Catawba (USA)

Map of Catawba

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2 * 1200 MW PWR constructed by Westinghouse; grid connection 1985/86

Facilities in Catawba

plantreactor typconstruction startoperation startshut down
Catawba-1PWR19741985
Catawba-2PWR19741986
1998-05-11

During a planned power reduction on May 7, plant operators determined that the temperature of the water in the AFW system´s holding tanks exceeded design limits. The temperature had risen to 110 °C, above the 59 °C stipulated in the plant´s final safety analysis report. All three AFW pumps were declared inoperable due to uncertainty regarding their operation with the higher water temperature.

1996-04-01

Duke Power Co.'s ambitious plan to replace steam generators at three nuclear power plants gets under way June 13 when change-outs begin at Catawba-1.
The replacements, to be done as part of a regular refueling outage at the unit, are scheduled for completion September 21. Other four-loop SG´s at McGuire-1 will be replaced beginning in January 1997, and at McGuire-2, in August 1997, also during the units´ regular refueling outages.
Duke ordered 12 SGs from Babcock & Wilcox International in 1992 for $510-million to replace Westinghouse SG´s that were showing stress corrosion cracking.
At Catawba-1, Duke plans to make 31 modifications to install the new SGs. The utility has estimated how long each will take to complete. Some, which require moving some instrumentation tubing, aren't supposed to take more than a few hours. Others, such as cutting the old reactor coolant pipes and welding the new SG nozzles to those pipes, will take several weeks.
The old generators will be stored on-site at each plant in specially-designed concrete buildings until the plants are decommissioned. Operating licenses for the three units run out in 1021, 2023 and 2024, although the new SGs have a 60-year design life.

1996-02-06

On 6 February 1996, the Catawba Unit 2 pressurized water reactor in Clover, South
Carolina automatically shut down from 100 percent power after main transformer
problems disconnected the reactor from the electrical grid. The loss of offsite power
signaled both of the emergency diesel generators to start and provide electricity to
vital equipment needed to cool the reactor core. One of the emergency diesel
generators started and powered its assigned equipment, but the second diesel generator
failed due to a faulty capacitor in its battery charger. Workers repaired this diesel
generator and connected it to its loads about 3 hours into the event. Workers repaired
the transformer and reconnected the reactor to its electrical grid about 37 hours into
the event.
The loss of offsite power deprived the reactor of all the equipment normally used to
cool the reactor core. The initial failure of one emergency diesel generator deprived
the reactor of half of the emergency equipment used to cool the reactor core during
accidents. The NRC calculated the severe core damage risk from this event to be
2.1 x 10-3 or 0.21% per reactor year and rated it Level 1 on the INES scale.
(source: Residual Risk)

1993-02-25

During testing it was discovered that in both units the discharge valve for the service water pumps would not open against pump discharge pressure. Cause: too low torque settings of the 32" motor operated butterfly valves. Because valves could not have opened ESSW would not be available in emergency situations. One of the pumps is running during normal operation supplying CCS. The other 3 pumps would be needed in many accident scenarios to cool EDGs, containment spray etc. But non of this would have worked because of the blocking of the discharge valves. Condition existed since initial operation.

1993-02-25

During testing it was discovered that the discharge valve for the service water pumps would not open against pump discharge pressure. Cause: too low torque settings of the 32" motor operated butterfly valves. because valves could not have opened ESSW would not be available in emergency situations. One of the pumps is running during normal operation supplying CCS. The other 3 pumps would be needed in many accident scenarios to cool EDGs, containment spray etc. But non of this would have worked because of the blocking of the discharge valves. Condition existed since initial operation.

1991-07-30

Design deficiency could have prevented the auxiliary feedwater system from meeting its design basis. The feedwater pump's discharge head was beeing overcome by other sump pumps´ discharge head in a common discharge header.

1991-05-16

HPCI: an orifice plate in one cold leg injection header had been installed backwards in 1990, required system flow could not be achieved.

1991-02-19

Design errors in the control room ventilation system would prevent it from pressurizing the CR under accidents conditions.

1991-02-19

Design errors in the control room ventilation system would prevent it from pressurizing the CR under accidents conditions.

1991-02-12

A refrigerant leak and a failed open cooling water valve caused the loss of both control room chillers.

1990-10-23

Control room ventilation failed, one train due to a damper actuator malfunction, in the other train chiller could not be started due to a switch failure.

1990-06-11

Operator error caused misalignment of RHR. This caused 20.000 liters of reactor coolant inventory to be discharged into the RWST. Reactor remained subcooled without any voids.

1990-03-21

Containment valve injection system: makeup supply line clogged with mud/debris.

1988-04-29

High failure rate of diesel starts resulting in Catawbas 4 diesel engines in 7 days.

1988-04-29

High failure rate of diesel starts resulting in Catawbas 4 diesel engines in 7 days tech .spec. surveillance test frequency.

1988-03-10

All AFW pumps inoperable because of degraded flow, Cause: Asiatic clams in both trains of nuclear service water.

1988-03-09

All AFW pumps inoperable because of degraded flow, cause asiatic clams in both trains of nuclear service water.

1988-03-09

Degradation of service water system by marin shellfish makes feedwater system inoperable

1986-06-13

On 13 June 1986, control room operators at the Catawba Unit 1 pressurized water
reactor in Clover, South Carolina received indications of a reactor coolant system leak
exceeding 1 gallon per minute. The normal makeup pumps could provide sufficient
water to the reactor coolant system to compensate for this leakage. Five hours after the
initial indication, the leak rate jumped to nearly 130 gallons per minute. This leak rate
exceeded the makeup capacity of the pumps. As the water level in the pressurizer
dropped due to more water leaving the reactor coolant system than was being added,
the operators manually shut down the reactor. The operators also took steps to reduce
the leak rate and measures to recover the pressurizer water level.
It was later determined that a weld on the letdown or bleed system piping had cracked
to cause the initial leak. The letdown system allows a continuous flow of about 45
gallons per minute of reactor cooling water to go to a system that purifies it and
adjusts its chemical parameters as necessary. Five hours later, the nameplate—a metal
label identifying the manufacturer and operating parameters—vibrated loose from a
power transformer and fell onto an electrical circuit board. The nameplate caused an
electrical short that, among other things, caused the flow control valve in the letdown
piping to fully open. The higher flow rate through the letdown piping caused the crack
to propagate.
The NRC calculated the severe core damage risk from this event to be 3 x 10-3 or 0.3%
per reactor year. The event was not rated on the INES scale.
(source: Residual Risk May 2007)