Research

Advanced modeling and simulation of neutronic and thermal-hydraulic behavior

PhD student Lulu Liu explains how the thermal neutron flux distribution within the core of a Pressurized Water Reactor is simulated with MIT’s code OpenMOC; high fidelity and detailed resolution (down to individual fuel pellets) come from solving the neutron transport equation with a highly efficient advanced method of the characteristics.Credit: Susan Young

Advanced modeling and simulation of neutronic and thermal-hydraulic behavior promises to provide designers of nuclear plants with high-fidelity quantification of the safety margins, as well as the best ways to improve fuel efficiency and operations. Taking advantage of recent advances in computer science, MIT researchers have developed new methods that dramatically accelerate the solution of the equations describing neutron transport—characterization of the movements made by neutrons in time and space from the moment the first atom of nuclear fuel undergoes fission—as well as fluid flow and heat transfer of the reactor coolant in the core. This work aids in the analysis and optimization of nuclear plants, including the ones already in operation, at a very fundamental level.

Thermal neutron flux distribution within the core of a Pressurized Water Reactor simulated with MIT’s code OpenMOC; high fidelity and detailed resolution (down to individual fuel pellets) come from solving the neutron transport equation with a highly efficient advanced method of the characteristics.
Novel hybrid models in Computational Fluid Dynamics (CFD) introduce radically improved resolution of turbulent flow within the reactor core.

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Research Team