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Home       > Electronic Reading       Room       > Document       Collections       > Generic       Communications >       Generic       Letters >       1997       > GL97005 
				UNITED STATES                         NUCLEAR REGULATORY COMMISSION                     OFFICE OF NUCLEAR REACTOR REGULATION                         WASHINGTON, D.C.  20555-0001                                  December 17, 1997NRC GENERIC LETTER 97-05:  STEAM GENERATOR TUBE INSPECTION TECHNIQUESAddresseesAll holders of operating licenses for pressurized-water reactors (PWRs),except those who have permanently ceased operations and have certified thatfuel has been permanently removed from the reactor vessel.PurposeThe U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to(1) emphasize to the addressees the importance of performing steam generatortube inservice inspections using qualified techniques in accordance with therequirements of Appendix B to  10 CFR Part 50, and (2) require certaininformation from addressees to determine whether they are in compliance withthe current licensing basis for their respective facilities given their steamgenerator tube inservice inspection practices.BackgroundSteam generator tubing constitutes a significant portion of the reactorcoolant pressure boundary (RCPB).  The design of the RCPB for structural andleakage integrity is addressed in either Title 10 of the Code of FederalRegulations, Part 50 (10 CFR Part 50), Appendix A or the licensing basis of afacility.  The General Design Criteria (GDC) of Appendix A state that the RCPBshall "have an extremely low probability of abnormal leakage" (GDC 14), "shallbe designed with sufficient margin to assure that the design conditions of thereactor coolant pressure boundary are not exceeded during any condition ofnormal operation" (GDC 15), and "shall be designed to permit periodicinspection and testing of important areas and features to assess theirstructural and leaktight integrity" (GDC 32).Once a plant is in operation, licensees are required by their technicalspecifications to perform periodic inservice inspections of the steamgenerator tubing and to repair or remove from service all tubes withdegradation exceeding the tube repair limits.  Eddy-current inspectiontechniques are the primary means by which licensees assess the condition ofthe steam generator tubes.  Such inspections are an important component of thedefense-in-depth measures to ensure the structural and leaktight integrity ofthe steam generator tubes.9712120014.                                                                GL 97-05                                                                December 17, 1997                                                                Page 2 of 6  The NRC issued Generic Letter (GL) 95-03, "Circumferential Cracking of SteamGenerator Tubes," on April 28, 1995.  One of the purposes of GL 95-03 was toemphasize the importance of utilizing qualified inspection techniques and equipment capable of reliably detecting steam generator tube degradation. Criterion IX, "Control of Special Processes," contained in Appendix B to10 CFR Part 50 states, in part, that "measures shall be established to assurethat special processes, including ... nondestructive testing, are controlledand accomplished by qualified personnel using qualified procedures."  Althoughthe main focus of GL 95-03 was to address circumferential steam generator tubecracking, the requirement of using qualified inspection techniques applies toall inspections for all forms of tube degradation.Criterion XI, "Test Control," requires, in part, that a test program beestablished to assure that all testing required to demonstrate thatstructures, systems, and components will perform satisfactorily in service isidentified and performed in accordance with written test procedures whichincorporate the requirements and acceptance limits contained in applicabledesign documents.  Licensees have traditionally relied upon eddy-current inspection techniques toassess the condition of their steam generator tubes.  Although the eddy-current method is a proven technique for detecting tube degradation, theability to depth size indications is possible only for specific modes ofdegradation.  Specifically, tube degradation from intergranular attack (IGA)and stress corrosion cracking (SCC), major modes of steam generator tubedegradation, are difficult to size with eddy-current inspection techniquesbecause of a number of complicating variables.  In one recent instance, alicensee employed a technique to size the depths of IGA tube degradation basedon tube specimens removed from two plants.  However, pulled tube data analyzedafter the initial application of the technique indicated that the method didnot adequately estimate the true depth of the indications consistent with thecriteria established for qualifying the sizing technique.Discussion(1)    Evaluation of Recent Inspection ExperienceIn general, plant technical specifications require the removal from service orthe repair of those steam generator tubes with degradation exceeding 40percent of the nominal tube-wall thickness.  