Information Notice No. 85-72: Uncontrolled Leakage of Reactor Coolant Outside Containment

                                                           SSINS NO.: 6835 
                                                            IN 85-72       

                                UNITED STATES
                           WASHINGTON, D.C. 20555

                               August 22, 1985

                                   OUTSIDE CONTAINMENT 


All boiling water reactors holding an operating license (OL) or a 
construction Permit (CP). 


This information notice is provided to alert recipients of a significant 
event involving an uncontrolled primary coolant leak outside containment. It
is expected that recipients will review the information for applicability to
their facilities and consider actions, if appropriate, to preclude similar 
events from occurring at their facilities. However, suggestions contained in
this information notice do not constitute requirements; therefore, no 
specific action or written response is required. 

Description of Circumstances: 

On June 12, 1985, a reactor scram occurred from 99% power at Oyster Creek. 
Nuclear Generating Station. The scram occurred following failure of the 
electric pressure regulator that subsequently caused a turbine bypass valve 
to open. This resulted in a reactor pressure decrease to the low pressure 
trip set point, causing the main steam isolation valves (MSIVs) to shut and 
the reactor to scram. 

As part of the scram sequence, the scram discharge volume (SDV) vent and 
drain valves are required to shut to contain the water released during a 
scram. However, in this event, the two drain valves did not fully close, 
allowing hot, reactor coolant to drain to the reactor building equipment 
drain tank. The hot fluid flashed in the drain system creating steam that 
flowed up through various drains in the 51-ft and 23-ft building levels. The
steam combined with the fumes from the building paint on the SDV drain 
piping and caused a portion of the reactor building deluge fire protection 
system to actuate and spray down the 51-ft level. Approximately 500 gal of 
reactor coolant flowed to the drain tank before the scram system could be 
reset, which took approximately 38 minutes. The fire protection deluge 
system actuated approximately 20 minutes after the scram and was shut off in
approximately 5 minutes. No safety equipment inside the reactor building was
adversely affected by actuation of the deluge system. 


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The failure of the SDV drain valves to properly close caused the following: 

1.   Uncontrolled reactor coolant leakage outside containment. 

2.   Temperatures of the control rod drive (CRD) seals exceeding the alarm 

3.   Actuation of the reactor building fire protection deluge system. 

4.   Radioactive contamination of the 23-ft level of the reactor building. 

Each SDV drain valve that failed had a different failure mechanism. The 
upstream valve stem/disc travel stopped approximately 1/8 inch before fully 
seating onto the valve seat. This was caused by the valve actuator not 
having the stroking length properly adjusted. The downstream valve had an 
improperly sized spring in the valve actuator. It is believed that the valve 
initially closed, but was then forced open when the system pressure exerted 
a force below the valve seat that exceeded the spring closing force of the 

The high CRD seal temperature alarms were received intermittently after the 
scram. The alarms are an indication of abnormal flow of reactor coolant 
within or out of the CRD system. Degradation, or possibly failure, of the 
seals could occur following prolonged exposure at elevated temperatures. As 
a result, abnormal leakage could occur that might adversely affect proper 
rod motion or rod scramming ability. 

The reactor building fire protection system actuated on the 51-ft level of 
the reactor building. Although no equipment was adversely affected by the 
deluge system spray the potential existed for damaging electrical equipment 
and possibly aggravating an already serious problem. 

Although the SDV vent and drain valves are stroke tested monthly in 
accordance with the inservice testing (IST) program, there were no criteria 
or requirements, for leak testing these valves. Following the initial 
installation of the downstream valve as part of a system backfit in 1984, no
postinstallation leak rate test of either valve against operating pressure 
was conducted. Both valve problems could have been detected by such a test. 

Information Notice No. 84-35, "BWR Post-Scram Drywell Pressurization" 
described an event of August 1982 at the Hatch Nuclear Plant Unit 2 where 
there was a similar leakage from the SDV. That event was also the subject of
an AEOD case study and was included in the 3rd quarter, 1983, "Report to 
Congress on Abnormal Occurrences." 

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                                                            August 22, 1985 
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No specific action or written response is required by this information 
notice. If you have any questions regarding this matter, please contact the 
Regional Administrator of the appropriate NRC regional office or this 

                                   Edward L. Jordan Director 
                                   Division of Emergency Preparedness 
                                     and Engineering Response 
                                   Office of Inspection and Enforcement 

Technical Contact:  David Powell, IE
                    (301) 492-8373

Attachment: List of Recently Issued Information Notices 

Page Last Reviewed/Updated Friday, May 22, 2015