Divertor Plates Research Articles

Study of detached plasma profile in the divertor simulation experimental module of tandem mirror GAMMA 10/PDX

A divertor simulation experimental module (D-module) in the tandem mirror GAMMA 10/PDX was used for the study of plasma detachment. In previous studies, it was difficult to measure far-upstream plasma parameters in the D-module, and only electrostatic probes on the target plate were used to perform electron temperature and density measurements. To study the detached plasma structure, a Thomson scattering (TS) system and a microwave interferometer system have been installed to measure the inside plasma parameters of the D-module, and a movable electrostatic probe has been placed at the inlet of the D-module to measure the inlet plasma density and temperature. The TS system in the central cell observed the electron temperature and density of the core plasma simultaneously. These measurements revealed the entire density and temperature structure from the core plasma to the divertor plate. The line average electron density measured by the microwave interferometer showed a rollover behavior during detachment. The results indicated that the ionization region was located around the center of the D-module, and it appears to move upstream along the axis.

Assessment of radiation damage of the first wall of a fusion neutron source DEMO-FNS with a blanket for the transmutation of minor actinides

The design of the demonstration fusion neutron source (DEMO-FNS) demonstration fusion–fission hybrid facility is carried out at the National Research Center ‘Kurchatov Institute’ in Russia. For the implementation of the DEMO-FNS project, it is important to assess the effect of the fast neutron spectrum on radiation-induced damage to the device materials, primarily the materials of the first wall, which is the most problematic unit of the device. Due to the lack of real hybrid fusion devices, the study can only be carried out using computer simulations of the experiment. A computer simulation using the Monte Carlo method was carried out to calculate the developed three-dimensional full-scale model of the DEMO-FNS reactor with a blanket for the transmutation of minor actinides. The MCNP-4 code was used with cross-sections from the FENDL-2.1 and ENDF/B-6 files, as well as with cross sections for calculating radiation displacements. The neutron spectrum in the first wall of DEMO-FNS was determined at the power of a volumetric deuterium–tritium neutron of 1.42 × 1019 n s−1. It was found that beryllium as a plasma-facing material has too short lifetime under the neutron load on the first wall and divertor plates of DEMO-FNS. The replacement beryllium to tungsten is problematic and requires special study. Copper alloys have sufficient resistance to the effects of neutron radiation of DEMO-FNS for one full power year, but will require several replacements when operating the device for more than 10 years. The steels used in fast reactors can meet the operational requirements of the first wall of DEMO-FNS. To solve the problem of choosing materials for the first wall of DEMO-FNS, new experimental researches on changes in the physical properties of these materials in the fast neutron spectrum are needed.

Ion temperature anisotropy in the tokamak scrape-off layer

Understanding tokamak exhaust-power heat loads on divertor plates depends critically on having a realistic model of the scrape-off layer (SOL) plasma. The Braginskii fluid model is often solved to understand the SOL plasma behavior. This model is based on the collisional limit for transport along the magnetic field . The ions and electron gyrofrequencies are assumed to be much larger than the Coulomb collision frequencies, which are nonetheless, sufficiently large to yield common parallel and perpendicular temperatures for each species, i.e. the temperatures are assumed to be isotropic. In certain circumstances such as encountered for the tokamak H-mode, the ion temperature can be quite anisotropic. In this work, the anisotropy effects are implemented in the two-dimensional (2D) transport code UEDGE. Various geometries (1D slab, 2D slab and a toroidal tokamak geometry) are used to study the 2D structure of ion temperature anisotropy and its effects on plasma transport in detail. Results show that the effects of ion temperature anisotropy on the plasma parallel transport are substantial near the magnetic X-point, which leads to different steady state density profiles in the divertor regions. The extra mirror force introduced by ion temperature anisotropy can be one of the main forces contributing to the plasma flow in the SOL.

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.

