Best estimate models of physical processes, best estimate plant states, and most probable system configurations and failure events provide the most realistic representation of plant behaviour and consequences during accidents. Deviations from these best estimate conditions can and will occur, resulting in uncertainty in the outcome of a best estimate analysis. In order to quantify the variability and uncertainty in the outcome of an accident, it is necessary to identify and characterize the components contributing to uncertainty and evaluate their impact on safety consequences. A primary objective is to define, through the use of an integrated probabilistic approach, the ranges of key plant parameters that assure that safety limits are met at a prescribed level of probability and confidence.
Research in this area investigates approaches to propagation of time-dependent variability and uncertainty in plant response during accidents using dynamic sensitivity analysis and quasilinearization methods. Of particular interest is the representation of nonlinear bifurcation behaviour within a framework of quasilinearized sensitivity analysis since this allows time-dependent solutions for multiple-parameter variations to be generated from a limited set of detailed computer simulations obtained from best estimate computer codes.
Boiling heat transfer from cylindrical nuclear fuel elements and fuel channel calandria tubes is an important process in CANDU reactor accidents. Specifically, changes in heat transfer regimes from nucleate boiling to film boiling influence both the operational safety margins and integrity of components during accidents.
Research in this area investigates fundamental aspects of limiting heat transfer processes in two-phase systems. The major focus is on investigating the fundamental processes of vapour bubble nucleation and transport from the walls of horizontal cylindrical heated tubes. This will be used to improve understanding of heat transfer limitations during transitions in boiling regimes and vapour formation during rapid depressurization of hot fluids.
Demonstrating the integrity of engineered barriers to fission product release is one primary objective of nuclear safety analysis. Where heat transfer to fluids occurs across an engineered barrier to fission product release – such as at the metal cladding (sheath) of a nuclear fuel element or at the pressure tube/calandria tubes of a fuel channel – a transition to a more limiting heat transfer regime (e.g. transition from nucleate boiling to film boiling) invariably poses a challenge to the integrity of the barrier. Additionally, transition from the film boiling regime to the nucleate boiling regime through the quenching process, in which a rapid increase in heat transfer and concomitant rapid reduction in the temperature of the wall occurs, is of importance in limiting the time during which integrity of a barrier is subject to challenge.
Research in this area is focused on mechanistic modelling of near-field processes including vapour bubble nucleation and transport processes and associated local heat transfer processes, such as conduction in metal walls and turbulent boundary layer convection, as they influence development and spread of vapour film patches (drypatches) on horizontal heated cylinders. Similarly, fundamental theoretical mechanistic modelling of the complementary transition from stable film boiling back to nucleate boiling (often referred to as the “quench” process) are being investigated.
Progression of events in LWR, CANDU and advanced reactors leading to severe fuel damage. Modelling severe accident phenomena: thermal-mechanical behaviour ( fuel damage; core degradation, core melt and relocation, heat transfer in molten corium pools, vapour explosions, in-vessel retention of corium); thermal-chemical behaviour (high-temperature oxidation of core structural materials, eutectic interactions, hydrogen generation and transport, fission product release and transport); ex-vessel behaviour (molten corium relocation into containment, molten core-concrete interactions, debris cooling, high-pressure melt discharge); atmospheric dispersion and off-site deposition of fission products.
Severe Accident Management Guidelines (SAMG), Probabilistic Risk Assessment (PRA) and complementary Risk Analysis methods to handle very low frequency and high consequence events.
Did you know…
Dr. Luxat is the NSERC/UNENE Industrial Research Chair in Nuclear Safety Analysis.
Ph.D., Electrical Engineering, University of Windsor, Ontario (1972)
M.Sc., Electrical Engineering, University of Cape Town, South Africa (1969)
B.Sc., Electrical Engineering, University of Cape Town, South Africa (1967)