[Federal Register Volume 69, Number 194 (Thursday, October 7, 2004)]
[Notices]
[Pages 60193-60196]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-22546]
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NUCLEAR REGULATORY COMMISSION
Proposed Generic Communication; Steam Generator Tube Integrity
and Associated Technical Specifications
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice of opportunity for public comment.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
issue a generic letter (GL) to request that addressees submit a
description of their program for ensuring steam generator (SG) tube
integrity for the interval between inspections and description of the
methodology used to assess the effects of non-pressure-related loads
such as bending on SG tube integrity. Addressees should also provide a
safety assessment demonstrating that the SG tubes will have adequate
structural and leakage integrity (with appropriate regulatory margins)
at the time of their next SG tube inspection, taking into account the
effects of non-pressure-related loads.
This Federal Register notice is available through the NRC's
Agencywide Documents Access and Management System (ADAMS) under
accession number ML042710075.
DATES: Comment period expires December 6, 2004. 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.
ADDRESSEES: 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 T-6D59), Rockville, Maryland,
between 7:30 am and 4:15 pm on Federal workdays.
FOR FURTHER INFORMATION CONTACT: Kenneth Karwoski, NRR at 301-415-2752
or by e-mail at [email protected] or Maitri Banerjee, NRR at 301-415-2277 or
by e-mail at [email protected].
SUPPLEMENTARY INFORMATION:
Draft NRC Generic Letter 2004-XX: Steam Generator Tube Integrity and
Associated Technical Specifications
Addressees
All holders of operating licenses for pressurized-water reactors
(PWRs), except those who have permanently ceased operations and have
certified that fuel has been permanently removed from the reactor
vessel and {the following plants that have already modified their
technical specifications to be consistent with those in the
Attachment{time} .
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this
generic letter (GL) to:
(1) Request that addressees submit a description of their program
for ensuring steam generator (SG) tube integrity for the interval
between inspections; and
(2) Request that addressees submit a description of the methodology
used to assess the effects of non-pressure-related loads such as
bending on SG tube integrity. Addressees should also provide a safety
assessment demonstrating that the SG tubes will have adequate
structural and leakage integrity (with appropriate regulatory margins)
at the time of their next SG tube inspection, taking into account the
effects of non-pressure-related loads.
Discussion
Steam generator tubes function as an integral part of the reactor
coolant pressure boundary (RCPB) and also serve to isolate radiological
fission products in the primary coolant from the secondary coolant and
the environment. For the purposes of this generic letter, tube
integrity means that the tubes are capable of performing these
functions in accordance with the plant design and licensing basis,
including applicable regulatory requirements.
During operation, licensees are required to monitor and maintain
the condition of the SG tubing with the objective of ensuring its
continued integrity. Specifically, licensees are required by 10 CFR
50.55a(b)(2)(iii), 10 CFR 50.55a(g), or by the plant technical
specifications to perform periodic inservice inspections and to repair
(e.g., sleeve) or remove from service (by installing plugs in the tube
ends) all tubes found to contain flaws exceeding
[[Page 60194]]
the plugging limit (i.e., tube repair criteria).
The current technical specification requirements for inspection and
repair of SG tubing were developed in the 1970s and define a
prescriptive approach for ensuring tube integrity. This prescriptive
approach involves inspecting the tubing at specified intervals,
implementing specified tube inspection sampling plans, and repairing or
removing from service by plugging all tubes found by inspection to
contain flaws in excess of specified flaw repair criteria. However, as
evidenced by operating experience, the prescriptive approach defined in
the technical specifications may not be sufficient to ensure that tube
integrity is maintained. For example, in cases of low to moderate
levels of degradation, the technical specifications only require that
3-to 21-percent of the tubes be inspected, irrespective of whether the
inspection results indicate that additional tubes need to be inspected
to reasonably ensure that tubes with flaws that may exceed the tube
repair criteria or which may impair tube integrity are detected. In
addition, the technical specifications (and Section XI of the American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code)
do not explicitly address the inspection methods to be employed for
different tube degradation mechanisms at specific tube locations, nor
are the specific objectives to be fulfilled by the selected methods
explicitly defined. Also, incremental flaw growth between inspections
can in many instances exceed what is allowed for in the specified tube
repair criteria. In such cases, the specified inspection frequencies
may not ensure reinspection of a tube before its integrity is impaired.
In short, current technical specification surveillance requirements may
not require licensees to actively manage their SG programs so as to
provide reasonable assurance that tube integrity is maintained. As a
result of the above, licensees have frequently found it necessary to
implement measures beyond the technical specification requirements to
ensure adequate tube integrity. These measures are frequently
accompanied by interaction with the NRC staff in an oversight or review
capacity to ensure that adequate tube integrity is being maintained.
