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==Subject:== | ==Subject:== | ||
Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 10 CFR 50.46 Annual Report 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.46 | Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 10 CFR 50.46 Annual Report 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.46 | ||
==References:== | ==References:== | ||
: 1) Letter from James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report," dated May 8, 2015 2) Letter from Russell K. Cox (AREVA INC.) to John Massari (Exelon Generation Company, LLC), "Transmittal of Input for TMI 50.46 Report for 2015," dated March 29, 2016 -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). The evaluation model (EM) application error regarding reactor coolant system (RCS) flow rate considered in the LOCA analyses resulted in a 0°F PCT impact and is discussed in Attachment 2, Section 13, Current LOCA Model Assessment. No other ECCS-related changes or modifications have occurred at TMI, Unit 1, that affect the assumptions of the ECCS system. | : 1) Letter from James Barstow (Exelon Generation | ||
U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Report May 6, 2016 Page 2 Two attachments are included with this letter that provide the current TMI, Unit 1, 1 O 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. No new regulatory commitments are established in this submittal. If any additional information is needed, please contact Frank Mascitelli at (610) 765-5512. 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 | : Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report," | ||
dated May 8, 2015 2) Letter from Russell K. Cox (AREVA INC.) to John Massari (Exelon Generation | |||
: Company, LLC), "Transmittal of Input for TMI 50.46 Report for 2015," dated March 29, 2016 -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). The evaluation model (EM) application error regarding reactor coolant system (RCS) flow rate considered in the LOCA analyses resulted in a 0°F PCT impact and is discussed in Attachment 2, Section 13, Current LOCA Model Assessment. | |||
No other ECCS-related changes or modifications have occurred at TMI, Unit 1, that affect the assumptions of the ECCS system. | |||
U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Report May 6, 2016 Page 2 Two attachments are included with this letter that provide the current TMI, Unit 1, 1 O 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. | |||
No new regulatory commitments are established in this submittal. | |||
If any additional information is needed, please contact Frank Mascitelli at (610) 765-5512. | |||
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 | |||
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 1 of 2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA) REPORT REVISION DATE: 05/06/2016 CURRENT OPERATING CYCLE: 21 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 ft 2 Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT) PCT = 1444.0ºF MARGIN ALLOCATION | |||
Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated September 7, 2010 (See Note 5) PCT = 225ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF | |||
Reactor Coolant System Flow Inconsistency (See Note 13) PCT = 0ºF Total PCT change from current assessments PCT = 0ºF Cumulative PCT change from current assessments lPCTl = 0ºF 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 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 2 of 2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA) REPORT REVISION DATE: 05/06/2016 CURRENT OPERATING CYCLE: 21 Evaluation Model: BWNT 2 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 = 1890ºF MARGIN ALLOCATION | |||
Report | Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2008 (See Note 2) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = +18ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF Reactor Coolant System Flow Inconsistency (See Note 13) PCT = 0ºF Total PCT change from current assessments PCT = 0ºF Cumulative PCT change from current assessments lPCTl = 0ºF | ||
2 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 6, 2015 | Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 1 of 3 | ||
: 1. Prior LOCA Model Assessment | : 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 0ºF 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 0ºF impact for LBLOCA and SBLOCA 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 0ºF PCT change. Reported change included the impact of an energy deposition factor error which resulted in a LBLOCA PCT impact of 0ºF. 3. 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 0ºF 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 0ºF impact for LBLOCA and SBLOCA 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 0ºF PCT change. Reported change included the impact of an energy deposition factor error which resulted in a LBLOCA PCT impact of 0ºF. | |||
: 3. 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. | |||
: 4. 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. | |||
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 | 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 1444ºF 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). | |||
