05000263/LER-2007-003

LER-2007-003,

Docket Number
Event date:	04-20-2007
Report date:	08-30-2007
2632007003R01 - NRC Website

Description On April 20, 2007, the plant was in MODE 4 (cold shutdown) for refueling outage 23 with the Mode switch in refuel for Control Rod Drive (CRD) [AA] exercise testing. Control rod [ROD] 26­ 35 was withdrawn per procedure. Per the procedure acceptance criterion the "00" position went out when the rod was withdrawn. Following CRD exercising, the control rod was re­ inserted. During these tests, it was recognized that a work order identified a problem with control rod 26-35 position indication. Having an equipment issue on the control rod position indication made the operators believe they should expect some abnormal behavior. The next control rod, 26-31, was selected and similarly exercised per procedure. When this control rod was withdrawn, the numerical indication changed from green to amber. One of the operators recalled that during withdrawal of control rod 26-35 the numerical indication remained green.

The crew completed testing of Control Rod 26-31 and fully inserted the control rod. The crew stopped, reviewed the acceptance criteria of the procedure, and notified the control room supervisor that the indication of the previous control rod was not correct.

Control rod testing was suspended while the crew evaluated the indications noted on control rod 26-35. Upon review of the bases for the Technical Specification (TS) it was discovered that the full in indication for TS compliance is the double dash or green light reed switch. The "00" position indication does not provide input for the one-rod-out interlock as specified in the procedure. TS Action 3.9.2.A was entered and the drive for control rod 26-35 was de­ activated.

Technical Specifications require that the full rod in position interlock be operable for each control rod for the conditions during the event. Upon discovery that control rod full in position for 26-35 was not operable, TS Action 3.9.2.A should have been entered. However, testing was performed on an additional control rod. The result was non-compliance with TS 3.9.2, "Refuel Position One-Rod-Out Interlock" since the interlock was inoperable and control rod withdrawal was not immediately suspended.

Event Analysis

Per 10 CFR 50.73 (a)(2)(i)(B), an Operation or Condition Prohibited by Technical Specifications requires a Licensee Event Report. There is no requirement for reporting in accordance with 10 CFR 50.72 for this event.

The event involves a safety system functional failure.

Safety Significance

The one-rod-out interlock is designed to ensure that movement of more than one control rod is restricted to prevent the reactor from becoming critical during refueling operations. During refueling operations, no more than one control rod is permitted to be withdrawn with fuel in the cell. The inoperability of the interlock for control rod 26-35 would have allowed control rod 26­ 35 and any other control rod to be withdrawn at the same time.

However two additional barriers exist to prevent two control rods from being withdrawn at the same time. All control rod movement is directed by procedures. Under the conditions for the event, no procedures direct multiple control rod withdrawal in Mode 4 and licensed operators are trained not to withdraw two control rods simultaneously in Mode 4.

If these barriers had failed and two control rods were withdrawn, it is possible for the reactor to become critical if the second control rod were close to control rod 26-35. These barriers did not fail and no more than one control rod was actually withdrawn at any time during this exercise evolution. At no time did the reactor become critical. This issue does impact nuclear safety due to the violation of Technical Specifications, but for the above reasons there were no industrial or radiological safety issues associated with this event.

The Probabilistic Risk Assessment (PRA) group performed an evaluation of the event and, based on discussions with Nuclear Engineering, concluded that the risk of core damage as a result of this event was minimal for the following reasons:

• The Reactor Manual Control System would have permitted the withdrawal of control rod 26-35 and any other rod. Withdrawal of three control rods could not have occurred.

• Nuclear Engineering performed a calculation with both control rods 26-35 and 26-31 fully withdrawn. The calculation indicated that the reactor would have become critical with both of these control rods fully withdrawn. No fuel damage would have occurred because the reactivity conditions of the potential criticality would have been bounded by the Control Rod Drop accident.

Cause

The cause of the event was an incorrect procedure acceptance criterion for satisfying the requirements of Tech Spec 3.9.4. A contributing cause to the event was that the Operators were not proficient in utilization of special operations procedures that are typically initiated once per cycle. A second contributing cause was that the operators did not fully understand the reason for the performance of the step and the impact of the equipment deficiency associated with control rod 26-35 position indication.

Corrective Action The following corrective actions are planned or have been completed:

• The position indication probe for control rod 26-35 was replaced and the control rod full in indication was returned to operable. (Completed)

• Individuals involved have been coached and counseled on their role in this event.

(Completed)

• The procedure was revised to correctly implement the technical specification.

(Completed)

• The station requested additional training for operators prior to outages that will require them to exercise these special operations procedures that do not get routinely utilized. (In progress) Failed Component Identification Probe, Rod Position Indication — General Electric Company — Part Number: 797E111G001

Previous Similar Events

A review of station events found no events that were considered to be related to this event.