Joe W. Durkee

Initial MCNP6 Release Overview

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 MCNPX Monte Carlo Radiation Transport Code

MCNPX (Monte Carlo N‐Particle eXtended) is a general‐purpose Monte Carlo radiation transport code with three‐dimensional geometry and continuous‐energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi‐processing computer platforms. MCNPX is based on MCNP4c and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low‐energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics, particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development.

The MCNP6 Delayed-Particle Feature

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.

Delayed-Gamma Simulation Using MCNPX

Abstract Monitoring issues related to activation and fission processes occur in many health physics, instrumentation and equipment design, nuclear forensics, and homeland security applications. Gamma radiation that is emitted during these processes as a result of the radioactive decay of reaction by-products [delayed gammas (DGs)] provides unique signatures that are useful for interrogation and information acquisition. Thus, it is of compelling interest to have a simulation tool that can be used to conduct studies to provide insights into the activation and fission processes. Beginning with version 2.5.0, MCNPX has been undergoing major upgrades to facilitate DG simulations. We illustrate the upgrades for a simple multiparticle reaction model involving 60Ni and for 235U photofission caused by 12-MeV photons.

Multiregion Concentration Diffusion Coefficient Determination Using Davidon’s Variable Metric Method

An analytic solution of the one-dimensional steadystate multiregion concentration diffusion decay equation is constructed. The solution is used to determine the diffusion coefficients of metallic fission products in high-temperature gas-cooled reactor fuel particles from experimental measurement of the concentrations using Davidons variable metric method for chi-square minimization. Typically, for two to four material regions with 50 measured concentration data points, the diffusion coefficients and their associated uncertainties can be determined rapidly (<8 s on the AMDAHL 470/V6). Using analytical solutions, the diffusion coefficients can be determined about25 times faster than using finite difference solutions. The methodology is applied to Zollers concentration measurements of /sup 137/Cs and /sup 90/Sr.

Material Protection, Accounting, and Control Technologies (MPACT) Advanced Integration Roadmap

The development of sustainable advanced nuclear fuel cycles is a long-term goal of the Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technologies program. The Material Protection, Accounting, and Control Technologies (MPACT) campaign is supporting research and development (R&D) of advanced instrumentation, analysis tools, and integration methodologies to meet this goal (Miller, 2015). This advanced R&D is intended to facilitate safeguards and security by design of fuel cycle facilities. The lab-scale demonstration of a virtual facility, distributed test bed, that connects the individual tools being developed at National Laboratories and university research establishments, is a key program milestone for 2020. These tools will consist of instrumentation and devices as well as computer software for modeling, simulation and integration.

Exact solution to the time-dependent one-speed multiregion neutron diffusion equation

Abstract Exact solutions are obtained to the time-dependent one-speed neutron diffusion equation in one-dimensional multiregion Cartesian and spherical geometries with multiplication and without delayed neutrons. These solutions enable the study of the one-speed space-time behavior of prompt neutrons in an arbitatry number of neutronically dissimilar material regions. Parametric benchmark calculations are presented.

Material Protection, Accounting, and Control Technologies (MPACT): Modeling and Simulation Roadmap

The development of sustainable advanced nuclear fuel cycles is a long-term goal of the Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technologies program. The Material Protection, Accounting, and Control Technologies (MPACT) campaign is supporting research and development (R&D) of advanced instrumentation, analysis tools, and integration methodologies to meet this goal. This advanced R&D is intended to facilitate safeguards and security by design of fuel cycle facilities. The lab-scale demonstration of a virtual facility, distributed test bed, that connects the individual tools being developed at National Laboratories and university research establishments, is a key program milestone for 2020. These tools will consist of instrumentation and devices as well as computer software for modeling. To aid in framing its long-term goal, during FY16, a modeling and simulation roadmap is being developed for three major areas of investigation: (1) radiation transport and sensors, (2) process and chemical models, and (3) shock physics and assessments. For each area, current modeling approaches are described, and gaps and needs are identified.

MCNPX Improvements for Threat Reduction Applications

Enhancements contained in the current MCNPX 2.6.0 Radiation Safety Information Computational Center (RSICC) release will be presented, including stopped‐muon physics, delayed neutron and photon generation, and automatic generation of source photons. Preliminary benchmarking comparisons with data taken with a muon beam at the Paul Scherrer Institute Spallation Neutron Source accelerator will be discussed. We will also describe current improvements now underway, including Nuclear Resonance Fluorescence (NRF), pulsed sources, and others. We will also describe very new work begun on a threat‐reduction (TR) user interface, designed to simplify the setup of TR‐related calculations, and introduce standards into geometry, sources and backgrounds.

Accelerator‐driven molten‐salt blankets: Physics issues

A number of nuclear physics issues concerning the Los Alamos molten‐salt, accelerator‐driven plutonium converter are discussed. General descriptions of several concepts using internal and external moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m3 per year. Beginning‐of‐life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics.