Oconee Nuclear Station Mark-B-HTP Fuel Transition ...methods to evaluate the mixed-core effects of...

Oconee Nuclear StationMark-B-HTP Fuel Transition Methodology

DPC-NE-2015

September 2007

Nuclear Engineering DivisionNuclear Generation Department

Duke Energy Carolinas, LLC

Abstract

This report describes the methodologies to be used by Duke Energy Carolinas,LLC (Duke) for performing the core reload design, the fuel assembly mechanicaland thermal-hydraulic analyses, and the UFSAR Chapter 15 non-LOCA transientand accident analyses, for the transition to the AREVA NP Mark-B-HTP fuelassembly design at the Oconee Nuclear Station. Included in this report are themethods to evaluate the mixed-core effects of the Mark-B-HTP fuel design withthe current Mark-B 11 fuel design. The main differences between the currentMark-B 11 design and the Mark-B -HTP design are that the Mark-B -HTP designhas larger fuel pellet and rod diameters, thicker cladding, a different spacer griddesign, a different axial pressure drop profile, and different critical heat fluxcorrelations. These methods are presented as revisions to Duke's existingmethodology reports that have been previously approved by the NRC. Themethodology revisions also include changes that are not associated with thechange in the fuel assembly design. These changes are enhancements to theexisting methods to improve analytical margins, to correct errors, and to provideeditorial clarification. Technical justification is provided for all of themethodology changes. A brief summary of the AREVA NP methods forperforming the LOCA analyses consistent with the requirements of 10 CFR50.46 and Appendix K is also presented. Required revisions to the OconeeTechnical Specifications and Bases, along with technical justification, arepresented.

DPC-NE-2015

MARK-B-HTP FUEL TRANSITION METHODOLOGY

TABLE OF CONTENTS

Pazes1.0 INTRODUCTION 1-3

2.0 MARK-B-HTP FUEL DESIGN SUMMARY 4-8

2.1 Design Description2.2 Operating Experience2.3 References

3.0 METHODOLOGY REPORTS OVERVIEW 9-12

3.1 Core Physics and Reload Design Methodology Reports3.2 Fuel Assembly Mechanical Design Methodology Report3.3 Core Thermal-Hydraulic Design Methodology Reports3.4 Transient and Accident Analysis Methodology Reports3.5 Inter-Relationships of Methodology Reports3.6 References

4.0 CORE PHYSICS AND RELOAD DESIGN METHODOLOGY REVISIONS 13-19

4.1 Revision 6 to NFS-1001A - Oconee Nuclear Station Reload Design Methodology4.2 Revision 3 to DPC-NE-1002-A - Oconee Nuclear Station Reload Design

Methodology II4.3 Mixed Core Effects4.4 References

5.0 FUEL ASSEMBLY MECHANICAL DESIGN METHODOLOGY REVISIONS 20-22

5.1 Revision I to DPC-NE-2008P-A - Fuel Mechanical Reload Analysis MethodologyUsing TACO3

5.2 References

ii

6.0 CORE THERMAL-HYDRAULIC DESIGN METHODOLOGY REVISIONS 23 -47

6.1 Revision 2 to DPC-NE-2003P-A - Oconee Nuclear Station Core Thermal-HydraulicMethodology Using VIPRE-01

6.2 Revision 4 to DPC-NE-2005P-A - Thermal-Hydraulic Statistical Core DesignMethodology

6.3 Appendix F to DPC-NE-2005P-A - Application of BHTP CHF Correlation6.4 Mixed Core Effects6.5 References

7.0 UFSAR CHAPTER 15 NON-LOCA TRANSIENT AND ACCIDENT 48-78ANALYSIS METHODOLOGY REVISIONS

7.1 Revision 4 to DPC-NE-3000-PA - Thermal-Hydraulic Transient AnalysisMethodology

7.2 Revision 3 to DPC-NE-3005-PA - UFSAR Chapter 15 Transient AnalysisMethodology

7.3 Appendix D to DPC-NE-3000-PA - Methodology Revisions for Mark-B-HTP Fuel7.4 Appendix E to DPC-NE-3000-PA - Expanded Oconee VIPRE-01 Methodology7.5 Mixed Core Effects7.6 References

8.0 AREVA NP LOCA ANALYSIS METHODOLOGY 79

8.1 Summary of Methodology8.2 References

9.0 TECHNICAL SPECIFICATION REVISIONS 80-829.1 Revision to -Technical Specification 2.1.1.2 - Reactor Core Safety Limits9.2 Revision to Technical Specification 5.6.5.b - Core Operating Limits Report (COLR)9.3 Revision to Technical Specification Bases B 2.1.1 - Reactor Core SLs9.4 Revision to Technical Specification Bases B 3.4.1 - RCS Pressure, Temperature, and

Flow Departure from Nucleate Boiling (DNB) Limits

10.0 CORE OPERATING LIMITS REPORT REVISIONS 8310.1 Reference 10- BAW-10164P-A10.2 LOCA Limits

iii

List of Tables

DPC-NE-2015

Table 2-1 Mark-B-HTP to Mark-B 11 Comparison

DPC-NE-2005-P, Revision 4

Table F-I Mark-B-HTP Fuel Assembly Data

Table F-2 VIPRE-01 BHTP Correlation Verification

Table F-3 CHF Test Database Analysis Results

Table F-4 Oconee SCD Statepoints

Table F-5 Oconee Statistically Treated Uncertainties

Table F-6 Oconee Statepoint Statistical ResultsBHTP Critical Heat Flux Correlation

Table F-7 Oconee Key Parameter Ranges


Table D-1 Mark-B-HTP Fuel AssemblyComponent Dimensions Used for Thermal-Hydraulic Analysis

iv

List of Figures

DPC-NIE-2015

Figure 2-1 Mark-B-HTP Design

Figure 2-2 Mark-B I I Design

Figure 7.5-1 Oconee [ ] Channel VIPRE-01 Model D


Figure 4-6 Mark-B-HTP 64 Channel VIPRE-01 Model



Figure F-I VIPRE-01 Predicted CHF Versus Measured CHFMark-B-HTP Data Base

Figure F-2 VIPRE-01 Predicted-to-Measured CHF vs. Mass VelocityMark-B-HTP Data Base

Figure F-3 VIPRE-01 Predicted to Measured CHF vs. PressureMark-B-HTP Data Base

Figure F-4 VIPRE-01 Predicted to Measured CHF vs. QualityMark-B-HTP Data Base


Figure D-1 Mark-B-HTP Fuel Assembly

Figure E-1 [ I Channel Oconee VIPRE-01 ModelSubchannel Detail

Figure E-2 [ ] Channel Oconee VIPRE-01 ModelLumped Channel Detail

Figure E-3 Hot Assembly Pin PowerDistributionGeneric Vendor Input (same as Figure 2.3-4)

Figure E-4 Adjusted Hot Assembly Pin Power DistributionFrom SIMULATE-3

V

Figure E-5(a)

Figure E-5(b)

Figure E-5(c)

List of Figures (cont.)

VIPRE-01 Pin Power Distribution Adjustment ProcessPin Delta-Power Values

VIPRE-01 Pin Power Distribution Adjustment ProcessAdjusted Delta-Power Values

VIPRE-01 Pin Power Distribution Adjustment ProcessRevised Pin Power Distribution

Vi

List of Acronyms

ANS American Nuclear SocietyASME American Society of Mechanical EngineersBOC Beginning of cycleCFM Centerline fuel meltCFR Code of Federal RegulationsCHF Critical heat fluxCOLR Core Operating Limits ReportDNB Departure from nucleate boilingDNBR Departure from nucleate boiling ratioEOC End of cycleEPRI Electric Power Research InstituteFSAR Final Safety Analysis ReportHFP Hot full powerHMP High mechanical performanceHTP High thermal performanceHZP Hot zero powerLCO Limiting condition for operationLOCA Loss-of-coolant-accidentMARP Maximum allowable radial peakMDNBR Minimum departure from nucleate boiling ratioNRC Nuclear Regulatory CommissionPT Pressure-temperatureRCP Reactor coolant pumpRCS Reactor Coolant SystemRTP Rated thermal powerSCD Statistical core designSCUF Statistically-combined uncertainty factorSDL Statistical design limitSRSS Square-root-sum-of-the-squareýSL Safety LimitUFSAR Updated Final Safety Analysis Report

vii

1.0 INTRODUCTION

The Oconee Nuclear Station is a three-unit pressurized water reactor of the Babcock & Wilcox(B&W) lowered-loop design, owned and operated by Duke Energy Carolinas, LLC (Duke). Thelicensed reactor power level is 2568 MWt. Duke has plans to transition to the AREVA NP Mark-B-HTP fuel design for core reloads beginning in 2008 or 2009. The current fuel assembly designis the AREVA NP Mark-B 11 design. The Mark-B-HTP design is different mainly in that the fuelpellets and cladding have larger diameters, the cladding is thicker, the spacer grids are a differentdesign, and there is.a different axial pressure drop profile. The Mark-B-HTP design is also moreresistant to grid-to-rod fretting due to several design features. The Mark-B-HTP design uses theAREVA NP BHTP critical heat flux (CHF) correlation, which has been previously reviewed andapproved by the NRC.

The purpose of this report is to submit for NRC review and approval the revisions to the NRC-approved methodology reports that Duke plans to use for the transition to the AREVA NP Mark-B-HTP fuel assembly design. These revisions include the methodologies for core reload design,fuel assembly mechanical and thermal-hydraulic analyses, and UFSAR Chapter 15 non-LOCAtransient and accident analyses. Also included in this report are the methods to evaluate themixed-core effects of the Mark-B -HTP fuel design with the current Mark-B II fuel design. Thereis also the possibility that the earlier Mark-B 10 design may be used as reinsert fuel in reloaddesigns along with the Mark-B-HTP and Mark-B 11 designs. Although not discussed further inthis methodology report, any use of Mark-B 10 fuel will be based on application of approvedmethods and similar mixed core models.

In this report Duke has also included methodology revisions that are not associated with thechange in the fuel assembly design. These revisions are enhancements to the methods to improveanalytical margins, or to make error corrections or editorial clarifications. Technical justificationis provided for all of the methodology changes. Some minor editorial revisions are also includedthat do not require technical justification.

The AREVA NP methods for performing the LOCA analyses consistent with the requirements of10 CFR 50.46 and Appendix K are also summarized. Other than the design details of the Mark-B-HTP fuel and the BHTP CHF correlation, the AREVA NP LOCA methodologies used for therevised LOCA analyses are the same as those used for the Mark-B 11 fuel design.

Revisions to the Oconee Nuclear Station Technical Specifications and associated Bases, and thetechnical justification for those revisions are included. Revisions to the Core Operating LimitsReport are also included.

The following paragraphs describe the contents of each chapter in this report. The references areprovided within each chapter later in the report.

Chapter 2.0, Mark-B-HTP Fuel Design Summary, describes the AREVA NP Mark-B-HTP fueldesign and compares it to the Mark-B 11 design currently in use at Oconee. Also included in thischapter is information on the industry operating experience with the Mark-B-HTP design. Theintent of this chapter is to capture the differences between the current and future designs, so thatthe new and revised methodologies that are proposed can be shown to address all of the designdifferences.

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Chapter 3.0, Methodology Reports Overview, provides an overview of Duke's NRC-approvedmethodology reports that are applicable to Oconee and their inter-relationships in the core reloaddesign and the UFSAR Chapter 15 safety analysis licensing basis. The intent of this chapter is tofacilitate an understanding of the various Duke methodology reports, so that the context of theproposed new and revised methodologies can be understood.

Chapter 4.0, Core Physics and Reload Design Methodology Revisions, describes the revisions toDuke's methodology reports associated with neutronics codes and core reload design. Dukeemploys the Studsvik Scandpower CASMO-3/SIMULATE-3 code system for Oconee physicsanalysis. Revisions to the methodology are described along with technical justifications andmarkups to the current revisions of the NRC-approved reports. The intent of this chapter is toconsolidate the new and revised neutronics and physics methodology in one chapter to facilitatethe NRC review of these related subjects.

Chapter 5.0, Fuel Assembly Mechanical Design Methodology Revisions, describes the revisionsto Duke's methods for analyzing the mechanical design of the fuel pellets and fuel rods. Dukeuses the AREVA NP TACO3 code as the primary mechanical and thermal analysis code.Revisions to the methodology are described along with technical justifications and markups to thecurrent revision of the associated report. The intent of this chapter is to capture all of the new andrevised fuel assembly mechanical design methodology in one chapter to facilitate the NRCreview.

Chapter 6.0, Core Thermal-Hydraulic Design Methodology Revisions, describes the revisions toDuke's methods for analyzing the thermal-hydraulic performance of the Mark-B-HTP design.Duke uses the EPRI VIPRE-01 code for thermal-hydraulic analyses. The AREVA NP BHTPC-F correlation is used to develop the correlation DNBR limit using VIPRE-01, and also todevelop the statistical DNBR limit using Duke's statistical core design (SCD) methodology. Themethodology for mixed core analysis is also presented. Revisions to the methodology aredescribed along with technical justifications and markups to the current revisions of theassociated reports. The intent of this chapter is to capture all of the new and revised core thermal-hydraulic design methodology in one chapter to facilitate the NRC review.

Chapter 7.0, UFSAR Chapter 15 Non-LOCA Transient and Accident Analysis MethodologyRevisions, describes the revisions to Duke's methods for analyzing UFSAR Chapter 15. Dukeuses the EPRI RETRAN-3D code for Oconee system transient and accident thermal-hydraulicanalyses, and the VIPRE-01 code for transient and accident core thermal-hydraulic analyses. TheStudsvik Scandpower SIMULATE-3K code is used for the three-dimensional rod ejectionaccident analysis. The VIPRE-01 methodology is revised to include a large detailed model thathas expanded capability relative to the current approved models. Mixed core modeling is alsodiscussed. The intent of this chapter is to consolidate the new and revised methodologiesassociated with UFSAR Chapter 15 non-LOCA transients and accidents in one chapter tofacilitate the NRC review of these related subjects.

Chapter 8.0, AREVA NP LOCA Analysis Methodology, summarizes the NRC-approvedAREVA NP Appendix K LOCA Evaluation Models that will be used to analyze the large breakand the small break LOCA with Mark-B-HTP fuel. The intent of this chapter is to identify theLOCA analysis methodologies that will be applied by AREVA NP, and to state that no newanalytical methods will be necessary.

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Chapter 9.0 describes the revisions to the Oconee Nuclear Station Technical Specifications andassociated Bases that are necessary to support the methodology revisions. The technicaljustification for these changes is included.

Chapter 10.0 describes the revisions to the Core Operating Limits Report that are associated withthe methodology revisions. This chapter provides a preview of the revisions to the COLR thatwill result from the use of AREVA NP Mark-B-HTP fuel.

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2.0 MARK-B-HTP FUEL DESIGN SUMMARY

2.1 Design Description

The AREVA NP Mark-B-HTP fuel assembly design (Table 2-1 and Figure 2-1) to be used at theOconee Nuclear Station is a 15x15 design with M5® cladding, instrument, and guide tubes. Theintermediate and top spacer grids are also made of M5®. The bottom spacer grid and the upperand lower end fittings are made of Inconel 718. The fuel rod nominal diameter is 0.430 inches,and the cladding nominal thickness is [ ] inches. Figure 2-2 shows the current Mark-B 11 Adesign. Table 2-1 compares the two designs. The Mark-B-HTP design is licensed to a fuel rodaverage burnup of 62 GWd/mtU (References 2-1 and 2-2). The values presented in Table 2-1 arerepresentative values. Future changes to the Mark-B-HTP fuel assembly design will be evaluatedunder the NRC-approved process described in Reference 2-1.

The main advantage of the Mark-B-HTP design, relative to the Mark-B 11 design, is the improvedperformance relative to grid-to-rod fretting induced cladding failure. The improved performanceis mainly a result of the intermediate grid design, thicker cladding, and the welded cageconstruction. The larger diameter fuel rod of the Mark-B-HTP design also allows for a largerpellet, and therefore a higher U0 2 loading. The Mark-B-HTP design has a higher overallpressure drop, which reduces flow in the Mark-B-HTP assemblies, in particular during a mixedcore configuration.

The critical heat flux correlation, BHTP, for the Mark-B-HTP design was submitted by AREVANP in References 2-3 and 2-4. The NRC safety evaluation reports are References 2-5 and 2-6.Below the first intermediate grid, the BWU-N CHF correlation is used (Reference 2-7).

2.2 Operating Experience

The HTP design, used initially in non-B&W reactors, has been used in over 5,200 fuel assembliesdelivered by AREVA NP to date, with Mark-B-HTP implementation in the B&W designedreactors beginning in 2003. The Mark-B-HTP design is the standard AREVA NP fuel design forthe Oconee-class plants, and has been loaded, or is planned for loading, into the followingreactors.

Reactor and Fuel Cycle Startup Date # Mark-B-HTP FAs

Crystal River-3 Cycle 14 Dec 2003 84

Crystal River-3 Cycle 15 Dec 2005 85

Arkansas Nuclear One Unit I Cycle 20 Dec 2005 56

Davis Besse Cycle 15 April 2006 76

Davis Besse Cycle 16 Jan 2008 88 (estimated)

Three Mile Island Unit I Cycle 17 Future Unknown

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Prior to the introduction of the Mark-B 11 design at Oconee, the AREVA NP Mark-B 10 designwas in use. The Mark-B 10 design had the same cladding diameter as the Mark-B-HTP design.