: 5. 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 225ºF. 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 SBLOCA or LBLOCA. | |||
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 2 of 3 | |||
: 7. Prior LOCA Model Assessment | : 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 0ºF PCT impact for both SBLOCA and LBLOCA. 8. 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 0ºF PCT impact for both SBLOCA and LBLOCA. | ||
All previous PCT assessments that are not applicable to Mark-B-HTP fuel have been removed. | : 8. Prior LOCA Model Assessment 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 1890ºF in accordance with our referenced calculation (86-9111507-000). | ||
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. Prior LOCA Model Assessment The 10 CFR 50.46 Report dated May 9, 2014, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA. 11. Prior LOCA Model Assessment The 30-day 10 CFR 50.46 Report dated December 22, 2014, reported a significant error due to thermal conductivity degradation (TCD) based on insufficient LOCA fuel temperature inputs in TACO-3/GDTACO computer codes. Correction of the TCD modeling in TACO-3/GDTACO results in a conservative increase of 393ºF in peak cladding temperature (PCT) for LBLOCA. The SBLOCA analyses are not sensitive to initial fuel temperature and therefore have an estimated PCT impact of 0ºF. | |||
All previous PCT assessments that are not applicable to Mark-B-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. Prior LOCA Model Assessment | |||
The 10 CFR 50.46 Report dated May 9, 2014, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA. | |||
: 11. Prior LOCA Model Assessment | |||
The 30-day 10 CFR 50.46 Report dated December 22, 2014, reported a significant error due to thermal conductivity degradation (TCD) based on insufficient LOCA fuel temperature inputs in TACO-3/GDTACO computer codes. Correction of the TCD modeling in TACO-3/GDTACO results in a conservative increase of 393ºF in peak cladding temperature (PCT) for LBLOCA. The SBLOCA analyses are not sensitive to initial fuel temperature and therefore have an | |||
estimated PCT impact of 0ºF. | |||
Additionally, TMI has implemented a 2 kw/ft penalty to LHR limits in Cycle 20 (10/20/14). The penalty has been applied through more restrictive operational imbalance limits and results in a reduction of PCT by 375ºF for LBLOCA. | |||
The overall cumulative impact for the error and the design input change is 0ºF for SBLOCA and 18ºF for LBLOCA. | The overall cumulative impact for the error and the design input change is 0ºF for SBLOCA and 18ºF for LBLOCA. | ||
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 | Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 3 of 3 | ||
: 12. Prior LOCA Model Assessment | : 12. Prior LOCA Model Assessment | ||
The 10 CFR 50.46 Report dated May 8, 2015, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA. 13. Current LOCA Model Assessment Reactor Coolant System (RCS) flow rate used in the TMI Unit 1 SBLOCA analysis (106.5% of design flow) is inconsistent with the RCS flow rate used in the departure-from-nucleate (DNB) analysis (104.5% of design flow). Additionally, a lower RCS flow rate was used in the LBLOCA analysis (102% of design flow) than that in the at-power minimum DNB analysis (104.5% of design flow). LOCA analyses performed using the AREVA LOCA evaluation model (EM) BAW-10192P-A, which is the applicable LOCA EM for TMI, Unit 1, are required to use the RCS flow rate that is used in the at-power, minimum DNB analysis. The SBLOCA and LBLOCA RCS flow rate inconsistency assessments estimated a 0ºF PCT impact. | The 10 CFR 50.46 Report dated May 8, 2015, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA. | ||
: 13. Current LOCA Model Assessment | |||
Reactor Coolant System (RCS) flow rate used in the TMI Unit 1 SBLOCA analysis (106.5% of design flow) is inconsistent with the RCS flow rate used in the departure-from-nucleate (DNB) analysis (104.5% of design flow). Additionally, a lower RCS flow rate was used in the LBLOCA analysis (102% of design flow) than that in the at-power minimum DNB analysis (104.5% of design flow). | |||
LOCA analyses performed using the AREVA LOCA evaluation model (EM) BAW-10192P-A, which is the applicable LOCA EM for TMI, Unit 1, are required to use the RCS flow rate that is used in the at-power, minimum DNB analysis. The SBLOCA and LBLOCA RCS flow rate inconsistency assessments estimated a 0ºF PCT impact. | |||
Exelon Generation TMl-16-046 May 6, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 | Exelon Generation TMl-16-046 May 6, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 | ||
==Subject:== | ==Subject:== | ||
Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 10 CFR 50.46 Annual Report 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.46 | Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 10 CFR 50.46 Annual Report 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.46 | ||
==References:== | ==References:== | ||
