ML20135G750
| ML20135G750 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/14/2020 |
| From: | Christopher Hunter NRC/RES/DRA/PRB |
| To: | |
| Littlejohn J (301) 415-0428 | |
| References | |
| LER 285-1990-009 | |
| Download: ML20135G750 (4) | |
Text
B-154 ACCIDENT SEQUENCE PRECURSOR PROGRAM EVENT ANALYSIS LER No.:
Event
Description:
Date of Event:
Plant:
285/90-009 Unavailability of AFW following postulated MSLB March 16, 1990 Fort Calhoun Summary During a design basis reconstitution effort, it was discovered that the increase in containment temperature following a main steam line break (MSLB) inside containment would pressurize the fluid between the normally closed auxiliary feedwater (AFW) isolation valves such that they could not be opened. This would effectively fault AFW following a MSLB. The conditional probability of core damage associated with this event is 1.1 x 10-5. The relative significance of the event compared to other postulated events at Fort Calhoun is shown below.
-LER 285/90-009 1E-7 IE-6 1E/1E-4 1E-3 1E-2 TripO 36Oh EP LOFW + 1 36h precursor cutoff -J
~
NTR AFW Event Description While performing the design basis reconstitution effort at Fort Calhoun, it was discovered that AFW flow to both steam generators (SGs) would be faulted following a main steam line break inside containment. Following such a break, there would be a rapid increase in containment temperature. Normally closed AFW containment isolation valves HCV-1I 107A, -1I 107B, -1I 108A, and -1I 108B are commanded open on low steam generator water level. This would occur at least 1 min into the accident. Because of the long length of uninsulated piping between the containment wall and the inner AFW isolation valves, thermal expansion would cause the pressure in the piping between each pair of isolation valves to increase to >3000 psig before the valves received open signals.
B-155 This pressure would prevent the AFW valves from opening, effectively faulting AFW following a MSLB.
This situation would also occur following a large loss-of-coolant accident (LOCA),
although the containment temperature is lower than with a MSLB.
Additional Event-Related Information The AFW piping is designed as USAS B3 1.7, and seismic category 1. The design pressure and temperature of the piping is 1660 psig at 550'F. There is a separate piping path for AFW supply to each of two SGs. Control valves HCV-1 107A, HCV-1 107B, HCV-l1 108A, and HCV-lI 108B are pneumatically-operated containment isolation valves in the AFW lines, and are closed during normal power operation. They are automatically operated by the AFW actuation signal, remote/manually-operated from the main control room panels, or manually-operated from the local control panel. The "A" suffix designates that the valve is inside containment, while the "B" suffix denotes outside containment.
ASP Modeling Assumptions and Approach The event has been modeled as a nonrecoverable unavailability of AFW given a MSLB inside containment. The potential core damage sequences associated with a MSLB were modeled using the following event tree.
~FAULT~FULED IFE&Ed Sq G
SLAE BORATION IBEDIHPR State No.
OK OK CD 101 CD 102 CD 103 OK OK CD 104 CD 105 CD 106 CD 107 ATWS 108 In these sequences, high-pressure injection (HPI) [via two of three charging and one of three high-pressure safety injection (HPSI) pumps] provides inventory makeup to compensate for reactor coolant contraction during the cooldown caused by the loss of steam generator (SG) inventory. In the event the impacted SG is not isolated, emergency
B-156 boration is required to prevent an uncontrolled return to criticality. Failure of BPI or SG isolation and emergency boration is assumed to result in core damage (sequences 103, 106, and 107). Following initial recovery from the break, decay heat is assumed to be removed using AFW or feed and bleed. Failure of AFW and feed and bleed (including high pressure recirculation) is assumed to also result in core damage (sequences 101, 102, 104, and 105). For a large MSLB, main feedwater is assumed to be unavailable.
Failure to trip (sequence 108) results in an anticipated transient without scram (ATWS) sequence and is not developed further.
The following conditional branch probabilities were used in the analysis:
Branch System NonRecovery Operator Failure MSLB (inside containment) 5 x 10-4yr 1.0 Reactor trip (RT) 2.8 x10-4*
- 0. 12*
Faulted SG Isolated I X1-1.0 Emergency boration 0.12 BPI (1 of 3 HPSI and 1.3 x 10-3 1.0 2 of 3 charging pumps)
AFW 1.0 1.0 Feed & bleed 1lX 10-2 1.0 0.01 HPR 1.5 X 10-4 1.0*
- Values are consistent with probability values used in other ASP calculations.
"*Branch probability is dominanted by operator error.
Analysis Results The conditional probability of severe core damage estimated for the event over a 1-yr period is 1. 1 X 10-5. The dominant sequence (highlighted on the following event tree),
involves failure to initiate feed and bleed to provide decay heat removal given unavailability of AEW, following successful SG isolation and HPI.
B-157 HPI AW FEED &
HR End Seq.
AW BLEED HR State No.
OK OK
- Ln~
CD 101 CD 102 CD 103 OK OK CD 104 CD 105 CD 106 CD 107 ATWS 108 Dominant core damage sequence for LER 285/90-009