ML063530324

Thirty-Day Report for Loss-of-Coolant Accident Evaluation Model Change

ML063530324
Person / Time
Site:	Cook
Issue date:	12/07/2006
From:	Jensen J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:6046-01
Download: ML063530324 (7)

Text

Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place Bridgman, Ml 49106 POWER0 AERcom A unit of American Electric Power December 7, 2006 AEP:NRC:6046-01 10 CFR 50.46 Docket No.: 50-315 U. S. Nuclear Regulatory Commission

• 7 ATTN: Document Control Desk Mail Stop O-Pl-17 Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Unit 1 THIRTY-DAY REPORT FOR LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL CHANGE

References:

1. Letter from Joseph N. Jensen, Indiana Michigan Power Company (I&M), to U. S.

Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C.

Cook Nuclear Plant Units 1 and 2, 10 CFR 50.46 Loss-of-Coolant Accident Reanalysis Schedule," submittal AEP:NRC:4046-01, Accession Number ML050040216, dated December 28, 2004.

2. Letter from Joseph N. Jensen,, I&M, to U. S. NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Thirty-day Report of Loss-of-Coolant Accident Evaluation Model Changes," submittal AEP:NRC:5046, Accession Number ML051300368, dated April 29, 2005.

Pursuant to 10 CFR 50.46, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP), is submitting a 30-day report of loss-of-coolant accident (LOCA) model changes resulting in a significant change in calculated peak fuel cladding temperature (PCT) for the CNP Unit 1 large break LOCA (LBLOCA) analysis. A significant change is defined as a change or error identified in the model which results in a calculated PCT greater than 50 degree Fahrenheit (°F) or a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50F.

Attachinent 1 to this letter describes an assessment against the CNP Unit 1 LBLOCA analysis of record. Attachment 2 provides the CNP Unit 1 LBLOCA analysis of record PCT value and error assessments. Attachinent 2 also demonstrates that the PCT value remains within the 2200'F PCT limit as required by 10 CFR 50.46(b)(1).

4 oDI

U. S. Nuclear Regulatory Conmnussion AEP: NRC: 6046-01 Page 2 The overall change to the Unit 1 LBLOCA analysis is classified as significant in accordance with 10 CFR 50.46(a)(3)(i). By Reference 1, I&M submitted a schedule for reanalysis of the Unit 1 and Unit 2 small break LOCA (SBLOCA) and the Unit 2 LBLOCA analyses of record. By Reference 2, I&M submitted a schedule for reanalysis of the Unit 1 LBLOCA analyses of record. I&M's current schedule for the reanalyses is provided in Attachmient 3.

Should you have any questions concerning this subject, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.

Sincerely, Site Support Sdr"Vicvks Vice President Attachments KAS/rdw c: J. L. Caldwell - NRC Region III K. D. Curry - AEP Ft. Wayne, w/o attachments J. T. King - MPSC, w/o attachments MDEQ - WHMD/RPMWS, w/o attachments NRC Resident Inspector P. S. Tam, NRC Washington, DC

ATTACHMENT 1 TO AEP:NRC:6046-01 ASSESSMENT AGAINST THE LOSS-OF-COOLANT ACCIDENT (LOCA)

ANALYSES OF RECORD Indiana Michigan Power Company's (I&M's) most recent annual 10 CFR 50.46 report for Donald C. Cook Nuclear Plant (CNP) has been submitted by Reference 1. The reported assessments on the Unit 1 large break loss-of-coolant accident (LBLOCA) analysis of record peak cladding temperature (PCT) in Attachment 2 of this letter is based upon the limiting fiesh 15x15 Upgrade Fuel introduced into the CNP Unit 1 Cycle 21 (U1C21) core. A new PCT assessment against the CNP Unit 1 LBLOCA analysis of record is described below. The new assessment is reflected in the PCT accounting in Attachment 2.

