ML20217L054
| ML20217L054 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/30/1998 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-06, GL-97-6, NUDOCS 9804070362 | |
| Download: ML20217L054 (9) | |
Text
Northern States Power Company Prairie Island Nuclear GeneratJng Plant 1717 Wakonade Dr. East Welch. Minnesota 55089 March 30,1998 10 CFR 50.54(f)
Generic Letter 97-06 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos.50-282 License Nos. DPR-42 50-306 DPR-60 Response to Generic Letter 97-06:
Steam Generator Internals Degradation On December 30,1997, the Nuclear Regulatory Commission issued the referenced Generic Letter requesting information. By this letter, Northern States Power Company is providing the required 90 day response for the P a rie Island Nuclear Generating Plant and is making no new commitments. Please cocitact Richard Pearson (612-388-1121) if you have questions regarding this issue.
Joel P Sorenser'
\\(p' Plant Manager Prairie Island b uclear Generating Plant
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c: Regional Administrator - Region Ill, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg
Attachment:
Response to NRC Generic Letter 97-06 980407o362 980330 PDR ADOCK 0500 2
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UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO.
50-282 50-306 GENERIC LETTER 97-06: Degradation of Steam Generator Internals Northern States Power Company, a Minnesota corporation, with this letter is submitting information requested by NRC Generic Letter 97-06.
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY BY
[Aoel P Sorensen Plant Manager Prairie Island Nuclear Generating Plant dz W
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On this#
day of f /
_ / 7 / O before me a notary public in and for said County, personalty ~ appeared Joel P Sorensen, Plant Manager, Prairie Island Nuclear Generating Plant; and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and bel' f the statements made in i are ue and that it is no interposed for delay.
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4 IESPONSE TO NRC GENERIC LETTER 97-06 Industry Response Background in response to the issuance of the proposed Generic Lettu, (GL) 97-06, " Degradation of Steam Generator Intemals" the industry, working through NEI, formed a task force in January 1997 to develop an action plan to coordinate an industry-wide response to the proposed GL. Participation on the task force included EPRI, licensees, and i
representatives of the owners' groups for each domestic steam generator design.
i Each owner's group, in response to the action plan, developed a program to assist their i
respective owners in assessing the susceptibility of tube damage 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 cooperation of Electricite de France (EdF),
developed the report, GC-109558, Steam GeneratorIntemals Degradation: Modes of Degradation Detectedin EdF Units. The EPRI report provides evaluations of the causal factors involved in the modes of degradation experienced in the French units.
The owners' groups used this report to gain insights into the applicability of the French experience to their own designs and operating history. This report was transmitted to the NRC via an NEl letter, dated December 19,1997.
Other attributes considered in developing the susceptibility assessment were design factors, fabrication and manufacturing techniques, plant operating history, including chemistry and related degradation, such as denting. Additionally, the owners' groups compiled and assessed collective visual, video and pertinent NDE inspection experience to further enhance their evaluations regarding the susceptibility to internals degradation.
The Westinghouse Owners Group has reviewed EPRI GC-109558 relative to the design of Series 51 steam generators and determined limited susceptibility. For plants with Series 51 steam generators, this conclusion is documented in report WCAP-15002,
" Evaluation of EDF Steam Generator Internals Degradation -Impact of Causal Factors on Westinghouse 51 Series Steam Generators," December 1997. The 51 Series Designs are the most similar to the EdF units. Prairie Island Units 1 and 2 have Westinghouse Model 51 Steam Generators. WCAP-15002, Revision 1 " Evaluation of EdF Steam Generator Internals Degradation-Impact of Causal Factors on Westinghouse 51 Series Steam Generator"(Proprietary), was transmitted to the NRC j
via NEl letter dated March 4,1998.
- WCAP-15002 docum'ents visualinspections of the plants from completed surveys of 37
Respo'nse to GL 974 NORTHERN STATES POWER COMPANY March 30,1998 Page 2 plant sites, it is concluded that the number of plants that have been inspected and the inspection results demonstrate that the causal factors identified by EdF do not jeopardize the continued operability of Westinghouse Series 51 steam generators.
Eddy current inspection of the tubes would detect any detrimental effects on the tubing due to wear caused by tube support plate (TSP) ligament degradation, loose parts, and secondary side flow distribution changes. Foreign object search and retrieval efforts are routinely performed to discover loose parts.
Within 90 days of the date of this generic letter, each addressee is required to provide a written report that includes the following information for its facility:
(1)
Discussion of any program in place to detect degradation of steam generator internals and a description of the inspection plans, including the inspection scope, frequency, methods, and equipment.
Per Table 5.2 of WCAP-15002, the Prairie Island steam generators (which contain a feed ring and carbon steel TSPs) are susceptible to:
erosion-corrosion in the moisture separator region, the tube support plate flow hole ligaments, and the feed ring holes (J-tubes are not installed at Prairie Island) cracking in the TSP ligaments, in the wrapper near supports (Iow susceptibility), and e
the transition cone girth weld.
wrapper drop (Iow susceptibility).
The following internal components have been inspected in each steam generator at Prairie Island.
visual inspection of swirl vane moisture separators and tangential exit nozzles visualinspection of feed ring hangers e
visualinspection of feed ring holes and plugs visual inspection of upper transition girth weld for signs of pitting e
remote visual inspection of feed ring and thermal sleeve internal diameter installation of sludge lancing equipment which confirms no substantial wrapper drop remote visual inspection of outer periphery of tube sheet after sludge lancing (Foreign Object Search and Retrieval) limited remote visual examination of upper TSP e
evaluation of bobbin coil eddy current data for support plate anomalies remote visualinspection of the wrapper support blocks and anti-rotation blocks e
ultrasonic inspections of the feed ring tee section.
e magnetic particle examination of the inside radius of the feedwater nozzle to shell weld region
K N,
Response to GL 97-06 NORTHERN STATES POWER COMPANY March 30,1998 Page 3 visual inspection of the thermal sleeve ends was conducted in each steam i
generator when the feedwater pipes were removed for feedwater pipe to nozzle weld modifications in 1992.
ultrasonic examination of vessel shell welds per ASME Section XI.
The only degradation observed is as folicwr Erosion of feed ring hole plugs appeared to be the result of porosity of the feed ring hole plug welds. After weld repair, no further degradation has been seen. Those inspections began in 1994.
A small amount of erosion has been observed in feed ring holes.
Erosion is occurring, but at an acceptable rate, on the leading edge of the feedwater ring thermal sleeve.
For the types of steam generator internals degradation observed at Prairie Island, it is expected that stesm generator internals degradation would be limited in extent such that the tubes will remain capable of sustaining the conditions of normal operation, including operational transients, design basis accidents, extemal events, and natural phenomena permitting the affected steam generator to perform its intended safety function.
Prairie Island Inservice insoection Plan The following inspection plan is in use at Prairie Island. Except where noted, these inspections will be completed in one steam generator each refueling outage. Trending of these inspection results may result in reduced inspection frequency in the future.
I Tube Support Plate Erosion-Corrosion and Cracking (both steam generators are inspected each outage):
1 Since the steam generator tube support plates in Prairie Island are made of carbon steel, a base line has been established employing low frequency bobbin results during the 1997 refueling outages. The technique employed is based on the EPRI Report on the " investigation of Applicability of Eddy Current to the Detection of Potentially Degraded Support Structures" dated May 1996, SG 05-003. The indications found were traced back in history to show that there is not an active degradation mechanism.
2.
In service inspection for tube support plate anomalies are conducted in the spirit of the latest Revision of the EPRI PWR Steam Generator Examination Guidelines.
