Heat Flux Mitigation Research Articles

Numerical study of potential heat flux mitigation effects in the TCV snowflake divertor

We report on EMC3-Eirene simulations of the plasma and neutral particle transport the TCV boundary layer of a series of snowflake (SF) equilibria characterized by the normalized poloidal flux coordinate of the secondary X-point x2. We refer to a snowflake plus (SF+) for , a snowflake minus (SF−) for and a single-null (SN) for . Four effects are identified that have the potential to mitigate the heat flux density at the outer strike point in a LFS SF−where x2 is located on the low field side of the primary X-point x1: (1) a scrape-off layer heat flux splitting, (2) an impurity radiation cloud forming at x2 (3) the increased connection length to the outer target and (4) increased transport between x1 and x2. The LFS SF− is thus expected to tolerate a larger power flux over the separatrix than a comparable SN configuration.

Particle-in-cell Simulation Concerning Heat-flux Mitigation Using Electromagnetic Fields

The Particle-in-Cell (PIC) method was used to study heat flux mitigation experiments with argon. In the experiment it was shown that a magnetic field allows to reduce the heat flux towards a target. PIC is well-suited for plasma simulation, giving the chance to get a better basic understanding of the underlying physics. The simulation demonstrates the importance of a self-consistent neutral-plasma description to understand the effect of heat flux reduction.

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak.

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

Analysis of the magnetohydrodynamic heat shield system for hypersonic vehicles

During hypersonic flight, the weakly-ionized plasma layer post shock can be utilized for flow control by externally applying a magnetic field. The Lorentz force, which is induced by the interaction between the ionized air and the magnetic field, decelerates the flow in the shock layer. Consequently, the thickness of the shock layer is increased and the convective heat flux can be mitigated. This so-called magnetohydrodynamic (MHD) heat shield system has been proved to be effective in heat flux mitigation by many researchers.Different from the dipole magnet conventionally used in previous researches on MHD heat shield, a normal columned solenoid-based MHD thermal protection system model is built in this paper. The present numerical analysis is mainly based on the low magneto-Reynolds MHD model, which neglects the induction magnetic field. Hall effect and the ion-slip effect are also neglected here because an insulating wall is assumed. With these hypothesis, a series of axisymmetric simulations on the flow field of Japanese Orbital Reentry Experimental Capsule (OREX) are performed to analyze the influence of different externally applied magnetic fields on the efficiency of MHD thermal protection. First, based on the dipole magnet field, the influence of magnetic induction density is analyzed. Second， differences between the efficiency of MHD thermal protection under three types of magnetic field, namely dipole magnet, solenoid magnet, and uniform magnet field are compared. Finally, the influence of the geometric parameters of solenoid magnet on the MHD thermal protection is analyzed. Results show that, saturation effect exists in the process of MHD heat flux mitigation and it confines the effectiveness of MHD heat shield system. Thermal protection capabilities under three types of magnetic field are ranked from weak to strong as dipole magnet, solenoid magnet, and uniform magnet field. Under the same magnetic induction intensity at the stagnation point, first， the increase of solenoid radius improves its effectiveness in MHD thermal protection; second, the influence of solenoid length on the efficiency of MHD thermal protection is weak, indicating that the solenoid length can be optimized with the remaining two factors, namely the exciting current density and the total weight of solenoid magnet. Finally, the closer distance between the solenoid and stagnation point has negative influence on MHD thermal protection for the stagnation and the shoulder area of the reentry capsule.

Modeling divertor concepts for spherical tokamaks NSTX-U and ST-FNSF

The compact nature of the spherical tokamak (ST) presents an economically attractive path to fusion commercialization, but concentrates power exhaust, threatening the integrity of plasma-facing components. To address this challenge, experimentally constrained divertor modeling in the National Spherical Torus Experiment (NSTX) is extrapolated to investigate divertor concepts for future ST devices. Analysis of NSTX Upgrade with UEDGE shows that the secondary snowflake X-point position can be adjusted for favorable neutral transport, enabling stable partial detachment at reduced core densities. For a notional ST-based Fusion Nuclear Science Facility, divertor concepts are identified that provide heat flux mitigation (<10 MW m−2) and low temperatures (<10 eV) compatible with high-Z targets. This research provides guidance for upcoming experiments and a basis for continued development of predictive capability for divertor performance in STs.

