ML13169A054
| ML13169A054 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/13/2013 |
| From: | Gatlin T South Carolina Electric & Gas Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML13169A054 (20) | |
Text
Thomas D. Gatlin Vice President,Nuclear Operations 803.345.4342 A SCANA COMPANY June 13, 2013 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir / Madam:
Subject:
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ECCS EVALUATION MODEL REVISIONS REPORT South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for South Carolina Public Service Authority, hereby submits the 2012 Emergency Core Cooling System (ECCS) Evaluation Model Revisions Annual Report for VCSNS. This report is being submitted pursuant to 10 CFR 50.46, which requires licensees to notify the NRC on at least an annual basis of corrections to or changes in the ECCS Evaluation Models.
Summary sheets describing changes and enhancements to the ECCS Evaluation Models for 2012 are included in Attachment I. Peak Clad Temperature (PCT) sheets are included in Attachment II.
If you have any questions, please call Bruce L. Thompson at (803) 931-5042.
Very truly yours, Thomas D. Gatlin TS/TDG/wm Attachments c: K, B. Marsh E. A. Brown S, A. Byrne NRC Resident Inspector J. B. Archie K. M. Sutton N. S. Carns NSRC J. H. Hamilton RTS (LTD 321, RR 8375)
J. W. Williams File (818.02-17)
W. M. Cherry PRSF (RC-13-0080)
V, M. McCree 7 4ooL Virgil C.Summer Station -Post Office Box 88 *Jenkinsville, SC .29065 . F(803) 345-5209
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 1 of 12 Attachment I Changes and Enhancements to the ECCS Evaluation Models for 2012
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 2 of 12 GENERAL CODE MAINTENANCE
Background
Various changes have been made to enhance the usability of codes and to streamline future analyses. Examples of these changes include modifying input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-1 3451.
Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The nature of these changes leads to an estimated Peak Cladding Temperature (PCT) impact of 0 degrees Fahrenheit.
Document Control Desk Attachment I LTD 321 RC-1 3-0080 Page 3 of 12 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND PEAKING FACTOR BURNDOWN
Background
Fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown were not explicitly considered in the VCSNS Best Estimate Large Break Loss-of-Coolant Accident (BE LBLOCA) Analysis of Record (AOR). Nuclear Regulatory Commission (NRC) Information Notice 2011-21 (Reference 1) notified addressees of recent information obtained concerning the impact of irradiation on fuel thermal conductivity and its potential to cause significantly higher predicted PCT results in realistic ECCS evaluation models. This evaluation provides an estimated effect of fuel pellet TCD and peaking factor burndown on the PCT calculation for the VCSNS BE LBLOCA AOR. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-1 3451 (Reference 2).
Affected Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect A quantitative evaluation for 10 CFR 50.46 reporting purposes, as discussed in Reference 3, was performed to assess the PCT effect of fuel pellet TCD and peaking factor burndown on the VCSNS BE LBLOCA analysis and concluded that the estimated PCT impact is 0 degrees Fahrenheit for Blowdown, 113 degrees Fahrenheit for Reflood 1 and 123 degrees Fahrenheit for Reflood 2. The peaking factor burndown included in the evaluation is provided in Table 1 and is conservative for the current cycle. SCE&G and its vendor, Westinghouse Electric Company LLC, utilize processes which ensure that the Loss-of-Coolant Accident (LOCA) analysis input values conservatively bound the as-operated plant values for those parameters and will be validated as part of the reload design process.
Table 1: Peaking Factors Assumed in the Evaluation of TCD Rod Burnup FdH (1),(2) FQ Transient (1) FQ Steady-State (MWd/MTU)________
0 1.70 2.50 2.00 30,000 1.70 2.50 2.00 30,000 1.62 2.45 2.00 60,000 1.40 1.96 1.60 62,000 1.40 1.96 1.60 (1) Includes uncertainties.
(2) Hot assembly average power follows the same burndown, since it is a function of FdH.
