Recent Publications

kobra: a new Vlasov code intended for plasma-wall modeling

Sebastian Konewko, Nathan Maestracci, Sven Van Loo

In a fusion device plasma-wall interactions \edit{on the sheath scale} can be modeled as a collisionless problem. When modeling these regions particle-in-cell codes suffer from statistical error originating from undersampling the velocity space. On the other hand, Vlasov codes do not have this issue as they evolve the full distribution function. Here, we present a new finite-volume Vlasov code, kobra, equipped with adaptive-mesh refinement to reduce computational effort. Currently, the code solves the Vlasov-Poisson equations. We validate our code in 1d1v and 1d2v using established benchmarks, i.e. the two-stream instability, Landau damping, the Dory-Guest-Harris instability, and also a classical electrostatic plasma sheath. We find that the code reproduces the theoretical properties of these problems well. More importantly, the adaptive grid provides a computational gain that is likely to scale to higher dimensional, plasma-wall simulations.

Plasma-Chemical Degradation of Molybdenum and Tantalum as First Mirror Materials for the EAST Tokamak

Viktor Afanasyev, Lidiia Lobanova, Maxim Semenov-Shefov, Shouxin Wang, Ling Zhang, Haiqing Liu

Kinetic modeling of lithium impurity transport in the HIDRA stellarator during controlled lithium evaporation

Steven Gula, Nina Mihajlov, Giovanni Diaz, Arnav Goyal, R Maingi, D Andruczyk, Davide Curreli

Post-operational imaging of the Hybrid Illinois Device for Research and Applications (HIDRA) stellarator following controlled lithium evaporation reveals distinct streak patterns of lithium deposition along the vacuum vessel wall. These localized structures exhibit strong alignment with magnetic field topology, suggesting that impurity transport in the scrape-off layer (SOL) is governed primarily by the three-dimensional magnetic geometry and associated electrostatic fields. In this work, large-scale kinetic simulations of trace lithium impurity transport are used to investigate the formation of these deposition structures. The magnetic field topology is reconstructed using a Biot-Savart law approach, from which a normalized flux-surface parameter is computed and used to prescribe the background helium plasma profiles. Electron density and plasma potential are modeled as normalized flux-surface-dependent fields scaled by representative scalar measurements and analytical sheath-based estimates, with the electric field obtained from ∇V p . These reconstructed fields are used to compute lithium ion trajectories and generate spatial maps of deposition on the vessel wall. The simulations reproduce streak-like deposition features aligned with magnetic flux tubes intersecting the wall, and the simulated deposition locations and orientations show qualitative agreement with experimental observations. These results demonstrate that magnetic field structure plays a dominant role in shaping impurity transport and deposition in HIDRA and establish a predictive framework for modeling impurity transport and surface deposition in plasma-surface interaction experiments involving liquid-metal plasma-facing components.

Positive sheath generation in front of fishscaled divertor plates

Choong-Seock Seock Chang, Seung-Hoe Ku, Xin Zhang, Trenton Brewer, Nicolas Lopez, Chris Marsden

Fish-scale divertor plates are becoming popular in the design of next-generation magnetic fusion reactors to make the edge of one divertor tile sits safely behind the shadowed profile of the preceding tile, hence to protect the leading edge from extreme heat deposition. Recently, ST40 tokamak has observed an extremely narrow and peaked divertor heat-load footprint near the separatrix strike point on the fish-scale plates, on top of the usual ion-drift-width scale footprint. This observation raised concern over the severely localized burn even when the well-known ion-drift-width scale burn issue can be resolved [X. Zhang et al., Nucl. Mater. Energy 41, 101772 (2024)]. In this report we demonstrate that the lost ion gyro-orbits to the tilted tile edges can lead to a significant positive-potential sheath at a practical magnetic field incidence angle that is much greater than the usual incidence angle for the ion gyro-sheath formation on flat strike-surface. This could make the kinetic electron heat-flow spilled over from the confined region down the divertor legs to gain a significant kinetic energy amplification and allow a highly localized heat-load peak under a low edge turbulence condition. Possible intrinsic mitigation mechanism is discussed. This issue should be considered carefully in the design of next generation magnetic fusion reactors.

