ML18016A375

Forwards Response to GL 97-06, Degradation of SG Internals.

ML18016A375
Person / Time
Site:	Harris
Issue date:	03/30/1998
From:	Robinson W CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-97-06, GL-97-6, HNP-98-039, HNP-98-39, NUDOCS 9804080199
Download: ML18016A375 (14)

Text

CATEGORY 1 REGULA'I~ Y INFORMATION DISTRIBUTIOL SYSTEM (RIDS)

ACCESSXON NBR:9804080199 DOC.DATE: 98/03/30 NOTARIZED: NO DOCKET FACIL:50-40( Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH.NAIF) i AUTHOR AFFXLXATION ROBINSON,W.R. Carolina Power E Light Co.

RECXP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Provides response to GL 97-06, "Degradation of SG Internals," for plant.

DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SXZE:

TITLE: OR Submittal: General Distribution NOTES:Application for permit renewal filed. 05000400 E

, RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL XD CODE/NAME LTTR ENCL PD2-1 LA 1 1 PD2-1 PD 1 1 0 FLANDERS,S 1 1 INTERNAL: ACRS 1 1 E E 1 1 NRR/DE/ECGB/A 1 1 NRR/DE/EMCB 1 1 NRR/DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS2 1 0 EXTERNAL: NOAC 1 1 NRC PDR D

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PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 14 ENCL 13

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Carolina Power & Light Company William R. Robinson PO Box 165 Vice President New Hill NC 27562 Harris Nuclear Plant SERIAL: HNP-98-039 MAR 30 1998 10 CFR 50.54(f)

United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HA'RRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 90-DAY RESPONSE TO NRC GENERIC LETTER 97-06, "DEGRADATIONOF STEAM GENERATOR INTERNALS"

Dear Sir or Madam:

Carolina Power & Light Company (CP&L) hereby responds to NRC Generic Letter 97-06, "Degradation of Steam Generator Internals," for the Harris Nuclear Plant (HNP).

Generic Letter 97-06, dated December 30, 1997, requested each licensee provide a written report within 90 days of the date of the Generic Letter that includes a discussion of any program in place to detect degradation of steam generator internals. This discussion should include a description of the inspection plans, inspection scope, frequency, methods, and equipment. If addressees currently have no program in place to detect degradation of steam generator internals, a justification is requested.

A written report providing the requested information for HNP is provided in the attachment to this letter.

Please refer any questions regarding this submittal to Mr. J. H. Eads at (919) 362-2646.

Sincerely, 9804080k%'P 980330 PDR ADOCK 05000400 P FDa AEC/aec Attachment 5413 Shearan Harris Road New Hill NC Tel 919 362-2502 Fax 919 362-2095

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Document Control Desk SERIAL: HNP-98-039 Page 2 W. R. Robinson, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are employees, contractors, and agents of Carolina Power Ec Light Company.

Notary ea My commission expires:

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yah Ry+o s c: Mr. J. B. Brady (NRC Senior Resident Inspector) >Urdu~ ~$

Mr. L. A. Reyes (NRC Regional Administrator, Region II)

Mr. S. C. Flanders (NRR Project Manager, HNP)

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Attachment to Serial: HNP-98-039 Page 1 of 6 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 90-DAY RESPONSE TO NRC GENERIC LETTER 97-06, "DEGRADATIONOF STEAM GENERATOR INTERNALS" Generic Letter 97-06, dated December 30, 1997, requested each licensee provide a written report within 90 days of the date of the Generic Letter. For the Harris Nuclear Plant (HNP), Carolina Power 8c Light Company (CP&L) provides the requested information as follows:

Re uested Item 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. The discussion 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. Ifthe addressee has performed such a review, include a discussion of the findings.

(b) Whether 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 components). If the addressee has performed such inspections, include a discussion of the findings.

(c) Whether degradation of steam generator internals has been detected at the facility, and how the degradation was assessed and dispositioned.

Res onse to Re uested Item I This response includes: (1) a discussion of industry experience related to the degradation of steam generator internals, (2) HNP's steam generator examination results to date, and (3) HNP's steam generator inspection plan. The inspection plan is subject to change based on site specific experience and evaluation of further industry experience.

