Information Notice 2002-02, Recent Experience with Plugged Steam Generator Tubes
| ML013480327 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/08/2002 |
| From: | Freeman M Operational Experience and Non-Power Reactors Branch |
| To: | |
| References | |
| TAC MB3376 IN-02-002 | |
| Download: ML013480327 (11) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
January 8, 2002
NRC INFORMATION NOTICE 2002-02:
RECENT EXPERIENCE WITH PLUGGED STEAM
GENERATOR TUBES
Addressees
All holders of operating licenses for pressurized-water reactors (PWRs), except those who have
permanently ceased operations and have certified that fuel has been permanently removed
from the reactor.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to inform
addressees about findings from recent inspections and examinations of steam generator tubes
at Three Mile Island Unit 1 (TMI-1). The NRC anticipates that recipients will review the
information for applicability to their facilities and consider taking actions, as appropriate, to
avoid similar problems. However, suggestions contained in this information notice do not
constitute NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances
On October 8, 2001, TMI-1 was shut down for a refueling outage. While inspecting the tubes in
the B Steam Generator, the licensee (AmerGen Energy Company, LLC) identified signs of wear
near the upper tubesheet on the outer surface of four tubes on the periphery of the tube bundle.
These wear indications did not appear to have been present during the prior steam generator
tube inspections, which were performed approximately 2 years earlier. Given the pattern and
location of the wear signs, the licensee suspected that a neighboring plugged tube had caused
the wear. The licensee removed the upper plug in the suspected tube and performed a video
inspection. The video inspection revealed that the tube was severed near the secondary face
of the upper tubesheet and in physical contact with the drilled hole where the tube passed
through the 15th support plate. Neighboring tubes were not in contact with the drilled holes at
the point where they passed through the 15th support plate.
The licensee removed the lower portion of the fractured surface of the severed tube to assess
the root cause. The preliminary laboratory investigation of the severed tube indicated signs of
high cycle fatigue, ductile failure, and outside diameter initiated intergranular attack (IGA).
In addition, the tube diameter was greater than the nominal tube diameter, indicating that the
severed tube had swollen.
The licensee determined that the most likely cause of failure was fatigue caused by flow- induced vibration of the swollen and restrained tube. The licensee attributed the swelling to
water leaking into the tube around the plugs while the unit was shut down, followed by
subsequent expansion of that water during plant startup. As the plant was heated, the water in
the tube expanded at a rate greater than that which could escape past the tube plugs, thereby
resulting in a pressure buildup and subsequent swelling of the tube. The effect is referred to as
the diode effect. The swelling caused the tube to become clamped at the drilled hole in the
15th tube support plate and at the upper tubesheet. Under the high-flow conditions that existed
in the area of the upper tubesheet, this clamping made the restrained tube more susceptible to
fatigue caused by flow-induced vibration. The licensee also concluded that the IGA on the
outside of the tube may have made it more susceptible to this phenomenon.
To address the concern that other plugged tubes might impact active tubes in future cycles, the
licensee embarked on a tube deplugging and inspection program. This program indicated that
some plugged tubes contained water, some were swollen, and some exhibited evidence of tube
degradation that was not present when they were originally plugged. The tubes that contained
water or were swollen were located throughout the tube bundle, not just on the periphery as
was the severed tube. The program also revealed that water or swelling only occurred in tubes
that were, or had been in the past, plugged with an Alloy 600 mechanical plug. In 1991, the
licensee began plugging tubes with a newer type of mechanical plug made of Alloy 690. Some
of the tubes that contained water or were swollen were plugged with an Alloy 690 mechanical
plug, but these were replacements for original Alloy 600 mechanical plugs. No tubes with
originally installed Alloy 690 mechanical plugs in the both upper and lower tubesheet showed
any signs of water or swelling.
In addition to the above findings, the tube deplugging and inspection program indicated that
one additional tube had circumferentially severed and one plugged tube had failed axially. The
circumferential sever occurred at the 15th tube support plate. Apparently, the tube was captured
within the tube support plate as there was no wear on any adjacent tubes. The axial failure
occurred in the freespan between the 15th tube support plate and the upper tubesheet.
Adjacent tubes were not impacted. Both tubes were swollen and were, or had been in the past, plugged with an alloy 600 mechanical plug.
To prevent plugged tubes that might sever in the future from impacting other tubes, the
licensee planned to either stabilize each plugged tube or surround it with stabilized tubes to
cage the non-stabilized plugged tube. This stabilizing would generally be achieved by
inserting cables into the tubes from the upper tubesheet through at least the 14th tube support
plate. Most of the tubes that exhibited swelling would be stabilized the full length of the tube.
