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                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C.  20555

                                August 23, 1995


NRC INFORMATION NOTICE 95-33:  SWITCHGEAR FIRE AND PARTIAL LOSS OF OFFSITE
                               POWER AT WATERFORD GENERATING STATION, UNIT 3


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 (IN) to alert addressees to a switchgear fire and subsequent partial
loss of offsite power at Waterford Generating Station, Unit 3.  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 June 10, 1995, Waterford 3 was operating at 100 percent power with an
operations staff consisting of a shift supervisor (SS), a control room
supervisor (CRS), and two reactor operators.  At 8:58 a.m. a generator trip
occurred in response to failure of a lightning arrester on a remote offsite
substation transformer.  The generator trip resulted in a fast transfer
activation.  All 6.9 kV and 4.16 kV buses transferred as designed except the
4.16 kV A2 bus.  A fire and electrical fault on the 4.16 kV A2 bus normal
power supply breaker caused a voltage and frequency perturbation on the 6.9 kV
A1 bus, which caused an underspeed condition on rector coolant pumps 1A and
2A.  This circumstance resulted in a reactor trip and a loss of offsite power
to the 4.16 kV nonsafety-related A2 bus and the associated 4.16 kV safety-
related A3 bus.  Emergency Diesel Generator A started and loaded to power the
A3 bus.  At 9:06 a.m., an auxiliary operator informed the control room of
heavy smoke within the turbine generator building.  At that time, the SS did
not activate the plant fire alarm or dispatch the fire brigade, but directed
two auxiliary operators to don protective gear and investigate whether a fire
existed.  At 9:35 a.m., the operators reported seeing flames above the A2
switchgear and the SS activated the fire brigade.  Operators requested
assistance from the local offsite fire department and declared an Unusual
Event in accordance with emergency response procedures.  The fire brigade was
unable to suppress the fire using portable fire extinguishers.  The offsite
fire department arrived on the scene at 9:58 a.m. and extinguished the fire
with water at 10:22 a.m., after the A2 bus was deenergized.  During the
cooldown transition from Mode 4 to Mode 5, operators discovered that the
isolation valves for both trains of shutdown cooling did not operate properly.

9508180092.                                                            IN 95-33
                                                            August 23, 1995
                                                            Page 2 of 4


The plant cooldown to Mode 5 was delayed approximately 38 hours while these
valves were repaired.

Discussion

During the period of June 13-16, 1995, the NRC conducted an augmented
inspection team (AIT) inspection to determine the causes, conditions, and
circumstances relevant to this event.  The results of this AIT inspection are
documented in NRC Inspection Report 50-382/92-15, dated July 7, 1995.  The AIT
identified three primary issues:  fire protection, fast bus transfer design,
and shutdown cooling valve inoperability.  These three issues are discussed in
greater detail in the following sections.

Fire Protection

Several recent events at U.S. nuclear power plants have included a fire
concurrent with a plant transient.  The fire at Waterford 3 highlights the
importance of (1) training for timely and effective response to initial
indications of a plant fire, (2) ensuring personnel are not assigned
potentially conflicting duties, and (3) plant staffing.

An auxiliary operator (a trained fire brigade member) noticed heavy smoke in
the turbine generator building and notified the control room.  The auxiliary
operator was asked if there was a fire in the room and responded that he did
not see flames because of the presence of heavy smoke.  The CRS did not
declare a fire until 29 minutes after receiving the report of heavy smoke. 
Activating the fire brigade required the SS to assume the responsibilities of
the CRS (the designated fire brigade leader), who was directing plant
personnel responding to the event.  Following the event, operators stated that
the loss of the CRS from the control room did not adversely affect their
ability to respond to this event and noted that a fire scenario, which
requires that the CRS leave the control room, is routinely used during
requalification training.

Before the local offsite fire department was allowed to extinguish the fire
with water, the fire brigade attempted to extinguish the fire using portable
carbon dioxide (CO2), halon, and dry chemical fire extinguishers.  The use of
portable extinguishers was not effective in extinguishing the fire.  When the
fire department arrived, it recommended the use of water to extinguish the
fire.  The fire brigade leader did not allow the use of water until about
20 minutes later.  The fire was finally extinguished by the offsite fire
department within 4 minutes of using water.  The use of water is consistent
with documented NRC staff positions.  The AIT determined that the operators
were reluctant to apply water to an electrical fire based on previous training
that had emphasized the use of water as a last resort on electrical fires.

