Plasma-facing Wall Research Articles

Temperature effects on the point defects formation in [111] W by neutron induced collision cascade

Tungsten is used as plasma-facing wall in ITER where is subjected to extreme operating conditions. In this work, we study the formation damage in [111] crystalline W by neutron bombardment in the temperature range of 300–900 K which is important for designing next generation of fusion reactors. Molecular Dynamics (MD) simulations are performed at a primary knock-on atoms (PKA) energy of 1 keV within the Gaussian Approximation Potential (GAP) framework. The analysis of the induced damage is done by the Fingerprinting and Visualization Analyzer of Defects (FaVAD) which is based on a rotation invariant mapping of the local atomic neighborhoods allowing to extract, to identify and to visualize the atomic defect formation dynamically during and after the collision cascades. The evaluation allows to classify several defect types and to quantify the defect evolution as function of time and initial sample temperature. We found that the production of Frenkel pair increases as a function of the temperature due to thermal activated mechanisms.

Molecular dynamics simulations of high-dose damage production and defect evolution in tungsten

Tungsten has been chosen as the plasma-facing wall material in fusion reactors, due to its high density and melting point. The wall material will not only be sputtered at the surface, but also damaged deep inside the material by energetic particles. We investigate the high-dose damage production and accumulation by computational means using molecular dynamics. We observe that the choice of interatomic potential drastically affects the evolution. The structure and stability of the obtained defect configurations are validated using a quantum-accurate Gaussian approximation potential.

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

Dust generation and accumulation in JET-ILW: morphology and stability of co-deposits on main plasma-facing components and wall probes

Dust particles and co-deposits were sampled for the first time from beryllium limiters and bulk tungsten divertor (both after ILW-3), and test mirrors from the main chamber after ILW-2 and ILW-3. The focus was on the morphology of molten particles and metal outgrowths. In parallel, the stability of beryllium layers under the impact of hot water was examined on limiters and Be coatings. The study performed by several microscopy methods including SEM, AFM, FIB, TEM and Be-sensitive EDX has revealed: (i) an asymmetric distribution of Be particles with the largest objects (125-550 μm) on side surfaces of the Be tile: (ii) the presence of highly porous particles, resembling blisters; (iii) very few thin flakes of co-deposits on the divertor, on the shadowed edge of lamella; (iv) the elemental composition and internal structure of the needle-shaped outgrowths on the mirrors: MoO; (v) no detectable impact of water on the beryllium morphology.

Modeling of hydrogen permeation behavior through tungsten deposition layer growing on nickel substrate by hydrogen plasma sputtering

Tungsten (W) is a candidate material of plasma-facing walls for fusion reactors. When a W redeposition layer is formed on the plasma-facing wall by sputtering, the behavior of tritium in the in-vessel components is possible to be affected. Therefore, it is necessary to understand the tritium permeation behavior through the W deposition layer as well as W bulk. In the previous study, the hydrogen permeating through the W deposition layer growing on a nickel substrate was measured experimentally. In this work, the TMAP simulation code was used to analyze the time variation in the measured hydrogen permeation flux and mass transfer parameters were obtained. The results suggested that the permeation flux initially increased with increasing temperature but decreased when the thickness of the W deposition layer was greater than the hydrogen injection range. The obtained recombination coefficient was several orders of magnitude larger than that reported for W bulk. The obtained values were used to predict hydrogen isotope permeation through the plasma-facing wall in the DEMO reactor. The results suggested that the W deposition formed on the plasma-facing wall of the DEMO reactor significantly reduces tritium permeation rate.

Effect of temperature distribution on tritium permeation rate to cooling water in JA DEMO condition

The estimation of tritium permeation rate through the plasma facing wall into coolant is required to discuss tritium balance in a D-T fusion plant, to design tritium recovery system and to perform safety assessments. In this work, tritium permeation rates in the blanket first wall and the divertor were estimated by numerical analysis for simplified multi-layer structures with considering the temperature distribution in recent JA DEMO condition. The permeation rate in the blanket first wall, which was a double layer consisting of tungsten and F82H, was estimated to be 0.69 g/day. The permeation rate in the divertor, which was a triple layer consisting of tungsten, copper and copper alloy or F82H, was estimated to be 0.013 g/day. When the permeation rate in tritium breeding region in the blanket can be reduced by three orders of magnitude due to a permeation barrier, total tritium permeation rate in the blanket and the divertor was estimated to be 0.71 g/day.

