[Federal Register Volume 59, Number 155 (Friday, August 12, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-19721]
[[Page Unknown]]
[Federal Register: August 12, 1994]
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NUCLEAR REGULATORY COMMISSION
Proposed Generic Communication; ``Voltage-Based Repair Criteria
For The Repair Of Westinghouse Steam Generator Tubes Affected By
Outside Diameter Stress Corrosion Cracking''
AGENCY: U.S. 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. A generic letter is an NRC document that: (1)
Requests licensees to submit analyses or descriptions of proposed
corrective actions, or both, regarding matters of safety, safeguards,
or environmental significance, or (2) requests licensees to submit
information to the NRC on other technical or administrative matters, or
(3) transmits information to licensees regarding approved changes to
rules or regulations, the issuance of reports or evaluations of
interest to the industry, or changes to NRC administrative procedures.
When issued, this generic letter will provide guidance for licensees
who may wish to request a license amendment to the plant technical
specifications to implement an alternate steam generator tube repair
limit applicable specifically to outside diameter stress corrosion
cracking at the tube-to-tube support plate intersections in
Westinghouse designed steam generators having drilled-hole tube support
plates. This generic letter is intended to provide relief while
maintaining an acceptable level of safety for licensees having steam
generators experiencing this particular degradation mechanism while the
NRC pursues a longer term resolution to the issue of steam generator
degradation through the development of a steam generator rule. 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. Additionally, the NRC is seeking
public comments on the following question which pertains to the
technical positions described in the proposed generic letter. The
voltage-based repair methodology and calculational approach
incorporates numerous conservatism throughout the calculation in part
to bound uncertainties that currently exist in the methodology. The NRC
is soliciting public comment on the propagation of uncertainties
through the leakage rate and radiological dose calculations under
postulated accident conditions and the appropriateness of the
conservatism that have been included in the analyses to account for
these uncertainties. Two examples of uncertainties in the voltage-based
repair methodology and calculational approach are: (1) Several
functional forms (in addition to the log-logistic curve used in the
proposed generic letter) can be fit to the available probability of
leakage data equally well from the standpoint of a statistical goodness
of fit, and (2) there is a paucity of definitive data describing iodine
releases into the reactor coolant system following a large
depressurization transient such as the postulated main steam line
break.
During development of the proposed generic letter, three individual
NRC staff members expressed technical concerns (including one member
filing a differing professional opinion) with the NRC positions
described in the generic letter, in response to an internal memorandum
requesting such comments. The differing professional opinion is
currently being processed in accordance with the established NRC
procedures. The NRC policies on differing professional opinions (DPOs)
or differing professional views (DPVs) were established to ensure
employees have the ability to freely express their DPOs or DPVs and to
underscore management's intention to address these concerns in a timely
and responsible manner. The NRC has decided to make these technical
concerns and the differing professional opinion publicly available as
part of the information available in the Public Document Rooms, and to
provide the public an opportunity to comment on these concerns as they
may relate to the draft generic letter. The NRC held internal technical
interactions with the three individuals regarding their technical
concerns and gave the concerns careful consideration (the concerns did
not necessarily result in revisions to the proposed generic letter nor
were they necessarily resolved to the individuals' satisfaction) during
development and review of the proposed generic letter. The technical
concerns and differing professional opinion are briefly summarized as
follows: (1) The first concern is that use of the eddy current voltage
repair criteria could result in leakage rates following a postulated
main steam line break that could ultimately exceed the make-up capacity
of the refueling water storage tank for the emergency core cooling
system supply and result in core damage. The concern stems from the
belief that there is no direct relation between leakage and measured
eddy current voltage. (2) The second concern is that there is no
physical basis for choosing a given probability of leakage (POL)
function versus other POL functions, when all functions fit the
available data from a statistical standpoint. (3) The third concern
stems from the application of a voltage-based repair criterion when
there is not a unique correlation between voltage amplitude and
physical parameters (i.e., length or depth) of a defect that can be
directly related to tube structural integrity or leakage. The DPO is
similar to technical concern: (1) With additional concerns raised
regarding the paucity of iodine spiking data for the calculation of
offsite doses for a postulated main steam line break with induced steam
generator tube leakage and the effectiveness of reducing reactor
coolant system iodine activity for reducing calculated offsite doses.
The summaries above are not intended to oversimplify the expressed
technical concerns or differing professional opinion. To fully
understand the concerns, it is recommended that each concern be read in
its entirety. The proposed generic letter and supporting documentation
were discussed in the 260th meeting of the Committee to Review Generic
Requirements (CRGR). At this meeting, the three individual NRC staff
members presented their technical concerns to the CRGR. The relevant
information used to support CRGR review of the proposed generic letter
will be available in the Public Document Rooms. In addition, the
proposed generic letter and supporting documentation were discussed in
a meeting of the Materials and Metallurgy Subcommittee of the NRC
Advisory Committee on Reactor Safeguards (ACRS) on August 3, 1994, as
well as a full ACRS meeting held on August 4, 1994.
The NRC will consider comments received from interested parties in
the final evaluation of the proposed generic letter. The NRC final
evaluation will include a review of the technical position and, when
appropriate, an analysis of the value/impact on licensees. Should this
generic letter be issued in final form by the NRC, it will become
available for public inspection in the Public Document Rooms.
DATES: Comment period expires September 12, 1994. 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, Washington, D.C.
20555. Written comments may also be delivered to Room T6-D59, 11545
Rockville Pike, Rockville, Maryland, 20852 from 7:30 a.m. to 4:15 p.m.,
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: Timothy A. Reed, (301) 504-1465.
SUPPLEMENTARY INFORMATION
NRC Generic Letter 94-XX: Voltage-Based Repair Criteria For The Repair
of Westinghouse Steam Generator Tubes Affected by Outside Diameter
Stress Corrosion Cracking
Addressees
All holders of operating licenses or construction permits for
nuclear power reactors having steam generators designed by Westinghouse
Electric Corporation (W).
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this
guidance for licensees who may wish to request a license amendment to
the plant technical specifications to implement an alternate steam
generator tube repair limit applicable specifically to outside diameter
stress corrosion cracking (ODSCC) at the tube-to-tube support plant
intersections in Westinghouse designed steam generators having drilled-
hole tube support plates (TSP). The NRC has previously allowed a few
licensees to implement alternate steam generator repair criteria for
this particular degradation mechanism on an operating cycle specific
basis. This generic letter does not restrict the approval of such
repair criteria to a cycle specific basis.
Current plant technical specifications require that flawed tubes be
removed from service by plugging or repaired by sleeving, if the depths
of the flaws exceed the repair limit, typically 40 percent through-
wall. This generic letter provides guidance on the implementation of an
alternate repair criterion to be applied to ODSCC at TSP locations.
