ITER Relevant Conditions Research Articles

Ablation Simulation of Tungsten-Alternative Plasma-Facing Components due to Edge Localized Modes and Hard Disruptions

Innovative materials are investigated using a simulated electrothermal (ET) plasma to characterize erosion characteristics under ITER-relevant off-normal conditions. The tungsten alternatives investigated are mono-crystalline silicon carbides and MAX Phase ceramics. Preliminary code simulations using the ETFLOW plasma code are presented to assess erosive behavior in preparation for future experiments at ORNL’s electrothermal high heat flux experiment and the DiMES experiments for induced disruption on the DIII-D tokamak. Results indicate that erosion properties for SiC and two commercially available MAX Phases, Ti3SiC2 and Ti2AlC, compare well with tungsten and other ITER relevant components. A material-specific ablation constant, measured as total mass removed per incident heat flux per second, serves as a means for directly comparing erosion properties. Tungsten possesses the highest ablation constant value when compared to carbon, beryllium and the alternative materials α-6H SiC, Ti3SiC2, and Ti2AlC. The ablation thickness, calculated from the ablation constant and the specific density of the material, provides a comparison of surface thickness lost during a given off-normal event. Carbon (4.25 cm3/MJ) and tungsten (5.98 cm3/MJ) possess the lowest values. The alternative materials Ti3SiC2 (7.32 cm3/MJ) and α-6H SiC (8.44 cm3/MJ) exhibit the next best values, with Ti2AlC being the least effective (9.35 cm3/MJ). SiC shows the best vapor shielding efficiency of the three alternative materials, with Ti3SiC2 and Ti2AlC giving similar efficiencies. Taking into account vapor shielding effects using both opacity and fractional models, SiC exhibits the best ablation characteristics of the three materials in terms of thickness loss, with Ti3SiC2 giving similar results and overall appearing the superior of the two MAX Phases.

DiMES PMI research at DIII-D in support of ITER and beyond

An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based on experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. These techniques may find applications in the future devices.

Testing of sandwich molybdenum mirrors with radiative cooling for H-alpha and visible spectroscopy in ITER

The full-scaled mock-ups of the first spherical (M1) and the second flat (M2) mirrors for H-alpha and visible spectroscopy (HA&VS) in ITER were repeatedly heated within the temperature range of 100–350°C in a vacuum chamber (0.2–5Pa). The mock-ups were made of single crystal molybdenum segments bonded to a polycrystalline molybdenum substrate through a Ti interlayer by a hot isostatic pressure (HIP) technique. The goal of the test was to verify the stability of the mirror shape in ITER relevant conditions. The curvature of the mirror was optically measured in-situ. As a result, no damage or remarkable bending on the mirrors was observed. The maximum relative change of the M1 curvature at 350°C was −0.23%. The maximum curvature of the M2 mirror at 350°C was 1/670m−1. After cooling down to room temperature, the shape of the mock-ups was fully restored. Numerical simulation of HA&VS optics performance shows that the observed bending of the mirrors does not exceed the tolerable limits.

In situ cleaning of diagnostic first mirrors: an experimental comparison between plasma and laser cleaning in ITER-relevant conditions

This paper presents an experimental comparison between the plasma cleaning and the laser cleaning techniques of diagnostic first mirrors (FMs). The re-deposition of contaminants sputtered from a tokamak first wall onto FMs could dramatically decrease their reflectance in an unacceptable way for the proper functioning of plasma diagnostic systems. Therefore, suitable in situ cleaning solutions will be required to recover the FMs reflectance in ITER. Currently, plasma cleaning and laser cleaning are considered the most promising solutions. In this work, a set of ITER-like rhodium mirrors contaminated with materials tailored to reproduce tokamak redeposits is employed to experimentally compare plasma and laser cleaning against different criteria (reflectance recovery, mirror integrity, time requirement). We show that the two techniques present different complementary features that can be exploited for the cleaning of ITER FMs. In particular, plasma cleaning ensures an excellent reflectance recovery in the case of compact contaminants, while laser cleaning is faster, gentler, and more effective in the case of porous contaminant. In addition, we demonstrate the potential benefits of a synergistic solution which combines plasma and laser cleaning to exploit the best features of each technique.

The preparation of the Shutdown Dose Rate experiment for the next JET Deuterium-Tritium campaign

The assessment of the Shutdown Dose Rate (SDR) due to neutron activation is a major safety issue for fusion devices and in the last decade several benchmark experiments have been conducted at JET during Deuterium-Deuterium experiments for the validation of the numerical tools used in ITER nuclear analyses. The future Deuterium-Tritium campaign at JET (DTE2) will provide a unique opportunity to validate the codes under ITER-relevant conditions through the comparison between numerical predictions and measured quantities (C/E). For this purpose, a novel SDR experiment, described in the present work, is in preparation in the frame of the WPJET3-NEXP subproject within EUROfusion Consortium. The experimental setup has been accurately designed to reduce measurement uncertainties; spherical air-vented ionization chambers (ICs) will be used for on-line ex-vessel decay gamma dose measurements during JET shutdown following DT operations and activation foils have been selected for measuring the neutron fluence near ICs during operations. Active dosimeters (based on ICs) have been calibrated over a broad energy range (from about 30keV to 1.3 MeV) with X and gamma reference beam qualities. Neutron irradiation tests confirmed the capability of active dosimeters of performing on-line decay gamma dose rate measurements, to follow gamma dose decay at the end of neutron irradiation as well as insignificant activation of the ICs.

Hybrid Formulation of Radiation Transport in Optically Thick Divertor Plasmas

AbstractKinetic Monte Carlo simulations of coupled atom‐radiation transport in optically thick divertor plasmas can be computationally very demanding, in particular in ITER relevant conditions or even larger devices, e.g. for power plant divertor studies. At high (∼ 1015 cm–3) atomic densities, it can be shown that sufficiently large divertors behave in certain areas like a black body near the first resonance line of hydrogen (Lyman α). This suggests that, at least in part, the use of continuum model (radiation hydrodynamics) can be sufficiently accurate, while being less time consuming. In this work, we report on the development of a hybrid model devoted to switch automatically between a kinetic and a continuum description according to the plasma conditions. Calculations of the photo‐excitation rate in a homogeneous slab are performed as an illustration. The outlined hybrid concept might be also applicable to neutral atom transport, due to mathematical analogy of transport equations for neutrals and radiation.

Surface hardening induced by high flux plasma in tungsten revealed by nano-indentation

Surface hardness of tungsten after high flux deuterium plasma exposure has been characterized by nanoindentation. The effect of plasma exposure was rationalized on the basis of available theoretical models. Resistance to plastic penetration is enhanced within the 100 nm sub-surface region, attributed to the pinning of geometrically necessary dislocations on nanometric deuterium cavities – signature of plasma-induced defects and deuterium retention. Sub-surface extension of thereby registered plasma-induced damage is in excellent agreement with the results of alternative measurements. The study demonstrates suitability of nano-indentation to probe the impact of deposition of plasma-induced defects in tungsten on near surface plasticity under ITER-relevant plasma exposure conditions.

Numerical analysis of TDS spectra under high and low flux plasma exposure conditions

A recently developed numerical model, based on the dislocation-driven nucleation of gas bubbles, is used to analyse experimental results on deuterium retention in tungsten under ITER relevant plasma exposure conditions. Focus is placed on understanding the relation between exposure temperature and flux on primary features of thermal desorption spectra: peak positions and intensities of the desorption flux. The model allows one to relate the peak positions with the size of plasma induced deuterium bubbles and envisage exposure conditions (temperature and flux) for their formation. Based on the performed analysis, dedicated experimental conditions to validate the model are proposed.

High power plasma interaction with tungsten grades in ITER relevant conditions

Experimental simulations of ITER-like transient events with relevant surface heat load parameters (energy density up to 1.1 MJ/m2 and the pulse duration of 0.25 ms) as well as particle loads (varied in wide range from 1023 to 1027 ion/m2 s) were carried out with a quasistationary plasma accelerator QSPA Kh-50. Particular attention was paid to elaboration of damage of tungsten as a main candidate material for ITER divertor surfaces and also as prospective material for DEMO design. Erosion features, cracks evolution and changes in their thickness with increasing exposition dose are studied for different W grades, including deformed material with elongated grains, as well as WTa5 and sintered tungsten.