: 1) Letter from James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report," dated May 8, 2015 2) Letter from Russell K. Cox (AREVA INC.) to John Massari (Exelon Generation Company, LLC), "Transmittal of Input for TMI 50.46 Report for 2015," dated March 29, 2016 -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). The evaluation model (EM) application error regarding reactor coolant system (RCS) flow rate considered in the LOCA analyses resulted in a 0°F PCT impact and is discussed in Attachment 2, Section 13, Current LOCA Model Assessment. No other ECCS-related changes or modifications have occurred at TMI, Unit 1, that affect the assumptions of the ECCS system. | : 1) Letter from James Barstow (Exelon Generation | ||
U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Report May 6, 2016 Page 2 Two attachments are included with this letter that provide the current TMI, Unit 1, 1 O 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. No new regulatory commitments are established in this submittal. If any additional information is needed, please contact Frank Mascitelli at (610) 765-5512. 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 | : Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report," | ||
dated May 8, 2015 2) Letter from Russell K. Cox (AREVA INC.) to John Massari (Exelon Generation | |||
: Company, LLC), "Transmittal of Input for TMI 50.46 Report for 2015," dated March 29, 2016 -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). The evaluation model (EM) application error regarding reactor coolant system (RCS) flow rate considered in the LOCA analyses resulted in a 0°F PCT impact and is discussed in Attachment 2, Section 13, Current LOCA Model Assessment. | |||
No other ECCS-related changes or modifications have occurred at TMI, Unit 1, that affect the assumptions of the ECCS system. | |||
U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Report May 6, 2016 Page 2 Two attachments are included with this letter that provide the current TMI, Unit 1, 1 O 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. | |||
No new regulatory commitments are established in this submittal. | |||
If any additional information is needed, please contact Frank Mascitelli at (610) 765-5512. | |||
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 | |||
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet | Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 1 of 2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA) REPORT REVISION DATE: 05/06/2016 CURRENT OPERATING CYCLE: 21 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 ft 2 Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT) PCT = 1444.0ºF MARGIN ALLOCATION | |||
Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated September 7, 2010 (See Note 5) PCT = 225ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF | Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated September 7, 2010 (See Note 5) PCT = 225ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF | ||
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 | Reactor Coolant System Flow Inconsistency (See Note 13) PCT = 0ºF Total PCT change from current assessments PCT = 0ºF Cumulative PCT change from current assessments lPCTl = 0ºF 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 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 2 of 2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA) REPORT REVISION DATE: 05/06/2016 CURRENT OPERATING CYCLE: 21 Evaluation Model: BWNT 2 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 = 1890ºF MARGIN ALLOCATION | |||
Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2008 (See Note 2) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = +18ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF Reactor Coolant System Flow Inconsistency (See Note 13) PCT = 0ºF Total PCT change from current assessments PCT = 0ºF Cumulative PCT change from current assessments lPCTl = 0ºF | |||
2 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 6, 2015 Assessment Notes Page 1 of 3 | |||
: 1. Prior LOCA Model Assessment | : 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 0ºF 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 0ºF impact for LBLOCA and SBLOCA 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 0ºF PCT change. Reported change included the impact of an energy deposition factor error which resulted in a LBLOCA PCT impact of 0ºF. 3. 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. 4. 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 0ºF 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 0ºF impact for LBLOCA and SBLOCA 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 0ºF PCT change. Reported change included the impact of an energy deposition factor error which resulted in a LBLOCA PCT impact of 0ºF. | |||
: 3. 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. | |||
: 4. 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 1444ºF 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). | |||
: 5. 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 225ºF. 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 SBLOCA or LBLOCA. | |||
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 | Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 2 of 3 | ||
: 7. Prior LOCA Model Assessment | : 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 0ºF PCT impact for both SBLOCA and LBLOCA. 8. 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 0ºF PCT impact for both SBLOCA and LBLOCA. | ||
