ML18190A404

NRR E-mail Capture - Verbal Authorization for Proposed Alternative Relief Request RR-4-15 Use of ASME Code Case N-513-4

ML18190A404
Person / Time
Site:	Summer
Issue date:	07/03/2018
From:	Shawn Williams Plant Licensing Branch II
To:	Bennett B, Thompson B, Dalick S South Carolina Electric & Gas Co
References
Download: ML18190A404 (4)

Text

NRR-DMPSPEm Resource From: Williams, Shawn Sent: Tuesday, July 3, 2018 9:00 PM To: DALICK, SARA BETH; THOMPSON, BRUCE L; BENNETT, BRUCE ALLEN Cc: Markley, Michael; Davis, Robert; Hilton, Eliza; Musser, Randy; Kalikian, Roger

Subject:

Verbal Authorization for Proposed Alternative Relief Request RR-4-15 use of ASME Code Case N-513-4 Attachments: Summer UNIT 1 VERBAL AUTHORIZATION FINAL.docx 1

Hearing Identifier: NRR_DMPS Email Number: 454 Mail Envelope Properties (BN6PR09MB1186E2B6953031BCF17238A290410)

Subject:

Verbal Authorization for Proposed Alternative Relief Request RR-4-15 use of ASME Code Case N-513-4 Sent Date: 7/3/2018 9:00:16 PM Received Date: 7/3/2018 9:00:18 PM From: Williams, Shawn Created By: Shawn.Williams@nrc.gov Recipients:

"Markley, Michael" <Michael.Markley@nrc.gov>

Tracking Status: None "Davis, Robert" <Robert.Davis@nrc.gov>

Tracking Status: None "Hilton, Eliza" <Eliza.Hilton@nrc.gov>

Tracking Status: None "Musser, Randy" <Randy.Musser@nrc.gov>

Tracking Status: None "Kalikian, Roger" <Roger.Kalikian@nrc.gov>

Tracking Status: None "DALICK, SARA BETH" <SARA.DALICK@scana.com>

Tracking Status: None "THOMPSON, BRUCE L" <bthompson@scana.com>

Tracking Status: None "BENNETT, BRUCE ALLEN" <BRUCE.BENNETT@scana.com>

Tracking Status: None Post Office: BN6PR09MB1186.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 4 7/3/2018 9:00:18 PM Summer UNIT 1 VERBAL AUTHORIZATION FINAL.docx 18783 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

Recipients Received:

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR PROPOSED ALTERNATIVE RELIEF REQUEST RR-4-15 USE OF ASME CODE CASE N-513-4 VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 DOCKET NO. 50-395 JULY 3, 2018 Technical Evaluation read by Roger Kalikian, Acting Chief of the Piping and Head Penetrations Branch, Office of Nuclear Reactor Regulation By letter dated July 3, 2018, 2018, as supplemented by an e-mail dated July 3, 2018, South Carolina Electric & Gas Company (SCE&G, the licensee), submitted a proposed alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, IWA-4000, at Virgil C. Summer Nuclear Station, Unit 1.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee submitted proposed alternative Relief Request RR-4-15 to implement ASME Code Case N-513-4 "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping,Section XI, Division 1," to disposition a pinhole leak in lieu of performing an ASME Code repair of a Service Water (SW) branch tee located on the ASME Code Class 3 20-inch diameter piping downstream of the B Component Cooling Water heat exchanger discharge valve XVB03123B-SW below a field weld (FW-1).

On July 2, 2018, the licensee discovered the leakage rate in the branch tee to be approximately 0.092 gallons per minute. The licensee has requested to use Code Case N-513-4 to evaluate the flaw in lieu of repair in accordance with ASME Code Section XI, IWA-4000. The licensee stated that it followed all requirements of Code Case N-513-4 and took no exceptions. The licensee noted that the NRC has not approved the use of Code Case N-513-4; therefore, a proposed alternative is required.

In addition to following all of the requirements of Code Case N-513-4, to ensure structural integrity of the subject elbow, the licensee considered the combined effect of the branch tee leakage and a nearby SW pipe leak. The licensee determined a combined allowable maximum leakage rate of 50.3 gallons per minute. The licensee stated that further degradation is acceptable as long as the average thickness of the remaining material outside the hole is greater than 0.051 inches within a diameter of 2.5 inches of the hole and a minimum wall thickness of 0.162 inches for all other areas of the pipe tee.

The NRC staff notes that Code Case N-513-4 requires frequent periodic inspections to determine if flaws are growing. The code case requires daily monitoring to confirm the analysis conditions used in the evaluation remain valid, thus ensuring that the structural integrity of the subject component is maintained. The NRC staff determines that the licensees approach of applying a maximum leakage rate that provides sufficient margin regarding the minimum required system flow and maximum allowable flooding, is acceptable because it will provide sufficient time for corrective measures to be taken before significant increases in leakage erodes defense-in-depth, which could lead to adverse consequences.

The NRC staff finds that performing an ASME Code repair will require the plant to shut down, which would lead to transients and potential risk without a compensating increase in the level of quality and safety.

The NRC finds that the licensees proposed alternative provides reasonable assurance that the structural integrity of the subject plant service water piping and its intended safety function will be maintained until the next scheduled refueling outage, or until the temporary acceptance criteria of Code Case N-513-4 are exceeded, or until the leak rate exceeds 50.3 gallons per minute, whichever event occurs first.

Authorization read by Michael Markley, Chief of the Plant Licensing Branch II-1, NRR As Chief of the Plant Licensing Branch II-1, Office of Nuclear Reactor Regulation, I concur with the conclusions of the Piping and Head Penetrations Branch.

The NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity of the subject plant service water piping. The NRC staff finds that complying with IWA-4000 of the ASME Code,Section XI, would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, on July 4, 2018, the NRC authorizes the use of proposed alternative Relief Request RR-4-15 at Virgil C. Summer Nuclear Station Unit 1, until the next refueling outage, which is scheduled to begin October 6, 2018, until the temporary acceptance criteria of Code Case N-513-4 are exceeded, or until the leak rate exceeds 50.3gallons per minute, whichever event occurs first.

The NRC staff notes that this approval does not imply or infer the generic use of Code Case N-513-4.

All other requirements in ASME Code,Section XI, for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding the proposed alternative while preparing subsequent written safety evaluation.