ML20135G964
| ML20135G964 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/14/2020 |
| From: | Christopher Hunter NRC/RES/DRA/PRB |
| To: | |
| Littlejohn J (301) 415-0428 | |
| References | |
| LER 1990-001-00 | |
| Download: ML20135G964 (6) | |
Text
B-275 ACCIDENT SEQUENCE PRECURSOR PROGRAM EVENT ANALYSIS LER No.: 366/90-001 Event
Description:
Reactor scram with HPCI inoperable Date of Event: January 12, 1990 Plant: Hatch 2 Summary Hatch 2 scrammed from 100% power following main steam isolation valve (MSIV) closure due to a false low condenser vacuum signal. The high-pressure coolant injection (HPCI) system failed after one successful injection cycle (low water level initiation to high water level shut-off) as a result of an open heater strip associated with a thermal overload relay for the injection valve. The conditional core damage probability estimated for this event is 6.0 x 10-5. The relative significance of this event compared to other postulated events at Hatch 2 is shown below.
LER 366/90-001 1E-7 1E-6 1E-5 FlE- IE-3 IE-2 cuofLrip ['WOP L WF + HMc precur" uoff_ 360 h HPCI
+ RCIC Event Description Hatch 2 was operating at 100% power at 16 10 h on January 12, 1990, when an isolation globe valve in the vacuum sensing line from the main condenser failed closed. The valve failed closed when the valve disc separated from its stem and isolated the sensing line, resulting in an indicated low vacuum condition in the main condenser. In response to this signal, the MSIVs began to close, generating a reactor scram when the valves reached less than 90% open. During the subsequent transient, the reactor core isolation cooling (RCIC) system was manually started. to control reactor water level since the steam-driven reactor feed pumps (RFPs) were not available with the MSIVs closed; however, the water level still reached the low water level initiation point for HPCI, which
B-276 started and began adding water to the vessel. Water level was *restored and increased to the high water level shut-off of HPCI/RCIC, at which point these systems tripped off as expected. When the water level again decreased, the operators attempted to restart HPCI to control level, but the injection valve failed to open. The valve failed to open because of a failed heater strip in a thermal overload relay associated with the valve. HPCI was declared inoperable, and level was maintained in the vessel using RCIC and the two control rod drive (CRD) pumps. The plant was stabilized in hot shutdown, about 1 h after the scram.
Additional Event-Related Information The HPCI system is a high-pressure injection system designed for small-break loss-of-coolant accident (LOCAs) that do not depressurize the reactor. HPCI is an independent system, uses a turbine-driven pump, and automatically initiates on low reactor water level. HPCI can deliver 4250 gpm of makeup water to the vessel through the feedwater piping.
ASP Modeling Assumptions and Approach This event has been modeled as a loss of feedwater (because of the MISV closure) with HPCI unavailable. Although HPCI started and initially injected, it failed to function once reactor vessel water level decreased. Multiple start-stop cycles of HPCI are required for operability (had RCIC and the CRD pumps been unavailable, high-pressure injection would have been failed once the HPCI injection valve failed to reopen). The nonrecovery probability for HPCI (local valve operation) and feedwater (MSIVs opened) was assumed to be 0.34. An additional analysis was performed to characterize the possible inconsistency between the observed plant response during the event and the transient event tree model used in the ASP Program for plants like Hatch. In this model
[which is consistent with NUREG- 1150 and other probabilistic risk assessment (PRA) models], either HPCI, RCIC or two CRD pumps can provide adequate high-pressure cooling following a loss of feedwater. In this event, RCIC and both CRD pumps were required to maintain water level, even though RCIC was manually started before its auto-
.actuation point, and HPCI, in conjunction with RCIC, was used to initially refill the reactor vessel to the high level trip point. It is possible, based on this observation, that RCIC or the two CRD pumps alone are not capable of providing core cooling following a loss of feedwater. This cannot be positively ascertained, since insufficient information is provided in the LER to predict the ultimate reactor vessel level had only the two CRD pumps been available after HPCI failed.
To model the sensitivity of the conditional core damage probability to the possibility that
B-277 both RCIC and both CRD pumps are required for core cooling (instead of either RCIC or the CRD pumps), the branch probability for RCIC was raised to the sum of the RCIC and CRD branches (6.3 x 10-3) and the branch probability for the CRD branch set to 1.0, effectively reconfiguring the model to require both RCIC and the CRD pumps if HPCI is unavailable.
Analysis Results The conditional probability of severe core damage estimated for this event is 6.0 x 10-5.
The dominant sequence to core damage (illustrated in the following evenc tree) involves failure of high-pressure core cooling and failure to depressurize using the automatic depressurization system (ADS) following a transient-induced LOCA.
The sensitivity analysis exploring the impact of the potential need for both RCIC and CRD indicates a core damage probability (conditional on the loss of feedwater, HPCI failure, and the assumption that both RCIC and the CRD pumps are required in lieu of HIPCI) of 1.5 x 10-4, 2.5 times the conditional probability estimated assuming that either RCIC or the CRD pumps can provide for core cooling, as modeled in a typical boiling-water reactor (BWR) PRA.
