Probabilistic safety assessment of support systems for CANDU stations
Journal Article
·
· Nuclear Technology; (United States)
OSTI ID:5559622
- Atomic Energy of Canada Ltd., CANDU Operations, Sheridan Park Research Community, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (CA)
This paper reports that Atomic Energy of Canada Limited (AECL) has performed probabilistic safety assessments (PSAs) of Canada deuterium uranium (CANDU) reactor support systems since 1975. AECL's experience in the use of PSAs on support system and the application of the PSA for the CANDU 3 are described. The PSA work reviews support system failures such as loss of service water and loss of instrument air as initiating events during full power and during plant shutdown conditions. The design changes resulting from this work, with respect to prevention of loss-of-coolant accidents and maintaining a long-term heat sink, are described.
- OSTI ID:
- 5559622
- Journal Information:
- Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 95:3; ISSN 0029-5450; ISSN NUTYB
- Country of Publication:
- United States
- Language:
- English
Similar Records
Probabilistic safety assessment of support systems on CANDU stations
MAAP-CANDU simulations of LOCA/LOECI accidents at Darlington NGS
Probabilistic Safety Assessments in Canada
Conference
·
Sat Dec 31 23:00:00 EST 1988
· Transactions of the American Nuclear Society; (USA)
·
OSTI ID:6527816
MAAP-CANDU simulations of LOCA/LOECI accidents at Darlington NGS
Journal Article
·
Mon Dec 30 23:00:00 EST 1996
· Transactions of the American Nuclear Society
·
OSTI ID:437038
Probabilistic Safety Assessments in Canada
Conference
·
Tue Dec 31 23:00:00 EST 1985
·
OSTI ID:5624455
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
CANDU TYPE REACTORS
DESIGN
DOUGLAS POINT ONTARIO REACTOR
ENGINEERED SAFETY SYSTEMS
EXPERIMENTAL REACTORS
HEAT SINKS
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
LOSS OF COOLANT
MAINTENANCE
NATURAL URANIUM REACTORS
PERFORMANCE TESTING
PHWR TYPE REACTORS
PRESSURE TUBE REACTORS
PROBABILISTIC ESTIMATION
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORE RESTRAINTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
RELIABILITY
RESEARCH AND TEST REACTORS
RESTRAINTS
SAFETY
SINKS
SUBCRITICAL ASSEMBLIES
TESTING
THERMAL REACTORS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
CANDU TYPE REACTORS
DESIGN
DOUGLAS POINT ONTARIO REACTOR
ENGINEERED SAFETY SYSTEMS
EXPERIMENTAL REACTORS
HEAT SINKS
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
LOSS OF COOLANT
MAINTENANCE
NATURAL URANIUM REACTORS
PERFORMANCE TESTING
PHWR TYPE REACTORS
PRESSURE TUBE REACTORS
PROBABILISTIC ESTIMATION
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORE RESTRAINTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
RELIABILITY
RESEARCH AND TEST REACTORS
RESTRAINTS
SAFETY
SINKS
SUBCRITICAL ASSEMBLIES
TESTING
THERMAL REACTORS