Fusion Energy Research Articles

Generation of shear flows induced by AE / EPM in LHD plasma

Abstract The generation of shear flows (SFs) by Alfven Eigenmodes (AEs) and energetic particle modes (EPMs) have important effects on the operation of future nuclear fusion reactors, because SFs regulate the saturation of the AEs/EPMs, the transport of EPs and thermal plasma, as well as the formation of transport barriers among other consequences. The aim of this study is the analysis of SFs generation during the saturation phase of AEs and EPMs in LHD plasma. Experiments performed in the 23rd and 24th LHD experimental campaigns are dedicated to explore the destabilization of AEs/EPMs in discharges with different heating patterns, thermal plasma and magnetic field configurations. In particular, the shots 176490 and 179697 show the destabilization of MHD bursts and energetic-ion-driven resistive interchange modes (EIC), respectively. Charge exchange spectroscopy measurements in both discharges indicate that the generation of SFs by AE/EPM is uncorrelated with the perturbation induced by the neutral beam injector (NBI). Nonlinear simulations performed using the gyro-fluid code FAR3d show the generation of zonal structures, especially SFs, induced during the saturation phase of Toroidal Alfven Eigenmodes (TAEs) triggered in the MHD burst as well as by the 1 / 1 EIC in the bursting phase. The simulations indicate that SFs are caused by the radial electric fields powered by energy transfers from the unstable AE/EPM towards the thermal plasma. The strongest SFs are measured during the EIC bursting phase once the 1 / 1 EPM overlaps with nearby resonances at the plasma periphery. Likewise, the largest SFs during the MHD burst are observed once TAEs radially overlap in the inner-middle plasma region.

Characterizing the negative triangularity reactor core operating space with integrated modeling

Abstract Negative triangularity (NT) has received renewed interest as a fusion reactor regime due to its beneficial power-handling properties, including low scrape-off layer power and a larger divertor wetted area that facilitates simple divertor integration. NT experiments have also demonstrated core performance on par with positive triangularity (PT) H-mode without edge-localized modes (ELMs), encouraging further study of an NT reactor core. In this work, we use integrated modeling to scope the operating space around two NT reactor strategies. The first is the high-field, compact fusion pilot plant concept MANTA (The MANTA Collaboration et al 2024 Plasma Phys. Control. Fusion 66 105006) and the second is a low field, high aspect ratio concept based on work by Medvedev et al (Medvedev et al 2015 Nucl. Fusion 55 063013). By integrating equilibrium, core transport, and edge ballooning instability models, we establish a range of operating points with less than 50 MW scrape-off layer power and fusion power comparable to positive triangularity (PT) H-mode reactor concepts. Heating and seeded impurities are leveraged to accomplish the same fusion performance and scrape-off layer exhaust power for various pressure edge boundary conditions. Scans over these pressure edge conditions accommodate any current uncertainty of the properties of the NT edge and show that the performance of an NT reactor will be extremely dependent on the edge pressure. The high-field case is found to enable lower scrape-off layer power because it is capable of reaching high fusion powers at a relatively compact size, which allows increased separatrix density without exceeding the Greenwald density limit. Adjustments in NT shaping exhibit small changes in fusion power, with an increase in fusion power density seen at weaker NT. Infinite-n ballooning instability models indicate that an NT reactor core can reach fusion powers comparable to leading PT H-mode reactor concepts while remaining ballooning-stable. Seeded krypton is leveraged to further lower scrape-off layer power since NT does not have a requirement to remain in H-mode while still maintaining high confinement. We contextualize the NT reactor operating space by comparing to popular PT H-mode reactor concepts, and find that NT exhibits competitive ELM-free performance with these concepts for a variety of edge conditions while maintaining relatively low scrape-off layer power.

