ML20217G117
| ML20217G117 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/30/1998 |
| From: | Ewing E ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-06, GL-97-6, W3F1-98-0057, W3F1-98-57, NUDOCS 9804020202 | |
| Download: ML20217G117 (15) | |
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Kmona, LA 70066 Tel 504 739 6242 C. Ewing, lli a Safety & Regda'ory Affairs W3F1-98-0057 A4.05 PR
! March 30,1998
. U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 90-Day Response to NRC Generic Letter 97-06, " Degradation of Steam Generator Internals" Gentlemen:
On December 30,1997, the NRC issued Generic Letter (GL) 97-06.- The generic letter reports foreign and U.S. experience associated with potential degradation mechanisms which may lead to tube support plate and tube bundle wrapper damage. It requests information related to licensee programs for detection of steam generator internals degradation. This information will enable the NRC staff to verify whether steam generator internals comply with and conform to the current licensing bases for the respective facility. By this letter, EOl is providing the requested 90-day response for the Waterford 3 Steam Electric Station.
Specifically, the GL requested a written response to the following questions:
(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.
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90-Day Response to NRC Generic Letter 97-06, " Degradation of Steam' Generator Internals" W3F1-98-0057 Page 2 March 30,1998 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 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.
(2)
If the 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.
Waterford 3 has a program in place for detection of steam generator internals degradation; therefore, in accordance with the Generic Letter request, a written report describing the program is included in the attached response. Accordingly, the report does not include a discussion and justification of the plans and schedule for establishing such a program, or why no program is needed. The information provided justifies that Waterford 3 steam generator internals comply with and conform to the currcnt licensing basis for the facility.
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90-Dhy Response to NRC Generic Letter 97-06, " Degradation of
. Steam' Generator Internals" W3F1-98-0057 Page 3 March 30,1998 This information is being submitted under oath and affirmation in accordance with.
10CFR50.54(f). Should you have any questions regarding this matter, please contact Mr.T.J. Gaudet at (504) 739-6666 or me at (504) 739-6242.
Very truly yours,
)
c_. 'N E.C. Ewing Director -
Nuclear Safety & Regulatory Affairs ECE/PRShtk
Enclosures:
Affidavit Response to Generic Letter 97-06 CEOG Member Plants SG Internals inspection Data cc:
E.W. Merschoff, NRC Region IV C.P. Patel, NRC-NRR J. Smith N.S. Reynolds NRC Resident inspectors Office m..._..
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l UNITED STATES OF AMERICA l
- NUCLEAR REGULATORY COMMISSION in the matter of
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Entergy Operations, incorporated
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Docket No. 50-382 Waterford 3 Steam Electric Station
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AFFIDAVIT Early Cunningham Ewing, being duly sworn, hereby deposes and says that he is Director, Nuclear Safety and Regulatory Affairs-Waterford 3 of Entergy Operations, Incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached 90 Day Response to NRC Generic Letter 97-06; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
w EarlyfunningEw]i)g Director, Nuclear Safety & Regulatory Affairs -
Waterford 3 1
STATE OF LOUISIANA
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PARISH OF ST, CHARLES
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Subscribed and sworn to before me, a Notary Public in,and for the Parish and State above named this l' o '3' day of /&c~ -
.1998.
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c L Notary Public
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My Commission expires d dM.
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' W3F1-98-0057 i
Page 1 of 10 Waterford 3 Response to Generic Letter 97-06
" Degradation of Steam Generator internals" q
Introduction Waterford 3's response provides the results of evaluations and inspections and identifies programs in place for monitoring potential steam generator secondary side internals and component degradation that may pose a risk to plant safety.
This response includes input from NEl, EPRI, and Combustion Engineering Owners Group (CEOG) specific to S/G internals degradation and evaluations performed to determine causal factors. Additional information is contained in the
" Background" section of this response. The CEOG information includes the following evaluations related to S/G internal degradation: Operability Assessment; Susceptibility of CE Designed S/Gs and Bounding Analysis.
The December 30,'1997 release of NRC Generic Letter 97-06, " Degradation of Steam Generator Internals," required a written response that includes the following:
(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.
(a)
Whether inspection records at the facility have been reviewed for indications of the 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 internals (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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Att:: chm nt 1 to W3F1-98-0057 Page 2 of 10 (c)
Whether degradation of steam generator intemals has been detected at the facility, and how the degradation was assessed and dispositioned.
(2)
If the addresses 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.
Response to item (1):
Discussion of any program in place to detect degradation of steam generator intemals and a description of the inspection plans, including the inspection scope, frequency, methods, and equipment.
The Waterford 3 S/G secondary side inspection program is implemented in accordance with Design Engineering procedure NOECP-257, " Steam Generator
- Secondary Side Inspections". The procedure provides instructions for inspecting the S/G internals for evidence of Flow Accelerated Corrosion (FAC) and loose parts. The inspection frequency as'a minimum is limited to one S/G during each 36 month period on a rotating basis. This is maintained on a Repetitive Task.
Waterford 3 has performed S/G secondary side inspections starting December 1986 and subsequently October 1988, March 1991, March 1994 and April 1997.
The following areas of the S/G secondary side internals are visually inspected where accessible or otherwise examined utilizing a video camera:
Below Steam Separator CAN Deck Tube Supports, I-Beams, Upper Eggerates and Batwings Anti-Vibration Bars (AVBs)
Feedring and Distribution Box Handholes, Tubesheet and Blowdown Lane and Piping Dryers, Dryer Supports and Drains
. Dryer Drain Channels and Steam Deflector Plate 1.; W
' Attachm:nt 1 to W3F1-98-0057 Page 3 of 10 Response to item 1 (a):
- Whetherinspection records at the facility have been reviewed forindications of
' the tube support plate signal anomalies from eddy current testing of the steam
. generator tubes that may be indicative of support plate damage orligament F cracking. If the addressee has performed such a review, include a discussion of the findings.
No, a review of the inspection records for this purpose was not performed.
Waterford 3 does not have drilled tube support plates and as such is not susceptible to failure mechanisms specifically associated with those supports
. due to the design differences. The CEOG evaluation and inspections of Waterford 3 provides assurance that the only credible steam generator internals components damage mechanism of poten_tial safety significance is flow accelerated corrosion of eggcrate type tube supports. A potential susceptibility to FAC has been identified for CE-designed plants with carbon steel eggerate type tube supports with heavily fouled tube bundles. As stated in Response (1),
. Waterford 3 has not identified conditions specific to FAC.
Response to item 1 (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 generatorintemals (e.g. support plates, tube bundle wrappers, or other components). If the addressee has performed such inspections, include a discussion of the findings.
During Refuel #6 (Spring 1994), Waterford 3 performed an extensive secondary side underwater in-bundle remote visual inspection of S/G #2 hot leg utilizing a remote video camera. The inspection examined the tenth partial eggcrate approximately 5 tubes in from the periphery. Inspection results identified minimal sludge deposition and no evidence of eggcrate deterioration.
Response to item 1 (c):
' Whether degradation of steam generatorintemals'has been detected at the
' facility, and how the' degradation was assessed and dispositioned.
S/G secondary side inspection results that have identified conditions adverse to Equality are documented and corrected via the Corrective Action process.
~
- Waterford 3 has addressed loose part issues related to feedring U-bolt nuts.
' information from eddy current data identifying ' potential loose parts in the blowdown lane have been confirmed as a result of visual inspections of the
- secondary tubesheetiThe U-bolt nuts have been located and retrieved from the t.
. to W3F1-98-0057 Page 4 of 10 blowdown lane. Corrective action to mitigate this condition' resulted in double-nutting and staking in locations of occurrence. During Refuel #8 (Spring 1997),
Waterford 3's corrective action required a feedring U-bolt nut be tack welded to mitigate the nut backing off the bolt thrads.
~ In addition, Waterford 3 collected sludge samples during the Refuel #8 (Spririg 1997) secondary side upper bundle flush and lancing activities which were analyzed to determine deposit loading and preliminary thermal performance.
The results of Waterford 3 S/G tube scale deposit analysis concurrent with design and operating data provides for a calculated fouling factor which has minimal effect on heat transfer. This information supports Waterford 3's S/G -
secondary internals good health based on steam pressure and minimal tube.
bundle fouling. Present secondary chemistry improvements and low feedwater iron transport numbers corroborate Waterford 3's lower susceptibility to eggcrate degradation.
Waterford 3 S/G secondary side inspections have not identified internal
' degradation caused by FAC.
Response to item (2):
If the addressee currently has no program in place to detect degradation of steam generatorintemals, include a discussion andjustification of the plans and schedule for establishing such a program, or wisy no program is needed.
Item (2) is not applicable to the Waterford 3 SES. As discussed in the previous c
sections, Waterford 3 has conducted routine inspections of S/G internals since -
1986 and plans to follow the industry guidan'ce to provide reasonable assurance that S/G tube integrity is maintained.
Backaround Information Addressina SIG Secondary Side Internals In_ response to the issuance of draft Generic Letter 97-06 on degradation of steam generator internals, NEl formed the Steam Generator Intemals Task Force in January 1997. The purpose of the task force was to develop a coordinated industry-wide response to the draft Generic Letter. Participation on the task force included EPRI, licensees, and representatives of the owners groups for each domestic steam generator design.
Nuclear' Steam Supply Systems Owners Groups initiated programs to assist members in assessing the susceptibility to tube damage or loss of decay heat removal (DHR) 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.
i W
f
4 to W3F1-98-0057 Page 5 of 10 EPRI responded to this need and with the cooperation of Electricite de France (EDF) developed the report, GC-109558, Steam Generator Internals Degradation: Modes of Degradation 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. This report was transmitted to the NRC via an NEl letter, dated December 19,1997.
In developing the susceptibility assessment, attributes considered were design
' factors; fabrication and manufacturing techniques; as well as plant operating history, including chemistry and related degradation, such as tube denting.
Additionally, the Owners Groups compiled and assessed information on their respective visual, video and pertinent NDE inspection experience to further enhance their evaluations regarding the susceptibility to internals degradation.
Furthermore, the NEl task force met with the NRC in May 1997, to gain a better understanding of the safety concerns discussed in the draft 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 units. These assessments provide reasonable assurance that steam generator tube integrity and DHR capability is not compromised by internals degradation.
It is the intention of the NEl SG Task Force to provide the respective Owners Groups reports via NEl for NRC Staff information.
Combustion Enaineerina Owners Grouo initiatives
- ~As a member of the CEOG Waterford 3 has participated in an evaluation of steam generator internals degradation experience in EDF and domestic CE-designed units. The CEOG through its Steam Generator Task Force (SGTF) has discussed steam generator inspections, evaluations, and issues programmatically in regularly scheduled meetings. The EDF steam generator internals degrs.dation experience has been discussed, and an evaluation has been conducted by ABB under CEOG SGTF funding after the NRC Information Notice 96-09 was issued. The finding of that evaluation was that the specific causal factors identified by EDF were not applicable to the CE design.S/Gs.
Subsequently in cooperation with the NEl SG intemals Task Force, the CEOG
' SGTF funded an evaluation of the broader question of all types of SG internals
. degradation.--The' evaluation conducted by the CEOG dispositions potential degradation mechanisms identified by review of experience on the basis of
- design,~ manufacturing, and operational practice. CE design plants can be
- divided into 3 different groups based on the tube supports design: carbon steel
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Attachm::nt 1 to W3F1-98-0057 Page 6 of 10
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eggerates with drilled plates at upper elevations; carbon steel eggcrates only; and stainless steel.eggerates only.
The CEOG program included the development of 3 reports which have been summarized in the text that follows. CEOG will provide the NRC, under NEl cover letter, additional details of the reports to support the NEl led industry agenda to address the susceptibility of CE designed units to identified S/G internal degradation mechanisms.- NEl intends on providing the completed reports to the NRC Staff by April 30,1998. CEOG member unit steam generator
. internals, which encompass the 3 different S/G tube supports designs identified
- above, have been inspected. A summary table of the inspections is provided in to this response.'
' The objectives and conclusions of the CEOG reports are summarized below.
1.
CEOG Report, CE NPSD-1092, " Evaluation of Degraded Secondary Internals - Operability Assessment" OBJECTIVES:
1.
Assess the applicability of EDF damage mechanisms to the CE design.
2.
Assess the applicability of tube support erosion-corrosion experience at Maine Yankee and SONGS 3 to other CE designed units.
3.
Assess the impact of applicable mechanisms on tube integrity and decay heat removal capability.
CONCLUSIONS:.
1.
The primary damage mechanism related to the wrapper support
)
failures in the French units are not directly applicable to the CE designed S/Gs.
2.
Support plate cracking is a residual effect of tube denting but is not detrimental to the safe operation of the S/G.
3.
Adequate margins against failure have been demonstrated for the FAC damage observed in the SONGS 3 eggerates.
4.
There are no reported tube wear indications directly related to tube support degradation.
5.
- Plants with degradation of tube supports, such as that observed at SONGS 3, can continue to operate safely because adequate margins against failure exist and possible tube damage can be detected in the normal eddy current testing examinations.
' Attachment 2, CEOG Member Plants S/G Intemals inspecten Data
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Attachm nt 1 to W3F1-98-0057 Page 7 of 10 4
9 6.
Eggcrate denting, such as detected in the lower eggerates of the Millstone 2 original steam generators, has been determined to be acceptable by safety analysis. Current chemistry practices can mitigate existing denting and preclude further degradation due to denting.
7.
Calvert Cliffs 2 which possibly has more support degradation than SONGS 3 can continue to operate safely, because adequate margins agsinst failure can be demonstrated and possible Flow induced Vibration tube wear is postulated to result in Leak Before Break.
8.
CE S/Gs designed with stainless steel tube supports are much less susceptible to erosion-corrosion than the S/Gs with carbon steel tube supports.
9.
. None of the degradation mechanisms reviewed pose a threat to the Reactor Coolant System pressure boundary integrity or the heat removal function of the S/G.
II.
CEOG Report, CE NPSD-1103, " Evaluation of Susceptibility of Internals Degradation in CE Designed Steam Generators" OBJECTIVES:
1.
Review the history of S/G Internals degradation in the CE designed units.
2.
Examine the susceptibility of CE designed units to internals degradation mechanisms that have occurred in CE designed and EDF S/Gs.
CONCLUSIONS:
1.
CE-designed S/Gs have not encountered a significant amount of internals degradation. Of the degradation that has occurred, appropriate mitigating action has been implemented to minimize the effect of this degradation.
2.
The most common form of S/G internals degradation has been from waterhammer events or erosion of components within the feedwater system. However, these degradation mechanisms are not considered to be safety significant.
3.
The S/G internals degradation mechanisms described in GL 97-06 are generally not applicable to CE-designed S/Gs. The only degradation mechanism applicable to CE units that could have safety significance is FAC of peripheral eggcrates.
4.
FAC of peripheral eggerates is primarily the result of secondary fluid flow redistribution caused by severe tube bundle fouling. Use of ammonia for pH controlin heavily fouled S/Gs increases the susceptibility to FAC.
Att: chm:nt 1 to W3F1-98-0057 -
Page 8 of 10 5.
S/Gs with stainless steel eggcrates (Palo Verde units and Palisades RSG) have a chromium content of at least 10.5% which increases the resistance to FAC by at least an order of magnitude and so are not considered susceptible to FAC of tube supports.
6.
Of the CE-designed S/Gs with carbon steel tube supports, only those units with severe tube bundle fouling as indicated by significant S/G secondary pressure loss may be susceptible to FAC of peripheral eggerates.
7.
No CEOG member plants have detected by NDE or_ visual inspections any FAC of drilled support plates to-date.
Ill.
CEOG Report, CE NPSD-1104," Evaluation of Degraded Secondary Internals - Bounding Analysis" OBJECTIVES:
1.
Assess the impact of each degradation issue applicable to CE designed plants.
2.
Determine the bounding cases for degradation issues affecting safety.
PRELIMINARY GENERAL CONCLUSIONS:
At the time that this response was developed the report CE NPSD-1104 was not finalized. Preliminary general conclusions are provided here but will not be final in accordance with ABB-CE QA requirements until April 1998.
1.
No CE designed plants are at risk for loss of tube integrity or decay heat removal function as a consequence of S/G secondary side internals degradation including FAC degradation for the limiting case.
7-Att: chm:nt 1 to W3F1-98-0057 l
Page 9 of 10 The Waterford 3 steam generators were designed and fabricated as ASME Code Section lil Class 1 vessels. The secondary structure supporting the 9,350 tubes in each S/G consists of the following items: stayed design tubesheet; shroud; 7 l
full horizontal carbon steel eggerates; 3 partial horizontal carbon steel eggerates; I
2 carbon steel batwings (Anti-Vibration Bars) and 7 vertical carbon steel eggcrate 1
supports.
FAC of eggcrates has been detected in one CE designed unit as reported in GL 97-06.. In general, FAC of tube supports is possible in units with carbon steel eggerates if tube bundle fouling causes redistribution of flow such that FAC threshold velocities are exceeded. The CEOG evaluations indicate that FAC without substantial fouling is unlikely. Experience has indicated that the onset of substantial fouling is evidenced by a reduction in the normal plant operating steam pressure. Experience has also shown that plants can experience steam pressure reduction and not experience the onset of FAC in the tube supports in the event of a substantial reduction in steam pressure, the CEOG supported by ABB-CE analysis recommends that an inspection for the onset of FAC of tube supports be conducted at the next scheduled outage. FAC occurs preferentially toward the periphery of the hot leg side of the tube bundle at the upper supports.
. If warranted by a detectable steam pressure loss, remote visual inspection of the uppermost eggerates will determine whether FAC has occurred. CEOG Report CE NPSD - 1103 provides recommendations for subsequent in-bundle visual inspections based on sludge loading conditions and thermal performance degradation.
. Conclusion Industry experience has been reviewed for S/G secondary internals degradation and it has been concluded that for Waterford 3 there is low susceptibility to the types of degradation discussed in the Generic Letter. Secondary side inspections and recent sludge analysis support calculated fouling factors and thermal performance.
Waterford 3's response provides the information requested by Generic Letter 97-06 and demonstrates that S/G secondary side internals comply with the h
licensing basis for Waterford 3. The assessments provided by CEOG as part of this evaluation provide reasonable assurance that S/G tube integrity has not been compromised by internal degradation.
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AttachmOnt 1 to
. W3F1-98-0057 Page 10 of 10
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References CE NPSD-1092, " Evaluation of Degraded Secondary Intemals - Operability
. Assessment" CE NPSD-1103, " Evaluation of Susceptibility of Internals Degradation in CE Designed Steam Generators" CE NPSD-1104," Evaluation of Degraded Secondary Internals Bounding Safety Analysis"-
NOECP-252, " Steam Generator Secondary Side Inspections," Rev. 2
' Dominion Engineering,' inc. Letter. L-4125-00-1, " Preliminary Steam Generator Thermal Performance Analysis for Waterford 3," Dated December 8,1997 Waterford 3 Steam Generator Secondary Side Visual Inspection, December 1986, CENC #1774 Waterford 3 Steam Generator Noise Investigation and Secondary Side Visual Inspection, October 1988, CENC #1851 Waterford 3 Steam Generator Secondary Side Visual Inspection, March 1991 Waterford 3 Steam Generator Secondary Side Visual Inspection, March 1994, CSE-94-174 Waterford 3 Steam Generator Secondary Side Visual Inspection, May 1997, CSE-97-171 s
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