ITER Plasma Facing Components Research Articles

Recent DIII-D progress toward validating models of tungsten erosion, re-deposition, and migration for application to next-step fusion devices

Fundamental mechanisms governing the erosion and prompt re-deposition of tungsten impurities in tokamak divertors are identified and analyzed to inform the lifetime of tungsten plasma-facing components in ITER and other future devices. Various experiments conducted at DIII-D to benchmark predictive models are presented, leveraging the DiMES removable sample exposure probe capability and the Metal Rings Campaign, in which toroidally symmetric rows of tungsten-coated tiles were installed in the DIII-D divertor. In tokamak divertors, the width of the electric sheath is of the order of the main ion Larmor radius, and a vast majority of sputtered tungsten impurities are typically ionized within the sheath. Therefore, W prompt redeposition is mainly governed by the ratio of the characteristic ionization mean-free path of neutral tungsten to the width of the sheath. In-situ monitoring of the prompt redeposition of tungsten impurities in divertors is demonstrated via the use of WII/WI line ratios and the ionizations/photon (S/XB) method in L-mode discharges. Even with this relatively limited set of emission measurements, net erosion measurements were found to be a consistent upper bound to an analytic scaling based on the ratio of the W ionization length, and the width of the magnetic sheath rather than the ratio of and the W+ gyro-radius. In the far-scrape-off layer (SOL) of the ITER divertor, however, it is calculated that the measurement of photon emissions associated with the ionization of tungsten impurities up to may be required. Finally, W deposition patterns on DiMES collector probes, interpreted via DIVIMP-WallDYN modelling, reveal the key roles of progressive W erosion/re-deposition staps and E × B drifts in regulating long-range high-Z material migration.

LIBS plasma in atmospheric pressure argon, nitrogen and helium: Spatio-temporal distribution of plume emission and Hα linewidth

Laser Induced Breakdown Spectroscopy (LIBS) method is considered for assessing the retention of hydrogen isotopes in the ITER plasma-facing components during the maintenance breaks when the reactor is filled with near atmospheric pressure nitrogen or inert gas. At these conditions, the broadening of the spectral lines of hydrogen isotopes and the reduction of line intensities complicates the distinguishing of hydrogen isotopes. The aim of the present study was to investigate the effect of atmospheric pressure nitrogen, argon and helium ambient gas on the spatio-temporal distribution of the LIBS plasma plume emission and linewidths of Hα line, representing the hydrogen isotopes. Nd:YAG laser with 8 ns pulse width was used to ablate the molybdenum (Mo) target with hydrogen impurity. The development of the formed plasma plume was investigated by time and space-resolved emission spectra in the 20 nm range around the 656.28 nm Hα line. For all gases used in the experiments, the intensity and linewidth of Hα line decreased with the delay time between the laser pulse and the spectral registration. At the same linewidth values, the highest intensities were obtained in the helium atmosphere while the lowest intensity was obtained in nitrogen. According to spatially resolved spectral measurements, the Hα line was most intense near the Mo target while the Mo lines peaked farther away. In the case of the helium atmosphere, the plasma plume emission was observed at a longer distance from the target and it decayed faster than in argon and nitrogen atmospheres. According to these results, helium is the most beneficial ambient gas for hydrogen isotope detection by atmospheric pressure LIBS. The use of argon ambient gas may be required when LIBS is used for the simultaneous determination of fuel and He retention in the wall material.

Hypervapotron heat sinks in ITER plasma-facing components—Process qualifications and production control toward series production

The hypervapotron heat sink designs have been adopted in the ITER plasma-facing components, namely for the Enhanced Heat Flux First Wall panels and Divertor Dome Plasma-Facing Units. The outcomes of recent discussion among the Domestic Agencies and ITER Organization on requirements for the qualifications and production controls are reported. In addition, the paper summarizes the hypervapotron design, manufacturing technology, and urgent R&D tasks identified in the discussion.

