Information Notice 2002-02, Recent Experience with Plugged Steam Generator Tubes
| ML021980191 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Three Mile Island  |
| Issue date: | 07/17/2002 |
| From: | Beckner W NRC/NRR/DRIP/RORP |
| To: | |
| Petrone C , NRC/NRR/RORP, 415-1027 | |
| References | |
| TAC M4964 IN-02-002, Suppl 1 | |
| Download: ML021980191 (11) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001 July 17, 2002 NRC INFORMATION NOTICE 2002-02, Supplement 1: 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 (IN) to inform
addressees of findings from recent inspections and examinations of steam generator tubes at
Oconee Nuclear Station Unit 1 (ONS-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.
Background
Potential severance of plugged steam generator tubes was discussed in IN 2002-02, Recent
Experience With Plugged Steam Generator Tubes, (ML013480327) as a result of inspection
findings at Three Mile Island Unit 1 (TMI-1) during the fall 2001 refueling outage. At TMI-1, a
plugged tube, located on the periphery of the tube bundle, severed near the secondary side of
the upper tubesheet and damaged four adjacent in-service (i.e., nonplugged) tubes. The
preliminary laboratory investigation of the severed tube found 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. As the tube heated during
plant startup, the water in the tube expanded faster than it could escape past the tube plugs, thereby resulting in a pressure buildup and subsequent swelling of the tube. The effect is
called 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
IN 2002-02 Sup 1 IGA on the outside of the tube might have made it more susceptible to severing and that the
plug type was probably a factor in the diode effect. In addition, the industry concluded that it
was unlikely the once-through steam generator tubes would sever in the lower tubesheet
region.
Description of Circumstances
On March 25, 2002, ONS-1 was shut down for a refueling outage. In addition to the standard
steam generator tube inspections, the licensee performed supplemental inspections of plugged
tubes in both steam generators. These supplemental inspections were performed to address
the TMI-1 plugged tube severance event (discussed in Information Notice 2002-02). During the
supplemental inspections the licensee removed a number of plugs, dewatered the tube, and
examined the tubes with eddy current, including tube diameter measurements to determine if
the tube was swollen. No tubes deplugged as a result of this inspection showed any evidence
of swelling or tube severance at the upper or lower tubesheet. The tubes were replugged prior
to plant restart.
While inspecting a deplugged tube in the B steam generator, row 77 tube 123 (B77-123), the
licensee identified signs of wear on the outside tube surface near the secondary face of the
lower tubesheet. The wear scar on the tube began at the lower tubesheet secondary face and
extended up approximately 6.4 inches and was approximately 28 percent through wall. This
wear scar was not detected when the tube was plugged in 1991. An explosive plug in the lower
tubesheet of Tube B78-124, which is adjacent to Tube B77-123 (see Figure 1), could not be
removed. Therefore, in accordance with the inspection program, the lower tubesheet explosive
plug was inspected. It was found to have defects and, per the program, the tube was required
to be captured/caged (i.e., surrounded by tubes that were plugged and stabilized). Since
previously plugged tube B78-123 was required to be a capture location for tube B78-124, it was
deplugged and inspected with eddy current probes. This tube also showed wear in the span
near the lower tubesheet area. Tube B78-123 had been plugged in 1999 for volumetric
indications in the lower span which could be traced back to 1995. The licensee compared the
2002 inspection data to the 1999 data. The indications appeared to be similar in size. The
licensee now believes the volumetric indications were wear indications due to impact from an
adjacent tube.
The circumferential location of the wear on these two tubes indicated that the wear was most
likely due to impact from an adjacent tube, B78-124. Subsequent visual inspection from the
secondary side of the steam generator indicated that tube B78-124 was completely severed at
the secondary side (i.e., top) of the lower tubesheet. Based on the flow direction and velocity in
the lower tubesheet region, the severed tube would be expected to impact three tubes
(B77-123, B78-123, and B79-127). Tube B79-127 was visually inspected from the secondary
side and a small wear mark was visible. Based on the location and preliminary characterization
of the severed tubes degradation, the licensee concluded it potentially represented a new
damage mechanism for once-through steam generators. The NRC obtained this information
from the licensee via several conference calls, summaries of which are being prepared and will
be made available on the docket.
The licensee removed several tubes, including the upper half of the fracture surface of the
severed tube, to assess the root cause of the failure. The preliminary laboratory investigation
of the severed tube indicated inside diameter IGA through the entire thickness of the tube wall.
