|Quad Cities Nuclear Power Station Unit 2|
|Reporting criterion:||10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident|
|ENS 52758||10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident|
|2652017001R00 - NRC Website|
|Person / Time|
|From:||Ohr K S|
Exelon Generation Co
Document Control Desk, Office of Nuclear Reactor Regulation
|Download: ML17194A817 (5)|
comments regarding burden estimate to the FOIA, Pnvacy and Information Collections Branch (T-5 F53), U S Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, 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.
3. LER NUMBER
2017 - 00 001
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
A. CONDITION PRIOR TO EVENT
Unit: 2 Reactor Mode: 1 Event Date: May 15, 2017 Event Time: 19:18 hours Mode Name: Power Operation Power Level: 100%
B. DESCRIPTION OF EVENT
On May 15, 2017 at 19:18 Central, station Operations personnel were performing Procedure QCOS 2300-05, "HPCI Pump Operability Test," which ensures the Motor Operated (MO) 2-2301-14, "HPCI Minimum Flow Valve," opens as pump flow decreases. During the test, the valve unexpectedly failed to open. Operators took steps to open the valve manually, but upon release of the control switch, the valve immediately returned to the closed position. Operators left the valve in the closed position. The Unit 2 HPCI system remained available but was declared inoperable.
Troubleshooting determined that the failure was due to a failure of the commercially dedicated HPCI Flow Indicating Switch (FIS) 2-2354. FIS 2-2354 had been recently replaced on February 2, 2017, under a corrective work order utilizing a commercially dedicated switch. The FIS was bench tested and post maintenance tested satisfactorily on February 2, 2017. The switch was again calibrated under a preventative maintenance work order during the next quarterly system work window on May 15, 2017, in accordance with a 92-day preventative maintenance frequency.
All results were satisfactory. The switch then failed during the subsequent quarterly system run on the same day. In the follow-up failure analysis Power Labs reported that the switch had an intermittent failure of the high side micro switch due to residual material remaining from the manufacturing process.
On May 15, 2017, at 23:27 Central, ENS #52758 was made to the NRC under 10 CFR 50.72(b)(3)(v)(D), to report this event as an event or condition that could have prevented the fulfillment of a safety function.
Given the impact on the HPCI system, this report is submitted for Unit 2 in accordance with the requirements of 10 CFR 50.73(a)(2)(v)(D), which requires the reporting of any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.
comments regarding burden estimate to the FOIA, Pnvacy and Information Collections Branch (T-5 F53), U S Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects Resource@nrc gov, and to the Desk Officer, Office of Information and Regulatory Affairs, 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
3. LER NUMBER
2017 00 001
C. CAUSE OF EVENT
The cause was determined to be failure of the HPCI Pump Discharge Flow Indicating Switch, specifically intermittent failure of the high side micro switch caused by residual material from the manufacturing process.
D. SAFETY ANALYSIS
System Design According to the Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2 Updated Final Safety Analysis Report (UFSAR) Section 220.127.116.11, the HPCI subsystem is designed to pump water into the reactor vessel under Loss of Coolant Accident (LOCA) conditions which do not result in rapid depressurization of the pressure vessel. The loss of coolant might be due to a loss of reactor feedwater or to a small line break which does not cause immediate depressurization of the reactor vessel. The sizing of the HPCI subsystem is based upon providing adequate core cooling during the time that the pressure in the reactor vessel decreases to a value that the Core Spray [BM] subsystem and/or the Low Pressure Coolant Injection (LPCI) [BO] subsystem become effective. The HPCI subsystem is designed to pump 5600 gallons per minute into the reactor vessel within a reactor pressure range of about 1120 pounds per square inch gage (psig) to 150 psig. Initiation of the HPCI subsystem occurs automatically on signals indicating reactor low-low water level or high drywell pressure. HPCI injection into the reactor vessel may be accomplished manually by the operator or without operator action by the HPCI automatic initiation circuitry. HPCI can also operate in a pressure control mode of consuming steam from the reactor vessel without providing full injection into the vessel (down to and including zero injection).
