ML092370095

Request for Additional Information Nuclear Energy Institute Topical Report Material Reliability Program (Mrp): Technical Basis for Preemptive Weld Overlays for Alloy 82/182 Butt Welds in Pressurized Water Reactors (MRP-169), (MRP-169) (TAC

ML092370095
Person / Time
Site:	Nuclear Energy Institute, PROJ0669
Issue date:	08/28/2009
From:	Mensah T M NRC/NRR/DPR/PSPB
To:	Riley J H Nuclear Energy Institute
Mensah T
References
TAC MD8005
Download: ML092370095 (7)

Text

August 28, 2009 Mr. James H. Riley, Director Engineering Nuclear Energy Institute Suite 400 1776 I Street, NW Washington, DC 20006-3708

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RE: NUCLEAR ENERGY INSTITUTE TOPICAL REPORT MATERIAL RELIABILITY PROGRAM (MRP): TECHNICAL BASIS FOR PREEMPTIVE WELD OVERLAYS FOR ALLOY 82/182 BUTT WELDS IN PRESSURIZED WATER REACTORS (MRP-169) (TAC NO. MD 8005)

Dear Mr. Riley:

By letter dated September 7, 2005 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML052520325), the Nuclear Energy Institute (NEI) submitted for U.S. Nuclear Regulatory Commission (NRC) staff review Topical Report (TR) Material Reliability Program (MRP): Technical Basis For Preemptive Weld Overlays For Alloy 82/182 Butt Welds In Pressurized Water Reactors (MRP-169). By letter dated August 3, 2006, the NRC issued a request for additional information (RAI) (ADAMS Accession No. ML062050337). By letter dated January 9, 2008, the NEI provided its response to the RAI (ADAMS Accession Nos. ML080780299 and ML080780301). By letter dated April 7, 2008, the NRC staff issued a RAI (ADAMS Accession No. ML080940280). By letter dated May 2, 2008, the NEI provided its response to the RAI (ADAMS Accession No. ML082610254).

In addition, by letter dated March 2, 2009 (ADAMS Accession No. ML090690630), the NEI provided an Appendix to TR MRP-169 which provides a set of design and analysis alternatives that can be used for cases in which non-destructive examination procedures can be qualified for the extended optimized weld overlay examination volume for circumferential flaws, but not for axial flaws. Upon review of the information provided, the NRC staff has determined that additional information is needed to complete the review.

The purpose of this letter is to formally transmit and request the NEI's written response to the enclosed RAI questions. On August 24, 2009, you and I agreed that the NRC staff will receive your response to the enclosed RAI questions within 45 days of issuance of this letter. If you have any questions regarding the enclosed RAI questions, please contact me at 301-415-3610.

Sincerely, /RA/

Tanya M. Mensah, Senior Project Manager Special Projects Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project Nos. 669 and 689

Enclosure:

RAI questions cc w/encl: See next page

The purpose of this letter is to formally transmit and request the NEI's written response to the enclosed RAI questions. On August 24, 2009, you and I agreed that the NRC staff will receive your response to the enclosed RAI questions within 45 days of issuance of this letter. If you have any questions regarding the enclosed RAI questions, please contact me at 301-415-3610.

Sincerely, /RA/

Tanya M. Mensah, Senior Project Manager Special Projects Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project Nos. 669 and 689

Enclosure:

RAI questions cc w/encl: See next page DISTRIBUTION

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ewillis@epri.com ADAMS ACCESSION NO.: ML092370095 OFFICE PSPB/PM PSPB/LA CPNB/BC PSPB/BC

NAME TMensah DBaxley (CHawes for) TChan SRosenberg (EBowman for)

DATE 08/27/09 08/27/09 08/27/09 08/28/09 OFFICIAL RECORD COPY REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT (TR) MATERIAL RELIABILITY PROGRAM (MRP):

TECHNICAL BASIS FOR PREEMPTIVE WELD OVERLAYS FOR ALLOY 82/182 BUTT WELDS IN PRESSURIZED WATER REACTORS (MRP-169)

NUCLEAR ENERGY INSTITUTE (NEI)

PROJECT NO. 689 The U.S. Nuclear regulatory Commission (NRC) staff generated the following comments and questions after its review of TR MRP-169, and supplemental provided as discussed in the NEI letters dated May 2, 2008 (ADAMS Accession No. ML082610254), and March 2, 2009 (ADAMS Accession No. ML090690630).

All section, page, table, or figure numbers cited in the questions below refer to items in TR MRP-169, unless specified otherwise.

Questions from NRC Staff

1. Section 4.1 of TR MRP-169 specifies the design of the optimized weld overlays (OWOL),

which have less thickness than the full structural weld overlays (FSWOL). Section 4.5 of TR MRP-169 discusses the implication of the weld overlay on leak before break analysis. As implied in Section 4.1, the OWOL is unable, by itself, to satisfy structural integrity design requirements. Instead, the OWOL design requires a portion of the underlying Alloy 82/182 dissimilar metal (DM) weld material to remain intact and carry a portion of the loads. This original weld material is susceptible to cracking. In order to understand potential limitations of OWOLs, the NRC staff has considered the possibility that either the OWOL design or installation process or the associated nondestructive examination (NDE) does not perform as expected and a crack grows in the original weld after the OWOL is applied. During the initial phases of crack growth, bending and residual stress variations and metallurgical inhomogeneity would lead to uneven growth. However, once a portion of a surface crack grew deep enough to encounter the crack resistant overlay material, it would stop growing in the depth direction at that azimuthal location. Other segments of the crack could continue to grow deeper until they also reach the overlay interface. This could continue until the remaining uncracked ligament of original weld material is insufficient to adequately reinforce the OWOL material, at which point the mitigated weld may fail without prior leakage during a design basis event.

