W3F1-2009-0006, RAI Response to Request for Alternative W3-ISI-006 to Extend the Second 10 Year Amse Code ISI and License Amendment Request NPF-38-280 to Support Request for Alternative W3-ISI-006
| ML090820487 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/19/2009 |
| From: | Murillo R Entergy Nuclear South, Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MD9669, TAC MD9671, W3-ISI-006, W3F1-2009-0006 | |
| Download: ML090820487 (5) | |
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Entergy Entergy Nuclear South Entergy Operations, Inc.
17265 River Road Killona, LA 70057-3093 Tel 504 739 6715 Fax 504 739 6698 rmurill@entergy.com Robert J. Murillo Licensing Manager Waterford 3 W3F1-2009-0006 March 19, 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Reference:
RAI Response to Request for Alternative W3-ISI-006 To Extend the Second 10 Year AMSE Code ISI and License Amendment Request NPF-38-280 To Support Request for Alternative W3-ISI-006 Waterford Steam Electric Station, Unit 3 (Waterford 3)
Docket No. 50-382 License No. NPF-38 Letter from Entergy to NRC, Request for Alternative W3-ISI-006 Proposed Alternative to Extend the Second 10-Year Inservice Inspection Interval for Reactor Vessel Internal Weld Examinations, dated September 18, 2008
Dear Sir or Madam:
This letter is being sent in response to the NRC request for additional information (RAI) received from the NRC NRR Waterford 3 Project Manager. The subject RAI is associated with Entergy Letter W3F1 -2008-0060 dated September 18, 2008 (Reference 1), which requested alternative W3-ISI-006 proposed alternative to extend the second 10-Year inservice inspection interval for Reactor Vessel internal weld examinations. Attachment 1 contains the Entergy response to the RAI.
There are no new commitments contained in this submittal.
If you have any questions or require additional information, please contact Mr. Robert J.
Murillo, Manager, Licensing at (504) 739-6715.
tcer I
- Rely, RJM/OPP sf
Attachment:
RAI Response to Request
W3F1`-2009-0006 Page 2 cc:
Mr. Elmo E. Collins, Jr.
Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box-822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004
Attachment I W3FI-2009-0006 RAI Response To Request to W3F1-2009-0006 Page 1 of 2 REQUEST FOR ADDITIONAL INFORMATION WATERFORD STEAM ELECTRIC STATION, UNIT 3 REQUEST FOR ALTERNATIVE W3-1SI-006 TO EXTEND THE SECOND 10-YEAR AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE INSERVICE INSPECTION INTERVAL AND LICENSE AMENDMENT REQUEST NPF-38-280 TO SUPPORT REQUEST FOR ALTERNATIVE W3-1SI-006 (TAC NOS. MD9671 AND MD9669)
In Section 3.4 of the final safety evaluation report issued May 8, 2008 (ML081060045), the staff notes that licensees which submit a request for an alternative based on Topical Report (TR) WCAP-1 6168,"Risk-lnformed Extension of the Reactor Vessel In-Service Inspection Interval," must submit the following plant-specific information:
Licensees must demonstrate that the RTmax-x and the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level (AT 30), must be calculated using the latest approved methodology documented in Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," or "other Nuclear Regulatory Commission (NRC)-approved methodology."
"Other NRC-approved methodology" includes equations 5, 6, and 7 as described in paragraph (g) of the proposed Title 10 of the Code of Federal Regulations (CFR) 50.61 a rule as published in the Federal Register, Vol. 72, No. 191, dated October 3, 2007.
Paragraph (f)(6) of proposed rule 10 CFR 50.61a contains a prescriptive approach to determining the validity of implementing equations 5 through 7 of paragraph (g) for the calculation of AT 30 values. Proposed rule paragraph 10 CFR 50.61 a(f)(6)(i) states that the licensee shall evaluate the results from a plant-specific or integrated surveillance program if the surveillance data has been deemed consistent as judged by the criteria set forth in 10.
CFR 50.61a(f)(6)(i) through (f)(6)(iv).
The licensee has implemented the use of the equations prescribed in 10 CFR 50.61 a(g) to determine AT 30 values for the Waterford 3 beltline materials. However, the licensee did not provide justification of the applicability of these 10 CFR 50.61 a(g) equations to Waterford Steam Electric Station, Unit 3 (Waterford 3), through the performance of surveillance checks as described in 10 CFR 50.61a(f)(6). Therefore, the staff requests the licensee to submit information consistent with the requirements of proposed rule 10 CFR 50.61 a(f)(6) that establishes the applicability of equations 5 through 7 as given in proposed rule 10 CFR 50.61 a(g) for calculating AT 30 values for the Waterford 3 beltline material.
Response
Paragraph (f)(6)(i)(B) of the proposed rule states the following:
If three or more surveillance data points measured at three or more different neutron fluences exist for a specific material, the licensee shall determine if the surveillance data show a significantly different trend than the embrittlement model predicts. This must be achieved by evaluating the surveillance data for consistency with the embrittlement model by following the procedures specified by paragraphs (f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of this section. If fewer than three surveillance data points exist for a specific to W3F1-2009-0006 Page 2 of 2 material, then the embrittlement model must be used without performing the consistency check.
Presently, only 2 points of surveillance data exist for Waterford 3. Therefore, the surveillance checks of paragraphs (f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) can not be performed and as stated in paragraph (f)(6)(i)(B), the embrittlement model must be used without performing the consistency check.