RF-Specific Tungsten Erosion and Global Transport in ITER under Neon Seeding

Atul Kumar, Dhyanjyoti Nath, Wouter Tierens, Jeremy D. Lore, Andrei Pshenov, Tom Wauters, Andrea Galvan, Davide Curreli, Syun'ichi Shiraiwa, Nicola Bertelli, et al.

Ion cyclotron radio-frequency heating (ICRH) is a key auxiliary heating system in ITER, but high-power RF operation can enhance plasma-material interactions through rectified RF sheath potentials on antenna structures and nearby plasma-facing components. We present the first predictive application of the STRIPE (Simulated Transport of RF Impurity Production and Emission) framework to assess RF sheath-driven tungsten (W) erosion and global impurity transport from the ITER ICRH antenna under ITER-relevant neon-seeded conditions. STRIPE couples SOLPS-ITER plasma backgrounds, full-wave RF sheath calculations, geometry-specific ion energy-angle distributions, sputtering physics, and three-dimensional impurity transport. Simulations predict RF sheath potentials of 1 to 3 kV on antenna limiter sidewalls, increasing gross W erosion by about a factor of 64 relative to thermal sheath conditions and producing a gross source of 3.34e18 W atoms per second. Erosion is governed by RF-modified ion energy-angle distributions together with local plasma flux rather than sheath voltage alone. About 10 percent of sputtered W is locally redeposited, giving a net source of 3.01e18 W atoms per second. The RF-induced antenna source remains about three orders of magnitude smaller than the thermal divertor source and more than two orders of magnitude smaller than the integrated thermal main-chamber source. After 100 ms, about 22 percent of the mobile W inventory resides within the SOLPS-covered confined-plasma region, corresponding to an annular W concentration of 1.70e-6. These results indicate that the ITER ICRH antenna is unlikely to dominate the total W source budget under the conditions considered and demonstrate the need for coupled modeling of RF waves, sheaths, sputtering, redeposition, and global impurity transport.

Effects of Long-Term Tritium Exposure on FKM, FFKM, and Epoxy Resin

Christopher J. Koch, Nasiba Abdurakhmanova, Natalie Wieber, Tim Krentz, Tyler Guin, Dale A. Hitchcock

Self-consistent integrated modeling of the critical processes governing D/T retention in PFMs

Tatyana Sizyuk

Self-consistently integrated models for deuterium plasma interactions with plasma-facing materials (PFMs) have been developed and integrated into the upgraded ITMC-DYN+ simulation package. These models capture critical processes such as bubble and blister growth from nanocavities, blister bursting, defects formation induced by deuterium supersaturation, and changes in deuterium diffusivity resulting from microstructural evolution. Benchmarking against experimental data on deuterium retention and defect evolution in tungsten demonstrates that our integrated approach accurately explains a range of experimental observations for the first time. Key findings include: (a) the dynamics of bubble and blister growth and their impact on trapping site accumulation, (b) the combined effects of cavity growth, bursting, and surrounding dislocation networks, (c) the formation of a deuterium supersaturated surface layer (DSSL), (d) the influence of DSSL on deuterium diffusivity and bulk transport, and (e) differences in deuterium retention between pre-damaged tungsten and tungsten simultaneously irradiated by deuterium and neutrons. This self-consistent analysis reveals synergistic effects of blistering and DSSL formation on deuterium retention and transport in tungsten under varying temperatures and damage rates.

Experimental study of tritium release from neutron-irradiated lithium metatitanate using the thermal desorption method

Yevgen Chikhray, Saulet Askerbekov, Timur Kulsartov, Asset Shaimerdenov, Magzhan Aitkulov, Zhanar Bugybay, Assyl Akhanov

Deuterium retention and surface modification of tungsten under exposure to deuterium-neon mixed plasmas and combined with transient heat loads

Yue Yuan, Ying Qin, Arkadi Kreter, A Terra, Sören Möller, Long Cheng, Di Hu, yuhao Li, Sijie Hao, Peng Zhang, et al.

