ML18096A004
| ML18096A004 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/06/2018 |
| From: | Jim Barstow Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18096A004 (8) | |
Text
200 Exelon Way Exelon Kennett Square. PA 19348 April 6, 2018 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 R.E. Ginna Nuclear Power Plant www.exeloncorp com Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244
Subject:
2018 1 O CFR 50.46 Annual Report 10 CFR 50.46
Reference:
Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "2017 1 O CFR 50.46 Annual Report," dated April 7, 2017 The purpose of this letter is to submit the 1 O CFR 50.46 annual reporting information for R.E. Ginna Nuclear Power Plant. The referenced letter is the most recent annual 1 O CFR 50.46 Report submitted to the U.S. Nuclear Regulatory Commission. Two attachments are included with this letter that provide the current Ginna 1 O CFR 50.46 status. Attachment 1 provides the Peak Cladding Temperature (PCT) "rack-up" sheets. Attachment 2, "Assessment Notes," contains a detailed description of each change/error reported. There are no commitments contained in this letter. If you have any questions, please contact Ron Reynolds at 610-765-5247. Respectfully, James Barstow Director -Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Peak Cladding Temperature Rack-Up Sheet for R.E. Ginna Nuclear Power Plant 2) Assessment Notes cc: USNRC Regional Administrator, Region I USNRC Project Manager, NRA USNRC Senior Resident Inspector, Ginna ATTACHMENT 1 10 CFR 50.46 "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 6, 2018 Peak Cladding Temperature Rack-Up Sheet for R.E. Ginna Nuclear Power Plant Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 6, 2018 Peak Cladding Temperature Rack-Up Sheet for Ginna PLANT NAME: Ginna Attachment 1 Page 1 of 2 ECCS EVALUATION MODEL: REPORT REVISION DATE: Small Break Loss of Coolant Accident (SBLOCA) 4/6/2018 CURRENT OPERATING CYCLE: 40 ANALYSIS OF RECORD Evaluation Model: NOTRUMP Calculation: Westinghouse CN-LIS-04-206, April 2005 Fuel: 422 Vantage+ Limiting Fuel Type: 422 Vantage+ Limiting Single Failure: Diesel Generator Failure to Start Limiting Break Size and Location: 2-inch Equivalent High Tavg Cold Leg Break Reference Peak Cladding Temperature (PCT) PCT= 1167.0°F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 1 O CFR 50.46 report dated April 30, 2007 (Note 1) f1PCT = 0°F 1 O CFR 50.46 report dated February 10, 2009 (Note 3) f1PCT = 0°F 1 O CFR 50.46 report dated March 4, 2011 (Note 5) f1PCT = 0°F 10 CFR 50.46 report dated March 27, 2012 (Note 6) f1PCT = 0°F 1 O CFR 50.46 report dated April 1, 2013 (Note 8) f1PCT = 0°F 1 O CFR 50.46 report dated April 9, 2014 (Note 9) f1PCT = 0°F 1 O CFR 50.46 report dated April 9, 2015 (Note 10) f1PCT = 0°F 1 O CFR 50.46 report dated April 7, 2016 (Note 11) f1PCT = 0°F NET PCT PCT =1167.0°F B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance (Note 13) f1PCT = 0°F Error in the Upper Plenum Fluid Volume Calculation (Note 13) f1PCT = 0°F Total PCT change from current assessments I f1PCT = 0°F Cumulative PCT change from current assessments I lf1PCTI= 0°F NET PCT PCT =1167.0°F Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 6, 2018 Peak Cladding Temperature Rack-Up Sheet for Ginna PLANT NAME: Ginna Attachment 1 Page 2 of 2 ECCS EVALUATION MODEL: REPORT REVISION DATE: Large Break Loss of Coolant Accident (LBLOCA) 4/6/2018 CURRENT OPERATING CYCLE: 40 ANALYSIS OF RECORD Evaluation Model: ASTRUM (2004) Calculation: Westinghouse CN-LIS-05-11, April 2005 Fuel: 422 Vantage+ Limiting Fuel Type: 422 Vantage+ Limiting Single Failure: Loss of one train of ECCS flow Limiting Break Size and Location: Cold Leg Split Break Reference PCT PCT= 1870.0°F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 1 O CFR 50.46 report dated April 30, 2007 (Note 1) 0°F 1 o CFR 50.46 report dated May 4, 2008 (Note 2) 37°F 1 O CFR 50.46 report dated February 10, 2009 (Note 3) 0°F 1 O CFR 50.46 report dated March 26, 201 O (Note 4) 0°F 1 O CFR 50.46 report dated March 4, 2011 (Note 5) 0°F 1 O CFR 50.46 report dated March 27, 2012 (Note 6) 0°F 1 O CFR 50.46 report dated