ML20217N391
| ML20217N391 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/30/1998 |
| From: | Jain S DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-06, GL-97-6, L-98-009, L-98-9, NUDOCS 9804090091 | |
| Download: ML20217N391 (15) | |
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gh Boav Valay Power Station Shippmgport, PA 15077 0004 SUSHIL C. JAIN (412)393 5512 Senior Vice President Fax (724) 643-8069 Nuclear Services Nuclear Power Division March 30,1998 L-98-009 U. S. Nuclear Regulatory Commission
. Attention: Document Control Desk
/ Washington, DC 20555-0001
Subject:
Beaver Valley Power Station, Unit No. I and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Generic Letter 97-06 Response The attachment to this letter provides the Beaver Valley Power Station, Unit No. I and No. 2 response to the Required Information section of Generic Letter 97-06,
" Degradation of Steam Generator Internals."
Questions concerning this response may be directed to Mr. J. Arias, Director, Safety & Licensing at (412) 393-5203.
Sincerely, Sushil C. Jain i
c:
Mr. D. S. Brinkman, Sr. Project Manager Mr. D. M. Kern, Sr. Resident Inspector s
Mr. H. J. Miller, NRC Region I Administrator
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.i AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA)
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COUNTY OF BEAVER
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Subject:
Beaver Valley Power Station, Unit No.1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Generic Letter 97-06 Response Before me, the undersigned notary public, in and for the County and Commonwealth aforesaid, this day personally appeared Sushil C. Jain, to me known, who being duly sworn according to law, deposes and says that he is Senior Vice President, Nuclear Services of the Nuclear Power Division, Duquesne Light Company, he is duly authorized to execute and file the foregoing submittal on behalf of said Comoany, and the statements set forth in the submittal are true and correct to the best of his knowledge, information and belief.
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Sushil C. Jain Subscribed and sworn to before me on thid/) day of @/sf 6/[, g'f[
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Notary Puhlin Sheita M. att re, Notary Public j My f s onExpiresYept 26, 98
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ATTACHMENT Beaver Valley Power Station, Unit No. I and No. 2 Response to NRC Generic Leuer 97-06, " Degradation of Steam Generator Internals" I.
BACKGROUND AND INTRODUCTION:
I In response to the issuance of a proposed Generic Letter on degradation of steam generator internals, Nuclear Energy Institute (NEI) formed the Steam Generator Internals Task Force in January 1997. The 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. Participation on the task force included EPRI, licensees and representatives of vendors and owners groups for each domestic steam generator design. This task force developed an action plan for the steam generator internals degradation issue.
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 understanding 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 report, GC-109558, Steam GeneratorInternals Degradation: Modes ofDegradation Detected in 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 steam generator designs and operating history. NEI transmitted this report to the NRC via an NEI letter dated December 19,1997. EPRI provided copies directly to the Steam Generator Management Program Technical Advisory Group representatives.
In addition to the review of the EdF degradation casual factors, the susceptibility assessments included consideration of design factors; fabrication and manufacturing techniques; plant opeating history, including chemistry; plant inspection experience; and related degradation, such as denting. As part of the inspection experience review, the owners groups complied and assessed collective visual, video and pertinent NDE inspection experience information to further enhance l
their evaluations regarding the susceptibility to internals degradation.
}
i The NEI Task Force met with the NRC in May 1997 to gain a better understanding of the safety concerns discussed in the proposed generic letter. As a result of these efforts, the owners groups developed preliminary safety and susceptibility assessments relative to the design and operating history of their fleet. These assessments provide reasonable assurance that degradation ofinternals has not compromised steam generator tube integrity or decay heat removal capability.
Attachment Response to Generic Letter 97-06 Page 2 The industry through the uwd Jons of the NEI Task Force has provided guidance and information necessarf. %3ces to adequately address the potentialissues regarding steam generator intert ah-Ap 7 don.
