ML20217E300
| ML20217E300 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/24/1998 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-06, GL-97-6, NUDOCS 9803300445 | |
| Download: ML20217E300 (8) | |
Text
- a.
Dave Morey Southern N. clear
- . O Vice President Operating Company Farley Project P.O. Box 1295 Birmingham. Alabama 35201 Tel 205.992.5131 SOUTHERN N
March 24,1998 Energy toScrveYourWorld" Docket Nos.: 50-348 10 CFR 50.54(f) 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Joseph M. Farley Nuclear Plant Response To Generic Letter 97-06. Decradation Of Steam Generator Intemals Ladies and Gentlemen:
On December 30,1997, the Nuclear Regulatory Commission issued Generic Letter 97-06,
" Degradation of Steam Generator Intemals". This letter was issued to: (1) communicate fm' dings of damage to steam generator intemals at foreign PWR facilities; (2) alert licensees to recent findings of damage to tube support plates at a U.S. PWR facility; (3) emphasize the importance of performing comprehensive examinations of steam generator intemals to ensure structural integrity is maintained in accordance with 10 CFR Part 50, Appendix B requirements; and (4) request information necessary to ensure that each facility is in compliance with its current licensing bases.
The nuclear power industry recently voted to adopt an initiative requiring each utility to meet the intent of the guidance provided in NEI 97-06," Steam Generator Program Guidelines," no later than the first refueling outage after January 1,1999. As discussed in NEI 97-06, each utility is required to monitor secondary-side steam generator components if their failure could L
prevent the steam generator from fulfilling its intended safety-related function. Farley Nuclear Plant intends to follow the requirements of NEI 97-06.
l Provided as an enclosure is the Farley Nuclear Plant specific information requested by Generic Letter 97-06.
1 There are no new commitments in this letter.
I 9803300445 900324
[
PDR ADOCK 05000348 P
PDR j
i 4
l 1
O, U. S. Nuclear Regulatory Commission Page 2
'If you have any questions, please advise.
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY Dave Morey Sworn to andsubscribed before me this olY day of b b Y l998
[
is k
Notary PubliF My Commission Expires: //th,' + -
4 ddd/
'u JEF/ REM /maf:GL9706B. DOC Enclosure cc:
Mr. L. A. Reyes, Region II Administrator Mr. J. I. Zimmerman, NRR Project Manager Mr. T. M. Ross, Plant Sr. Resident Inspector Dr. D. E. Williamson, State Department of Public Health l
l 1
4 e
ENCLOSURE s
l l
I
1
_)
4 Generic Letter 97-06 Response
Background
In response to the issuance of a proposed Generic Letter on degradation of steam generator internals, NEI formed the Steam Generator Internals Task Force in January 1997. The i
. purpose of the task force was to develop a coordinated industry-wide response to the secondary-side degradation issues identified in the proposed Generic Letter. Panicipation on the task force included EPRI, licensees, and representatives of the vendors and owners groups for each domestic steam generator design. The task force developed an action plan.
- Each owners group initiated a program to assist its respective owners in assessing the susceptibility of tube damage and loss of decay heat removal capability due to secondary-side
- degradation. An integral component in this assessment was an appreciation of the applicability of the degradation found in the French units to domestic steam generators. EPRI
- responded to this need and with the assistance of Electricite de France (EdF) developed the j
report, GC-109558, Steam GeneratorInternals Degradation: Modes ofDegradation
' Detectedin EdF Units. NEI transmitted this report to the NRC via an NEI letter, dated
- December 19,1997.
- In addition to the review of the EdF degradation casual factors, the owrers group r
susceptibility assessments included consideration of design factors; fabrication and manufacturing techniques; plant operating history, including chemistry; plant inspection experience; and related degradation, such as denting. As part of the inspection experience review, the owners groups compiled and assessed collective visual, video and pedinent NDE
- inspection experience information to further enhance their evaluations regarding the susceptibility to internals degradation.
