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Dynamics of heavy-Rydberg ion-pair formation in K(14p,20p)-SF{sub 6}, CCl{sub 4} collisions

The dynamics of formation of heavy-Rydberg ion-pair states through electron transfer in K(np)-SF{sub 6}, CCl{sub 4} collisions is examined by measuring the velocity, angular, and binding energy distributions of the product ion pairs. The results are analyzed with the aid of a Monte Carlo collision code that models both the initial electron capture and the subsequent evolution of the ion pairs. The model simulations are in good agreement with the experimental data and highlight the factors such as Rydberg atom size, the kinetic energy of relative motion of the Rydberg atom and target particle, and (in the case of attaching targets that dissociate) the energetics of dissociation that can be used to control the properties of the product ion-pair states.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Spherical Radiative Transfer in C++ (SRTC++): A Parallel Monte Carlo Radiative Transfer Model for Titan

We present a new computer program, SRTC++, to solve spatial problems associated with explorations of Saturn’s moon Titan. The program implements a three-dimensional structure well-suited to addressing shortcomings arising from plane-parallel radiative transfer approaches. SRTC++'s design uses parallel processing in an object-oriented, compiled computer language (C++) leading to a flexible and fast architecture. We validate SRTC++ using analytical results, semianalytical radiative transfer expressions, and an existing Titan plane-parallel model. SRTC++ complements existing approaches, addressing spatial problems like near-limb and near-terminator geometries, non-Lambertian surface phase functions (including specular reflections), and surface albedo nonuniformity.

97 MATHEMATICS AND COMPUTING↗

Effect of moisture content and distribution on radiation transport in concrete - 14426

Criticality Emergency Planning (CEP) requires a detailed knowledge of the Facility design and construction in order to undertake radiation transport studies in support of criticality consequence calculations. Such calculations are used at AWE to determine mitigated doses, CEP exclusion zones and in support of Criticality Warning Systems (CWS) detector head placement studies. At AWE, such calculations have traditionally been performed by assuming that structural concrete has a fixed and evenly distributed moisture profile. In order to validate this assumption, work has been undertaken using the finite element analysis tool Femmasse to simulate moisture movement within the concrete and calculate moisture distributions in a 65 cm thick concrete wall as a function of aging and specific environmental conditions. This work identified that considerable variation in the moisture profile may arise, with an equilibrium condition being reached after circa 20 years, where the overall moisture content is reduced to a minimum. Furthermore, the calculations identified that the overall moisture content used in extant CEP assessments was considerably lower than that estimated in this work. In order to determine the effect of the moisture distribution profiles, the Monte Carlo transport code MCNP5 was used to study the transmission of radiation through aged concrete and thus assess its performance as a radiation shield in the event of a criticality. Various moisture distributions were assessed based on the findings of the Femmasse calculations and were compared with that assessed using traditional assumptions. Results were obtained for neutron and photon doses. The paper provides an overview of the findings and makes recommendations on the suitability of using evenly distributed moisture profiles in concrete in support of mitigated doses, CEP and CWS detector head placement studies. (authors)

36 MATERIALS SCIENCE↗

Axisymmetric Plume Simulations with NASA's DSMC Analysis Code

A comparison of axisymmetric Direct Simulation Monte Carlo (DSMC) Analysis Code (DAC) results to analytic and Computational Fluid Dynamics (CFD) solutions in the near continuum regime and to 3D DAC solutions in the rarefied regime for expansion plumes into a vacuum is performed to investigate the validity of the newest DAC axisymmetric implementation. This new implementation, based on the standard DSMC axisymmetric approach where the representative molecules are allowed to move in all three dimensions but are rotated back to the plane of symmetry by the end of the move step, has been fully integrated into the 3D-based DAC code and therefore retains all of DAC s features, such as being able to compute flow over complex geometries and to model chemistry. Axisymmetric DAC results for a spherically symmetric isentropic expansion are in very good agreement with a source flow analytic solution in the continuum regime and show departure from equilibrium downstream of the estimated breakdown location. Axisymmetric density contours also compare favorably against CFD results for the R1E thruster while temperature contours depart from equilibrium very rapidly away from the estimated breakdown surface. Finally, axisymmetric and 3D DAC results are in very good agreement over the entire plume region and, as expected, this new axisymmetric implementation shows a significant reduction in computer resources required to achieve accurate simulations for this problem over the 3D simulations.