Although the appropriate fire alarms had activated in the control room, the
control room crew was not aware of the alarms because of (1) other auditory
alarms caused by the event and (2) the lack of a visual fire alarm signal on a
front panel of the control room.  Control room operators did not refer to the
fire alarm panel when the auxiliary operator reported seeing heavy smoke.  In
this instance, the ineffectiveness of the fire alarms did not directly affect 
                                                            IN 95-33
                                                            August 23, 1995
                                                            Page 3 of 4


the response to the fire because an auxiliary operator alerted the control
room to heavy smoke in the turbine building.  Nevertheless, fire alarms that
are inaudible under actual operational conditions and lack redundant visual
signals can inhibit prompt identification of, and response to, plant fires. 
Also, it is important for operators to refer to the fire alarm panel upon any
verbal report of a potential fire, in order to ensure that the fire is not
wider spread than visually reported.  NRC fire protection requirements and
guidelines specify that fire drills include an assessment of fire alarm
effectiveness.

IN 91-77 "shift staffing at Nuclear Power Plants", reminded licensees that
Section 50.54(m) of Title 10 of the Code of Federal Regulations 
(10 CFR 50.54 (m)) addresses minimum staffing levels for licensed personnel. 
It does not address availability of personnel for performing all actions
specified in the licensee's administrative procedures required during an
event.  NRC fire protection requirements and guidelines provide flexibility in
assigning personnel to a fire brigade (e.g., the brigade leader may possess
either an operator's license or an equivalent knowledge of plant safety-
related systems).  The potential exists for personnel to be assigned duties
that, during certain events, may present concurrent and conflicting demands. 
Such conditions could significantly delay or degrade the response of those
individuals.

Fast Bus Transfer Design

The Waterford 3 fast bus transfer design consists of an automatic transfer of
safety and nonsafety-related station auxiliary loads from the normal power
supply (from the main generator through the unit auxiliary transformer) to the
alternate power supply (from the offsite transmission network through the
startup transformer).  All supply breakers are General Electric, Magne-Blast
type.  During a fast bus transfer, the normal supply breakers are designed to
open in five cycles and the alternate supply breakers are designed to close in
seven cycles, resulting in a two-cycle deadband on the respective buses.  To
prevent simultaneous closing of both the supply breakers, some other fast bus
transfer designs include mechanical or electrical interlocks.  The Waterford 3
design does not include interlocks.  

During this event, when the fast bus transfer was initiated, the A2 bus normal
supply breaker did not open in five cycles but the alternate supply breaker
closed within seven cycles.  As a result, (1) the A2 bus was connected to both
the offsite transmission network and the main generator, (2) both supply
breakers to the A2 bus received overcurrent trip signals, (3) while the A2 bus
alternate supply breaker adequately isolated the offsite transmission network,
the A2 bus normal supply breaker did not isolate the main generator, (4) the
A2 switchgear cubicle for the normal supply breaker caught fire, and (5) the
cable bus for the normal supply breaker also caught fire.

Shutdown Cooling Valves

During the plant cooldown to Mode 5, the shutdown cooling isolation valves
failed to operate properly when operators attempted to align low-temperature
overpressure protection relief valves in preparation for placing shutdown  .                                                            IN 95-33
                                                            August 23, 1995
                                                            Page 4 of 4


cooling into service.  The Loop 1 shutdown cooling suction header isolation
valve (SI-405B) failed to fully open and automatically closed after
approximately 15 minutes.  The Loop 2 shutdown cooling suction header
isolation valve (SI-405A) fully opened; however, several hours later, the
valve hydraulic pump was observed to be running continuously instead of
cycling as designed.  These two valves isolate low-pressure portions of the
shutdown cooling system from the reactor coolant system and must be opened in
order to complete plant cooldown below 200 �F (Mode 5).  Troubleshooting
revealed that both valves contained inadequate hydraulic oil levels in the
valve actuator reservoirs.  The cause of the low levels was inadequate
instructions for a periodic maintenance task for the valves.

Related Generic Communications

BUL 75-04, "Cable Fire at Browns Ferry Nuclear Power Station," dated March 24,
1975.

BUL 75-04A, "Cable Fire at Browns Ferry Nuclear Power Station," dated April 3,
1975

BUL 75-04B, "Cable Fire at Browns Ferry Nuclear Power Station," dated  
November 3, 1975.

IN 85-80, "Timely Declaration of an Emergency Class, Implementation of an
Emergency Plan, and Emergency Notifications," dated October 15, 1985.

IN 91-57, "Operational Experience on Bus Transfers," dated September 19, 1991.

IN 91-77, "Shift Staffing at Nuclear Power Plants," dated November 26, 1991.

IN 93-44, "Operational Challenges During a Dual-Unit Transient," dated     
June 15, 1993.

IN 93-81, "Implementation of Engineering Expertise on Shift," dated     
October 12, 1993.

This information notice requires no specific or written response.  If you have
any questions about the information in this notice, please contact the
technical contact listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager.


                                    /s/'d by DMCrutchfield

                                    Dennis M. Crutchfield, Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical contact:  Eric J. Benner, NRR               
                    (301) 415-1171
```