Impact of MHD mixed convection on heat and mass transfer in a typical DCLL breeder blanket

Abstract A typical dual-coolant lead lithium (DCLL) liquid blanket employs eutectic liquid lithium lead alloy (PbLi) as coolant and tritium breeder for Tokamak nuclear fusion reactor, and is responsible for transporting heat and tritium. The blanket is subjected to both strong magnetic field and strong radially non-uniform neutron volume heating. The flow of liquid metal is influenced by the combined effects of magnetohydrodynamics (MHD), buoyancy, and inertial effects, resulting in a complex MHD mixed convection phenomenon. This study employs direct numerical simulation (DNS) based on finite volume method to analyze the flow, heat transfer, and mass transfer within blanket unit. The results demonstrate that mixed convection has a significant impact on the MHD pressure drop and flow characteristics. The complex MHD mixed convection gives rise to a considerable number of vortices, particularly in regions of the turn and the downward poloidal duct. The formation of low-speed zones and jets within the blanket has the potential to significantly impact heat and mass transfer, leading to the accumulation of heat and tritium. On the other hand, the markedly elevated tritium production rate is a significant contributing factor to the obviously elevated tritium concentration in the vicinity of the first wall.

Reconstruction of the fast-ion deuterium distribution in a tritium-rich plasma in the JET DTE2 campaign

Abstract An important step on the way to future fusion power plants was the 2021 deuterium-tritium experimental campaign (DTE2) at the Joint European Torus (JET), in which crucial DT physics was investigated. In this study, we have reconstructed the fast-ion deuterium distribution function in JET discharge 99971 which broke the former fusion energy record. It is the first time that the fast-ion distribution has been reconstructed from experimental data in a DT discharge. The reconstruction shows that the fast-ion deuterium distribution is anisotropic, with a bias towards co-going ions (p>0). The fast-ion deuterium distribution likely peaks in energy (E) at around 60-70 keV and has a marginal high-energy tail (approx. above 180 keV). Furthermore, an orbit analysis shows that the fast-ion distribution is composed of mostly co-passing orbits (50 %), trapped orbits (21 %) and counter-passing orbits (27 %), as well as a small population of potato orbits (1.7 %) and counter-stagnation orbits (0.3 %). The orbit-type constituents of the neutron measurements are distributed in similar fractions.

Indication of p + 11B reaction in Laser Induced Nanofusion experiment

The NanoPlasmonic Laser Induced Fusion Energy (NAPLIFE)1 project proposed fusion by regulating the laser light absorption via resonant nanorod antennas implanted into hydrogen rich urethane acrylate methacrylate (UDMA) and triethylene glycol dimethylacrylate (TEGDMA) copolymer targets. In part of the tests, boron-nitride (BN) was added to the polymer. Our experiments with resonant nanoantennas accelerated protons up to 225 keV energy. Some of these protons then led to p + 11B fusion, indicated by the sharp drop of observed backward proton emission numbers at the 150 keV resonance energy of the reaction. The generation of alpha particles was verified by CR-39 (Columbia Resin #39) nuclear plastic track detectors.

A case for gross electricity producing compact fusion pilot plants

Abstract A case for compact gross electricity producing pilot plant is presented.
The feasibility of such a plant with a moderate fusion power that is capable of delivering
gross electricity to the grid is investigated. The physics and engineering considerations
of such power plants are elucidated. We show that for a fusion power of about 300 MW
with fusion gain of 5, a moderate plasma β with improved confinement regime is
required to prevent excessive transport power loss. The sensitivity analysis indicates a
wide enough parameter range where, the fusion power and fusion gain can meet their
target values. The constraints arising from the shielding, magnets and maintenance
are discussed. The feasibility of steady-state gross electricity production of 160 MW
is discussed using a helium-cooled solid breeder blanket with an intermediate energy
storage system. It is argued that such a plant has all key technical elements of DEMO,
albeit at a smaller scale, thereby providing strong technical basis for DEMO.

