05000247/LER-2015-003, Regarding Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure
| ML16041A018 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/03/2016 |
| From: | Coyle L Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-16-010 LER 15-003-00 | |
| Download: ML16041A018 (8) | |
| 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 |
| 2472015003R00 - NRC Website | |
text
En tergy Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 Lawrence Coyle Site Vice President NL-16-010 February 3, 2016 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738
SUBJECT:
Licensee Event Report # 2015-003-00, 'Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure" 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) 2015-003-00. The attached LER identifies an event where there was a manual reactor trip due to indication of multiple dropped control rods. This event 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-IP2-201 5-05458.
NL-1 6-010 Page 2 of 2 There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Regulatory Assurance at (914) 254-6710.
Sincerely, Attachment: LER-201.
cc:
Mr. Daniel H. Dorman, Regional Administrator, NRC Region I NRC Resident Inspector's Office Ms. Bridget Frym ire, New York State Public Service Commission
Abstract
On December 5,
- 2015, control room operators initiated a manual reactor trip (RT) after observing indications consistent with multiple dropped control rods (CR) following an alarm for the trip of Motor Control Center (MCC)-24/24A.
No Control Rod indication was available due to MCC-24 being de-energized.
All primary safety sYstems functioned properly except the primary rod control cabinet power supply (PSI) which was in a
degraded condition prior to the event and failed to function as required.
The plant was stabilized in hot standby.
There was no radiation release.
The Auxiliary FW system automatically started as designed.
The direct cause of the event was loss of MCC-24 due to an internal fault at the line side leads at cubicle 2H where they connect to the bucket stab assemblies (load side fault).
This caused the supply breaker feed to open per design and clear the fault.
The de-energization of MCC-24 removed the functioning backup Control Rod (CR) power supply and the remaining degraded primary power supply failed to function as required.
The apparent cause was an unanticipated loss of power to the CR system due to the degradation of the primary CR power supply (PSl) which failed to function when the operating power supply (PS2) was lost.
MCC-24/24A was lost due to a design error that allowed the positioning of a mounting plate too close and obstructing the line side wiring resulting in contact.
Vibration over time resulted in degraded wiring insulation which eventually shorted.
Corrective actions included inspection and testing of the MCC-24 bus and control wiring.
The degraded Rod Contrbl power supply (PSl) was replaced.
Maintenance procedures will be revised to provide imore in-depth inspection criteria.
The event had no effect on public health and safety.
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There were no significant potential safety consequences of this event.
The Reactor Protection System (RPS) is designed to actuate a RT for any anticipated combination of plant conditions, when necessary, to ensure a minimum departure from nucleate boiling (DNB) ratio (DNBR) equal to or greater than the applicable safety analysis limit DNBR.
The RPS monitors parameters related to safe operation and trips the reactor to protect the reactor core against fuel rod cladding damage caused by DNB, and to protect against reactor coolant system damage caused by high system pressure.
DNB is prevented by the RPS by monitoring plant variables affecting DNB [i.e.,
thermal power, coolant flow, coolant temperature, coolant pressure, core power distribution (hot channel factors)] and initiating a RT when applicable limits are reached.
Plant parameters used to protect against DNB include the Overtemperature Delta Temperature
- trip, the Low Pressurizer Pressure trip to prot'ect against excessive core voids that could lead to DNB, and the Overpower Delta Temperature trip to protect against excessive power (fuel rod rating protection) all of which initiate a RT.
In addition, a manual RT can be initiated by control room operators based on two independent systems that are provided to sense dropped rods; 1) a rod bottom position detection system and,
- 2) a system that uses ex-core power range detectors which senses sudden reduction in ex-core neutron flux.
Dropped Rods will rapidly depress the local neutron flux which will be detected by one of the four ex-core detectors.
The reactivity control system is composed of RCCAs divided into control banks and shutdown banks.
The control banks are used in combination with chemical shim (boric acid) control to provide control of reactivity changes.
The shutdown banks are provided to supplement the control banks of RCCAs to make the reactor at least 1.3 percent subcritical following RT from any, credible operating condition assuming the most reactive RCCA is in the fully withdrawn position.
Sufficient shutdown capability is provided so that the minimum DNBR is equal to or greater than the applicable safety analysis limit DNBR, assuming the most reactive rod to be in the fully withdrawn position for the most severe anticipated cooldown transient associated with a single active failure.
The RPS is designed so that the most probable modes of failure in each protection channel result in a signal calling for the protective trip.
The RPS design is of sufficient redundancy and independence to assure that no single failure or removal from service of any component or channel will result in loss of the protection function.
The protection system design is to fail into a safe state or state established as tolerable.
Rapid reactivity shutdown is provided by the insertion of RCCAs by gravity fall.
Duplicate series-connected circuit breakers supply all power to the control rbd drive mechanisms.
The reactor uses magnetic-type control rod drive mechanisms which must be energized for the RCCAs to remain withdrawn from the core.
The RCCAs fall by gravity into the core upon loss of power to the control rod drive mechanism coils.
RT breakers (RTB) which provide power to the control rod drive mechanism coils and are opened by undervoltage coils on both RTBs (normally energized), become de-energized by any of several RT signals.
The electrical state of the devices providing signals to the circuit breaker undervoltage trip coils is such as to cause these coils to trip the breaker in the event of RT or power loss.
RT is implemented by interrupting power to the magnetic latch mechanisms on all control rod drives allowing the RCCAs to insert into the core by gravity.
In addition to automatic RT by the RPS, manual RT is also available.
Manual RT for multiple dropped rods is required by plant procedures and operator training includes scenarios of multiple dropped rods.
The manual RT actuating devices are independent of the automatic trip circuitry and are not subject to failures that could make the automatic circuitry inoperable.
All components in the RCS were designed to withstand the effects of cyclic loads due to reactor system temperature and pressure changes.
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 S~s.
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.