Numerical simulation of erosion of misaligned W/Cu monoblocks caused by ITER-like type-Iedge localized modes

The thermal performance of the divertor W/Cu monoblock tiles under the ITER-like transient events has been one of the main concerns for ITER plasma facing components. Owing to the assembly tolerances during installation, the leading edge caused by misalignment between toroidal neighboring tiles will receive the extremely high cumulative heat flux and be damaged. In this work, we develop a two-dimensional heat conductivity model, including evaporation, radiation, melting process, and coupling cooling water condition, to investigate the thermal erosion of two shapes of tiles of W/Cu monoblock (unshaped and beveled tiles) with misalignment in a range from 0 to 0.3 mm, within the allowable maximum misalignment for ITER. To reflect the geometrical effects of castellated divertor tiles on the properties of its adjacent plasma, the energy flux density distribution arriving at the castellated divertor tile surface is evaluated first by using a two-dimension-in-space and three-dimension-in-velocity particle-in-cell plus Monte Carlo collisions code, and the obtained energy flux distribution is then used as input for the heat conduction model. The simulation results show that the maximum temperature of the unshaped tile with no misalignment is lower than that of the beveled tile under the steady-state inter-ELM heat flux, which increases more quickly than that of the beveled tile with misalignment increasing and will be larger than that of the beveled tile when misalignment is not less than 0.105 mm. Two shapes of the divertor tiles would melt and vaporize under typical heat flux density of a transient event of type-I edge localized modes (ELMs) for ITER, deposition energy of 1 MJ·m<sup>–2</sup> in a duration of 600 μs. The highest temperature, the maximum melting thickness, the maximum vaporization thickness of the unshaped tile with no misalignment are higher than those of the beveled tile except the melting volume ratio. The thermal erosion of the unshaped tile increases more remarkably than that of the beveled tile with misalignment increasing, and the melting volume ratio of the unshaped tile will exceed that of the beveled tile at a misalignment of 0.17 mm. In comparison with the unshaped tile, the beveled tile is more resistant to such a high heat flux of an ELM and misalignment.

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Development of refractory metals for application as plasma-facing armour material remains among priorities of fusion research programmes in Europe, China and Japan. Improving the resistance to high temperature recrystallization, enhancing material strength to sustain thermal fatigue cracking and tolerance to neutron irradiation are the key indicators used for the down selection of materials and manufacturing processes to be applied to deliver engineering materials. In this work we investigate the effect of neutron irradiation on mechanical properties and microstructure of several tungsten grades recently developed. Neutron irradiation campaign is arranged for screening purposes and therefore is limited to the fluence relevant for the ITER plasma facing components. At the same time, the neutron exposure covers a large span of irradiation temperatures from 600 up to 1000 °C. Four different grades are included in the study, namely: fine-grain tungsten strengthened by W-carbide (W–4wt.% W2C), fine-grain tungsten strengthened by Zr carbides (W–0.5% ZrC), W alloyed with 10 at.% chromium and 0.5 at.% yttrium (W–10Cr–0.5Y) and technologically pure W plate manufactured according to the ITER specification by Plansee (Austria). The strengthening by W2C and ZrC particles leads to an enhanced strength, moreover, the W–0.5ZrC material exhibits reduced DBTT (compared to ITER specification grade) and is available in the form of thick plate (i.e. high up-scaling potential). The W–10Cr–0.5Y grade is included as the material offering the self-passivation protection against the high temperature oxidation.