The NRC staff, with external stakeholder involvement, embarked on
efforts to improve the SG tube integrity regulatory framework as
discussed in SECY-03-0800, ``Steam Generator Tube Integrity (SGTI)--
Plans for Revising the Associated Regulatory Framework.'' As a result
of these efforts, the NRC and industry generically developed modified
technical specifications for addressing steam generator tube integrity.
These generically developed technical specifications were recently
incorporated into one facility's technical specifications. (Proposals
to change the plant-specific technical specifications are reviewed in
accordance with the license amendment review process to confirm their
acceptability). These modified technical specifications are attached to
this generic letter for your information. The approach taken in the
modified technical specifications in the Attachment is conceptually
similar to the approach outlined in the industry initiative referred to
as NEI 97-06, ``Steam Generator Program Guidelines.'' The modified
technical specifications in the Attachment are performance-based in
that they are focused on ensuring that the tubing satisfies performance
criteria that are commensurate with tube integrity. This approach can
be readily adapted to new or unexpected degradation mechanisms and
advances in nondestructive examination technology. This approach also
includes programmatic elements to ensure that tubes are being
adequately monitored and maintained relative to the performance
criteria.
The requirements pertaining to the integrity of the SG tubes are
contained within Title 10 of the Code of Federal Regulations (10 CFR).
Specifically, the general design criteria (GDC) \1\ described in
Appendix A to 10 CFR part 50 contain, in part, requirements related to
the RCPB (e.g., GDC 14, GDC 30, GDC 32). In addition to the GDC, 10 CFR
50.55a specifies that components that are part of the RCPB must meet
the requirements for Class 1 components in Section III and XI of the
American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel Code unless the plant technical specifications for surveillance
differ from those specified in the ASME Code, in which case the
technical specifications govern.
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\1\ Or, for PWR facilities licensed before the issuance of 10
CFR part 50, Appendix A, similar requirements in the plant-specific
principal design criteria.
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The requirements pertaining to the content of a plant's technical
specifications are given in 10 CFR 50.36, ``Technical Specifications.''
All currently operating PWR licensees have technical specifications
governing the surveillance of the SG tubes. These technical
specifications also include operational leakage limits so that if
significant leakage develops, the plant is shut down. The plugging
limits in the technical specifications were developed to ensure that
degraded tubes: (1) Maintain factors of safety against gross rupture
consistent with the plant design basis (i.e., consistent with the
stress limits of the ASME Code, Section III); and (2) maintain leakage
integrity consistent with the plant licensing basis while, at the same
time, allowing for potential flaw size measurement error and flaw
growth between inservice inspections.
As part of the plant licensing basis, applicants for PWR licenses
are also required to analyze the consequences of postulated design-
basis accidents. Typical accidents analyzed are the SG tube rupture,
the locked-rotor, control rod ejection, and a main steamline break.
These analyses consider the potential primary-to-secondary leakage
through the tubes during these events and must show that the offsite
radiological doses do not exceed 10 CFR part 100 limits (or some
fraction thereof) and GDC 19 of Appendix A to 10 CFR part 50.
Irrespective of technical specification requirements for SG tube
inspection and repair, licensees are also required by 10 CFR part 50,
Appendix B Criterion XVI, ``Corrective Action,'' to ensure that
conditions adverse to quality are promptly identified and corrected. In
the case of significant conditions adverse to quality, the measures
shall assure that the cause of the condition is determined and
corrective action taken to preclude repetition.
The staff is requesting information as to: (1) Actions licensees
are taking or will take to ensure tube integrity is being maintained,
and (2) contemplated changes to the technical specifications to reflect
these actions.
As discussed above, the approach in the attached technical
specifications is performance-based. There are three performance
criteria for the SG tubes: (1) A structural integrity performance
criterion, (2) a primary-to-secondary leakage performance criterion for
normal operation, and (3) a primary-to-secondary leakage performance
criterion for postulated accident conditions.
During public interactions with stakeholders on the structural
integrity performance criterion, the staff became aware that the
effects of various non-pressure-related loads such as bending loads may
not be fully addressed in industry guidance documents for assessing the
integrity of degraded SG tubes. Non-pressure-related loads were
assessed in the original design of the SG tubes so as to ensure that
nondegraded tubes would have adequate integrity for the full range of
operating conditions. As a result, this generic letter requests
addressees to discuss how they have
[[Page 60195]]
assessed the effects of non-pressure-related loads in their assessments
of tube integrity and to discuss whether all tubes will have adequate
structural integrity at the time of their next SG tube inspection,
taking all loading conditions on the tube into account.
Requested Information
Addressees are requested to provide the following information to
the NRC within 60 days of the date of this generic letter:
1. A description of the actions they are taking or will take to
ensure tube integrity is being maintained and contemplated changes to
the technical specifications to reflect these actions.