Safety Impact The safety impact of this condition was minimal. Valve MO 2-2301-14 is a normally closed valve that is used to maintain minimum flow through the HPCI pump to the Suppression Pool when the injection isolation valves are shut.
The bypass valve is automatically opened on low pump flow and closed on high flow whenever the steam supply valve to the HPCI turbine is open. According to UFSAR Section 18.104.22.168.3, the HPCI minimum flow system is provided for pump protection. The minimum flow valve is automatically opened on low pump flow and closed on high flow whenever the steam supply valve to the turbine is open. Even though the minimum flow valve failed to open, the HPCI System remained capable of performing its intended design/safety function and would not have hindered the system from fulfilling any required safety function or injection over the required 10 minute mission time.
According to UFSAR Table 6.2-7, MO 2-2301-14 is considered a primary containment isolation valve [ISV] with a normal position of closed. MO 2-2301-14 is the only primary containment isolation valve present in line 2-2340-4"-DX.
The failure of the FIS caused MO 2-2301-14 to fail in and remain in the closed position. Since the line could effectively be isolated utilizing the primary containment isolation valve, the primary containment integrity could be assured, therefore, the primary containment system remained capable of performing its intended design/safety function.
With MO 2-2301-14 failing to open, HPCI was declared inoperable and TS 3.5.1 Condition G was entered. Required Action G.2 is to restore HPCI System to OPERABLE status, with a completion time of 14 days. The FIS was replaced, post maintenance tested and the HPCI system declared operable within the 14 day completion time. There were no other issues or problems identified with valve MO 2-2301-14 after replacement of the FIS. No other repairs were required or performed.
Since HPCI is a single train safety system, this notification is being made in accordance with 10 CFR 50.73(a)(2)(v)(D) (Event or Condition that Could Have Prevented Fulfillment of a Safety Function).
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 e-mail to Infocollects.Resource@nrc gov, and to the Desk Officer, Office of Information and Regulatory Affairs, 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
3. LER NUMBER
2017 - 00 001 The engineering analysis that was performed demonstrated this event did not constitute a Safety System Functional Failure (SSFF). (Reference NEI 99-02, Revision 7, Regulatory Assessment Performance Indicator Guideline, Section 2.2, "Mitigating Systems Cornerstone, Safety System Functional Failures, Clarifying Notes, Engineering Analyses.") As such, this event will not be reported in the NRC Performance Indicator (PI) for SSFF since an engineering analysis was performed which determined that the system was capable of performing its safety function during this event.
Risk Insights The plant Probabilistic Risk Assessment (PRA) model gives no credit for failure of the HPCI Minimum Flow Valve failure to open and does not include it in the model; hence, the failure of the valve to open did not contribute to an increase in risk.
In conclusion, due to the fact that the HPCI system was available to perform its safety function, the overall safety significance and impact on risk of this event were minimal.
E. CORRECTIVE ACTIONS
1. The failed FIS was replaced with a new commercially dedicated and bench tested FIS and post maintenance tested satisfactorily.
1. Site Procurement will purchase the FIS from an approved Appendix B program vendor rather than purchasing a commercially dedicated part.
2. The Calibration Data Sheet for the FIS will be revised to include cycling new switches ten (10) times during bench testing.
F. PREVIOUS OCCURRENCES
The station events database, LERs, and INPO Consolidated Event System (ICES) were reviewed for similar events at QCNPS. This event was caused by an intermittent failure of the high side micro switch of the HPCI FIS which contained residual material from the manufacturing process.
No previous occurrences were identified as applicable to the circumstances of this event.
G. COMPONENT FAILURE DATA
Failed Equipment: Flow Indicating Switch Component Manufacturer: Barton Component Model Number: 289A Component Part Number: N/A This event has been reported to ICES as Report No. 415432