All previous PCT assessments that are not applicable to Mark-B-HTP fuel have been removed. | : 8. Prior LOCA Model Assessment 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 1890ºF in accordance with our referenced calculation (86-9111507-000). | ||
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. Prior LOCA Model Assessment The 10 CFR 50.46 Report dated May 9, 2014, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA. 11. Prior LOCA Model Assessment The 30-day 10 CFR 50.46 Report dated December 22, 2014, reported a significant error due to thermal conductivity degradation (TCD) based on insufficient LOCA fuel temperature inputs in TACO-3/GDTACO computer codes. Correction of the TCD modeling in TACO-3/GDTACO results in a conservative increase of 393ºF in peak cladding temperature (PCT) for LBLOCA. The SBLOCA analyses are not sensitive to initial fuel temperature and therefore have an estimated PCT impact of 0ºF. | |||
All previous PCT assessments that are not applicable to Mark-B-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. Prior LOCA Model Assessment | |||
The 10 CFR 50.46 Report dated May 9, 2014, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA. | |||
: 11. Prior LOCA Model Assessment | |||
The 30-day 10 CFR 50.46 Report dated December 22, 2014, reported a significant error due to thermal conductivity degradation (TCD) based on insufficient LOCA fuel temperature inputs in TACO-3/GDTACO computer codes. Correction of the TCD modeling in TACO-3/GDTACO results in a conservative increase of 393ºF in peak cladding temperature (PCT) for LBLOCA. The SBLOCA analyses are not sensitive to initial fuel temperature and therefore have an | |||
estimated PCT impact of 0ºF. | |||
Additionally, TMI has implemented a 2 kw/ft penalty to LHR limits in Cycle 20 (10/20/14). The penalty has been applied through more restrictive operational imbalance limits and results in a reduction of PCT by 375ºF for LBLOCA. | |||
The overall cumulative impact for the error and the design input change is 0ºF for SBLOCA and 18ºF for LBLOCA. | The overall cumulative impact for the error and the design input change is 0ºF for SBLOCA and 18ºF for LBLOCA. | ||
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 | Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 3 of 3 | ||
: 12. Prior LOCA Model Assessment | : 12. Prior LOCA Model Assessment | ||
The 10 CFR 50.46 Report dated May 8, 2015, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA. 13. Current LOCA Model Assessment Reactor Coolant System (RCS) flow rate used in the TMI Unit 1 SBLOCA analysis (106.5% of design flow) is inconsistent with the RCS flow rate used in the departure-from-nucleate (DNB) analysis (104.5% of design flow). Additionally, a lower RCS flow rate was used in the LBLOCA analysis (102% of design flow) than that in the at-power minimum DNB analysis (104.5% of design flow). LOCA analyses performed using the AREVA LOCA evaluation model (EM) BAW-10192P-A, which is the applicable LOCA EM for TMI, Unit 1, are required to use the RCS flow rate that is used in the at-power, minimum DNB analysis. The SBLOCA and LBLOCA RCS flow rate inconsistency assessments estimated a 0ºF PCT impact.}} | The 10 CFR 50.46 Report dated May 8, 2015, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA. | ||
: 13. Current LOCA Model Assessment | |||
Reactor Coolant System (RCS) flow rate used in the TMI Unit 1 SBLOCA analysis (106.5% of design flow) is inconsistent with the RCS flow rate used in the departure-from-nucleate (DNB) analysis (104.5% of design flow). Additionally, a lower RCS flow rate was used in the LBLOCA analysis (102% of design flow) than that in the at-power minimum DNB analysis (104.5% of design flow). | |||
LOCA analyses performed using the AREVA LOCA evaluation model (EM) BAW-10192P-A, which is the applicable LOCA EM for TMI, Unit 1, are required to use the RCS flow rate that is used in the at-power, minimum DNB analysis. The SBLOCA and LBLOCA RCS flow rate inconsistency assessments estimated a 0ºF PCT impact.}} | |||
Revision as of 08:45, 30 June 2018
| ML16130A065 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 05/06/2016 |
| From: | Jim Barstow Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TMl-16-046 | |
| Download: ML16130A065 (9) | |
Text
Exelon Generation TMl-16-046 May 6, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 10 CFR 50.46 Annual Report 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.46
References:
- 1) Letter from James Barstow (Exelon Generation
- Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report,"
dated May 8, 2015 2) Letter from Russell K. Cox (AREVA INC.) to John Massari (Exelon Generation
- Company, LLC), "Transmittal of Input for TMI 50.46 Report for 2015," dated March 29, 2016 -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). The evaluation model (EM) application error regarding reactor coolant system (RCS) flow rate considered in the LOCA analyses resulted in a 0°F PCT impact and is discussed in Attachment 2, Section 13, Current LOCA Model Assessment.
No other ECCS-related changes or modifications have occurred at TMI, Unit 1, that affect the assumptions of the ECCS system.
U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Report May 6, 2016 Page 2 Two attachments are included with this letter that provide the current TMI, Unit 1, 1 O 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.