Overview of recent progress on steady state operation of all-metal plasma facing wall device QUEST

• QUEST has been equipped with the hot wall to control fuel particle balance. • The plasma pulse duration was longer after the installation of the hot wall. • The model demonstrated the quantitative effect of the hot wall installation. • The 6 h discharge was demonstrated. • The presence of the unexplained hydrogen refill was confirmed during the 6 h discharge. QUEST (Q-shu university experiment with steady state spherical tokamak) is a midsize spherical tokamak capable of steady-state operation, comprising all-metal plasma-facing walls and a hot wall (HW) to address issues pertaining to fuel particle balance. The HW was installed summer 2014. Quantitative analysis pertaining to the HW at 373 K is carried out, and clarify the quantitative impact of shot history that obviously appears in wall stored hydrogen just before the discharge at the wall temperature. The model indicates the plasma-induced deposition layer play an essential role in fuel particle balance. A clear temperature dependence of fuel recycling was observed using outgassing just after plasma termination and played an essential role in regulation of particle balance. Consequently, long duration discharges lasting more than 1 h has been obtained at wall temperature, T W < 423 K in the range of 40–60 kW of RF power and, similarly, a 6 h discharge could be achieved at T W = 473 K using 20 kW of injected RF power.

Evaluation of hydrogen retention behavior in tungsten exposed to hydrogen plasma in QUEST

Abstract The W (tungsten) samples were placed at Top, Equator and Bottom plasma facing walls of QUEST (Q-shu University Experiment with Steady-State-Spherical Tokamak) device and exposed to 754 shots of hydrogen plasma during 2017A/W (Autumn/Winter) campaign. Thereafter, their surface morphologies and chemical states were evaluated by TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). The XPS results showed that a thick carbon layer about 3–18 nm has formed throughout the wall surface. Among them, the Bottom wall had the deposition layer with the thickness of 2 nm, which was thinner than the top wall, namely erosion-dominated. On the other hand, a thick C layer about 18 nm was deposited on the Equator wall. The additional 1 keV D2+ was implanted into these samples and the D (deuterium) retention enhancement was estimated. The D2 TDS (Thermal Desorption Spectroscopy) spectra for all the samples had two major desorption stages at 400 K and 650 K, namely the desorption of D trapped by irradiation damages and deposition layer. The erosion/deposition profile would be caused by wall position and plasma condition, like a current start-up experiment. The desorption temperature of H2 (hydrogen) was shifted toward higher temperature side compared to that exposed to previous plasma campaign (2016 A/W), suggesting that H was mainly accumulated in the deposition layer with forming C–H bonds.

Hydrogen permeation behavior through tungsten deposition layer

Understanding of hydrogen isotope behaviors in plasma-facing wall is important from viewpoints of fuel control and tritium safety. Tungsten (W) is a candidate material of plasma-facing components. In this study, hydrogen diffusion coefficient and hydrogen solubility constant in the W deposition layer formed by hydrogen plasma sputtering were obtained by gas-driven permeation experiments. The hydrogen diffusion coefficient in the W deposition layer was smaller than that in bulk W, and the hydrogen solubility constant was larger than that in bulk W. Then, hydrogen permeation behavior through the W deposition layer growing on nickel plate by hydrogen plasma sputtering was observed in the temperature range from 77 to 145 °C. Initially, the hydrogen permeation flux steeply increased but then gradually decreased with increasing the thickness of the deposition layer. From the analysis of the permeation behavior by TMAP code, the recombination coefficient on the surface of the deposition layer was evaluated. The obtained recombination coefficient was smaller than that on bulk W.