Neon (Ne) seeding is essential in ITER to mitigate divertor heat loads, but it significantly influences plasma-surface interactions and fuel retention in tungsten (W). Moreover, transient heat loads from edge-localized modes (ELMs) further alter surface morphology and fuel retention behavior. This work investigates the combined effects of Ne seeding and ELM-like transient heat loads on surface modification and deuterium (D) retention in rolled W. Experiments were carried out in the linear plasma device PSI-2 using pure D plasma and D+10% Ne plasma at 500 K, with an ion energy of 40 eV, and a fluence of 1 × 1026 D m-2. Transient loads were applied using a pulsed laser (1 ms, 0.76 GW m-2, 1000 cycles), either successively or simultaneously with plasma exposure. For plasma-only exposures, Ne seeding caused pronounced erosion and fine cracks on blister caps, resulting in approximately 32% lower total D retention compared with pure D plasma. Transient heat loads generated temperature and stress fields, thereby causing surface roughening and dynamic recrystallization, as well as increasing D retention in both the near-surface (< 4 µm) and deeper regions. When combined with Ne seeding, the effects depended on the loading sequence. With sequential transient heat loads followed by plasma exposure, D retention decreased by about 44% in the near-surface and by 27% in total retention. In contrast, under simultaneous THL + plasma exposure, Ne seeding increased near-surface and total D retention by approximately 62% and 16%, respectively, and produced an additional broad D release peak at around 1200 K. This peak could be attributed to the formation of high-binding-energy Ne-vacancy complexes (Nei-V complexes) induced by the synergistic interaction of D-Ne and transient heat loading. These results demonstrate that both impurity seeding and transient heat loads critically determine D retention behavior, emphasizing the importance of accounting for their synergistic effects when predicting fuel retention and optimizing operation scenarios in future fusion devices.

Overview of Achievements and Outlook of the IFMIF/EVEDA Project

Kazuo Hasegawa, Atsushi Kasugai, Keitaro Kondo, Kai Masuda, Satoshi Sato, Kentaro Ochiai, Hervé Dzitko, Fabio Cismondi, Yann Carin, Dominique Gex, et al.

The Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) project have been conducted as one of the three projects (IFMIF/EVEDA, IFERC and JT60SA) within the Broader Approach (BA) agreement between EURATOM and the Japanese government since 2007. The IFMIF is intended to deliver accelerator-based deuterium-lithium (D-Li) neutrons at energies and intensities to sufficient to enable the qualification of candidate materials for future fusion energy reactors, such as DEMO. The primary objective of the IFMIF/EVEDA project is twofold: (i) to develop a detailed engineering design of the IFMIF and (ii) to validate its major components, namely the Accelerator Facility, the Lithium Target Facility and the Test Facility. During Phase I of the BA, which concluded in March 2020, the Engineering Validation Activity (EVA) for the Lithium Target Facility and the Test Facility were successfully completed through the construction and testing of prototypes. In contrast, the EVA for the Accelerator Facility, implemented through the Linear IFMIF Prototype Accelerator (LIPAc), remains on-going. The current phase (Phase II) focuses on the continued commissioning of the LIPAc and the enhancement of some sub-systems to support the development of the Fusion Neutron Source Design (FNSD). This article presents an overview of the progress achieved in the LIPAc commissioning and FNSD activities and outlines the future directions of the activities.

Molecular dynamics simulation of migration and mechanism transition of 1/2 <111> dislocation loops in tungsten

Zhaofan Wang, Chao Yin, Qi Xiong, Zhe Liu, Ze Chen, Shifeng Mao, Minyou Ye

Spectroscopic characterisation of boronisation glow discharges in ASDEX Upgrade

Ralph Dux, Sehoon An, Volker Rohde

Helium bubble characterization in tritium aged austenitic stainless steels

Daniel Morrall, Tim Krentz, Natalie Wieber, Gavin Mattingly, Dale Hitchcock

First wall erosion induced by charge-exchange neutrals on EAST

Rui Ding, Jin Guo, Lei Mu, Guoliang Xu, Yaowei Yu, Yuming Liu, Rong Yan, Hai Xie, Dahuan Zhu, Junling Chen, et al.

