Proposed Generic Communication; Degradation of Steam Generator Internals

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

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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

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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]

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