The Advanced Fuel Cycle Initiative (AFCI) program will rely on the use of accurate calculations and simulations of criticality and shielding for the separation process of the longlived isotopes that present a significant safety hazard in commercial spent fuel. To help design and verify the safety of the separation process, the neutronics code MCNPX will be used to model the distribution of neutron flux within the fuel blanket and to determine the neutron multiplication, keff. However, the cross section libraries and computational methods used by MCNPX at these neutron energies still have some uncertainty and will require validation.
Currently MCNPX relies on a hand created input deck, which can be time consuming to produce and prone to errors. One way to solve this problem is to create a graphical user interface (GUI) to help the user create the input deck. Also, to achieve accurate results in a short period of time there is a need to increase the efficiency of the parallel version of MCNPX. We propose to involve UNLV students and faculty in this endeavor to create a GUI, to increase the speed of MCNPX on parallel clusters of computers, and to continue application of MCNPX to solve practical AFCI problems.
The project proposed for year three of the project will consist of three parts:
• Optimization and validation of MCNPX on multiple platforms using Message Passing Interface (MPI).
• Create a graphical user interface (GUI) to help users generate input files for MCNPX.
• Continue MCNPX simulations in support of AFCI work.
Monte Carlo N-Particle eXtended (MCNPX); Neutron flux; Neutrons; Nuclear reactors; Particles (Nuclear physics); Radioactive wastes — Transmutation; Spallation (Nuclear physics); Spent reactor fuels; Transmutation (Chemistry)
Nuclear | Nuclear Engineering | Oil, Gas, and Energy
Radiation Transport Modeling using Parallel Computational Techniques.
Available at: http://digitalscholarship.unlv.edu/hrc_trp_reactor/3