Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Fusion Power”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Ion collector design for an energy recovery test proposal with the negative ion source NIO1

Commercial viability of thermonuclear fusion power plants depends also on minimizing the recirculation power used to operate the reactor. The neutral beam injector (NBI) remains one of the most important method for plasma heating and control. For the future fusion power plant project DEMO, a NBI wall plug efficiency at least of 0.45 is required, while efficiency of present NBI project is about 0.25. The D{sup −} beam from a negative ion source is partially neutralized by a gas cell, which leaves more than 40% of energy in residual beams (D{sup −} and D{sup +}), so that an ion beam energy recovery system can significantly contribute to optimize efficiency. Recently, the test negative ion source NIO1 (60 keV, 9 beamlets with 15 mA H{sup −} each) has been designed and built at RFX (Padua) for negative ion production efficiency and the beam quality optimization. In this paper, a study proposal to use the NIO1 source also for a beam energy recovery test experiment is presented and a preliminary design of a negative ion beam collector with simulations of beam energy recovery is discussed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Overview of JET results for optimising ITER operation

The JET 2019–2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major neutral beam injection upgrade providing record power in 2019–2020, and tested the technical and procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle (α) physics in the coming D–T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed shattered pellet injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design and operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D–T benefited from the highest D–D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

An Overview of Recent Standard and Accelerated Molecular Dynamics Simulations of Helium Behavior in Tungsten

One of the most critical challenges for the successful adoption of nuclear fusion power corresponds to plasma-facing materials. Due to its favorable properties in this context (low sputtering yield, high thermal conductivity, high melting point, among others), tungsten is a leading candidate material. Nevertheless, tungsten is affected by the plasma and fusion byproducts. Irradiation by helium nuclei, in particular, strongly modifies the surface structure by a synergy of processes, whose origin is the nucleation and growth of helium bubbles. In this review, we present recent advances in the understanding of helium effects in tungsten from a simulational approach based on accelerated molecular dynamics, which emphasizes the use of realistic parameters, as are expected in experimental and operational fusion power conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2014

RF H and CD systems for DEMO - Challenges and opportunities

The aim of driving a sufficient amount of plasma current with an appropriate radial current density profile is considered as one of the key challenges for a tokamak fusion power plant in steady state operation. Furthermore, efficient heating to enable transition to regime of enhanced confinement and to achieve breakeven plasma temperatures as well as MHD control and plasma breakdown assistance are required. In the framework of the EFDA Power Plant Physics and Technology (PPPT) activities, the ability of the Electron cyclotron (EC), Ion Cyclotron (IC) and Lower Hybrid (LH) systems to fulfil these requirements, was studied for a demonstration fusion power plant (DEMO). As boundary condition, a 1D description of the plasma for a pulsed DEMO based on system code studies combined with transport analysis was developed. The predicted 1D plasma parameters were used to calculate the current drive (CD) efficiency of each system and eventually optimised it. As an example, the EC current drive efficiency could be increased strongly by top launch compared to equatorial launch at least by a factor of two. For the IC system, two possible windows of operation for standard and higher frequencies were highlighted, whereby again top launch leads to higher CD-efficiencies. The efficiencies predicted for DEMO for the RF current drive systems will be presented. Finally, gaps in the feasibility of RF systems under DEMO relevant conditions will be identified.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

