[Federal Register Volume 61, Number 252 (Tuesday, December 31, 1996)]
[Notices]
[Pages 69118-69120]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-33248]


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NUCLEAR REGULATORY COMMISSION

Proposed Generic Communication; Steam Generator Tube Inspection 
Techniques (M96401)

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of opportunity for public comment.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to issue 
a generic letter concerning steam generator tube inspection practices 
at pressurized-water reactor facilities. The purpose of the proposed 
generic letter is to (1) emphasize to 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) request certain information from addressees to verify 
whether or not steam generator tube inservice inspection practices 
comply and conform with the current licensing basis for their 
respective facilities. The NRC is seeking comment from interested

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parties regarding both the technical and regulatory aspects of the 
proposed generic letter presented under the Supplementary Information 
heading.
    The proposed generic letter was endorsed by the Committee to Review 
Generic Requirements (CRGR) on December 17, 1996. The relevant 
information that was sent to the CRGR will be placed in the NRC Public 
Document Room. The NRC will consider comments received from interested 
parties in the final evaluation of the proposed generic letter. The 
NRC's final evaluation will include a review of the technical position 
and, as appropriate, an analysis of the value/impact on licensees. 
Should this generic letter be issued by the NRC, it will become 
available for public inspection in the NRC Public Document Room.

DATES: Comment period expires January 30, 1997. Comments submitted 
after this date will be considered if it is practical to do so, but 
assurance of consideration cannot be given except for comments received 
on or before this date.

ADDRESSES: Submit written comments to Chief, Rules Review and 
Directives Branch, U.S. Nuclear Regulatory Commission, Mail Stop T-6D-
69, Washington, DC 20555-0001. Written comments may also be delivered 
to 11545 Rockville Pike, Rockville, Maryland, from 7:30 am to 4:15 pm, 
Federal workdays. Copies of written comments received may be examined 
at the NRC Public Document Room, 2120 L Street, N.W. (Lower Level), 
Washington, D.C.

FOR FURTHER INFORMATION CONTACT: Phillip J. Rush, (301) 415-2790.

SUPPLEMENTARY INFORMATION:

NRC Generic Letter 96-XX: Steam Generator Tube Inspection 
Techniques

Addressees

    All holders of operating licenses for pressurized water reactors 
(PWRs), except those licenses that have been amended to possession-only 
status.

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) request certain information from addressees to verify 
whether or not steam generator tube inservice inspection practices 
comply and conform with the current licensing basis for their 
respective facilities.

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 a requirement under Title 10 of the 
Code of Federal Regulations, Part 50 (10 CFR Part 50), Appendix A. 
Specifically, 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.
    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 eddy-current methods are a proven technique for detecting tube 
degradation, there has been only limited success in demonstrating the 
capability to accurately depth size indications from the eddy-current 
signals. 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. Through 
recent inspections and discussions of eddy-current practice with 
various licensees, the NRC has become aware that several utilities are 
allowing degraded steam generator tubes to remain in service on the 
basis of estimates of IGA and SCC degradation depths using eddy-current 
methods.

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, 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,

[[Page 69120]]

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 mode of degradation, e.g., cracklike, IGA, corrosion 
induced thinning, or wear and the orientation for cracklike 
degradation, and (3) accurately sizing the depth of the indication. 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)-one 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., probe) 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 the structural and leakage integrity margins for steam 
generator tubes provided in the General Design Criteria (Appendix A to 
10 CFR Part 50) are satisfied, the NRC has concluded that it is 
appropriate for the addressees to review the types of steam generator 
tube degradation that are being left in service based on sizing, the 
inspection method being used to perform the sizing for each type of 
degradation, and the technical basis for the acceptability of each 
inspection method.

Requested Information

    Within 60 days of the date of this generic letter, all addressees 
are requested to provide the following information: (1) Whether it is 
their practice to leave steam generator tubes with defects in service, 
based on sizing, and (2) if the response to item (1) is affirmative, 
those licensees are requested to submit a written report that includes, 
for each type of steam generator degradation mechanism, a description 
of the associated nondestructive examination method being used and the 
technical basis for the acceptability of the technique used.

Required Response

    Within 30 days of the date of this generic letter, addressees are 
required to submit a written response indicating: (a) Whether or not 
the requested information will be submitted, and (b) whether or not the 
requested information will be submitted within the requested time 
period. Addressees who respond in the affirmative to item (1) under 
Requested Information and choose not to submit the requested 
information, or are unable to satisfy the requested completion date, 
must describe in their response any alternative course of action that 
is proposed to be taken, including the basis for the acceptability of 
the proposed alternative course of action.
    NRC staff will review the responses to this generic letter and if 
concerns are identified, affected addressees will be notified.
    Address written material 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).

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). The information requested will enable 
the NRC staff to determine whether addressees' steam generator tube 
inspection practices comply and conform with the current licensing 
basis for their respective facilities. In particular, it would help 
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. 
Additionally, 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.

    Dated at Rockville, Maryland, this 23rd day of December, 1996.

    For the Nuclear Regulatory Commission.
David B. Matthews,
Acting Director, Division of Reactor Program Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 96-33248 Filed 12-30-96; 8:45 am]
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