Bulletin 88-02: Rapidly Propagating Fatigue Cracks in Steam Generator Tubes
OMB No.: 31500011
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
February 5, 1988
NRC BULLETIN NO. 88-02: RAPIDLY PROPAGATING FATIGUE CRACKS IN STEAM
For Action - All holders of operating licenses or construction permits
for Westinghouse (W)-designed nuclear power reactors with steam
generators having carbon steel support plates. Steam generators in this
category include Westinghouse models 13, 27, 44, 51, D1, D2, D3 and D4,
and the Westinghouse model E steam generators at South Texas Unit 1.
For Information - All other holders of operating licenses or construction
permits for Westinghouse (W) and Combustion Engineering (CE) designed
nuclear power reactors.
The purpose of this bulletin is to request that holders of operating licenses
or construction permits for Westinghouse (W)-designed nuclear power reactors
with steam generators having carbon steel support plates implement actions
specified herein to minimize the potential for a steam generator tube rupture
event caused by a rapidly propagating fatigue crack such as occurred at North
Anna Unit 1 on July 15, 1987.
Description of Circumstances:
On July 15, 1987, a steam generator tube rupture event occurred at North Anna
Unit 1 shortly after the unit reached 100% power. For several days prior to
the event, operators had observed erratic air ejector radiation monitor read-
ings. Grab samples were taken prior to the tube rupture for purposes of
performing environmental release calculations. Subsequent analysis of this
data indicated that increasing primary to secondary leakage had occurred over
a 24- to 36-hour period before the tube rupture event. This leakage had been
below the limit given in the Technical Specifications. The ruptured tube was
located in Row 9 Column 51 in steam generator "C." The rupture location in
this model 51 steam generator was at the top support plate on the cold leg
side of the tube. The rupture extended circumferentially 360ø around the
The cause of the tube rupture has been determined to be high cycle fatigue.
The source of the loads is believed to be a combination of a mean stress level
in the tube and a superimposed alternating stress. (The mean stress is
produced by denting of the tube at the uppermost tube support plate, and the
alternating stress is the result of out-of-plane deflection of the U-bend
portion of the
. NRCB 88-02
February 5, 1988
Page 2 of 5
tube above the uppermost support plate, caused by flow-induced vibration.)
Denting also shifts the maximum tube bending stress to the vicinity of the
uppermost tube support plate. These loads are sufficient to produce fatigue
in an all volatile treatment (AVT) water chemistry environment.
The specific mechanism for the flow-induced vibration has been determined to
be a fluid-elastic instability. The fluid-elastic mechanism has a significant
effect on tube response in cases where the fluid-elastic stability ratio
equals or exceeds 1.0. The stability ratio, SR, is defined as:
SR = V eff / V c
where V eff is the effective crossflow velocity and V c is the critical
velocity beyond which the displacement response to the tube increases rapidly.
The most significant contributors to the occurrence of a high fluid-elastic
stability ratio are believed to have been (1) a reduction in damping at the
tube-to-tube support plate intersection caused by denting and (2) locally high
flow velocities caused by non-uniform antivibration bar (AVB) penetrations
into the tube bundle. The presence of an AVB support will restrict tube
motion and thus preclude the deflection amplitude required for fatigue. The
original design configuration required AVBs to be inserted to Row 11.
However, inspections have shown that some AVBs in the North Anna Unit 1 steam
generators penetrate to Row 8, exceeding the minimum AVB design depth.
However, no AVB support was present for the Row 9 Column 51 tube that
Based on available information, the staff concludes that the presence of all
the following conditions could lead to a rapidly propagating fatigue failure
such as occurred at North Anna:
(1) denting at the upper support plate
(2) a fluid-elastic stability ratio approaching that for the tube that
ruptured at North Anna
(3) absence of effective AVB support
Within 45 days following receipt of this bulletin, addressees having
Westinghouse steam generators with carbon steel support plates shall submit a
written report detailing the status of their compliance with the actions
specified below for purposes of minimizing the potential for rapidly
propagating fatigue failure such as occurred at North Anna 1. The report
shall include an appropriate schedule for completion of the analyses described
under item C below, if applicable.
. NRCB 88-02
February 5, 1988
Page 3 of 5
A. The most recent steam generator inspection data should be reviewed for
evidence of denting at the uppermost tube support plate. Inspection
records may be considered adequate for this purpose if at least 3% of the
total steam generator tube population was inspected at the uppermost
support plate elevation during the last 40 calendar months. "Denting"
should be considered to include evidence of upper support plate corrosion
and the presence of magnetite in the tube-to-support plate crevices,
regardless of whether there is detectable distortion of the tubes. The
results of this review shall be included as part of the 45-day report.
Where inspection records are not adequate for this purpose, inspections
of at least 3% of the total steam generator tube population at the
uppermost support plate elevation should be performed at the next
refueling outage. The schedule for these inspections shall be included
as part of the 45-day report and the results of the inspections shall be
submitted within 45 days of their completion. Pending completion of
these inspections, an enhanced primary-to-secondary leak rate monitoring
program should be implemented in accordance with paragraph C.1. below.
B. For plants where no denting is found at the uppermost support plate, the
results of future steam generator tube inspections should be reviewed for
evidence of denting at the uppermost support plate. If denting is found
in the future, the provisions of item C below should be implemented.
