TMI-14-066, 10 CFR 50.46 Annual Report

10 CFR 50.46 Annual Report

ML14129A206
Person / Time
Site:	Three Mile Island
Issue date:	05/09/2014
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TMI-14-066
Download: ML14129A206 (8)

Text

10 CFR 50.46 TMI-14-066 May 9, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

10 CFR 50.46 Annual Report

Reference:

1) Letter from D. P. Helker (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "1 0 CFR 50.46 Annual Report,"

dated May 10, 2013 The purpose of this letter is to submit the 10 CFR 50.46 reporting information for Three Mile Island Nuclear Station (TMI), Unit 1. The most recent annual 50.46 Report for TMI, Unit 1 (Reference 1) provided the cumulative Peak Cladding Temperature (PCT) errors for the most recent fuel designs.

Since the Reference 1 report was issued, no vendor notifications of Emergency Core Cooling System (ECCS) model errors/changes applicable to TMI, Unit 1 have been issued.

No other ECCS-related changes or modifications have occurred at TMI, Unit 1 that affect the assumptions of the ECCS system.

Two attachments are included with this letter that provide the current TMI, Unit 1, 10 CFR 50.46 status. Attachment 1 ("Peak Cladding Temperature Rack-Up Sheets"

)

provides updated information regarding the PCT for the limiting SBLOCA and LBLOCA analyses. Attachment 2 (.. Assessment Notes") contains a detailed description for each change or error reported.

10 CFR 50.46 Annual Report May 9, 2014 Page2 No new regulatory commitments are established in this submittal. If any additional information is needed, please contact Tom Loomis at (610) 765-5510.

Respectfully, James Barstow Director- Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Peak Cladding Temperature Rack-Up Sheets

2) Assessment Notes cc: USNRC Administrator, Region I USNRC Project Manager, TMI, Unit 1 USNRC Senior Resident Inspector, TMI, Unit 1

ATTACHMENT 1 10 CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments Assessments as of May 9, 2014 Peak Cladding Temperature Rack-Up Sheets TMI, Unit 1

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 9, 2014 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 1 of 2 PLANT NAME: Three Mile Island Nuclear Station. Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)

REPORT REVISION DATE: 5/9/14 CURRENT OPERATING CYCLE: 20 ANALYSIS OF RECORD (AOR)

Evaluation Model: BWNT 1 Calculation: AREVA NP, 86-9111507-000, August 2009 (Mark-B-HTP with Enhanced Once-Through Steam Generators (EOTSGs))

Fuel: Mark-B-HTP Limiting Fuel Type: Mark-B-HTP Limiting Single Failure: Loss of One Train of ECCS Limiting Break Size and Location: 0.07 ft2 Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT)

MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) ~PCT =0°F Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) ~PCT =ooF Annual10 CFR 50.46 Report dated May 14, 2010 (See Note 4) ~PCT =ooF 30-Day 10 CFR 50.46 Report dated September 7, 2010 (See Note 5) ~PCT =225°F Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) ~PCT = ooF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) ~PCT = ooF Annual10 CFR 50.46 Report dated May 11, 2012 (See Note 8) ~PCT =ooF Annual10 CFR 50.46 Report dated May 10, 2013 (See Note 9) ~PCT =ooF NETPCT PCT =1669°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 10) ~PCT =ooF NETPCT PCT = 1669°F 1

The BWNT EM is based on RELAP5/MOD2-B&W.

