Future Fusion Devices Research Articles

Surface modification of ZrC dispersion-strengthened W under low energy He plasma irradiation

ZrC dispersion-strengthened W exhibits high strength/ductility, low ductile-to-brittle transition temperature, and excellent thermal shock resistance, making it a promising candidate plasma-facing material for future fusion devices. In this study, surface modification of 0.5 wt.% ZrC dispersion-strengthened W (WZrC) under low energy and high fluence He plasma irradiation at high temperature was presented. Under the energy of 90 eV and fluence ranging from 6 × 1024 He·m−2–2 × 1026 He·m−2 He irradiation at 920 °C, a typical fuzz nanostructure appeared on the W matrix of WZrC. The thickness of fuzz layer is proportional to the square root of He irradiation fluence. The fuzz showed comparable thickness and structural features to pure W, indicating limited effects of the particle’s addition on resistance to high fluence He irradiation at high temperatures. Under continuous He injection, the fuzz would grow extending onto the particle area, making the particle obscured. Besides, the erosion behavior of particles under He plasma irradiation has been investigated, which is thought to be dominated by a sputtering process. Under the He influence of 6 × 1024 He·m−2, only nanopores were observed in the surface region. With fluence increasing to 5 × 1025 He·m−2, the surface became relatively uneven with larger holes. W aggregated in spots and distributed on the surface of the particle, which might be the result of subthreshold sputtering and deposition. When fluence further increased to 2 × 1026 He·m−2, the particles were eroded completely and covered by the extended fuzz, forming cavities. In addition, distinctive layered nanotendrils were observed above the cavities, which were characterized to consist of inner W-riched skeletons and outer Zr-riched layers. It indicates that the layered nanotendrils should be the result of fuzz extension combined with particle sputtering/deposition.

High-parameters experimental linear plasma device for fusion wall conditioning studies

In future, HELPD would provide valuable platform for investigating the issue in the W impurity control, towards the wall conditioning requirements in tokamak.The tungsten (W) impurity caused by sputtering of the first wall is considered as an issue influencing the plasma safe operation of EAST, ITER and other future fusion devices. As a supplement to the experiments on the tokamak, the linear plasma device (LPD) serves as an effective divertor simulator for the W impurity control research. This work presents the design and construction of High-parameters Experimental Linear Plasma Device (HELPD) in detail, including its layout and operation parameters. Besides the function of plasma exposure, the target sample could be exchanged in a separate material pretreatment vessel for wall conditioning, such as lithium coating and glow discharge cleaning (GDC) without breaking vacuum conditions. And the preliminary results show that the device could produce a steady state plasma of 1∼10 eV, and its flux could reach 1022 m−2s−1 with 10 kW helicon wave. In the future, HELPD would provide valuable platform for investigating the issue in the W impurity control，towards the wall conditioning requirements in tokamak.

Low-frequency whistler waves driven by energetic electrons in plasmas of solely electron cyclotron wave heating

Whistler waves are a type of low-frequency electromagnetic wave common in nature, which is usually associated with energetic electron phenomena. This study presents experimental observations of low-frequency whistler wave instabilities driven by energetic electrons through wave–particle interactions on EXL-50. The energetic electrons are generated by electron cyclotron waves (ECWs) through stochastic heating [Wang et al., J. Plasma Phys. 89, 905890603 (2023)] and do not match the characteristics of the runaway electrons [Shi et al., Nucl. Fusion 62, 086047 (2022)]. In the steady-state plasma of the Energy iNNovation XuanLong-50 (EXL-50), whistler waves within the 30–120 MHz frequency range were observed during electron cyclotron resonance heating. These waves displayed multiple frequency bands, and the frequencies of waves were directly proportional to the Alfvén velocity. Furthermore, it was interesting to find that superposition of lower hybrid wave into ECW resulted in the suppression of these whistler waves. The experimental results may indicate that the whistler waves are driven by energetic electrons (excluding runaway electrons). These discoveries carry significant implications for several areas of research, including the investigation of wave–particle interactions, the development of radio frequency wave current drivers, their potential impact on the electron dynamics in future fusion devices, and even the presence of unusually low-frequency whistler waves in Earth's radiation belts.

