Heat Deposition Profile Research Articles

Dose voxel resolution determination for Monte Carlo electron beam simulation in thin targets.

Monte Carlo particle simulation has become the primary tool for designing low-energy miniature x-ray tubes due to the difficulties of physically prototyping these devices and characterizing their radiation fields. Accurate simulation of electronic interactions within their targets is necessary for modelling both photon production and heat transfer. Voxel-averaging can conceal hot spots in the target heat deposition profile that can threaten the integrity of the tube. This research seeks a computationally-efficient method of estimating voxel-averaging error in energy deposition simulations of electron beams penetrating thin targets to inform the appropriate scoring resolution for a desired accuracy level. An analytical model to estimate voxel-averaging along the target depth was developed and compared to results from Geant4 via its wrapper, TOPAS. A 200 keV planar electron beam was simulated to impinge tungsten targets of thicknesses between 1.5- and 12.5- . For each target, the model was used to calculate the energy deposition ratio between voxels of varying sizes centered on the longitudinal midpoint of the target. Model-calculated ratios were compared to simulation outputs to gauge the model's accuracy. Then, the model was used to approximate the error between the point value of electron energy deposition and a voxel-based measurement. The model underestimates error to within 5% for targets less than 7.5- in thickness with increasing error for greater thicknesses. For the 1.5- target, calculations of the point-vs.-voxel energy deposition show an 11% averaging effect between the midpoint and a 1.5- voxel. Energy deposition profiles along the target depth were also calculated in the Monte Carlo for reference. A simple analytical model was developed with reasonable accuracy to guide Monte Carlo users in estimating the appropriate depth-voxel size for thin-target x-ray tube simulations. This methodology can be adapted for other radiological contexts to increase robustness in point-value estimations. This article is protected by copyright. All rights reserved.

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.

Comparison of private flux region instability in conventional and super-X divertor configurations

Understanding turbulence in the divertor leg of tokamaks is essential to predict the heat deposition profile on the divertor plate. This in turn is important for evaluating advanced divertor configurations, such as the super-X divertor. Within the divertor region, the private flux region is of interest because it is relatively unaffected by turbulence extending from the outboard midplane, so instabilities in this region could have a particularly pronounced effect on transport. These instabilities are modeled using the Arbitrary Topology Equation Reader (ArbiTER) eigenvalue code. Eigenmodes are examined further by comparing physics models to determine the fundamental mechanisms behind their formation and quantifying the effect of individual terms. This analysis is conducted on both conventional and super-X divertors to compare these effects. The resulting analysis reveals the presence of a geodesic curvature driven instability that is significantly more pronounced in the super-X configuration.

Liquid metal MHD flow influence on heat transfer phenomena in fusion reactor blankets

DEMO reactor is projected as a technological demonstrator and a component test reactor for several key systems such as the breeding blanket. The dual coolant lithium lead (DCLL) is one promising breeding blanket candidate that performs the functions of shielding, part of the cooling and breeding with the same liquid metal fluid. PbLi is also in charge of transporting the bred tritium to the Tritium Extraction System. Structural cooling of the blanket walls is performed by pressurized helium flow which also transfers heat to the balance of plant. Keeping a proper distribution of thermal loads between the two coolants is a key aspect to reach a plant high efficiency. The thermal power transferred from the liquid metal to the helium channels depends on the flow regime. While liquid metal flows under the magnetic field, the magnetohydrodynamic (MHD) phenomena will occur, influencing the flow regime and the velocity profile. The heat is volumetrically deposited in the liquid metal coolant due to the interaction between the highly energetic neutrons originated in the plasma and the coolant. The heat deposition profile is roughly exponential causing temperature differences across the same channel cross section. Temperature gradients induce buoyant effects altering the expected MHD profile and affecting the heat transfer rate from the liquid metal to the surrounding walls cooled with helium. It is of utmost interest to identify the effects of the main flow parameters on the heat transfer coefficient under such complex conditions. The work studies computationally a domain consisting of a liquid region and a surrounding solid region electrically and thermally coupled using latest EU DCLL design characteristics. A fully developed flow is assumed where buoyancy is modelled under the Boussinesq approximation. Joule effect has been considered negligible compared to the neutron deposition. The influence of the buoyant MHD phenomena to the heat transfer coefficient is analysed parametrically with a CFD solver implemented in OpenFOAM®. The parameters investigated here are the Reynolds, Hartmann and Grashof numbers, the Grashof ratio, and the wall conductance ratio (cw).