An overview of recent physics results from NSTX

The National Spherical Torus Experiment (NSTX) is currently being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive operation in high-performance plasmas, assessing plasma–material interface (PMI) solutions to handle the high heat loads expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection (CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet–Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Toroidal Alfvén eigenmode avalanches and higher frequency Alfvén eigenmodes can affect NB-driven current through energy loss and redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium–oxygen affinity, which increases deuterium pumping and reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding. Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing unstable regime. Two Alfvén eigenmode-driven fast ion transport models have been developed and successfully benchmarked against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned for mid-2015.

Energy deposition onto HL-2A divertor plates in ELMy H-mode discharges using infrared thermography

Using infrared (IR) thermography, power loads onto the divertor plates have been investigated in ELMy H-mode plasmas on HL-2A. In the ELMy H-mode discharges, ELMs are the largest contributors to the divertor target energy load. Analysis of energy balance shows that up to 45% of the energy losses are deposited onto the divertor targets during ELMs and about 30% are found as plasma radiation. Moreover, divertor heat flux mitigation has been achieved during an ELMy H-mode phase by using Supersonic Molecular Beam Injection (SMBI), characterized by a sharp increase of ELM frequency and a reduction in peak heat flux. The increased plasma radiation energy losses, especially the doubled plasma radiation in the divertor region, should be responsible for the reduction of integrated energy deposition onto divertor targets.

Multi‐Fluid Transport Modeling of NSTX Upgrade Standard and Snowflake Divertor Configurations

AbstractIn NSTX Upgrade (NSTX‐U), power exhaust is expected to challenge available material and heat removal technologies, which are limited to ∼10 MW/m2. Preliminary analysis of heat flux mitigation strategies, including the snowflake divertor configuration and radiative divertor operation, is performed with the multi‐fluid edge transport code, UEDGE. Divertor recycling between 95% and 99%, and power exhaust between 7 and 9 MW are explored. Compared to the standard divertor, the particular snowflake divertor studied here strongly reduces outer target heat flux, but tends to have higher (&gt;10 MW/m2) inner target heat flux. This suggests that snowflake divertor optimization should be pursued in future work. Both neon and argon impurity seeding scenarios mitigate high heat fluxes in both standard and snowflake configurations, but argon can cause radiative collapse of the core plasma at only 3% concentration. Divertor cryopumping is also considered, and modeling shows ∼50% more particle removal with the snowflake divertor. (© 2014 WILEY‐VCH Verlag GmbH &amp; Co. KGaA, Weinheim)

Divertor heat flux mitigation using Supersonic Molecular Beam Injection in type-III ELMy H mode plasmas of HL-2A tokamak

Abstract Divertor heat flux mitigation has been achieved using the Supersonic Molecular Beam Injection (SMBI) in type-III ELMy H-mode plasmas on the HL-2A tokamak. After an SMBI, ELM mitigation occurs and is characterized by an ELM frequency increase by a factor of 2–3 and an amplitude decrease by 30–40%. The corresponding confinement has a little degradation (stored energy drops is smaller than 7%). The burst rate of smaller-scale filament (density amplitude σ ) increases by 20–30%, and the divertor heat flux decreases by 40–50%, which is consistent with an increase in radiation power loss. This observation indicates that SMBI should be a good candidate for ELM mitigation.