Industry Experience An evaluation of the types of degradation of steam generator internals experienced in the Electricite de France pressurized water reactors (PWRs) was provided in Electric Power Research Institute (EPRI) report GC-109558, "Steam Generator Internals Degradation: Modes of Degradation Detected in EDF Units," submitted to the NRC by letter dated December 19, 1997. The Westinghouse Owners Group (WOG) has completed reviews of EPRI GC-109558 for Series 51 and Model F design steam generators. An evaluation has not been completed by the WOG for the Model D4 steam generators installed at the HNP. An evaluation report for the

Attachment to Serial: HNP-98-039 Page 2 of 6 Model D4 steam generators is expected by June 1998. Inspection recommendations for HNP have been defined on an interim basis and will be evaluated and modified as necessary in response to site specific results and industry experience. 4 The secondary side internal degradation types found in Westinghouse steam generators are identified in Table 1 at the end of this document. The HNP steam generators belong to the "Preheat Steam Generators with Carbon Steel Tube Support Plates (TSPs)" category in Table 1.

HNP Inspection Results A steam generator secondary side inspection summary for HNP is provided below.

1. During each of the refueling outages since plant startup, HNP has performed sludge lancing followed by a top of tubesheet visual inspections to verify the effectiveness of sludge lance activities and to look for, and retrieve where possible, loose parts. During these examinations, no indications of secondary side internals degradation have been discovered.

These examinations have included visual inspection of the internal blowdown piping.

2. During the last steam generator examination, performed in 1997 during Refueling Outage 7 (RFO7), a sampling of tubes (approximately 300) was evaluated for possible support plate damage utilizing the low frequency on the eddy current bobbin probe. Guidance was provided in the site specific analyst guidelines to perform this evaluation, including an eddy current reporting code, to allow the analysts to report any anomalous signals. No anomalous signals were detected during this evaluation. Had anomalous signals been reported, further interrogation, with either more advance eddy current techniques, or a visual examination would have been considered.

3. During the last steam generator examination (RFO7), HNP discovered the presence of a loose part in the preheater of steam generator A. It was necessary to enter the preheater to make efforts to find and remove the suspected loose part. Erosion of the preheater support ribs was found during this examination. The findings of this examination have been documented in a site engineering evaluation. Based on the information gathered to date, it can be qualitatively concluded that any separated rib pieces that may occur would be small, and that the erosion/corrosion pattern occurring in the ribs would be a gradual enlargement of the 1" holes with eventual coalescence of the flow holes. Tube wear rates at the top of the B tube support plate for an assumed object would not result in tube wear to a depth of greater than 40%

through-wall in less than several operational cycles.

4. During RFO4 (1992), HNP performed a secondary side visual inspection that included the.

first three tube support plates and the tube support plate wedge blocks. This examination "

revealed no secondary side degradation.

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Attachment to Serial: HNP-98-039 Page 3 of 6

5. During RFO3 (1991), HNP performed a fairly extensive secondary side visual inspection.

This examination included the following scope:

Top of tubesheet including blowdown hardware Top of the flow distribution baffle and first support plate Bottom of the preheater water box (2"'upport plate)

Top (11'") support plate, including the flow holes and tube hole ligaments Anti-vibration bar area The inspection involved entering the upper tube bundle through the primary moisture separators. One of the main purposes of this examination was to better understand the mechanisms causing anti-vibration bar wear in the HNP steam generators. The examination also included the additional areas described above. No secondary side degradation was found during this examination.

6. HNP has not performed chemical cleaning.

In addition to the visual examinations performed at the HNP, CP&L has monitored the results of secondary internals inspections performed at Commonwealth Edison's Byron Unit 1 and Braidwood Unit 1. Like HNP, Byron 1 and Braidwood 1 both had, as original plant equipment, Westinghouse Model D4 Steam Generators. These two units have performed extensive visual examinations in support of a 3.0 volt Interim Plugging Criteria (IPC). The Commonwealth Edison examinations performed at Byron 1 and Braidwood 1 included visual inspection of the vertical support bar welds, tube wrapper alignment, top-most tube support plate stay-rod nuts, and eddy current examinations focused on tube support plate integrity in the areas of the anti-rotation devices. These examinations revealed no internals degradation for the Byron 1 or Braidwood 1 Westinghouse D4 Steam Generators.

HNP Inspection Plan The following describes the inspection plan that has been implemented at HNP. Except where noted, future inspections will be completed during each refueling outage. Inspection scope and frequency may be adjusted as necessary based on site specific experience and evaluation of industry results of these inspections.

