ML19098A689

2019 10CFR 50.46 Annual Report

ML19098A689
Person / Time
Site:	Ginna
Issue date:	04/08/2019
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML19098A689 (9)

Text

200 Exelon Way Exelon Generation Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.46 April 8, 2019 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 NRC Docket No. 50-244

Subject:

2019 10 CFR 50.46 Annual Report

Reference:

1. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "2018 10 CFR 50.46 Annual Report," dated April 6, 2018 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 letter is the most recent annual 10 CFR 50.46 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 610-765-5247.

Respectfully, James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

2018 10 CFR 50.46 Annual Report R.E. Ginna Nuclear Power Plant April 8, 2019 Page 2 Attachments: 1) Peak Cladding Temperature Rack-Up Sheets for R.E. Ginna Nuclear Power Plant

2) Assessment Notes cc: U.S. NRC Administrator, Region I U.S. NRC Project Manager, Ginna U.S. NRC 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 8, 2019 Peak Cladding Temperature Rack-Up Sheets for R.E. Ginna Nuclear Power Plant

Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 8, 2019 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: 4/8/2019 CURRENT OPERATING CYCLE :41 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 10 CFR 50.46 report dated April 30, 2007 (Note 1) ~PCT= 0°F 10 CFR 50.46 report dated February 10, 2009 (Note 3) ~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 April 1, 2013 (Note 8) ~PCT= 0°F 10 CFR 50.46 report dated April 9, 2014 (Note 9) ~PCT= 0°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= 0°F 10 CFR 50.46 report dated April 6, 2018 (Note 13) ~PCT= 0°F NET PCT PCT =1167.0°F B. CURRENT LOCA MODEL ASSESSMENTS U02 Fuel Pellet Heat Capacity (Note 14) ~PCT= 0°F Total PCT chanae from current assessments L: ~PCT= 0°F Cumulative PCT change from current assessments L: l~PCTI= 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 8, 2019 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: 4/8/2019 CURRENT OPERATING CYCLE:41 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) tiPCT = 0°F 10 CFR 50.46 report dated May 4, 2008 (Note 2) tiPCT =+37°F 10 CFR 50.46 report dated February 10, 2009 (Note 3) tiPCT = 0°F 10 CFR 50.46 report dated March 26, 2010 (Note 4) ~PCT= 0°F 10 CFR 50.46 report dated March 4, 2011 (Note 5) tiPCT = 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) tiPCT = +75°F 10 CFR 50.46 report dated April 9, 2014 (Note 9) tiPCT = +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 (Note11 ) tiPCT = +1°F 10 CFR 50.46 report dated April 7, 2017 (Note 12) tiPCT = 0°F 10 CFR 50.46 report dated April 6, 2018 (Note 13) tiPCT = 0°F NET PCT PCT =2119.0°F B. CURRENT LOCA MODEL ASSESSMENTS Error In CCTF Model Used In WCOBRA/TRAC Validation For tiPCT = 0°F UPI Plants (Note 14)

Vapor Temperature Resetting (Note 14) ~PCT= 0°F Total PCT chanQe from current assessments I ~PCT= 0°F Cumulative PCT change from current assessments I ItiPCT 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 8, 2019 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 8, 2019 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) (Ml 14232A331) dated August 21, 2014. The SER (ML14232A331) evaluated the 10 CFR 50.46 reporting criteria explicitly.

Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 8, 2019 Attachment 2 Assessment Notes Page 2 of 3

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.

1O. 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. 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 steady-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 10 CFR 50.46. There were no impacts or assessments to SBLOCA.

13. Prior LOCA Model Assessment The 10 CFR 50.46 report dated April 6, 2018, reported General Code Maintenance tor the SBLOCA and LBLOCA which each led to a PCT impact of 0°F. The SBLOCA also reported one error pertained to the upper plenum fluid volume calculation with an estimated PCT impact of 0°F. The LBLOCA reported three assessments with the first involving an evaluation of inconsistent application of numerical ramp applied to the entrained liquid I vapor interfacial drag coefficient, the second involving an evaluation of inappropriate resetting of transverse liquid mass flow, and the third involving an evaluation of steady-state fuel temperature calibration method. All three errors each resulted in an estimated PCT impact of 0°F for 10 CFR 50.46.

Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of April 8, 2019 Attachment 2 Assessment Notes Page 3 of 3

14. Current LOCA Model Assessment For the current LBLOCA and SBLOCA analyses, there was one error assessed to the SBLOCA analysis. A typographical error was discovered in the implementation of the U02 fuel pellet heat capacity as described by Equation C-4 of WCAP-8301 for fuel rod heat-up calculations within the Appendix K Large Break and Small Break LOCA evaluation models. The erroneous formulation results in an overprediction of heat capacity that increases with fuel temperature. The corrected formulation results in a maximum decrease in heat capacity on the order of approximately 1.2% for existing analyses of record. The small over-prediction in U02 fuel pellet heat capacity has been evaluated to have a negligible effect on the small break LOCA analysis results due to the small magnitude of the change, leading to an estimated PCT impact of 0°F.

There were two errors assessed to the LBLOCA analysis. The first error involved error in CCTF model used in WCOBRA/TRAC validation for Upper Plenum Injection plants.

WCAP-14449-P-A, Revision 1, documents the extension of the Best-Estimate Large-Break Loss-of- Coolant Accident methodology using WCOBRA/TRAC to 2-loop Westinghouse Pressurized Water Reactors (PWRs) equipped with Upper Plenum Injection (UPI), and supports the application of the ASTRUM methodology to 2-loop PWRs equipped with UPI. WCOBRA/TRAC calculations for two Cylindrical Core Test Facility (CCTF) upper plenum injection tests were performed as part of the code validation. The test report for these tests indicated that the accumulator and high pressure coolant injects into the lower plenum for some time, then ramps down to zero and injection switches entirely into the cold legs. In the WCOBRA/TRAC model for these tests, the accumulator and high pressure coolant flow into the lower plenum was not modeled consistent with the test conditions. Therefore, the total amount of injected flow into the lower plenum was higher in the WCOBRA/TRAC calculations than in the tests. Calculations using the modified accumulator and high pressure coolant flow confirm that the error does not impact the conclusions made relative to the simulations, leading to an estimated peak cladding temperature impact of 0°F.

The second evaluation involved the WCOBRA/TRAC codes which when the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. It was discovered that this vapor temperature resetting logic results in an inconsistency between the conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions. Engineering judgement supported by sensitivity calculations showed that correcting this error had minimal impact on LOCA transient calculations, leading to an estimated peak cladding temperature impact of 0°F.

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.