Generic Letter 97-Oo (6L), " Degradation of Steam Generator Internals" was issued December 30,1997 to: (1) again alert addressees to the previously communicated findings of damage to steam generator internals, namely, tube support plates and tube bundle wrappers, at foreign PWR facilities; (2) alert addressees to recent fmdings of damage to steam generator tube support plates at a U.S. PWR facility; (3) emphasize to addressees the importance of performing comprehensive examinations of steam generator internals to ensure steam generator tube structural integrity is maintained in accordance with the requirements cf Appendix B to 10 CFR Part 50; and (4) require all addressees to submit information that will enable the NRC staff to
- verify whether addressees' steam generator internals comply with and conform to the current licensing bases for their respective facilities.-
The Westinghouse Owners Group has reviewed EPRI GC-109558 relative to the design of Series 51 steam generators and determined limited potential susceptibility to most types ofsteam generator degradation identified in the GL. For plants _with Series 51 steam generators, this conclusion is documented in report WCAP-15002, Rev.1, " Evaluation of EDF Steam Generator Internals Degradation -Impact of Causal Factors on Westinghouse 51 Series Steam Generators",
December 1997. The 51 Series SGs include Westinghouse model designations 51, SIM, SIF, and 54F The 51 Series Designs are the most similar to the EdF units.
WCAP-15002, Rev.1, documents visual inspections conducted at domestic plants of Westinghouse design. Based on the number ofplants that have been inspected and the inspection results, it is concluded that the causal factors identified by EdF do not jeopardize the continued operability ofWestinghouse Series 51 steam generators. Eddy current inspection of the tubes would detect any detrimental effects on the tubing due to wear caused by TSP ligament degradation, loose parts, and secondary side flow distribution changes. In addition, Foreign Object Search and Retrieval (FOSAR) efforts are normally conducted to discover loose parts and thus mitigate their potential for initiating tube degradation.
The information requested from the licensees by the Generic Letter includes:
(1) A discussion of any program in place to detect degradation of steam generator internals and descriptive inspection plans, including the inspection scope, frequency, methods and equipment. The GL requires discussions to include the following information for each facility:
(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.
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l Attachment l
Response to Generic Letter 97-06 Page 3.
(b) Whether visual or video camera inspections on the secondary side of the steam generators have been performM at the facility to gain information on the condition of the steam generator internals (e.g., support plates, tube bundle wrappers, or i
other components),
(c) Whether degradation of steam generator internals has been detected at the facility, and how the degradation was assessed and dispositioned.
(2) If the addressee currently has no program in place to detect degradation of steam generator l
internals, discussion and justification of the plans and schedule for establishing such a -
program, or why no program is needed.
A response to item 1 of the GL has been completed for Beaver Valley Units 1 and 2 which is t provided in Section II of this attachment. A response to Item 2 of the GL is included in Section III.Section IV provides a Safety Assessment of the French experience as it may pertain to domestic Westinghouse designed and fabricated steam generators.
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Attachment Response to Generic Letter 97-06 Page 4 M.
Restwnse to GL Item i for Series 51 Steam Generators Item 1 (1) A discussion ofanyprogram inplace to detect degradation ofsteam generator internals and descriptive inspection plans, including the inspection scope, frequency, methods and equipment.
Background
As discussed in WCAP-15002, Rev.1, surveys were sent to all Westinghouse Owners Group (WOG) utilities requesting the results of all steam generator, secondary side inspections and relevant tube inspections for tube support plate conditions. Completed surveys were received for 37 of 49 plants. For the 51 Series steam generators, responses were received for 18 plants. Of these,16 responded as having inspected or reviewed inspection data for tube support plate (TSP) ligament indications and 11 having performed steam generator (SG) secondary side entries that give confidence of not having wrapper drop. TSP ligament indications are reponed for 468 ligaments in TSPs made of carbon steel with round tube holes and flow holes. The total number of tubes involved is on the order of 129,000 tubes with roughly 3.6 million ligaments.
The modes of degradation detected include many cases of flow-assisted corrosion, or erosion-corrosion, and of premature cracking that results from either surface fatigue or from corrosion cracking that is associated with surface conditions such as pitting or geometric concentrations.
This reported degradation is not associated with the tube support plates. For the most part, the surveys do not report detection of several modes of degradation experienced in the EdF units.