2.3 References

2-1 Safety Criteria and Methodology for Acceptable Cycle Reload Analyses, BAW-10179P-A, Revision 6, AREVA NP, August 2005

2-2 Extended Burnup Evaluation, BAW-10186P-A, Revision 2, AREVA NP, June 2003

2-3 BHTP DNB Correlation Applied with LYNXT, BAW-10241(P)(A), Framatome ANP,September 2004

2-4 BHTP DNB Correlation Applied with LYNXT, BAW-10241(P)(A), Revision 1,Framatome ANP, July 2005

2-5 Letter, H. N. Berkow (NRC), to J. F. Mallay (Framatome ANP), (SER for BAW-10241P), September 29, 2004

2-6 Letter, H. N. Berkow (NRC), to R. L. Gardner (Framatome ANP), (SER for BAW-10241P, Revision 1), July 25,2005

2-7 The BWU Critical Heat Flux Correlations, BAW-10199P-A, Framatome Cogema Fuels,August 1996 (and including Addendum 1, December 2000)

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Table 2-1

Mark-B -HTP to Mark-B 11 Comparison

FUEL ASSEMBLY:Nominal loading (kg U)[ ] [ ]

Fuel rod pitch (inch)[ ]

Envelope (inch) 8.536

FUEL ROD:

Number per assembly 208

Fuel column length (inch) 143.0

End cap material M5

Cladding material M5Cladding OD (inch) 0.416 0.430

Cladding ID (inch)

Pellet diameter (inch)

STRUCTURE:Type [ Floating I Welded cage

GUIDE TUBE:


Material M5

OD (inch)

ID (inch) [_ ]

INSTRUMENT TUBE:


Material Zirc-4 M5

OD (inch)

ID lower / upper (inch)

UPPER END FITTING:Hold down method | 6-Leaf cruciform spring

LOWER END FITTING:

Debris protection oplug in grid FUELGUARD

A

Note: Dimensions are representative values

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Figure 2-1

Mark-B-HTP Design

Mark-B-HTP Fuel Assembly

Removable Upper EndFitting

" Alloy 718 CruciformSprings

* Crimped Top-Hat

Nut

* M5® Fuel Rods

* M50 Guide Tubes

* M50 Instrument Tube

* FUELGUARDTM LowerEnd Fitting

M50 HTP Grids (7x) withCurved Flow Channels

Alloy 718 Lower HMPGrid (Straight FlowChannel)

Page 7

Figure 2-2

Mark-B I1 Design

Mark-B 1lA Fuel, Assembly

Removable Upper End . Alloy 718 TopFitting

• Alloy 718 CruciformSprings Zirc-4 Interme

" "Quick Disconnect" VaneridFeatre • Vane Grids

Feature

* M5® Fuel Rods

M53 Guide Tubes j,

* Zirc-4 Instrument Tube

Lower End Fitting" Open Structure * Zirc-4 Vanefe,• Plug-in-Grid Debris Ao

Resistance Alloy 718 Low

Grid

diate Mixing

ss Grid

er Grid

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3.0 METHODOLOGY REPORTS OVERVIEW

This chapter provides an overview of Duke's NRC-approved methodology reports and their inter-relationships in core reload design and in the UFSAR licensing basis transient and accidentanalyses for the Oconee Nuclear Station. Chapters 4.0 through 7.0 describe the proposedrevisions to these methodology reports in detail. Duke's Oconee-related methodology reports fallinto the categories of core physics and reload design, fuel assembly mechanical design, corethermal-hydraulic design, and transient and accident analysis. Revisions to all of the followingreports except DPC-NE-1004-A, "Nuclear Design Methodology Using CASMO-3/SIMULATE-3P," and DPC-NE-3003-PA, "Mass and Energy Release and Containment ResponseMethodology," are included in this report. A detailed review has concluded that DPC-NE-1004-A and DPC-NE-3003-PA do not require revision to support the transition to the Mark-B-HTP fueldesign. The proposed revisions to the other reports constitute the next sequential revision numberafter the revision numbers listed below.

Number Title Revision Date

_CofrePhysics and Reload ~Design; ~ ~ Q _____'

NFS-1001A Oconee Nuclear Station Reload Design 5 Jan-01Methodology

DPC-NE-1002-A Oconee Nuclear Station Reload Design 2 Oct-85Methodology II

DPC-NE-1004-A Nuclear Design Methodology Using CASMO- I Dec-973/SIMULATE-3P

F"ue•l Assmbll Mechanical Dleskigiv . ,, - - ,_ ,;___,,

DPC-NE-2008P-A Fuel Mechanical Reload Analysis 0 Apr-95Methodology Using TACO3

_Coreý,Tlieriial-Hyrui Design

Oconee Nuclear Station Core Thermal-DPC-NE-2003P-A Hydraulic Methodology Using VIPRE-01 1 Sep-00DPC-NE-2005P-A Thermal-Hydraulic Statistical Core DesignDPC-NE-200_P-A Methodology 3 Sep-02

DPC-NE3000PA Thermal-Hydraulic Transient Analysis 3 Sep-04DPC-NE-3000-PA_ Methodology

DPC-NE-3003-PA Mass and Energy Release and Containment I Sep-04DPC-NE-3003-PA___ Response Methodology __Sep-04

DPC-NE-3005-PA UFSAR Chapter 15 Transient Analysis 2 May-05I Methodology

3.1 Core Physics and Reload Design Methodology Reports

NFS-1001 A, "Oconee Nuclear Station Reload Design Methodology" (Reference 3-1), describesthe fuel cycle design process, the development of core physics parameters, the maneuveringanalysis, and the development of core safety limits, reactor protection setpoints, and technical

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specification limiting conditions for operation. The process to confirm that core physicsparameters which are important to the accident analysis remain valid for each reload cycle is alsodescribed.

DPC-NE-1002-A, "Oconee Nuclear Station Reload Design Methodology II" (Reference 3-2), ismainly a revision to NFS-1001A to include minor changes based on additional experienceapplying the methods, and for the change to the Mark-BZ fuel assembly design. Also, theCASMO computer code is included as an alternative for generating nuclear data.

DPC-NE-1004-A, "Nuclear Design Methodology Using CASMO-3/S1MULATE-3P" (Reference3-3), presents the methodology for calculating nuclear physics data using the CASMO-3/SIMULATE-3P code package. A new set of reliability and uncertainty factors is developed.These nuclear analysis codes have replaced the codes described in NFS-1001A and DPC-NE-1002-A for Oconee reload design applications. Since the lattice for the Mark-B-HTP fuelassembly design is similar to the lattice for the Mark-B designs that preceded the current Mark-B 11 design, Duke has previously established the accuracy of the nuclear design codes to beapplied to the Mark-B-HTP design. Therefore no revisions to the methodology in DPC-NE-1004-A are necessary for Mark-B-HTP implementation. Mark-B-HTP fuel segments in theCASMO-3 cross section interpolating library will be created.

3.2 Fuel Assembly Mechanical Design Methodology Report

DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3", (Reference3-4), describes Duke's application of the AREVA NP TACO3 code for calculating linear heatrate to melt, pin pressure, cladding strain, and fuel rod input parameters for LOCA analyses. Theapplication of the AREVA NP CROV code for cladding creep collapse is also described.

The NRC conducted an on-site audit of Duke's use of the AREVA fuel assembly mechanicaldesign methodology and issued an associated safety evaluation report to Duke by letter datedApril 3, 1995 (Reference 3-5). The audit included a review of DPC-NE-2008 and relateddocumentation. Pursuant to NUREG-0390, Duke submitted copies of the proprietary (DPC-NE-2008P-A) and the non-proprietary (DPC-NE-2008-A) versions of DPC-NE-2008 including theaudit safety evaluation report under cover of letter dated November 13, 1997. The "-A"designation was added to indicate that DPC-NE-2008 had been approved, albeit through an on-site audit.

3.3 Core Thermal-Hydraulic Design Methodology Reports

DPC-NE-2003P-A, "Oconee Nuclear Station Core Thermal-Hydraulic Methodology UsingVIPRE-01" (Reference 3-6), presents Duke's VIPRE-01 models for performing steady-state corethermal-hydraulic analyses and the methods that are used to calculate departure from nucleateboiling (DNBR) limits and associated Reactor Protective System setpoints and core powerdistribution limits.

DPC-NE-2005P-A, "Thermal -Hydraulic Statistical Core Design Methodology" (Reference 3-7),presents the VIPRE-01 methodology for the statistical combination of the uncertainties related tothe calculation of the DNBR into a statistical DNBR limit for each fuel assembly design andcritical heat flux correlation.

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3.4 Transient and Accident Analysis Methodology Reports

DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology" (Reference 3-8),describes the RETRAN-3D system transient thermal-hydraulic models, and the VIPRE-01transient core thermal-hydraulic models, that are used to simulate non-LOCA UFSAR Chapter 15transients and accidents. Validation of these models is included in the report. Appendix C ofDPC-NE-3000-PA documents a previous evaluation of the RETRAN-3D SER conditions andlimitations. This evaluation remains valid for the proposed methodology revisions.

DPC-NE-3003-PA, "Mass and Energy Release and Containment Response Methodology"(Reference 3-9), describes the methodology for analyzing the mass and energy release from loss-of-coolant accidents (LOCAs) and main steam line breaks, and the resulting containment pressureand temperature response. The description of the RELAP5 model used for LOCA mass andenergy release analysis is not fuel assembly design specific. Similarly, the description of theRETRAN-3D model used for main steam line break analysis is not fuel assembly design specific.DPC-NE-3000-PA is referred to by DPC-NE-3003-PA, and the revisions to the RETRAN-3Dmodel for the Mark-B-HTP fuel design will be captured in that report. Therefore, this report doesnot need to be revised for the Mark-B-HTP fuel transition and is not discussed further.

DPC-NE-3005-PA, "USFAR Chapter 15 Transient Analysis Methodology" (Reference 3-10),describes the methodologies for simulating the UFSAR Chapter 15 non-LOCA transients andaccidents. The RETRAN-3D code is used for the system transient simulation, and the VIPRE-01code is used for the transient DNBR analysis. The CASMO-3/SIMULATE-3 code system is usedfor calculating physics parameters and simulating power distributions. The SIMULATE-3K codeis used for the three-dimensional kinetics analysis for the rod ejection accident. Themethodologies for determining the safety analysis physics parameters and the safety analysissetpoints are also included. For each UFSAR Chapter 15 event the initial condition and boundaryconditions are specified: Appendix A of DPC-NE-3005-PA documents a previous evaluation ofthe RETRAN-3D SER conditions and limitations. This evaluation remains valid for the proposedmethodology revisions.

3.5 Inter-Relationships of Methodology Reports

The nine methodology reports summarized above are inter-related in many ways based in largepart on the design limits and regulatory limits that must be met for a specific fuel pellet design,fuel rod design, fuel assembly design, and core loading pattern. These can be separated intosteady-state design requirements concerned with normal operation of the reactor, and regulatoryrequirements associated with licensing basis transients and accidents.

The steady-state design requirements include mechanical design limits based on the DPC-NE-2008P-A methodology, and thermal design limits based on the DPC-NE-2003P-A and DPC-NE-2005P-A methodologies. These include limits such as maximum power level, maximum burnup,pin pressure, cladding corrosion thickness, centerline fuel melt (CFM), core power distribution,and conditions (heat flux, mass flux, pressure, temperature, void fraction) that preclude theoccurrence of DNB. These limits are established to meet NRC regulations. The methodologiesin NFS-1001A and DPC-NE-1002-A employ these limits in the process of establishing theloading pattern for a reload core, the cycle length, and core operating limits.

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The methodologies associated with licensing basis transients and accidents, DPC-NE-3000-PAand DPC-NE-3005-PA, establish additional limits based on NRC acceptance criteria that NFS-1001A and DPC-NE-1002-A must also meet. These include the DNBR limit, the CFM limit, thefuel peak enthalpy limit for the rod ejection accident, and limits on the values of all of the keysafety analysis physics parameters that determine the transient response of the reactor. Thelicensing basis transient and accident analyses determine the acceptability of the ReactorProtective System setpoints, some of which are also used in the DPC-NE-2008P-A methodologyfor steady-state thermal limits. And finally, the LOCA kW/ft limits on the core power

.distribution provided by AREVA NP complete the set of limits that must be observed in the corereload design process.

The relationships between the above methodologies exist regardless of the fuel assembly design.However, the transition to the AREVA NP Mark-B-HTP fuel design requires that all of themethodologies be updated to account for all of the design differences. In addition, any effectsdue to mixed cores must also be accounted for in the applications of the methodologies.

3.6 References

3-1 Oconee Nuclear Station Reload Design Methodology, NFS-1001A, Revision 5, January2001

3-2 Oconee Nuclear Station Reload Design Methodology II, DPC-NE-1002-A, Revision 2,October 1985

3-3 Nuclear Design Methodology Using CASMO-3/SIMULATE-3P, DPC-NE-1004-A,Revision 1, December 1997

3-4 Fuel Mechanical Reload Analysis Methodology Using TACO3, DPC-NE-2008P-A, April1995

3-5 Letter, H. N. Berkow (NRC), to M. S. Tuckman (Duke), April 3, 1995, (Subject: DukePower Company's Use of TACO3 and the Fuel Rod Gas Pressure Criterion for theOconee, McGuire, and Catawba Nuclear Stations)

3-6 Oconee Nuclear Station Core Thermal-Hydraulic Methodology Using VIPRE-01, DPC-NE-2003P-A, Revision 1, September 2000

3-7 Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005P-A, Revision 3,September 2002

.3-8 Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3000-PA, Revision 3,September 2004

3-9 Mass and Energy Release and Containment Response Methodology, DPC-NE-3003-PA,Revision 1, September 2004

3-10 UFSAR Chapter 15 Transient Analysis Methodology, DPC-NE-3005-PA, Revision 2,May 2005

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4.0 CORE PHYSICS AND RELOAD DESIGN METHODOLOGY REVISIONS

This chapter details the proposed methodology revisions to Duke's NRC-approved core physicsand reload design methodology reports that are applicable to the Oconee Nuclear Station. Eachrevision is described in detail along with the technical justification. A markup to the currentversion of the methodology report is also presented. Deleted text is designated by astrikethrough, and new text is designated by bold font.

4.1 Revision 6 to NFS-1001A, Oconee Nuclear Station Reload Design Methodology

NFS-1001A, "Oconee Nuclear Station Reload Design Methodology" (Reference 4-1), describesthe fuel cycle design process, the development of core physics parameters, the maneuveringanalysis, and the development of core safety limits, reactor protection setpoints, and technicalspecification limiting conditions for operation. The process to confirm that core physicsparameters which are important to the accident analysis remain valid for each reload cycle is alsodescribed. The following proposed Revision 6 consists of revisions associated with 1) thetransition to the AREVA NP Mark-B-HTP fuel design, 2) new methods to address technicalissues, 3) corrections to address errors in the existing methodology, and 4) editorial revisions.

Revision 4-1 Section 2.1, Fuel Design (p. 7)

Description: The Mark-B-HTP fuel design and associated methodology references will be addedto the list of fuel assembly designs. This revision is an editorial change that will be madefollowing NRC review and approval.

Revision 4-2 Section 7.2.2.1 Calculation of Power-Power Imbalance Limits for Center FuelMelt / Clad Strain Criterion (p. 21)

Description: The fuel densification power spike factor, which is specified as a value of 1.08, isreplaced with an axially-dependent factor. Also, a reference to a different methodology report isused for a more complete description of the penalty factor and its implementation.

4. Densification power spike factor which varies with axial location of the peak in thecore (Reference 15). For .urrent fuel lesigns a fac•t.r of 1.08 is ut.ilizeA

Technical Justification: The densification power spike factor is provided by the fuel vendor as anaxially-dependent table of values. In the current methodology the maximum value of 1.08 (8%)has been applied as a conservative approach. This approach was necessary since the software thatapplies the power spike factor to the core power distribution was not capable of using an axially-dependent table of values. The software has been revised and is now able to use the actual valuesprovided by the vendor. The use of an upper bound value is discontinued. Reference 4-2, DPC-NE-1002-A, Oconee Nuclear Station Reload Design Methodology II, has additional contentdescribing the application of the densification power spike factor. It is proposed to maintain thatadditional content in Reference 4-2 rather than to duplicate it in NFS-1001A.

Page 13

Revision 4-3 Section 7.2.2.1 Calculation of Power-Power Imbalance Limits for Center FuelMelt/ Clad Strain Criterion (p. 21)

Description: A pin power peaking penalty for the effect of fuel assembly bow is included.

The effect of fuel assembly bow on the pin power distribution is accounted for by apenalty factor that is dependent on the location of the pin within the assembly.Although fuel assembly boing 1 i..nidered to have the potential fer enhancing thepower: peaks, no explicit alaneis rie-quir-ed fer- assembly bow an the basis that the,other conservatism faetor.s (nuclear uncertainty faet.., nhot hannel factor,and densificatin pike fatr) are adequate to offse..t the, e... ff.etAS..... .. ohe aeblybow power- spike factor-withouit an additional allowance.

Technical Justification: In the original methodology the effect of fuel assembly bow on the pinpower distribution was judged to be adequately offset by other existing conservative factors.With the revised methodology a pin peaking penalty is explicitly included to address this effect.The penalty value is associated with the location of the pin in the assembly, with peripheral pinshaving a larger peaking penalty than interior pins.

Revision 4-4 Section 7.2.2.2 Calculation of Power-Power Imbalance Limits for DNBRCriterion (p. 22)


For each power distribution, the calculated maximum total peaking factors of each of theassemblies is increased by the radial nuclear uncertainty factor, and a factor to accountfor the effect of fuel assembly bow, and the resulting adjusted peak is compared to theallowable peaking factor for that axial peaking factor and axial peak location.

Technical Justification: In the original methodology the effect of fuel assembly bow on the pinpower distribution was judged to be adequately offset by other existing conservative factors.With the revised methodology a pin peaking penalty is explicitly included to address this effect.The penalty value is associated with the location of the pin in the assembly, with peripheral pinshaving a larger peaking penalty than interior pins.

Revision 4-5 Section 7.4.1 Determination of LOCA-Limited Power Distribution Limits (p.

2_7


Insert this sentence at the end of the second paragraph on p. 27

The effect of fuel assembly bow on the pin power distribution is accounted for by apenalty factor that is dependent on the location of the pin within the assembly.

Technical Justification: In the original methodology the effect of fuel assembly bow on the pinpower distribution was judged to be adequately offset by other existing conservative factors.

,With the revised methodology a pin peaking penalty is explicitly included to address this effect.

Page 14

The penalty value is associated with the location of the pin in the assembly, with peripheral pinshaving a larger peaking penalty than interior pins.

Revision 4-6 Revision 6 to NFS-1001A, Chapter 10, References (p. 46)

Description: The list of references is updated to the latest revisions and dates, and to include newreferences. This revision is editorial only and no technical justification is necessary.

