Recent Publications

Change of the Tungsten neoclassical particle transport characteristic due to the coexisting of the lower-Z impurity in the tokamak plasmas

Chengkang Pan

The high-Z impurity Tungsten neoclassical particle transport in the tokamak plasmas with the coexisting of the lower-Z impurity is investigated. The lower-Z impurity could reduce or enhance the Tungsten neoclassical particle transport. The Tungsten neoclassical transport characteristic will be changed with the lower-Z impurity content large enough. The dependence of the Tungsten neoclassical transport on the bulk ion radial gradients predicted by the existing impurity neoclassical transport theory will be broken. The existing impurity transport theory will over-predict (under-predict) the Tungsten neoclassical particle transport for the bulk ion radial gradients less (larger) than the critical value. The effect of the lower-Z impurity on the high-Z impurity neoclassical transport is through the parallel friction force coupling and it should be included in the Tungsten neoclassical particle transport calculation.

Mesh-based multiphysics coupling acceleration for fusion neutronics clustering for fusion blanket applications

Jin Whan Bae, Arpan Sircar, Katarzyna Borowiec, Vittorio Badalassi, Cami Collins

Accurate modeling of particle transport within fusion blankets is essential for predicting performance metrics such as heat deposition and the tritium breeding ratio (TBR). However, high-fidelity coupling of thermal fluids from computational fluid dynamics (CFD) to neutronics simulations often incurs significant computational costs due to the complexity of surface intersection calculations in Monte Carlo codes. This paper presents an accelerated multiphysics coupling method for neutronics that utilizes hierarchical agglomerative clustering to map complex material property distributions to a neutronics model. Implemented within the Fusion Reactor Design and Assessment (FREDA) framework, the method leverages existing Python packages to automate the creation of clustered geometries for OpenMC. The approach is demonstrated on a sector model of an ARC-class tokamak with an immersion molten salt blanket, and an simple geometry with varying isotopic concentrations. Results show that the clustering method significantly reduces computational burden without compromising fidelity, providing a foundation for agile iteration of neutronics simulations involving multiple coupled material properties.

The status and design challenges of the heating and current drive systems for DTT

Gustavo Granucci, Silvio Ceccuzzi, Afra Romano, Andrea Murari, Gian Luca Ravera, Piero Agostinetti, Sofia Bertolami, Falk Braunmüller, Alessandro Bruschi, Daniele Busi, et al.

This paper reports the main design keys and the challenging issues of the Heating and Current Drive System (HCD) of the Divert Tokamak Test (DTT) facility that is under construction at ENEA site in Frascati with the aim to perform studies on the power exhaust in a flexible and easily modifiable environment. The selected HCD systems for DTT are those with the most consolidated technology and expected to be relevant for the future reactor. The status of each system is reported, both in terms of design and procurement, which are well advanced for the system required in the first phase of the DTT exploitation: electron and ion cyclotron resonant heating. The third system is neutral beam injector, based on negative ion acceleration, which will be installed in DTT in a second phase, after the first five years of operation. The full heating power will be reached with the doubling of the radiofrequency power to reach the 45 MW at plasma needed to test the divertor with a power density at reactor level.

Modeling of divertor heat flux limits and lithium vapor shielding in NSTX-U using UEDGE code