Experimental Activities with the IELLLO Lithium-Lead Facility

Fusion power offers the prospect of an almost inexhaustible source of energy for future generations. Although this may be true, it also presents so far serious scientific and engineering challenges. Breeding Blanket is one of key components of future fusion reactors, since it directly involves tritium breeding and energy extraction, both of them critical to development of fusion power. To deal with the development of Breeding Blankets, ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) designed and built a lead lithium loop (IELLLO: Integrated European Lead Lithium Loop): the experiment aims to support R and D activities for the conceptual and preliminary design of the European test blanket systems. IELLLO was completed, installed and operated at ENEA Brasimone Research Centre. The purpose of this work is to describe the experimental activities that have been carried out concerning the qualification of the facility, and relative instrumentations, for lead lithium applications. A series of thermo-hydraulic investigations on IELLLO facility has been elaborated: lead lithium pressure transducers, heat exchanger, permanent magnet pumping system, mass flow meters and air cooler have been tested. The absolute pressure meter in contact with Pb-16Li was qualified in pressure range up to 10 bar and temperature up to 400 deg. C. The installation of new kinds of Pb-16Li differential pressure transducers will allow to check the pump characteristic curve and accurately calculate valves pressure drops. These factors could give a better information for the thermo-hydraulic qualification of the facility. Vortex Mass Flow Meter has shown good results in terms of accuracy. The performance has been validated by thermal balance with a dedicated heating system. Nevertheless, the installations of different types of mass flow meters will enable to test IELLLO at ITER relevant conditions, and at lower mass flow rates than those tested up to now. The new pumping system with a PM Pump, was qualified for the first time in Pb-16Li, and the procedure for the operation and maintenance of the pump was developed. The pump is able to supply a mass flow rate in the range from 1.2 to 2.4 kg/s with a temperature of Pb-16Li up to 550 deg. C in the test section. Significant results have been obtained also for the heat exchanger, which has shown high efficiencies, even more than expected. Likewise, the Air Cooler has been verified to be able to cool Pb-16Li in a relatively short time, even at the worst temperature and mass flow conditions. (authors)

42 ENGINEERING↗
DOE OSTI2026

Valuing Maintenance Strategies for Fusion Plants as Part of a Future Electricity Grid

Scheduled maintenance is likely to be lengthy and therefore consequential for the economics of fusion power plants. The maintenance strategy that maximizes the economic value of a plant depends on internal factors such as the cost and durability of the replaceable components, the frequency and duration of the maintenance blocks, and the external factors of the electricity system in which the plant operates. Here, this paper examines the value of fusion power plants with various maintenance properties in a decarbonized United States Eastern Interconnection circa 2050. Seasonal variations in electricity supply and demand mean that certain times of year, particularly spring to early summer, are best for scheduled maintenance. Seasonality has two important consequences. First, the value of a plant can be 15% higher than what one would naively expect if value were directly proportional to its availability. Second, in some cases, replacing fractions of a component in shorter maintenance blocks spread over multiple years is better than replacing it all at once during a longer outage, even through the overall availability of the plant is lower in the former scenario.

power plant, maintenance, seasonality, electricity↗
DOE OSTI2023

Control-oriented core-SOL-divertor model to address integrated burn and divertor control challenges in ITER

The real-time regulation of a burning plasma’s temperature and density, or burn control, will be necessary to produce high fusion power in future tokamaks like ITER. This is made more challenging due to the plasma’s nonlinear characteristics and the interdependence between the core-plasma and edge-plasma regions. For example, a raising plasma temperature leads to increasing reactivity and therefore to more alpha-particle heating, which further increases temperature. Furthermore, a raise of the fusion power increases the heat flow through the scrape-off-layer (SOL), which can compromise the integrity of the divertor without proper safeguards. For control design, a model-based approach is attractive because it can directly incorporate the nonlinear, coupled, burning-plasma dynamics into the design. To facilitate this design approach, a control-oriented core-SOL-divertor (CSD) model is presented in this work. In this CSD model, a core-plasma model captures the nonlinear dynamics of the core’s density and temperature, and a SOL-divertor model defines the plasma conditions at the separatrix and divertor including the heat load on the target plates. The core-plasma and SOL-divertor models are coupled through the exchange of various variables. In particular, the SOL-divertor model yields the separatrix temperature and the influx of recycled particles into the core-plasma. Further, these variables influence the power and particle balances captured by the core-plasma model. In return, the core-plasma model determines the intensity of the heat and particles fluxes across the separatrix, and this outflow strongly impacts the SOL-divertor model. Therefore, the power and density of the core-plasma, which can be readily modulated through external heating systems and pellet injection, can be viewed as control knobs for the SOL-divertor region in addition to the gas puffing. In simulations of the CSD model, it is demonstrated how external actuation can be utilized to meet burn control and divertor control objectives simultaneously.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