Stewart, B. D.↗

Development of Estimation Method for Radioactivity Concentration Distribution of Radioactive Wastes in a Box-shaped Container - 17034

We are developing a new estimation method suitable for determining radioactivity concentration distribution of radioactive wastes stored in a square box-shaped container. This method uses the Maximum A Posteriori (MAP) method which is a way to solve the inverse problem, and it estimates the radioactivity concentration distribution by comparing simulated data with measured data. In the measurement, gamma rays emitted from the radioactive wastes are measured by several gamma ray spectrometers which are set at different positions in relation to each other. On the other hand, the gamma ray spectrum data of each spectrometer are simulated by a Monte Carlo simulation code. The method uses an iterative calculation to solve the concentration distribution with measured spectrum data and simulated spectrum data. We conducted a simulation test to check the algorithm of our estimation method. And we also conducted an experiment using activated objects. We found that the concentration of the wastes in a box-shaped container could be estimated within a 20 % error for the simulation, and within a 40 % error for the experiment and we confirmed that feasibility of our method was favorable. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Review of Monte Carlo simulations for backgrounds from radioactivity

For all experiments dealing with the rare event searches (neutrino, dark matter, neutrino-less double-beta decay), the reduction of the radioactive background is one of the most important and difficult tasks. There are basically two types of background, electron recoils and nuclear recoils. The electron recoil background is mostly from the gamma rays through the radioactive decay. The nuclear recoil background is from neutrons from spontaneous fission, (α, n) reactions and muoninduced interactions (spallations, photo-nuclear and hadronic interaction). The external gammas and neutrons from the muons and laboratory environment, can be reduced by operating the detector at deep underground laboratories and by placing active or passive shield materials around the detector. The radioactivity of the detector materials also contributes to the background; in order to reduce it a careful screening campaign is mandatory to select highly radio-pure materials. In this review I present the status of current Monte Carlo simulations aimed to estimate and reproduce the background induced by gamma and neutron radioactivity of the materials and the shield of rare event search experiment. For the electromagnetic background a good level of agreement between the data and the MC simulation has been reached by the XENON100 and EDELWEISS experiments, using the GEANT4 toolkit. For the neutron background, a comparison between the yield of neutrons from spontaneous fission and (α, n) obtained with two dedicated softwares, SOURCES-4A and the one developed by Mei-Zhang-Hime, show a good overall agreement, with total yields within a factor 2 difference. The energy spectra from SOURCES-4A are in general smoother, while those from MZH presents sharp peaks. The neutron propagation through various materials has been studied with two MC codes, GEANT4 and MCNPX, showing a reasonably good agreement, inside 50% discrepancy.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Numerical study of error propagation in Monte Carlo depletion simulations

Improving computer technology and the desire to more accurately model the heterogeneity of the nuclear reactor environment have made the use of Monte Carlo depletion codes more attractive in recent years, and feasible (if not practical) even for 3-D depletion simulation. However, in this case statistical uncertainty is combined with error propagating through the calculation from previous steps. In an effort to understand this error propagation, a numerical study was undertaken to model and track individual fuel pins in four 17 x 17 PWR fuel assemblies. By changing the code's initial random number seed, the data produced by a series of 19 replica runs was used to investigate the true and apparent variance in k{sub eff}, pin powers, and number densities of several isotopes. While this study does not intend to develop a predictive model for error propagation, it is hoped that its results can help to identify some common regularities in the behavior of uncertainty in several key parameters. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Integration of the QMSFRG Database into the HZETRN Code

