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Shores, Erik Frederick

Publications and source records attributed to Shores, Erik Frederick.

Calculated Gamma Output from a 6-kilogram Sphere of Neptunium

We previously modeled a 6 kg neptunium sphere with pyDMTK 2.0.0b, a python-based intrinsic radiation (INRAD) modeling tool, on the MOONLIGHT machine. Here we report results from version 2.0.1b on SNOW, another TriLab Linux Capacity Cluster (TLCC) resource on the Laboratory’s Turquoise network. We also present gamma output from MCNP 6.2.0 in terms of discrete line strengths, full gamma spectra, and dose rate maps for visualization. Results from both models agree with recent gamma measurements.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Reactivity of a Cube, Cylinder, and Sphere

A water-reflected cube of dry plutonium oxide is found to be more reactive than a sphere of the same mass and composition. MCNP6 ® criticality results for a cube, sphere, and cylinder of identical density, mass, and volume are discussed in the context of several hydrogenous reflectors. Similar geometric comparisons are made for a fast metal system based on 240 Pu Jezebel.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

l3d2vtk: An MCNPTools Utility to Enable LNK3DNT File Visualization & Post-Processing

This paper describes recent work to develop and incorporate a utility program into the MCNPTools package that enables LNK3DNT file visualization & post-processing. A LNK3DNT file is a binary Cartesian, cylindrical, or spherical structured mesh file that describes material and material relative density distributions by mesh voxel. This mesh file format is principally used by the PARTISN discrete-ordinates radiation-transport code. However, the MCNP ® Monte Carlo radiation-transport code can use LNK3DNT files to specify geometry, and the MCNP code can produce LNK3DNT files representing a user-defined region of the MCNP constructive solid geometry (CSG). Because of the use of LNK3DNT files as MCNP input and output and with other computer codes, improving the ability to visualize and/or quantitatively post-process LNK3DNT files benefits end users. In particular, it is helpful to visually verify that the material and density distributions are as expected. Also, an analyst can orient viewers with respect to other in formation, e.g., by co-plotting a voxelized representation of MCNP CSG with mesh tally and/or particle track results. However, until now, the only capability to interactively interrogate LNK3DNT files is through a 2-D slice plotter provided by the Monte Carlo Application ToolKit (MCATK), which is not currently distributed outside of Los Alamos National Laboratory. The utility created as part of this work converts LNK3DNT binary files to an ASCII XML-based StructuredMesh VTK (.vts) file. The geometries supported are 1-D spherical and cylindrical ( r ), 1-D Cartesian ( x ), 2-D cylindrical ( r-z and r-θ ) and Cartesian ( x-y ), and 3-D cylindrical ( r-z-θ ) and Cartesian ( x-y-z ). This file format is chosen to minimize external dependencies (e.g., on an HDF5 library) and to provide a file that can be interrogated by the user in a text editor or standard visualization applications (e.g., ParaView or VisIt). However, other file formats may provide more utility such as XDMF, which can use HDF5 and is discussed further as something to explore as future work. This paper is organized as follows: a brief introduc tion to the MCNPTools library and utility programs is given, the LNK3DNT-to-VTK (l3d2vtk) utility functionality is de scribed, and then an example post-processing workflow using the ParaView application is given.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

l3d2vtk: An MCNPTools Utility to Enable LNK3DNT File Visualization & Post-processing [Slides]

This presentation discusses the objective of the project which is to provide a utility (among other utilities) that improves MCNP analysis post-processing and visualization capabilities. Much recent work has been done to improve MCNP post-processing and visualization. It also discusses the goals of the project, which are to create a product that is portable, open, easily interrogated by a user, no dependencies, including a mesh tally-to-VTK conversion, an EEOUT-to-VTK conversion, and an interactive, 3-D, visualization environments (e.g., ParaView). The presentation also discusses the reasons regarding why this is being done, which are to provide physical insight into behaviors and trends, to facilitate error mitigation (easier to “see” errors than “read” them), and to improve communication with collaborators. It also discusses why this is being done now as analyses are getting ever-more complicated and are more flexible in terms of build, test, and deployment capabilities.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Cyclops I: Modeling Error Study for Inertially Confined Detonation of Insensitive High Explosive (IHE) PBX-9502

A validation study is performed using the hydrodynamics code PAGOSA to assess the accuracy of reflected shock, burn front propagation, and Photon Doppler Velocimetry (PDV) predictions using several burn models and strength models. Experimental data collected from proton radiography (pRad) is compared to burn front propagation times and reflected shock patterns. Burn front isochrones are computed and the lag time for the burn front is quantified for a given burn model. The effect of strength models and damage are assessed by comparison of simulations to experimental PDV data.

97 MATHEMATICS AND COMPUTING↗

Tabulated Neutron Emission Rates for Plutonium Oxide

This work tabulates neutron emission rates for 80 plutonium oxide samples as reported in the literature. Plutonium-­238 and plutonium-­239 oxides are included and such emission rates are useful for scaling tallies from Monte Carlo simulations and estimating dose rates for health physics applications.

61 RADIATION PROTECTION AND DOSIMETRY↗

Predicted Radiation Output from Several Kilograms of Plutonium Oxide

Neutron and gamma signatures from 3.44 kg of plutonium oxide were predicted with MCNP6 in anticipation of a passive diagnostic measurement campaign. Calculated radiation fields are visualized through use of “FMESH” type tallies. This work is effectively an extension of previous efforts and application of MISC for the photon source term is the main advancement. Large sources of modeling uncertainty include physical can dimensions and thus internal volume and density estimates. Material age is unknown and will also affect the passive gamma signature. The measurement campaign will provide information for model refinement and subsequent simulations. The authors have requested passive data collection on the pair of cans with the largest and smallest mass and such data will be valuable for benchmarking purposes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