Simultaneous reproduction of experimental profiles, fluxes, transport coefficients, and turbulence characteristics via nonlinear gyrokinetic profile predictions in a DIII-D ITER similar shape plasma

Experimental conditions obtained on the DIII-D tokamak in the ITER Similar Shape (ISS) have been compared extensively with nonlinear gyrokinetic simulation using the CGYRO code [Candy et al., J. Comput. Phys. 324, 73–93 (2016)] with comparisons spanning ion and electron heat fluxes, electron and impurity particle transport, and turbulent fluctuation levels and characteristics. Bayesian optimization techniques [Rodriguez-Fernandez et al., Nucl. Fusion 62(7), 076036 (2022)], combined with nonlinear gyrokinetics, have been used to obtain simultaneously Qi, Qe, and Γe flux-matched profiles that are found to be in good agreement with experimental profile measurements. Synthetic diagnostics were used to compare measured beam emission spectroscopy and correlation electron cyclotron emission turbulent fluctuations with nonlinear simulation. Although some disagreements exist, nonlinear simulations are found to be in generally good agreement with measured fluctuation levels, spectral shapes, and measured radial trends in low-k δne/ne and δTe/Te. Low (Li and C) and mid-Z (Ca) impurity transport was also compared with these flux-matched simulations. Fully stripped, low-Z impurities are well reproduced by the gyrokinetic modeling while clear disagreement exists in comparisons with mid-Z impurities. Nonlinear gyrokinetic investigation into the Z dependence of impurity transport in the ISS conditions is also performed, demonstrating clear trends of impurity diffusion with impurity Z (both D∝Z and D∝1/Z) that vary with the radial location studied. These trends are shown to result from the local dominance of ion temperature gradient or ∇n driven trapped electron mode turbulence and may contribute to the disagreement between simulation and experiment in mid-Z impurity transport. The results of this work represent one of the most complete validation studies of the gyrokinetic model performed to date and provide an example of new capabilities for predicting performance in future fusion devices.

The angle of incidence dependence of the sputtering energy threshold of tungsten in future nuclear fusion device

The angle of incidence dependence of the sputtering energy thresholds (Eth) for tungsten targets irradiated by Helium, Tritium, and Deuterium ions was studied using two different simulation programs (RDS-BASIC and SDTrimSP). For each of the studied irradiation systems, the sputtering energy threshold was found to slightly decrease (3 to 6 eV) by increasing the angle of incidence of the irradiating ion beam. This decrease was found to vary from one irradiation system to another and give a minimum Eth value which was found to occur at a certain angle range for each system. Departing from this minimal Eth angles toward higher angles was found to result in an abrupt and more dramatic increase in the Eth values which can exceed the normal incident Eth values, especially at grazing angles. Our results were also compared and discussed in light of some existing theories.

Magnetic diagnostics layout design for CFETR plasma equilibrium reconstruction

Plasma equilibrium reconstruction provides essential information for tokamak operation and physical analysis. An extensive and reliable set of magnetic diagnostics is required to obtain accurate plasma equilibrium. This study designs and optimizes the magnetic diagnostics layout for the reconstruction of the equilibrium of the plasma according to the scientific objectives, engineering design parameters, and limitations of the Chinese Fusion Engineering Test Reactor (CFETR). Based on the CFETR discharge simulation, magnetic measurement data are employed to reconstruct consistent plasma equilibrium parameters, and magnetic diagnostics’ number and position are optimized by truncated Singular value decomposition, verifying the redundancy reliability of the magnetic diagnostics layout design. This provides a design solution for the layout of the magnetic diagnostics system required to control the plasma equilibrium of CFETR, and the developed design and optimization method can provide effective support to design magnetic diagnostics systems for future magnetic confinement fusion devices.

