ITER Divertor Research Articles

Design and vibration test of first mirror mount assembly for ITER divertor VUV spectrometer

The divertor vacuum ultraviolet (VUV) spectrometer is a diagnostic system in the ITER tokamak, monitoring impurity content and behavior in the divertor region. The first mirror of the spectrometer, made of silicon carbide (SiC), is exposed to harsh environmental conditions, including high temperatures and significant inertial loads from electromagnetic disruption events. To ensure its reliable performance, we have designed and tested a robust mirror holder assembly. This paper introduces a novel design of the first mirror holder assembly for the ITER divertor VUV spectrometer and presents the results from comprehensive vibration tests conducted on a full-scale mock-up. The design features a double-holder structure with spring plate assemblies to accommodate thermal expansion and resist vibrational loads. The mock-up underwent a series of resonance search, sine dwell, and random vibration tests, replicating the expected loads during vertical displacement events in ITER. The mirror holder assembly and the dummy mirror successfully withstood the vibration tests without damage, validating the design for the ITER environment. The results demonstrate the robustness and reliability of the mirror holder assembly, ensuring the accurate and reliable operation of the divertor VUV spectrometer in ITER.

Reflectivity measurement of a silicon carbide mirror sample for ITER divertor vacuum ultraviolet spectrometer.

The first mirror is the front-end optic component that reflects light emitted from the plasma to the diagnostic system in fusion plasmas. Silicon carbide (SiC), known for its relatively high mechanical strength and radiation tolerance, has been selected as the substrate material for the first mirror in the ITER divertor vacuum ultraviolet (VUV) spectrometer. To measure the reflectivity of the ellipse cylindrical SiC mirror to be manufactured, a device for reflectivity measurement in the VUV wavelength range was developed. First, the reflectivity of a sample SiC mirror (15mm diameter × 10mm thick) was measured across the ITER-required incidence angles, and the results are reported in this study. A hollow cathode lamp with helium gas was used as the VUV light source in the wavelength range of 23-60nm, and a dedicated VUV spectrometer to select specific wavelengths was developed. The spectrometer utilized laminar-type replica diffraction gratings (Shimadzu 30-006) and two back-illuminated charge-coupled devices (BI-CCD, Andor DO 940P-BEN) for the grating and detector, respectively. A cropping technique with aperture was employed to precisely localize the VUV light's reflection onto the SiC mirror surface. The experimentally measured reflectivity values of SiC at the required incidence angles of VUV light were compared with theoretically calculated reflectivity curves. The oxidation layer (SiO2) formed on the SiC surface and the incidence angle of VUV light to the BI-CCD chip (E2V) would be the factors affecting the accuracy of the reflectivity.

Performance assessment of spectroscopic measurements of hydrogen and beryllium influxes and ionization fronts in ITER divertor plasmas using the divertor impurity monitor

Characteristics of Hα and Be II emissions in ITER divertor plasmas have been evaluated, and the applicability of a spectroscopic evaluation method of influxes and ionization front positions by using the divertor impurity monitor (DIM) has been systematically verified for the first time based on plasma profile datasets with parameter scans. In addition, practical values of the ionization/photon factor (S/XB) for DIM to evaluate influxes have been determined. In attached divertor plasmas and the onset of plasma detachment, H influx integrated over the entire area of the divertor can be evaluated within ±30 % accuracy using Hα intensities measured by the DIM and S/XB value of 32, which is higher than the S/XB in most divertor plasmas in previous devices owing to high electron density. The ionization front position can also be evaluated to within a few mm by using peak channel position. However, when electron temperature further decreases from the detachment onset, the ionizing plasma component (caused by electron impact excitation) in the total Hα emission decreases and the spatial profiles of emission and ionization diverge, making spectroscopic evaluation difficult. In all conditions, Hα intensities could be measured with a more than sufficient number of photons at the required time resolution using the current DIM design. For Be influx integrated over the entire divertor area, it has been found that Be I is not suitable for the DIM since its intensity is similar to the background continuous spectra. Replacing Be I with Be II solves this problem. Even in detached divertor plasmas, it is possible to evaluate Be1+ ionization flux within ±30 % accuracy using Be II intensities and S/XB, the value of which depends on the DIM’s fields of view. However, the evaluated Be1+ ionization flux is lower than Be influx by more than 80 %, meaning that the evaluation of Be influx using Be II would be an underestimate. But, Be II can be measured with a more than sufficient number of photons.

