ML081410457

CFR 50.46 Report

ML081410457
Person / Time
Site:	Three Mile Island
Issue date:	05/15/2008
From:	David Helker AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
5928-08-20103
Download: ML081410457 (8)

Text

AmerGen Energy Company, LLC www.exeloncorp.com AmerGt An Exelon Company M

20o Exelon Way Kennett Square, PA 19348 10 CFR 50.46 May 15, 2008 5928-08-20103 U.S. Nuclear Regulatory Commission Attn: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Three Mile Island, Unit 1 Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

10 CFR 50.46 Report

Reference:

1) Letter from David P. Helker (AmerGen Energy Company, LLC) to U. S.

Nuclear Regulatory Commission, "Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Required by 10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,'" dated May 16, 2007 The purpose of this letter is to submit the 10 CFR 50.46 reporting information for Three Mile Island (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.

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 Small Break Loss of Coolant Accident (SBLOCA) and Large Break Loss of Coolant Accident (LBLOCA) analysis. 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 Tom Loomis at (610) 765-5510.

Respectfully, David P. Helker Manager - Licensing Attachments: 1) Peak Cladding Temperature Rack-Up Sheets , t

2) Assessment Notes

/o4

5928-08-20103 May 15, 2008 Page 2 cc: S. J. Collins, USNRC Administrator, Region I P. J. Bamford, USNRC Project Manager, TMI Unit 1 D. M. Kern, USNRC Senior Resident Inspector, TMI Unit 1 File No. 00068

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 14, 2008 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 14, 2008 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: 05/14/08 CURRENT OPERATING CYCLE: 17 ANALYSIS OF RECORD (AOR)

Evaluation Model: BWNT 1 Calculation: Framatome ANP 86-5011294-00, March 2001 AREVA NP, 86-9049246-000, June 2007 Fuel: Mark-B9, Mark-B12, Mark-B-HTP Limiting Fuel Type: Mark-B9 Limiting Single Failure: Loss of One Train of ECCS Limiting Break Size and Location: 0.05 ft2 Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT) PCT= 1454OF MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated June 6, 2002 (see note 3) APCT = 0 °F 10 CFR 50.46 report dated June 19, 2003 (see note 4) APCT = 0 OF 10 CFR 50.46 report dated June 1, 2004 (see note 5) APCT = 0 OF 10 CFR 50.46 report dated May 16, 2005 (see note 6) APCT = 0 OF 10 CFR 50.46 report dated May 9, 2006 (see note 7) APCT = 0 OF 10 CFR 50.46 report dated May 16, 2007 (see note 8) APCT = 0 °F NET PCT PCT = 1454 OF B. CURRENT LOCA MODEL ASSESSMENTS I RELAP5 bypass pin pressure calculation limitation (see note 9) 1 APCT = 0 °F NET PCT PCT = 1454 °F

'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 14, 2008 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: 05/14/08 CURRENT OPERATING CYCLE: 17 ANALYSIS OF RECORD (AOR) 2 Evaluation Model: BWNT Calculation: Framatome ANP 86-5002073-02, July 1999 (Mark-B9)

Framatome ANP 86-5011294-00, March 2001 (Mark-B12)

AREVA NP 86-9049246-000, June 2007 (Mark-B-HTP)

Fuel: Mark-B9, Mark-B12, Mark-B-HTP Limiting Fuel Type: Mark-B9 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= 2083 OF MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated June 5, 2000 (see note 1) APCT = 0 OF 10 CFR 50.46 report dated June 11, 2001 (see note 2) APCT = 0 OF 10 CFR 50.46 report dated June 6, 2002 (see note 3) APCT = 0 OF 10 CFR 50.46 report dated June 19, 2003 (see note 4) APCT = 0 'F 10 CFR 50.46 report dated June 1, 2004 (see note 5) APCT = -25 °F 10 CFR 50.46 report dated May 16, 2005 (see note 6) APCT = 0 OF 10 CFR 50.46 report dated May 9, 2006 (see note 7) APCT = 0 OF 10 CFR 50.46 report dated May 16, 2007 (see note 8) APCT = 0 OF NET PCT PCT = 2058 OF B. CURRENT LOCA MODEL ASSESSMENTS I EDF application error correction (see note 10) 1 APCT = 0 °F I NET PCT PCT = 2058 °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 Assessments Assessments as of May 14, 2008 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 14, 2008 Attachment 2 Assessment Notes Page 1 of 2

1. Prior LOCA Model Assessment The 10 CFR 50.46 report dated June 5, 2000 reported new LBLOCA and SBLOCA analyses to support operations at 20% steam generator tube plugging conditions for Mark-B9 fuel.

2. Prior LOCA Model Assessment The 10 CFR 50.46 report dated June 11, 2001 reported evaluations for LBLOCA and SBLOCA model changes which resulted in 0 OF PCT change.

3. Prior LOCA Model Assessment The 10 CFR 50.46 report dated June 6, 2002 reported new LBLOCA analyses to support operations with Mark-B12 fuel. For SBLOCA, an increase in SBLOCA PCT of 42 OF for Mark-B9 fuel was reported due to increase in emergency feedwater temperature. This analysis is applicable to both Mark-B12 fuel and Mark-B9 fuel.

4. Prior LOCA Model Assessment The 10 CFR 50.46 report dated June 19, 2003 reported evaluation for LBLOCA model change which resulted in 0 OF PCT change. SBLOCA was not impacted.

5. Prior LOCA Model Assessment The 10 CFR 50.46 report dated June 1, 2004 reported evaluation for LBLOCA and SBLOCA model changes which resulted in 0 OF PCT change. An error correction in containment pressure input resulted in a reduction in PCT for the LBLOCA analysis.

6. Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 16, 2005 reported evaluations for LBLOCA model changes which resulted in a 0 OF PCT change. LOCA oxygen/hydrogen recombination was considered and the PCT effect was determined to be 0 OF. SBLOCA was not impacted.

7. Prior LOCA Model Assessment The 10 CFR 50.46 report dated May 9, 2006 reported evaluations for LOCA model changes which resulted in a 0 OF PCT change. Reported changes included operation with no APSR pull and batch 18 fuel design changes. These were applicable for SBLOCA and LBLOCA.

8. 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 0 OF PCT change. The reported evaluation considered the effect on the containment pressure response for LOCA due to GS1-191 related reactor building sump screen replacement. The evaluation resulted in 0 OF impact for both LBLOCA and SBLOCA PCTs.

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 14, 2008 Attachment 2 Assessment Notes Page 2 of 2

9. RELAP5 bypass pin pressure calculation limitation An unexpected behavior was observed in the fuel pin internal pressure of a pin channel for demonstration 3025 MWt power uprate SBLOCA analyses. The faulted channel showed no change in the pin pressure; in contrast, the pin pressures in the other channels continued to change as expected.

The code logic of RELAP5/MOD2-B&W was reviewed to determine what had caused the unexpected behavior. It was discovered that the code logic would bypass the pin pressure calculations after the uppermost pin segment in an unruptured pin channel becomes plastic.

The logic flaw did not appreciably alter the results and therefore the PCT impact of this error was determined to be 0 OF.

10. EDF Application Error An error affecting the energy deposition factors (EDF) utilized in LOCA analyses was identified. The source of the error was an incorrect interpretation of gamma energy fractions reported from the ORIGEN2 code. The correction of the gamma energy fractions resulted in an increase in the energy deposited within the fuel during the LBLOCA transient for high burnup and low power conditions (e.g., end-of-life analyses). The effect of this error was evaluated and the PCT impact was determined to be 0 OF.