There is no evidence of post chemical cleaning inspections discovering any significant material losses. NOTE: Neither Beaver Valley unit has implemented chemical cleaning to date. There is no evidence of any wrapper having dropped. There is no evidence of TSP ligament cracking or thinning that is progressive and continuing. TSP ligament cracking or missing pieces ofligaments have been observed, but only in units with carbon steel support plates with drilled round tube holes and flow holes. These conditions are generally traceable to initial inspections and are not progressing based on sequentialinspection data. Many of the conditions are probably related to original TSP drilling alignment. There are cases ofindications in TSPs that have been linked to patch plate welds.
Plants with significant hour-glassing of the tube support plates as a result of the denting process have exhibited ligament cracking throughout the thickness of the support plate between the flow holes in the plate or the flow holes in the tube lane. If denting remained uncontrolled, as subsequent support plate corrosion occurs, the potcatial exists for fragments of the support plate material to become completely free of the main TSP stmeture. However, these plate segments generally remain locked in place because of the in-plane forces that give rise to denting, as well as
Attachment
. Response to Generic Letter 97-06 Page5 the deformation that contains the individual piece. Operating plants with active denting are under periodic monitoring by the utility and have long-standing criteria and review by the NRC In addition, the EdF experiences reported are not related to support plate degradation that has progressed to the tube denting stage. Note: Neither Beaver Valley unit has experienced any significant denting at tube suppon plates.
. The secondary side internal degradation types found in Westinghouse, 51 Series steam generators
- are identified in Table 1 J. Beaver Valley Unit I has Model 51 steam generators with carbon steel drilled TSPs and Unit 2 has Model SIM steam generators with carbon steel drilled support plates.
Attachment Response to Generic Letter 97-06 Page 6 Table 1.0
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Susceptibility to Secondary Side Internals Degradation I
in Westinghouse 51 Series SG Designs Reference WCAP-15002 for Details i
Dearadation Tvoe Level of Susceptibility j
Erosion Corrosion:
Moisture Separator X
TSP Quatrefoil Ligaments N
,o TSP Flow Hole / Ligaments L
Feed Ring /J-Tubes X
Cracking:
TSP Ligaments Near Wedges @
N TSP Ligaments Near Patch Plates XN Carbon Steel TSP Ligaments (Random Areas)
XN Wrapper Near Supports @
N Transition Cone Girth Weld X
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Other:
A Wrapper Drop N
X = Observed in some SGs N = Not Susceptible to EDF Causal Factors L = Low Susceptibility to EDF Causal Factors (1)
Various indications of degradation may t,a artifacts of manufacturing related to patch plate plug l
welds and/or drilling alignment.
l (2)
Various Westinghouse design features are beneficial relative to some of the steam generator l
design features of foreign manufacturers.
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Attachment Response to Generic Letter 97-06 Page 7 Inservice Inspection Results Beaver Valley Unit 1 Tube Support Plate Ligannents - Screening of 100% of the eddy current bobbin coil data for tube support plate ligament degradation was implemented during the last refueling.
outage (Unit I twelftinefueling outage - October 1997). The technique utilized is defined in EPRI Report SG-96-05-003, Investigation ofApplicability ofEddy Current to the Detection ofPotentially DegradedSupport Structures, dated May 1996. Eddy current data for approximately 121,000 tube support plate intersections were screened. Of this total, twenty-one (21) locations exhibited potential suppon plate ligament degradation.
Ten (10) of the twenty-one (21) locatig)s were associated with the p (12) indications were at the seventh (7 hot l hot leg support and three (3) were at the first (gsupport, six (6) were at t 1 ) hot leg support.
All potentially degraded locations were further interrogated with + Point Rotating Pancake Coil (RPC) eddy current probes. Additionally, historical bobbin data was reviewed for the twenty-one (21) locations. This review demonstrated that the potential ligament degradation identified during the Unit I twelfth refueling outage examination remained unchanged from prior inspections and thus, did not represent an active inservice degradation mechanism. Based on a conservative evaluation of the eddy current data, the largest possible ligament gap at any location was 115 circumferential extent. This condition was determined to be acceptable since analyses demonstrate the minimum gap
-l necessary to permit the tube to move into a larger displacement mode is 146 when assuming lower tolerance tube size. Only one location exhibited any form of degradation from the bobbin and + Point RPC eddy current data. This location had a Distorted Signal 81 Indication (DSI) at the first (1 ) hot leg support plate intersection which was indicative of TSP outside diameter stress corrosion cracking (ODSCC) and not associated with the ligament condition. However, this tube was preventatively plugged as a conservative measure even though the signal amplitude did not exceed the licensed GL 95-05 alternate repair criteria for ODSCC.
