ML031560413
| ML031560413 | |
| Person / Time | |
|---|---|
| Site: | Ginna  (DPR-018) |
| Issue date: | 05/29/2003 |
| From: | Mecredy R Rochester Gas & Electric Corp |
| To: | Clark R Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML031560413 (6) | |
Text
Robert C. Mecredy Vice President Nuclear Operations May 29, 2003 Mr. Robert L. Clark Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Response to Request for Additional Information Associated with the 2002 Steam Generator Inservice Inspection Report Rochester Gas and Electric Corporation R.E. Ginna Nuclear Power Plant Docket No. 50-244
References:
(1)
Letter from R.L. Clark, NRC, to R.C. Mecredy, RG&E,
Subject:
Requlest for Additional Information Regarding R.E. Ginna Nuclear Power Plant 2002 Steam Generator Inservice Inspection Report (TAC No. MB6467),
dated May 1, 2003.
Dear Mr. Clark:
By the above reference, the NRC staff requested additional information regarding the results of the 2002 steam generator inservice inspection at the R. E. Ginna Nuclear Power Plant. of this letter provides the requested information. There are no commitments associated with this response.
If you should have any questions regarding this submittal, please contact Mr. Tom Harding, 585-771-3384.
Ver 47uly yours, Robert C. Mecred Attachments: 1.
Response to NRC Request for Additional Information (RAI) Dated May 1, 2003 An equal opportunity employer 89 East Avenue I Rochester, NY 14649 tel (585) 546-2700 www.rge.com An Energy East Company Always at Your Service
Subject:
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Mr. Robert Clark (Mail Stop 0-8-C2)
Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector Response to NRC Request for Additional Information (RAI) Dated May 1, 2003 The response to the RAI will be structured as follows. The items in bold italics below are the questions provided by the NRC in the RAI dated May 1, 2003. A response to each item is then provided by RG&E.
- 1.
In Section 8.2 of Attaclment 1 to your October 17, 2002, submittal it was indicated that some thzermally treated alloy 690 tubes were pluggedfor preventive reasons su chi as proximity to baffle plate edges. Please clarify thle concer n witht respect to te baffle plate edges.
Response
Section 8.2 of Attachment 1 to the October 17, 2002 submittal was a discussion of the Rochester Gas & Electric (RG&E) review of industry experience with Alloy 690TT tubing performed as part of the degradation assessment process.
The referenced tube was in a replacement steam generator manufactured by another vendor at a plant belonging to another utility. This tube was not considered to be relevant to the Ginna degradation assessment since there are no flow distribution baffle plates in the Ginna steam generators. Potential mechanical degradation mechanisms specific to the Ginna replacement steam generators are addressed in detail in Section 10 of Attachment 1 to our October 17, 2002, submittal.
RG&E is unable to comment further on this issue since we do not have detailed information on the design of this specific steam generator or the information that was used to make the plugging decision.
- 2.
Section 9.2.1 ofAttachment 1 to your October 17, 2002, letter discusses a tube proximity issue. From te description provided, it appears that thie outermost tube in a column and thle tube below it may be in close proximity as a result of the fabrication procedure. However, in other sections (Section 2.4 and Section 9.2.3) the report appears to be discussing tie potentialfor wear to occur as a result of the tube proximity issue in all tubes in a given column. Provide information that would clarify if the tube proxim ity issue is one that would be applicable to all tubes in a column or a subset of tubes in a column. Also, provide infornation clarifying whietter the wear observed in McGuire 1 and St. Lucie 1 occurred only in te outermost columyis, tte tube in an adjacent column, or along tte entire column and how this infornationi applies to Ginna.
Response
The "tube proximity" issue discussed in section 9.2.1 is different from the "localized typical" wear discussed in section 9.2.3 of Attachment 1 to our October 17, 2002, submittal.
The tube proximity issue is only applicable to peripheral tubes and is a result of not properly establishing the profile of the outermost tube prior to setting and welding "J-tabs" in the u-bend structure, as discussed in section 9.2.1. This can result in a shift in the outermost tube profile when the generator is upended, and potential reduction in the gap between the outermost tube and the tube below it. As discussed in section 10.2.4 of to our October 17, 2002, submittal, several tubes at Ginna are known to exhibit this phenomenon. These were inspected during the 1997, 1999, and 2002 refueling outages using bobbin and rotating probes, and no wear has been detected.
