TMI-12-181, Response to Request for Additional Information - 10 CFR 50.46 30-Day Report

Response to Request for Additional Information - 10 CFR 50.46 30-Day Report

ML12349A175
Person / Time
Site:	Three Mile Island
Issue date:	12/12/2012
From:	Jesse M Exelon Generation Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
TMI-12-181
Download: ML12349A175 (5)

Text

10 CFR 50.46 TMI-12-181 December 12, 2012 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 No. 50-289

Subject:

Response to Request for Additional Information - 10 CFR 50.46 30-Day Report

References:

1) Letter from M. D. Jesse (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 30-Day Report,"

dated March 21, 2012

2) Letter from P. Bamford (U.S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Generation Company, LLC), "Three Mile Island Nuclear Station, Unit 1 - Request for Additional Information Regarding 30-Day Report for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46 (TAC NO. ME8237),"

dated November 13, 2012

3) Letter from P. Salas (AREVA NP Inc.) to U.S. Nuclear Regulatory Commission, "Generic RAI Response to a 30-Day 10 CFR 50.46 Report of Significant PCT Change," dated December 6,2012 In the Reference 1 letter Exelon Generation Company, LLC (Exelon) submitted a 30-day 10 CFR 50.46 report for Three Mile Island Nuclear Station (TMI), Unit 1. This letter discussed an AREVA NP Inc. (AREVA) notification concerning two Evaluation Model (EM) error corrections. In the Reference 2 letter the U.S. Nuclear Regulatory Commission requested additional information. Attached is Exelon's response to this request. As part of this response, AREVA has prepared a response to Question 1, which is contained in Reference 3.

No new regulatory commitments are established in this submittal. If any additional information is needed, please contact Tom Loomis at (610) 765-5510.

Response to Request for Additional Information - 10 CFR 50.46 30-Day Report December 12, 2012 Page 2 Respectfully, ichael . Jesse /

Director - Licensing e(Re tory Affairs Exelon Generation C any, LLC

Attachment:

Response to Request for Additional Information cc: USNRC Administrator, Region I USNRC Project Manager, TMI, Unit 1 USNRC Senior Resident Inspector, TMI, Unit 1

ATTACHMENT Response to Request for Additional Information

Response to Request for Additional Information Page 1 of 2 Question 1:

1. There are two changes to PCT for large break loss~of~coolant accident (LBLOCA) analysis discussed in the report submitted by the licensee. The first change is an Evaluation Model (EM) application error in the determination of the end of Emergency Core Cooling System (ECCS) bypass, which resulted in an 80°F decrease in PCT. The second change is an EM modeling change to include the effects of the upper plenum column weldments, which resulted in an 80°F increase in PCT.

Provide the analysis that lead to each change having an 80 degree change in PCT.

Response

See letter from P. Salas (AREVA NP Inc.) to U.S. Nuclear Regulatory Commission, "Generic RAI Response to a 30-Day 10 CFR 50.46 Report of Significant PCT Change," dated December 6, 2012.

Question 2:

2. Paragraph 50.46(a)(3)(ii) of 10 CFR states: " ... If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 50.46 requirements .. ,11 The PCT for LBLOCA for TMI-1 has changed by an absolute value of 160°F since the analysis was performed. Simply reporting the changes and errors in the methodology does not satisfy the intent of the regulation, Justify not providing a schedule for reanalysis or taking other action to show compliance with Section 50.46.

Response

The response to Question 1 provides additional detail regarding the analytical bases for the peak cladding temperature (PCT) error estimates, which were based upon explicit RELAP5/MOD2-B&W code runs for a Babcock & Wilcox (B&W)-designed nuclear steam supply system (NSSS). One error that was corrected in the evaluation models was specific to the determination of the end of emergency core cooling system (ECCS) bypass. A separate error correction to the ECCS evaluation model was made based on the effects of the upper plenum column weldments.

As evidenced by the information provided in the response to Question 1, both of these error corrections have been analyzed in detail. Furthermore, the error corrections in the ECCS evaluation model do not result in any challenge to the 10 CFR 50.46(b) acceptance criteria.

As the individual error corrections have been identified for the applicable evaluation model, and there are no other known changes identified at this time, the overall evaluation model is considered bounding and complete.

The corrected ECCS evaluation model, as discussed in the response to Question 1, will be used for any future analyses.

Response to Request for Additional Information Page 2 of 2 In summary, the response to the NRC staff request establishes the following:

• The error-adjusted PCTs at Three Mile Island Nuclear Station, Unit 1 remain considerably below the 10 CFR 50.46(b) acceptance criteria with greater than 300°F PCT margin.

• The SBLOCA analyses are not affected by the ECCS evaluation model errors.

• The response provides additional information regarding the nature of the PCT error evaluations, which are supported by explicit analyses using the B&W plant ECCS evaluation model.

Based on the actions discussed above, Three Mile Island Nuclear Station, Unit 1 remains in compliance with 10 CFR 50.46, and further LBLOCA reanalysis is not warranted.