05000333/LER-2004-001
| Docket Number | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3332004001R00 - NRC Website | |
Event Description:
On October 7, 2004, the James A. FitzPatrick Nuclear Power Plant (JAF) was in a refueling outage with the reactor cavity flooded and the spent fuel pool gates removed. Reactor Power was at 0% and the Mode Switch was in the Refuel Position. At approximately 2055, during the performance of a scheduled calibration of reactor vessel pressure indicator, 02-3PI-60A, a pressure perturbation in the instrument sensing line resulted while returning the pressure indicator to service. The pressure perturbation caused the reactor vessel level transmitters that share this sensing line to sense a low reactor vessel water level. Since the reactor cavity was flooded up for refueling, the signal generated by the reactor vessel level transmitters in response to the pressure perturbation was invalid. The invalid reactor vessel low water level signal resulted in actuation of the Emergency Diesel Generator (EDG) [EK] and Emergency Core Cooling Systems (ECCS). The initiation logic functioned as designed. EDGs 'A', 'B', 'C' and 'D' [EK] started but did not close in and power the emergency buses since there was no loss of offsite power. Residual Heat Removal (RHR) [BO] 'A', 'B', 'C', and 'D' pumps and Core Spray (CS) [BM] 'A' and 'B' pumps started and injected into the core. In addition, signals were received for High Pressure Coolant Injection (HPCI) [BJ], Reactor Core Isolation Cooling (RCIC) [BN], Standby Gas Treatment (SBGT) [BH] initiation, Reactor Water Recirculation (RWR) pump trip and Alternate Rod Insertion (ARI); however, these systems were removed from service for outage activities so there was no response. Upon start of the EDGs and the ECCS pumps, Operations personnel immediately verified plant conditions, confirmed that there was no need for injection, secured ECCS injection and secured the EDGs [EK]. The ECCS injection was completely secured within 60 seconds after the start of the injection.
As the ECCS injection began, personnel on the Refuel Floor observed an influx of turbid water that began to rise in the reactor cavity. They placed the refuel floor equipment in a safe condition and exited the area. As a result of the ECCS injection, water level in the reactor cavity and reactor internals storage pit increased and overflowed into the RB floor drains, from the north and west curbs of the reactor internals storage pit. The narrow design of the passage between the reactor cavity and the spent fuel pool restricted the flow of water into the spent fuel pool (SFP) and no overflow from the SFP was noted. The water that flowed over the north and west curbs of the reactor internals storage pit onto the refuel floor was directed to the RB Floor drain system on the north side of the RB 369 foot elevation. The RB floor drains on the north side of the 369 foot elevation are directed to downcomers (vertical runs of piping) on the north wall of the RB. Similar arrangements of floor drains connected to downcomers exist on each elevation of the RB until the 300 foot elevation where the downcomers enter a cross-over (long horizontal run of piping) that directs flow to the downcomers on the south side of the RB, that in turn, direct drainage to the RB floor drain sumps, located below the 227 foot elevation floor in the east and west crescent areas. The floor drain system, unable to handle the in-flow rate, backed-up above this cross-over, and overflowed onto the area around the various RB north side floor drains. The east and west crescent area floors were contaminated by the overflow from the floor drain sumps. The overflow also entered the north stairway and flowed down the stair way, eventually joining the flow to the floor drains at the lower RB elevations. This flow down the stair way was the source of contamination on the 272 foot elevation. The overflow resulted in contamination of limited portions of the reactor building. No personnel contaminations or unexpected exposures occurred as a result of this incident.
When the injection was secured, the Radiation Protection (RP) department took conservative and proactive actions to restrict access to the Reactor Building and promptly deployed RP technicians to survey the reactor building to determine the extent of contamination. The RP surveys determined that limited portions of each RB elevation were contaminated.
Event Description (continued):
The event was entered in the plant's corrective action program for further evaluation, root cause determination, and documentation of corrective action. The event was also discussed with the Senior NRC Resident Inspector and NRC Region 1 Staff.
This event was the subject of a telephone report made by JAF to the NRC Emergency Operations Center on Nov. 22, 2004 which met the reporting requirements of 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A). In order to provide a more thorough summary of the event, JAF is submitting this voluntary LER.
Cause of Event:
Technicians had removed the remote reactor vessel pressure indicator from service for performance of a scheduled calibration activity. During the restoration portion of the calibration activity, a pressure perturbation was induced in the sensing line shared by the reactor vessel pressure indicator and reactor vessel level transmitters. The pressure perturbation resulted in an invalid reactor vessel low level signal that started the EDGs and initiated ECCS injection.