This criterion does not set limits on the depth of the cracks to ensure
tube integrity margins; instead, it relies on correlating the eddy
current voltage amplitude from a bobbin coil probe with the more
specific measurement of burst pressure and leak rate.
This generic letter is intended to provide relief while maintaining
an acceptable level of safety for licensees having steam generators
experiencing this particular degradation mechanism while the staff
pursues a longer term resolution to the issue of steam generator
degradation through the development of a steam generator rule. Although
this generic letter allows licensees to pursue various options
regarding the implementation of the voltage-based criteria (e.g., tube
support place deflection analyses, probability of detection versus
voltage dependence), licensees should recognize that pursuing such
options could have significant scheduler implications since the NRC
staff would be required to review and approve the associated
information and analyses. Regarding the correlations and supporting
data utilized to implement the generic letter guidance, the staff will
review this information on an as-required basis to enable updated
correlations and new data to be used for implementation of the generic
letter guidance. The NRC staff will make publicly available an updated
list of approved correlations and models on a periodic basis.
Background
The thin-walled tubing of the steam generator constitutes more than
half of the reactor coolant pressure boundary (RCPB). Maintenance of
the structural and leakage integrity of the RCPB is a requirement under
Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50),
Appendix A. Specific requirements governing the maintenance of steam
generator tube integrity are contained in the plant technical
specifications and Section XI of the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code). These
include requirements for periodic inservice inspection of the tubing,
flaw acceptance criteria (i.e., repair limits for plugging or
sleeving), and primary-to-secondary leakage limits. These requirements
coupled with the broad scope of plant operational and maintenance
programs, have formed the basis for assuring adequate steam generator
tube integrity.
Flaw acceptance criteria, termed plugging/repair limits, are
specified in the plant technical specifications. The purpose of the
technical specification repair limits is to ensure that tubes accepted
for continued service will retain adequate structural and leakage
integrity during normal operating, transient, and postulated accident
conditions, consistent with General Design Criteria (GDC) 14, 15, 30,
31, and 32 of 10 CFR part 50, appendix A. Structural integrity refers
to maintaining adequate margins against gross failure, rupture, and
collapse of the steam generator tubing. Leakage integrity refers to
limiting primary-to-secondary leakage to within acceptable limits.
The traditional strategy for accomplishing the objectives of the
General Design Criteria related to steam generator tube integrity has
been to establish a minimum wall thickness requirement in accordance
with the structural criteria of Regulatory Guide 1.121, ``Bases for
Plugging Degraded PWR Steam Generator Tubes.'' Development of minimum
wall thickness requirements to satisfy Regulatory Guide 1.121 was
governed by analyses for uniform thinning of the tube wall in the axial
and circumferential directions. The assumption of uniform thinning
conservatively bonds the degrading effects of all flaw types occurring
in the field, and is the basis of the standard 40 percent depth-based
plugging limit incorporated into the technical specifications. However,
the 40 percent repair limit is conservative for highly localized flaws
such as pits and short cracks. In particular, the 40 percent depth-
based repair limit is conservative for outside diameter stress
corrosion cracking (ODSCC) that occurs at the tube support plates.
The new voltage-based repair limit does not incorporate a minimum
wall thickness requirement. The voltage-based repair limit allows the
possibility that tubes with up to 100 percent through-wall cracks,
which may develop between successive steam generator inspections, can
remain in service, subject to certain restrictions. These restrictions
ensure structural integrity and leakage limits consistent with the
applicable GDC of 10 CFR part 50 appendix A and the limits of 10 CFR
Part 100. Although the voltage-based repair limit ensures adequate
structural integrity and leakage limits, the NRC staff recognizes that
overall margins have been reduced when compared to the margins
associated with the existing 40% depth-based repair limit.\1\ Because
of the increased likelihood of through-wall cracks developing in
service, the staff has included provisions for augmented steam
generator inspections and more restrictive operational tube leakage
limits in the generic letter guidance.
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\1\During development of the proposed generic letter, three
individual NRC staff members expressed technical concerns (including
one member filing a differing professional opinion) with the NRC
positions described in the generic letter. The technical concerns
and differing professional opinion are publicly available in the
Public Document Rooms. The NRC will consider public comments from
interested parties on the technical concerns as they relate to the
positions proposed in the draft generic letter.
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In taking the interim action described in this letter, the NRC
staff wishes to emphasize that, while use of the specific voltage-based
repair methods described herein is approved as an acceptable short-term
measure for dealing with ODSCC tube degradation, this action should not
be construed to discourage the use by licensees of better or further
refined data acquisition techniques, eddy current technology, and eddy
current data analysis as they become available; and the staff strongly
encourages the industry to continue its efforts to improve the
nondestructive examination of steam generator tubes. The staff
continues to believe that inspection methods and repair criteria based
on physical dimensions (e.g., length and depth) of defects are the most
desirable when they can be achieved.
Discussion
1. Overview of the Voltage Repair Limit Approach
In order to use the voltage repair criteria, licensees should
complete the following actions:
--Perform an enhanced inspection of tubes, particularly at the tube
support plate (TSP) intersections,
--Utilize nondestructive examination (NDE) data acquisition and
analysis procedures that are consistent with the methodology used to
develop the voltage-based repair limits,
--Repair or plug tubes that exceed the voltage limits,
--Determine the beginning of cycle voltage distribution,
--Project the end-of-cycle (EOC) distribution,
--For the projected EOC voltage distribution, calculate leakage and
conditional burst probability (and repair additional tubes if
necessary).
2. Generic Letter Applicability
The criteria in this generic letter are only applicable to ODSCC
located at the tube-to-tube support plate intersections in Westinghouse
designed steam generators. These criteria are not applicable to other
forms of steam generator tube degradation, nor are they applicable to
ODSCC that occurs at other locations within a steam generator. The
voltage-based repair criteria can be applied only under the following
constraints:
(1) The repair criteria of this generic letter apply only to
Westinghouse designed steam generators with 1.9 cm [\3/4\ inch] and 2.2
cm [\7/8\ inch] diameter tubes and drilled hole tube support plates,
(2) The repair criteria of this generic letter apply only to
predominantly axially oriented ODSCC confined within the tube-to-tube
support plate intersection (refer to Section 1.a of Enclosure 1 for
further guidance) and,
(3) Certain intersections are excluded from the application of the
voltage-based repair criteria as discussed in Section 1.b of Enclosure
1.
3. Voltage Repair Limit
The voltage repair limits are:
(1) for 2.2 cm [\7/8\ inch] diameter tubes:
Indications below 2.0 volts as measured by bobbin coil may
remain in service;
Indications between 2.0 and 5.6 volts as measured by
bobbin coil can remain in service if motorized rotating pancake coil
(MRPC) inspections do not confirm the indications; and
Indications between 2.0 and 5.6 volts as measured by
bobbin coil that are confirmed by MRPC and indications exceeding 5.6
volts as measured by bobbin coil must be repaired.