Nucleation and growth of hydrogen bubbles on dislocations in tungsten under high flux low energy plasma exposure

A new mechanism for the nucleation and growth of hydrogen (H) bubbles on dislocations under plasma exposure of tungsten was recently proposed on the basis of direct ab initio calculations. Density functional theory calculations demonstrated that H atoms are strongly bound to a screw dislocation core and exhibit fast one-dimensional migration along its line. Once the number of hydrogen atoms trapped on a dislocation segment exceeds eight, the emission of a jog occurs thereby converting a pure HN cluster into a HN+1-jog configuration. On the basis of these results a kinetic model was formulated to evaluate the conditions (i.e., range of temperature and flux exposure) for the transformation of pure H clusters into supercritical hydrogen–vacancy clusters attached to the dislocation line. In this work, a parametric study employing the kinetic nucleation model was performed to derive the hydrogen bubble formation energy function that offers the best agreement with available experimental results. The obtained results allow one to rationalize the depth and temperature dependence of the experimentally observed hydrogen deposition after high flux low energy plasma exposure for ITER relevant conditions.

Sawtooth control in JET with ITER relevant low field side resonance ion cyclotron resonance heating and ITER-like wall

New experiments at JET with the ITER-like wall show for the first time that ITER-relevant low field side resonance first harmonic ion cyclotron resonance heating (ICRH) can be used to control sawteeth that have been initially lengthened by fast particles. In contrast to previous (Graves et al 2012 Nat. Commun. 3 624) high field side resonance sawtooth control experiments undertaken at JET, it is found that the sawteeth of L-mode plasmas can be controlled with less accurate alignment between the resonance layer and the sawtooth inversion radius. This advantage, as well as the discovery that sawteeth can be shortened with various antenna phasings, including dipole, indicates that ICRH is a particularly effective and versatile tool that can be used in future fusion machines for controlling sawteeth. Without sawtooth control, neoclassical tearing modes (NTMs) and locked modes were triggered at very low normalised beta. High power H-mode experiments show the extent to which ICRH can be tuned to control sawteeth and NTMs while simultaneously providing effective electron heating with improved flushing of high Z core impurities. Dedicated ICRH simulations using SELFO, SCENIC and EVE, including wide drift orbit effects, explain why sawtooth control is effective with various antenna phasings and show that the sawtooth control mechanism cannot be explained by enhancement of the magnetic shear. Hybrid kinetic-magnetohydrodynamic stability calculations using MISHKA and HAGIS unravel the optimal sawtooth control regimes in these ITER relevant plasma conditions.

Thermal Discrete Element Analysis of EU Solid Breeder Blanket Subjected to Neutron Irradiation

Due to neutron irradiation, solid breeder blankets are subjected to complex thermo-mechanical conditions. Within one breeder unit, the ceramic breeder bed is composed of spherical-shaped lithium orthosilicate pebbles, and as a type of granular material, it exhibits strong coupling between temperature and stress fields. In this paper, we study these thermo-mechanical problems by developing a thermal discrete element method (Thermal-DEM). This proposed simulation tool models each individual ceramic pebble as one element and considers grain-scale thermo-mechanical interactions between elements. A small section of solid breeder pebble bed in a helium-cooled pebble bed (HCPB) is modelled using thousands of individual pebbles and subjected to volumetric heating profiles calculated from neutronics under ITER-relevant conditions. We consider heat transfer at the grain scale between pebbles through both solid-to-solid contacts and the interstitial gas phase, and we calculate stresses arising from thermal expansion of pebbles. The overall effective conductivity of the bed depends on the resulting compressive stress state during the neutronic heating. The Thermal-DEM method proposed in this study provides access to the grain-scale information, which is beneficial for HCPB design and breeder material optimization, and a better understanding of overall thermo-mechanical responses of the breeder units under fusion-relevant conditions.

