05000387/LER-2017-005

LER-2017-005, Automatic Reactor Protection System Trip on High Neutron Flux

Susquehanna Steam Electric Station Unit 1
Event date:	06-08-2017
Report date:	1-0-2017
Reporting criterion:	10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(iv)(A), System Actuation
Initial Reporting
ENS 52795	10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release, 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation, 10 CFR 50.72(b)(3)(iv)(A), System Actuation, 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge
3872017005R01 - NRC Website

LER 17-005-01 for Susquehanna, Unit 1 Regarding Automatic Reactor Protection System Trip on High Neutron Flux

ML17277A320
Person / Time
Site:	Susquehanna
Issue date:	10/04/2017
From:	Berryman B Susquehanna, Talen Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-7637 LER 17-005-01
Download: ML17277A320 (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.

Susquehanna Steam Electric Station Unit 1 05000387 2017 - 005 - 01

CONDITIONS PRIOR TO EVENT

Unit 1 — Mode 1, approximately 100 percent Rated Thermal Power Unit 2 — Mode 1, approximately 100 percent Rated Thermal Power There were no structures, systems, or components that were inoperable at the start of the event that contributed to the event.

EVENT DESCRIPTION

On June 8, 2017, at 1527 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.810235e-4 months <br />, the reactor automatically scrammed due to a loss of Main Turbine Electro- Hydraulic Control (EHC) [EMS System Code TG] logic power causing a high neutron flux, Reactor Protection System (RPS) trip. Workers were restoring from making a power supply swap for the EHC, from permanent Magnetic Generator (PMG) to a house power supply. A Maxi Grabber was being removed from the PMG positive lug and came in contact with a grounding screw, resulting in a short and causing a loss of EHC pressure control.

After the momentary short cleared, the meter indications showed the power supply returned to the normal operating voltage. The 30 VDC (volts DC) bus voltage transient resulted in temporary saturation of the pressure regulatory output and caused the bypass valve fast open logic to actuate and closure of the control valves. Once the control valves closed beyond the bypass valve capacity, the reactor pressure increased resulting in the high neutron flux reactor scram.

All control rods inserted and both reactor recirculation pumps tripped due to reaching reactor water low level 2. Reactor water level lowered to -49 inches causing Level 3 (+13 inches) and Level 2 (-38 inches) isolations. High Pressure Coolant Injection (HPCI) [EIIS System Code BJ] and Reactor Core Isolation Cooling (RCIC) [EIIS System Code BN] automatically initiated and were overridden by control room operations after reactor pressure vessel water level was restored to the normal band with feedwater. HPCI and RCIC injected to the Reactor Coolant System during reactor level stabilization. Division 1 RHR was manually placed in suppression pool cooling. All isolations and initiations occurred as expected. No main steam relief valves opened. Pressure was controlled via main turbine bypass valve operation. The safety systems operated as expected.

Secondary Containment [EllS System Identifier: NG] Zone 1, 2, and 3 differential pressure lowered to 0" WG due to a trip of the normal operation of the Reactor Building Ventilation system that resulted from the Unit 1 Level 2 isolation. The differential pressure was restored to Zones 1, 2, and 3 by the initiation of Standby Gas Treatment System on the Unit 1 Level 2 initiation.

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Susquehanna Steam Electric Station Unit 1 05000387 NUMBER NO.

This event was reported by notification EN 52795, on June 8, 2017 at 1910 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.26755e-4 months <br />, under the four and eight- hour non-emergency reporting requirements pursuant to 10CFR 50.72(b)(2) and (b)(3). This event is also reportable in accordance with 10CFR 50.73(a)(2)(iv)(A) as an event that resulted in an automatic actuation of the RPS, including reactor scram, HPCI and RCIC. Although no system safety functional failure occurred, this event is also reportable pursuant to 10CFR 50.73(a)(2)(v)(C) as a condition that could have prevented fulfillment of a safety function.

CAUSE OF EVENT

The scram was caused directly by a DC+ (direct current, positive) test lead (Maxi Grabber) that inadvertently contacted with the PMG Power Supply grounding screw, causing a short and momentary loss of EHC logic power, which then caused an EHC transient. This resulted in the high flux RPS trip.

The root cause for this event is an insufficient focus on the High Risk Activity of adjusting the EHC power supply, and inadequate risk mitigating actions for that activity.

ANALYSIS/SAFETY SIGNIFICANCE

There were no actual, or potential for safety consequences from this event. The neutron monitoring system trip protects the fuel against high heat generation rates. The sequence of events and systems operation during this event remained consistent with existing safety analysis and design basis for these systems, the scram occurred before any thermal limits were reached, and in a manner bounded by the limiting events described in the Final Safety Analysis Report (FSAR). Required isolations and initiations occurred as would be expected for this event. No main steam relief valves opened and pressure was controlled via main turbine bypass valve operation. The safety systems operated as expected.

An engineering evaluation was performed and concluded that secondary containment could have performed its safety function of isolating as assumed in the accident analysis and also of re-establishing 0.25 inches vacuum (drawdown) within the assumed accident analysis time (10 minutes). Therefore, the subject event did not cause a loss of safety function. This event will not be counted as a safety system functional failure (SSFF) for the NRC performance indicator based on the engineering analysis that shows there was no loss of ability to fulfill the safety function.

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1. The Integrated Risk Management Procedure will be revised to include a Senior Reactor Operator (SRO) in the quorum for a Risk Management Challenge Board (RMCB) concerning an Operational High Risk activity.

The Integrated Risk Management Procedure will be revised to include individuals directly involved with the high risk portion of an activity as required in the quorum for a RMCB.

PREVIOUS SIMILAR EVENTS

No other similar events were identified with the same underlying concern or reason for occurrence.