Accurate nuclear interaction data bases are needed for describing the transport of space radiation in matter including space craft structures, atmospheres, and tissues. Transport models support the identification and development of new material concepts for human and electronic part protection. Quantum effects are manifested in nuclear reactions in several ways including interference effects between terms in the multiple scattering series, the many-body nuclear wave functions (for e.g. the roles of shell structure and Fermi momentum) and nuclear clustering. The quantum multiple scattering fragmentation model (QMSFRG) is a comprehensive model for generating nuclear interaction databases for galactic cosmic ray (GCR) transport. Other nuclear databases including the NUCFRG model and Monte-Carlo simulation codes such as FLUKA, LAHET, HETC, and GEANT ignore quantum effects. These codes fail to describe many important features of nuclear reactions and are thus inaccurate for the evaluation of materials for radiation protection. Previously we have shown that quantum effects are manifested through constructive interference in forward production spectra, the effects of Fermi momentum on production spectra, cluster nuclei knockout, and the nuclear response function. Quantum effects are especially important for heavy ions with mass numbers less than 20 that dominate radiation transport in human tissues and for the materials that are expected to be superior in space radiation protection. We describe the integration of the QMSFRG model into the HZETRN transport code. Integration milestones include proper treatment of odd-even charge-mass effects in nuclear fragmentation and the momentum distribution of nucleon production from GCR primary heavy ions. We have also modified the two-body amplitudes in the model to include nuclear medium effects. In order to include a comprehensive description of the GCR isotopic composition in materials, we have described the isotopic composition of the GCR by extending the 59-isotope version of HZETRN to an 120-isotope version. The isotopic composition of most primary GCR elements (including H, He, C, N, O, Ne, Mg, Si, Ar, Ca, Cr, and Fe) are included in the extended model. We discuss results for the high-energy neutron composition inside materials, and the charge and mass distribution for benchmark GCR problems.

Cucinotta, F. A.↗

Passive Safety Feature of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. The facility has an electron accelerator driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assembly is composed of WWR-M2 Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The water coolant operates at room temperature during normal operation, when the facility is operating at full power using 100-KW electron beam power. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/BVII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight temperature increase and the corresponding density decrease in the water coolant would cause relatively large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

Monte Carlo interpretation of the AMMON/REF experiment in EOLE for the JHR reactor calculations

A new experiment, named AMMON and dedicated to the analysis of the JHR neutron physics and to the qualification of the associated neutron calculation tools, is currently in progress in the EOLE zero-power experimental reactor at CEA Cadarache. The first core configuration, so called 'reference configuration' (AMMON/REF) has already been loaded: it consists of an experimental zone of 7 JHR assemblies with U{sub 3}Si{sub 2} 27% {sup 235}U enriched fuel curved plates surrounded by a driver zone with 623 standard PWR UOx fuel pins. The analysis of the AMMON/REF measurements is presented in this paper. It is based on calculations performed with the 3-dimensional reference Monte Carlo TRIPOLI4.7 code and the JEFF3.1.1 European library: it highlights a good agreement between calculation and experiment concerning reactivity and power distribution in the experimental zone. Reactivity prediction is very satisfactory, despite the presence of a large aluminum quantity in the core: C-E=+ 376 pcm {+-} 290. The radial and axial plate power profiles are also in good agreement. Assembly power distribution is predicted within the experimental uncertainties (1{sigma}). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

