Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
SSINS No.: 6835
IN 84-76
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
October 19, 1984
Information Notice No. 84-76: LOSS OF ALL AC POWER
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or
construction permit (CP).
Purpose:
This notice is provided to alert recipients of a significant degradation in
the level of safety caused by loss of all ac power. It is expected that
recipients will review this information for applicability to their
facilities and consider actions, if appropriate, to preclude similar
problems occurring at their facilities. However, suggestions contained in
this notice do not constitute NRC requirements and, therefore, no specific
action or written response is required.
Description of Circumstances:
On July 26, 1984, Susquehanna Steam Electric Station Unit 2 was operating at
30% power and Unit 1 was operating at 100% power when the licensee began
preparations for a "Loss of Turbine Generator and Offsite Power" startup
test on Unit 2. Test conditions required that the electrical supplies to the
units be separated so that all Unit 1 engineered safety system buses were to
be fed from the Unit 1 startup transformer, all Unit 2 engineered safety
system buses were to be fed from the Unit 2 startup transformer, and all
feeder breakers from the Unit 1 startup transformer to the Unit 2 engineered
safety system buses were to be racked out. The tie breaker between Unit 1
and Unit 2 auxiliary buses also was required to be racked out. All loads
common to both units were to be placed on Unit 1 supplies.
After establishing the required test configurations and prerequisites, the
startup test was initiated by simultaneously opening the Unit 2 main
generator output breakers and the Unit 2 startup transformer feeder breaker.
As expected, the reactor tripped, turbine bypass valves opened and
containment isolation occurred. However, the four emergency diesel
generators did not start and all four feeder breakers to the Unit 2
engineered safety system buses remained closed. The breakers should have
opened and the diesel generators should have started automatically when the
startup transformer feeder breaker opened. The operator opened these
breakers from the control room. When the diesels still did not start, the
operator manually started all four diesels from the control room. Diesel
generator D tripped on overvoltage, and B tripped on overvoltage and
underfrequency. Diesel generators A and C idled but did not close onto their
associated buses. Diesel generator A exhibited large frequency
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IN 84-76
October 19, 1984
Page 2 of 3
oscillations and was manually tripped by the operator. The operator tried to
manually close diesel generator C breaker to the associated bus, but the
breaker would not close. The operator then closed the startup transformer
breaker and attempted to close the feeder breakers to the engineered safety
system buses, but they would not close.
At this point, operators were instructed to rack in the feeder breakers from
the Unit 1 startup transformer to the four Unit 2 engineered safety system
buses. As each breaker was restored to operability, the related engineered
safety system bus preferred feeder breaker closed and the related tripped
diesel generator (A, B, and D) automatically started. Power was restored to
the first engineered safety system bus in approximately 11 minutes and the
last bus in approximately 18 minutes. After restoring power, diesel
generators A, B, and D were shut down manually from the control room as they
had high priority alarms.
During the loss of all ac power to Unit 2, a significant portion of the
instrumentation in the control room failed downscale. The dc powered
instrumentation available to the operator included two narrow range level
instruments for monitoring reactor water level, high pressure coolant
injection (HPCI) and reactor core isolation cooling (RCIC) supply pressure
indicators for monitoring reactor pressure, and a source range monitor. The
full core display provided erroneous indication that a significant number of
rods had not inserted into the core, which initially confused the operators.
The shutdown was confirmed, based on the indication from source range
monitor instrumentation and the reactor pressure trend. The control room had
no indication of suppression pool temperature and no indication of reactor
water level below zero on the narrow range instrument. Personnel stationed
at the local instrument racks were able to provide reactor water level
information to the control room.
Discussion:
The causes of the event include operator error, inadequate operator
training, imprecise procedures, ineffective independent verification, and
inadequate implementation of corrective actions for previously identified
problems. NRC and licensee investigations have revealed that the event was
initiated as a result of incorrect performance of the process utilized to
rack out the feeder breakers from the Unit 1 startup transformer to the Unit
2 engineered safety system buses. The normal practice for racking out a
breaker is to ensure the breaker is open, enter the breaker cubicle, and
open the knife switch supplying dc power for breaker control. The breaker is
then racked out. When the operator went to rack out a breaker, he was
confronted with two dc knife switches and opened the wrong switch, thereby
removing dc power to the engineered safety system logic circuitry for each
bus rather than the dc control power to the breaker. The error was repeated
on all four buses.
The consequences of the error were as follows:
- no automatic transfer capability of engineered safety buses to
alternate power sources
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IN 84-76
October 19, 1984
Page 3 of 3
- no automatic diesel generator start on loss of bus voltage
- no ability to re-energize the buses from an offsite source from
the control room
- no bus load shedding capability
- no degraded grid protection
- no breaker overcurrent or differential current protection
- no core spray or residual heat removal pump automatic or manual
start capability even with power available, hence, disabling the
low pressure emergency core cooling systems
The investigation further revealed that the knife switch for dc control
power to the breaker was labeled "BREAKER CONTROL SWITCH AND TRIP CIRCUIT
FUSES" and the one for dc power to the engineered safety system logic
circuitry was labeled "DC CONTROL." It was the "DC CONTROL" knife switch
that the operator opened, and the error was not detected by the startup
engineer assigned to verify the operator actions. During the investigation
of the event, it was learned that on two previous occasions during the
preoperational testing, the "DC CONTROL" switch was improperly operated. The
licensee corrective action following these events was to provide operator
training; however, the operator who racked out the breakers for the startup
testing had not received that training.
Following the event, the licensee initiated immediate and long-term
corrective action programs. Immediate actions, which included revising
labeling and painting of knife switches and providing training and revising
procedures to preclude similar events, were completed prior to the unit
restart. The longterm corrective actions include: improvement of the
independent verification program; upgrading of existing electrical operating
procedures and developing new ones, as required; determination of the
adequacy of instrumentation available on loss of ac power; and evaluation of
the present design for compliance with Regulatory Guide 1.47, Bypass and
Inoperable Status Indication for Nuclear Power Plant Safety Systems.
Although the diesel generator problems which occurred during the event added
to the complexity of the event, licensee actions to resolve these problems
are beyond the scope of this notice.
No specific action or written response is required by this information
notice. If you have any questions regarding this matter, please contact the
Regional Administrator of the appropriate regional office or this office.
Edward L. Jordan, Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: R. N. Singh, IE
(301) 492-8985
Attachment: List of Recently Issued IE Information Notices