(2) For 1.9 cm [\3/4\-inch] diameter tubes.
Indications below 1.0 volt as measured by bobbin coil may
remain in service;
Indications between 1.0 and 2.7 volts as measured by
bobbin coil can remain in service if MRPC inspections do not confirm
the indications; and
Indications between 1.0 and 2.7 volts as measured by robin
coil that are confirmed by MRPC and indications exceeding 2.7 volts as
measured by bobbin coil must be repaired.
The voltage-based repair limits of this generic letter were
determined considering the entire range of design basis events that
could challenge tube integrity. The voltage repair limits ensure
structural integrity and leakage limits for all postulated design basis
events. The structural criteria are intended to ensure that tubes
subjected to the voltage repair limits will be able to withstand a
pressure of 1.4 times a maximum possible main steam line break (MSLB)
differential pressure postulated to occur at the end of the operating
cycle consistent with the criteria of Regulatory Guide 1.121. The
leakage criteria ensure that for tubes subject to the voltage repair
limits, induced leakage under worst-case MSLB conditions calculated
using licensing basis assumptions, will not result in offsite dose
releases that exceed the applicable limits of 10 CFR Part 100.
Requested Actions
Implementation of the guidance in this generic letter is voluntary.
If a licensee chooses to implement these criteria, the following should
be included in the proposed program:
(1) Implementation of the inspection guidance discussed in Section
3 of Enclosure 1. The inspection guidance ensures that the techniques
used to inspect steam generators are consistent with the techniques
used to develop voltage-based repair limit methodology.
(2) Calculation of leakage per the guidance of Section 2.b of
Enclosure 1. This calculation, in conjunction with the use of licensing
basis assumptions for calculating offsite releases, enables licensees
to demonstrate that the applicable limits of 10 CFR Part 100 continue
to be met.
(3) Calculation of conditional burst probability per the guidance
of Section 2.a of Enclosure 1. This is a calculation to assess the
voltage distribution left in service against a threshold value.
(4) Implementation of the operational leakage limits identified in
Section 5 of Enclosure 1. The operational leak limit is a defense-in-
depth measure that provides a means for identifying leaks during
operation to enable repair before such leaks result in tube failure.
(5) Review of leakage monitoring measures including the procedures
for timely detection, trending, and response to rapidly increasing
leaks. The objective is to ensure that should a significant leak be
experienced in service, it will be detected and the plant shut down in
a timely manner to reduce the likelihood of a potential rupture.
(6) Acquisition of tube pull data per the guidance of Section 4 of
Enclosure 1. It is necessary to acquire pulled tube data to confirm the
degradation mechanism.
(7) Reporting of results per the guidance of Section 6 of Enclosure
1.
(8) Submittal of a technical specification (TS) amendment request
that commits to the above and provides TS pages per the guidance of
Enclosure 2 including the associated consideration of no significant
hazards consideration (10 CFR 50.92) and supporting safety analysis.
Licensees that plan to adopt this TS amendment are encouraged to
follow the guidance given in Enclosures 1 and 2. The staff requests
that licensees following the guidance of this generic letter submit
their TX amendment request at least 90 days prior to the beginning of
the refueling outage during which the alternate repair criteria are to
be implemented.
Backfit Discussion
Licensee action to propose TS changes under the guidance of this
generic letter is voluntary; therefore, such action is not a backfit
under the provisions of 10 CFR 50.109.
Paperwork Reduction Act Statement
[To Be Provided in the Final Generic Letter]
Enclosures:
1. Guidance for a Proposed License Amendment to Implement an
Alternate Steam Generator Tube Repair Limit for Outside Diameter Stress
Corrosion Cracking at the Tube Support Plate Intersections
2. Model Technical Specifications
3. List of Recently Issued NRC Generic Letters
Guidance for a Proposed License Amendment To Implement an Alternate
Steam Generator Tube Repair Limit for Outside Diameter Stress Corrosion
Cracking at the Tube Support Plate Intersections
1. Introduction
This guidance provides the NRC staff position on the implementation
of the voltage-based repair criteria in steam generators designed by
Westinghouse for outside diameter stress corrosion cracking (ODSCC)
located at the tube-to-tube support plate intersections. This guidance
is not applicable to other forms of steam generator tube degradation
nor is it applicable to ODSCC that occurs at other locations with the
steam generator. The voltage-based repair criteria have been developed
for, and are currently applicable only to, Westinghouse-designed steam
generators with 2.2 cm [\7/8\-inch] or 1.9 cm [\3/4\-inch] diameter
tubes with drilled hole tube support plates (TSPs). Application of the
alternate repair criteria to other vendor designed steam generators
would require both the development and NRC staff review and approval of
a comparable data base and the associated correlations for each vendor
steam generator type.
The NRC staff wants to emphasize that while the NRC has approved
the implementation of the voltage-based repair methods described in
this generic letter as a short-term measure, this guidance should not
be construed to discourage the development and use of better
acquisition techniques, eddy current technology, and eddy current data
analysis. The staff strongly encourages the industry to continue to
improve the NDE of steam generator tubes.
1.a ODSCC
The voltage-based repair criteria are applicable only to
indications at support plate intersections where the degradation
mechanism is dominantly axial ODSCC with no significant cracks
extending outside the thickness of the support plate.
For purposes of this guidance, OSDCC refers to degradation whose
dominant morphology consists of axial stress corrosion cracks which
occur either singularly or in networks of multiple cracks, sometimes
with limited patches of general intergranular attack (IGA).
Circumferential cracks may sometimes occur in the IGA affected regions
resulting in a grid-like pattern of axial and circumferential cracks,
termed cellular corrosion. Cellular corrosion is assumed to be
relatively shallow (based on available data from tube specimens removed
from the field), transitioning to dominantly axial cracks as the
cracking progresses in depth. The circumferential cracks are assumed
(based on available data) not to be of sufficient size to produce a
discrete, crack-like circumferential indication during field
nondestructive examination (NDE) inspections. Thus, the failure mode of
ODSCC is axial and the burst pressure is controlled by the geometry of
the most limiting axial crack or array of axial cracks.
It is also assumed for purposes of this guidance that the ODSCC is
confined to within the thickness of the tube support plate, based on
available data from tube specimens removed from the field. Very shallow
microcracks are sometimes observed on these specimens to initiate at
locations slightly outside the thickness of the tube support plate;
however, these microcracks are small compared to the cracks within the
thickness of the support plate and are too small to produce an eddy
current response.
