05000499/LER-2007-001
| South Texas | |
| Event date: | 03-14-2007 |
|---|---|
| Report date: | 05-10-2007 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 4992007001R00 - NRC Website | |
This event is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B). Technical Specification 3.7.1.2 requires that four independent steam generator auxiliary feedwater pumps and associated flow paths be operable. However, an auxiliary feedwater pump (AFW) and its associated flow path were inoperable longer than the allowed outage time. Consequently, STP Unit 2 was in a condition prohibited by Technical Specifications.
B. PLANT OPERATING CONDITIONS PRIOR TO EVENT
STP Unit 2 was in Mode 1 at 84% power conducting coast down operations in preparation for an upcoming refueling outage.
C. STATUS OF STRUCTURES, SYSTEMS, AND COMPONENTS THAT WERE INOPERABLE
AT THE START OF THE EVENT AND THAT CONTRIBUTED TO THE EVENT
No other inoperable structures, systems, or components contributed to the event.
D. NARRATIVE SUMMARY OF THE EVENT
On March 5, 2007, Unit 2 AFW Pump 23 was started for a post maintenance test. During the test, it was noted that the pump discharge flow was not as expected. Investigation determined that the closed Long Path Recirculation Isolation Valve 2-AF-0092 was leaking by its seat.
The operational impact of this condition is that the design bases flow to the steam generator was not achieved.
During troubleshooting of this condition, an installed "knocker hand wheel" was used to verify that valve 2-AF-0092 was closed. After this operation of the valve, leakage past the valve seat increased further. Investigation determined that there was no lubrication on the portion of the valve stem just below the actuator. When the valve actuator was disassembled, the stem nut was found broken into two pieces.
A preventive maintenance requirement to lubricate the portion of the valve stem just below the actuator was cancelled in 1993. It was incorrectly thought that the valve had a sealed gear case and lifetime lubrication so that any loss of lubrication from the stem would be gradual and any unusual operation of the valve would be noted and reported prior to failure.
The valve stem moves up and down when engaged by the rotating stem nut during valve operation. Rotational movement of the valve stem is restrained by an anti-rotation device.
During inspection prior to the start of maintenance on March 5, 2007, it was noted that the anti-rotation device was digging into a metal guide when the valve was in its final position.
2-AF-0092 is stroked during monthly surveillance tests. It was concluded that the valve must not have fully closed during the last surveillance test on February 9, 2007, and could have been in this condition since a surveillance test performed on January 10, 2007. The rolled metal on the anti-rotation device and the lack of lubrication on the valve stem combined to make the valve feel closed. The unusual pump discharge flow noted on March 5, 2007 was not checked, and therefore not noted, during the surveillance tests performed on February 9, 2007 and January 10, 2007. Further, it was determined that the stem nut failed due to brittle IRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION -2001) IR_ N kRRATIVE (If more space is required, use additional copies of NRC Form 366A) (17) I overload when the valve was closed during troubleshooting in March using the "knocker hand wheel.If 2-AF-0092 was repaired and AFW Pump 23 and its associated flow path were restored to service on March 9, 2007. On March, 14, 2007 it was determined that the inoperable condition of the valve existed for a period of time longer than the allowed outage time of the Technical Specifications.
E. METHOD OF DISCOVERY OF EACH COMPONENT FAILURE, SYSTEM FAILURE, OR
PROCEDURAL ERROR
This condition was identified during post maintenance testing of the Unit 2 AFW Pump 23.
II. COMPONENT OR SYSTEM FAILURES
A. FAILURE MODE, MECHANISM, AND EFFECTS OF EACH FAILED COMPONENT
Long Path Recirculation Isolation Valve 2-AF-0092 was inspected. The inspection revealed that the valve stem just below the actuator had no lubrication. The additional torque caused by friction between the stem nut and the stem caused enough friction between the anti-rotation device and its guide that it made the metal roll when the valve was almost closed. The rolled metal on the anti-rotation guide and the lack of lubrication on the valve stem combined to make the valve feel closed even though it wasn't. The effect was a failure of 2-AF-0092 to fully shut when manually operated. After subsequent disassembly of the actuator, the stem nut was found to be broken into two pieces. This condition resulted in an increased flow past the valve seat. The failure was a Maintenance Rule Functional Failure.