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 9, 2014 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 2 of 2 PLANT NAME: Three Mile Island Nuclear Station. Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)

REPORT REVISION DATE: 5/9/14 CURRENT OPERATING CYCLE: 20 ANALYSIS OF RECORD (AOR)

Evaluation Model: BWNT2 Calculation: AREVA NP, 86-9111507-000, August 2009 (Mark-B-HTP with EOTSGs)

Fuel: Mark-B-HTP Limiting Fuel Type: Mark-B-HTP Limiting Single Failure: Loss of One Train of ECCS Limiting Break Size and Location: Guillotine Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT) PCT = 1890°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) ~PCT =ooF Annual 10 CFR 50.46 Report dated May 15, 2008 (See Note 2) ~PCT =ooF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) ~PCT = ooF Annual10 CFR 50.46 Report dated May 14, 2010 (See Note 4) ~PCT = ooF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) LlPCT =ooF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) LlPCT =ooF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) LlPCT =ooF Annual10 CFR 50.46 Report dated May 10, 2013 (See Note 9) LlPCT =ooF NET PCT PCT = 1890°F B. CURRENT LOCA MODEL ASSESSMENTS None (See Note 10) ~PCT =ooF NETPCT 2

The BWNT EM is based on RELAP5/MOD2-B&W.

ATTACHMENT 2 10 CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments Assessments as of May 9, 2014 Peak Cladding Temperature Rack-Up Sheets TMI, Unit 1 Assessment Notes

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 9, 2014 Attachment 2 Assessment Notes Page 1 of 2

1. Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 16, 2007 reported an evaluation for a LOCA model change which resulted in a ooF PCT change. The reported evaluation considered the effect on the containment pressure response for LOCA due to GSI-191 related reactor building sump screen replacement. The evaluation resulted in ooF impact for L8LOCA and S8LOCA PCTs.

2. Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 15, 2008 reported an evaluation for LOCA model change which resulted in a ooF PCT change. The reported change included the impact of an energy deposition factor error which resulted in a L8LOCA PCT impact of 0°F.

3. Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 15, 2009 reported no evaluations or PCT penalties for either S8LOCA or L8LOCA.

4. Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 14, 2010 reported a change to the reference PCT value for L8LOCA due to the final discharge of all Mark-89 fuel.

Also identified in this report was a new S8LOCA analysis, implemented beginning with the Cycle 18 operation. This S8LOCA analysis was evaluated with the mixed core of Mark-812 and Mark-8-HTP and a new PCT of 1444°F was calculated for the limiting Mark-8-HTP fuel type, which bounds the Mark-812 fuel type. This analysis also includes consideration of the effect of reduced EFW wetting associated with the Enhanced Once-Through Steam Generators (EOTSGs).

5. Prior LOCA Model Assessment The 10 CFR 50.46 report dated September 7, 2010 reported an evaluation for the S8LOCA analysis due to a non-bounding axial power shape from middle-of-cycle to end-of-cycle conditions. This resulted in a PCT increase of 225°F. The large break LOCA is not affected in this report.

6. Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 13, 2011 reported no evaluations or PCT penalties for either S8LOCA or L8LOCA.

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 9, 2014 Attachment 2 Assessment Notes Page 2 of 2

7. Prior LOCA Model Assessment The 10 CFR 50.46 report dated March 21, 2012 reported two changes to the TMI LOCA model. One consisted of an error in the ECCS Bypass Calculation that affected the LBLOCA analysis. The second change consisted of correcting the Upper Plenum Column Weldment Model which affected both the SBLOCA and LBLOCA analysis. The results of both of these changes were a 0°F PCT impact for both SBLOCA and LBLOCA.

8. Prior LOCA Model Assessment With the Cycle 19 reload, all Mark-812 fuel types were discharged from the core. Currently, the limiting fuel type is Mark-8-HTP for both SBLOCA and LBLOCA. The limiting PCT for LBLOCA has been updated to 1890°F in accordance with our referenced calculation (86-9111507-000). All previous PCT assessments that are not applicable to Mark-8-HTP fuel have been removed.

The 10 CFR 50.46 report dated May 11, 2012 reported no evaluations or PCT penalties for either SBLOCA or LBLOCA.

9. Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 10, 2013 reported no evaluations or PCT penalties for either SBLOCA or LBLOCA.

10. Current LOCA Model Assessment No model changes or errors were identified for this annual report.