A coordinated DIII-D campaign exposed and comparatively assessed 44 advanced plasma-facing materials from 12 institutions, including four public-private fusion partnerships, to support fusion pilot plant wall and divertor material down-selection. Samples were exposed using the Divertor Materials Evaluation System (DiMES) under Ohmic, L-mode, and H-mode conditions with edge-localized modes, at 0.2-2.5 MW m$^{-2}$ on flush geometries and 10-15 MW m$^{-2}$ on 10$^{\circ}$ angled geometries. Engineered tungsten architectures retained integrity; long-fiber Wf/W showed the clearest crack-arrest behavior. W-Re and K-doped W showed near-ITER-W-like responses, while additively manufactured W-Ta showed heat-flux-sensitive mass losses of 0.64 mg for the flat sample and 2.19-2.87 mg for angled samples. After irradiation to 0.3 dpa at 550$^{\circ}$C, neutron-irradiated ITER-grade W retained 2.8 times more deuterium than pristine W, while TiB$_2$ showed the lowest D$_2$ release in the Ohmic set. VTaHfMo was the most stable refractory multi-principal-element alloy. NbC and (Nb$_{0.5}$Ta$_{0.5}$)C retained integrity with 0.02-0.03 mg mass loss, whereas ZrC lost 7 mg. CVD SiC retained macroscopic integrity but exhibited an effective Si erosion yield of 0.5, about 5-10 times above prior DIII-D trends. Renewable boron pebble rods underwent controlled recession; 13% of released boron was ionized near the outer strike point and up to 50% was recovered locally. Initial in-situ chromium gross-erosion measurements yielded values of order $10^{-2}$. Together, these results provide cross-material benchmarks for fusion pilot plant down-selection and future AI/ML-assisted plasma-facing-material development.

We demonstrate a new renewable boron aggregate material that can withstand heat loads of 40 MW m-2 without cooling and featuring low sublimation of below 100 Torr-L/s/m2 and sputtering yield of 0.04 from 40 eV D, similar to solid boron. The retention of deuterium (D) is 3.2 x 1020 D/m2, 200x less than solid boron, 100x less than carbon, and comparable to tungsten at similar fluence and energy. The material sheds under high heat loads, at a rate of 0.35 cm/s, exposing the cool surface to the plasma, allowing heat removal and tritium extraction away from the plasma and eliminating the need for cooling channels. This recession rate can be compensated for, as the material can be extruded from a paste at a rate of 1 cm/s.

The thermo-mechanical response of an ATJ graphite sample to controlled runaway electron (RE) dissipation, realized in DIII-D, is modelled with a novel work-flow that features the RE orbit code KORC, the Monte Carlo particle transport code Geant4 and the finite element multiphysics software COMSOL. KORC provides the RE striking positions and momenta, Geant4 calculates the volumetric energy deposition and COMSOL simulates the thermoelastic response. Brittle failure is predicted according to the maximum normal stress criterion, which is suitable for ATJ graphite owing to its linear elastic behavior up to fracture and its isotropic mechanical properties. Measurements of the conducted energy, damage topology, explosion timing and blown-off material volume, impose a number of empirical constraints that suffice to distinguish between different RE impact scenarios and to identify RE parameters which provide the best match to the observations.

Carbon pebble rods are a promising candidate for use in high heat flux regions of magnetic fusion energy reactor walls. Under high (10 to 50 MW/m2) heat loads, carbon pebble rods release hot pebbles from the exposed surface, carrying away heat as the pebble rod surface recedes. In this work, we show that the surface recession rate during heating can be adjusted by changing the mechanical strength of the extruded rods, modifying the heat removal rate; this is accomplished here by varying the fill fraction of the inter-pebble matrix. A three-dimensional finite element model is presented that captures many experimental observations, including the sphere temperature and the surface recession rate. The model predicts that pebble release is caused by thermally driven crack propagation through the matrix and that the matrix strength against breaking is the single most important material parameter setting the pebble release rate; this prediction is supported by experimental results.

