ML17290A317
| ML17290A317 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 09/19/2017 |
| From: | Randy Hall Plant Licensing Branch II |
| To: | Mcelroy G Southern Nuclear Operating Co |
| References | |
| MF9812, MF9813 | |
| Download: ML17290A317 (4) | |
Text
1 NRR-PMDAPEm Resource From:
Hall, Randy Sent:
Tuesday, September 19, 2017 7:39 AM To:
Ken McElroy Cc:
Joyce, Ryan M. (RMJOYCE@southernco.com)
Subject:
Hatch 1 and 2 - Draft Request for Additional Information RE: Request For Alternative No. HNP-ISI-ALT-05-05 (CAC Nos. MF9812, MF9813)
Attachments:
Draft RAI - Hatch 1 and 2 - Code Case N-702 Evaluation.docx September 19, 2017 Mr. G. Ken McElroy Licensing Manager Southern Nuclear Operating Company, Inc.
Dear Ken,
By letter dated June 5, 2017 (Agencywide Documents Access and Management System Accession No. ML17156A831), the Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted two requests for alternatives to the ASME Code requirements for the Edwin I. Hatch Nuclear Plant, Units 1 and 2 (HNP).
The first alternative request, HNP-ISI-ALT-05-05, if authorized pursuant to 10 CFR 50.55a(z)(1), would allow SNC to implement an alternative examination of a minimum of 25% of the specified reactor vessel nozzle-to-vessel welds and the inside radius sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702, in lieu of the code-required 100% examinations. The second alternative request, HNP-ISI-ALT-05-06, is being reviewed separately by the NRC staff (under CAC Nos.
MF9851 and MF9852), and will be addressed in separate correspondence.
The NRC staff has reviewed the information provided by SNC in your submittal regarding Request for Alternative HNP-ISI-ALT-05-05, and has determined that the additional information identified in the attached draft Request for Additional Information (RAI) is needed to complete its evaluation.
Please let me know if you would like a clarification call with the NRC staff to discuss the draft RAI. The call is intended to ensure that the draft RAI questions are understandable, the regulatory basis for the questions is clear, and to determine whether any of the information requested was previously docketed. Following a clarification call or your confirmation that a call is not necessary, the staffs RAIs will be documented as an Official Agency Record. SNC is requested to provide a written response to the RAI within 30 days of the date of the clarification call, or your confirmation that a call is not needed.
Randy Randy Hall, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation USNRC (301) 415-4032 Randy.Hall@nrc.gov
Hearing Identifier:
NRR_PMDA Email Number:
3773 Mail Envelope Properties (Randy.Hall@nrc.gov20170919073900)
Subject:
Hatch 1 and 2 - Draft Request for Additional Information RE: Request For Alternative No. HNP-ISI-ALT-05-05 (CAC Nos. MF9812, MF9813)
Sent Date:
9/19/2017 7:39:03 AM Received Date:
9/19/2017 7:39:00 AM From:
Hall, Randy Created By:
Randy.Hall@nrc.gov Recipients:
"Joyce, Ryan M. (RMJOYCE@southernco.com)" <RMJOYCE@southernco.com>
Tracking Status: None "Ken McElroy" <gkmcelro@southernco.com>
Tracking Status: None Post Office:
Files Size Date & Time MESSAGE 2155 9/19/2017 7:39:00 AM Draft RAI - Hatch 1 and 2 - Code Case N-702 Evaluation.docx 40531 Options Priority:
Standard Return Notification:
No Reply Requested:
No Sensitivity:
Normal Expiration Date:
Recipients Received:
Enclosure REQUEST FOR ADDITIONAL INFORMATION REQUEST FOR ALTERNATIVE NO. HNP-ISI-ALT-05-05 REACTOR PRESSURE VESSEL NOZZLE-TO-SHELL WELDS AND NOZZLE INNER RADIUS EXAMINATIONS FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2 SOUTHERN NUCLEAR OPERATING COMPANY DOCKET NOS. 50-321 AND 50-366 CAC NOS. MF9812 and MF9813 By application dated June 5, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17156A831), Southern Nuclear Operating Company (the licensee) submitted alternative request number HNP-ISI-ALT-05-05 in accordance with Paragraph 50.55a(z)(1) of Title 10 of the Code of Federal Regulations (10 CFR) for a proposed alternative to the requirements of 10 CFR 50.55a, Codes and standards, for Edwin I. Hatch Nuclear Plant (HNP), Units 1 and 2. Specifically, the licensee proposes an alternative to the inservice inspection requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) applicable to the reactor pressure vessel (RPV) nozzle-to-vessel shell welds and nozzle inner radius, based on ASME Code Case N-702, Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds. The NRC staff requires a response to the following requests for additional information (RAIs) to complete the review of alternative HNP-ISI-ALT-05-05.
RAI 1
One of the technical basis report for ASME Code Case N-702 is BWR Vessel and Internals Project (BWRVIP)-108, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI Technical Report 1003557, October 2002. In its safety evaluation of BWRVIP-108, dated December 19, 2007 (ADAMS Accession No. ML073600374), the NRC staff documented the BWRVIPs supplemental probabilistic fracture mechanics (PFM) evaluation that showed that the limiting probability of failure (PoF) is 1.98x10-6 per year for normal operation, compared to 1.19x10-7 per year for a low temperature over-pressure (LTOP) event. However, the PFM analysis in to Enclosure 1 of the licensees submittal reported PoF values only for the LTOP event. The NRC staff requests the licensee to report the PoF values for normal operation or discuss how the PoF values for LTOP are more limiting than those for normal operation.
RAI 2
Section 3.1 Fatigue Cycles of Attachment 1 to Enclosure 1 of the licensees submittal described the fatigue cycles used in the PFM analysis. The NRC staff requests the licensee to confirm the following: (1) the fatigue cycles of all other transients applicable to the HNP, Units 1 and 2 recirculation inlet and outlet nozzles were lumped into the selected bounding transients, and (2) fatigue crack growth in the PFM analysis was performed using the fatigue cycles projected to 60 years of operation.
RAI 3
Section 4.2.2 Fluence of Attachment 1 to Enclosure 1 of the licensees submittal indicated that since fluence values were not available for the HNP, Units 1 and 2 recirculation inlet and outlet nozzles, the peak fluence at the inside RPV surface from the closest RPV circumferential weld (C4) and the lower vessel shell were used in the PFM analysis. The NRC staff requests the licensee to confirm that RPV circumferential weld C4 and the lower vessel shell selected for fluence values are closer to the active core in the RPV axial direction than the HNP, Units 1 and 2 recirculation inlet and outlet nozzles.