Advanced materials development, manufacturing, and modeling capabilities for innovative reactor designs support nuclear security and mission-focused science through enhanced technology for safer and more efficient and secure production of nuclear energy. The research in this project has established: 1) a state-of-the-art neutron-based hydrogen mapping and cross-section measurement capability as well as detailed crystallographic characterization of hydrogen atoms at LANSCE, and 2) a multi-physics framework for simulating behavior of moderator materials and other material performance in advanced nuclear reactors. Through the course of this project, we successfully developed and demonstrated measurement techniques for hydrogen distribution and atomistic-scale behavior of hydrogen atoms using pulsed neutron techniques. In parallel, advanced multi-physics simulation tools to predict the behavior of hydrogen atoms, e.g. in a moderator for a nuclear reactor, through materials performance, neutron transport, and thermal mechanical behavior were enhanced. Multi-discipline areas across the laboratory were involved in the project as the integration of improved experimental capabilities with enhanced modeling and simulation through MST, NEN, SIGMA, and XCP division subject matter experts.

Monte Carlo N-Particle or MCNP® is a general-purpose Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. This MCNP Version 6.2 follows the MCNP6.1.1 beta version and has been released in order to provide the radiation transport community with the latest feature developments and bug fixes for MCNP. Since the last release of MCNP major work has been conducted to improve the code base, add features, and provide tools to facilitate ease of use of MCNP version 6.2 as well as the analysis of results. These release notes serve as a general guide for the new/improved physics, source, data, tallies, unstructured mesh, code enhancements and tools. For more detailed information on each of the topics, please refer to the appropriate references or the user manual which can be found at http://mcnp.lanl.gov. This release of MCNP version 6.2 contains 39 new features in addition to 172 bug fixes and code enhancements. There are still some 33 known issues the user should familiarize themselves with (see Appendix).

Cosmic neutron and photon fluxes are known to scale exponentially with elevation. Consequently, cosmic neutron elevation scaling was implemented for use with the background-source option shortly after its introduction into MCNP6, whereby the neutron flux weight factor was adjusted by the elevation scaling factor when the user-specified elevation differed from the selected background. dat grid-point elevation. At the same time, an elevation scaling factor was suggested for the cosmic photon flux, however, cosmic photon elevation scaling is complicated by the fact that the photon background consists of two components: cosmic and terrestrial. Previous versions of the background. dat file did not provide any way to separate these components. With Rel. 4 of this file in 2015, two new columns were added that provide the energy grid and differential cosmic photon flux separately from the total photon flux. Here we show that the cosmic photon flux component can now be scaled independently and combined with the terrestrial component to form the total photon flux at a user-specified elevation in MCNP6.Cosmic background fluxes also scale with the solar cycle due to solar modulation. This modulation has been shown to be nearly sinusoidal over time, with an inverse effect - increased modulation leads to a decrease in cosmic fluxes. This effect was initially included with the cosmic source option in MCNP6 and has now been extended for use with the background source option when: (1) the date is specified in the background. dat file, and (2) when the user specifies a date on the source definition card. A description of the cosmic-neutron/photon date scaling feature will be presented along with scaling results for past and future date extrapolations. (C) 2017 The Authors. Published by Elsevier 2017 B.V.

Secondary electrons in the form of delta-rays, also referred to as knock-on electrons, have been a feature of MCNP for electron and positron transport for over 20 years. While MCNP6 now includes transport for a suite of heavy-ions and charged particles from its integration with MCNPX, the production of delta-rays was still limited to electron and positron transport. In the newest release of MCNP6, version 6.2.0, delta-ray production has now been extended for all energetic charged particles. The basis of this production is the analytical formulation from Rossi and ICRU Report 37. This paper discusses the MCNP6 heavy charged-particle implementation and provides production results for several benchmark/test problems.