The localized typical wear discussed in section 9.2.3, and the additional expansion criterion discussed in section 2.4, address a different potential degradation mode. This is the wear that has been seen at St. Lucie and McGuire, which have replacement steam generators from the same vendor as Ginna. This wear afflicts multiple tubes within the columns adjacent to a specific u-bend fan bar, possibly as a result of some local distortion in the bar or how it is attached at the u-bend superstructure which results in an increased tube-to-support gap. This wear can occur deep within the bundle and is not associated with the tube proximity issue at the bundle periphery. The main reason for calling this out as a different mechanism is because of the potential to afflict the columns adjacent to a column where wear indications are found. Because Ginna samples every other column to achieve 50% bobbin coverage, it was considered prudent to expand the inspection to adjacent uninspected column(s) if any u-bend wear indications were found.
No wear of any kind was found in the Ginna replacement steam generators during the 2002 inspection.
- 3.
In Section 10.6.5.a ofAttactment 1 toyour October 17, 2002, letter tere vas some discussion about "tot-leg accessible indications. " Please clarifyv what is meant by ttat statement, especially if it has any implications tat some tubes cannot be inspected because tey are not accessible.
Response
The statement does not mean that there are tubes that cannot be inspected. Every in-service tube in each Ginna steam generator is accessible.
RG&E currently uses a single manipulator per steam generator for inspections to minimize schedule, dose, and inspection cost. This manipulator is located in the hot leg channel head since any temperature-driven degradation mechanism would be expected to be seen in the hot leg first. For rotating probes, the motor unit cannot be passed over the u-bend and therefore extension units must be used. These limit the ability to effectively inspect the entire leg opposite the leg in which the manipulator is mounted using rotating probes.
The referenced statement is intended to communicate that a 20% sample of manufacturing burnish mark (MBM) calls greater than five volts, accessible using rotating probes from the hot leg, is performed. Since any potential degradation at MBM sites is very likely to be temperature driven, hot leg degradation would be expected prior to cold leg degradation.
There is nothing to prevent performance of rotating probe inspections in the cold leg, although this would require relocation of the manipulator from the hot leg channel head to the cold leg channel head.
Ginna used array coil probes on a limited number of tubes during the 2002 inspection outage and may use these more extensively in the future. These probes are capable of providing an enhanced inspection and can access the entire tube from a single channel head.
- 4.
Concerning te scored tube hole discussed in Section 10.2.2 of Attaclmenit I to your October 17, 2002, letter, please clarify wletler any special examinatiot requirements have beenz, or will be, implementedfor this tube.
Response
The subject tube was inspected using a +Point rotating probe during the 2002 outage to obtain a baseline for future comparison. The eddy-current data indicated a benign tube geometry along the length of the tube within the tubesheet. As a potential precursor to future degradation, it is expected that this tube will be inspected in the area of interest at each subsequent inspection.
A recent peer-review of the Ginna steam generator program recommended that all potential degradation precursors, including the scored tube hole, be addressed consistently in the degradation assessment, similar to the way in which MBMs are currently treated. This recommendation is currently being tracked for inclusion in the degradation assessment to be performed prior to the next inspection.
S.
On page 9 of Attachment 2 to your October 17, 2002, letter there was reference to rotating probe examinations of "gripper locations. " Please clarify wlat gripper locations are.
Response
During a pre-mobilization meeting prior to the 2002 outage, RG&E reviewed the vendor equipment, including the grippers used to secure the inspection manipulator into the tube sheet. Each gripper has a mandrel which spreads metal fingers when retracted. These fingers grip the inside of the tube and may possibly mark the inside diameter surface.
The manipulator gripper tube locations have been consistent from outage to outage, with two grippers on the manipulator and two manipulator locations required for complete tubesheet coverage. RG&E inspected the subject tubes, and the vendor now uses an elastomeric ring around the gripper as a precaution to further minimize the possibility of marking the tube inside diameter surface.