Confirmation that the degradation mechanism is dominantly axial
ODSCC should be accomplished by periodically removing tube specimens
from the steam generators and by examining and testing these specimens
as specified in Section 4 of this guidance. The acceptance criteria
should consist of demonstrating that the dominant degradation mechanism
affecting the burst and leakage properties of the tube is axially
oriented ODSCC. In addition, results of inservice inspections with
motorized rotating pancake coil (MRPC) probes should be evaluated in
accordance with Section 3.b of this guidance to confirm the absence of
detectable crack-like circumferential indications and detectable ODSCC
indications extending outside the tube support plate thickness.
1.b Exclusion of Intersections
The voltage repair criteria of this guidance do not apply to
intersections meeting the criteria discussed below:
1.b.1 The repair criteria do not apply to support plate intersections
where the tubes may potentially collapse or deform following a
postulated loss-of-coolant accident plus safe shutdown earthquake event
(e.g., intersections located near the wedge supports at the upper tube
support plates). Licensees should perform or reference an analysis that
identifies which intersections are to be excluded.
1.b.2 The repair criteria do not apply to tubes support plate
intersections having dent signals greater than 5 volts as measured with
the bobbin probe.
1.b.3 The repair criteria do not apply to intersections where there are
mixed residuals of sufficient magnitude to cause a 1-volt ODSCC
indication (as measured with a bobbin probe) to be missed or misread.
1.b.4 The repair criteria do not apply to the tube-to-flow distribution
baffle plate intersections.
2. Tube Integrity Evaluation
Licensees should perform an evaluation prior to plant restart to
confirm that the steam generator tubes will retain adequate structural
and leakage integrity until the next scheduled inspection. The first
portion of this evaluation, referred to as the conditional burst
probability calculation, assesses the voltage distribution left in
service against a threshold value of 1 x 10-2 probability of
rupture under postulated main steam line break (MSLB) conditions. The
conditional burst probability calculation is intended to provide a
conservative assessment of tube structural integrity during a
postulated MSLB occurring at end-of-cycle2 (EOC). It is used to
determine whether the NRC needs to focus additional attention on the
particular voltage repair limit application. If the calculated
conditional burst probability exceeds 1 x 10-2, the licensee
should notify the NRC per the guidance provided in Section 6 of this
guidance.
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\2\For the purposes of this guidance, ``cycle'' refers to the
operating cycle between two scheduled steam generator inspections.
Operating cycle and inspection cycle are used interchangeably.
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The second portion of the tube integrity evaluation is intended to
assure that total leak rate from the affected steam generator (SG)
during a postulated MSLB occurring at EOC would be less than that which
could lead to radiological releases in excess of the licensing basis
for the plant. If calculated leakage exceeds the allowable limit
determined by the licensing basis dose calculation, licensees can
either repair tubes, beginning with the largest voltage indications
until the leak limit is met, or reduce reactor coolant system specific
iodine activity (refer to example technical specification (TS) pages of
Enclosure 2). The analyses discussed above may incorporate or reference
previous analyses, or portions thereof, to the extent that they
continue to bound the conditions of the steam generator as determined
by inspection.
For plants in which the technical specifications do not require the
pressurizer power-operated relief valves (PORVs) to be operable during
power operation, these tube integrity analyses should be conducted for
an assumed differential pressure across the tube walls equal to the
pressurizer safety valve set point plus 3 percent for the valve
accumulation less atmospheric pressure in faulted steam generators. For
plants in which the technical specifications do require the PORVs to be
operable, the assumed differential pressure may be based on the PORV
set point in lieu of the safety valve set point with similar
adjustments.
2. a Conditional Probability of Burst During MSLB
For this generic letter, the conditional probability of burst
refers to the probability that the burst pressures associated with 1 or
more indications in the faulted steam generator will be less than the
maximum pressure differential associated with a postulated MSLB assumed
to accur at EOC. A methodology should be submitted for NRC review and
approval for calculating this conditional burst probability. After the
NRC approves a method for calculating conditional probability of burst,
licensees may reference the approved method. This methodology should
involve: (1) Determining the distribution of indications as a function
of their voltage response at beginning of cycle (BOC) as discussed in
Section 2.b.1, (2) projecting this BOC distribution to an EOC voltage
distribution based on consideration of voltage growth due to defect
progression between inspections as discussed in Section 2.b.2(2) and
voltage measurement uncertainty as discussed in Section 2.b.2(1), and
(3) evaluating the conditional probability of burst for the projected
EOC voltage distribution using the correlation between burst pressure
and voltage discussed in Section 2.a.1. The solution methodology should
account for uncertainties in voltage measurement (Section 2.b.2(1)),
the distribution of potential voltage growth rates applicable to each
indication (Section 2.b.2(2)), and the distribution of potential burst
pressures as a function of voltage (Section 2.a.1). Monte Carlo
simulations constitute an acceptable approach for accounting for these
various sources of uncertainty.
2.a.1 Burst Pressure Versus Bobbin Voltage
An empirical model, for \7/8\-inch or \3/4\-inch diameter tubing as
applicable, should be used to relate burst pressure to bobbin voltage
response for purposes of estimating the conditional probability of
burst during a postulated MSLB. This model should explicitly account
for burst pressure uncertainty as indicated by scatter of the
supporting test data and should also account for the parametric (i.e.,
slope and intercept) uncertainty of the regression fit of the data. The
supporting data for \7/8\-inch diameter and \3/4\-inch diameter tubing
should include all applicable data consistent with the industry
recommendations in letter dated April 22, 1994, to Jack Strosnider,
NRC, from David A. Steininger, EPRI, ``Exclusion of Data from Alternate
Repair Criteria (ARC) Databases Associated with \7/8\ inch Tubing
Exhibiting ODSCC'' (Reference 1) and letter dated June 9, 1994, to
Brian Sheron, NRC, from David J. Modeen, Nuclear Energy Institute
(Reference 2) respectively, with certain exceptions. Specifically, data
excluded under criteria 2a and 2b in References 1 and 2 should not be
excluded pending staff review and approval of these criteria.
