is sufficient to* evaluate whether there is a trend in the bias. In .

• the following equations, "x" is the independent variable representing the parameter of interest ( e.g.,

enrichment). The variable 'Y' represents kctf. Variables "a" and "b" are coefficients for the function where "b" is the slope _and "a" is the intercept The function Y(x) represents /vi,(x).

I

• Per Reference Al, the equations used to produce a weighted fit of.a straight line to the data are given in this section.

Y(x) =a+bx Equation A-14 where, WCAP-18414-NP Revision 0

A-7 Once the data has been fit to a line, a determination as to the "goodness of fit" must be made. Per Reference A 1, two steps should be_ employed when determining the goodness of fit The first step is to plot the data against the independent variable which 'allows for a visual evaluation of the ~ffectiveness of the regressiori fit.

The second step is to numerically determine a goodness of fit after the linear .relations are fit W the data

. This adds a useful ~ure because visual in~tion of the data plot will not necessarily reveal just how good the fit is to the data. Per Reference Al, the linear correlation coefficient is one standard method used to numerically measure the linear association between the random variables x and y.

The sample correlaticm coefficient between x and y (linear-correlation coefficient) is a* quantitative measure of the degree to which a linear association exists between two variabl~. For weighted data, the linear correlation coefficient is:

L;f (x1- .f)CY1- Ji)

T = -;=::;======:-;::=== Equation A-15

. j:£~ (Xi- £) JL-tr CY1- Y) 2 2 where, The weighted mean for the independent parameter is:

Equation A-16

• The weighted mean for the dependent paramet~r (y) is keff-The value of r1 is the coefficient of determination. It can be interpreted as the percentage of variance of one variable that is predictable from the other variable. The closer r1 approach~ the value.of l, the better the fit of the data to the linear equation. Note that the value.of a sample correl'ation coefficient' r shows .

only the extent to *which x and y are linearly associated. It does not by itself imply that any sort of causal relationship exists between x and y.

In addition to the linear correlation coefficient, the Student's t test is used to determine if the trend in the

• linear fit of the data is statistically significant. A trend is statistically significant when the slope of the linear regression fit (b) is equal to some specified value (bo). For the purposes of this validation suite, the null hypothesis, Ho. bo = 0 is that no statistically sigruficant trend exists (slope is zero) with an alternative*

• I hypothesis of Hi: bot= 0, at a significance lev~l of a= 0.05.

In order to determine if the null hypothesis is supported, t..,_,on, is calculated and compared to the Student's t distribution (ta11.11-1). The fscoro for the slope of a regression line is given by:

WCAP-18414-NP Revision 0

A-8 (b- /Jo),/n-2 tscare = I SS5 Equation A-17

~l:C:rri52 where, SSE is the sum of the squares of the resjduals:

• SSE= :E[kefft - (a+ bxt)]2 Equation A-18 The null hypothesis is rejected if ltscorel > t<¥2, 11-1.

When Ho is rejected and a statistically significant trend is_ determined, the trended value of a bias and its

. ~iated uncertainty are used when it is more restrictive than the untrended value. of the bias. In the area where untrend~ bias yields more restrictive value, the untrended bias and its associated uncertainty are used.

Per Reference Al, when a relationship between a calculated k..trand an independent variable can be determined (the trend exists), a one-sided lower tolerance band may be used: This conservative method provides a fitted curve above which the true population of kcff is expected to lie. The equation for the one-sided lower toleranc*e band from Reference* A 1 is:

. { ...(2,n-2) r.!. (r-x)2 ] .* (n-2) }

• Ki(x) = Kia ( x ) - Spftt 21'~ i;;+"Cr,:...x' 2 + ZzP-l L, A)

_2

,{1-y,n.:.2 EquationA-19.

Kft,(X) is the function derived in the trend analysis described above. Because a positive bias may not be conservative, the following equation must be used for all values of x where Ktit(x) > 1:

Ki(X) =. 1 - Spftt { 2F(Z,n-Z) a

[.!.n +.'°(r,-x'2 (x-x)2 ] + z.

