Information Notice 2004-16, Tube Leakage Due to a Fabrication Flaw in a Replacement Steam Generator
| ML041460357 | |
| Person / Time | |
|---|---|
| Issue date: | 08/03/2004 |
| From: | Reis T NRC/NRR/DIPM/IROB |
| To: | |
| Karwoski K , NRR/DE/EMCB, 301-415-2752 | |
| References | |
| IN-04-016 | |
| Download: ML041460357 (6) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
August 3, 2004
NRC INFORMATION NOTICE 2004-16:
TUBE LEAKAGE DUE TO A FABRICATION FLAW
IN A REPLACEMENT STEAM GENERATOR
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 recent operating experience with replacement steam generators. In
particular, the potential for tubes to be damaged during fabrication and packaging. The NRC
anticipates that recipients will review the information for applicability to their facilities and
consider taking actions, as appropriate, to avoid similar issues. However, no specific action or
written response is required.
Description of Circumstances
Both steam generators at Palo Verde Nuclear Generating Station (PVNGS), Unit 2, were
replaced in December 2003. These new steam generators incorporate many of the design
enhancements present in other replacement steam generators (e.g., Alloy 690 tubes, stainless
steel tube supports). The tube bundle in the new steam generators is consistent with the
original Combustion Engineering design with a horizontal run between the hot and cold leg
rather than the more typical U-shape.
When the plant was started up following the replacement of the steam generators in December
2003, the licensee observed a small primary-to-secondary leak measuring approximately 0.6 gallons per day (gpd) (2.3 liters per day (lpd)). Over the following 2 months, the leak rate varied
between 0.4 and 0.7 gpd (1.5 and 2.6 lpd) until February 19, 2004, when the leak rate
increased from approximately 0.7 to 11 gpd (2.6 to 41 lpd) in a 38-minute timeframe. Although
the leak rate did not exceed the technical specification limit, the plant was shut down to identify
the source of the leak.
While the plant was shut down, the secondary side of the steam generator was pressurized to
600 pounds per square inch (psi) (4137 kilopascal (kPa)) to assist in the identification of the
leaking tube or tubes. During this pressure test, leakage was easily observed coming from a
peripheral tube. This tube was subsequently inspected with both a bobbin and a rotating probe.
These inspections did not reveal evidence of inservice degradation. These inspections did
confirm the presence of a dent near a vertical support in the middle of the horizontal run of the
tube that was detected in the preservice examination. This dent signal was considered
anomalous because it differed from a typical dent signal in that it exhibited some flawlike
characteristics (i.e., it had a vertical component). A comparison of the preservice bobbin and rotating probe inspection data to the data obtained during the outage revealed no significant
differences in the dent signal. Although the dent signal was anomalous, there was no distinct
indication of material volume loss.
Since the eddy current inspections of the affected tube did not provide conclusive evidence of a
through wall flaw, additional testing was performed. This testing included primary and
secondary side visual inspections and an in situ pressure test. The visual inspections
confirmed the presence of a dent which did not appear to be due to the fabrication of the
support structure since the dent was not located directly next to a support strap and did not
appear to be the result of impact or leverage. During an in situ pressure test of the entire tube,
0.08 gpm (0.3 lpm) leakage was observed at the differential pressure associated with
postulated accident conditions (e.g., a main steam line break), and the tube did not burst at
three times the differential pressure associated with normal operating conditions. These tests
confirmed the tube had adequate structural integrity. In addition, the leakage from this tube
was well below the allowable leakage under postulated accident conditions. Following the in
situ pressure test, the leaking tube was plugged and stabilized.
In response to the findings regarding the leaking tube, the rotating probe data for all dent
signals obtained during the preservice examination were reviewed to ascertain whether similar
anomalous dent signals existed. In addition, rotating probe inspections were performed at
dents whose voltages exceeded a specific voltage (e.g., 2 to 5 volts) if these dents had not
been inspected with a rotating probe during the preservice inspection. Based on these efforts, one additional tube was identified with an anomalous signal, but was not conclusively similar to
the other indication with respect to the vertical presentation of the eddy current signal. This
tube was plugged during the preservice examination because the dent obstructed the passage
of the normal-sized bobbin probe and there was a concern regarding the future inspectability of
this location.
Upon identification of the leaking tube, additional efforts were made to determine the root cause
of the leak. These efforts included reviewing steam generator manufacturing records and
developing mock-up specimens to simulate the anomalous eddy current signal in the leaking
tube.
During the review of the manufacturing records of the steam generator, it was determined that
one tube was scrapped during the fabrication of the replacement steam generators since it was
damaged (or pierced) by a packing screw. Screws are used in the packing crate in which the
tubes are shipped from the tubing manufacturer to the steam generator fabrication facility. The
affected portion of this tube was sent back to the tubing manufacturer and corrective actions
were taken; however, at the time of the discovery of this damaged tube, all of the tubes in one
of the Unit 2 steam generators were installed and the other steam generator was in the process
of being fabricated.
To simulate the anomalous dent signal in the leaking tube, a series of dents was fabricated in a
mock-up facility. The simulation included impact dents from a nail, a screw, and a drill bit. A
wood screw, similar to that used in the tube manufacturers crate, was driven through a piece of wood and into the sample tube. Eddy current testing was performed on these specimens and
the damage caused by the wood screw yielded a similar anomalous signal to that found in the
leaking tube at Palo Verde.
