[Federal Register Volume 61, Number 252 (Tuesday, December 31, 1996)]
[Notices]
[Pages 69116-69118]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-33250]


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NUCLEAR REGULATORY COMMISSION

Proposed Generic Communication; Degradation of Steam Generator 
Internals

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 the degradation of steam generator 
internals at foreign pressurized-water reactor facilities. The purpose 
of the proposed generic letter is to (1) re-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) 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 
(3) request all addressees to submit information that will enable the 
NRC staff to verify whether or not the condition of addressees' steam 
generator

[[Page 69117]]

internals comply and conform with the current licensing basis for their 
respective facilities. The NRC is seeking comment from interested 
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: Stephanie M. Coffin, (301) 415-2778.

SUPPLEMENTARY INFORMATION:

NRC Generic Letter 96-XX: Degradation of Steam Generator Internals

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) re-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) 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 (3) request all 
addressees to submit information that will enable the NRC staff to 
verify whether or not the condition of addressees' steam generator 
internals comply and conform with the current licensing basis for their 
respective facilities.
Background
    The NRC issued Information Notice (IN) 96-09 and IN 96-09, 
Supplement 1 to alert addressees to findings of damage to steam 
generator internals at foreign PWR facilities.
Description of Circumstances
    Foreign authorities have reported various steam generator tube 
support plate damage mechanisms. The affected steam generators are 
similar, but not identical, to Westinghouse Model 51 steam generators. 
As previously documented in IN 96-09 and IN 96-09, Supplement 1, one 
damage mechanism involved the wastage of the uppermost support plate 
caused by the misapplication of a chemical cleaning process. A second 
damage mechanism involved broken tube support plate ligaments at the 
uppermost, and sometimes at the next lower, tube support plates. The 
support plate ligaments broke near a radial seismic restraint and near 
an antirotation 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 during the final 
thermal treatment of the monobloc steam generators, which in turn was 
caused by 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 involved wastage not associated with chemical cleaning and 
affected tube support plates at various elevations. This damage 
mechanism is active (progressive) and apparently involves a corrosion 
or erosion-corrosion mechanism of undetermined origin.
    The staffs of potentially affected foreign reactors are currently 
inspecting steam generators for evidence 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 large quantities of auxiliary feedwater may 
have been a key factor in the steam generator wrapper drop phenomenon 
observed at a foreign PWR facility. These cooling transients are 
believed to have been particularly severe for two units as a result of 
the use of a special operating procedure to accelerate the transition 
from hot to cold shutdown. The weight of the wrapper assembly and 
support plates is borne by six tenons mounted on the steam generator 
shell. The wrapper is nominally free to expand axially relative to the 
shell. However, it is postulated that an interference fit developed 
between the wrapper and the seismic restraints (mounted to the shell) 
as a result of differential thermal expansion associated with the 
cooling transients at the seventh 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 
localized wrapper failure at this location and downward displacement of 
the wrapper (20 millimeters, maximum). Poor quality wrapper support 
welds may also have contributed to this failure. Repairs have been 
implemented at the affected foreign PWR facility. Wrapper dropping is 
being monitored in all steam generators of similar design. The 
monitoring is through online instrumentation and through visual 
inspections during outages. In addition to the wrapper dropping 
problem, cracking of the wrapper above the original upper support was 
discovered at the same foreign unit. The cause of the cracking is not 
yet known.
Discussion
    The reported foreign experience highlights the potential for 
degradation mechanisms that may lead to tube support plate and tube 
bundle wrapper damage. The steam generator tube support plates support 
the tubes against lateral displacement and vibration and minimize 
bending moments in the tubes in the event of an accident. Support plate 
damage can impair their ability to perform this function and, thus, 
could potentially lead to the impairment of tube integrity. Vibration-
induced fatigue could present a potential problem if tube support 
plates lose integrity, particularly in areas of high secondary side 
crossflows. As previously noted in IN 96-09, tube support plate signal 
anomalies found during eddy current testing of the steam generator 
tubes may be indicative of support plate damage or ligament cracking. 
Certain visual and video camera inspections on the secondary side of a 
steam generator may also provide useful information concerning the 
degradation of steam

[[Page 69118]]

generator internals. The NRC staff will continue to monitor information 
on tube support plate and tube bundle wrapper damage as it becomes 
available from foreign authorities.
    This letter also alerts addressees to 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. Criterion XI of 
Appendix B, ``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 the applicable design documents. The applicable steam 
generator tube design documents include General Design Criteria (GDCs) 
14, 15, 30, 31, and 32 of 10 CFR Part 50, Appendix A and Section III of 
the ASME Boiler and Pressure Vessel code. Criterion XVI of Appendix B, 
``Corrective Action,'' requires in part that measures be established to 
assure that conditions adverse to quality are promptly identified and 
corrected.

Requested Information

    Within 60 days of the date of this generic letter, each addressee 
is requested to provide a written report that includes the following 
information for its facility:
    (1) Discussion of the program in place, if any, to detect 
degradation of steam generator internals and a description of the 
inspection plans, including the inspection scope, frequency, methods, 
equipment and criteria, and plans for corrective action in the event 
degradation is found.
    The discussion should include the following information:
    (a) Whether past 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 
provide 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, the written response should 
include a discussion of the plans for establishing such a program, or a 
justification as to why no such program is needed.
    Addressees are encouraged to work closely with industry groups on 
the coordination of inspections, evaluations, and repair options for 
all types of steam generator degradation that may be found.
    The NRC is aware that the industry has developed generic industry 
guidance on performing steam generator inspections, and that this 
guidance is continually being updated. If an addressee intends to 
follow the guidance developed by the industry for this issue, reference 
to the relevant generic guidance documents is acceptable, and 
encouraged, as part of the response, as long as the referenced 
documents have been officially submitted to the NRC. However, 
additional plant-specific information will be needed.

Required Response

    Within 30 days of the date of this generic letter, each addressee 
is required to submit a written response indicating: (1) Whether or not 
the requested information will be submitted and (2) whether or not the 
requested information will be submitted within the requested time 
period. Addressees who 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 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 the provisions 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, and 10 CFR 50.54(f), this generic letter transmits an 
information request for the purpose of verifying compliance with 
applicable existing regulatory requirements. Specifically, the 
requested information will enable the NRC staff to determine whether or 
not the condition of the addressees' steam generator internals 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 are 
met, namely, (1) Criterion XI, ``Test Control,'' concerning the 
establishment of effective test programs for systems, structures and 
components, and (2) Criterion XVI, ``Corrective Action,'' which 
requires that measures shall be established to assure that conditions 
adverse to quality, such as failures, malfunctions, deficiencies, 
deviations, defective material and equipment, and nonconformances are 
promptly identified and corrected. 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. 96-33250 Filed 12-30-96; 8:45 am]
BILLING CODE 7590-01-P