MD Shahinul Islam, Maxim V Umansky, Vlad Soukhanovskii

In this work, UEDGE simulations coupled self-consistently with a two-dimensional wall transport solver (Wall-Li) are used to investigate two closely related issues important for NSTX-U operation: (1) plasma-surface interactions of graphite plasma-facing components (PFCs) under increasing input power for standard and snowflake divertors and (2) lithium PFC vapor shielding and its dependence on upstream plasma conditions for a standard lower single null divertor with a 5~mm layer of lithium on the graphite tiles. Simulations with graphite target PFCs show a monotonic increase in surface temperature with increasing heat flux incident on the target plates. For a broad range of operating conditions, surface temperatures exceed the graphite sublimation temperature (>1200°C, when graphite starts massively sublimating) once peak heat fluxes exceed 7~MW/m2. This occurs despite enhanced carbon radiation. A snowflake-minus divertor configuration is analyzed to assess divertor heat handling on the secondary strike point (SP\#2) under conditions where both magnetic field-line incidence and cross-field transport are uncertain. Simulation results indicate that the heat flux to SP\#2 is dominated by enhanced CM transport rather than parallel conduction, leading to increased deposited heat flux with increasing CM strength and incidence angle. Lithium vapor shielding is investigated using a self-consistent model in which lithium sourcing depends on local plasma and surface conditions. The model demonstrates that lithium vapor shielding provides passive thermal cooling, maintaining surface temperatures below 700°C even as the core power increases. However, lithium accumulation upstream increases once surface temperatures exceed 600°C. Increasing the core density augments the drag force on impurity ions due to the parallel plasma flow in the scrape-off layer, confining lithium closer to the target and thereby extending the operational window for effective vapor shielding. Overall, these results highlight the necessity of (1) thick lithium layers to prevent strike-point depletion, (2) active cooling mechanisms (such as fast-flowing liquid lithium) to suppress excessive surface temperatures and evaporation, as lithium operation is constrained by surface temperature, and/or (3) deuterium gas puffing near the divertor to further increase SOL flow and density

Overview of the European Breeding Blanket programme for Helias 5-B: Dual Coolant Lead-Lithium design, integration and tools

Iole Palermo, Javier Alguacil, Gaetano Bongiovi, Juan Pablo Catalan, Ilenia Catanzaro, Iván Fernández-Berceruelo, Salvatore Giambrone, Guillermo Gómez Fonfría, Jose Ángel Noguerón, Vicente Manuel Queral Mas, et al.

CIEMAT, Universita degli Studi di Palermo, UNED

As part of EUROfusion’s mission to bring stellarators to technological maturity, the Stellarator Power Plant Studies (SPPS) WPPRD began in 2021 to develop a HELIAS-class power plant. Building on DEMO tokamak experience, European teams are designing a Dual Coolant Lead-Lithium (DCLL) breeding blanket (BB) for HELIAS. This concept uses liquid PbLi as breeder/coolant and decoupled helium cooling for the first wall (FW). Two key adaptations address HELIAS’s complex geometry: a detached FW using Capillary Porous System (CPS) technology, and a quasi-toroidal segmentation (QTS) with PbLi flow aligned to magnetic field lines. QTS reduces magnetohydrodynamic (MHD) pressure drop by up to two orders of magnitude, potentially eliminating electrical insulation needs. Remote handling (RH) is rethought for 3D stellarators, where traditional vertical-port extraction is impractical. Alternatives include enlarged fixed coils, movable coils for temporary large ports, and detachable vessel periods. The detached FW strategy shifts maintenance from large BB segments to smaller, easily replaceable FW panels, extending BB lifetime. CPS designs with Li in tungsten matrices were analysed thermally, hydraulically, and neutronically, showing potential to lower displacement-per-atom (dpa) damage while maintaining tritium breeding. Be-based moderators behind the FW CPS matrix improved TBR while reducing back-structure damage. To accelerate design and analysis, HeliasGeom and SHANE tools were developed to rapidly generate realistic 3D parametric geometries for CAD, neutronic, and thermal-hydraulic coupling. Preliminary 3D studies addressed TBR optimisation, shielding, MHD in non-uniform fields (via GridapMHD), and multi-scale thermal–mechanical assessments. These innovations collectively advance stellarator blanket technology, integration, and maintainability toward viable power plant concepts.

The physics of ballooning-limited ELM-free regimes in EUROfusion tokamaks

Mike G Dunne, Michael Faitsch, Olivier Sauter, Eleonora Viezzer, Benoit Labit, Athina Kappatou, David Keeling, Branka Vanovac, Itziar Balboa, Petra Bilkova, et al.

