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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

ACCELERATING FUSION REACTOR NEUTRONICS MODELING BY AUTOMATIC COUPLING OF HYBRID MONTE CARLO/DETERMINISTIC TRANSPORT ON CAD GEOMETRY

Detailed radiation transport calculations are necessary for many aspects of the design of fusion energy systems (FES) such as ensuring occupational safety, assessing the activation of system components for waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid Monte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily shielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of FES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required the translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion Generator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified to support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and eliminating the need for this translation step. This was done by adding a new ray tracing routine to ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC) software library. This new capability is demonstrated with a prompt dose rate calculation for an ITER computational benchmark problem using both the Consistent Adjoint Driven Importance Sampling (CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters produced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry. Significant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as a factor of 59.6).

Biondo, Elliott D↗
DOE OSTI2017

Development of the Advanced D1S for Shutdown Dose Rate Calculations in Fusion Reactors

The assessment of the shutdown dose rate due to neutron activation is fundamental for shielding design, materials selection, licensing and planning of maintenance operations in high performances fusion devices. The shutdown dose rate calculations require the combined use of radiation transports and inventory codes. The typical approach consists in a two steps procedure. A pure neutron transport simulation is performed in a first step to calculate the spatial distribution of neutron spectra. The results are used as input for an activation calculation that provides the decay gamma source used in a second step for a pure gamma transport simulation (rigorous-two step, R2S). The Direct 1-Step (D1S) approach has been developed more than ten years ago by ENEA and the ITER team: this technique relies on a coupled neutron-decay gamma transport in a single simulation. It is based on the use of a modified version of the MCNP5 Monte Carlo code with specially tailored nuclear cross-section data for transport and FISPACT inventory code for temporal behaviour. The 'Advanced-D1S' is an improved version of D1S in which additional computational features have been introduced, such as the mesh tally capability and continuous time evolution assessment of the dose rate in single run dose. It has been validated through measurements at JET tokamak and comparison with R2S codes and a future benchmark experiment during next high performance deuterium-tritium campaign is in preparation at JET. It has also been extensively used for ITER calculations and recently applied to DEMO assessment. Present limitations and in progress code improvement strategy are described and discussed in this paper.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Sticking Coefficients of Fusion Reactor Impurities from Molecular Dynamics Simulations for the Design of Cryopumps

A cryopump can be utilized as an impurity removal component of a direct internal recirculation (DIR) system for the fusion fuel cycle. The DIR facilitates a low fuel inventory by continuously pumping unburnt fuel while removing impurities from the fusion exhaust stream. A cryopump can target multiple impurity species by maintaining a temperature lower than the gas triple-point temperature that promotes desublimation. The desublimation/condensation of gases in cryopumps can be characterized by the sticking coefficient, which is defined as the probability for a gas particle to stick to a (cryo-)surface upon collision. The sticking coefficient is one of the important design/operation parameters for cryopumps, and it depends on a variety of surface and gas properties. Here, in this study, molecular dynamics simulations were utilized to estimate the sticking coefficients of typical fusion gas impurity species N 2 , CO 2 , and CH 4 over a Cu surface for a range of gas temperatures and surface coverages. The molecular dynamics study showed that the sticking coefficients for gases decrease with an increase in gas temperature. The presence of a single full monolayer of condensate on the metallic surface showed an adverse effect on the sticking of gases; however the sticking improved with two full monolayers of condensate on the surface. The sticking of gases over the mixed condensate on a surface was more favorable than the condensate of the same species for N 2 and CH 4 , with an exception for CO 2 , which showed a decrease in sticking over the mixed condensate.

cryopump↗
NASA NTRS1995

Polarized Nuclei in a Simple Mirror Fusion Reactor

The possibility of enhancing the ratio of output to input power Q in a simple mirror machine by polarizing Deuterium-Tritium (D- T) nuclei is evaluated. Taking the Livermore mirror reference design mirror ratio of 6.54, the expected sin(sup 2) upsilon angular distribution of fusion decay products reduces immediate losses of alpha particles to the loss cone by 7.6% and alpha-ion scattering losses by approx. 50%. Based on these findings, alpha- particle confinement times for a polarized plasma should therefore be 1.11 times greater than for isotropic nuclei. Coupling this enhanced alpha-particle heating with the expected greater than 50% D- T reaction cross section, a corresponding power ratio for polarized nuclei, Q(sub polarized), is found to be 1.63 times greater than the classical unpolarized value Q(sub classical). The effects of this increase in Q are assessed for the simple mirror.

