Gregg W. McKinney

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.

Present and future capabilities of MCNP

Several new capabilities have been added to MCNP4C including: (1) macrobody surfaces; (2) the superimposed mesh importance functions, so that it is no longer necessary to subdivide geometries for variance reduction; and (3) Xlib graphics and DVF Fortran 90 for PCs. There are also improvements in neutron physics, electron physics, perturbations, and parallelization. In the more distant future we are working on adaptive Monte Carlo code modernization, more parallelization, visualization, and charged particles.

MCNP Enhancements, Parallel Computing, and Error Analysis for BNCTa

The Boron Neutron Capture Therapy (BNCT) treatment planning procedure used by the Harvard/MIT BNCT Program relies on MCNPa to calculate dose rates throughout a patient specific model for NCT irradiations. Since MCNP transport calculations are a time consuming portion of the treatment planning process, their acceleration greatly improves treatment planning efficiency. Source code augmentations and implementation of parallel computing on Windows NT computers have greatly decreased the time needed for these dosimetry calculations. A statistical uncertainty analysis verified that the appropriate number of source particles, which varies with the irradiation beam orientation, was being used for treatment planning.

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.

DNDO Report: Developing a Photon Clutter Scaling Algorithm for Implementation into the NINESIM Tool

The Nuclear Inspection Node Event SIMulator (NINESIM) is a tool that is capable of modeling a node that monitors passing traffic for radiological and nuclear materials. This task requires the consideration of a number of factors. One such factor is clutter, or the level of traffic passing through a node. Clutter might affect the performance of radiation detectors due to attenuation or scattering that particles undergo as they move through the material. This paper discusses the results from MCNP simulations that were performed to determine in what scenarios clutter had a significant impact on detector performance, where the receiver operating characteristic (ROC) curve was used as the metric to evaluate detector performance. Clutter was found to have an insignificant impact on neutron detection, therefore, clutter may be ignored when modeling neutron detectors. However, there were a number of scenarios where clutter was found to affect the performance of photon detectors; the amount of traffic was found to be directly proportional to the effectiveness of photon detectors. The results of this work were then used to develop a photon clutter scaling algorithm that will be implemented into NINESIM. The Domestic Nuclear Detection Office, and possibly other agencies such as the Defense Threat Reduction Agency, will benefit from this work by eventually obtaining a tool that can enhance their ability in countering the illicit movement of radiological and nuclear materials. DNDO Report: Developing a Photon Clutter Algorithm for Implmentation into the NINESIM Tool INTRODUCTION Nuclear terrorism is one of the most significant national security concerns that exist. In order to combat this threat, the United States established the Domestic Nuclear Detection Office (DNDO) within the U.S. Department of Homeland Security. This office is tasked with overseeing and improving the Global Nuclear Detection Architecture (GNDA), which is a comprehensive system with the express purpose of detecting the illicit movement of nuclear and radiological material. The GNDA is composed of nodes, each of which is a complicated, interconnected logistical system of capabilities. In order to optimize the design of an arbitrary GNDA node, the DNDO is funding a project called the Nuclear Inspection Node Event SIMulator (NINESIM). DESCRIPTION OF THE RESEARCH PROJECT The goal of the NINESIM project is to develop a discrete event simulation tool that can model the movement of traffic through an arbitrary GNDA node. A few factors that the tool considers include geographic location, quality of personnel, and the type of radiation detector deployed. A user can then adjust parameters to determine the optimal design of a given node. A specific factor that must be considered when developing NINESIM is the effect of clutter on radiation detection. The clutter, or level of traffic passing through a node, might affect the ability of a radiation detector to detect photons and neutrons emitted by sources. Neglecting this scaling factor could lead to poor predictions of detector performance in certain scenarios. A modified version of MCNP6 that had a memory reduction capability was used

Enhancements to the MCNP6 background source

Abstract 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.

Testing the Delayed Gamma Capability in MCNP6

Abstract 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., 137Ba), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% (28Mg, 28Al, 42K, 47Ca, 47Sc, 60Co). 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.

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.