05000247/LER-2013-001

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LER-2013-001, Manual Reactor Trip as a Result of Decreasing Steam Generator Water Levels Caused by the Trip of Both Heater Drain Tank Pumps During AOV Diagnostic Testing
Indian Point 2
Event date: 02-13-2013
Report date: 04-15-2013
Reporting criterion: 10 CFR 50.73(a)(2)(iv)(B), System Actuation

10 CFR 50.73(a)(2)(iv)(A), System Actuation

10 CFR 50.73(a)(2)(v), Loss of Safety Function
2472013001R00 - NRC Website

Note: The Energy Industry Identification System Codes are identified within the brackets { }

DESCRIPTION OF EVENT

On February 13, 2013, while at approximately 85% reactor power during power reduction, operators initiated a manual reactor trip (RT) {JC} at 13:55 hours, as a result of lowering steam generator (SG) {AB} levels. All control rods {AA} fully inserted and all required safety systems functioned properly. The plant was stabilized in hot standby with decay heat being removed by the main condenser {SG}. The Auxiliary Feedwater System {BA} automatically started as expected due to SG low level from shrink effect. There was no radiation release. The Emergency Diesel Generators {EK} did not start as offsite power remained available. The event was recorded in the Indian Point Energy Center corrective action program (CAP) as CR-IP2-2013-00721. A post trip evaluation was initiated and completed on February 13, 2013.

Prior to the event, Instrumentation and Control (I&C) personnel were performing air operated valve (AOV) diagnostics per procedure 0-IC-PC-AOV (Use of Air Operator Valve Diagnostics) which directed lifting the Current/Pressure (I/P) lead to Heater Drain Tank (HDT) {SN} level control valve LCV-1127B {LCV}. When the lead was lifted valves LCV-1127C and LCV-1127D failed open. LCV-1127C and LCV-1127D are large dump valves for drain lines to the condensers. The open dump valves resulted in draining the HDT causing a low level in the HDT that resulted in a trip signal to the heater drain pumps {P}. Loss of heater drain pump flow to the suction of the main feedwater {SJ} pumps was sensed by the feedwater pump suction header pressure transmitter and provided an input signal to the feedwater pump speed control system {JK} (low suction pressure cutback) reducing FW pump discharge flow to the SGs.

On February 13, 2013, Control Room operators received a hotwell low level alarm followed by a hotwell low low level alarm at 13:50 hours. At 13:52 hours, both HDT pumps tripped and operators entered abnormal operating procedure 2-A0P-FW-1 (Loss of Main Feedwater) and commenced turbine load reduction. As a result of decreasing SG levels and the inability to maintain SG levels, operators initiated a manual RT at 13:55 hours and entered emergency operating procedure EOP 2-E-0 (Reactor Trip or Safety Injection) and transitioned to 2-ES-0.1 (Reactor Trip Response) at 14:03 hours.

Equipment that did not perform properly included 1) neutron source range detector N-31 (failed low), 2) neutron intermediate range detector N-35 (pegged low), and 3) 22 Main FW pump High Pressure stop valve did not close.

During normal operation approximately 65 percent of main feedwater (FW) pump suction water is supplied by condensate flow from the low pressure heaters {SM} and approximately 35 percent is provided by Heater Drain Tank (HDT) pumps discharge. The HDT and its two heater drain pumps are located on the 15 foot evaluation of the Turbine Building {NM}. The primary function of the extraction steam and heater drains and vents system is to extract steam from the turbines at various stages, condense it in the FW heaters and return it to the feed cycle through a cascading drain system. The heater drain pumps take suction from the HDT and pump the contents of the HDT to the suction of the main FW pumps.

The HDT system collects high temperature drains {SN} from the three number 26 FW heaters, the three number 25 FW heaters, and the six moisture separator drain tanks. FW heaters 25A, 25B and 25C drain through a loop seal to the HDT without level control.

The HDT is vented back to the 25 FW heaters as drainage is by gravity since the pressure in these heaters and the HDT are essentially equal.

Cause of Event

The direct cause of the manual RT was lowering SG levels and the inability to maintain SG levels. The decreasing SG levels were due to reduced FW flow from low FW suction pressure. The low FW suction was a result of the loss of HDT pump flow due to the trip of both HDT pumps. Loss of both HDT pumps was caused by a trip signal due to low HDT level. All three HDT large dump valves failed open after de-terminating the I/P leads on valve LCV-1127B in accordance with Work Order (WO) 52202988 during performance of Preventive Maintenance (PM) with AOV Diagnostics per procedure 0-IC-PC-AOV. The open HDT large dump valves drained the HDT resulting in a low level in the HDT.

The root cause was inadequate procedure design and content. The procedural steps for equipment setup (specifically Section 4.2) in procedure 0-IC-PC-AOV are conditional and allow changes in testing work scope to be made in the field without proper review prior to performing work. Additionally, the Caution note in the procedure did not provide adequate instructions to field personnel on how to verify no other components were affected. Also, the procedure layout was not adequate for reviewers to determine which test methodology was to be used (installed I/P or test I/P).

