Effect of transient heating loads on beryllium

Beryllium will be used as a plasma facing material for ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of ITER first wall. The results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility are presented. The Be/CuCrZr mock-ups were exposed to upto 100 shots by deuterium plasma streams with pulse duration of 0.5 ms at ∼250 °C and average heat loads of 0.5 and 1 MJ/m². Experiments were performed at 250 °C. The evolution of surface microstructure and cracks morphology as well as beryllium mass loss are investigated under erosion process.     

Simulation of residual thermostress in tungsten after repetitive ELM-like heat loads

Brittle destruction of tungsten armour under action of edge localised modes of plasma instabilities (ELMs) in ITER is an important issue determining the lifetime of the divertor. Besides, cracking of the armour produces tungsten dust with characteristic size of 1–10 μm flying from the armour surface with velocities up to 10 m/s. Influx of the tungsten dust into the ITER confinement decreases the temperature of the plasma, reduces the thermonuclear gain and even may run the confinement into disruption.This paper describes experiments in QSPA-Kh50 plasma gun and modeling, which has been performed for providing more insight into the physics of tungsten cracking under action of ELMs and for confirmation of the important result on stabilization of the crack development at the tungsten armour surface, predicted in our previous paper – the same authors, 2010.The threshold value of energy density deposition for start of tungsten cracking has been measured as 0.3 MJ/m² after 5–10 shots. From analytical considerations three times smaller threshold value has been predicted with increasing number of shots.     

Evaluation of energy and particle impact on the plasma facing components in DEMO

We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacement events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Under steady-state operation heat transfer into the coolant must remain below the critical heat flux (CHF) to avoid the possible severe degradation of the coolant heat removal capability. From the plasma side it is particularly demanding to keep the bulk plasma contamination during the reactor long operational discharges below the fatal level. The possible damage of the FW materials due to the plasma sputtering erosion is estimated. The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions.     

Recovery of Technetium with Active Carbon Column in Partitioning Process of High-Level Liquid Waste

Blistering and exfoliation of several tungsten alloys, which cause surface damage, were investigated using 3-MeV He-ion bombardment at room temperature (RT), 400, and 550°C. The alloy W-0.3TiC, which was fabricated by the mechanical alloying method and had an ultrafine grain structure, a K-doped W alloy, and pure W metal were examined to explore a way of suppressing the surface damage. In RT irradiation, surface exfoliation occurred at a fluence of (1–2) × 10²² He/m² in all the tested specimens. In the case of 550°C irradiation, surface exfoliation was observed above 2 × 10²² He/m² irradiation in pure W and K-doped W, but no surface exfoliation was observed in W-0.3TiC up to a fluence of 2 × 10²³ He/m². The results showed that W-0.3TiC showed a higher resistance to surface exfoliation by He-ion bombardment and the level of resistance was temperature-dependent. The surface morphology, cross-sectional morphology, and microstructure were characterized by transmission electron microscopy. Helium thermal desorption spectrometry was carried out to determine the mechanism whereby the surface attained resistance to the damage through He-ion bombardment. The improvement in the resistance to the surface exfoliation could be attributed to the ultrafine grain structure and the intergranular enhanced He diffusion behavior of the MA-processed material.     

Feasibility study of an actively cooled tungsten divertor in Tore Supra for ITER technology testing

In order to reduce the risks for ITER Plasma Facing Components (PFCs), it is proposed to equip Tore Supra with a full tungsten divertor, benefitting from the unique long pulse capabilities, the high installed RF power and the long experience with actively cooled high heat flux components of the Tore Supra platform. The transformation from the current circular limiter geometry to the required X-point configuration will be achieved by installing a set of copper poloidal coils inside the vacuum vessel. The new configuration will allow for H-mode access, providing relevant plasma conditions for PFC technology validation. Furthermore, attractive steady-state regimes are expected to be achievable. The lower divertor target design will be closely based on that currently envisaged for ITER (W monoblocks), while the upper divertor region will be used to qualify the main first wall heat sink technology adopted for the ITER blanket modules (CuCrZr copper/stainless steel) with a tungsten coating (in place of the Be tiles which ITER will use). Extended plasma exposure will provide access to ITER critical issues such as PFC lifetime (melting, cracking, etc.), tokamak operation on damaged metallic surfaces, real time heat flux control through PFC monitoring, fuel retention and dust production.     

