The MCNPX radiation transport code

MCNPX is a general-purpose Monte Carlo N-Particle eXtended radiation transport code. It is a complete super-set of MCNP 4C that includes three-dimensional geometry, continuous-energy physics with extensive data libraries, and a wide variety of source models, tally options, variance reduction, and graphics capabilities. MCNPX is extended to nearly all particles (34 different particle types), nearly all energies (up to 1 TeV), and nearly all applications, with no penalty in computational speed. MCNPX is also a complete super-set of CINDER-90 and borrows heavily from LAHET, ITS, CEM2K, INCL4, and old versions of FLUKA, Bertini, HETC, and Isabel. Additional MCNPX capabilities include fission multiplicity modeling enhancements, positron sources, spontaneous fission neutron sources, multiple source particles, neutron coincidence detector modeling, residual nuclei tallies, and light-ion recoil. MCNPX is written in Fortran-90 and supports both the PVM and MPI parallel computing standards. New results are presented that highlight the following three recent MCNPX capabilities. First, the MCNPX mesh tally plot allows visualization of fluxes, energy deposition, sources, populations, and many other edits on a color mesh superimposed over the problem geometry. Second, variance reduction techniques may be done with pulse-height tallies (PHTs), thus increasing the efficiency of these calculations by orders of magnitude. Third, PHTs have a unique anticoincidence capability. (authors)