Information Notice No. 79-27, Steam Generator Tube Ruptures at Two PWR Facilities
IN79027
November 16, 1979
MEMORANDUM FOR: B. H. Grier, Director, Region I
J. P. O'Reilly, Director, Region II
J. G. Keppler, Director, Region III
K. V. Seyfrit, Director, Region IV
R. H. Engelken, Director, Region V
FROM: Norman C. Moseley, Director, Division of Reactor
Operations Inspection, Office of Inspection and
Enforcement
SUBJECT: Information Notice No. 79-27, STEAM GENERATOR TUBE
RUPTURES AT TWO PWR FACILITIES
The enclosed Information Notice No. should be dispatched November 16, 1979,
to all power reactor facilities holding operating licenses and construction
permits.
Norman C. Moseley, Director
Division of Reactor Operations
Inspection
Office of Inspection and Enforcement
Enclosures:
1. Draft Transmittal Letter
2. Information Notice No. 79-27
CONTACT: H. A. Wilber, TP
49-28180
.
(Draft letter to all power reactors facilities with operating licenses and
construction permits.)
Information Notice No. 79-27
Gentlemen:
The enclosed Information Notice No. provides information with regard to the
sequence of events that followed incidents involving steam generator tube
ruptures at two PWR units.
Sincerely,
Signature
(Regional Director)
Enclosures:
1. Information Notice No. 79-27
2. Recently Issued IE
Information Notices
.
SSINS No.: 6870
Accession No.:
7910250488
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON D.C. 20555
November 16, 1979
Information Notice No. 79-27
STEAM GENERATOR TUBE RUPTURES AT TWO PWR PLANTS
Description of Circumstances:
In recent months two incidents involving steam generator tube ruptures have
occurred. In both instances, the units were cooled down and placed in the
residual heat removal mode with existing procedures.
Event of June 25, 1979 at the Doel 2 Nuclear Power Plant in Belgium
The first event occurred on June 25, 1979, at the Doel 2 nuclear power plant
in Belgium. The Doel unit is a 390 Mwe Westinghouse two-loop reactor. The
event consisted of a rupture of several tubes in the loop B steam generator.
The resultant leakage between the primary and secondary systems was
estimated to be 125 gpm. The event started when the plant was heated up
after a shutdown caused by a malfunction of the main steam isolation valve.
At the time of the incident the primary coolant pressure was: 2233 psi and
the temperature: 491F. The reactor remained subcritical throughout the
event.
The first indication of abnormal behavior was a rapid decrease of the
primary system pressure (approximately: 28 psi/min.). This was followed by
the sequence of events listed below:
Time, min.
1. Increase of charging flow demand, requiring startup of a 1.8
second charging pump.
2. Automatic isolation of the CVCS letdown line. 2.4
3. Shut off of the pressurizer heaters due to low liquid level 2.4
in the pressurizer.
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Time. min.
4. Closing of block valves in the pressurizer relief line. 4.6
5. Rapid increase of water level in the damaged steam 9.4
generator (loop B). The steam generator was isolated.
6. Startup of the third charging pump and realignment of the
suction of all charging pumps from the CV tank to the
refueling water storage tank.
7. Shut off of the main coolant pump in loop B. This was done 17.4
in order to reduce heat generation in the primary coolant system.
8. Safety Injection Signal on low pressure in pressurizer 19.2-19.5
followed by: startup of diesels, containment isolation,
and high pressure safety injection, resulting in increase of
the primary system pressure.
9. Manual startup of the pressurizer spray in an attempt to 28
decrease primary system pressure.
10. Pressurizer fills up solid with water. Level indicator off 33
scale. There was no release of primary coolant from the
pressurizer because the block valve was closed and the
pressurizer did not exceed safety valve settings.
11. Automatic startup of auxiliary feedwater flow to both 44
steam generators.
12. Flow of auxiliary feedwater to the damaged steam generator 50
is stopped.
13. Beginning of depressurization of the primary coolant system. 60-88
SI pumps are stopped and the isolation valves in the CV
letdown line are opened.
14. Startup of the residual heat removal system. 195
Discussion
The operator's action during the accident were directed towards:
a. maintaining primary coolant subcooled,
b. minimizing leakage rate between the primary and secondary coolant
system.
c. preventing radioactive fluid from escaping from the damaged steam
generator.
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Information Notice No. 79-27 November 16, 1979
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Sufficiently high degree of subcooling in the primary coolant system was
achieved by reducing heat generation in the primary system (switching off
one (1) main coolant pump "B") and by controlling, to the extent possible,
primary coolant pressure.
