Text

Submittal of 30-Day 10 CFR 50.46 Report
	ML14356A342
Person / Time
Site:	Three Mile Island
Issue date:	12/22/2014
From:	Jim Barstow; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TMl-14-142
	Download: ML14356A342 (11)
	v • d • e

10 CFR 50.46 TMl-14-142 December 22, 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 Nos. 50-289 Subject 10 CFR 50.46 30-Day Report

References:

1) Letter from James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report," dated May 9, 2014(ML14129A206)

2) Notification Letter FAB 14-00624 from Russell K. Cox (AREVA) to Robert Jaffa (Exelon Generation Company, LLC), "Evaluation of TMl-1 for Condition Report 2014-6492 for Potential Reporting Under 10 CFR 50.46," dated November 25, 2014

3) Notification Letter FAB 14-00658 from Russell K. Cox (AREVA) to Robert Jaffa (Exelon Generation Company, LLC), "Revised Evaluation of TMl-1 for Condition Report 2014-6492 for Potential Reporting Under 10 CFR 50.46," dated December 9, 2014 The purpose of this letter is to submit a 30-Day 10 CFR 50.46 Report for Three Mile Island Nuclear Station, Unit 1 (TMI). The most recent annual 50.46 Report for TMI (Reference 1) provided the cumulative Peak Cladding Temperature (PCT) errors. This report provided the results of the most recent Large Break Loss-of-Coolant Accident (LBLOCA) analysis.

The peak clad temperature (PCT) reported for the LBLOCA was 1890.0°F.

Subsequent to the issuance of Reference 1, AREVA notified Exelon Generation Company, LLC, (Exelon) of a deficiency in AREVA ECCS-LOCA analysis, as documented in AREVA Condition Report 2014-6492 (References 2 and 3). The deficiency is related to the modeling of thermal conductivity in the codes TAC03 and GDTACO, which are currently

U.S. Nuclear Regulatory Commission TMI 10 CFR 50.46 30-Day Report December 2014 Page 2 part of the approved LOCA evaluation model for B&W plants. Investigation of this concern was initiated by the NRC via NRC Information Notice (IN) 2009-23 "Nuclear Fuel Thermal Conductivity Degradation." Recent comparisons of the fuel temperatures from these codes with fuel temperatures from the code COPERNIC2 (a modern fuel performance code that explicitly models fuel thermal conductivity degradation (TCD) as a function of burnup) now indicates that the TAC03 and GDTACO codes thermal conductivity model, augmented by adjustments to compensate for TCD not being modeled, leads to an under prediction of the peak cladding temperature (PCT) during a LBLOCA Therefore, an additional fuel temperature uncertainty has been recommended for TAC03 and GOTACO to properly account for the lack of TCD modeling.

Application of the additional fuel temperature uncertainty in TAC03/GDTACO results in a conservative increase of 393"F in PCT for LBLOCA The SBLOCA analyses are not sensitive to initial fuel temperature and, therefore, have an estimated PCT impact of 0°F.

In order to preserve the existing TMI LOCA analysis of record for PCT, AREVA recommended, and TMI implemented, a 2 kw/ft penalty to LHR limits in Cycle 20 on October 20, 2014. Imposition of this penalty assured that the TCD increase was offset and the resultant PCT for uranium dioxide fuel rods will be equal to or less than the peak cladding temperature prior to the Evaluation Model (EM) correction. However, for Gadolinia fuel rods, the 2 kw/ft effectively accounts for the TCD volume-average fuel temperature increase but does not completely offset the increase for all weight percentages. Therefore, this change provides a PCT benefit of 375°F. The penalty has been applied through more restrictive operational imbalance limits.

The cumulative PCT impact of the TCD error and the change to the LOCA LHR limits is 18°F and the absolute value cumulative impact for the error and the design input change is 768°F. As discussed in 10 CFR 50.46(a)(3)(i), this 30-Day Report is required because the absolute magnitude of the temperature change is greater than 50°F.

AREVA's recommendation to Exelon with respect to a LBLOCA reanalysis for TMI is to perform a full LBLOCA reanalysis with the revised EM that uses a COPERNIC2 based TCD uncertainty increase to the TAC03 and GDTACO inputs at MOL and EOL The recommended reanalysis consists of formal analysis with the Enhanced Once-Through Steam Generators (EOTSG) with a full core of Mark-B-HTP fuel. The reanalysis will address the significant EM error corrections to cover the ECCS bypass error correction and column weldment modeling changes.

Three attachments are included with this letter that provide the TMI 10 CFR 50.46 status. ("Peak Cladding Temperature Rack-Up Sheet") provides updated information regarding the PCT for the limiting LOCA analysis evaluations for TMI. Attachment 2

("Assessment Notes") contains a detailed description for each change or error reported. contains the commitment to perform LBLOCA reanalysis.

The following commitment is made by this letter:

U.S. Nuclear Regulatory Commission TMI 10 CFR 50.46 30-Day Report December 2014 Page 3 Exelon will perform a full LBLOCA reanalysis for TMI Unit 1 by March 31, 2017. The effects of fuel pellet thermal conductivity degradation will be accounted for by use of a fuel temperature uncertainty adjustment factor based on COPERNIC2.

If you have any questions, please contact Frank Mascitelli at 610-765-5512.