Criterion IX in Appendix B to10 CFR Part 50 requires that nondestructive testing be completed usingqualified procedures.  Therefore, licensees must be able to demonstratethrough the qualification process that an inspection technique used for sizingsteam generator tube indications can measure the through-wall penetration ofcracks and other forms of degradation with an accuracy commensurate with the"bases" of the tube repair limits in the technical specifications.Theoretically, there is a relationship between the depth of penetration of adefect and the eddy-current signal response; in practice, however, therelationship between signal voltage or phase angle and the degradation depthis influenced by many other variables.  Oxide deposits, .                                                                GL 97-05                                                                December 17, 1997                                                                Page 3 of 6variability of tube material properties and geometry, degradation morphology,human factors, and eddy-current data analysis and acquisition practices aresome of the factors that can significantly alter a depth estimation of steamgenerator tube degradation.  The NRC is aware that the depth of severalspecific forms of volumetric steam generator tube degradation can be sizedwith a reasonable degree of accuracy; however, qualifying techniques forsizing of some forms of degradation, e.g., IGA and SCC, has been problematic.In order to successfully disposition steam generator tube degradation inaccordance with the repair limits in the technical specifications andAppendix B to 10 CFR Part 50, the inspection process must be capable of(1) detecting indications of tube degradation, (2) characterizing theindications,  e.g., cracklike, IGA, manufacturing burnish mark, or wear andthe orientation for cracklike degradation,  and (3) accurately sizing thedepth of degradation.  The term "inspection process" refers to the use of oneor a combination of nondestructive inspection techniques to evaluate aspecific mode of steam generator tube degradation.  This evaluation couldpotentially include three inspection methods (e.g., eddy-current probes)þonefor detection, one for characterization, and a third to size the indication. However, the successful qualification of the inspection process requires aqualification of each method (i.e., probes, cables, software, etc.) for themode of degradation being evaluated in the steam generator tube examinations. Experience has demonstrated that for effective qualification the data setdemonstrating the capability of the inspection process should consist, to theextent practical, of service-degraded tube specimens (i.e., specimens removedfrom operating steam generators), supplemented, as necessary, by tubespecimens containing flaws fabricated using alternative methods provided thatthe nondestructive examination parameter responses from these flaws are fullyconsistent with actual inservice degradation of the same flaw geometry.(2)    Safety AssessmentSteam generator tube degradation is managed through a combination of severaldefense-in-depth measures including inservice inspection, tube repaircriteria, primary-to-secondary leak rate monitoring, water chemistry, operatortraining, and analyses to ensure safety objectives are met.  In addition, onthe basis of NRC conclusions regarding the potential consequences of steamgenerator tube failure events in NUREG-0844, "NRC Integrated Program for theResolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding SteamGenerator Tube Integrity," the risk from the potential rupture of one or moretubes is small.  However, since tube ruptures represent a failure of one ofthe principal fission product boundaries and present a pathway for a releaseto the environment bypassing the containment, all reasonable precautionsshould be taken to prevent such an occurrence.To verify compliance with Appendix B to 10 CFR Part 50 and the technicalspecifications, and to maintain a reasonable level of assurance thatstructural and leakage integrity margins for steam generator tubes aresatisfied, the NRC has concluded that it is appropriate for the addressees toreview the types of steam generator tube indications that are being left inservice based on sizing, the inspection method being used to perform thesizing for each type of indication, and the technical basis for theacceptability of each inspection method. .                                                                GL 97-05                                                                December 17, 1997                                                                Page 4 of 6  Required InformationWithin 90 days of the date of this generic letter, addressees are required tosubmit a  written response that includes the following information:       (1)   whether it is their practice to leave steam generator tubes with             indications in service based on sizing,       (2)   if the response to item (1) is affirmative, those licensees should             submit a written report that includes, for each type of             indication, a description of the associated nondestructive             examination method being used and the technical basis for the             acceptability of the technique used.Address the required written information to the U.S. Nuclear RegulatoryCommission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, underoath or affirmation under the provisions of Section 182a, Atomic Energy Act of1954, as amended, and 10 CFR 50.54(f).  