Information Notice No. 84-35: BWR Post-scram Drywell Pressurization

                                                            INS No.:  6835 

                                UNITED STATES
                           WASHINGTON, D.C. 20555 

                               April 23, 1984 



All boiling water power reactor facilities holding an operating license (OL)
or construction permit (CP). 


This information notice is provided as a notification of events that 
resulted in drywell pressure increases following a reactor scram and the 
subsequent unavailability of systems that could be used to reduce drywell 
pressure. We expect recipients of this notice will review the information 
for applicability to their facilities and consider actions, if appropriate, 
to preclude a similar problem occurring at their facilities. However, 
suggestions contained in this information notice do not constitute NRC 
requirements and, therefore, no specific action or response is required. 

Description of Circumstances: 

Edwin I. Hatch Plant Unit 2 

In August 1982, Hatch Nuclear Plant Unit 2 sustained a reactor scram and 
group 1 isolation from full power conditions when an inboard main steam 
isolation valve (MSIV) failed closed. After resetting the scram, the opera-
tors controlled vessel level and pressure with the high pressure coolant 
injection (HPCI) system and the safety relief valves (SRVs). On opening the 
"A" SRV a second time the operators noticed that drywell pressure was in-
creasing rapidly. When the drywell pressure reached 2.0 psig a second 
reactor scram and loss of coolant accident (LOCA) signal were initiated. The 
scram signal reopened the scram inlet and outlet valves while the LOCA 
signal initiated the emergency core cooling system and shut off the drywell 
coolers. Drywell pressure continued to rise and peaked at about 4.0 psig. 

Some time after the second scram signal was received, it was discovered that
the scram discharge volume (SDV) drain line isolation valve had not fully 
closed. With the sustained reactor scram signal, the reactor water continued
to flow out of the open scram valves, through the SDV drain, and into the 
reactor building equipment drain (RBED) sump where the hot pressurized 
reactor water flashed to steam. The steam then flowed back through open 
drains into the reactor building causing a high area temperature trip of the 
reactor core isolation cooling (RCIC) system and elevated air temperatures 
in other parts of the reactor building. 


                                                            IN 84-35 
                                                            April 23, 1984 
                                                            Page 2 of 3 

The only immediate way of isolating the flow of reactor water through the 
scram valves was by resetting the reactor scram signal, which would close 
the scram valves. This could not be readily accomplished because the 2.0 
psig drywell pressure scram signal was still present. The operator 
unsuccessfully attempted to reduce the drywell pressure with the 
ventilation, system and decided that personnel would have to restore the 
drywell coolers to operation in order to rapidly reduce the drywell 
pressure. To accomplish this, an electrical technician had to be sent to the 
equipment cabinets to bypass the trip signal for the drywell coolers since 
no control room trip override switch was provided on Unit 2. (Such a switch 
does exist at Hatch Unit 1). 

During this event it took nearly one hour and 45 minutes from the time the 
coolers tripped to the time they could be restored. It took nearly another 
hour before the coolers reduced the drywell pressure below 2 psig, at which 
time the scram could finally be reset and the steam release to the reactor 
building terminated. 

Quad Cities Station Unit 2 

In June 1982, the Unit 2 reactor scrammed from approximately 95 percent 
power. The reactor scrammed from a low water level signal following the trip
of a reactor feed pump due to the loss of one of two major 4 kV buses. Prior
to the trip of the turbine generator, and the subsequent loss of the other 
major 4 kV bus, the operator ran the vessel level up to the high level trip 
point of the feedwater pumps. The operator then controlled reactor pressure 
by intermittently opening main steam relief valves. Approximately 30 minutes
after the scram the drywell pressure exceeded 2 psig and a second scram sig-
nal was received along with an emergency core cooling system (ECCS) initia-
tion signal and the trip of the drywell coolers and the reactor building 
closed cooling water (RBCCW) pumps. 

Both 4 kV buses were returned to service 40 minutes after the trip and sup-
pression pool cooling was established 50 minutes after the trip. The trip 
signal to the drywell coolers and RBCCW pumps was bypassed about l hour and 
35 minutes after the trip and the drywell pressure was reduced to below 2 
psig. Once the drywell pressure was reduced with the drywell coolers, the 
second scram signal could be reset. 


During the Hatch event it is believed that the high drywell pressure was 
caused by the "A" SRV tailpipe vacuum breaker sticking in the open or par-
tially open position (see Information Notice No. 83-26, "Failure of Safe-
ty/Relief Valve Discharge Line Vacuum Breakers"). The high drywell pressure 
experienced during the Quad Cities event was apparently caused by leaky gas-
kets on the flanged elbows in the relief valve discharge lines. Such 
bypasses or leakage pathways in relief valve piping may reduce the 
effectiveness of the pressure suppression system and are of themselves 
significant events. 

This Information Notice is focussed on the subsequent difficulties in making
equipment available to reduce drywell pressure to normal following leakage 


                                                            IN 84-35 
                                                            April 23, 1984 
                                                            Page 3 of 3 

from SRV piping. During both events, systems normally used to reduce drywell
pressure were tripped by a high drywell pressure condition and could not be 
readily reset. Considerable time was spent before drywell pressure was re-
duced because electrical jumpers had to be installed in order to bypass the 
trip signals. 

At Hatch Unit 2, no design changes were made to the drywell chiller trip 
logic to retain drywell cooling in the presence of a LOCA signal. Instead, 
the licensee's corrective action included training of site personnel on 
bypassing signals in general along with operator training on the functions 
of systems that would allow signals to be bypassed, if needed, on an 
emergency basis. 

At Quad Cities Unit 2 the high drywell pressure trip logic has since been 
modified to allow for continuous operation of the drywell coolers and the 
RBCCW pumps when normal power is available to the emergency buses. 

A more detailed description of the event and corrective actions taken after 
the event at Hatch Unit 2 is discussed in Power Reactor Events, Volume 5, 
No. 4, Nuclear Regulatory Commission, January, 1984. 

Because of the potential seriousness of this type of event, licensees may 
wish to consider design changes to the high drywell pressure logic to 
prevent tripping the drywell coolers when offsite power is available, and/or 
provide convenient override arrangements (e.g., switches) to permit rapid 
restarting of drywell coolers when a high drywell pressure condition still 
exists. Before making such design changes, appropriate analyses should be 
performed in consultation with NRC. 

No written response to this notice is required. If you have any questions 
regarding this matter, please contact the Regional Administrator of the 
appropriate NRC Regional Office or this office. 

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

Technical Contact:  Eric W. Weiss, IE 
                    (301) 492-4973 

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