ZP-1 (n-2) }

..:i.

Equation A-20 L, A) A1-y",n-2 where, p The desired confidence fovel (0.95)

F?.n-z) = The F distribution percentile with degree of fit, n-2 degrees of freedom. The degree of fit is 2 for a linear fit.

n The number of critical experiment ket-rvalues x = The*independent fit variable x, The independent parameter in the data set corresponding to the ith kc££ value x The weighted mean of the independent variables WCAP-18414-NP Revision 0

. \

. A-9 Z2P-1 = The symmetric percentile of the normaJ distribution that contains the P fraction y = 1-p 2

Xi-y,n:-2 = The upper Chi-square percentile Fot a weighted analysis:

Equation A-21

' 1

.- I:-;;rt

_ _:i_

X - 1 Equation A-22 r-uf Equation A-23 n

u-2 =-1 Equation A-24 L~

l Equation A-25 Within the equation for ~:

Bias(x) = Ktit-1.0 If Kttt<l.O Equation A-26 0.0 , if Kttt"i!l.O Arid the uncertainty in_the bias is:

95/95 Bias ~nc~ainty(x) = Sp/It { 2F;2..n-2

) [! + L~x-~z] +

n x 1--:x Zzp-1' ;n- 2

) }'

X1-y,11-2 Equation A-27 When Ho is rejected and a statistically significant trend is determined, the trended value of a bias and its as~ciated uncertainty should be used while it is.more restrictive than the untrended value of the bias*. In the area where an imtrended bias yields a more restrictive value, the untrended bias and its associa~

uncertainty shall. be used.

• A.2.4 *Non-Parametric Treatment If the data fails the test for normality, a non-parametric treatment of the data will be necessary. Per Reference Al, the determination of KL, the lower limit of the 95/95 tolerance interval is as follows:

WCAP-18414-NP Revision 0

A-10 Ki= k;j!}1- uncertainty fork~ - NPft! Equation A-28

where, k~ is the minimum (smallest) normaJiz.ed kcrr in a dataset, uncertainty for k~n is the pooled Monte Carlo and experimental uncertainty, and NPM is the non-parametric margin, which is added to accmmt for the small sample size.

The non-parametric treatment outlined in Reference Al uses the order statistics-to represent the characteristics of a dataset after it has been ranked (ordered) from the smallest observed kctr(kzy;) to the largest observed 1cetr (ktj'f). The following ~tion is the general equation that determines the percent confidence that a fraction of the population is-above the lowest observed valu~:

p = [1 - r:;;-o1 C,c:~1) (1 ~ q/ qn-j ] X100%, Equation A-29 where:

q is the desired population fraction (normally 0.9~)

n is the number of data val'ues in one data set m is the rank order indexing from the smallest sample value to the largest (m = 1 for the smallest sample value; m = 2 for the second small~ sample value, etc.). *  :

The smallest observed keft' has the rank order index I and the largest observed ketr has the rank order index 1

equal tq the number of observations. Thus, for a desired ~pulation fracti~n of 95% and k~ (rank. ~rder index I), the percent confidence that a fraction of the population of n data points is above the lowest observed value is:

p = (1 - 0.95n) X 100% Equation A-30 Similarly, for a desired population fraction (qfof95% and the 2 00 lowest ketr(rank order index m=2), the percent confidence that a fraction. of the population.of n data points is above the second lowest observed value is:

P ~ [1 - [0.95' + (,:;)! *(1 - 0.95)

• 0.95*- J]

1 X JOO% Equation A-31 Although 59 experiments would be required to reach a 95/95 tolerance limit for rank order 1 as stated in Reference Al, the recommended non-parametric margin (NPM) correction is 0.0 for confidence values greater than 90 percent, as also indicated in Table 2.2 of Reference A 1.