Flux-driven integrated modelling of main ion pressure and trace tungsten transport in ASDEX Upgrade

Neoclassical and turbulent heavy impurity transport in tokamak core plasmas are determined by main ion temperature, density and toroidal rotation profiles. Thus, in order to understand and prevent experimental behaviour of W accumulation, flux-driven integrated modelling of main ion heat and particle transport over multiple confinement times is a vital prerequisite. For the first time, the quasilinear gyrokinetic code QuaLiKiz is applied for successful predictions of core kinetic profiles in an ASDEX Upgrade H-mode discharge in the turbulence dominated region within the integrated modelling suite JETTO. Neoclassical contributions are calculated by NCLASS; auxiliary heat and particle deposition profiles due to NBI and ECRH are prescribed from previous analysis with TRANSP. Turbulent and neoclassical contributions are insufficient in explaining main ion heat and particle transport inside the q = 1 surface, necessitating the prescription of further transport coefficients to mimic the impact of MHD activity on central transport. The ion to electron temperature ratio at the simulation boundary at stabilizes ion scale modes while destabilizing ETG modes when significantly exceeding unity. Careful analysis of experimental measurements using Gaussian process regression techniques is carried out to explore reasonable uncertainties. In following trace W impurity transport simulations performed with additionally NEO, neoclassical transport under consideration of poloidal asymmetries alone is found to be insufficient to establish hollow central W density profiles. Reproduction of these conditions measured experimentally is found possible only when assuming the direct impact of a saturated MHD mode on heavy impurity transport.

Particle and heat flux diagnostics on the C-2W divertor electrodes.

A suite of diagnostics was developed to measure particle and heat fluxes arriving at the divertor electrodes of the C-2W experiment at TAE Technologies. The divertor electrodes consist of 4 concentric rings, each equipped with a bolometer, electrostatic energy analyzer, and thermocouple mounted at two opposing azimuthal locations. These probes provide measurements of the power flux to the divertor electrodes as well as measurements of the ion current density, ion energy distribution, and total energy deposition. The thermocouples also provide calibration points for inferring the heat deposition profile via thermographic imaging of the electrodes with a fast infrared camera. The combined measurements enable the calculation of the energy lost per escaping electron/ion pair, which is an important metric for understanding electron heat transport in the open field lines that surround the field-reversed configuration plasma in C-2W.

Parametric dependencies of the experimental tungsten transport coefficients in ICRH and ECRH assisted ASDEX Upgrade H-modes

The profiles of the W transport coefficients have been experimentally calculated for a large database of identical ASDEX Upgrade H-mode discharges where only the radio-frequency (RF) power characteristics have been varied [Angioni et al., Nucl. Fusion 57, 056015 (2017)]. Central ion cyclotron resonance heating (ICRH) in the minority heating scheme has been compared with central and off-axis electron cyclotron resonance heating (ECRH), using both localized and broad heat deposition profiles. The transport coefficients have been calculated applying the gradient-flux relation to the evolution of the intrinsic W density in-between sawtooth cycles as measured using the soft X-ray diagnostic. For both ICRH and ECRH, the major player in reducing the central W density peaking is found to be the reduction of inward pinch and, in the case of ECRH, the rise of an outward convection. The impurity convection increases, from negative to positive, almost linearly with RF-power, while no appreciable changes are observed in the diffusion coefficient, which remains roughly at neoclassical levels independent of RF power or background plasma conditions. The ratio vW/DW is consistent with the equilibrium ∇nW/nW prior to the sawtooth crash, corroborating the separate estimates of diffusion and convection. These experimental findings are slightly different from previous results obtained analysing the evolution of impurity injections over many sawtooth cycles. Modelling performed using the drift-kinetic code NEO and the gyro-kinetic code GKW (assuming axisymmetry) overestimates the diffusion coefficient and underestimates the experimental positive convection. This is a further indication that magneto-hydrodynamic/neoclassical models accounting for 3D effects may be needed to characterize impurity transport in sawtoothing tokamak plasmas.

Studies of electron diffusion in photo-excited Ni using time-resolved X-ray diffraction

We show that the heat deposition profile in a laser-excited metal can be determined by time-resolved X-ray diffraction. In this study, we investigated the electron diffusion in a 150 nm thick nickel film deposited on an indium antimonide substrate. A strain wave that mimics the heat deposition profile is generated in the metal and propagates into the InSb, where it influences the temporal profile of X-rays diffracted from InSb. We found that the strain pulse significantly deviated from a simple exponential profile, and that the two-temperature model was needed to reproduce the measured heat deposition profile. Experimental results were compared to simulations based on the two-temperature model carried out using commercial finite-element software packages and on-line dynamical diffraction tools. To reproduce the experimental data, the electron–phonon coupling factor was lowered compared to previously measured values. The experiment was carried out at a third-generation synchrotron radiation source using a ...

