United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 95-04: Excessive Cooldown and Depressurization of the Reactor Coolant System Following a Loss of Offsite Power


January 19, 1995

                               OF THE REACTOR COOLANT SYSTEM FOLLOWING A
                               LOSS OF OFFSITE POWER


All holders of operating licenses or construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to an excessive cooldown and depressurization of
the reactor coolant system (RCS) and the main steam system following a loss of
offsite power at the McGuire Nuclear Station Unit 2.  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 December 27, 1993, while McGuire Unit 2 was operating at 100-percent power,
525-kV bus line 2B was lost because of a failed insulator (this was the event
initiator).  Because the expected turbine runback failed to initiate, breakers
for bus line 2A opened on overcurrent protection, resulting in a loss of
offsite power to the unit.  Reactor coolant pumps tripped when offsite power
was lost, resulting in core cooling by natural circulation.  Emergency diesel
generators successfully started and provided power to the vital buses.  The
reactor tripped on "Power Range High Flux Rate" 36 seconds into the event, and
the turbine immediately tripped because of the reactor trip.  After the
reactor trip, the RCS rapidly cooled down and depressurized because of a
reduction of energy input and an increase in energy removal, by full
unthrottled auxiliary feedwater (AFW) flow and several steam release paths. 
The steam paths included the AFW pump, open steam line relief and safety
valves, open steam dumps, and open drain lines.  A safety injection signal was
received on low pressurizer pressure 7 minutes and 32 seconds into the event,
followed by another safety injection signal on low steam line pressure and a
main steam isolation signal 1 second later.  The main steam isolation valve
(MSIV) for steam generators A and B failed to close fully.  The continuing
steam loads, including the open MSIV, caused secondary system pressure to drop
rapidly, and unthrottled AFW flow continued to lower steam pressure and
temperature.  Continued secondary cooling caused a continuous drop in the RCS
temperature.  In addition, because forced circulation was lost with the loss
of the reactor coolant pumps, a large temperature differential existed across 

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the reactor core, indicated by a rapid reduction in the RCS cold-leg
temperature.  Sixteen minutes and 48 seconds into the event, operators
throttled AFW flow to zero for all four steam generators, allowing the
stabilization of pressure in steam generators A, C, and D.  Pressure in steam
generator B recovered temporarily but continued to drop because of the open
MSIV.  In addition, water level in steam generator B was decreasing. 
Approximately 1.5 hours into the event, the indicated wide-range water level
in steam generator B reached zero.  Concurrently, operators were cycling
pressurizer power-operated relief valves to reduce the primary pressure and
thereby reduce the differential pressure across the tubes of steam generator B
to less than 11.03 MPa [1600 psi] (as suggested by McGuire emergency operating
procedures for a dry steam generator).  Differential pressure across the tubes
of steam generator B reached approximately 13.65 MPa [1980 psi].  Cycling the
power-operated relief valves caused the pressure and water level to increase
in the pressurizer relief tank, rupturing the rupture disk.  Ice condenser
doors opened in response to the rupture of the rupture disk.  Offsite power
was restored approximately 1.5 hours into the event, and the vital buses were
realigned to offsite power approximately 2.5 hours into the event.


In its followup evaluation, Duke Power Company (the licensee) concluded that
steam loads and AFW flow caused the rapid secondary side depressurization and
cooldown before the main steam isolation.  The licensee performed detailed
modeling of the relative contributions to the cooldown of the unthrottled AFW
flow, the AFW pump turbine steam load, and the open steam line drains, and
determined that the open steam line drains were the primary contributor.  At
McGuire, AFW actuation logic for a loss of offsite power automatically starts
the turbine-driven AFW pump.  A loss of offsite power causes a loss of control
power for both main feedwater pumps, which trips the pumps and causes an
automatic start of the motor-driven AFW pumps.  AFW injects at the maximum
rate when started automatically.  At the time the safety injection actuated,
the operators had not reached a point in the emergency operating procedures
where throttling of the AFW was allowed.  In addition, no specific guidance
was given to the operators for monitoring the RCS cold-leg temperature or
throttling the AFW to slow a cooldown indicated by the cold-leg temperature. 
In natural circulation, reactor coolant cooling is best tracked by monitoring
the cold-leg temperature.  Emergency operating procedures (EOPs) have been
changed to instruct operators when checking for an uncontrolled cooldown that
the cold-leg temperature will be used if reactor coolant pumps are off to
ensure that operators control AFW earlier in this type of event.  A training
package was issued to all reactor operators on the use of the EOP foldout
page.  This information should reduce the time required to complete EOPs and
to arrive at the steps for manual main steam isolation and throttling of the
AFW.  The licensee has also modified steam line drains upstream and downstream
of the MSIVs to fail closed on a loss of power.  The modifications will slow
the secondary side depressurization rate and the RCS cooldown rate.  The
change of the EOP is consistent with Revision 1B of the Westinghouse Emergency
Response Guidelines.  These changes were discussed by the operations support
manager in a lecture during requalification training of the licensed
operators.  The changes will ensure more effective AFW control during an
excess cooldown process following a reactor and turbine trip..                                                            IN 95-04
                                                            January 19, 1995
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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.

                                    Original signed by

                                    Brian K. Grimes, Director
                                    Division of Project Support
                                    Office of Nuclear Reactor Regulation

Technical contacts:  Eric J. Benner, NRR
                     (301) 504-1171

                     Chu-Yu Liang, NRR
                     (301 504-2878

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