05000388/LER-2015-001
| Susquehanna Steam Electric Station, Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 3882015001R01 - NRC Website | |
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
CONDITIONS PRIOR TO EVENT
Unit 1 — Mode 1, 99 percent Rated Thermal Power Unit 2 — Mode 1, 100 percent Rated Thermal Power Other than the drifted switch itself, there were no structures, systems, or components that were inoperable at the start of the event that contributed to the event.
EVENT DESCRIPTION
Susquehanna Technical Specification (TS) 3.3.5.1 requires operability of four Reactor Steam Dome Pressure - Low channels to provide the injection permissive for the Core Spray system [EIIS System Identifier: BM] (Function 1d) and the Low Pressure Coolant Injection system (LPCI) [EIIS System Identifier:
BO] (Function 2d). The Reactor Steam Dome Pressure-Low signals are initiated from four pressure switches [EIIS System/Component Identifier: JE/PS] that sense the reactor dome pressure. The pressure switches are set to actuate between the Upper and Lower Allowable Values on decreasing reactor dome pressure. The Upper Allowable Value is low enough to ensure that the reactor dome pressure has fallen to a value below the Core Spray and RHR/LPCI maximum design pressures to preclude over pressurization.
The Lower Allowable Value is high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46.
On January 21, 2015 during quarterly calibration of The Reactor Steam Dome Pressure — Low switches, PIS-B21-2N021A was found outside of the Technical Specification allowable value. As found was 447.8 psig vs. an upper allowable limit of 446.7 psig. PIS-B21-2N021A was subsequently restored with an as left value of 433.0 psig.
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (-1-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
Following are the setpoints and upper and lower allowable values for each of the Unit 2 switches:
PIS-B21-2N021A PIS-B21-2N021B PIS-B21-2N021C PIS-B21-2N021D Setpoint 433.7 432.7 441.5 442 Upper Allowable 446.7 445.7 454.5 455 Lower Allowable 420.7 419.7 428.5 429 On February 27, 2015, the history of PIS-B21-2N021A and the switches in the other three channels were evaluated and the following was identified:
Date PIS-B21-2N021A PIS-B21-2N021B PIS-B21-2N021C PIS-B21-2N021D 1/21/2015 447.8 * 439.7 443 445.8 10/19/2014 435.11 432.19 438.5 440.81 8/8/2014 413.4 * 423.4 439.9 433.2 4/25/2014 435.4 433.4 442.8 439 2/8/2014 451.4 * 444.6 447.5 447.5 10/25/2013 436.7 435.6 446.5 446.4 7/26/2013 410.5 * 418.6 * 435.7 435.5 4/12/2013 438.1 429.1 446 444 1/25/2013 447 * 442 443.6 447.7 10/25/2012 434.7 433.1 440.8 443.3 7/27/2012 422.4 421.2 433.8 435.5 5/11/2012 427.9 429 440.6 438 1/19/2012 452 * 445.7 449.3 446 * Values outside Technical Specification allowable comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
Following are the setpoints and upper and lower allowable values for each of the Unit 1 switches:
The history of the Unit 1 switches were evaluated and the following was identified:
- Values outside Technical Specification allowable comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by interne( e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
NUREG-1022, Revision 3 states that discrepancies found in TS surveillance tests should be assumed to occur at the time of the test unless there is firm evidence, based on a review of relevant information (e.g., the equipment history and the cause of failure), to indicate that the discrepancy occurred earlier. Based on the above information, firm evidence is considered to exist to indicate that the condition existed prior to the time of discovery. Based on this conclusion, the condition is considered to be a condition prohibited by Technical Specifications and reportable in accordance with 10 CFR 50.73(a)(2)(i)(B). In addition, on July 26, 2013, two Unit 2 channels (A and B) were found outside of the Technical Specification allowable values. With both the A and B channels inoperable at the same time, the safety functions of both core spray and LPCI were impacted. Core spray and LPCI remained capable of performing their design function as analyzed in the FSAR accident analysis. Based on this information, the condition is considered reportable as a common cause inoperability of independent trains or channels (10 CFR 50.73(a)(2)(vii)), a condition that could have prevented fulfillment of a safety function (10 CFR 50.73(a)(2)(v)(D)), and a single cause that could have prevented fulfillment of safety functions of trains or channels in different systems (10 CFR 50.73 (a)(2)(ix)(A)) for Unit 2.
CAUSE OF EVENT
The direct cause was instrument set point drift caused by unrecognized seasonal temperature and humidity effects and mechanical hysteresis.
Temperature and humidity are issues with the bellows / linkages of the pressures switch materials and micro-switches. The manufacturer did not test these pressure switches to show the effects of temperature and humidity. The Barton 288A switches are not temperature / humidity stabilized. It is believed that these instruments exhibit a hysteresis due to the mechanical action of the bellows when used in applications above their normal operating range.
The apparent cause was determined to be less than adequate design. The design failed to consider the effects of mechanical hysteresis of the subject devices when operated above their normal operating range and did not recognize the seasonal temperature effects to the set points.
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
ANALYSIS/SAFETY SIGNIFICANCE
Actual consequences: There were no actual safety consequences associated with this event. In July 2013, the Unit 2 A and B switches were found out for the Technical Specification allowable values. The as-found values for the A and B switches, as corrected for head pressure, were 396.8 psig and 405.9 psig. The Technical Specification lower allowable value is greater than or equal to 407 psig.
From FSAR Table 6.3-2C, the analytical limit used in the LOCA analysis for the switches is 400 psig. From Calculation EC-080-1007, Revision 1, the uncertainty due to accuracy and calibration for the low pressure switches is 6.41 psig. Therefore, to compare the as-found setpoints on July 26, 2013 to the analytic limit used in the LOCA analysis, the uncertainty must be subtracted from the as-found values that have been corrected for head pressure. This results in a value of 390.39 psig for Switch A and 399.49 for Switch B.
Based on this information, the injection pressure permissive for Core Spray and LPCI on July 26, 2013 was 399.49 psig instead of the analytical limit of 400 psig used in the LOCA analysis. The Susquehanna fuel vendor, AREVA, was contacted to determine the impact of the as-found injection permissive of 399.49 psig on the peak cladding temperature (PCT) in the LOCA analysis. The AREVA analysis showed a maximum increase of 0.1 degrees F on PCT for the limiting break and a 0 degree F increase in the PCT for the limiting small break LOCA. The PCT for the limiting break is 1844 degrees F; therefore, the resulting PCT remains less than the LOCA PCT acceptance criterion of 2200 degrees F.
Potential consequences: The automatic initiation for the CS/LPCI injection valves would be lost if complete failure of the A and B OR C and D pressure indicating switches occurred concurrently. The ability to open the low pressure ECCS injection valves would remain available; however, opening would require a manual operator action.
As-found set point values drifting out of the allowable tolerance could result in a delayed automatic response if A and B OR C and D pressure indicating switches were concurrently out of as-found tolerance low.
CORRECTIVE ACTIONS
Key corrective actions include the following:
1. As an interim corrective action, the calibration frequency for the reactor steam dome — low switches on both units has been increased from quarterly to once every 45 days. The increased calibration frequency will continue until the switches are replaced with switches that are correct for the application.
2. Replace the PIS-B21-1N021A, B, C, and D and in PIS-B21-2N021A, B, C, and D with switches that are correct for the application.
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
COMPONENT FAILURE INFORMATION
The switches that drifted are Barton Model 288A pressure switches. The current company name is Cameron International Corporation.
PREVIOUS SIMILAR EVENTS
Trip Setpoints Out of Calibration," dated July 18, 2014.