Current profile reconstructions are obtained for high current ( I p similar or equal to 550 kA) post-disruption runaway electron (RE) plateau plasmas in DIII-D. Two novel methods of measuring the RE current profile in high-current RE plateaus are introduced and compared: localization of the q = 2 rational surface using visible synchrotron emission (SE) imaging and the measurement of the polarization angle of line-integrated Ar-II line emission. The two methods are found to be consistent with each other within the data uncertainties. Different simulations of the RE current profile are compared with the measurements: the toroidal fluid RE model is found to best fit the data, within the measurement uncertainties. In addition to introducing two novel methods to measure the RE current profile and validating present simulation capabilities, this work demonstrates that instabilities can grow at q = 2 and q = 1 surfaces without necessarily causing a RE final loss instability. Numerical simulations are also presented to elucidate the role of these instabilities on synchrotron emission.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

In this work we utilize the recently upgraded Thomson scattering diagnostic to resolve density and temperature plasma profiles after pure deuterium and mixed neon/deuterium shattered pellet injections (SPIs) on DIII-D. This allows us to study individual components of the staggered scheme proposed for disruption mitigation on ITER, consisting of a low-Z material SPI followed by a delayed high-Z SPI. Obtained spatio-temporal density profiles exhibit very different dynamics after dominantly neon and pure deuterium SPIs. The neon SPI causes a fast radiative plasma collapse in a few milliseconds and results in almost flat density profile once the impurity mixes with the plasma during and after the thermal quench (TQ). The deuterium SPI leads to a disruption delayed by ten and more milliseconds, but very limited core fueling can be observed before the disruption. Even during and after the TQ, the edge deuterium density significantly exceeds the core density. 1D transport modeling suggests that this poor core fueling can be explained by strong outward grad-B-induced drift of the injected deuterium. Preliminary simulations show that larger pellet shards and greater injected quantity can be used to improve the penetration of the low-Z material into the core. These results call for optimization and further evaluation of the staggered SPI on ITER.

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant beta phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent 'pedestal-to-divertor' integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T-e and radiation which need control. High-beta(P) scenarios with q(min) > 2, q(95)-7.9, beta(N) > 4, beta(T)-3.3% and H-98y2 > 1.5 are sustained with high density ((n) over bar = 7E19 m(-3), f(G)-1) for 6 tau(E), improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

Subcritical energetic electrons (SEEs) produced by the runaway electron (RE) avalanche source at energies below the runaway threshold are found to be the primary contributor to surface heating of plasma-facing components (PFCs) during final loss events. This finding is supported by theoretical analysis, computational modeling with the Kinetic Orbit Runaway electrons Code (KORC), and qualitative agreement with DIII-D experimental observations. The avalanche source generates significantly more secondary electrons below the runaway threshold, which thermalize rapidly when well-confined. However, during a final loss event, the RE beam impacts the first wall, and SEEs are deconfined before they can thermalize. Additionally, because the energy deposition length decreases faster than energy, the deposited energy density, and thus the maximum PFC surface temperature change, is larger for SEEs than REs. KORC simulations employ an analytic first wall to model particle deconfinement onto a non-axisymmetric wall composed of individual tiles. PFC surface heating is calculated using a 1D model extended to include an energy-dependent deposition length scale. Simulations of DIII-D qualitatively agree with infrared (IR) imaging only when SEEs from the avalanche source are included. These results demonstrate that SEEs are the dominant contributor to PFC surface heating and indicate that the avalanche source plays a critical role in the PFC damage caused during final loss events. The prominence of SEEs also has important implications for interpreting IR imaging, one of the primary diagnostics for RE-wall interaction diagnosis, despite REs dominating the energy and current density. This result improves predictions of wall damage due to post-disruption REs to estimate material lifetime and design RE mitigation systems for ITER and future reactors.

Experimental trends in thermal plasma partial recombination resulting from massive D 2 injection into high-Z (Ar) containing runaway electron (RE) plateaus in DIII-D and JET are studied for the purpose of achieving sufficiently low electron density ( n e ≈ 10 18 m − 3 ) to increase RE final loss MHD levels. In both DIII-D and JET, thermal electron density n e is found to drop by ∼100 × when the thermal plasma partially recombines, with a minimum at a vacuum vessel-averaged D 2 density in the range 10 20 − 10 21 m − 3 . RE effective resistivity also drops after partial recombination, indicating expulsion of the Ar content. The n e level after partial recombination is found to increase as RE current is increased. The amount of initial Ar in the RE plateau is not observed to have a strong effect on partial recombination. Partial recombination timescales of order 5 ms in DIII-D and 15 ms in JET are observed. These basic trends and timescales are matched with a 1D diffusion model, which is then used to extrapolate to ITER and SPARC tokamaks. Within the approximations of this model, it is predicted that ITER will be able to achieve sufficiently low n e values on time scales faster than expected RE plateau vertical drift timescales (of order 100 ms), provided sufficient D 2 or H 2 is injected. In SPARC, it is predicted that achieving significant n e recombination will be challenging, due to the very high RE current density. In both ITER and SPARC, it is predicted that achieving low n e will be easier with Ar as a background impurity (rather than Ne).

