Analysis of the coupling of direct Monte Carlo reactor kinetics to termohydraulics solvers

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PhD típus: 
Fizikai Tudományok Doktori Iskola
Légrády Dávid
Email cím:
egy. docens
Tudományos fokozat: 

Dynamics of nuclear reactors is a two-way interaction between the neutron population creating heat release by nuclear fission and the heat transfer process that change nuclear interaction physics. Altough accurate simulation is required for nuclear plant safety analysis, daily routine still resort to modelling in a few neutron energy groups and one dimensional heat transfer. Given the immensly complex nature of neutron fission chains and heat transfer involving fluids, leaping for higher precision is a matter of computer resources. The GUARDYAN dedicated Monte Carlo reactor kinetics code is under development at the Institute of Nuclear Technique with features of virtually approximation-free neutronics, utilization of Graphical Processing Units and a detailed verification involving both comparisons to other Monte Carlo codes and to measurements.

The convergence of a Monte Carlo simulation is a stochastic issue: the lower the variance the more converged the solution is. Heat transfer calculations usually involve  deterministic numerical differential equiation solvers to the continuous differential equations with their convergence being a matter of numerical stability. In multiphysics calculations coupling thermal hydraulics or solid mechanics deterministic solutions to stochastic simulations presents differential equations driven by stochastic quantities and stability of convergence shares features of both realms. 

For a succesfull scientific work in this field the candidate should cover

  • the general theoretical convergence analysis of numerical differential equation solver driven by stochastic quantities
  • numerical convergence analysis in simplified neutron transport-heat transfer systems
  • development of a methodology of error prediction of converged states
  • testing the developed methodology on realistic systems usng the GUARDYAN code extended with the coupling features
  • development of a practical full-scale power reactor application scheme including the utilisation of deterministic neutron transport solvers, precalculation of converged states and advanced variance reduction
  • estimation of computer resources needed of the coupled GUARDYAN-thermohydraulic code for power reactor transients



Monte Carlo methods, good command of the English language, affinity to programing, strong mathematical background in probability theory and numerical solutions of differential equations, experience in deterministic neutron transport solutions

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