United States Nuclear Regulatory Commission - Protecting People and the Environment

ACCESSION #: 9604190298
                       LICENSEE EVENT REPORT (LER)

FACILITY NAME:  James A. FitzPatrick Nuclear Power Plant PAGE: 1 OF 6

DOCKET NUMBER:  05000333

TITLE:  Potential Common Mode Failure of Circuit Breakers in Both
        Safety Divisions Due to Design or Installation Error

EVENT DATE:  02/12/96   LER #:  96-002-00   REPORT DATE:  04/12/96

OTHER FACILITIES INVOLVED:                          DOCKET NO:  05000

OPERATING MODE:  N   POWER LEVEL:  100

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR
SECTION:
50.73(a)(2)(v), OTHER: 10CFR21

LICENSEE CONTACT FOR THIS LER:
NAME:  Mr. W. Verne Childs, Senior Licensing TELEPHONE:  (315) 349-6071
       Engineer

COMPONENT FAILURE DESCRIPTION:
CAUSE:  B   SYSTEM:  EB   COMPONENT:  52   MANUFACTURER:  G080
REPORTABLE NPRDS:  Y

SUPPLEMENTAL REPORT EXPECTED:  NO

ABSTRACT:

On 2/12/96 while at 100 percent rated power, Residual Heat Removal
Service Water (RHRSW) pumps A and C in safety division 1 failed to start
upon operator demand.  Investigation revealed a high resistance
electrical contact in the pump motor circuit breaker close coil circuit.
Evaluation of the failure determined that the electrical contact had high
resistance due to repeated interruption of current approximately three
times rated.  Failure of the contacts occurred after 2,163 and 3,233
operating (close) cycles for pumps A and C respectively compared to a
manufacturers design of 10,000 cycles.  Since the contact failures
occurred after a fraction of the design cycles, the event is considered
to be a condition requiring a report under 10CFR21.  In addition, since
potential failures could affect any 4 kV circuit breaker in either or
both safety divisions, this also requires a report under 10CFR50.73.  The
contacts were replaced in the failed circuit breakers and other
safety-related circuit breakers with more than 1500 operating cycles.
Additional corrective actions include permanent jumpering of the contacts
and procedure changes to check the contacts during preventive maintenance
until jumpering is complete.

END OF ABSTRACT

TEXT                                                          PAGE 2 OF 6

EIIS Codes are in []

EVENT DESCRIPTION

On February 12, 1996 at 0027 hours during normal plant operation at 100
percent rated power while performing monthly pump and valve operability
tests required by Technical Specification (TS) Surveillance Requirement
4.5.B.1.c.1 Residual Heat Removal Service Water (RHRSW) [BI] pump C
failed to start upon operator demand from the Main Control Room [NA].

Operations personnel were dispatched to the safety-related 4160 VAC
switchgear [EB] which contains the circuit breaker for RHRSW pump C to
investigate the cause of the failure of the circuit breaker to close.  No
abnormal conditions such as blown fuses, circuit breaker protective relay
trip device indicators (flags) or circuit breaker to cubicle misalignment
were found.  Indicators on the circuit breaker indicated that the closing
springs were fully charged and that the circuit breaker was ready for
closure upon demand.  Operators withdrew the circuit breaker from the
cubicle (racked-out) and then racked-in the circuit breaker without
noting any abnormal conditions.  Another attempt was made to start RHRSW
pump C from the Main Control Room and again the circuit breaker did not
close.

Operations personnel initiated a Deficiency/Event Report (DER) to
document entry into the Limiting Condition for operation and verified the
remaining components of the containment cooling mode of Residual Heat
Removal/Low Pressure Coolant Injection (RHR/LPCI) [BO) were operable as
required by TS 3.5.B.2/4.5.B.2 which allow continued operation of the
plant with one RHRSW pump inoperable for 30 days.

At 0844 hours Operations personnel attempted to start RHRSW pump A as
part of a routine Primary Containment [NH] pressure suppression chamber
(torus) pool cooling evolution.  RHRSW pump A did not start as expected.
Preliminary investigation did not reveal any cause for the failure to
start.

Failure of RHRSW pump A to start resulted in both pumps in subsystem A
(Loop A) of the containment cooling mode of RHR/LPCI being inoperable.
Continued reactor operation is permitted in this condition for seven days
by TS 3.5.B.3/4.5.B.3 provided the redundant containment cooling
subsystem (Loop B) is verified operable immediately and daily thereafter.
Operators completed verification of the operability of containment
cooling loop B at 0930 hours and demonstrated by actual pump starting
that RHRSW pumps B and D operated properly.