2. Nuclear Design Methodology Using CASMO-3/SIMULATE-3P, DPC-NE-1004A, "Revision 1, Duke Power Company, December 1997

3. UFSAR Chapter 15 Transient Analysis Methodology, DPC-NE-3005-PA, Revision 3,Duke Power Company, (approval date)

5. Oconee Nuclear Station Core Thermal-Hydraulic Methodology Using VIPRE-01,DPC-NE-2003P-A, Revision 2, Duke Power Company, (approval date)

6. Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005P-A,Revision 4, Duke Power Company, (approval date)

15. Oconee Nuclear Station Reload Design Methodology II, DPC-NE-1002-A,Revision 3, Duke Power Company, (approval date)

4.2 Revision 3 to DPC-NE-1002-A - Oconee Nuclear Station Reload Design Methodology II

DPC-NE-1002-A, "Oconee Nuclear Station Reload Design Methodology II" (Reference 4-2), ismainly a revision to NFS-1001A to include minor changes based on additional experienceapplying the methods, and for the change to the Mark-BZ fuel assembly design. Also, theCASMO computer code is included as an alternative for generating nuclear data. The followingproposed Revision 3 consists of revisions associated with 1) deletion of superseded computercodes, 2) new methods to address technical issues, 3) corrections to address errors in the existingmethodology, and 4) editorial revisions.

Revision 4-7 Entire report

Description: The computer codes EPRI-CELL, EPRI-NODE-P, PDQ, and related utility codes,and the-TACO2 code have been superseded by other codes described in other NRC-approvedmethodology reports, and are deleted.

All content related to these codes will be deleted in Revision 3

Technical Justification: The EPRI-CELL, EPRI-NODE, and PDQ codes have been supersededby the CASMO-3 and SIMULATE-3 codes as detailed in other methodology reports, and will bedeleted in Revision 3 throughout. TACO2 has been superseded by TACO3, and that change willbe made in Revision 3. The NRC has approved all of these new methodologies.

Page 15

Revision 4-8 Section 7.A (Section 7.2.2.1 of NFS-1001A) (p. 7-2)

Description: The values of the nuclear uncertainty factors are updated and a new reference isadded.

1. Nuclear uncertainty factor = 1.065 1.056 (Reference 13)a. Bias=-h63% 1.3%b. 95/95 Total Peaking Uncertainty = 4--9% 3.8%c. Bounding Radial-Local Uncertainty = 2-.2% 2.0%

Technical Justification: The values of the nuclear uncertainty factors that are being deleted areassociated with the PDQ and NODE computer codes that were used in the original methodology.These computer codes are no longer used, and so the uncertainty values are updated based on thecurrent SIMULATE-3 code and the associated methodology. These uncertainty values were allpreviously established in methodology report DPC-NE-1004-A (Reference 4-3).


Description: The radial-local factor is deleted based on methodology report DPC-NE-1004-A.

4. R-adial lWcal factor: varies with location of the assembly in the cor-e (typical valuie is1.10). (Thisi is not an uncertainty factOr but a dlirect muitltipl1ier eni theIA44A mai mu-M totalpeak. This is calculated by the W cor-e PDQ07 2 D model.)

The nuclear uncertainty factor accounts for the uncertainty in the calculated peak due tothe limitations of the analytical models. and the radial loce-al fa•cto is applied to accout...for- the fact that the calculations are per-fer-med using an assemfbly by assembly moidelr-athe-r th-an by using a pin by pin moadel.

Technical Justification: The radial-local factor was discontinued with the implementation of theCASMO-3/SIMULATE-3 methodology in DPC-NE-1004-A.


Description: The fuel densification power spike factor, which is specified as an upper boundvalue of 8%, is replaced with an axially-dependent factor.

3. Densification power spike factor (F-SPIKE(z)): varies with axial location of the peakin the core. As an upper bou.nd, an 8% uncertaintyill be applied for- all axial

Technical Justification: The densification power spike factor is provided by the fuel vendor as anaxially-dependent table of values. In the current methodology the bounding value of 1.08 (8%)has been applied as a conservative approach. The use of an upper bound value is discontinued.

Page 16

Revision 4-11 Section 7.A (Section 7.2.2.1 of NFS-1OOIA) (p. 7-2); Section 7.C (Section 7.4.1of NFS-1001A) (p. 7-4)

Description: A power peaking penalty for the effect of fuel rod bow is included.

(p. 7-2) The effect of fuel rod bow on the pin power distribution is accounted for bya penalty factor (F-RB) that is a function of burnup. Although fuel red bewing isconsider-ed to have the potential for- enhancing the power- peaks, no) explicit allowAance isprovided for the rod bov' power spike factor on the basis that the other conser-vatismfactor~s (nuclear: uncer-tainty factorF and enlgineering bet channiel factor-) are adequate tooffset the effect of the red bow~ power- spike factor without an additionlal allowAancee.

(p. 7-4 - new last sentence) The effect of fuel rod bow on the pin power distributionis accounted for by a penalty factor (F-RB) that is a function of burnup.

Technical Justification: In the original methodology the effect of fuel rod bow on the pin powerdistribution was judged to be adequately offset by other existing conservative factors. With therevised methodology a pin peaking penalty is explicitly included to address this effect. The rodbow penalty value, which is determined by the fuel vendor, is a function of burnup. The F-RBfactor is included in the SCUFcIm and SCUFLOCA equations in the revisions that follow.


Description: The values in the equation for centerline fuelmelt uncertainty are updated.

Replace:

SCUFCm = I + 0.0163 + V0.04292 + 0.0222 + 0.0142 + 0.082 = 1.111

With:

SCUFCFM(Z) = 1 + 0.013 + V0.0382 + 0.0202 + 0.0142 + (F- SPIKE(z)) 2 + (F-RB) 2

Technical Justification: The values in the centerline fuel melt uncertainty equation are updated,and the total uncertainty, value becomes axially-dependent. The updated values were previouslyestablished in methodology report DPC-NE-1004-A, with the exception of the densificationpower spike factor (F-SPIKE(z)) as described in Revision 4-10 above. The rod bow penaltyfactor (F-RB) is also described above. The total numerical value cannot be revised due to the twonew parameters being variables.

Revision 4-13 Section 7.4.1 (p. 7-4)

Description: The values of the nuclear uncertainty factors are updated and a new reference isadded.

1. Nuclear uncertainty factor = -. 065 1.056 (Reference 13)a. Bias = .63% 1.3%b. 95/95 Total Peaking Uncertainty = 4.2•9% 3.8%

Page 17

c. Bounding Radial-Local Uncertainty = 2-.2% 2.0%

Technical Justification: These uncertainty values were all previously established in methodologyreport DPC-NE-1004-A (Reference 4-3).

Revision 4-14 Section 7.C (Section 7.4.1 of NFS-1001A) (p. 7-4)

Description: The values in the equation for LOCA uncertainty are updated.

Replace:

SCUFLOCA = 1 + 0.0163 + VO.04292 + 0.0222 + 0.022 + 0.0142 = 1.070

With:

SCUFLOCA = I + 0.013 + V0.0382 + 0.0202 + 0.0142 + (F- RB) 2

Technical Justification: These uncertainty values were all previously established in methodologyreport DPC-NE-1004-A (Reference 4-3). The 0.02 factor in the original equation, that is nowbeing deleted, is the power level uncertainty. The power level uncertainty is applied separately asa multiplier since it is applicable to the entire core. This is a more conservative approach than theoriginal methodology. The rod bow penalty factor (F-RB) is described in the revision above.The revised numerical value is not included since the rod bow penalty is a function of burnup.

Revision 4-15 Section 7.C (Section 7.4.1 of NFS-1001A) (p. 7-4)

Description: The radial local factor is deleted since it is no longer applicable.

in addition, the radial leoal faet.r is applied to acceunt for local pin peaking. The radiallocal factor varies with location of the assembly in the .... re (typical value is 1.10).

Technical Justification: Use of the CASMO-3/STMULATE-3 code system includes a pinreconstruction model. Therefore, the original text describing the use of a radial-local factor is nolonger applicable.

Revision 4-16 Revision 3 to DPC-NE-1002-A, Chapter 10 References


1. Oconee Nuclear Station Reload Design Methodology, NFS-1001A, Revision 6, DukePower Company, (approval date)

13. Nuclear Design Methodology Using CASMO-3/SIMULATE-3P, DPC-NE-1004A, Revision 1, Duke Power Company, December 1997

Page 18

4.3 Mixed Core Effects

This section addresses any core physics or reload design methodology issues that are related tothe mixed core configuration with AREVA NP Mark-B 11 and Mark-B-HTP fuel assemblydesigns co-resident in reload core designs.

The active fuel region for the Mark-B-HTP fuel assembly is offset [ ] relative to the Aactive fuel region of the Mark-B 11 fuel assembly. An evaluation of the offset in active fuellength difference will be performed with SIMULATE-3 to characterize the impact on peakingfactors introduced by the active fuel region offset. The analysis will determine any impact onlocal and integral peaking factors. As a result, mixed core FAh and Fq peaking penalties will beapplied, as necessary, to calculated peaks for the Mark-B 11 and/or the Mark-B-HTP fuelassemblies for the mixed core reload designs, prior to comparison to thermal limits.

4.4 References

4-1 Oconee Nuclear Station Reload Design Methodology, NFS-1001A, Revision 5, January2001

4-2 Oconee Nuclear Station Reload Design Methodology II, DPC-NE-1002-A, Revision 2,October 1985

4-3 Nuclear Design Methodology Using CASMO-3/SIMULATE-3P, DPC-NE-1004-A,Revision 1, December 1997

Page 19

5.0 FUEL ASSEMBLY MECHANICAL DESIGN METHODOLOGY REVISIONS

This chapter details the proposed methodology revisions to Duke's fuel assembly mechanicaldesign methodology reports that are applicable to the Oconee Nuclear Station. Each revision isdescribed in detail along with the technical justification. A markup to the current version of themethodology report is also presented. Deleted text is designated by a strikethrough, and new textis designated by bold font. DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis MethodologyUsing TACO3", (Reference 5-1), describes Duke's application of the AREVA NP TACO3 code(Reference 5-2) for calculating linear heat rate to melt, pin pressure, and cladding strain. Theapplication of the AREVA NP CROV code (Reference 5-3) for cladding creep collapse is alsodescribed.

AREVA-NP topical reports are referenced for cladding stress, cladding fatigue, and claddingcorrosion methodologies that are applied by Duke.

5.1 Revision I to DPC-NE-2008P-A - Fuel Mechanical Reload Analysis Methodology UsingTACO3

Revision 5-1 Section 2.4, Cladding Creep Collapse (p. 6)

Description: The entire Section 2.4 text and references for the cladding creep collapsemethodology are updated to the current NRC-approved content and associated references. Nochanges to the methodology are associated with the transition to the AREVA NP Mark-B-HTPdesign. This revision is editorial only and no technical justification is necessary.

2.4 Cladding Creep Collapse

An analysis is performed to demonstrate that the cladding will not collapse underthe combined effects of neutron flux, temperature, and cladding stress. The CROVcomputer code (Reference 8) is used to demonstrate that the effective full powerhours (or equivalent burnup) to cladding collapse is greater than the actual incoreresidence time. Conservative analysis inputs include [

]. Other conservatisms include []. Internal pin pressure and cladding

temperature inputs are calculated by TACO3 using a []. NRC acknowledged Duke's application of the CROV A

cladding creep collapse methodology in Reference 9.

Revision 5-2 New Section 2.5, Cladding Corrosion

Description: A new Section 2.5 is added to describe the NRC-approved cladding corrosionanalysis methodology. No changes to the methodology are associated with the transition to theAREVA NP Mark-B-HTP design. This revision is editorial only and no technical justification isnecessary.

Page 20

2.5 Cladding Corrosion

Cladding oxide thickness is calculated using the COROS02 corrosion model andmethodology described in References 10 and 11. The cladding oxide thickness isdetermined for [

]

The NRC Safety Evaluation Report for Reference 10 characterizes the COROS02corrosion model as [

A

If an assembly contains a rod whose predicted oxide thickness exceeds the limit, itcan be designated a lead corrosion assembly and continue to operate. Reference 10defines the following requirements for lead corrosion assemblies: 1) corrosionmeasurements will be taken on this assembly after it has been discharged from thecore, 2) the total number of lead corrosion assemblies is limited to 8 per cycle, and3) the total number of lead corrosion and other demonstration assemblies is limitedto 12 per cycle.

Revision 5-3 New Section 2.6, Cladding Stress

Description: A new Section 2.6 is added to describe the NRC-approved cladding stress analysismethodology. No changes to the methodology are associated with the transition to the AREVANP Mark-B-HTP design. This revision is editorial only and no technical justification isnecessary.

2.6 Cladding Stress

The cladding stress analysis is performed using the AREVA NP methodology inReferences 10, 11, 12, and 13, using inputs from the TACO3 code (Reference 3).

Revision 5-4 New Section 2.7, Cladding Fatigue

Description: A new Section 2.7 is added to describe the NRC-approved cladding fatigue analysismethodology. No changes to the methodology are associated with the transition to the AREVANP Mark-B-HTP design. This revision is editorial only and no technical justification isnecessary.

2.7 Cladding Fatigue

The cladding fatigue analysis is performed using the AREVA NP methodology inReferences 10, 11, 12, and 13.

Page 21

Revision 5-5 Revision I to DPC-NE-2008P-A, Chapter 4.0 References

Description: The references are updated to the latest revisions and dates, and to include newreferences. This revision is editorial only and no technical justification is necessary.

2. Oconee Nuclear Station Reload Design Methodology II, DPC-NE-1002-A,Revision 3, (approval date)

4. Fuel Rod Gas Pressure Criterion, BAW-10183P-A, B&W Fuel Company, July1995

6. Deleted (superseded by Reference 8)

8. Program to Determine In-Reactor Performance of BWFC Fuel Cladding CreepCollapse, BAW-10084P-A, Revision 3, B&W Fuel Company, August 1995

9. Letter, H. N. Berkow (NRC) to M. S. Tuckman (Duke), Subject: DukePower use of CROV Computer Code, June 19, 1995

10. Extended Burnup Evaluation, BAW-10186P-A, Revision 2, AREVA NP,June 2003

11. Letter D. LaBarge (NRC) to W. R. McCollum, Jr. (Duke), Subject: Use ofFramatome Cogema Fuels Topical Report on High Burnup - OconeeNuclear Station, Units 1, 2 and 3, March 1, 1999

12. Safety Criteria and Methodology for Acceptable Cycle Reload Analyses,BAW-10179P-A, Revision 6, AREVA NP, August 2005

13. Evaluation of Advanced Cladding and Structural Material (M5) in PWRReactor Fuel, BAW-10227P-A, Revision 1, AREVA NP, June 2003

5.2 References

5-1 Fuel Mechanical Reload Analysis Methodology Using TACO3, DPC-NE-2008P-A, April 1995

5-2 TACO3 Fuel Pin Thermal Analysis Computer Code, BAW-10162P-A, B&WFuel Company, October 1989

5-3 Program to Determine In-Reactor Performance of BWFC Fuel Cladding CreepCollapse, BAW-10084P-A, Rev. 3, B&W Fuel Company, August 1995.

5-4 Letter, H. N. Berkow (NRC), to M. S. Tuckman (Duke), Subject: Duke PowerCompany's Use of TACO3 and the Fuel Rod Gas Pressure Criterion for theOconee, McGuire, and Catawba Stations, April 3, 1995

5-5 Letter, H. N. Berkow (NRC) to M. S. Tuckman (Duke), Subject: Duke Power useof CROV Computer Code, June 19, 1995

Page 22

6.0 CORE THERMAL-HYDRAULIC DESIGN METHODOLOGY REVISIONS

This chapter details the proposed methodology revisions to Duke's NRC-approved core thermal-hydraulic design methodology reports that are applicable to the Oconee Nuclear Station. Eachrevision is described in detail along with the technical justification. A markup to the currentversion of the methodology report is also presented. Deleted text is designated by astrikethrough, and new text is designated by bold font.

6.1 Revision 2 to DPC-NE-2003P-A - Oconee Nuclear Station Core Thermal-HydraulicMethodology using VIPRE-01

DPC-NE-2003P-A, "Oconee Nuclear Station Core Thermal-Hydraulic Methodology usingVIPRE-01" (Reference 6-1), presents Duke's VIPRE-01 models for performing steady-state corethermal-hydraulic analyses, and all of the methods that are used to calculate DNBR limits andassociated Reactor Protective System setpoints and core power distribution limits. Revision 2consists of the following revisions.

Revision 6-1 Section 3.0, Station Description (p. 3)

Description: The Mark-B-HTP fuel assembly design is described by adding the followingcontent and the associated reference at the bottom of Section 3.0.

The AREVA NP Mark-B-HTP fuel design is described in Table F-1 of Reference 11.

Technical Justification: The Mark-B-HTP fuel design will be modeled with VIPRE-01 and themethodology of DPC-NE-2003P-A. Rather than include the design information in severaldifferent methodology reports, a reference to DPC-NE-2005P-A, Table F-I, is used. Table F-i isa new table that is described by a revision in Section 6.2 below.

Revision 6-2 Section 5.0, VIPRE-01 Data (p. 8)

Description: The Mark-B-HTP fuel assembly and the BHTP and BWU-N CHF correlations areincluded by adding the following content and the associated references at the bottom of Section5.0.

The Mark-B-HTP fuel assembly design input parameters are described in Table F-1of Reference 11. The applicable critical heat flux correlation is the BHTPcorrelation (Reference 13). The BHTP correlation limit and the statistical coredesign limit are provided in Reference 11, Appendix F. Below the first intermediategrid the BWU-N correlation (Reference 14) is used.

Technical Justification: The Mark-B-HTP fuel design will be modeled with VIPRE-01 and themethodology of DPC-NE-2003P-A (Reference 6-1). Rather than include the design informationin several different methodology reports, a reference to DPC-NE-2005 (Reference 6-2), Table F-1, is used. Table F-i is a new table that is described by a revision in Section 6.2 below. TheAREVA NP BHTP critical heat flux correlation (Reference 6-3) will be used by Duke for DNBR

Page 23

analyses for the Mark-B-HTP fuel, with the exception of below the first intermediate grid wherethe BWU-N correlation (Reference 6-4) will be used. References 6-3 and 6-4 have beenapproved by the NRC. The BHTP correlation limit and the statistical core design limit developedby Duke with the VIPRE-01 code are documented in DPC-NE-2005P-A, which is detailed by arevision in Section 6.2 below.

Revision 6-3 Section 5.7, Inlet Flow Distribution (p. 12)

Description: A new reference is included for the vessel mixing flow tests that are the basis for theinlet flow distribution factors. This is an editorial revision and no technical justification isnecessary.

... in B&W's 1/6-scale Vessel Model Flow Test, r-ef--_t (Reference 15).

Revision 6-4 Section 5.8.2, Turbulent Mixing (p. 14)

Description: The turbulent mixing coefficient is a design specific parameter. The value given isonly applicable to the Mark-BZ design. A clarification is added.