Modeling of tungsten divertor target erosion induced by impurity during edge‐localized modes by using a kinetic model

AbstractThe burst of edge‐localized modes (ELMs) leads to an increase in the energy and particle fluxes to the divertor target. Tungsten (W) is chosen as the primary candidate material for plasma‐facing components (PFCs) in the future fusion devices, and EAST has already upgraded all divertors to use W. Therefore, understanding tungsten target erosion during ELMs and finding the correlation between erosion rate and key ELM parameters are crucial for steady‐state operation. In this work, based on the Vlasov–Poisson model (VPM), we develop a one‐dimensional kinetic parallel transport code to investigate the parallel transport of particles in the EAST device during ELMs and the resulting target erosion. The EAST experiment (#102182) is simulated by VPM code. The simulation results are compared with experimental data as well as free‐stream model (FSM) calculation, showing the accuracy of the code. Considering the presence of lithium (Li) impurities in EAST discharge, the erosion of the W target is simulated. The results indicate that during the burst of ELM, the total average tungsten erosion rate, , is determined by both deuterium (D) and Li ions. D ions dominate the erosion when the ELM frequency (fELM) is low (ranging from 50 to 175 Hz), while Li impurities become more important than D+ in high‐frequency ELMs (fELM > 175 Hz). As fELM increases, the time‐averaged erosion of the W target first increases and then decreases. Therefore, the reduction of W erosion benefits from high‐frequency ELMs, with impurity ions being the primary contributor to the erosion.

Influence of microstructure on thermal fatigue effect of laminated tungsten based plasma-facing material

The response of tungsten (W) to thermal shock loading, as the best candidate for plasma-facing material (PFM), is an important issue in the research of future fusion devices. Under thermal loading, thermal irradiation damage, including brittle cracking and fatigue cracking, occurs on the surface of tungsten based plasma-facing material (W-PFM). In this work, a new scheme to suppress the thermal irradiation damage to W-PFM, i.e. the laminated structure W-PFM scheme, is proposed. Thermal fatigue experiments of laminated structure W composed of W foils with different thickness and heat treatment processes are carried out by using an electron beam device. The samples are subjected to thermal pulses with a power density of 48 MW/m<sup>2</sup> for 5000 cycles. The results indicate that the crack damage to the surface of the laminated structure W decreases with the decrease of the thickness of W foils under the same heat treatment conditions. The main cracks are produced on the surface of laminated structure W after cyclic thermal loads have been all approximately parallel to the foil thickness direction. Only the main cracks appear on the surfaces of W foils with a smaller thickness, while crack networks develop on the surfaces of W foils with a larger thickness , in addition to the main cracks with a larger width. In the rolled state, the laminated structure W has the lowest degree of surface plastic deformation for the same thickness. The thermal fatigue crack damage to the surface is quantitatively analyzed by using computer image processing software and analysis software, and scanning electron microscope images of the thermal damage area are finally selected. It is found that the de-stressed state W has the smallest crack area and the smallest number of cracks for the same thickness, indicating that the de-stressed state W has the strongest resistance to irradiation damage. The experimental results also show that in addition to the effect of microstructure, both the uniaxial stress state and the crack-blocking mechanism of the laminated structured W-PFM contribute to the improvement of its thermal fatigue performance.

First results of high density H-mode operation in metal-wall EAST tokamak

In metal-wall EAST superconducting tokamak, H-mode operation with plasma density close to the Greenwald density limit nGW has been achieved with radio frequency and NBI heating for the first time. Both gas puffing from horizontal plane and HFS pellet fueling were used for density ramp-up during the experiment. The confinement of H-mode gradually deteriorates with plasma density increasing. And the H-L transition can be observed after heating power dropping or pellet injection. In the density range of (0.6–1)×nGW, the divertor detachment occurs and causes an obvious confinement degradation. The maximum accessible density ne, max in H-mode phase deviates from the Greenwald scaling. It has been observed that the fraction ne, max/nGW is almost independent of the total heating power, but a high heating power is helpful to extend the duration of high density H-mode. And ne,max/nGW has an increase relation with the safety factor q95 varied by changing plasma current. Besides, it is also found that the discharges with low plasma current have a higher ne, max/nGW than those with high plasma current. All these findings will provide a good reference to the high density plasma operation for future metal-wall fusion devices.

