05000220/LER-2017-003

LER-2017-003, Automatic Reactor Scram due to Reactor Vessel Low Water Level

Nine Mile Point Unit 1
Event date:	09-06-2017
Report date:	11-2-2017
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(B), System Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation
Initial Reporting
ENS 52950	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation, 10 CFR 50.72(b)(3)(iv)(A), System Actuation
2202017003R00 - NRC Website

LER 17-003-00 for Nine Mile Point, Unit 1, Regarding Automatic Reactor Scram due to Reactor Vessel Low Water Level

ML17325A994
Person / Time
Site:	Nine Mile Point
Issue date:	11/02/2017
From:	Kreider R E Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NMP1 L3184 LER 17-003-00
Download: ML17325A994 (7)

Estimated burden per. response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported 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 Infodollects.Resource@nrc.goy, and to the Desk Officer, Office of Information and Regulatory Affairs, 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 003 00

A. PRE-EVENT PLANT CONDITIONS:

Prior to the event, Nine Mile Point Unit 1 (NMP1) was in Run and operating at 100% reactor power.

B. EVENT:

On 9/6/2017, at approximately 11:57, NMP1 experienced a reactor scram due to reactor water low level. Following the scram Unit 1 also experienced a low-low reactor water level condition causing automatic Reactor Pressure Vessel (RPV) and containment isolation. Significant automatic responses that occur due to low-low reactor water level conditions include Main Steam Isolation Valve (MSIV) isolation, Reactor Water Clean Up (RWCU) isolation, core spray initiation signal, reactor recirculation pump trip, and emergency condenser vent and drain valve isolation.

The HPCI mode of feed water first initiated on the reactor water low level signal. Following this HPCI initiation, vessel level continued to lower until low-low level was received. Due to the scram, trip of recirculation pumps, and HPCI initiation, vessel level rose to greater than 85" requiring Reactor Operators to perform procedurally driven actions for securing the feed water pumps.

Additionally, following MSIV isolation on low-low level, vessel pressure control was established on the emergency condenser system briefly by manually placing the 11 emergency condenser in service. This resulted in an expected rise in vessel level. Approximately five minutes after placing the 11 emergency condenser in service Reactor Operations reopened MSIVs, transferred pressure control from the 11 emergency condenser to the turbine bypass valves, and removed 11 emergency condenser from service.

Following the restoration of RWCU, opening of the MSIVs, and recovery of RPV water level, the HPCI mode of feed water was reset and feed water level control was transferred to the manual mode.

Troubleshooting following the scram identified a failed power supply within Proportional Amplifier PAM- ID23E which caused the Flow Control Valve for #13 Feedwater Pump to close.

The module was sent for failure analysis. During this analysis, it was identified that a Schottky diode internal to the power supply of the Proportional Amplifier had failed, resulting in the output from the module to drop out.

Nine Mile Point Unit 2 (NMP2) was unaffected by the scram at NMP1.

Operations performed the ENS notification (#52950) required by 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A) for the manual reactor scram and specified system activations.

This event has been entered into the plant's corrective action program as IR 4049445.

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 Inf000llects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104). Office of Management and Budget. Washington, DC 20503. If a means 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 003 - 00

• C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO

THE EVENT:

No other systems, structures, or components contributed to this event.

ACTIONS:

The dates, times, and major occurrences and operator actions for this event are as follows.

September 6, 2017 1156:57 —Computer Point K331, FW CNTR SF/FF ERROR SIG received on the PPC alarm typer 1157 —13 Feedwater Flow Control Valve rapidly closes 1157 —Annunciator F2-3-3, REACT VESSEL LEVEL HI-LO received for low RPV level 1157:15 —Reactor automatically scrams on low RPV level, HPCI initiation signal received 1157:21 —11 and 12 Feedwater Pumps start in HPCI mode 1157:34 —Lo-Lo Vessel Level reached, vessel and containment isolation signals received 1157:34 —All Reactor Recirculation Pumps trip 1157:34 —Reactor Water Cleanup isolates 1157:35 —Core Spray auto start signal 1157:35 —All MSIVs shut 1157:35 —Lo-Lo isolation signal clear 1158 —Reactor Operator executes N1-SOP-1 level control actions, secures both Feedwater Pumps due to high RPV level 1158:45 —11 Emergency Condenser placed into service for RPV pressure control 1203:31 —MSIVs reopened, Main Condenser re-established for RPV pressure control 1204:14 —11 Emergency Condenser removed from service 1217 —Core Spray is secured