UEDGE-CRUMPET simulations indicate the impact of the molecular hydrogenic isotopologue effect under high-recycling LFS divertor conditions in DIII-D to be negligible for electron density and temperature profiles at the LFS target plate. A 30% decrease in molecular content, accompanied by a 10% increase in atomic content, is predicted for deuterium compared to hydrogen. The predicted isotopologue effect on the radiative power balance, validated with calibrated spectroscopy, is found to be small despite a 20% increase in LFS divertor molecular band emission for deuterium compared to hydrogen. The predictions and measurements show a negligible contribution of molecularly-induced atomic and direct molecular emission to the total radiative power balance under high-recycling conditions, consistent with previous EDGE2D-EIRENE investigations. The UEDGE-CRUMPET simulations were performed using effective hydrogen and deuterium rates considering molecular breakup and excitation processes for H2 and D2, calculated by the CRUMPET collisional-radiative model.

To avoid or mitigate runaway electron (RE) beams in ITER, RE-driven instabilities are actively studied as a complimentary technique to massive material injection. In this work we report experimental dependencies of Alfvenic instabilities driven by REs during the current quench in DIII-D on plasma and material injection parameters. These instabilities, observed in the frequency range of 0.1-3 MHz, correlate with increased RE loss and thus may play a role in non-sustained RE beams. It was found that as the toroidal magnetic field (B-T) decreases, the RE population becomes more energetic, the energy of instabilities increases, and no RE beam is observed when the maximum energy of REs exceeds 15 MeV (or when B-T is below 1.8 T). Analysis of disruptions at plasma core temperature (T-e) of 1 keV and 8 keV shows that the RE population is much less energetic (with the maximum energy of only about 3 MeV) when Te is high, and no instabilities are observed in this case. Besides disruptions above caused by Ar injection, cases with Ne and D-2 injections were also studied. Both Ne and D-2 injections cause no sustained RE beams, however, for different reasons. Measurements of the instability polarization indicate that it is of predominantly compressional nature at the edge, which is consistent with modeling suggesting excitation of compressional Alfven eigenmodes. However, drive of global Alfven eigenmodes is also possible at low frequencies.

This work presents first experiments toward the development of continuously renewable (extrudable) pebble-based carbon rods for use as plasma-facing components in extreme steady-state plasma flux environments. The primary envisioned application of this work is a first wall that can survive long-term in future magnetic fusion power reactors while also improving recovery of the reactor fuel (tritium and deuterium atoms). Bench tests applying extreme steady-state front-surface heat loads of up to 50 MW/m(2) are presented. Continuous pebble rod front-surface recession and intact pebble recovery are successfully demonstrated, at a rate of order 0.2 cm/s. Numerical simulations of the pebble rod front-surface recession are able to match observations reasonably well, indicating that the recession mechanism can be understood as occurring due to pebble thermal expansion and resulting shock and cracking of the inter-pebble binder. Tests of the pebble rod extrusion demonstrate that friction between the rods and the stainless steel extrusion channel is tolerably low (<50 N for the expected channel length) over a wide range of temperatures. Front-surface outgassing rates below 1000 Torr L/s/m(2) are achieved, believed to be sufficiently low for use in magnetic fusion reactors. Initial parametric scans over pebble rod size and binder fraction to vary front-surface recession rates are presented.

Systematic test particle tracing simulations for runaway electrons (REs) are performed for six post-thermal quench equilibria from DIII-D and ITER, where large scale, kink-like n = 1 (n is the toroidal mode number) magnetohydrodynamic (MHD) instabilities are found. The modeled particle guiding center orbits allow extraction of the effective diffusion–convection coefficients of REs in the presence of large three-dimensional (3D) perturbations up to 10% of the equilibrium toroidal field. With a fixed spatial distribution of the field perturbation, the RE transport coefficients along the plasma radial coordinate track reasonably well with the surface-averaged perturbation level. A substantial variation in the value of the transport coefficients—by three orders of magnitude in most cases, however, occurs with varying launching location of REs along the plasma radius. Large 3D perturbations almost always lead to comparable diffusion and convection processes, meaning that diffusion alone is insufficient to describe the particle motion. At lower (but still high) level of perturbation, the RE convection is found to be dominant over diffusion. A similar observation is made when the perturbation is too strong. In the presence of large perturbation, the dependence of the RE transport on the particle energy is sensitive to the spatial distribution of the perturbation. Based on numerically obtained RE transport coefficients, an analytic fitting model is proposed to quantify the particle diffusion and convection processes due to large MHD events in post-thermal quench plasmas. The model is shown to reasonably well reproduce the direct test particle tracing results for the RE loss fraction and can, thus, be useful for incorporating into other kinetic RE codes in order to simulate the RE beam evolution in the presence of large 3D perturbations.

