ML20140A202
| ML20140A202 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah, Susquehanna  |
| Issue date: | 05/19/2020 |
| From: | NRC/RES/DRA/PRB |
| To: | |
| Littlejohn J (301) 415-0428 | |
| References | |
| LER 1992-010-00 | |
| Download: ML20140A202 (6) | |
Text
B-142 B.20 LER Number 328/92-010 Event
Description:
Emergency Diesel Generator and Residual Heat Removal Pump Inoperable Date of Event:
July 17, 1992 Plant:
Sequoyah Nuclear Plant, Unit 2 B.20.1 Summary During performance of a surveillance procedure on the 2B-B Residual Heat Removal (RHR) pump, it was found that the miniflow control valve continuously cycled open and closed when it should have remained opened. While the 2B-B RHR pump was inoperable, the 2A-A emergency diesel generator (EDG) was inoperable for 17 h and the 2A-A centrifugal charging pump was inoperable for 6 h. The conditional core damage probability estimated for this event is 1.9 x 10'. The relative significance of this event compared to other postulated events at Sequoyah, Unit 2 is shown in Fig. B.42.
L[R 328/92-010 ME7 precwuor m~toff 1W-5 1E,4 IE-3 1&-2 LOFW + 1 MTRAFW LOOP Fig. B.42.
Relative event significance of LER 328/92-010 compared with other potential events at Sequoyah 2.
B.20.2 Event Description On July 17, 1992, with the unit at 100% power, a quarterly surveillance procedure on the 2B-B RHR pump was conducted. During the test, it was discovered that the pump's miniflow line motor control valve was continuously cycling open and closed when it should have remained open.
Further investigation revealed that the valve had been miswired on July 1, 1992, during performance of the flow switch quarterly preventive maintenance procedure. Between July 1, 1992 and July 17, 1992, there were 10 instances where Train A safety equipment had been out of service. Only two of these LER NO: 328/92-010
B-143 instances were of a significant duration; EDG 2A-A was out of service for 17 h, and centrifugal charging pump (CCP) 2A-A was out of service for 6 h.
The wiring for the other RHR trains was verified to be correct.
B.20.3 Additional Event-Related Information The Sequoyah Units have miniflow lines for each of the RHR pumps. This flow path consists of the pump, a flow sensor, the RHR heat exchanger, and a recirculation line that returns to the pump suction.
The recirculation line contains a motor-operated flow control valve that varies its position, based on the pump discharge flow signal, to maintain total pump flow between 500 and 1500 gal/min. Manual control and indication of the valve's position is available in the control room.
During an accident, the pump would be aligned for reactor coolant system (RCS) injection. However, the pump would be in the recirculation mode until RCS pressure drops below the pump deadhead pressure, or the RHR system is realigned to the safety injection pump suction during the recirculation phase.
The recirculation valve does not have any thermal overloads and may fail after 15 min of continuous operation. With the valve closed and RCS pressure greater than the RHR pump deadhead pressure, insufficient flow through the pump could damage the pump because of overheating. With the valve fully open, flow to the RCS would be insufficient to ensure accident mitigation under large break LOCA conditions. Because the valve continuously cycled opened and closed, the actual time to failure of the RHR pump is more difficult to predict.
The two CCPs fulfill part of the emergency core cooling system (ECCS) function.
The discharge pressure of the pumps (2670 psig) is greater than normal RCS pressure. The two high pressure safety injection (HPSI) system pumps have a discharge pressure of 1650 psig. All four pumps are used during initial injection and during long term recirculation cooling. During the recirculation mode, the lA-A RHR pump supplies the lA-A safety injection (SI) pump and both CCPs. The lB-B RHR pump supplies only the 1B-B SI pump.
B.20.4 Modeling Assumptions The event was modeled as a potential LOOP assuming the 2B-B RHR train and the 2A-A EDG were inoperable for 17 h. Equipment associated with the train 2A-A EDG (2A-A AFW pump, 2A-A SI pump, 2A-A RHR pump) is rendered inoperable due to loss of electrical power. Both trains of high-pressure recirculation were inoperable because both trains of RHR were inoperable.
The current Accident Sequence Precursor (ASP) models do not account for the separate high head CCPs and intermediate head systems (SI) that Sequoyah uses for the ECCS function. Inoperability of one train of RHR and one train of charging is not normally analyzed in the ASP program. Therefore the 6-hour CCP train/RHR train inoperability was not considered a precursor, and, as a result, was not analyzed.
For the 17 h RHR train/EDG inoperability, the HPI system model was modified to incorporate the CCPs.
The modification was performed as follows.
LER NO: 328/92-010
B-144 p(HPI system)
= [p(HPI train 1) x p(HPI train 2)] x [p(CCP train 1) x p(CCP train 2) + p(CCP valves)]
= [0.01 x 1.0] x [0.01 x 1.0 + 0.0011]
= 1.11 x 10-4 p(CCP valves)
= 4 x [vlvI x (vlv2 + BETA v)]
= 4 x [0.003 x (0.003 + 0.088)]
= 0.001092 B.20.5 Analysis Results The conditional probability of core damage estimated for this event is 1.9 x 10-.
The dominant sequence, highlighted on the event tree in Fig. B.43, involves a postulated LOOP with failure of on-site emergency power, and failure to recover offsite power prior to a RCP seal LOCA.
LOOP I Ld I
L O
HPI I
I I
I CHAL RESEAT LOCA (LONG)
HPR PORV IOPEN SEQ END NO STATE OK OK 41 CD 42 CD OK OK 43 CO 44 CD 45 CD OK 46 CD 47 CD 48 CD OK 49 CD 50 CD OK 51 CD 52
-CD 53 CO OK 54 CD 55 CD 40 ATWS Fig. B.43.
Dominant core damage sequences for LER 328/92-010 I
LER NO: 328/92-010
B-145 LER NO: 328/92-010
B-146 LER NO: 328/92-010
B-147 LER NO: 328/92-010