Engineering topics

Canik, John

Publications and source records attributed to Canik, John.

8 records

DOE OSTI2024

Sparse-grid discontinuous Galerkin methods for the Vlasov–Poisson–Lenard–Bernstein model

Sparse-grid methods have recently gained interest in reducing the computational cost of solving high-dimensional kinetic equations. In this paper, we construct adaptive and hybrid sparse-grid methods for the Vlasov–Poisson–Lenard–Bernstein (VPLB) model. This model has applications to plasma physics and is simulated in two reduced geometries: a 0x3v space homogeneous geometry and a 1x3v slab geometry. Here we use the discontinuous Galerkin (DG) method as a base discretization due to its high-order accuracy and ability to preserve important structural properties of partial differential equations. We utilize a multiwavelet basis expansion to determine the sparse-grid basis and the adaptive mesh criteria. We analyze the proposed sparse-grid methods on a suite of three test problems by computing the savings afforded by sparse-grids in comparison to standard solutions of the DG method. The results are obtained using the adaptive sparse-grid discretization library ASGarD.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2021

Integrated model predictions on the impact of substrate damage on gas dynamics during ITER burning-plasma operations

Divertor design and choice of plasma-facing materials (PFM) will be essential to the success of next-generation fusion reactors as they operate under more powerful scenarios. Understanding and controlling interactions between the plasma and PFM is essential to making these choices. Within these plasma–material interactions and especially in tungsten (W), the interplay between the most abundant plasma species (hydrogen isotopes and helium, He) with the wall material alters fuel retention. However, this interplay is yet to be sufficiently understood to confidently project fuel retention levels to future fusion devices. The paper presents a series of integrated simulations of fusion plasmas and their interaction with tungsten. Specifically, this study assesses the impact of He plasma pre-exposure on hydrogenic species retention during 100 s of burning plasma operations (BPO) in ITER. Multiple pre-exposure scenarios are considered, including sub-surface damage resulting from exposures in the linear device PISCES and from early ITER He-operation. The predictions from these consecutive He-BPO exposures show that fuel content and spatial distribution in the material are largely determined by the He-induced damage, as manifest in: (i) changes in surface temperature expected during BPO have little effect on fuel retention in the presence of He-induced damage; (ii) gas content stabilizes quickly in substrates pre-exposed in PISCES, at levels set by the concentration of pre-existing vacancies, while it continues to increase in substrates initially pristine or pre-exposed to ITER He plasmas; (iii) the presence of He and He–V clusters in the near-surface region locally increases hydrogenic retention, but decreases its permeation; this results in hydrogenic species that remain closer to the surface in pre-damaged substrates, while the bulk content is higher for initially pristine cases. In summary, the interaction and binding of D and T with the pre-existing He–V clusters modifies retention and permeation of hydrogen species during ITER BPO.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Control of the Plasma-Material Interface for Long Pulse Optimization in EAST Final Report

This final report describes the funded activity to understand and control the plasma-material interface to improve long pulse discharge control and performance on the EAST (Experimental Advanced Superconducting Tokamak) facility, located in Hefei, China. The primary focus of the research was long pulse recycling control via optimization of lithium (Li) delivery systems, including flowing liquid lithium plasma facing components (PFCs), because Li persists as the main wall conditioning technique. On EAST, lithium was primarily introduced using crucible evaporation, pellet injection, and gravitationally dropped powder. Overall, the introduction of lithium into the EAST plasma had strong and persistent effects on the ELM behavior during long pulse EAST discharges demonstrating overall improved performance. Additionally, a flexible impurity injection tool, a multi-chamber impurity powder dropper, was developed and deployed on EAST, to extend powder injection to elements other than Li. Finally, improvements to the Soft X-ray diagnostic system were introduced to improve measurements of impurity transport and electron temperature.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Divertor heat flux challenge and mitigation in SPARC

Owing to its high magnetic field, high power, and compact size, the SPARC experiment will operate with divertor conditions at or above those expected in reactor-class tokamaks. Power exhaust at this scale remains one of the key challenges for practical fusion energy. Based on empirical scalings, the peak unmitigated divertor parallel heat flux is projected to be greater than 10 GW m -2 . This is nearly an order of magnitude higher than has been demonstrated to date. Furthermore, the divertor parallel Edge-Localized Mode (ELM) energy fluence projections (~11–34 MJ m -2 ) are comparable with those for ITER. However, the relatively short pulse length (~25 s pulse, with a ~10 s flat top) provides the opportunity to consider mitigation schemes unsuited to long-pulse devices including ITER and reactors. The baseline scenario for SPARC employs a ~1 Hz strike point sweep to spread the heat flux over a large divertor target surface area to keep tile surface temperatures within tolerable levels without the use of active divertor cooling systems. In addition, SPARC operation presents a unique opportunity to study divertor heat exhaust mitigation at reactor-level plasma densities and power fluxes. Not only will SPARC test the limits of current experimental scalings and serve for benchmarking theoretical models in reactor regimes, it is also being designed to enable the assessment of long-legged and X-point target advanced divertor magnetic configurations. Experimental results from SPARC will be crucial to reducing risk for a fusion pilot plant divertor design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