August 16, 2012 (Note 7) +134°F 1 O CFR 50.46 report dated April 1, 2013 (Note 8) +75°F 1 O CFR 50.46 report dated April 9, 2014 (Note 9) +2°F 1 O CFR 50.46 report dated April 9, 2015 (Note 10) 0°F 10 CFR 50.46 report dated April 7, 2016 (Note 11) +1°F 10 CFR 50.46 report dated April 7, 2017 (Note 12) 0 °F NET PCT PCT =2119.0°F B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance (Note 13) 0 °F Evaluation of Inconsistent Application of Numerical Ramp Applied to 0 °F the Entrained Liquid I Vapor lnterfacial Drag Coefficient (Note 13) Evaluation of Inappropriate Resetting of Transverse Liquid Mass Flow (Note 13) Evaluation of Steady-State Fuel Temperature Calibration Method (Note 13) Total PCT change from current assessments 0°F Cumulative PCT change from current assessments L I I = 0°F NET PCT PCT =2119.0°F ATTACHMENT 2 10 CFR 50.46 "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 6, 2018 Assessment Notes R.E. Ginna Nuclear Power Plant Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 6, 2018 Assessment Notes 1. Prior LOCA Model Assessment Attachment 2 Page 1 of 3 The 1 O CFR 50.46 report dated April 30, 2007, reported new licensing basis peak cladding temperature (PCT) for small break loss of coolant accident (SBLOCA) and large break loss of coolant accident (LBLOCA) analyses to support fuel assembly transition from OFA to 422 Vantage+ and extended power uprate. The new licensing basis PCT reported for SBLOCA and LBLOCA are 1167°F and 1870°F, respectively. 2. Prior LOCA Model Assessment The 1 O CFR 50.46 report dated May 4, 2008, reported an evaluation for LBLOCA related to HOTSPOT fuel relocation error which resulted in a 37°F PCT assessment. 3. Prior LOCA Model Assessment The 1 O CFR 50.46 report dated February 10, 2009, reported evaluations for SBLOCA and LBLOCA model changes which resulted in 0°F PCT change. 4. Prior LOCA Model Assessment The 1 O CFR 50.46 report dated March 26, 2010, reported evaluations for LBLOCA model changes which resulted in 0°F PCT change. 5. Prior LOCA Model Assessment The 1 O CFR 50.46 report dated March 4, 2011, reported evaluations for SBLOCA and LBLOCA model changes which resulted in 0°F PCT change. 6. Prior LOCA Model Assessment The 1 O CFR 50.46 report dated March 27, 2012, reported evaluations for SBLOCA and LBLOCA model changes which resulted in 0°F PCT change. 7. Prior LOCA Model Assessment The 30-day 1 O CFR 50.46 report dated August 16, 2012, reported evaluations for fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown, and design input change assessments which resulted in a 134°F PCT impact for LBLOCA. 8. Prior LOCA Model Assessment The 1 o CFR 50.46 report dated April 1, 2013, reported evaluations for SBLOCA model changes which resulted in 0°F PCT impact. A LBLOCA assessment for the evaluation of an elevated initial containment and accumulator temperature was submitted in a License Amendment Request for NRC review and approval. The assessment resulted in a 75°F PCT impact. This increase in temperature was approved in an NRC Safety Evaluation Report (SER)
Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 6, 2018 Assessment Notes Attachment 2 Page 2 of 3 (ML 14232A331) dated August 21, 2014. The SER (ML 14232A331) evaluated the 10 CFR 50.46 reporting criteria explicitly. 9. Prior LOCA Model Assessment The 1 O CFR 50.46 report dated April 9, 2014, reported evaluations for SBLOCA model changes which resulted in 0°F PCT impact. A LBLOCA assessment was reported related to revised heat transfer multiplier distribution which resulted in a 2°F PCT assessment. 10. Prior LOCA Model Assessment The 1 O CFR 50.46 report dated April 9, 2015, reported general code maintenance for both LBLOCA and SBLOCA. An error in Decay Group Uncertainty Factors against the LBLOCA model was reported. Additionally, it reported errors in Fuel Rod Gap Conductance, Radiation Heat Transfer Model, and SBLOCTA Pre-DNB Cladding Surface Heat Transfer Coefficient Calculation for the SBLOCA model. All changes resulted in 0°F PCT impact. 11. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 7, 2016, reported General Code Maintenance for the SBLOCA which led to a PCT impact of 0°F. Additionally, Ginna began inserting reconstituted fuel with 5 stainless steel filler rods starting in Cycle 39. The effects to SBLOCA are 0°F and the effects to LBLOCA are 1°F for as long as reconstituted fuel with 5 stainless steel filler rods remain in the core. 12. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 7, 2017 reported General Code Maintenance for the LBLOCA to enhance the usability of codes and to streamline future analyses which led to a PCT impact of 0°F. There were two errors assessed to the LBLOCA analysis related to the calculation of high temperature oxidation within a realistic LBLOCA calculation and to the use of the American Society of Mechanical Engineers (ASME) steam tables to calculate the state upper head liquid temperature as a function of the pressure and specific enthalpy in the ASTRUM software program. Both errors each resulted in an estimated PCT impact of 0°F for 1 O CFR 50.46. There were no impacts or assessments to SBLOCA. 13. Current LOCA Model Assessment For the current LBLOCA and SBLOCA analyses, 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 resulted in an estimated PCT impact of 0°F.
Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 6, 2018 Assessment Notes Attachment 2 Page 3 of 3 There was one error assessed to the SBLOCA analysis. An error was found in the fluid volume calculation in the upper plenum where the support column outer diameter was being used instead of the inner diameter. The correction of this error lead to a reduction in the upper plenum fluid volume used in the Appendix K Small Break LOCA analysis. The corrected values represent a less than 1 % change in the total RCS fluid volume and will be incorporated on a forward-fit basis, based on the evaluated impact on the current licensing basis analysis results and is to have a negligible effect on small break LOCA analysis results, leading to an estimated PCT impact of 0°F. There were three evaluations assessed to the LBLOCA analysis. The first evaluation involved a numerical ramp which was used to account for the disappearance of the entrained liquid phase was applied to the entrained liquid I vapor interfacial drag coefficient. The numerical ramp was applied such that the interfacial drag coefficient used in the solution of the entrained liquid and vapor momentum equations was not consistent. WCOBRA/TRAC was updated to apply the numerical ramp prior to usage of the interfacial drag coefficient in the momentum equations, such that a consistent interfacial drag coefficient was used in the entrained liquid and vapor momentum equations. It was determined, based on the code validation results, the impact of correcting the error is estimated to have a 0°F impact on PCT. The second evaluation involved the WCOBRA/TRAC routine which evaluates the mass and energy residual error of the time step solution, the transverse liquid mass flow is reset as the liquid phase disappears. The routine is updated to remove the resetting of the transverse liquid mass flow since the routine is to only evaluate the residual error based on the time step solution values. Based on the code validation results and limited applicability of the logic removed, correcting the error is estimated to have a 0°F impact on PCT. The final evaluation concerned how the ASTRUM steady-state fuel pellet temperature calibration method involves solving for the hot gap width (AG FACT} to calibrate the fuel temperature for each fuel rod. In some infrequent situations, small non-conservatisms can occur in the calibration process such that the resulting fuel pellet temperature will be slightly lower than intended and outside the acceptable range defined by Table 12-6 of WCAP-16009-P/NP-A. A review of licensing basis analyses concluded that the potential non-conservatisms in the fuel pellet temperature calibration did not occur for the limiting analysis cases. Therefore, an estimated PCT impact of 0°F is assigned for 1 O CFR 50.46 reporting purposes. Therefore, there is no PCT impact to the LBLOCA analysis from the evaluations and there is no PCT impact to the SBLOCA analysis from the errors.