2.b Total Leak Rate During MSLB
A methodology should be submitted for NRC review and approval for
calculating the total primary-to-secondary leak rate in the faulted
steam generator during a postulated MSLB assumed to occur at EOC. After
the NRC approves a leakage calculation methodology, licensees may
reference the approved method. This methodology involves: (1)
Determining the distribution of indications as a function of their
voltage response at beginning of cycle (BOC) as discussed in Section
2.b.1, (2) projecting this BOC distribution to an EOC voltage
distribution based on consideration of voltage growth due to defect
progression between inspection as discussed in Section 2.b.2(2) and
voltage measurement uncertainty as discussed in Section 2.b.2(1), and
(3) evaluating the total leak rate for the projected EOC voltage
distribution using a probability of leakage (POL) model as discussed in
Section 2.b.3(1) and conditional leak rate model as discussed in
Section 2.b.3(2). The solution methodology should account for
uncertainties in voltage measurement (Section 2.b.2(1)), the
distribution of potential voltage growth rates applicable to each
indication (Section 2.b.2(2)), the uncertainties in the probability of
leakage as a function of voltage (Section 2.b.3(1)), and the
distribution of potential conditional leak rates as a function of
voltage (Section 2.b.3(2)). Monte Carlo simulations are an acceptable
method for accounting for these sources of uncertainty provided that
the calculated total leak rate reflects an upper 95% quantile value.
[Note: draft NUREG-1477, Section 3.3, page 3-21, presents an expression
for Tl (i.e., working bound for total leak rate) which is based on
the premise that leak rate is independent of voltage. This expression
does not account for parametric uncertainty in either the POL or
conditional leak rate model. Thus, the draft NUREG-1477 equation should
not be used unless appropriate modifications are made to the equation
to account for these parametric uncertainties.]
2.b.1 Distribution of Bobbin Indications as a Function of Voltage at
BOC
The frequency distribution by voltage of bobbin indications
actually found during inspection should be scaled upward by a factor of
1/POD to account for non-detected cracks which can potentially leak or
rupture under postulated MSLB conditions during the next operating
cycle. POD stands for probability of detection of ODSCC flaws. This
adjusted frequency distribution minus detected indications for tubes
that have been plugged or repaired should constitute, for purposes of
the tube integrity analyses, the assumed frequency distribution of
bobbin indications at BOC as a function of voltage. This can also be
expressed as:
Nl=(1/POD)(Nd)--Nr
Nl=assumed frequency distribution of bobbin indications
Nd=frequency distribution of indications actually detected
Nr=frequency distribution of repaired indications
POD=probability of detection of ODSCC flaws
POD should be assumed to have a value of 0.6, or as an alternative,
and NRC approved POD function can be used if such a function becomes
available.
Nd includes all flaw indications detected by the bobbin coil,
regardless of whether these indications are confirmed by MRPC
inspection.
2.b.2 Projected End-of-Cycle (EOC) Voltage Distribution
As discussed above, the calculation of both conditional burst
probability and leakage (during a postulated MSLB) requires the
generation of the projected EOC voltage distribution. To project an EOC
voltage distribution from the BOC voltage distribution determined
above, requires consideration of: (1) Eddy current voltage measurement
uncertainty and (2) the addition of voltage growth to account for
defect progression. Monte Carlo techniques are an acceptable means for
sampling eddy current measurement uncertainty and the voltage growth
distribution to determine the projected EOC voltage distribution. Eddy
current measurement uncertainty and voltage growth are discussed below.
2.b.2(1) Eddy Current Voltage Measurement Uncertainty
Uncertainty in eddy current voltage measurements stems primarily
from two sources:
(a) Voltage response variability (i.e., test repeatability error)
which stems primarily from probe wear
(b) Voltage measurement variability among data analysts (i.e.,
measurement repeatability error)
Each of these uncertainties should be quantified. An acceptable
characterization of these uncertainties is contained in EPRI TR-100407,
Revision 1, Draft Report August 1993, ``PWR Steam Generator Tube Repair
Limits-Technical Support Document for Outside Diameter Stress Corrosion
Cracking at the tube Support Plates'' (Reference 3), Sections 2.4.1,
2.4.2, and D.4.2.3, with the exception that no distribution cutoff
should be applied to the voltage measurement variability distribution.
(However, the assumed 15 percent cutoff for the voltage response
variability distribution in Reference 3 is acceptable.)
2.b.2(2) Voltage Growth Due to Defect Progression
Potential voltage growth rates during the next inspection cycle
(i.e. operating cycle between two scheduled steam generator
inspections) should be based on voltage growth rates observed during
the last one or two inspection cycles. For a given inspection, previous
inspections results at tube support plate intersections currently
exhibiting a bobbin indication should be re-evaluated consistent with
the date analysis guidelines in Section 3 below. In cases where data
acquisition guidelines employed during previous inspection differ from
those discussed in Section 3, adjustments to the evaluation of the
previous data should be made to compensate for the difference. Voltage
growth rates should only be evaluated for those intersections where
bobbin indications can be identified at tow successive inspections.
The distribution of observed voltage growth rates (based on the
change in voltage on an intersection-to-intersection basis) should be
determined for each of the last one or two inspection cycles. When only
the current or only the current and previous inspections employed data
acquisition guidelines similar to those discussed in Section 3, only
the growth rate distribution for the previous cycle should be used to
estimate the voltage growth rate distribution for the next inspection
cycle. If both the two previous inspections employed such similar
guidelines, the most limiting of the two previous growth rate
distributions should be used to estimate the voltage growth rate
distribution for the next inspection cycle. However, the two
distributions should be combined if one or both the distributions is
based on a minimal number (i.e., <200) of indications.
It is acceptable to use a statistical model fit of the observed
growth rate distribution as part of the integrity analysis. It is also
acceptable that the voltage growth distribution be in terms of
volts rather than percent volts provided the
conservatism of this approach continues to be supported by operating
experience. Finally, negative growth rates should be included as zero
growth rates in the assumed growth distribution.
2.b.3 Calculation of Projected MSLB Leakage
Once the projected EOC voltage distribution is determined, the
leakage for the postulated MSLB is calculated utilizing the EOC voltage
distribution and the use of two models: (1) The probability of leakage
model and (2) the conditional leak rate model. As previously discussed
in Section 2.b, Monte Carlo techniques are an acceptable approach for
accounting for the uncertainties implicit in these models. These models
are discussed below.
2.b3(1) Probability of Leakage as a Function of Voltage
The Probability of leakage (POL) model should utilize the log-
logistic functional form. This model should explicitly account for
parameter uncertainty of the POL functional fit of the data (i.e.,
``model fit'' uncertainty). The supporting data sets for 2.2 cm (\7/8\-
inch) diameter and 1.9 cm (\3/4\-inch) diameter tubing should include
all applicable data consistent with the industry recommendations in
References 1 and 2, respectively, with certain exceptions. Namely, data
excluded under criteria 2a and 2b in References 1 and 2 should not be
excluded pending staff review and approval of these criteria.
2.b.3(2) Conditional Leakage Rate under MSLB Conditions
The conditional leak rate model should incorporate a linear
regression fit to the log of the leak rate data, for 2.2 cm (\7/8\-
inch) and 1.9 cm (\3/4\-inch) diameter tubing respectively, as a
function of the log of the bobbin voltage and should account for both
data scatter and parameter uncertainty of the linear regression fit.
