ML20217E365
| ML20217E365 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/25/1998 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2524, GL-97-06, GL-97-6, NUDOCS 9803310019 | |
| Download: ML20217E365 (22) | |
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Dass-Besse Nuclear Pdwer Station gg?V 5501 North State Route 2 Oak Harbor.Onio 43449-9760 m.
Jchn K. Wood.
419-249-2300 '
VrePresident Nuctoar Fax: 419-321-8337 Docket Number 50-346 License Number NPF-3 Serial Number 2524 March 25, 1998 United States Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555-0001
Subject:
Response to NRC Generic Letter 97-06, Degradation of Steam Generator Internals Ladies and Gentlemen:
. On December 30,1997, the Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 97-% (Toledo Edison Letter Log Number 5189). That letter requested licensees, such as those for the Davis-Besse Nuclear Power Station (DBNPS), to respond within 90 days and 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.
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. If the 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 intemals (e.g., support plates, tube bundle wrappers, or other components). If the addressee has performed such inspections, include a discussion of the findings.
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- Docket Number 50-346 License Number NPF-3
, Serial Number 2524 Page 2
- (c).Whether degradation of steam generator internals has been detected at the facility, and how the degradation was assessed and dispositioned.
(2)" If the licensee 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.
The GL also stated that licensees are encouraged to work closely with industry groups on the coordination _ ofinspections, evaluations, and repair options for all types of steam generator degradation that may be found.
Joledo Edison (TE) has worked closely with the Nuclear Energy Institute (NEI) Steam Generator
/ Internals Task Force since it was formed in January,1997 in response to a proposed Generic
" Letter (GL) on degradation of steam generator internals. The purpose of the task force was to develop a coordinated industry wide response to the secondary side degradation issues identified in the proposed GL. Participants 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 that is discussed below.
. 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 report, GC-109558, Steam Generator Internals 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 own steam generator designs and operating history.
The NEI transmitted this report to the NRC via an NEl letter, dated December 19,1997.
In addition to the review of the EdF degradation causal factors, the 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 3
assessed collective visual video and pertinent non-destructive examination (NDE) inspection experience information to further enhance their evaluations regarding the susceptibility to internal degradation.
- The NEI task force met with the NRC in May 1997, to gain n better understanding of the safety.
concerns discussed in the generic letter. As a result of these efforts, the owners groups developed preliminary safety and susceptibility a,sessments relative to the design and operating history of -
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- Docket Number 50-346 -
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. their fleet. These assessments provide reasonable assurance that degradation ofinternals has not -
m compromised steam generator tube integrity, or decay heat removal capability.
The industry, through 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. Site specific details regarding steam generator internals -
- degradation for the DBNPS are provided below.
To assist plants in responding to the subject generic letter, steam generator vendors, working.
with the NEI Steam Generator Internals Task Force, initiated programs to determine the
. susceptibility of various designs to secondary-side degradation and its potential impact on tube integrity.' As a result of these programs, owners groups developed vendor-specific responses to the generic letter. The Babcock & Wilcox (B&W) Owners Group (B&WOG) response is compiled from plant specific information from the various B&W plants that make up the Owners
- Group. Toledo Edison endorses the B&WOG response to GL 97-06 and is forwarding it to the
. NRC as an attachment to this letter.
In addition to the B&WOG response, TE provides the following plant specific information:
In accordance with the Operating License Technical Specifications, eddy current examinations are performed each refueling outage to evaluate the status of tubes in the periphery of the Steam Generators which would be susceptible to damage from movement of the abandoned Auxiliary Feedwater (AFW) supply header. No evidence of movement or new indications of tube degradation as a result of movement of this header has occurred since the internal header was stabilized in the early 1980s. This internal header is also visually inspected per Technical Specifications every ten years. This visual inspection last occurred in the Sixth Refueling L Outage, January 1990. There was no evidence of movement or degradation of the Auxiliary Feedwater header at that time. The AFW thermal sleeves were also inspected with the internal header and also displayed no degradation.
During the Eighth Refueling Outage, March 1993, the main feedwater nozzles were inspected and replaced to improve the efficiency of the steam generators. During this maintenance, the original main feedwater nozzle thermal sleeves were inspected for degradation. No notable degradation was observed in these components.
Portions of the secondary side of the Steam Generators were visually inspected during the Tenth
' Refueling Outage in April 1996, (approximately 3856 EFPD). Inspections between the lower tube sheet and the 1st tube support plate, between the 4th and 5th tube support plates,'and
' between the 6th and 7th tube support plates were conducted. With the exception of some L deposits, no degradation was observed.
