ML041460357

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 
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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 manufacturer’s crate, was driven through a piece of 
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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 
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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-tosecondary 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 NRC’s 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
____________________________________________________________________________________

OL = Operating License

CP = Construction Permit

Attachment

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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.

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