2. A description of the methodology used to assess the effects of
non-pressure-related loads such as bending on SG tube integrity. In
addition, addressees should provide a safety assessment demonstrating
that the SG tubes will have adequate structural and leakage integrity
at the time of their next SG tube inspection, taking into account the
effects of non-pressure-related loads.
Required Response
In accordance with 10 CFR 50.54(f), addressees are required to
submit written responses to this generic letter. Two options are
available:
(a) Addressees may choose to submit written responses providing the
information requested above within the requested time period.
(b) Addressees who cannot meet the requested completion date or who
choose an alternate course of action are required to so notify the NRC
in writing as soon as possible but no later than 30 days from the date
of this generic letter. The response must address any alternative
course of action proposed, including the basis for the acceptability of
the proposed alternative course of action, and the basis for finding
that the SGs remain operable. If the information requested in the
previous section of this GL will be subsequently provided, the response
must set forth the schedule for submitting the information.
The required written response should be addressed to the U.S.
Nuclear Regulatory Commission, ATTN: Document Control Desk, 11555
Rockville Pike, Rockville, Maryland 20852, under oath or affirmation
under the provisions of section 182a of the Atomic Energy Act of 1954,
as amended, and 10 CFR 50.54(f). In addition, a copy of the response
should be sent to the appropriate regional administrator.
Reasons for Requested Information
This GL requests addressees to submit information. The requested
information will enable the NRC staff to determine whether addressees'
SG tube integrity programs provide reasonable assurance of tube
integrity consistent with their design and licensing basis and
applicable regulatory requirements (10 CFR part 50, Appendix
A1; 10 CFR part 50, Appendix B). In addition, the requested
information will enable the NRC staff to determine whether SG tube
integrity is being maintained under all loading conditions consistent
with the design and licensing basis and applicable regulatory
requirements (10 CFR part 50, Appendix A1).
The NRC staff will review the responses to this GL in order to
determine whether additional actions are necessary.
Backfit Discussion
Under the provisions of section 182a of the Atomic Energy Act of
1954, as amended, and 10 CFR 50.54(f), this GL transmits an information
request for the purpose of verifying compliance with applicable
existing requirements. Specifically, the requested information will
enable the NRC staff to determine whether the applicable requirements
discussed above are being met. No backfit is either intended or
approved in the context of issuance of this GL. Therefore, the staff
has not performed a backfit analysis.
Federal Register Notification
To be done after the public comment period.
Small Business Regulatory Enforcement Fairness Act
The NRC has determined that this action is not subject to the Small
Business Regulatory Enforcement Fairness Act of 1996.
Paperwork Reduction Act Statement
This GL contains information collection requirements that are
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). These information collections were approved by the Office of
Management and Budget, approval no. 3150-0011, which expires on
February 28, 2007.
The burden of these mandatory information collections on the public
is estimated to average 200 hours per response, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
information collection. Send comments regarding this burden estimate or
any other aspect of these information collections, including
suggestions for reducing the burden, to the Records and FOIA/Privacy
Services Branch (T-5 F52), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or by Internet electronic mail to
[email protected]; and to the Desk Officer, Office of Information
and Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management
and Budget, Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
Sample Technical Specifications
Steam Generator (SG) Program
A SG Program shall be established and implemented to ensure that SG
tube integrity is maintained. The SG Program shall include the
following provisions:
a. Provisions for condition-monitoring assessments. A condition-
monitoring assessment is an evaluation of the ``as-found'' condition of
the tubing with respect to the performance criteria for structural
integrity and accident-induced leakage. The ``as-found'' condition
refers to the condition of the tubing during a SG inspection outage, as
determined from the inservice inspection results or by other means,
prior to the plugging of tubes. Condition-monitoring assessments shall
be conducted during each outage during which the SG tubes are inspected
or plugged to confirm that the performance criteria are being met.
b. Performance criteria for SG tube integrity. SG tube integrity
shall be maintained by meeting the performance criteria for tube
structural integrity, accident-induced leakage, and operational
LEAKAGE.
1. Structural integrity performance criterion: All inservice SG
tubes 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 specifications) and design-basis accidents. This includes
retaining a safety factor of 3.0 against burst under the normal steady
state full-power operation primary-to-secondary pressure differential
and a safety factor of 1.4 against burst applied to the design-basis
accident primary-to-secondary pressure differentials. Apart from the
above requirements, additional loading conditions associated with the
design-basis accidents, or combination of accidents in accordance with
the
[[Page 60196]]
design and licensing basis, shall also be evaluated to determine if the
associated loads contribute significantly to burst or collapse. In the
assessment of tube integrity, those loads that do significantly affect
burst or collapse shall be determined and assessed in combination with
the loads due to pressure with a safety factor of 1.2 on the combined
primary loads and 1.0 on axial secondary loads.