Document Control Desk Attachment I LTD 321 RC-1 3-0080 Page 4 of 12 References
- 1. NRC Information Notice 2011-21, McGinty, T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation," December 13, 2011. (NRC ADAMS # ML113430785)
- 2. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"
October 1992.
- 3. OG-12-386, "For Information Only - Input Supporting the PWROG LBLOCA Program Regarding Nuclear Fuel Thermal Conductivity Degradation (PA-ASC-1073, Revision 0)
(Proprietary/Non-Proprietary)," September 18, 2012.
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 5 of 12 PAD 4.0 IMPLEMENTATION
Background
The BE LBLOCA AOR for VCSNS utilized fuel rod design inputs from PAD Version 3.4. To isolate the effect of fuel rod design input from PAD code version differences, the impact of using fuel rod design input from PAD Version 4.0 was estimated prior to explicitly considering fuel rod design input which includes fuel pellet TCD and peaking factor burndown and is based on the PAD Version 4.0 code. The implementation of PAD Version 4.0 into the 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model was described in Reference 1 as a forward-fit, Discretionary Change in accordance with Section 4.1.1 of WCAP-1 3451 (Reference 2). The plant-specific implementation of PAD Version 4.0 into the BE LBLOCA AOR for VCSNS is considered a design input change into the BE LBLOCA analysis.
SCE&G and its vendor, Westinghouse Electric Company LLC, utilize processes which ensure that LOCA analysis input values conservatively bound the as-operated plant values for those parameters.
Affected Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect A qualitative evaluation, as discussed in Reference 3, was performed to estimate a PCT effect resulting from a change in fuel rod design input parameters from PAD Version 3.4 and PAD Version 4.0. The evaluation concluded that the estimated PCT impact is negative 83 degrees Fahrenheit for Blowdown, negative 118 degrees Fahrenheit for Reflood 1, and negative 118 degrees Fahrenheit for Reflood 2 for 10 CFR 50.46 reporting purposes.
References
- 1. LTR-NRC-01-6, Letter from H. A. Sepp (Westinghouse) to J. S. Wermiel (NRC), "U. S.
Nuclear Regulatory Commission, 10 CFR 50.46 Annual Notification and Reporting for 2000,"
March 13, 2001.
- 2. WCAP-1 3451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"
October 1992.
- 3. OG-12-386, "For Information Only - Input Supporting the PWROG LBLOCA Program Regarding Nuclear Fuel Thermal Conductivity Degradation (PA-ASC-1073, Revision 0)
(Proprietary/Non-Proprietary)," September 18, 2012.
Document Control Desk Attachment I LTD 321 RC-1 3-0080 Page 6 of 12 PCT ASSESSMENT OF TRANSVERSE MOMENTUM CELLS FOR ZERO CROSS-FLOW BOUNDARY CONDITION ERROR IN THE V. C. SUMMER (CGE) BEST-ESTIMATE LARGE BREAK LOCA ANALYSIS
Background
An error was identified in the input of the transverse momentum cells for zero cross-flow boundary condition. Based on the nature of the input error, it was determined that no WCOBRA/TRAC calculations are necessary. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-1 3451 (Reference 1).
Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect This error was determined to have no impact on the calculated results; thus, the estimated PCT impact is 0 degrees Fahrenheit for all time periods for 10 CFR 50.46 reporting purposes.
Reference
- 1. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"
October 1992.
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 7 of 12 HOTSPOT BURST TEMPERATURE CALCULATION FOR ZIRLO CLADDING
Background
A problem was identified in the calculation of the burst temperature for ZIRLO(1 ) cladding in the HOTSPOT code when the cladding engineering hoop stress exceeds 15,622 psi. This problem results in either program failure or an invalid extrapolation of the burst temperature versus engineering hoop stress table. This problem has been evaluated for impact on existing analyses, and its resolution represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The evaluation of existing analyses demonstrated no impact on the overall PCT results, leading to an estimated effect of 0 degrees Fahrenheit.
(1)ZIRLO is a registeredtrademarkof Westinghouse Electric Companv LLC, its affiliates and/or its subsidiaries in the United States ofAmerica and may be registeredin other countries. All rights reserved. Unauthorizeduse is strictly prohibited.
