Probabilistic accident analysis: ATWS
Using operational experience for plant parameters, the probability that the peak reactor coolant system pressure will exceed allowable pressure has been evaluated for the Loss of Feedwater (LOFW) Anticipated Transient Without Scram (ATWS) for the Standard 3800 MWt, 205 fuel assembly, Babcock and Wilcox reactor. The probabilistic approach to ATWS is reviewed including current estimates of anticipated transients' frequency, probability of scram system failure, and probability of excessive peak pressure. The result of the present study is a best estimate probability of 0.18 that an allowable pressure of 3750 psia would be exceeded if an ATWS were to occur and 0.6 that 3200 psia would be exceeded. Considering all uncertainties and using estimates of transient frequency and scram system failure, it is found that the frequency of significant ATWS events is smaller than 1.1 x 10/sup -6/ per reactor per year. Thus it is concluded that the ATWS safety goal of approximately 10/sup -6/ per reactor per year has been met with the current design. A proposed redundant Backup Scram System (BUSS) to the present generation reactor protection system would reduce the estimate to approximately 7.8 x 10/sup -7/ per reactor per year. Therefore, compliance with th ATWS safety goal is demonstrated.
- Research Organization:
- Babcock and Wilcox Co., Lynchburg, VA (USA)
- OSTI ID:
- 6266881
- Report Number(s):
- EPRI-NP-1090
- Country of Publication:
- United States
- Language:
- English
Similar Records
Analysis of a SBLOCA initiated by an ATWS event
Containment venting as a mitigation technique for BWR Mark I plant ATWS
Correlation for predicting reactor power during a BWR ATWS
Conference
·
Mon Dec 31 23:00:00 EST 1984
·
OSTI ID:5761556
Containment venting as a mitigation technique for BWR Mark I plant ATWS
Conference
·
Tue Dec 31 23:00:00 EST 1985
·
OSTI ID:7047577
Correlation for predicting reactor power during a BWR ATWS
Conference
·
Tue Dec 31 23:00:00 EST 1985
· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:6992950
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ATWS
BW STANDARD REACTOR
HIGH PRESSURE
MATHEMATICS
PROBABILITY
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTORS
STATISTICS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ATWS
BW STANDARD REACTOR
HIGH PRESSURE
MATHEMATICS
PROBABILITY
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTORS
STATISTICS
WATER COOLED REACTORS
WATER MODERATED REACTORS