3 As a result of an NEI meeting with the NRC in May 1997, the owners groups developed preliminary safety and susceptibility assessments relative to the design and operating history of their fleet. Westinghouse Owner's Group WCAP-15002," Evaluation of EdF Steam j
Generator Internals Degradation - Impact of Casual Factors on Westinghouse 51 Series Steam Generator " dated December 1997, provides reasonable assurance that degradation ofinternals 4
- has not compromised steam generator tube integrity nor decay heat removal capability.
' Moreover, steam generator tube eddy current inspection'can detect any detrimental effects due to wear caused by tube support plate ligament degradation, loose parts, or secondary-side flow
- distribution changes. In addition, steam generator secondary-side foreign object search and
. retrieval (FOSAR) efforts can help identify problems.
4 The focused efforts of the NEI task force has provided guidance and information necessary for licensees to adequately address the potential issues regarding steam generator internals degradation.
This response provides the information for Farley Nuclear Plant requested by Generic Letter 97-06.-
Generic Letter 97-06 Response Page 2 Joseph M. Farley Nuclear Plant Item 1 (1) Discussion ofanyprogram in place to detect degradation ofsteam generator internals
- and a description ofthe inspection plans, including the inspection scope, frequency, methods, andequipment.
The discussion should include thefollowing information:
a) Whether inspection records at thefacility have been reviewedfor indications of tube support plate signal anomaliesfrom eddy-current testing ofthe steam generator
' tubes that may be indicative ofsupportplate damage or ligament cracking. Ifthe addressee hasperformedsuch a review, include a discussion ofthefindings.
b) Whether visual or video camera inspections on the secondary-side ofthe steam generators have been performed at thefacility to gain information on the condition of the steam generator internals (e.g., supportplates, tube bundle wrappers, or other components). Ifthe addressee has performedsuch inspections, include a discussion of '
thefindings.
I c) Whether degradation ofsteam generator mternals has been detected at thefacility, and how the degradation was assessed and dispositioned.
RESPONSE
Inspection Plan As described in WCAP-15002, " Evaluation of EdF Steam Generator Internals Degradation -
4 Impact of Casual Factors on Westinghouse 51 Series Steam Generator " dated December 1997, the following secondary components were identified as potential areas for degradation.
The associated inspection plan for each degradation for Farley Nuclear Plant is described below. Except where noted, these inspections will be completed each refueling outage.
Tube Support Plate Erosion-Corrosion and Cracking:
As discussed in the PWR Steam Generator Examination Guidelines: Revision 5, the critical area for mechanical or thermally induced support plate cracking is defined as 3 tube columns around the periphery and 2 rows / columns around the patch plate regions in each support plate.
The critical area for ligament erosion / corrosion is the entire tube bundle.
Farley Nuclear Plant has conducted inspections for cracked steam generator tube support
. plates for both Units I and 2. The approach taken for both units involved conducting a 20%
sample bobbin eddy current inspection throughout each steam generator at each tube support i plate concentrating on patch plate and periphery regions. Rotating coil (Plus Point) g[-
characterization was performed on bobbin indications. A review of prior bobbin eddy cunent data was conducted to establish the possible time of existence for any indications detected.
i l.: '
^
q Generic Letter 97-06 Response Page 3 Joseph M. Farley Nuclear Plant Data reviewed during the Unit 2 eleventh refueling outage inspection in the Fall of 1996
. indicated a total of 29 tube support plate indications. A total of seven locations contained
- ligament indications in the patch plate region. A review of prior eddy current data determined these indications have been in existence since as early as 1986. Possible " missing" ligament
. material was reponed by the rotating coil from three patch plate locations. None of these
- tubes contained indications requiring implementation of the Attemate Repair Criteria. Two tubes adjacent to missing ligament material were conservatively plugged even though the tubes' were not found to be defective. No visual inspections were performed relative to the missing ligament material.
Data recorded during the Unit I fourteenth refueling outage inspection in the Spring of 1997 identified a total of 39 tube suppod plate indications confirmed through Plus Point examination. None of the reported indications showed any loss in ligament suppod. The confirmed indications were evaluated relative to their impact on tube integrity, alternate repair criteria, and tube support plate loading considerations. Based on the completed engineering evaluation, no tube stabilization or plugging was required.