Corrective Actions

The following corrective actions have been or will be performed under Entergy's Corrective Action Program to address the cause and prevent recurrence:

  • Maintenance procedures 0-IC-PC-AOV and 0-VLV-404-AOV will be revised to: 1) Eliminate conditional steps for equipment setup (Section 4.2) that allows changes to work scope to be made in the field without proper review prior to performing work, 2) Eliminate the subject Caution block, and 3) Include signature blocks that state Instrumentation & Control technician or mechanic has reviewed the I/P signal and instrument air drawings and has validated that lifting the I/P signal lead or disconnecting the instrument tubing will not affect any other valve or component.
  • Maintenance procedure EN-MA-143 will be revised to eliminate conditional steps for equipment setup that allows changes to work scope to be made in the field without proper review prior to performing work and the Caution note will be revised to provide clear direction to field personnel performing AOV diagnostics for this component.
  • Maintenance procedure 0-VLV-491-ACT will be revised to eliminate conditional steps for equipment setup that allows changes to work scope to be made in the field without proper review prior to performing work and the Caution note will be revised to provide clear direction to field personnel performing AOV diagnostics for this component.
  • Work Order packages requiring the lifting of leads will contain electrical prints/drawings related to those leads being lifted in the work package. Proper expectations were developed and communicated for all planners to consistently apply the new work package creation criterion.
  • AOV PMs will be converted, where I/P and/or other components are calibrated, to work items with equipment lists to ensure all components are clearly within scope of the PM.

Event Analysis

The event is reportable under 10CFR50.73(a)(2)(iv)(A). The licensee shall report any event or condition that resulted in manual or automatic actuation of any of the systems listed under 10CFR50.73(a)(2)(iv)(B). Systems to which the requirements of 10CFR50.73(a)(2)(iv)(A) apply for this event include the Reactor Protection System (RPS) including RT and AFWS actuation. This event meets the reporting criteria because a manual RT was initiated at 13:55 hours, on February 13, 2013, and the AFWS actuated as a result of the RT. On February 13, 2013, a 4-hour non-emergency notification was made to the NRC at 16:41 hours, for an actuation of the reactor protection system {JC} while critical and included an 8-hour notification under 10CFR50.72(b)(3)(iv)(A) for a valid actuation of the AFW System (Event Log #47999).

As all primary safety systems functioned properly there was no safety system functional failure reportable under 10CFR50.73(a)(2)(v).

Past Similar Events

A review was performed of the past three years for Licensee Event Reports (LERs) reporting a RT as a result of main FW transient. The review identified LER-2010-007.

LER-2010-007 reported an automatic RT on September 3, 2010, due to a turbine trip as a result of a high SG level after transition to single FW pump operation. The root cause was inadequate design control of the proportional band and reset settings of the main FW pump speed controller. The cause of the event reported in LER-2010-007 is different from this event therefore, the corrective actions for that event would not have prevented this event.

Safety Significance

This event had no effect on the health and safety of the public.

There were no actual safety consequences for the event because the event was an uncomplicated reactor trip with no other transients or accidents. Required primary safety systems performed as designed when the RT was initiated. The AFWS actuation was an expected reaction as a result of low SG water level due to SG void fraction (shrink), which occurs after a RT and main steam back pressure as a result of the rapid reduction of steam flow due to turbine control valve closure.

There were no significant potential safety consequences of this event. Operators for this event anticipated a possible low SG level and actuated a manual RT. The manual actuating devices are independent of the automatic trip circuitry and are not subject to failures which make the automatic circuitry inoperable. There are two manual trip buttons, one located on flight panel FCF and the other on safeguards supervisory panel SBF2. Either one of these buttons will directly energize the trip coils of the reactor trip and bypass breakers in addition to de-energizing the undervoltage coils of the reactor trip and bypass breakers. The Reactor Protection System (RPS) is designed to actuate a RT for any anticipated combination of plant conditions to include low SG level. The reduction in SG level and RT is a condition for which the plant is analyzed. A low water level in the SGs initiates actuation of the AFWS. Redundant safety SG level instrumentation was available for a low SG level actuation which automatically initiates a RT and AFWS start providing an alternate source of FW. The AFW System has adequate redundancy to provide the minimum required flow assuming a single failure.

The analysis of a loss of normal FW (UFSAR Section 14.1.9) shows that following a loss of normal FW, the AFWS is capable of removing the stored and residual heat plus reactor coolant pump waste heat thereby preventing either over pressurization of the RCS or loss of water from the reactor. For this event, rod control was in automatic and all rods inserted upon initiation of a RT. The AFWS actuated and provided required FW flow to the SGs. RCS pressure remained below the set point for pressurizer PORV or code safety valve operation and above the set point for automatic safety injection actuation. Following the RT, the plant was stabilized in hot standby.