I i
Respo'nse to dL 97-06 NORTHERN STATES POWER COMPANY March 30,1998 Page 4 The critical area for mechanical or thermally induced support plate cracking is defined as 3 tubes around the periphery and 2 rows around the patch plate regions in each support plate. An initial sample of 100% of these tubes has been completed.
The critical area for ligament erosion / corrosion is the entire bundle. An initial sample of 20% of the tubes is recommended. An initial sample of 100% of the tubes has been completed.
Wrapper Drop:
1.
Verify that the sludge lane equipment can be inserted without interference.
2.
If interference with the sludge lance equipment is detected, the lower wrapper support blocks should be visually inspected. A remote visualinspection of all wrapper blocks in 1997 showed no degradation of the wrapper block supports.
Wrapper Cracking:
No inspection is recommended unless evidence of wrapper misposition or tube damage in the periphery of the first tube support plate is detected, if degradation is i
detected, a visual inspection of the lower wrapper support blocks will be conducted.
No wrapper cracking was observed at the wrapper support blocks at Prairie Island during remote visual inspection in 1997.
Upper Package:
Primary and Secondary moisture separators, feedring (feed ring hanger, feed ring holes and plugs, T-section, reducer, backing ring and thermal sleeve)- these components are visually inspected in one steam generator each refueling outage.
Transition Cone Girth Weld:
in addition to the ASME Section XI inservice inspection requirements, a visual e
I inspection for pitting is conducted on the inside diameter of the upper transition cone girth weld in one steam generator each refueling outage.
Feedwater Nozzle:
Degradation of.the thermal sleeve may affect the feedwater nozzle. The thermal sleeve is inspected using remote visual inspection in one steam generator each refueling outage. Loose parts monitoring and in service inspection requirements for the feedwater nozzle will be continued.
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Respo'nse to dL 97-06
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NORTHERN STATES POWER COMPANY March 30,1998 Page 5 -
1 For song term programmatic guidance relative to monitoring secondary-side I
degradation, Prairie Island plans to follow the requirements of the NEl Steam Generator Initiative as outlined in NEl 97-06 Steam' Generator Program Guideline, Section 3.9
" Maintenance of Steam Generator Secondary-Side Integrity".'
The di'Acussion should include the following information:
(a)
Whether inspection records at the facility have been reviewed -
for indications of tube support plate signal anomalies from eddy-current testing of the steam generator tubes that may be indicative of support plate damage or ligament cracking. If the addressee has performed such a review, include a discussion of the findings.
The bobbin coil eddy current inspection records for both steam generators in both units have been reviewed for indications of TSP signal anomalies that may be indicative of support plate damage or ligament cracking. This special evaluation of bobbin coil data' for tube support plate anomalies was done in response to NRC Information Notice 96-038 Supplement 1 and met the requirements of the WOG recommendations in j
Westinghouse letter WOG-97-186, " Transmittal of NEl Sponsored Steam Generator intemals Degradation Interim inspection Guidelines".
During the January 1997 Unit 2 refueling outage and the October 1997 Unit i refuehng
- outage, all TSP intersechons were examined with the bobbin coil eddy current probes.
The bobbin coil data was analyzed to determine if indications of possible TSP ligament anomalies were present. When bobbin coil data identified anomalies at TSPs, rotating coil technology (RPC) was used to reexamine the suspect locations.
No TSP anomalies were found in the Unit 2 steam generators.
During the 1997 Unit 1 steam generators inspection, examination of the bobbin coil possible support plate indications with RPC identified 6 TSP intersections in 11 SG and 3 TSP intersections in 12 SG with indications of possible TSP ligament " cracks".
i Review of the boboin coil history determined that these indications were present in 1996 and 1988 and had not significantly changed. The + Point" eddy current probe characterization showed that none of these indications reflected significant missing ligaments. The largest possible ligament gap was less than 50 degrees.
These indications were similar to those indications found in the eddy current qualification mockup which contained EDM simulations of cracked TSP ligaments s
t
Respo'nse to GL 97-06 NORTHERN STATES POWER COMPANY March 30,1998 Page 6.
