05000331/LER-2009-004
| Document Numbersequentia Revmonth Day Year Year Month Day Year 05000L Number No. | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3312009004R00 - NRC Website | |
I.�Description of Event:
On October 8, 2009, while operating at 100% power, the following Limiting Conditions for Operations were in effect:
1. TS 3.7.5, Required Action A.1, 30 day completion time for B Control Building Chiller 2. TS 3.7.2, Required Action A.1, 7 day completion time for B River Water Supply Loop 3. TS 3.3.3.1, Function 6, Required Action A.1, 30 day completion time for position indication of MO 1949B, B Residual Heat Removal Heat Exchanger Vent 4. TS 3.7.1, Required Action A.1, 30 day completion time for 1P22D, D Residual Heat Removal Service Water Pump.
At 1014, Surveillance Test Procedure (STP) 3.3.3.2-09B, Reactor Water Level and Pressure Instruments (Loop B) Calibration, was commenced. Following the performance of a pre-job briefing, Instrument and Control (IC) technicians proceeded to the instruments to be calibrated. As the technicians progressed through the STP they eventually reached STP step 7.1.62 where they were directed to close Reactor Vessel Pressure Transmitter (PT) Isolation valve, PT4564-V-92. Following the completion of Step 7.1.62, the technicians proceeded to and completed step 7.1.63 which required removal of a test fitting tubing cap, connection of test equipment, and the opening of PT4564-V-91, PT4564 Drain/Test Isolation Valve. Upon completion of opening V-91, the technicians noted that the test equipment hand pump hose suddenly pressurized. At 1420 on October 8, 2009, an automatic reactor shutdown occurred. When the technicians heard the scram solenoids de-energize and observed the hand pump tube jump, one of the technicians returned to PT4564-V-92 and manipulated it an additional 1/8 turn in the closed direction.
II. Assessment of Safety Consequences:
There were no structures, systems, or components (SSCs) that failed during this event. Primary Containment Isolation System (PCIS) Groups 2, 3 and 4 were received due to reactor water level dropping below 170 inches due to the expected level shrink immediately following the reactor scram.
All isolations went to completion. Reactor water level was subsequently returned to normal as expected under post scram conditions.
This event did not result in a Safety System Functional Failure.
Therefore, the reactor scram did not result in any radiological or nuclear concern which would impact the health and safety of the public.
III. Cause of Event:
A Root Cause Evaluation (RCE 1086) was completed for this event. The RCE determined that the shutdown was caused by the failure to close an instrument isolation valve for a Reactor Vessel Pressure Transmitter PT4564. The failure to close this valve resulted in creating a sensed low reactor water level on Reactor Protection System (RPS) channels A2 and B2, and thus resulted in the automatic reactor shutdown. The specific root cause and contributing factors are as follows:
Root Cause:
Defenses in depth were inadequate to prevent the plant transient, when the IC Technician did not completely close PT4564 V-92 during the performance of Step 7.1.62 of STP 3.3.3.2-09B.
Contributing Factors:
The requirements of the DAEC Technical Procedure Writing Standards related to Critical Steps were not implemented in STP 3.3.3.2-09B.
At the time of issuance of INPO 06-003, Human Performance Reference Manual, administrative procedures for implementing the site's Operating Experience (OE) Program did not require formal review of INPO guideline documents. The document was reviewed outside the OE Program, and actions were implemented to meet the intent of INPO 06-003. The implementation of the critical steps missed the specific area of removing the instrument from service, and focused primarily on the issues identified in OE on returning the instrument to service.
IV. Corrective Actions:
Immediate Actions Taken:
Immediately following the reactor shutdown the site's Maintenance Department initiated a Human Performance investigation, collected written statements from the IC technicians involved with the STP and held an IC Department stand-down to share information regarding the event and to reinforce in-plant job performance expectations.
The plant shutdown was documented in corrective action document CAP 70334, and from this corrective action document Condition Evaluation (CE) 7739, Corrective Action (CA) 53475, and Root Cause Evaluation (RCE) 1086 were initiated.
CE 7739 was written to conduct an initial evaluation under the corrective action program to document the probable cause, immediate, and interim corrective actions for reactor shutdown.
CA 53475 was written to require, as an interim corrective action, IC Maintenance, prior to commencing any STPs, to conduct a review to identify any STPs that include instrument calibrations with shared reference legs. Any that are identified will have the steps that isolate instruments marked as requiring additional verification.
In addition, plant management has determined that all upcoming surveillance testing that is performed on reactor vessel level shared instrument legs (piping) will be regarded as 'high risk' until such time as the corrective actions for RCE 1086 have been completed.
Corrective Actions to Address Root Cause and Contributing Factors:
1. Revise the STPs identified in CE 7778 to require pre-pressurization of the instrument test lines.
2. Revise STP 3.3.3.2-09B to identify critical steps as defined in Technical Procedure Writing Standards.
V. Additional Information:
Previous Similar Occurrences:
A review of LERs over the previous 5 years revealed the following similar occurrences:
During Performance of a Surveillance Test EIIS System and Component Codes:
Primary Containment Isolation System--JM Engineered Safety Features Actuation System--JE Level Transmitter--LT Level Indicating Switch--LIS Reporting Requirements:
This event is reportable under 10 CFR 50.73(a)(2)(iv)(A).