Simulation of vibration-induced effect on plasma current measurement using a fiber optic current sensor.

An accurate measurement of the plasma current is of paramount importance for controlling the plasma magnetic equilibrium in tokamaks. Fiber optic current sensor (FOCS) technology is expected to be implemented to perform this task in ITER. However, during ITER operation, the vessel and the sensing fiber will be subject to vibrations and thus to time-dependent parasitic birefringence, which may significantly compromise the FOCS performance. In this paper we investigate the effects of vibrations on the plasma current measurement accuracy under ITER-relevant conditions. The simulation results show that in the case of a FOCS reflection scheme including a spun fiber and a Faraday mirror, the error induced by the vibrations is acceptable regarding the ITER current diagnostics requirements.

Plasma exposure of different tungsten grades with plasma accelerators under ITER-relevant conditions

This paper presents the results of tungsten irradiation experiments performed with three plasma facilities: the QSPA Kh-50 quasi-steady-state plasma accelerator, the PPA pulsed plasma gun and the magneto-plasma compressor. Targets made of different kinds of tungsten (sintered, rolled and deformed) were irradiated with powerful plasma streams at heat fluxes relevant to edge-localized modes in ITER. The irradiated targets were analyzed and two different meshes of cracks were identified. It has been shown that the major cracks do not depend on the tungsten grade. This has been attributed to ductile-to-brittle transition effects. Meshes of inter-granular micro-cracks were detected for energy loads above the melting threshold and these were probably caused by the re-solidification process. The blister-like and cellular-like structures were observed on sample surfaces exposed to helium and hydrogen plasmas.

Construction of PREMUX and preliminary experimental results, as preparation for the HCPB breeder unit mock-up testing

One of the European blanket designs for ITER is the Helium Cooled Pebble Bed (HCPB) blanket. The core of the HCPB-TBM consists of so-called breeder units (BUs), which encloses beryllium as neutron multiplier and lithium orthosilicate (Li4SiO4) as tritium breeder in form of pebble beds. After the design phase of the HCPB-BU, a non-nuclear thermal and thermo-mechanical qualification program for this device is running at the Karlsruhe Institute of Technology.Before the complex full scale BU testing, a pre-test mock-up experiment (PREMUX) has been constructed, which consists of a slice of the BU containing the Li4SiO4 pebble bed. PREMUX is going to be operated under highly ITER-relevant conditions and has the following goals: (1) as a testing rig of new heater concept based on a matrix of wire heaters, (2) as benchmark for the existing finite element method (FEM) codes used for the thermo-mechanical assessment of the Li4SiO4 pebble bed, and (3) in situ measurement of thermal conductivity of the Li4SiO4 pebble bed during the tests.This paper describes the construction of PREMUX, its rationale and the experimental campaign planned with the device. Preliminary results testing the algorithm used for the temperature reconstruction of the pebble bed are reported and compared qualitatively with first analyses completed with the FEM codes.

Experimental and numerical assessment of normal heat flux first wall qualification mock-ups under ITER relevant conditions

The ITER first wall (FW) panel consists of beryllium in the form of tiles covering its surface, high strength copper alloy as the heat sink material and stainless steel as the structural material. Small-scale normal heat flux FW mock-ups, provided by Fusion for Energy, are tested in the electron beam facility JUDITH 2 at Forschungszentrum Jülich to determine the performance of this design under thermal fatigue. The mock-ups are loaded cyclically under a surface heat flux of 2 MW m−2 with ITER relevant water coolant conditions. In this study, three-dimensional finite element method thermo-mechanical analyses are performed with ANSYS to simulate the thermal fatigue behaviour of the mock-ups. The temperature results indicate that the beryllium surface temperature is below the maximum allowed temperature (600 °C) of beryllium to be tested. The thermal mechanical results indicate that copper rupture and debonding between Be and copper are the drivers of the failure of a mock-up. In addition, the experimental data, e.g. the surface temperature measured using an infrared camera and the bulk temperature measured using thermocouples, are reported. A comparative study between experimental and simulation results is performed.