Development of potassium doped tungsten plate for fusion reactor applications

Since fusion reactors including DEMO and beyond are expected to be operated for a longer period of time than ITER, plasma-facing materials (PFM) will suffer by longer term heat load and non-negligible neutron irradiation damage. Therefore, for tungsten (W) materials as PFM, improvement of low temperature brittleness and suppression of recrystallization embrittlement and neutron irradiation embrittlement are required. To address these issues, various modified W materials have been developed by applying alloying with solid solution elements and second-phase dispersion techniques such as solid particles and bubbles. As a result, potassium-doped W (K-doped W), in which K bubbles are dispersed, was found to be effective for all of the above issues. However, the materials used in these evaluations were relatively small and not large enough to fabricate, for example, ITER divertor monoblocks (12 mm thick). In the present study, K-doped W plate with a sufficient thickness (more than 12 mm) for application to fusion reactor components like divertor monoblocks were developed and the fundamental properties were evaluated.

Effects of Gamma-ray and neutron irradiation on infrared optical properties of ZnSe and ZnS for ITER divertor thermography

Gamma-ray and neutron irradiation tests were performed by using antireflective-coated and non-antireflective-coated samples of ZnSe and ZnS to examine the impact on their infrared optical properties. In neutron irradiation tests, ZnSe and ZnS samples showed minimal transmittance reduction even up to 1.44 × 1016 n/cm2 (evaluated by dpa value, the values are 5.77 × 10−6, 2.07 × 10−5 for ZnSe and ZnS, respectively), about 5.8 times higher than the expected neutron fluence in ITER. Gamma irradiation at 100 kGy did not affect any of the samples, however, at 5.6 MGy (2.6x the expected dose of ITER), non-antireflective-coated ZnSe and ZnS exhibited induced absorption bands. Antireflective coating did partially mitigate this effect, but ZnS, even with antireflective coating, showed up to 2.5 % degradation over the wide wavelength range of 1.5–4.5 µm, whereas ZnSe was less affected. Both ZnSe and ZnS exhibited a modest 2 % absorption around 3 µm. Surface analysis by SEM, laser microscope, and XPS linked the 3-µm degradation to gamma-irradiation-induced Zn(OH)2 formation that was confined to the surface. Thus, transmittance degradation of ZnSe over the lifespan of ITER was estimated to be a maximum of approximately 7.8 %. The practical use of ZnSe in the IRTh system requires further evaluation with Si and CaF2 lenses and for the optical design to be finalized. Nevertheless, these findings suggest that ZnSe is a promising candidate for the IRTh system's lenses.

Impact of gas injection location and divertor surface material on ITER fusion power operation phase divertor performance assessed with SOLPS-ITER *