Culham Centre for Fusion Energy, Max-Planck-Institut für Plasmaphysik, Ecole Polytechnique Federale de Lausanne, Academy of Sciences of the Czech Republic, University of Seville

The development of operational scenarios without large Type-I ELMs is of utmost importance for the stable operation and longevity of future tokamaks. The EUROfusion tokamak exploitation program has therefore made the understanding of ELM-free regimes a major topic of exploration across all its contributing devices (ASDEX Upgrade, JET, MAST-Upgrade, TCV, and WEST). An integrated program to investigate a range of Type-I ELM-free regimes has been developed covering the enhanced D-alpha (EDA), magnetic perturbations (MP), negative triangularity (NT), quasi-continuous exhaust (QCE), quiescent H-mode (QH), the baseline small ELMs (SE), I-mode, and X-point radiator (XPR) regimes. This contribution focuses on the development and understanding of the NT and QCE regimes on ASDEX Upgrade, JET, and TCV. The importance of transport via ballooning modes in both regimes is highlighted, as well as the progress in developing access models based on ideal-MHD. In the case of the QCE, this can also be expressed as a minimum separatrix density, which corresponds well to experimentally measured separatrix densities. Particular focus is paid to the performance of the QCE in terms of the achieved pedestal top values, which, when appropriately normalised, do not differ significantly from ELMy H-mode plasmas. This, combined with the predicted minimum separatrix density for the 15~MA ITER baseline plasma, highlight the relevance of the QCE as a potential operational scenario for both ITER and future reactors.

Progress of LHCD experiment at 4.6 GHz towards long-pulse and high-power operation on EAST

Miaohui Li, Weiwei Zhang, Junlin Chen, Chenbin Wu, Bojiang Ding, Mao Wang, Liang Liu, Lianmin Zhao, Wendong Ma, Yong Yang, et al.

Chinese Academy of Sciences - Hefei Institutes of Physical Sciences, Institute of Plasma Physics Chinese Academy of Sciences, CEA

Recent achievements of lower hybrid current drive (LHCD) experiment at 4.6 GHz towards long-pulse and high-power operation on EAST tokamak are reported. The duration of long-pulse plasmas has been extended to 1056 s with 1.1 MW LH power in I-mode and 1066 s with 0.92 MW in H-mode. The operational domains in plasma current and line-averaged density for fully non-inductive discharges with LH alone and with the combined LH and electron cyclotron (EC) waves are presented. The dependence of LHCD efficiency on plasma density in both L- and H-mode discharges characterized by residual loop voltage Vloop = 0 is quantified. It is found that the LHCD efficiency is improved significantly by EC heating due to the increase of electron temperature. The dominant issues in long-pulse and high-power operation are summarized and discussed, including the power coupling imbalance, the hot spot and arc events in front of the antenna, and the deteriorated plasma heating effect with high LH power. Finally, prospects with a new 4 MW LHCD system at 4.6 GHz which is under development are given.

First campaign with alternative divertor configurations in ASDEX Upgrade

Tilmann Lunt, Felix Albrecht, Matthias Bernert, Dominik Brida, Ralph Dux, Michael Faitsch, Tabea Gleiter, Sebastian Josef Hörmann, Joey Kalis, Bernd Kurzan, et al.

Max-Planck-Institut für Plasmaphysik

After a major hardware extension ASDEX Upgrade has now established and characterized a variety of alternative divertor configurations (ADCs) with heating powers reaching up to 20 MW and plasma currents up to 1 MA. These high performance conditions in configurations with arbitrarily small field line incidence angles were made possible due to the high precision in the tile alignment of 300 μm as well as a careful optimization of the error fields caused by the current feeds. The formation of an X-point radiator (XPR) was observed in a low-field side snowflake minus (LFS SF - ) configuration at a very low impurity concentration determined by the intrinsic sources only, i.e. without additional seeding of impurities. Compared to a typical lower single-null (SN) configuration the ELMs were found to be substantially smaller in size and higher in frequency and therefore barely detectable. While the primary strike line is fully detached the secondary one in the far-SOL shows heat fluxes of a similar magnitude as the primary strike line in a previous upper SN phase. These fluxes might be reduced by installing configuration-optimized baffles and/or by increasing the plasma current and thereby reducing λ q . In fact significantly smaller far-SOL heat fluxes were found when increasing the plasma current from 800 to 1000 kA. According to the divertor Langmuir probes the peak heat flux is then by a factor of two smaller than the ones in the SN reference.