Noever, David A.↗
DOE OSTI2014

Characterization of the axial plasma shock in a table top plasma focus after the pinch and its possible application to testing materials for fusion reactors

The characterization of plasma bursts produced after the pinch phase in a plasma focus of hundreds of joules, using pulsed optical refractive techniques, is presented. A pulsed Nd-YAG laser at 532 nm and 8 ns FWHM pulse duration was used to obtain Schlieren images at different times of the plasma dynamics. The energy, interaction time with a target, and power flux of the plasma burst were assessed, providing useful information for the application of plasma focus devices for studying the effects of fusion-relevant pulses on material targets. In particular, it was found that damage factors on targets of the order of 10{sup 4} (W/cm{sup 2})s{sup 1/2} can be obtained with a small plasma focus operating at hundred joules.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Divertorlets concept for low-recycling fusion reactor divertor: experimental, analytical and numerical verification

The 'divertorlets' concept is a potential non-evaporative liquid metal solution for heat removal at low recycling regime. A toroidal divertorlets prototype was built and tested in LMX-U at Princeton Plasma Physics Laboratory to evaluate the performance of this configuration. Here, details of the design, experimental results, comparison with analytical theory and MHD numerical simulations of toroidal divertorlets are covered. Experiments, analytical model and simulations showed agreement and allowed the projection of operation properties at higher magnetic flux densities (reactor-like operation), proving the concept to be a compelling solution for divertor applications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Ion Cyclotron Heating in a Levitated Dipole Fusion Reactor

OpenStar Technologies is pursuing the levitated dipole (LD) as a highly modular, loosely-coupled system that leverages their expertise in high temperature superconductor (HTS) technology. The next generation experiment at OpenStar, Tahi (Ma¯ori for “first”), will demonstrate the generation and confinement of fast ions in a levitated dipole for the first time. Ion cyclotron range of frequency (ICRF) heating is a leading candidate for energetic ion formation in Tahi. A frequency in the 10 MHz range will be used for H minority heating or D majority heating with waves launched from an antenna located above the floating coil. Unlike a tokamak, where the targeted cyclotron resonance is typically a vertical path through the center of the plasma, in a dipole the resonance location follows a C-shaped path from the separatrix to the center of the plasma. The value of B also varies significantly within the confined plasma resulting in a large number of cyclotron harmonics present in the low field region. Furthermore, levitated dipoles contain a “first closed flux surface” surrounding the floating coil, in addition to the traditional separatrix /last closed flux surface. Simulation e ff orts using full-wave ICRF codes show that ICRF heating of a levitated dipole reactor is feasible using a pair of toroidal current straps phased to launch the appropriate parallel ( i.e. poloidal) refractive index.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Study of the Interaction of Plasma with Beryllium That Is a Candidate Material for the First Wall of a Fusion Reactor

The experimental conditions for treatment of a candidate material of the first wall, beryllium, by the plasma have been determined. The plasma parameters obtained using the simulator with a plasma-beam setup during the irradiation of samples as well as the results of material studies obtained by scanning electron microscopy and energy-dispersion analysis of the beryllium surface layer have been considered.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

A Source for Fuel Supply to a Fusion Reactor Core

We present the results of studies of the plasma source based on the coaxial accelerator with the slothole channel geometry for plasma acceleration and working gas inlet into the accelerator via the electrodynamic valve. The plasma parameters at the output of the accelerator are measured. The slot-hole channel of the accelerator created higher jet pressure, as compared to the coaxial channel, especially at large distances from the source. The jet pressure reached 10{sup 6} N/m{sup 2} at a distance of 0.7 m. The source created moderately pure plasma for a current below 80 kA. The density was (2.5–5) × 10{sup 22} m{sup –3}, which was higher than the density obtained with the coaxial gun.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Exploring Acceleration Schemes for Nonlinear Heat Equations in Fusion Reactor Modeling

Plasma modeling inside a tokamak reactor poses a challenging and often times computationally expensive endeavor. The process of running these simulations has the potential to increase the predictive capabilities of these models. Here we investigate the effect particular acceleration schemes, mainly Anderson Acceleration, have on a nonlinear transport equation solver, Tango, developed by Parker et al. We show that Anderson acceleration offers a method for decreasing the computational cost through a test problem developed by Shestakov et al, which shows promise towards utilizing Anderson Acceleration in complex transport simulations

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Q1 Report for FY25 Theory and Simulation Performance Target: Development of an integrated modeling framework for fusion reactor design and assessment