Density wavenumber spectrum measurements, synthetic diagnostic development, and tests of quasilinear turbulence modeling in the core of electron-heated DIII-D H-mode plasmas

Measurements of the turbulent density wavenumber spectrum, δnˆe(k⊥) , using the Doppler Back-Scattering (DBS) diagnostic are reported from DIII-D H-mode plasmas with electron cyclotron heating as the only auxiliary heating method. These electron-heated plasmas have low collisionality, νe∗<1 , Te/Ti>1 , and zero injected torque—a regime expected to be relevant for future fusion devices. We probe density fluctuations in the core (ρ ≈ 0.7) over a broad wavenumber range, 0.5⩽k⊥⩽16 cm−1 ( 0.1⩽k⊥ρs⩽5 ), to characterize plasma instabilities and compare with theoretical predictions. We present a novel synthetic DBS diagnostic to relate the back-scattered power spectrum, Ps(k⊥) —which is directly measured by DBS—to the underlying electron density fluctuation spectrum, δnˆe(k⊥) . The synthetic DBS Ps(k⊥) spectrum is calculated by combining the SCOTTY beam-tracing code with a model δnˆe(k⊥) predicted either analytically or numerically. In this work we use the quasi-linear code Trapped Gyro-Landau Fluid (TGLF) to approximate the δnˆe(k⊥) spectrum. We find that TGLF, using the experimental profiles, is capable of closely reproducing the DBS measurements. Both the DBS measurements and the TGLF-DBS synthetic diagnostic show a wavenumber spectrum with variable decay. The measurements show weak decay (k −0.6) for k < 3.5 cm−1, with k −2.6 at intermediate-k ( 3.5⩽k⩽8.5 cm−1), and rapid decay (k −9.4) for k > 8.5 cm−1. Scans of physics parameters using TGLF suggest that the normalized ∇Te scale-length, R/LTe , is an important factor for distinguishing microturbulence regimes in these plasmas. A combination of DBS observations and TGLF simulations indicate that fluctuations remain peaked at ITG-scales (low k) while R/LTe -driven TEM/ETG-type modes (intermediate/high k) are marginally sub-dominant.

Experimental investigation of axial tensile and fatigue behaviors of HTS round strands

For the development of high-temperature superconducting (HTS) magnet systems of future fusion devices, a novel HTS round strand based on a stacking structure was designed and manufactured using second generation (2G) HTS tapes. Different mechanical loads during operation can result in irreversible degradation of the strand. The axial tension and fatigue loads need particular attention. Therefore, it is important to investigate the electromechanical behavior of the round strand under various axial tension and cyclic loads. In this paper, the axial tensile and fatigue tests were conducted at 77 K, self-field. Taking 95% critical current (Ic) retention as the criterion, the results of the tensile tests revealed that the average tensile stress and strain were as high as 344 MPa and 0.47%, respectively. Fatigue characteristics were also investigated as a function of axial tensile stress. No significant performance degradation was observed up to 100,000 loading cycles with stress amplitudes ranging from 20 MPa to 200 MPa. Ic degradation occurs after 16,000 loading cycles with 380 MPa as the maximum stress. Furthermore, the microscopic defects of the round strand samples due to fabrication imperfections and mechanical loading were investigated using metallographic microscope and scanning electron microscope. These results presented in this paper are useful for comprehending and improving the mechanical behaviors of the strand in high-field and large-scale fusion magnet systems.

