Information Notice No. 85-72: Uncontrolled Leakage of Reactor Coolant Outside Containment
SSINS NO.: 6835
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
August 22, 1985
Information Notice No. 85-72: UNCONTROLLED LEAKAGE OF REACTOR COOLANT
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.
August 22, 1985
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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."
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
Attachment: List of Recently Issued Information Notices
Page Last Reviewed/Updated Friday, May 22, 2015