The paper overviews the analytical studies performed at Politehnica University of Bucharest on the analysis of late phase severe accident phenomena in a Canada Deuterium Uranium (CANDU) plant. The calculations start from a dry debris bed at the bottom of calandria vessel. Both SCDAPSIM/RELAP code and ansys-fluent computational fluid dynamics (CFD) code are used. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty on calandria vessel failure: metallic fraction of zirconium inside the debris, containment pressure, timing of water depletion inside calandria vessel, steam circulation in calandria vessel above debris bed, debris temperature at moment of water depletion inside calandria vessel, calandria vault nodalization, and the gap heat transfer coefficient.
Skip Nav Destination
Article navigation
April 2017
Research-Article
Analysis of Late Phase Severe Accident Phenomena in CANDU Plant
D. Dupleac
D. Dupleac
Power Plant Engineering Faculty,
Politehnica University,
313 Splaiul Independentei, Sector 6,
Bucharest 060042, Romania
e-mail: danieldu@cne.pub.ro
Politehnica University,
313 Splaiul Independentei, Sector 6,
Bucharest 060042, Romania
e-mail: danieldu@cne.pub.ro
Search for other works by this author on:
D. Dupleac
Power Plant Engineering Faculty,
Politehnica University,
313 Splaiul Independentei, Sector 6,
Bucharest 060042, Romania
e-mail: danieldu@cne.pub.ro
Politehnica University,
313 Splaiul Independentei, Sector 6,
Bucharest 060042, Romania
e-mail: danieldu@cne.pub.ro
Manuscript received July 30, 2016; final manuscript received December 5, 2016; published online March 1, 2017. Assoc. Editor: Arun Nayak.
ASME J of Nuclear Rad Sci. Apr 2017, 3(2): 020904 (8 pages)
Published Online: March 1, 2017
Article history
Received:
July 30, 2016
Revised:
December 5, 2016
Citation
Dupleac, D. (March 1, 2017). "Analysis of Late Phase Severe Accident Phenomena in CANDU Plant." ASME. ASME J of Nuclear Rad Sci. April 2017; 3(2): 020904. https://doi.org/10.1115/1.4035416
Download citation file:
Get Email Alerts
Advancement Towards the Experimental Measurement of the Lbe Thermal Properties Using DSC Technique
ASME J of Nuclear Rad Sci
Study on the Step by Step Process and Performance of Laser Welding for the Spent Fuel Pool Floor
ASME J of Nuclear Rad Sci (October 2023)
Linear and Nonlinear Stability Analysis of Molten Salt Natural Circulation Loop
ASME J of Nuclear Rad Sci (October 2023)
Experimental Analysis of Bubble Behavior and Critical Heat Flux During Pool Boiling on Vertical Circular Tubes
ASME J of Nuclear Rad Sci (October 2023)
Related Articles
Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications
ASME J of Nuclear Rad Sci (April,2017)
Experimental Study on Melt Coolability Capability of Calandria Vault Water During Severe Accident in Indian PHWRs for Prolonged Duration
ASME J of Nuclear Rad Sci (July,2018)
Analyses of Feedwater Trip With SBO Sequence of VVER1000 Reactor
ASME J of Nuclear Rad Sci (October,2016)
Dynamic Behavior of Transportation Pressure Relief Valves Under Simulated Fire Impingement Conditions
J. Pressure Vessel Technol (February,2000)
Related Proceedings Papers
Related Chapters
PSA Level 2 — NPP Ringhals 2 (PSAM-0156)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Summary of Water Hammer-Induced Pipe Failures
Fluid Mechanics, Water Hammer, Dynamic Stresses, and Piping Design
Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)