United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 92-16, Supplement 1: Loss of Flow from the Residual Heat Removal Pump During Refueling Cavity Draindown

                               UNITED STATES 
                          WASHINGTON, D.C.  20555 

                                May 7, 1992 

                                             HEAT REMOVAL PUMP DURING 
                                             REFUELING CAVITY DRAINDOWN 

All holders of operating licenses or construction permits for nuclear power 


The U.S. Nuclear Regulatory Commission (NRC) issued Information Notice (IN)
92-16 to alert addressees to an event at the Vogtle Electric Generating 
Plant, Unit 1, on October 26, 1991, involving the loss of flow from the 
residual heat removal (RHR) pump during a draindown of the refueling cavity.  
The staff discussed inadequacies in the draindown procedure and noted a 
failure mode in which a common vent path could affect all of the level 
instruments for the reactor vessel.

The NRC is issuing this supplement to IN 92-16 to alert addressees to 
another mechanism by which licensees may experience a loss of RHR while 
conducting draindown operations.  This supplement describes an event at the 
Prairie Island Nuclear Generating Plant, Unit 2, in which RHR flow was lost 
during reactor vessel draindown.  Unanticipated level instrument 
performance, due to plant conditions not foreseen in the level instrument 
design, led to draining the reactor coolant system (RCS) further than 
planned.  This supplement also highlights the value of outage risk reduction 
efforts in mitigating the consequences of a loss of RHR event.  It is 
expected that recipients will review the information for applicability to 
their facilities and consider actions, as appropriate, to avoid similar 
problems.  However, suggestions contained in this information notice 
supplement are not NRC requirements; therefore, no specific action or 
written response is required.

Description of Circumstances

On February 20, 1992, at about 5:00 p.m., the Northern States Power Company 
(the licensee) began to drain down the RCS to the mid-loop level to install 
dams on the steam generator nozzles.  The licensee was maintaining the RCS 
temperature at 133�F with the 22 RHR pump in service.  The licensee had 
established a vent path from the RCS to the pressurizer relief tank (PRT) 
through a power-operated relief valve that was locked open.  To prevent 
drawing a vacuum on the pressurizer and PRT while draining, the licensee was 
maintaining an overpressure of 3-6 psig in the system by periodically adding 
nitrogen to the PRT.


                                                      IN 92-16, Supplement 1 
                                                      May 7, 1992 
                                                      Page 2 of 3 

The licensee was monitoring the reactor vessel level using a tygon tube 
which was vented to the containment atmosphere.  The operators were 
performing calculations to correct the observed level in the tygon tube for 
the effects of the nitrogen pressure.  The licensee also had two electronic 
level sensors installed to provide indication in the control room.  These 
instruments were designed to use a PRT pressure sensor input to 
automatically correct the sensed level for the effects of any system 
pressure.  After draining down the RCS for several hours, the operators 
began to suspect a problem with the electronic level instruments, which had 
not come on scale as anticipated.  The operators continued the draindown, 
relying solely on the corrected level in the tygon tube, while attempting to 
diagnose the suspected problem with the electronic instruments.  

Between approximately 11:00 and 11:10 p.m., the electronic instruments came 
on scale indicating a low level, the operators received indications of 
decreasing RHR flow, and they stopped the draindown.  About one minute 
later, the operators stopped the 22 RHR pump when they noticed oscillations 
in the RHR flow and the pump motor electrical current, which indicated that 
a vortex was forming in the pump suction.

The operators immediately took actions to recover the RCS level and restore 
shutdown cooling.  They increased the RCS level to the reactor vessel flange 
using both charging pumps and the 21 RHR pump to transfer water from the 
refueling water storage tank to the RCS.  They then realigned the RHR system 
for shutdown cooling using the 21 RHR pump.  The loss of forced shutdown 
cooling flow lasted approximately 21 minutes.

The level in the RCS, as measured by the electronic instruments, dropped 
approximately 8 inches below the center line of the reactor vessel nozzle.  
During the period that shutdown cooling flow was lost, the core exit 
temperature rose from 133�F to a maximum recorded value of 221.5�F.  RCS 
samples drawn following the event showed that no fuel had been damaged. 


The nitrogen pressure on the RCS was a significant contributor to this 
event.  The licensee found that the draindown procedure did not adequately 
address the capabilities and operation of the installed level 
instrumentation and the effects of the nitrogen pressure on these 
instruments.  The procedure allowed a nitrogen pressure of up to 6 psig in 
the RCS.  However, both electronic instruments were unable to provide 
accurate level indication whenever nitrogen pressure exceeded about 3.4 

An NRC Augmented Inspection Team (AIT) dispatched to the site found that the 
draindown procedure did not provide sufficient guidance on the processes 
required to achieve a stable mid-loop condition.  The procedure lacked 
guidance on verifying level instrument performance and on the accuracy 
required when correcting the observed tygon tube level for the effect of the 
nitrogen pressure.  In the absence of such guidance, the operators did not 
act conservatively and continued to drain down the RCS without determining 
the cause of the unanticipated performance of the electronic instruments.  
Furthermore, the operators had frequently been rounding the pressure values 
to the nearest whole pound per 

                                                      IN 92-16, Supplement 1
                                                      May 7, 1992 
                                                      Page 3 of 3 

square inch when calculating corrected tygon tube levels, resulting in 
corrected levels that were in error by up to about 1 foot. 

The AIT identified a design vulnerability in the level indication system in 
that the same PRT pressure instrument was used in correcting both electronic
level instruments and used in the calculation to correct the tygon tube 
level.  If not properly calibrated, this instrument would result in 
erroneous level indications for all instruments. 

The licensee had taken steps to address loss of shutdown cooling events.  
These actions included administrative requirements during reduced inventory 
operations to maintain all offsite and onsite power sources available and to
maintain sufficient secondary inventory in at least one steam generator to 
provide an alternate means of decay heat removal.  The licensee had 
developed comprehensive abnormal and emergency operating procedures to deal 
with reduced inventory events.  The operators responded effectively in 
implementing these procedures to stabilize the plant.  These measures were 
effective in minimizing the risk of plant damage due to this event. 

This information notice requires no specific action or written response.  If 
you have any questions about the information in this notice, please contact 
one of the technical contacts listed below or the appropriate Office of 
Nuclear Reactor Regulation (NRR) project manager.

                              Charles E. Rossi, Director
                              Division of Operational Events Assessment
                              Office of Nuclear Reactor Regulation

Technical contacts:  R. Schaaf, NRR
                     (301) 504-1170

                     B. Jorgensen, RIII
                     (708) 790-5689

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