WCAP-18414-NP Revision 0

A-II Within the equation for KL:

min min Bias = 1'etf - 1.0 if ketf <LO Equation A-32 0.0 if k:t}';;,:.1.0 And the uncertainty in the bias is:

Equation A-33

where,

<rcalc 2 is the Monte Carlo uncertainty frpm the selected rank order case r 2

<Texp is the experimental uncertainty from the selected rank order case A rank order of 2 may be used in determination of.the percent of confidence that a fraction of the population is above the lowest observed value if the sample size is greater than 93. This effectively means that the second lowest k.,1ris used for the determination of the Bias and Bias uncertainty. Employing the aforementioned methods will produce a particularly conservative bias and bias uncertainty and negate the need for any further trending analysis for a non-parametric data set.

A.3 DESCRIPTION OF CRITICAL EXPERIMENTS Many studied series of the critical experiments allow using a simplified model with some zones homogenized.or omitted. Only the complete model provided in Section 3.0 of each evaluated series of experiments is used for keff determination.

A.3.1 [

]a,c WCAP-18414-NP Revision 0

A-12

[

A.3.2 [

[

A.3.3 [

]a,c

[_

]B,C WCAP-18414-NP Revision 0

A-13

]a,c A3.4 [

WCAP-18414-NP Revision 0

A-14

]11,C A.3.5 [

1a.c

]a,c WCAP-18414-NP

• Revision 0

A-15 A.3.6 [

1a.c

]a.c A3.7 [

]a,t

[ '

]o.,c WCAP-18414-NP Revision 0

\,_'

A-16

[

]a,c A.3.8 [

[ .

]a.o WCAP-18414-1';-IP Revision 0

A-17 A.3.9 [

1a.c

[

. ]a,c A.3.10 [

]a,c

[

WCAP-18414-NP*.

Revision 0

A-18

[

]a,c A.3.11 [

]a,c WCAP-18414-NP .

Revision 0

A-19 A3.12 [

WCAP-18414-NP Revision 0

A-20

]a,c A.3.13 [

]a,c

' [

]a,c A.3.14 [

]11,C WCAP-18414-NP Revision 0

A-21

[

\

]"'"

WCAP-18414-NP Revision 0

A-22

]o.,c 11,C Table A~t Benchmark Values ofknrand Respective Uncertainties

)

A.3.15 [

]a,c

]8,C WCAP-18414-NP Revision 0

A-23

[

A.3.16 I ]a,c NUREG/CR-6361 ~eference A3) is intended as a guide for performing criticality benchmark

• *calculations for LWR fuel apJ?lications. It documents 180. critical experiments and includes recommendations for selecting suitable experiments and determining the calculational bias and bias uncertainty. When selec~ng experiments, pref~rence is given to Reference A2 because it is more current than Reference A3. [

WCAP-18414-NP Revfaion 0

A-24 TableA-2 a,c

].;c -

A.3.17 HTC Experiments I -

The HTC experiments are a series of experimerrts perfonned with mixed oxide rods designed to have a U and Pu isotopic composition. representative to that of U(4.5%)0i PWR fuel with 37;500 MWd/MTU bumup. No fission produ~ are focluded in_the composition. Up to this point, all the experiments modeled in this suite represent fresh fuel; the HTC experiments are included to ensure the validation suite covers *spent fuel as well. The HTC critical experiment set was taken from two phases:

• - Phase 1 - Water-Moderated and Reflected Simple Arrays ~eference AS)

WCAP-18414-NP Revision 0

A-25

• Phase 2- Reflected Simple Arrays Moderated by Water Poisoned with Gadolinium or Boron

-(Reference A6)

• Phase 3 - Pool Storage (Reference A7)

Reference A4 is an ORNL evaluation of the HTC experiments. [

]a,c WCAP-18414-NP Revision 0

A-26

]a,c WCAP-18414-NP

• Revision 0

A-27 A.4 RAW CALCULATION RESULTS

[

]a.c Definitions of ka.ic, ke:q,, knonnaI and their associated uncertainties are explained in Section A.2.