Physics and technology maturity level required for the K-DEMO design points

The conceptual design study for the Korean fusion demonstration reactor (K-DEMO) has been conducted, placing an emphasis on the imperative to reduce the reactor's dimension as a means to cost minimization. The design space of K-DEMO was delineated based on maturity level of physics and technology. Advanced technology features such as a use of a tungsten carbide (WC) shield, a tritium breeding blanket concept of He cooled lithium lead (HCLL), the maximum allowable magnetic field at the TF coil, Bmax = 16 T with a Nb3Sn superconducting material, etc. were adopted. With specified design criteria, including a net electric power ≥ 300 MW, a fusion gain, Q > 20.0, a neutron wall loading < 2.0 MW/m2, an indicator of divertor power handling capability (ratio of power to divertor to major radius), Pdiv/R0 < 25 MW/m, and the capability for steady-state operation, a design space of the K-DEMO was established based on the energy confinement scaling law of IPB98[y,2] under physics level of Greenwald density fraction, ne/nG < 1.2, normalized plasma beta (ratio of plasma pressure to magnetic pressure normalized by plasma current divided by the product of minor radius and toroidal magnetic field), βN < 3.0, confinement enhancement factor, H < 1.3, and a direct cost ≤ 7.5 B$. After an exploration of system parameters, prospective design points for K-DEMO were identified, characterized by a major radius, R0 ∼ 6.8 m, an aspect ratio, A = 3.1, a toroidal magnetic field at plasma center, BT ≥ 6.5 T, and a fusion power, Pfusion ∼ 1,500 MW. When β-independent energy confinement scaling law was applied, the design points were accessible with smaller ne/nG, βN, H, Pfusion, and larger BT. From a sensitivity analysis of the minimum major radius to the input parameters, strong sensitivities to Greenwald density fraction, ne/nG, normalized plasma beta, βN, confinement enhancement factor, H, edge safety factor, qedge, and elongation, κ were found. Additionally, the operational envelope in physics and technology parameters was established with system parameters associated with the design points.

Optimization design of multi-scale TPMS lattices based on geometric continuity fusion and strain energy driven

3D lattice structure is a kind of advanced metamaterial structure with excellent performance, such as lightweight, high specific strength and stiffness, shock damping, heat dissipation, and electromagnetic shielding capabilities. With the development of additive manufacturing technology, it has been widely concerned and developed rapidly in recent years. In order to further integrate the performance of lattice structures with mesoscale units into multi-scale optimization, this paper proposes an optimization design method for the multi-scale TPMS (Triply Periodic Minimal Surfaces) lattices based on geometric continuity fusion and strain energy driven. Regarding the challenge of geometric continuity problem of complex transition boundaries in multi-TPMS lattices, interface interpolation and density factor optimization methodologies are studied. At the same time, the mapping mechanism between lattice configuration and mechanical properties of multi-TPMS lattices is discussed in view of the requirements of multi-scale design from macro to micro configurations. Then, the multi-TPMS lattice density and configuration fields in the design domain are co-designed with the strain energy driven. Simulation and experimental results show that the mechanical properties of the multi-scale TPMS lattice designed by the proposed method are significantly improved compared with the traditional single-configuration gradient lattice, including the stiffness of the cantilever beam optimized structure is improved by 20.5% and the flexural stiffness of the three-point flexural beam optimized structure is improved by 51.3%.

Atomistic simulation of chemical short-range order on the irradiation resistance of HfNbTaTiZr high entropy alloy