FIB line marking as a tool for local erosion/deposition/fuzz formation measurements in ASDEX Upgrade during the He campaign

Tungsten (W) is the prime candidate for the plasma facing material in present and future fusion devices. When exposed to a plasma containing He ions, W can exhibit creation of sub-surface nano-bubbles leading to formation of nano-tendrils called fuzz. Formation of W fuzz was confirmed by many laboratory experiments including He loading in linear plasma devices. There is, however, limited experience related to the parameter space for fuzz formation and re-erosion in a tokamak environment. For this reason, a dedicated experiment was carried out in ASDEX Upgrade (AUG) during its 2022 He campaign to study the evolution of fuzz-like W structures.Twelve tungsten samples were mechanically polished to a mirror-like surface finish. Six of them were subsequently exposed to a He plasma in the PSI-2 linear device in order to establish a fuzz layer with a thickness of 600 – 800 nm. Before exposure in AUG, all samples were pre-characterized by Focused Ion Beam (FIB) cross-sectioning. In total 48 FIB cross-sections with line markings for quantification of local erosion, deposition and fuzz formation were prepared – 4 on each sample. The samples were placed in two parallel poloidal rows spanning a range of 20 cm around the outer strike line position (OSP). They were subsequently exposed in AUG to a series of 14 consecutive discharges, 8 in H-mode and 6 in L-mode.Detailed analysis by means of electron microscopy revealed on the samples regions of erosion, deposition and fuzz formation. Below the H-mode OSP, homogeneous co-deposits containing W and O, with a thickness up to 400 nm, were found. In the close vicinity of the H-mode OSP, significant erosion of pre-exposure PSI-2 fuzz was observed. The erosion reached up to 100 – 250 nm, depending on the location. In addition, the initially polished samples did not show any newly formed fuzz in that zone. Above the H-mode OSP, new fuzz was formed with a thickness of up to 1 μm. There was no Mo found in the newly formed fuzz. Pre-exposure PSI-2 fuzz was either removed or modified.

Measurement of divertor neutral pressure in EAST tokamak by ASDEX pressure gauge

In the tokamak plasma operation, divertor neutral particles play a key role in the study of fuel recycling, heat load mitigation, plasma confinement, and so on. Therefore, the in-situ measurement of neutral pressure in the divertor region is essential for the divertor physics studies. Two ASDEX pressure gauges were installed in the lower divertor dome region and passive plate region, respectively, for the in-situ pressure measurement. Experimental results show that, the ASDEX pressure gauge diagnostic is more sensitive, and its measurement on divertor neutral pressure is about 100 milliseconds faster than the external gauge of the divertor. Moreover, the ASDEX pressure gauge diagnostic is applicable for long pulse and high power plasma operation, and the in-situ pressure measurement of the divertor during over 200 s and higher than 10 MW discharges were successfully achieved in EAST. In addition, ASDEX pressure gauge diagnostic sensitively reacts to the changes in plasma parameters, and the neutral pressure is found to be significantly influenced by the plasma heating power. These results provide powerful technical support for the key divertor physical studies in EAST tokamak and the future fusion devices.

Modelling of deuterium diffusion and thermal desorption in tungsten exposed to high-flux deuterium plasma

In nuclear fusion, hydrogen isotopes transportation in tungsten has been identified as a crucial process which could affect various key processes in the operation of a fusion device, such as tritium self-sustaining and operational safety. To enhance the understanding of deuterium diffusion and trapping in millimeter-thick tungsten under high-flux plasma conditions, we carried out a systematic simulation investigation on the deuterium plasma exposure and subsequent thermal desorption using the TMAP7 code based on experimental results. The simulation reveals that the defect concentration with a low de-trapping energy , as estimated solely from simulating thermal desorption, is lower than its actual density due to the temperature history of plasma exposure and subsequent cooling processes. Moreover, the growth rate of deuterium mobile concentration exhibits nonmonotonic behavior over time during plasma exposure, which strongly depends on the ratio of defect concentration with a high de-trapping energy to that with a low de-trapping energy. Additionally, we observed that the maximum deuterium diffusion depth does not follow a proportional relationship with the square root of exposure time, which is attributed to the growing number of plasma-induced defects and the reducing injection coefficient with increasing fluences. This work provides valuable insights into the understanding of deuterium transportation in tungsten during high-flux plasma exposure which could contribute to understanding and modelling of hydrogen isotopes recycling in future fusion devices.