E. METHOD OF DISCOVERY:

This event was discovered by Reactor Operators when the 13 Feedwater Flow Control Valve began to rapidly close and Reactor Vessel Level Hi-LO Alarm was received for low RPV level.

F. SAFETY SYSTEM RESPONSES:

All safety systems responded per design.

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 dces not display a currently valid OMB control number, the I NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020

II. CAUSE OF EVENT:

The cause of the scram was due to a failed power supply within the Proportional Amplifier, PAM-ID23E.

This power supply failure was determined to be due to a Schottky diode internal to the power supply of the Proportional Amplifier which resulted in the output from the module dropping out causing the #13 Feedwater Pump Flow Control Valve to close.

III. ANALYSIS OF THE EVENT:

The automatic reactor scram is reportable under 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.73(a)(2)(iv)(A). It is defined under paragraph 10 CFR 50.73(a)(2)(iv)(A) as any event or condition that resulted in manual or automatic action of any of the specified systems listed in 10 CFR 50.73(a)(2)(iv)(B).

The equipment failure associated with this event is a power supply within the Proportional Amplifier, PAM-ID23E. This power supply failure was determined to be due to a Schottky diode internal to the power supply of the Proportional Amplifier which resulted in the output from the module dropping out causing the #13 Feedwater Pump Flow Control Valve to drive closed. All other plant systems performed per design. Plant parameters, other than the RPV water level, remained within normal values throughout the event. There was no loss of offsite power to the onsite emergency buses, both trains of The HPCI mode of feed and condensate system initiated as designed, and the core spray initiation signal was received but injection was not required.

Based on the above discussion, it is concluded that the safety significance of this event is low and the event did not pose a threat to the health and safety of the public or plant personnel.

After review and following the guidance contained within NEI 99-02 it was determined that this event did not require additional operator actions beyond that of a "normal" scram. The scram did not involve the unavailability of or inability to recover main feedwater nor did it involve the lifting of electromagnetic relief valves for pressure control. This event did not present additional challenges to plant operations staff and does not meet the criteria required to be classified as a complicated scram.

This event does affect the NRC Regulatory Oversight Process Indicator for unplanned scrams per 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> of critical operation.

IV. CORRECTIVE ACTIONS:

A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL

STATUS:

Replaced failed Feedwater Level Control module PAM-ID23E.

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 003 - 00

• B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

To prevent recurrence, a corrective design change is scheduled for 2019 which will implement and utilize fault tolerant digital controllers for the feedwater flow control valves.

A. FAILED COMPONENTS:

The root cause was a failed power supply within the Proportional Amplifier, PAM-ID23E. This power supply failure was determined to be due to a Schottky diode internal to the power supply of the Proportional Amplifier which resulted in the output from the module dropping out causing the #13 Feedwater Pump Flow Control Valve to drive closed. This failure was premature as the power supply was new as of March 2017.

B. PREVIOUS LERs ON SIMILAR EVENTS:

error proportional amplifier. The cause of this failure was in a new power supply and the actions from 2012 would not have prevented this event C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION

IN THIS LER:

IEEE 803 IEEE 805

IDENTIFIER IDENTIFICATION

Main Steam Isolation Valve ISV SB Proportional Amplifier AMP JB Reactor Pressure Vessel RPV AD Feedwater Level Control System N/A JB High Pressure Coolant Injection System N/A BJ Feedwater System N/A SJ Reactor Water Clean Up System N/A CE Reactor Protection System N/A SC Containment Isolation System N/A JM Emergency Condenser System N/A BL Core spray system N/A BG

D. SPECIAL COMMENTS:

None