Numerical Calculation on Recycling Ratio of Tritium from Tungsten Wall Used in Current CFETR Design

The recycling of tritium from plasma facing wall is an important neutral fuel source (in atomic and molecular form) for plasma confinement and particle control. In this study, the recycling process at tungsten wall based on current CFETR design was modeled. Monte Carlo code SRIM was used to model the implantation of energetic tritium ions into pure tungsten and to get back-scattering fraction of ions and the distribution of implanted tritium ions. The diffusion process of atoms in materials, with recombination at surface as boundary condition, was simulated using numerical approach for both stead and transient state. The total recycling ratio was contributed by fast process (implantation and back scattering) and slow process (diffusion and recombination) and its value nearly equals to 1 for stead state. Temporal dependence of total recycling ratio mainly depended on the slow process and was limited by diffusion coefficient in the bulk near surface and existence of traps in material. For tungsten material with good surface condition, the time of 90% recycling was characterized as 1 ms and affected by temperature, recombination coefficient and concentration of traps while the thickness of material had less affection. Isotope effect that recycling ratio of tritium was larger than that of deuterium at the same situation was also found in the simulation and this effect may affect particle balance and fueling in D-T plasma operation. A collection of theoretical models to estimate the recycling ratio and its time dependence were also summarized and validated by the simulation results.

Vibrationally resolved electron impact electronic excitation of BeH

Beryllium is being adopted for plasma facing walls in fusion reactors. This has led to the observation of emissions from the A 2Π state of beryllium hydride. Use of these emissions to monitor Be erosion requires electron impact excitation rates. Cross sections for electron impact vibrational excitation within the X 2Σ+ state and vibrationally resolved electronic excitation to the A 2Π state are reported for BeH, BeD and BeT. Electron collisions are studied at a range of internuclear separations using the UK molecular R-matrix (UKRmol+) codes. Electronic excitation is studied both within the Franck–Condon approximation and by explicit averaging of the T-matrix elements. It is found that (a) inclusion of the effect of higher partial waves using the Born approximation leads to significant increases in the cross sections and (b) the Franck–Condon approximation underestimates the importance of collisions for which the vibrational state changes during electronic excitation.

Transport parameters and permeation behavior of hydrogen isotopes in the first wall materials of future fusion reactors

Abstract The first wall of a future magnetic fusion device is essentially defined as the plasma-facing surface of the breeding blankets, which is supposed to be subjected to bi-directional hydrogen isotopes permeation: in one direction by edge plasma-driven permeation (PDP) of deuterium as well as tritium into blankets, and in the other direction by breed tritium gas-driven permeation (GDP) into the edge plasma. Deuterium and tritium PDP will complicate the recovery of tritium from the blanket, while tritium GDP will lead to an unwanted increase of particle recycling in the first wall region, which could even affect core confinement performance. Reduced activation ferritic/martensitic (RAFM) steels are widely proposed as candidate structural materials for the blanket of a DEMO reactor, the surface coatings made of tungsten are necessary to protect the plasma-facing wall from sputtering under high-energy particle bombardment. Therefore, the characterization of hydrogen isotopes transport through a multi-layer W + RAFM wall is of crucial importance to evaluate major reactor design issues including tritium retention, particle recycling and breeding feasibility, etc. This paper is intended to provide a review over the transport parameters of hydrogen isotopes in RAFM steels, including permeability, diffusivity, solubility and surface recombination coefficient. In addition, the present research status of hydrogen isotopes permeation and retention behavior of tungsten coated RAFM steels is briefly introduced.