Virtual-reality visualization of loss points of 1 MeV tritons in the Large Helical Device, LHD

Intersection points of 1 MeV tritons and the plasma facing wall are visualized in the vacuum vessel of Large Helical Device (LHD) with the divertor plates by the virtual-reality (VR) system. The trajectory of the energetic triton is evaluated by means of the collisionless Lorentz orbit code, and the strike point is calculated by the winding number algorithm. The intersection points of the tritons and the plasma facing wall are visualized as spheres with the rendered internal vacuum wall and divertor plates in the VR space, and it is possible to directly observe the strike points distributed on the wall and on the plates. It is found that many tritons intersect with the divertor plates and that some strike on the vacuum vessel. To evaluate the effectiveness of this VR visualization, we ask people to score the effect of experience and the level of understanding after finding the place where many particles attack on the plasma facing wall on the two-dimensional plane or in the VR space. This visualization helps us to determine where the material probes should be placed on the plasma facing wall in the real LHD for experimental analysis of the accumulated tritium on the plasma facing wall. This development of VR visualization will make a significant contribution to the fusion plasma research.Graphic abstract

Thermal-hydraulic modelling and analysis of ITER tungsten divertor monoblock

• A mathematical model has been developed to investigate the thermal-hydraulic performance of ITER tungsten divertor monoblock. • The model predicts the thermal response of the divertor structural materials and coolant tube. • The model accounts for the melting, vaporization, and solidification of the upper layer of the divertor facing plasma. • The model is verified against a previous calculation in the literature for DEMO divertor at an incident surface heat flux of 10 MW/m 2 . • The model is used to simulate the divertor materials response subjected to a VDE. The divertor is a fundamental component of fusion power plants, being primarily responsible for power exhaust and impurity removal via guided plasma exhaust. Due to its position and functions, the divertor has to sustain very high heat flux arising from the plasma (up to 20 MW/m 2 ), while experiencing an intense nuclear deposited power, which could jeopardize its structure and limit its lifetime. Therefore, attention has to be paid to the thermal–hydraulic design of its cooling system. In this work a mathematical model has been developed to investigate the steady state and transient thermal–hydraulic performance of ITER tungsten divertor monoblock. The model predicts the thermal response of the divertor structural materials and coolant tube. The coolant tube is divided into specified axial regions and the divertor plate is divided into specified radial zones, and then a two-dimensional heat conduction calculation is performed to predict the temperature distribution for both steady and transient states. A two-dimensional numerical finite difference technique is adapted in Cartesian coordinate system where the implicit scheme is used for transient calculation. The model also accounts for the melting, vaporization, and re-solidification of the upper layer of the divertor facing plasma. The selected heat transfer correlations cover all possible operating conditions of ITER under both normal and off-normal situations. The model is verified against a previous calculation in the literature for DEMO divertor at an incident surface heat flux of 10 MW/m 2 . The model is then used to predict the steady state thermal behaviour of the divertor under incident surface heat fluxes ranges from 2 to 20 MW/m 2 for a bare cooling tube and cooling tube with swirl-tape insertion. It calculates the maximum tube surface heat flux and the minimum critical heat flux ratio for all cases as well. The model is also used to simulate the divertor materials response subjected to high heat flux during a vertical displacement event (VDE) where 60 MJ/m 2 plasma energy is deposited over 500 ms.

Characteristics of double-peaked particle deposition at divertor target plates in the EAST tokamak

Increasing the heat and particle deposition on the divertor target plates is an effective approach for reducing the extremely high heat load. Future tokamak fusion devices are expected to have a deposition width on the order of a millimeter. Recently, a double-peaked distribution (DPD) pattern of particle deposition at the divertor target plates has been observed in both deuterium and helium plasma discharges in the Experimental Advanced Superconducting Tokamak, which can significantly spread the deposition width of heat and particle fluxes, and clearly shows a toroidal symmetric distribution. It has been found that the DPD behavior occurs not only in plasma discharges with lower hybrid wave (LHW) heating, but also with electron cyclotron resonance heating or neutral beam injection (NBI) heating alone. In addition, the DPD shows an obvious in–out asymmetry between the divertor target plates, which strongly depends on the toroidal field direction, i.e. it distinctly appears on the outer divertor plate in unfavorable B t conditions, while shifting to the inner divertor plate under favorable B t conditions. Meanwhile, the transition between the single-peaked distribution and the DPD is normally correlated with the line-averaged plasma density and the auxiliary heating power. The statistical results for the pure LHW heating discharge scheme show that the DPD behavior is significantly affected by both the plasma density and the heating power, i.e. the lower the ratio of the plasma density to the heating power, the more likely it is that DPD behavior will occur. The critical boundary lines between the DPD and non-DPD patterns are obtained with the line expressions ∼ 2P LHW,abs + 1.4 and ∼ 3.33P LHW,abs + 1 for the unfavorable and favorable B t cases, respectively. Finally, the underlying physics is also discussed, suggesting that the electric field drifts and the local sheath boundary condition may exert a strong influence on the DPD.