The potential tube support plate ligament degradation identified during the Unit I twelfth refueling outage is postulated to be associated with either the patch plate rejoining process or potentially misdrilling of flow holes.
Wrapper Support Blocks - A remote video examination of four (4) wrapper support blocks in each of two (2) steam generators was conducted during the Unit I twelfth refueling outage examination. No evidence ofwrapper cracking or support block wear was observed with any of the eight (8) blocks examined. No evidence of wrapper drop was observed.
Attachment Response to Generic Letter 97-06 Page 8 s
Wrapper Drop - Foreign Object Search And Retrieval'(FOSAR) operations are performed every outage and sludge lancing operations have been performed every outage since the Unit I eighth refueling outage. No obstructions with the wrapper have been
~ identified during the conduct of these activities.
J-tube and Feedring - Ultrasonic thickness measurements have been routinely performed cn the carbon steel J-tubes and feedring since the ' Unit I third refueling outage (1983).
These examinations are performed every refueling outage. J-tube degradation has occurred due to Flow Assisted Corrosion (FAC). Select J-tubes have been replaced with ones fabricated from Inconel 600 or repaired by weld build up. All remaining carbon steel J-tubes are scheduled for replacement during the next refueling outage (Unit I refueling
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outage 13 in 1999). Replacement J-tubes will be fabricated from Inconel 600. Ultrasonic Test (UT) measurements conducted on the feedring have not revealed any significant degradation. Remote visual examination of the feedring inside diameter to interrogate the condition of the weld joint backing rings was conducted during the Unit I twelfth refueling outage. Some minor FAC was noted on the backing ring leading edges but all rings accessible for examination were intact except for one that was partially removed during original fabrication.
Feedring Thermal Sleeves - The feedwater nozzle elbows were replaced during the' Unit I ninth refueling outage (1993) with elbows that incorporated thermal sleeves with high molybdenum content to mitigate the cracking of said elbows due to thermal stratification. At that time, the feedring thermal sleeves were visually cumined and did not exhibit significant erosion / corrosion or FAC degradation.
Primary Separators, Swirl Vanes and Dryers - Visual examination and remote visual examinations were conducted during the Unit I ninth refueling outage. No significant
- FAC or erosion / corrosion degradation was observed. No weld degradation was observed.
Steam Generator Shell Welds - The shell welds are examined in accordance with ASME Section XI requirements and frequency. No relevant indications have been identified.
Beaver Valley Unit 2 Tube Support Plate Ligaments - Screening of 100% of the eddy current data is scheduled for the next refueling outage (Unit 2 refueling outage 7 in September of 1998).
Wrapper Support Blocks - The susceptibility assessment documented in WCAP 15002 l
has concluded that the domestic Model 51 and SIM steam generators are not susceptible to wrapper drop or wrapper cracking at the lower supposts due to differences in key design features between the domestic Westinghouse designed and fabricated steam generators and those fabricated for EdF. The inspection results from Beaver Valley Unit 1
AttachmentJ Response to Generic Letter 97-06 Page 9 -
further demonstrate non-susceptibility to this' degradation mechanism. Therefore, examination of the wrapper support blocks at Unit 2 is not planned. ~However, if interference with the wrapper and sludge lancing and/or FOSAR equipment should occur or for any other reason misposition of the wrapper is suspected, examination of the wrapper support blocks will be conducted.
Wrapper Drop - Foreign Object Search And xetrieval (FOSAR) operations are performed every outage and sludge lancing operations have been performed every outage since the Unit 2 second refueling outage. No obstructions with the wrapper have been identified during the conduct of these activities. (See Inservice Inspection Plan -
Section III).