Analysis of mixed-material layers deposited on the toroidal array probes during the FY 2012 LHD plasma campaign

Processes occurring on the plasma-facing walls in the Large Helical Device (LHD) and in-vessel material migration were investigated using a technique of material deposition probes. Ten Si plates were located on the outer side of the first-wall surface in each 36° toroidal section (Nos. 1–10). Mixed-material layers deposited on the probes during the fiscal year 2012 plasma campaign contained carbon, boron, hydrogen, and metals from which the plasma-facing elements were made. Metallic impurities and hydrogen content in deposited layers were examined by Rutherford backscattering spectrometry (RBS) and elastic recoil detection analysis (ERDA), respectively. The cross-sectional observations of the deposited mixed-material layers were performed with the help of a scanning electron microscope equipped with an energy dispersive X-ray (EDX) spectrometer. The results of the combined RBS/ERDA analysis and TEM-EDX examination allow tracing the processes occurring on plasma-facing components during the plasma campaign.

Particle balance investigation with the combination of the hydrogen barrier model and rate equations of hydrogen state in long duration discharges on an all-metal plasma facing wall in QUEST

The fuel particle balance during long duration discharges in the Q-shu University Experiment with steady state spherical tokamak (QUEST) was investigated. QUEST has all-metal plasma facing walls (PFWs) that were temperature controlled during the experiments. The presence of a transport barrier for hydrogen (H) at the interface between a plasma-induced deposition layer and metallic substrate was confirmed by nuclear reaction analysis with exposing deuterium plasma. An effective method to evaluate global hydrogen flux to PFWs is proposed, taking advantage of the nature of wall saturation. The outgoing flux of fuel particles from the PFWs just after the plasma termination was proportional to the square of wall-stored H, which indicates that enhanced recombination of solved hydrogen played an essential role in dynamic retention and was in agreement with predictions from the H-barrier model. A simple calculation based on the combination of wall modelling and rate equations of the H states denoted a significant impact of wall modelling on the time response of the plasma density. Hence, a proper wall model including the effects of the deposition layer creating the H barrier is required to be developed, even for all-metal PFW devices.

A wall-aligned grid generator for non-linear simulations of MHD instabilities in tokamak plasmas

Block-structured mesh generation techniques have been well addressed in the CFD community for automobile and aerospace studies, and their applicability to magnetic fusion is highly relevant, due to the complexity of the plasma-facing wall structures inside a tokamak device. Typically applied to non-linear simulations of MHD instabilities relevant to magnetically confined fusion, the JOREK code was originally developed with a 2D grid composed of isoparametric bi-cubic Bézier finite elements, that are aligned to the magnetic equilibrium of tokamak plasmas (the third dimension being represented by Fourier harmonics). To improve the applicability of these simulations, the grid-generator has been generalised to provide a robust extension method, using a block-structured mesh approach, which allows the simulations of arbitrary domains of tokamak vacuum vessels. Such boundary-aligned grids require the adaptation of boundary conditions along the edge of the new domain. Demonstrative non-linear simulations of plasma edge instabilities are presented to validate the robustness of the new grid, and future potential physics applications for tokamak plasmas are discussed. The methods presented here may be of interest to the wider community, beyond tokamak physics, wherever imposing arbitrary boundaries to quadrilateral finite elements is required.

Initial Neutronics Investigation of a Liquid-Metal Plasma-Facing Fusion Nuclear Science Facility

The use of a liquid-metal (LM) plasma-facing component (LM-PFC) in fusion reactor designs has some advantages as well as some disadvantages as compared to traditional designs that use a solid plasma-facing wall. Neutronics analysis of these potential LM-PFC concepts is important in order to ensure that radiation limits are met and that system performance meets expectations.A three-dimensional (3-D) neutronics analysis parametric study considering four LM first-wall (FW) candidates, (PbLi, Li, Sn, and SnLi) was performed with a thin (2.51-cm) LM-PFC design. The 3-D neutronics study used a fusion reactor based on the Fusion Energy Systems Study (FESS) Fusion Nuclear Science Facility (FNSF) (FESS-FNSF) that served as the baseline for comparison. FESS-FNSF is a deuterium-tritium–fueled tokamak with 518 MW of fusion power. A partially homogenized 3-D computer-aided-design model of the LM-PFC FNSF design was analyzed using the DAG-MCNP5 transport code.The results show that all candidate LM designs are acceptable with 4% to 13% increases in the tritium breeding ratio compared to the baseline case. The peak displacements per atom at the FW decrease 2% to 15%. For all four LM designs examined, the magnet heating and fast neutron fluence are well below acceptable limits. Overall, the Li LM design is the best candidate from a neutronics perspective.