Based upon vendor predictions of mixing test results, a mixing coefficient of [ ] will Abe used for all Occnee Nuclear Stati on core thermal-hydraulic analyses for the Mark-BZfuel design. For the Mark-Bll design refer to Reference 11, Appendix D. For theMark-B-HTP design refer to Reference 11, Appendix F.

Technical Justification: The stated value for the turbulent mixing coefficient is only applicable tothe Mark-BZ fuel assembly design. The clarification makes that correction and refers toReference 6-2, Appendices D and F, for the values for the Mark-B 11 and Mark-B-HTP designs,respectively. Appendix F is detailed by a revision in Section 6.2 below.

Revision 6-5 Section 5.9, Reference Design Power Distribution (p. 17)

Description: A different reference design power distribution is used for the Mark-B-HTP fueldesign.

Insert a new second paragraph in Section 5.9

For the Mark-B-HTP fuel design the reference design power distributions areshown in Figures 4-6 and 4-7. The pin power values in Rods 2 and 4 have beenreduced relative to the original reference power distribution to ensure that the hotsubchannel remains Subchannel 1.

Page 24


D

Page 25


D

Technical Justification: The VIPRE models shown in Figures 4-1 through 4-5 are designed toobtain the minimum DNBR on Rod 1 in Subchannel 1. This is confirmed by the DNBR resultswhen the model is run for a range of statepoint conditions. For a given fuel design this resultdepends on all of the model input associated with the dimensions and design data for a specificfuel assembly, and the CHF correlation used. To maintain the hot subchannel in Subchannel I forthe Mark-B -HTP design, the pin peaks for Rods 2 and 4 were reduced from [ D

respectively.

Revision 6-6 Section 5.11, Hot Channel Factor (p. 19)

Description: For the Mark-B-HTP design, the hot channel factors are given in Appendix F toReference 11 (Section 6.3 below). The following new paragraph is added at the bottom ofSection 5.11.

The hot channel factors for the Mark-B-HTP design are given in Appendix F toReference 11. The application of these factors is the same as described above for theother designs.

Page 26

Technical Justification: Rather than repeat the hot channel factors for the Mark-B-HTP design intwo separate methodology reports, a reference to Appendix F (Section 6.3 below) of DPC-NE-2005P (Reference 6-2) is used. The hot channel factors are given in Table F-5.

Revision 6-7 Section 6.4, Pressure-Temperature (p. 24); Section 6.4.2, Core Power (p. 25);Figure 6-1 (p. 55)

Description: The methodology has been revised to no longer include the ] reactor coolant

pump operating initial condition.

Section 6.4:

To ensure that the PT envelope provides DNB protection, PT curves are determined forI ] reactor coolant (4C-) pump operation.

Section 6.4.2:

The maximum power level for [ ] pump operation is based on the flux/flow tripsetpoint including the appropriate flow measurement uncertainty. The PT curves arecalculated for the maximum power levels for ] pump operation.

Figure 6-1:

Will be revised to delete the [ ] limit line

Technical Justification: The original methodology included analyses with [ ] reactor coolantpumps in operation at 50% full power. The Oconee Technical Specifications were revised todelete the [ I operating conditions. Therefore, the [ I operating conditions Dcontent can be deleted. This includes text in Sections 6.4 and 6.4.2, and also Figure 6-1 ofReference 6-1.

Revision 6-8 Revision 2 to DPC-NE-2003P-A, Section 7.0 References


1. Oconee Nuclear Station Reload Design Methodology II, DPC-NE-1002P-A,Revision 3 (approval date)

2. VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores, EPRI NP-251 1-CCM,Revision 4, EPRI, February 2001

9. UFSAR Chapter 15 Transient Analysis Methodology, DPC-NE-3005-PA,Revision 3, (approval date)

10. RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis ofComplex Fluid Flow Systems, EPRI-NP-7450(A), EPRI, Revision 5, July2001

Page 27

11. Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005-PA,Revision 4, (approval date)

13. BHTP DNB Correlation Applied with LYNXT, BAW-10241(P)(A), Revision1, Framatome ANP, July 2005

14. The BWU Critical Heat Flux Correlations, BAW-10199P-A, FramatomeCogema Fuels, August 1996 (and including Addendum 1, December 2000)

15. Reactor Vessel Model Flow Tests, BAW-10037, Revision 2, Babcock andWilcox, November 1972

6.2 Revision 4 to DPC-NE-2005P-A - Thermal-Hydraulic Statistical Core DesignMethodology

DPC-NE-2005P-A, "Thermal-Hydraulic Statistical Core Design Methodology" (Reference 6-2),presents the VIPRE-01 methodology for the statistical combination of the uncertainties related tothe calculation of the departure from nucleate boiling ratio (DNBR) into a statistical DNBR limitfor each fuel assembly design and critical heat flux correlation. This report is structured toinclude new fuel assembly designs by adding appendices. For the Mark-B-HTP design,Appendix F is added as described below and is presented as Section 6.3. Revision 4 consists ofthe following revision.

Revision 6-9 Appendix F, Application of BHTP CHF Correlation to the Mark-B-HTP FuelDesign

Description: A new Appendix F, "Application of BHTP CHF Correlation to the Mark-B-HTPFuel Design", is added. The methodology in' this appendix is essentially identical to priorappendices, such as Appendix D for the current Mark-B 11 design, and includes all of the Mark-B-HTP design information and the development of the DNBR limits and statistical DNBR limitsusing VIPRE-OI with the BHTP CHF correlation. Appendix F to DPC-NE-2005 is included hereas Section 6.3. The technical justification is included in the text of the appendix.

Page 28

6.3 ADDendix F to DPC-NE-2005P

DPC-NE-2005P

Thermal-HydraulicStatistical Core Design Methodology

APPENDIX F

Oconee Nuclear Station Specific Data

Mark-B-HTP Fuel

Application of the BHTP CHF Correlation tothe Mark-B-HTP Fuel Design

September 2007

Page 29

DPC-NE-2005-P

Appendix F

This appendix contains the plant-specific data and limits for the Oconee Nuclear Station with

Mark-B-HTP fuel using the BHTP critical heat flux correlation. Below the first intermediate grid

the BWU-N critical heat flux correlation is used. The thermal-hydraulic statistical core design

analysis was performed as described in the main body of this report.

Plant-Specific Data

This analysis is for the Oconee Nuclear Station (two-loop Babcock and Wilcox PWR) as

described in Reference F-1. The parameter uncertainties and statepoint ranges were selected to

bound the Oconee unit and cycle-specific values. This analysis models the 0.430 inch diameter

fuel rod Mark-B-HTP fuel assembly design.

Thermal-Hydraulic Code and Model

The VIPRE-01 thermal-hydraulic computer code described in Reference F-2 and the Oconee

eight and nine channel models approved in Reference F-i are used in this analysis. Due to the

fuel assembly design change, some specific data supplementary to Table 3-1 in Reference F-I

requires updating. This data is listed in Table F-I in this appendix. Table F-i includes fuel rod,

control rod, and instrument guide tube diameters, the number and design of the grids, and the fuel

rod length.

The VIPRE-01 models approved in Reference F-I are used to analyze the Mark-B-HTP fuel with

the following exceptions:

1) The Mark-B-HTP fuel assembly dimensional information is listed in Table F-1.

2) The turbulent mixing factor has been changed to [ ] for the Mark-B-HTP fuel

assembly design due to the presence of HTP grid mixing features, and [ ] for A

the HMP grid. The numerical value was determined and provided by the fuel

supplier.

3). The bulk void fraction model was changed from the Zuber-Findlay model to the

EPRI model. The Zuber-Findlay bulk void model is applicable only to qualities

Page 30

below approximately 0.7 and is discontinuous at a quality equal to 1.0 (Reference F-

2). The EPRI bulk void model is essentially the same as the Zuber-Findlay bulk void

model except for the equation used to calculate the drift velocity (Reference F-2).

This eliminates the discontinuity at a quality equal to 1.0. Therefore, the EPRI model

provides a full range (i.e., void fraction range, 0 - 1.0) of applicability required for

performing DNB calculations. Also, for overall model compatibility, the subcooled

void model was changed from LEVY, as specified in Reference F-I, to the EPRI

correlation for the Mark-B-HTP fuel.

Critical Heat Flux Correlation

The NRC-approved BHTP critical heat flux correlation in Reference F-3 is used for all Mark-B-

HTP analyses, with the exception that below the first intermediate grid the BWU-N critical heat

flux correlation (Reference F-4) is used. The BHTP correlation was developed by AREVA NP

for application to the Mark-B-HTP fuel design. The analysis in Reference F-3 was performed

with the LYNXT thermal-hydraulic computer code (Reference F-5). This correlation was

programmed into the VIPRE-01 thermal-hydraulic computer code and the Mark-B-HTP data base

analyzed in its entirety. The results of this analysis are shown in Tables F-2 and F-3. The

resulting average P/M value, data standard deviation, and CHF correlation limit are within 1% of

the values reported in Reference F-3 (also shown in Table F-2 under LYNXT column).

Figures F-1 through F-4 graphically show the results of this evaluation. Figure F-i shows there is

no bias of VIPRE-01 predicted CHE values to measured values for the data base. Figures F-2

through F-4 show there is no bias with the VIPRE-01 calculated P/M ratios with respect to mass

velocity, pressure, or thermodynamic quality. These figures compare closely with the same

parameter representations in Reference F-3.

Based on the results shown in Tables F-2 and F-3 and Figures F-1 through F-4, the BHTP CHF

correlation licensed in Reference F-3, can bd used in DNBR calculations with VIPRE-01 for

Mark-B-HTP fuel.

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Statistical Core Design Analysis

Statepoints

The statepoint conditions evaluated in this analysis are listed in Table F-4. These statepoints

represent the range of conditions to which the statistical DNB analysis limit will be applied. The

range of key parameter values analyzed is listed in Table F-7.

Key Parameters and Uncertainties

The key parameters and their uncertainty magnitude and associated distribution used in this

analysis are listed in Table F-5. The uncertainties were selected to bound the values calculated

for each parameter. The uncertainties have not changed from Reference F-1 except for the rod

power hot channel factor (Fq), core flow measurement, and CHF correlation. The uncertainty for

Fq has changed due to fuel design changes. The core flow measurement uncertainty was

increased to ensure that it is bounding. This results in a more conservative SDL. The DNBR

correlation uncertainty is a bounding value as compared to Reference F-3. It is noted that the Fz

parameter uncertainty distribution in Table F-5 is treated as a normal distribution. The nuclear

uncertainty database for Fz is actually characterized as nearly normal, with the data more skewed

than normal. It is judged that the data can be treated as normal for the purposes of the SCD

methodology.

DNB Statistical Design Limits

The statistical DNBR limit for each statepoint evaluated is listed in Table F-6. Section 1 of Table

F-6 contains the 500 case runs and Section 2 contains the 5,000 case runs. All of the DNBR

distributions are normally distributed. The maximum statistical DNBR value in Table F-6 (full

core of Mark-B-HTP fuel) for 5,000 propagations is [ ]. Therefore, the statistical design

limit, using the BHTP ClIF correlation for Mark-B-HTP fuel at Oconee, is [ ] for the range of D

parameters given in Table F-7. This result is also valid below the first intermediate grid using the

BWU-N correlation.

Transition Cores

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The transition core model determines the impact of the geometric and hydraulic differences

between the resident Mark-B II fuel and the new Mark-B-HTP design. The 9 channel model

described in Reference F-I is used to evaluate the impact of transition cores containing Mark-B-

HTP fuel. Referring to Figure 4-5 in Reference F-I, Mark-B-HTP fuel is modeled instead of

Mark-B6/7, and Mark-B 11 fuel is modeled instead of Mark-B5. Therefore, Channels I - 7 are

modeled as Mark-B-HTP fuel, Channel 8 is modeled as Mark-B II fuel, and Channel 9 is

modeled as Mark B-HTP fuel. The transition core analysis models each fuel type in those

respective locations with the correct geometry. The form loss coefficients and geometry for each

fuel design are input so the effect of crossflow between fuel designs is calculated.

A transition core penalty is evaluated by determining the DNBR impact on a Mark-B-HTP

limiting assembly when analyzed with the 9 channel model. Once determined, several methods

are available to conservatively account for the penalty. One method of accounting for the

reduction in DNB performance due to the hydraulic effects of the conservatively modeled

transition core is to explicitly apply a penalty to the Mark-B-HTP fuel generic peaking limits

based on a full Mark B-HTP core. Another option is to calculate maximum allowable peaking

limits specifically modeling the transition core loading pattern in the detailed 64 channel model

approved in Reference F-I. These methods will be used, as necessary, to determine the DNB

effect of transition cores.

To evaluate the impact of the transition core on the statistical DNB analysis, the most limiting

statistical DNB statepoint (Statepoint 17 in Table F-6) was evaluated using the 9 channel model.

This statepoint is designated 26-T in Table F-6. At 5,000 cases, the statistical DNBR for

statepoint 26T is lower than the statistical design limit, [ ]. Therefore, the statistical design D

limit [ ] is bounding for Mark-B-HTP/Mark-B II transition cores, as well as for full Mark-B-

HTP cores.

References

F-I Oconee Nuclear Station Core Thermal-Hydraulic Methodology Using VIPRE-01, DPC-NE-2003P-A, Revision 2, (approval date)

F-2 VIPRE-01: A Thermal-Hydraulic Code For Reactor Cores, EPRI NP-251 I-CCM-A, Vol.1-4, Revision 4, February 2001

F-3 BHTP DNB Correlation Applied with LYNXT, BAW-10241(P)(A), Revision 1,Framatome ANP, July 2005

Page 33

F-4 The BWU Critical Heat Flux Correlations, BAW-10199P-A, Framatome Cogema Fuels,August 1996 (and including Addendum 1, December 2000)

F-5 LYNXT Core Transient Thermal-Hydraulic Program, BAW-10156-A, Revision 1, B&WFuel Company, August 1993

Page 34

Table F-I

Mark-B-HTP Fuel Assembly Data

Dimensions (nominal inches)

Fuel rod diameter 0Guide tube diameterInstrument guide tube diameterFuel rod pitchFuel assembly pitch 8Fuel rod length

Design Characteristics

.430

.58755.0

]]] A

Material Quantity Location Type

Grids Inconel 718 I Lower HNMP non-mixing

HTP non-mixing

Fuel Rods

Guide Tubes

InstrumentTube

M5®

M5®

M5®

M5®

7

208

Intermediateand upper

16

I

Page 35

Table F-2

VIPRE-01 BHTP Correlation Verification

VIPRE-OI/LYNXT Statistical Results

n, # of data points

P/M, average predicted to measured CHF

cy (M/P/N)

DNBR correlation limit

VIPRE-01

[ A,D

]A,D

]A,D

1.120

LYNXT

]A

1.]A

[ A

1.132

Table F-3

C-F Test Database Analysis Results

Parameter Ranges

Pressure (psia)

Mass flux (Mlbm/hr-ft2)

Thermodynamic quality at CHF

Thermal-hydraulic computer code

Spacer grid

Correlation design limit DNBR

1385 to 2425

0.492 to 3.549

Less than 0.512

VIPRE-01

Mark-B-HTP 15x15

1.132

Page 36

Table F-4

Oconee SCD Statepoints

Statepoint Power(')

Number (% RTP)

I

"2

34

5

678

9

10

1112

13

14

15

16

17

18

19

20

21

22

23

24

25(3) 26

RCS Flow (2)

% DFPressure.(sia)

Core InletTemperature

(OF)Axial Peak(F, @ Z)

Radial PeakFAH

Notes: D1) 100% RTP = 2568 MWt2) 100% design flow = 352,000 gpm3) Statepoint 17 was rerun with the mixed Mark-B-HTP/Mark-B 11 core Oconee

9 channel model as per Reference F-I

Page 37

Table F-5

Oconee Statistically Treated Uncertainties

Parameter

Reactor System

Core Power

Core Flow

Pressure

Temperature

Nuclear

FAH

Fz

z

Fq

Hot Channel FlowArea

DNBR

DNBR

TMear

Measurement

Measurement

Measurement

Measurement

Type ofDistribution

Normal

Normal

Normal

Normal

Uncertainty

+/-2.0%FP

+/-4.2% design

+/-30.0 psi

+/-2.0°F

StandardDeviation

+/- 1.0%FP

+/-2.1% design

+/-15.0 psi

+/-1.0 0 F

Calculation

Calculation

Calculation

Calculation

Measurement

Correlation

Code

Normal

Normal

Uniform

Normal

Uniform

Normal

Normal

+/-6 inches

[ ]

+/-2.84%

+/-2.91%

A

[ ] A,D

JDI

Page 38

Table F-5 (cont.)

Oconee Statistically Treated Uncertainties

JustificationParameter

System Pressure

Inlet Temperature

Core Power

Core Flow

Radial Power

This uncertainty accounts for random uncertainties in various instrumentationcomponents. Since the random uncertainties are normally distributed, the squareroot of the sum of the squares (SRSS) that results in the pressure uncertainty isalso normally distributed.

Same approach as pressure uncertainty.

The core power uncertainty was calculated by statistically combining variousrandom uncertainties associated with the measurement of core power. Since therandom uncertainties are normally distributed, the SRSS that results in the corepower uncertainty is also normally distributed.

Same approach as core power uncertainty.

This uncertainty accounts for the error associated in the physics code's calculationof radial assembly power (FAh) and the measurement of the assembly power. Thisuncertainty distribution is normal.

Axial Peak Power This uncertainty accounts for the axial peak (Fz) prediction uncertainty of thephysics codes. The uncertainty is assumed normally distributed.

Axial Peak Location This uncertainty accounts for the possible error in interpolating on axial peaklocation (Z) in the maneuvering analysis. The uncertainty is based on the axialnode length in the' physics code. The uncertainty distribution is conservativelyapplied as uniform.

Rod Power HCF This uncertainty accounts for the increase in rod power due to manufacturingtolerances. The uncertainty in calculating the peak pin from assembly radial peakis also statistically combined with the manufacturing tolerance uncertainty toarrive at the correct value. The uncertainty is normally distributed andconservatively applied as one-sided in the analysis to assure the MDNBR channellocation is consistent for all cases.

Hot Channel Flow Area This uncertainty accounts for manufacturing variations in the instrument guidetube subchannel flow area. This uncertainty is uniformly distributed and isconservatively applied as one-sided in the analysis to ensure the MDNBR channellocation is consistent for all cases.

