United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 87-23: Loss of Decay Heat Removal during Low Reactor Coolant Level Operation

                                                      SSINS No.:  6835 
                                                        IN 87-23

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

                                  May 27, 1987


Information Notice No. 87-23:  LOSS OF DECAY HEAT REMOVAL DURING  
                                   LOW REACTOR COOLANT LEVEL OPERATION 


Addressees:

All holders of an operating license or a construction permit for pressurized-
water reactor facilities.

Purpose: 

This notice provides information regarding the loss of decay heat removal 
capability at pressurized water reactors resulting from the loss of RHR pump 
suction during plant operations with low reactor coolant levels. It is ex-
pected that recipients will review this information for applicability to their 
reactor facilities and consider actions, if appropriate, to prevent similar 
problems.  Suggestions contained in this notice do not constitute NRC require-
ments; therefore, no specific action or written response is required.

Description of Circumstances:

On April 10, 1987 the Diablo Canyon Unit 2 reactor experienced a loss of decay 
heat removal capability in both trains.  The reactor coolant system had been 
drained down to the mid-height of the hot-leg piping in preparation for the 
removal of the steam generator manways.  During the 85 minute period that the 
heat-removal capability was lost, the reactor coolant heated from 87� F to 
boiling, steam was vented from an opening in the head, water was spilled from 
the partially unsealed manways, and the airborne radioactivity levels in the 
containment rose above the maximum permissible concentration of noble gases 
allowed by 10 CFR 20.  The reactor, which was undergoing its first refueling, 
had been shut down for seven days at the time and the containment equipment 
hatch had been opened.

Erroneous level instrumentation, inadequate knowledge of pump suction 
head/flow requirements, incomplete assessment of the behavior of the air/water 
mixture in the system and poor coordination between control room operations 
and containment activities all contributed to the event.  Under the conditions 
that existed, the system that indicated the level of coolant in the reactor 
vessel read "high" and responded poorly to changes in the coolant level.  In 
addition, the intended coolant level, established for this operation, was 
later determined to be below the level at which air entrainment due to 
vortexing was predicted to commence.  At the time of the event, the plant 
staff believed that the coolant level was six inches or more above the level 
that would allow vortexing.  

8705200749
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                                                                 May 27, 1987
                                                                 Page 2 of 5


The event began at about 8:43 pm, when a test engineer in preparation for a 
planned containment penetration local leak rate test, began draining a section 
of the reactor coolant pump leakoff return line, which he believed to be 
isolated.  However, because of a leaking boundary valve, this action caused 
the volume control tank fluid to be drained through the intended test section 
to the reactor coolant drain tank.  The control room operators, who were not 
aware that the engineer had begun conducting the test procedure, increased 
flow to stop the fluid reduction from the volume control tank.  A few minutes 
later the operators were informed that the reactor coolant drain tank level 
was increasing but they could not determine the source of the leakage.  
Although the actual level of coolant in the reactor vessel was apparently 
dropping below the minimum intended level, the indication of level in the 
vessel remained within the desired control band.  At 9:25 p.m. the electrical 
current of the active RHR pump (No. 2-2) was observed to be fluctuating.  The 
2-1 pump was started and the 2-2 pump was shut down.  However, the current on 
the 2-1 pump also fluctuated, so it was immediately shut down as well.  

The operators did not immediately raise the water level in the reactor because 
they still did not know either the source of the leakage, the true vessel 
level, or the status of the work on the steam generator manways.  Operators 
were sent to vent the RHR pumps.  One pump was reported to be vented at 10:03 
p.m.  At 10:21 p.m. an attempt was made to start this RHR pump, but the 
current fluctuated and it was shut down again.  During this period the 
operators did not know the temperature of the coolant in the reactor vessel 
because the core exit thermocouples had been disconnected in preparation for 
the planned refueling.  By 10:30 p.m. airborne activity levels in the 
containment were increasing and personnel began to evacuate from the 
containment building.  

At 10:38 p.m. when the operators learned that the steam generator manways had 
not been removed, action was initiated to raise the reactor vessel water level 
by adding water from the refueling water storage tank.  About 10 minutes later 
the test engineer identified the source of the leakage and stopped it.  By 
10:51 p.m., the vessel level had been raised sufficiently to restart one of 
the RHR pumps.  The indicated RHR pump discharge temperature immediately rose 
to 220� F.  At this time the reactor vessel was slightly above atmospheric 
pressure and steam was venting from an opening in the reactor vessel head.  

