05000247/LER-2013-003, Manual Reactor Trip Due to Decreasing Steam Generator Water Levels Due to Loss of Main Feedwater (FW) Flow Caused by a Loss of Instrument Air to the FW Regulating Valves

Manual Reactor Trip Due to Decreasing Steam Generator Water Levels Due to Loss of Main Feedwater (FW) Flow Caused by a Loss of Instrument Air to the FW Regulating Valves

ML13247A173
Person / Time
Site:	Indian Point
Issue date:	08/27/2013
From:	Ventosa J Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-13-113 LER 13-003-00
Download: ML13247A173 (5)

LER-2013-003, Manual Reactor Trip Due to Decreasing Steam Generator Water Levels Due to Loss of Main Feedwater (FW) Flow Caused by a Loss of Instrument Air to the FW Regulating Valves

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 10 CFR 50.73(a)(2)(v), Loss of Safety Function
2472013003R00 - NRC Website

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-Entergy Indian Point Energy Center 450 Broadway, GSB3 P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 John A. Ventosa Site Vice President NL-13-113 August 27, 2013 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738

SUBJECT:

Licensee Event Report # 2013-003-00, "Manual Reactor Trip Due to Decreasing Steam Generator Water Levels Due to Loss of Main Feedwater (FW) Flow Caused by a Loss of Instrument Air to the FW Regulating Valves" 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) 2013-003-00. 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 and two Main Steam Isolation Valves closed, 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-1IP2-2013-02717.

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 Ms. Bridget Frymire, New York State Public Service Commission

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Abstract

On July 3,'2013, operators initiated a manual reactor trip as a result of lowering steam generator (SG) levels due to the loss of feedwater (FW) from the trip of both main FW pumps.

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.

The Auxiliary Feedwater System automatically started as expected.

Investigations determined the decreasing SG levels were due to a loss of main FW flow as a result of the closure of the FW regulating valves.

The FW regulating valves closed due to a loss of instrument air (IA) pressure.

The IA pressure was lost when a two inch copper IA tubing in the 22 Main Transformer moat separated at a soldered coupling.

Prior to the event piping lines including the IA line buried in the main transformer moat were excavated and temporary supports installed.

The apparent cause was poor legacy workmanship assembling the IA tubing coupling during original plant construction.

The IA tubing was not fully inserted into the coupling resulting in reduced joint strength.

Corrective actions included reassembly and soldering of the IA joint with full insertion, acoustic emission and snoop testing on repaired coupling.

Axial and thrust restraints were installed on the IA line in the moat.

A caution was placed in the Buried Piping Program database associated with buried copper tubing identifying the potential for the separation of soldered joints when the line is excavated and the need for restraints or other contingencies to minimize the probability of a line separation.

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 NRC Form 366A) (17)

This event meets the reporting criteria because a manual RT was initiated at 07:41 hours, on July 3, 2013, and the AFWS actuated as a result of the RT.

Two of four MSIVs closed (23 and 24 MSIVs) due to the loss of IA thereby qualifying as multiple MSIV actuations.

On July 3, 2013, a 4-hour non-emergency notification was made to the NRC at 10:46 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 and closure of multiple MSIVs (Event Log #49171).

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 Licensee Event Reports (LERs) reporting a RT as a result of main FW reduction.

The review identified LER-2009-002.

LER-2009-002 reported a manual RT on April 3,

2009, due to decreasing SG levels caused by loss of the 21 Main FW pump and failure of the main turbine to automatically runback.

The direct cause was failure of the autostop oil tubing/Swagelock fitting on the main FW autostop oil header.

The root causewas improper tubing installation due to poor worker practices.

This LER is similar as it involved tubing and a failure due to improper installation.

The corrective actions for LER-2009-002 included replacement of the fractured tubing, reconfiguration to meet installation requirements and training on Swagelock fitting installation.

The corrective actions for that event would not have prevented this event as the issue was with a Swagelock fitting and stainless steel tubing and was internal to equipment and not a support issue with excavated buried copper tubing soldered to a coupling.

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.

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.