1CAN121203, Unit 1, Response to Request for Additional Information Related to 10 CFR 50.46 30-Day Report

Unit 1, Response to Request for Additional Information Related to 10 CFR 50.46 30-Day Report

ML12353A489
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/17/2012
From:	Pyle S Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN121203
Download: ML12353A489 (5)

Text

1CAN121203 December 17, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Response to Request for Additional Information Related to 10 CFR 50.46 30-Day Report Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51

References:

1.

Entergy letter dated March 20, 2012, 10 CFR 50.46 Report -

Significant Change in Peak Cladding Temperature (1CAN031202)

(ML12080A120)

2.

NRC email dated November 8, 2012, RAI on the Review of 10 CFR 50.46 30-day Report (ML12313A440)

Dear Sir or Madam:

Entergy Operations, Inc. (Entergy), submitted 30-day report to inform the NRC of two errors in the Arkansas Nuclear One, Unit 1 (ANO-1) Emergency Core Cooling System evaluation model. The errors resulted in no net changes in to the Peak Clad Temperature calculated for the Large Break Loss-of-Coolant Accident. Each individual error as well as the sum of the absolute values of the two errors is greater than the requirement of 10 CFR 50.46.

On reviewing the submittal, the NRC staff requested additional information to continue the review and issued Reference 2. Attached is the response to the request.

This submittal contains no regulatory commitments.

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Licensing Arkansas Nuclear One

1CAN121203 Page 2 of 2 Should you have any questions, please contact me.

Sincerely, Original signed by Stephenie L. Pyle SLP/rwc

Attachment:

Response to Request for Additional Information Regarding Arkansas Nuclear One, Unit 1 (ANO-1) 30-Day Report cc:

Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Kaly Kalyanam MS O-8 B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852

Attachment to 1CAN121203 Response to Request for Additional Information Regarding Arkansas Nuclear One, Unit 1 (ANO-1) 30-Day Report

Attachment to 1CAN121203 Page 1 of 2 Response to Request for Additional Information Regarding Arkansas Nuclear One, Unit 1 (ANO-1) 30-Day Report

1. There are two changes to peak cladding temperature (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.

The response to this request was provided by AREVA NP for the Babcock & Wilcox (B&W) licensees in letter dated December 6, 2012 (AREVA Letter NRC:12:062, Generic RAI Response to a 30-Day 10 CFR 50.46 Report of Significant PCT Change). The AREVA report is applicable to ANO-1.

2. 10 CFR 50.46(a)(3)(ii) 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... "

The PCT for LBLOCA for Arkansas Nuclear One, Unit 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.

The response to Question 1 provides additional detail regarding the analytical bases for the PCT error estimates, which were based upon explicit RELAP5/Mod2-B&W code runs for B&W-designed nuclear steam supply systems.

One error that was corrected in the EMs was specific to the determination of the end of ECCS bypass. A separate error correction to the ECCS EM 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 EM do not result in any challenges to the 10 CFR 50.46(b) acceptance criteria. As the individual error corrections have been incorporated into the applicable EM, and there are no other known changes to be made at this time; therefore, the overall EM is considered adequate.

Attachment to 1CAN121203 Page 2 of 2 In summary, the response to the NRC staff requests establishes the following:

The error-adjusted PCTs for ANO-1 remain considerably below the 10 CFR 50.46(b) acceptance criteria with approximately 190F PCT margin.

The Small Break LOCA analyses are not affected by the ECCS EM 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 EM.

Based on the above, there are no adverse impacts to safety. Therefore, further LBLOCA reanalysis for ANO-1 is not warranted.