TEXT                                                          PAGE 3 OF 6

Maintenance personnel performed troubleshooting of the circuit breakers
and determined that in each case a switch contact in the circuit breaker
closing circuit had a high electrical resistance that prevented
energization of the circuit breaker closing coil (solenoid).  The
switches were replaced and the circuit breakers were tested with
satisfactory results.  RHRSW pumps A and C were declared operable at 1805
hours on February 12, 1996 (17 hours and 38 minutes after discovery of
the initial problem with RHRSW pump C).

In an effort to determine the cause of the high switch contact resistance
that resulted in failure of the circuit breakers for RHRSW pumps A and C
to operate properly, evaluation of the failed switches, circuit breaker
operating and maintenance history, and industry operating experience
related to the circuit breakers was conducted.  The evaluation revealed:

1.   That the circuit breakers for RHRSW pumps A and C which experienced
     the failure to close upon demand had been operated (closed) 3,233
     and 2,163 times, respectively, and the contact resistance was 300 to
     1000 ohms and 200 to 400 ohms, respectively.  The contact blocks for
     each of these circuit breakers and for the other safety-related
     circuit breakers with more than 1,500 close cycles were replaced.

2.   An Equipment Failure Evaluation of the contact blocks which caused
     the circuit breakers for RHRSW Pumps A and C to fail to close
     revealed the following:

     -    The published electrical current interrupting rating in the
          manufacturers catalog for the contacts of concern was 2.2
          amperes (direct current, inductive load) while the circuit
          design results in interruption of an electrical current of
          approximately 6.0 amperes (direct current, inductive load).

     -    Disassembly of a failed contact block showed evidence of arcing
          in the form of metal beads (similar to weld splatter), contact
          burning and an oxide layer on the contacts.  Visual inspection
          of contacts removed from other circuit breakers with more than
          1,500 operating cycles indicated a definite correlation between
          the number of cycles and the condition of the contacts.  The
          contacts from a circuit breaker with 1,536 cycles were in good
          condition compared to contacts with more than 2,250 cycles.

     -    Preventive maintenance procedures (based on manufacturers
          maintenance recommendations) for the circuit breakers did not
          include a measurement of the contact resistance for the
          contacts of concern or address replacement of the contact block
          based on the number of operating cycles (circuit breaker
          closures) or age.

     -    The manufacturers design life of the circuit breakers is 10,000
          circuit breaker close cycles.

TEXT                                                          PAGE 4 OF 6

3.   Plant drawings and manufacturers drawings are not in complete
     agreement with respect to the contact block of concern.

     The contact block of concern is provided by the manufacturer for
     applications where the customer desires an indicator lamp which
     indicates that the closing springs are fully charged and thus the
     circuit breaker is ready for closure.  This "springs charged"
     indication is not included in the FitzPatrick plant design.  The
     elementary diagrams (Reference 1) for the circuit breaker for RHRSW
     pumps and all other safety-related 4160 VAC loads do not include the
     contact block while the manufacturers connection diagram (Reference
     2) indicates that when the contact block is not furnished the
     contacts are to be jumpered.  The terminology and sense of the
     notation (that is, the terminology and sense of "when not furnished
     the contacts are to be jumpered") is quite different than the
     notation used on the manufacturers elementary diagram (Reference 3)
     which indicates that the contacts of concern are to be "jumpered
     when not used" (emphasis added).

During the time period from the initial failure, February 12, 1996, to
March 21, 1996, switch failure events were evaluated to determine whether
or not the failures required a report under 10CFR21.  On March 21, 1996
it was concluded that because of the potential for failures in both
safety divisions; the potential for the defect to result in the failure
to allow either automatic or manual (operator demand) closure of any
safety-related 4 kV circuit breaker, and the observed failures at a
fraction of the 10,000 cycle design life, the defect resulted in a
"substantial safety hazard" as defined in 10CFR21.

The Plant Manager and Vice President of Nuclear Operations were informed
of this determination on March 21, 1996 and the NRC Emergency Operations
Center was also informed.

EVENT CAUSE

The failure of the circuit breakers for RHRSW pumps A and C to close upon
demand was due to a design and installation error (Cause Code B).  The
contacts failed at a fraction of the 10,000 cycle design life due to the
repeated interruption of the closing coil current which is approximately
three times the direct current, inductive load, interruption rating.

TEXT                                                          PAGE 5 OF 6

EVENT ANALYSIS

The event requires a report under 10 CFR 50.73(a)(2)(v).  That is, the
design error alone could have prevented the fulfillment of the safety
function of systems needed to: 1) shut down the reactor and maintain it
in a safe shutdown condition, 2) remove residual heat, 3) control the
release of radioactive material, and 4) mitigate the consequences of an
accident.