TGLF Recalibration for ITER Standard Case Parameters FY2015: Theory and Simulation Performance Target Final Report

This work was motivated by the observation, as early as 2008, that GYRO simulations of some ITER operating scenarios exhibited nonlinear zonal-flow generation large enough to effectively quench turbulence inside r /a ~ 0.5. This observation of flow-dominated, low-transport states persisted even as more accurate and comprehensive predictions of ITER profiles were made using the state-of-the-art TGLF transport model. This core stabilization is in stark contrast to GYRO-TGLF comparisons for modern-day tokamaks, for which GYRO and TGLF are typically in very close agreement. So, we began to suspect that TGLF needed to be generalized to include the effect of zonal-flow stabilization in order to be more accurate for the conditions of reactor simulations. While the precise cause of the GYRO-TGLF discrepancy for ITER parameters was not known, it was speculated that closeness to threshold in the absence of driven rotation, as well as electromagnetic stabilization, created conditions more sensitive the self-generated zonal-flow stabilization than in modern tokamaks. Need for nonlinear zonal-flow stabilization: To explore the inclusion of a zonal-flow stabilization mechanism in TGLF, we started with a nominal ITER profile predicted by TGLF, and then performed linear and nonlinear GYRO simulations to characterize the behavior at and slightly above the nominal temperature gradients for finite levels of energy transport. Then, we ran TGLF on these cases to see where the discrepancies were largest. The predicted ITER profiles were indeed near to the TGLF threshold over most of the plasma core in the hybrid discharge studied (weak magnetic shear, q > 1). Scanning temperature gradients above the TGLF power balance values also showed that TGLF overpredicted the electron energy transport in the low-collisionality ITER plasma. At first (in Q3), a model of only the zonal-flow stabilization (Dimits shift) was attempted. Although we were able to construct an ad hoc model of the zonal flows that fit the GYRO simulations, the parameters of the model had to be tuned to each case. A physics basis for the zonal flow model was lacking. Electron energy transport at short wavelength: A secondary issue – the high-k electron energy flux – was initially assumed to be independent of the zonal flow effect. However, detailed studies of the fluctuation spectra from recent multiscale (electron and ion scale) GYRO simulations provided a critical new insight into the role of zonal flows. The multiscale simulations suggested that advection by the zonal flows strongly suppressed electron-scale turbulence. Radial shear of the zonal E×B fluctuation could not compete with the large electron-scale linear growth rate, but the k x -mixing rate of the E×B advection could. This insight led to a preliminary new model for the way zonal flows saturate both electron- and ion-scale turbulence. It was also discovered that the strength of the zonal E×B velocity could be computed from the linear growth rate spectrum. The new saturation model (SAT1), which replaces the original model (SAT0), was fit to the multiscale GYRO simulations as well as the ion-scale GYRO simulations used to calibrate the original SAT0 model. Thus, SAT1 captures the physics of both multiscale electron transport and zonal-flow stabilization. In future work, the SAT1 model will require significant further testing and (expensive) calibration with nonlinear multiscale gyrokinetic simulations over a wider variety of plasma conditions – certainly more than the small set of scans about a single C-Mod L-mode discharge. We believe the SAT1 model holds great promise as a physics-based model of the multiscale turbulent transport in fusion devices. Correction to ITER performance predictions: Finally, the impact of the SAT1model on the ITER hybrid case is mixed. Without the electron-scale contribution to the fluxes, the Dimits shift makes a significant improvement in the predicted fusion power as originally posited. Alas, including the high-k electron transport reduces the improvement, yielding a modest net increase in predicted fusion power compared to the TGLF prediction with the original SAT0 model.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Comparisons between tokamak fueling of gas puffing and supersonic molecular beam injection in 2D simulations

