05000325/LER-2005-003
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No 05000 | |
| Event date: | 04-20-2005 |
|---|---|
| Report date: | 06-16-2005 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3252005003R00 - NRC Website | |
INTRODUCTION
On April 20, 2005, Unit 1 was in Mode 5, having been shut down on April 15, 2005, for a fuel bundle replacement outage. Following completion of maintenance to the position indicating probe (PIP) for control rod [AA] 46-15, a PIP position simulator, being used to provide full-in indication to the one- rod-out interlock, was not removed prior to performing single control rod subcriticality testing and operability testing for control rod 46-15. Following maintenance to the PIP for control rod 46-15, plant operators determined that performance of single control rod subcriticality testing was a prerequisite for performing operability testing of control rod 46-15. Operators incorrectly used Technical Specification 3.0.5 to return control rod 46-15 to service before beginning performance of the single control rod subcriticality testing. Following single control rod subcriticality testing for two control rods and encountering problems with the operability testing for control rod 46-15, operators continued with testing of one additional control rod before it was determined that the PIP position simulator was still installed, thus resulting in a failure to comply with Technical Specification requirements. This condition was determined to be reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by the Technical Specifications.
EVENT DESCRIPTION
Initial Conditions On April 15, 2005, Unit 1 was shut down for a fuel bundle replacement outage. Following diagnostic testing, it was determined that the PIP for control rod 46-15 required replacement. Control rod 46-15 was fully inserted and disarmed.
Technical Specification 3.9.4, "Control Rod Position Indication," requires the "full-in" position indication channel for each control rod to be operable while in Mode 5. In addition, Technical Specification 3.9.2, "Refuel Position One-Rod-Out Interlock," requires the refuel position one-rod-out interlock to be operable when in Mode 5 with the mode switch in the refuel position when any control rod is withdrawn.
In order to allow fuel replacement activities to continue, a PIP position simulator was installed. The PIP position simulator provided a full-in indication to the one-rod-out interlock for control rod 46-15, allowing the refueling bridge to move over the core.
Discussion The PIP for control rod 46-15 was replaced on April 18, 2005. On April 19, 2005, at the beginning of the night work shift, a senior reactor operator (SRO) from the outage organization was dispatched to the control room to assist in the performance of control rod testing in accordance with plant procedure OPT-90.2, "Friction Testing of Control Rods." At that time, the PIP for control rod 46-15 was still considered inoperable and the limiting condition for operation (LCO) of Technical Specification 3.9.4 was being satisfied by Required Actions A.2.1 and A.2.2 of Technical Specification 3.9.4 (i.e., control rod 46-15 was fully inserted and disarmed). A review by the SRO determined that the work order to repair or replace the PIP for control rod 46-15 was still open, testing to demonstrate the operability of control rod 46-15 had not yet been performed, and equipment controls to disarm control rod 46-15 remained in effect. The SRO contacted the Instrumentation & Control (I&C) group to determine the status of the work order.
An I&C Supervisor incorrectly determined, based on review of the work order and discussion with an I&C technician involved, that replacement of the PIP had been completed and that the PIP position simulator had been removed. No field verification was performed to confirm the status of the work order. This resulted in incorrect information being provided to the SRO assigned to assist with the performance of control rod testing, who processed the work order as complete.
Based on completion of the work order for control rod 46-15, the SRO prepared to perform testing to demonstrate the operability of control rod 46-15 in accordance with plant procedure OPT-90.2, "Friction Testing of Control Rods." The SRO correctly determined that Technical Specification 3.0.5 would allow re-arming of control rod 46-15 for the purpose of testing that control rod. Technical Specification 3.0.5 allows equipment that has been removed from service or declared inoperable to comply with Technical Specification Actions to be returned to service, under administrative controls, solely to perform testing required to demonstrate operability of the equipment.
However, a prerequisite of plant procedure OPT-90.2 requires that a reactor subcriticality test be performed in accordance with plant procedure OFH-11, "Refueling," if this testing is being performed following an initial core reload. The reactor subcriticality test is referenced as the "single control rod subcriticality test" in step 7.2.15 of plant procedure OFH-11. The reactor engineer recommended that two control rods, 14-27 and 26-15, be tested to satisfy the prerequisite in plant procedure OPT-90.2 for the reactor subcriticality test.
The SRO assisting with control rod testing then asked the unit SRO to independently assess the use of Technical Specification 3.0.5. Both SROs incorrectly determined that Technical Specification 3.0.5 allowed re-arming of control rod 46-15 for the purpose of completing prerequisite single rod subcriticality testing to establish operability of control rod 46-15. In fact, the single control rod subcriticality testing could have been performed without use of Technical Specification 3.0.5 to re-arm control rod 46-15. The one-rod-out interlock would have allowed individual selection and withdrawal of the control rods necessary for the single control rod subcriticality test and, therefore, use of Technical Specification 3.0.5 was not necessary.
At 0330 hours0.00382 days <br />0.0917 hours <br />5.456349e-4 weeks <br />1.25565e-4 months <br /> on April 20, 2005, control rod 46-15 was re-armed for the testing. Single control rod subcriticality testing was then individually performed on control rods 14-27 and 26-15 by withdrawing each control rod from position 00 to position 48 and then inserting each control rod from position 48 back to position 00. At 0515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br />, testing for control rod 46-15 began. When operators attempted to move control rod 46-15, the green full-in indicator light remained lit and the position indication stayed on position 00. Additional attempts were made to move control rod 46-15 with the same results.
