NRC-10-0060, Response to Request for Additional Information Regarding Relief Requests for RR-A37 & RR-A30 for Inservice Inspection/Nondestructive Examination Program Third Ten-Year Interval

Response to Request for Additional Information Regarding Relief Requests for RR-A37 & RR-A30 for Inservice Inspection/Nondestructive Examination Program Third Ten-Year Interval

ML102280294
Person / Time
Site:	Fermi
Issue date:	08/13/2010
From:	Plona J Detroit Edison, DTE Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-10-0060
Download: ML102280294 (11)

Text

Joseph H. Plona Site Vice President 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.5910 Fax: 734.586.4172 DTE Energy" 10 CFR 50.55a August 13, 2010 NRC-10-0060 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001

References:

1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

2) Detroit Edison's Letter to NRC, "Submittal of the Inservice Inspection / Nondestructive Examination Program Relief Requests for the Third Ten-year Interval," NRC-10-0004, dated January 20, 2010

Subject:

Response to Request for Additional Information Regarding Relief Requests RR-A37 and RR-A30 for the Inservice Inspection /

Nondestructive Examination Program Third Ten-Year Interval In Reference 2, Detroit Edison submitted proposed relief requests for the third ten-year interval of the Inservice Inspection / Nondestructive Examination Program. In an email from Mr. Mahesh Chawla to Mr. Alan Hassoun dated April 27, 2010, the NRC requested additional information for Relief Request RR-A37, Alternative Requirements for Examination of Boiling Water Reactor (BWR) Nozzle Inner Radius Sections and Nozzle-to-Shell Welds. This request was discussed in subsequent telephone conversations between NRC staff and Detroit Edison personnel on May 4, and June 10, 2010.

In another email from Mr. Mahesh Chawla to Mr. Alan Hassoun dated June 29, 2010, the NRC requested additional information for Relief Request RR-A30, Continuation of Risk-Informed Inspection (RI-ISI) Application on Circumferential Welds in Class 1 Piping. This request was discussed in a subsequent telephone conversation between NRC staff and Detroit Edison personnel on July 1, 2010.

The additional information requested by the NRC staff is enclosed.

There are no new commitments included in this document.

USNRC NRC-10-0060 Page 2 Should you have any questions or require additional information, please contact Mr.

Rodney W. Johnson of my staff at (734) 586-5076.

Sincerely, Enclosure cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 4, Region III Regional Administrator, Region III Supervisor, Electric Operators, Michigan Public Service Commission

Enclosure to NRC-10-0060 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 Response to Request for Additional Information (RAI) Regarding Relief Requests RR-A37 and RR-A30, for the Inservice Inspection / Nondestructive Program Third Ten-Year Interval

Enclosure to NRC-10-0060 Page 1 NRC Request for Additional Information Regarding Relief Request RR-A37 The Nuclear Regulatory Commission (NRC) staff has reviewed the information provided by Detroit Edison (the licensee) for Fermi Unit 2 in its letter dated January 20, 2010, and has determined that additional information is necessary to complete the review of Relief Request No.

RR-A37. Based on the staff's review, please provide a response which addresses the following request for additional information (RAI) questions.

1. In the December 19, 2007 letter from Matthew A. Mitchell to Rick Libra, BWRVIP Chairman, in reference to the "Safety Evaluation Of Proprietary EPRI Report, "BWR Vessel And Internals Project, Technical Basis For The Reduction Of Inspection Requirements For The Boiling Water Reactor Nozzle-To-Vessel Shell Welds And Nozzle Inner Radius (BWRVIP-108)," in the first paragraph in section, "5.0 PLANT-SPECIFIC APPLICABILITY" it states, "Licensees who plan to request relief from the ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-108 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-108 report to their units in the relief request by showing that all the following general and nozzle-specific criteria are satisfied..."

The Licensee has not requested the use of ASME Code Case N-702 as an alternative.

Please explain this discrepancy since relief can only be granted for the use of ASME Code Case N-702 using the BWRVIP-108 report as the technical basis for the use of ASME Code Case N-702.

Response

Based on discussions between Detroit Edison representatives and NRC staff on May 4 and June 10, 2010, Detroit Edison is clarifying the intent to apply the alternative sample population criteria provided for in Code Case N-702 using the BWRVIP-108 report as the technical basis to reduce the ASME'Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius selections.

2. The fourth sentence in the first paragraph of Section 5 "Proposed Alternative and Basis for Relief" of Relief Request RR-A37 states, "The alternative requirements will be applied to the control rod drive return nozzle which has been cut and capped at Fermi 2 and therefore is not subject to the thermal fatigue issues which might otherwise be a concern for operational control rod drive return lines."

Enclosure to NRC-10-0060 Page 2 Whereas, ASME Code Case N-702, states in part, "This Case excludes BWR feedwater nozzles and control rod drive return line nozzles."

In order to expand the scope of ASME Code Case N-702 to include capped control rod drive return line nozzles in your application you need to provide an analysis as extensive as the one performed in BWRVIP-108, considering all stresses, including the reduced thermal fatigue stresses and using assumptions relevant to this line. Alternatively, you may remove this item from the request.

Response

Based on discussions between Detroit Edison representatives and NRC staff on May 4 and June 10, 2010, Detroit Edison is hereby withdrawing the request for including the capped CRD return nozzle in Relief Request RR-A37.

Enclosure to NRC-10-0060 Page 3 NRC Request for Additional Information Regarding Relief Request RR-A30 The Nuclear Regulatory Commission (NRC) staff has reviewed the information provided by Detroit Edison for Fermi 2 in its letter dated January 20, 2010, and has determined that additional information is necessary to complete the review of Relief Request RR-A30. Based on the staff's review, please provide a response which addresses the following questions.

