ML071360337
| ML071360337 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 05/16/2007 |
| From: | David Helker AmerGen Energy Co |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| 5928-07-20099 | |
| Download: ML071360337 (7) | |
Text
1 OCFR50.46 May 16,2007 5928-07-20099 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island, Unit 1 (TMI Unit
- 1) Facility Operating License No. DPR-50 NRC Docket No. 50-289
Subject:
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" In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), AmerGen Energy Company, LLC (AmerGen), is submitting the annual report of the Emergency Core Cooling System (ECCS) Evaluation Model changes and errors for TMI Unit 1. Attachment 1, "Peak Cladding Temperature Rack-Up Sheets," provides updated information regarding the peak cladding temperature (PCT) for the limiting small break and large break loss-of-coolant accident (LOCA) analyses evaluations for TMI Unit 1. 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 David J. Distel at (610) 765-5517.
Respectfully, A David P. Helker Manager - Licensing Attachments:
- 1) Peak Cladding Temperature Rack-Up Sheets 2) Assessment Notes 5928-07-20099 May 16,2007 Page 2 cc: S. J. Collins, USNRC Administrator, Region I P. Bamford, USNRC Project Manager, TMI Unit 1 D. M. Kern, USNRC Senior Resident Inspector, TMI Unit 1 File No. 00068 Attachment 1 TMI Unit 1 Docket No. 50-289 License No. DPR-50 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 Peak Cladding Temperature Rack-Up Sheets 5928-07-20099 Attachment 1 Page 1 of 2 lOCFR50.46 report dated June 6,2002 (see note
- 3) 1 OCFR50.46 report dated June 19,2003 (see note
- 4) 1 OCFR50.46 report dated June 1, 2004 (see note
- 5) 10CFR50.46 report dated May 16, 2005 (see note 6) 10CFR50.46 report dated May 09, 2006 (see note
- 7) Attachment 1 APCT = 0°F APCT = 0°F APCT = 0°F APCT = 0°F APCT = 0°F Three Mile Island Unit I 10CFR50.46 Report Peak Cladding Temperature Rack-up Sheets PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE:
16 Three Mile Island Unit 1 Small Break Loss of Coolant Accident (SBLOCA) 031 301 07 ANALYSIS OF RECORD (AOR)
Evaluation Model: BWNT ' Calculation:
Framatome ANP 86-501 1294-00, March 2001 Fuel: Mark-B9, Mark-B12 Limiting Fuel Type: Mark-B12 Limiting Single Failure: Loss of One Train of ECCS Steam Generator Tube Pluqging (SGTP): 20% Limiting Break Size:
0.05 ft Break in Cold Leg Pump Discharge Piping Reference Peak Cladding Temperature (PCT) PCT
= 1454.0"F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS NET PCT PCT = 1454.0"F B. CURRENT LOCA MODEL ASSESSMENTS I Impact of Heat Sink on Safety Analyses (see note
5928-07-20099 Attachment 1 Page 2 of 2 Fuel Type: Limiting Fuel Type: Reference PCT Attachment 1 Mark-B9 Mark-B12 Mark-B9 Mark-B12 2083°F 1 989°F Three Mile Island Unit 1 10CFR50.46 Report Peak Cladding Temperature Rack-up Sheets 10CFR50.46 report dated June 5,2000 (see note 1) 1 OCFR50.46 report dated June 11,2001 (see note
- 2) 1 OCFR50.46 report dated June 6,2002 (see note
- 3) 10CFR50.46 report dated June 19,2003 (see note
- 4) 1 OCFR50.46 report dated June 1,2004 (see note
- 5) 10CFR50.46 report dated May 16, 2005 (see note
- 6) 10CFR50.46 report dated May 09, 2006 (see note
- 7) PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE:
03/30/07 CURRENT OPERATING CYCLE:
16 Three Mile Island Unit 1 Larae Break Loss of Coolant Accident (LBLOCA)
APCT = 0°F NIA APCT = 0°F NIA APCT = 0°F APCT = 0°F APCT = -25°F APCT= -35°F APCT = 0°F APCT = 0°F APCT = 0°F APCT = 0°F APCT = 0°F APCT = 0°F Evaluation Model: BWNT2 Calculation: Framatome ANP 86-5002073-02, July 1 999 (Mark-B9)
Framatome ANP 86-501 1294-00, March 2001 (Mark-B12) Limiting Single Failure: Loss of One Train of ECCS Steam Generator Tube Plugging (SGTP): 20% Limiting Break Size: Guillotine Break in Cold Leg Pump Discharge Piping MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS NET PCT PCT = 2058°F 1954°F B. CURRENT LOCA MODEL ASSESSMENTS I Impact of Heat Sink on Safety Analyses (see note
- 8) 1 APCT = 0°F IAPCT=O"F 1 NET PCT PCT = 2058°F 1954°F
- The BWNT EM is based on RELAP5/MOD2-B&W.
Attachment 2 TMI Unit 1 Docket No. 50-289 License No. DPR-50 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 Assessment Notes 5928-07-20099 Attachment 2 Page 1 of 1 Attachment 2 Three Mile Island Unit 1 lOCFR50.46 Report Assessment Notes 1. Prior LOCA Model Assessment The 1 OCFR50.46 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 lOCFR50.46 report dated June 11, 2001 reported evaluations for LBLOCA and SBLOCA model changes which resulted in 0°F PCT change.
- 3. Prior LOCA Model Assessment The 1 OCFR50.46 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°F 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 1 OCFR50.46 report dated June 19,2003 reported evaluation for LBLOCA model change, which resulted in 0°F PCT change.
SBLOCA was not impacted.
- 5. Prior LOCA Model Assessment The 1 OCFR50.46 report dated June 1,2004 reported evaluation for LBLOCA and SBLOCA model changes which resulted in 0°F 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 lOCFR50.46 report dated May 16, 2005 reported evaluations for LBLOCA model changes which resulted in a 0°F PCT change. LOCA oxygen/hydrogen recombination was considered and the PCT effect was determined to be 0°F. SBLOCA was not impacted.
- 7. Prior LOCA Model Assessment The 10CFR50.46 report dated May 09, 2006 reported evaluations for LOCA model changes which resulted in a 0°F PCT change. Reported changes included operation with no APSR pull and batch 18 fuel design changes. These were applicable for SBLOCA and LBLOCA. 8. Impact of Heat Sink on Safety Analyses The 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 0°F impact for both LBLOCA and SBLOCA PCTs.