Laser induced breakdown spectroscopy for hydrogen detection in molybdenum at atmospheric pressure mixtures of argon and nitrogen

Present study investigated the effect of argon and nitrogen mixtures at atmospheric pressure on the hydrogen Hα line emission intensity during the Laser Induced Breakdown Spectroscopy (LIBS) measurements. LIBS is foreseen for the assessment of the fuel retention in ITER plasma facing components during the maintenance breaks when the reactor is filled with near atmospheric pressure nitrogen or dry air. The use of argon flow during LIBS measurements has shown to increase the LIBS signal at atmospheric conditions which aids the distinguishing of hydrogen isotopes. LIBS was used to ablate a small amount of molybdenum (Mo) target and the development of formed plasma plume was investigated by time-resolved emission spectra and photographs. Spectral measurements showed that the Hα line intensity and electron density decreased faster with the increased fraction of nitrogen in the background atmosphere. The electron temperature of created plasma was also lower in the nitrogen atmosphere. The spatial development of plasma plume was comparable in both argon and nitrogen atmospheres. According to these results, the reduction of Hα line intensity in the presence of nitrogen can be explained by higher fraction of energy consumption in dissociation and excitation processes of molecular gas and increased quenching of excited H states. Furthermore, the results demonstrate that a small amount of N2 impurities in Ar atmosphere does not have detrimental effect on LIBS measurements of hydrogen isotopes. Most of the local thermodynamic equilibrium (LTE) conditions required for the applicability of CF-LIBS were fulfilled in the time-window between 1-5 µs for Ar and 1-3 µs for N2 gas. One notable exception was the time-stability criterion for hydrogen atoms. The effect of this deviation from LTE on the applicability of CF-LIBS requires further studies.

Approximate atomic models for fast computation of the Fokker–Planck equation in fusion plasmas with high-Z impurities and suprathermal electrons

With the choice of tungsten as a material for the ITER plasma facing components, the suprathermal electron interaction with non-fully ionized impurities emerged as an important issue in plasma modeling. Microwave heating and current drive systems, especially lower hybrid current drive, can generate a significant population of suprathermal electrons in the plasma. Also, in the case of the runaway electron generation and mitigation by massive gas injection, the collisions with impurities can have a significant impact on the electron drag force. A correct description of the fast electrons collisions with non-fully ionized impurities requires calculation of the atomic form factor. This can be done with ab initio models that are accurate, though time consuming in practical applications. In this paper, we compare existing approximations of the form factors, based on the Thomas–Fermi or Pratt–Tseng models. Ab initio density functional theory (DFT) calculations are used as a reference method to determine the accuracy of the compared models. Based on this analysis, we propose some modifications of the existing models, tuned with numerical parameter optimization, which provide a higher accuracy while maintaining a short computation time. These modifications include multiple exponents in the Pratt–Tseng model and fitting the parameters of the form factor equation to the DFT-based results. Some applications of the presented models to the calculation of the elastic and inelastic collision frequencies for Fokker–Planck equation are presented, showing a good agreement between the results obtained with the DFT method and the proposed models.

Structural analysis of the Interspace Afocal Module of the Wide Angle Viewing System diagnostic for ITER

The Interspace Afocal Module (IAM) is one integral component of the ITER Visible/Infrared Wide Angle Viewing System (WAVS), which is an optical diagnostic aimed at monitoring the ITER plasma facing components for machine protection. The diagnostic comprises 15 lines of sight (LoS), strategically distributed in the ITER Equatorial Ports (EP) 3, 9, 12 and 17. Design of WAVS in EP12 is critical, since it has to be fully operational for the first plasma of ITER; it is at its Preliminary stage, carried out by the Consortium constituted by CEA, CIEMAT and Bertin Technologies, within the Framework Partnership Agreement financed by F4E.The WAVS in EP12 comprises three LoS that pass through the IAM, consisting in a refractive optical system that relays the pupil forward and controls the beam diameter. It includes two doublets of lenses for each LoS, made of Shappire and Calcium Fluoride, with diameters up to 128 mm. The afocal sets (up to 1.9 m in length) are tightly held in place by the support structure of the IAM, which is directly attached to the Interspace Support Structure (ISS).The IAM structural requirements are highly demanding, given the optical performance that has to be assured under thermal and inertial loads in normal operation, including seismic SL1. In addition, it also has to withstand loads in Category III and IV, including higher seismic events SMHV and SL2, or accidental loads such as the loss of coolant or fire events, since the IAM structure is classified as ITER Safety Relevant, due to its attachment to the ISS.The paper summarizes the structural analyses performed by CIEMAT to validate the mechanical behaviour of the IAM Preliminary design in EP12, to guarantee both the structural integrity and the optical performance, in accordance with the RCC-MR Code.