Plasma fueling with high efficiency and deep injection is very important to enable fusion power performance requirements. It is a powerful and efficient way to study neutral transport dynamics and find methods of improving the fueling performance by doing large scale simulations. Two basic fueling methods, gas puffing (GP) and supersonic molecular beam injection (SMBI), are simulated and compared in realistic divertor geometry of the HL-2A tokamak with a newly developed module, named trans-neut, within the framework of BOUT++ boundary plasma turbulence code [Z. H. Wang et al., Nucl. Fusion 54, 043019 (2014)]. The physical model includes plasma density, heat and momentum transport equations along with neutral density, and momentum transport equations. Transport dynamics and profile evolutions of both plasma and neutrals are simulated and compared between GP and SMBI in both poloidal and radial directions, which are quite different from one and the other. It finds that the neutrals can penetrate about four centimeters inside the last closed (magnetic) flux surface during SMBI, while they are all deposited outside of the LCF during GP. It is the radial convection and larger inflowing flux which lead to the deeper penetration depth of SMBI and higher fueling efficiency compared to GP.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Follow the power—pathways to steady-state tokamak reactors

Abstract Diagramming pathways of dimensionless power is a potent method for extrapolating between operating points on present-day tokamaks and future burning plasma devices. The heat transport power, current drive power, H-mode threshold power and other ‘plasma physics’ powers can be expressed in dimensionally correct (or normalized) form as Pa 3/4 , where P is the power and a is the plasma minor radius, with the relative gyroradius ( ρ * ) dependence ranging from gyro-Bohm-like for transport ( P a 3 / 4 ∝ ρ * − 3 / 2 ), Bohm-like for current drive ( P a 3 / 4 ∝ ρ * − 5 / 2 ) and worse than Bohm-like for H-mode threshold ( P a 3 / 4 ∝ ρ * − 3 ). The D–T fusion power cannot be normalized in the same fashion since it is governed by nuclear physics, but at fixed B T it scales like P a 3 / 4 ∝ ρ * − 9 / 2 . Other ‘mixed physics’ powers can be incorporated into the dimensionless power framework by holding B T fixed in the same manner. Diagramming these dimensionless powers vs ρ * shows how the pathway to a steady-state reactor can be optimized relative to various operational boundaries. Using a steady-state hybrid discharge with β N = 3.2 from DIII-D as the starting point, a multi-parameter optimization finds an attractive pathway to steady-state operation on ITER using 76 MW of current drive power (fusion gain of Q fus = 8), along with a pathway to Q fus = 20 in a JET-sized steady-state reactor with B T = 10 T.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Exploration of the Super H-mode regime on DIII-D and potential advantages for burning plasma devices

A new high pedestal regime (“Super H-mode”) has been predicted and accessed on DIII-D. Super H-mode was first achieved on DIII-D using a quiescent H-mode edge, enabling a smooth trajectory through pedestal parameter space. By exploiting Super H-mode, it has been possible to access high pedestal pressures at high normalized densities. While elimination of Edge localized modes (ELMs) is beneficial for Super H-mode, it may not be a requirement, as recent experiments have maintained high pedestals with ELMs triggered by lithium granule injection. Simulations using TGLF for core transport and the EPED model for the pedestal find that ITER can benefit from the improved performance associated with Super H-mode, with increased values of fusion power and gain possible. Similar studies demonstrate that the Super H-mode pedestal can be advantageous for a steady-state power plant, by providing a path to increasing the bootstrap current while simultaneously reducing the demands on the core physics performance.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