The actual failures involved the circuit breakers for RHRSW pumps A and
C.  Both pumps are in the same safety division (containment cooling loop
A) and the redundant safety division containment cooling loop pumps were
demonstrated to be operable.  However, when the observed failures at a
fraction of the design cycles are considered, it appears that a potential
for multiple circuit breaker failures in both safety divisions may have
existed.  For example, if during a design basis Loss of Coolant Accident
(LOCA) with coincident loss of off-site power, failure of Emergency
Diesel Generator [EK] load circuit breakers to close in one safety
division and the failure of the Core Spray System [BM] pump motor circuit
breaker to close in the other safety division could result in a complete
loss of accident mitigating low pressure Emergency Core Cooling System
(ECCS) injection to the reactor vessel.  Other combinations of potential
circuit breaker failures result in different potential ECCS failures or
on-site emergency AC power failures including station blackout.  The
Authority also considers the failures to be substantial safety hazard
which requires a report under 10CFR21.

CORRECTIVE ACTIONS

1.   The contact blocks in circuit breakers for RHRSW pumps A and C were
     replaced.  [Complete]

2.   The contact blocks in other safety-related circuit breakers with a
     history of more than 1,500 close cycles were replaced.  [Complete]

3.   Procedures for preventive maintenance of the circuit breakers will
     be revised to include a check of the resistance of the contacts of
     concern and to require replacement when the resistance is excessive.
     [Due Date September 1, 1996]

TEXT                                                          PAGE 6 OF 6

4.   The safety-related 4 kV circuit breakers which require automatic or
     manual closure to perform the intended accident mitigation function
     will be changed by permanent jumpering the contact of concern prior
     to startup following the 1996 refuel outage to eliminate the
     potential failure mode.  [Due Date December 10, 1996]

5.   The safety-related 4 kV circuit breakers which are normally closed
     will be changed by permanent jumpering the contact of concern during
     the next scheduled bus outage to eliminate the potential failure
     mode.  [Due Date November 1, 1998]

ADDITIONAL INFORMATION

Failed Components:

     Component Name:               General Electric Magne-Blast Circuit
                                   Breaker
     Model Number:                 AMH-4.76-250
     Manufacturers NPRDS Code:     G080

Previous Similar Events:

     None

References:

     1.   ESK-5BG, Revision 12, D.C.  Emergency Diagram, 4160 V Circuit,
          Residual Heat Removal Service Water Pump 10P-1A

     2.   Vendor Drawing 1.41-168, Revision C, (GE Drawing 0121D4634)
          Wiring Diagram - 4 kV Switchgear, Breaker 10520 and 10620

     3.   Vendor Drawing 1.41-106B, Revision B, (GE Drawing 0108B1458,
          Revision 1) Elementary Diagram - Inspection Box 4 kV Switchgear

ATTACHMENT TO 9604190298                                      PAGE 1 OF 2

                              Attachment 1

                               LER-96-002

                           Commitment Status

Number                         Commitment                        Due Date

JAFP-96-0163-01     Revise maintenance procedure to              9/1/96
                    include check of 4 kV circuit breaker
                    close coil circuit resistance.

JAFP-96-0163-02     The safety-related 4 kV circuit              12/10/96
                    breakers which require automatic or
                    manual closure to perform the intended
                    accident mitigation function will be
                    changed by permanent jumpering the
                    contact of concern to eliminate the
                    potential failure mode.

JAFP-96-0163-03     The safety-related 4 kV circuit              11/1/98
                    breakers which are normal closed will
                    be changed by permanent jumpering the
                    contact of concern during the next
                    scheduled bus outage to eliminate the
                    potential failure mode.

ATTACHMENT TO 9604190298                                      PAGE 2 OF 2

James A.  FitzPatrick
Nuclear Power Plant
P.O.  Box 41
Lycoming, New York 13093
315-342-3840

New York Power                                    Michael J.  Colomb
Authority                                         Plant Manager

April 12, 1996
JAFP-96-0163

United States Nuclear Regulatory Commission
Document Control Desk
Mail Station P1-137
Washington, D.C.  20555

SUBJECT:  DOCKET NO.  50-333
          10CFR21 REPORT
          LICENSEE EVENT REPORT: LER-96-002

          Potential Common Mode Failure of Circuit Breakers in Both
          Safety Divisions Due to Design or Installation Error

Dear Sir:

This report is submitted in accordance with 10CFR50.73(a)(2)(v) and in
accordance with 10CFR21.

Questions concerning this report may be addressed to Mr.  W.  Verne
Childs at (315) 349-6071.

Very truly yours,

MICHAEL J.  COLOMB

MJC:WVC:las
Enclosure

cc:  USNRC, Region 1
     USNRC Resident Inspector
     INPO Records Center

*** END OF DOCUMENT ***


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