[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]


-----------------------------------------------------------------------


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.

-----------------------------------------------------------------------

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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    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