Information Notice No. 90-61: Potential for Residual Heat Removal Pump Damage Caused by Parallel Pump Interaction

                                UNITED STATES
                           WASHINGTON, D.C.  20555

                             September 20, 1990

                                   DAMAGE CAUSED BY PARALLEL PUMP 


All holders of operating licenses or construction permits for nuclear power 


This information notice is intended to alert addressees to the potential for 
flow stoppage caused by the interaction of the parallel pumps in residual 
heat removal systems that have discharge check valves located upstream of 
the recirculation lines.  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 do not constitute NRC requirements; therefore, no 
specific action or written response is required.

Description of Circumstances:

In December 1989, the staff at Unit 1 of the Sequoyah Nuclear Power Plant 
found that one of the residual heat removal pumps was running without flow 
(deadheading) during simultaneous surveillance testing of both pumps, a 
condition that can damage a pump from overheating.  In this test, both 
residual heat removal pumps, operating in the safety injection mode, were to 
draw water from the refueling water storage tank suction lines and discharge 
it back through the "minimum flow" recirculation lines (see Figure of the 
residual heat removal system, attached).  In the residual heat removal 
system at Sequoyah, each of the two trains has a check valve located on the 
discharge side of the pumps and upstream of the recirculation lines.  A 
normally open crossover line connects the two trains downstream of the 
recirculation lines.  Because one of the pumps had a higher discharge 
pressure than the other, the pressure of the stronger pump acting through 
the crossover line forced the discharge check valve of the weaker pump to 
close.  This stopped the flow from the weaker pump.  In its analysis of the 
event at Sequoyah, the Tennessee Valley Authority (the licensee) found that 
operating a pump with no flow for longer than 11 minutes may cause pump 


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The NRC Office of Analysis and Evaluation of Operational Data has published 
a study of this issue entitled "Potential for Residual Heat Removal System 
Pump Damage," AEOD/E90-06.  This study reviewed the RHR designs of nineteen 
randomly selected plants and found five of them with system piping 
configurations similar to that of the RHR system at the Sequoyah plant.  The 
study indicates that the potential for the adverse pump-to-pump interaction, 
which can cause the discharge check valve to close, may not be detected 
during a surveillance test in which one pump is tested at a time.

Discussion of Safety Significance

In addition to providing core cooling when the reactor is shut down, the 
residual heat removal system provides low-pressure coolant injection (safety 
injection) during an accident.  If a loss-of-coolant accident followed by 
the actuation of the safety injection system were to occur, both pumps would 
start running.  However, until the primary system pressure has decreased to 
a level that is below the pump shutoff head of 184 psi, the pumps could not 
inject into the reactor.  For a small break, the amount of time required to 
decrease the reactor pressure to the low-pressure injection point could 
cause deadheaded pumps to overheat.  For this reason, the minimum flow 
bypass line valves are designed to open during this phase to permit 
sufficient flow through the pumps to cool them.  During normal operation, 
the crossover line between the two residual heat removal trains is kept 
open.  This is to assure a cooling water supply to the reactor from either 
of the RHR pumps under adverse conditions, such as a break in one of the 
lines to the reactor cold legs.  Consequently, the conditions identified in 
the Sequoyah test would probably exist during a small break accident and 
could cause the failure of the weaker pump.  

NRC Bulletin 88-04, "Potential Safety-Related Pump Loss," previously 
addressed the issue of the deadheading of a weaker pump during two-pump 
minimum flow operation.  In its analysis that was performed in response to 
this bulletin, the Sequoyah staff concluded that the deadheading problem did 
not exist at Sequoyah.  This conclusion was based on the use of a value of 
11.1 psi for the differential pressure between the two pumps, which had been 
derived from the average values from several tests.  This value was 
considered to be too low to cause the deadheading problem.  Following the 
discovery of the deadheading problem in 1989, the Sequoyah staff 
recalculated the differential pressure between the two pumps based on 
individual pump pressures and concluded that the actual value was 17 psi, 
which was sufficient to block the flow from the weaker pump.

The staff at Sequoyah has made an interim change in its emergency operating 
procedures to prevent the damaging of a residual heat removal pump during a 
safety injection actuation.  This change requires that one of the pumps be 
stopped and placed in the standby mode if the reactor coolant system 
pressure remains above 180 psi for longer than a specified time after the 
initiation of safety injection.  However, this procedure has the 
disadvantage of requiring operator action within a short time during an 
emergency situation.  As a permanent corrective action, the Sequoyah staff 
will install check valves in each train downstream of the recirculation 
lines.  With these valves in place, 


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any backflow resulting from crossover from a stronger pump would close the 
new check valve in the lower pressure train and isolate the weak pump's 
recirculation line from the stronger pump. 

This information notice requires no specific action or written response.  If 
you have any questions about the information in this notice, please contact 
the technical contact listed below or the appropriate NRR project manager.

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

Technical Contact:  Chuck Hsu, AEOD
                    (301) 492-4443

1.  Figure of Residual Heat Removal System - Minimum Flow Lineup
2.  List of Recently Issued NRC Information Notices


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