TMI-12-079, 10 CFR 50.46 Annual Report
| ML12135A380 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/11/2012 |
| From: | Jesse M Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TMI-12-079 | |
| Download: ML12135A380 (8) | |
Text
10 CFR 50.46 TMI-12-079 May 11,2012 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
References:
1)
Letter from M. D. Jesse (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report," dated May 13, 2011 2)
Letter from M. D. Jesse (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 30-Day Report," dated March 21, 2012 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, one vendor notification of Emergency Core Cooling System (ECCS) model error/changes applicable to TMI, Unit 1 was issued (Reference 2). 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 11, 2012 Page 2 No new regulatory commitments are established in this submittal. If any additional information is needed, please contact Tom Loomis at (610) 765*5510.
Respectfully, Michael D.
Director*
Regulatory Affairs Exelon Generation Company, LLC Attachments:
1)
Peak Cladding Temperature Rack*Up Sheets 2)
Assessment Notes cc:
W. Dean, USNRC Administrator, Region I P. J. Bamford, USNRC Project Manager, TMI, Unit 1 D. L Werkheiser, USNRC Senior Resident Inspector, TMI, Unit 1
ATIACHMENT 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 11, 2012 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 11, 2012 Peak Cladding Temperature Rack-Up Sheet Page 1 of 2 PLANT NAME:
ECCS EVALUATION MODEL:
REPORT REVISION DATE:
CURRENT OPERATING CYCLE:
ANALYSIS OF RECORD (AOR)
Three Mile Island Nuclear Station, Unit 1 Small Break Loss of Coolant Accident (SBLOCA) 5/11/12 19 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 te 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)
LlPCT = OaF Annual 10 CFR 50.46 Report dated Mav 15, 2009 (See Note 2)
LlPCT = OaF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 3)
LlPCT = OaF 30-Day 10 CFR 50.46 Report dated September 7, 2010 (See Note 4)
LlPCT = 225°F Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 5)
LlPCT = OaF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 6)
LlPCT = OaF NETPCT PCT =1669°F B. CURRENT LOCA MODEL ASSESSMENTS 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 11, 2012 Peak Cladding Temperature Rack-Up Sheet Page 2 of 2 PLANT NAME:
ECCS EVALUATION MODEL:
REPORT REVISION DATE:
CURRENT OPERATING CYCLE:
ANALYSIS OF RECORD (AOR)
Three Mile Island Nuclear Station, Unit 1 Large Break Loss of Coolant Accident (LBLOCA) 5/11/12 19 Evaluation Model: BWNT2 Calculation: Framatome ANP 86-5011294-00, March 2001 (Mark-B12)
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)
MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS PCT =1890°F Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) llPCT = OaF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 2) llPCT = OaF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 3) llPCT = OaF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 5)
LiPCT = OaF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 6)
LiPCT = OaF NETPCT PCT = 1890°F B. CURRENT LOCA MODEL ASSESSMENTS 2 The BWNT EM is based on RELAP5/MOD2-B&W.
AITACHMENT2 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 11, 2012 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 11, 2012 Assessment Notes 1.
Prior LOCA Model Assessment Page 1 of 2 The 10 CFR 50.46 report dated May 16, 2007 reported an evaluation for a LOCA model change which resulted in a O°F 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 O°F impact for LBLOCA and SBLOCA PCTs.
2.
Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 15, 2009 reported no evaluations or PCT penalties for either SBLOCA or LBLOCA.
3.
Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 14, 2010 reported a change to the reference PCT value for LBLOCA due to the final discharge of all Mark-B9 fuel.
Also identified in this report was a new SBLOCA analysis, implemented beginning with the Cycle 18 operation. This SBLOCA analysis was evaluated with the mixed core of Mark-B12 and Mark-B-HTP and a new PCT of 1444°F was calculated for the limiting Mark-B-HTP fuel type, which bounds the Mark-B12 fuel type. This analysis also includes consideration of the effect of reduced EFW wetting associated with the Enhanced Once-Through Steam Generators (EOTSGs).
4.
Prior LOCA Model Assessment The 10 CFR 50.46 report dated September 7, 2010 reported an evaluation for the SBLOCA 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.
5.
Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 13, 2011 reported no evaluations or PCT penalties for either SBLOCA or LBLOCA.
6.
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 O°F PCT impact for both SBLOCA and LBLOCA.
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 11, 2012 Assessment Notes 7.
Current LOCA Model Assessment Page 2 of 2 With the Cycle 19 reload, all Mark-B12 fuel types were discharged from the core. Currently, the limiting fuel type is Mark-B-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-B-HTP fuel have been removed.
No new model changes or errors were identified for this annual report.