Use of this approach is subject to demonstrating that the linear
regression fit is valid at the 5% level with a ``p-value'' test. If
this condition is not satisfied, the linear regression fit should be
assumed to have zero slope (i.e., the linear regression fit should be
assumed to be constant with voltage).
The supporting data sets for 2.2 cm (\7/8\-inch) diameter and 1.9
cm (\3/4\-inch) diameter tubing should include all applicable data
consistent with the industry recommendations in References 1 and 2,
respectively, with certain exceptions. Specifically, data excluded
under criteria 2a, 2b, 3a, 3b, and 3c in References 1 and 2 should not
be excluded pending staff review and approval of these criteria. In
addition, an MSLB leak rate of 2496 liters/hour should be utilized for
the data point obtained from V.C. Summer tube R28C41 pending staff
review and approval of the revised leakage estimate for this tube
described in Reference 2.
2.b.4 Calculation of Offsite and Control Room Doses
For the MSLB leak rate calculated above, offsite and control room
doses should be calculated utilizing currently accepted licensing basis
assumptions. Licensees should note that Enclosure 2 of this generic
letter provides example TS pages for reducing reactor coolant system
specific iodine activity limits. Reactor coolant system iodine
activities may be reduced to .35 microcuries per gram does equivalent
I-131. Licensees wishing to reduce iodine activities below this level
should provide a justification supporting the request that addresses
the release rate data described in Reference 6. Reduction of reactor
coolant iodine activity is an acceptable means for accepting higher
projected leakage rates and still meeting the applicable limits of
Title 10 of the Code of Federal Regulations Part 100 utilizing
licensing basis assumptions.
3. Inspection Criteria
The inspection scope, data acquisition, and data analysis should be
performed in a manner consistent with the methodology utilized to
develop the voltage limits (e.g., the methodology described in
Reference 4, Appendix A, and Reference 5, Appendix A) with the
exceptions and clarifications noted below.
3.a Bobbin Coil Inspection Scope and Sampling
3.a.1 The bobbin coil inspection should include 100 percent of the
hot-leg TSP intersections and cold-leg intersections down to the lowest
cold-leg TSP with known ODSCC. The determination of TSPs having ODSCC
should be based on the performance of at least a 20 percent random
sampling of tubes inspected over their full length.
3.b Motorized Rotating Pancake Coil (MRPC) Inspection
MRPC\3\ inspections should be conducted as given below for purposes
of obtaining additional characterization of ODSCC flaws found with the
bobbin probe and to inspect intersections with significant bobbin
interference signals (due to copper, dents, large mix residuals) which
may impair the detectability of ODSCC with the bobbin probe or which
may unduly influence the bobbin voltage measurement. With respect to
ODSCC flaw characterization, a key purpose of the MRPC inspections is
to ensure the absence of detectable crack-like circumferential
indications and detectable indications extending outside the thickness
of the tube support plate. The voltage-based repair criteria are not
applicable to intersections exhibiting such indications, and special
reporting requirements pertaining to the finding of such indications
are described in Section 6.
---------------------------------------------------------------------------
\3\For the purposes of this guidance, MRPC also includes the use
of comparable or improved nondestructive examination techniques.
---------------------------------------------------------------------------
3.b.1 MRPC inspection should be performed for all indications
exceeding 1.5 volts as measured by bobbin coil for 2.2 cm [\7/8\-inch]
diameter tubes or 1.0 volt as measured by bobbin coil for 1.9 cm [\7/
8\-inch] diameter tubes.
3.b.2 The voltage-based criteria of this guidance are not
applicable to intersections with copper deposits, dent signals greater
than 5 volts, and large mixed residuals.
3.b.3 All intersections with bobbin coil signals indicative of
copper deposits should be inspected with MRPC. Any indications found at
such intersections with MRPC should cause the tube to be repaired.
3.b.4 All intersections with dent signals greater than 5 volts
should be inspected with MRPC. Any indications found at such
intersections with MRPC should cause the tube to be repaired.
3.b.5 All intersections with large mixed residuals should be
inspected with MRPC. For purposes of this guidance, large mixed
residuals are those that could cause a 1-volt bobbin signal to be
missed or misread. Any indications found at such intersections with
MRPC should cause the tube to be repaired.
3.b.6 A minimum sample of 100 intersections should be inspected
with MRPC to meet the criteria of this part.
3.c Data Acquisition and Analysis
3.c.1 The bobbin coil calibration standard should be calibrated
against the reference standard used in the laboratory as part of the
development of the voltage-based approach by direct testing or through
use of a transfer standard.
3.c.2 Bobbin coil probes should be calibrated based on four 100
percent through-wall holes.
3.c.3 Once the probe has been calibrated on the 100 percent
through-wall hole, the voltage response of new bobbin coil probes for
the 20 percent to 80 percent American Society of Mechanical Engineers
(ASME) through-wall holes should not differ from the nominal voltage by
more than +/-10 percent.
3.c.4 Probe wear should be controlled by either an inline
measurement device or through the use of a periodic wear measurement.
When utilizing the periodic wear measurement approach, if a probe is
found to be out-of-specification, all tubes inspected since the last
successful calibration should be reinspected with the new calibrated
probe.
3.c.5 Data analysts should be trained and qualified in the use of
the analyst's guidelines and procedures. Data analyst performance
should be consistent with the assumptions for analyst measurement
variability (Section 2.b.2(1)) utilized in the tube integrity
evaluation (Section 2).
3.c.6 Quantitative noise criteria (resulting from electrical
noise, tube noise, calibration standard noise) should be included in
the data analysis procedures. Data failing to meet these criteria
should be rejected, and the tube reinspected.
3.c.7 Data analysts should review the mixed residuals on the
standard itself and take action as necessary to minimize these
residuals.
3.c.8 Smaller diameter probes can be used to inspect tubes where
it is impractical to utilize a full-sized probe provided that the
probes and procedures have been demonstrated on a statistically
significant basis to give an equivalent voltage response and detection
capability when compared to the full size probe. This demonstration can
be done on a plant-specific or generic basis.
4. Tube Removal and Examination/Testing
Implementation of voltage-based plugging criteria should include a
program of tube removals for testing and examination as described
below. The purpose of this program is to confirm axial ODSCC as the
dominant degradation mechanism as discussed in Section 1.a and to
provide additional data to enhance the burst pressure, probability of
leakage, and conditional leak rate correlations described in Sections
2.a.1, 2.b.3(1), and 2.b.3(2), respectively.