[

WCAP-18414-NP Revision 0

A-28 Table A-3 [ a,c

\VCAP-18414-"NP September 2019 Revision 0

A-29 Table A-3 ]a.c (cont.) u.c .

)

WCAP-18414-NP September 2019 R~vision 0

A-30 Table A-3 ]a.c (cont.) u.c WCAP-18414-NP September 2019 *.

Revision 0

A-31

• ,Table A-3 ]a.c (cont.) [1.0

/

WCAP-18414-NP September 2019 Revision 0

A-32 Table A-3 1a,c (cont) n.c

• WCAP-18414-NP September 2019 Revision 0

A-33 Table A-4 ]a,c ... n,c .

WCAP-18414-NP September 2019 Revision 0

A-34 1------r---T:_a_bl_e_A_-_4__[_ _.--.......------,---....----....---.----...-------T-- -.-----,--"""T""--]-""'.--(c_o_n_t.)-r--~l 11,c WCAP-18414-NP September 2019 Revision 0

A-35 Table A-4 ]a.c (cont.) u,c '

(

WCAP-18414-NP. September 2019 Revision 0

A-36 Table A:-4 [ *

• 1a,c (cont.) U,C

• WCAP-18414-NP . September 2019 Revision 0

A-37 TableA-5 111,(1 a.c WCAP-18414-NP September 2019 Revision 0

A-38 Table A-5 * ]a.e (cont.) a.c

. I WCAP-18414-NP September 2019 Revision 0

A-39 TableA-5 1a,c (cont) n.c WCAP-18414-NP September 2019 Revision 0

A-40 Table A-5 1n,c (cont.) !l,C WCAP, 18414-NP September 20 f 9 Revision 0

A-41 Table A-5 ]a.c (cont.) *1* !l,C WCAP-18414-NP September 2019 Revision 0

A-42 Table A-6*

1a.c n,c WCAP-18414-NP September 2019 Revision 0

A-43

-I._~-T_a_b_le_A_-6_ _ _~~--~-----.----'--..---'---'---,].-a.c_ _ ~-----.---~-~--~---1 11.C WCAP.18414-NP September 2019

• Revision 0

A-44 Table A-6 1a.c ll,C WCAP-18414-NP September 2019 Revision 0

- A-45 Table A-6*

u,c WCAP. 18414-NP September 2019 Revision 0

A-46 n,c WCAP-18414-NP September 2019 Revision 0

A-47 Table A-6

]a.c U,O WCAP-18414-NP September 2019 Revision 0

A-48 A.5 DATA SET NORMALITY ASSESSMENT A.5.1 [

[

111,C Table A . [

]a,c a,c

]11,C A.5.2 [ ]a,c

']_'able A-8 1a.c I a,c WCAP-18414-NP September 2019 Revision 0

A-49

[

]""'

A.5.3. [ 1a.c

[

]a,c Table A-9 1a.c I a,c

]a.c Note that the quintiles, D'q(n), of the distribution of the D' statistic i_n_ Reference A8 are provided only for even n. For the odd n, linear interpolation is used between adjacent values.

A.5.4 [ p,c

]a,c a,c WCAP-18414-NP 'September 2019 Revision 0

A-50

[

]a.<:

WCAP-18414-NP September 2019 Revision 0

A-51 A.6 TRENDING ANALYSIS Of the four main datasets evaluated, trendtng analysis will only be performed on the [

r,c The regression fits and go_odness of fit tests "described in Section A2.1 are applied [

1*

]""'

A.6.1 [ 1a.c

[ .

]""'

TableA-11 [

]a,c I Statistical Data Parameter c WCAP-18414-NP September: 2019 Revision 0

A-52 a.c FigureA-1 ]a,c a,c Figure A-2 [

WCAP-18414-NP September 2019

• Revision 0

A-53 a,c Figure A-3 [ ]a,c a,c Figure A,.-4 [

WCAP-18414-NP September 2019 Revision 0

A-54

[

r,c A6.1.1 ]a,c

]~

Table A-12 [

a,c

]a,c WCAP-18414-NP September 2019 Revision 0

A-55 TableA-13 [

a,c A.6.1.2 ]a,c

]a,c WCAP-18414-NP September 2019 Revision 0

. A-56 Table A-14 a,c

]a,c WCAP-18414-NP September 2019 Revision 0

A-57 a,.c Figure A-5 1-..: .