FY2018 Fusion Theory and Simulation Milestone on Plasma Materials Interaction (PMI) (Final Report)

The interaction of the boundary plasma with the material surfaces in magnetically confined plasmas is among the most critical problems in fusion energy science. In FY2018, perform high performance computing simulations with coupled boundary plasma physics and materials surface models to predict the fuel recycling and tritium retention of the ITER divertor for D-T burning plasma conditions, accounting for erosion, re-deposition and impurity transport in the plasma boundary, and an initial evaluation of the influence of material deposition on the recycling and retention.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Advancing Fusion with Machine Learning Research Needs Workshop

The pursuit of fusion energy has required extensive experimental and theoretical science activities to develop the knowledge needed that will enable design of successful fusion power plants. Even today, following decades of research in many key areas including plasma physics and material science, much remains to be learned to enable optimization of the tokamak or other paths to fusion energy. Data science methods from the fields of machine learning and artificial intelligence (ML/AI) offer opportunities for enabling or accelerating progress toward the realization of fusion energy by maximizing the amount and usefulness of information extracted from experimental and simulation output data. Jointly supported by the Department of Energy Offices of Fusion Energy Science (FES) and Advanced Scientific Computing Research (ASCR), a workshop was organized to identify Priority Research Opportunities (PRO’s) for application of ML/AI methods to enable accelerated solution of fusion problems. The resulting “Advancing Fusion with Machine Learning Research Needs Workshop,” held in Gaithersburg, MD, April 30 – May 2, 2019, brought together ~ 60 experts in fields spanning fusion science, data science, statistical inference and mathematics, machine learning, and artificial intelligence, along with DOE program managers and technical experts, to identify key PRO’s.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Surface Erosion of Plasma-Facing Materials Using an Electrothermal Plasma Source and Ion Beam Micro-Trenches

Erosion characteristics of tungsten-alternative plasma-facing materials (PFMs) were tested under high heat flux conditions in the electrothermal plasma source facility at Oak Ridge National Laboratory. The PFMs of interest are high-purity β-3C chemical vapor deposition silicon carbide (SiC) and the MAX phases Ti 3 SiC 2 and Ti 2 AlC [MAX = chemical formula M n+1 AX n , where M is an early transition metal (such as Ti or Ta), A is an A-group element (such as Si or Al), and X is carbon or nitrogen]. Additionally, an erosion analysis method was developed using a combination of focused ion beam microscopy and scanning electron microscopy, carving micro-trench geometries into polished sample surfaces. Samples of SiC, Ti 3 SiC 2 , and Ti 2 AlC were exposed to the electrothermal plasma source alongside tungsten and monocrystalline silicon. Samples were exposed to a Lexan polycarbonate (C 16 H 14 O 3 ) electrothermal plasma stream in a He environment, at a specified impact angle, with infrared camera diagnostics. Edge localized mode–relevant heat fluxes of 0.9 to 1 GW/m 2 over 1-ms discharges were generated on the target surfaces. Tungsten samples exhibited pronounced melt-layer formation and deformation, with measured molten pits 2 to 10 μm in diameter and melt-layer depths of up to 7 μm deep. Surface erosion rates for Ti 3 SiC 2 and Ti 2 AlC ranged from 80 to 775 μm/s and 85 to 470 μm/s, respectively. Both MAX phases exhibited extreme surface fracture and material ejection, with damage depths past 4 μm for Ti 2 AlC and 11 μm for Ti 3 SiC 2 . Lastly, SiC displayed the best performance, in one case surviving 15 consecutive electrothermal plasma exposures with an average erosion rate of about 29 μm/s and no surface fracturing. SiC erosion rates ranged from 23 to 128 μm/s.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Radial transport modeling of high density deuterium plasmas in proto-MPEX with the B2.5-Eirene code

High density (≥6 × 10 19 m –3 ), low temperature (2–6 eV) helicon discharges in the Prototype Material Plasma Exposure eXperiment (Proto-MPEX) are analyzed with the coupled multifluid plasma, kinetic neutrals code B2.5-Eirene. The interpretative analyses are constrained by data from multiple diagnostics, including Langmuir probes, Mach probes, filterscopes, infrared TV system, Thomson scattering, and baratrons. The objectives of the transport simulations include: investigation of the effects of heating, fueling, and plasma production; pumping, and assumed radial transport models on the calculated density and temperature distributions; plasma flow profiles and power balance. The primary objective in this report is to investigate the effects of the radial transport model in full plasma (the entire length of the plasma column in Proto-MPEX) data-constrained simulations. Results from three assumed forms of the radial transport coefficients are presented, including spatially constant, radially decreasing, and Bohm (D,χ ~T e /|B|). The results from each of the three transport coefficient sets agree qualitatively with the core (near axis) data. With the implicit T e dependence, the Bohm coefficients tend to decrease as functions of radius, although not as strongly as the centrally peaked set. The axial variation in the Bohm coefficients is largely due to the axial structure of the magnetic field. Furthermore, the agreement of the simulations and the diagnostic data with the Bohm set indicates that transport in the plasma column of Proto-MPEX is dominated by Bohm diffusion.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