Tube Su ort Plate Erosion-Corrosion and Crackin

1. As discussed above, during the last steam generator examination, a sampling of tubes (approximately 300) was evaluated for possible tube support plate damage utilizing the low frequency on the bobbin probe. HNP is planning to expand this examination to'comply with the recommendations made by the Westinghouse Owners Group. Historical data will be available for use during each examination, including baseline data. Ifanomalous tube

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Attachment to Serial: HNP-98-039 Page 4 of 6 support plate signals are found during future examinations, then historical data will be reviewed to determine ifan active damage mechanism exists.

2. Inservice inspection will be conducted in accordance with Revision 5 of the EPRI PWR Steam Generator Examination Guidelines.

The critical area for mechanical or thermally induced support plate cracking is tentatively defined as 3 tubes around the periphery and 2 rows around the patch plate regions in each tube support plate. The critical area for ligament erosion/corrosion is the entire bundle. At a minimum, an initial sample of 20% of the tubes will be completed. HNP has performed periphery bobbin examinations for the past several outages, primarily to detect loose parts.

This examination will be expanded, during our next Steam Generator eddy current inspection, to include the recommended scope described above for detection of support plate degradation.

Wra er Dro:

To date, HNP has performed sludge lancing each refueling outage since plant startup. Due to the low amount of sludge removed in previous outages, sludge lancing is not currently planned for RFO8 (1998). Sludge lancing will be performed in future outages as necessary. Each outage that sludge lancing is performed, it will be verified that the sludge lance equipment can be inserted without interference. Ifinterference with the sludge lance equipment is detected, the lower wrapper support blocks will be visually inspected.

Wra er Crackin:

No inspection is recommended unless evidence of wrapper misposition or tube damage in the periphery of the first tube support plate is detected. Ifdegradation is detected, a visual inspection of the lower wrapper support blocks will be conducted.

U~PP k Primary and Secondary moisture separators:

1. During the upper package examination that was performed during RFO3 (1991), access to the upper package was through the primary separator(s). No degradation of these components was detected during this examination.

2. HNP has not performed a visual inspection of the secondary dryers. During RFO8 (1998),

HNP will attempt to visually inspect the secondary dryers.

The main challenge to Steam Generator tubing from degradation of the primary and secondary separators is due to loose parts. HNP will continue to monitor, through low frequency bobbin examination, for loose parts on the top tube support plate.

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. Attachment to Serial: HNP-98-039 Page 5 of 6 Transition Cone Girth Weld:

The steam generator shell will be inspected in accordance with the ASME Boiler and Pressure Vessel Code,Section XI, inservice inspection requirements.

Preheater Water Box Erosion/Corrosion A loose object originating from the water box rib could potentially cause wear to 40% depth in a steam generator tube located in the preheater T-slot in approximately 2.2 years. As a result, eddy current inspections of peripheral and T-slot tubes within the preheater will be undertaken at each scheduled outage to detect any tubes with significant tube wall degradation. HNP included the entire T-slot region in the eddy current examination during RFO7, and plans to include these tubes in future examinations as part of the periphery examination.

Re uested Item 2 Ifthe addressee currently has no program in place to detect degradation of steam generator internals, include a discussion and justification of the plans and schedule for establishing such a program, or why no program is needed.

Res onse to Re uested Item 2 The HNP program in place to detect degradation of steam generator internals is provided in the response to requested item 1.

Attachment to Serial: HN -9S-039 Page 6 of 6 Table 1 Secondary Side Internal Degradation Types In Westinghouse Design SGs SG Category: Feed Ring Preheat Feed Ring Preheat Carbon Steel Carbon Steel Stainless Stainless Degradation Type TSPs TSPs Steel TSPs Steel TSPs Erosion-Corrosion:

Moisture Separator X X Water Box NA x<'>

TSP Flow Hole/Ligaments NA NA Feed Ring/J-Tubes X X NA Cracking:

TSP Ligaments X L

rapper Near Supports"'ransition Cone Girth Weld X x<'>

Other:

Wrapper Drop" Lien d:

X= Observed in some steam generators S = Susceptible L = Low Susceptibility NA = Not Applicable (1) Various indications of possible tube degradation may be artifacts of manufacturing anomalies related to patch plate welds and drilling alignment.

(2) Various Westinghouse design features are beneficial relative to some steam generator design features of foreign manufacturers.

(3) In SG replacements with the original shell.

(4) This mechanism does not apply to the Model D3 because of the Alloy 600 inlet manifold design used.