The ITER divertor design and performance assessment, primarily based on the SOLPS-4.3 burning plasma database (Pitts R. et al 2019 Nucl. Mater. Energy 20 100696), assumes the use of beryllium (Be) as the divertor surface material and the injection of gas from the main chamber top. However, the current ITER baseline favors gas injection from the more toroidally symmetric sub-divertor region. This paper evaluates the implications of these assumptions for divertor performance in the ITER fusion power operation phase. The impact of the divertor surface material and the gas injection location on the main ions mirrors the hydrogen only low power phase scenario shown in Park J.-S.et al (2020 Nucl. Fusion 61 016021). However, during burning plasma operation, extrinsic impurity seeding will be required. In the case of neon (Ne), studied here, impurity retention is influenced by both the divertor surface material and the fueling location. Neon leakage increases due to more energetic reflection from tungsten than beryllium, but equivalent divertor performance can be achieved by adjusting the neon seeding rate. While the impurity seeding location does not affect the distributions of impurity or radiation, the fueling location does. Top fueling provides local ionization sources mainly in the mid-SOL under detached conditions, enhancing divergences of the flux there (source-driven flow), bringing stagnation points close to the fueling location, and equilibrating flows towards both targets. In contrast, the global flow pattern (in the absence of fluid drifts) in the case of sub-divertor fueling is biased towards the inner target. Impurity flows, driven by force balance, largely mirror those of the main ion flow, including the stagnation point. The case with top fueling enhances Ne retention and corresponding radiation in the outer divertor, effectively reducing the total and peak target heat fluxes by 20%–40%, compared to the case with divertor fueling. Meanwhile, the case with outer target fueling also achieves similar reductions by enhancing plasma-neutral interactions. These results suggest the possibility that the selection of the fueling location and throughput can be used as an actuator to control impurity divertor retention and divertor radiation asymmetry.

Stability of beryllium-tungsten coatings under annealing up to 1273 K

Tungsten and beryllium were foreseen to be the plasma facing materials (PFMs) for the first wall and divertor in ITER. Particles eroded from PFMs surfaces will be transported by the plasma and co-deposited in different locations of the vessel walls. Elemental mixing in the exposed surfaces enables the formation of distinct phases that may influence the stability of PFMs in a wide temperature range. In the present work, Be and W coatings were deposited on W and Be plates, respectively, and annealed in vacuum in the 673–1073 K temperature range. Additionally, Be75W25 coatings deposited on Be plates were annealed in the 873–1273 K range. The Be-W chemical reactivity, phase formation and coatings stability were followed by ion beam analysis, X-ray diffraction and electron microscopy. The reactivity is too weak up to 973 K. It becomes fast at 1073 K with a competitive formation of Be2W and Be12W. At 1173 K, a strong Be22W formation occurs, and Be22W is the only stable phase at 1273 K. The growth of BeO occurs at the coatings surface and increases with temperature. Despite the presence of distinct phase structures, the experiment evidences that the growth of tungsten beryllides enhances the mechanical stability of the coatings from 973 K to 1073 K. In opposition, Be22W formation at 1173 K and 1273 K induced fracture and split off from the Be plates.

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.

Assessment of the ITER divertor bolometer diagnostic performance

The ITER bolometer diagnostic will provide measurements of the total radiation emitted from the plasma, a part of the overall energy balance. It will consist of some 550 lines of sight (LOS) of bolometer detector, bundled in multiple individual cameras, which will be located in the gaps between blanket modules on the vacuum vessel wall, in five divertor cassettes, in two upper port plugs and in one equatorial port plug (Meister et al., 2017). The LOS are optimised as much as possible with design constraints to capture the details of different plasma regions in which the intensity and local distribution of radiation will vary over a large range. This is especially true in the divertor, where up to 70 % of the total thermal exhaust power from the core will have to be radiated during burning plasma operation. This radiation will be tightly concentrated meaning that not only will tomographic reconstruction of the distribution be challenging, but the bolometer cameras themselves need to be properly designed to handle the resulting heat fluxes. This paper first presents an assessment of the photonic heat fluxes falling on bolometers in the divertor region during high performance operation (where the heat fluxes will be highest). These heat fluxes are computed using ray-tracing from radiation distributions obtained with the SOLPS (plasma boundary region) and JINTRAC (plasma core region) codes and will serve as essential input to the thermal design studies of bolometers. The same SOLPS and JINTRAC simulations are used as models for the second focus of the paper which examines how closely the radiation distribution obtained from tomographic inversion of synthetic signals from the bolometer LOS matches the model input. Some effort has also been devoted to a study of how the reconstructions are affected by loss of individual LOS.