DNB Correlation This uncertainty accounts for the CHF correlation's ability to predict DNB. Theuncertainty distribution is applied as normally distributed..

Code/Model This uncertainty accounts for the thermal-hydraulic code uncertainties andoffsetting conservatisms. This uncertainty also accounts for the small DNB.prediction differences between the various model sizes. The uncertaintydistribution is normally distributed.

Page 39

Table F-6

Oconee Statepoint Statistical ResultsBHTP Critical Heat Flux Correlation

500 Case Runs

Coefficient of StatisticalDNBRStatepoint #

!

234567891011121314151617181920212223242526

Mean 'Y Variation

D

Page 40

Table F-6 (cont.)

Oconee Statepoint Statistical ResultsBHTP Critical Heat Flux Correlation

5000 Case Runs

Coefficient ofVariation

StatisticalDNBRStatepoint #

9-T1 I-T15-T16-T17-T18-T20-T24-T25-T26-T

Mean

D

Page 41

Table F-7

Oconee Key Parameter Ranges

Parameter Maximum Minimum

Core power (%RTP)

RCS flow (% Design)

Pressure (psia)

T inlet ('F)

FAH, Fz, ZD

Note: All values listed in this table are based on the currently analyzed statepoints. Ranges are

subject to change based on future statepoint conditions.

Page 42

Figure F- I

VIPRE-01 Predicted CHF Versus Measured ClFMark-B-HTP Data Base

A, D

Page 43

Figure F-2

VIPRE-01 Predicted-to-Measured CHF vs. Mass VelocityMark-B-HTP Data Base

A, D

Page 44

Figure F-3

VIPRE-01 Predicted to Measured CHF vs. PressureMark-B-HTP Data Base

A, D

Page 45

Figure F-4

VIPRE-01 Predicted to Measured CHF vs. QualityMark-B-HTP Data Base

A, D

Page 46


This section addresses any core thermal-hydraulic design methodology issues that are related tothe mixed core configuration with AREVA NP Mark-B 11 and Mark-B-HTP fuel assemblydesigns co-resident in reload core designs.

The transition core model determines the impact of the geometric and hydraulic differencesbetween the resident Mark-B 11 fuel and the new Mark-B-HTP design. The 9 channel modeldescribed in Reference 6-1 is used to evaluate the impact of transition cores containing Mark-B-HTP fuel. Referring to Figure 4-5 of Reference 6-1, Mark B-HTP fuel is modeled instead ofMark-B6/7, and Mark-B 11 fuel is modeled instead of Mark-B5. Therefore, Channels 1-7 aremodeled as Mark-B-HTP fuel, Channel 8 is modeled as Mark-B 11 fuel, and Channel 9 ismodeled as Mark-B-HTP fuel. The transition core analysis models each fuel type in thoserespective locations with the correct geometry. The form loss coefficients and geometry for eachfuel design are input so the effect of crossflow between fuel designs is calculated.

A transition core penalty is evaluated by determining the DNBR impact on a Mark-B-HTPlimiting assembly when analyzed with the 9 channel model. Once determined, several methodsare available to conservatively account for the penalty. One method of compensating for thereduction in DNB performance due to the hydraulic effects of the conservatively modeledtransition core is to explicitly apply a penalty to the Mark-B-HTP fuel generic peaking limitsbased on a full Mark-B-HTP core. Another option is to calculate maximum allowable peakinglimits specifically modeling the transition core loading pattern in the detailed 64 channel modelapproved in Reference 6-1. These methods will be used, as necessary, to determine the DNBeffect of transition cores.

To evaluate the impact of the transition core on the statistical DNB analysis (using the statisticaldesign methodology of Reference 6-2) the most limiting statistical DNB statepoint was evaluatedusing the 9 channel model. At 5,000 cases, the statistical DNBR for this statepoint is less than thestatistical design limit of [ ]. Therefore, the statistical design limit of [ I is bounding for DMark-B-HTP/Mark-B 11 transition cores; as well as, full Mark-B-HTP cores.

6.5 References

6-1 Oconee Nuclear Station Core Thermal-Hydraulic Methodology Using VIPRE-01, DPC-NE-2003-PA, Revision 1, September 2000

6-2 Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005-PA, Revision 3,September 2002



Page 47

7.0 UFSAR CHAPTER 15 NON-LOCA TRANSIENT AND ACCIDENT ANALYSISMETHODOLOGY REVISIONS

This chapter details the proposed methodology revisions to Duke's NRC-approved UFSARChapter 15 non-LOCA transient and accident analysis methodology reports that are applicable tothe Oconee Nuclear Station. Each revision is described in detail along with the technicaljustification. A markup to the current version of the methodology report is also presented.Deleted text is designated by a strikethrough, and new text is designated by bold font.

7.1 Revision 4 to DPC-NE-3000-PA - Thermal-Hydraulic Transient Analysis Methodology

DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology" (Reference 7-1),describes the RETRAN-3D system transient thermal-hydraulic models, and the VIPRE-01transient core thermal-hydraulic models, that are used to simulate non-LOCA UFSAR Chapter 15transients and accidents. Validation of these models is included in the report. Revision 4 consistsof the following revisions.

Revision 7-1 New Appendix D, Methodology Revisions for Mark-B-HTP Fuel

Description: A new Appendix D (Section 7.3 below) is added to describe the RETRAN-3D andthe VIPRE-01 modeling of the Mark-B-HTP fuel design and the critical heat flux equations thatare used to predict the DNBR.

See Section 7.3 - Appendix D to DPC-NE-3000-P

Technical Justification: New Appendix D (Section 7.3 below), "Methodology Revisions forMark-B-HTP Fuel", provides a description of-the RETRAN-3D and VIPRE-01 models that areused to simulate the Mark-B-HTP fuel design. The appendix also describes the critical heat flux(CHF) correlations that are used for the Mark-B-HTP design. These are the AREVA NP BHTPcorrelation (Reference 7-2) for most of the DNBR calculations. Below the first intermediate gridthe BWU-N correlation is used (Reference 7-3). For some of the main steam line break analysesthe Modified-Barnett CHF correlation (Reference 7-4) is used so that the low pressure values atthe limiting statepoint are within the range of the CHF correlation.

The Mark-B-HTP fuel design is different than the Mark-B I1 design mainly due to larger fuelpellet and rod diameters, thicker cladding, a non-mixing vane grid design, and a different axialpressure drop distribution. The RETRAN-3D and VIPRE-01 models for the Mark-B-HTP designare developed in the same manner as the previous models for the Mark-B 11 and earlier fueldesigns as detailed in DPC-NE-3000-PA. This involves using controlled design data supplied byAREVA NP to develop the input files. The Mark-B-HTP fuel design has a higher pressure dropthan the Mark-B 11, and will result in a reduction in core flow.

Page 48

The CHF correlation that is generally used for the Mark-B-HTP design is the AREVA NP BHTPcorrelation. The range of conditions for which the BHTP correlation is valid is:

Pressure (psia) 1385 to 2425Mass flux (Mlbnmhr-ft2) 0.492 to 3.549Quality Less than 0.512

Using the VIPRE-01 code, Duke has verified that the BHTP correlation design limit is 1.132.Applying Duke's statistical design methodology, the statistical design limit is [ ] for the Dfollowing range of conditions (this includes use of the BWU-N correlation below the firstintermediate grid):

Core power (%RTP) 76.0 to 140.0RCS flow (% of design) 60.0 to 115.0Pressure (psia) 1600 to 2242Core inlet temperature (F) 505.0 to 572.8

Due to some of the main steam line break cases reaching pressure values lower than the BHTPcorrelation applicability limits shown above, the Modified-Barnett CHF correlation is used. TheModified-Barnett C-IF correlation is valid for the following range of conditions:

Pressure (psia) 150 to 725Mass flux (Mlbm/hr-ft2) 0.03 to 1.70Enthalpy rise (ABtu/Ibm) 6.0 to 373.0Heated length (inches) 32.9 to 174.8Axial heat flux Uniform

AREVA NP has obtained NRC approval for the Modified-Barnett CHF correlation forapplication to main steam line break in Reference 7-5. The 95/95 correlation DNBR limit is1.135 (References 7-6 and 7-7) for the Modified-Barnett CHF correlation. Duke will use the1.135 DNBR limit established by AREVA NP. This correlation will be used for main steam linebreak analyses when the BHTP correlation cannot be used.

It is noted that there is a range of pressure values for which neither the BHTP nor the Modified-Barnett CHF correlations are valid. Duke will conservatively reduce the pressure values to the725 psia upper range limit for the Modified-Barnett if a limiting statepoint falls between theapplicable pressure ranges.

The hot channel power factor, Fq, for the Mark-B-HTP design is computed statistically from the

average or overall variation on rod diameter, enrichment, and fuel weight per rod. It is applied tothe heat generation rate in the pin; thus it will have an effect on all terms that are computed fromthis heat rate with the exception of the heat flux for DNB ratio computation. The value of Fq

used is [ ] for Mark-B-HTP fuel. I A

Revision 7-2 New Appendix E - Expanded Oconee VIPRE-01 Methodology

Description: A new Appendix E (Section 7.4 below) is added to describe the expanded OconeeVIPRE-01 methodology, which consists of a new large VIPRE-01 model with enhanced

Page 49

modeling capability, and a revised approach to the core power distribution used in VIPRE-01models.

See Section 7.4 - Appendix E to DPC-NE-3000-P

Technical Justification: Section 2.3 of DPC-NE-3000-PA (Reference 7-1) details the genericOconee VTPRE-01 models used for transient core thermal-hydraulic analysis. These models areused to predict the location of minimum DNBR for most of the UFSAR Chapter 15 transients andaccidents, as well as other thermal results such as fuel pellet and cladding temperatures. DPC-NE-3005-PA (Reference 7-8) details other VIPRE-01 models with specific application toparticular UJFSAR Chapter 15 events. These models were developed to provide conservativepredictions of the minimum DNBR as well as to be computationally efficient. A nodalizationreduction approach was employed to justify the use of simplified models with a limited numberof subchannels. A generic conservative core radial and pin power distribution, provided by thefuel vendor, was also used.

The purpose of Appendix E (Section 7.4 below) is to describe an expanded VIPRE-01methodology that includes two features. The first feature is a new larger VIPRE-01 nodalizationmodel that enables modeling of most of the hot assembly as well as parts of three adjacent fuelassemblies. With this new model essentially any pin power distribution predicted by theSIMULATE-3 (Reference 7-9) or SIMULATE-3K (Reference 7-10) physics codes can bespecifically modeled with VIPRE-01. Also, the thermal-hydraulic effects of the inter-assemblygap can be modeled. This large VIPRE-01 model provides modeling capabilities that are notpossible with the models described in Reference 7-1, Section 2.3, and Reference 7-8. Advancesin computing technology make this expanded methodology possible. The second feature is theoption to use core and pin power distributions other than the generic core and pin powerdistributions shown in Reference 7-1, Figures 2.3-3 and 2.3-4. Advances in fuel assembly designsuch as radial zoned enrichment and different burnable poison rod loading patterns, and issuessuch as fuel assembly bow, indicate a need for the capability to model actual core and pin powerdistributions.

Reference 7-1, Section 2.3, Figures 2.3-5, 2.3-6, and 2.3-7, show the current generic OconeeVIPRE-01 models used for transient core thermal-hydraulic analysis. Other special-purposeVIPRE-01 models are described in Reference 7-8. Figures E-I and E-2 show the [ ] channelVIPRE-01 model that is proposed as an option in addition to the Reference 7-1, Section 2.3 andReference 7-8 models.

DThe numbering

Page 50

convention used for the subchannels is to start in the approximate center of the hot assembly andnumber the channels sequentially in a clockwise direction spiraling out to the outer lumpedassemblies.

This modeling approach is consistent with past practices. Development of input for this VIPRE-01 model is the same as the approach for the models detailed in Reference 7-1, Section 2.3.Selection of code options is also the same. The main difference is the inclusion of the modelingof the inter-assembly gap as a different width than the intra-assembly gaps.

Reference 7-1, Section 2.3, Figures 2.3-3 and 2.3-4, show the current core radial powerdistribution and the pin power distribution used to develop the VIPRE-01 input. These wereprovided by the fuel vendor and have been used as bounding conservative inputs. Use of corepower distribution input from the SIMULATE-3 physics code used for core reload design andUFSAR Chapter 15 statepoint analysis is proposed as an alternative to the generic vendor pinpower distribution.

A reconstructed pin power distributionfrom SIMULATE-3 is shown in Figure E-4. The Figure E-4 power distribution has been adjustedso that the peak pin is 1.714 for comparison with Figure E-3. It is noted in Figure E-4 that thelocation of the peak pin is no longer near the center of the fuel assembly, and that there is [ D

] Figure E-4 is far moretypical of an actual pin power distribution than the conservative generic vendor pin powerdistribution that has been used previously for VIPRE-01 modeling for Oconee.

The proposed method for using SIMULATE-3 -based core radial and pin power distributions forinput to Oconee VIPRE-01 models, as an option rather than the generic vendor inputs inReference 7-1, Section 2.3, is as follows. First, a SIMULATE-3 power distribution typical ofcurrent reload designs is selected. Then, for each pin a delta-power is calculated by subtractingits peaking factor from the maximum pin peaking factor in that assembly. Next, a multiplier isselected and applied to each delta-power value, to produce an adjusted smaller delta-power valuefor each pin. These adjusted delta power values are then subtracted from the maximum pinpeaking factor to create a revised higher pin peaking factor for each pin. This process results in amore conservative pin power distribution to use in VIPRE-01 due to both more energy input andalso by nature of being flatter. This process is illustrated in Figure E-5. The pin powerdistribution in Figure E-4 is used to develop the delta power values for each pin (Figure E-5 (a).These delta values are adjusted by applying a multiplier of 0.75 (example multiplier value) toproduce the adjusted delta power values in Figure E-5(b). The revised pin power distribution isobtained by subtracting each of these values from the maximum pin peaking factor of 1.714 (fromFigure E-4), as shown in Figure E-5(c). Results from this process would then be used for pinpower distribution input for the Oconee VIPRE-01 models in Reference 7-1, Section 2.3.

The value of the multiplier used to flatten the pin power distribution will be selected so that theflatness of the actual power distribution in the hot assembly for each reload design will bebounded by the flatness of the pin power distribution input to VIPRE-01. Defining a specificvalue of the multiplier is not proposed as an element of the methodology due to the need forflexibility due to the variation in local reactivity associated with the various possible fuelassembly designs (assembly enrichment, radial zoned pin enrichment, burnable poison rodloading concentration/number/placement, exposure, etc.).

Page 51

7.2 Revision 3 to DPC-NE-3005-PA - UFSAR Chapter 15 Transient Analysis Methodology

DPC-NE-3005-PA, "USFAR Chapter 15 Transient Analysis Methodology" (Reference 7-8),describes the methodologies for simulating the UFSAR Chapter 15 non-LOCA transients andaccidents. The RETRAN-3D code is used for the system transient simulation, and the VIPRE-01code is used for the transient DNBR analysis. The CASMO-3/SIMULATE-3 code system is usedfor calculating physics parameters and simulating power distributions. The SIMULATE-3K codeis used for the three-dimensional kinetics analysis for the rod ejection accident. Themethodologies for determining the safety analysis physics parameters and the safety analysissetpoints are also included. For each UFSAR Chapter 15 event the initial conditions andboundary conditions are specified.

Revision 7-3 - Section 2.3.1 Code Description (p. 2-10)

Description: Duke has implemented source code revisions to the EPRI VIPRE-01 MOD2 code.The list below includes only new methodology and not error corrections or user convenienceitems. (the revisions that have not been previously submitted for NRC review are designated inbold font)

The yersion of the VIPRE-01 code currently used in the analyses is a Duke version ofVIPRE-O0/MOD2. The Duke version of the code includes additional features andeditorial changes so that the constitutive equations, correlations, and solution schemes ofthe VTPRE-01/MOD2 code have been preserved. These additional features and editorialchanges are described below:

* Add the following critical heat flux (CHF) correlations:1. BWC CHF correlation2. BWCMV CHF correlation3. BWU-Z and BWU-N CHF correlations4. BHTP CHF correlation

" Add the ability to print the friction, form loss elevation, acceleration and crossflow pressure drops for specified channels

* Add the option to allow the user to use either a linear interpolation or spline fitfor the input nodal axial power profile

" Add the option to generate a summary file of the minimum DNBR value* Add the option to allow the user to input the power hot channel factor (Fq) and

the local heat flux hot channel factor (Fq") to a subchannel in order toconservatively calculate the DNBR in that subchannel

" Enhance the logic used when VIPRE-01 is utilized to iterate on a parameter, suchas radial power, to converge to a MDNBR limit

" Add the option to allow the code to switch from the BWU-Z correlation tothe BWU-N correlation to analyze the non-mixing grid spans of the Mark-Bll fuel assembly design

" Add the option to allow the BWU-Z correlation to use the actual length ofthe top mixing vane grid spacing of the Mark-B11 fuel assembly design

Page 52

0 Add the option to allow the code to switch from the BHTP correlation to theBWU-N correlation to calculate DNBR within the non-HTP grid span of theMark B-HTP fuel assembly design

* Add the option to allow the BHTP correlation to use the actual length of thetop grid spacing of the Mark B-HTP fuel assembly design

* Add the option to allow the use of constant or axially-variable turbulentmixing coefficient for the various grid designs of the Mark B-HTP fuelassembly

* Add the calculation of fuel average enthalpy

Technical Justification: The code revisions highlighted in bold font (with the exception of thelast bulleted item) are required to correctly analyze AREVA NP fuel designs with the BWU orthe BHTP CIF correlations. These designs are complicated by different spacer grids being usedin the same fuel assembly thus requiring the use of different CHIF correlations. In the currentMark-B 11 design, the BWU-Z correlation is used except for the lower segment of the fuelassembly where the BWU-N correlation must be used due to the lowest grid being a non-mixingvane grid. For the Mark-B-HTP design, the lower grid is also a different design requiring the useof the BWU-N correlation. These modifications to the VIPRE-01 code are all necessary tocorrectly apply the AREVA NP CHF correlations. The last bulleted item, the calculation of fuelaverage enthalpy, was added to automate calculation of this key result for the rod ejectionanalysis.

Revision 7-4 Section 2.3.2 Simulation Models (p. 2-12)

Description: The [ ] channel VIPRE-01 models described in Reference 7-1, Section2.3, and the [ J channel VIPRE-01 model developed in Reference 7-1, Appendix E, areincluded as models that can be applied for Oconee UFSAR Chapter 15 transients and accidents.This is a new paragraph at the end of Section 2.3.2.