There were no indications of fatigue or significant ductile tearing. There were three areas of
IN 2002-02 Sup 1 smeared metal from rubbing against the three adjacent worn tubes. After removing the upper
half of the fracture surface of the severed tube, the licensee examined the remaining section of
tube B78-124 with a bobbin coil probe and a rotating probe. The licensee determined that slight
tube swelling had occurred between the 12th and 15th support plates, but no swelling was
observed near the point of severance. This was very different than the experience at TMI-1.
The licensee determined the most likely cause of failure was inside diameter IGA, which implies
a corrosive environment inside the tube. This tube had a unique history. It was 1 of 12 tubes
that had first-of-a-kind (FOAK) instrumentation installed in 1971 prior to commercial operation.
The purpose of this instrumentation was to determine various temperature distributions in the
once-through steam generator design. To conduct this testing, thermocouples were installed in
12 tubes in the B steam generator. In accordance with the FOAK instrumentation installation
procedure, the tube which would eventually sever was plugged with an explosive plug in the
lower tubesheet in 1971. Testing was completed during the first cycle of operation in 1974 and
an explosive plug was installed in the upper tubesheet at that time. It is likely the tube was
approximately 50 percent full of water when the upper plug was installed. The upper tubesheet
plug was backed up by a welded plug in 1993 due to concerns about degradation of upper
tubesheet explosive plugs. Based on a number of observations and hypotheses, the licensee
believes the unique conditions of operation associated with FOAK instrumentation are a primary
contributor to the corrosive environment and eventual tube failure. The secondary-side flow
conditions at the lower tubesheet are in the radial direction toward the center of the tube
bundle. The licensees analysis of the flow conditions shows that velocities around the severed
tube are large and would develop forces sufficient to pin the severed end of a tube against the
adjacent tubes and cause wear.
The licensee pressure-tested tubes B77-123 and B78-123 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
maximum test pressure of three times the differential pressure encountered during normal
operating conditions was reached for both tubes with no leakage or burst. Therefore, neither
tube challenged the steam generator tube structural or leakage performance criteria discussed
by the Nuclear Energy Institute in NEI 97-06, Steam Generator Program Guidelines.
Except for the above findings, the supplemental tube deplugging and inspection program did
not identify any other tubes with evidence of tube severance, significant degradation, or
swelling at the lower or upper tubesheet. Inspection of tubes surrounding the plugged tubes
found no other wear indications that suggested the presence of a severed tube. All tube
locations adjacent to the 12 FOAK instrumentation tubes were plugged and stabilized at the
lower tubesheet such that any FOAK instrumentation tube severance in the lower tubesheet
would not result in further degradation of any adjacent in service-tubes.
Discussion
The event at ONS-1 is another example of the potential for a plugged tube to affect the integrity
of adjacent tubes. Although this phenomenon does not appear to be widespread, it may
become more frequent as more tubes are plugged and as the length of time plugged tubes are
in service increases. Isolated occurrences of this phenomenon may be risk-significant.
IN 2002-02 Sup 1 Based on the TMI-1 event and earlier events (discussed in IN 2002-02), the severance of the
plugged tube at ONS-1 was not expected. The industry did not expect the tube span near the
lower tubesheet to be susceptible to the diode effect observed at TMI-1. The licensee for
ONS-1 has pointed out that there are differences between the ONS-1 and TMI-1 event. For
example: the ONS-1 tube did not swell in the region of the severance; the preliminary
metallurgical examination showed no signs of high cycle fatigue and very minimal signs of
ductile tearing; the wear rate on the adjacent tubes appears to have been slower than at TMI-1;
and the root cause of the failure appears to be the unique history of the tube, not the diode
effect seen at TMI-1. However, as the licensee has stated, its conclusions about the root cause
of the event are based on circumstantial evidence drawn from several observations and
hypotheses. Conclusive evidence was not available.
The events at ONS-1 illustrate a mechanism by which a plugged tube can sever and impact
adjacent tubes. At ONS-1 the impacted tubes were inactive (i.e., plugged), but the severed
tube could have impacted active tubes. Since the wear indications in tube B78-123 could be
traced back to 1995, it appears that the wear rates at ONS-1 were relatively slow. However, this was a function of the location of the tube within the tube bundle. If the severed tube had
been at a different location, the wear rates could have been higher. There was no evidence
that the degradation was related to the tubes location within the tube bundle (only that it was
possibly limited to tubes which in FOAK instrumentation was installed). The ONS-1 results
show the importance of either evaluating plugged tubes or stabilizing the entire length of
plugged tubes in all regions of the tube bundle to ensure they do not compromise the integrity
of adjacent active tubes (i.e., the reactor coolant pressure boundary).