Shutdown dose rate assessment of the interspace components of wide angle viewing system diagnostic for ITER

Shutdown dose rate (SDDR) calculations have been performed to give support to the design of the Interspace (IS) VIS/IR WAVS diagnostic components of the Equatorial Port 12 (EP12), in the framework of the F4E-FPA-407 SG04 project. The Vis/IR WAVS comprises the viewing systems in four Equatorial Ports: 3, 9, 12 and 17, reminding that the present results are relative to the EP12 specifically. The Vis/IR WAVS is an optical diagnostic aimed at monitoring in visible and infrared radiation the ITER plasma facing components for machine protection.This paper describes the methodology used and shows the results obtained. The objective of the assessment is to evaluate whether the design meets the ALARA criteria in the IS area. In this work, once the materials have been selected to contribute as less as possible to the activation, the main criterion is that the contribution of all diagnostics installed in the port should not overcome the target of the biological dose rate of 100 μSv/h in the IS area after 106s of decay time.The main conclusion of the analysis is that the contribution of the VIS/IR WAVS IS components to the SDDR meets with the ALARA criterion in the IS. The contribution of the VIS/IR WAVS IS components to the SDDR in the IS, concerning the contribution of the whole ITER model, is about 9.6% in the left corridor and 5.5% in the right corridor. Besides, the percentage of contribution of the SS316L structure with respect to the total SDDR induced by the IS VIS/IR WAVS components is 99.3% in the left corridor and 99.5% in the right corridor. However, it has to be taken into account that these values have been obtained using a local model, therefore these values will be lower than the realistic values. Therefore, these values should be taken only as relative values.

Post-mortem analysis of tungsten plasma facing components in tokamaks: Raman microscopy measurements on compact, porous oxide and nitride films and nanoparticles

Raman microscopy is one of the methods that could be used for future post-mortem analyses of samples extracted from ITER plasma facing components. This study shows that this technique is useful for studying tungsten-based materials containing impurities including oxides and nitrides. Here, we apply pulsed laser deposition and DC argon glow discharges to produce tungsten-containing synthetic films (compact, and porous), and nanoparticles, and investigate the influence of their morphology on the measured Raman spectra. The amounts of oxygen and/or nitrogen in the films are also investigated. Comparative data are obtained via x-ray Photoelectron Spectroscopy, Atomic Force Microscopy, Electron Microscopies (Scanning and Transmission), Energy Dispersive x-ray spectroscopy, and Time-of-Flight Elastic Recoil Detection Analysis. The power density of the laser beam used to perform the Raman microscopy is varied by up to 4 orders of magnitude (0.01–20 mW μm−2) so as to investigate the thermal stability of films and nanoparticles. As a first result, we give evidence that Raman microscopy is sensitive enough to detect surface native oxides. Secondly, more tungsten oxides are detected in porous materials and nanoparticles than in compact films, and the intensities of the Raman band correlate to their oxygen content. Thirdly, the thermal stability of these films (i.e. structural and chemical modification under laser heating) is poor when compact films contain a sufficiently large amount of nitrogen. This finding suggests that nitrogen can be substituted by oxygen during Raman laser induced heating occurring in ambient air. Finally, our methodology can be used to rapidly characterize the morphology and chemistry of the samples analyzed, and also to create oxides at the micrometer scale.