1. Per Regulatory Guide (RG) 1.193, Rev 2 (Oct 2007), Code Case N-578 is listed as an unacceptable Section XI Code Case. Please explain the use of Code Case N-578 in Relief Request RR-A30. Additionally, Code Case N-578 was revised by the ASME Code and version N-578-1 was approved on March 28, 2000. In light of the fact that Code Case N-578 does not include item number R1.20 for Elements not Subject to a Damage Mechanism, should the reference to this Code Case be to N-578-1?

Response

The Fermi 2 Risk Informed Inservice Inspection Program is based on EPRI TR-112657 Rev. B-A which included consideration of Code Case N-578. Detroit Edison also included the updated ASME Code Case Item number R1.20 in the Risk Informed Inservice Inspection (RI-ISI) Program for the Third Ten-Year Interval for welds that have no degradation mechanism. Therefore, it is also appropriate to reference Code Case N-578-1.

2. How does the update from the ASME Code, 1989 Edition, No Addenda to the 2001 Edition through the 2003 Addenda impact the Risk-Informed Inservice Inspection (RI-ISI) program?

Response

The update from ASME Section XI 1989 Edition, No Addenda to the 2001 Edition through the 2003 Addenda has no impact on the RI-ISI Program because the weld sample population is selected in accordance with the RI-ISI Program based on the applicable degradation mechanism(s) and risk category. The only requirement of the ASME Code that applies is the UT exam methodology which has not changed as a result of the update.

3. How does the proposed 3rd 10-year interval RI-ISI program impact the augmented inspection program that is implemented at Fermi 2 in accordance with the guidelines contained in the BWR Vessel and Internals Project Report No. BWRVIP-75?

Enclosure to NRC-10-0060 Page 4

Response

This topic was first addressed in the version of Relief Request RR-A30 that was submitted and approved for the Second Interval per NRC letter to Detroit Edison dated September 10, 2001 (ML012400331). The approach remains the same for the Third Interval. As such, BWRVIP-75-A will be implemented concurrently with the RI-ISI Program for the examination of Class 1 welds that are subject to intergranular stress corrosion cracking (IGSCC). BWRVIP-75-A specifies examination extent and frequency requirements for austenitic stainless steel welds that are classified as Categories "A" through "G", depending on their susceptibility to IGSCC. In accordance with EPRI TR-112657, piping welds identified as Category "A" are considered resistant to IGSCC, and, as such are assigned a low failure potential provided no other damage mechanisms are present. Therefore, the RI-ISI application subsumes the Class 1 IGSCC Category "A" welds. The BWRVIP-75-A augmented inspection program for the other piping welds subject to IGSCC is not affected by the RI-ISI submittal and welds are examined at the percentage established in BWRVIP-75-A.

4. Item 3 on Page 3 of RR-A30 describes welds that were added to the RI-ISI population to replace welds that would have had limited examination coverage. The added welds were described as "single side access welds." Please explain what is meant by "single side access welds," using figures if necessary. What coverage is expected on these welds? If coverage on each weld is less than essentially 100% (essentially 100%, as clarified by ASME Code Case N-460, Alternative Examination Coverage for Class 1 and Class 2 Welds, is greater than 90% coverage of the examination volume), please explain why the combination of coverage for two welds is an adequate/appropriate substitution for one weld that was originally selected for the RI-ISI program.

Response

Intergranular stress corrosion cracking (IGSCC) is a cracking mechanism of the base material (not the weld) in the heat affected zone (HAZ) of sensitized material resulting from activities such as welding or grinding that causes chromium carbide precipitation while the material is at elevated temperatures. Sensitization depletes the affected base material grain of chromium and results in an area along the edge of individual grains that can be attacked by corrosion (see micrograph below). The weld structure of an austenitic stainless steel weldment has a coarse dendritic grain structure that cannot always be reliably penetrated by ultrasonic waves. Therefore, when stainless steel welds are examined they must be scanned from both sides of the weld to claim 100% coverage.

Sometimes component configuration such as pipe to fittings (valves, tees, caps) have surface profiles such as tapers that do not provide a good flat base material scanning surface to perform the examination. Hence, the term single side examination. Because of the difficulty penetrating the austenitic weldment, the base material in the required exam volume is examined from only one side and can only be assumed to be 50%

Enclosure to NRC-10-0060 Page 5 complete. The intent of the risk program is to effectively examine not less than 10% of the welds with applicable degradation mechanisms. In the case mentioned, one weld examined from two sides is technically equivalent to two welds being examined from one side.

i.

I IGSCC in Base Material Heat Affected Zone The attached weld profile sheets provide an example of the substitutions described on page 3 of the Enclosure to Reference 2. Report No.94-066 shows the original 4 inch sweepolet to cap weld selected for examination (FW-RD-2-A1-W1). Since this weld could not be effectively examined due to the cap configuration, two other adjacent single sided welds (S.W-RS-2-A3-W4 and SW-RS-2-A3-W5, Reports RF-11-68 and RF-11-69) were selected instead of the original weld.

When a weld cannot be examined, it is common to replace this examination with another weld in the same system that is similar in size, material and has the same degradation mechanism. There are very few welds that met these criteria except for "single-sided" 4 inch welds in the same system. Single-sided welds are accessible, but only from one side of the weld. In other words only 50% coverage can be achieved because the weld can only be examined from one side. As such, they do not meet the 90% minimum offered by Code Case N-460. Due to the lack of other suitable candidate welds, examinations were performed from one side of two welds rather than from both sides of one weld.

Since the pipe size is the same for all the welds involved, an equivalent amount of weld volume is examined using this approach.

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