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Introduction
Published in Alireza Haghighat, Monte Carlo Methods for Particle Transport, 2020
The above brief history indicates that early development of Monte Carlo particle transport techniques were mainly conducted by the scientists at LANL. As a result, LANL has been the main source of general-purpose Monte Carlo codes, starting with MCS (Monte Carlo Simulation) in 1963 and followed by MCN (Monte Carlo Neutron) in 1965, MCNG (Monte Carlo coupled Neutron and Gamma) in 1973, and MCNP (Monte Carlo Neutron Photon) in 1977. MCNP has been under continuous development and the latest version, MCNP6, was released in 2013 [80]. The progress made over the past 50 years demonstrates the sustained effort at LANL on development, improvement, and maintenance of Monte Carlo particle transport codes. There has also been simultaneous development of and improvement in nuclear physics parameters, i.e., cross sections, in the form of the cross-section library referred to as the Evaluated Nuclear Data File (ENDF). Currently, ENDF/B-VIII, the 8th version, is in use.
FERMI: Fusion Energy Reactor Models Integrator
Published in Fusion Science and Technology, 2023
V. Badalassi, A. Sircar, J. M. Solberg, J. W. Bae, K. Borowiec, P. Huang, S. Smolentsev, E. Peterson
The MCNP 6.2 unstructured mesh transport calculation outputs cell-averaged tallies, superimposed Cartesian and cylindrical mesh tallies, and tetrahedral mesh tallies. Cell tallies are used for metrics that do not require a fine spatial resolution, such as TBR. For energy deposition data, which require finer spatial resolution, tetrahedral mesh tallies are used for coupling with OpenFOAM. A coupled neutron-photon transport case is run in which the neutron source is modeled using an MCNP source subroutine (described in Sec. II.A). The Evaluated Nuclear Data File (ENDF)-B/VII.1 data libraries44 are used for the calculation. The tetrahedral energy deposition mesh tallies (“F6 tallies”) are in units of MeV/g. The value for each mesh cell is then multiplied by the density of the material, which is then multiplied by the number of neutrons emitted per second for the ARC-class tokamak power level of 525 MW. The resulting value is energy deposition per volume per second, or W/cm3. A 2D cutaway of the derived power deposition is shown in Fig. 12. This processed result is collected in a machine-readable format and sent to OpenFOAM as input. The energy deposition per volume per second is close to the values reported in other ARC-class tokamak simulation reports.45
DeCE: the ENDF-6 data interface and nuclear data evaluation assist code
Published in Journal of Nuclear Science and Technology, 2019
The ENDFIO library offers an easy access to the evaluated nuclear data file. The function ENDFRead copies data in the file specified by to the object. The ENDF object data are printed by a function ENDFWrite in the ENDF-6 appropriate format. Since each section has a different data structure depending on , one has to consult the ENDF-6 manual [5] for extracting requisite information if needed. Otherwise, these functions, ENDFRead and ENDFWrite, should work properly for all the sections.