[Federal Register Volume 69, Number 185 (Friday, September 24, 2004)]
[Notices]
[Pages 57367-57368]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-21424]


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NUCLEAR REGULATORY COMMISSION


Notice of Clarification to Steam Generator Tube Integrity Event 
Reporting Guideline in NUREG-1022, ``Event Reporting Guidelines 10 CFR 
50.72 and 50.73''

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of clarification in reporting guideline for steam 
generator tube integrity event.

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SUMMARY: The U.S. Nuclear Regulatory Commission has made a 
clarification in the reporting guideline for serious steam generator 
tube degradation contained within Revision 2 to NUREG-1022, ``Event 
Reporting Guidelines 10 CFR 50.72 and 50.73.'' The NRC will issue an 
errata to NUREG-1022, Revision 2. The purpose of this clarification is 
to ensure that the NRC receives timely notification of serious steam 
generator tube degradation.

SUPPLEMENTARY INFORMATION: On February 18, 2004, the NRC staff issued a 
Federal Register notice (69 FR 7661) that requested comments on the 
staff's intent to issue errata to Revision 2 of NUREG-1022, ``Event 
Reporting Guidelines 10 CFR 50.72 and 50.73.'' The errata would 
indicate that steam generator tube degradation is considered serious if 
either of the two criteria specified in Section 3.2.4(A)(3) of NUREG-
1022, Revision 2, is not satisfied.
    Steam generator tube degradation is currently characterized in 
Section 3.2.4(A)(3) of NUREG-1022 as being seriously degraded if the 
tubing fails to meet the following two performance criteria:
    (A) Steam generator tubing shall retain structural integrity over 
the full range of normal operating conditions (including startup, 
operation in the power range, hot standby, and cooldown and all 
anticipated transients included in the design specification) and design 
basis accidents. This includes retaining a margin of 3.0 against burst 
under normal steady state full power operation and a margin of 1.4 
against burst under the limiting design basis accident concurrent with 
a safe shutdown earthquake.
    (B) The primary to secondary accident induced leakage rate for the 
limiting design basis accident, other than a steam generator tube 
rupture, shall not exceed the leakage rate assumed in the accident 
analysis in terms of total leakage rate for all steam generators and 
leakage rate for an individual steam generator. The licensing basis 
accident analyses typically assume a 1 gallon per minute primary to 
secondary leak rate per steam generator, except for specific types of 
degradation at specific locations where the tubes are confined, as 
approved by the NRC and enumerated in conjunction with the list of 
approved repair criteria in the licensee's design basis documents.
    The first performance criterion is commonly referred to as the 
structural integrity performance criterion and the second criterion is 
commonly referred to as the accident induced leakage performance 
criterion. As written, NUREG-1022 could be read to indicate that the 
principal safety barrier (i.e., the steam generator tubes in this case) 
would only be considered seriously degraded if it had neither 
structural nor leakage integrity. Accordingly, if the steam generator 
tubes lacked only one of structural or leakage integrity, they would 
not be considered seriously degraded. This is contradictory to existing 
NRC regulations which require, in part, that the reactor coolant 
pressure boundary (which includes the steam generator tubes) be 
designed to permit periodic inspection and testing of important areas 
and features to assess both their structural and leaktight integrity 
(refer to General Design Criterion 32 of Appendix A to 10 CFR part 50) 
and be designed and tested so as to have an extremely low probability 
of abnormal leakage, of rapidly propagating failure, and of gross 
rupture (refer to General Design Criterion 14 of Appendix A to 10 CFR 
part 50). The regulations, therefore, indicate that both structural and 
leakage integrity criteria must be satisfied, and not meeting either 
one of the two performance criteria should constitute serious 
degradation of the principal safety barrier.
    In response to the Federal Register notice, one public comment was 
received from Progress Energy (ML040850494). The comment was that the 
notice did not indicate whether the new criteria would require the re-
evaluation of the reportability of existing steam generator tube 
degradation that was previously evaluated based on the criteria that 
were in effect before issuance of the errata. The commenter also 
indicated that