Line identification of extreme ultraviolet (EUV) spectra from low-Z impurity ions in EAST tokamak plasmas

Extreme ultraviolet (EUV) spectra emitted from low-Z impurity ions in the wavelength range of 10–500 Å were observed in Experimental Advanced Superconducting Tokamak (EAST) discharges. Several spectral lines from K- and L-shell partially ionized ions were successfully observed with sufficient spectral intensities and resolutions for helium, lithium, boron, carbon, oxygen, neon, silicon and argon using two fast-time-response EUV spectrometers of which the spectral intensities are absolutely calibrated based on the intensity comparison method between visible and EUV bremsstrahlung continua. The wavelength is carefully calibrated using well-known spectra. The lithium, boron and silicon are individually introduced for the wall coating of the EAST vacuum vessel to suppress mainly the hydrogen and oxygen influxes from the vacuum wall, while the carbon and oxygen intrinsically exist in the plasma. The helium is frequently used as the working gas as well as the deuterium. The neon and argon are also often used for the radiation cooling of edge plasma to reduce the heat flux onto the divertor plate. The measured spectra were analyzed mainly based on the database of National Institute of Standards and Technology. As a result, spectral lines of He II, Li II–III, B IV–V, C III–VI, O III–VIII, Ne II–X, Si V–XII, and Ar X–XVI are identified in EAST plasmas of which the central electron temperature and chord-averaged electron density range in T e0 = 0.6–2.8 keV and n e = (0.5–6.0) × 1019 m−3, respectively. The wavelengths and transitions of EUV lines identified here are summarized and listed in a table for each impurity species as the database for EUV spectroscopy using fusion plasmas.

Predictive modeling of a lithium vapor box divertor in NSTX-U using SOLPS-ITER

Abstract The unmitigated heat flux in attached operation of a fusion power plant is predicted to be destructive to any solid divertor surface. Detachment, whereby the plasma pressure drops significantly before reaching the divertor target thus greatly reducing the heat flux and sputtering, will be necessary to ensure adequate lifetime of plasma facing components (PFCs). The lithium vapor box divertor aims to detach the divertor plasma via evaporating and condensing lithium surfaces. By evaporating lithium near or at the divertor plate and condensing it closer to the main chamber, a lithium vapor density gradient can be created. This density gradient ties energy losses to poloidal distance between the target and the detachment point. The radiation zone is then prevented from reaching the X-point as the lithium ionization rate decreases when the detachment front moves away from the divertor target. Here we present Scrape Off Layer Plasma Simulator (SOLPS) simulations of a lithium vapor box divertor using an NSTX-U equilibrium and PFC geometry. The parameters for the core boundary conditions, gas puff intensity, and heat and particle transport coefficients are chosen to match experimental values. Acceptable agreement with experimental Scrape-Off Layer (SOL) widths is found, indicating a reasonable choice of transport coefficients. In predictive simulations, lithium is added via evaporation at the target. Predictions for peak heat fluxes and upstream impurity concentration are given for a variety of evaporation rates. Target electron temperature is predicted to be able to be reduced to recombination levels (below 1 eV) for lithium evaporation rates of 1 ⋅ 1 0 23 Li/s, indicating detachment. Peak heat flux at the lower outer target could be reduced by as much as a factor of six while maintaining upstream lithium fractions below 2%. The prevention of lithium from reaching the midplane is shown to be due to an increase in frictional forces acting on the lithium from a deuterium gas puff. Lithium is also shown to be redeposited close to the evaporator which is favorable for initial tests and future capillary porous systems.

Design and operation of the in-vessel cryopump for the new upper divertor in ASDEX Upgrade

To study alternative divertor configurations in the ASDEX Upgrade (AUG) tokamak a new upper divertor, Div-IIo, with two in-vessel coils and a cryopump (CP) will be installed in 2022. For efficient density control in the new divertor configuration during plasma discharges, the CP has to guarantee an effective pumping speed for deuterium (D2) of at least Seff = 15 m3/s. The CP will be installed behind the inner divertor with a toroidal pumping gap of width 30 mm between the inner and outer divertor plates. It consists of 7 modules with a major radius of R = 1.2 m and a total active length of 5.4 m.To investigate reactor relevant D2-He gas mixtures during upper single null discharges, the liquid helium cryo-panels are coated with activated charcoal as cryosorbent.For quick regeneration of the charcoal from the accumulated gases, heating wires on the cryo-panels are installed.To reduce the pumping speed during high density discharges, an in-vessel cryo valve between two modules is installed. The pumping speed can be reduced by 35%.The cryo-panels operate with saturated liquid helium (LHe) at a temperature of 4.4 K. The thermal radiation shield consisting of reflector and chevron-baffle operate with saturated liquid nitrogen (LN2) at 80 K. To supply the new upper CP with LHe and LN2 the existing LHe-LN2-supply of the lower CP will be used. Therefore, an additional distributor box and a multi-channel transfer line to the upper CP have been installed in 2019.

