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DOE OSTI2020

Fiber-Optic Silicon Fabry-Perot Interferometric Bolometer: The Influence of Mechanical Vibration and Magnetic Field

Plasma radiation plays an important role in the operation and control of magnetic confinement fusion and is typically measured by bolometers. Here, we study the influence of mechanical vibration and magnetic field, which are typically present in magnetic confinement fusion devices, on the performance of a fiber-optic bolometer (FOB) based on silicon Fabry-Perot interferometer (FPI) demodulated by a scanning diode laser. It is found that both the vibration and magnetic field could cause large noise due to the birefringence of the silicon FPI. To mitigate the birefringence effects, we demonstrated two effective methods: polarization maintaining FOB and polarization scrambling. We show that two methods can be effective in mitigating the effect caused by the sensor birefringence: (1) replacing the regular single-mode fiber in the FOB system with polarization-maintaining fiber and (2) using a polarization scrambler after the laser source. For both methods, no significant increase in the noise was observed in the vibration and magnetic field tests. Finally, our research highlights the importance of the polarization management for the FOB systems toward their practical applications in magnetic-confinement fusion systems.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Predicting the critical gradient of ITG turbulence in fusion plasmas

The quasilinear mixing-length approach to efficient prediction of transport in fusion devices is improved to account for the ‘Dimits’ upshift between linear and nonlinear critical pressure gradients in zonal-flow-saturated turbulence regimes. This modification uses the frequency mismatch between modes interacting turbulently to track changes in saturation efficiency. Near criticality, energy is transferred exclusively to stable eigenmodes, rapidly increasing the efficacy of the nonlinearity. Finally, the modified quasilinear model is able to predict below-threshold turbulent ion-temperature-gradient-driven transport accurately and also yields significantly improved predictions for trapped-electron-mode transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Developments and Challenges in the Design of the ITER DRGA

We report the ITER Diagnostic Residual Gas Analyzer (DRGA) will measure the distribution of gas species, i.e., deuterium (D), tritium (T), and impurities, in the divertor exhaust stream and in the plasma periphery, with time resolution relevant to fusion plasma–wall particle dynamics. The uniqueness of the DRGA, over previous implementations of plasma dynamics residual gas analysis, is an integrated approach, combining mass and low-temperature plasma-activated optical spectroscopy, in a differentially pumped analysis station. A further unique feature of the ITER divertor-specific DRGA is an ~8-m separation of the analysis station from the sampled pumping duct, while still maintaining a ~1-s response time for hydrogen isotopic concentrations. ITER DRGA final design activities are strongly benefiting from testing of prototypical DRGA components and methods on present fusion devices, most currently on JET and W7-X. DRGA systems are implemented on both these devices and include sensors (and pumping methods) that are directly relevant to the ITER DRGA design. The recent JET-DTE2 campaign has provided the first experience on operating the combined ITER DRGA sensors with D-T plasmas. While enhancing system design for ITER, this experience on operating devices has also revealed additional engineering challenges, which further guide the continuing final design project. Meanwhile, the recent determination that the ITER DRGA, with slight optimization, will resolve the helium isotopes well enough to support an ITER pre-DT, He-3-based heating scheme, has greatly increased ITER Research Program interest in the DRGA and its implementation well ahead of the DT phase.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Error field correction strategies in preparation to MAST-U operation

In magnetic fusion devices, unwanted non-axisymmetric magnetic field perturbations, known as error fields (EF), can have detrimental effects on plasma stability and confinement. Such EFs may originate from several sources, such as axi-symmetric coil misalignments, coil feeds, 3D structures in the wall surrounding the plasma, presence of ferromagnetic materials near the plasma surface, blanket materials in future devices. To minimize their impact on plasma performance and on the available operational space, it is important to identify the EF sources and develop EF control strategies. MAST Upgrade (MAST-U) is a spherical tokamak which will operate in the near future after a series of enhancements from the previous MAST experiment. To deliver a machine with EF amplitude low enough to allow a high quality experimental programme, systematic analysis of the intrinsic EF sources has been carried out for poloidal eld (P) and divertor (D) coils. To deliver a machine with EF amplitude low enough to allow a high quality experimental programme, systematic analysis of the intrinsic EF sources has been carried out for poloidal field (P) and divertor (D) coils, whose magnetic eld measurements were available when writing this paper. Such measurements reveal that P and D coils are 3D deformed and thus are responsible for intrinsic EFs, with mainly n=1 and n=2 toroidal mode numbers. In preparation to MAST-U operation, both passive and active EF control strategies have been adopted for n=1 and n=2 EFs compensation. Passive EF control consisted of installing finely each D and P coil within MAST-U device so as to minimize the intrinsic n=1 EF amplitude. The optimal coil alignment has been determined based on magnetic eld measurements and the corresponding 3D electro-magnetic modelling, and envisaged coil shift and tilt of 3.2 mm and 0.7 mrad, respectively, in the case of P coil named P4. Conversely, active control will be use during MAST-U operation to reduce the n=2 EF which is associated mainly with P4 and P5, as well. Since these coils have been re-used from the MAST device, studies attempting n=2 EF control, based on MAST plasmas, have been modelled utilizing the MARS-F code to interpret experimental results in MAST and to give hints for future n=2 EF control studies in MAST-U. Here, a model-based control set for n=2 EF control has been identified which would allow for the minimization of rotation braking, of the resonant magnetic field at the q=2 and of the plasma displacement in MAST-U, simultaneously.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Data on erosion and hydrogen fuel retention in Beryllium plasma-facing materials

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m 2 . Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.