The mission of the Domestic Nuclear Detection Office is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism.A delayed particle capability was initially added to MCNPX 2.6. A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. To meet the objectives of detection scenario modeling, the capability was designed to sample a particular time for emitting particular multiplicity of a particular energy. Because the sampling process of selecting both time and energy is interdependent, to linearize the time and emission sampling, atom densities are computed at several discrete time steps, and the time-integrated production is computed by multiplying the atom density by the decay constant and time step size to produce a cumulative distribution function for sampling the emission time, energy, and multiplicity. The delayed particle capability was initially given a time-bin structure to help reasonably reproduce, from a qualitative sense, a fission benchmark by Beddingfield, which examined the delayed gamma emission. This original benchmark was only qualitative and did not contain the magnitudes of the actual measured data but did contain relative graphical representation of the spectra. A better benchmark with measured data was later provided by Hunt, Mozin, Reedy, Selpel, and Tobin at the Idaho Accelerator Center; however, because of the complexity of the benchmark setup, sizable systematic errors were expected in the modeling, and initial results compared to MCNPX 2.7.0 showed errors outside of statistical fluctuation.Presented here is a more simplified approach to benchmarking, utilizing closed form analytic solutions to the granddaughter equations for particular sets of decay systems. We examine five different decay chains ( two-stage decay to stable) and show the predictability of the MCNP6 delayed gamma feature. Results do show that while the default delayed gamma calculations available in the MCNP6 1.0 release can give accurate results for some isotopes ( e.g., Ba-137), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% (Mg-28, Al-28, K-42, Ca-47, Sc-47, Co-60). With the MCNP6 1.1 Beta release, the tenth entry on the DBCN card allows improved calculation within <5% as compared to the closed form analytic solutions for immediate parent emissions and transient equilibrium systems. While the tenth entry on the DBCN card for MCNP6 1.1 gives much better results for transient equilibrium systems and parent emissions in general, it does little to improve daughter emissions of secular equilibrium systems. Hypotheses were presented as to why daughter emissions of secular equilibrium systems might be mispredicted in some cases and not in others.

My internship program was the Domestic Nuclear Detection Office Summer Internship Program. I worked at Los Alamos National Laboratory with Trevor A. Wilcox and Gregg W. McKinney in the NEN-5 group. My project title was “MCNP Physical Model Interoperability & Validation”. The goal of my project was to write a program to predict the solar modulation parameter for dates in the future and then implement it into MCNP6. This update to MCNP6 can be used to calculate the background more precisely, which is an important factor in being able to detect Special Nuclear Material. We will share our work in a published American Nuclear Society (ANS) paper, an ANS presentation, and a LANL student poster session. Through this project, I gained skills in programming, computing, and using MCNP. I also gained experience that will help me decide on a career or perhaps obtain employment in the future.

Essentially a set of slides, the presentation begins with the MCNP6 cosmic-source option, then continues with the MCNP6 transport model (atmospheric, terrestrial) and elevation scaling. It concludes with a few slides on results, conclusions, and suggestions for future work.

The particle transport code MCNP has been used to produce a background radiation data file on a worldwide grid that can easily be sampled as a source in the code. Location-dependent cosmic showers were modeled by Monte Carlo methods to produce the resulting neutron and photon background flux at 2054 locations around Earth. An improved galactic-cosmic-ray feature was used to model the source term as well as data from multiple sources to model the transport environment through atmosphere, soil, and seawater. A new elevation scaling feature was also added to the code to increase the accuracy of the cosmic neutron background for user locations with off-grid elevations. Benchmarking has shown the neutron integral flux values to be within experimental error.

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.

Transport codes like MCNPX (Monte Carlo N-Particle eXtended) which are based primarily on experimentally derived data libraries at energies below approximately 100 MeV are reliable and predictive in transporting and producing neutrons and gammas. However, they are deficient in the microscopic sense because of the inability to produce correlated secondary particles. Neutrons will produce the average number of gammas at each collision regardless of the specific neutron reaction. Consequently, capture gammas may be produced by elastic collisions and inelastic gammas may be produced by capture. To remedy this deficiency the Cascading Gamma-ray and Multiplicity (CGM) V3.4 code has been integrated into MCNPX V2.7.0. The CGM code is based on Hauser-Feshbach and optical models. This paper details an overview of the CGM code integration into MCNPX followed by a test case highlighting the gamma spectrum and multiplicities from the CGM/MCNPX code. The test case provides gamma spectrum and multiplicities for the (n,gamma) reaction on Fe-56 using thermal neutrons. A second test case highlights gamma correlation for the (n,gamma) reaction on Fe-56 using thermal neutrons using three coincidence detectors. Published by Elsevier Ltd.

With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

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.

The interaction of radiation with matter can cause activation or fission reactions producing unstable residuals that decay with the emission of delayed-neutron and/or delayed-gamma radiation. This delayed radiation can be exploited for a variety of purposes, including homeland security, health physics, instrumentation and equipment design, and nuclear forensics. Here we report on capability that has been developed to provide automated simulations of delayed-neutron and/or delayed-gamma radiation using MCNP6. We present new high-fidelity delayed-gamma simulation results for models based on the neutron-fission experiments conducted by Beddingfield and Cecil to illustrate and validate this powerful feature.