Impact of fine divertor geometrical features on the modelling of JET corner configurations

The modelling of JET corner configurations, in which the strike points are positioned deep in the corners of the divertor, is extremely challenging for edge plasma fluid modelling tools. To circumvent this technical limitation, a geometrical approximation has been proposed, consisting in considering an artificial minor modification of the geometry of the divertor targets plates. In this paper, we investigate how significantly this approximation impacts the output of transport simulations. Using the SOLEDGE2D-EIRENE transport code which has the unique capability to be able to cope with both the full and the approximated geometry, we have performed a density scan in H-mode for pulses in which the outer strike-point lies in the corner of the divertor. We report here how simulations in the artificial geometry differ from the ones in unaltered geometry. At the exception of low density cases, mid-plane profiles in the closed field lines region and the near scrape-off layer are little impacted. Further out however, in flux-surfaces that are concerned by the geometrical modification, we find that modifying the geometry leads to a strong overestimate of the plasma density. The density perturbation is not local and concerns the whole flux surfaces. Although the divertor geometry is modified only on the outer side, the largest impact is found at the inner divertor where densities are systematically overestimated by a factor that can exceed 10 in low density cases in the far Scrape-Off Layer (SOL) and temperature underestimated by 10 to 20 eV in most of the studied density range. The near SOL and strike point peak values are also impacted in the same direction with density changes by a factor of 2. As a consequence, the threshold to detachment of the inner divertor is found lower in the approximate geometry than in the unaltered one. Due to the large flux expansion between the outer and the inner target, the difference in plasma is especially sensitive at the top of the inner divertor baffle, which could have consequence on the evaluation of physical sputtering at that critical location. • Minor modification of the divertor geometry impacts the edge plasma equilibrium. • The impact is not local and can be larger on the opposite side of the divertor. • Care must be taken when modelling divertor geometry with approximate description. • Mean-field codes with up to the wall capability should be used in such cases.

TECXY simulations of Ne seeding in JET high power scenarios

• Numerical simulations of the high power and high current JET baseline scenario discharges. • Validation of the TECXY edge plasma model with experimental results for JET-DT baseline scenarios. • Determining the optimal plasma conditions with the lowest target plate temperature and the lowest effective charge for four different upstream densities with neon seeding. • Analysis of the nickel impurity on the JET plasma radiation in the Scrape-of Layer. Preparation of D-T experiments on JET device raises a question about the mitigation of assumed high power entering the SOL. JET DT scenarios aim to achieve good plasma confinement and the heat loads reduction to the divertor at the same time. Therefore, the divertor corner magnetic field geometry, strike point swiping and impurity seeding are considered to reduce expected high heat fluxes to the divertor plates. The aim of the paper is to analyse the influence of the neon impurity seeding on the plasma transport and its efficiency of the power mitigation in the JET tokamak as well as to perform validation of applied edge plasma model. In this contribution numerical simulations have been performed for two high power (34 MW), neon seeded DD JET discharges in the H-mode with different upstream densities and the same corner divertor configuration prepared as possible candidate for JET DT scenarios. The edge plasma transport have been described by two-dimensional multifluid TECXY code based on Braginskii plasma transport equations with assumed classical parallel transport of the plasma and anomalous perpendicular transport defined by ad hoc heat and particle transport coefficients. TECXY results show impact of the neon seeding on the reduction of the power flowing to the divertor. Scan with the neon concentration carried out for four different upstream densities allow us to determine optimal plasma conditions with the lowest target plate temperature and the lowest effective charge. Performed studies with use of the TECXY code and they comparison to experimental results give the opportunity to perform validation of applied TECXY edge plasma model and show the optimal range of plasma parameters like the upstream density and neon concentration, for which the radiation power in the SOL is the highest.