Measurements of the angle-dependent reflectivity of plasma-facing components and assessment of the impact on the estimations of coverage of the IVVS measurements of the ITER VV

Abstract The In-Vessel Viewing System (IVVS) is a metrology instrument designed for deployment inside the ITER Vacuum Vessel (VV) to assess any damage, erosion or displacements of the Plasma-Facing Components (PFCs) through the lifetime of the ITER experiment. The latest developments of the IVVS have identified the Incidence-Angle-Dependent Reflectivity (IADR) of the PFCs to be critical to the assessment of the coverage of the IVVS within the VV. Some ITER PFC prototypes have now been manufactured and a specific instrument has been developed to measure their IADR on site, in an IVVS-like geometry with 100 dB dynamic range. This work presents the first measurements of manufactured ITER PFCs’ IADR together with an assessment of the impact of the IADR on the effective coverage of the viewing channel of the IVVS operating through amplitude-modulated lidar.

A framework for the assessment and control of ITER main chamber heat loads

A control-oriented approach, based on real-time equilibrium reconstruction, has been developed to monitor power fluxes to ITER plasma-facing components (PFC). The model describes the deposited heat flux as a poloidal flux function with two main input parameters: the power exhausted across the plasma boundary and the scrape-off layer heat flux width. A module containing weighting factors accounts for the real PFC 3D geometry. These factors are obtained using a new sophisticated GUI interface, SMITER, hosting a magnetic field line tracing code permitting the import and appropriate meshing of full PFC CAD descriptions. This new methodology has been used to examine two critically important first wall (FW) power loading issues for ITER: the start-up phase at low current and power, but requiring direct contact with the inboard wall, and the case of an upward plasma shift in a diverted configuration under high power, high current H-mode burning plasma conditions, which imposes very high power loads on the upper chamber FW panels.

Finite element analysis of hydrogen retention in ITER plasma facing components using FESTIM

Abstract The behaviour of hydrogen isotopes in ITER monoblocks was studied using the code FESTIM (Finite Element Simulation of Tritium In Materials) which is introduced in this publication. FESTIM has been validated by reproducing experimental data and the Method of Manufactured Solutions was used for analytical verification. Following relevant plasma scenarios, both transient heat transfer and hydrogen isotopes (HIs) diffusion have been simulated in order to assess HIs retention in monoblocks. Relevant materials properties have been used. Each plasma cycle is composed of a current ramp up, a current plateau, a current ramp down and a resting phase before the following shot. 100 cycles are simulated. The total HIs inventory in the tokamak during resting phases reaches 1.8 × 10 − 3 mg whereas during the implantation phases it keeps increasing as a power law of time. Particle flux on the cooling channel of the monoblock is also computed. The breakthrough time is estimated to be t = 1 × 10 5 s which corresponds to 24 cycles. Relevance of 2D modelling has been demonstrated by comparing the total HIs inventory obtained by 2D and 1D simulations. Using 1D simulations, a relative error is observed compared to 2D simulations which can reach -25% during the resting phase. The error during implantation phases keeps increasing.

Monte-Carlo simulations of electrostatic self-charging of tritiated tungsten and beryllium particles

The electrostatic self-charging rate of tokamak dust is investigated using Geant4, a toolkit for the simulation of the passage of particles through matter. To do so, the particles geometrical characteristics, the β disintegration energy spectrum and the deepness of tritium infusion are taken into account. The investigated materials are tungsten and beryllium, the plasma facing components (PFC) of ITER, considered as spherical particles from 20 nm to 200 μm in diameter, both tritiated. Two cases of tritium distribution in the particles are examined. On the one hand, tritium is homogeneously distributed over the whole sphere; on the other hand, tritium is homogeneously distributed within the external 100 nm layer of the sphere. The self-charging rate is assessed through the calculation of the particle exiting electron rate. Based on a tritium inventory of 10 GBq/g, relevant for ITER tokamak environment, our results show that, for a single tungsten or beryllium particle of 10 μm in diameter, the self-charging rate when the tritium is homogeneously distributed within the whole sphere is respectively 2.4 and 1.9 positive elementary charges per second. In the configuration where the tritium absorption is confined in the external 100 nm layer, the charge magnitude raises up to 37.1 and 8.4 respectively.