J-tube and Feedring - The carbon steel J-tubes were replaced with ones fabricated from Inconel 600 prior to commercial operation. No UT examinations are performed on the j
Inconel J-tubes. A remote visual examination of the internals of one feedring (SG) was conducted during the Unit 2 sixth refueling outage (September 1996) to interrogate the j
weld joint backing rings and the internal feedring/J-tube interface. Small portions of two backing rings were eroded. One of the two backing rings was repaired by complete removal by grinding. The other was left in the "as found" condition. FOSAR examinations of the secondary top of tubesheet did not identify any loose parts from the
. feedring backing rings. The feedring/J-tube interfaces did not exhibit erosion /wrrosion or FAC degradation. Similar examinations are scheduled for all SGs N the Unit 2 seventh refueling outage (September 1998).
Feedring Thermal Sleeves - Erosion of the feedring thermal sleeve can result in leakage at the feedring/feedwater nozzle interface. No leakage was identified while conducting the feedring examination during the Unit 2 sixth refueling outage. "Best Effort" attempts will be made to examine the feedring thermal sleeves in conjunction with the backingring exams scheduled for the Unit 2 seventh refueling outage.
Steam Generator Shell Welds - The shell welds are examined in accordance with ASME Section XI requirements and frequency. No relevant indications have been identified.
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Attachment
~ Response to Generic Letter 97-06 Page 10 UL Resoonse to GL hem 2 Item 2 (2) If the ad&essee currently has noprogram inplace to detect degradation ofsteam generator internals, include a discussion andjustification of the plans andschedulefor establishing such a program, or why noprogram is needed Inservice Inspection Plan Examinations have been performed at both Beaver Valley Units 1 and 2 to assess the condition of various steam generator secondary internal components. Requirements for examinations pertaining to secondary side internal components are delineated in the Beaver Valley Steam Generator Program and thus a " program" does currently exist. The results of the examinations performed to date were detailed above. However, it is recognized that the comprehensive attributes of the Inservice Inspection Plan detailed below need to be formalized in Beaver Valley's Steam Generator Program document. This formalization will occur concurrent with DLC's implementation of NEI Initiative 97-06, Steam Generator Program Guidelines, scheduled for January 1,1999. Except where noted, these inspections will be completed each refueling outage.
Inspection scope and frequency may be adjusted as necessary based on site specific experience and evaluation ofindustry results of these inspections.
- Tube Suoport Plate Erosion-Corrosion and Cracking:
- 1. As the steam generator tube support plates in Beaver Valley Units 1 and 2 are made of carbon steel, a base line will be established employing low frequency bobbin results from a past outage or current outage. The technique to be employed is defined in the EPRI Report on the
" Investigation of Applicability of Eddy Current to A Detection ofPotentially Degraded Support Structures" dated May 1996, SG-96-0WO3. (Qualification of this eddy current technique to Appendix H of the EPRI Steam Generator Examination Guidelines is not required since this technique is not examining tube integrity). Ifindications are found, they will be traced back in history to establish if this is an active degradation mechanism.
- 2. Inservice inspection will be conducted in accordance with the EPRI PWR Steam Generator Examination Guidelines.
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. The critical area for ligament erosion / corrosion is the entire bundle. An initial sample of 20% (minimum) of the tubes will be completed.
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i Attachment Response to Genericleter 97 Page 11 1
i Wraoner Drop:
- l. It will be verified that the sludge lance equipment can be inserted without interference. The frequency of sludge lancing at both Beaver Valley units is every refueling outage.-
- 2. Ifinterference with the sludge lance equipment is detected, the lower wrapper support blocks will be visually inspected.
Wraoper Crackinn:
No inspection is recommended unless evidence of wrapper misposition or tube damage that may be associated with wrapper drop or misposition in the periphery of the first tube support plate is detected. If degradation is detected, a visual inspection of the Nwer wrapper support blocks will be conducted.
Upper Package:
Examination of the Beaver Valley Unit I carbon steel J-tubes will be performed every outage until said J-tubes are replaced with ones fabricated from FAC resistant material. Examination of the Unit 2 feedring weld backing rings will be performed every outage until all backing rings have been inspected and repaired as necessary.
The upper intemals examination program will be formalized and included in the Beaver Valley Steam Generator Program document as part of the implementation ofNEI Initiative 97-06, Steam GeneratorProgram Guidelines.
