Information Notice 2003-13, Steam Generator Tube Degradation at Diablo Canyon
| ML032410215 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/28/2003 |
| From: | Beckner W NRC/NRR/DIPM |
| To: | |
| Dozier J, NRR/IROB 415-1014 | |
| References | |
| IN-03-013 | |
| Download: ML032410215 (11) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
August 28, 2003 NRC INFORMATION NOTICE 2003-13:
STEAM GENERATOR TUBE DEGRADATION AT
DIABLO CANYON
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 is issuing this information notice to inform
addressees about findings from a recent steam generator tube inspection at the Diablo Canyon
Power Plant, Unit 2 (DCPP-2). 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, no specific action or written response is required.
Description of Circumstances
DCPP-2 has four Westinghouse model 51 steam generators (SGs), with 7/8 inch outside
diameter (OD), mill-annealed Alloy 600 tubing and drilled hole carbon steel tube support plates.
The model 51 steam generator has 45 rows of tubes, with row 1 having the smallest bend radii
in the U-bend area.
During Operating Cycle 11, a small steam generator tube leak (less than or equal to
approximately 6.5 gallons per day) was present at DCPP-2. During the 2003 refueling outage, Pacific Gas & Electric (PG&E), the licensee for DCPP-2, performed SG secondary side
pressure tests to locate the source of the SG leakage. Several potentially leaking SG tubes
were identified and subsequent eddy current testing identified two contributing degradation
modes: circumferential primary water stress corrosion cracking (PWSCC) in the U-bend region
and axial outside diameter stress corrosion cracking (ODSCC) at the tube-to-tube support plate
intersections.
Circumferential Indications
The licensee inspected the U-bend region of all of the active tubes in all four SGs with a
rotating eddy current probe after finding leaking circumferential flaws in the U-bend region of a
row 5 tube and identifying other circumferential indications in the U-bend region of other high
row (above row 3) tubes during the initial examinations. In all, 12 tubes in rows 3 through 10
were identified with circumferential indications in the U-bend region. The indications were
short, about 0.25 inch long. They were on the tube flank and originated from the inside of the
tube. PG&E concluded that PWSCC was the most probable cause of cracking based on the
susceptibility of mill-annealed Alloy 600 tubing, the residual stresses from bending the tubes, and the exposure of the tubes to primary coolant. It was not practical to remove one of these
flawed tubes for destructive analysis because of the tubes location in the SG. All tubes with
circumferential indications in the U-bend region were pressure-tested in situ and met the
structural integrity performance criteria. The estimated potential for accident induced leakage
from these tubes when combined with estimated potential for accident induced leakage from
other tubes in the SG did not exceed the accident leakage integrity performance criteria.
Axial ODSCC
All of the axial ODSCC indications identified in the potentially leaking tubes had been detected
during the previous SG tube inspection. The tubes were left in service in accordance with the
voltage-based alternate repair criteria (ARC) discussed in Generic Letter (GL) 95-05, Voltage- Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter
Stress Corrosion Cracking, and the plants technical specifications. During subsequent bobbin
coil inspections of the SG tubes during the 2003 refueling outage, the licensee detected an
unexpected number of large bobbin voltage indications (i.e., greater than 3 volts), including one
indication which had a substantially larger voltage than any other indication. This indication
measured 21.5 volts. The licensee investigated the cause of the 21.5 volt indication and the
cause of the unexpected number of high voltage (>3 volts) indications.
The licensee found that the tube with the 21.5 volt indication had been previously identified with
a 2 volt indication during the 2001 refueling outage. The tube was left in service at that time
since the bobbin voltage (2 volts) met the plants technical specification repair criteria for
remaining in service. PG&E had also inspected the indication with a rotating eddy current
probe during the 2001 outage and the 2003 outage. During the root cause investigation in
2003, PG&E used the rotating probe eddy current data to estimate the length and depth of the
flaw for both outages. The licensee estimated that the flaw was nearly through wall at the time
of the 2001 refueling outage and that the flaw fully penetrated the tube wall during the ensuing
operating cycle. This through wall penetration occurred over a significant portion of the crack
length. Industry data shows that flaw voltage response increases sharply upon initial through
wall penetration and with subsequent lengthwise growth of the through wall component. Thus, the licensee concluded that the large increase in voltage observed for this indication over the
operating cycle was the result of the crack penetrating entirely through wall over a significant
length rather than being the result of a significant increase in growth rate of the physical
dimensions of the flaw.
The licensee partly attributed the unexpected number of large voltage indications (other than
the 21.5 volt indication) to the voltage growth rate distribution used in earlier projections. These
earlier projections were based on the industry methodology outlined in the EPRI report Steam
Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates
Database for Alternate Repair Limits. The methodology described in the EPRI report
recognizes that voltage growth rates can be higher for indications with larger initial voltages. To
determine the growth rate distribution, the growth rate data is binned based on this initial
voltage. The industry methodology specifies that each bin (i.e., range of initial voltages) should
have a certain number of growth rate data points. To satisfy this minimum number-of-data- points requirement, the licensee expanded the range of the highest growth rate bin to include the growth rates for indications with lower initial voltages. This reduced the probability that a
large growth rate would be selected for indications in the upper voltage bin and resulted in
nonconservative projections of the bobbin voltage.
