Information Notice No. 93-28: Failure to Consider Loss of DC Bus in the Emergency Core Cooling System Evaluation May Lead to Nonconservative Analysis

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

                                 April 9, 1993

                               LEAD TO NONCONSERVATIVE ANALYSIS 


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 that a single failure of a 125-Vdc bus has been
identified as potentially the worst case single failure in the emergency core
cooling system (ECCS) evaluation for certain boiling water reactors (BWR-3 and
BWR-4).  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 July 30, 1992, the Nebraska Public Power District, licensee for the Cooper
Nuclear Station, notified the NRC that the worst case single failure in the
ECCS had not been correctly identified in the analysis of the loss-of-coolant
accident (LOCA) in conjunction with a loss of offsite power.  To satisfy
regulatory requirements, the most limiting single failure, which results in
the most severe calculated consequences, must be considered in performing the  
LOCA analysis.  The previous licensee analysis assumed that the worst case
single failure was a failure of the low pressure coolant injection (LPCI)
system injection valve in the ECCS train that is connected to one
recirculation loop, concurrent with a pipe break in the other recirculation
loop.  In July 1992, the licensee recognized that failure of a 125-Vdc bus
that provides the control power for the LPCI injection valve serving the
unbroken recirculation loop is the worst single failure for this accident. 
The licensee discovered the problem while performing a plant design basis


The low-pressure trains of the ECCS are shown conceptually in Figures 1 and 2. 
Trains A and B each include a core spray pump and two residual heat removal
(RHR) pumps that function as LPCI pumps during a LOCA.  The two RHR pumps in


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each train share common piping and a common injection valve to the associated
recirculation loop.  When offsite power is lost, two emergency diesel
generators (EDGs) supply power to the pump motors and two 250-V batteries
supply motive power to the valve motors.  Two 125-V batteries provide control
power to the valve motors and the output circuit breakers for the EDGs.  

In the previous ECCS analysis, as shown in Figure 1, the licensee assumed a
guillotine break in one recirculation loop and failure of the LPCI injection
valve for the other recirculation loop.  If this event were to occur, coolant
from RHR pumps B1 and B2 would not reach the reactor vessel because the
coolant would flow out through the broken recirculation loop, and RHR pumps A1
and A2 would not pump coolant to the reactor vessel because of the failure of
the LPCI injection valve to open.  However, it was concluded that both core
spray pumps would pump coolant to the reactor vessel and the temperature of
fuel cladding would increase but remain less than the regulatory limit.

The licensee has now determined that the worst single failure would be the
failure of one of the two 125-Vdc buses.  As shown in Figure 2, failure of the
125-Vdc bus for train A would prevent closure of the output breaker for EDG A. 
Motors for RHR pumps A1 and B2 and core spray pump A would not receive 
emergency power and would not supply coolant to the reactor vessel.  Likewise,
the LPCI injection valve for recirculation loop A would fail to open and RHR
pump A2, which receives power from EDG B, would be unable to pump coolant to
the reactor vessel.  If the break were in recirculation loop B, then RHR pump
B1 would pump coolant to the broken loop, leaving only core spray pump B to
pump coolant to the reactor vessel, which is insufficient to perform the
intended ECCS function.  To avoid the possibility that the temperature of the
fuel cladding might exceed the regulatory limit, the licensee reduced the
reactor power level pending completion of modifications to correct the
problem.  The licensee took this action based on analyses by the General
Electric Company.

Although the combined probability of occurrence of a guillotine rupture, loss
of offsite power, and failure of the 125-Vdc bus for the ECCS train serving
the intact recirculation loop is very low, the licensee has modified the 
control power for the LPCI injection valves and recirculation pump discharge
valves so that they are powered from the 250-Vdc buses, while leaving the 
125-Vdc control power for the EDG output breakers unchanged.

Related Generic Communications

The General Electric Company submitted a report to the NRC as an attachment to
a letter dated November 1, 1978, addressing the effect of a failure of a
direct current power supply on BWR-3 and BWR-4 reactors.  General Electric
concluded that peak cladding temperature would be greater than previously
anticipated for a small break LOCA, but less than the regulatory limit, and 
that the peak clad temperature would be unaffected for large breaks.  The
report indicated that at least two ECCS pumps would be available.

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The NRC sent letters to licensees on April 25, 1980, requesting that 
addressees confirm the validity of the conclusions made by the General
Electric report.  The letters asked that responses include lists of ECCS
equipment that would be available for breaks in the suction and discharge
piping connected to pumps in the recirculation loops.  Based on the recent
analysis performed at Cooper, the conclusion in the General Electric report
that at least two ECCS pumps would remain available may not be true for all
scenarios involving direct current power supply failure. 

The NRC recently issued Information Notice 93-11, "Single Failure
Vulnerability of Engineered Safety Features Actuation Systems," to alert
licensees to a design deficiency identified at Millstone Nuclear Power
Station, Unit 2, that causes a spurious engineered safety feature actuation
when one train of DC electrical power is deenergized.  In addition, in Generic
Letter 89-18, "Systems Interactions in Nuclear Power Plants," the NRC
highlighted concerns regarding actuation system designs, including electrical
power system designs, that could cause adverse system interactions.  

This information notice requires no specific action or written response.  If  
you have any question 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.

                                         ORIGINAL SIGNED BY

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

Technical contacts:  David L. Skeen, NRR
                     (301) 504-1174

                     Elmo E. Collins, RIV
                     (817) 860-8291

1.  Figure 1:  Low Pressure ECCS Trains...
    Figure 2:  Break in Recirculation Loop...
2.  List of Recently Issued Information Notices


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