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 8 of 12 HOTSPOT ITERATION ALGORITHM FOR CALCULATING THE INITIAL FUEL PELLET AVERAGE TEMPERATURE
Background
The HOTSPOT code has been updated to incorporate the following corrections to the iteration algorithm for calculating the initial fuel pellet average temperature: (1) bypass the iteration when the input value satisfies the acceptance criterion; (2) prevent low-end extrapolation of the gap heat transfer coefficient; (3) prevent premature termination of the iteration that occurred under certain conditions; and (4) prevent further adjustment of the gap heat transfer coefficient after reaching the iteration limit. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.
Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect Sample calculations and engineering judgment lead to an estimated PCT impact of 0 degrees Fahrenheit.
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 9 of 12 WCOBRA/TRAC AUTOMATED RESTART PROCESS LOGIC ERROR
Background
A minor error was identified in the WCOBRA/TRAC Automated Restart Process (WARP) logic for defining the Double-Ended Guillotine (DEG) break tables. The error has been evaluated for impact on current licensing-basis analysis results and will be incorporated into the plant-specific analyses on a forward-fit basis. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.
Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect These errors were evaluated to have a negligible impact on the LBLOCA analysis results, leading to an estimated PCT impact of 0 degrees Fahrenheit.
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 10 of 12 ROD INTERNAL PRESSURE CALCULATION
Background
Several issues which affect the calculation of rod internal pressure (RIP) have been identified for certain BE LBLOCA evaluation models (EMs). These issues include the sampling of rod internal pressure uncertainties, updating HOTSPOT to consider the effect of transient RIP variations in the application of the uncertainty, and generating RIPs at a consistent rod power.
These issues have been evaluated to estimate the impact on existing LBLOCA analysis results.
The resolution of these issues represents a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.
Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The effects described above are either judged to have a negligible effect on existing LBLOCA analysis results or have been adequately incorporated into the thermal conductivity degradation evaluations, leading to an estimated PCT impact of 0 degrees Fahrenheit.
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 11 of 12 WCOBRA/TRAC THERMAL-HYDRAULIC HISTORY FILE DIMENSION USED IN HSDRIVER
Background
A problem was identified in the dimension of the WCOBRAITRAC thermal-hydraulic history file used in HSDRIVER. The array that is used to store the information from the WCOBRA/TRAC thermal-hydraulic history file is dimensioned to 3000 in HSDRIVER. It is possible for this file to contain more than 3000 curves. If that is the case, it is possible that the curves would not be used correctly in the downstream HOTSPOT execution. An extent-of-condition review indicated that resolution of this issue does not impact the PCT calculation for prior LBLOCA analyses.
This represents a Discretionary Change in accordance with Section 4.1.1 of WCAP-1 3451.
Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model for Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect As discussed in the Background section, resolution of this issue does not impact the PCT calculation for prior LBLOCA analyses, which leads to a PCT impact of 0 degrees Fahrenheit.
Document Control Desk Attachment I LTD 321 RC-13-0080 Page 12 of 12 NOTRUMP-EM EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION
Background
An evaluation has been completed to estimate the effect of fuel pellet TCD on PCT for plants in the United States with analyses using the 1985 Westinghouse Small Break Loss-of-Coolant Accident (SBLOCA) Evaluation Model with NOTRUMP (NOTRUMP-EM). This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.
Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect Based on the phenomena and physics of the SBLOCA transient, in combination with limited sensitivity calculations, it is concluded that TCD has a negligible effect on the limiting cladding temperature transient, leading to an estimated PCT impact of 0 degrees Fahrenheit.