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Docket Number 50-346~
- License Number NPF-3 Serial Number 2524 Page 4.
Should you have any questions or require additional information, please contact Mr.' James L.
Freels, Manager-Regulatory Affairs, at (419) 321-8466.
Very truly yours, u
- FW aj
~' Attachment cc:
A. B. Beach, Regional Administrator, NRC Region III S. J. Campbell, NRC Region III, DB-1 Senior Resident Inspector A. G. Hansen, DB-1 NRC/NRR Project Manager Utility Radiological Safety Board 4
i -
' Docket' Number 50-346.
License Number NPF-3.
~ Serial Number 2524 Page 5
RESPONSE
TO NRC GENERIC LETTER 97-06 FOR THE DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 This letter is submitted pursuant to 10 CFR 50.54(f) and contains information pursuant to NRC Generic Letter 97-06, " Degradation of Steam Generator Internals" for the Davis-Besse Nuclear Power Station, Unit Number 1.
By:
John K[ood, Vice President - Nuclear Sworn to and subscribed before me this 25th day of March,1998.
4 s >&
1 Notary Public, State ofMhio Nora Lynn Flood My Commission expires September 4, 2002, i
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B&WOG Response to GL 97-06 INTRODUCTION:
Generic Letter (GL) 97-06, Degradation of Steam Generator Internals was issued 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 findings 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 of Appendix B to 10 CFR Part 50; and (4) require all addressees to submit information that will enable the NRC staff to verify whether addressecs' steam generator intemals comply with and conform to the current licensing bases for their respective facilities.
This response provides the B&W Owners Group (B&WOG) member utilities' information relative to the information requested by the Generic Letter. The B&WOG includes the following plants:
Arkansas Nuclear One Nuclear Power Plant Unit 1 (Arkansas 1, ANO-1)
Crystal River Nuclear Power Plant Unit 3 (Crystal River 3, CR-3) e Oconee Nuclear Power Plant Unit 1,2,3 (Oconee 1,2,3 ; ONS-1,2,3) e Three Mile Island Nuclear Power Plant Unit 1 (Three Mile Island 1, TMI-1)
Davis Besse Nuclear Power Plant Unit 1 (Davis Besse 1, DB-1) e RESPONSE TO REOUESTED INFORMATION:
The information requested by the GL is in italics; the response in normal font.
(1)
Discussion ofanyprogram in place to detect degradation ofsteam generator internals and a description ofinspection plans, including the inspection scope, frequency, methods, and equipment.
The B&WOG plants as a whole currently have no formal program to inspect / monitor steam generator internals degradation. However, as a result of other SG activities, such as sludge lancing and chemical cleaning, a significant number of secondary side inspections have been conducted at each of the B&WOG member plants. Visual inspections have been performed from the lower tubesheet to the upper tubesheet and all tube support plates over a range of EFPY from pre-service to 17.1 EFPY (the oldest B&W-designed OTSG)
(a)
Whether inspection records at thefacility have been reviewedfor indications oftube supportplate signal anomaliesfrom eddy-current testing ofthe steam generator tubes that may be indicative ofsupportplate damage or ligament cracking. Ifthe addressee has performed such a review, include a discussion ofIhefindings:
Page 1 of17
L j-There is currently no qualified eddy current technique to detect degradation in.
B&W broached hole TSPs. Currently available eddy current techniques are, however, considered adequate to detect the presence of tube support plates.-
The many recent 100% bobbin examinations at the BWOG plants have resulted in no instances of missing TSP indications.: Therefore, it is concluded that the TSPs are located properly and show no signs of gross degradation. Eddy -
- current techniques are now being developed for the detection and characterization of TSP degradation.
'(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 ofsteam generator internals (e.g., supportplates, tube bundle wrappers, or other components). Ifthe addressee hasperformedsuch inspections, include a discussion ofthefindings:
Table 1 contains a summary of the secondary side internal inspections performed at each of the B&WOG member plants. This table presents the steam generator, EFPY, date, location, purpose, and results for each internals inspection performed. This table shows that a significant number of secondary side inspections have been conducted at each of the B&WOG member plants.
These inspections span from pre-service to 17 EFPY, and include all 15 tube support plates, the upper and lower tubesheets, and one recent inspection of the upper wrapper welds. These inspections are typically performed in conjunction with cleaning processes or tube repair operations.