Experiences and lessons from the mitigation of leading edge induced melting on actively cooled ITER-like W/Cu monoblocks for divertor target in EAST

EAST is the first tokamak to feature fully actively water-cooled ITER-like W/Cu monoblocks on the divertor target. In 2014 and 2021, the graphite tiles in the upper and lower divertor were upgraded to W/Cu monoblocks, respectively. With the increase in plasma parameters, severe melting phenomena were inspected at the leading edges of upper divertor from 2017 to 2020. Through theoretical analysis and numerical simulation, two main factors, radial misalignment and chamfer structure have been identified to mainly impact the leading-edge induced melting of MBs. It was found that assembly misalignments increased gradually during the plasma discharges from 2015 to 2017 and the maximum misalignment can reach up to 3 mm, which play an important role that is responsible for the melting during 2018 to 2020. In 2020, engineering assembly improvements were implemented, resulting in a significant improvement compared to the previous situation, with all misalignments brought below 1 mm, and thus effectively prevent melting at the leading edge on upper divertor. In addition, the modification of chamfer structure to 1.5 × 17 mm at inter-CMs of the lower divertor further mitigated the issue of leading-edge-induced melting compared to upper divertor. Such experiences and lessons from the mitigation of leading edge induced melting provide important references for future fusion devices.

Effect of material properties on the laser-induced desorption of hydrogen from tungsten

One of the priority tasks for sustainable operations of future fusion devices is the systematic control of hydrogen isotope content in plasma-facing components. A remote non-damaging analysis of the gas content can be realised by applying laser-induced desorption (LID). LID is based on local heating of the studied surface, followed by the detection of desorbed particles. Interpretation of the LID measurements relies on the definition of effective desorption region from which particles are being released. This region strongly depends on the properties of a laser pulse, an analysed material, and defect centres containing hydrogen isotopes. To investigate the role of these properties on LID, a series of numerical simulation was conducted for the case of hydrogen release from a 10 μm tungsten layer, initiated under nanosecond, microsecond, and millisecond heat loads. The simulations were performed using a common reaction-diffusion model with sequentially varied tungsten thermal conductivity, detrapping energy of hydrogen from traps, concentration of hydrogen traps, and their initial filling ratio. Additionally, the Soret effect was considered in order to estimate its impact on the LID efficiency. Findings reveal that the peak surface temperature reached during heating can characterise the LID efficiency under a particular heat pulse. The LID efficiency dependence on the peak surface temperature weakly varies with material thermal properties, but can alter with the change of trap parameters. The detrapping energy of hydrogen from traps considerably influences the desorption efficiency, while concentration and the initial filling ratio of traps have a moderate effect. The Soret effect is shown to noticeably affect the LID efficiency per a laser pulse only at near-melting temperatures of tungsten. Finally, results demonstrate that the highest desorption efficiency from thick layers can be reached using laser pulses with millisecond duration regardless of surface effects, when the release under short-term pulses can be limited by them.

Sputtering yield reduction for nano-columnar W surfaces under D ion irradiation

This study investigates the sputtering properties of nano-columnar tungsten surfaces under 2 keV D2+ irradiation. It was conducted both by performing experiments in a highly sensitive quartz crystal microbalance setup as well as by numerical methods using two simulation codes called SPRAY and SDTrimSP-3D. A key question was whether the strong sputtering yield reduction observed in previous studies under 2 keV Ar+ irradiation is maintained when using a much lighter ion species like deuterium. For the latter, a substantially larger projected range in the material has to be expected. Therefore, a total confinement of the ion-solid interaction within the W nano-columns cannot be assumed a priori. However, both the experiments and the numerical simulations showed in good agreement that the geometrically induced sputtering yield reduction is still observed to the same extent despite the increased range of ions in the material. Thus, a potential application of such nano-columnar tungsten surfaces as plasma facing components in future nuclear fusion devices is not affected.