A Parallelized Oxidation-Driven Surface Recession Framework in DSMC Code, SPARTA

Spacecrafts rely on ablative thermal protection systems (TPS) made of composites consisting of a carbon-based reinforcement and a polymeric matrix. These materials are designed to withstand high-temperature oxidation and surface recession during re-entry into the Earth's atmosphere. However, ablation occurs due to a complex interplay of thermal, mechanical, and chemical factors, making it challenging to determine the individual impact of each on the TPS's overall degradation. In this study, we have developed an ablation model that can leverage a finite rate carbon oxidation model to predict material recession and surface states more accurately. Stochastic PArallel Rarified-gas Time-accurate Analyzer (SPARTA), a direct-simulation Monte Carlo (DSMC) code, is modified to allow oxidation-driven ablation of implicitly defined carbon surfaces. In SPARTA, implicit surfaces are generated from the grid corner point values via a marching cubes algorithm, therefore creating a new set of surface elements every time ablation is performed. The finite-rate oxidation model developed by Gopalan et. al can perform both gas-surface and pure-surface reactions and is now adapted to tally surface data on a per grid cell basis. The ablation functionality was also adjusted so once the reactions have occurred, the number of reactions leading to CO formation can be converted to corner point reduction values; therefore, carbon removal is directly proportional to surface recession. We also briefly discuss some unique challenges associated with parallelizing this dynamic surface state and geometry. Finally, we analyze the performance of this parallelized implicit chemistry model with simple 2D and 3D benchmark cases by producing surface state statistics, area changes over time, and visualization across a range of surface temperatures and processors with and without load-balancing.

DSMC↗

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↗

RAPTURE User's Manual

This User's Guide serves as a brief introduction to the RAPTURE radiation effects analysis code. It includes an overview of the input format, RAPTURE's error- and consistency-checking of the user-provided input files, the automatic-differentiation and convergce-checking schemes employed by RAPTURE, and the RAPTURE output files. A variety of example problems are included in this Guide which collectively demonstrate RAPTURE's current capabilities and provide a suite of test problems and template input files for the user. This Guide includes, for each problem, the problem description, RAPTURE input files, and comparison of the RAPTURE solution with solutiong generated with the Monte Carlo transport code ITS, the legacy deterministic code ADEPT, and, where possible, published experimental results. An appendix includes a description of all keywords and options in the RAPTURE input file.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Coupled Monte Carlo Probability Density Function/ SPRAY/CFD Code Developed for Modeling Gas-Turbine Combustor Flows

The success of any solution methodology for studying gas-turbine combustor flows depends a great deal on how well it can model various complex, rate-controlling processes associated with turbulent transport, mixing, chemical kinetics, evaporation and spreading rates of the spray, convective and radiative heat transfer, and other phenomena. These phenomena often strongly interact with each other at disparate time and length scales. In particular, turbulence plays an important role in determining the rates of mass and heat transfer, chemical reactions, and evaporation in many practical combustion devices. Turbulence manifests its influence in a diffusion flame in several forms depending on how turbulence interacts with various flame scales. These forms range from the so-called wrinkled, or stretched, flamelets regime, to the distributed combustion regime. Conventional turbulence closure models have difficulty in treating highly nonlinear reaction rates. A solution procedure based on the joint composition probability density function (PDF) approach holds the promise of modeling various important combustion phenomena relevant to practical combustion devices such as extinction, blowoff limits, and emissions predictions because it can handle the nonlinear chemical reaction rates without any approximation. In this approach, mean and turbulence gas-phase velocity fields are determined from a standard turbulence model; the joint composition field of species and enthalpy are determined from the solution of a modeled PDF transport equation; and a Lagrangian-based dilute spray model is used for the liquid-phase representation with appropriate consideration of the exchanges of mass, momentum, and energy between the two phases. The PDF transport equation is solved by a Monte Carlo method, and existing state-of-the-art numerical representations are used to solve the mean gasphase velocity and turbulence fields together with the liquid-phase equations. The joint composition PDF approach was extended in our previous work to the study of compressible reacting flows. The application of this method to several supersonic diffusion flames associated with scramjet combustor flow fields provided favorable comparisons with the available experimental data. A further extension of this approach to spray flames, three-dimensional computations, and parallel computing was reported in a recent paper. The recently developed PDF/SPRAY/computational fluid dynamics (CFD) module combines the novelty of the joint composition PDF approach with the ability to run on parallel architectures. This algorithm was implemented on the NASA Lewis Research Center's Cray T3D, a massively parallel computer with an aggregate of 64 processor elements. The calculation procedure was applied to predict the flow properties of both open and confined swirl-stabilized spray flames.

