05000530/LER-2006-002
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No. 05000 | |
| Event date: | 03-05-2006 |
|---|---|
| Report date: | 04-28-2006 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| Initial Reporting | |
| ENS 42387 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| 5302006002R00 - NRC Website | |
1. REPORTING REQUIREMENT(S):
This LER (50-530/2006-002-00) is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) to report a reactor protection system (RPS) (EIIS: JC) initiated reactor trip which occurred on March 5, 2006 at approximately 07:10 Mountain Standard Time (MST).
On March 5, 2006 at 09:12 MST, APS made notification of the event to the Nuclear Regulatory Commission (NRC) via the event notification system.
(Reference: ENS call 42387)
2. DESCRIPTION OF STRUCTURE(S), SYSTEM(S) AND COMPONENT(S):
The core protection calculator/control element assembly calculator (CPC/CEAC)(EIIS: JC) system monitors pertinent reactor core conditions to provide CEA withdrawal prohibit (CWP) signals to the control element drive mechanism control system (CEDMCS) (EIIS: AA) and provides an accurate, reliable means of initiating a reactor trip. The CPC/CEAC system is an integral part of the plant protective system in that it provides departure from nucleate boiling ratio (DNBR) and local power density (LPD) trip signals to the reactor protection system (RPS) (EIIS: JC). Trip signals are provided to the reactor protection system whenever the minimum DNBR or fuel design limit LPD is approached during reactor operation.
Each CEAC receives reed switch assembly inputs for all control element assemblies (CEAs) (EIIS: AA). The CEACs compare the positions of all CEAs within each CEA subgroup and determine penalty factors based upon CEA deviations within a subgroup. A penalty factor is transmitted via four fiber-optic data links to the CPCs. The CPCs also compute penalties for CEA group out-of-sequence and deviations between subgroup conditions.
The reactor protection system (RPS) provides a rapid and reliable shutdown of the reactor to protect the core and the reactor coolant system pressure boundary from potentially hazardous operating conditions. Shutdown is accomplished by the generation of reactor trip signals. The trip signals open the reactor trip switchgear (RTSG) breakers (EIIS: AA), de-energizing the control element drive mechanism (CEDM) coils (EIIS: AA), allowing all CEAs to drop into the core by the force of gravity.
3. INITIAL PLANT CONDITIONS:
On March 5, 2006 Palo Verde Unit 3 was in Mode 1 (Power Operations), operating at approximately 100 percent power. At the start of the event an unexpected alarm was received for Control Element Assembly Calculator (CEAC) number 1 (CEAC 1). No other major structures, systems, or components were inoperable at the start of the event that contributed to the event.
4. EVENT DESCRIPTION:
On March 5, 2006 at approximately 07:10 MST the Unit 3 reactor tripped from 100% power due to Low DNBR trips on all four CPCs. At 07:04 MST, a CEA Sensor CEAC 1 alarm was received. The Control Room staff addressed the alarm response procedure and determined CEA number 60 input to CEAC 1 was indicating 160.6" withdrawn. CEA number 60 indicated 151.1" withdrawn on CEAC 2. The staff entered procedure 72A0-9SB01 "CEAC Inoperable" and the reactor trip occurred before CEAC INOP codes could be placed into the CPCs.
The control room staff entered the emergency operations procedures and diagnosed a reactor trip, the Control Room Supervisor (CRS) entered site procedure 40EP-9E002 "Reactor Trip".
The event was classified by the Shift Manager as an Uncomplicated Reactor Trip. The Shift Manager and Shift Technical Advisor (STA) reviewed site procedure EPIP-99 and no event classification was required. The plant was stabilized in Mode 3.
The NRC Operations Center was notified of the reactor trip in accordance with 10CFR50.72(b)(2) at 09:12 MST. (ENS 42387)
5. ASSESSMENT OF SAFETY CONSEQUENCES:
The low DNBR trip is provided to prevent the DNBR in the core from exceeding the fuel design limit in the event of design bases anticipated operational occurrences. The reactor trip occurred when all four channels of CPCs calculated a DNBR value that exceeded the low DNBR trip setpoint. The cause of the reactor trip was a hardware induced CEA position deviation error that resulted in a large penalty factor being generated in control element assembly calculator CEAC1. The CPCs calculated DNBR based on the penalty factor generated in CEAC1. The actual DNBR safety limit was not approached nor exceeded.
Primary and secondary pressure boundary limits were not approached due to the reactor tripping from a steady state condition, followed by a "quick open" of the steam bypass control system (EIIS: JI). The transient did not cause any violation of the specified acceptable fuel design limits. Therefore, there were no safety consequences or implications as a result of this event. This event did not adversely affect the safe operation of the plant or health and safety of the public. Unit 3 plant performance and plant protection system evaluations were performed to determine plant responses to transients experienced subsequent to the plant trip. The plant performance evaluation included a safety function impact analysis for each of the safety functions and included an assessment of equipment malfunctions, abnormal alarms and/or events observed during the event. The evaluations revealed that the plant responded as required, the reactor trip was uncomplicated, no safety limits were exceeded, and the event was bounded by current safety analyses.
There are no actual safety consequences as a result of this condition, the condition would not have prevented the fulfillment of the safety function, and the condition did not result in a safety system functional failure as defined by 10 CFR50.73 (a) (2) (v).
6. CAUSE OF THE EVENT:
The direct cause of the reactor trip was an erroneous position indication signal for CEA #60 as sensed by CEAC #1 due to a faulty CEA Positional Isolation Amplifier (CPIA) board. The root cause of the CPIA board failure is categorized as a probable cause of "random electronic failure" as defined by site procedure 70DP-OEE01 (Equipment Root Cause of Failure Analysis) of the U6 operational amplifier.
7. CORRECTIVE ACTIONS:
On March 6, 2006 repairs were made to the CEAC number 1 by replacing and satisfactorily retesting the faulted circuit board that contained the integrated operational amplifier for CEA number 60.
The Unit was started and synchronized to the grid on March 6, 2006.
Additional detailed equipment root cause of failure analysis is in progress as an Equipment Root Cause of Failure Analysis (ERCFA) Level II investigation (CRDR 2873800). If any conclusions from that investigation differ significantly from the information provided above, a supplement to this LER will be provided.
8. PREVIOUS SIMILAR EVENTS:
In the past three years, Palo Verde reported reactor shutdowns initiated by Reactor Protection System but none associated with the same root cause.