Direct comparison of 3D non-linear JOREK simulations of shattered pellet injection with ASDEX Upgrade experiments

Weikang Tang, Matthias Hoelzl, Paul Heinrich, Di Hu, F J Artola, Pascal de Marne, Mathias Dibon, Mike G Dunne, Ondřej Ficker, Peter Halldestam, et al.

Max-Planck-Institut für Plasmaphysik, ITER Organization, Beihang University, Institute of Plasma Physics Czech Academy of Sciences, CEA

Shattered pellet injection (SPI) as primary mitigation method for major disruptions in ITER has a large parameter space available for optimization including the total amount of injected material, the size of the individual pellet fragments, the material composition, and the timing of multiple injections. This flexibility needs to be exploited to simultaneously minimize thermal heat loads, electromagnetic vessel forces, and formation of relativistic electrons and their impacts on plasma facing components. In this article, we apply 3D non-linear magnetohydrodynamic modelling to SPI experiments in the ASDEX Upgrade tokamak, going beyond our previous work [Tang et al Nucl. Fusion 65 116003 (2025)] by resolving some discrepancies between simulations and experiment and carrying out direct qualitative and quantitative comparisons to experimental measurements. The key element that enables the transition is the incorporation of the parallel heat-flux limit, which is done here in a simplified form. The work increases the confidence of reproducing key processes of disruption mitigation in direct 3D non-linear simulations in view of future predictive studies for ITER.

Multiscale assessment of tritium behavior in preliminary fusion pilot plant design using surrogate models in TMAP8

Lin Yang, Pierre-Clément Simon, Emre Yildirim, Jose Manuel Trueba Cutillas, Matthew Robinson, Masashi Shimada

Idaho National Laboratory, Tokamak Energy Ltd

The complexity and significance of multiscale phenomena in fusion energy systems make advanced modeling necessary for designing, optimizing, and safely deploying fusion plants. Tritium accountancy is one of those challenges for deuterium–tritium fusion systems. Its availability is constrained by its short half-life (12.33 years) and limited natural abundance, which require fusion plants to breed tritium onsite. Therefore, accurate tritium accountancy is essential for effective resource management, safety, and economics in fusion plants. Through the U.S. Department of Energy milestone program, Tokamak Energy Ltd. is developing a fusion pilot plant design and evaluating tritium retention and loss in key components and their effect on the fuel cycle. To rapidly explore design trade-offs and quantify design decisions on tritium management, this study presents a multiscale analysis to investigate tritium diffusion, trapping, and recovery in key plasma-facing components. To enhance computational efficiency, we integrate surrogate models at the component-level within a fuel cycle model at the system-level, enabling rapid evaluation of tritium recycling dynamics and inventory under various operational scenarios. The goal of this study is twofold: (1) demonstrate the feasibility of utilizing surrogate models to increase the accuracy of fuel cycle modeling, and (2) rapidly evaluate the performance of fusion technologies to accelerate design iterations. This multiscale model provides the tritium transport and retention behavior and supports the plasma-facing components design optimization in normal and bake-out operations. The work is implemented using the Tritium Migration Analysis Program, Version 8 (TMAP8), an open-source application for tritium transport analysis in fusion systems.

Nonlinear dynamic evolution of energetic particle mode due to nonadiabatic wave-particle interaction

Pengjuan Su, Fulvio Zonca, Matteo Valerio Falessi, Jialei Wang, Yasushi Todo, Zhiyong Qiu

ENEA Frascati Research Centre, National Institute for Fusion Science, Chinese Academy of Sciences Hefei Institutes of Physical Science Institute of Plasma Physics

Adopting dedicated phase-space diagnostics, we analyze the nonlinear dynamic evolution of an energetic particle mode (EPM) due to nonlinear wave–particle interactions, including nonlinear frequency chirping and secular particle motion due to trapping and de-trapping. By tracking representative particle orbits, we show that the nonlinear evolution is dominated by continuous turnover of the distinct resonant populations via self-consistent trapping and de-trapping, rather than by adiabatic frequency sweeping of a fixed cohort. The resulting clump motion in phase space follows the instantaneous low-frequency resonance contour, leading to pronounced downward frequency chirping. The measured EPM frequency chirping rate scales linearly with the mode amplitude, as predicted by general theory.