B-278 OK OK OK II CORE DAMAGE OK 12 CORE DAMAGE OK Ox 13 DORE DAMAG OK OK 14 CORE DAMAGE OK OK
is CORE DAMAGE OK CORE DAMAGE CORE DAMAGE IA CORE DAMAGE
_________________________________ 20 CORE DAMA03E ON ON 21 CORE DAMAGE ON ON 22 CORE DAMAGE Ox ONt 2D ORE DAMAGE OK 26 CORE DAMAGE OK 26 ORE DAMAGE 27 CORE DAMAGE
__________________________________ 28 CORE DAMAGE ON ON 20 OR DAMAGE ON E OK 0 CORE DAMAGE OK E ON
"--- - 31 CORE DAMAGE cc ON 32 DORE DAMAGE ON OK 33 CORE DAMAGE ON ON 38 CORE DAMA03E ON 38 CORE DAMAGE ON 36 CORE DAMAGE 3G ORE DAMAGE
__________________ 38 GORE DAMAG2E 89 ATWS Dominant core damage sequence for LER 366/90-001
B-279 CONDITIONAL CORE DAMAGE PROBABILITY CALCULATIONS Event Identifier: 366/90-001 Event
Description:
Reactor scram with HPCI inoperable Event Date: 01/12/90 Plant: Hatch 2 INITIATING EVENT NON-RECOVERABLE INITIATING EVENT PROBABILITIES TRANS 1 .OE+DD SEQUENCE CONDITIONAL PROBABILITY SUMS End State/Initiator Probability CD TRANS 6.OE-05 Total 6.OE-05 ATWS TRANS 3.OE-05 Total 3.OE-05 SEQUENCE CONDITIONAL PROBABILITIES (PROBABILITY ORDER)
Sequence End State Prob N Rec**
28 trans -rx.ahutdown PCS/TRANS srv.chall/trans.-scram srv.close CD 5.3E-05 8.2 E-02 FW/PCS.TRANS HPCI srv.ads 11 trans -rx.shutdown PCS/TRANS srv.chall/trans.-scrain -srv.close CD 3.5E-06 7.6E-02
-FW/PCS.TRANS rhr(sdc) rhr(spcool) /rhr(sdc) 99 trans rx.'slutdown ATWlS 3.CE-05 1.0z+00
- non-recovery credit for edited case SEQUENCE CONDITIONAL PROBABILITIES (SEQUENCE ORDER)
Sequence End State Prob N Rec**
11 trans -rx.shutdown PCS/TRANS srv.chall/trans.-scram -srv.close Co 3.5E-06 7. 6E-02
-FW/PCS.TRANS rhr(sdc) rhr(spcool) /rhr(adc) 28 trans -rx.shutdown PCS/TRANS srv.chall/trans.-scram srv.close CD 5.3E-05 8 .2E-02 FW/PCS.TRANS HPCI arv.ads 99 trans rx.shutdown - ATWS 3 .DE-05 1.0E+00
- - non-recovery credit for edited case SEQUENCE MODEL: c:\asp\1989\bwrcaeal.cmp BRANCH MODEL: c:\asp\1989\hatch.sll PROBABILITY FILE: c:\asp\1989\bwr-csll.pro No Recovery Limit BRANCH FREQUENCIES/PROBABILITIES Event Identifier: 366/90-001
B-280 Branch System Non-Recov Opr Fail trans 6.1E-04 l.OE+00 loop 1.6E-05 3.6E-01 loca 3.3E-06 5. OE-01 rx .shutdown 3.OE-05 1. 0E+00 rx .shutdown/ep 3.5E-04 1.*OE+00 PCS/ TRANS 1.7E-01 > 1.02+00 1. 02+00 Branch Model: 1.OF.1 Train 1 Cond Prob: 1.7E-01 > Unavailable arv.chall/trans.-scram 1.02+00 1. 02+00 srv.chall/loop. -scram 1. OE+00 l.OE+00 srv.close 3.6E-02 1. 02+00 emerg.power 5.4E-04 8.0E-01 ep. rec 1.6E-01 1. 02+00 FW/PCS .TRANS 4.6E-01 > 1.02+00 3.4E-01 Branch Model: 1.0F.1 Train 1 Cond Prob: 4.6E-01 > Unavailable fw/pcs.loca 1 .02+00 3.4-01 HPCI 2.9E-02 > 1.02+00 7.02-01 > 3.4E-01 Branch Model: 1.OF.1 Train 1 Cond Prob: 2.9E-02 > Failed rcic 6 .OE-02 7.02-01 crd 1 .02-02 1.02+00 1. 0E-02 srv.ads 3.7E2-03 7.1E-01 1. 02-02 lpcs 3.02-03 3.4-01 lpci (rhr) /lpcs 1.02-03 7. 1E-01 rhr (sdc) 2A.E-02 3.4-01 1. 02-03 rhr(sdc) /-lpci 2.02-02 3.4E-01 1. 02-03 rhr(sdc) /lpci 1 .02+00 1. 02+00 1. 02-03 rhr (spcool) /rhr (sdc) 2.02-03 3.42-01 rhr(spcool) /-lpci.rhr(sdc) 2.OE-03 3.4-01 rhr(spcool) /lpci.rhr(sdc) 9.3E-02 1. 02+00 rhrsw 2 .OE-02 3.42-01 2. 02-03
- branch model file
- forced Mina rick 08-06-1991 17:29:42 Event Identifier: 366/90-001