1CAN101305, Unit 1, Response to Additional Request for Additional Information Related to 10 CFR 50.46 30-Day Report

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

ML13296A744
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/22/2013
From:	Pyle S Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN101305
Download: ML13296A744 (6)

Text

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Licensing Arkansas Nuclear One 1CAN101305 October 22, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Response to Additional 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)

3. AREVA letter dated December 6, 2012, Generic RAI Response to a 30-Day 10 CFR 50.46 Report of Significant PCT Change, (AREVA Letter NRC:12:062) (ML12342A381)

4. Entergy letter dated December 17, 2012, Response to Request for Additional Information Related to 10 CFR 50.46 30 Day Report (1CAN121203) (ML12353A489)

5. NRC email dated March 21, 2013, Request for Additional Information - TAC No. ME9719 (ML13080A350)

6. AREVA letter dated March 28, 2013, Generic RAI Response to a 30-Day 10 CFR 50.46 Report of Significant PCT Change, (AREVA Letter NRC:13:014)

7. Entergy letter dated April 10, 2013, Response to Request for Additional Information (1CAN041301) (ML13101A267)

8. NRC email dated August 20, 2013, RAI No. 3 regarding ANO-1 30-day report for ECCS model changes pursuant to the requirements of 10 CFR 50.46(a)(1)(i) (ML13232A354)

1CAN101305 Page 2 of 3

Dear Sir or Madam:

Entergy Operations, Inc. (Entergy) submitted a 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 (Reference 1). The errors resulted in no net changes 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 (RAI) to continue the review and issued Reference 2. References 3 and 4 provided the responses to the request.

Reference 5 provided the NRCs subsequent RAI. In response to this RAI, AREVA has prepared a generic response to address these questions. This response is provided by Reference 6. ANO-specific response was provided by Reference 7.

Subsequently, the NRC transmitted an additional RAI regarding References 1 and 4 (Reference 8). The purpose of this submittal is to provide a response to the RAI.

This submittal contains no regulatory commitments.

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 30-Day Response for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46(a)(1)(i)

1CAN101305 Page 3 of 3 cc: Mr. Steven A. Reynolds 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 1CAN101305 Response to Request for Additional Information Regarding 30-Day Response for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46(a)(1)(i)

Attachment to 1CAN101305 Page 1 of 2 Response to Request for Additional Information Regarding 30-Day Response for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46(a)(1)(i)

Request for Additional Information By letter dated March 20, 2012 ((Agencywide Documents Access and Management System (ADAMS) Accession No. ML12080A120), Entergy Operations, Inc. the licensee for Arkansas Nuclear One, Unit 1 (ANO-1) sent a notice reporting a change or error discovered in an evaluation model or in the application of such a model that affects the peak cladding temperature calculation. This report was submitted pursuant to the requirements of 10 CFR 50.46, which requires, in part, that licensees report a change in the evaluation model used resulting in a significant change in PCT (greater than 50°F). The intent of this requirement is to enable the staff to establish the safety significance of this change (See FR Volume 53, No. 180, pp. 35996-36005).

On October 15, 2012, the staff issued a Request for Additional Information (RAI) to the licensee (ADAMS Accession Number ML12279A036) regarding the 30-day report which stated the following:

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.

The licensee responded to the RAI on December 17, 2012 (ADAMS Accession Number ML12353A489). The RAI response does not include a proposed schedule for providing a reanalysis. In the response, the licensee states that the PCT error evaluations are supported by explicit analyses using the B&W plant ECCS evaluation model. Since a schedule for reanalysis was not provided, justify how generic analysis for the B&W plant ECCS evaluation model constitutes taking other action to show compliance with Section 50.46. In particular, while the submitted RAI response addresses the acceptance criteria contained in 10 CFR 50.46(b), the response does not address the requirement, at 10 CFR 50.46(a)(1)(i), to calculate ECCS cooling performance in accordance with an acceptable evaluation model. In light of the presently reported, significant, estimated effects of errors and changes, explain how the present ECCS cooling performance has been calculated in accordance with an acceptable evaluation model, such that any other action, as provided in 10 CFR 50.46(a)(3), has been taken to show compliance with 10 CFR 50.46 requirements, including those contained in 10 CFR 50.46(a)(1).

1CAN101305 Page 2 of 2

Response

The evaluation that supports the 2012 10 CFR 50.46 Loss of Coolant Accident (LOCA)

Report for Babcock & Wilcox (B&W) Plants denotes other actions taken to show compliance with 10 CFR 50.46 requirements. The evaluation demonstrated the requirements of 10 CFR 50, Appendix K for a conservative model were fully met based on a reported estimated net zero change in peak clad temperature (PCT). The evaluation concluded the actual net PCT would decrease; therefore, the existing model results remain conservative and acceptable. As the reported net PCT did not change, local oxidation and whole core hydrogen generation from the original model are unaffected and remain in compliance. In addition, the coolable core geometry and long-term cooling impacts remain unchanged and fully meet 10 CFR 50.46(b) requirements. Emergency Core Cooling System (ECCS) cooling performance was calculated with errors corrected and this result comes from an acceptable evaluation model that complies with 10 CFR 50.46(a)(1)(i) and 50.46(a)(1)(ii). The analytical technique used approximates realistically the behavior of the reactor system during a loss-of-coolant accident. As such, there is a high level of probability that the criteria will not be exceeded.

In lieu of submitting a proposed schedule for providing a reanalysis, the actions already taken as described above are considered sufficient to satisfy the intent of 10 CFR 50.46(a)(3)(ii), specifically or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements.