NRC Generic Letter 1997-06

NRC Generic Letter 1997-006: Degradation of Steam Generator Internals

ML031110075
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Washington Public Power Supply System, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Clinch River, Crane
Issue date:	12/30/1997
From:	Roe J W Office of Nuclear Reactor Regulation
To:
References
GL-97-006, NUDOCS 9712180168
Download: ML031110075 (7)

UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555December 30, 1997NRC GENERIC LETTER 97-06: DEGRADATION OF STEAM GENERATOR INTERNALS

Addressees

All holders of operating licenses for pressurized-water reactors (PWRs), except those who havepermanently ceased operations and have certified that fuel has been permanently removedfrom the reactor vessel.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1) again alertaddressees to the previously communicated findings of damage to steam generator internals,namely, tube support plates and tube bundle wrappers, at foreign PWR facilities; (2) alertaddressees to recent findings of damage to steam generator tube support plates at a U.S. PWRfacility; (3) emphasize to addressees the importance of performing comprehensiveexaminations of steam generator internals to ensure steam generator tube structural integrity ismaintained 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 whetheraddressees' steam generator internals comply with and conform to the current licensing basesfor their respective facilities.BackgroundThe NRC issued Information Notice (IN) 96-09 and IN 96-09, Supplement 1 to alert addresseesto findings of damage to steam generator internals at foreign PWR facilities.

Description of Circumstances

Foreign authorities have reported various steam generator tube support plate damagemechanisms. The affected steam generators are similar, but not identical, to WestinghouseModel 51 steam generators. As previously documented in IN 96-09 and IN 96-09, Supple-ment 1, one damage mechanism involved the wastage of the uppermost support plate causedby the misapplication of a chemical cleaning process. A second damage mechanism involvedbroken tube support plate ligaments at the uppermost, and sometimes at the next lower, tubesupport plates. The support plate ligaments broke near a radial seismic restraint and near anantirotation key; the damage apparently dates back to initial startup of the affected plants.According to foreign authorities, the ligaments may have broken because of excessive stress?Da PrODOL C)O00005 D U 11 2C;7 /zaheH4/!:,( a>AO --L/

GL 97-06December 30, 1997 during the final thermal treatment of the monobloc steam generators, which in turn was causedby inadequate clearance for differential thermal expansion between the support plates,wrapper, and seismic restraints.As previously documented in IN 96-09, Supplement 1, a third damage mechanism involvedwastage not associated with chemical cleaning and affected tube support plates at variouselevations. This damage mechanism is active (progressive) and apparently involves acorrosion or erosion-corrosion mechanism of undetermined origin.The staffs of potentially affected foreign reactors are currently inspecting steam generators forevidence of the various damage mechanisms, both visually and with eddy-current testing.Tubes without adequate lateral support are being plugged.IN 96-09, Supplement 1, also documented that cooling transients involving the injection of largequantities of auxiliary feedwater may have been a key factor in the steam generator wrapperdrop phenomenon observed at a foreign PWR facility. These cooling transients are believed tohave been particularly severe for two units as a result of the use of a special operatingprocedure to accelerate the transition from hot to cold shutdown. The weight of the wrapperassembly and support plates is borne by six tenons mounted on the steam generator shell. Thewrapper is nominally free to expand axially relative to the shell. However, it is postulated thatan interference fit developed between the wrapper and the seismic restraints (mounted to theshell) as a result of differential thermal expansion associated with the cooling transients at theseventh support plate elevation. This interference fit prevented axial expansion of the wrapper,which led to excessive vertical bearing loads at the tenon supports, thus causing localizedwrapper failure at this location and downward displacement of the wrapper (20 millimetersmaximum). Poor quality wrapper support welds may also have contributed to this failure.Repairs have been made at the affected foreign PWR facility. Wrapper dropping is beingmonitored in all steam generators of similar design. The monitoring is performed through bothonline instrumentation and visual inspections during outages. In addition to the wrapperdropping problem, cracking of the wrapper above the original upper support was discovered atthe same foreign unit. The cause of the cracking is not yet known.At a U.S. PWR facility, degradation of steam generator tube eggcrate supports was discoveredthrough secondary side visual inspections performed during the spring 1997 refueling outage.The licensee identified erosion corrosion as the damage mechanism; the cause is not yetknown. The damage appears to be confined to the periphery and the untubed staywell regionof the tube bundle. The eggcrate degradation at the periphery extends inward to the first one ortwo rows of tubes. The degradation at the staywell region primarily affects the supportstructures within the untubed section. Damage to the eggcrate supports was found in bothsteam generators on the hot and cold leg sides although the damage was more extensive onthe hot leg side. No degradation of eggcrate supports was identified within the tube bundle.DiscussionThe reported foreign and U.S. experience highlights the potential for degradation mechanismsthat may lead to tube support plate and tube bundle wrapper damage. The steam generatortube support plates support the tubes against lateral displacement and vibration and minimize GL 97-06December 30, 1997 bending moments in the tubes in the event of an accident. Support plate damage can impairthe support plate's ability to perform this function and, thus, could potentially lead to theimpairment of tube integrity. Vibration-induced fatigue could present a potential problem if tubesupport plates lose integrity, particularly in areas of high secondary side cross flows. Aspreviously noted in IN 96-09, tube support plate signal anomalies found during eddy-currenttesting of the steam generator tubes may be indicative of support plate damage or ligamentcracking. Certain visual and video camera inspections on the secondary side of a steamgenerator may also provide useful information concerning the degradation of steam generatorinternals. The NRC staff will continue to monitor information on tube support plate and tubebundle wrapper damage as it becomes available.This letter also alerts addressees to the importance of performing comprehensive examinationsof steam generator intemals to ensure steam generator tube structural integrity is maintained inaccordance with the requirements of Appendix B to 10 CFR Part 50. More specifically,Criterion XVI of Appendix B, "Corrective Action," requires, in part, that measures be establishedto assure that conditions adverse to quality are promptly identified and corrected.Required InformationWithin 90 days of the date of this generic letter, each addressee is required to provide a writtenreport that includes the following information for its facility:(1) Discussion of any program in place to detect degradation of steam generator internalsand 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 oftube support plate signal anomalies from eddy-current testing of the steamgenerator tubes that may be indicative of support plate damage or ligamentcracking. If the addressee has performed such a review, include a discussion ofthe findings.(b) Whether visual or video camera inspections on the secondary side of the steamgenerators have been performed at the facility to gain information on thecondition of steam generator intemals (e.g., support plates, tube bundlewrappers, or other components). If the addressee has performed suchinspections, include a discussion of the findings.(c) Whether degradation of steam generator internals has been detected at thefacility, and how the degradation was assessed and dispositioned.(2) If the addressee currently has no program in place to detect degradation of steamgenerator intemals, include a discussion and justification of the plans and schedule forestablishing such a program, or why no program is neede GL 97-06December 30, 1997

Addressees

are encouraged to work closely with industry groups on the coordination ofinspections, evaluations, and repair options for all types of steam generator degradation thatmay be found.The NRC is aware that the industry has developed generic guidance on performing steamgenerator inspections, and that this guidance is continually being updated. If an addresseeintends to follow the guidance developed by the industry for this issue, reference to the relevantgeneric guidance documents is acceptable, and encouraged, as part of the response, as longas the referenced documents have been officially submitted to the NRC. However, additionalplant-specific information will be needed.NRC staff will review the responses to this generic letter and if concerns are identified, affectedaddressees will be notified.Address the required written responses to the U.S. Nuclear Regulatory Commission, ATTN:Document Control Desk, Washington, D.C. 20555-0001, under oath or affirmation under theprovisions of Section 182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f).

Backfit Discussion

Under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and10 CFR 50.54(f), this generic letter transmits an information request for the purpose of verifyingcompliance with applicable existing regulatory requirements. Specifically, the requestedinformation will enable the NRC staff to determine whether the condition of the addressees'steam generator internals comply with and conform to the current licensing bases for theirrespective facilities. In particular, the information would help the staff to ascertain whether theregulatory requirements pursuant to Appendix B to 10 CFR Part 50 are met.No backfit is either intended or approved in the context of issuance of this generic letter.Therefore, the staff has not performed a backfit analysis.

Federal Register Notification

A notice of opportunity for public comment was published in the Federal Register onDecember 31,1996 (61 FR 69116).

Paperwork Reduction Act Statement

This generic letter contains information collections that are subject to the Paperwork ReductionAct of 1995 (22 U.S.C. 3501 et seq.). These information collections were approved by theOffice of Management and Budget, approval number 3150-0011, which expires onSeptember 30, 2000.The public reporting burden for this collection of information is estimated to average 80 hour9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />sper response, including the time for reviewing instructions, searching existing data sources,gathering and maintaining the data needed, and completing and reviewing the collection of GL 97-06December 30, 1997 information. The U.S. Nuclear Regulatory Commission is seeking public comment on thepotential impact of the collection of information contained in the generic letter and on thefollowing issues:(1) Is the proposed collection of information necessary for the proper performance of thefunctions of the NRC, including whether the information will have practical utility?(2) Is the estimate of burden accurate?(3) Is there a way to enhance the quality, utility, and clarity of the information to be collected?(4) How can the burden of the collection of information be minimized, including the use ofautomated collection techniques?Send comments on any aspect of this collection of information, including suggestions forreducing this burden, to the Information and Records Management Branch, T-6F33,U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the Desk Officer,Office of Information and Regulatory Affairs, NEOB-10202 (3150-0011), Office of Managementand Budget, Washington, DC 20503.The NRC may not conduct or sponsor, and a person is not required to respond to, a collectionof information unless it displays a currently valid OMB control number.If you have any questions about this matter, please contact the technical contact listed below orthe appropriate Office of Nuclear Reactor Regulation (NRR) project manager.Jack W. Roe, ing DirectorJ1 Division of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contact: Stephanie M. Coffin, NRR301-415-2778E-mail: smclnrc.govLead Project Manager: George F. Wunder, NRR301-415-1494E-mail: gfw~nrc.gov

Attachment:

List of Recently Issued NRC Generic Lettersh4"k"" T.i &' -

AttachmentGL 97-06December 30, 1997Page 1 of ILIST OF RECENTLY ISSUED GENERIC LETTERSGENERICI FTTFRDATE OFSUBJECT ISSUANCEISSUED TO......97-0596-06,Sup. 191-18,Rev. 197-0497-03Steam Generator TubeInspection TechniquesAssurance of EquipmentOperability and ContainmentIntegrity During Design-BasisAccident ConditionsInformation to LicenseesRegarding NRC InspectionManual Section on Resolutionof Degraded and Nonconform-ing ConditionsAssurance of Sufficient NetPositive Suction Head forEmergency Core Coolingand Containment HeatRemoval PumpsAnnual Financial SuretyUpdate Requirements forUranium Recovery Licensees12/17/9711/13/9710/08/9710/07/9707/09/97All holders of OLs forpressurized-water reactors,except those who havepermanently ceasedoperations and have certifiedthat fuel has been perman-ently removed from thereactor vesselAll holders of OLs for nuclearpower reactors except thosewho have permanentlyceased operations and havecertified that fuel has beenpermanently removed fromthe reactor vesselAll holders of OLs for nuclearpower and NPRs, includingthose power reactorlicensees who have per-manently ceased operations,and all holders of NPRlicenses whose license nolonger authorizes operationAll holders of OLs for nuclearpower plants, except thosewho have permanentlyceased operations and havecertified that fuel has beenpermanently removed fromthe reactor vesselUranium recovery licenseesand state officialsOP = Operating LicenseCP = Construction PermitNPR = Nuclear Power Reactors GL 97-06December 30, 1997 information. The U.S. Nuclear Regulatory Commission is seeking public comment on thepotential impact of the collection of information contained in the generic letter and on thefollowing issues:(1) Is the proposed collection of information necessary for the proper performance of thefunctions of the NRC, including whether the information will have practical utility?(2) Is the estimate of burden accurate?(3) Is there a way to enhance the quality, utility, and clarity of the information to be collected?(4) How can the burden of the collection of information be minimized, including the use ofautomated collection techniques?Send comments on any aspect of this collection of information, including suggestions forreducing this burden, to the Information and Records Management Branch, T-6F33,U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the Desk Officer,Office of Information and Regulatory Affairs, NEOB-1 0202 (3150-0011), Office of Managementand Budget, Washington, DC 20503.The NRC may not conduct or sponsor, and a person is not required to respond to, a collectionof information unless it displays a currently valid OMB control number.If you have any questions about this matter, please contact the technical contact listed below orthe appropriate Office of Nuclear Reactor Regulation (NRR) project manager.original signed by D.B. MatthewsJack W. Roe, Acting DirectorDivision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contact: Stephanie M. Coffin, NRR301-415-2778E-mail: smcl@nrc.govLead Project Manager:George F. Wunder, NRR301-415-1494E-mail: gfwinrc.gov

Attachment:

List of Recently Issued NRC Generic Letters*SEE PREVIOUS CONCURRENCEDOCUMENT NAME: 97-06.GLTo receive a copy of this document. Indicate In the box: 'C' -Copy w/oattachment/enclosure 'E' -Copy w/attachment/enclosure 'N' -No copyOFFICE ITECH CONT M fyC l C:PECl3:DsNAME SMCoffin* MRafky* SARichardsRDATE 11/10/9 11/08/97 11/24/97OFFICIAL RECORD COPY(A)D: E IJWRoe El1ijja~lmq