05000250/LER-2014-001
| Turkey Point Unit 3 | |
| Event date: | 1-3-2014 |
|---|---|
| Report date: | 3-4-2014 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2502014001R00 - NRC Website | |
DESCRIPTION OF THE EVENT
On January 3, 2014 with the Unit 3 reactor [AC, RCT] in Mode 1 at 100% power during performance of the channel calibration surveillance procedure for Main Steam Line Pressure Transmitter [JE, PT] PT-3-495, the instrument channel was found outside procedural acceptance criteria due to PT drift. PT-3-495 was replaced, calibrated successfully, and returned to service on January 4, 2014.
PT-3-495 drifted such that the output was higher than the actual value of steam pressure and its associated trip function would not have protected the safety analysis limit rendering the channel inoperable. Subsequent review determined the instrument channel is considered inoperable from the time of the due date for a missed surveillance test of March 9, 2013 to January 3, 2014.
During the period of inoperability, the instrument channel associated with PT-3-495 remained in service exceeding the allowed outage time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with the required action to place the channel in the tripped condition not taken and the shutdown actions of Technical Specification (TS) 3.0.3 not entered. Therefore, this event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by the TS.
CAUSE OF THE EVENT
The root cause is attributed to a deviation from the normal process of using a dedicated work order (WO) to satisfy Surveillance Requirements (SR), which created an error likely situation for which no compensatory action was established. As result, Instrumentation and Control (I&C) supervision failed to validate WO activities credited for satisfying TS requirements.
ANALYSIS OF THE EVENT
Background
PT-3-495 is an input to Engineered Safety Features Actuation System TS Table 3.3-2, Functional Unit 1.e, High Differential Pressure Between the Steam Line Header and any Steam Line. PT-3-495 provides the Steam Generator (SG) [SB, SG] 'C' Channel III pressure signal for this function. SG pressure is equivalent to Steam Line pressure referred to in the TS. This function is required in Modes 1, 2 and 3 (Mode 3 if Pressurizer [AB, PZR] pressure is greater than or equal to 2000 psig). This transmitter also provides the steam pressure correction for the Steam Flow/Feedwater Flow mismatch coincident with SG Water Level - Low reactor trip signal (TS Table 3.3-1, Functional Unit 12). Any error in measured steam flow affects the accuracy of the steam flow computation.
The function provided by PT-3-495 is used in the analysis for secondary side high energy line breaks. The transmitter provides the pressure signal (one of three channels) for the individual steam line associated with the 'C' SG to the instrument channel. The function initiates safety injection (SI) [BQ], feedwater line isolation, and steam line isolation.
Updated Final Safety Analysis Report, Chapter 14, Section 14.2.5 describes the Steam Line Rupture accident analysis. In this analysis, steam release from more than one SG will be prevented by automatic closure of the Main Steam Isolation Valves (MSIV) [SB, ISV] in the steam lines either by high containment [NH] pressure signals or by high steam flow coincident with low steam line pressure. For a break upstream of the MSIVs, MSIV closure is not required due to the presence of the Main Steam Check Valves [SB, SHV], which prevent blowdown of the non-faulted SGs. In this case, SI actuation would occur from high differential steam pressure between the faulted steam line and the main steam header.
SI is actuated by high steam line differential pressure in any loop. The differential pressure signals are derived by two out of three low steam line pressure signals in one line compared with main steam header pressure. The SI signal will subsequently trip the reactor and actuate main feedwater isolation. A manual block of this signal is allowed when Pressurizer pressure is below 2000 psig. The SI System provides borated water to the reactor core during a steam line break in order to counter any positive reactivity insertion due to cool down of primary coolant. Main feedwater isolation and steam line isolation are initiated to limit primary system cool down and to limit the mass and energy releases through the break.
Analysis On April 26, 2011, PT-3-495 was successfully calibrated. An I&C supervisor incorrectly approved the next surveillance test as complete. The approval was based on the review of WOs associated with replacement of instrument channel modules; however, while the WOs did specify post maintenance tests they did not include calibration of the associated transmitter. Normally, a dedicated WO would be used for the surveillance test.
As a result, the TS SR for calibration was not met. This missed SR was subsequently recognized and the test was performed on January 3, 2014 which determined the channel had been inoperable.
While a specific date cannot be established for when the transmitter drifted out of specification, the date the missed surveillance was due of March 9, 2013 is used consistent with the guidance in NUREG-1022, Revision 3, Section 3.2.2 for the start of inoperability. The transmitter was in a degraded condition likely due to age and was replaced.
Reportability The TS Limiting Condition for Operation (LCO) 3.3.2, Table 3.3-2, Functional Unit 1.e requirement for three operable channels was not met during the period of inoperability. Because the required action to place the channel in trip was not met within the allowed outage time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the shutdown actions of TS 3.0.3 were not entered a condition prohibited by the TS existed, reportable in accordance with 10 CFR 50.73(a)(2)(i)(B)
ANALYSIS OF SAFETY SIGNIFICANCE
While Steam Generator Pressure Channel III could respond to the lower pressure of a steam line break, it could not do so within its required range and so was inoperable. Safety significance remained low during the period Channel III was inoperable because both redundant Channels II and IV (there is no Channel I) remained available and were proven to be operable by the performance of their respective surveillances.
Regarding the failure of Channel III, the assumed failure of PT-3-495 has no noticeable impact on Core Damage Frequency (CDF). Any delta CDF is well below the NRC threshold of 1E-6 per year for low risk significance.
CORRECTIVE ACTIONS
Corrective actions are in accordance with condition report 1938191 and include the following:
1. PT-3-495 was replaced, the channel was tested satisfactorily and returned to service.
2. The surveillance tracking program procedure will be revised to state that an independent verification is to be performed and documented in the surveillance tracking program prior to approval that a surveillance test has been completed when crediting non-dedicated WOs.
3. A review of WOs was conducted to ascertain the extent of condition.
ADDITIONAL INFORMATION
component function identifier (if appropriate)].
FAILED COMPONENTS IDENTIFIED: Rosemount Pressure Transmitter 1153GB PREVIOUS SIMILAR EVENTS: None