DPC-NE-20 15-POconee Nuclear Station

Mark-B-HTP Fuel Transition Methodology

NRC / Duke Energy MeetingOne White Flint North

March 3, 2008

DhukearEnergy.

Objectives of Meeting

*Explain why Duke is transitioning to the ARE VA NP Mark-B-HTP design at.Oconee

*Explain the proposed licensing approach

*Provide an overview of Duke's NRC-approved methodologyreports

*Present the more significant methodology revisions includedin the DPC-NE-20 15-P methodology report

*Respond to the Staff's questions

Dfuke

Why Transition to Mark-B-HTP?

* The Mark-Ri 1 fuel design currently in use at Oconee is experiencingcladding failures due to flow-induced. vibration. A study of alternatives tosolve this problem has concluded that a transition to the Mark-B-HTPdesign is the best option.

* The Mark-B-HTP fuel design is becoming the standard fuel design for theOconee class plants. It is currently in use at ANO-1, CR-3, and DB.

" Oconee 2 Cycle 24 will be the first reload of Mark-B-HTP fuel. This cyclestarts up in December 2008.

* The October 22, 2007 submittal requested NRC approval by September30, 2008

F DueNRC! Duke Meeting -March 3, 20083

Proposed Licensing Approach

* Duke has previously licensed two major fuel transitions and one minorfuel transition. The Mark-B-HTP fuel is a minor transition.

* Revisions to seven NRC-approved methodology reports are required tosupport the transition to the Mark-B-HTP fuel design

* The DPC-NE-2015-P "Mark-B-HTP Fuel Transition Methodology"'report consolidates all of the revisions to these seven reports under oneumbrella report (rather than submit seven separate revisions)

0 Duke is requesting that the NRC SE for DPC-NE-2015-P state that it alsoapproves the included revisions for each of the seven reports

* Duke will then publish approved versions of all eight of these reports* This approach was presented at the NRC/Duke meeting on 12/21/2005 as

Duke's. intended licensing approach for fuel transitions* There is a precedent for one SE to approve revisions to two methodology

reports (9/24/2003 letter, L. N. Olshan (NRC) to R. A. Jones (Duke))

DueNRC / Duke Meeting - March 3, 20084

The Seven NRC-Approved Duke

Methodology Reports

That Are Revised in DPC-NE-20 15-P

Number Title Revision Date

Cote P~hy~sis and' Reload Desg __________

NFS-1001A Oconee Nuclear Station Reload 5 Jn0____________jDesign Methodology ____ _____

DPC-NE- 1 002-A Oconee Nuclear Station Reload 2 Oct-85_____________ jDesign Methodology 11____________

FuelAssemibly Mecthanical Dsg __________

DPCNE2008PA Fuel Mechanical Reload Analysis 0Ar9____________jA Methodology Using TAC03 0___ _______

Core;.T.ermal-Hydra'lic 011sOconee Nuclear Station Core

DPC-NE-2003 P-A Thermal-Hydraulic Methodology 1 Sep-00_____________Using VIPRE-01

DPC-NE-2005P-A Thermal-Hydraulic Statistical Core 3.Sep-02_____________Design Methodology ____ _____

Transien tan'd'Accidenf Analysis< -______

DPCNE300PA IThermal-Hydraulic Transient 3Sp0______________________ Analysis Methodology 3_______ ____________

DPC-NE-3005-PA UFSAR Chapter 15 Transient 2 May-05____________________ Analysis Methodology__________

Scope of Duke Analytical Methodolog'ies

" Fuel assembly mechanical analysis using ARIEVA NP codes (TACO3,CROV, etc)'

" Core physics and reload design analyses, using Studsvik CASMIO-3 andSIMULATE-3 codes

" Core thermal-hydraulic (DNBR) analyses using VIPRIE-01" Non-LOCA Chapter 15 system transient analyses using RIETRAN-3D* Three-dimensional rod ejection analyses using SIMULATE-3K

>The LOCA ~anays n~ is jp~ iod~ed by AIREVA NIPan doe not ýnvove an~y

ME` DukeNRC IDuke Meeting - March 3, 2008 6

Mark-B-HTP Fuel Design Summary

* 15 x 15 lattice of 0.430 diameter fuel rods0 M5® cladding, guide tubes, and instrument tube* HTP non-mixing vane grids* Higher pressure drop than Mark-Bit (will reduce flow)0 Bottom of fuel pellet stack is higher than Mark-Bi11* Currently in service at ANO-1, CR-3, and DB

PDukeWhEnergys NRC / Duke Meeting - March 3, 20087

Mark-B-HTP, Fuel Assembly

Mark-B-HTP Fuel

.M510 HTP Grids (7x) withCurved Flow Channels

7.Alloy 7 18 Lower HMPGrid (Straight FlowChannel)

*Removable Upper EndFitting

" Alloy 718 Cruciform'Springs

" Crimped Top-HatNut

*M5ý' Fuel Rods" M50 Guide Tubes" M5,, Instrument Tube

" FUEL GUARD T" LowerEnd Fitting

PDukeOrEnergy® NRC / Duke Meeting - March 3, 20088 8

Technical Specification Revisions

Revision. to Technical Specification 2.1.1.2 - Reactor Core Safety Limits

In MODES 1 and 2, the departure from- nucleate boiling ratio shall bemaintained greater than the limit of 1.18 for the BWC correlation, 1.19for the BWU correlation, ainid 1.132 ifbir the BHTFI cire~zloni.Operation within these limits is ens ur .ed by compliance with the AxialPower Imbalance Protective Limits and RCS Variable Low PressureProtective Limits as specified in the Core Operating Limits Report.

