ML080700175
| ML080700175 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/03/2008 |
| From: | Duke Energy Carolinas, Duke Power Co |
| To: | NRC/RGN-II |
| References | |
| DPC-NE-2015-P | |
| Download: ML080700175 (18) | |
Text
DPC-NE-20 15-P Oconee Nuclear Station Mark-B-HTP Fuel Transition Methodology NRC / Duke Energy Meeting One White Flint North March 3, 2008 Dhuke arEnergy.
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 methodology reports
- Present the more significant methodology revisions included in 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 experiencing cladding failures due to flow-induced. vibration. A study of alternatives to solve this problem has concluded that a transition to the Mark-B-HTP design is the best option.
- The Mark-B-HTP fuel design is becoming the standard fuel design for the Oconee 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 cycle starts up in December 2008.
- The October 22, 2007 submittal requested NRC approval by September 30, 2008 F DueNRC!
Duke Meeting -March 3, 20083
Proposed Licensing Approach
- Duke has previously licensed two major fuel transitions and one minor fuel transition. The Mark-B-HTP fuel is a minor transition.
- Revisions to seven NRC-approved methodology reports are required to support 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 one umbrella report (rather than submit seven separate revisions) 0 Duke is requesting that the NRC SE for DPC-NE-2015-P state that it also approves 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 011s Oconee 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 and SIMULATE-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 rods 0 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 PDuke WhEnergys NRC / Duke Meeting - March 3, 20087
Mark-B-HTP, Fuel Assembly Mark-B-HTP Fuel
.M510 HTP Grids (7x) with Curved Flow Channels 7.Alloy 7 18 Lower HMP Grid (Straight Flow Channel)
- Removable Upper End Fitting
" Alloy 718 Cruciform
'Springs
" Crimped Top-Hat Nut
- M5ý' Fuel Rods
" M50 Guide Tubes
" M5,, Instrument Tube
" FUEL GUARD T" Lower End Fitting PDuke OrEnergy 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 be maintained greater than the limit of 1.18 for the BWC correlation, 1.19 for the BWU correlation, ainid 1.132 ifbir the BHTFI cire~zloni.
Operation within these limits is ens ur.ed by compliance with the Axial Power Imbalance Protective Limits and RCS Variable Low Pressure Protective 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~ Limits Report (COLR)
(1I1) IBAW=Ft1U64=PA, IR3LAi5/MGD2-B&W - Ain Advanced Computeir Pinogram 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 Pages
1.0 INTRODUCTION
1-3 2.0 MARK-B-HTP FUEL DESIGN
SUMMARY
4-8 2.1 Design Description 2.2 Operating Experience 2.3 References 3.0 METHODOLOGY REPORTS OVERVIEW 9-12 3.1 Core Physics and Reload Design Methodology Reports 3.2 Fuel Assembly Mechanical Design Methodology Report 3.3 Core Thermal-Hydraulic Design Methodology Reports 3.4 Transient and Accident Analysis Methodology Reports 3.5 Inter-Relationships of Methodology Reports 3.6 References 4.0 CORE PHYSICS AND RELOAD DESIGN METHODOLOGY REVISIONS 13-19 4.1 Revision 6 to NFS-lO1001A - Oconee Nuclear Station Reload Design Methodology 4.2 Revision 3 to DPC-NE-1I002-A - Oconee Nuclear Station Reload Design Methodology II 4.3 Mixed'Core Effects 4.4 References 5.0 FUEL ASSEMBLY MECHANICAL DESIGN METHODOLOGY REVISIONS 20-22 5.1 Revision.1 to DPC-NE-2008P-A - Fuel Mechanical Reload Analysis Methodology Using TAC03 5.2 References
