Talk by Dr. David Millman (Computer Science Dept, MSU)

2/7/2019  Wilson Hall 1-144  3:10-4:00pm


Monte Carlo (MC) codes for solving the Neutron Transport Equation often represent a domain using a Constructive Solid Geometry (CSG) representation.  In particular, the geometry of the domain is many "components" defined by set operations on half-spaces defined by quadrics.  While a CSG representation is excellent for an engineer modeling a nuclear reactor, it can add substantial complexity into a solver.  For example, many MC solvers spend as much as 1/3 of their total computation performing point location and distance to surface calculations or use an entire super computer to compute the volume of every component.  In this talk, we will introduce some of the geometric problems that occur when solving the Transport Equation with MC methods and how we can use techniques from Computational Geometry to address these problems.