United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 94-43: Determination of Primary-to-Secondary Steam Generator Leak Rate

UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C.  20555

June 10, 1994


NRC INFORMATION NOTICE 94-43:  DETERMINATION OF PRIMARY-TO-SECONDARY STEAM
 GENERATOR LEAK RATE


Addressees

All holders of operating licenses or construction permits for
pressurized-water reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert licensees to problems that may exist in methods used to
determine steam generator primary-to-secondary leak rates that could lead to
the calculation of unconservative leak rates.  It is expected that recipients
will review the information for applicability to their facilities and consider
actions, as appropriate, to avoid similar problems.  However, suggestions
contained in this information notice are not NRC requirements; therefore, no
specific action or written response is required.

Description of Circumstances

In July 1992, the licensee for Palo Verde Nuclear Generating Station Unit 2
(Palo Verde 2) noted a small steam generator tube leak by means of detectable
levels of activity in the secondary system.  By early February 1993, the
licensee observed an increasing trend in steam generator blowdown radiation
monitor activity.

On March 4, 1993, primary-to-secondary leakage in steam generator no. 2
increased rapidly.   At the same time, there was a small increase of 275 to
345 kilopascal (kPa) [40 to 50 psi] in reactor coolant system pressure because
of a charging pump surveillance test.  Using samples of steam generator
blowdown water analyzed for iodine-131 (I-131), the licensee estimated
primary-to-secondary leak rates of approximately 38 liters per day (lpd)
[10 gallons per day (gpd)].  Retrospective calculations based on readings from
the condenser vacuum exhaust radiation monitor and a xenon-133 (Xe-133) gas
grab sample, indicated that the leak rate had spiked to approximately 397 lpd
(105 gpd), gradually decreased, and stabilized at approximately 76 lpd
(20 gpd) two days later.

On March 14, 1993, while operating at 98-percent power, Palo Verde 2
experienced a steam generator tube rupture (SGTR) causing a
primary-to-secondary leakage rate of approximately 908 liters per minute
(lpm) [240 gallons per minute (gpm)].  Before the tube rupture event, the
licensee had been measuring leak rates of approximately 15 to 38 lpd (4 to 10
gpd) in

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Unit 2 steam generator no. 2 using samples from the blowdown line.  An NRC
augmented inspection team (AIT) was dispatched to the site to review the
event.  This event is further described in NRC Information Notice 93-56 dated
July 22, 1993.

Before the SGTR event, the licensee had used several different methods to
calculate primary-to-secondary leakage.  The method the licensee used most
often involved measuring radionuclide activity (usually I-131) in the steam
generators using samples from the blowdown line.  Station procedures did not
specify which method to use or require a verification of the leak rate by
performing a comparison.

On the basis of earlier tests, Combustion Engineering (CE) had concluded that
the steam generator blowdown samples were diluted by a factor ranging from
5 to 10 by feedwater "spilling over" the center divider plate in the steam
generator.  CE sent this information to the licensee in a letter dated
December 10, 1992.  The CE letter stated that the downcomer samples were more
representative of steam generator bulk water than the blowdown samples.

The NRC staff reviewed the accuracy of calculating primary-to-secondary
leakage using samples from the blowdown line and concluded that this method
was inaccurate and had caused the licensee to underestimate the actual leakage
by as much as a factor of ten.  The underestimation occurred because feedwater
had diluted the steam generator blowdown samples.

Discussion

Failure to recognize the potential problems associated with the methods used
to calculate primary-to-secondary leak rates can lead to unconservative
estimates of the leak rate.  Early indications of a primary-to-secondary leak
can be obtained from several different locations, including the condenser
off-gas radiation monitors, main steam line radiation monitors (particularly
those sensitive to N-16), and chemistry samples from the secondary side of
the steam generator.  It is important to understand the limitations of any
method used in order to take appropriate actions to mitigate the consequences
of a tube leak or rupture.