High-heat-flux performance of monoblock target prepared with advanced W-K plate

Fan Feng, Youyun Lian, Jianbao Wang, Jiupeng Song, Mengxia Liang, Yuzhong Jin, Xiang Liu

Southwestern Institute of Physics, Xihua University

Potassium-doped tungsten (W-K) is a promising plasma-facing material because nanoscale K bubbles may improve microstructural stability without introducing solid second phases. In this work, large-scale rolled W-K plates containing ~90 ppm K were fabricated by powder metallurgy, hot rolling and stress-relief annealing, and were machined into ITER-like water-cooled monoblock mock-ups. The rolled plates showed a tensile strength of 1225 MPa at 50 °C , ductility exceeding 20% at 200 °C, and a recrystallization temperature of ~1500 °C . Transient electron-beam thermal-shock tests on the RD-TD plane demonstrated that the as-rolled and 1400 °C -annealed W-K remained crack-free after 100 pulses of 1 ms up to 0.66 GW/m², whereas specimens annealed at ≥1500 °C exhibited reduced cracking thresholds. Under steady-state high-heat-flux fatigue, W-K monoblocks maintained structural integrity at 20 MW/m² for up to 1500 cycles, although surface roughening, intergranular fissures and local melting developed with increasing cycle number. At 25 MW/m², severe roughening/erosion occurred after 500 cycles as the apparent surface temperature exceeded 2300 °C . A key mechanistic finding is that crack density and crack depth are governed by different factors: crack density increased mainly with accumulated thermal cycles, whereas crack depth was controlled predominantly by peak surface temperature. Stable K-bubble dispersion is suggested to retard grain-boundary migration and suppress microcrack nucleation, contributing to the high thermal-shock and HHF tolerance of rolled W-K monoblocks.

Impact of toroidal-field-coil generated error field over ADITYA Ohmic Breakdown and Start-up

Shishir Purohit, Joydeep Ghosh, Manoj Kumar Gupta, Kumarpalsinh A Jadeja, Malay Bikas Chowdhuri, Ranjana Manchanda, Kaushal Patel, S B Bhatt, Moti Makwana, C. N. Gupta, et al.

Institute for Plasma Research

The conventional Ohmic plasma breakdown and current ramp-up in the ADITYA tokamak have been investigated over a wide range of input parameters to understand the error field generated by the toroidal field coils. The plasma breakdown and ramp-up phases are quantified by the breakdown time (τb) and the CIII emission intensity rollover time. A good number of discharges having τb in the range of 0.8 – 2 ms, the applied loop voltage in the range of 20 - 25 V, pre-fill fuel pressure in the range of 10−5 to 10−4 Torr, and toroidal magnetic field (BT ) in the range of 0.75 – 1.1 T have been analyzed. For the constant sets of E/P (V/m Torr) and equilibrium field during breakdown, it has been observed that the decreasing trend of τb with increasing BT ceases at a particular value of BT after which τb remains constant with BT . With the errors generated by other coils remaining constant, τb should continuously decrease with BT as the increase in BT increases the connection length, thereby facilitating early breakdown. However, τb remaining constant with increasing BT (after a threshold value) indicates that there exists an error field due to toroidal field coils, which negates the gains associated with connection length at higher toroidal fields. The magnitude of BT error field is estimated to be, < BT F C err > /BT ∼ 4 – 7 x 10−3 in ADITYA. Further investigation revealed that although τb ceases to decrease with increasing BT , beyond a threshold BT , the rollover of carbon (CIII spectral emission) occurs more rapidly at higher BT . This indicates a faster temperature rise with higher BT in the current ramp-up phase of the discharges and suggests a weak connection between breakdown and current ramp-up phases of a discharge.

