Diagnostic tools for understanding the edge plasma behavior in fusion devices are essential. The main focus of the present work is to present the infra-red (IR) diagnostics installed on Tokamak Energy’s spherical tokamak (ST40) and the IR thermographic inversion tool, Functional Analysis of Heat Flux (FAHF). FAHF is designed for multi-2D thermographic inversions within the divertor tiles using the finite difference method and an explicit time stepping scheme. ST40’s re-entrant endoscope allows the acquisition of IR data with the highest available effective spatial resolution. With these data, FAHF calculates the plasma perpendicular heat flux density on the divertor—a crucial quantity for edge plasma analysis. Although FAHF demonstrates significant sensitivity to user-selected settings, precise heat flux values are recoverable by ensuring a sufficiently high resolution. Implications for the optimal resolution of both the code and the IR diagnostic system are discussed. FAHF’s simplifications are shown to give an error within 10% with respect to COMSOL Multiphysics® simulations. Finally, by means of comparison with Langmuir probe heat flux data, the accuracy of the FAHF heat fluxes is estimated to be satisfactory. As such, FAHF is proven to be a precise and accurate tool for IR thermographic inversions in ST40.

In this work, we cross-compare the state-of-the-art edge plasma codes SOLPS-ITER, SOLEDGE2D, and UEDGE in a reactor-relevant neon-seeded Divertor Tokamak Test scenario at nominal power, extending the simplified test-bed of Moscheni et al (2022 Nucl. Fusion 62 056009). Converged solutions targeting the same separatrix density and radiated power are obtained by adjusting the pumping albedo and the neon puffing rate. This higher-power scenario is generally characterised by substantial disagreement between the three codes, up to 78%-178% in peak heat fluxes. Discrepancies found in Moscheni et al (2022 Nucl. Fusion 62 056009) are indeed exacerbated, and new ones arise. Underlying causes include the over-penetration of neutrals implied by the unified ion-neutral temperature of UEDGE (observed in Moscheni et al (2022 Nucl. Fusion 62 056009)), here resulting in a 38%-114% over-estimation of core plasma densities. The particular set of EIRENE atomic-molecular reactions adopted is found to stiffly restrict the achievable code solutions, which results in the predicted effective charge Zeff changing from similar to 5 to similar to 8 at the outer mid-plane separatrix. The strong link between Zeff, main ion density and unified ion temperature emphasises the need of proper assessments of impurity cross-field transport, with its implications on core contamination and wall erosion. The advantages of extended plasma meshes are found to come with associated modelling intricacies, yet to be fully characterised but seemingly impacting on the activity around the secondary X-point of single-null magnetic topologies. An appreciable impurity particle imbalance, generated by the neon ion density floor, is noted-speculatively contributing to SOLEDGE2D's different radiation emission distribution, and expected to be even more deleterious for high-Z impurities in all the codes. Potential drivers of further discrepancies are the different Braginskii formulations of collision times and momentum sources in presence of impurities, and the SOLPS-ITER extra terms gaining importance around the detachment front. Outstanding questions unanswered in this work prompt further investigations.

The Heat Flux Engineering Analysis Toolkit (HEAT) (Looby et al., 2022) was used to simulate the heat flux, qtarget, on ST40 molybdenum divertors. Results were compared with experimental infrared (IR) data. Two shots, 11419 and 11376 at time instants 119 ms and 120 ms, with lower-biased disconnected double null geometries, were studied. Single-λ and multi-λ heat flux profiles were used as input in HEAT simulation. The simulated qtarget was used in OpenFOAM to calculate the corresponding temperature, T, on the divertors. Results showed that a good agreement on the simulated qtarget, T and shape of the heat flux on the divertor with the IR data was achieved when a multi-λ heat flux profile was used for 11419, and single-λ heat flux profile for 11376. This indicates that HEAT can be reliable in analyzing and understanding the heat loading of plasma facing components in tokamak devices.