Document Control Desk Attachment II LTD 321 RC-1 2-0080 Page 1 of 7 Attachment II Peak Clad Temperature (PCT) Rackup Sheets
Document Control Desk Attachment II LTD 321 RC-12-0080 Page 2 of 7 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: V. C. Summer Utility Name: South Carolina Electric & Gas Revision Date: 3/1/2013 Composite Analysis Information EM: CQD (1996) Analysis Date: 2/3/2003 Limiting Break Siize: Guillotine FQ: 2.5 FdH: 1.7 Fuel: Vantage + SGTP (%): 10 Notes: Delta 75 Replacement Steam Generator Uprate Core Power 2900 MWt Clad Temp (*F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1988 1 PCT ASSESSMENTS (Delta PCT)
A. PRIOR ECCS MODEL ASSESSMENTS
- 1. Backfit Through 2001 Reporting Year 0 2
- 2. Revised Blowdown Heatup Uncertainty Distribution 5 3 B. PLANNED PLANT MODIFICATION EVALUATIONS
- 1. Fan Cooler Performance Increase 2 2
- 2. Upflow Conversion Evaluation -29 4 C. 2012 ECCS MODEL ASSESSMENTS
- 1. PAD 4.0 Implementation -118 5
- 2. Evaluation of Fuel Pellet Thermal Conductivity 123 5 (a)
Degradation and Peaking Factor Burndown
- 3. Transverse Momentum Cells for Zero Cross-flow Boundary 0 5 (b)
Condition Error D. OTHER
- 1. WCAP-16043, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Virgil C. Summer Nuclear Station," June 2003.
- 2. CGE-03-12, "10 CFR 50.46 Annual Notification and Reporting for 2002," March 2003.
- 3. CGE-05-20, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.
- 4. LTR-LIS-08-578, Revision 2, "10 CFR 50.46 Reports for the V. C. Summer (CGE) Upflow Conversion Large Break LOCA Evaluation and Assessment of Transverse Momentum Cells with a Zero Cross-flow Boundary Condition Error," January 2009.
- 5. LTR-LIS-12-372, "V. C. Summer, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown," September 20, 2012.
Notes:
(a) This evaluation credits peaking factor burndown, see Reference 5.
(b) This input error was originally reported in Reference 4. That evaluation is superseded by the report in Reference 5.
Document Control Desk Attachment II LTD 321 RC-12-0080 Page 3 of 7 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: V. C. Summer Utility Name: South Carolina Electric & Gas Revision Date: 3/1/2013 Blowdown Analysis Information EM: CQD (1996) Analysis Date: 2/3/2003 Limiting Brezak Size: Guillotine FQ: 2.5 FdH: 1.7 Fuel: Vantage + SGTP (%): 10 Notes: Delta 75 Replacement Steam Generator Uprate Core Power 2900 MWt Clad Temp (*F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1860 1 PCT ASSESSMENTS (Delta PCT)
A. PRIOR ECCS MODEL ASSESSMENTS
- 1. Backfit Through 2001 Reporting Year 0 2
- 2. Revised Blowdown Heatup Uncertainty Distribution 49 3 B. PLANNED PLANT MODIFICATION EVALUATIONS
- 1. Fan Cooler Performance Increase 0 2
- 2. Upflow Conversion Evaluation -7 4 C. 2012 ECCS MODEL ASSESSMENTS
- 1. PAD 4.0 Implementation -83 5
- 2. Evaluation of Fuel Pellet Thermal Conductivity 0 5 (a)
Degradation and Peaking Factor Burndown
- 3. Transverse Momentum Cells for Zero Cross-flow 0 5 (b)
Boundary Condition Error D. OTHER
References:
- 1. WCAP-16043, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Virgil C. Summer Nuclear Station," June 2003.
- 2. CGE-03-12, "10 CFR 50.46 Annual Notification and Reporting for 2002," March 2003.
- 3. CGE-05-20, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.
- 4. LTR-LIS-08-578, Revision 2, "10 CFR 50.46 Reports for the V. C. Summer (CGE) Upflow Conversion Large Break LOCA Evaluation and Assessment of Transverse Momentum Cells with a Zero Cross-flow Boundary Condition Error," January 2009.
- 5. LTR-LIS-12-372, "V. C. Summer, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown," September 20, 2012.
Notes:
(a) This evaluation credits peaking factor burndown, see Reference 5.