Operators concluded that control rod 46-15 was difficult to move and then used Section 8.1, "Control Rod Difficult to Withdraw, Control Rod at Position 00," of plant procedure OP-07, "Reactor Manual Control System Operating Procedure," to attempt movement of the control rod.
To verify the control rod drive system was functioning properly, operators decided to stop testing control rod 46-15 and to perform friction testing of control rod 26-35. Testing of control rod 26-35 began at 0536 hours0.0062 days <br />0.149 hours <br />8.862434e-4 weeks <br />2.03948e-4 months <br />. At 0548 hours0.00634 days <br />0.152 hours <br />9.060847e-4 weeks <br />2.08514e-4 months <br />, operators successfully withdrew control rod 26-35 to position 48 and reinserted the control rod to position 00.
Having confirmed proper functioning of the control rod drive system, Operators then returned to control rod 46-15 and again made numerous attempts to move the control rod from position 00. These attempts were not successful. At 0609 hours0.00705 days <br />0.169 hours <br />0.00101 weeks <br />2.317245e-4 months <br />, operators concluded an undetermined problem existed with control rod 46-15 and terminated control rod testing. Investigation revealed the PIP position simulator was still installed for control rod 46-15. At that time, operators reassessed compliance with Technical Specifications. Operators determined that compliance with the Required Actions of Technical Specifications 3.9.2 and 3.9.4 had not been satisfied when control rods 14-27, 26-15, and 26-35 had been withdrawn for the single control rod subcriticality testing, but that the facility was currently in compliance with Technical Specifications 3.9.2 and 3.9.4. The PIP position simulator was subsequently removed. When the PIP position simulator was removed and the circuit restored, the control room position indication displayed position 00 for control rod 46-15. Testing of control rod 46-15 was then completed satisfactorily.
EVENT CAUSE
The cause of this occurrence was the misapplication of Technical Specification 3.0.5. Technical Specification 3.0.5 allows for the return of equipment to service in order to verify operability.
Although performance of the single control rod subcriticality test for control rods 14-27 and 26-15 was a procedural prerequisite for testing control rod 46-15, the single control rod subcriticality test was not required in order to demonstrate the operability of control rod 46-15 and re-arming of control rod 46-15 was not necessary to perform the single control rod subcriticality testing. Therefore, the allowance of Technical Specification 3.0.5 should have been used only to return control rod 46-15 to service for operability testing and should not have been used to return control rod 46-15 to service for performance of single control rod subcriticality testing of control rods 14-27, 26-15, and subsequent friction testing of control rod 26-35.
The assessment and determination of the acceptability to use Technical Specification 3.0.5 was conducted independently by two senior reactor operators. However, interviews of other licensed operators determined that a larger population of operators could have misapplied the allowance of Technical Specification 3.0.5 in this situation.
A contributing cause of this event was the failure to remove the PIP position simulator. Because the PIP position simulator was installed instead of a jumper, operators received normal position indications (i.e., a green "full-in" light and the display of position 00 on the four rod display) from control rod 46-15. Installation of a jumper instead of the PIP position simulator would have provided a green "full-in" light to operators, but the four rod position indication would have been blank. With the green "full-in" light and the display of position 00 on the four rod display, operators initially concluded that, with normal indications, the problem with control rod 46-15 was a difficult-to-move control rod.
CORRECTIVE ACTIONS
1. To prevent recurrence of this event, training on the proper use of Technical Specification 3.0.5 will be provided to licensed operators. This is scheduled to be completed by July 18, 2005.
2. Lessons learned for this event will be reviewed with licensed operators and I&C personnel.
3. Refresher training on the proper use of Technical Specification 3.0.5 and outage-related LCOs will be provided to licensed operators.
4. Appropriate I&C personnel involved with this event have been counseled.
5. The controls governing the installation and removal of PIP jumpers will be changed from the plant work order process to being controlled through a plant procedure. Applicable plant procedures will be revised to prohibit use of the PIP position simulator as a jumper for the one- rod-out interlock.
SAFETY ASSESSMENT
The significance of this occurrence is considered minimal. Although manipulation of control rods with the PIP position simulator installed potentially allowed withdrawal of more than one control rod, the control rod manipulations were conducted using plant procedures that only allowed individual control rod manipulation. Therefore, only two control rods could have been withdrawn at the same time. An analysis was performed, assuming that control rod 46-15 and a face adjacent control rod (i.e., control rod 46-19 was used) were fully withdrawn simultaneously. The analysis demonstrates that the reactor would have remained subcritical at 0.611% shutdown margin. This analysis bounds the potential combined withdrawal of control rods 46-15 and 26-35.
PREVIOUS SIMILAR EVENTS
A review of events occurring within the past three years identified an occurrence which was reported in Specification. This LER discussed an event where operators placed equipment control tags on the wrong hydraulic control unit during replacement of a control rod drive, resulting in the failure to comply with TS 3.10.4 Required Actions. The root cause and corrective actions associated with herein.
COMMITMENTS
No regulatory commitments are contained in this report. Those actions discussed in this submittal will be implemented in accordance with corrective action program requirements.