High entropy alloys (HEAs) have been considered as one of the potential structural material candidates for fourth-generation nuclear reactors and fusion reactors due to their excellent irradiation resistance. Current studies have shown that the chemical short-range order (CSRO) usually exists in HEAs, which has a significant effect on the mechanical properties and irradiation resistance of HEAs. Refractory high entropy alloys (RHEAs), as a new class of HEAs have better mechanical properties at high temperatures than face-centered cubic (FCC) HEAs, and therefore have better prospects of application in the nuclear field. In this study, CSRO and its effect on the irradiation resistance of HfNbTaTiZr are analyzed via molecular dynamics (MD) and Monte Carlo (MC). The primary cascade simulations, multi-cascade simulations and surface bombardment simulations are carried out to simulate the generation and accumulation of irradiation damage. The results of the primary cascade simulations and surface bombardment simulations of CSRO models show that the presence of CSRO induces cascade splitting into subcascades. The presence of subcascades reduces the thermal peak enhancement effect and thus lowers the recombination rate of Frenkel pairs (FPs) in the damage zone when FPs concentrations are low. However, the creation of subcascades increases the size of the damage zone caused by the cascade. Thus, when the concentrations of FPs are high, the larger area of damage zone allows more of the already existing FPs to be included, thus promoting their recombination, i.e., impedes their accumulation when concentrations are high. These subcascades lower the recombination of FPs at low FPs concentrations but inhibit their accumulation at high FPs concentrations. The presence of CSRO is also beneficial in inhibiting the growth of point defect clusters, which further improves the resistance of HfNbTaTiZr to dislocation generation. Furthermore, the presence of CSRO facilitates the irradiation-induced phase transition. But it is found that HfNbTaTiZr shows suppression of HCP cluster growth. And the tendency to break down large HCP clusters into smaller ones is demonstrated in the CSRO model. From our calculations we also find that the irradiation-induced HCP atoms have a higher potential energy relative to the matrix. The potential energy difference between those energetic HCP atoms and the matrix can lead to generating a great number of insurmountable barriers pervading the matrix and largely suppressing the long-term mobility of FPs, thus limiting their aggregation and growth into clusters.

Advancing sustainability in Electron and laser beam powder Bed Fusion technologies via Innovation: Insights from patent analysis

Additive manufacturing (AM) technologies continuously evolve in materials and operational processes. However, challenges related to energy consumption, material reuse efficiency, and the integration of Circular Economy (CE) principles may impede the overall sustainability of AM processes. As many industrial sectors adopt AM technologies, there is a growing need for innovative tools, methods, and frameworks that guide companies toward more sustainable and circular production practices. This shift involves improving resource efficiency and fostering new Circular Business Models that emphasize material recovery, product lifecycle extension, and waste minimization. This study assesses current and future trends in the development of sustainable AM technologies for processing metal feedstock, based on a comprehensive patent analysis covering the period from 2004 to 2024. We specifically focus on the sustainability impacts of Electron and Laser Beam Powder Bed Fusion (PBF-EB, PBF-LB) and Direct Energy Deposition (DED), given their widespread industrial applications. Patents were categorized into three groups: materials, processes, and components. Furthermore, AM technologies were analyzed according to their role within the production cycle: materials preparation, pre-processing, manufacturing, and post-processing. This categorization allows for a detailed understanding of how innovations in AM contribute to more sustainable production and consumption practices by improving energy efficiency, material usage, circularity, and overall environmental impact.

Machine learning powered predictive modelling of complex residual stress for nuclear fusion reactor design

Fusion In-vessel components, assembled and maintained using laser welding, one of the most promising techniques, exhibit complex distributions of residual stress, microstructures, and material properties. These residual stresses can compromise structural integrity and lifespan of critical components. Although using advanced experimental measurements can evaluate the residual stress for individual case, extending the measurements to massive number of components are costly and time-consuming. Traditional machine learning (ML) models struggle to account for the heterogeneity and anisotropy of these stress distributions. Here, we develop a novel ML framework based on the Eurofer97 steel, the structural material for in-vessel components. The ML framework is trained on high-resolution residual stress data derived from recently-developed evaluation techniques. Combining with microstructures, the model enables prediction of heterogenous and anisotropic residual stress distribution. It successfully predicts the compressive residual stress in fusion zone (∼−200 MPa) balanced by tensile residual stress in heat affected zone (∼300 MPa), aligning closely with experimental results with the R-squared value of 0.989 and the mean square error of 10−4. Unlike experiments that take hours, the ML model provides predictions within seconds. It offers valuable insights into residual stress prediction for various joints, enhancing the reliability and lifetime prediction of in-vessel components.