(b) This input error was originally reported in Reference 4. That evaluation is superseded by the report in Reference 5.
Document Control Desk Attachment II LTD 321 RC-12-0080 Page 4 of 7 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: V. C. Summer Utility Name: South Carolina Electric & Gas Revision Date: 3/1/2013 Reflood 1 Analysis Information EM: CQD (1996) Analysis Date: 2/3/2003 Limiting Break S ize: Guillotine FQ: 2.5 FdH: 1.7 Fuel: Vantage + SGTP (%): 10 Notes: Delta 75 Replacement Steam Generator Uprate Core Power 2900 MWt Clad Temp (*F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1808 1 PCT ASSESSMENTS (Delta PCT)
A. PRIOR ECCS MODEL ASSESSMENTS
- 1. Backfit Through 2001 Reporting Year 0 2
- 2. Revised Blowdown Heatup Uncertainty Distribution 5 3 B. PLANNED PLANT MODIFICATION EVALUATIONS
- 1. Fan Cooler Performance Increase 1 2
- 2. Upflow Conversion Evaluation -44 4 C. 2012 ECCS MODEL ASSESSMENTS
- 1. PAD 4.0 Implementation -118 5
- 2. Evaluation of Fuel Pellet Thermal Conductivity 113 5 (a)
Degradation and Peaking Factor Burndown
- 3. Transverse Momentum Cells for Zero Cross-flow 0 5 (b)
Boundary Condition Error D. OTHER
References:
- 1. WCAP-16043, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Virgil C. Summer Nuclear Station," June 2003.
- 2. CGE-03-12, "10 CFR 50.46 Annual Notification and Reporting for 2002," March 2003.
- 3. CGE-05-20, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.
- 4. LTR-LIS-08-578, Revision 2, "10 CFR 50.46 Reports for the V. C. Summer (CGE) Upflow Conversion Large Break LOCA Evaluation and Assessment of Transverse Momentum Cells with a Zero Cross-flow Boundary Condition Error," January 2009.
- 5. LTR-LIS-12-372, "V. C. Summer, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown," September 20, 2012.
Notes:
(a) This evaluation credits peaking factor burndown, see Reference 5.
(b) This input error was originally reported in Reference 4. That evaluation is superseded by the report in Reference 5.
Document Control Desk Attachment II LTD 321 RC-1 2-0080 Page 5 of 7 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: V. C. Summer Utility Name: South Carolina Electric & Gas Revision Date: 3/1/2013 Reflood 2 Analysis Information EM: CQD (1996) Analysis Date: 2/3/2003 Limiting Break Size: Guillotine FQ: 2.5 FdH: 1.7 Fuel: Vantage + SGTP (%): 10 Notes: Delta 75 Replacement Steam Generator Uprate Core Power 2900 MWt Clad Temp (*F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1988 1 PCT ASSESSMENTS (Delta PCT)
A. PRIOR ECCS MODEL ASSESSMENTS 1, Backfit Through 2001 Reporting Year 0 2 2, Revised Blowdown Heatup Uncertainty Distribution 5 3 B. PLANNED PLANT MODIFICATION EVALUATIONS 1, Fan Cooler Performance Increase 2 2 2, Upflow Conversion Evaluation -29 4 C. 2012 ECCS MODEL ASSESSMENTS 1, PAD 4.0 Implementation -118 5 2, Evaluation of Fuel Pellet Thermal Conductivity 123 5 (a)
Degradation and Peaking Factor Burndown 3, Transverse Momentum Cells for Zero Cross-flow 0 5 (b)
Boundary Condition Error D. OTHER
References:
- 1. WCAP-16043, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Virgil C. Summer Nuclear Station," June 2003.
- 2. CGE-03-12, "10 CFR 50.46 Annual Notification and Reporting for 2002," March 2003.
- 3. CGE-05-20, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.
- 4. LTR-LIS-08-578, Revision 2, "10 CFR 50.46 Reports for the V. C. Summer (CGE) Upflow Conversion Large Break LOCA Evaluation and Assessment of Transverse Momentum Cells with a Zero Cross-flow Boundary Condition Error," January 2009.
