United States Nuclear Regulatory Commission - Protecting People and the Environment

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

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF NUCLEAR REACTOR REGULATION
                          WASHINGTON, D.C.  20555 

                             February 25, 1992 


NRC INFORMATION NOTICE 92-16:  LOSS OF FLOW FROM THE RESIDUAL HEAT REMOVAL 
                               PUMP DURING REFUELING CAVITY DRAINDOWN


Addressees

All holders of operating licenses or construction permits for nuclear power 
reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information 
notice to alert addressees to a recent event involving the loss of flow from 
the residual heat removal pump during refueling cavity draindown.  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 are not 
NRC requirements; therefore, no specific action or written response is 
required.

Description of Circumstances

On October 26, 1991, the Vogtle Electric Generating Plant, Unit 1, was in 
Mode 6 (Refueling) with the reactor vessel head removed.  The Georgia Power 
Company (the licensee) had reloaded the core and reinstalled the upper 
internals.  The licensee was using the 1B residual heat removal (RHR) pump 
to provide shutdown cooling and the 1A RHR pump to drain the refueling 
cavity by taking suction from one of the reactor coolant system (RCS) hot 
legs and discharging to the refueling water storage tank (RWST).  The RCS 
temperature was approximately 87�F.  The water level in the refueling cavity 
was at 210 feet 4 inches.  Operations personnel were preparing to lower the 
level to 192 feet, 2 feet below the reactor vessel head flange, to allow the 
reactor vessel head to be reinstalled.  The mid-loop elevation of the RCS 
for Unit 1 is 187 feet.  An assistant plant operator (APO) in the Unit 1 
containment was directed to establish a watch at a tygon tube to monitor the 
RCS level during draindown and mid-loop operations.  During the outage, the 
licensee had installed a permanent sight glass in the Unit 1 containment for 
monitoring the RCS level.  This new sight glass had neither been tested nor 
aligned for the operators to use.  The APO assumed that the new sight glass 
was operable and established communications with the control room at the 
permanent sight glass, rather than at the tygon tube, to monitor the 
draindown.  The licensee then started the draindown. 

When the day shift ended, a night shift plant equipment operator (PEO) 
relieved the day shift APO who was monitoring the permanent sight glass.  
The PEO discovered that the valves for the permanent sight glass were not 
aligned correctly.  The PEO informed the control room and the operators 
stopped the 

9202190317 
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draindown while the problem was investigated.  The PEO and APO then filled 
and vented the sight glass without using a procedure.  In their attempt to 
place the permanent sight glass in service, the upper isolation valve, which 
was not readily visible, was not opened as required. 

The licensee resumed the cavity draindown and, approximately 2 hours later, 
received a control room annunciator which indicated a high level, 192 feet 6 
inches, in the reactor vessel.  The control room operator observed that the 
control room level indicator was at the top of scale (100 percent) and 
tapped on the indicator, causing it to drop to a reading of 60 percent (190 
feet 9 inches).  The licensee again stopped the draindown. The PEO 
monitoring the sight glass level reported that reactor vessel water level 
appeared to be even with the reactor vessel head flange (194 feet), which 
agreed with the level indicated by the permanent sight glass and the 
temporary tygon tube.  The licensee assumed that the control room level 
indicator was inaccurate and continued the draindown, believing that it had 
three reliable indications of the RCS level, i.e., visual vessel water 
level, the permanent sight glass, and the temporary tygon tube. 

When the level in the RCS reached approximately 193 feet, as indicated by 
the sight glass, a control room operator observed discharge pressure, flow, 
and motor current oscillations for the 1B RHR pump, indicating that the 
coolant was forming a vortex on the suction side of the pump or that the 
pump was cavitating.  The operators closed the discharge valve for the 
1B RHR pump, thus putting the 1B RHR pump on the miniflow line.  Although 
the electrical current reading for the motor of the 1B RHR pump became more 
stable, the discharge pressure remained low.