Implications of the Tore-Supra WEST-Project on Radio-Frequency Additionnal Heating Systems

This year Tore-Supra celebrated its 25 years of operation. During this long time, a number of technologies have been developed. First of all, it was mandatory to develop reliable superconducting magnets at ~ 1.8 K, with superfluid helium as an efficient coolant. For the production of steady state discharge, three types of radio frequency (RF) additional heating systems have been developed: 1) lower hybrid current drive; 2) ion cyclotron resonance heating; and 3) electron cyclotron resonance heating. To cope with long lasting discharges (up to 380 s ×2.8 MW) and large RF additional heating power (12.3 MW ×3 s), actively cooled (AC) plasma facing components were deployed in Tore-Supra for the first time in a tokamak environment. Tore-Supra is now being modified into a D-shape axisymmetric tokamak with AC tungsten main chamber walls and a divertor, the WEST project (W-for tungsten-environment in steady-state tokamak). This new facility has the objective to offer ITER a test bed for validating the relevant AC metallic technologies in D-shape H-mode plasmas. In contrast to other metallic devices, such as JET and ASDEX Upgrade, WEST will rely only on the RF additional power systems. A set of plasma scenarios have been identified, ranging from a high total RF power scenario up to 15 MW in 30 s, to a high fluence scenario of 1000 s with up to 10 MW of injected RF power. These scenarios are able to reproduce ITER relevant conditions of steady state heat loads of 10-20 MW/m2, to test tungsten AC divertor technologies with relevant power heat fluxes and particle fluence.

GDMT-T: Superconducting Linear Device for PMI Studies

The prospect of a new facility for plasma-surface interaction studies is presented. The unique features of this facility named GDMT-T are ability to generate both stationary plasma flux and transient events which simulate ITER ELM-mode operation and strong magnetic field on the target important for studies of dust and impurity migration. The experimental program of GDMT-T will include testing of plasma-facing components at ITER-relevant conditions and investigations of electron beam generation and beam-plasma interaction for application in the next generation of linear fusion facilities.

Overview of design and thermal–hydraulic analysis of Indian solid breeder blanket concept

India has developed two concepts of breeding blanket for the DEMO reactor: one is Lead Lithium Ceramic Breeder (LLCB), and the other one is Helium-cooled Ceramic Breeder (HCCB) concept. Indian HCCB concept is having edge on configuration of helium-cooled solid breeder with RAFMS structure. Li2TiO3/Li4SiO4 and beryllium are used as the tritium breeder and neutron multiplier, respectively. 2D thermal–hydraulic simulation studies using ANSYS have been performed based on the heat load obtained from neutronics calculations to confirm heat removal under ITER pulsed operation. Transient thermal analysis has been simulated in ANSYS for the ITER relevant operational conditions. Thermal analysis provides important information about the temperature distribution in different materials used and their temperature–time histories. Result of thermal–hydraulic simulations shows that in each cycle, the maximum temperature of all materials remains same. The peak temperatures of all materials are well within their limiting value. Concept designs of HCCB blanket and its thermal hydraulic analysis will be presented in this paper.

Low-recycling conditions and improved core confinement in steady-state operation scenarios in JET (Joint European Torus)

In this paper, we discuss the phenomena that link particle recycling from the vessel walls in the L-mode during discharge start-up and the core confinement in the H-mode during the subsequent main heating phase. We consider available data of JET experiments that aimed at approaching fully non-inductive ITER-relevant steady-state conditions and show that the high electron temperature produced at the edge by a low recycling during start-up tends to favour the build-up of high normalized β(βN) regimes in the H-mode, the confinement being improved in a large plasma volume. To provide an insight into this complex phenomenon we have modelled the relation between particle recycling in the scrape-off layer and the evolution of plasma transport, plasma current density and shear as well as the stability properties for such experimental conditions. The results confirm the existence of a link between the confinement in the H-mode phase and the values at the edge of electron temperature, bootstrap current density and local magnetic shear during start-up. Such a link could favour these regimes to be self-sustained in time.