- 5. LTR-LIS-12-372, "V. C. Summer, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown," September 20, 2012.
Notes:
(a) This evaluation credits peaking factor burndown, see Reference 5.
(b) This input error was originally reported in Reference 4. That evaluation is superseded by the report in Reference 5.
Document Control Desk Attachment II LTD 321 RC-12-0080 Page 6 of 7 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: V. C. Summer Utility Name: South Carolina Electric & Gas Revision Date: 3/1/2013 Analysis Information EM: NOTRUMP Analysis Date: 9/12/2006 Limiting Break Size: 3 Inch FQ: 2.45 FdH: 1.62 Fuel: Vantage + SGTP (%): 10 Notes:
Clad Temp (OF) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1775 9 (a)
A. PRIOR ECCS MODEL ASSESSMENTS
- 1. None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS
- 1. Upflow Conversion 148 10,11 C. 2012 ECCS MODEL ASSESSMENTS
- 1. None 0 D. OTHER
References:
- 1. CGE-94-205, "South Carolina Electric and Gas Company, Virgil C. Summer Station, 10 CFR 50.46 Notification and Reporting Information," February 8, 1994.
- 2. CGE-94-228, "South Carolina Electric and Gas Company, Virgil C. Summer Station, SBLOCTA Axial Nodalization," October 27, 1994.
- 3. CGE-95-201, "South Carolina Electric and Gas Company, Virgil C. Summer Station, 10 CFR 50.46 Notification and Reporting Information," February 3, 1995.
- 4. CGE-96-202, "South Carolina Electric and Gas Company, Virgil C. Summer Station, 10 CFR 50.46 Annual Notification and Reporting," February 9, 1996.
- 5. CGE-96-213, "South Carolina Electric and Gas Company, Virgil C. Summer Station, 10 CFR 50.46 Small Break LOCA Notification and Reporting," July 8, 1996.
- 6. CGE-00-044, "South Carolina Electric and Gas Company, Virgil C. Summer Nuclear Station, 10 CFR 50.46 Appendix K (BART / BASH / NOTRUMP) Evaluation Model, Mid-Year Notification and Reporting for 2000," June 30, 2000.
- 7. CGE-03-80, "10 CFR 50.46 Mid-Year Notification and Reporting for 2003," January 2004.
November 2006.
- 9. LTR-LIS-06-662, Transmittal of V. C. Summer SBLOCTA PCT Rackup Sheets for HHSI Throttle Valve Replacement," November 2006.
- 10. WCAP-16980-P, Revision 1, "Reactor Internals Upflow Conversion Program Engineering Report V. C. Summer Nuclear Station Unit 1," December 2008.
- 11. LTR-LIS-09-18, "10 CFR 50.46 Report for the V. C. Summer (CGE) Upflow Conversion
Document Control Desk Attachment II LTD 321 RC-12-0080 Page 7 of 7 Program Small Break LOCA Evaluation," January 2009.
Notes:
(a) The Rebaseline Analysis includes the impacts of the following model assessments:
1-LUCIFER Error Corrections (Ref. 1) 2-Effect of SI in Broken Loop (Ref. 1) 3-Effect of Improved Condensation Model (Ref. 1) 4-Axial Nodalization, RIP Model Revision and SBLOCTA Error Corrections Analysis (Ref. 2) 5-Boiling Heat Transfer Error (Ref. 3) 6-Steam Line Isolation Logic Error (Ref. 3) 7-NOTRUMP Specific Enthalpy Error (Ref. 4) 8-SALIBRARY Double Precision Error (Ref. 4) 9-SBLOCTA Fuel Rod Initialization Error (Ref. 5) 10-NOTRUMP Mixture Level Tracking / Region Depletion Errors (Ref. 6) 11-NOTRUMP Bubble Rise / Drift Flux Model Inconsistency Corrections (Ref. 7) 12-Refined Break Spectrum (Ref. 8) 13-High head safety injection (HHSI) flow increase (Ref. 9)