Two actions were taken to prevent radioactive fluid from escaping from the
leaky steam generator. As soon as the leak was detected, the secondary side
of the steam generator was isolated and the setpoints of the safety valves
were raised to their maximum value.
In general, the accident was handled in accordance with the existing
procedures and no radioactive releases or equipment damage was experienced.
All safety systems functioned as designed with exception of the air operated
valves in the CV letdown line and in the line to the cooling system of the
main pump thermal shields. The cause of this problem was that the
containment isolation signal interrupted the supply of compressed air to
these valves and rendered them inoperative until the air was manually
restored. This malfunction of the valves resulted in a delay of primary
system cooldown and depressurization (item 13) and caused the primary
coolant pumps to operate for a while without proper cooling. However, none
of these events produced any detrimental consequences.
Conclusions
The accident was successfully terminated using the presently existing
procedures which, with only one exception, proved to be adequate. In the
future, the procedure dealing with containment isolation will have to be
revised.
The leak was reported to be located in the U-Bend of the first row tubes.
The suspected cause was stress corrosion due to ovalization of the short
bend radius tubes.
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Information Notice No. 79-27 November 16, 1979
Page 4 of 9
Event of October 2, 1979 at the Prairie Island 1 Nuclear Power Plant
The second event occurred on October 2, 1979 at Prairie Island Nuclear
Generating Plant Unit No. 1, a 530 Mwe Westinghouse two-loop reactor. The
event consisted of mechanical wear due to a foreign object until a tube
failure occurred in the "A" Steam Generator; the resultant leakage was
calculated to be about 390 gpm. At the time of the incident, the plant was
operating at 100% power. The following information was taken from the
licensee's event report No. 79-27 dated October 16, 1979 and from NRC
inspections of the event.
Date Time (CDT) Event
Oct 2 1414 High Radiation alarm on the air ejector discharge
gaseous radiation monitor
1420 Overtemperature T Turbine Runback due to
decreasing pressure (Maximum rate was approximately
100 psi/minute.)
1421 Low Pressurizer pressure (< 2139.9 psig)
1421 (approx) Commenced load reduction
1422 Low pressurizer level (< 18.3%)
1423 Started second charging pump (#11)
1424 (approx) Started third charging pump (#13)
1424:09 Reactor trip for "Low Pressurizer Pressure" (< 1900
psig)
1424:14 Safety injection (SI) occurred due to "Low
Pressurizer Pressure (< 1815 psig)
1424:33 Minimum RCS water inventory; RCS pressure begins
increasing
1426 11 Reactor Coolant Pump stopped
1427 12 Reactor Coolant Pump stopped
1430 Emergency Alert declared
1432:29 11 Steam Generator level increased above the "Lo Lo
Level" setpoint (13%) on the narrow range after
having gone offscale low after the trip (It is
normal for SG Level to go offscale low on a trip;
recovery in this case was much more rapid than
usual)
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Information Notice No. 79-27 November 16, 1979
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Date Time Event
1438 SI Reset
1441 Loop A MSIV closed to isolate No. 11 Steam
Generator
1456 Pressurizer Level returned on scale
1456 Stopped 12 SI pump
1456-57 Began depressurization of the RCS using the
pressurizer PORV. (The valve was cycled 6 to 8
times to reduce pressure to required value)
1500 (approx) Site Emergency declared
1502 Pressurizer level reached the high level setpoint
(> 55%)
1506 11 SI Pump stopped
1507 Pressurizer Relief Tank rupture disc relieved
1515 RCS pressure at 910 psig (same as 11 SG pressure )
Leak apparently stopped
1550 Commenced normal cooldown
2200 Site Emergency terminated.
Oct 3 0640 RHR placed in service to continue cooldown to cold
shutdown
1300 RCS at cold shutdown
The radiological aspects of the event are summarized below:
RADIOACTIVE RELEASED FROM THE PLANT
Airborne
The monitor on the exhaust of the steam jet air ejectors (SJAE) alarmed at
1514 hours EDT about 10 minutes prior to the reactor trip. The monitor was
off-scale
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shortly thereafter; the highest range of the monitor is equivalent to
approximately 0.004 Ci/sec release rate at an exhaust flow of about 20 cfm.
The monitor was thought to have been filled with water.
Based on the initial full-scale reading of the SJAE monitor, and analysis of
several grab samples taken from the SJAE exhaust, it is estimated that
approximately 30 curies of noble gases (primarily xenon) were released
throughout the incident with the majority of the release being within the
first 2 hours. No iodine levels were measured.