Tapered inner-cladding fiber design for uniform heat deposition in Ytterbium-doped fiber amplifiers

A method for designing double-clad fiber with tapered inner cladding and uniform core is proposed for linear pump power profile, i.e. uniform heat deposition, in the ytterbium-doped fiber amplifier. The analytical formula for the inner-cladding diameter profile along the fiber is given. The inner-cladding diameter near the pump injection port is determined purely by the diameter of the doped region, the number density of the doped ions, the absorption cross section at the pump wavelength and the length of the fiber. The simplified linearly varying inner-cladding diameter is proven to have a smoother heat deposition profile with lower maximum thermal load in both the co-pumping scheme and the counter-pumping scheme.

Formularization of the confinement enhancement factor as a function of the heating profile for FFHR-d1 core plasma design

A quantitative estimation of the confinement enhancement due to the heating profile effect is introduced to the helical fusion DEMO reactor design of FFHR-d1, based on the experimental results of the Large Helical Device. By applying this to the direct profile extrapolation (DPE) method, radial profiles in the reactor are extrapolated from experimental results. In reactor plasmas, the heat deposition profile of alpha heating is expected to be peaked in the core region as in the case of tangential neutral beam (NB) injection on low-density plasmas. The height of the pressure profile normalized by the gyro-Bohm-type parameter dependence increases with the power (∼0.6) of the peaking factor of the heat deposition profile, as long as the core confinement degradation observed in low-density plasmas is ignored. According to this observation, the confinement enhancement factor expected under the self-ignition condition ranges from ∼1.1 to ∼1.7, for example, depending on the used data. Degradation of the global energy confinement observed in high-density NB-heated plasmas is mitigated and the gyro-Bohm-type parameter dependence reappears after introducing the confinement enhancement due to the heating profile effect. Finally, typical example profiles in FFHR-d1 are provided by the DPE method for future analyses.

CORONAL PLUMES IN THE FAST SOLAR WIND

The expansion of a coronal hole filled with a discrete number of higher density coronal plumes is simulated using a time-dependent two-dimensional code. A solar wind model including an exponential coronal heating function and a flux of Alfven waves propagating both inside and outside the structures is taken as a basic state. Different plasma plume profiles are obtained by using different scale heights for the heating rates. Remote sensing and solar wind in situ observations are used to constrain the parameter range of the study. Time dependence due to plume ignition and disappearance is also discussed. Velocity differences of the order of approximately 50 km/s, such as those found in microstreams in the high-speed solar wind, may be easily explained by slightly different heat deposition profiles in different plumes. Statistical pressure balance in the fast wind data may be masked by the large variety of body and surface waves which the higher density filaments may carry, so the absence of pressure balance in the microstreams should not rule out their interpretation as the extension of coronal plumes into interplanetary space. Mixing of plume-interplume material via the Kelvin-Helmholtz instability seems to be possible within the parameter ranges of the models defined here, only at large di stances from the Sun, beyond 0.2-0.3 AU. Plasma and composition measurements in the inner heliosphere, such as those which will become available with Solar Orbiter and Solar Probe Plus, should therefore definitely be able to identify plume remnants in the solar wind.

Edge Plasma Transport in the Helical Divertor Configuration in LHD

AbstractTo understand the edge plasma transport in the complex magnetic field lines structure in the LHD heliotoron, the heat and particle fluxes profiles on divertor plates were investigated experimentally and theoretically by using EMC3‐EIRENE code. The heat flux profiles on a divertor plate in different magnetic configuration were measured and the main channel of the heat transport was revealed to be the field lines with long connection lengths which connect the divertor plate and the stochastic field lines layer. A comparison of the heat and the ion fluxes profiles on the divertor plates between the experimental data and the results of the calculation using EMC3‐EIRENE code was conducted. It was revealed that the change of the cross‐field transport in the edge plasma region causes not only the change of the heat and particle flux profiles on divertor plates, but also the changes of the intrinsically non‐uniform global heat and particle deposition profiles on the helical diveror. The cross‐field transport seems to be larger in the higher edge electron temperature discharge in the experimental conditions in this paper (© 2010 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Design and manufacturing of the ITER ECRH upper launcher mirrors