6.0 CORE THERMAL-HYDRAULIC DESIGN METHODOLOGY REVISIONS 2-4 23-47 6.1 Revision 2 to DPC-NE-2003P-A - Oconee Nuclear Station Core Thermal-Hydraulic Methodology Using VIPRE-0 1 6.2 Revision 4 to DPC-NE-2005P-A - Thermal-Hydraulic Statistical Core Design Methodology 6.3 Appendix F. to DPC-NE-2005P-A - Application of BHTP CHF Correlation 6.4 Mixed Core Effects 6.5 References 7.0 UFSAR CHAPTER 15 NON-LOCA TRANSIENT AND ACCIDENT 48-78 ANALYSIS METHODOLOGY REVISIONS 7.1 Revision 4 to DPC-NE-3000-PA - Thermal-Hydraulic Transient Analysis Methodology 7.2 Revision 3 to DPC-NE-3005-PA - UFSAR Chapter 15 Transient Analysis Methodology 7.3 Appendix D to DPC-NE-3000-PA - Methodology Revisions for Mark-B-HTP Fuel 7.4 Appendix E to DPC-NE-3000-PA - Expanded Oconee VIPRE-01 Methodology 7.5 Mixed Core Effects 7.6, References 8.0 AREVA NP LOCA ANALYSIS METHODOLOGY 79 8.1 Summary of Methodology 8.2 References 9.0 TECHNICAL SPECIFICATION REVISIONS 80-82
- 9. 1, Revision to Technical Specification 2.1.1.2 - Reactor Core Safety Limits 9.2 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 SLs 9.4 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 83 10.1 Reference 10-BAW-10164P-A 10.2 LOCA Limits
Methodology Report Revision Characterization
- 1) Revisions associated with the Mark-B-HTP fuel design and associated critical heat flux correlations
- 2) Mixed core modeling techniques
- 3) Revisions to improve analytical margins / enhancements
- 4)
Error corrections
- 5) Deletion of superseded or historical content
- 6) Clarifications and editorial changes Obfnergya NRC! Duke Meeting -March 3, 2008 13
Revisions to NFS-1001A and DPC-NE-1002-A Core Physics and Reload Design Methods
- 1) The fuel densification power spike factor is revised from a value of 1.08 to an axially-dependent value provided by ARIEVA-NP
- 2) A pin power peaking penalty due to the effect of fuel assembly bow is included
- 3) A mixed core penalty (FAh and FQ) is applied to account for the difference in the bottom elevation of the fuel pellet stack DukEery NRC /Duke Meeting - Mairch 3, 2008 14
Revisions to DPC-NE-2008P-A Fuel Mechanical Analysis Editorial revisions to update the cladding creep collapse, cladding corrosion, cladding stress, and cladding fatigue analysis methodologies and references.
Ob~nrgy@NRC IDuke Meeting - March 3, 2008 15
Revisions to DPC-NE-2003P-A and DPC-NE-2005P-A Core Thermal-Hydraulic Analysis
- 1. Add the ARE VA NP BHTP critical heat flux correlation that is applicable to the Mark-B-HTP design. Note that the BWU-N correlation is used below the first intermediate grid
- 2. The pin power distribution used in the VIPRE-01 thermal-hydraulic models is revised
- 3. New Appendix F to DPC-NE-2005P that details the application of Duke's Statistical Core Design methodology to the Mark-B-HTP fuel.
- 4. Description of mixed core modeling approach and the justification that the statistical design limit remains the same for Mark-B-IITP fuel in mixed core configurations with Mark-B 11 fuel.
FW~egaNRC!
Duke Meeting -March 31,2008 16
Revisions to DPC-NE-3000-PA RETRAN-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 the critical 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 steam line break accident. The 1.135 DNBR limit established by ARIEVA NP is used.
- 2. New Appendix E describing an expanded VIPRIE-01 model to enable modeling the inter-assembly' gap and mixed cores
- 3. Also in Appendix E, a description of the use of SIMULATE-3 pin power distributions in VIPRE-01 models rather than historical vendor pin power distributions DueNRC IDuke Meeting - March 3, 2008 17
Revisions to -DPC-NE-3005-PA UFSAR Chapter 15 Transient and Accident Analyses
- 1. Revisions to the VIPRE-Ol, code to allow use of the BHTP correlation,. and to calculate fuel pellet average -enthalpy
- 2. Use of more detailed VIPRIE-01 models for predicting DNBR
- 3. Add the three critical heat flux correlations previously mentioned
- 4. For the rod ejection analysis, delete the adjustment of the initial radial power distribution since this was determined to not necessarily be conservative
- 5. Mixed core modeling using the VIPRIE-01 code Duke NC/Dk etn aci,081 k7Energy R!Dkoetn-ac
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