05000286/LER-2014-004-01, Regarding Automatic Reactor Trip as a Result of Meeting the Trip Logic for Over Temperature Delta Temperature During Reactor Protection System Pressurizer Pressure Calibration
| ML16224B012 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 08/01/2016 |
| From: | Coyle L Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-16-075 LER 14-004-01 | |
| Download: ML16224B012 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv)(B), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2862014004R01 - NRC Website | |
text
- ~=~* Entergx NL-16-075 August 1, 2016 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-i F1 Rockville, MD 20852-2738 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
SUBJECT:
Licensee Event Report# 2014-004-01, "Automatic Reactor Trip as a Result of Meeting the Trip Logic for Over Temperature Delta Temperature During Reactor Protection System Pressurizer Pressure Testing" Indian Point Unit No. 3 Docket No. 50-286
Reference:
- 1. LER-2014-004-00 submitted by letter NL-14-126, dated October 10, 2014 J
Dear Sir or Madam:
Pursuant to 10 CFR 50. 73(a)(1 ), Entergy Nuclear Operations Inc. (ENO) hereby '
provides Licensee Event Report (LER) 2014-004-01. The attached LER is a revision to an LER submitted by Reference 1, that identified an event where the reactor automatically tripped, which is reportable under 1 O 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-IP3-2014-01903. The root cause was identified as indeterminate. As a result of further troubleshooting, monitoring and vendor analysis the root cause was determined and additional corrective actions identified.
NL-16-075 Docket No. 50-286 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, cc:
Mr. Daniel H Dor an, Regional Administrator, NRG Region I NRG Resident specter's Office, Indian Point 3 Ms. Bridget Frymire, New York State Public Service Commission
Abstract
On August 13, 2014, Instrumentation and Control (I&C) technicians started performance of an 8-hour scheduled surveillance 3-PC-OL04A (Pressurizer Pressure Loop P-455 Channel Calibration) with Loop I in test and tripped.
The test was approved by I&C and operations to be stopped for a break with the bistables still tripped, Channel I in test.
During the break, an automatic reactor trip (RT) occurred as a result of meeting the trip logic for Overtemperature Delta Temperature (OTDT).
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 due to SG shrink effect.
No work was JJeing performed at the time of the RT and no actual OTDT existed.
The direct cause of the RT was a spurious signal spiking on channel 3 of the OTDT circuitry with another channel tripped for testing.
The two possible root causes were 1) a random failure of the OTDT static gain unit (Foxboro Integrator/Converter module QM-431D), 2) loose wiring connection on distribution block DB-4 (output of static gain unit QM-431) due to workmanship issue.
Corrective actions included replacement of static gain module QM-431D and associated PR N-43 isolation amplifies NM306 and NM307, and static gain modules QM-421D, and QM-411D, replacement of three OPDT trip bistables (TC-421 A/B, TC-431 A/B, TC-441 A/B), replacement of setpoint module TM-432B (other setpoint modules had been previously replaced), and replacement of Loop 3 T(avg) E/I converter TM-432R.
Procedure IP-S~M-WM-140 revised to include expectation on minimizing break times when a channel is tripped.
The event had no effect on public health and safety.
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Event Analysis
U.S. NUCLEAR REGULATORY COMMISSION LEA NUMBER (6)
I SEQUENTIAL I REVISION NUMBER NUMBER 004 01 PAGE (3) 5 OF 5
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 an automatic RT was initiated at 11:57 hours, on August 13, 2014, and the AFWS actuated as a result of the RT.
On August 13, 2014, a 4-hour non-emergency notification was made to the NRC at 13:06 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 #50361)
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 testing.
No LERs were identified.
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.
For this RT there was no actual OTDT condition.
There were no significant potential safety consequences of this event.
The 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.
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 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.