05000286/LER-2014-001, Regarding Automatic Reactor Trip as a Result of Steam Flow/Feedwater Flow Mismatch with Low 33 Steam Generator (SG) Water Level Due to the Failure of the 33 SG Feedwater Flow Controller
| ML14077A067 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/04/2014 |
| From: | Ventosa J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-14-021 LER 14-001-00 | |
| Download: ML14077A067 (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 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2862014001R00 - 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 Administration NL-14-021 March 4, 2014 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738
SUBJECT:
Licensee Event Report # 2014-001-00, "Automatic Reactor Trip as a Result of Steam Flow/Feedwater Flow Mismatch with Low 33 Steam Generator (SG) Water Level Due to the Failure of the 33 SG Feedwater Flow Controller" Indian Point Unit No. 3 Docket No. 50-286 DPR-64
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2014-001-00. The attached LER identifies an event where the reactor automatically 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-IP3-2014-00054.
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, cc:
Mr. William Dean, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 3 Ms. Bridget Frymire, New York State Public Service Commission t%~I
Abstract
On January 6, 2014, the 33 Steam Generator (SG) steam flow (SF) Feedwater flow (FF) mismatch and SG level control deviation alarms annuniciated.
Operators noticed the selected feedwater (FW) flow (FF) channel at zero and swapped to the alternate channel and noted both FW flow channels at zero.
With the 33 SG level at approximately 10% and lowering, operators started actions to initiate a manual reactor trip (RT) when an automatic RT occurred on a SF/FF mismatch coincident with low 33 SG level.
All control rods fully inserted and all required safety systems functioned properly with the exception of the intermediate range excore neutron flux detector N-35 which was undercompensated.
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.
The Emergency Diesel Generators did not start as offsite power remained available and stable.
Investigations determined the decreasing SG levels was due to reduced main FW flow as a result of a closed 33 Feedwater Regulating Valve (FRV).
The closure of the 33 FRV-437 was due to no output from flow controller FC-437.
The apparent cause was station personnel failed to apply lateral thinking and teamwork and question what other options were available besides procuring an identical controller.
Corrective actions included replacing FC-437 and FC-417.
A project to improve procurement and ease replacement of NUS modules will be developed.
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)
This event meets the reporting criteria because an automatic RT was initiated at 21:15 hours, on January 6, 2014, and the AFWS actuated as a result of the RT.
On January 6, 2014, a 4-hour non-emergency notification was made to the NRC at 21:55 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 #49698).
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 reduction.
No LERs were identified that reported a RT due to a FW events.
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.
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.
In addition, Operators for this event anticipated a possible low SG level and could have initiated 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.
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.