Background
TMI-1 uses once-through steam generators, in which the primary reactor coolant enters the top
of the steam generator, passes through the upper tubesheet, flows down through approximately
15,500 tubes, passes through the lower tubesheet, and exits the steam generator at the
bottom. Secondary-side feedwater enters the steam generator in the center, flows downward in
the annular region between the tube bundle and the steam generator shell, turns upward and
flows around the outside of the tubes where it is converted to steam. The steam continues upward and then turns downward in the annular region between the tube bundle and the steam
generator shell and exits in the center. Secondary-side flow is characterized by counterflow
conditions in the tube bundle and crossflow conditions near the upper and lower tubesheets.
The highest flowrates on the secondary side of the tubes occur at the periphery of the upper
tubesheet as the steam flow changes from counterflow to crossflow and turns downward in the
annular region. Similar flowrates occur near the lower tubesheet, although the flow is mostly
liquid in this area.
The tubes are attached to an upper and lower tubesheet and supported by 15 tube support
plates, which have holes that permit the passage of the tubes. The holes in the tube support
plates are broached with three points of contact, with the exception of the holes in the periphery
of the 15th tube support plate which are round drilled holes. In accordance with the NRCs
requirements, usually in Technical Specifications, the licensee inspects the tubes during plant
outages to assess the structural and leakage integrity of the tubes. Tubes with degradation
above specified limits are plugged or repaired using methods approved by the NRC.
Plugging involves removing a tube from service by installing plugs in both ends of the tube to
prevent the flow of primary coolant through it. The tube plugs are typically either mechanically
expanded against the tube wall or welded to the tube. Once a tube is removed from service, it no longer requires inspection because it is no longer part of the reactor coolant pressure
boundary.
The tube that severed at TMI-1 and caused wearing of adjacent tubes was Tube 130, located in
the B Steam Generator at Row 66 (B66-130). This tube was plugged with an Alloy 600
mechanical plug in 1986 as a result of IGA near the fifth tube support plate. At the time it was
originally plugged, there was no observable degradation at the point where the tube eventually
severed. The original plug was replaced in 1997 with the newer Alloy 690 mechanical plug as
part of a program to replace many of the Alloy 600 plugs in the upper tubesheet.
Eddy current testing of the four adjacent tubes affected by severed tube B66-130 indicated that
the tube walls had worn through 37 percent of the wall thickness for the least affected tube, to
92 percent of the wall thickness for the most affected tube. The overall length of the wear scars
ranged from approximately 2.8 inches to 8.3 inches. Typically the bobbin coil probe, which is
the eddy current probe most relied upon by industry to detect wear indications, is expected to
detect flaws of this range. However, the licensee determined that those analysts reviewing
bobbin coil data did not detect the wear indication on the least affected tube. Only those
analysts reviewing rotating pancake coil data were able to identify the wear on the least
affected tube at TMI-1.
The licensee also pressure-tested the four affected tubes to determine the margin to failure.
Specifically, the licensee increased pressure until either the tube failed or pressure reached
three times the differential pressure encountered during normal operating conditions. The most
affected tube failed at a point near the design differential pressure for a main steamline break.
One other tube failed above the main steamline break design pressure, but very near three
times the normal operating differential pressure. These two tubes challenged the steam
generator tube structural performance criteria discussed by the Nuclear Energy Institute in NEI 97-06, Steam Generator Program Guidelines. The remaining two tubes had no leakage
when held at three times normal operating differential pressure for 2 minutes.
Discussion
Licensees have plugged tubes using a variety of different plug types and materials since steam
generators were placed in service. The industry has experienced leaking plugs for at least
20 years, and probably back to the initial use of mechanical plugs. If the phenomenon at TMI-1 were widespread, numerous instances of freespan wear scars adjacent to plugged tubes
should have been observed throughout the years. Although this is not believed to be the case, indicating that the phenomenon is not widespread, the potential for plugged tubes to affect the
integrity of adjacent tubes may increase with time. If isolated occurrences of this phenomenon
were to occur, they might be risk-significant.
Known experience suggests that the most likely consequence is an axial failure of the plugged
tube unless a circumferential flaw or high crossflow velocities are present. Although the long- term potential for axial failures to cause a plugged tube to sever is unknown, such tubes are
probably a lesser contributor to risk than plugged tubes that contain circumferential flaws
resulting from corrosion or fatigue. Tubes with detected circumferential flaws are normally
stabilized when plugged; however, undetected circumferential flaws, typically less than
40 percent of the wall thickness, will sometimes be present in tubes that are plugged for other
causes. In addition, the potential for the development of corrosion-related circumferential flaws
in plugged tubes is not clearly known, since the cooler tube temperatures may reduce the
potential for corrosion, while higher axial loads in plugged tubes may work to increase the
potential for circumferential stress corrosion cracking.
Nonetheless, the events at TMI-1 illustrate an effect in which a plugged tube can sever and
affect adjacent tubes. For two of the tubes adjacent to the severed tube, the extent of wear
was such that the structural performance criteria were challenged. The degradation of the
active tubes appeared to occur during one cycle. The three circumstances that apparently
contributed to the severed tube include tube swelling, flow-induced vibration, and tube
degradation from IGA. The results at TMI-1 indicate the importance of either evaluating
plugged tubes or stabilizing plugged tubes to ensure that they do not compromise the integrity
of adjacent active tubes, i.e., the reactor coolant pressure boundary.