Respectfully, James Barstow Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Peak Cladding Temperature Rack-Up Sheet (TMI)

2) Assessment Notes (TMI)

3) Summary of Regulatory Commitments cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, TMI USNRC Project Manager, TMI R R Janati, Bureau of Radiation Protection

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 as of December 22, 2014 Peak Cladding Temperature Rack-Up Sheet Three Mile Island Nuclear Station, Unit 1

Three Mile Island Unit 1 10 CFR 50.46 Report Attachment 1 Peak Cladding Temperature Rack-up Sheet Page 1of2 PLANT NAME: Three Mile Island Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)

REPORT REVISION DATE: 12/22/2014 CURRENT OPERATING CYCLE: 20 ANALYSIS OF RECORD (AOR)

Evaluation Model: BWNT 1 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: 0.07 ft 2 Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT) PCT= 1444.0"F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS Annual 10 CFR 50.46 Report dated May 16, 2007 (See Note 1) LiPCT = 0 °F Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) LiPCT = 0 °F Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) ~PCT= 0 °F 30-Day 10 CFR 50.46 Report dated September 7, 2010 (See ~PCT= 225°F Note 5)

Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) LiPCT = 0 °F 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) LiPCT = 0 °F Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) LiPCT = 0 °F Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) LiPCT = 0 °F Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) LiPCT = 0 °F NET PCT PCT= 1669.0°F B. CURRENT LOCA MODEL ASSESSMENTS Thermal Conductivity Degradation Error (See Note 11) ~PCT= 0°F Total PCT change from current assessments 2: ~PCT= 0°F Cumulative PCT change from current assessments 2: l~PCTI = 0°F NET PCT PCT = 1669.0°F 1

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

Three Mile Island Unit 1 10 CFR 50.46 Report Attachment 1 Peak Cladding Temperature Rack*up Sheet Page 2 of 2 PLANT NAME:

ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)

REPORT REVISION DATE:

CURRENT OPERATING CYCLE:

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) L'.PCT = 0 °F Annual 10 CFR 50.46 Report dated May 15, 2008 (See Note 2) L'.PCT = 0 °F Annual 10 CFR 50.46 Report dated May 15, 2009 (See Note 3) L'.PCT = 0 °F Annual 10 CFR 50.46 Report dated May 14, 2010 (See Note 4) L'.PCT = 0 °F Annual 10 CFR 50.46 Report dated May 13, 2011 (See Note 6) LiPCT = 0 °F 30-Day 10 CFR 50.46 Report dated March 21, 2012 (See Note 7) LiPCT = 0 °F Annual 10 CFR 50.46 Report dated May 11, 2012 (See Note 8) LiPCT = 0 °F Annual 10 CFR 50.46 Report dated May 10, 2013 (See Note 9) LiPCT = 0 °F Annual 10 CFR 50.46 Report dated May 9, 2014 (See Note 10) LiPCT = 0 °F NET PCT PCT= 1890°F B. CURRENT LOCA MODEL ASSESSMENTS Thermal Conductivity Degradation Error (See Note 11) L'.PCT = +393°F LHR limit reductions of 2 kW/ft (See Note 11) =

L'.PCT -375°F Total PCT change from current assessments L L'.PCT = +18°F Cumulative PCT change from current assessments L IL'.PCTI =

768°F NET PCT PCT= 1908°F 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 Assessment Notes Three Mile Island Nuclear Station, Unit 1

Three Mile Island Unit 1 10 CFR 50.46 Report Attachment 2 Assessment Notes Page 1of2

1. Prior LOCA Model Assessment The 10 CFR50.46 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 GSl-191 related reactor building sump screen replacement The evaluation resulted in 0 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 0 °F PCT change. Reported change included the impact of an energy deposition factor error which resulted in a LBLOCA PCT impact of 0

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-89 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-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 EDW 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 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 SBLOCA or LBLOCA

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

Three Mile Island Unit 110 CFR 60.46 Report Attachment 2 Assessment Notes Page 2 of 2

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 S8LOCA and L8LOCA The limiting PCT for L8LOCA 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 S8LOCA or L8LOCA

9. Prior LOCA Model Assessment The 10 CFR 50.46 Report dated May 10, 2013, reported no evaluations or PCT penalties for either S8LOCA or L8LOCA

10. Prior LOCA Model Assessment The 10 CFR 50.46 Report dated May 9, 2014, reported no evaluations or PCT penalties for either S8LOCA or L8LOCA

11. Current LOCA Model Assessment A significant error was discovered where the existing adjustments to the methodology and increases in the LOCA fuel temperature inputs that compensate for TCD modeled in TAC0-3/GDTACO computer codes used from TMl-1 LOCA analyses have been found to be insufficient when compared to recent COPERNIC2 results. Correction of the TCD modeling in TAC0-3/GDTACO results in a conservative increase of 393"F in peak cladding temperature (PCT) for L8LOCA The S8LOCA analyses are not sensitive to initial fuel temperature and, therefore, have an estimated PCT impact of OF.

In order to maintain sufficient LOCA PCT margin, AREVA recommended, and TMI has implemented, a 2 kw/ft penalty to LHR limits in Cycle 20 (October 20, 2014). The penalty has been applied through more restrictive operational imbalance limits and results in a reduction of PCT by 375°F.

The overall cumulative impact for the error and the design input change is 0°F for S8LOCA and 18°F for L8LOCA

ATTACHMENT 3 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 Summary of Regulatory Commitments Three Mile Island Nuclear Station, Unit 1

ATTACHMENT 3

SUMMARY

OF REGULATORY COMMITMENTS The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)

COMMITTED COMMITMENT TYPE COMMITMENT DATE OR ONE~TIME ACTION PROGRAMMATIC "OUTAGE" (Yes/No) (Yes/No)

Exelon will perform a full LBLOCA March 31, 2017 Yes No reanalysis for TMI Unit 1 by March 31, 2017. The effects of fuel pellet thermal conductivity degradation will be accounted for by use of a fuel temperature uncertainty adjustment factor based on COPERNIC2.