The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. This paper discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models’ theories are included as well.

The goal of this work was to improve accuracy and efficiency of two Monte-Carlo transport codes (MCNP and DIANE) with an emphasis on γ+electron physics. The approach involved intercode comparisons + measurements for gamma/e- energy deposition in a cylinder with a photon source and different materials (C, Pb) and the bombardment of 15-MeV electrons on thick targets (Al, Be, Pb). Comparisons of the codes DIANE and MCNP6 showed good agreement (differences < 3%) for gamma-electron energy deposition in a 2D cylinder, except for the first 0.1 μm of lead (difference < 10%). Comparisons with measurements showed generally good agreement, often better than 10%; best-performing codes/options are problem-dependent; and single-event discrepancies are in active use in reviewing electron elastic scattering.

The goal of this work was to improve accuracy and efficiency of two Monte-Carlo transport codes (MCNP and DIANE) with an emphasis on γ+electron physics. The approach involved intercode comparisons + measurements for gamma/e- energy deposition in a cylinder with a photon source and different materials (C, Pb) and the bombardment of 15-MeV electrons on thick targets (Al, Be, Pb). Comparisons of the codes DIANE and MCNP6 showed good agreement (differences < 3%) for gamma-electron energy deposition in a 2D cylinder, except for the first 0.1 μm of lead (difference < 10%). Comparisons with measurements showed generally good agreement, often better than 10%; best-performing codes/options are problem-dependent; and single-event discrepancies are in active use in reviewing electron elastic scattering.

MCNP6 is simply and accurately 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.

For the past several years, a major effort has been undertaken at Los Alamos National Laboratory (LANL) to develop the transport code MCNP6, the latest LANL Monte-Carlo transport code representing a merger and improvement of MCNP5 and MCNPX. We emphasize a description of the latest developments of MCNP6 at higher energies to improve its reliability in calculating rare-isotope production, high-energy cumulative particle production, and a gamut of reactions important for space-radiation shielding, cosmic-ray propagation, and accelerator applications. We present several examples of validation and verification of MCNP6 compared to a wide variety of intermediate-and high-energy experimental data on reactions induced by photons, mesons, nucleons, and nuclei at energies from tens of MeV to about 1 TeV/nucleon, and compare to results from other modern simulation tools.

MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of those two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Decision Applications Division, Radiation Transport and Applications Team (D-5), respectively, have combined their code development efforts to produce the next evolution of MCNP. While maintenance and 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 16 new features not previously found in either code. These new features include the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to transport electrons down to 10.0 eV, to model complete atomic relaxation emissions, and to generate or read mesh geometries for use with the LANL discrete ordinates code Partisn. The first release of MCNP6, MCNP6 Beta 2, is now available through the Radiation Safety Information Computational Center, and the first production release is expected in calendar year 2012. High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, the regression test suite, its code development process, and the underlying high-quality nuclear and atomic databases.

Researchers at Los Alamos National Laboratory are considering the possibility of using the Low Energy Demonstration Accelerator, constructed at Los Alamos Neutron Science Center for the Accelerator Production of Tritium Project, as a neutron source. Evaluated nuclear data are needed for the p+ 7 Li reaction, to predict neutron production from thin and thick lithium targets. In this paper we describe evaluation methods that make use of experimental data, and nuclear model calculations , to develop an ENDF-formatted data library for incident protons with energies up to 150 MeV. The important 7 Li(p,n 0) and 7 Li(p,n 1) reactions are evaluated from the experimental data, with their angular distributions represented using Lengendre polynomial expansions. The decay of the remaining reaction flux is estimated from GNASH nuclear model calculations. This leads to the emission of lower-energy neutrons and other charged particles and gamma-rays from preequi-librium and compound nucleus decay processes. The evaluated ENDF-data are described in detail, and illustrated in numerous figures. We also illustrate the use of these data in a representative application by a radiation transport simulation with the code MCNPX.

A new algorithm for energy-loss straggling in MCNP is demons trated. An approximate but accurate energy-loss moment-preserving differential cross sectio n is used in conjunction with single event Monte Carlo simulation through each condensed history step to show that highly accurate energy spectra, leakage currents, and dose profiles can be obtained . This new approach provides a viable and even preferred alternative to the Landau straggling model c urrently employed in MCNP.

We review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, we emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. We also briefly touch on the current situation in regard to photon, electron, and proton transport tables.