Four of the 16 ITER upper port plugs will be devoted to electron cyclotron resonance heating (ECRH) in order to control magnetohydrodynamic (MHD) instabilities [1,2]. In order to achieve the stabilisation of the neoclassical tearing modes (NTM) and sawtooth oscillation, a deposition of a very localized and peaked current density profile over a broad poloidal steering range is required. In the present optical configuration eight 2 MW mm-wave beams enter each of the four upper launchers (UL) through waveguides into the vacuum vessel. Each beam line comprises consecutive corrugated waveguide sections with two mitre bends, orientating the poloidal and toroidal directions and three sections of quasi-optical transmission [3]. The beam waist locations and beam shaping properties in free space propagation are defined by two additional mirrors, the first being a static focusing mirror and the second a plane poloidally steerable mirror. Each mirror reflects a group of 4 mm-wave beams. The three types of UL mirrors (mitre bend, focusing and steering) absorb heat generated essentially by three sources: the ohmic loss of the RF beam reflected at the mirror surfaces and the nuclear and thermal radiation coming from the plasma. While the average heat load is within reasonable engineering limits, three elements condition the actual mirror design, the peak ohmic heat load (Gaussian or Bessel type heat deposition profiles), the electromagnetic forces generated in vertical disruption events (VDE), and the ITER cooling water requirements. This paper provides an overview of the different upper port-plug mirror designs and cooling schemes and an outlook on the prototype manufacturing activities and the future test program. The optimized mm-wave layout within the ECH port plugs is also presented.

Steady-state operation and high energy particle production of MeV energy in the Large Helical Device

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. High temperatures of approximately 1–2 keV, and steady-state plasma sustainment operations have been reported. Recently the steady-state operation regime was greatly extended in the Large Helical Device (LHD). A high-temperature plasma was created and maintained for 54 min with 1.6 GJ in the 2005FY experimental programme. The three-dimensional heat-deposition profile of the LHD helical divertor was modified, and during long-pulse discharges it effectively dispersed the heat load using a magnetic axis swing technique developed at the LHD. A sweep of only 3 cm in the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies and partially by electron cyclotron of fundamental resonance frequency. By accumulating the small flux of charge-exchanged neutral particles during the long-pulse operation, a high energy ion tail which extended up to 1.6 MeV was observed. This is the first experimental evidence of high energetic ion confinement of MeV range in helical devices. The long-pulse operations lasted until a sudden increase in radiation loss occurred, presumably because of metal wall flakes dropping into the plasma. The sustained line-averaged electron density and temperature were approximately 0.8 × 1019 m−3 and 2 keV, respectively, at a 1.3 GJ discharge (#53776) and 0.4 × 1019 m−3 and 1 keV at a 1.6 GJ discharge (#66053). The average input power was 680 kW and 490 kW, and the plasma duration was 32 min and 54 min, respectively. These successful long operations show that the heliotron configuration has a high potential as a steady-state fusion reactor.

Thirty-minute plasma sustainment by real-time magnetic-axis swing for effective divertor-load-dispersion in the Large Helical Device

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. A high temperature of approximately 2keV, and steady-state plasma-sustainment operation of the Large Helical Device (LHD) [O. Motojima, K. Akaishi, H. Chikaraishi et al., Nucl. Fusion 40, 599 (2000)] is reported. High-temperature plasmas were created and maintained for more than 30min with a world record injected heating power of 1.3GJ. The three-dimensional heat-deposition profile of the LHD helical divertor was modified and during long-pulse discharges it effectively dispersed the heat load using a magnetic-axis swing technique developed at the LHD. A sweep of only 3cm of the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The modification of the heat-load profile was explained well by field-line tracing. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies. The operation lasted until a sudden increase of radiation loss occurred, presumably because of wall metal flakes dropping into the plasma. The sustained line-averaged electron density was approximately 0.7–0.8×1019m−3. The average input power was 680kW, and the plasma duration was 31min 45s. This successful long operation shows that the heliotron configuration has a high potential as a steady-state fusion reactor.