[[Page 57368]]

retroactive application of the new event reporting criteria to 
previously evaluated events would add burden to the licensees but would 
not provide timely notification to the NRC. Based on this comment and 
the reasons set forth below, the staff recommends that the errata 
clarify that retroactive notification is necessary only required if 
either of the criteria were exceeded during the last steam generator 
tube inspections.
    The errata to NUREG-1022 are intended to clarify existing 
requirements rather than to establish new requirements or criteria; 
however, the NRC recognizes that the wording in NUREG-1022 may have 
resulted in confusion regarding whether a report was required, given 
the condition of the tubes. As a result, the staff assessed the purpose 
of the report, other steam generator tube inspection reports received, 
and the potential value of evaluating previous inspection results. 
These items are discussed further below.
    The main purpose of the event report is to notify the staff, in a 
timely manner, of significant degradation of the steam generator tubes. 
This report allows the staff to review the corrective actions taken, to 
assess the generic implications of the findings, and to take any 
regulatory action that may be appropriated. From a practical 
perspective, the staff and public are informed of the results of the 
steam generator tube inspections following each inspection through 
reports submitted to the NRC in accordance with technical specification 
reporting requirements. These reports are typically submitted to the 
NRC within one year of the inspection. As a result, if a licensee were 
to experience significant degradation of the steam generator tubes, the 
staff and public would have the opportunity to identify this through 
the review of these reports. In addition, it is highly likely that if 
significant degradation was observed, it would have been assessed as 
part of the reactor oversight process. For this reason, retroactive 
notification of previous occurrences when either criterion was exceeded 
is not likely to provide any new information. This logic holds for all 
previous inspections except for the last steam generator tube 
inspections since these results may not have been reported and/or the 
NRC may not have completed its review of these reports. As a result, 
the staff concludes that the last steam generator tube inspection 
results should be reviewed and if either criterion was exceeded, this 
should be reported in accordance with 10 CFR 50.72 and 50.73. Given 
that the industry's steam generator initiative (referred to as NEI 97-
06) has essentially the same criteria and all pressurized water 
reactors have committed to follow this initiative, no significant 
burden should be imposed on any licensee in assessing whether the 
criteria were exceeded during the last steam generator tube inspection.

ADDRESSES: Submit written comments to the Chief, Rules and Directives 
Branch, Division of Administrative Services, Office of Administration, 
U.S. Nuclear Regulatory Commission, Mail Stop T6-D59, Washington, DC 
20555-0001, and cite the publication date and page number of this 
Federal Register notice. Written comments may also be delivered to NRC 
Headquarters, 11545 Rockville Pike (Room T6-D59), Rockville, Maryland, 
between 7:30 a.m. and 4:15 p.m. on Federal workdays.

FOR FURTHER INFORMATION, CONTACT: Samuel S. Lee at (301) 415-1061 or by 
e-mail to <a href="/cdn-cgi/l/email-protection#97e4e4fbd7f9e5f4b9f0f8e1"><span class="__cf_email__" data-cfemail="a8dbdbc4e8c6dacb86cfc7de">[email protected]</span></a>, or Kenneth J. Karwoski at (301) 415-2752 or by 
e-mail to <a href="/cdn-cgi/l/email-protection#91fafbfaa0d1ffe3f2bff6fee7"><span class="__cf_email__" data-cfemail="3d5657560c7d534f5e135a524b">[email protected]</span></a>.

    Dated in Rockville, Maryland, this 27th day of August, 2004.

    For the Nuclear Regulatory Commission.
Francis M. Costello,
Acting Chief, Reactor Operations Branch, Division of Inspection Program 
Management, Office of Nuclear Reactor Regulation.
[FR Doc. 04-21424 Filed 9-23-04; 8:45 am]
BILLING CODE 7590-01-P


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