05000247/LER-2013-001, Regarding 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
| ML13114A263 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 04/15/2013 |
| From: | Ventosa J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-13-049 LER 13-001-00 | |
| Download: ML13114A263 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2472013001R00 - NRC Website | |
text
-Entergy Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 John A. Ventosa Site Vice President NL-13-049 April 15, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001
SUBJECT:
Licensee Event Report # 2013-001-00, "Manual Reactor Trip as a Result of Decreasing Steam Generator Water Levels Caused by the Trip of Both Heater Drain Tank (HDT) Pumps During AOV Diagnostic Testing" Indian Point Unit No. 2 Docket No. 50-247 DPR-26
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2013-001-00. The attached LER identifies an event where the reactor was manually tripped, which is reportable under 10 CFR 50.73(a)(2)(iv)(A). As a result of the reactor trip, the Auxiliary Feedwater System was actuated, which is also reportable under 10 CFR 50.73(a)(2)(iv)(A). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP2-2013-00721.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 254-6710.
Sincerely, cc:
Mr. William Dean, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 2 Ms. Bridget Frymire, New York State Public Service Commission LEREvents@inpo.org
.J.
Abstract
On February 13, 2013, operators initiated a manual reactor trip (RT) as a result of lowering steam generator (SG) levels.
All control rods 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 automatically started as expected due to SG low level from shrink effect.
An investigation determined the decreasing SG levels was due to reduced main feedwater (FW) flow from a loss of Heater Drain Tank (HDT) pumps.
The HDT pumps tripped during valve diagnostics on HDT level control valve LCV-1127B which resulted in HDT Large Dump valves failing open.
The open HDT large dump valves resulted in low HDT level and trip of the HDT pumps.
The HDT large Dump valves failed open when the current/pressure (I/P) lead was lifted during air operated valve (AOV) diagnostics per procedure 0-IC-PC-AOV.
Loss of HDT flow to the main feedwater pumps (FWPs) caused the FWPs speed controller cutback to reduce FW flow to the SGs.
The root cause (RC) was inadequate procedure design and content.
Corrective actions from the RC will be to revise Maintenance procedures 0-IC-PC-AOV and 0-VLV-404-AOV to: 1) 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,
- 2) Eliminate the subject Caution block, and 3)
Include signature blocks for review of drawings and validation that lifting a I/P lead or disconnecting the instrument tubing will not affect any other valve or component.
The event had no effect on public health and safety.
(If more space is required, use additional copies of (if more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17)
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.