A theory of E×B staircase stability: favorable role of fast ions

Gyungjin Choi, Mikhail Angel Torio, Taik-Soo Hahm

KAIST, Seoul National University

We present an analytic theory for the E×B staircase stability against intrinsic collapse with a bump-hole pair growth. The obtained scale-dependent critical E×B shearing rate required to sustain the staircase indicates that finer-scale flow components are inherently more susceptible to collapse. Extending our model to include fast ions, we find a drastic reduction in the critical shearing rate. It originates from the reduction of the radial drift wave group velocity due to fast ion population. The result suggests a highly favorable role of fast ions in maintaining the sharp, step-wise E×B staircase structure, consistent with recent experimental observations of long-lived staircases in KSTAR FIRE mode plasmas.

Magnetic topology discontinuous changes and discrete shifts of divertor heat flux induced by edge-localized modes in the HL-2A tokamak

Jinming Gao, Xiao Lan Zou, Guoliang Xiao, Wulyu Zhong, Tianbo Wang, Dongmei FAN, Tengfei Sun, Y B Dong, Min Jiang, Zhou Yulin, et al.

Southwestern Institute of Physics, IRFM, Dalian University of Technology

Magnetic reconnection is a fundamental physical process in plasma physics. Edge-localized mode (ELM) physics and its impact on divertor heat loads are critical issues for fusion plasmas. We report the first observation of discontinuous changes in magnetic topology—manifested as sudden shifts of the divertor heat load peak—during ELM crashes in tokamak plasmas. Nonlinear ELM evolution drives magnetic reconnection, generating energetic electrons that form helical current filaments. These filaments induce an abrupt poloidal extension of topological fractures, evidenced by non-gradual changes in the poloidal correlation length of ELM-induced magnetic perturbations. These results present a new aspect of ELM physics, which should be taken into account in the modelling. Also our findings on nonlinear interactions of electron dynamics with the magnetic field topology reveal previously unexplored mechanisms governing reconnection physics.

Revisit two-mode couplings of the ablative Rayleigh-Taylor instability in inertial confinement fusion

Xian Jiang, Shenming Xu, Dehua Zhang, Tao Tao, Jingfei Xin, Rui Yan, Hang Ding, Jian Zheng

University of Science and Technology of China, BYD Automobile Industry Co. Ltd

Two-dimensional two-mode couplings of the ablative Rayleigh-Taylor instability (ARTI) under inertial-confinement-fusion (ICF) relevant parameters are revisited via numerical simulations, with a focus on the regimes where the wavenumbers of the initial modes are close to the cutoff wavenumber. In the quasi-linear stage in which the generated modes approximately grow exponentially with time, both the self-growth of the generated mode (as if in the single-mode regime) and the driven-growth due to mode couplings are found to be important. A series of linearly unstable modes are found to be generated by coupling of two linearly stable initial modes, and then compete for the dominant mode according to a proposed criterion. A model including the contribution of both the self-growth and the driven-growth with an empirically modified coupling coefficient is proposed and shown to accurately predict the growth of the generated mode. The findings have implications for assessing the impact of ARTI in ICF designs.

Radial phase variation and energy flow of Alfvén gap modes

Xinran Xu, Jian Bao, Wenlu Zhang, Chao Dong, Jintao Cao, Ding Li

Chinese Academy of Sciences Institute of Physics

Radially curved mode structures of Alfvén eigenmodes are commonly observed in fusion experiments associated with energy transport, which indicate the radial phase variation arising from non-ideal magnetohydrodynamics (MHD) and global effects. In this work, based on MAS global eigenvalue simulations with Landau-fluid bulk plasmas and non-perturbative gyrokinetic energetic ions, we investigate the physical mechanisms responsible for the curved poloidal mode structure tail of RSAE and the rapid phase change of radially coupled RSAE and TAE in DIII-D plasmas. The former one is due to kinetic interaction between RSAE and Alfvén continuum with enhanced mode conversion to kinetic Alfvén waves, and the latter one can be explained by global effects of higher-order radial eigenstate or RSAE-TAE hybrid modes with multiple poloidal harmonics. An improved energy transport model based on global mode structures is formulated and implemented in the MAS framework, which clearly demonstrates the radial phase variation and EI non-perturbative effects on the radial Poynting vector.