4.a Number and Frequency of Tube Pulls
Pulled tube specimens for at least six tube support plate
intersections should be obtained for each plant either during the plant
steam generator inspection outage that implements the voltage repair
limits or during the inspection outage preceding initial application of
voltage-based repair criteria. Additional pulled tube specimens should
be obtained periodically after the initial application of voltage-based
plugging criteria on a frequency of six tube intersections every two
steam generator inspections outages. In some cases, it may be necessary
for the staff to request plant specific tube pulls due to special
circumstances involved with a particular plant specific application of
the voltage-based repair limits.
Alternatively, the request to acquire pulled tube specimens may be
met by participating in an industry sponsored tube pull program
endorsed by the NRC that meets the objectives of this guidance. Such a
program would have to satisfy the following objectives: (1) To confirm
the degradation mechanism for plants utilizing the generic letter for
the first time, (2) to continue monitoring the ODSCC mechanism over
time, and (3) to enhance the burst pressure, probability of leakage,
and conditional leak rate correlations. [Note; the industry has
proposed such a program in letter dated May 10, 1994, to Brian Sheron,
NRC, from David J. Modeen, Nuclear Energy Institute (Reference 5),
which is currently under NRC staff review.]
4.b Selection Criteria
Selection of tube pulls should consider the following criteria:
4.b.1 There should be an emphasis on removing tube intersections
with large voltage indications.
4.b.2 Where possible the removed tube intersections should cover a
range of voltages, including intersections with no detectable
degradation.
4.b.3 As a minimum, selected intersections should be such as to
ensure that the total data set includes at least a representative
number of intersections with MRPC signatures indicative of a single
dominant crack as compared to intersections with MRPC signatures
indicative of two or more dominant cracks about the circumference.
4.c Examination and Testing
Removed tube intersections should be subjected to leak and burst
tests under simulated MSLB conditions to confirm that the failure mode
and leakage rates are consistent with that assumed in development of
the voltage-based criteria. In addition, these data may be used to
enhance the supporting data sets for the burst pressure and leakage
correlations subject to NRC review and approval as stated in 4.d,
below. Subsequent to burst testing, the intersections should be
destructively examined to confirm that the degradation morphology is
consistent with the assumed morphology for ODSCC.
4.d General Criteria for Burst and Leakage Models and Supporting Test
Data
This guidance allows only the use of NRC approved burst and leakage
models and correlations; this includes NRC approval of the data that
supports the models and correlations.
5. Operational Leakage Limits
5.a The operational leakage limit should be reduced to 150 gallons
per day (gpd) through each steam generator.
5.b Licensees should review their leakage monitoring measures to
ensure that should a significant leak be experienced in service, it
will be detected and the plant shut down in a timely manner to reduce
the likelihood of a potential rupture. Specifically, the effectiveness
of these procedures for ensuring the timely detection, trending, and
response to rapidly increasing leaks should be assessed. This should
include consideration of the appropriateness of alarm set points on the
primary-to-secondary leakage detection instrumentation and the various
criteria for operator actions in response to detected leakage.
5.c Steam generator tubes with known leaks should be repaired
prior to returning the steam generators to service following a steam
generator inspection outage.
6. Reporting Requirements
6.a Threshold Criteria for Requiring Prior Staff Approval To Continue
With Voltage-Based Criteria
This guidance allows licensees to implement the voltage-based
repair criteria on a continuing basis after the NRC staff has approved
the initial TS amendment. However, there are several situations for
which the NRC staff must receive prior notification before a licensee
can continue with the implemtation of the voltage-based repair
criteria:
6.a.1 If the actual measured voltage distribution would have
resulted in an estimated leakage during the previous operating cycle
greater than the leakage limit (determined from the licensing basis
calculation), then the licensee should notify the NRC of this
occurrence and provide an assessment of its significance prior to
returning the steam generators to service.
6.a.2 If (1) indications are identified that extend beyond the
confines of the TSP or (2) indications are identified that appear to be
circumferential in nature, then the NRC staff should be notified prior
to returning the steam generators to service.
6.a.3 If the calculated conditional probability of burst based on
the projected EOC voltage distribution exceeds 1X10-2, licensees
should notify NRC and provide an assessment of the significance of this
occurrence prior to returning the steam generators to service. This
assessment should address the safety significance of the calculated
conditional probability.
6.b Information To Be Provided Following Each Restart
The following information should be submitted to the NRC staff
within 90 days of each restart following a steam generator inspection:
(a) The results of metallurgical examinations performed for tube
intersections removed from the steam generator.
(b) The following distributions should be provided in both tabular
and graphical form. This information is to enable the staff to assess
the effectiveness of the methodology, determine whether the degradation
is changing significantly, determine whether the data supports higher
voltage repair limits, and to perform confirmatory calculations:
(i) EOC voltage distribution--all indications found during the
inspection regardless of MRPC confirmation
(ii) Cycle voltage growth rate distribution (i.e., from BOC to EOC)
(iii) Voltage distribution for EOC repaired indications--
distribution of indications presented in (i) above that were repaired
(i.e., plugged or sleeved)
(iv) Voltage distribution for indications left in service at the
beginning of the next operating cycle regardless of MRPC confirmation--
obtained from (i) and (iii) above
(v) Voltage distribution for indications left in service at the
beginning of the next operating cycle that were confirmed by MRPC to be
crack-like or not MRPC inspected
(vi) Non-destructive examination uncertainty distribution used in
predicting the EOC (for the next cycle of operation) voltage
distribution
(c) The results of the tube integrity evaluation described in
Section 2. Note that these calculations must be completed prior to
restart to ensure that an adequate number of tubes have been repaired
to meet the leakage limit and ensure continued tube integrity.
7. References
1. Letter dated April 22, 1994, to Jack Strosnider, NRC, from David
A. Steininger, EPRI, ``Exclusion of Data from Alternate Repair Criteria
(ARC) Databases Associated with \7/8\ inch Tubing Exhibiting ODSCC''.
2. Letter dated June 9, 1994, to Brian Sheron, NRC, from David J.
Modeen, Nuclear Energy Institute.
3. EPRI TR-100407, Revision 1, Draft Report August 1993, ``PWR
Steam Generator Tube Repair Limits-Technical Support Document for
Outside Diameter Stress Corrosion Cracking at the Tube Support
Plates''.
4. WCAP-12985, Revision 1. ``Kewaunee Steam Generator Tube Plugging
Criteria for ODSCC at Tube Support Plates,'' Westinghouse Electric
Corporation, January 1993, Westinghouse Proprietary Class 2.
5. WCAP-13522, ``V.C. Summer Steam Generator Tube Plugging Criteria
for Indications at Tube Support Plates,'' Westinghouse Electric
Corporation, Westinghouse Proprietary Class 2.
6. J.P. Adams and C.L. Atwood, ``The Iodine Spike Release Rate
During a Steam Generator Tube Rupture,'' Vol. 94, pg. 361, (1991).
