Features of MCNP6

Abstract

a b s t r a c t MCNP6 can be described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory's X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and Nonproliferation Division, Radiation Transport Modeling Team (NEN-5) respectively, have combined their code development efforts to produce the next evolution of MCNP. While maintenance and major bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. In fact, the initial release of MCNP6 contains numerous new features not previously found in either code. These new features are summarized in this document. Packaged with MCNP6 is also the new production release of the ENDF/B-VII.1 nuclear data files usable by MCNP. The high quality of the overall merged code, usefulness of these new features, along with the desire in the user community to start using the merged code, have led us to make the first MCNP6 production release: MCNP6 version 1. High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, our automated nightly software debugger tests, the underlying high quality nuclear and atomic databases, and significant testing by many beta testers.

Ionizing Radiation and Complex DNA Damage: Quantifying the Radiobiological Damage Using Monte Carlo Simulations

“…Therefore, in recent years, there have been efforts to implement track-structure models into some general-purpose MC codes to enable simulations at both the macro and microscopic scale. The most notable examples are the Monte Carlo N-Particle version 6 (MCNP6) [71], PENELOPE (modification to PENELOPE/penEasy, resulting in the LionTrack code) [72][73][74], and the Geant4-DNA package of Geant4 [75][76][77]. Apart from MCNP6, which uses a simple interpolation of high-energy atomic models down to the eV energy range, thus neglecting molecular aggregation and condensed-phase effects, the Geant4-DNA and modified PENELOPE/penEasy extensions are based on elaborate physics models specifically developed for liquid water.…”

Maintaining a critical spectra within Monteburns for a gas-cooled reactor array by way of control rod manipulation

“…In the MCNP model, these temperatures were accounted for by the combination of using TMP cards evaluated to the exact temperature and using ACE (a compact ENDF) formatted cross sections at the nearest available temperature. The TMP cards in MCNP6 were used to adjust the thermal cell temperature necessary for the free-gas thermal treatment of low-energy neutron scatter [18,32]. The ACE formatted cross sections used for each of the regions were taken from an ENDF/B VII.0 temperature dependent library processed in-house at 50 K increments using the NJOY processing code [33].…”

“…In MCNP, below 6 eV, if available for a particular species, and selected by the user, S(α,β) cross sections for the entire material replace the individual isotope cross sectioned in order to better capture the effect of the bound species on the interaction kinematics. Due to the extensive amount of graphite, this particular reactor was extremely thermal; and therefore we used the graphite S(α,β) kernel evaluated at 500 K (the in-house 50 K cross section simply uses the S(α,β) available in the official MCNP release) [18,32,33]. Each criticality calculation was run with 10,000 histories per cycle, 30 settle cycles and 270 active cycles.…”

Calculation of Control Rods Reactivity Worth of RSG-GAS First Core Using Deterministic and Monte Carlo Methods

“…The first core was chosen for two reasons, namely first, the completeness of the experimental data, and second, the fuel elements, absorber and beryllium reflectors were still fresh and therefore the uncertainties raised from the fuel burnup and absorber depletion calculations, and from the calculation of lithium poisoning in the beryllium reflectors can be eliminated. The continuous-energy Monte Carlo code, MCNP6 [4], was used for the entire calculations. In the present work, as the nuclear data libraries for the code, the Evaluated Nuclear Data Library, ENDFVII.1 [5] is used to check their accuracy by comparing to the experimental data.…”