A model of the W/Cu interface in the ITER cooling monoblocks from density functional theory

ITER divertor is built with tungsten monoblocks that contain a tungsten–copper interface. There, hydrogen isotopes could possibly accumulate leading to safety and mechanical issues. As a consequence, the tungsten–copper interface has to be modeled and characterized, which is here performed at the atomic-scale by means of density functional theory calculations. In order to build the model, we selected the tungsten and copper orientations that minimizes the mismatch between both networks; this results in the W(001)/Cubcc(001) and W(001)/Cufcc(001)R45° interfaces. After relaxation, both systems converge to the same Wbcc(001)/Cuhcp(112̄0) structure, which is consistent with previous experimental observations. Such reconstruction of the copper network has the effect of changing the charge density in the copper part of the interface, with possible effects on hydrogen interaction.

Evaluation of ITER divertor shunts as a synthetic diagnostic for detachment control

Reliable diagnostics that measure the detached state of the ITER divertor plasma will be necessary to control heat flux to the divertor targets during steady state, burning plasma operation. This paper conducts an initial exploration into the feasibility of the divertor shunt diagnostic as a lightweight, robust, and real-time detachment sensor. This diagnostic is a set of shunt lead pairs that measure the voltage drop along the divertor cassette body, from which the plasma scrape-off layer (SOL) current is calculated. Using SOLPS-ITER simulations for control-relevant ITER plasma scenarios, the thermoelectric current magnitude along the SOL is shown to decrease significantly with the onset of partial detachment at the outer divertor target. Electromagnetic modelling of a simplified divertor cassette is used to develop a control-oriented inductance-resistance circuit model, from which SOL currents can be calculated from shunt pair voltage measurements. The sensitivity and frequency-response of the resulting system indicates that the diagnostic will accurately measure SOL thermoelectric currents during ITER operation. These currents will be a good measure of the detached state of the divertor plasma, making the divertor shunt diagnostic a potentially extremely valuable and physically robust sensor for real-time detachment control.

Feasibility of raised inner strike point equilibria scenario in ITER for detritiation from beryllium co-deposits

In ITER, tritium retention primarily occurs through co-deposition with beryllium. To avoid exceeding the strict tritium inventory limit, efficient tritium recovery techniques are essential. Baking is the ITER baseline for tritium recovery, but its effectiveness in removing tritium from thick beryllium layers is limited. A raised strike point scenario is considered an alternative method for removing tritium from the ITER inner vertical divertor target by heating components via plasma flux. This paper presents SOLPS-ITER code simulations conducted under various conditions, assessing the divertor performance and tritium outgassing of the raised strike point scenario. As the strike point is raised, recycled neutrals are not efficiently baffled by the dome and scrape-off layer, significantly changing the neutral trajectory and ionization source distribution. This improves detachment accessibility but worsens core-edge compatibility compared to the baseline scenario. However, in the partially detached condition, the impact of raising the strike point, perpendicular transport, and q 95 on target heat flux is not significant, as it primarily scales with the input power. Target heat flux is translated to target surface temperature using a simplified heat transfer model that considers the 3D target monoblock geometry and active cooling condition, excluding Be layer thermal properties. For partially detached divertor conditions, the bulk tungsten monoblock surface temperature remains below the baking temperature, which is insufficient for efficient tritium outgassing under the actively cooled ITER divertor condition. However, considering the potential thermal contact resistance between the beryllium and tungsten layers, which may significantly impact temperature distribution, the temperature of the beryllium layer can be raised to a level sufficient for efficient tritium outgassing. Therefore, the raised strike point scenario can be considered as an alternative in-vessel tritium removal technique.

Recent progress of plasma exhaust concepts and divertor designs for tokamak DEMO reactors

The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER-level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Praddiv) for larger Psep/Rp ≥ 25MWm−1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm−2 for the large fraddiv = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard-enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm−2.Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 °C and 200 °C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike-point when the high qtarget of ∼10 MWm−2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high-Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 °C/ 5 MPa (EU-DEMO) and 290 °C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design.

Development of the optical remote handling connector for ITER Divertor Operational Instrumentation

Divertor Operational Instrumentation (DOI) measures mechanical, electromagnetics and thermal parameters of the ITER Divertor Cassettes. The maintenance of the Cassettes requires their periodic removal and replacement by remote handling tools. The remote handling connector (RHC) of the DOI includes electrical and optical connections for the instrumentation cables, which must be disconnected during maintenance. The available space for the DOI RHC design is limited and the environmental conditions inside the Vacuum Vessel include high vacuum 10^-5 Pa, high temperature up to 350 °C and neutron and gamma-ray irradiation up to 1 GGy by the end of ITER operation. The access of remote handling tools is limited in the operating space surrounding DOI RHC. A study of existing connectors designed for plasma diagnostics electrical cables indicated that the concepts could not be adapted as such for the DOI RHC. The main concerns were related to the increased limitation of space, and the need to incorporate optical contacts for the fiber-optic cables that should be aligned with high precision to minimize optical losses. A dedicated design was required and developed. A novel extending DOI RHC concept was selected as it minimizes the required design changes to the surrounding systems. The concept consists of a cassette socket with an extending link located on the Divertor Cassette and a permanently installed vacuum vessel socket. The concept relies on the actuation of linear motion by the bolting tools of Cassette Toroidal Mover (CTM) manipulator and the flexibility of the cabling during the actuation. A physical full-size mock-up of the DOI RHC was manufactured to test the design solutions. The functionality of the mechanical structure in regards to remote handling operations were tested at remote handling connector platform (RHCP) at VTT Tampere Finland. Suppliers of optical connectors and ferrules were contacted to discuss the requirements of DOI RHC and commercial off-the-shelf optical components. Since there is no commercially available solution for attaching the fiber to the ferrule which would meet the aforementioned requirements, initial soldering tests were performed. The developed DOI RHC design provides input to the following design phases.

Coupled simulations with SOLPS-ITER and B2.5-Eunomia for detachment experiments in Magnum-PSI

Heat loads of 10 MW m−2 are expected for steady state operation at ITER and up to 20 MW m−2 in slow transient situations. Plasma linear devices like Magnum-PSI can recreate situations close as those expected to be achieved at ITER divertor, providing easier access for diagnostics than in a tokamak. Numerical models are still necessary to complement experiments and to extrapolate relevant information to fusion devices, as the relevant atomic and molecular processes. SOLPS-ITER (formerly known as B2.5-Eirene) is typically employed to solve the plasma and neutral distribution in a coupled way for tokamak devices. For Magnum-PSI, B2.5 has been coupled with a different neutral module, named Eunomia, developed mostly for linear devices. Nevertheless, there is an interest in using SOLPS-ITER for simulating Magnum-PSI, as it would ease the process of relating linear device results with tokamaks. A previous work found significant differences in the implementation of relevant plasma-neutral processes in Eirene and Eunomia. A wide range of plasma scenarios are compared between B2.5-Eunomia and SOLPS-ITER. Although both codes produce results close to experimental Thomson scattering density and temperature near the target once the electric potential at the source is adjusted, these are achieved with completely different plasma and neutral distributions. Anomalous transport coefficients, which are other of the free-parameters in Magnum-PSI simulation, are set equal between the two codes. When studied in a wide range of neutral pressures, SOLPS-ITER shows a trend closer to experiments, as well as providing a converged solution at neutral pressures higher than 4 Pa for which B2.5-Eunomia was unable to provide a converged solution. Additional measurements of the neutral distribution in the target chamber as well as the electric potential at the source are required to determine which code is producing results closer to the experiment.