The [ ] channel VIPRE-01 models described in Section 2.3 of Reference 2-3 may also be applied for Oconee transient core thermal-hydraulic analyses.Normally the [ I channel model is sufficient, but the [ ] channel modelshave expanded modeling capability and may be applied in some situations. Also, the[ ] channel Oconee VIPRE-01 model described in Appendix E of Reference 2-3may be applied for situations where additional modeling detail is required.

Technical Justification: This revision includes the [ ] channel Oconee VIPRE-01models in addition to the other VIPRE-01 models already described in Section 2.3.2. The currentSection 2.3.2 text did not contemplate a need for using the more detailed [ ] channelmodels already described in Section 2.3 of Reference 7-1, and the [ ] channel model is new.Any of these models may be used instead of the [ ] channel model to address issues that can'tbe modeled with the [ ] channel model, or to obtain better analytical results due to less Dsimplification of the actual configuration being analyzed.

Revision 7-5 Table 4-1 (p. 4-5)

Description: The initial power level for three reactor coolant pump operation is revised from80% to 75% full power.

Page 53

Replace: 78% 80% FP *** with 73% - 75% FP

Delete the *** footnote

Technical Justification: The original methodology assumed an initial power level for threereactor coolant pump operation of 80% full power (FP). This assumption was made to support apossible revision of the technical specifications to allow operation at 80% FP rather than thecurrent value of 75% FP. This technical specification revision to 80% FP is no longercontemplated, and so the current technical specification limit of 75% full power will be used.The associated footnote is also deleted since it is no longer needed.

Revision 7-6 Section 9.2.2, VIPRE-01 (p. 9-2)

Description: The BHTP critical heat flux correlation is added for application to the Mark-B-HTPfuel design, and the BWC and BWU-Z correlations are designated as applicable to the Mark-B 10and Mark-B 11 designs, respectively.

The critical heat flux (CHF) correlations used are the BWC CHF correlation for theMark-B10 fuel design (Reference 9-4), and the BWU-Z C-I-F correlation (Reference 9-5) for the Mark-Bll fuel design. For the Mark-B-HTP fuel design the BHTPcorrelation (Reference 9-8) is applicable. The VIPRE analysis employs the SCDmethodology for the loss of flow core thermal-hydraulic analyses.

New Reference

9-8 BHTP DNB Correlation Applied with LYNXT, BAW-10241(P)(A), Revision1, Framatome ANP, July 2005

Technical Justification: The BWC critical heat flux correlation (Reference 7-12) is applicable tothe Mark-B 10 fuel design. The BWU correlation (Reference 7-3) is applicable to the Mark-B 11fuel design. The BHTP correlation (Reference 7-2) is applicable to the Mark-B-HTP fuel design.All of these C-F correlations have been previously reviewed and approved by the NRC. Theapplication of the BWU correlation by Duke has been previously reviewed and approved by theNRC in Reference 7-13. The application of the BHTP correlation by Duke is detailed inAppendix F to DPC-NE-2005-P (Section 6.3).

Revision 7-7 Section 9.3.1.1, Initial Conditions (Core Power) (p. 9-3)

Description: The initial core power level for three reactor coolant pump operation is revised from80% to 75% FP.

Nominal rated thermal power is assumed for the four-pump initial condition and ,0 75%FP for the three-pump initial condition ...

Technical Justification: The original methodology assumed an initial power level for threereactor coolant pump operation of 80% full power (FP). This assumption was made to support apossible revision of the technical specifications to allow operation at 80% FP rather than the

Page 54

current value of 75% FP. This technical specification revision to 80% FP is no longercontemplated, and so the current technical specification limit of 75% full power will be used.


Description: The BHTP critical heat flux correlation is added for application to the Mark-B-HTPfuel design, and the BWC and BWU-Z correlations are designated as applicable to the Mark-B 10and Mark-B 11 designs, respectively.

The critical heat flux (CHF) correlations used are the BWC CHF correlation for theMark-B10 fuel design (Reference 10-4), and the BWU-Z CHF correlation (Reference10-5) for the Mark-Bll fuel design. For the Mark-B-HTP fuel design the BHTPcorrelation (Reference 10-8) is applicable. The VIPRE analysis employs the SCDmethodology for the loss of flow core thermal-hydraulic analyses.

New Reference


Technical Justification: The BWC critical heat flux correlation (Reference 7-12) is applicable tothe Mark-B 10 fuel design. The BWU correlation (Reference 7-3) is applicable to the Mark-B 11fuel design. The BHTP correlation (Reference 7-2) is applicable to the Mark-B-HTP fuel design.All of these CHF correlations have been previously reviewed and approved by the NRC. Theapplication of the BWU correlation by Duke has been previously reviewed and approved by theNRC in Reference 7-13. The application of the BHTP correlation by Duke is detailed inAppendix F to DPC-NE-2005-P (Section 6.3).

Revision 7-9 Section 10.3.1.1, Initial Conditions (Core Power) (p. 10-3)

Description: The initial core power level for three reactor coolant pump operation is revised from80% to 75% FP.

Nominal rated thermal power is assumed for the four-pump initial condition and 80 75%FP for the three-pump initial condition...

Technical Justification: The original methodology assumed an initial power level for threereactor coolant pump operation of 80% full power (FP). This assumption was made to support apossible revision of the technical specifications to allow operation at 80% FP rather than thecurrent value of 75% FP. This technical specification revision to 80% FP is no longercontemplated, and so the current technical specification limit of 75% full power will be used.

Revision 7-10 Section 13.3, Boundary Conditions (Manual Operator Actions) (p. 13-5)

Description: The Reactor Coolant System boron concentration can also be determined by acalculation, in addition to a sample measurement.

Page 55

Cooldown of the RCS is halted upon reaching 450'F. RCS b•r..n sampling is completedat-4is-tiffe. A determination of the RCS boron concentration is completed at thistime. An operator action delay time of 90 minutes is assumed.

Technical Justification: The current methodology requires a boron sample measurement todetermine the Reactor Coolant System boron concentration prior to continuing with thecooldown. This action is taken to ensure that adequate subcritical margin is maintained. Therevision uses the words "determination of the RCS boron concentration" rather.than "RCS boronsampling" to include the alternative of calculating the boron concentration. This alternativemethod (performing a calculation of the boron concentration) is desired to avoid the radiologicaldose that personnel may be subjected to during the sampling process.

Revision 7-11 Chapter 14, Rod Ejection Analysis (pages throughout Chapter 14), and tables andfigures referenced from these pages)

Description: The ARROTTA code is deleted throughout because it is no longer used (only theSIMULATE-3K code is used).

All content related to ARROTTA will be deleted in Revision 1. "SIMULATE-3K"will replace "ARROTTA" in some text locations.

Technical Justification: The original methodology used the EPRI ARROTTA code and theSIMULATE-3K code for the three-dimensional kinetics simulation in the rod ejection analysis.ARROTTA is no longer used, and is therefore deleted.

Revision 7-12 Section 14.2.4.1, Peak Pellet Enthalpy (Model Description) (p. 14-3)

Description: The VIPRE []

The+

Technical Justification: The original methodology [] This is an D

incorrect statement that needs to be deleted.

Revision 7-13 Section 14.2.4.1, Peak Pellet Enthalpy (Power Distribution) (p. 14-3)

Description: The axial power distribution [

For hot zero- p-wer (HZP) conditions, the+

Technical Justification: The original methodology states that the [] In the

revised methodology the time-dependent axial shape from SIMULATE-3K is used. [ D

Page 56

] Therefore this statement is deleted.

Revision 7-14 Section 14.2.4.1, Peak Pellet Enthalpy (Fuel Enthalpy Calculation) (p. 14-5)

Description: The VIPRE-01 code has been modified to calculate U0 2 enthalpy as a function oftemperature.

VIPRE-01 has been modified to include does-not-pe+4er- fuel enthalpy calculations.Thus, the fuel enthalpy for a given fuel temperature during the transient is calculatedseparately frem VIPRE based on the equation obtained from MATPRO (Reference 14-8).

Technical Justification: In the original methodology the UO2 enthalpy value was calculatedmanually based on VIPRE temperature results using the equation from MATPRO (Reference 7-15). VIPRE has been modified to do this calculation internal to the code using the same equationfrom MATPRO. This revision is a user convenience.

Revision 7-15 Section 14.2.4.2, DNBR Evaluation (Fuel Conduction Model) (p. 14-8)

Description: The statement that low fuel temperatures are bounding is erroneous and is revised.Also, the word "conductivities" has been revised to "conductances" to correct the terminology.

Both sets of MARP limits assume initial gas gap c conductances that yieldboundingly low high fuel temperatures, Low fiuel temperatures .esult in a high gapeendueti-ity -which translates into a higher heat flux during the transient.

Technical Justification: The erroneous statement is corrected to state that high fuel temperaturesare bounding. The methodology has been correctly applied - only the statement in the report wasin error. The word "conductances", the correct terminology for the parameter being described,replaces "conductivities".

Revision 7-16 Section 14.2.4.1, Peak Pellet Enthalpy (Heat Transfer Correlations) (p. 14-4);Section 14.2.4.2, DNBR Evaluation (Heat Transfer Correlations) (p. 14-8); Section14.2.4.3, Coolant Expansion Rate (Heat Transfer Correlations) (p. 14-10)

Description: The critical heat flux correlations are revised to delete the BWC correlation and toadd the BWU and BHTP correlations.

Section 14.2.4.1The critical heat flux correlations used to define the peak of the boiling curve and topredict the DNBR is are the BWU CHF correlation (Reference 14-10) for Mark-B11fuel, and the BHTP CHF correlation (Reference 14-11) and the BWU-N CHFcorrelation (Reference 14-10) for the Mark-B-HTP fuel. The same CGP correlationswill be used that is used to predict the DNB-RR. The, minimum DNIRR value. for- th _peaof the curv~e is set to 1.24, which is the correlationi limit of 1. 18 plus 5% mar-gini.

Page 57

Section 14.2.4.2The critical heat flux correlations used to define the peak of the boiling curve and topredict the DNBR are is the BWU CHF correlation (Reference 14-10) for Mark-Bllfuel, and the BHTP CHF correlation (Reference 14-11) and the BWU-N correlation(Reference 14-10) for the Mark-B-HTP fuel. tWC correlation. The frinimum DNBRvalue for which transition boiling occurs is set to 1.24.

Section 14.2.4.3The critical heat flux correlations used to define the peak of the boiling curve are is theBWU CHF correlation (Reference 14-10) for Mark-Bll fuel, and the BHTP CHFcorrelation (Reference 14-11) and the BWU-N correlation (Reference 14-10) for theMark-B-HTP fuel. is the BWC, co.relation for, the results presented (the sam..e orrelatioused to predict the DNBR). The minimum DNBR value for- the peakc of the curvei stt1. 21 ,A,wh iceh i S t hee corrleIa tioan i wit of 1.18 pl1u s 5 % margin

Technical Justification: The BWC critical heat flux correlation (Reference 12) has beensuperseded by the BWU correlation (Reference 7-3) for the Mark-B 11 fuel design, and by theBHTP correlation (Reference 7-2) for the Mark-B-HTP fuel design. Both of these CHFcorrelations have been previously reviewed and approved by the NRC. The application of theBWU correlation by Duke has been previously reviewed and approved by the NRC in Reference7-13. The application of the BHTP correlation by Duke is detailed in Appendix F to DPC-NE-.2005-P (Section 6.3 of this submittal).

Revision 7-17 Section 14.2.4.3, Coolant Expansion Rate Calculation (Radial PowerDistribution) (p. 14-10)

Description: The number of radial power distribution forcing functions [ D] and is deleted. The normalization of these data is extraneous detail and is deleted. This

change is editorial only and no technical justification is necessary.

Radial Power DistributionsThe transient assembly radial power distributions generated by ARROT4T-, erSIMULATE-3K are used for the analysis. [

D

Revision 7-18 Section 14.3, Nuclear Analysis (p. 14-12)

Description: The number of radial nodes per fuel assembly is corrected from one to four.

Model geometry is typically one nede four nodes per fuel assembly in the radialdirection.

Technical Justification: The original methodology using the ARROTTA code used one radialnode per fuel assembly. Using the SIMULATE-3K code, four radial nodes per assembly havebeen used and will continue to be used. Therefore, this statement is incorrect. Since only theSIMULATE-3K code will be used, the number of radial nodes per assembly is corrected to four.

Page 58

Revision 7-19 Section 14.3.1, SIMULATE-3K Analysis (Ejected Rod Location and Velocity)(p. 14-13)

Description: The modeling of the motion of the ejected rod with constant velocity is deleted.

The core location of the ejected control rod is chosen specifically foreach analysis toproduce the most conservative results. Figure 14-1 shows the core configuration andlocation of the ejected control rod for the various analyses. For both HZP transients thecontrol rod in core location L-10 is ejected from a fully inserted position in 0.15 secondsat constant veleeity.

Technical Justification: The original methodology specifies that the motion of the ejected rodshould be modeled with constant velocity. The important aspect of the modeling of the ejectedrod is the ejection time of 0.15 second. The additional detail regarding constant velocity isexcessive detail and is deleted.

Revision 7-20 Section 14.3.1, SIMULATE-3K Analysis (p. 14-14)

Description: The [] based on additional experience with applying the methodology.

Initial Pc'wer Di~tributien,

Technical Justification: The original methodology included a modeling approach that wasintended to introduce conservatism that would ensure that future core designs would remainbounded for the rod ejection accident. This approach involved [

] Additional experience with applyingthe methodology has shown that this approach will not necessarily produce conservative resultsfor both the hot zero power and hot full power initial condition cases, and has also shown that theinitial core power distribution is not a key parameter. The revised rod ejection analysismethodology maintains sufficient conservatism without this approach.

Page 59

Revision 7-21 Section 14.3.1, SIMULATE-3K Analysis (Moderator Temperature Coefficient)(p. 14-14)

Description: As an additional method, the moderator temperature coefficient is adjusted to-aconservative value by changing the boron concentration.

The initial condition moderator temperature coefficient is

] While thiscoefficient has little effect on the peak power response of this transient, this conservativeadjustment is-made to ensure consistency with other postulated transients which arelimited by the moderator temperature coefficient. As an additional method, the boronconcentration can be increased to add conservatism.

Technical Justification: The original methodology for [ D] As an additional

method conservatism can be obtained by changing the boron concentration. This approach isincluded as an additional method.

Revision 7-22 Section 14.3.3, Results (p. 14-16)Description: The initial core power level for three reactor coolant pump operation is revised from

82% to 77% FP for future analyses.

Insert the following at the end of the first paragraph:

The demonstration analysis results that follow assume an initial core power level of82% FP for the three reactor coolant pump at power cases. This assumption wasbased on a possible future technical specification revision to 80% FP, plus 2%uncertainty. This uprating is no longer contemplated, and so future analyses willassume 75% + 2% = 77% FP for the initial power level.

Technical Justification: The original methodology and the demonstration analyses assumed aninitial power level for three reactor coolant pump operation of 82% full power (FP) includinguncertainty. This assumption was made to support a possible revision of the technicalspecifications to allow operation at 80% FP rather than the current value of 75% FP. Thistechnical specification revision to 80% FP is no longer contemplated, and so the current technicalspecification limit of 75% full power plus a 2% allowance for uncertainty (77% FP) will be used.

Revision 7-23 Revision 3 to DPC-NE-3005-PA, Chapter 14 References (p. 14-23)

Description: The references are updated to the latest revisions and dates, and to include newreferences, This revision is editorial only and no technical justification is necessary.

14-5 VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores, EPRI NP-2511-CCM, Revision 4, EPRI, February 2001

Page 60

14-6 RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis ofComplex Fluid Flow Systems, EPRI-NP-7450(A), EPRI, Revision 5, July2001

14-9 Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3000-PA,Revision 4, (approval date)

14-10 The BWU Critical Heat Flux Correlations, BAW-10199P-A, FramatomeCogema Fuels, August 1996 (and including Addendum 1, December 2000)


14-12 Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005-PA, Revision 4, (approval date)

14-13 TACO3 - Fuel Pin Thermal Analysis Computer Code, BAW-10162P-A,B&W Fuel Company, October 1989


Description: The critical heat flux correlations used for the steam line break analyses are updatedfor the current Mark-B 11 fuel design and for the future Mark-B -HTP fuel design.

For the with offsite power analysis, the critical heat flux (CHF) correlations used toevaluate the DNBR are the Westinghouse W-3S (Reference 15-3, Appendix D) for theMk-B 10 of: M&,B Ik fuel types,-and the BWU correlation (References 15-7 and 15-8) forMark-B 11 fuel. For the Mark-B-HTP fuel design the CHF correlation is theModified-Barnett correlation as described in Appendix D of Reference 15-2.

For the without offsite power analysis, the [ ] VIPRE-01 model described in DReference 15-2 is used to calculate the transient local coolant properties and DNBR. The DBWC (Reference 15-6) and BWU CHF correlations (References 15-7 and 15-8) are usedto perform the DNBR calculations for the Mk-B 10 and Mk-B 11 fuel assembly types,respectively. For the Mark-B-HTP fuel the BHTP correlation as described inAppendix D of Reference 15-2 is used. The VIPRE-01 analysis employs the SCDmethodology for the offsite power lost case.

Technical Justification: For the current Mark-B 10 and Mark-B 11 fuel designs the critical heatflux (CHF) correlations used in the steam line break analysis with offsite power are the W-3Scorrelation (Reference 7-14, Appendix D) and the BWU-Z correlation (Reference 7-3),respectively. For the steam line break analysis without offsite power the CHF correlations are theBWC correlation (Reference 7-12) and the BWU-Z correlation, respectively.

For the Mark-B-HTP fuel design the CHIF correlation used in the steam line break analysis withoffsite power is the Modified-Barnett correlation (References 7-4 and 7-5). This correlation ispresented in the new Appendix D to DPC-NE-3000-P in Section 7.3 below. For the steam linebreak analysis without offsite power the BHTP correlation (Reference 7-2) is used. Thiscorrelation is also presented in the new Appendix D. The technical justification for Appendix Dis not repeated here.