The two most common methods for determining primary-to-secondary leak rates
are (1) sampling the secondary side of the steam generator for iodine and
(2) sampling feedwater for tritium.  Limitations associated with these methods
include hideout of the iodine within the steam generator, sample dilution by
incoming feedwater, changes to blowdown rate, and unaccounted tritium losses
in the secondary system.  Hideout occurs when iodine from the primary coolant
selectively concentrates in structures within the steam generator after the
iodine transfers to the secondary coolant.  In System 80 steam generators
designed by CE, feedwater enters the steam generator on the cold-leg side and
is prevented from contacting the hot-leg side by a center divider plate.
However, the potential exists for feedwater to spill over the divider plate,
enter the hot-leg blowdown line, and dilute the sample.
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Using tritium to calculate primary-to-secondary leak rates involves a simple
mass balance of tritium in the secondary plant.  Difficulties arise when
significant tritium loss mechanisms are either not accounted for or are not
well understood.  One of these mechanisms is the loss of tritium in water
vapor through the condenser off-gas exhaust and leaking seals.

Additionally, methods of leak rate determination can lag significantly behind
the actual leak rate or can vary during power transients due to hideout
return.  The lag time is due to the inherent time delay in the circulation of
the water from the steam generator to the sample point.

Main steam line monitors designed for detecting N-16 are able to discriminate
between N-16 and other radiation and respond relatively rapidly in real time
to a steam generator tube leak.  These monitors also have a continuous readout
in the control room, making it possible to track and trend the monitor
response.

Another method for detecting primary-to-secondary steam generator tube leakage
is through the use of on-line condenser off-gas monitors.  When such monitors
are in service, they respond very rapidly to radiation from noble gases
associated with primary-to-secondary steam generator leaks.  The response of
off-gas monitors can be continuously observed and their alarm setpoints can be
set to quickly respond even for small leaks, allowing operators to respond to
the alarm and take appropriate corrective actions.  Samples of the condenser
off-gas can be collected and subsequently analyzed to assist with leak rate
calculations; krypton (Kr) and xenon (Xe) are generally used for this purpose.
The common choice of isotope for this method is Xe-133.  The accuracy of this
method is generally unaffected by the limitations that affect the other
methods.

On August 22, 1993, the operators at McGuire Nuclear Station Unit 1, shut down
the unit based on the early indications of a primary-to-secondary leakage.
While the plant was operating at full power, operators received radiation
monitor alarms from steam line nitrogen-16 (N-16) detectors and the condenser
air ejector radiation monitor.  Chemistry samples confirmed
primary-to-secondary leakage of 666 lpd (176 gpd) from steam generator A.
Based on the licensee's administrative limit of 190 lpd (50 gpd) leakage in
any steam generator, the licensee shut down the plant.  Subsequent licensee
testing revealed that there were eight tubes leaking with leak rates ranging
from 7 lpd to 700 lpd (2 gpd to 185 gpd).  The 700 lpd (185 gpd) leak was
coming through the upper joint of a sleeved tube.  This illustrates an
appropriate use of indications from one method to detect and other methods to
quantify a primary-to-secondary steam generator leak.  NRC Information Notice
(IN) 94-05, "Potential Failure of Steam Generator Tubes with Kinetically
Welded Sleeves," provided further details on this event.  IN 91-43, "Recent
Incidents Involving Rapid Increases in Primary-to-Secondary Leak Rate," and
IN 88-99, "Detection and Monitoring of Sudden and/or Rapidly Increasing
Primary-to-Secondary Leakage," also discuss events where condenser air
ejector monitors detected abnormal steam generator leakage.

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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.

/S/'D BY BKGRIMES


                            Brian K. Grimes, Director
                            Division of Operating Reactor Support
                            Office of Nuclear Reactor Regulation

Technical contacts:  Thomas Koshy, NRR
                 (301) 504-1176

                 James Reese, RIV
                 (510) 975-0237

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