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LER-2005-001, Steam Generator Feedwater Pump Trip Leading to Manual Reactor Trip and Auxiliary Feedwater Actuation
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Event date: 3-22-2005
Report date: 05-20-2005
Reporting criterion: 10 CFR 50.73(a)(2)(iv)(B), System Actuation

10 CFR 50.73(a)(2)(iv)(A), System Actuation
2512005001R00 - NRC Website

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 4 05000251


On March 22, 2005, at

0346 hours
0.004 days
0.0961 hours
5.720899e-4 weeks
1.31653e-4 months

the 4A Steam Generator Feed Pump (SGFP) [EIIS: Si, P] tripped on over-current due to a fault within the pump motor. As a result, a turbine runback from 100% power to 78% power occurred and Steam Generator [EHS: AB, SG] levels decreased to 15%. At that time, the reactor was manually tripped in accordance with station procedures. The operating crew verified the reactor and turbine were tripped, and that emergency power to the 4 KV buses and safety injection were not required. The plant was verified to be stable prior to transitioning from power operation to hot standby.

All control rods inserted as expected. Auxiliary Feedwater [EIIS: BA] automatically actuated as expected when Steam Generator narrow range levels decreased below 10%. The turbine [EIIS: SB, TRB] runback reduced the energy removal rate from the secondary side of the plant. This initially resulted in elevated Reactor Coolant System (RCS) [EIIS: AB] temperatures and resulted in an increase in the pressurizer [EHS: AB, PZR] pressure and level and decrease in steam generator levels.

Secondary system steam relief was achieved via atmospheric dump valves [EIIS: SB, V] and steam dump valves [EHS: SB, V] to the condenser [EHS: SG, COND]. Once steam generator levels recovered, auxiliary feedwater flow was throttled back and subsequently secured in accordance with station procedures. Plant systems required for safe shutdown of the plant functioned as designed.

There were no safety systems out of service prior to the event. All plant parameters responded within the design envelope for this type of transient.


At the time of the event, the plant was in Mode 1 at 100% power. The design basis Loss of Normal Feedwater Flow Event is analyzed in UFSAR section 14.1.11. The event is described in the UFSAR as a reduction in capability of the secondary system to remove heat generated in the reactor core. If an alternate supply of feedwater were not supplied to the plant, core residual heat following reactor trip could heat the reactor coolant to the point where water relief from the pressurizer would occur, resulting in a loss of inventory from the Reactor Coolant System. For this particular event, the initial conditions were well within the assumed condition of the Loss of Normal Feedwater Flow event analyzed in the UFSAR. Only one train of normal feedwater was lost and the auxiliary feedwater system actuated and served to maintain steam generator levels. UFSAR minimum and maximum analyzed values were not exceeded during this transient. The Auxiliary Feedwater System auto-initiated as required, due to the expected decrease in steam generator levels below 10% narrow range. The RCS pressure remained below the setpoint for pressurizer PORV or Code safety valve actuation.

Plant systems functioned as designed with some minor secondary side equipment performance exceptions and some minor primary side indication and alarm performance exceptions. These exceptions had no adverse effect on the operating crews' ability to safety shutdown the reactor and stabilize the plant. This event did not adversely affect the health and safety of the public.



The 4A Steam Generator Feed Pump motor breaker tripped on instantaneous over-current due to a motor failure. This initiated a turbine runback and decreasing steam generator water levels, which lead to Unit 4 being manually tripped. The motor, a General Electric, 7000 HP, 2 Pole, 4000 Volt, Frame 8611s, was last overhauled in 2002. During that overhaul, the motor leads were found to be degraded and were replaced. During the lead replacement, the existing 250 MCM flexible multi-strand motor leads were spliced to lead extensions which were a stiffer 250 MCM Cable. The connection of multi-strand wire to the stiffer MCM cable was a non-standard connection.

The motor was subjected to failure analysis through disassembly and inspection. One cable (terminal #2) had partially melted the connector tubing on the side of the original flexible motor cable and approximately one half inch of copper wires was melted. The terminal #3 cable had signs of an explosive short circuit that flared the cable strands of copper wire. An apparent tracking flashover area was located around 10 o'clock on the winding head where an arc had occurred. Both failures on the cables were located on the side of the original equipment manufacturer (OEM) motor cable, close to the splice connection. The cause of the failure is an improper splice that resulted in mechanical vibration of the motor leads that fractured the heat damaged strands due to mechanical fatigue. Potential contributing factors included: 1) superficial oil and dust contamination that developed tracking discharges around the insulated splice, leading to gradual melting of strands of copper wire and arcing; and 2) high electrical resistance due to the crimped splice connection which eventually caused heat damage. With the reduced copper cross section, a high temperature was developed melting copper wire strands until the current was conducted by plasma. The high temperature developed by the plasma melted the splice end and burnt the adjacent terminal lead cover insulation, creating an explosive phase to phase short circuit that tripped the protection relay.


A review of the reporting requirements of 10 CFR 50.72 and 10 CFR 50.73 and NRC guidance provided in NUREG-1022, Revision 2, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73, was performed for the subject condition. As a result of this review, the condition is reportable as described below.

10CFR50.73(a)(2)(iv)(A) states that the licensee shall report any event or condition that resulted in a manual or automatic actuation of any of the systems listed in 10CFR50.73(a)(2)(iv)(B). Systems to which the requirements of 10CFR50.73(a)(2)(iv)(A) apply include the Reactor Protection System (RPS) [EIIS: JC] actuation resulting in a manual trip (10 CFR 50.73(a)(2)(iv)(B)(1). Unit 4 was manually tripped in response to decreasing steam generator levels. The event is also reportable in accordance with 10 CFR 50.73 (a)(2)(iv)(A), due to automatic AFW system actuation (10 CFR 50.73(a)(2)(iv)(B)(6)).



This event had no significant effect on the health and safety of the public. A manual reactor trip was initiated in response to the turbine runback. All safety systems operated as designed; and no unexplained aspects of this transient, when compared to the UFSAR, were noted. The auxiliary feedwater system automatically actuated. And the condenser continued to function as a heat sink and condenser steam dumps remained operable throughout this event. Post-trip reviews established that plant parameters were within UFSAR analyzed minimum and maximum values for a loss of feedwater event. No radiological release occurred. Therefore, the event had very low safety significance.


Short Term — The SGFP 4A motor was removed for offsite inspection and repair.

Long Term — The repaired motor is being replaced during the 2005 Unit 4 Cycle 22 refueling outage. In addition, the motor repair Specification E-008 will be revised to include lessons learned from this event.


To evaluate the extent of condition, the maintenance history of other Unit 3 and Unit 4 SGFP motors was reviewed. None of these repairs employed a spliced connector. The terminations for the new leads went back to the coil connections in all cases.

identifier, second component function identifier (if appropriate)].

No similar events were identified.