As part of the secondary side cleaning processes, pre-cleaning and post-cleaning visual inspections were performed on the secondary side of the OTSGs. These cleaning processes include sludge lancing, chemical cleaning and water slap. The post-cleaning visuals provided the clearest view of the tube support plates inspected, typically the 3'dthrough 6*,9*, and 10* support plates. These inspections, which were performed to ensure that the process was effective in removing the deposits, showed no signs of any structural damage to the tube support plates.
In recent years, fiberscopic inspections of the secondary side have been conducted following tube pull operations. These inspections utilize the open path left by the pulled tube to visually inspect the condition of the support plates.' Because the tubes are typically pulled from the lower tubesheet, most of the inspections encompassed the lower tubesheet through either the seventh, 1
eighth, or ninth tube support plates. In one case the tube hole was inspected from the upper tube end to the 11* tube support plate as well. None of these inspections have found any tube support plate structural damage.
Page 2 of 17
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in 1997, the oldest OTSG (Oconee 1) was visually inspected to determine the condition of the upper wrapper welds. For this inspection, two of the main feedwater nozzles were removed to examine the upper wrapper welds from below. No damage was found as a result of this inspection.
In summary, visual inspections have been performed on the internals of all the B&WOG member utilities' OTSGs. These inspections include numerous inspections of the tube support plates and one recent inspection of the upper wrapper welds. These inspections have found no structural damage to these internal components.
(c)
Whether degradation ofsteam generator internals has been detected at the facility, and how the degradation was assessed and dispositioned.
With the exception of the internal AFW header, no degradation of tube support plates, internal support structures, or other internal components that may affect tube integrity has been detected at any B&W-designed plant.
There are two configurations of auxiliary feedwater header assemblies which are used on the steam generators of B&W-designed 177FA plants. The first type uses an external distribution header mounted outside the OTSG with nozzles penetrating the shell and shroud. The second type uses an internal distribution header mounted inside the OTSG. Only two of the operating B&WOG member plants used internal AFW headers, Oconee 3 and Davis i
Besse 1.
In 1981 and 1982, tube leaks were experienced by Davis Besse 1 and Oconee i
3, respectively. As a result of eddy currer.t and visual examinations, it was I
determined that the internal headers and the brackets which attached them to the wrapper were damaged. This degradation resulted in inovement of the internal header during plant operation, damaging some peripheral tubes. The AFW internal headers were subsequently stabilized and functionally replaced by extemal headers at these plants. No movement or new indications of tube degradation have been noted at either plant since the internal AFW supply headers were stabilized.
Eddy current examinations of peripheral tubes of the DB-1 steam generators are performed each outage to ensure tube integrity. The internal header is I
visually inspected per Technical Specifications every ten years at Davis Besse i
Unit 1.
Oconee 3 inspected the internal AFW headers as part of a commitment to the NRC in 1982. In addition, visual examinations were performed the two subsequent refueling outages and then during the 2"d 10-year ISI outage.
Oconee 3 has conducted 100% eddy current bobbin coil inspections since Page 3 of 17
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a 1995, and has not detected any new degradation of the tubes in the periphery of the steam generators in the areas that are susceptible to damage from movement of the AFW supply header. Another visual inspection is planned for d
the 3 10-year ISI.
(2)
Ifthe addressee currently has no program in place to detect degradation ofsteam generator internals, include a discussion andjustification ofthe plans and schedule for establishing such aprogram, or why no program is needed.
Prior to issuance of GL 97-06, U.S. nuclear utilities, the Electric Power Research Institute and the Nuclear Energy Institute (NEI) deseloped an action plan to assess the susceptibility of SG internals to secondary side degradation. Based upon this action plan, the B&WOG member utilities have developed a process by which a formal program to detect the degradation of steam generator internals will be developed. This process was started in 1997 and is scheduled for completion by December 1998.
The major tasks that comprise the B&WOG process are discussed below. Preliminary results of the work completed are also presented. All available data from NDE inspections, tube pull evaluations, and secondary-side visual inspections supports compliance with and conformance to the current licensing basis for the B&WOG member utilities.
1.0 Owners Group Deeradation Experience The purpose of this task is to identify any internals degradation detected at operating OTSG plants. This task focuses on internal components that may have an effect on tube integrity, and includes the review of relevant visual and ECT data.
Preliminary review of secondary side visual inspection data and available ECT data has shown no generic internals degradation in the B&W plants (Table 1).
