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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)
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Full-Text Articles in Nuclear Engineering
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth
Reactor Campaign (TRP)
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. …