Electron Pressure Profiles in High-Density Neutral Beam Heated Plasmas in the Large Helical Device

In the Large Helical Device (LHD), electron pressure profiles in gas-fueled high-density discharges tend to have a similar shape, as if these were frozen. This frozen profile is insensitive to variations in the magnetic field strength and moderate changes in the neutral beam heat deposition profile. At the same time, however, the absolute value of the electron pressure itself increases with the heating power, the electron density, and the magnetic field strength. In this study, a reference model for the electron pressure is proposed which consists of the frozen profile and parametric dependences derived from experimental observations. It is possible to define an operational regime where this typical profile appears by comparing the electron pressure profiles with this model. In the standard configuration, at which the maximum plasma stored energy in LHD has been obtained, the frozen profile appears in the plateau to the Pfirsh-Schluter regimes. As the collisionality decreases to the collisionless regime, the electron pressure becomes smaller than the prediction of the model and the deterioration is significant in the plasma core region. This tendency is enhanced in the configuration with the outward-shifted magnetic axis. The global energy confinement time, τE, in the high-collisionality regime has a weaker density dependence together with the mitigated power degradation, scaling as τE∝nebar0.28P-0.43 (nebar and P are the line-averaged density and the heating power, respectively), compared with the International Stellarator Scaling 95, where τE∝nebar0.51P-0.59.

Dependence of particle transport on heating profiles in ASDEX Upgrade

The behaviour of the density profiles in ASDEX Upgrade can be described well with the assumption D ∝ χ and a pinch of the order of the neoclassical Ware pinch. The latter is estimated from slowly equilibrating density profiles. The assumption D ∝ χ has been succesfully tested by varying the heat deposition profile, making use of on-/off-axis ICRH and ECRH: due to the generally observed self-similarity of the temperature profile, such variations in the heat flux profile have a strong effect on the χ-profile and on the D-profile if the above assumption is correct. The corresponding variations in the density profiles have indeed been observed. The model is also capable of describing the decay of the density profile after injecting a train of pellets. The anomalous transport of impurities is also increased with central heating, and the corresponding flattening of the density profile leads to a significant reduction of the neoclassical impurity pinch. Central ICR or ECR heating are therefore now routinely used to control density peaking and its negative effect on the stability of neoclassical tearing modes as well as to control the impurity transport in ASDEX Upgrade.

Impact of heat deposition profile on global confinement of NBI heated plasmas in the LHD**Original title in FEC2002, Lyon: response of temperature and density profiles to heat deposition profile and its impact on global scaling in LHD.

Energy confinement and heat transport of net-current-free NBI heated plasmas in the large helical device (LHD) are discussed with emphasis on density and power deposition profile dependences. Although the apparent density dependence of the energy confinement time has been demonstrated in a wide parameter range in LHD, the loss of this dependence has been observed in the high density regime under specific conditions. Broad heat deposition due to off-axis alignment and shallow penetration of neutral beams degrades the global energy confinement while the local heat transport maintains a clear temperature dependence, lying between Bohm and gyro-Bohm characteristics. The central heat deposition tends towards an intrinsic density dependence like τE∝(n̄e/P)0.6 from the state where density dependence is lost. The broadening of the temperature profile due to the broad heat deposition profile contrasts with the invariant property that has been observed widely as profile resilience or stiffness in tokamak experiments. The confinement improvement as a result of the inward shift of the magnetic axis is obvious in the core region, which emphasizes the improvement of transport because of the geometry being unfavourable for the central heating of NBI in this configuration. The edge pressure, clearly, does not depend on the magnetic axis position. Unlike a tokamak H-mode, the edge pressure is determined by transport and can be increased by increasing the heating power.

Significance of the radial electric field to divertor load asymmetries

JET divertor load measurements using embedded thermocouples have indicated a large target asymmetry and non-uniformities in the loads and that, under H-mode conditions, the load is dominated by the ion component. ASCOT simulations of the ion contribution to JET divertor loads are consistent with the presence of a large outward radial electric field E r in the scrape-off layer (SOL) next to the separatrix. Such a field would provide an explanation for observed features of the divertor loads in JET. The required field for getting the correct in/out asymmetry is within the range of the estimated SOL E r in JET, but reproducing the experimental heat deposition profiles requires higher field values.

Differential laser absorption and thermal emission for remote identification of opaque surface coatings

We experimentally and theoretically studied the phenomenon of thermal emission from nonvolatile liquid surface coatings following heating with a pulsed CO(2) laser. The effects of thermal diffusion across the liquid-air and liquid-substrate interfaces as well as the full absorption spectrum of the liquid are addressed theoretically. The differential temporal and intensity characteristics of the thermal emission signal from the heated surface coating, resulting from the differential heat deposition profile for on- and off-resonance excitation, are shown to be useful for the purposes of identifying different surface contaminants. The application of this technique to standoff thermal imaging of contaminated surfaces is discussed.