United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 87-60: Depressurization of Reactor Coolant Systems in Pressurized-Water Reactors

                                                        IN 87-60

                                  UNITED STATES
                          NUCLEAR REGULATORY COMMISSION
                             WASHINGTON, D.C.  20555

                                December 4, 1987

                                   SYSTEMS IN PRESSURIZED-WATER REACTORS


All holders of operating licenses or construction permits for pressurized 
water reactors.


This notice is being provided to alert addressees of potential problems 
resulting from the loss of pressure control in the reactor coolant system 
(RCS) which could affect the operator's ability to depressurize the reactor 
coolant system in a timely manner during a steam generator tube rupture 
accident, or to control the plant during natural circulation cooldown.  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:

Two events have occurred which demonstrate the importance of maintaining the 
capability to depressurize the RCS in emergencies.  

The importance of maintaining effective pressure control in mitigating a steam 
generator tube rupture event was positively demonstrated during the North Anna 
Unit 1 tube rupture which occurred on July 15, 1987.  A double ended rupture 
of a single tube occurred in the "C" steam generator causing an initial break 
flow of around 600 gpm.  

The plant was manually tripped from 100% power at about five minutes into the 
event.  This was followed in about 20 seconds by an automatic safety injection
actuation.  After positively identifying and isolating the steam generator 
with the rupture, the operators initiated a rapid cooldown to 480 degrees F in 
order to establish an adequate subcooling margin.  This was accomplished by 
dumping steam from the undamaged steam generators to the main condenser using 
steam dump valves.  A few minutes later a rapid RCS depressurization was 
commenced by fully opening the two pressurizer spray valves.  As this pressure 
reduction began to tail off, the operators briefly opened a pressurizer PORV 
causing an additional rapid 40 psi drop in RCS pressure.  The primary to 
secondary leakage was promptly terminated when the RCS pressure was equalized 
with the pressure of the steam generator having the ruptured tube at about 30 

.                                                            IN 87-60
                                                            December 4, 1987
                                                            Page 2 of 4

minutes into the event.  During the remainder of the cooldown the primary 
pressure was maintained below the pressure of the steam generator with the 
rupture to minimize secondary contamination and to facilitate cooldown of the 
steam generator using backfill.

Because they were able to maintain good primary pressure control, the 
operators were able to minimize the radiological release during this event.  
None of the secondary atmospheric relief valves were actuated.  The release 
was limited to the contamination of the secondary system before the steam 
generator with the rupture could be isolated.  The total release was estimated 
at 159 mCi for the entire event.  

On August 26, 1986, a reactor trip occurred at Salem Unit 2 when a technician 
inadvertently grounded a 120 VAC instrument bus, causing a spurious loss-of- 
reactor-coolant-pump reactor trip signal.  The voltage spike generated by the 
grounding also generated a spurious low-steam line pressure signal which, in 
conjunction with a high-steam flow indication, initiated a safety injection 
signal.  About 30 seconds later, a series of vital bus transfers were 
generated by the protective relaying logic.  During these transfers, two of 
the vital buses were without power simultaneously for about two seconds, which 
resulted in the generation of a station blackout signal.  However, offsite 
power was actually available and the reactor coolant pumps continued to 
operate.  The coincident safety injection and station blackout signals 
disconnected all vital power buses and automatically sequenced selected safety 
injection loads onto the emergency buses powered by the already operating 
diesel generators.  The number two vital bus remained deenergized because the 
diesel generator for this bus had been taken out of service for maintenance.  
However, in accordance with the plant design, this automatic sequencing did 
not load the component cooling water pumps onto the emergency buses.  

The reactor operators secured the reactor coolant pumps after 5 minutes of 
operation because component cooling water was not available to cool the motor 
bearings and the thermal barrier.  The high-pressure safety injection pumps 
continued to operate after the receipt of the safety injection signal.  The 
resulting rise in reactor coolant system pressure caused a power-operated 
relief valve (PORV) to lift numerous times.  Normal pressurizer spray was not 
available to control the primary system pressure rise once the reactor coolant 
pumps were tripped.  