Overview on R&D and design activities for the ITER core charge exchange spectroscopy diagnostic system

The ITER core charge exchange recombination spectroscopy (core CXRS) diagnostic system is designed to provide experimental access to various measurement quantities in the ITER core plasma such as ion densities, temperatures and velocities. The implementation of the approved CXRS diagnostic principle on ITER faces significant challenges: First, a comparatively low CXRS signal intensity is expected, together with a high noise level due to bremsstrahlung, while the requested measurement accuracy and stability for the core CXRS system go far beyond the level commonly achieved in present-day fusion experiments. Second, the lifetime of the first mirror surface is limited due to either erosion by fast particle bombardment or deposition of impurities. Finally, the hostile technical environment on ITER imposes challenging boundary conditions for the diagnostic integration and operation, including high neutron loads, electro-magnetic loads, seismic events and a limited access for maintenance. A brief overview on the R&D and design activities for the core CXRS system is presented here, while the details are described in parallel papers.     

Erosion and re-deposition processes in JET tiles studied with ion beams

Plasma facing components in fusion reactor chambers will operate under extreme conditions. Among the processes with implications on the material lifetime are erosion and re-deposition due to plasma interactions.This work will address the behaviour of both JET divertor and outer poloidal limiters (OPL) under plasma irradiation. The limiters comprise about 50 pairs of tiles in a poloidal stack, each of which has a plasma facing surface about 25 mm (poloidal) by 350 mm (toroidal) and is about 50 mm thick. The divertor tiles are located at the bottom of the chamber and withstand high fluxes of radiation and heat. Standard carbon-fibre composite (CFC) tiles coated with a thin layer of W overlaid with a 10 μm layer of C were studied with RBS/PIXE to understand the erosion/re-deposition processes occurring in these regions of the reactor chamber. High resolution surface morphology was assessed through SEM with and without tilting of the sample. The retention of hydrogen isotopes in the tiles were studied combining NRA and ERDA techniques – this is mostly ²H from the fuelling gas, but ³H is also present as a result of ²H–²H fusion reactions, and ¹H coming from the atmospheric exposure.     

Non-destructive examination of the bonding interface in DEMO divertor fingers

Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m² in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La₂O₃ (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers.     

Operational characteristics of the high flux plasma generator Magnum-PSI

In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 10²³–10²⁵ m⁻² s⁻¹ hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.     

Thermal effects and component cooling in Magnum-PSI

Magnum-PSI is a linear plasma generator, built at the FOM-Institute for Plasma Physics Rijnhuizen. Subject of study will be the interaction of plasma with a diversity of surface materials. The machine is designed to provide an environment with a steady state high-flux plasma (up to 10²⁴ H⁺ ions/m² s) in a 3 T magnetic field with an exposed surface of 80 cm² up to 10 MW/m². Magnum-PSI will provide new insights in the complex physics and chemistry that will occur in the divertor region of the future experimental fusion reactor ITER and reactors beyond ITER. The conditions at the surface of the sample can be varied over a wide range, such as plasma temperature, beam diameter, particle flux, inclination angle of the target, background pressure and magnetic field. An important subject of attention in the design of the machine was thermal effects originating in the excess heat and gas flow from the plasma source and radiation from the target.     

High heat flux testing of mock-ups for a full tungsten ITER divertor

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m² in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing.After 1000 cycles at 10 MW/m², the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m² or 500 cycles at 20 MW/m².However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m² followed by 1000 cycles at 20 MW/m².The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m² in steady-state conditions.     

Experimental and theoretical investigation of droplet emission from tungsten melt layer

Tungsten in form of macrobrush structure is foreseen as one of candidate materials for the ITER divertor and the dome. Melting of tungsten and the following melt motion and melt splashing are expected to be the main mechanisms of damage which determine the lifetime of plasma facing components. New experimental investigations of droplet emission from the W melt layer for the Edge Localised Mode (ELM)-like heat loads have been carried out at the plasma gun facility quasistationary plasma accelerators (QSPA-T). In these experiments the threshold for droplet emission and the distributions of velocity on emission angles and amplitude of the ejected droplets were determined. In the paper the main physical mechanism (the Kelvin–Helmholtz instability) of the melt splashing under the heat loads being applied at QSPA-T and those anticipated after the ITER transients is analyzed. These numerical simulations demonstrated a reasonable agreement with the experimental data on the droplet sizes and droplet velocities and allowed the projections upon the W melt splashing at ITER conditions.     