The licensee pulled two tubes with axial ODSCC indications for destructive examination: the
tube with the largest axial ODSCC flaw voltage and a tube with a large voltage axial ODSCC
flaw. The laboratory burst and leakage test results indicated that both tubes had adequate
integrity.
Discussion
DCPP-2 identified two steam generator tube issues during the 2003 refueling outage:
circumferential indications in the U-bend region out to row 10, and large, increases in bobbin
voltage associated with axial ODSCC at the tube-to-tube support plate intersections.
Historically, industry practice has been to inspect the U-bend region of low row tubes for
indications of circumferential cracking and to expand the inspection to higher row tubes based
on the results. This practice was developed based on the understanding that the low row tubes
have higher residual stress levels in the U-bend region due to the tighter bend radii than the U-
bend region of higher row tubes. The higher residual stress level makes the U-bend region of
low row tubes more susceptialble to cracking. Prior to the experience at Diable Canyon, operating experience from steam generator inspections has validated this approach. The
experience from DCPP-2 suggests that the U-bend region of higher row tubes may have a
similar susceptibility to cracking in the U-bend region as the U-bend region of lower row tubes.
PG&Es experience with axial ODSCC indications yields two insights. First, collection and
further evaluation of rotating probe inspection data from axial ODSCC indications may help
identify indications that could be prone to significant bobbin voltage growth. Second, when
developing voltage-dependent growth rates through the use of generic industry guidance or
other methodologies, it is important that the methodology result in conservative growth rates.
The previously described examples of SG tube degradation illustrates the need for maintaining
robust steam generator inspection programs. An effective program should sample for
degradation based on both operating experience and engineering assessments of potentially
susceptible locations and should be able to conservatively predict degradation growth.
Related Generic Communications
The following documents describe other recent reactor operating experience with steam
generator tubes:
IN 2003-05, Failure to Detect Freespan Cracks in PWR Steam Generator Tubes, dated
June 5, 2003
IN 2002-02 and IN 2002-02 supplement 1, Recent Experience With Plugged Steam
Generator Tubes dated January 8, 2002 and July, 17, 2002 IN 2002-21, Axial Outside-Diameter Cracking Affecting Thermally Treated Alloy 600
Steam Generator Tubing dated June 25, 2002 IN 2001-16, Recent Foreign and Domestic Experience with Degradation of Steam
Generator Tubes and Internals, dated October 31, 2001 NRC Generic Letter 97-05, Steam Generator Tube Inspection Techniques, dated
December 17, 1997
Inspection Report 50-323/03-09, Diablo Canyon Power Plant - NRC Special Team Inspection
Report dated May 8, 2003 (Adams ML031290198)
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, Branch Chief
Reactor Operations Branch
Division of Inspection Program Management
Office of Nuclear Reactor Regulation
Technical contacts:
Martin Murphy, NRR
301-415-3138
301-415-1014 E-mail: mcm2@nrc.gov
E-mail: jxd@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
- See previous concurrence
OFFICE
OES:IROB:DIPM
Tech Editor
EMCB:DE
EMCB:DE
BC:EMCB:DE
NAME
IJDozier*
PKleene*
MCMurphy*
LALund*
WHBateman*
DATE
07/15/2003
07/15/2003
08/12/2003
08/12/2003
08/20/2003 OFFICE
SC:OES:IROB:DIPM
BC:IROB:DIPM
NAME
TReis
WDBeckner
DATE
08/28/2003
08/28/2003
______________________________________________________________________________________
OL = Operating License
CP = Construction Permit
Attachment 1 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information
Date of
Notice No.
Subject
Issuance
Issued to
_____________________________________________________________________________________
89-69, Sup 1
Shadow Corrosion Resulting in
Fuel Channel Bowing
08/25/2003
All holders of operating licenses
for boiling water reactors (BWRs),
except those who have
permanently ceased operations
and have certified that fuel has
been permanently removed from
the reactor vessel.
2003-12
Problems Involved in
Monitoring Dose to the Hands
Resulting from the Handling of
Radiopharmaceuticals
08/22/2003 All holders of 10 CFR Parts 32,
33, and 35 licenses.
2003-11 Leakage Found on Bottom-
Mounted Instrumentation
Nozzles
08/13/2003
All holders of operating license or
construction permits for nuclear
power reactors, except those that
have permanently ceased
operations and have certified that
fuel has been permanently
removed from the reactor.
2003-10
Criticality Monitoring System
Degradation at BWX
Technologies, Inc., Nuclear
Products Division, Lynchburg, VA
08/04/2003
All U.S. Nuclear Regulatory
Commission (NRC) licensees
authorized to possess a critical
mass of special nuclear material.
2002-26, Sup 1
Additional Failure of Steam
Dryer after a Recent Power
Uprate
07/21/2003
All holders of operating license or
construction permits for nuclear
power reactors, except those that
have permanently ceased
operations and have certified that
fuel has been permanently
removed from the reactor.
Note:
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