UK coalition gears up to demonstrate commercial viability of fusion energy

The government is aiming to build a prototype plant in the next two decades.

Erosion enhancement by impurity entrainment in the highly collisional plasmas of Magnum-PSI

Abstract Extrinsic impurity seeding is planned for ITER to avoid high heat fluxes to the tungsten divertor targets. Nevertheless, excessive sputtering by impurities would lead to a reduction in fusion power output and divertor lifetime. Unlike today's fusion devices, the entrainment of impurities is expected to play an increasing role in ITER's highly collisional near-surface plasma. Impurity entrainment refers to the acceleration of impurities with the plasma flow by collisional drag, leading to elevated impact energies. To investigate impurity entrainment under ITER-like plasma conditions ($n_e &gt; 2 \cdot 10^{20}$ m$^{-3}$, $T_e &lt;$ 5 eV), argon-seeded hydrogen plasmas were formed with the linear plasma generator Magnum-PSI. Doppler shifts in spatially-resolved emission profiles of argon (Ar II - 480.6 nm) and hydrogen (H$_\mathrm{\beta}$ - 486.1 nm) reveal the entrainment of seeded argon towards the upstream hydrogen velocity ($\sim$9 cm from surface), corresponding to $\sim$0.39 of the common-system sound speed. In addition, the sputtering yield of tungsten by argon bombardment for the argon-seeded hydrogen plasma was compared to a pure argon plasma by monitoring tungsten (W I - 400.9 nm) emission. These sputtering experiments indicate further entrainment of the argon impurities to $\sim$0.65 of the common-system sound speed at the sheath edge, leading to a large increase in their impact energy. Extrapolation of the Magnum-PSI results to ITER, using existing SOLPS-ITER simulations to determine the divertor plasma conditions, indicates more than an order of magnitude increase in gross erosion of the divertor targets due to impurity entrainment. However, the net erosion rate could still be kept under control if high re-deposition rates are achieved, which is treated in the companion paper \cite{Cornelissen2023}. The combination of a low sputtering threshold energy and high impact energy of impurities constrains the fine balance between radiative cooling with impurity seeding and avoiding extensive divertor erosion.

Novel high temperature tritium blanket designs for confined spaces in spherical tokamak fusion reactors

Tritium self-sufficiency is one of the fundamental challenges for commercially viable deuterium–tritium nuclear fusion power stations. The combination of key high temperature radiation shielding materials that possess dense, high neutron absorption cross-section, and moderation properties, and tritium breeding materials could involve interesting design spaces for the central column challenge in spherical tokamaks. Potential tungsten alloys can be used for two functions: radiation shielding and structural material, providing a new design space window for spherical tokamak central column breeding space. In this paper, we present two novel high temperature concepts for the inboard side of the breeder blanket in a confined space, such as a spherical tokamak. A tungsten–rhenium–hafnium-carbide lithium-based design was found to offer the best TBR given a parameter optimisation based on shielding and thermal requirements. A silicon-carbide lead-lithium breeder design was also investigated. The highest TBR was found to be 0.135 in a 3D neutronics calculation with a W-24.5Re-2HfC (structural and shielding, wt%), Li (90% Li-6 enriched breeder), and tungsten pentaboride (W2B5) (shielding) option. Although this TBR is lower than unity, it will contribute to the reactor’s global TBR.

Detachment scalings derived from 1D scrape-off-layer simulations

Fusion power plants will require detachment to mitigate sputtering and keep divertor heat fluxes at tolerable levels. Controlling detachment on these devices may require the use of real-time scrape-off-layer modeling to complement the limited set of available diagnostics. In this work, we use the configurable Hermes-3 edge modeling framework to perform time-dependent, fixed-fraction-impurity 1D detachment simulations. Although currently far from real-time, these simulations are used to investigate time-dependent effects and the minimum physics set required for control-relevant modeling. We show that these simulations reproduce the expected rollover of the target ion flux — a typical characteristic of detachment onset. We also perform scans of the input heat flux and impurity concentration and show that the steady-state results closely match the scalings predicted by the 0D time-independent Lengyel–Goedheer model. This allows us to indirectly compare to SOLPS simulations, which find a similar scaling but a lower value for the impurity concentration required for detachment for given upstream conditions. We use this result to suggest a series of improvements for the Hermes simulations, and finally show simulations demonstrating the impact of time-dependence.