The airborne releases do not appear to have exceeded the applicable
Technical Specification limit (120 Ci/Hr) on maximum allowable release rate
averaged over an hour period. The release rate decreased after the isolation
of the steam generator, continuing to decrease with time. After the first
hour the release rate was ~ 0.002 Ci/sec and was in the range of 2-500
Ci/sec after the second hour.
Liquid
Analysis of samples of water from the turbine building sumps showed only one
isotope detectable, Xe-133 at the concentration of ~ 5 x 10-5 uCi/ml. During
the course of the incident, water was pumped from the sumps for offsite
release at a rate of about 250 gallons per minute for approximately 3
minutes, resulting in a total release of about 140 uCi of noble gases
(Xe-133) dissolved in water. No regulatory limits were exceeded for this
release, considering an MPC of about 2 X 10 -4 uCi/ml normally used for
noble gases dissolved in water.
OFFSITE RADIOLOGICAL IMPACT
During the first 4 hours after the steam generator tube rupture, the winds
were blowing generally from the east to the west. Using site meterological
data, the dispersion factor (X/Q) at the site boundary was estimated to be 4
X 10 -5
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sec/M3. Conservatively assuming the total estimated release of ~30 curies of
noble gases over the 4-hour period, the dose to an individual continuously
present at the site boundary would be about 0.05 millirem, slightly above
the normal background dose rate.
After the first 2 hours, the release rate had dropped to the point where
calculated dose rates offsite were well below natural background radiation
levels.
Environmental surveys were carried out by licensee and State teams operating
out to a distance of about 5 miles from the site. Air samples and direct
radiation surveys made by these survey teams yielded negative results (i.e.,
background readings). Surveys performed by the NRC inspectors at the site
confirmed the licensee and State results.
At ~ 2000 hours EDT, the State of Minnesota conducted an aerial survey over
the site at altitudes from 400 to 2000 feet. The survey detected only
background levels using a portable survey instrument (CDV-700).
RADIOACTIVITY IN THE PLANT
Area direct radiation monitors in the plant and direct radiation surveys
showed no significant increase in radiation levels.
Analysis of air samples taken in the turbine building showed concentrations
of krypton and rubidium daughters in the range of 10-10 to 10-9 uCi/cc (MPC
of 10-6 uCi/cc) and xenon at a concentration of 10-6 uCi/cc (MPC of 10-5
uCi/cc).
The direct radiation monitor in containment (instrument seal table) showed
no increase after the trip (~2 mrem/hr). The noble gas monitor in
containment increased by a factor of ~10 (from 1000 to 12,000 cpm)
indicating 3 X 10-3 uCi/cc gaseous activity in containment.
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PLANT PERSONNEL EXPOSURES
No personnel overexposures resulted from the occurrence. A total of about
200 plant contractor personnel were involved in evacuation from the site as
a result of the declaration of a site emergency condition. These personnel
were working in the auxiliary building and turbine building. All personnel
had been "badged" with personnel monitors and were surveyed for
contamination before they departed the site.
Cause of Event
Licensee examination of the steam generator tube determined that a single
tube (out of 3388 in the steam generator) had ruptured. The size of the
rupture was 2 inches long and 3/8 inches wide in the wall of the 7/8-inch
diameter tube.
Plant personnel found a coil spring lodged near the ruptured tube. The
spring apparently had rubbed against the tube during operation, causing the
tube to wear away and eventually rupture. An adjacent tube was also worn by
the spring vibration.
The spring is believed to have been part of a hose used to loosen and remove
sludge products from the tube support sheet during an early refueling
outage.
Action Taken to Prevent Recurrence
The ruptured tube, the additional worn tube and surrounding tubes have been
plugged. The spring has been removed from the steam generator.
The licensee has completed eddy current examination of approximately 6 per
cent of the tubes in the steam generator with failed tubes and approximately
3 per cent of the second Unit 1 steam generator. Both steam generators were
examined to assure there are no other visible objects that could cause tube
damage. While in
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both events a cold shutdown was achieved with existing procedures, there was
a common concern expressed on the effects of isolating the air supplies to
valves inside containment on the maintenance of reactor coolant inventory
and pressure.
This Information Notice No. is provided as an early notification of a
possibly significant matter that is still under review by the NRC staff. It
is expected that recipients will review the information for possible
applicability to their facilities. No specific action or response is
requested at this time. If NRC evaluations so indicate, further licensee
actions may be requested or required.
No written response to this Information Notice No. is required. If you have
any questions regarding this matter please contact the Director of the
appropriate NRC Regional Office.
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Page Last Reviewed/Updated Tuesday, March 09, 2021