Qualification and post-mortem characterization of tungsten mock-ups exposed to cyclic high heat flux loading

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, high heat flux tests were performed in the electron beam facility FE200, Le Creusot, France. Thereby, in total eight small-scale and three medium-scale monoblock mock-ups produced with different manufacturing technologies and different tungsten grades were exposed to cyclic steady state heat loads. The applied power density ranges from 10 to 20 MW/m² with a maximum of 1000 cycles at each particular loading step. Finally, on a reduced number of tiles, critical heat flux tests in the range of 30 MW/m² were performed.Besides macroscopic and microscopic images of the loaded surface areas, detailed metallographic analyses were performed in order to characterize the occurring damages, i.e., crack formation, recrystallization, and melting. Thereby, the different joining technologies, i.e., hot radial pressing (HRP) vs. hot isostatic pressing (HIP) of tungsten to the Cu-based cooling tube, were qualified showing a higher stability and reproducibility of the HIP technology also as repair technology. Finally, the material response at the loaded top surface was found to be depending on the material grade, microstructural orientation, and recrystallization state of the material. These damages might be triggered by the application of thermal shock loads during electron beam surface scanning and not by the steady state heat load only. However, the superposition of thermal fatigue loads and thermal shocks as also expected during ELMs in ITER gives a first impression of the possible severe material degradation at the surface during operational scenarios at the divertor strike point.     

Effects of Low Energy and High Flux Helium／Hydrogen Plasma Irradiation on Tungsten as Plasma Facing Material

The High-Z material tungsten (W) has been considered as a plasma facing material in the divertor region of ITER (International Thermonuclear Experimental Reactor). In ITER, the divertor is expected to operate under high particle fluxes (> 1023 m-2s-1) from the plasma as well as from intrinsic impurities with a very low energy (< 200 eV). During the past dacade, the effects of plasma irradiation on tungsten have been studied extensively as functions of the ion energy, fluence and surface temperature in the burning plasma conditions. In this paper, recent results concerning blister and bubble formations on the tungsten surface under low energy (< 100 eV) and high flux (> 1021 m-2s-1) He/H plasma irradiation are reviewed to gain a better understanding of the performance of tungsten as a plasma facing material under the burning plasma conditions.     

Modelling of thermally enhanced erosion of beryllium

This paper presents two concurring models for the thermally enhanced erosion of metals. The modelling particularly deals with the erosion of beryllium by a helium plasma as an example system. Molecular dynamics (MD) simulations are used to reduce the number of free parameters in the models. A model of sublimation of ad-atoms created during ion impact was earlier proposed as an explanation of thermally enhanced erosion. Using MD calculations the parameter space for this model was reduced to a single free parameter, the areal surface defect density δ_Def. Using the reduced parameter space a very low δ_Def has to be assumed in order to reproduce the experimental observations. Therefore a new model is proposed here that is very similar to the ad-atom model but is based on a different mechanism to create weakly bonded surface atoms. The paper shows that inclusion of He atoms during exposure to high flux (10²² m⁻² s⁻¹) of low energy He (50 eV) leads to the formation of weakly bonded atoms in the surface. The comparison of both models with experimental data and their applicability to other projectile/target systems is discussed.     

Recent developments in environmental aspects of D-He fuelled fusion devices

The stress on fusion safety has stimulated worldwide research in the late 1980s for fuel cycles other than D-T. With advanced cycles, such as D-D, D-³He, p-¹¹B, and ³He-³He, it is not necessary to breed and fuel large amounts of tritium. The D-³He fuel cycle in particular is not completely aneutronic due to the side D-D reactions. Neutron wall loadings, however, can be kept low (by orders of magnitude) compared to D-T fuelled plants with the same output power, eliminating the need for replacing the first wall and shielding components during the entire plant lifetime. Other attractive safety characteristics include low activity and decay heat levels, low-level waste, and low releasable radioactive inventory from credible accidents.There is a growing international effort to alleviate the environmental impact of fusion and to support the most recent trend in radwaste management that suggests replacing the geological disposal option with more environmentally attractive scenarios, such as recycling and clearance. We took the initiative to apply these approaches to existing D-³He conceptual designs: the ARIES-III power plant and the Candor experiment. Furthermore, a comparison between the radiological aspect of the D-³He and D-T fuel cycles was assessed and showed notable differences. This report documents the comparative assessment and supports the safety argument in favour of the D-³He fuel cycle.     