An additional issue of importance is the wear scars on tube B78-123. Tube B78-123 had been
plugged in 1999 for volumetric indications in the lower span which could be traced back to
1995. The licensee now believes the volumetric indications were wear indications due to impact
from an adjacent tube. The wear scars could have been considered to be a precursor to the
condition discovered during the spring 2002 steam generator inspection. This highlights the
importance of fully assessing tube degradation (e.g., volumetric indications) with no conclusive
cause.
The NRC staff is continuing to evaluate the generic implications of the ONS-1 and TMI-1 occurrences.
IN 2002-02 Sup 1 This information notice requires no 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 Office of Nuclear Reactor Regulation (NRR) project manager.
/RA/
William D. Beckner, Program Director
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contacts: Kenneth J. Karwoski, NRR Jerome J. Blake, RII
301-415-2752 404-562-4607 E-mail: kjk1@nrc.gov E-mail: jjb1@nrc.gov
Cheryl B. Khan, NRR
301-415-2751 E-mail: cdb@nrc.gov
1. Figure 1 Steam Generator B Tubesheet Pattern
2. List of Recently Issued NRC Information Notices
IN 2002-02 Sup 1 This information notice requires no 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 Office of Nuclear Reactor Regulation (NRR) project manager.
/RA/
William D. Beckner, Program Director
Operating Reactor Improvements Program
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contacts: Kenneth J. Karwoski, NRR Jerome J. Blake, RII
301-415-2752 404-562-4607 E-mail: kjk1@nrc.gov E-mail: jjb1@nrc.gov
Cheryl B. Khan, NRR
301-415-2751 E-mail: cdb@nrc.gov
Attachment:
1. Figure 1 Steam Generator B Tubesheet Pattern
2. List of Recently Issued NRC Information Notices
DISTRIBUTION:
IN File
- See previous concurrence
ADAMS ACCESSION NO.: ML021980191 Template: NRR-052 DOCUMENT NAME: G:RORP\OES\Staff Folders\Petrone\Oconee Information Notice - Version 5.wpd
OFFICE RSE:RORP:DRIP TECH EDITOR EMCB:DE EMCB:DE RGNII:DRS:MB
NAME CDPetrone* PKleene* KJKarwoski* CDKhan* JJBlake
DATE 07/09/2002 07/02/2002 07/12/2002 07/12/2002 / /2002 OFFICE EMCB:DE BC:EMCB:DE SC:TSS:RORP:DRIP PD:RORP:DRIP
NAME LLund* WHBateman* TReis WDBeckner
DATE 07/10/2002 07/15/2002 07/16/2002 07/17/2002 OFFICIAL RECORD COPY
Attachment 1 IN 2002-02 Sup 1 OCONEE NUCLEAR STATION UNIT 1
STEAM GENERATOR B TUBESHEET PATTERN
FIGURE 1
Attachment 2 IN 2002-02 Sup 1 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information Date of
Notice No. Subject Issuance Issued to
_____________________________________________________________________________________
2002-23 Unauthorized Administration of 07/16/2002 All Medical Licensees.
Byproduct Material for Medical
Use
2002-22 Degraded Bearing Surfaces in 06/28/2002 All holders of operating licenses
GM/EMD Emergency Diesel for pressurized- or boiling-water
Generators nuclear power reactors, including
those that have ceased
operations but have fuel on site.
2002-21 Axial Outside-Diameter 06/25/2002 All holders of operating licenses
Cracking Affecting Thermally for pressurized-water reactors
Treated Alloy 600 Steam (PWRs), except those who have
Generator Tubing permanently ceased operations
and have certified that fuel has
been permanently removed from
the reactor.
2002-20 Microwave Parameter Intrusion 06/20/2002 All holders of operating licenses
Detection System Installation or construction permits for
nuclear power reactors, decommissioning reactors with
fuel on site, category 1 fuel
facilities, uranium conversion
facilities, independent spent fuel
installations, and gaseous
diffusion plants that are
authorized to receive Safeguards
Information and have a need to
know.
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issued by subscribing to the NRC listserver as follows:
To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the following
command in the message portion:
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______________________________________________________________________________________
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