Divertor heat load distribution measurements with infrared thermography in the LHD helical divertor

We evaluated the two dimensional (2D) distribution of the divertor heat flux in LHD. The Infrared (IR) thermography was performed to measure the surface temperature at a divertor plate. The 3D heat conduction equation was solved using the finite element method (FEM) by taking into account the practical divertor geometry including the heat sink behind the graphite tile and the cooling system. The FEM analysis successfully reconstructs heat load distribution from the measured temperature pattern. Significant difference was found between the temperature and the heat load patterns. The FEM analysis shows that the highest heat deposition is observed at the strike line with 5–10 MWm−2 of ∼ 10 mm width in NBI heated discharge. In addition to the strike line there is also found the lower heat deposition region of ∼ 1 MW m−2 with wide channel width ∼ 30 mm. The detailed heat transport analysis inside the divertor components shows that the heat transport process is different between the strike line and the other region due to the heat channel width and to the divertor component structure. The comparison between the heat flux obtained by the thermography and that by the Langmuir probes shows reasonable agreement, except that the peak value of the heat load is higher in the IR thermography than in the Langmuir probe. By relaxing the assumption that electron temperature, Te, equal to be ion temperature, Ti, in the sheath model for the Langmuir probe analysis, the agreement becomes better for the case with Ti > Te.

Influences of heating and plasma density on impurity production and transport during the ramp-down phase of JET ILW discharge

The aim of this paper is to study the influences of plasma heating and plasma density on impurity production and transport during the plasma-termination phase. We have analyzed the ramp-down (RD) phase of a set of representative high-current JET ITER-like wall discharges: #92 437 (disrupted) and #92 442 (soft landing), characterized by a high plasma current of I p = 3.5MA. Analysis is performed for different time slots within the RD phase, corresponding to different levels of electron line density and auxiliary heating power. Since the deuterium gas fluxes are different, the influence of the separatrix density is also analyzed. The main conclusion from the simulations is the observation that for the same average-electron density, a decrease of the separatrix density leads to an increase of the plasma temperature at the divertor plate, leading to increased W production and consequently to a larger W concentration and radiation in the core. When the central electron temperature approaches the 2 keV level, corresponding to the maximum W and Ni cooling rate, the radiation in the plasma’s center is enhanced. Ni radiation is more important in the RD phase.

Current structure in the scrape-off layer of a tokamak in a quiescent state

Current structure in the scrape-off layer (SOL) of a tokamak is analyzed. It is demonstrated that poloidal currents measured in the experiments are a combination of several current types with different physical nature. Besides the known Pfirsch–Schlüter currents and thermoelectric currents, so-called plate closing currents flowing to/from the divertor plates are also analyzed. The latter close radial currents in the SOL and below/above the X-point in the SOL and private flux region (PFR). In particular, the current flowing to the outer plate in the PFR, opposite to the thermoelectric current, is predicted for the standard single-null configuration and favorable direction of drift. In addition, a pair of currents should flow to and away from the outer plate. In the single-null configuration, they are often masked by a larger thermoelectric current. However, for the connected double null (CDN) case, where the thermoelectric current is strongly reduced due to smaller temperature asymmetry, these currents dominate. The suggested physical model is supported by the results of simulations performed with the SOLPS-ITER transport code. Simulations were done for ASDEX Upgrade (AUG), L- and H-modes, single-null configurations, and for Globus-M H-mode, both disconnected and CDN configurations. Results of the simulations are compared with probe measurements for AUG and Globus-M tokamaks, and reasonable agreement has been found. The role of parallel currents in the formation of the potential maximum/minimum in the vicinity of the X-point for strongly detached regimes is also analyzed.

Impurity control study of radiative tungsten divertor in EAST tokamak

Abstract EAST was equipped with W-coated divertor plates and dome for the upper divertor, the remaining plasma facing components include the plasma facing components at lower divertor consisted of Carbon materials, at the baffles and main chamber walls consisted of Molybdenum materials. In order to study the control of W impurity on EAST with W divertor by an additional radiation losses, the radiative tungsten divertor with Ar impurity seeding for the control of tungsten impurity is studied by SOLPS-ITER simulation. The computational region includes the core, SOL and private flux region. The real geometry and material of plasma facing components (PFCs) are taken into account. The plasma species in the simulation include D 0 , D + 1 , e - , He 0 - He + 2 , C 0 - C + 6 , Mo 0 - Mo + 42 , W 0 - W + 74 and Ar 0 - Ar + 18 . The simulation results show that with the increase of Ar seeding rate at the dome the plasma temperature at the target plates and the concentration of W impurity in the plasma decrease. The simulation plays an important role in the study of radiative tungsten divertor towards the higher heating power and steady state operation.