During the Unit I fourteenth and Unit 2 eleventh refueling outages, the impacts of the reported indications observed on pressure boundary integrity and operational concerns for both the Unit I and 2 inspection results were evaluated relative to Alternate Repair Criteria.
The structural evaluation for in-plane and out-of-plane loading showed no significant effect on the overall structural integrity of the tube support plate. The structural evaluation also identified no localized plate deformations that could lead to loss of flow area under LOCA +
SSE loads. Based on these inservice inspections results, there was no evidence of tube degradation or wear associated with the reported tube support plate hole indications.
Continued surveillance for potential missing ligament material in Unit 2 and for further
' indications in Unit 1 is cunently scheduled for the next refueling outage for each unit.
However, the need and schedule for subsequent inspections will be based on an evaluation of the current inspection results in accordance with the Westinghouse Owners Group inspection
. program.
~ Wrapper Dron:
The wrapper drop is best detected by the ability to install sludge lance equipment. Farley Nuclear Plant routinely conducts sludge lance activities and has verified that the sludge lance equipment can be inserted without interference. No instances of failure to install the sludge lance equipment due to wrapper drop have been detected at Farley Wranoer Crackinn:
Since Farley Nuclear Plant has detected no evidence of wrapper misposition or tube damage in the periphery of the first tube support plate for either Unit 1 or 2, no inspection for wrapper cracking is being performed.
1 s.
' Generic lietter 97-06 Response Page 4
. Joseph M. Farley Nuclear Plant c
Upoer Packane:
The Farley Nuclear Plant steam generator upper package includes the primary and secondary
. moisture separators, and the feedring (including J-tubes, carbon steel feedring adjacent to J-tubes, T-section, reducer, backing ring, and thermal sleeve).
The Farley Nuclear Plant steam generator upper package is inspected and monitored in
. accordance with the Flow Accelerated Corrosion (FAC) management program. Both visual and ultrasonic NDE techniques are utilized on the above upper package components.
These various inspections have detected no evidence of significant structural deterioration.
Feedwater Thermal Liner:
Thermal liner degradation is most effectively monitored via loose parts monitoring. Plant Farley utilizes FOSAR for locating and retrieving loose parts, which may include thermal liner components. There are no Section XI examination requirements or examinations performed on the liner. Some visual intemal examination of the secondary side internal.
i components have been performed as part of our FAC program. These various inspections j
have detected no evidence of significant structural deterioration.
j Transition Cone Girth Weld:
i Transition cone welds (six total, one per steam generator) are inspected per Section XI. Plant Farley augments the Section XI requirement by examining one steam generator transition cone weld each 40-month period. Manual ultrasonic examinations have been performed in accordance with standard ASME Section XI techniques. These various inspections have detected no evidence of significant structural deterioration.
I Feedwater Nozzle:
Feedwater nozzle / reducer welds (six total, one per steam generator) are inspected at Plant Farley per Section XI and by augmented inspection by performing ultrasonic examination each outage. The feedwater nozzle to reducer has been examined with automated equipment and with manual UT techniques. These various inspections have detected no evidence of significant structural deterioration.
. General Farley Nuclear Plant has conducted numerous visual inspections on the secondary-side of the
. steam generators. Included in the past scope of exams are: Flowslot photography, Secondary
' Inspection Device (SID), CECIL inspection and sludge removal technology, FOSAR i
activities, Welch; Allyn camera inspections, In-bundle inspections, and upper steam drum
. inspections. These various inspections have detected no evidence of significant structural deterioration.
I e _:
G:neric Letter 97-06 Response Page 5 Joseph M. Farley Nuclear Plant Item 2 (2). Ifthe addressee currently has noprogram in place to detect degradation ofsteam generator internals, include a discussion andjustification oftheplans andschedulefor establishing such a program, or why no program is needed.
Response
Farley Nuclear Plant has a program to detect degradation of steam generator internals; therefore, Item 2 of Generic Letter 97-06 does not apply.
.i e