This report describes the work and activities carried out towards the completion of each of the following milestones in FY25 Q1: 1. Demonstrate workflow for generating self-consistent CESOL plasma profiles + first wall and divertor loading prediction and generate the CAT plasma and neutron loading needed for further engineering analysis: $/circ$ Initially run CESOL with SOLPS and map heat flux using simple HEAT method: ▪ Generate SOLPS grid for CAT reference case, ▪ Implement and apply ‘lower’ fidelity method of using HEAT-like analytic method to map SOLPS heat flux from charged particles to first wall, and ▪ Provide initial heat flux + neutron loading for further engineering analysis. 2. Demonstrate multiphysics magnet analysis: $/circ$ Demonstrate validation of the Elmer workflow by comparing induced stresses due to EM forces generated by TF coils using multiple codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Q3 Report for FY25 Theory and Simulation Performance Target: Development of an integrated modeling framework for fusion reactor design and assessment

This report describes the work and activities carried out towards the completion of each of the following milestones in FY25 Q3: 1. Demonstrate workflow for generating self-consistent CESOL plasma profiles + first wall and divertor loading prediction and generate the CAT plasma and neutron loading needed for further engineering analysis. • Benchmark between two first wall heat flux mapping methods, identify importance of various heat flux sources and physics impact of using fully coupled CESOL vs post-analysis evaluation. 2. Generate medium fidelity parametrized CAD. • Generate parametrized CAD components for the CAT example case via either user-defined modules called within the geometry generation or by defeatured/parametrized CAD, including DCLL blanket matched to divertor boundary and magnets. Define materials, labels, and boundary conditions for passing the mesh to CFD tools. 3. Demonstrate multiphysics magnet analysis. • Demonstrate magnet analysis workflow called from the FREDA workflow, and 4. Demonstrate nuclear analysis. • Add model to OpenFOAM and/or other codes possibly including Diablo to account for tritium diffusion in solids.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Q2 Report for FY25 Theory and Simulation Performance Target: Development of an integrated modeling framework for fusion reactor design and assessment

This report describes the work and activities carried out towards the completion of each of the following milestones in FY25 Q2: 1. Demonstrate workflow for generating self-consistent CESOL plasma profiles + first wall and divertor loading prediction and generate the CAT plasma and neutron loading needed for further engineering analysis: $\circ$ Run CESOL with BOUT++/Hermes-3 and immersed boundary condition to directly map to wall: • Run BOUT++/Hermes-3 through the IPS workflow to find radial particle and energy diffusivities to match either the Eich or the physics-based scaling of the SOL heat flux width, and • Expand source of first wall heat flux to include charged particles, neutrals, and radiation from the core+edge. 2. Generate medium fidelity parametrized CAD: $\circ$ Develop the TRACER tool to read an existing CAD, regenerate the geometry based on vertex location and connectivity information, define vertex translation and parameters needed for scaling the CAD, and $\circ$ Utilize the FreeGS code to determine CAT PF coil placement, including minimizing the number of coils, coil current, and electromechanical stresses. 3. Utilize plasma loading for engineering analysis: $\circ$ Couple the plasma loading to input for OpenFOAM and demonstrate initial test of thermal analysis of CAT first wall loading with typical DCLL blanket component cooling boundary conditions. 4. Demonstrate nuclear analysis: $\circ$ Apply initial analysis of tritium transport in DCLL blanket by evaluating spatially resolved tritium generation rates, tritium diffusion and convection.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Development of a Flexible Design for Irradiation of Miniature Tensile and Charpy Test Specimens in the High Flux Isotope Reactor

Fusion promises a sustainable, clean, safe source of energy that could play a major role in solving the great challenge of meeting the world’s growing power needs. Much of fusion technology in its current state requires further maturation to support a usable fusion reactor platform. Given that the environment in and around a sustained fusion reaction is very harsh, selecting the best materials for such a reactor is crucial. The European fusion materials community spent many years developing one such material, a low-activation martensitic structural steel named Eurofer that is intended primarily for use in the fusion blanket modules. Although this material has reached technological maturity, nuclear performance data are required to qualify Eurofer for these applications. Karlsruhe Institute of Technology (KIT), a member of the EUROfusion consortium, has sponsored an irradiation program to establish irradiation performance data for various forms of the Eurofer alloy family. This document describes the design and development of flexible irradiation capsule designs that can be used to irradiate specimens to temperatures ranging from ~200 – 400°C. This report also details as-built data for the screening and high dose phases of the KIT irradiation campaign of capsules that were submitted for irradiation in the High Flux Isotope Reactor in FY18.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