The licensee again stopped the draindown by shutting down the 1A RHR pump 
and realigning its suction to the refueling water storage tank (RWST) to 
refill the refueling cavity.  Shortly after beginning to refill the RCS, the 
licensee noted that the discharge pressure of the 1B RHR pump began to 
improve.  When the flow of the 1B RHR pump reached approximately 2600 
gallons per minute, the licensee again observed indications of vortex 
formation or cavitation.  The licensee reduced the flow from the 1B RHR pump 
to 1800 gallons per minute and found that the pump operated satisfactorily 
with no indication of vortex formation or cavitation.  The licensee used the 
1A RHR pump to refill the refueling cavity from the RWST and stopped 
refilling when the sight glass indicated a level of 194 feet 10 inches.  The 
licensee increased the flow from the 1B RHR pump to approximately 
3000 gallons per minute and found that the pump operated satisfactorily with 
no further indication of vortex formation or cavitation.  

When operators performed a walkdown inspection of the tygon tube and the 
sight glass level indicators, they found the upper isolation valve for the 
sight glass closed with a tag on it which indicated that the new sight glass 
had not been released for use.  The licensee later determined that a similar 
tag had also been installed on the lower isolation valve but apparently had 
fallen off the valve.  
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The licensee also discovered that a high efficiency particulate absorber 
(HEPA) filter unit was connected, by means of a flexible duct, to the 
opening from which a pressurizer safety valve had been removed to provide a 
vent path for all level instrumentation.  The licensee found that the HEPA 
unit was running and the flexible duct was collapsed, apparently caused by 
the vacuum created by the running HEPA filter unit and the RCS draindown.  
This resulted in an inadequate vent path from the pressurizer.  (LER 
50-424/91-09 and NRC Inspection Report 50-424,425/91-30)

Discussion 

False high RCS level indications led to the RCS level being inadvertently 
lowered to the point at which the coolant formed a vortex in the RHR pump 
suction line.  The false high level indications were caused by an inadequate 
vent path from the pressurizer and by the closed upper isolation valve for 
the sight glass.  When conditions in the pressurizer changed, it affected 
all of the reactor vessel level instruments, because their reference legs 
connected to the pressurizer.  The system installed at Vogtle did not meet 
the intent of two independent continuous water level indications as 
discussed in Generic Letter 88-17, "Loss of Decay Heat Removal."

Procedures for the initial RCS draindown during refueling operations 
provided sufficient steps to ensure that the level instrumentation was 
installed properly and the vent paths were adequate.  However, the 
procedures for the subsequent draindowns did not include sufficient steps to 
reverify these actions.  Administrative controls were inadequate in 
addressing the reviews and documents required for attaching HEPA filter 
units to plant equipment.  In this case, the HEPA filter unit was installed 
without a temporary modification or a work order, and consequently the 
control room was not aware of the installation.

During the event, the 1B RHR pump was not available to provide recirculation 
shutdown cooling for approximately 16 minutes.  Core temperature as 
indicated at the RHR pump discharge increased from approximately 87�F to 
107�F.  There was no radiological release to the environment.  The licensee 
reviewed available data further and found that the coolant on the suction 
side of the 1B RHR pump had formed a vortex but the pump did not cavitate.  

Air may have begun entering the 1A RHR pump shortly before the pump's 
discharge valve was closed.  This resulted in a slightly reduced discharge 
pressure and flow.  The coolant in the RCS reached the lowest level, 186 to 
187 feet, when the discharge valve for the 1A RHR pump's heat exchanger was 
closed.  After the event, the licensee performed an inservice test on the 1A 
and 1B RHR pumps and found that the performance of neither pump was 
degraded. 

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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:  Doug Starkey, Region II
                     (404) 554-9901 

                     Pierce Skinner, Region II
                     (404) 331-6299

Attachment:  List of Recently Issued NRC Information Notices






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