Page 61

7.3 Appendix D to DPC-NE-3000-P - Methodology Revisions for Mark-B-HTP Fuel

APPENDIX D

METHODOLOGY REVISIONS FOR MARK-B-HTP FUEL

This appendix contains non-LOCA thermal-hydraulic transient analysis methodology revisionsrelated to the Mark-B-HTP fuel assembly design. This fuel design is characterized by largerdiameter fuel pins, larger diameter fuel pellets, non-mixing vane grids, and a different axialpressure drop distribution relative to the current Mark-B 11 fuel design. The information includedin this appendix is in addition to that presented in Sections 2.1.2.1, 2.2.3.1, and 2.3 of the mainbody of this report.

Fuel Assembly Description

The Maik-B-HTP fuel assembly consists of spacer grids, end fittings, fuel rods, and guide tubes.The lower end HMP grid is made of Inconel Alloy 718, while the six intermediate spacer gridsand the upper grid are made of M5®. The intermediate spacer grids are comprised of the HTPnon-mixing vane grid type. Each fuel assembly is a 15 by 15 array containing 208 fuel rods, 16control rod guide tubes, and one incore instrument guide tube, all M5®. The fuel rod consists ofdished-end, cylindrical pellets of uranium dioxide. The fuel assembly and fuel rod dimensions,and other related fuel parameters used in the thermal-hydraulic analyses are given in Table D-1.A drawing of the Mark-B-HTP fuel assembly is shown in Figure D-1. These materials anddesign parameters are used in developing the RETRAN-3D and the VIPRE-01 models for theMark-B-HTP fuel designs.

VIPRE Models

The various VIPRE models described in Section 2.3 of the main body of this report are used forthe Mark-B-HTP thermal-hydraulic analyses with the VIPRE-01 thermal-hydraulic computercode described in Reference D-1. The models are updated to reflect the Mark-B-HTP fuelparameters listed in Table D-I.

Code Option and Input Selections

Thermal-Hydraulic Correlations:No changes to the correlations used in the main body of the report and Appendix A.

Turbulent Mixing Correlations:The turbulent mixing factors for the Mark-B-HTP design, [ ] for the HTP grid and [ Afor the HMP grid, were provided by AREVA NP based on scaled testing.

Page 62

Pressure Losses:The pressure losses are calculated in VIPRE-01 as described in the main body of this report usingthe spacer grid form loss coefficients for the Mark-B-HTP fuel assembly provided by AREVANP.

Critical Heat Flux Correlation

The CHF correlation that is generally used for the Mark-B-HTP design is the AREVA NP BHTPcorrelation (Reference D-2). Below the first intermediate grid the BWU-N correlation is used(Reference D-3). The range of conditions for which the BHTP correlation is valid is:

Pressure (psia) 1385 to 2425Mass flux (Mlbm/hr-ft2) .0.492 to 3.549Quality Less than 0.512

Using the VIPRE-01 code, Duke has verified that the BHTP correlation design limit is 1.132.Applying Duke's statistical design methodology, the statistical design limit is [ I for the Dfollowing range of conditions (this limit is also applicable below the first intermediate grid):

Core power (%RTP) 76.0 to 140.0RCS flow (% of design) 60.0 to 115.0Pressure (psia) 1600 to 2242Core inlet temperature (F) 505.0 to 572.8

Due to some of the main steam line break cases reaching pressure values lower than the BHTPcorrelation applicability limits shown above, the Modified-Barnett CHF correlation (ReferenceD-4) is used. The Modified-Barnett CHF correlation is valid for the following range ofconditions:

Pressure (psia) 150 to 725Mass flux (Mlbmlhr-ft2) 0.03 to 1.70Inlet subcooling (ABtu/lbm) 6.0 to 373.0Heated length (inches) 32.9 to 174.8Axial heat flux Uniform

AREVA NP has obtained NRC review and approval for the Modified-Barnett CHF correlationfor application to main steam line break in Reference D-5. The 95/95 correlation DNBR limit is1.135 (References D-6 and D-7) for the Modified-Barnett CHF correlation. Duke will use the1.135 DNBR limit established by AREVA NP. This methodology will be used for main steamline break analyses when the BHTP correlation cannot be used.

It is noted that there is a range of pressure values for which neither'the BHTP nor the Modified-Barnett CHF correlations are valid. Duke will conservatively reduce the pressure values to the725 psia upper range limit for the Modified-Barnett if a limiting statepoint falls between theapplicable pressure ranges.

Page 63

Hot Channel Factors

The hot channel power factor, Fq, is computed statistically from the average or overall variation

on rod diameter, enrichment, and fuel weight per rod. It is applied to the heat generation rate inthe pin; thus it will have an effect on all terms that are computed from this heat rate with theexception of the heat flux for DNB ratio computation. The value of Fq used is [ ] for AMark-B-HTP fuel.

Table D- 1

Mark-B-HTP Fuel AssemblyComponent Dimensions Used for Thermal-Hydraulic Analysis

Fuel pin diameter 0.430 in.

Control rod guide tube diameter [ ] in.

Instrument guide tube diameter [ in. A

Effective pin pitch [ ] in.

Assembly flow area 40.398 in. 2

Assembly wetted perimeter 309.903 in.

Unit channel flow area 0.178 in.2

Unit channel wetted perimeter 1.354 in.

Unit channel heated perimeter 1.354 in.

Control rod guide tube channel flow area 0.159 in.2

Control rod guide tube channel wetted perimeter 1.433 in.

Control rod guide tube heated perimeter 1.015 in.

Instrumentation guide tube channel flow area 0.166 in.2

Instrumentation guide tube channel wetted perimeter 1.403 in.

Instrumentation guide tube channel heated perimeter 1.015 in.2

Peripheral channel flow area (half channel) 0.116 in.

Peripheral channel wetted perimeter (half channel) 0.677 in.

Peripheral channel heated perimeter (half channel) 0.677 in.

Comer channel flow area (quarter channel) 0.074 in.2

Corner channel wetted perimeter (quarter channel) 0.338 in.

Corner channel heated perimeter (quarter channel) 0.338 in.

Active fuel length 143.0 in.

Page 64

Figure D- I

Mark-B-HTP Fuel Assembly

" M50 HTP Grids (7x) withCurved Flow Channels

" Alloy 718 Lower HMPGrid (Straight FlowChannel)

* Removable Upper EndFitting

" Alloy 718 CruciformSprings

" Crimped Top-HatNut

* M5® Fuel Rods

" M5® Guide Tubes

" M5® Instrument Tube

+ FUELGUARDTM LowerEnd Fitting

Page 65

References

D-1 VIPRE-01: A Thermal-Hydraulic Code For Reactor Cores, EPRI NP-25 11 -CCM-A,Vol. 1-4, Revision 4, EPRI, February 2001

D-2 BHTP DNB Correlation Applied with LYNXT, BAW-10241(P)(A), Revision 1,Framatome ANP, July 2005

D-3 The BWU Critical Heat Flux Correlations, BAW-10199P-A, Framatome Cogema Fuels,August 1996 (and including Addendum 1, December 2000)

D-4 A Correlation of Rod Bundle Critical Heat Flux for Water in the Pressure Range 150 to725 psia, IN-1412, Idaho Nuclear Corporation, July 1970

D-5 SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, EMF-2310(P)(A), Revision 1, AREVA NP, May 2004

D-6 Letter, S. A. Varga (NRC) to J. A. Jones (CP&L), Amendment No. 71 to FOL for H. B.Robinson, July 23, 1982

D-7 Letter, F. Akstulewicz (NRC), to J. F. Mallay (Siemens), Subject: Acceptance forReferencing of Topical Report EMF-84-093(P), Revision 1, February 16, 1999

Page 66

7.4 Appendix E.to DPC-NE-3000-P - Expanded Oconee VIPRE-01 Methodology

APPENDIX E

EXPANDED OCONEE VIPRE-01 METHODOLOGY

Section 2.3 of this report details the generic Oconee VIPRE-01 models used for transient corethermal-hydraulic analysis. These models are used to predict the location of minimum DNBR formost of the UFSAR Chapter 15 transients and accidents, as well as other thermal results such asfuel pellet and cladding temperatures. DPC-NE-3005-PA (Reference E-1) details other VIPRE-01 models with specific application to particular UFSAR Chapter 15 events. These models weredevelopedto provide conservative predictions of the minimum DNBR as well as to becomputationally efficient. A nodalization reduction approach was employed to justify the use ofsimplified models with a limited number of subchannels. A generic conservative core radial andpin power distribution, provided by the fuel vendor, was also used.

The purpose of this appendix is to describe an expanded VIPRE-01 methodology that includestwo features. The first feature is a new larger VIPRE-01 nodalization model that enablesmodeling of most of the hot assembly as well as parts of three adjacent fuel assemblies. With thisnew model essentially any pin power distribution predicted by the SIMULATE-3 (Reference E-2)or SIMULATE-3K (Reference E-3) physics codes can be specifically modeled with VIPRE-01.Also, the thermal-hydraulic effects of the inter-assembly gap can be modeled. This large VIPRE-01 model provides modeling capabilities that are not possible with the models described inSection 2.3 and Reference E-1. Advances in computing technology make this expandedmethodology possible. The second feature is the option to use core and pin power distributionsother than the generic core and pin power distributions shown in Figures 2.3-3 and 2.3-4.Advances in fuel assembly design such as radial zoned enrichment and different burnable poisonrod loading patterns, and issues such as fuel assembly bow, indicate a need for the capability tomodel actual core and pin power distributions rather than use of generic power distributioninputs.

Use of the new methodology in this appendix as an option for licensing applications is proposedalong with continued use of the previously approved methodology. There are no changes beingproposed in the code options used with the new models.

Expanded VIPRE Model

Section 2.3, Figures 2.3-5, 2.3-6, and 2.3-7, shows the current generic Oconee VIPRE-01 modelsused for transient core thermal-hydraulic analysis. Other special-purpose VIPRE-01 models aredescribed in Reference E-1. Figures E-1 and E-2 show the [ ] channel VIPRE-01 model that Dis proposed as an option in addition to the Section 2.3 and Reference E-1 models.

Page 67

I

DThe numbering

convention used for the subchannels is to start in the approximate center of the hot assembly andnumber the channels sequentially in a clockwise direction spiraling out to the outer lumpedassemblies.

Development of input for this VIPRE-01 model is the same as the approach for the modelsdetailed in Section 2.3. Selection of code options is also the same. The main difference is theinclusion of the modeling of the inter-assembly gap as a different width than the intra-assemblygaps.

Revised Core Power Distribution Input

Section 2.3, Figures 2.3-3 and 2.3-4 show the current core radial power distribution and the pinpower distribution used to develop the VIPRE-01 input. These were provided by the fuel vendorand have been used as bounding conservative inputs. Advances in fuel assembly design such asradial zoned enrichment and different burnable poison rod loading patterns, and issues such asfuel assembly bow, indicate a need for the capability to model'actual core and pin powerdistributions rather than use of generic power distribution inputs. Use of core power distributioninput from the SIMULATE-3 physics code, the code used for core reload design and UFSARChapter 15 statepoint analysis, is proposed.

D

] Figure E-4 is far more typical of an actual pin power distribution than theconservative generic vendor pin power distribution that has been used previously for VIPRE-01modeling for Oconee.

The proposed method for using SIMULATE-3-based core radial and pin power distributions forinput to Oconee VIPRE-01 models, as an option instead of the generic vendor inputs in Section2.3, is as follows. First, a SIMULATE-3 power distribution typical of current reload designs isselected. Then, for each pin a delta-power is calculated by subtracting its peaking factor from the

Page 68

maximum pin peaking factor in that assembly. Next, a multiplier is selected and applied to eachdelta-power value, to produce an adjusted smaller delta-power value for each pin. These adjusteddelta power values are then subtracted from the maximum pin peaking factor to create a revisedhigher pin peaking factor for each pin. This process results in a more conservative pin powerdistribution to use in VIPRE-01 due to both more energy input and also by nature of being flatter.This process is illustrated in Figure E-5. The pin power distribution in Figure E-4 is used todevelop the delta power values for each pin (Figure E-5 (a). These delta values are adjusted byapplying a multiplier of 0.75 (example multiplier value) to produce the adjusted delta powervalues in Figure E-5(b). The revised pin power distribution is obtained by subtracting each ofthese values from the maximum pin peaking factor of 1.714 (from Figure E-4), as shown inFigure E-5(c). Results from this process are then used for pin power distribution input for theOconee VIPRE-01 models in Section 2.3, and in VIPRE-01 models in Reference E-1.

The value of the multiplier used to flatten the pin power distribution is selected so that theflatness of the actual power distribution in the hot assembly for each reload design is bounded bythe flatness of the pin power distribution input to VIPRE-01.

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Figure E-VI

[ ]Channel Oconee VIPRE-01 ModelSubchannel Detail

D

Page 70

Figure E-2

[ Channel Oconee VIPRE-01 ModelLumped Channel Detail

D

Page 71

Figure E-3

Hot Assembly Pin Power Distribution

Generic Vendor Input (same as Figure 2.3-4)

A, D

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Figure E-4

Adjusted Hot Assembly Pin Power DistributionFrom SIMULATE-3

1.714 1.663 1.706 1.694 1.688 1.687 1.686 1.686 i.686 1.688 1.690 1.696 1.708 1.662 1.706

1.667 1.674 1.651 1.640 1.637 1.635 1.637 1.635 1.637 1.637 1.640 1.643 1.654 1.674 1.658

1.707 1.648 1.629 1.619 1.617 •j Nx 1.619 1.618 1.620 1.621 1.624 1.634 1.650 1.698

1.689 1.631 1.613 1.608 1.608 1.613 1.613 1.614 1.611 1.613 1.620 1.635 1.683

1.678 1.623 1.606 1.602 1.606 1.608 1.617 1.616 1.618 1.611 1.611 1.610 1.615 1.629 1.674

1.673 1.617 , 1.598 1.603 ,, : 1.627 1.633 1.628 -, , 1.609 1.606 1.624 1.670

1.671 1.617 1.601 1.601 1.609 1.624 1.663 1.635 1.664 1.627 1.615 1.608 1.610 1.624 1.668

1.671 1.615 1.600 1.600 1.607 1.629 1.634 __, 1.635 1.631 1.612 1.607 1.608 1.622 1.667

1.670 1.616 1.601 1.600 1.608 1.622 1.661 1.633 1.662 1.625 1.613 1.607 1.609 1.623 1.667

1.672 1.616 1.596 1.600 .. 1.623 1.629 1.624 •4 1.605 1.603 1.622 1.669

1.677 1.622 1.604 1.598 1.601 1.602 1.610 1.610 1.611 1.605 1.606 1.605 1.611 1.626 1.671

1.688 1.628 1.609 1.600 1.599 1.604 1.605 1.605 1.602 1.605 1.614 1.631 1.680

1.704 1.643 1.622 1.610 1.606 1.607 1.606 1.608 1.610 1.615 1.626 1.644 1.694

1.663 1.666 1.641 1.627 1.623 1.619 1.620 1.619 1.621 1.621 1.626 1.630 1.643 1.665 1.652

1.707 1.652 1 .691 1.676 1.667 1.665 1.663 1 .662 1.664 1.667 1.670 1.679 1.693 1.651 1.699

Figure E-5 (a)

VIPRE-01 Pin Power DistributionAdjustment Process

Pin Delta-Power Values

0.000 0.051 0.008 0.020 0.026 0.027 0.028 0.029 0.028 0.026 0.025 0.018 0.006 0.052 0.008

0.047 0.040 0.063 0.074 0.077 0.079 0.077 0.079 0.077 0.077 0.074 0.071 0.060 0.041 0.056

0.007 0.067 0.085 0.095 0.097 1 0.095 0.096 0.094 1 0.093 0.090 0.080 0.065 0.016

0.025 0.083 0.101 0.107 0.106 0.101 0.101 0.100 0.103 0.102 0.094 0.079 0.031

0.036 0.091 0.108 0.112 0.108 0.106 0.097 0.098 0.096 0.103 0.103 0.104 0.099 0.085 0.041

0.041 0.097 0.116 0.111 0.087 0.081 0.086 0.105 0.108 0.090 0.044

0.043 0.097 0.113 0.114 0.105 0.090 0.051 0.079 0.050 0.087 0.099 0.106 0.104 0.090 0.046

0.043 0.099 0.114 0.114 0.107 0.085 0.080 0.079 0.083 0.102 0.107 0.106 0.092 0.047

0.044 0.098 0.113 0.114 0.106 0.092 0.053 0.081 0.052 0.089 0.101 0.107 0.105 0.091 0.047

0.042 0.098 0.118 0.114 0.091 0.085 0.090 0.109 0.111 0.092 0.045

0.037 0.092 0.111 0.116 0.113 0.112 0.104 0.104 0.103 0.109 0.108 0.109 0.103 0.088 0.043

0.026 0.086 0.105 0.114 0.115 0.110 0.110 0.109 0.112 0.110 0.100 0.083 0.034

0.010 0.071 0.092 0.104 0.108 0.108 0.108 0.106 0.104 0.099 0.088 0.071 0.020

0.051 0.048 0.073 0.087 0.091 0.095 0.094 0.095 0.093 0.093 0.088 0.084 0.071 0.049 0.062

0.007 0.062 0.023 0.038 0.047 0.049 0.051 0.052 0.050 0.047 0.044 0.035 0.021 0.063 0.015

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Figure E-5 (b)

VIPRE-Ol Pin Power DistributionAdjustment Process

Adjusted Delta-Power Values

0.000 0.038 0.006 0.015 0.020 0.020 0.021 0.021 0.021 0.020 0.018 0.014 0.005 0.039 0.006

0.035 0.030 0.047 0.056 0.057 0.059 0.058 0.059 0.058 0.058 10.055 0.053 0.045 0.030 0.042

0.006 0.050 0.064 0.071 0.073 -F0.071 0.072 0.071 0.070 0.067 0.060 0.048 0.012

0.019 0.062 0.076 -: 0.080 0.079 0.076 0.076 0.075 0.077 ý0.076 - _0.070 0.059 0.0230.027. 0.068 0.081 0.084 0.081 0.080 0.073 0.074 0.072 0.077 0.077 0.078 0.074 0.064 0.0300.031 0.073 .< 0.087 0.083 0.065 0.061 0.064 . .0.079 0.081 .i 0.067 0.0330.032 0.073 0.084 0.085 0.079 0.068 0.038 0.059 0.037 0.065 0.075 0.079 0.078 0.068 0.0350.032 0.074- 0.086 0.086 0.080 0.064 0.060 .,.0.059 0.062 0.076 0.080 0.079 0.069 0.035

0.033 0.073 0.085 0.086 0.080 0.069 0.039 0.060 0.039 0.067 0.076 0.080 0.079 10.068 0.035