2. Accident-induced leakage performance criterion: The primary-to-
secondary accident-induced leakage rate for any design-basis accident,
other than a SG tube rupture, shall not exceed the rates assumed in the
accident analysis for total leakage rate from all SGs and leakage rate
from an individual SG. Accident-induced leakage is not to exceed
[licensee to insert value] gallons per day through each SG and
[licensee to insert value] gallons per day through all SGs.
3. The operational LEAKAGE performance criterion is specified in
limiting condition for operation (LCO) [licensee to insert reference to
appropriate LCO. For limits currently greater than 150 gallons per day,
the LCO limit should be lowered to a value less than or equal to 150
gallons per day.]
c. SG tube repair criteria. Tubes found by inservice inspection to
contain flaws with a depth equal to or exceeding 40 percent of the
nominal tube wall thickness shall be plugged.
d. Provisions for SG tube inspections. Periodic SG tube inspections
shall be performed. The number and portions of the tubes inspected and
the method of inspection shall be performed with the objective of
detecting flaws of any type (for example, volumetric flaws, axial and
circumferential cracks) that may be present along the length of the
tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-
tubesheet weld at the tube outlet, and that may satisfy the applicable
tube repair criteria. The tube-to-tubesheet weld is not part of the
tube. In addition to meeting requirements d.1, d.2, and d.3 below, the
inspection scope, inspection methods, and inspection intervals shall be
such as to ensure that SG tube integrity is maintained until the next
SG inspection. An assessment of degradation shall be performed to
determine the type and location of flaws to which the tubes may be
susceptible and, based on this assessment, to determine which
inspection methods need to be employed and at what locations.
1. Inspect 100 percent of the tubes in each SG during the first
refueling outage following SG replacement.
2. Inspect 100 percent of the tubes at sequential periods of [for
licensees with thermally treated Alloy 690 tubes, insert ``144, 108,
72, and thereafter 60 effective full-power months;'' for licensees with
thermally treated Alloy 600 tubes, insert ``120, 90, and thereafter 60
effective full-power months;'' for licensees with mill-annealed Alloy
600 tubes, insert ``60 effective full-power months;'']. The first
sequential period shall be considered to begin after the first
inservice inspection of the SGs. In addition, inspect 50 percent of the
tubes by the refueling outage nearest the midpoint of the period and
the remaining 50 percent by the refueling outage nearest the end of the
period. No SG shall operate for more than [for licensees with thermally
treated Alloy 690 tubes, insert ``72 effective full-power months or
three refueling outages;'' for licensees with thermally treated Alloy
600 tubes, insert ``48 effective full-power months or two refueling
outages;'' for licensees with mill-annealed Alloy 600 tubes, insert
``24 effective full-power months or each refueling outage'' (whichever
is less)] without being inspected.
3. If crack indications are found in any SG tube, then the next
inspection for each SG for the degradation mechanism that caused the
crack indication shall not exceed 24 effective full-power months or one
refueling outage (whichever is less). If definitive information, such
as from examination of a pulled tube, diagnostic nondestructive
testing, or an engineering evaluation indicates that a cracklike
indication is not associated with a crack or cracks, then the
indication need not be treated as a crack.
e. Provisions for monitoring operational primary-to-secondary
leakage.
Steam Generator (SG) Tube Inspection Report
A report shall be submitted within 180 days of the initial entry
into MODE 4 following completion of the inspection. The report shall
include:
a. The scope of inspection performed on each SG.
b. Active degradation mechanisms found.
c. Nondestructive examination techniques utilized for each
degradation mechanism.
d. Location, orientation (if linear), and measured sizes (if
available) of service-induced indications.
e. Number of tubes plugged during the inspection outage for each
active degradation mechanism.
f. Total number and percentage of tubes plugged to date.
g. The results of condition monitoring, including the results of
tube pulls and in-situ testing.
End
Documents may be examined, and/or copied for a fee, at the NRC's
Public Document Room at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland. Publicly available records will be
accessible electronically from the Agencywide Documents Access and
Management System (ADAMS) Public Electronic Reading Room on the
Internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html.
If you do not have access to ADAMS or if you have problems in accessing
the documents in ADAMS, contact the NRC Public Document Room (PDR)
reference staff at 1-800-397-4209 or 301-415-4737 or by e-mail to
[email protected].
Dated in Rockville, Maryland, this 30 day of September, 2004.
For the Nuclear Regulatory Commission.
Francis M. Costello,
Acting Branch Chief, Reactor Operations Branch, Division of Inspection
Program Management, Office of Nuclear Reactor Regulation.
[FR Doc. 04-22546 Filed 10-6-04; 8:45 am]
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