The pressure perturbation was caused by inadequacies in the calibration procedure and the technique the technician used to open the instrument isolation valve for instrument restoration. The procedural guidance did not elaborate that the purpose of pressurizing the instrument to match indicated reactor pressure, prior to opening the instrument isolation valve, was to establish a 0 psig differential pressure across the instrument isolation valve, thereby minimizing the potential for pressure perturbations. In addition, the procedure did not provide any guidance on the need for additional offset pressure, to achieve the desired 0 psig differential, when vessel level was above the instrument sensing line (e.g. during flood-up). When the technician performed the procedure, an unintended differential pressure was established across the instrument isolation valve because the reactor cavity was flooded up for refueling, placing an additional pressure on the sensing line that was not addressed in the procedure. When the technician cracked the isolation valve open too rapidly, the equalization of the differential pressure across the valve resulted in a pressure perturbation in the common sensing line.
The initial investigation also identified a limitation in operating procedures for the ECCS systems. The procedures did not allow isolation of the system initiation logic even though the systems were not required to be operable per the plant Technical Specifications.
The event investigation identified that the inadequate guidance in the calibration procedure was a result of inadequate guidance in the procedures for writing and controlling procedures. The guidance on placing cautions in operating and technical procedures and on ensuring that plant procedures do not rely on a single barrier to prevent an unintended plant transient was deficient. This deficiency resulted in the calibration procedure containing inadequate guidance for performance during all conditions that could be encountered during a refueling outage and a lack of appropriate caution statements on critical steps.
Inadequate use of operating experience (OE) also contributed to this event. The management expectation to use and discuss relevant OE in pre-job briefings for each maintenance activity is well known. However, the discussion of OE during the pre-job brief for the calibration of the reactor vessel pressure indicator was based on the past experience of the job supervisor and technician involved in the task and focused on the consequences of bumping other instruments in the same rack The preparation for the briefing did not include a formal search of OE. A formal search of OE would probably have identified two previous scrams at JAF, one in 1990 (LER-90 026-01) and another in 1991 (LER-91-007-01), caused by poor valving techniques during the restoration of a feedwater level transmitter after calibration, and so, the opportunity to identify relevant OE was missed. For this reason, failure to properly review relevant operating experience is considered to be a contributing cause in this event.
Event Analysis:
The response of plant equipment was in accordance with the design. The EDGs [En RHR [BO], and CSP [BM] started in the appropriate sequence. Additionally, initiation signals were received for High Pressure Coolant Injection (HPCI) [BJ], Reactor Core Isolation Cooling (RCIC) [BN], Standby Gas Treatment (SBGT) [BH] initiation, Reactor Water Recirculation (RWR) pump trip and Alternate Rod Insertion (ARI). However, these systems were removed from service in accordance with the outage schedule, and so, did not actuate.
The ECCS injection flow caused the disturbance of corrosion products in the reactor vessel and injection piping.
These corrosion products became suspended in the water and resulted in a loss of clarity.
The control room operators verified that plant conditions did not require an ECCS injection and secured the injection within 60 seconds of initiation. Their prompt recognition and response to this event limited the amount of water that overflowed from the reactor cavity and limited the associated spread of contamination.
Personnel on the refuel floor recognized the onset of off normal conditions, placed equipment in a safe condition and exited the area in a timely manner. The timely recognition of this change in environment by those personnel in the RB at the time of the event prevented any personal contamination events as a result of the ECCS injection.
RP proactively and conservatively restricted access to the reactor building and promptly deployed RP technicians to determine the extent of contamination. RP identified the contaminated areas in the RB and initiated the appropriate access controls by setting up boundaries and appropriately posting the contaminated areas.
The maintenance department performed an initial investigation of the event and quickly identified the technician's valving technique as a probable cause of the event.
The outage team implemented a plan to expeditiously restore water clarity in the reactor cavity and effectively decontaminate the areas on the RB affected by the overflow.
The event was entered into the plant corrective action system and a root cause investigation team performed a thorough review of the event. The root cause team identified deficiencies with the actions of the personnel performing the calibration activity and the procedures used to control the activity. The team also identified corrective actions to resolve the identified deficiencies.
Consequences Of Event:
The event resulted in the invalid actuation of all low pressure Emergency Core Cooling System (ECCS) pumps (four Residual Heat Removal (RHR) [BO] pumps and two Core Spray (CS) [BM] pumps). All six low pressure ECCS pumps were aligned to draw from the torus. While the water quality in the torus has acceptable water chemistry, it is not as pure as normal reactor cavity make-up water. Additionally, the rapid introduction of torus water to the core causes corrosion products that have settled out in low flow regions of the vessel and injection piping to become suspended in the reactor water. Since the event occurred during refueling operations with the reactor vessel head removed, and the vessel and the cavity flooded up for refueling, these impurities were dispersed throughout the water in the reactor, reactor cavity, spent fuel pool and reactor internals storage pit.
This resulted in poor water clarity delaying refueling activities.
Because the reactor cavity was flooded for refueling, the injection also resulted in overflow from the north and west curbs of reactor internals storage pit to the RB Floor Drain System. The overflow to the RB floor drains initially exceeded the capacity of the system resulting in contaminated water being spread onto limited portions of the RB floors on the north side of the building and to the floors of the east and west crescent areas.
