05000461/LER-2017-007

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LER-2017-007, Manual Reactor SCRAM due to Loss of Feedwater Heating
Clinton Power Station, Unit 1
Event date: 06-10-2017
Report date: 11-09-2017
Reporting criterion: 10 CFR 50.73(a)(2)(iv)(A), System Actuation
Initial Reporting
ENS 52800 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
4612017007R01 - NRC Website
LER 17-007-00 for Clinton, Unit 1 re Manual Reactor SCRAM due to Loss of Feedwater Heating
ML17223A187
Person / Time
Site: Clinton Exelon icon.png
Issue date: 08/09/2017
From: Stoner T R
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
SRRS 5A.108, U-604366
Download: ML17223A187 (5)


comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LER NUMBER

2017 - 01 007

PLANT AND SYSTEM IDENTIFICATION

General Electric -- Boiling Water Reactor, 3473 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in text as [XX].

EVENT IDENTIFICATION

Manual Reactor SCRAM due to Loss of Feedwater Heating A. Plant Operating Conditions before the Event Unit: 1 Event Date: 6/10/17 Mode: 1 Mode Name: Power Operation Event Time: 2256 CDT Reactor Power: 98 percent

B. Description of Event

On June 10, 2017, at 2256 CDT, Clinton Power Station (CPS) experienced a complete loss of the 'A' feedwater (FW) heater [HX] string. The operators received numerous FW trouble alarms on FW string 'A' and low pressure (LP) heater [HTR] 1A/1B bypass valve (1CB004) automatically opened.

The operators entered procedure CPS 4005.01, "Loss of FW Heating," which directs the operators to restore and maintain reactor power at or below the original power level and within stability control and power/flow map limits by adjusting reactor recirculation flow, control rods, or CRAM array. The operators lowered core flow and inserted all CRAM rods. The operators observed that FW temperature had dropped by greater than 100°F. Procedure CPS 4005.01 directs the operators to place the reactor mode switch [JS] into the shutdown position and enter procedure CPS 4100.01, "Reactor Scram." With the unit at approximately 93 percent power, the operators placed the mode switch in shutdown at 2306 on June 10, 2017 and entered procedure CPS 4100.01.

The components of the CPS power conversion system are designed to produce electrical power utilizing the steam generated by the reactor [RCT], condense that steam into water, and return the water to the reactor as heated feedwater. A portion of the main turbine [TRB] steam is extracted for FW heating. CPS has two trains of cascading FW heaters. Under normal, full power conditions, the extraction steam valves [V] to each of the FW heaters are open such that steam is condensed in the body of the heater. In addition, the normal heater drain valves are normally open and the emergency FW heater drain valves to the main condenser [COND] are closed. A high-high level in a FW heater will isolate the input sources to the heater (i.e., extraction steam valve(s) and the upstream normal FW heater drain valve(s)), reducing the reactor FW temperature.

A walkdown of panel [PL] 1PAO8J (Miscellaneous Sensors & Transducers Power Supply Cabinet), which houses Moore trip units for both the 'A' and 'B' FW heating strings, identified that there were no lights on rack CA-1 and the indicator for fuse [FU] FU-89 was open.

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LER NUMBER

2017 - 01 007 Troubleshooting using an ohmmeter found that high resistance in the circuit was eliminated after pulling Moore trip unit 1LYHD103A, which is the trip unit that provides automatic actions on a Hi-Hi level for FW heater 4A. This indicated that the loss of power to rack CA-1 was caused by fuse FU- 89 opening in response to a shorted condition on the Moore trip unit 1LYHD103A. When fuse FU- 89 opened, power was also lost to the Moore trip units for FW heaters 1A, 2A, 3A, 5A, and 6A resulting in the loss of heating to the 'A' FW heater string.

C. Cause of the Event

The root cause of the manual reactor scram due to the loss of the 'A' feedwater heater string is that the design of the feedwater heater level control trip circuitry was not adequate to prevent scrams due to an unevaluated single point vulnerability. The first contributing cause is a technical error in an analysis that incorrectly determined that there was no single component failure that will cause a FW temperature drop greater than 100°F. The second contributing cause is that the designer did not adequately consider the potential for high heat conditions inside panel 1PAO8J due to lack of adequate cooling; the high heat conditions in the panel have resulted in shortened life and reduced reliability of the Moore trip units.

D. Safety Consequences This event is reportable under the provisions of 10 CFR 50.73(a)(2)(iv)(A) due to the manual actuation of the reactor protection system.

An assessment of the safety consequences and implication of this event determined that the manual reactor scram ensured the plant remained in a safe and stable condition and no operating limits were exceeded.

The design basis loss of feedwater heating transient for CPS is based on a maximum temperature transient of 100°F. Should this event occur at a lower reactor power level, the severity of the transient would be reduced commensurate with the reduction in FW heating.

E. Corrective Actions

Prior to startup, CPS modified the circuit card locations in the panel that contains the Heater Drain system Moore trip units and thereby diversified the power supplied so that the trip units have less dependency on common fuses. In addition, the blown fuse FU-89 was replaced. Additional corrective actions included installation of temporary cooling and temperature loggers in the Panel 1PAO8J to monitor for elevated temperature condition.

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LER NUMBER

2017 - 01 007 Further actions include developing an engineering evaluation to determine if there are additional single component failures, operator errors, or events for the FW heating system that could result in a decrease in FW temperature of greater than 100°F. In addition, a permanent modification that eliminates the high heat conditions in the panel is being tracked by the corrective action program.

F. Previous Similar Occurrences Communication Between the Architect Engineer and the Nuclear Steam Supply System Supplier.

On July 28, 1988, CPS experienced a partial loss of FW heating. The FW temperature drop, excluding the change caused by a reduction in power, was greater than 102°F, but less than 112°F. The design basis loss of FW heating transient for CPS is based on a maximum temperature transient of 100°F. The cause of the loss of FW heating was the inappropriate setting of the FW heater level controllers. The cause of exceeding the design basis is attributed to the failure of the FW heating system design to meet design requirements. This was caused by a lack of adequate communication between the Nuclear Steam Supply System (NSSS) supplier and the architect engineer regarding the NSSS design requirements for the FW heating system.

Feedwater heating system design changes, including changes to the level trip setpoint for closing the extraction steam valves and replacing power supply fuses, were made to ensure that the design basis is met.

G. Component Failure Data

Failed card was determined to be a Moore Industries DCA alarm card.

Model Number: DCA/4-20ma/DH1L2/45dC/-AD-100HB1 (PC) Serial Number: 2412651