0.031 0.073 0.089 0.085 > 0.068 0.064 0.067 __ 0.082 0.083 0.069 0.0340.028 0.069 10.083 0.087 0.085 0.084 0.078 0.078 0.077 0.082 0.081 0.082 0.078 0.066 0.0320.020 0.064 0.079 ~K 0.086 0.086 0.082 0.082 0.082 0.084 0.082 0.075 0.063 0.026

0.007 0.053 0.069 0.078 0.081 i% 0.081 0.081 0.080 0.078 0.075 0.066 0.053 0.0150.039 0.036 0.055 0.065 0.069 0.071 0.070 0.071 0.070 0.069 0.066 0.063 0.053 0.037 0.046

0.005 0.047 0.017 0.029 0.035 0.037 0.038 0.039 0.037 0.035 0.033 0.026 0.016 0.047 0.011

Figure E-5 (c)

VIPRE-Ol Pin Power DistributionAdjustment Process

Revised Pin Power Distribution

1.714 1.676 1.708 1.699 1.694 1.694 1.693 1.693 1.693 1.694 1.696 1.701 1.709 1.675 1.708_1.679 1.684 1.667 1.658 1.657 1.655 1.656 1.655 1.656 1.656 1.659 1.661 1.669 1.684 1.672

1.708 1.664 1.650 1.643 1.641 ' 1.643 1.642 1.643 ,. 1.644 1.647 1.654 1.666 1.7021.695 1.652 1.638 .' 1.634 1.635 1.638 1.638 1.639 1.637 1.638 1.644 1.655 1.6911 .687 1.ý646 1 .633 1.630 1 .633 1 .634 1 .641 1 .640 1.642 1 .637 1.637 1.636 1 .640 1 .650 1.684

1.683 1.641 1.627 1.631 .1.649 1.653 1.650 ~. 1.635 1.633 1.647 1.6811.682 1 .641 1 1.630 1.629 1 1.635 1 .646 1 .676 1 .655 11.677 1 .649 1 .639 1.635 1 .636 1 .646 1.679

1.682 1 .640 1.628 1.628 1 .634 1 .650 1 .654 .1 .655 1 .652 1 .638 1.634 1 .635 1.645 1.6791.681 1.641 1.629 1.628 1.634 1.645 1.675 1.654 1.675 1.647 1.638 1.634, 1.635 1.646 1.679

1.686 1.645 1.631 1.6 27 1.629 1.630 1.636 1.636 1.637 1.632 1.633 1.632 1.636 1.648 1.6821.694 1.650 11.635 . .1.628 1.628 1.632 1.632 1.632 1.630- 1.632 1~ 1.639 1.651 1.688

1.707 1.661 1.645 1.636 1.633 Y2>1.633 1.633 16634 ~ .1.636 1.639 1.648 1.661 1.699

1.675 1.678 1.659 1.649 1.645 1.643 1.644 11.643 1.644 1.645 1.648 1.651 1.661 1.677 1.668

1.709 1.667 1.697 1.685 1.679 1.677 1.676 11.675 1.677 1.679 1.681 1.688 1.698 1.667 1.703

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References

E-1 UFSAR Chapter 15 Transient Analysis Methodology, DPC-NE-3005-PA, Revision 3,(approval date)

E-2 Nuclear Design Methodology Using CASMO-3/SIMULATE-3P, DPC-NIE-1004-A,Revision 1, December 1997

E-3 SIMULATE-3 Kinetics Theory and Model Description, SOA-96-26, Studsvik ofAmerica, April 1996

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This section addresses any UFSAR Chapter 15 non-LOCA transient and accident methodologyissues that are related to the mixed core configuration with AREVA NP Mark-B 11 and Mark-B-HTP fuel assembly designs co-resident in reload core designs.

The methodology for addressing mixed core effects is presented in this section. The concern isthe possibility that the presence of both the Mark-B 11 fuel and the Mark-B-HTP fuel in thereactor will introduce the need for special modeling. The main issue is the cross flow that willoccur between the two fuel assembly types due to design differences.

With regard to the modeling of the fuel assemblies in the Oconee RETRAN-3D model, there isno need to model mixed core effects., This conclusion is based on the similarity of the currentMark-B 11 fuel design and the future Mark-B-HTP fuel design, the fact that the reactor will beoperated at the same thermal-hydraulic conditions regardless of the fuel type, and the coarsemodeling (three axial nodes represent the entire core) of the fuel assemblies in RETRAN-3D.The detailed modeling of the fuel will be done using the VIPRE-01 models using input (coreaverage power or heat flux, core inlet flow, core exit pressure, core inlet temperature) fromRETRAN-3D. The RETRAN-3D input to the detailed VIPRE-01 modeling of Mark-B-HTP fuelwill use either existing RETRAN-3D analysis results with a full core of Mark-B 11 fuel modeled,or new analyses with a full core of Mark-B-HTP fuel modeled.

With regard to the modeling of the mixed core effects in VIPRE-01 during UFSAR Chapter 15transients and accidents, two approaches will be used. The first approach is to use the mixed corepenalty developed as described in Section 6.4. This penalty can then be applied as a peakingpenalty or a DNBR penalty. An additional approach using more detailed modeling of mixedcores has been developed. For each of the Oconee VIPRE-01 models described in Section 2.3, inAppendices D and E, and in DPC-NE-3005-PA, the mixed core effect can be explicitly modeledby including the number and location of each fuel assembly type in each VIPRE-01 model. Forexample, in a VIPRE-01 model that has different fuel assemblies modeled, the core loadingpattern for a mixed core can be used to specifically model the spatial relationship of each fuelassembly type. Also, in lumped channels that combine more than one fuel assembly type, theexact number of each fuel assembly type can be combined when calculating the input for eachlumped channel. Figure 7.5-1 shows the [ ] channel Oconee VIPRE-01 model (same asReference 7-1, Figure 2.3-5). In this model Channels [ J represent the hot assembly,Channels [ ] are individual fuel assemblies, and Channel [ ] represents theremaining fuel assemblies. To explicitly model a mixed core configuration, the core loadingpattern for the mixed core is used to map each fuel assembly type into these [ ] channels.Then, the input for each lumped channel and each rod is calculated based on the dimensions anddesign of the specific fuel assembly type for the percentage of each type that is assigned to thatlumped channel. [

D] The same explicit modeling of each mixed core configuration can be implemented

in each Oconee VIPRE-01 model. These more detailed approaches for addressing mixed coreissues will continue to provide conservative DNBR results.

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Figure 7.5 -1

Oconee [ ] Channel VIPRE-01 Model

D

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7.6 References

7-1 Thermal-Hydraulic Transient Analysis Methodology, DPC-NE-3000-PA, Revision 3,September 2004



7-4 A Correlation of Rod Bundle Critical Heat Flux for Water in the Pressure Range 150 to725 psia, IN-1412, Idaho Nuclear Corporation, July 1970

7-5 SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, EMF-2310(P)(A), Revision 1, AREVA NP, May 2004

7-6 Letter, S. A. Varga (NRC) to J. A. Jones (CP&L), Amendment No. 71 to FOL for H. B.Robinson, July 23, 1982

7-7 Letter, F. Akstulewicz (NRC), to J. F. Mallay (Siemens), Subject: Acceptance forReferencing of Topical Report EMF-84-093(P), Revision 1, February 16, 1999

7-8 UFSAR Chapter 15 Transient Analysis Methodology, DPC-NE-3005-PA, Revision 2,May 2005

7-9 SIMULATE-3: Advanced Three Dimensional Two-Group Reactor Analysis Code,STUDSVIK/SOA-92/01, Studsvik of America, April 1992

7-10 SIMULATE-3 Kinetics Theory and Model Description, SOA-96/26, Studsvik ofAmerica, April 1996

7-11 TACO3 - Fuel Pin Thermal Analysis Computer Code, BAW-10162P-A, B & W Fuel

Company, October 1989

7-12 BWC Correlation of Critical Heat Flux, BAW-10143P-A, April 1985

7-13 Thermal-Hydraulic Statistical Core Design Methodology, DPC-NE-2005P-A, Revision 3,September 2002

7-14 VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores, EPRI NP-251 1-CCM,Revision 4, EPRI, February 2001

7-15 Donald L. Hagrmann, et al., MATPRO-Version 11 (Revision 2): A Handbook ofMaterials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior,NUREG/CR-0479, August 1981

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8.0 AREVA NP LOCA ANALYSIS METHODOLOGY

8.1 Summary of Methodology

AREVA NP will be performing the LOCA analyses using the NRC-approved LOCA EvaluationModel that is documented in UFSAR Section 15.14. The AREVA NP LOCA Evaluation Modeltopical report is BAW-10192P-A, "BWNT LOCA - BWNT Loss-of-Coolant AccidentEvaluation Model for Once-Through Steam Generator Plants" (Reference 8-1). BAW-10164P-A,"RELAP5/MOD2-B&W - An Advanced Computer Program for Light Water Reactor LOCA andNon-LOCA Transient Analysis" (Reference 8-2), is the AREVA NP topical report that describesthe RELAP5 code that is the primary Evaluation Model code for LOCA simulation. BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR ReactorFuel" (Reference 8-3) is the AREVA NP topical report that details the modeling of M5® claddingin the LOCA analyses.

For LOCA analysis of the Mark-B-HTP fuel design AREVA NP has obtained NRC approval(Reference 8-4) of Revision 6 to BAW-10164. Revision 6 added the BHTP critical heat fluxcorrelation, the correlation that is applicable to the Mark-B-HTP fuel design, to the RELAP5code. This version of the RELAP5/MOD2-B&W topical report will be used for the Mark-B-HTPLOCA analysis for Oconee.

AREVA NP will also analyze or evaluate the effects of the mixed core configurations that willexist as the Mark-B -HTP fuel design replaces the existing Mark-B 11 fuel design.

The results of the AREVA NP Mark-B-HTP LOCA analyses for Oconee with the HTP fueldesign will.be evaluated against the 10 CFR 50.46 Final Acceptance Criteria.

8.2 References

8-1 BWNT LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Once-ThroughSteam Generator Plants, BAW-10192P-A, AREVA NP, June 1998

8-2 RELAP5/MOD2-B&W - An Advanced Computer Program for Light Water ReactorLOCA and Non-LOCA Transient Analysis, BAW-10164P-A, Revision 6, AREVA NP,June 2007

8-3 Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel,BAW-10227P-A, Revision 1, AREVA NP, June 2003

8-4 Letter, Nieh, H. K. (NRC) to R. Gardner (AREVA NP), Final Safety Evaluation forAREVA NP, Topical Report BAW-10164(P), Revision 6, June 25, 2007

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9.0 TECHNICAL SPECIFICATION REVISIONS

The transition to the AREVA NP Mark-B-HTP fuel design will result in the following proposedrevisions to the Oconee Technical Specifications and Technical Specification Bases. Thetechnical justification for each revision is included. Deleted text is designated by a strikethroughand new text is designated by bold font.

9.1 Revision to Technical Specification 2.1.1.2 - Reactor Core Safety Limits

Technical Specification 2.1.1.2, Reactor Core Safety Limits, will be revised to add the BHTPcritical heat flux correlation, the correlation that is applicable to the Mark-B-HTP fuel assemblydesign, and the correlation limit, as follows:

In-MODES I and 2, the departure from nucleate boiling ratio shall be maintained greaterthan the limit of 1.18 for the BWC correlation, a*4 1.19 for the BWU correlation, and1.132 for the BHTP correlation. Operation within this nlfpi4s these limits is ensured bycompliance with the Axial Power Imbalance Protective Limits and RCS Variable LowPressure Protective Limits as specified in the Core Operating Limits Report.

Technical Justification: The existing Technical Specification 2.1.1.2 includes the critical heatflux correlations and the correlation DNBR limits that are applicable to the AREVA NP fuelassembly designs that are currently in use at Oconee. With the transition to the AREVA NPMark-B-HTP fuel assembly design the BHTP critical heat flux correlation and the correlationDNBR limit (1.132) must be added to Technical Specification 2.1.1.2. It is noted that the BWU-N correlation is used below the first intermediate grid for the Mark-B-HTP design. It is alsonoted that the Modified-Barnett critical heat flux correlation is used for the Mark-B-HTP designfor one of the main steam line break analysis cases. The BWU-N and the Modified-Barnettcorrelations are not included in Technical Specification 2.1.1.2 because there is a precedent toinclude only the primary correlation and not to include secondary correlations with limitedapplication.

9.2 Revision to Technical Specification 5.6.5.b - Core Operating Limits Report (COLR)

Technical Specification 5.6.5.b, Core Operating Limits Report (COLR), will be revised to add theAREVA NP topical report BAW- 10164-PA as follows:

(11) BAW-10164-PA, RELAP5/MOD2-B&W - An Advanced Computer Programfor Light Water Reactor LOCA and non-LOCA Transient Analysis

Technical Justification: The AREVA NP topical report BAW-10164-PA, Revision 6,"RELAP5/MOD2-B&W - An Advanced Computer Program for Light Water Reactor LOCA andnon-LOCA Transient Analysis", was approved by NRC in the Safety Evaluation dated June 25,2007. This revision adds the BHTP critical heat flux correlation that is applicable to the Mark-B-HTP fuel design to the RELAP5 code. This revised version of RELAP5 will be used by AREVANP for the Oconee LOCA analyses with the Mark-B-HTP fuel design. The results of the LOCAanalyses will be used to set the linear power density limits for the Oconee reload designs.

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A

9.3 Revision to Technical Specification Bases B 2.1.1 - Reactor Core SLs

Bases Section B 2.1.1 is revised (pp. B 2.1.1-1, B 2.1.1-4) to include the BHTP critical heat fluxcorrelation that is applicable to the Mark-B-HTP fuel design, and to add the associated reference.

DNB is not a directly measurable parameter during operation, but neutron power andReactor Coolant System (RCS) temperature, flow and pressure can be related to DNBusing a critical heat flux (CHF) correlation. The BWC (Ref. 2), aii4 the BWU (Ref. 4),and the BHTP (Ref. 5) CHF correlations have been developed to predict DNB foraxially uniform and non-uniform heat flux distributions. The BWC correlation applies toMark-BZ fuel. The BWU correlation applies to the Mark-B 11 fuel. The BHTPcorrelation applies to Mark-B-HTP fuel. The local DNB heat flux ratio (DNBR),defined as the ratio of the heat flux that would cause DNB at a particular core location tothe actual local heat flux, is indicative of the margin to DNB. The minimum value of theDNBR, during steady-state operation, normal operational transients, and anticipatedtransients is limited to 1.18 (BWC), and 1.19 (BWU), and 1.132 (BHTP).

REFERENCES

5. BAW-10241(P)(A), Revision 1, BHTP DNB Correlation Applied with LYNXT,Framatome ANP, July 2005

Technical Justification: The existing Technical Specification Bases B 2.1.1 includes the criticalheat flux correlations and the correlation DNBR limits that are applicable to the AREVA NP fuelassembly designs that are currently in use at Oconee. With the transition to the AREVA NPMark-B-HTP fuel assembly design, the BHTP critical heat flux correlation and the correlationlimit (1.132) must be added to Technical Specification Bases B 2.1.1. It is noted that the BWU-Ncorrelation is used below the first intermediate grid for the Mark-B-HTP design. It is also notedthat the Modified-Barnett critical heat flux correlation is used for the Mark-B-HTP design for oneof the main steam line break analysis cases. The BWU-N and the Modified-Barnett correlationsare not included in Technical Specification 2.1.1.2 because there is a precedent to include onlythe primary correlation and not to include secondary correlations with limited application. Theassociated reference, BAW-10241(P)(A), Revision 1, BHTP DNB Correlation Applied withLYNXT, Framatome ANP, July 2005, is included as new Reference 5.

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9.4 Revision to Technical Specification Bases B 3.4.1 - RCS Pressure, Temperature, andFlow Departure from Nucleate Boiling (DNB) Limits

Bases Section B 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling(DNB) Limits, (APPLICABLE SAFETY ANALYSES) will be revised to add the BHTP criticalheat flux correlation, the correlation that is applicable to the Mark-B-HTP fuel assembly design,and the correlation limit, as follows:

The requirements of LCO 3.4.1 represent the initial conditions for DNB limited transientsanalyzed in the plant safety analyses (Ref. 1). The safety analyses have shown thattransients initiated from the limits of this LCO will meet the DNBR criterion of Ž 1.18 forBWC correlation, > 1.19 for BWU correlation, > 1.132 for BHTP correlation, or anequally valid limit when the statistical DNBR limit is employed (SCD methodology). Thisis the acceptance limit for the RCS DNBR parameters.

Technical Justification: The existing Technical Specification Bases B 3.4.1 includes the criticalheat flux correlations and the correlation DNBR limits that are applicable to the AREVA NP fuelassembly designs that are currently in use at Oconee. With the transition to the AREVA NPMark-B-HTP fuel assembly design, the BHTP critical heat flux correlation and the correlationlimit (1.132) must be added to Technical Specification Bases B 3.4.1. It is noted that the BWU-Ncorrelation is used below the first intermediate grid for the Mark-B-HTP design. It is also notedthat the Modified-Barnett critical heat flux correlation is used for the Mark-B-HTP design for oneof the main steam line break analysis cases. The BWU-N and the Modified-Barnett correlationsare not included in Technical Specification Bases B 3.4.1 because there is a precedent to includeonly the primary correlation and not to include secondary correlations with limited application.

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10.0 CORE OPERATING LIMITS REPORT REVISIONS

The Core Operating Limits Report will be revised to include any changes resulting from thetransition to the Mark-B-HTP fuel design. The following revisions have been identified to benecessary.

10.1 Reference 10 - BAW-10164P-A

COLR Reference 10, "BAW-10164P-A, Rev. 4 RELAP5/MOD2-B&W - An AdvancedComputer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis, SERdated April 9, 2002", will be revised to Revision 6 with the June 25, 2007 SER date. Thisrevision will be added to the Oconee COLR for each Oconee reload cycle that includes the Mark-B-HTP fuel design.

10.2 LOCA Limits

The COLR includes limits on linear power density (kW/ft) that are a result of the AREVA NPLOCA analysis. These limits are dependent on the fuel assembly design, the elevation in thecore, and the fuel rod bumup. New LOCA limits will be provided by AREVA NP for the Mark-B-HTP fuel design. These limits will be added to the Oconee COLR for each Oconee reloadcycle that includes the Mark-B-HTP fuel design.

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