CORSICA modelling of ITER hybrid operation scenarios

The hybrid operating mode observed in several tokamaks is characterized by further enhancement over the high plasma confinement (H-mode) associated with reduced magneto-hydro-dynamic (MHD) instabilities linked to a stationary flat safety factor ($q$ ) profile in the core region. The proposed ITER hybrid operation is currently aiming at operating for a long burn duration (>1000 s) with a moderate fusion power multiplication factor, $Q$ , of at least 5. This paper presents here candidate ITER hybrid operation scenarios developed using a free-boundary transport modelling code, CORSICA, taking all relevant physics and engineering constraints into account. The ITER hybrid operation scenarios have been developed by tailoring the 15 MA baseline ITER inductive H-mode scenario. Accessible operation conditions for ITER hybrid operation and achievable range of plasma parameters have been investigated considering uncertainties on the plasma confinement and transport. ITER operation capability for avoiding the poloidal field coil current, field and force limits has been examined by applying different current ramp rates, flat-top plasma currents and densities, and pre-magnetization of the poloidal field coils. Various combinations of heating and current drive (H&CD) schemes have been applied to study several physics issues, such as the plasma current density profile tailoring, enhancement of the plasma energy confinement and fusion power generation. A parameterized edge pedestal model based on EPED1 added to the CORSICA code has been applied to hybrid operation scenarios. Finally, fully self-consistent free-boundary transport simulations have been performed to provide information on the poloidal field coil voltage demands and to study the controllability with the ITER controllers.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Density of Additively-Manufactured, 316L SS Parts Using Laser Powder-Bed Fusion at Powers Up to 400W

Selective laser melting is a powder-based, additive-manufacturing process where a threedimensional part is produced, layer by layer, by using a high-energy laser beam to fuse the metallic powder particles. A particular challenge in this process is the selection of appropriate process parameters that result in parts with desired properties. In this study, we describe an approach to selecting parameters for high density (>99%) parts using 316L stainless steel. Though there has been significant success in achieving near-full density for 316L parts, this work has been limited to laser powers <225W. We discuss how we can exploit prior knowledge, design of computational experiments using a simple model of laser melting, and single-track experiments to determine the process parameters for use at laser powers up to 400W. Our results show that, at higher power values, there is a large range of scan speeds over which the relative density remains >99%, with the density reducing rapidly at high speeds due to insufficient melting, and less rapidly at low speeds due to the effect of voids created as the process enters keyhole mode.

316L stainless teel↗
DOE OSTI2011

UCI Final Report for SciDAC GPS-TTBP (3/1/2008-2/28/2011) Gyrokinetic Particle Simulation of Turbulent Transport in Burning Plasmas

“We propose to further develop the global, particle-in-cell, gyrokinetic toroidal code (GTC) for simulations of plasma turbulence and transport, and to advance understanding and our ability to control radial transport of heat, momentum, and particles in magnetically confined plasmas with temperatures up to those required for nuclear fusion power production regimes called burning plasmas. Building on the excellent physics capabilities and high performance computing, GTC will be extensively applied to simulate experimentally relevant parameter regimes including electromagnetic fluctuations, kinetic electrons, multiple ion species, realistic collision operators, and shaped plasmas. Collaborative code development will be pursued to accelerate implementations of new physics capabilities including long time steady state simulation with profile evolution, and core-edge coupling. The new simulation capabilities and access to tera- and petascale computers will allow us to address new physics and parameter regimes important for burning plasma experiments such as International Thermonuclear Experimental Reactor (ITER).” Among the three key goals and objectives (code development, comprehensive physics simulation, and core-edge coupling), we have achieved fully the first two goals, namely, 1. “to further develop the global, particle-in-cell, gyrokinetic toroidal code (GTC) for simulations of plasma turbulence and transport, and to advance understanding and our ability to control radial transport of heat, momentum, and particles in magnetically confined plasmas with temperatures up to those required for nuclear fusion power production regimes called burning plasmas.” 2. “GTC will be extensively applied to simulate experimentally relevant parameter regimes including electromagnetic fluctuations, kinetic electrons, multiple ion species, realistic collision operators, and shaped plasmas.” Regarding the third objective, 3. “Collaborative code development will be pursued to accelerate implementations of new physics capabilities including long time steady state simulation with profile evolution, and core-edge coupling.” While we have productively engaged in collaborative code development and made satisfactory progress on implementing new physics including long time steady state simulation with profile evolution, we have not achieved the goal of core-edge coupling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Self-consistent core-pedestal ITER scenario modeling