Two camera systems are installed on the DIII-D tokamak at the toroidal positions of 90° (90° system) and 225° (225° system), respectively. The cameras have two types of relay optics, namely, a coherent optical fiber bundle and a periscope system. The periscope system provides absolute intensity calibration stability while sacrificing resolution (10 lp/mm), while the fiber system provides high resolution (16 lp/mm) while sacrificing calibration stability. The periscope is available only for the 90° system. The optics of the 225° system were designed for view stability, repeatability, and easy maintenance. The cameras are located inside optimized neutron, x ray and magnetic shielding in order to reduce electronics damage, reboots, and magnetic and neutron interference, increasing the overall system reliability. An automated filter wheel, providing remote filter change, allows for remote wavelength selection. A software suite automates camera acquisition and data storage, allowing for remote operation and reduced operator involvement. System metadata is used to streamline the data analysis workflow, particularly for intensity calibration. The spatial calibration uses multiple observable wall features, resulting in a reconstruction accuracy ≤2 cm.

Molecular deuterium line emission is observed in both the visible and ultraviolet (UV) wavelength ranges after massive (> 100 Torr-L) injection of D2 gas into post-disruption runaway electron (RE) dominated plasmas in the DIII-D tokamak. D2 UV line emission is found to be the dominant source of radiated power, surpassing D Lyα. Interpretive modeling with a collisional-radiative model (CRM) indicates that D2 radiation surpasses D radiation because Lyα is strongly trapped, while D2 UV lines are mostly untrapped. The CRM also indicates that the D2 line emission is completely dominated by RE impact (rather than thermal electron impact), so the D2 line emission can serve as a good diagnostic for the spatial localization of REs. Analysis of D2 visible lines indicates that the D2 molecules in the plasma are thermally equilibrated with the background plasma, with vibrational, rotational, and kinetic temperatures all near 0.3 eV. D2 spectroscopy therefore serves as a convenient diagnostic of background plasma temperature. Measurement of D2 radiated power also appears to serve as a useful diagnostic for constraining neutral transport modeling.

The first dynamic (time-dependent) measurements of impurity ion radial (cross field) and parallel (along-field) diffusion coefficients for post-disruption runaway electron plateaus are presented. Small (∼1 mm diameter) carbon or silicon pellets are fired into the edge of steady-state runaway electron (RE) plateaus, and the resulting radial and toroidal transport of singly charged impurity ions (C+ or Si+) is monitored with spatially distributed visible spectrometers. Consistent with previous steady-state particle balance estimates of Ar+ radial transport, radial (cross field) diffusion coefficients D⊥≈2–5 m2/s are obtained, about 10× larger than expected from neo-classical theory. Parallel diffusion coefficients D∥≈30–80 m2/s are estimated, also much (≈50×) larger than classical. It is speculated at present that these large diffusion coefficients may be due to turbulent transport. Indications of fairly significant (almost 2×) toroidal variation in electron density are seen in the RE plateaus, and this appears to cause some toroidal variation in impurity radial diffusion rates. Indications of slow (≈1 Hz) toroidal rotation in the impurity ions are observed, although the uncertainty in this measurement is large.

A series of experiments have been executed at JET to assess the efficacy of the newly installed shattered pellet injection (SPI) system in mitigating the effects of disruptions. Issues, important for the ITER disruption mitigation system, such as thermal load mitigation, avoidance of runaway electron (RE) formation, radiation asymmetries during thermal quench mitigation, electromagnetic load control and RE energy dissipation have been addressed over a large parameter range. The efficiency of the mitigation has been examined for the various SPI injection strategies. The paper summarises the results from these JET SPI experiments and discusses their implications for the ITER disruption mitigation scheme.

Small (∼1 mm) neon pellet fragments are fired into DIII-D H-mode plasmas, and resulting trajectory-averaged photon efficiencies (neutral neon ionization events for every photon emitted) of S/XB≈85 are estimated for Ne-I 640 nm by dividing the estimated initial pellet fragment mass by the measured number of emitted Ne-I photons. The experiments are modeled by running the Lagrangian particle (LP) fluid/magneto-hydrodynamic pellet code to estimate axial ablation plume neon density profiles and temperature profiles at each pellet position. These solutions are then fed into the PrismSPECT collisional-radiative code, which calculates resulting neon charge states and photon emission rates, giving a profile-average of S/XB≈109. The burnthrough plasma minor radius predicted by LP (ρ≈0.63) is reasonably close to the experimental observation ρ≈0.6. The modeling indicates that local S/XB is not constant along the pellet trajectory but tends to increase with increasing ablation rate. Non-equilibrium kinetics are predicted to be very important, while line trapping is predicted to be relatively unimportant (for Ne-I 640 nm S/XB).