No new regulatory commitments are established in this submittal.
If any additional information is needed, please contact Frank Mascitelli at (610) 765-5512.
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
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 1 of 2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA) REPORT REVISION DATE: 05/06/2016 CURRENT OPERATING CYCLE: 21 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 ft 2 Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT) PCT = 1444.0ºF MARGIN ALLOCATION
Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated September 7, 2010 (See Note 5) PCT = 225ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF
Reactor Coolant System Flow Inconsistency (See Note 13) PCT = 0ºF Total PCT change from current assessments PCT = 0ºF Cumulative PCT change from current assessments lPCTl = 0ºF 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 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 2 of 2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA) REPORT REVISION DATE: 05/06/2016 CURRENT OPERATING CYCLE: 21 Evaluation Model: BWNT 2 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 = 1890ºF MARGIN ALLOCATION
Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2008 (See Note 2) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = +18ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF Reactor Coolant System Flow Inconsistency (See Note 13) PCT = 0ºF Total PCT change from current assessments PCT = 0ºF Cumulative PCT change from current assessments lPCTl = 0ºF
2 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 6, 2015 Assessment Notes Page 1 of 3
- 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 0ºF 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 0ºF impact for LBLOCA and SBLOCA 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 0ºF PCT change. Reported change included the impact of an energy deposition factor error which resulted in a LBLOCA PCT impact of 0ºF.
- 3. 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.
- 4. 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 1444ºF 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).
- 5. 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 225ºF. 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 SBLOCA or LBLOCA.
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 2 of 3
- 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 0ºF PCT impact for both SBLOCA and LBLOCA.
- 8. Prior LOCA Model Assessment 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 1890ºF 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.
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. Prior LOCA Model Assessment
The 10 CFR 50.46 Report dated May 9, 2014, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA.
- 11. Prior LOCA Model Assessment
The 30-day 10 CFR 50.46 Report dated December 22, 2014, reported a significant error due to thermal conductivity degradation (TCD) based on insufficient LOCA fuel temperature inputs in TACO-3/GDTACO computer codes. Correction of the TCD modeling in TACO-3/GDTACO results in a conservative increase of 393ºF in peak cladding temperature (PCT) for LBLOCA. The SBLOCA analyses are not sensitive to initial fuel temperature and therefore have an
estimated PCT impact of 0ºF.
Additionally, TMI has implemented a 2 kw/ft penalty to LHR limits in Cycle 20 (10/20/14). The penalty has been applied through more restrictive operational imbalance limits and results in a reduction of PCT by 375ºF for LBLOCA.
The overall cumulative impact for the error and the design input change is 0ºF for SBLOCA and 18ºF for LBLOCA.
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 3 of 3
- 12. Prior LOCA Model Assessment
The 10 CFR 50.46 Report dated May 8, 2015, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA.
- 13. Current LOCA Model Assessment
Reactor Coolant System (RCS) flow rate used in the TMI Unit 1 SBLOCA analysis (106.5% of design flow) is inconsistent with the RCS flow rate used in the departure-from-nucleate (DNB) analysis (104.5% of design flow). Additionally, a lower RCS flow rate was used in the LBLOCA analysis (102% of design flow) than that in the at-power minimum DNB analysis (104.5% of design flow).
LOCA analyses performed using the AREVA LOCA evaluation model (EM) BAW-10192P-A, which is the applicable LOCA EM for TMI, Unit 1, are required to use the RCS flow rate that is used in the at-power, minimum DNB analysis. The SBLOCA and LBLOCA RCS flow rate inconsistency assessments estimated a 0ºF PCT impact.
Exelon Generation TMl-16-046 May 6, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 10 CFR 50.46 Annual Report 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.46
References:
- 1) Letter from James Barstow (Exelon Generation
- Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report,"
dated May 8, 2015 2) Letter from Russell K. Cox (AREVA INC.) to John Massari (Exelon Generation
- Company, LLC), "Transmittal of Input for TMI 50.46 Report for 2015," dated March 29, 2016 -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). The evaluation model (EM) application error regarding reactor coolant system (RCS) flow rate considered in the LOCA analyses resulted in a 0°F PCT impact and is discussed in Attachment 2, Section 13, Current LOCA Model Assessment.
