Information Notice No. 95-10: Potential for Loss of Automatic Engineered Safety Features Actuation

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                            WASHINGTON, D.C.  20555

                               February 3, 1995

                               SAFETY FEATURES ACTUATION


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 the potential for loss of the automatic
actuation function of engineered safety features (ESF) as a result of
electrical faults in some non-class 1E input signals.  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 February 2, 1995, the licensee for the Diablo Canyon facility reported to
the NRC a condition that could result in the failure of one train of their
solid state protection system (SSPS) during a main steamline break in the
turbine building (10 CFR 50.72 report number 28318).  The licensee postulated
a break of a main steamline at the turbine stop valve in the turbine building. 
If the steamline breaks completely, it is free to rotate approximately 10
degrees.  The 10-degree rotation of the steamline could result in the steam
jet from the faulted steamline striking an electrical junction box.  The
junction box contains terminations for non-safety input signals to the SSPS,
turbine stop valve position indication (four circuits, two circuits for each

The force of the steam jet impinging on the junction box is postulated to
destroy the box and result in electrical faults in the affected non-safety
inputs to the SSPS.  The high current resulting from the electrical faults
would cause 15-ampere fuses to open, interrupting 120-V ac power supply to the
faulted circuits.  Since dc power supplies for SSPS logic and ESF train
actuation relays are supplied by the same 15-ampere fuses, opening of the
fuses would also interrupt power to the SSPS logic channels and possibly one
ESF train actuation relay bank.  This would render one SSPS train inoperable. 
If a single failure of the other SSPS train is considered, as is required in 

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the high energy line break analysis methodology, both trains of the SSPS would
be rendered inoperable and no ESF actuations would be automatically available
to mitigate the consequences of the steamline break.  The reactor trip
circuitry would be de-energized resulting in a reactor trip.  Manual action
could be initiated to operate individual pieces of equipment.    

NRC inspectors determined that other non-class 1E circuits that provide input
to the SSPS were not properly isolated.  These circuits include turbine auto
stop oil (three circuits), seismic trip (constructed to class 1E standard),
12-kV undervoltage, 12-kV underfrequency, and reactor coolant pump breaker
position indication.  An electrical fault in any of these circuits could cause
loss of power to SSPS logic circuit in the same way described above. 

Although a single main steamline break would likely render only one SSPS train
inoperable, either train could be rendered inoperable depending upon the
location of the steamline break.  The licensee declared the ESF portion of the 
solid state protection system inoperable and entered Technical Specification
3.3.2 for inoperable ESF instrumentation and then Technical Specification
3.0.3 limiting condition for operation to start shutdown of both units within
1 hour.

On February 1, 1995, the licensee for the Salem facility notified the NRC that
it had been determined that the design of the SSPS at its facility was similar
to that at the Diablo Canyon facility (10 CFR 50.72 report number 28321).  The
Salem licensee concluded that a main steamline break could have the same
effect on non-class 1E circuits as that postulated at the Diablo Canyon
facility.  In addition, the licensee concluded that a seismic event could
challenge both trains of SSPS since both junction boxes associated with both
trains of SSPS are located in the turbine building (the Diablo Canyon licensee
is continuing to evaluate seismic and other vulnerabilities of these non-class
1E circuits).  The circuits that are potentially affected at Salem include
turbine stop valve position indication, auto-stop oil pressure switches, and
reactor coolant pump breaker position indication.  The circuit faults
initiated by the steamline break or seismic event could result in loss of
power to SSPS logic circuitry similar to that postulated by the Diablo Canyon
licensee.  The resulting impact would be either a partial or total loss of the
automatic actuation function of the SSPS.  The reactor trip circuitry would be
de-energized, resulting in a reactor trip.  Manual action would be required to
mitigate the consequences of a main steamline break event.  The licensee
declared the SSPS inoperable and began a shutdown of Unit 1 as required by
Technical Specifications.  (Unit 2 was already shut down.)


The licensees are undertaking similar corrective actions.  The electrical
supply to the SSPS dc power supplies will be taken from a point electrically
upstream of the 15-ampere fuse referred to above.  This modification will
ensure that opening of the fuse, because of faults in the non-class 1E
circuits, does not cause a loss of power to the class 1E dc power supplies. 
Figure 1 is a one-line diagram illustrating the as-found condition of a single
channel of the SSPS at Diablo Canyon.  Two out of four channels are     .                                                            IN 95-10
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potentially affected in SSPS Train A and two out of four channels are
potentially affected in SSPS Train B. 

The Diablo Canyon licensee plan for repairs includes drafting a formal
modification procedure, curtailing work near the vulnerable junction boxes,
deferring train-related maintenance and surveillance during the modification
period, maintaining constant power level, testing the procedure on a mock-up,
de-energizing one channel at a time while modifying that channel (the reactor
trip bypass breaker will be closed during the modification), training 
operators on safety considerations during the repairs, and postmodification
testing.  The licensee estimated that the increase in core damage frequency
during the repair period would be less than 2E-7.  (The Salem licensee repair
plan was not available when this information notice was prepared.)

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.

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

Technical contacts:  E. Nick Fields, NRR
                     (301) 415-1173

                     Cliff Doutt, NRR 
                     (301) 415-2847

1.  Figure 1 (See File IN95010.WP1 for Figure 1)

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