Charge-exchange neutrals (CXNs), in particular of hydrogen isotopes deuterium and tritium, are expected to contribute notably to first wall erosion in future fusion reactors. To understand the CXN-induced first wall erosion under different discharge conditions in deuterium, dedicated experiments with a set of new diagnostics have been performed on EAST. Measurements of CXN energy spectrum by the low-energy neutral particle analyzer (LENPA) shows that the integrated CXN flux at the first wall positively correlated with the heating power and line-averaged electron density (n_"e" ), and increased by more than one magnitude from ohmic to high power discharges in the database. Deeper plasma fueling by supersonic molecular beam injection (SMBI) leads to a lower edge neutral pressure and thereby a ~50% lower CXN flux. The CXN flux in the intra-ELM phase is ~2 times higher than that in the inter-ELM phase. Measurements of material erosion rate by the quartz crystal microbalance (QMB) show that higher heating power can lead to stronger material erosion by CXNs. The erosion rate increases with n_"e" at first due to the higher CXN flux and then saturates due to the lower incident energy. The 3D-GAPS code is applied to model the CXN-induced erosion based on the LENPA-measured CXN energy spectrums, which shows good agreement with post-mortem analysis of exposed samples and QMB measurements.

Stress corrosion mechanism of CLF-1 RAFM steel in flowing Pb-17Li at 450℃: synergistic effects of tensile stress and microstructural evolution

Zhenchao Sun, Yu Guo, Xiujie Zhang, Teng Zhang, Wei Qian, Yao Zhao, Lei Wang, Xinting Lv, Yiming Wang, Zhengdong Li

The compatibility of RAFM steel with liquid Pb-17Li is a critical concern for its application as a structural material in liquid metal blankets of fusion reactors. This study investigated the stress corrosion mechanism of CLF-1 RAFM steel under a tensile stress of 250 MPa for 5000 hours in flowing Pb-17Li at 450 °C. The results indicate that stress-induced strain disrupts the continuity of the protective oxide layer, shortening the incubation period and increasing corrosion initiation sites, thereby intensifying the corrosion severity and susceptibility. The corrosion morphology is jointly influenced by the flow velocity of Pb-17Li and the steel’s microstructure. At high flow velocities, the martensitic lath structure is eroded, resulting in an etched surface. Conversely, lower velocities preserve the more corrosion-resistant laths, producing lamellar corrosion structures. Furthermore, shear stress promotes microstructural coarsening, which eliminates certain grain and sub-grain boundaries. Although this locally improves corrosion resistance, the resulting long and straight grain boundaries instead facilitates the penetration of liquid metal, thereby accelerating corrosion failure. These findings clarify the stress corrosion mechanism of RAFM steel in Pb-17Li and provide new insights for the modeling of liquid Pb-17Li corrosion and design of liquid metal blankets.

Low deuterium retention in chemical vapor deposited tungsten with columnar grain structures under combined effects of displacement damage and helium seeding

Ting Wang, Arkadi Kreter, Peng Bi, Hanqing Wang, Y. Mao, Hao Wang, Yue Yuan, Long Cheng, Li-Qun Shi, Jun Tang, et al.

Controlling hydrogen isotope retention is a critical challenge for plasma-facing materials (PFMs) in fusion reactors. Chemical vapor deposited tungsten (CVD-W), featuring columnar grain structures, has attracted increasing attention as a candidate PFM. However, its surface blistering and deuterium (D) retention behavior under fusion-relevant, complex irradiation conditions remain insufficiently understood. In this work, CVD-W and ITER-like forged W both with grains elongated normal to the exposed surface were irradiated with pure D and mixed D+5% He plasma, with and without prior W self-ion irradiation at damage levels of 0.2 and 2 dpa. Compared with ITER-like W, CVD-W exhibits substantially enhanced tolerance to D-induced surface blister formation and markedly reduced D retention under pure D plasma exposure. Regardless of displacement damage, He seeding, or their combination, CVD-W consistently retains less D than ITER-like W across all irradiation sequences studied. This persistently low D retention in CVD-W is primarily associated with its lower defect density (e.g. fewer grain boundaries) and a blister-resistant <001> surface texture that limits blister-related D trapping. Furthermore, displacement damage and He seeding effectively suppress surface blistering, but exert opposing individual effects on D retention in both materials. Their combined effect shows a clear dependence on D fluence and material microstructure. This work highlights the advantages of CVD-W in suppressing hydrogen isotope retention under complex irradiation environments and provides valuable insights for the selection and microstructural design of advanced PFMs with improved resistance to hydrogen-induced blistering and reduced retention.