Commitments to implement these actions shall be submitted when the
results of A above are submitted.
C. For plants where denting is found, the NRC staff requests that the fol-
lowing actions be taken:
1. Pending completion of the NRC staff review and approval of the
program described in C.2 below or completion of inspections
specified in item A above to confirm that denting does not exist, an
enhanced primary-to-secondary leak rate monitoring program should be
implemented as an interim compensatory measure within 45 days of the
date of receipt of this bulletin.* Implementation of this program
shall be documented as part of the 45-day report. The enhanced
monitoring program is intended to ensure that if a rapidly
propagating fatigue crack occurs under flow-induced vibration, the
plant power level would be reduced to 50% power or less at least 5
hours before a tube rupture was predicted to occur. The
effectiveness of this program should be evaluated against the
assumed time-dependent leakage curve given in Figure 1.
*While this bulletin was being prepared, licensees for a few plants committed
to an enhanced primary-to-secondary leak rate monitoring program at the
staff's request. These plants had been identified on a preliminary basis by
Westinghouse as being potentially susceptible to rapidly propagating fatigue
cracks. These enhanced programs should be upgraded as necessary to comply
with this paragraph. However, no relaxation from current commitments should
be made without prior approval by the NRC staff.
. NRCB 88-02
February 5, 1988
Page 4 of 5
This program should consider and provide the necessary leakage
measurement and trending methods, time intervals between measure-
ments, alarms and alarm setpoints, intermediate actions based on
leak rates or receipt of alarms, administrative limits for
commencing plant shutdown, and time limitations for (1) reducing
power to less than 50% and (2) shutting down to cold shutdown.
Appropriate allowances for instrument errors should be considered.
Finally, the program should make provision for out of service
radiation monitors, including action statements and compensatory
2. A program should be implemented to minimize the probability of a
rapidly propagating fatigue failure such as occurred at North Anna
Unit 1. The need for long-term corrective actions (e.g., preventive
plugging and stabilization of potentially susceptible tubes,
hardware, and/or operational changes to reduce stability ratios)
and/or long-term compensatory measures (e.g., enhanced leak rate
monitoring program) should be assessed and implemented as necessary.
An appropriate program would include detailed analyses, as described
in subparagraphs (a) and (b) below, to assess the potential for such
a failure. Alternative approaches and/or compensatory measures
implemented in lieu of the actions in subparagraphs (a) or (b) below
should be justified.
Although the 45-day report shall provide a clear indication of
actions proposed by licensees, including their status and schedule,
a detailed description of this program and the results of analyses
shall be submitted subsequently, but early enough to permit NRC
staff review and approval prior to the next scheduled restart from a
refueling outage. Where the next such restart is scheduled to take
place within 90 days, staff review and approval will not be
necessary prior to restart from the current refueling outage. An
acceptable schedule for submittal of the above information should be
arranged with the NRC plant project manager by all licensees to
ensure that the staff will have adequate time and resources to
complete its review without adverse impact on the licensee's
schedule for restart.
(a) The analysis would include an assessment of stability ratios
(including flow peaking effects) for the most limiting tube
locations to assess the potential for rapidly propagating
fatigue cracks. This assessment would be conducted such that
the stability ratios are directly comparable to that for the
tube which ruptured at North Anna.
(b) The analysis would include an assessment of the depth of pene-
tration of each AVB. The purpose of this assessment is
twofold: (1) to establish which tubes are not effectively
supported by AVBs and (2) to permit an assessment of flow
. NRCB 88-02
February 5, 1988
Page 5 of 5
(Note: Most steam generators have at least two sets of AVBs.
This applies only to the set that penetrates most deeply into
the tube bundle.) The methodology used to determine the depth
of penetration of each individual AVB shall be described in
detail in the written report. The criteria for determining
whether a tube is effectively supported by an AVB shall also be
identified. (Note: An AVB that penetrates far enough to
produce an eddy current signal in a given tube may not
penetrate far enough to provide a fully effective lateral
support to that tube.)
If addressees cannot perform this suggested approach or meet this suggested
schedule, they should justify to the NRC their alternative approaches and
The written reports shall be submitted to the appropriate Regional
Administrator under oath or affirmation under provisions of Section 182a,
Atomic Energy Act of 1954, as amended. In addition, the original of the cover
letter and a copy of the reports shall be transmitted to the U.S. Nuclear
Regulatory Commission, Document Control Desk, Washington, D.C. 20555 for
reproduction and distribution.
This request for information was approved by the Office of Management and
Budget under blanket clearance number 31500011. Comments on burden and
duplication may be directed to the Office of Management and Budget, Reports
Management, Room 3208, New Executive Office Building, Washington, D.C. 20503.
The NRC intends to review the information collected under this bulletin and
determine the adequacy of specific actions proposed by each licensee. The
information will be analyzed and placed in the NRC Public Document Rooms.
If you have any questions about this matter, please contact one of the
technical contacts listed below or the Regional Administrator of the
appropriate regional office.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts: Emmett Murphy, NRR
Keith Wichman, NRR
1. Figure 1 Leak Rate Versus Time Chart
2. List of Recently Issued NRC Bulletins
Page Last Reviewed/Updated Thursday, March 25, 2021