In fact, the only internals degradation found is related to the internal auxiliary feedwater (AFW) headers. As noted in the response to question 1, part c, only Davis Besse Unit I and Oconec Unit 3 had OTSGs of this design, and these internal AFW headers were stabilized and functionally replaced with external feedwater headers in the early 1980's. No further tube damage associated with the internal AFW header has been detected during subsequent tube inspections at these two plants.
Page 4 of 17
l 2.0 Owners Group Degradation Assessment This part of the process is broken down into subtasks and described below:
Evaluate Susceptibility of OTSG Relative to EdF Experience The B&WOG member utilities are in the process of assembling and summarizing design documentation relative to steam generator secondary side components for all operating B&W plants. Existing analyses that relate to possible degradation mechanisms are also being collected and summarized.
This information is being used to determine the susceptibility of the OTSG to secondary side degradation relative to the experience of EdF plants.
EPRI released GC-109558 " Steam Generator Internals Degradation: Modes of Degradation Detected in EdF Units"in December 1997. According to this document, several types of degradation were found in the EdF SG internals.
These degradation modes, and the preliminary results of evaluations to determine possible applicability to the OTSG are presented below.
- 1. Flow assisted corrosion (FAC) of the top tube support plate (TSP) at Fessenheim 2 was noted during routine eddy current testing (ECT) in 1995.
This corrosion was determined to be the result ofimproper placement of hoses during a chemical cleaning performed in 1992. The corrosion is not progressing, and inspections indicate that no other French units have experienced similar attack.
Four of the seven B&WOG plants have had their steam generators chemically cleaned. The EPRI SGOG chemical cleaning solvent used in these cleanings is a different solvent than the one utilized by EdF. The EPRI SGOG solvent has been put through extensive qualification testing prior to its first application and continues to be tested prior to most applications. FAC has been studied as part of the qualification testing and the flow rates during steam generator cleanings are designed to remain below the critical FAC rates. Post ::hemical cleaning visur.1 inspections have shown no TSP degradation at any of the four plants.
- 2. In response to the Fessenheim 2 experience, all other operating units in France with drilled hole carbon steel TSPs, i.e., all units with Model 51 A and 51M steam generators, were inspected for TSP damage. Both ECT and i
television (TV) visual inspections were performed. Review ofinspection data has shown that the ligament cracking was present at the first inspection for which interpretable data is available, i.e., at either pre-service or early in-service inspections. The cracking has not changed with time. It is believed that the cracking is the result of mechanical loads applied to the TSPs during manufacture, shipping or early operation.
Page 5 of 17
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There is currently no eddy current technique qualified to detect TSP degradation in B&W broached hole TSPs. Currently available non-qualified eddy current techniques indicate that the' tube support plates are
' located properly, and show no indications of degradation.' These and other eddy current techniques are now being evaluated for EPRI Appendix H
. qualification. Also, no TSP degradation has been observed in any of the numerous visual inspections performed over the service lives of the OTSGs (see Table 1).
- 3. FAC induced thinning of the top TSP has been detected by ECT and TV in three units (Gravelines 2,3, and 4) that operated from startup in the early 1980s until recently using ammonia water chemistry. The FAC occurred at the periphery of the TSP at the location of the largest radius U-bends. It -
occurred mostly on the hot leg side, but some was also observed on the cold leg side. The French conclude that the FAC is associated with ammonia water chemistry.
All of the OTSG plants have operated on All Volatile Treatment (AVT) feedwater chemistry. Early in plant life, the chemistry consisted of condensate polished feedwater treated with hydrazine and ammonia for oxygen and pH control, respectively. Later, the pH control additive was switched to alternate amines, such as morpholine. The objective was to reduce FAC in the balance of plant system piping where such problems had been observed. Based on NDE inspections, tube pull evaluations, and the numerous secondary side inspections listed in Table 1, FAC has not been identified as a damage mechanism in the OTSGs at this time. The primary contributors to FAC - flow velocity, quality, and solution pH - will be evaluated in more depth as part of the B&WOG process.
- 4. Wrapper drop was detected in 1994 at two steam generators in Blayais 3.
The drop occurred because of the failure of the wrapper supports located at the first TSP elevation. Two scenarios for the wrapper drop had been proposed by the French: (1) failure of the supports mainly due to thermal
]
expansion loads developed during transients associated with the use of cold auxiliary feedwater to expedite plant cool-down, aggravated by poor quality of the welds joining the support blocks to the wrapper and possibly by fatigue induced crack propagation, and (2) failure of the supports mainly -
due to fatigue cracking of the support to wrapper welds, aggravated by poor quality of the welds and by axial load development by thermal transients.
