05000250/LER-2004-005
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber Na 05000 | |
| Event date: | 12-3-2004 |
|---|---|
| Report date: | 2-1-2005 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2502004005R00 - NRC Website | |
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 3 05000250
DESCRIPTION OF THE EVENT
The Turkey Point Unit 3 Cycle 21 refueling outage began on September 26, 2004. On October 16, 2004, abnormal indications were noted on the PT-3-445 pressurizer [EIIS: AB, PZR] pressure indications in the control room. Troubleshooting, including cable [EIIS: CBL] characterization utilizing the CHAR 200 system, revealed that the cable for pressurizer pressure transmitter [EIIS: AB, PT] PT-3-445 control loop installed in conduit [EIIS: CND] 3C226-2 was damaged inside the reactor containment building [EIIS: NH].
The cable was subsequently removed from the conduit and significant damage to the cable jacket and insulation [MIS: ISL] was observed over approximately a four-foot length of cable. The cable damage occurred where the cable passed directly over Reactor Coolant System (RCS) [EIIS: AB] 3B hot leg piping.
Inspection and testing was performed for all conduits/cables in close proximity to identified hot pipes inside containment to identify all degraded cables. Of the cables found degraded, the degradation of cables associated with level transmitters [EIIS: LT] LT-3-460 and LT-3-461, pressurizer level channels 2 and 3, respectively, would not have assured their ability to function after an accident. For post accident conditions conduits/cables could be subjected to extreme environmental conditions, including saturated steam and chemical spray. Inspection of the degraded cables associated with these instruments indicates that the cables would not have been able to survive the 31-day accident mission time. Therefore, the respective channels for instruments LT-3-460 and LT-3-461 were considered inoperable for their post accident monitoring function. Due to the nature of the cable degradation (heat induced aging), the time frame in the past that the respective channels for instruments LT-3-460 and LT-3-461 could have become inoperable for accident monitoring is indeterminate.
After a thorough evaluation, this condition was determined to be reportable on December 3, 2004 in accordance with 10 CFR 50.73(a)(2)(i)(B).
Unit 3 was defueled at the time the abnormal pressurizer pressure indications were observed. Unit 4 was operating in Mode 1 at 100% power.
BACKGROUND
Three pressurizer level channels in a two-out-of-three logic for high pressurizer level are used for reactor [EIIS: RCT] trip [EIIS: JC]. This function is not relied upon as a primary trip function in the plant safety analysis. It may perform as a backup trip for any significant heatup transient, which results in a large specific volume change for RCS primary coolant. The degradation of cables associated with pressurizer level transmitters LT-3-460 and LT-3-461 does not affect their reactor trip function.
During normal operation, isolated output signals from these channels are used for volume control, increasing or decreasing water level. A level control failure could fill or empty the pressurizer at a slow rate (on the order of one half hour or more). Therefore, ample time and alarms exist for operator action in the event of increasing or decreasing water level in the pressurizer.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 3 05000250 High Level A reactor trip on pressurizer high level is provided to prevent rapid thermal expansions of reactor coolant fluid from filling the pressurizer: the rapid change from high rates of steam relief to water relief can be damaging to the safety valves and the relief piping and pressure relief tank [FM: TK]. However, a level control failure cannot actuate the safety valves because the high pressure reactor trip is set below the safety valve set pressure. With the slow rate of charging available, overshoot in pressure before the trip is much less than the difference between reactor trip and safety valve set pressures. Therefore, a control failure does not require protection system action. In addition, ample time and alarms are available for operator action.
Low Level Ample time and alarms exist for operator action in the event of a decreasing water level in the pressurizer.
Regulatory Guide (RG) 1.97, Revision 3, divides all instrumentation used for post accident monitoring into five functional types. Type A variables provide the primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety function for design basis accident events. Primary information is information that is essential for the direct accomplishment of the specified safety functions; it does not include those variables that are associated with contingency actions that may also be identified in written procedures.
Table 1 of RG 1.97, Revision 3 provides design and qualification criteria for post accident monitoring instrumentation. The criteria are divided into three categories depending on the importance to safety of the specific variable. In general, Category 1 provides for full qualification, redundancy, and continuous real time display and requires onsite (standby) power.
Category 1 instrumentation located in harsh environments should comply with the requirements of 10 CFR 50.49. Instruments provided and designed to function following a major loss-of-coolant accident initiate or otherwise govern the operation of engineered safety features. Pressurizer pressure and level, steam generator [EIIS: SG] level and main steam [EIIS: SB] flow are typical examples of sensors that are located inside containment because an equivalent signal cannot be obtained from a sensor location outside containment.