Following successful campaigns on the compact high field spherical tokamak ST40, resistive gold foil bolometers have now been installed to measure the radiated power profile. Positioned on the midplane, two bolometer cameras offer perspectives of the horizontal and vertical planes, while the third camera, situated above the midplane, provides a vertical view of the top divertor plates, including the X-point and the strike points. These cameras use a commercial off-the-shelf data acquisition platform with modules specifically designed for bolometers in a Wheatstone bridge configuration, providing both phase sensitive detection and in situ calibration. This paper presents an overview of the design and functionality of all three cameras and the commissioning of the horizontal plane camera. The commissioning of the horizontal camera shows data from a range of plasma pulses, presenting a positive correlation between the relative profiles measured from the bolometer camera and an absolute extreme ultraviolet diode camera viewing along similar lines of sight.

Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MW/m2. The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (≥1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (≥0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.

The scrape-off layer parallel heat flux decay lengths measured at ST40, a high field, low aspect ratio spherical tokamak, have been observed to bifurcate into two groups. The wide group matches closely with the scale of ion poloidal Larmour radius and follows existing H-mode scalings, while the narrow group falls up to 10 times below scalings, on the scale of ion total Larmour radius. The onset of the narrow scrape-off layer width is observed to be associated with suppressed magnetic fluctuations, suggesting reduced electromagnetic turbulence levels in the SOL.

The scrape-off layer parallel heat flux decay lengths measured at ST40, a high field, low aspect ratio spherical tokamak, have been observed to bifurcate into two groups. The wide group follows established H-mode scalings (ranging between 2 to 8 mm) while the narrow group falls up to 10 times below these scalings (between 0.2 and 0.8 mm), being comparable to the ion total Larmor radius rather than the ion poloidal Larmor radius. The heat flux profiles of the latter group can only be described by a multi-exponential function, rather than the single exponential function convoluted with a Gaussian. The onset of the narrow scrape-off layer width is observed to be associated with suppressed magnetic fluctuations, suggesting reduced electromagnetic turbulence levels in the SOL.

ST40 is a compact, high-field (B-T0 <= 2.1T) spherical tokamak (ST) with a mission to expand the physics and technology basis for the ST route to commercial fusion. The ST40 research programme covers confinement and stability; solenoid-free start-up; high-performance operating scenarios; and plasma exhaust. In 2022, ST40 obtained central deuterium ion temperatures of 9.6 +/- 0.4 keV, demonstrating for the first time that pilot plant relevant ion temperatures can be reached in a compact, high-field ST. Analysis of these high-ion temperature plasmas is presented, including a summary of confinement, transport and microstability characteristics, and energetic particle instabilities. Recent scenario development activities have focused on establishing diverted H-mode plasmas across a range of toroidal fields and plasma currents, along with scenarios with high non-inductive current fractions. In future operations, beginning in 2025, a 1 MW dual frequency (104/137 GHz) electron cyclotron (EC) system will be installed to enable the study of EC and electron Bernstein wave plasma start-up and current drive. Predictive modelling of the potential performance of these systems is presented.

Embedding diagnostics in future pilot plants will be a challenging task, because of space-and irradiation-related concerns. Relying on high-fidelity synthetic diagnostics would then be valuable. The 3D Monte-Carlo ray-tracing code CHERAB allows the development of numerous synthetic spectroscopic diagnostics. Focus of the present work is the introduction of new CHERAB models. The forward modelling of a synthetic D camera in ST40, the privately funded, high-field spherical tokamak, owned and operated by Tokamak Energy Ltd, and the comparison against experimental data is chosen as a testbed for quality assessment. Main output of the study then consists of estimates of the neutral particle densities throughout the chamber, of crucial relevance within edge plasma studies. Starting from simple analytical models, a 2D D source in the poloidal plane is generated. However, the centre column limited plasmas in ST40 display an intrinsically-3D D emission, mostly localised around the discrete poloidal limiters on the centre column, not captured by any axisymmetric source model. Hence, a novel methodology is introduced in CHERAB to approximate the 3D non-toroidally-symmetric pattern via a piece-wise emission distribution. Irrespective of the geometry of the emission and size of the tokamak, the pronounced non-homogeneity in the edge plasma emission requires sub-millimetric (similar to power fall-off length) spatial resolution to guarantee an accurate estimate of the peak emission. Minimising the associated burden via implementation of a non-uniform source sampling algorithm, which is a modification of the standard CHERAB uniform sampling, results in a >10-fold reduction of the computational cost. The significantly-shortened simulation time also makes the inclusion of more sophisticated models affordable. Of potential appeal in view of highly-detached divertors, the approximation of optically thin plasma is dropped, and photon-plasma interactions are accounted for. Brand-new CHERAB models able to take into account phenomena of photon absorption and scattering are so introduced.