Lithium droplet transport in tokamak edge plasmas

A. Diaw, J. D. Lore, S. Smolentsev

A lithium droplet transport and evaporation model has been developed within the direct simulation Monte Carlo code OpenEdge. This model integrates gravity, collisional ion drag, orbital-motion-limited charging, energy-balance evaporation, and an anisotropic rocket recoil force using a Strang-split integrator. Validation against analytical drag-gravity solutions and independent RK45 evaporation integration demonstrates relative errors below 10−5 for droplet radii of 1.5, 2.5, and 3.5 mm. Simulations of ensembles containing 105 droplets, launched from inner and outer divertor surfaces in SOLPS-ITER plasma background for the CAT tokamak reactor concept, indicate that transport outcomes are determined by initial size, velocity, and launch location. Outer-divertor droplets predominantly redeposit locally, whereas inner-divertor droplets reach the low-field-sidewall. Smaller droplets lose most of their mass to evaporation before reaching the core, while larger droplets retain their mass and redeposit on nearby tiles. Both one-way and iterative two-way coupling frameworks map the evaporated lithium onto the SOLPS-ITER mesh as volumetric sources, facilitating self-consistent evaluation of lithium droplet impacts on edge-plasma performance.

Influence of tritium aging and helium-3 content on fracture resistance in austenitic stainless steel GTA welds and heat-affected zones

Natalie Wieber, Timothy Krentz, Anastasia Mullins, Dorian Balch, Joseph Ronevich

Simulation study of tungsten erosion and impurity transport under the assumption of a full-tungsten wall in EAST plasmas

Wu YiHan, Xuele Zhao, Chaofeng Sang, Yilin Wang, Chen Zhang, Qingrui Zhou, Dezhen Wang, Rui Ding, Baoguo Wang, Qingquan Yang, et al.

Future fusion reactors are expected to employ full-tungsten plasma-facing components, and the generation and core accumulation of tungsten impurities under full-tungsten wall environment is one of the key issues that limit the stable operation of tokamak devices. In this work, the upgraded impurity transport code IMPEDGE is coupled with the extended-grid version of SOLPS-ITER to investigate the erosion and transport of tungsten impurity in full-tungsten wall environment. The background plasma is based on experimental data from EAST discharge #140718, with all plasma-facing components assumed to be tungsten. The simulations show that, under full-tungsten wall assumption, both the tungsten impurity erosion and its core concentration remain at relatively high levels even in the detached regime, with dominant contributions originating from the far scrape-off layer (far-SOL) of the lower divertor and the upper divertor region. The main reason is that, although the electron temperature at the near scrape-off layer (near-SOL) of the lower divertor decreases significantly during detachment, that in the far-SOL region remains difficult to further reduce (~10 eV), resulting in substantial tungsten erosion. The total impurity penetration probability (defined as the probability for emitted particles to enter the plasma core) also remains at a relatively high level, mainly because the lower plasma density in the far-SOL weakens the impurity screening effect. In the downstream region, the reduced ion density decreases the friction force acting on impurities, allowing them to more easily escape from the region and migrate upstream. In the upstream region, the lower electron density increases the ionization mean free path, causing impurities to be ionized closer to the last closed flux surface and thereby enhancing impurity penetration. These results indicate that impurity screening in the near-SOL divertor region can be significantly enhanced by impurity-seeded detachment. However, erosion and penetration from the far-SOL region still dominate the core impurity accumulation, suggesting that more effective impurity control strategies for the far-SOL region are required in future studies.

Monovacancy-enhanced trapping physical mechanism for interstitial oxygen in molybdenum and tungsten

Mengxin Wang, Aoyu Mo, Xinyu Xiu, Luzhe Chen, Quan-Fu Han, Yue-Lin Liu