Assessment Against the LBLOCA Analysis of Record 15x15 Upgrade Fuel

Background

I&M and members of the Nuclear Regulatory Commission (NRC) Staff began discussion of the planned introduction of the 15x15 Upgrade Fuel for CNP Unit 1 to address fuel fietting wear issues at a February 3, 2005, meeting. Subsequent discussions were noted in an NRC summary of a July 25, 2005, telephone conference that was documented in Reference 2. As planned, the 15x15 Upgrade Fuel has been introduced into the CNP U1C21 core in accordance with 10 CFR 50.59. The CNP Unit 1 LBLOCA analysis was evaluated as part of the engineering work performed to support the U1C21 core reload. The PCT assessment associated with this evaluation is greater than 50 degrees Fahrenheit ('F) and is being reported in accordance with 10 CFR 50.46(a)(3)(ii).

Affected Evaluation Models 1981 Westinghouse LBLOCA Evaluation Model with BASH Estimated Effect The impact on PCT was estimated using plant-specific calculations with the BASH evaluation model. As indicated in the PCT accounting in Attachment 2, the effect of the change to the 15x15 Upgrade Fuel is a 59°F reduction. The previously determined 31'F penalty associated with transition core effects due to intermediate flow mixer (IFM) grids no longer applies, since all fuel assemblies that comprise the U1C21 core have IFMs. Additionally, the previously identified 37'F penalty associated with the "Spacer Grid Blocked Area Ratio/Open Area Fraction" does not apply to the 15x15 Upgrade Fuel. The net result of these assessments and removed penalties is a 127 0 F reduction in the calculated PCT with assessments compared to that previously reported in Reference 1.

to AEP:NRC:6046-01 Page 2 Conclusion This transmittal satisfies the 30-day reporting requiren-ient of 10 CFR 50.46(a)(3)(ii). demonstrates that the PCT value remains within the 2200°F PCT limit specified in 10 CFR 50.46(b)(1).

References

1. Letter from Joseph N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Annual Report and Thirty-day Report of 10 CFR 50.46 Loss-of-Coolant Accident Evaluation Model Changes," Accession Number ML062340193, submittal AEP:NRC:6046, dated August 11, 2006.

2. Letter from L. Raghavan, NRC, to Mano K. Nazar, I&M, "Donald C. Cook Nuclear Plant Units 1 and 2 - Summary of Telephone Conference Re: Fuel Upgrade TAC Nos.

MC5646 and MC5647)," Accession Number ML052650307, dated September 28, 2005.

ATTACHMENT 2 TO AEP:NRC:6046-01 DONALD C. COOK NUCLEAR PLANT (CNP) UNIT 1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT (LOCA)

PEAK CLAD TEMPERATURE (PCT)

SUMMARY

Attachiment 2 to AEP:NRC:6046-01 Page I TABLE 1 CNP UNIT 1 LARGE BREAK LOCA Evaluation Model: BASH FQ= 2.15 FAHZ= 1.55 SGTP = 15% Break Size: Cd = 0.4 Operational Parameters: RHR System Cross-Tie Valves Closed, 3250 MWt Reactor Power' LICENSING BASIS Analysis-of-Record, December 2000 PCT = 038'F MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. LOCBART Cladding Emissivity Errors -11017 B. Planned 50.59 Plant Change Evaluations

1. Reduced Containment Spray Temperature +23"F

2. 15X15 Upgrade Fuel -59 0F C. OTHER

1. Rebaseline Using PAD 4.0 +57'F D. LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2048'F

1. The 3250 MWt power level used in the reanalysis is acceptable because it bounds the Unit 1 3304 MWt steady state power limit in the operating license after adjusting for recapture of feedwater flow measurement and power calorimetric uncertainty.

ATTACHMENT 3 TO AEP:NRC:6046-01 REGULATORY COMMITMENTS The followbig table identifies those actions conmitted to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Conmmission (NRC) for the NRC's information and are not regulatory commnitments.

Commitment Date A new Unit 1 small break loss-of-coolant accident (SBLOCA) analysis March 2007 of the safety injection cross-tie valves closed case will be provided.

A new Unit 1 large break loss-of-coolant accident (LBLOCA) analysis December 2007 will be provided.,

A riew Unit 2 SBLOCA analysis of both the safety, injection cross-tie March 2009 valves closed case and the safety injection cross-tie valves opeh case will be provided.

A new Unit 2 LBLOCA analysis will be provided. March 2009