An IMAS-integrated workflow for energetic particle stability

The confinement of energetic particles (EPs) generated by fusion reactions and external heating methods is crucial for the performance of future fusion devices. However, EP transport can occur due to their interaction with electromagnetic perturbations, affecting heating efficiency and overall performance. Robust reduced models are needed to analyze stability and transport. This paper presents an automated IMAS-based workflow for analyzing the time-dependent stability of EP-driven modes, focusing on the linear properties of Toroidal Alfvén Eigenmodes (TAEs) in general tokamak geometry. The workflow utilizes efficient computational methods and reduced models to deliver fast and reproducible results. A demonstration of the workflow’s effectiveness was performed, identifying key linear properties of TAEs in various simulated ITER scenarios. This approach represents a critical step toward developing tools for analyzing EP transport and optimizing the performance of future fusion reactors.

Investigation of helium exhaust dynamics at the ASDEX Upgrade tokamak with full-tungsten wall

An efficient removal of He ash by active pumping in future fusion devices is necessary to avoid fuel dilution and not degrade the core confinement properties. Therefore, a deep understanding of the underlying physics mechanisms is mandatory. Helium exhaust has been experimentally investigated at the ASDEX Upgrade tokamak. This is an ideal test environment, thanks to the ITER-like divertor geometry, an extensive diagnostics coverage and the presence of plasma-facing components made of tungsten. The exhaust efficiency, characterized by the He compression in the divertor, was found to improve with increasing divertor neutral pressure but to degrade with detachment. A multi-reservoir particle balance model was developed to interpret the observed exhaust dynamics, accounting for plasma transport and wall retention. The limited performance of the pumping system and the efficient helium retention capability of the tungsten wall were identified to have the strongest impact on the exhaust dynamics.

Material Plasma Exposure eXperiment High Heat Flux Microwave Absorber Design, Manufacture, and Articles Test

The Material Plasma Exposure eXperiment (MPEX) at Oak Ridge National Laboratory is in the final design phase. MPEX will be capable of exposing neutron-irradiated materials to plasmas for the study of plasma-material interaction. This facility will provide information about the complex effects of plasmas on materials and contribute to examining new materials that can withstand high heat fluxes and high ion fluences for future fusion devices. MPEX plasma is heated by 70-GHz or 105-GHz electron Bernstein wave/electron cyclotron heating (ECH), and the high-frequency microwaves are prone to scattering microwave power, which can have detrimental effects, especially on diagnostic components. A large portion of the injected ECH power is expected to be absorbed by plasma, but the remainder requires that microwave absorbers be placed immediately upstream and downstream of the ECH launcher to minimize stray microwaves leaving the ECH region. These microwaves can inadvertently heat components that cannot be shielded or otherwise protected. The microwave absorber design is based on an array of pyramid-shaped ceramic tiles brazed to a water-cooled explosion-bonded heat sink and a stainless steel plate to produce one tile module. Computational fluid dynamics and structural analyses were performed to optimize and validate the design. Multiple test coupons were produced to validate the process for brazing the two different tile materials to the Glidcop AL-15 baseplate. The articles were tested to evaluate the reliability and thermal performance through exposure to an electron beam with a heat flux of up to 1.5 MW/m2. Nondestructive testing was performed before and after testing to identify voids or separations that may have been introduced by the high heat flux. This paper discusses the details of high heat flux microwave absorber design, manufacturing details and associated challenges, and test results, demonstrating the effectiveness of the proposed design.