Pressurizer water level is a RG 1.97 Type A, Category 1 parameter. Due to cable degradation, environmental conditions in containment post-accident may affect the monitoring function of channels 2 and 3.
CAUSE OF THE EVENT
The root cause of cable degradation is conduits/cables routed in close proximity to RCS hot leg pipes with limited heat dissipation capability. This resulted in localized hot spots that subjected the cables to DOCKETFACILITY NAME (1) LER NUMBER (6) PAGE (3)NUMBER (2) prolonged temperatures above the cable insulation design rating. Contributing factors include gaps in RCS piping insulation; uninsulated RCS pipe stubs; enclosed area (i.e., low ceiling) and a Normal Containment Cooling (NCC) [EIIS: VA] ventilation register that was closed due to a broken linkage.
ANALYSIS OF THE EVENT
Unit 3 was shutdown for refueling when abnormal pressurizer pressure indications were observed. Trouble shooting revealed that the instrument cable associated with pressurizer pressure transmitter PT-3-445 had shorted conductors, which resulted in a spurious pressurizer pressure signal. The damaged section of the cable was in conduit 3C226-2, which was routed less than 1 inch above the 3B RCS hot leg pipe insulation.
All cables in conduit 3C226-2 were removed and found damaged where the conduit was routed over the hot leg pipe. Testing and removal of cables in four conduits immediately adjacent to 3C226-2 identified degraded cables in the same location above the 3B RCS hot leg.
Based on the damaged cables identified above the 3B hot leg, the following actions were taken:
- Walkdowns and drawing reviews were conducted on Unit 3 to identify other conduits routed in close proximity to the RCS hot, cold and intermediate legs and other hot piping in containment.
- Testing (CHAR cable characterization and/or megger test) conducted on 13 cables, identified as being in conduits above the RCS 3B hot leg, showed that most cables were degraded.
- All cables in conduits in close proximity to the RCS hot legs, RCS cold legs, pressurizer surge line and some specifically selected hot pipes in containment were removed and visually inspected.
- Visual inspection of the cables showed that the sections of cable in close proximity (above) the RCS hot legs were degraded; however, the sections of cables in close proximity to (primarily above) the RCS cold and intermediate legs and other hot pipes were in satisfactory condition.
- Visual inspection of RCS piping, pressurizer surge line, blowdown lines, letdown lines, Feedwater [EIIS: SJ] piping and Main Steam [EIS: SB] lines inside containment were conducted to identify any insulation deficiencies. Gaps in RCS piping insulation and uninsulated RCS pipe stubs were identified.
Some minor insulation deficiencies were identified on other hot pipes.
- NCC vents were inspected to ensure they were open.
Based on these investigations, the following was determined:
- The conduit (3C226-2) associated with pressurizer pressure transmitter PT-3-445 was in contact with pipe insulation via an insulation buckle.
DOCKETFACILITY NAME (1) LER NUMBER (6) PAGE (3)NUMBER (2)
- Visual inspection determined that most in-service cables with sections in close proximity above RCS hot legs were degraded.
- Visual inspection determined that no in-service cables with sections in close proximity to the RCS cold legs, intermediate legs, or other hot pipes were degraded. All these cable sections were in satisfactory condition.
- An EPRI plant support engineer performed indenter testing of some cables. This testing identified that degradation was heat related and not radiation related. In addition, this testing confirmed that sections of cable in close proximity to RCS cold legs, intermediate legs, and other hot pipes were in good condition.
- Wrap-around mandrel testing of the removed section of cable 3SCM1AJTE3-433A-T3115/1 (Cable Code LT1) from conduit 3D1005, located above the 3C RCS hot leg, determined the cable was capable of performing its design basis function.
- Inspection of piping insulation identified several gaps in RCS piping insulation and uninsulated RCS pipe stubs in the general area where conduits passed over the RCS legs. These capped pipe stubs resulted from RTD bypass piping removal implemented in 1991.
- NCC ventilation register inspection identified one register closed due to broken linkage near the 3B RCS hot leg.
- The following cable types were identified as degraded. The 60-year temperature design rating for the cables is provided. The exact temperature and duration (i.e., acute or cumulative affect) that the cables were exposed to is not known.