During its formal root cause evaluation, the licensee for PVNGS Unit 2 confirmed that the tube
packing crate used wood spacers and cross brace materials that were assembled using
common screws as the tubes were loaded into the crate. The design of this packing material
placed the screws in close proximity to specific locations on some tubes, and the location, shape, and size of the deformation in the leaking tube are consistent with damage that would
occur if a screw penetrated completely through the packing material and came in contact with
the tube.
As a result of the findings, the licensee took many corrective actions, including performing
inspection of selected tubes, plugging and stabilizing the leaking tube, adding additional quality
control inspectors at the steam generator fabrication facility (since replacement steam
generators for Unit 1 are being fabricated at this facility), modifying the receipt inspections
performed (including procedural changes) on the tubes at the fabrication facility, evaluating/
modifying the packing procedure/design, identifying the tubes that were shipped in package
locations where packing screw damage was possible, and initiating additional mock-up testing
to improve the capability to identify and characterize volumetric flaws located within a dent (e.g.,
puncture-type defects).
After concluding that there was reasonable assurance of tube integrity, the licensee returned
the plant to service. The primary-to-secondary leak rate following startup was near the
detection threshold (i.e., less than 0.1 gpd (0.4 lpd)). In addition, following plant startup, six
additional tubes were found at the fabrication facility during the unpacking of tubes for the Palo
Verde Unit 1 replacement steam generators that had been pierced by a packing crate screw.
These tubes were not installed in any of the steam generators being fabricated.
Discussion:
Steam generators have been replaced at many U.S. plants, and a number of other plants plan
to replace in the next several years.
The finding of tubes damaged during the fabrication of the Palo Verde replacement steam
generators illustrates the importance of monitoring the fabrication process. This includes the
packing procedures for the tubes and the receipt inspections performed on these tubes once
they arrive at the steam generator fabrication facility.
In addition, the findings at Palo Verde illustrate the importance of fully evaluating the
implications of all abnormal conditions (i.e., conditions adverse to quality) identified during the
fabrication process and communicating these results to all affected parties within an
organization. In this instance, the personnel performing the preservice inspection at Palo Verde
were not specifically notified of the identification of a tube that been damaged by a packing
screw. As a result, they did not consider the potential for this type of flaw to exist in their review of the inspection data. By communicating non-conforming conditions observed during
fabrication to the individuals responsible for the preservice examination, nondestructive
examination techniques can be selected and the personnel trained to ensure potential defects
are reliably detected and evaluated.
The findings at Palo Verde also illustrate the inspection challenges in finding flaws (such as
from a screw) when they are located within a dent. These inspection challenges include
determining the appropriate voltage threshold at which rotating probe examinations should be
performed on a dent to detect flaws and determining the capability of the rotating probe to
reliably identify flaws (e.g., holes) in a dent.
Lastly, the findings at Palo Verde indicate that the source of small amounts of primary-to- secondary leakage from volumetric defects can be determined through secondary side
pressure tests.
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 the technical contact listed
below or the appropriate project manager in the NRCs Office of Nuclear Reactor Regulation
(NRR).
/RA/
Terrence Reis, Acting Chief
Reactor Operations Branch
Division of Inspection Program Management
Office of Nuclear Reactor Regulation
Technical Contacts: Charles Paulk, Region IV
Kenneth Karwoski, NRR
817-860-8236
301-415-2752 E-mail: cjp@nrc.gov
E-mail: kjk1@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
See previous concurrence
OFFICE
OES:IROB:DIPM
EMCB:DE
TECH EDITOR
RIV
EMCB:DE
NAME
BRini*
KKarwoski*
BAR/for/*
CPaulk*
Alund*
DATE
06/30/2004
06/23/2004
06/30/2004
06/23/2004
06/23/2004 OFFICE
D:DE
C:EMCB:DE
A:SC:OES:IROB:DIPM
A:C:IROB:DIPM
NAME
RBarrett*
WBateman*
AWMarkley*
TReis
DATE
07/09/2004
06/29/2004
07/29/2004
08/03/2004
______________________________________________________________________________________
OL = Operating License
CP = Construction Permit
Attachment LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information
Date of
Notice No.
Subject
Issuance
Issued to
_____________________________________________________________________________________
2004-15 Dual-Unit Scram at Peach
Bottom Units 2 and 3
07/22/2004
All holders of operating licenses
for nuclear power reactors except
those who have permanently
ceased operation and have
certified that fuel has been
permanently removed from the
reactor vessel.
2004-14
Use of less than Optimal
Bounding Assumptions in
Criticality Safety Analysis at
Fuel Cycle Facilities
07/19/2004
All licensees authorized to
possess a critical mass of special
nuclear material.
2004-13
Registration, Use, and Quality
Assurance Requirements for
NRC-Certified Transportation
Packages
06/30/2004
All materials and
decommissioning reactor
licensees.
2004-12
Spent Fuel Rod Accountability
06/25/2004
All holders of operating licenses
for nuclear power reactors, research and test reactors, decommissioned sites storing
spent fuel in a pool, and wet
spent fuel storage sites.
2004-11
Cracking in Pressurizer Safety
and Relief Nozzles and in
Surge Line Nozzle
05/06/2004
All holders of operating licenses 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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