Energetics of vacancy clusters in TiVTaW low-activation high-entropy alloy investigated by ab initio calculations

High-entropy alloys (HEAs) have attracted attention as potential candidates for nuclear fusion reactor materials due to their superior mechanical properties and irradiation resistance. The TiVTaW low-activation HEA exhibits an especially large variation in the cohesive energies of the pure metals of its alloying elements, leading to the characteristic of energetically heterogeneous atomic environments that differ at each atomic site, a focal point of this study. To evaluate the influence of this site-to-site variation on vacancy cluster formation, in this study, density functional theory calculations were conducted on small vacancy clusters consisting of the nearest neighbor configurations in equimolar TiVTaW. The monovacancy formation energies in TiVTaW were found to span a significantly broader range compared to those in other HEAs. Moreover, the range of vacancy cluster formation energies in TiVTaW was observed to be nearly identical to the range of the vacancy cluster formation energies of its alloying elements in the pure metals. Our calculations also showed that clusters with five or fewer vacancies exhibit a lower mean binding energy compared to the pure metals of the alloying elements, indicating that vacancy clusters in TiVTaW are less stable than those in pure metals. Additionally, the distribution of cluster binding energies for sizes less than five-vacancy clusters was found to range from −1.60 to 3.12 eV, which includes unstable clusters with negative binding energies. These energetic characteristics arising from the heterogeneous atomic environments unique to HEAs can influence void nucleation under irradiation. Furthermore, based on a thermodynamic model, local maxima in the free energy change correlating with cluster size were observed during the nucleation of vacancy clusters, indicating the generation of metastable small vacancy clusters. These findings provide a new perspective on irradiation damage mechanisms in general HEAs.

Parametric dependencies of ion temperature profile peaking in ASDEX Upgrade high-beta scenarios

Abstract Advanced Tokamak (AT) scenarios are attractive candidates for future nuclear fusion power plants. These scenarios often feature peaked temperature profiles, suggesting a local reduction of turbulent transport. The mechanisms behind this are, as yet, not fully understood. Parameters that are thought to be connected to these transport reductions include the q-profile and the ExB-shear ωExB. Another parameter considered to be important is the fast ion content, which can reduce transport in multiple different ways.&amp;#xD;This work presents AT experiments performed at the ASDEX Upgrade tokamak (AUG) in which these three parameters are varied to investigate their respective effect on the observed peaked ion temperature profiles. To disentangle potentially competing effects, simulations using the codes GENE and TGLF have been performed. It is found that the ExB-shear does not play a role in the AT scenarios performed at AUG, which have relatively low rotation, with vtor=100-250km/h. Experimentally, a significant dependence of R/LTi on the electron cyclotron current drive (ECCD) settings was found, indicating that either the current profile (and, therefore, the q-profile) or Te/Ti has a significant impact on the observed suppression of turbulent transport. In support of the first option, both GENE and TGLF show a q-profile dependence of R/LTi. Furthermore, according to GENE, electromagnetic (EM) fast ion effects are essential in reproducing the experimental results. Scans with TGLF (which does not include such effects), suggest that these fast ion effects become relevant above a threshold of R/LTi= 4-5. In the presence of ICRF, additional electrostatic (ES) fast ion effects seem to come into effect, according to GENE, albeit to a smaller degree than the EM effects.