Cable Code LT1 and LP1: Anaconda, 60 year temperature rating - 67°C (152.5 °F) Cable Code 060 and 061: General Electric, 60 year temperature rating - 70°C (158°F) As previously stated, inspection and testing was performed for all conduits/cables in close proximity to identified hot pipes inside containment to identify all degraded cables. The following is a brief discussion of each hot piping area evaluated:
RCS Hot Legs (3A, 3B and 3C) Visual inspection showed most of the in-service cables with sections above the RCS hot legs were degraded.
Piping insulation gaps and uninsulated capped pipe stubs were found on the RCS hot legs in close proximity to some conduits. Furthermore, the conduits are routed above the hot legs and below the ceiling, about an 8 inch air space that can trap heat. For 3B hot leg (where most cables were damaged) the NCC ventilation FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 3 05000250 register was found closed further limiting airflow. All cables routed in close proximity to RCS hot legs have been replaced with new cable.
RCS Cold Legs (3A, 3B and 3C) Visual inspection showed no in-service cables with sections in close proximity above the RCS cold legs were degraded. All these cable sections were in satisfactory condition. Piping insulation gaps and uninsulated capped pipe stubs were found on the RCS cold legs in close proximity to some conduits. The conduits are also routed above the cold legs and below the ceiling, about an 8-inch airspace. However, there is an opening in the ceiling at the reactor coolant pump (RCP) [EIIS: P] end of the cold leg that provides elevated airflow and does not trap heat. All cables routed in close proximity to RCS cold legs have been replaced with new cable. The cables were replaced since they were removed from the conduits for inspection.
RCP 3A temperature element cables in conduit 3D026 were replaced. Inspection of the RCP 3A temperature element cables showed jacket and insulation damage at the RCP termination location only. The damage is attributed to mechanical wear from working the cables during outages and is not associated with high temperature heat degradation. Inspection of the removed RCP 3A temperature element cables showed no heat-related degradation at the cold leg location.
RCS Intermediate Legs (3A, 3B and 3C) The identified conduits in close proximity to the RCS intermediate legs are below the insulation by 9 to 18 inches. There are no obstructions to airflow or heat dissipation for the RCS intermediate legs. The cables in conduits 3C1345 and 3C1368 that were approximately 9 inches below the intermediate leg were visually inspected at a pull box, which was located approximately 3 feet from the upper side of the leg, and were found in satisfactory condition. Therefore, cables in conduits below the RCS intermediate legs are considered to be in satisfactory condition.
Unit 3 Pressurizer Surge Lines Visual inspection showed no in-service cables with sections in close proximity to/above the pressurizer surge lines were degraded. All these cable sections were in satisfactory condition. There are no obstructions to airflow or heat dissipation for the pressurizer surge lines. All cables routed in close proximity to the pressurizer surge lines were replaced with new cable. The cables were replaced since they were removed from the conduits for inspection.
Unit 3 Blowdown Lines The two cables in conduit 3D011, which is 3 inches from an uninsulated pipe and is considered the worst case in terms of conduits in close proximity to uninsulated piping, were removed for inspection. Visual inspection showed the cables to be in satisfactory condition. There are no obstructions to airflow or heat dissipation for the SG blowdown lines. Therefore, cables in other conduits in close proximity to SG DOCKETFACILITY NAME (1) LER NUMBER (6) PAGE (3)NUMBER (2) blowdown lines are considered to be in satisfactory condition. The cables removed for inspection were replaced with new cables.
Unit 3 Main Steam Lines and Feedwater Lines Cables in conduits in close proximity to the Main Steam lines and Main Feedwater lines were not removed for inspection based on the following:
- Main Steam and Main Feedwater lines have lower process flow temperatures and lower insulation surface temperatures.
- Main Steam and Main Feedwater lines are in open areas were heat is dissipated.
- Main Steam and Main Feedwater lines have calcium-silicate insulation that is generally considered more effective than reflective insulation.
Unit 3 Pressurizer Block Valves The conduits/cables for these valves are run on the outside surface of the pressurizer wall. Wiring for these motor-operated valves, located above the pressurizer, is inspected every refueling outage. No further action was required.
Normal Containment Cooling Register The NCC supply register in the area of the 3B RCP and 3B hot leg has upper and lower louvers [EIIS: LV].
By inspection, the upper louvers were found closed. Attempts to re-open the louvers were unsuccessful as the louvers were stuck and the damper linkage was no longer connected. These dampers are not typically adjusted and there is no specific surveillance in place to verify their position. A documented design position is not available. The duration the upper louvers were closed is indeterminate; however, based on the condition of the louvers and damper linkage, they are presumed closed for a significant period. Considering design flow rates, the low flow velocity at the affected conduit area is not expected to provide significant cooling. However, open louvers help to ensure adequate circulation in the area to prevent a stagnant air build-up directly below the concrete ceiling.