Simulation of tokamak SOL and divertor region including heat flux mitigation by gas puffing

Two-dimensional (2D), scrape-off layer (SOL)-divertor transport simulations are performed using the integrated plasma-neutral-impurity code KTRAN developed at Seoul National University. Firstly, the code is applied to reproduce a National Spherical Torus eXperiment (NSTX) discharge by using the prescribed transport coefficients and the boundary conditions obtained from the experiment. The plasma density, the heat flux on the divertor plate, and the D α emission rate profiles from the numerical simulation are found to follow experimental trends qualitatively. Secondly, predictive simulations are carried out for the baseline operation mode in Korea Superconducting Tokamak Advanced Research (KSTAR) to predict the heat flux on the divertor target plates. The stationary peak heat flux in the KSTAR baseline operation mode is expected to be 6.5 MW/m2 in the case of an orthogonal divertor. To study the mitigation of the heat flux, we investigated the puffing effects of deuterium and argon gases. The puffing position is assumed to be in front of the strike point at the outer lower divertor plate. In the simulations, mitigation of the peak heat flux at the divertor target plates is found to occur when the gas puffing rate exceeds certain values, ∼1.0 × 1020 /s and ∼5.0 × 1018 /s for deuterium and argon, respectively. Multi-charged impurity transport is also investigated for both NSTX and KSTAR SOL and divertor regions.

Erratum to: Simulation of Tokamak SOL and Divertor Region Including Heat Flux Mitigation by Gas Puffing

Erratum to: Simulation of Tokamak SOL and Divertor Region Including Heat Flux Mitigation by Gas Puffing

Snowflake divertor configuration studies in National Spherical Torus Experiment

Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4–6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width λq was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5–1 MW/m2 (from 4–7 MW/m2 in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, Prad in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

Evaluation of heat and particle controllability on the JT-60SA divertor

The JT-60SA divertor design has been established on the basis of engineering requirements and physics analysis. Heat and particle fluxes under the full input power of 41MW can give severe heat loads on the divertor targets, while the allowable heat load is limited below 15MW/m2. Dependence of the heat flux mitigation on a D2 gas-puff is evaluated by SONIC simulations for high density (ne_ave∼1×1020m−3) high current plasmas. It is found that the peak heat load 10MW/m2 with dense (ned>4×1020m−3) and cold (Ted, Tid⩽1eV) divertor plasmas are obtained at a moderate gas-puff of Γpuff=15×1021s−1. Divertor plasmas are controlled from attached to detached condition using the divertor pump with pumping-speed below 100m3/s. In full non-inductive current drive plasmas with low density (ne_ave∼5×1019m−3), the reduction of divertor heat load is achieved with the Ar injection.

Aerodynamic Heating of Reentry Body Equipped with Onboard-Surface Hall Magnetohydrodynamic Generator

The generator performance and the aerodynamic heating of a reentry body equipped with an onboard-surface Hall-type magnetohydrodynamic power generator are examined under some different anode electrode configuration cases by means of two-dimensional numerical simulation. The reentry body with a nose radius of 1.35 m has a pair of electrodes on the wall surface. Flight altitude and velocity are set to 60 km and 5.6 km/s, respectively. The strength of an applied magnetic field is about 0.5 T. Numerical results demonstrate that the onboard-surface Hall-type magnetohydrodynamic power generator can effectively mitigate the wall heat flux at the stagnation point as well as extract the electrical power exceeding 1 MW when a ring-shaped anode electrode is placed on the wall surface away from the stagnation point. On the other hand, if a bowl-shaped anode electrode is placed to cover a wide wall surface including the stagnation point, an effective mitigation of the wall heat flux at the stagnation point by the magnetohydrodynamic interaction becomes impossible, although the electrical power exceeding 1 MW can be obtained. This is because a Hall electric field, strong enough to induce the strong magnetohydrodynamic interaction, cannot be obtained near the stagnation point, due to the electrically coupled circuit between the plasma and the bowl-shaped anode electrode.