Transition Cone Girth WsJst Inspect in accordance with Section XI Inservice Inspection requirements and frequency.
Feed Water Nozzle:
ASME Section XI Inservice Inspection requirements for the feed water nozzle will be continued.
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4 Attachment Response to Generic Letter 97-06 Page 12 IV.
Safety Assessment The following safety concerns have been postulated relative to the French steam generator internals, degradation experience. These are:
Loss of support leading to wear and possible primary-to-secondary leakage or inadequate burst margins.
More significant tube support plate deformation during a postulated LOCA +SSE event resulting in unacceptable steam generator tube collapse or secondary-to-primary in-leakage.
The generation of a loose object in the secondary side of a steam generator which may result in tube wear or itapacting and possibly primary-to-secondary leakage.
Based on a review of Table 1.0, the only degradation types that may occur domestically that may result in the loss of tube support plate integrity are: TSP flow hole / ligaments erosion-corrosion, TSP ligament cracking near the patch plates, and TSP ligament cracking in random areas. There are no observations of post chemical cleaning inspections discovering any significant material losses in any domestic unit. Neither Beaver Valley unit has pedormed chemical cleaning to date.
There are no observations of any wrapper having dropped. There are no observations of TSP ligament cracking or thinning that is progressive and continuing. TSP ligament cracking or missing pieces ofligaments have been observed, but only in units with carbon steel TSPs with drilled round holes and flow holes. All utilities with 51 Series steam generators with carbon steel support plates inspect a significant percentage of steam generator tubes every outage with a bobbin probe, eddy current examination. If sections of the tube support plate are missing, this would be readily detectable due to a lack of eddy current response at the tube npport plate elevation and actions can be taken to address the absence of the support. Future application of.
the voltage-based plugging criteria will also consider the presence of any missing ligaments. The alternate plugging criteria would not be applied at these locations.
There is no increased susceptibility to ligament cracking near the wedge supports in the 51 Series steam generator designs as either there are no flow holes extending to the periphery at the wedge locations or the wedges are not welded to the TSPs, as is the case with the EDF SIM steam generator. At Beaver Valley Unit I with Model 51 steam generators the flow holes do not extend to the periphery and at Unit 2 with Model SIM steam generators the wedges are not welded to the TSPs and there are three (3) wrapper anti-rotation keys which further reduce potential fabrication related stresses in the wedge areas. Existing calculations evaluating the effects of LOCA + SSE loadings on the tube bundle continue to apply in determining whether certain tubes should be excluded from application of the voltage-based plugging criteria or whether certain tubes should be removed from service in plants which do not currently apply such a criteria but which may have steam generator tubes experiencing cracking at the tube support plate intersections.
p Attachment j
Response to Generic Letter 97-06 l
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'Another occurrence resulting from steam generator internals degradation that may affect a steam z generator from performing its intended safety function is the potential for tube wear and primary-to-secondary leakage due to the generation of a loose object on the secondary side of the steam generator. This may occur due to erosion-corrosion of the moisture separators, feed ring /J-tube, or tube support plate flow holes, or the occurrence of tube support plate ligament cracking. If primary-to-secondary leakage should occur due to tube wear from a loose object, the expected consequences would be bounded by.a single tube rupture event and, therefore, would remain within the current licensing bases of a plant. Regardless, it is Duquesne Light's position that loose objects should be removed from the steam generator, whenever possible. In addition, since it is Beaver Valley practice to perform 100% bobbin coil ET examination at every refueling outage, tubes exhibiting damage due to possible loose pa.ts will be identified, further interrogated and evaluated and repaired if found to be defective. Eddy current inspection and foreign object search and retrieval (FOSAR) activities which are conducted during each refueling outage ensure the maintenance of tube integrity during plant operation.
For the types of steam generator internals degradation observed at Beaver Valley Units 1 and 2,
. it is expected that steam 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, external events, and natural phenomena permitting the affected steam generator to perform its intended safety function.
Referente 1.
WCAP-15002, Rev.1, " Evaluation of EDF Steam Generator Internals Degradation -
Impact of Causal Factors on Westinghouse Series 51 Steam Ger erators"