Effect of nonuniform density structure on burn-up ratio of multi-shock-compressed DT fuel in fast ignition

Tomoyuki Johzaki, HIdeo Nagatomo, Yasuhiko Sentoku, Shinsuke Fujioka

The University of Osaka, Hiroshima University

The burn characteristics of multi-shock-compressed DT fuel under fast ignition conditions are investigated using two-dimensional burn simulations. In solid-sphere implosions, high fuel areal density can be achieved through multi-shock compression with reduced susceptibility to hydrodynamic instabilities. However, the resulting compressed-fuel density structure is highly nonuniform, consisting of a localized high-density core surrounded by a low-density fuel region that contains most of the fuel mass. In such a configuration, ignition can be initiated in the central high-density core, but the burn wave rapidly attenuates as it propagates into the surrounding low-density region, preventing self-sustained burn propagation. Consequently, the burn-up ratio is significantly lower than that of uniformly compressed fuel with comparable total fuel areal density. By examining the effects of implosion timing and artificially modified surrounding-density profiles, we show that the burn-up ratio is governed not only by the total fuel areal density, but also by the density structure around the ignition region. These results indicate that achieving a high burn-up ratio requires a sufficiently dense fuel region along the burn-wave propagation path, rather than a localized high-density core alone.

Non-dimensional confinement scaling in similar negative triangularity plasmas on the DIII-D and TCV tokamaks

Alessandro Marinoni, Colin Chrystal, Stefano Coda, Reinart Coosemans, Claudio Marini, Mario Podesta, Olivier Sauter, Matteo Agostini, Max E Austin, Emily A Belli, et al.

General Atomics, Ecole Polytechnique Federale de Lausanne, Columbia University, University of California San Diego, Consorzio RFX

Similarity experiments were performed on the DIII-D and TCV tokamaks to explore the scaling of energy confinement in negative triangularity plasmas using non-dimensional variables. Near up-down symmetric plasmas with large top-bottom averaged negative triangularity were created in a lower single null configuration, with the shape of the separatrix being closely matched between the two devices. The normalized energy confinement is found to weakly improve at increasing collisionality and, between the two devices, shows a machine size scaling behavior between Bohm and gyro-Bohm. Engineering scaling on a large DIII-D dataset is in agreement with the non-dimensional experiment.

Deuterium retention of boron powder from deuterium gas or ion exposure

Shota Abe, Adam Q. Kuang, Christopher P Chrobak, Alessandro Bortolon, Camilo Jaramillo-Correa, Bruce E Koel

Princeton Plasma Physics Laboratory, Princeton University, Commonwealth Fusion Systems LLC

We report retention of deuterium (D), a proxy for tritium (T), from D2 neutral gas and D+ ion exposures on commercial boron (B) powder, a proxy for B dust potentially formed in fusion reactors. D removal behavior from B powder is reported as a means of estimating a potential T inventory in B dust in advanced fusion reactors, such as SPARC and ITER, the latter currently plans to employ B wall conditioning by glow-discharge boronization. B wall conditioning forms surface coatings of chemical compounds on tungsten (W) plasma-facing components (PFCs), thus suppressing plasma contamination by impurities such as oxygen, carbon, and tungsten. However, B-based slag or dust particles are expected to form. Such B dust particles can retain hydrogen isotope species, thereby causing a T inventory issue. In this work, the commercial B powder was exposed to D2 neutral gas or D+ ions and analyzed by temperature-programmed desorption to quantify D retention and the desorption temperature. The experiment confirmed D retention from D2 neutral gas exposure. A strong D2 desorption peak at 700 K, corresponding to B-D bonding, was observed for all D2 gas and D+ ion exposure cases. For D2 gas exposures, D retention was significantly enhanced at a B powder temperature of 550 K. B powder bakeout under vacuum at 600 K for 22 hours after D2 exposure exhibited efficient D removal. In contrast, the B powder bakeout at 420 K, even for 1 day, did not remove D efficiently. D retention yields from D+ ion exposure, emulating charge-exchange involving D atoms, were determined. Experiments confirmed that D retention was suppressed by oxidation of the B powder surface, which naturally occurs in reactor environments.