In Situ Impact Analysis of Very High Heat Flux Transients on Nonlinear p-n Diode Characteristics and Mitigation Using On-Chip Single- and Two-Phase Microfluidics

The reliable operation of advanced integrated circuits (ICs) is impacted by the presence of randomly generated heat-flux transients. These transients can induce severe nonlinearities in the device response and can result in the degradation of ohmic contacts for p-n diodes as well as gate dielectric for metal-oxide field-effect transistors. In this paper, we focus on the impact and mitigation of steady-state and transient heat flux, with levels ranging from 0 to 250 W/cm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> . The impact analysis exercise focused on the heat-flux-induced nonlinear variation of the p-n diode power exponent factor (alpha). For the steady-state heat-flux scenario, the alpha- <i xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">V</i> <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">FB</sub> (forward bias voltage) characteristics yield a maxima point (alpha <i xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">m</i> ) which is observed to decrease monotonically with an increase in diode temperature. The alpha <i xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">m</i> value is further useful for deriving other parameters, such as saturation current ( <i xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">Is</i> ), series resistance ( <i xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">Rs</i> ), and the diode ideality factor (beta) . The transient heat-flux scenario yields a transient power exponent factor (alpha <i xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">T</i> , maxima-alpha <i xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">mT</i> ) which is distinct from the steady-state case. The alpha <i xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">mT</i> shows an inverse dependence on instantaneous diode temperature. Transient mitigation is evident when the diode power exponent parameter is recovered under the application of single-phase as well as two-phase on-chip fluid flows. Finally, while our primary focus has been on transient mitigation, we have also looked at the feasibility of localizing transient heat sources based on temperature profiles generated using an on-chip distributed resistance temperature detector sensor array. In real-life ICs, the systematic localization and characterization of heat sources will be of interest in order to provide information on the origin of transients, thus leading to modifications in circuit design or process integration steps.

Divertor heat flux mitigation in high-performance H-mode discharges in the National Spherical Torus Experiment

Experiments conducted in high-performance 1.0 and 1.2 MA 6 MW NBI-heated H-mode discharges with a high magnetic flux expansion radiative divertor in NSTX demonstrate that significant divertor peak heat flux reduction and access to detachment may be facilitated naturally in a highly shaped spherical torus (ST) configuration. Improved plasma performance with high βt = 15–25%, a high bootstrap current fraction fBS = 45–50%, longer plasma pulses and an H-mode regime with smaller ELMs has been achieved in the strongly shaped lower single null configuration with elongation κ = 2.2–2.4 and triangularity δ = 0.7–0.8. Divertor peak heat fluxes were reduced from 6–12 to 0.5–2 MW m−2 in ELMy H-mode discharges using the inherently high magnetic flux expansion fm = 15–25 and the partial detachment of the outer strike point at several D2 injection rates. A good core confinement and pedestal characteristics were maintained, while the core carbon concentration and the associated Zeff were reduced. The partially detached divertor regime was characterized by an increase in divertor radiated power, a reduction in ion flux to the plate and a large neutral compression ratio. Spectroscopic measurements indicated the formation of a high-density, low-temperature region adjacent to the outer strike point, where substantial increases in the volume recombination rate and C II, C III emission rates were measured.

Experimental Verification of Heat-Flux Mitigation by Electromagnetic Fields in Partially-Ionized-Argon Flows

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Divertor heat flux mitigation in the National Spherical Torus Experiment

Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4–6MWm−2to0.5–2MWm−2 in small-ELM 0.8–1.0MA, 4–6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

Numerical Analysis of Reentry Trajectory Coupled with Magnetohydrodynamics Flow Control

Effects of the magnetohydrodynamics flow control on Earth reentry flight characteristics are examined by a coupled numerical analysis of the reentry flight trajectory and the magnetohydrodynamics flows with thermochemical reactions. Initial flightaltitude andvelocity ofabluntbody withanoseradius of1.35 mare 75.3km and7:2 km=s,respectively.Numericalresultsshowthatathighaltitudessuchasabout75km,themainfactorofthe mitigation of wall heat flux by the magnetohydrodynamics flow control is the increase of the thickness of the shock layer by applying the magnetic field. At low altitudes such as about 55 km, on the other hand, the main factor of its mitigation isthedecrease of flight velocitydueto thechangeof flighttrajectory byapplyingthe magnetic field.It can also befound from a rough estimation thatabout 75%of its mitigation amount at the altitude of about 60 km, where the aerodynamic heating is severest, is attributed to the decrease of flight velocity by applying the magnetic field.