Model Technical Specifications
The model technical specifications are based on the ``Standard
Technical Specifications (STS) for Westinghouse Pressurized Water
Reactors,'' NUREG-0452, Revision 4a. The indicated changes are
identified in italics. Note that the model technical specification
changes described below also include an example change to reduce
reactor coolant system specific activity. The model technical
specifications identified below should be adopted consistent with the
licensing basis. It should be noted that in the improved STS, some of
these surveillance requirements have been relocated to the
Administrative Controls section.
3/4.4.5 Reactor Coolant System
4/4.5.2 Steam Generator Tube Selection and Inspection
[add the following paragraphs]
b.4. Tubes left in service as a result of application of the tube
support plate plugging criteria shall be inspected by bobbin coil probe
during all future refueling outages.
d. Implementation of the steam generator tube/tube support plate
plugging criteria requires a 100 percent bobbin coil inspection for
hot-leg tube support plate intersections and cold-leg intersections
down to the lowest cold-leg tube support plate with known outside
diameter stress corrosion cracking (ODSCC) indications. The
determination of tube support plate intersections having ODSCC
indications shall be based on the performance of at least a 20 percent
random sampling of tubes inspected over their full length.
4.4.5.4 Acceptance Criteria
a. As used in this specification:
6. Plugging Limit\4\ means the imperfection depth at or beyond
which the tube shall be removed from service and is equal to 40 percent
of the nominal wall thickness. This definition does not apply to tube
support plate intersections for which the voltage-based plugging
criteria are being applied. Refer to 4.4.5.4.a.10 for the plugging
limit applicable to these intersections.
---------------------------------------------------------------------------
\4\For plants that have approved sleeving, ``plugging'' can be
replaced with ``repair'' to allow tubes to be either plugged or
sleeved when indications exceed applicable repair limits.
---------------------------------------------------------------------------
10. Tube Support Plate Plugging Limit is used for the disposition
of a steam generator tube for continued service that is experiencing
outside diameter stress corrosion cracking confined within the
thickness of the tube support plates. At tube support plate
intersections, the repair limit is based on maintaining steam generator
tube serviceability as described below:
a. Degradation attributed to outside diameter stress corrosion
cracking within the bounds of the tube support plate with bobbin
voltage less than or equal to [Note 1] will be allowed to remain in
service.
b. Degradation attributed to outside diameter stress corrosion
cracking within the bounds of the tube support plate with a bobbin
voltage greater than [Note 1] will be repaired or plugged except as
noted in 4.4.5.4.a.10.c below.
c. Indications of potential degradation attributed to outside
diameter stress corrosion cracking within the bounds of the tube
support plate with a bobbin voltage greater than [Note 1] but less than
or equal to [Note 2] may remain in service if a rotating pancake coil
inspection does not detect degradation. Indications of outside diameter
stress corrosion cracking degradation with a bobbin voltage greater
than [Note 2] volts will be plugged or repaired.
d. [If applicable] Certain intersections as identified in
[reference report] will be excluded from application of the voltage-
based repair criteria as it is determined that these intersections may
collapse or deform following a postulated LOCA + SSE event.
e. If a result of leakage due to a mechanism other than ODSCC at
the tube support plate intersection, or some other cause, an
unscheduled mid-cycle inspection is performed, the following repair
criteria apply instead of 4.4.5.4.10.c. If bobbin voltage is within
expected limits, the indication can remain in service. The expected
bobbin voltage limits are determined from the following equation:
TN12AU94.000
where:
V=measured voltage
VBOC=voltage at BOC
t=time period of operation to unscheduled outage
CL=cycle length (full operating cycle length where operating cycle is
the time between two scheduled steam generator inspections)
VSL=4.5 volts for \3/4\-inch tubes and 9.6 volts for \7/8\-inch
tubes
Note 1.--1.0 volt for \3/4\-inch diameter tubes or 2.0 volts
for \7/8\-inch diameter tubes.
Note 2.--2.7 volts for \3/4\-inch diameter tubes or 5.6 volts
for \7/8\-inch diameter tubes.
4.4.5.5 Reports
d. For implementation of the voltage-based repair criteria to tube
support plate intersections, notify the staff prior to returning the
steam generators to service should any of the following conditions
arise:
1. If estimated leakage based on the actual measured end-of-cycle
voltage distribution would have exceeded the leak limit (for the
postulated main steam line break utilizing licensing basis assumptions)
during the previous operation cycle.
2. If circumferential crack-like indications are detected at the
tube support plate intersections.
3. If indications are identified that extend beyond the confines of
the tube support plate.
4. If the calculated conditional burst probability exceeds
1 x 10-2, notify the NRC and provide an assessment of the safety
significance of the occurrence.
Reactor Coolant system
3/4.4.6 Reactor Coolant System Leakage
3.4.6.2 Reactor Coolant System leakage shall be limited to:
a. No Pressure boundary Leakage,
b. 1 GPM UNIDENTIFIED LEAKAGE,
c. Primary-to-secondary leakage through all steam generators shall
be limited to 150 gallons per day through any one steam generator,
d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and
3. ____ GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure
of 2235 +/- 20 psig.
f. 1 GPM leakage at a Reactor Coolant System pressure of 2235 +/-
20 psig from any Reactor Coolant System Pressure Isolation Valve
specified in Table 3.4-1.
For licensees who want to reduce RCS specific iodine activity, the
following TS pages apply:
Reactor Coolant System
3/4.4.8 Specific Activity
3.4.8 The specific activity of the primary coolant shall be
limited to:
a. Less than or equal to [reduced value] microcurie per gram DOSE
EQUIVALENT I-131, and
b. Less than or equal to 100/E microcuries per gram.
APPLICABILITY: MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1, 2, AND 3*:
a. With the specific activity of the primary coolant greater than
[reduced value] microcurie per gram DOSE EQUIVALENT I-131 for more than
48 hours . . .
* * * * *
MODES 1, 2, 3, 4, and 5:
a. With the specific activity of the primary coolant greater than
[reduced value] microcurie per gram DOSE EQUIVALENT I-131 or greater
[Revised Figure 3.4-1 to lower the line by a factor corresponding
to the reduction in specific activity. The lowered line should parallel
the original]
Reactor Coolant System
BASES
3/4.4.5 STEAM GENERATORS
[To be provided in the final generic letter]
Dated at Rockville, Maryland, this 8th day of August 1994.
For the Nuclear Regulatory Commission.
Elizabeth L. Doolittle,
Acting Chief, Generic Communications Branch, Division of Operating
Reactor Support, Office of Nuclear Reactor Regulation.
[FR Doc. 94-19721 Filed 8-11-94; 8:45 am]
BILLING CODE 7590-01-M