Development of tungsten armored high heat flux plasma facing components for ITER like divertor application

The dome and reflector plate are the parts of divertor plasma facing components (PFCs) of ITER tokamak which are mainly aimed for the removal of heat load of maximum 5 MW/m2 in steady state condition. The dome is a curved tungsten armoured component and the reflector plate is a straight component. These components have multi-layered joints made of various materials such as tungsten (W), OFHC copper (Cu), copper alloy (CuCrZr) and stainless steel (SS316LN). Joining of such multi-layered joints is known to be problematic due to joining of several dissimilar materials. In this paper, we report the indigenous development of medium size dome and reflector plate via vacuum brazing route for ITER like tokamak application. In order to evaluate the performance of the dome against ITER-like scenarios (maximum heat flux removal of 5MW/m2), the dome has been successfully tested for 1000 number of steady-state thermal cycles at incident heat fluxes of 3.87 MW/m2 in the High Heat Flux Test Facility (HHFTF) at IPR. Subsequent testing of additional 200 thermal cycles was also done at incident heat flux of 6 MW/m2. During the High heat flux (HHF) tests, surface temperature of W tiles reached 640oC and the beam power was restricted at 6MW/m2 to limit the temperature below 450oC at the CuCrZr heat sink. Total 1200 steady-state thermal cycles have been completed. At 6 MW/m2, the absorbed heat flux was 4 MW/m2. Engineering analysis on the HHFT of the dome has been performed using Finite element method (FEM) and Computational Fluid Dynamics (CFD) to simulate and to correlate with the experimental data. Ultrasonic immersion technique – Non destructive testing (NDT) was used to inspect the brazed joint quality of the dome before and after the HHFT. The results of the experimental details, engineering analysis and methodology adopted to fabricate the medium size dome and reflector plate are presented here.

Thermal desorption spectrometry of beryllium plasma facing tiles exposed in the JET tokamak

The phenomena of retention and de-trapping of deuterium (D) and tritium (T) in plasma facing components (PFC) and supporting structures must be understood in order to limit or control total T inventory in larger future fusion devices such as ITER, DEMO and commercial machines. The goal of this paper is to present details of the thermal desorption spectrometry (TDS) system applied in total fuel retention assessment of PFC at the Joint European Torus (JET). Examples of TDS results from beryllium (Be) wall tile samples exposed to JET plasma in PFC configuration mirroring the planned ITER PFC is shown for the first time. The method for quantifying D by comparison of results from a sample of known D content was confirmed acceptable. The D inventory calculations obtained from Ion Beam Analysis (IBA) and TDS agree well within an error associated with the extrapolation from very few data points to a large surface area.

Comprehensive 3-D simulation and performance of ITER plasma facing and nearby components during transient events—Serious design issues

A key obstacle to a successful magnetic fusion energy production in Tokamak reactors is performance during abnormal events. Abnormal events include plasma disruptions, edge-localized modes (ELMs), vertical displacement events, and runaway electrons. While tremendous efforts are being made to find ways to mitigate such events, a credible reactor design must be able to tolerate a few of these transient events. We have recently enhanced our comprehensive HEIGHTS (High Energy Interaction with General Heterogeneous Target Systems) simulation package to enable detailed 3-D investigation of the overall aspects of plasma-material interaction processes during all the transient events. Advanced models and numerical tools were developed to efficiently couple major key processes during the transient events, and in particular disruptions and giant ELMs. These include dynamic interaction, deposition, and scattering of the escaping core plasma particles with the evolving and propagating secondary divertor vapor/plasma in the strong magnetic field. These details are critical for assessing the damage to all interior components, including the hidden structure and the first wall which were not directly exposed to these transient events and never thought to be affected as a result. Despite developing numerous efficient numerical techniques and solution methods, such calculations take several months on current supercomputers to complete. Our present results show, for the first time, that unmitigated transient events could cause significant melting and vaporization damage to most interior and hidden components, including the first wall that were not directly exposed to these events. The current ITER divertor design may not work properly and need to be significantly modified or redesigned to prevent this damage.

Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

Activation analyses play a vital role in nuclear reactor design. Activation analyses, along with nuclear analyses, provide important information for nuclear safety and maintenance strategies. Activation analyses also help in the selection of materials for a nuclear reactor, by providing the radioactivity and dose rate levels after irradiation. This information is important to help define maintenance activity for different parts of the reactor, and to plan decommissioning and radioactive waste disposal strategies. The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential, due to the presence of a high-energy neutron environment which makes decisive demands on material selection. This study comprises two parts; in the first part the activation characteristics, in a fusion radiation environment, of several elements which are widely present in structural materials, are studied. It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment. The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions. The structural materials selected for this study, i.e. India-specific Reduced Activation Ferritic−Martensitic steel (IN-RAFMS), P91-grade steel, stainless steel 316LN ITER-grade (SS-316LN-IG), stainless steel 316L and stainless steel 304, are candidates for use in ITER either in vessel components or test blanket systems. Tungsten is also included in this study because of its use for ITER plasma-facing components. The study is carried out using the reference parameters of the ITER fusion reactor. The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port. The presence of elements like Nb, Mo, Co and Ta in a structural material enhance the activity level as well as the dose level, which has an impact on design considerations. IN-RAFMS was shown to be a more effective low-activation material than SS-316LN-IG.

Overview of manufacturing technologies under development in Russia for ITER plasma facing components

Manufacturing of the plasma-facing components (PFC) for the ITER is under qualification now (Merola et al., 2014) [1]. To start production of 179 first wall (FW) panels and 58 divertor domes (DD), for which the RF DA is responsible, it is necessary to manufacture their full-scale prototypes. In preparation for their manufacturing a wide range of technologies has been developed, which are continuously progressing (Mazul et al., 2011, 2012, 2016) [2–4]. Considering that changes were made in the DD design in view of the change-over to an all-tungsten divertor, as well as on the basis of the obtained results of manufacturing of the FW panel semi- prototype, the necessity arose to optimize some technological manufacturing stages. Thus, for the DD the composition of structural materials has been changed, the new assembly structure has been developed and analyzed; some technological operations have been optimized. The technological operations for FW panel have been also optimized, namely, the brazing temperature of the beryllium armor has been decreased, preparation works are under way on replacement of explosion welding by diffusion bonding for the steel-bronze joint. Most of the technological stages should be accompanied by non- destructive testing (ultrasonic, X-ray), the methods of which have been developed as well. The RF DA is also responsible for thermal testing of divertor elements. Due to application of the X-ray matrix the methods have been improved for recording of density of the applied heat flux produced by the electron beam. The ways for and the results of optimization of the technological and testing procedures are also presented below.

Current Status of Korea Heat Load Test Facility KoHLT-EB for Fusion Reactor Materials

The Korea heat load test facility using electron beam (KoHLT-EB) was constructed and commissioned to perform high heat flux testing on plasma facing components (PFCs), such as ITER blanket first wall, the tungsten divertor, and heat transfer experiments on other cooling devices. Also it will be used for the thermo-hydraulic performance test of ITER helium cooled ceramic reflector test blanket module in Korea. Qualification tests will be performed to evaluate the use of the high heat flux test facility on PFC and fusion reactor materials. For the thermal fatigue test of PFCs, two types of tungsten mockups were fabricated. One is a coated tungsten on the ferritic–martensitic steel (FMS). The other is a hot isostatic pressing bonded tungsten with FMS. Also, the integrity and cooling performance should be tested under the same or similar operation conditions of the ITER PFCs. After the completion of the preliminary mockup test and the facility qualification, the high heat flux test facility, KoHLT-EB, will provide the performance test for the various PFCs in fusion reactor materials and PFCs.