Temperature Monitoring Temperature monitoring equipment has been installed on conduits that are in close proximity above the RCS hot leg piping. The temperature data loggers will record the temperatures during the next operating cycle (Cycle 21) to establish the qualified life for the new replacement cables. In addition, temperature readings were taken at normal operating pressure and temperature (NOP/NOT) as Unit 3 returned to service.
Temperature readings taken at NOP/NOT confirmed local temperatures are within cable ratings for Cycle 21.
DOCKETFACILITY NAME (1) LER NUMBER (6) PAGE (3)NUMBER (2) 2004 - 005 � 00 All degraded cables and cables in close proximity to RCS hot legs were replaced with new cable that has a 28-year qualified life at 194.7°F. The long-term qualified life for the replacement cables beyond the current operating cycle will be based on temperature data logger information obtained throughout the cycle. Data from these loggers was collected on December 16, 2004 during a short notice outage. The data covers the period from 11/27/2004 to 12/16/2004. During this period, Unit 3 was in Mode 1 approximately 12 days with RCS hot leg temperatures of at least 560°F including a 3-day period at 100% power (RCS hot leg 600°F). The data supports the highest recorded conduit temperature of 124°F previously taken during NOP/NOT with a highest recorded temperature of 135°F. Based on the temperatures monitored on the RCS hot legs, the highest expected temperature at the conduits is approximately 152°F. The calculated 28-year qualified life of the new cable installed at the RCS hot leg locations remains valid. The data for the full Unit 3 Cycle 21 will be reviewed to make the final determination of the qualified life of the new cable.
Operability Of thirty cables found degraded eight cables and related instruments associated with six functions required evaluation for operability. Three affected instruments have accident monitoring as well as reactor trip functions. Reactor trip and engineered safety function actuation would not have been affected by any of the degraded cables. For post-accident conditions the conduits/cables could be subjected to extreme environmental conditions, including saturated steam and chemical spray. Inspection of the degraded cables associated with the instruments performing a post-accident function indicates that three cables might not have been able to survive the 31-day accident mission time. Therefore, the respective channel for instruments LT-3-476, LT-3-460 and LT-3-461 are considered inoperable with respect to their post-accident monitoring function. Due to the nature of cable degradation (heat induced aging), the time when the respective channel for instruments LT-3-476, LT-3-460 and LT-3-461 could have become inoperable for accident monitoring is indeterminate.
Level transmitter LT-3-476 supports steam generator level loop A, channel 3. Post-accident monitoring generator to be operable. Channels 1 and 2 for steam generator level loop A were unaffected by cable degradation and operable. However, during the time that channel 1 or 2 was out-of-service for surveillance in the past, only one channel would have been operable. In such a condition, Technical Specification 3.3.3.3, Table 3.3-5, item 21, Action 31 would have been applicable. Review of past surveillance records (July 2001 to present) show that the channel is tested and restored to service on the same day, well within the seven-day allowed outage time of Action 31. Therefore, Technical Specifications were satisfied for LT 3-476 inoperability.
Reportability A review of the reporting requirements of 10 CFR 50.72 and 10 CFR 50.73 and NRC guidance provided in "Event Reporting Guidelines," 10 CFR 50.72 and 10 CFR 50.73 (NUREG-1022, Rev. 2) was performed for the subject condition. As a result of this review, the identified condition is reportable as described below.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 3 05000250 2004 -� 005 CO Technical Specification 3.3.3.3, Table 3.3-5, item 6 requires two pressurizer level channels to be operable in Modes 1, 2 and 3. Pressurizer level channels 2 and 3 would not have been operable for accident monitoring.
Pressurizer level channel 1 was operable. If one of the two required channels is inoperable, Action 31 allows 7 days to restore it to operable status or the unit must be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The time frame in the past that the two pressurizer level channels became inoperable is indeterminate. Therefore, Technical Specification 3.3.3.3, Table 3.3-5, Action 31 was exceeded in the past.
Action 32 applies when both required channels are inoperable and allows 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to return at least one channel to operable status. Review of past surveillance records (May 2001 to present) show that the channel is tested and restored to service on the same day. Therefore, Action 32 was not exceeded in the past.
