ML17097A094
| ML17097A094 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/07/2017 |
| From: | Jim Barstow Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML17097A094 (8) | |
Text
Exelon Generation !*
200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.46 April 7, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRG Docket No. 50-244
Subject:
2017 1o CFR 50.46 Annual Report
Reference:
Letter from J. Barstow (Exelon Generation Company, LLC) to U.S.
Nuclear Regulatory Commission, "2016 10 CFR 50.46 Annual Report,"
dated April 7, 2016 The purpose of this letter is to submit the 10 CFR 50.46 annual reporting information for R.E. Ginna Nuclear Power Plant. The referenced document is the most recent annual 10 CFR 50.46 Report report submitted to the U.S. Nuclear Regulatory Commission.
Two attachments are included with this letter that provide the current Ginna 10 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 61 0-765-5247.
r Respectfully,
,J<-*J-r LJcr-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: U.S. NRG Administrator, Region I U.S. NRG Project Manager, Ginna U.S. NRG 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 7, 2017 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 7, 2017 Attachment 1 Peak Cladding Temperature Rack-Up Sheet for Ginna Page 1 of 2 PLANT NAME: Ginna ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)
REPORT REVISION DATE: 4/7/2017 CURRENT OPERATING CYCLE: 39 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 T avg Cold Leg Break Reference Peak Cladding Temperature (PCT) PCT= 1167.0°F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated April 30, 2007 (Note 1) dPCT = 0°F 10 CFR 50.46 report dated February 10, 2009 (Note 3) dPCT = 0°F 10 CFR 50.46 report dated March 4, 2011 (Note 5) dPCT = 0°F 10 CFR 50.46 report dated March 27, 2012 (Note 6) dPCT = 0°F 10 CFR 50.46 report dated April 1, 2013 (Note 8) dPCT = 0°F 10 CFR 50.46 report dated April 9, 2014 (Note 9) dPCT = 0°F 10 CFR 50.46 report dated April 9, 2015 (Note 10) dPCT = 0°F 10 CFR 50.46 report dated April 7, 2016 (Note 11) dPCT = 0°F NET PCT PCT:1167.0°F B. CURRENT LOCA MODEL ASSESSMENTS None (Note 12) dPCT = 0°F Total PCT change from current assessments L: dPCT = 0°F Cumulative PCT change from current assessments L: ldPCTI= 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 7, 2017 Attachment 1 Peak Cladding Temperature Rack-Up Sheet for Ginna Page 2 of 2 PLANT NAME: Ginna ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)
REPORT REVISION DATE: 4/7/2017 CURRENT OPERATING CYCLE: 39 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 10 CFR 50.46 report dated April 30, 2007 (Note 1) ~PCT= 0°F 1o CFR 50.46 report dated May 4, 2008 (Note 2) ~PCT=+ 37°F 10 CFR 50.46 report dated February 10, 2009 (Note 3) ~PCT= 0°F 10 CFR 50.46 report dated March 26, 201 O (Note 4) ~PCT= 0°F 10 CFR 50.46 report dated March 4, 2011 (Note 5) ~PCT= 0°F 10 CFR 50.46 report dated March 27, 2012 (Note 6) ~PCT= 0°F 10 CFR 50.46 report dated August 16, 2012 (Note 7) ~PCT= +134°F 10 CFR 50.46 report dated April 1, 2013 (Note 8) ~PCT= +75°F 10 CFR 50.46 report dated April 9, 2014 (Note 9) ~PCT= +2°F 10 CFR 50.46 report dated April 9, 2015 (Note 10) ~PCT= 0°F 10 CFR 50.46 report dated April 7, 2016 (Note 11) ~PCT= +1°F NET PCT PCT =2119.0°F B. CURRENT LOCA MODEL ASSESSMENTS General Code Maintenance (Note 12) ~PCT= 0°F Error in Oxidation Calculations (Note 12) ~PCT= 0°F Error in Use of ASME Tables (Note 12) ~PCT= 0°F Total PCT change from current assessments I ~PCT= 0°F Cumulative PCT change from current assessments I l~PCTI= 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 7, 2017 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 7, 2017 Attachment 2 Assessment Notes Page 1 of 3
- 1. Prior LOCA Model Assessment The 10 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 10 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 10 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 10 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 10 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 10 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 10 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 10 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 7, 2017 Attachment 2 Assessment Notes Page 2 of 3 (ML14232A331) dated August 21, 2014. The SER (ML14232A331) evaluated the 10 CFR 50.46 reporting criteria explicitly.
- 9. Prior LOCA Model Assessment The 10 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 10 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. Current LOCA Model Assessment For the current LBLOCA analysis, 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.
There were two errors assessed to the LBLOCA analysis. The first error was related to the calculation of high temperature oxidation within a realistic large break loss-of-coolant accident (LOCA) calculation. It was determined that correcting the high temperature oxidation calculation in WCOBRA/TRAC is estimated to have a negligible impact on the Best Estimate (BE) LBLOCA peak cladding temperature (PCT) analysis results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes. The second error was related to the use of the American Society of Mechanical Engineers (ASME) steam tables to calculate the steady-state upper head liquid temperature as a function of the pressure and specific enthalpy in the ASTRUM software program. The steam table applicable to steam/gas is used to determine the upper head fluid temperature. However, the water in the upper head is in the subcooled liquid state during normal operation (and the steady-state calculation). Therefore, the steam table applicable to
Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 7, 2017 Attachment 2 Assessment Notes Page 3 of 3 liquid should be used to determine the upper head fluid temperature. It was determined that the temperatures calculated by the ASME steam tables applicable to the steam/gas side and the liquid side are very similar within the typical upper head pressure and liquid specific enthalpy ranges. Therefore, this error was evaluated to have a negligible impact on the ASTRUM BE LBLOCA analysis results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes.
Therefore, there is no PCT impact to the LBLOCA analysis from the evaluations and assessments within this report. Furthermore, the SBLOCA analysis was not impacted by any assessment within this report.