Ion beam studies of H and He in metals

The use of ion beams to study hydrogen and helium in metals is demonstrated. The ³He (d,p)⁴He nuclear reaction previously has been used together with ion channeling to determine the lattice locations of ion-implanted D and ³He in tungsten. Preliminary results applying these techniques to helium bubble and blister formation in tungsten are also presented and show that changes attributed to helium bubble formation are observed in tungsten at a He fluence as low as 6 × 10¹⁶ He/cm². The retention of ion-implanted deuterium in W, Au, and Pd surfaces is shown to be greatly enhanced by prior He ion-induced lattice damage. The amount of the damage trapping is also found to depend on whether the metal is in single crystal or polycrystalline form.     

Investigation of plasma exposed W–1% La2O3 tungsten in a high ion flux,low ion energy,low carbon impurity plasma environment for the International Thermonuclear Experimental Reactor

Previous investigations of tungsten for the International Thermonuclear Experimental Reactor (ITER) were focusing on using energetic ion beams whose energies were over 1 keV. This study presents experimental results of exposed W–1% La₂O₃ in high ion flux (10²² m^–2), low ion energies (about 110 eV) steady-state deuterium plasmas at elevated temperatures (873–1250 K). The tungsten samples are floating during plasma exposure. Using a high-pressure gas analyzer, the residual carbon impurities in the plasma are found to be about 0.25%. No carbon film is detected on the surface by the EDX analysis after plasma exposure. An infrared pyrometer is also used as an in situ detector to monitor the surface emissivities of the substrates during plasma exposure. Using the scanning electron microscopy, microscopic pits of sizes ranging from 0.1 to 5 μm are observed on the plasma exposed tungsten surfaces. These pits are believed to be the results of erupted deuterium gas bubbles, which recombine underneath the surface at defect locations and grain boundaries, leading to substrate damage and erosion loss of the substrate material. Low temperature plasma exposure of a tungsten foil indicates that deuterium gas (D₂) is trapped inside the substrate. Macroscopic blisters are observed on the surface. The erosion yield of the W–1% La₂O₃ increases with temperature and seems to saturate at around 1050 K. Scattered networks of bubble sites are found 5 μm below the substrate surface. High temperature plasma exposure appears to reduce the population as well as the size of the pits. The plasma exposed W–1% La₂O₃ substrates, exposed above 850 K, retain about 10¹⁹ D/m², which is two orders of magnitude less than those retained by the tungsten foils exposed at 400 K.     

The use of an electrothermal plasma gun to simulate the extremely high heat flux conditions of a tokamak disruption

Disruption damage conditions for future large tokamaks like ITER are nearly impossible to simulate on current tokamaks. The electrothermal plasma source SIRENS has been designed, constructed, and operated to produce high density (> 10²⁵/m³), low temperature (1–3 eV) plasma formed by the ablation of the insulator with currents of up to 100 kA (100 s pulse length) and energies up to 15 kJ. The source heat fluence (variable from 0.2 to 7 MJ/m²) is adequate for simulation of the thermal quench phase of plasma disruption in future fusion tokamaks. Different materials have been exposed to the high heat flux in SIRENS, where comparative erosion behavior was obtained. Vapor shield phenomena has been characterized for different materials, and the energy transmission factor through the shielding layer is obtained. The device is also equipped with a magnet capable of producing a parallel magnetic field (up to 16 T) over a 8 msec pulse length. The magnetic field is produced to decrease the turbulent energy transport through the vapor shield, which provides further reduction of surface erosion (magnetic vapor shield effect).     

Mechanism of He release from He implanted samples deduced from post-implantation He release measurements

A combination of post-implantation, room temperature, He release measurements and surface erosion investigation by scanning electron microscopy was used for the study of the possible mechanisms of He release from implanted samples. He release was greatly accelerated at fluences exceeding a critical value. The critical fluence for fast He release was found to be smaller than that needed for the onset of surface erosion and was independent of surface erosion, (i.e. different samples with markedly different amounts of surface erosion exhibited the same He release). Post-implantation He release could be explained in terms of atomic diffusion processes. It was suggested that at high He concentrations this diffusion takes place via micro-channels created by micro-erosion processes that are independent of the known macro-erosion processes (such as flaking, cracking and blistering).