WCAP-18414-NP September 2019 Revision 0

A-58 TableA-15 ]a,c I Parameter a,c

]a,c WCAP-18414-NP September 2019 Revision 0

A-59 A.6.2 [

I,

[

Table A-16 . ]:a,c a,c WCAP-18414-NP September 2019 Revision 0

A-60 A.7 AREA OF APPLICABILITY

[

TableA-17 [ a,c WCAP-18414-NP September 2019 Revision 0

A-61 TableA-17 [ 1-.c a,c

]a.<:

WCAP-18414-NP September 2019 Revision 0

A-62 A.8 VALIDATION

SUMMARY

[

]a.c Table A-18 Summary of Biases and Bias Uncertainties Determination a,c

. [

with th.e trended bias uncertainty determined as

[ '.

]a,c WCAP-18414-NP September 2019 Revision 0

A-63 A.9 REFERENCES Al. NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," Science Applications International Corporation, January 2001.

A2. "'International Handbook of Evaluated Criticality Safety Benchmark Experiments,"

NENNSC/DOC (95) 03, September 2013.

A3. NUREG/CR-6361; "Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages," U.S. Nuclear Regulatory Commission, 1997.

A4. NUREG/CR-6979, "Evaluation of the Fren~h Haut Taux de Combustion (HTC) Critical Experiment Data," U.S. Nuclear Regulatory Commission, September 2008.

AS. F.Femex, "Programme HTC-Phase 1: Reseaux de crayons dans l'eau pure (Water-moderated and reflected simple arrays) Reevaluation des experiences," DSU/SEC/T/2005-33/D.R., Institut de Radioprotection et de Sfirete Nucleaire, 2008.

A6. F.Femex, "Programme HTC - Phase 2: Reseaux simples en eau empoisonnee (bore et gadolinium) (Reflected simple arrays moderated by poisoned water with gadolinium or boron)

Reevaluatibn.des experiences," DSU/SECfr/2005-38/D.E.., Jnstitut de Radioprotection et de

. SOrete Nucleaire, May 2008.

-A7.

• F.Femex, "Programme HTC:.:.. Phase 3: Configurations "stockage en piscjne" (Pool storage)

Reevaluation des experiences," DSU/SECfr/2005-37/D.R., Institut de Radioprotection et de Sfi!et:e Nucleaire, May_2008.

A8. D. Lurie et al. "Applying Statistics," NUREG-1475, Revision I, U.S. Nuclear Regulatory Commission, March 201 I.

A9. R.I. Smith et al., "Clean Critical Experiment Benchmarks for Plutonium Recycle in LWR's,"

Volume I, EPRI NP-196, Electric Power Research Institute, April 1976.

AI 0. S.R. Bierman~ "Criticality Experiments with Neutron Flux Traps Containing Voids,"

PNL-7167, Pacific Northwest Laboratory, April 1990.

Al I. L.W. Newman, "Urania-Gadolinia: Nuclear Model Development and Critical Experiment Benchmark," DOE/ET/34212-41 (BAW-1810), Babcock & Wilcox, April 1984.

Al2. T.C. Engelder et al., "Spectral Shift Control Reactor Basic Physics Program, Criti~l Experiments on Lattices Moderated by Dz0-H20 Mixtures," BAW-1231, Babcock & Wilcox Company, December 1961.

A13. T.C. Engelder et al., "Spectral Shift Control Reactor Basic Physics Program, Measurement and Analysis of Uniform Lattices of Slightly Enriched UOi Moderated by DiO-H20 Mixtures,"

BAW-1273, Babcock & Wilcox Company, November 1963.

WCAP-18414-NP Septembe( 2019 Revision 0