05000286/LER-2013-005, Regarding Automatic Actuation of Safety Injection and Engineered Safety Features During Reactor Protection System Functional Testing
| ML13157A002 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 05/23/2013 |
| From: | Ventosa J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-13-083 LER 13-005-00 | |
| Download: ML13157A002 (5) | |
| 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 |
| 2862013005R00 - NRC Website | |
text
B En t egy 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-13-083 May 23, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001
SUBJECT:
Licensee Event Report # 2013-005-00, "Automatic Actuation of Safety Injection and Engineered Safety Features During Reactor Protection System Functional Testing" 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) 2013-005-00. The attached LER identifies an event where the safety injection system actuated along with engineered safety features during Reactor Protection System functional testing, which is reportable under 10 CFR 50.73(a)(2)(iv)(A). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP3-2013-02115.
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, JAV/cbr 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 LEREvents@inpo.org
Abstract
On March 27, 2013, while in Mode 3 during startup from a refueling outage, an inadvertent safety injection (SI) occurred during reactor protection logic channel functional testing.
The test procedure noted that pressurizer pressure must be above 1930 psig to perform the test but signals may be installed to simulate the condition.
In accordance with Appendix A of the test, helipot installation is directed for four pressurizer (PZR) pressure channels to allow PZR pressure signals to be simulated.
A helipot was first installed in PT-455 and then simulated pressure raised above 2000 psi.
A second helipot was installed for PT-456 and simulated pressure was raised above 2000 psi.
SI is automatically enabled with 2 channels above 1900 psig but one of the two out of three pressure signals remained below the 1720 psig signal required for SI actuation.
During this time the plant is susceptible to an SI actuation if one of the inputted signals is lost.
Before pressure could be simulated in the remaining transmitter PT-457, the test lead connected to PT-456 failed open causing the system to register a 2/3 PZR low pressure satisfying the logic for SI actuation.
The root causes were 1) The test procedure did not contain steps to preclude a single point failure, 2) Manufacturing defect of a test equipment lead, 3)
Insufficient evaluation of the risk of when the test was being performed.
Corrective actions include removing the defective lead from service, inspecting and testing remaining leads.
Procedure 3-PT-MI3B will be revised to remove single point vulnerability.
A Training Evaluation & Action Request will be initiated to create a Case Study for this event which will include information associated with performing in depth risk assessments when surveillance tests are added to or removed during an outage.
A TEAR will incorporate the Study into operations training, and a review will be performed on acceptability of testing in Mode 3.
The event had no significant 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 an SI actuation was initiated.
The SI signal also initiated a safeguards equipment signal, auxiliary feedwater system actuation, containment ventilation isolation, and containment phase A isolation.
The safeguards equipment signal initiated start of the Emergency Diesel Generators (EDGs),
Reactor Protection System (RPS) actuation, and containment fan cooler unit (FCU) actuation.
The EDGs started but did not load as offsite power remained available.
An immediate notification was provided by EN 48854 on March 27, 2013 at 7:30 hours.
Past Similar Events A review of the past three years of Licensee Event Reports (LERs) for events that involved inadvertent SI actuation did not identify any events.
Safety Significance
This event had no significant effect on the health and safety of the public. There were no actual safety consequences for the event because there were no accidents or transients during the time of the event and the SI actuation was not in response to an actual condition or event requiring mitigation.
Due to the SI signal, the SI pumps automatically started and water from the Refueling Water Storage Tank was injected into the RCS cold legs.
Actual pressurizer level was maintained less than 100% and the steam bubble along with RCS pressure control was maintained.
The pressurizer safety valves did not lift during the transient.
For this event SI was terminated with RCS temperature at approximately 390 degrees F which is above the 10CFR50, Appendix G limit of 380 degrees F.
The primary to secondary side differential pressure limit of 1700 psid was not exceeded as a result of the event.
Following the inadvertent SI, the pressurizer heatup rate exceeded the 100 degree F1 hour limit associated with Technical Requirements Manual (TRM)
TRO 3.4.D during system restoration.
An evaluation of this condition concluded the structural integrity of the pressurizer was acceptable.