Discussion:

The NRC has documented numerous instances in the past where decay heat removal 
systems have been disabled because pump suction was lost while the plant was 
being operated at low reactor coolant water levels.  Information Notice No. 
86-101 describes four such events that occurred in 1985 and 1986.  NRC Case 
Study Report AEOD/C503 describes six such events that occurred in 1984, five 
that occurred in 1983, and seven that occurred in 1982.  Information Notice No. 
81-09 described an event at Beaver Valley in March 1981. The case study report 
further indicates that a total of 32 such events occurred from 1976 through 
1984.  The documentation includes descriptions of a total of 23 events that 
have occurred since 1981 involving loss of decay heat removal capability 
resulting from a loss of pump suction while operating at reduced water levels.

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                                                                 May 27, 1987
                                                                 Page 3 of 5


For all but four of these 23 events the primary cause of the loss of pump 
suction and loss of decay heat removal capability was attributed to incorrect, 
inaccurate, or inadequate level indication.  Two events were attributed to 
loss of pump suction because of vortexing brought on by the simultaneous 
operation of both pumps.  In many of these events procedural errors were also 
a contributing factor.  In at least nine of the cases, the redundant pump was 
lost because air was entrained when the operators, not understanding the cause 
of the problem, switched to the second pump.  There are repeated references to 
difficulties in getting the pumps vented quickly after air binding had 
occurred and to the operators' inability to take immediate action to raise 
reactor vessel levels until the safety of personnel working on the primary 
systems could be assured.  The length of time that decay heat removal was 
completely lost varied from eight minutes to two hours and averaged almost an 
hour.  In at least three previous cases, boiling is known to have occurred. 

A number of actions have been recommended previously to prevent the loss of 
RHR pump suction during low vessel level operations.  These include:

     Providing accurate level instrumentation designed for reduced vessel 
     water level operations.  

     Providing alarms in the control room for low decay heat removal flow and 
     low water level. 
     
     Including in the procedures specific requirements for frequent monitoring 
     and strict limits on level.

     Considering in the procedures the possibility of vortex formation and air 
     entrainment, including a precaution against starting a second RHR pump 
     until the cause of the loss of the first pump is determined and 
     corrective actions have been taken.  

     Training the operators on the correlation between water level and pump 
     speed at the onset of vortexing and air entrainment.

     Careful planning, coordination, and communication with control room 
     personnel regarding all ongoing activities which could affect the primary 
     system inventory.

The NRC review of the Diablo Canyon event indicated that vortexing and air 
entrainment may occur at higher water levels than anticipated.  In addition, 
operation at mid-hot-leg levels can lead to unanticipated conditions which may 
not have been adequately considered in instrumentation design and procedure 
preparation.  

The NRC staff's initial assessment of this event has identified the potential 
for a significant loss of decay heat removal capability both from a total loss 
of the RHR system and from a loss of the steam generator heat sink due to air 
blanketing of the steam generator tubes.  Correct operator actions then become 
critical for plant recovery.  

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                                                                 May 27, 1987
                                                                 Page 4 of 5


NRC communications in the past have expressed serious concern with failures to 
maintain adequate decay heat removal capability.  Information Notice No. 81-09 
pointed out that loss of shutdown cooling capability had been found to be a 
potentially significant contributor to the total risk.  AEOD/C503 and other 
sources indicate that the time available to restore shutdown cooling before 
core uncovery can occur is not necessarily large.  At four days after shutdown 
from long-term power operation, with the vessel drained down to the RHR 
suction loss level, the vessel water can heat to the boiling point in about 
1/2 hour.  Under such conditions boiloff to the core uncovery level can occur 
in less than two hours.

Following the loss of decay heat removal capability on April 10, 1987 at 
Diablo Canyon, PG&E took a number of actions to prevent loss of RHR suction 
during low level operation and to improve recovery should such a loss occur.  
These actions included the following: 

     Evaluation of the reactor vessel level indicating system to determine the 
     level at which vortexing would occur and the effect of vortexing on the 
     level measurement. 
     
     Enhancements of the instrumentation to include accurate level 
     measurement, alarm capability and core exit temperature measurement 
     during low level operation. 
     
     Enhancement of procedures to include requirements for verifying proper 
     RHR pump suction before starting the second RHR pump. Also included are 
     precautions specifying minimum vessel levels as a function of RHR flow. 
     
     Improvements in work planning, control and communication to include a 
     restriction of the work scope to items that do not have the potential to 
     reduce RCS inventory. 
     
     Improvement of operator training including a discussion of the potential 
     causes of RHR flow loss, as well as recovery procedures.

The NRC is currently considering additional generic action on this issue.

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                                                                 May 27, 1987
                                                                 Page 5 of 5


This information notice requires no specific action or written response.
If you have any questions about this matter, please contact the Regional 
Administrator of the appropriate regional office or this office. 




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


Technical Contacts:  Donald C Kirkpatrick, NRR
                     (301) 492-8166

                     Warren C. Lyon, NRR
                     (301) 492-7605


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