Currents structure in the scrape-off layer of a tokamak

Abstract Currents structure in the scrape-off layer (SOL) of a tokamak is analyzed. It is demonstrated that poloidal currents measured in the experiments are a combination of several current types of different physical nature. Besides known Pfirsch-Schluter (PS) currents and thermoelectric currents, so-called PCC (plate closing currents) flowing to/from the divertor plates are also analyzed. The latter close radial currents in the SOL and below/above the X-point. In particular, current flowing to the outer plate in the private flux region (PFR), opposite to thermolelectric current is predicted for the standard single-null configuration and favorable direction of ∇B drift (∇B drift of ions is directed towards active X-point). In addition, a pair of currents to and away from the outer plate should flow. In the single-null configuration they are often masked by a larger thermoelectric current, however for the connected double null (CDN) case, where thermoelectric current is strongly reduced due to less temperature asymmetry, these currents dominate. The suggested physical model is supported by results of simulations performed with SOLPS-ITER transport code. Simulations were done for ASDEX-Upgrade (AUG), L-mode , single-null configurations, and for Globus-M H-mode, both disconnected and connected double null configurations. Results of the simulations are compared with probe measurements for AUG and Globus-M tokamaks, and reasonable agreement has been found.

High-heat flux tests of tungsten divertor mock-ups with steady-state plasma and e-beam

Abstract Mock-ups of the ITER divertor plate are tested with the combination of steady-state plasma and e-beam loads: (1) thermocyclic tests with powerful electron beam load of up to 49 MW/m2 in the e-beam facility; and then (2) subsequent testing in the PLM plasma device with steady-state plasma loads of 0.5–1 MW/m2 and more. Such tests simulate the variable load on divertor plates in the ITER. Tungsten samples are irradiated with helium plasma in experiments on the PLM plasma device with discharge duration of up to 200 min. The tests of ITER-grade tungsten VM-P with the combination of steady-state plasma and thermocyclic e-beam loads led to erosion, cracking, and nanostructured “fuzz” structure growth on the material surface. Post-mortem scanning electron microscopy revealed a stochastic nanostructured surface with dimension of structural elements less than 100 nm. The growth of a nanostructured surface with a “fuzz”-type structure and high porosity is observed. The results of such tests are of interest to estimate the erosion of tungsten in fusion reactors, including ITER, fusion neutron source FNS and DEMO.

Study of lattice thermal conductivity of tungsten containing bubbles by molecular dynamics simulation

Abstract Exposed to high fluxes of helium/hydrogen isotope particles and heat, tungsten divertor plates will suffer damage thus degrading its performance such as its thermal conductivity. This paper presents a study on the effect of bubbles on the lattice thermal conductivity of tungsten at the atomic level using molecular dynamics simulations. The present study finds that empty bubbles in tungsten lead to a decrease in the lattice thermal conductivity of tungsten. Furthermore, He/D filled bubbles aggravate this decrease. The physical origin of this behavior is discussed. It is also found that the decrease in lattice thermal conductivity depends strongly on both the impurity density in the bubbles and the bubble size.

Mechanisms of crack generation in high-pure tungsten exposed to high power density plasma

The mechanisms of crack formation in the tungsten divertor plates irradiated by hydrogen plasma under ITER-type steady-state and transient events are investigated. The stresses and defects of samples‘ initial structure are annealed as a result of steady-state irradiation with Falcon ion source, which generates 2 keV hydrogen ion flux of ~ 1022 m-2s−1 and power density ~ 1.7 MW × m−2. Quasi-stationary plasma accelerator QSPA Kh-50 provides pulsed hydrogen plasma loads which operating mode simulates transient events (type I ELMs) in ITER: ion impact energy is about 400 eV, maximum plasma pressure is 0.32 MPa, pulse duration is 0.25 ms. The diffraction maxima profiles, positions, intensities, half-widths, and intensity distributions are analyzed. The origin of crack formation is found to be in the plastic deformation of surface layers by the twinning mechanism. The latter is reinforced by the interaction of twins with the hydrogen-filled micro-pores. The micro-pores appear as a result of hydrogen saturation of subtraction loops formed due to the coalescence of vacancy complexes.