The purpose of these integrated ITER simulations is to identify dependencies that can impact the performance of ITER. The dependence of fusion power production, temperature and density pedestals, and core profiles on varying magnetic-q, edge density fueling strength, neoclassical transport, magnetic field strength, and alpha heating, are examined. It is noted that ion modes usually dominate the interior, while electron modes dominate the barrier. The electron mode without flow shear is substantially stronger. The q increase tends to cause more kinetic ballooning modes at the edge and the q decrease means that the peeling is getting worse, but the ballooning is getting better. The large edge q tends to provide hollow density while small edge q gives normal density profile. This is consistent with the observation that the density of particles is inversely proportional to the density of the plasma current. The rise in the source of the edge particle increases the density of the edge and the reaction rate close to the edge increases temperature fluxes, resulting in weak pedestal barriers. When neoclassical transport is turned off or the strength of B-field is increased, the pronounced edge barriers are identified. In addition, a distinct effect on the magnetohydrodynamic (MHD) modes is observed due to the increased B-field. The interesting aspect is that the barrier collapses about q = 2 as the stability of MHD has decreased to replace the barrier with a continuous slope but with roughly the same fusion power. This would actually be good for a reactor. The slope of the H-mode pedestal is found to be reduced due to the alpha heating. It is expected that the problem of edge-localized modes causing damage to the first wall will be eased. The effects of circular, elongated, and general geometry on plasma profiles are also compared and contrasted. The circular geometry has a hollow density profile and no edge or internal thermal transport barrier (ETB or ITB). However, both ETBs and ITBs can be found in the elongated geometry. In general geometry, there is a weaker ETB, comparable ITBs, but a higher edge particle transport barrier than in elongated geometry. Higher density near the edge, on the other hand, causes more wall erosion, so elongated geometry might be the best option.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Multispecies density peaking in gyrokinetic turbulence simulations of low collisionality Alcator C-Mod plasmas

Peaked density profiles in low-collisionality AUG and JET H-mode plasmas are probably caused by a turbulently driven particle pinch, and Alcator C-Mod experiments confirmed that collisionality is a critical parameter. Density peaking in reactors could produce a number of important effects, some beneficial, such as enhanced fusion power and transport of fuel ions from the edge to the core, while others are undesirable, such as lower beta limits, reduced radiation from the plasma edge, and consequently higher divertor heat loads. Fundamental understanding of the pinch will enable planning to optimize these impacts. We show that density peaking is predicted by nonlinear gyrokinetic turbulence simulations based on measured profile data from low collisionality H-mode plasma in Alcator C-Mod. Multiple ion species are included to determine whether hydrogenic density peaking has an isotope dependence or is influenced by typical levels of low-Z impurities, and whether impurity density peaking depends on the species. We find that the deuterium density profile is slightly more peaked than that of hydrogen, and that experimentally relevant levels of boron have no appreciable effect on hydrogenic density peaking. The ratio of density at r/a = 0.44 to that at r/a = 0.74 is 1.2 for the majority D and minority H ions (and for electrons), and increases with impurity Z: 1.1 for helium, 1.15 for boron, 1.3 for neon, 1.4 for argon, and 1.5 for molybdenum. The ion temperature profile is varied to match better the predicted heat flux with the experimental transport analysis, but the resulting factor of two change in heat transport has only a weak effect on the predicted density peaking.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Investigating the Potential and Limitations of Fast Scanning Calorimetry to Simulate the Laser Power Bed Fusion Coalescence Process

Laser-Based Powder Bed Fusion (PBF-LB) • PBF-LB is part of one of seven additive manufacturing techniques • Layer wise printing that selectively coalesces discrete powder particles into a solid object o Advantages: Low anisotropy, use of engineering grade materials o Weaknesses: lack of method standardization, challenges in controlled post print quality

Blackman, Malik A. [Georgia Inst. of Technology, A↗