Although safety injection was not needed, the charging pumps continued to 
inject water into the primary system through the emergency core cooling system 
(ECCS) piping.  The isolation valves had assumed their safeguards (open) 
position following initiation of the safety injection signal.  Since the vital 
bus that powered the ECCS isolation valves was deenergized, the control room 
operators could not isolate the ECCS flow without taking local manual control 
of the isolation valves.  The operators elected not to shutdown the charging 
pumps because they were needed to supply injection water to the reactor 
coolant pump seals.  In addition, the operators were unable to initiate the 
auxiliary pressurizer spray even with the charging pumps running because the 
spray isolation valve, also connected to the deenergized vital bus, was closed 
as part of the automatic safeguards alignment and could not be opened 

.                                                            IN 87-60
                                                            December 4, 1987
                                                            Page 3 of 4

The operators manually energized the component cooling water pumps after 7 
minutes.  However, it took more than 20 minutes for the operators to secure 
safety injection, start a reactor coolant pump, and reestablish normal 
pressure control.  


Reactor coolant system pressure control is necessary for the timely recovery 
from steam generator tube rupture accidents; i.e., to minimize the discharge 
of reactor coolant into the faulted steam generator and the subsequent loss of 
coolant outside containment, such as occurred during the Ginna accident 
(January 25, 1982).  Pressure control also is important to facilitate natural 
circulation cooldown.  Generally, the normal pressurizer spray system is used 
to control or reduce reactor coolant system pressure.  However, this system 
requires the operability of the reactor coolant pumps and the pressurizer 
spray control valves.  In the Salem event, the reactor coolant pumps were 
secured and one of the normal pressurizer spray lines had been isolated for 
about three months because of excessive leakage.  

Emergency operating procedures for many plants utilize the PORVs for depres-
surizing the primary system following a steam generator tube rupture accident 
if the normal pressurizer spray system is not available.  In the Salem event, 
one of the PORVs had been isolated for about 2 weeks prior to the event, also 
because of excessive leakage.  Although an isolated PORV could probably be 
unblocked if it was seriously needed for pressure reduction, the PORV 
isolation represents an additional loss of redundancy and reliability.  If the 
normal pressurizer spray system is out of service and the PORVs are 
unavailable, the auxiliary pressurizer spray system on plants having such a 
system can be used to depressurize the primary system.  However, during the 
Salem event the auxiliary pressurizer spray system was also unavailable 
because its isolation valve was closed and could not be repositioned from the 
control room due to the loss of its vital bus.  This vital bus was not 
re-energized immediately because the diesel generator supplying power to this 
bus was out of service for preventive maintenance.  

The availability of the pressurizer spray system, the PORVs for some plants 
and the auxiliary pressurizer spray system are generally not assured by the 
limiting conditions for operation contained in the Technical Specifications.  
Nevertheless, as these events demonstrate, these systems can be important to 
the safety of the plant under certain emergency conditions.  Consequently, it 
is important that out of service periods for repairs or maintenance be 
minimized for these systems.  In the case of the PORVs the reliability of the 
closing capability as well as the assurance of availability for pressure 
control is important.  During the Ginna accident the PORV stuck open causing a 
loss-of-coolant to the containment and the formation of coolant voids in the 
reactor vessel head and the tube bundle of the faulted steam generator.  At 
Indian Point Unit 2 (LER 247/85-002) and Callaway Unit 1 (LER 483/84-064) all 
of the PORVs were found to have been isolated during normal operation, 
inhibiting their ability to provide pressure control and to promptly mitigate 
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                                                            December 4, 1987
                                                            Page 4 of 4

a potential accident.  Further information regarding this issue can be found 
in AEOD/E708, "Depressurization of Reactor Coolant Systems in PWRs," an engi-
neering evaluation report issued by the NRC Office for the Analysis and 
Evaluation of Operational Data.  

No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the technical 
contact listed below or the Regional Administrator of the appropriate regional 

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

Technical Contacts:  Sanford Israel, AEOD
                     (301) 492-4437

                     Donald C. Kirkpatrick, NRR
                     (301) 492-8166

Attachment:  List of Recently Issued NRC Information Notices

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