Estimation of fuel particle balance in steady state operation with hydrogen barrier model

Abstract This research investigated fuel particle balance during long duration discharge in an all-metal plasma facing wall (PFW) through intensive QUEST execution. A simple wall model including the plasma-induced deposition layer that creates hydrogen (H) barriers, called the H barrier model, was established. A simple calculation, based on a combination of H state rate equations and the H barrier model, was applied to real plasma in the early phase of its longest discharge. The model accurately reconstructed the evolutions of electron density and wall-stored H over time, proper values are chosen for the parameters that are difficult to determine experimentally. Comparative calculations that used the H barrier and a fully reflective models, predicted significant impacts of wall models on the plasma density time response and value of electron density, indicating that a proper wall model should be developed for all-metal PFW devices.

High heat flux limits of the fusion reactor water-cooled first wall

The water-cooled WCLL blanket is one of the possible candidates for the blanket of the fusion power reactors. The plasma-facing first wall manufactured from the reduced-activation ferritic-martensitic steel Eurofer97 will be cooled with water at a typical pressurized water reactor (PWR) conditions. According to new estimates, the first wall will be exposed to peak heat fluxes up to 7 MW/m2 while the maximum operated temperature of Eurofer97 is set to 550 °C. The performed analysis shows the capability of the designed flat first wall concept to remove heat flux without exceeding the maximum Eurofer97 operating temperature only up to 0.75 MW/m2. Several heat transfer enhancement methods (turbulator promoters), structural modifications, and variations of parameters were analysed. The effects of particular modifications on the wall temperature were evaluated using thermo-hydraulic three-dimensional numerical simulation. The analysis shows the negligible effect of the turbulators. By the combination of the proposed modifications, the permitted heat flux was increased up to 1.69 MW/m2 only. The results indicate the necessity of the re-evaluation of the existing first wall concepts.

Designs and technologies for plasma-facing wall protection in EU DEMO

The EU DEMO plasma is almost completely enveloped by large breeding blanket segments for tritium breeding and power extraction. Shaping of the blanket plasma-facing wall in 3D may prove to be essential, but this strategy alone is not sufficient to protect against anticipated transient plasma events. The high heat flux wall-limiter approach used in ITER is not thought to be viable in a tritium self-sufficient power reactor, and so in EU DEMO wall protection using discrete limiters is pursued. Two types of discrete limiter are described in this paper. One is an equatorial port limiter designed to handle the power during the plasma start-up phase, making use of water-cooled tungsten/CuCrZr monoblock technology. The second is the upper limiter, featuring a plasma-facing component designed specifically for extreme transient loading due to a vertical displacement event. The heat flux channelling and thermal barrier features of this design are shown to considerably reduce CuCrZr pipe temperature, and so reduce the likelihood of catastrophic failure. A preliminary neutronic calculation has shown that the impact of these discrete limiters on overall tritium breeding ratio is relatively low.

Recrystallization kinetics of 50% hot-rolled 2% Y2O3 dispersed tungsten

Tungsten is considered as the most promising material for the divertor of fusion reactors and the plasma-facing first wall due to its excellent properties. With the growing demand for higher and better mechanical properties, the addition of rare earth oxides Y2O3 was selected as a dispersion strengthener to enhance the tungsten. The recrystallization behavior of yttria oxide dispersion strengthened tungsten plate hot-rolled to a medium plastic strain by 50% thickness reduction was investigated by isothermal annealing under the temperature range from 1250 to 1350 °C. Degradation of mechanical properties caused by recovery and recrystallization during annealing were quantified by Vickers hardness measurement. The entire recovery and recrystallization process were described nicely by serval classical models. The introduction of Y2O3 particles into tungsten retards the recrystallization process compared to that of pure tungsten.