The B&W-designed OTSGs have a different wrapper design than the EdF plants with the noted degradation. The wrapper consists of two shells, one upper and one lower, separated by a small gap. The lower wrapper of the-Page 6 of17 mm m_
/
4 OTSGs is supported by the lower tubesheet. The upper wrapper is supported by an annular ring which is welded to the shell.
During 1997, the oldest operating B&W-designed plant (Oconee-1) conducted a visual inspection of the upper wrapper assembly, welds, and internal components in the vicinity of the welds. During this inspection, no signs of movement or slippage of the wrapper were noted. There was no evidence of damage to the welds or components in the vicinity of the wrapper assembly which' would indicate that the wrapper had shifted or dropped. Wrapper drop is therefore not considered 'a significant near-term problem for the OTSGs. However, during the continuing evaluation of SG internals, this potential degradation mechanism will be investigated further.
- 5. Fatigue cracks emanating from support blocks have been detected in the same two Blayais 3 steam generators that experienced wrapper drop, and in one steam generator of similar design at Blayais 2. The fatigue cracks appear to be the results of flow induced vibration of the wrapper. The French indicate that they are evaluating the possible occurrence of this mechanism at other units and designs, since the root cause of this cracking has not been demonstrated to be Blayais 2 and 3 specific.
As noted above, inspection of the upper wrapper supports at Oconee 1 did not identify signs of degradation. This potential damage mechanism continues to be evaluated for poss;ble future impact on the OTSGs.
- 6. Some cases of TSP wedge block cracking have been observed at EdF plants. The causes of the wedge block cracking are believed to be the same as those causing TSP ligament cracking. Some tie rod lock welds have been found to be cracked. These cracks are considered to have no safety significance.
From visual inspections, no degradation has been noted. However, during the evaluation of SG internals degradation, this degradation mechanism will be investigated further.
Assess Susceptibility to Other Potential Damage Mechanisms Preliminary information indicates that recent internals damage found at the San Onofre Nuclear Generating Station (SONGS) has been attributed to FAC associated with fouling.
The steam generators at SONGS are Combustion Engineering recirculating steam generators which have few design and performance similarities to an OTSG. The concentrating mechanisms, bulk water chemical concentrations, and pH control additives behavior are variables that are typically different in Page 7 of17
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7,_c the two types of steam generators. The B&WOG member utilities are aware of the FAC damage at SONGS, and will evaluate it further as part of the BWOG process.
The objective of this subtask is also to address damage mechanisms which have not yet been observed in operating plants, but which may be possible -
based on operating experience not considered during original plant design.
Structural, chemical, and thermal hydraulic performance experience will be reviewed to identify these potential forms of degradation. Any mechanisms which are identified as requiring further investigation will be identified and additional work will be conducted as appropriate.
3.0 Inspection Reauirements and Methodolony Based on the results of task 2, internal components identified as being susceptible to internals degradation which could affect tube integrity will be identified. For each of these components, a recommended inspection scope and methodology to monitor for potential future secondary side degradation will be determined.
4.0 Industry Response The final deliverable for this project will be a document that will serve as the i
overall project summary report for the B&WOG member utilities. All areas of the OTSG identified as being potentially susceptible to degradation (if any) will be discussed in the report, along with recommended inspection procedures and frequencies, and disposition criteria.
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SUMMARY
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l The B&W Owners Group has completed a preliminarily review of EPRI GC-109558, " Steam Generator Internals Degradation: Modes of Degradation Detected in EdF Units" relative to the design and operation of Once Through Steam Generators. For each category of degradation, the B&WOG has concluded that the OTSGs are not significantly at risk for the same degradation in the near term. The future susceptibility of the OTSG to these or other forms of degradation continues to be evaluated as part of the B&WOG process to develop a forrnal SG internals program.
3 Table 1 provides a summary of secondary side visual inspections conducted at each B&W-designed plant. The B&WOG concludes diat the number of plants that have been inspected and the visual inspection results demonstrate with a high degree of confidence that there is currently no significant degradation of SG Internals in OTSGs. Currently two of the B&WOG plants monitor periphery tubes for damage from internal AFW headers. The other B&WOG plants do not have this design feature and thus are not susceptible to this type of SG internals degradation.
Based on results to date, it is concluded that no near term inspections of the internals are required, and that compliance with and conformance to the current licensing bases for the B&WOG member utilities has been maintained.
When the program is complete, recommendations may be made for future periodic inspections if needed.
These recommendations will be included in a report which will be distributed to all the B&WOG member utilities upon completion of this process.
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