In a FSWOL the corrosion and primary water stress-corrosion cracking (PWSCC) resistance of the overlay material can be credited to prevent crack growth into the overlay in the event that a large pre-existing crack was missed by NDE, or in the event that design deficiencies or misapplication of the FSWOL resulted in unanticipated tensile residual stress fields. If large cracks occur in the original DM weld material under a FSWOL, the FSWOL can withstand full design loading without failing; and the PWSCC resistant material preserves the FSWOL load carrying ability and minimizes the likelihood of pipe rupture. In contrast, if the same deficiency in design or application affects the OWOL, the OWOL material, precisely because it is resistant to PWSCC, can cause small circumferential cracks in the original dissimilar metal weld to grow deep around the entire circumference, in which case the OWOL may become unable to withstand its design loading. In light of this possibility, please explain why application of an OWOL to a DM weld is an appropriate mitigation method and why its application will not invalidate previously approved leak-before- break analyses.

2. By letter dated May 2, 2008, the NEI responded to the NRC staff's request for additional information. Under Stress Analysis Question 1, the NRC staff asked the NEI to justify a target stress at the inside surface of 10 ksi. NEI responded that the 10 ksi maximum tensile stress criterion provides protection against primary water stress corrosion cracking (PWSCC).

American Society of Mechanical Engineers (ASME) Code,Section XI, Code Case N-770 1 has established that as part of an effective stress improvement mitigation technique, a compressive stress state was required on the wetted surface of all susceptible material for DM weld application. This is consistent with the NRC staff position and was developed, in part, due to the uncertainties in precise finite element stress modeling of the wetted surface of DM welds. Furthermore, the NRC staff position was not established to define a stress level at which crack initiation could not occur, rather to provide a conservative stress value that along with calculated stress levels throughout the volume of the weld provide a basis for reasonable assurance of structural integrity for a stress improved DM weld.

The NEI's response does not provide sufficient basis to demonstrate that increasing the wetted surface stress limit to 10 ksi would be equivalent to the NRC staff position. The NEI statement that stress corrosion cracking will not initiate on a surface that is below yield stress is not a sufficient basis for this conclusion due to large uncertainties in attempting to precisely model the wetted surface condition of in-service DM welds. Please provide additional basis, including supporting data, analyses and operational experience, to support allowing a wetted surface stress threshold of 10 ksi.

1 ASME Code,Section XI, Code Case N-770, Alternative Examination Requirements and Acceptance Standards for Class 1 PWER Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation ActivitiesSection XI, Division 1, Appendix I.

Nuclear Energy Institute Project No. 689 Electric Power Research Institute Project No. 669 cc:Mr. Anthony Pietrangelo, Senior Vice President & Chief Nuclear Officer Nuclear Generation Nuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 arp@nei.org Mr. Jack Roe, Director Security Nuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 jwr@nei.org Mr. Charles B. Brinkman Washington Operations ABB-Combustion Engineering, Inc. 12300 Twinbrook Parkway, Suite 330 Rockville, MD 20852 brinkmcb@westinghouse.com Mr. James Gresham, Manager Regulatory Compliance and Plant Licensing Westinghouse Electric Company P.O. Box 355 Pittsburgh, PA 15230-0355 greshaja@westinghouse.com

Ms. Barbara Lewis Assistant Editor Platts, Principal Editorial Office 1200 G St., N.W., Suite 1100 Washington, DC 20005 Barbara_lewis@platts.com Mr. Alexander Marion, Vice President Nuclear Operations Nuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 am@nei.org Mr. John Butler, Director Operations Support Nuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 jcb@nei.org Mr. James H. Riley, Director Engineering Nuclear Energy Institute 1776 I Street, NW Washington, DC 20006-3708 jhr@nei.org Mr. Chris Larsen Vice President and Chief Nuclear Officer EPRI 3412 Hillview Avenue Palo Alto, CA 94304-1338 cblarsen@epri.com Mr. David J. Modeen Director, External Affairs

EPRI 1300 W. T. Harris Boulevard Charlotte, NC 28262-8550 dmodeen@epri.com

Dr. Sean Bushart EPRI 3412 Hillview Avenue Palo Alto, CA 94304-1338 sbushart@epri.com Mr. Kurt Edsinger EPRI 3412 Hillview Avenue Palo Alto, CA 94304-1338 kedsinge@epri.com Mr. Ken Canavan EPRI 1300 W.T. Harris Boulevard 3/19/08 Charlotte, NC 28262-8550 kcanavan@epri.com Mr. Greg Selby EPRI 1300 W. T. Harris Boulevard Charlotte, NC 28262-8550 gselby@epri.com Mr. David Steininger EPRI 3412 Hillview Avenue Palo Alto, CA 94304-1338 dsteinin@epri.com Mr. Neil Wilmshurst EPRI 1300 W. T. Harris Boulevard Charlotte, NC 28262-8550 nwilmshu@epri.com