High cycle fatigue was the dominant, and perhaps sufficient, cause of the severed tube at
TMI-1. Fatigue was also involved in the severed tube in the original steam generators at the
R. E. Ginna Nuclear Power Plant in 1982. In addition, all known circumferential failures of
active tubes have also involved high cycle fatigue (e.g., Oconee 2, Rancho Seco, North Anna 1, Mahomet). These high cycle fatigue failures involved flow-induced vibration associated with
high crossflow velocities, although repeated impacts from a loose part may also have
contributed to severing the plugged tube at Ginna. Nominally, plugged and active tubes are not
expected to experience fatigue under flow-induced vibration; however, off-nominal conditions
may enhance the potential for excessive vibration and, thus, fatigue. Such off-nominal
conditions could include swelling of tubes, denting, localized flow peaking effects, or adjacent
loose parts. The licensee did not determine if only a specific population of plugged tubes was susceptible to
this phenomenon and stabilize or cage only these tubes, rather they elected to stabilize or cage
all plugged tubes, as discussed above. Nonetheless, the licensee did attempt to address the
flow velocities and stability ratios of concern, the extent to which tubes were clamped as a result
of swelling, denting, and/or other phenomena, whether the plug type (e.g., welded, mechanical)
played a role in determining which tubes swell, and the effects that degradation (e.g., an axial
crack, a fishmouth rupture, a circumferential crack, etc.) in a plugged tube may have on the
flow induced vibration analysis.
In response to the eddy current findings at TMI-1, the NRC independently reviewed some of the
bobbin coil data. For the tube least affected by the severed tube, the NRC determined that the
differential channel showed a very small flaw signal and the absolute channel showed evidence
of signal drift. The NRC concluded that the wear indication may have been missed because of
the relatively small voltage and atypical signal behavior. The absolute drift signal rotated in the
clockwise direction as channel frequencies were decreased whereas a typical flaw signal would
be expected to rotate in the counter-clockwise direction.
The staff is continuing to evaluate the generic implications of the TMI-1 occurrence.
This information notice does not require any specific action or written response. If you have
any questions about the information in this notice, please contact one of the technical contacts
listed below or the appropriate project manager in the NRCs Office of Nuclear Reactor
Regulation (NRR).
/RA/
William D. Beckner, Program Director
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contacts:
Emmett L. Murphy, NRR
M. Scott Freeman, RII
301-415-2710
864-882-6927 E-mail: elm@nrc.gov
E-mail: msf1@nrc.gov
Kenneth J. Karwoski, NRR
Michael C. Modes, RI
301-415-2752
610-337-5198 E-mail: kjk1@nrc.gov
E-mail: mcm@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
- See previous concurrence
OFFICE
REXB
Tech Editor
EMCB
C:EMCB
NAME
MSFreeman*
PGarrity*
LALund*
WHBateman*
DATE
12/20/2001
12/20/2001
12/21/2001
01/04/2002 OFFICE
RORP
PD:RORP
NAME
TKoshy*
WDBeckner*
DATE
01/07/2002
01/08/2002
______________________________________________________________________________________
OL = Operating License
CP = Construction Permit
Attachment 1 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information
Date of
Notice No.
Subject
Issuance
Issued to
_____________________________________________________________________________________
2002-01
Metalclad Switchgear Failures
and Consequent Losses of
Offsite Power
01/08/2002
All holders of Licenses for nuclear
power reactor.
2001-19
Improper Maintenance and
Reassembly of Automatic Oil
Bubblers
12/17/2001
All holders of operating licenses
for nuclear power reactors, except those who have
permanently ceased operations
and have certified that fuel has
been permanently removed from
the reactor vessel.
2001-18
Degraded or Failed Automated
Electronic Monitoring, Control,
Alarming, Response, and
Communications Needed for
Safety and/or Safeguards
12/14/2001 All uranium fuel conversion, enrichment, and fabrication
licensees and certificate holders
authorized to receive safeguards
information. Information notice is
not available to the public
because it contains safeguards
information.
2001-17
Degraded and Failed
Performance of Essential
Utilities Needed for Safety and
Safeguards
12/14/2001 All uranium fuel conversion, enrichment, and fabrication
licensees and certificate holders
authorized to receive safeguards
information. Information notice is
not available to the public
because it contains safeguards
information.
2001-08, Sup. 2
Update on Radiation Therapy
Overexposures in Panama
11/20/2001 All medical licensees.
2001-16
Recent Foreign and Domestic
Experience with Degradation of
steam Generator Tubes and
Internals
10/31/2001
All holders of operating licenses
for pressurized-water reactors
(PWR), except those who have
permanently ceased operations
and have certified that fuel has
been permanently removed from
the reactor.