Bubble Formation in ITER-Grade Tungsten after Exposure to Stationary D/He Plasma and ELM-like Thermal Shocks

Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, bubble formation in ITER-grade tungsten was detected when exposed to fusion relevant conditions. For this study, ITER-grade tungsten was exposed to simultaneous ELM-like transient heat loads and D/He (6%) plasma in the linear plasma device PSI-2. Bubble formation was then investigated via SEM micrographs and FIB cuts. It was found that for exposure to 100.000 laser pulses of 0.6 GWm−2 absorbed power density (Pabs), only small bubbles in the nanometer range were formed close to the surface. After increasing Pabs to 0.8 and 1.0 GWm−2, the size of the bubbles went up to about 1 µm in size and were deeper below the surface. Increasing the plasma fluence had an even larger effect, more than doubling bubble density and increasing bubble size to up to 2 µm in diameter. When using deuterium-only plasma, the samples showed no bubble formation and reduced cracking, showing such bubble formation is caused by exposure to helium plasma.

Compression Tests at High Strain Rate on 3D-Printed CuCrZr Alloy Specimens - Material Model Calibration

CuCrZr alloy is used to produce actively cooled components for high heat flux elements of beamlines and for heat sink of plasma facing components in nuclear fusion devices such as ITER and DEMO. It has an excellent thermal conductivity and specific mechanical strength, together with a high electrical conductivity that is giving high push to its use. Recently, CuCrZr alloy was also considered as an attractive material for Additive Manufacturing, leading to extend its application in the field of strain rate studies. As a matter of fact, its strain rate dependency is playing an important role for vertical target plasma-facing units components uses as heat sink in the ITER divertor or as structural material for actively cooled plasma facing components. This paper describes the results obtained by quasi-static and dynamic compression tests carried out on CuCrZr specimens produced by laser Powder Bed Fusion (PBF), with Selective Laser Melting (SLM) technology. Quasi-static tests have been conducted by means a servo-hydraulic tensile machine, while a direct tension-compression split Hopkinson bar has been used to perform the tests at high strain rate. Since dedicated heat treatments are required to obtain optimal combination of strength, ductility, and conductivity, some of the specimens have been heated up to 560 – 580 °C for 4 – 5 h and then cooled in air. Eventually, the calibration of the most appropriate constitutive models for 3D-printed CuCrZr alloy deformed at high strain rate has been carried out by means an inverse analytical procedure.

Development of Methods for Determining Near-Surface Plasma Parameters During Tests of Fusion Reactor First-Wall Prototypes Using the PLM Device

Plasma-surface interaction and high heat flux on plasma-facing materials in magnetic fusion devices cause surface ablation and degradation, while the influx of eroded materials into plasma can have a shielding effect. The reduction of the power load due to the plasma detachment effect over tungsten fuzz is an important phenomenon to be investigated for the ITER divertor problem. Measuring near-wall plasma parameters is a challenging task, requiring the development of improved and advanced techniques, including high-resolution spectroscopic methods. In this paper, we present study results of steady-state plasma over tungsten fuzz formed in plasma linear multicusp (PLM). The PLM device is a linear plasma trap composed of an eight-pole multicusp magnetic field with steady-state plasma discharge with parameters similar to the scrape-off layer and divertor plasma in a tokamak. We used spectroscopic measurements to estimate spatial distributions of plasma radiation in the vicinity of the sample surface exposed to the plasma column. Thus, we obtained information on the temperature and composition of the boundary layer plasma and the temperature of the sample surface. Helium plasma exhibits ionization-type nonequilibrium even at atmospheric pressure, necessitating the use of specific methods to estimate its electron temperature Te. When the helium ion spectral line He II 468.5 nm is present in the spectra, its intensity ratio to one of the atomic lines He I can be described by using coronal approximation. Spectrum analysis has shown that emitting helium ions are highly sensitive indicators of average electron energy = 3kTe/2. Therefore, utilizing intensity ratios of the strongest emitting lines in the ultraviolet-visible near-infrared range, He II 468.6 nm and several He I lines with well-known electron excitation functions were found to be a reliable Te measurement method in the case of magnetized low-pressure helium plasma. We also propose a method for determining the concentrations of the metallic admixture in the plasma on the data on relative intensities of its spectral lines.