Source record↗

Radiation Protection for the First Korea Heavy-ion Medical Accelerator (KHIMA) Facility - Paper 7

The number of particle therapy facilities is increasing worldwide due to the encouraging results obtained at research facilities such as the Lawrence Berkeley National Laboratory (USA), the Helmholtz Center for Heavy Ion Research GSI (Germany), and the Heavy Ion Medical Accelerator (HIMAC) in Chiba (Japan). The first hospital-based heavy-ion therapy facility is under construction in Busan, Korea, at the Korea Institute of Radiological and Medical Sciences [KIRAMS]. The project was originally designed to develop a compact superconducting cyclotron for hadron therapy able to deliver forefront beams of heavier ions, essentially carbon ions of energy, 430 MeV/u. Radiation protection issues play an important role in every stage of the project such as civil construction planning, control system design, and special technological plants. Monte Carlo transport codes are the best tools to solve these issues and the FLUKA2011 and MCNPX2.7.0 codes are used in this work. Beam loss distributions of cyclotron installations in combination with passive or active beam shaping methods have a particularly high implication on the protection layout of the facility. Methods for shielding and activation calculations are described with an emphasis on carbon ion beams. All the tools and databases require verification. Benchmarking is an instrumental process for verification. An overview of benchmarking efforts will also be given and the need for further benchmarks is discussed. (authors)

43 PARTICLE ACCELERATORS↗

Neutron counter based on beryllium activation

The fusion reaction occurring in DD plasma is followed by emission of 2.45 MeV neutrons, which carry out information about fusion reaction rate and plasma parameters and properties as well. Neutron activation of beryllium has been chosen for detection of DD fusion neutrons. The cross-section for reaction {sup 9}Be(n, α){sup 6}He has a useful threshold near 1 MeV, which means that undesirable multiple-scattered neutrons do not undergo that reaction and therefore are not recorded. The product of the reaction, {sup 6}He, decays with half-life T{sub 1/2} = 0.807 s emitting β{sup −} particles which are easy to detect. Large area gas sealed proportional detector has been chosen as a counter of β–particles leaving activated beryllium plate. The plate with optimized dimensions adjoins the proportional counter entrance window. Such set-up is also equipped with appropriate electronic components and forms beryllium neutron activation counter. The neutron flux density on beryllium plate can be determined from the number of counts. The proper calibration procedure needs to be performed, therefore, to establish such relation. The measurements with the use of known β–source have been done. In order to determine the detector response function such experiment have been modeled by means of MCNP5–the Monte Carlo transport code. It allowed proper application of the results of transport calculations of β{sup −} particles emitted from radioactive {sup 6}He and reaching proportional detector active volume. In order to test the counter system and measuring procedure a number of experiments have been performed on PF devices. The experimental conditions have been simulated by means of MCNP5. The correctness of simulation outcome have been proved by measurements with known radioactive neutron source. The results of the DD fusion neutron measurements have been compared with other neutron diagnostics.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Correlated prompt fission data in transport simulations

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

OpenEdge: A collaborative, open-source, multi-purpose direct simulation Monte Carlo for plasma simulation in magnetic fusion environments

OpenEdge is a collaborative, open-source, object-oriented Direct Simulation Monte Carlo (DSMC) code, designed specifically for plasma simulations in magnetic fusion environments. Here, the code features include advanced structures, robust capabilities, and an effective parallelization strategy, all of which significantly enhance performance. It includes specialized modules for managing complex particle interactions, including collisions, ionization/recombination, and reflection/sputtering. Benchmarks and performance analyses have confirmed its efficiency and scalability. Versatile and adaptable, OpenEdge is applied across a broad spectrum of plasma-material interaction studies and charged particle transport in various fusion research settings.

Boundary plasma↗