This information will enable theCommission to determine whether a license should be modified, suspended orrevoked.  In addition, submit a copy of the written information to theappropriate regional administrator.NRC staff will review the responses to this generic letter and if concerns areidentified, affected addressees will be notified.  Backfit DiscussionThis generic letter only requests information from the addressees under theprovisions of Section 182a of the Atomic Energy Act of 1954, as amended, and10 CFR 50.54(f).  This generic letter does not constitute a backfit as definedin 10 CFR 50.109(a)(1) since it does not impose modifications of or additionsto structures, systems or components or to design or operation of anaddressee's facility.  It also does not impose an interpretation of theCommission's rules that is either new or different from a previous staffposition.  The staff, therefore, has not performed a backfit analysis.Reason for Information RequestThis generic letter transmits an information request pursuant to theprovisions of Section 182a of the Atomic Energy Act of 1954, as amended, and10 CFR 50.54(f) for the purpose of verifying compliance with applicableregulatory requirements.  Specifically, the requested information will enablethe NRC staff to determine whether addressees, given their steam generatortube inspection practices, are in compliance with current licensing basis fortheir respective facilities.  In particular, this information will help toascertain whether or not the regulatory requirements pursuant to Appendix B to10 CFR Part 50, namely, Criterion IX, "Control of Special Processes," andCriterion XI, "Test Control," are met.  .                                                                GL 97-05                                                                December 17, 1997                                                                Page 5 of 6Federal Register NotificationA notice of opportunity for public comment was published in the FederalRegister (61 FR 69118) on December 31, 1996.  Comments were received from oneindustry organization and one licensee.  Copies of the staff evaluation ofthese comments have been made available in the public document room.Paperwork Reduction Act StatementThis generic letter contains information collections that are subject to thePaperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).  These informationcollections were approved by the Office of Management and Budget, approvalnumber 3150-0011, which expires September 30, 2000. The public reporting burden for this collection of information is estimated toaverage 75 hours per response, including the time for reviewing instructions,searching existing data sources, gathering and maintaining the data needed,and completing and reviewing the collection of information.  The U.S. NuclearRegulatory Commission is seeking public comment on the potential impact of thecollection of information contained in the generic letter and on the followingissues:       1. Is the proposed collection of information necessary for the proper          performance of the functions 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 of automated collection techniques?Send comments on any aspect of this collection of information, includingsuggestions for reducing this burden, to the Information and RecordsManagement Branch, T-6 F33,  U.S. Nuclear Regulatory Commission, Washington,DC 20555-0001, or by Internet electronic mail at bjs1@nrc.gov; and to the DeskOfficer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0011),Office of Management and Budget, Washington, DC 20503.The NRC may not conduct or sponsor, and a person is not required to respondto, a collection of information unless it displays a currently valid OMBcontrol number..                                                                GL 97-05                                                                December 17, 1997                                                                Page 6 of 6If you have any questions about this matter, please contact the technicalcontact listed below or the appropriate Office of Nuclear Reactor Regulation(NRR) project manager.                                       signed by                                       Jack W. Roe, Acting Director                                       Division of Reactor Program Management                                       Office of Nuclear Reactor RegulationTechnical contact:  Phillip Rush, NRR                    301-415-2790                    E-mail:  pjr1@nrc.govLead Project Manager:  Alexander W. Dromerick, NRR                       301-415-3473                       E-mail:  awd@nrc.gov