[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