Part 50.73(a)(2)(i)(B) of Title 10 CFR states that the licensee shall report "Any operation or condition which was prohibited by the plant's Technical Specifications except when:
(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions:
or (3) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event.
The allowed outage time of Technical Specification 3.3.3.3, Table 3.3-5, Action 31 was exceeded in the past for operability of two required pressurizer level channels (item 6). This condition was prohibited by the Technical Specifications, does not meet any of the exceptions identified above and so is reportable under 10 CFR 50.73(a)(2)(i)(B).
ANALYSIS OF SAFETY SIGNIFICANCE
Based on the analysis described below, it is concluded that the health and safety of the public were not affected by the degraded cable condition.
The degraded cables would not have affected reactor protection functions and engineered safety feature actuations.
Operability of pressurizer level channels 2 and 3 cannot be assured during post-accident conditions due to the degraded cables. Assuming both pressurizer level channels fail, pressurizer level channel 1 (LT-3-459) would need to be relied upon. With channel 1 available, control room operators still have an environmentally qualified pressurizer level transmitter to use to make decisions at various points throughout the Emergency Operating Procedures (EOP). Pressurizer level is used throughout the EOPs to assess adequacy of RCS inventory control. Additionally, the cold-calibrated pressurizer level channel (LT-3-462), although not used for EOP purposes and not environmentally qualified, might be available to back up the validity of the indication of LT-3-459. It is possible that with conflicting pressurizer level indication, the operator may conclude that pressurizer level information can not be used. Without pressurizer level DOCKETFACILITY NAME (1) LER NUMBER (6) PAGE (3)NUMBER (2) information, some recovery actions to an accident may be delayed since pressurizer level is used for safety injection termination criteria. Assurance of adequate core cooling can be obtained from the core exit thermocouples and the subcooled margin monitor [EIIS: AB, MON]. The degraded cables do not affect the subcooled margin monitor. Reactor coolant system level information is also available from the Reactor Vessel Level Monitoring System [EIIS: IP]. Thus, accident mitigation would be minimally impacted by the potential unavailability of pressurizer level.
Considering other plant functions impacted by the degraded cables, it was identified that post-accident environmental conditions may have resulted in spurious opening of one or both of the power-operated relief valves (PORV) [EIIS: AB, RV]. This is the result of pressurizer control and protection cables being degraded. Such spurious opening of the PORVs would be evident and require operator mitigating actions to close the block valves. These cables are not designed to be environmentally qualified, therefore, actions required for spurious opening are proceduralized. This is not considered to significantly complicate response to an accident. The safety injection system would provide adequate core cooling and reactor coolant system inventory control. Therefore, the health and safety of the public were not affected by the degraded cable condition discussed in this report and any increase in risk would be very small.
CORRECTIVE ACTIONS
1. All cables in close proximity to the RCS hot legs, RCS cold legs and pressurizer surge lines were removed and have been replaced with new cables. Note: RCS cold leg and pressurizer surge line cables were not degraded, but were replaced due to removal for inspection.
2. Gaps and other deficiencies in piping insulation were repaired. Insulation was installed on the uninsulated piping stubs.
3. All NCC ventilation registers in the vicinity of the RCS hot legs and cold legs were verified open prior to unit operation. The containment closeout inspection procedure was revised to ensure ventilation registers are in their proper position in the vicinity of the RCS hot and cold legs.
4. Temperature monitoring equipment (data loggers) was installed on conduits that are in close proximity above the RCS hot legs. The data loggers will record temperatures during the current operating cycle (Cycle 21) to establish the qualified life for the new cables.
5. The potential for cable degradation in the operating Unit 4 was evaluated to determine the impact on system and component operability. No conditions were identified that affected the required operability of systems and components. Unit 4 cable inspection and needed replacement will be performed during the next refueling outage currently scheduled for April 2005 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 3 05000250 NUMBERNUMBER
ADDITIONAL INFORMATION
second component function identifier (if appropriate)].
FAILED COMPONENTS IDENTIFIED:
The degraded cable is not considered failed since environmental conditions exceeded its qualification.
SIMILAR EVENTS
The original cable associated with pressurizer pressure transmitter PT-3-444 control loop was found defective and shorted during Unit 3's previous refueling outage (Cycle 20). This cable was replaced with a spare cable in the same conduit (3C226-2). The limited review at the time did not determine the cause of the cable failure. The original cable associated with pressurizer pressure transmitter PT-3-445 was found to be defective during the recent outage along with all other cables in conduit 3C226-2. This was a missed opportunity to identify the condition earlier.