In this work, we study the effect of installing a liquid metal divertor (LMD) using a capillary-porous structure in the EU DEMO tokamak within the same envelope of the baseline solid divertor. We used the SOLPS-ITER code to model the scrape-off layer (SOL) plasma and neutrals, coupled to a target thermal model to enable the self-consistent calculation of the LM target erosion rate, and adopting a fluid neutral model for the sake of simplicity. First calculations considering only D and Li (or Sn) showed a significant reduction of the steady state target heat load with respect to simulations considering only D, thanks to vapor shielding . Nevertheless, the computed peak target heat flux (∼31 MW m −2 and ∼44 MW m −2 for Li and Sn, respectively) was still larger than/borderline to the power handling limit of the LMD concepts considered. Moreover, the impurity concentration in the pedestal—a proxy for the core plasma dilution/contamination—was computed to be above/close to tolerability limits suggested by previous COREDIV calculations. These results indicate that the operational window of an LMD for the EU DEMO, without any additional impurity seeding, might be too narrow, if it exists, and that Sn looks more promising than Li. A second set of calculations was then performed simulating Ar seeding in the SOL, to further reduce the target heat load, and consequently the metal erosion rate. It was found that the mitigation of the plasma heat load due to Ar radiation in the SOL effectively replaces the radiation associated to vapor shielding in front of the target, thus allowing to operate the LMD in a regime of low target erosion. The resulting operational window was found to be significantly wider, both in terms of tolerable peak target heat flux and of acceptable core plasma contamination.

As reactor-level nuclear fusion experiments are approaching, a solution to the power exhaust issue in future fusion reactors is still missing. The maximum steady-state heat load that can be exhausted by the present technology is around 10 MW m −2 . Different promising strategies aiming at successfully managing the power exhaust in reactor-relevant conditions such that the limit is not exceeded are under investigation, and will be tested in the Divertor Tokamak Test (DTT) experiment. Meanwhile, the design of tokamaks beyond the DTT, e.g. EU-DEMO/ARC, is progressing at a high pace. A strategy to work around the present lack of reactor-relevant data consists of exploiting modelling to reduce the uncertainty in the extrapolation in the design phase. Different simulation tools, with their own capabilities and limitations, can be employed for this purpose. In this work, we compare SOLPS-ITER, SOLEDGE2D and UEDGE, three state-of-the-art edge codes heavily used in power exhaust studies, in modelling the same DTT low-power, pure-deuterium, narrow heat-flux-width scenario. This simplified, although still reactor-relevant, testbed eases the cross-comparison and the interpretation of the code predictions, to identify areas where results differ and develop understanding of the underlying causes. Under the conditions investigated, the codes show encouraging agreement in terms of key parameters at both targets, including peak parallel heat flux (1%–45%), ion temperature (2%–19%), and inner target plasma density (1%–23%) when run with similar input. However, strong disagreement is observed for the remaining quantities, from 30% at outer mid-plane up to a factor 4–5 at the targets. The results primarily reflect limitations of the codes: the SOLPS-ITER plasma mesh not reaching the first wall, SOLEDGE2D not including ion-neutral temperature equilibration, and UEDGE enforcing a common ion-neutral temperature. Potential improvements that could help enhance the accuracy of the code models for future applications are also discussed.