No other ECCS-related changes or modifications have occurred at TMI, Unit 1, that affect the assumptions of the ECCS system.
U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Report May 6, 2016 Page 2 Two attachments are included with this letter that provide the current TMI, Unit 1, 1 O 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.
No new regulatory commitments are established in this submittal.
If any additional information is needed, please contact Frank Mascitelli at (610) 765-5512.
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
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 1 of 2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA) REPORT REVISION DATE: 05/06/2016 CURRENT OPERATING CYCLE: 21 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 ft 2 Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT) PCT = 1444.0ºF MARGIN ALLOCATION
Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated September 7, 2010 (See Note 5) PCT = 225ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF
Reactor Coolant System Flow Inconsistency (See Note 13) PCT = 0ºF Total PCT change from current assessments PCT = 0ºF Cumulative PCT change from current assessments lPCTl = 0ºF 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 6, 2016 Attachment 1 Peak Cladding Temperature Rack-Up Sheet Page 2 of 2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA) REPORT REVISION DATE: 05/06/2016 CURRENT OPERATING CYCLE: 21 Evaluation Model: BWNT 2 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 = 1890ºF MARGIN ALLOCATION
Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2008 (See Note 2) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) PCT = 0ºF Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) PCT = 0ºF 30-Day 10 CFR 50.46 Report dated December 22, 2014 (See Note 11) PCT = +18ºF Annual 10 CFR 50.46 Report dated May 8, 2015 (Note 12) PCT = 0ºF Reactor Coolant System Flow Inconsistency (See Note 13) PCT = 0ºF Total PCT change from current assessments PCT = 0ºF Cumulative PCT change from current assessments lPCTl = 0ºF
2 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 6, 2015 Assessment Notes Page 1 of 3
- 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 0ºF 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 0ºF impact for LBLOCA and SBLOCA 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 0ºF PCT change. Reported change included the impact of an energy deposition factor error which resulted in a LBLOCA PCT impact of 0ºF.
- 3. 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.
- 4. 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 1444ºF 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).
- 5. 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 225ºF. 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 SBLOCA or LBLOCA.
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 2 of 3
- 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 0ºF PCT impact for both SBLOCA and LBLOCA.
- 8. Prior LOCA Model Assessment 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 1890ºF 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.
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. Prior LOCA Model Assessment
The 10 CFR 50.46 Report dated May 9, 2014, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA.
- 11. Prior LOCA Model Assessment
The 30-day 10 CFR 50.46 Report dated December 22, 2014, reported a significant error due to thermal conductivity degradation (TCD) based on insufficient LOCA fuel temperature inputs in TACO-3/GDTACO computer codes. Correction of the TCD modeling in TACO-3/GDTACO results in a conservative increase of 393ºF in peak cladding temperature (PCT) for LBLOCA. The SBLOCA analyses are not sensitive to initial fuel temperature and therefore have an
estimated PCT impact of 0ºF.
Additionally, TMI has implemented a 2 kw/ft penalty to LHR limits in Cycle 20 (10/20/14). The penalty has been applied through more restrictive operational imbalance limits and results in a reduction of PCT by 375ºF for LBLOCA.
The overall cumulative impact for the error and the design input change is 0ºF for SBLOCA and 18ºF for LBLOCA.
Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 6, 2015 Assessment Notes Page 3 of 3
- 12. Prior LOCA Model Assessment
The 10 CFR 50.46 Report dated May 8, 2015, reported no evaluations or PCT penalties for either SBLOCA or LBLOCA.
- 13. Current LOCA Model Assessment
Reactor Coolant System (RCS) flow rate used in the TMI Unit 1 SBLOCA analysis (106.5% of design flow) is inconsistent with the RCS flow rate used in the departure-from-nucleate (DNB) analysis (104.5% of design flow). Additionally, a lower RCS flow rate was used in the LBLOCA analysis (102% of design flow) than that in the at-power minimum DNB analysis (104.5% of design flow).
LOCA analyses performed using the AREVA LOCA evaluation model (EM) BAW-10192P-A, which is the applicable LOCA EM for TMI, Unit 1, are required to use the RCS flow rate that is used in the at-power, minimum DNB analysis. The SBLOCA and LBLOCA RCS flow rate inconsistency assessments estimated a 0ºF PCT impact.