Information Notice No. 84-76: Loss of All AC Power

                                                           SSINS No.:  6835 
                                                           IN 84-76        

                               UNITED STATES 
                          WASHINGTON, D.C. 20555  

                              October 19, 1984 

Information Notice No. 84-76:   LOSS OF ALL AC POWER 


All nuclear power reactor facilities holding an operating license (OL) or 
construction permit (CP). 


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  


                                                         IN 84-76          
                                                         October 19, 1984  
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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. 


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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                                                          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 

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