Energy-iNRC IDuke Meeting -March 3, 2008 9

Technical Specification Revisions (cont.)

Revision to Technical Specification- 5.6.5.b - Core Operating~ LimitsReport (COLR)

(1I1) IBAW=Ft1U64=PA, IR3LAi5/MGD2-B&W - Ain Advanced ComputeirPinogram for ULghit Waei R1eactor LOCA and rnon-LODCA Tfiannenit

Note: T. S. base's revisions are included for information

Duk~ery NRC /Duke Meeting - March 3, 2008 10

DPC-NE-2015-P

MARK-B-HTP FUEL TRANSITION METHODOLOGY

TABLE OF CONTENTS

Pages1.0 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-lO1001A - Oconee Nuclear Station Reload Design Methodology4.2 Revision 3 to DPC-NE-1I002-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 .1 to DPC-NE-2008P-A - Fuel Mechanical Reload Analysis MethodologyUsing TAC03

5.2 References

6.0 CORE THERMAL-HYDRAULIC DESIGN METHODOLOGY REVISIONS 2-423-47

6.1 Revision 2 to DPC-NE-2003P-A - Oconee Nuclear Station Core Thermal-HydraulicMethodology Using VIPRE-0 1

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 Tec~hnical 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

Methodology Report Revision Characterization

1) Revisions associated with the Mark-B-HTP fuel design andassociated critical heat flux correlations

2) Mixed core modeling techniques3) Revisions to improve analytical margins / enhancements4) Error corrections

5) Deletion of superseded or historical content6) Clarifications and editorial changes

Obfnergya NRC! Duke Meeting -March 3, 2008 13

Revisions to NFS-1001A and DPC-NE-1002-ACore Physics and Reload Design Methods

1) The fuel densification power spike factor is revised from a value of1.08 to an axially-dependent value provided by ARIEVA-NP

2) A pin power peaking penalty due to the effect of fuel assembly bow isincluded

3) A mixed core penalty (FAh and FQ) is applied to account for thedifference in the bottom elevation of the fuel pellet stack

DukEery NRC /Duke Meeting - Mairch 3, 2008 14

Revisions to DPC-NE-2008P-AFuel Mechanical Analysis

Editorial revisions to update the cladding creep collapse, claddingcorrosion, cladding stress, and cladding fatigue analysismethodologies and references.

Ob~nrgy@NRC IDuke Meeting - March 3, 2008 15

Revisions to DPC-NE-2003P-A and DPC-NE-2005P-ACore Thermal-Hydraulic Analysis

1. Add the ARE VA NP BHTP critical heat flux correlation that isapplicable to the Mark-B-HTP design. Note that the BWU-Ncorrelation is used below the first intermediate grid

2. The pin power distribution used in the VIPRE-01 thermal-hydraulicmodels is revised

3. New Appendix F to DPC-NE-2005P that details the application ofDuke's Statistical Core Design methodology to the Mark-B-HTP fuel.

4. Description of mixed core modeling approach and the justificationthat the statistical design limit remains the same for Mark-B-IITPfuel in mixed core configurations with Mark-B 11 fuel.

FW~egaNRC! Duke Meeting -March 31,2008 16

Revisions to DPC-NE-3000-PARETRAN-3D and VIPRE-OlI Models Used. for UFSAR

Chapter 15 Transient and Accident Analyses

1. New Appendix D describing the Mark-B-HTP fuel design and thecritical heat flux correlations that are used

-The BHTP correlation

-The BWIJ-N correlation below the first intermediate grid-The.Modified-Barnett correlation for low pressure main steamline break accident. The 1.135 DNBR limit established byARIEVA NP is used.

2. New Appendix E describing an expanded VIPRIE-01 model to enablemodeling the inter-assembly' gap and mixed cores

3. Also in Appendix E, a description of the use of SIMULATE-3 pinpower distributions in VIPRE-01 models rather than historicalvendor pin power distributions

DueNRC IDuke Meeting - March 3, 2008 17

Revisions to -DPC-NE-3005-PAUFSAR Chapter 15 Transient and Accident Analyses

1. Revisions to the VIPRE-Ol, code to allow use of the BHTPcorrelation,. and to calculate fuel pellet average -enthalpy

2. Use of more detailed VIPRIE-01 models for predicting DNBR3. Add the three critical heat flux correlations previously mentioned4. For the rod ejection analysis, delete the adjustment of the initial

radial power distribution since this was determined to not necessarilybe conservative

5. Mixed core modeling using the VIPRIE-01 code

Duke NC/Dk etn aci,081k7Energy® R!Dkoetn-ac ,081