The effect of recrystallization annealing on the tungsten surface carbidization in a beam plasma discharge

<abstract> <p>Tungsten was chosen as the plasma facing material (PFM) of the ITER divertor. However, graphite and carbon-graphite materials are used as PFM in some research thermonuclear facilities, including the Kazakhstan materials science tokamak. This circumstance determines the interest in continuing the study of the formation of mixed layers under plasma irradiation. This article is devoted to the study of the effect of preliminary recrystallization annealing on the carbidization of the tungsten surface in a beam-plasma discharge (BPD), which is one of the ways to simulate the peripheral plasma of a tokamak. Experiments on preliminary isochoric and isothermal annealing of tungsten samples were carried out in the mode of direct heating of tungsten samples by an electron beam. The carbidization of tungsten samples after annealing was carried out in a methane atmosphere in the BPD at a temperature of 1000 ℃ for a duration of 3600 s. Optical microscopy (OM) and X-ray diffraction were used to analyze the structure of the tungsten surface. It has been established that differences in the structure arising during recrystallization annealing affect the transfer of carbon atoms in the near-surface area of tungsten and the formation of tungsten carbides (WC or W<sub>2</sub>C).</p> </abstract>

A reduced model for the ITER divertor based on SOLPS solutions for ITER Q = 10 baseline conditions: B. A reduced model based on reversed-direction two point modeling * ∗This manuscript has been partially authored by UT-Battelle, LLC, under Contract DE-AC05-00OR22725 with the US Department of Energy. This material is based upon work supported the Scientific Discovery through Advanced Computing

Edge codes such as SOLPS coupled to neutral codes such as EIRENE have become so comprehensive and sophisticated that they now constitute, in effect, ‘code-experiments’ that, as for actual experiments, can benefit from interpretation using simple models and conceptual frameworks, i.e. reduced models. The first task is the identification of options for the reduced model control parameters that are best suited for control of the action of the divertor, i.e. for control of target power loading and sputter–erosion, primarily. A strong correlation between the electron temperature at the divertor target, T e,t, and the neutral deuterium D2 density at the target, n D2,t, flux-tube resolved, has recently been reported for a number of code studies including SOLPS-4.3 modeling of a set of ∼50 ITER baseline cases: Q DT = 10, q 95 = 3, P SOL = 100 MW, metallic walls, and Ne seeding (Pitts et al 2019). Part A of the present study reports new results for largely the same ITER cases, confirming the strong correlation reported earlier between local values of T e,t, and (i) n D2,t, and (ii) normalized volumetric losses of power and pressure in the divertor. Strong correlations have now also been found, and are reported here for the first time, between T e,t and all of the divertor target quantities of practical interest. A physical explanation for this surprising result has not as yet been fully identified; nevertheless it has encouraging implications for reduced modeling of the ITER divertor. For such ITER conditions, (i) the global Ne injection rate, InjNe (Ne s−1), and (ii) the electron temperature at the location on the target where the peak power deposition occurs, T e,t@q⊥,pk (eV), are found to be promising reduced model control parameters. In this part B, a reduced model for the ITER divertor is developed and described in detail, based on reversed-direction two point modeling, Rev2PM. The input to the reduced model is a value of the variable pair for a chosen case and the output are values of the various target as well as divertor-entrance quantities of practical interest, e.g. q ⊥,pk, the electron density at the X-point, n e,Xpt, etc. The reduced model was quantitatively characterized using one half of the code cases; it was then used to successfully predict (replicate) the code values of e.g. n e,Xpt for the other half.