This work presents an overview of the integrated strategy developed, as part of the DEMO Key Design Integration Issue 1 (KDII1), to protect the EU-DEMO first wall (FW) from planned and unplanned plasma transients by employing discrete limiters. The present Breeding Blanket (BB) FW design, which aims at minimizing the loss of neutrons while travelling to the breeding zone, is able to withstand steady state heat fluxes up to ≈1-1.5 MW/m² [1], which is not sufficient to guarantee its integrity for most plasma-FW direct contact. This is different from ITER, which has a FW designed for peak heat loads up to 4.6 MW/m2 [2], and it does not have the DEMO BB breeding related requirement. A series of documents was compiled in the DEMO Pre-Conceptual Design Phase, in support of the KDII1. The work presented here was presented at the 2020 DEMO Gate 1 (G1) review, and collects also the comments of the panel and the relative additional studies triggered by them. The design process, presented in this paper was adopted to systematically evaluate the impact of design changes, or new physics inputs, on the FW protection strategy and integration issues. It includes compiling the list of transients, and performing the relative plasma simulations, the design of discrete limiters and the evaluation of their capability to reduce the heat flux density on the FW, and finally a preliminary analysis of the heat loads effects on the Plasma Facing Components (PFC). All these aspects, together with preliminary limiter design, where considered since the beginning, in an integrated way.

The pre-conceptual layout for an electron cyclotron system (ECS) in DEMO is described. The present DEMO ECS considers only equatorial ports for both plasma heating and neoclassical tearing mode (NTM) control. This differs from ITER, where four launchers in upper oblique ports are dedicated to NTM control and one equatorial EC port for heating and current drive (H&CD) purposes as basic configuration. Rather than upper oblique ports, DEMO has upper vertical ports to allow the vertical removal of the large breeding blanket segments. While ITER is using front steering antennas for NTM control, in DEMO the antennas are recessed behind the breeding blanket and called mid-steering antennas, referred to the radially recessed position to the breeding blanket. In the DEMO pre-conceptual design phase two variants are studied to integrate the ECS in equatorial ports. The first option integrates waveguide bundles at four vertical levels inside EC port plugs with antennas with fixed and movable mid-steering mirrors that are powered by gyrotrons, operating at minimum two different multiples of the fundamental resonance frequency of the microwave output window. Alternatively, the second option integrates fixed antenna launchers connected to frequency step-tunable gyrotrons. The first variant is described in this paper, introducing the design and functional requirements, presenting the equatorial port allocation, the port plug design including its maintenance concept, the basic port cell layout, the transmission line system with diamond windows from the tokamak up to the RF building and the gyrotron sources. The ECS design studies are supported by neutronic and tokamak integration studies, quasi-optical and plasma physics studies, which will be summarized. Physics and technological gaps will be discussed and an outlook to future work will be given.

We developed a first single null divertor scenario for EU-DEMO with the SOLPS-ITER code, including all charge states of D, He and Ar in the simulation and kinetic analysis of the neutral gas, but still leaving out drifts. Our results suggest that a partially detached divertor condition can indeed be obtained, with a corresponding peak heat flux on targets to ≈2 MW m−2, within the expected tolerable steady-state limits. The plasma re-attaches ≈30 cm (measured along the plate) from the low field side strike point deep in the SOL, with a consequent raise in the electron temperature up to ≈16 eV outside the detached region, which requires further analysis to assess a possible role of sputtering, especially considering W self-sputtering. The plasma we developed exhibits Z eff ≈ 1.9 in the core, and a pumping speed a few hundred’s m3 s−1 to maintain steady-state; this points to the need of further studies to better characterize the He accumulation in the core region.

Database for reproducing the calculations presented in the publication "Parametric Study of the Radiative Load Distribution on the EU-DEMO First Wall Due to SPI-Mitigated Disruptions", submitted to Fusion Engineering and Design. Work carried out within the framework of the EUROfusion Consortium.

The EU-DEMO First Wall (FW) will be a relatively thin structure. In order not to damage this layer, heat loads distributed onto the wall should be carefully controlled. In the case of transient events, as for example plasma disruptions, the steady-state heat load limit (∼1-2MW/m2) can be largely exceeded for a timespan sufficiently long to cause damages. Therefore, when the control system detects an upcoming disruption, Shattered Pellet Injection (SPI) or Massive Gas Injection (MGI) mitigation techniques can be employed to inject impurities and switch off the plasma safely. In the present work, the Monte-Carlo ray-tracing code CHERAB is used to compute the radiative heat load distribution on the EU-DEMO Plasma Facing Components (PFCs) due to a mitigated plasma disruption. By applying ad-hoc techniques to improve the quality of the Monte Carlo calculation, we obtain a peak radiative load of ∼490MW/m2 on the PFCs, which is ∼25% lower than previous estimates.