' (Westinghouse mockup built for EPRI to qualify ET for examination of TSP ligament cracks). There was no indication of missing TSP ligaments nor of large circumferential extent of the indications. Three indications in 11 SG are at patch plate locations. Three of the indications in 11 SG and the three indications in 12 SG are in the outer periphery suspect locations identified in Westinghouse letter WOG-97-186. The bobbin coil data for these possible support plate ligament indications was reviewed from 1988 and found to be unchanged. Also there was no detectable tube, degradation at the locations of the possible degraded TSP ligaments. The affected tubes were left !n service.
c Westinghouse provided an " Assessment of Tube Integrity: Effect of Ligament Gaps in Tube Support Plau for Prairie Island 1; October 1997" as documented in
- Westinghouse letter NSD-E-TAP-0095 dated November 15,1997 which stated that it is likely that these conditions reflect steam generator as-built conditions which resulted from misalignment of drilling of flow holes or tube holes as has been visually verified at Diablo Canyon.
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(b)
Whe+her visual or video camera inspections on the secondary side of the steam generators have been performed at the facility to gain information on the condition of steam generator internals (e.g., support plates, tube bundle wrappers, or other i
components)._ If the addressee has performed such Inspections, include a discussion of the findings.
4 Remote visual examinations from the outside of the tube bundle wrapper of all of the
. steam generator wrapper support lugs and wrapper support block window areas and anti-rotation keys were conducted on each steam generator in 1997.
i Only one significant anomaly was found. An apparent discontinuity at a wrapper support lug window was found at one of the 21 SG wrapper support lugs. Each of the six wrapper supports consist of one support lug which is welded to the SG shell and two support blocks, one above and one below the support lug, which are welded to the
. wrapper. During fabrication, the wrapper support blocks are inserted from the inside of the wrapper through two windows at each iocation. The design weld is a 1/4 inch fillet weld around the support blocks on the inside of the wrapper. There is a 3/4 inch ligament between the two windows which is normally continuous. The discontinuity observed was at the ends of the ligament and it appeared that the ligament had been
. removed and then replaced during the fitting process resulting in a weld or partial weld observed during the inspection. The inspection was done on the outside of the wrapper.
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Respo'nse to GL 97-06 NORTHERN STATES POWER COMPANY March 30,1998 Page 7 1
Upon further review, there were two locations in 21 steam generator and two locations in 22 steam generator that had indications that the ligament was removed and replaced during the support block installation process. There was no indication of wrapper block rotation or crack like inoications emanating from the corners of the blocks which would indicate degradation of the fillet welds. Westinghouse reviewed the video tape of the examination and determined the anomalous condition of the wrapper support block ligaments is believed to be consistent with the normal manufacturing process.
All four steam generators have been inspected for signs of wrapper drop at the handhole elevations. No sign of wrapper drop was found. Measurements of the wrapper height at the handhole have been taken.
Limited remote visual examinations of the top TSPs have been concucted in each steam generator. There is no indication of wastage at the TSP flow holes.
Each refueling outage, an upper internals inspection is conducted in at least one steam generator inspection of feed ring hangers, feed ring flow holes, and feed ring backing rings has found no significant anomalies.
J (c)
Whether degradation of steam generator internals has been detected at the facility, and how the degradation was assessed and dispositioned.
Some porosity was observed in the feed ring hole plug weld inspections which required weld repair. The welds were repaired to prevent potential loose parts. Some erosion was observed in weld backing rings during the feedwater nozzle repairs conducted in 1992. The backing rings were repaired by welding to prevent loose parts. No other degradation of the steam generator internals has been detected.
(2)
If the addressee currently has no program in place to detect degradation of steam generator inte;nals, include a discussion and justification of the plans and schedule for establishing such a program, or why no program is needed.
The Prairie Island program to detect degradation of steam generator internals is described above.
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