Classical-quantum simulation of non-equilibrium Marshak waves

In the radiation hydrodynamic simulations used to design inertial confinement fusion (ICF) and pulsed power experiments, nonlinear radiation diffusion tends to dominate CPU time. This raises the interesting question of whether a quantum algorithm can be found for nonlinear radiation diffusion which provides a quantum speedup. Recently, such a quantum algorithm was introduced based on a quantum algorithm for solving systems of nonlinear partial differential equations (PDEs) which provides a quadratic quantum speedup. Here, we apply this quantum PDE (QPDE) algorithm to the problem of a non-equilibrium Marshak wave propagating through a cold, semi-infinite, optically thick target, where the radiation and matter fields are not assumed to be in local thermodynamic equilibrium. The dynamics is governed by a coupled pair of nonlinear PDEs which are solved using the QPDE algorithm, as well as two standard PDE solvers: (i) Python's py-pde solver; and (ii) the KULL ICF simulation code developed at Lawrence-Livermore National Laboratory. We compare the simulation results obtained using the QPDE algorithm and the standard PDE solvers and find excellent agreement.

Prediction of performance and turbulence in ITER burning plasmas via nonlinear gyrokinetic profile prediction

Abstract Burning plasma performance, transport, and the effect of hydrogen isotope (H, D, D-T fuel mix) on confinement has been predicted for ITER baseline scenario (IBS) conditions using nonlinear gyrokinetic profile predictions. Accelerated by surrogate modeling (Rodriguez-Fernandez et al 2022 Nucl. Fusion 62 076036), high fidelity, nonlinear gyrokinetic simulations performed with the CGYRO code (Candy et al 2016 J. Comput. Phys. 324 73), were used to predict profiles of Ti , Te , and ne while including the effects of alpha heating, auxiliary power (NBI + ECH), collisional energy exchange, and radiation losses inside of r / a = 0.9. Predicted profiles and resulting energy confinement are found to produce fusion power and gain that are approximately consistent with mission goals ( P fusion = 500 MW at Q = 10) for the baseline scenario and exhibit energy confinement that is within 1σ of the H-mode energy confinement scaling. The power of the surrogate modeling technique is demonstrated through the prediction of alternative ITER scenarios with reduced computational cost. These scenarios include conditions with maximized fusion gain and an investigation of potential resonant magnetic perturbation (RMP) effects on performance with a minimal number of gyrokinetic profile iterations required (3–6). These predictions highlight the stiff ITG nature of the core turbulence predicted in the ITER baseline and demonstrate that Q &gt; 17 conditions may be accessible by reducing auxiliary input power while operating in IBS conditions. Prediction of full kinetic profiles allowed for the projection of hydrogen isotope effects around ITER baseline conditions. The gyrokinetic fuel ion species was varied from H, D, and 50/50 D-T and kinetic profiles were predicted. Results indicate that a weak or negligible isotope effect will be observed to arise from core turbulence in IBS conditions. The resulting energy confinement, turbulence, and density peaking, and the implications for ITER operations will be discussed.

Heterogeneous Online Computational Platform for GEM-Based Plasma Impurity Monitoring Systems

The fusion energy research field presents many intricate challenges that require resolution. Many diagnostic systems employed in experiments are approaching the limits of current technology. Implementing efficient measurements requires using an appropriate set of tools to facilitate the optimal utilization of hardware. Fusion energy measurements must provide low latency processing with the capacity for future improvements and the ability to handle complex data flows efficiently. The presented work addresses these requirements and describes the implementation of a high-performance, low-latency software platform with convenient development for soft X-ray (SXR) plasma impurities emission tracing—the Asynchronous Complex Computation Platform (AC2P). This article presents the architectural design, implementation details, and performance and latency measurements based on the raw data acquired from the WEST tokamak and laboratory tests. AC2P provides the tools to develop low-latency, multi-core, multi-device complex data flow graph scale-up solutions for measuring impurities in hot plasmas. The system has been designed to operate online during experiments, calculate the energy distribution, position and occurrence time of SXR photons, monitor the data stream’s quality and archive any abnormalities for subsequent offline verification and algorithm improvement. This article presents AC2P, which operates as part of the SXR measurement system on the WEST tokamak.