B. CAUSE OF EACH COMPONENT OR SYSTEM FAILURE
The amount of friction between the threads of the valve stem and the stem nut increased because the valve stem was not lubricated adequately. As the amount of friction between the threads increased, the amount of torque on the stem and stem nut increased. When the stem tried to turn, it was restrained by the anti-rotation device being restrained from turning by the guide on the valve yoke. When the anti-rotation device was pushed hard against the guide, galling occurred between the metals of the device and guide which increased the torque needed to move the stem down even more.
The additional torque stresses due to lack of stem lubrication combined with the increased tensile stress due to using the "knocker hand wheel" and due to having to overcome the rolled metal on the anti-rotation device guide were enough to cause brittle overload failure of the stem nut.
Extent of condition testing and evaluation determined that this event did not have a common cause.
Incremental Change in Core Damage Probability was calculated with no credit given for the AFW Pump 23 degraded flow capability. In addition, a sensitivity case was calculated to demonstrate the risk reduction when crediting degraded AFW Pump 23 flow capability. The results are as follows:
Exposure Exposure Base Sensitivity Exposure Dates Time Time Case Case 1 (Days) (Hours) ICCDP ICCDP 2/9/07 to 3/9/07 28.5 683.7 1.16E-06 3.0E-07 1/10/07 to 3/9/07 57.6 1383.3 2.34E-06 6.0E-07 Deterministic calculations have shown that the degraded flow (469 gpm) from the AFW Pump 23, that represents the actual flow path condition during the inoperable period of time, was capable of removing reactor decay heat and sensible heat for RCS cool downs. The conservative sensitivity case shows that the core damage risk is reduced below 1E-06. Therefore, this event was of very This event resulted in no personnel injuries, no offsite radiological releases, and no damage to other safety-related equipment.
IV. CAUSE OF THE EVENT
The cause of Long Path Recirculation Isolation Valve 2-AF-0092 not fully closing on February 9, 2007 and the stem nut failing on March 6, 2007 is that no periodic preventive maintenance existed to lubricate the stem.
V. CORRECTIVE ACTIONS
1. Long Path Recirculation Isolation Valve 2-AF-0092 was repaired and lubricated on March 9, 2007.
2. The long path recirculation isolation valves for each AFW System train in both units were tested satisfactorily to verify they did not have enough seat leakage to adversely impact their functionality.
were cleaned, lubricated and inspected. The auxiliary feedwater flow path long path recirculation isolation valves in Unit 1 will be cleaned, lubricated and inspected by July 24, 2007.
4. Surveillance procedures were revised to include testing to verify that the auxiliary feedwater flow path long path recirculation isolation valves do not have excessive seat leakage.
5. Plant Generation Risk and Graded Quality Assurance High and Medium Risk-ranked components in the AFW System will be reviewed by May 31, 2007 to determine the adequacy of current preventive maintenance scope and frequencies.
VI. PREVIOUS SIMILAR EVENTS
STP reviewed Equipment History for previous failure of valve stem nuts. Of seven valves identified with either failed or severely damaged stem nuts, none were ranked as High or Medium Risk- Significant. The Equipment History review determined eight conditions where valves had stems or stem nuts that lacked lubrication. Only one of these valves is ranked as risk significant and this valve has preventive maintenance for performing lubrication.
The Equipment History review identified six occasions where main steam power-operated relief valves failed to close or were hard to operate. An apparent cause evaluation determined that part of the cause was that the lubricant used in the valve actuator would break down and harden at the high temperatures experienced by the valves. It was concluded that the lubricant in the auxiliary feedwater system valves was not expected to separate and harden in the milder environment in which these valves operate.
VII. ADDITIONAL INFORMATION
STP plans to conduct a review of risk significant components to determine which components do not have active preventive maintenance activities. The critical attributes that made the component IIII
- IRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION 1-2001) 2. DOCKET1. FACILITY NAME 6. LER NUMBER 3. PAGE I� N kRRATIVE (If more space is required, use additional copies of NRC Form 366A) (17) I risk significant will be considered when determining whether preventive maintenance activities should be created.