United States Nuclear Regulatory Commission - Protecting People and the Environment

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. 
.

Information Notice No. 79-27                          November 16, 1979 
                                                            Page 2 of 9 

                                                                 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.
.

Information Notice No. 79-27                          November 16, 1979 
                                                            Page 3 of 9 

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. 
.

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) 
.

Information Notice No. 79-27                          November 16, 1979 
                                                            Page 5 of 9 

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 
.

Information Notice No. 79-27                          November 16, 1979 
                                                            Page 6 of 9 

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 
.

Information Notice No. 79-27                        November 16, 1979 
                                                       Page 7 of 9 

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. 
.

Information Notice No. 79-27                          November 16, 1979 
                                                            Page 8 of 9 

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 
.

Information Notice No. 79-27                          November 16, 1979 
                                                            Page 9 of 9 

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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