Event Analysis (continued):
Consequences Of Event (continued):
During the second fuel shuffle, there were anomalous discolorations noted on two fuel channels. Reactor Engineering concluded that this was a likely consequence of the ECCS injection. To confirm this, samples of the material causing the discoloration were wiped off and elemental and isotopic analyses were performed. The elements and isotopes identified were typical of what would normally be observed in primary system crud.
Global Nuclear Fuels (GNF) concurred with ENN's conclusion that this crud would not affect fuel channel performance and that these channels did not require replacement and manual removal of the crud was not warranted.
Due to the circulation of corrosion products that became suspended in the reactor cavity water during the ECCS injection, dose rates on the Fuel Pool Cooling, Residual Heat Removal, and Alternate Decay Heat Removal systems have increased somewhat, as has the resulting dose to plant personnel. This is being addressed by a systematic clean-up and radioactive source removal effort. Reactor reassembly required additional time as the injection created a need for decontamination efforts, in the reactor cavity, beyond those originally planned.
Safety Significance:
While this event had limited consequences related to the radiological conditions in the plant, there was no impact on the health and safety of the public. Contamination, as discussed in this report, was confined to limited portions of the reactor building. There was no measurable release of radiation to the public or outside the reactor building due to this event.
Risk Significance:
The water inventory in the reactor cavity was more than adequate to provide the required core cooling. The Alternate Decay Heat Removal (DHR) system, credited for meeting the Technical Specification requirements for decay heat removal, remained operable throughout the event. The EDGs started and forced paralleled as designed, and if a Loss of Offsite Power had occurred they would have energized the emergency buses. The inadvertent start of the EDGs and of all low pressure ECCS pumps did not appreciably alter the overall plant risk during this event.
Extent of Condition:
Since the technique used by the technician in restoring the instrument was a cause of this event, a stand down briefing was held with all l&C technicians about the critical importance of using good valving techniques when restoring instruments to service. Personnel were also briefed on the procedural deficiencies that were identified as contributing causes so that such deficiencies could be identified on any procedures being worked, pending completion of the root cause corrective actions. No similar procedural deficiencies were identified during the remainder of the refueling outage. Further review of procedures is discussed in the corrective action section of this report.
Extent of Condition (continued):
Based on the corrective actions to (1) allow isolation of the initiation logic for ECCS subsystems when they are not required to be operable by technical specifications; (2) provide refresher training on valving techniques; (3) revise maintenance and operating procedures to include appropriate cautions when performing critical steps that could result in plant impact; (4) revise maintenance and operating procedures to ensure that either multiple barriers are in place to prevent individual actions from impacting the plant or require specific management review and approval; and (5) review maintenance and operating procedures to ensure that plant impacts have been identified and properly addressed, an event of this nature is unlikely to recur.
Corrective Actions:
The corrective actions discussed below are programmatic enhancements to ongoing practices which are captured in the plant corrective action program and do not represent regulatory commitments.
Immediate Corrective Actions:
The EDG start and ECCS injection were diagnosed as being the result of an invalid signal. The ECCS injection was secured and the EDGs were shutdown. Equipment on the refuel floor was proactively and conservatively placed in a safe condition and personnel exited the area. Access to the reactor building was restricted and survey teams were deployed to determine the extent of contamination within the reactor building. As contaminated areas were identified, access was restricted using barriers and postings in accordance with the plant procedures. A recovery plan to address water clarity in the reactor cavity and decontamination was developed and implemented. Plant operating procedures were revised to allow lock out of the ECCS subsystem initiation logic when operability is not required by plant Technical Specifications. All maintenance shops conducted a work stand-down and briefing on the event, and emphasized the need to use all of the human performance tools available to prevent similar occurrences and reinforced management expectations to make effective use of OE in pre-job briefs.
Corrective Actions:
The Procedure Writing Manual and Control of Procedures procedure will be revised to prevent situations where a single individual is the only barrier to an action with immediate and irreversible impact on plant operation, unless senior management approval for such a situation is obtained.
The instrument surveillance procedure associated with the remote reactor vessel pressure indicator will be revised to incorporate the lessons learned from this event regarding the need for cautions and additional guidance. Additionally the Maintenance l&C, Maintenance Electrical, Maintenance Mechanical, Operations, Radiation Protection, and System Engineering (Reactor Engineer &Test Engineer) organizations will each review a sample of their department procedures to ensure that they have been appropriately evaluated for plant impact.
Refresher training on valving techniques will be conducted for the appropriate group of technicians.
The Procedure Writing Manual will be revised to provide guidance to assess potential plant impacts when developing new procedures. In addition, requirements to (1) document the reason a procedure is considered to have a potential for plant impact within the procedure and (2) to require that cautions are placed in appropriate sections to alert the procedure user of impacts on plant operations due to direct or indirect actions, will be added to the Procedure Writing Manual and to the administrative procedure for the "Control of Procedures".
Clean up from this event includes decontamination of the areas affected by the overflow as well as flushing and hydrolazing systems affected by the transportation of corrosion products during the injection.
Safety System Functional Failure Review:
Although the actuation signal was invalid, all safety systems responded as designed. There were no safety system functional failures associated with this event.