A modified version of MCNP5 has been developed to treat continuous-energy proton transport. This work is summarised in companion papers by Hughes et al. and Bull et al. (in these proceedings). An intrinsic part of this development effort has involved testing, verification and validation of a capability for simulating proton radiographs. This paper presents the results of calculations simulating various different test objects and the effects of alternative physics models. The significant physics processes include elastic scattering, multiple coulomb scattering, collisional energy-loss and straggling, magnetic fields and attenuation owing to nuclear interactions. Comparisons with experimental data are presented.

With the introduction of continuous-energy heavy charged particle transport in MCNP5, the need for tracking charged particles in a magnetic field becomes increasingly important. Two methods for including magnetic field effects on charged particles are included in the proton transport version of the code. The first technique utilises transfer maps produced by the beam dynamics simulation and analysis code COSY INFINITY. This method is fast and accurate; however, its use is limited to void cells only and to ensembles of particles with a fairly small energy spread. The second technique, particle ray tracing, is based on an algorithm adopted from the MARS transport code. This method can be applied to both void and material cells and is valid over a very large range of particle energies. Results from tracking particles in a quadrupole 'identity lens' using the two techniques are compared.

In July of 2004, an updated version of MCNP5{trademark} (MCNP5-RSICC-1.30) was released to the Radiation Shielding Information Computational Center. This updated version has three new features, thirteen bug fixes and several minor coding improvements. The new features are: support for 8 byte integers, specialized tally treatment of large lattices, and mesh tally enhancements. Of the thirteen bug fixes, only four resulted in incorrect answers in specific circumstances. In addition to the standard RSICC distribution of the MCNP5 source, executables and patches, the patch file (only) is available on the MCNP website: http://www-xdiv.lanl.gov/x5/MCNP/theresources.html. The three new MCNP5 features are discussed. Several new improvements have also been made to the manual and development environment. All of the features, bug fixes, coding improvement issues and related documentation are now maintained in Sourceforge. Fortran and C source code and regression test problems are now under version control with CVS.

We present the current forms of the algorithms used to model collisional energy-loss straggling for electrons and for heavy charged particles in MCNP5 and MCNPX. We also describe a recent improvement in the logic for applying the straggling models for heavy charged particles, and show illustrative results for the improved logic.

We review recent evaluations of neutron and proton reaction cross sections up to 150 MeV in the LA150 Library, for use in computer code simulations of accelerator-driven systems. An overview is provided of the nuclear theory together with measured cross section data used in the evaluations. The possible use of bismuth activation foils for high-energy neutron spectrometry is also discussed. We describe recent developments to the MCNPX radiation transport code, which merges MCNP and LAHET in one code and uses the LA150 evaluated data. A number of benchmark comparisons against integral experiments are described, for thick-target neutron production, neutron transmission through macroscopic slabs, and neutron kerma coefficients. The benchmarks help validate the transport code and the evaluated data for use in ADS simulations of neutron production in a spallation target (n/p), radiation shielding, heating, and damage. A brief summary is also given of future data needs for subcritical transmuters and spallation targets, in accelerator transmutation of waste technologies.

The Monte Carlo particle transport code MCNPX and its associated data have been the focus of a major development effort at Los Alamos for several years [1]. The system has reached a mature state, and has become a significant tool for many intermediate- and high-energy particle transport applications. A recent version has been released to the Radiation Safety Information Computational Center (RSICC). A recent report [2] provides an overview of the code and an extensive set of references for the component physics modules used in the code.

We describe evaluation methods that make use of experimental data, and nuclear model calculations, to develop an ENDF-formatted data library for the reaction p + Li7 for incident protons with energies up to 150 MeV. The important 7-Li(p,n_0) and 7-Li(p,n_1) reactions are evaluated from the experimental data, with their angular distributions represented using Lengendre polynomial expansions. The decay of the remaining reaction flux is estimated from GNASH nuclear model calculations. The evaluated ENDF-data are described in detail, and illustrated in numerous figures. We also illustrate the use of these data in a representative application by a radiation transport simulation with the code MCNPX.

We describe recent evaluations of neutron and proton interaction cross sections up to 150 MeV in the LA150 Data Library, for use in MCNPX computer code simulations of accelerator-driven systems (ADS). An overview is provided of the nuclear theory used. A number of benchmark comparisons against experiments are described, for thin and thick-target neutron production, neutron transmission through macroscopic slabs, and neutron heating and damage. Future data needs for waste transmutation are also briefly described.