ML061290333
| ML061290333 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/09/2006 |
| From: | David Helker AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 5928-06-20450 | |
| Download: ML061290333 (8) | |
Text
1 OCFR50.46 May 9,2006 5928-06-20450 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., 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, David P. Helker Manager - Licensing Attachments: 1) Peak Cladding Temperature Rack-Up Sheets
- 2) Assessment Notes
5928-06-20450 May 9,2006 Page 2 cc: S. J. Collins, USNRC Administrator, Region I F. E. Saba, USNRC Project Manager, TMI Unit 1 D. M. Kern, USNRC Senior Resident Inspector, TMI Unit 1 File No. 00068 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
5928-06-20450 Page 1 of 2 10 CFR 50.46 report dated June 6, 2002 (see note 3) 10 CFR 50.46 reDort dated June 19.2003 (see note 4)
APCT = 0 OF APCT = 0 OF Three Mile Island Unit 1 10 CFR 50.46 Report Peak Cladding Temperature Rack-up Sheets 10 CFR 50.46 report dated June 1, 2004 (see note 5) 10 CFR 50.46 report dated May 16,2005 (see note 6)
PLANT NAME:
ECCS EVALUATION MODEL:
REPORT REVISION DATE:
04/25/06 CURRENT OPERATING CYCLE: 16 Three Mile Island Unit 1 Small Break Loss of Coolant Accident (SBLOCAZ APCT = 0 OF APCT = 0 OF ANALYSIS OF RECORD (AOR)
Cycle 15 Fuel Cycle Design Change to leave APSRs in at EOC (see note 7)
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 Plugging (SGTP): 20%
Limiting Break Size: 0.05 ft Break in Cold Leg Pump Discharge Piping APCT =
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
- 6. CURRENT LOCA MODEL ASSESSMENTS Evaluation of Batch 18 Fuel Design Change (see note 8)
I APCT = o OF NET PCT PCT = 1454.0"F
' The BWNT EM is based on FELAPS/MOD2-B&W.
5928-06-20450 Page 2 of 2 Fuel Type:
Limiting Fuel Type:
Reference PCT Mark-B9 Mark-61 2 Mark-69 Mark431 2 2083 OF 1989 OF Three Mile Island Unit 1 10 CFR 50.46 Report Peak Cladding Temperature Rack-up Sheets 10 CFR 50.46 report dated June 5, 2000 (see note 1) 10 CFR 50.46 report dated June 11,2001 (see note 2) 10 CFR 50.46 report dated June 6, 2002 (see note 3) 10 CFR 50.46 report dated June 19,2003 (see note 4) 10 CFR 50.46 report dated May 16, 2005 (see note 6) 10 CFR 50.46 report dated June 1, 2004 (see note 5)
PLANT NAME:
ECCS EVALUATION MODEL:
REPORT REVISION DATE:
04/25/06 Three Mile Island Unit 1 Larae Break Loss of Coolant Accident (LBLOCA)
APCT = 0 O F N/A APCT = 0 OF N/A APCT = 0 OF APCT = 0 O F
APCT = 0 OF APCT = 0 OF APCT = 0 OF APCT = 0 OF APCT = -25 OF APCTz -35 OF CURRENT OPERATING CYCLE: 16 Cycle 15 Fuel Cycle Design Change to leave APSRs in at EOC (see note 7\\
Evaluation Model: BW NT2 Calculation: Framatome ANP 86-5002073-02, July 1999 (Mark-B9)
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 Framatome ANP 86-501 1294-00, March 2001 (Mark-B12)
APCT =
MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS NET PCT PCT =
2058°F 1954 OF
- 6. CURRENT LOCA MODEL ASSESSMENTS I
1 Evaluation of Batch 18 Fuel Design Change (see note 8)
I APCT = 0 OF NET PCT PCT =
2058°F 1954 OF
- The BWNT EM is based on RELAPS/MOD2-B&W.
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
5928-06-20450 Page 1 of 2 TMI Unit 1 10 CFR 50.46 Report Assessment Notes
- 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 O 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 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. Operation with no APSR Pull The original reload design and licensing for Cycle 15 operation was performed based on removal of the axial power shaping rods (APSR) near the end of the cycle. The original design was modified such that the APSRs would not be removed. Thus the effect on the LOCA evaluations supporting the LHR limits applied to Cycle 15 core maneuvering was investigated.
It was found that all checks relating to the LOCA evaluation remained applicable and bounding and that the LHR limits used in the core maneuvering analyses remained acceptable.
5928-06-20450 Page 2 of 2
- 8. Fuel Design Changes Reload Evaluation Reload licensing evaluations for Cycle 16 were performed. Fuel design changes in the Batch 18 fuel were evaluated in detail. The effect of the large change in the plenum volume was determined to have an unfavorable effect at BOL, however this was offset by available fuel temperature margin as allowed by the reload licensing methodology BAW-10179P-A.
Therefore, it was concluded that the previous Mark-B12 LOCA analyses remained applicable and bounding to the Batch 18 fuel.