Erosion–deposition↗

DOE OSTI2022

Toroidal magnetic field ripple in the presence of misaligned toroidal field coils on the COMPASS-U tokamak

Homogeneity of the toroidal magnetic field is a critical aspect of the tokamak fusion device. It is affected by the design of toroidal magnetic field coils and their manufacturing and assembly misalignments. The magnetic field of the COMPASS-U project is analysed using toroidal field ripple quantity. Here ANSYS modelling of bitter-plate-designed toroidal field coils is performed to prepare an accurate simplistic representation of the coils using a virtual current line. Toroidal field ripple is analysed using a Gaussian distribution of possible misalignments. The worst-case toroidal field ripple corresponding to the technical drawings only slightly exceeds the chosen limit of 0.5%.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Effect of radial pressure corrugations and profile shearing on turbulence in Fusion plasmas

Microturbulence can produce stationary fine-scale radial corrugations on the plasma density and temperature gradients in magnetic confinement fusion devices. We show that these structures play a significant role in regulating turbulent transport. We focus on the pedestal, studying electron-temperature-gradient (ETG) mode destabilisation and saturation in the presence of radial corrugations on the electron temperature gradient that could result from microtearing turbulence. A linear dispersion relation is derived for a shearless slab case, which indicates that in the presence of a sinusoidal background corrugation, each ETG mode splits into three distinct eigenvalues, with one being the original, one being more unstable and one being less unstable. However, despite the presence of more unstable linear modes, nonlinear gyrokinetic simulations of ETG with corrugated background electron temperature show a reduction of fluxes. Our investigation reveals a radial variation of the phase velocity of the modes that is proportional to the diamagnetic drift velocity and the local pressure gradient. The associated profile shearing breaks the turbulent eddies apart, reducing the transport level. This profile shearing resulting from fine-scale pressure corrugations could be a ubiquitous turbulence saturation mechanism not just in Fusion plasmas, but in Astrophysics and other areas.

FOS: Physical sciences↗

DOE OSTI2025

Electron Cyclotron Emission–Based Separatrix Identification in ITER with OMFIT Synthetic Modeling

Accurate determination of the separatrix location is essential for understanding edge plasma behavior and optimizing confinement in tokamaks, especially in next-generation devices such as ITER. In this study, a synthetic microwave diagnostics module was developed and implemented in the OMFIT framework to assess the feasibility of an electron cyclotron emission–based separatrix detection method in ITER plasmas. Simulations were carried out using ITER H-mode equilibrium scenarios with different plasma density profiles and different pedestal widths. Here, the results show that the electron emission temperature profiles consistently exhibit an inversion pattern near the edge, with a well-defined minimum point that could serve as a proxy for the separatrix location. However, unlike in the DIII-D, the minimum point in ITER is systematically offset by approximately 2 cm into the scrape-off layer, independent of density or pedestal width, which is within the radial resolution range (2 to 5 cm) determined by the 500-MHz channel spacing. While the method does not provide the exact separatrix location, it offers a reliable indicator of the boundary region and has potential applications for real-time boundary monitoring in ITER and other future fusion devices.

Electron cyclotron emission↗

DOE OSTI2024

Development of Integrated Education, Training and Research Program with Interdisciplinary Applications II

Plasma models have been formulated including realistic spatial profiles of both flow and radio frequency induced ponderomotive force. With these inclusions the picture of stability of various plasma and fluid instabilities, as expected, changed drastically with ground-breaking consequences. The inhomogeneous parallel flow and the radio frequency waves can actually shown to stabilize turbulence. This is different from the prevalent notion that both parallel flow shear and radio frequency waves are responsible for the excitation (destabilization) of plasma turbulence. This has several ground-breaking consequences:- (1) the stabilization by parallel flow clearly goes against the conventional notion of the origin of ionospheric oscillation which invokes parallel flow destabilization as the origin, (2) the stabilization by parallel flow opens us a new avenue for improved mode formation in fusion devices - which mostly rely on the perpendicular flow shear stabilization for improved mode formation but the perpendicular flow is damped in a tokamak - so the improved mode formed by the parallel flow can sustain longer and has more prospect for ignition, (3) the complete stabilization of the ITG mode (and consequent suppression of transport) not only explain many unknown phenomena in the space physics but it also raises a prospect for transport barrier formation by the RF waves but not by the RF induced flow (as most works suggest) which is never observed in a tokamak of that magnitude to create a barrier. These are indeed ground-breaking consequences.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Transport and confinement physics: Chapter 2 of the special issue: on the path to tokamak burning plasma operation

Progress in physics understanding and theoretical model development of plasma transport and confinement (TC) in the ITPA TC Topical Group since the publication of the ITER Physics Basis (IPB) document (Doyle et al 2007 Nucl. Fusion 47 S18) was summarized focusing on the contributions to ITER and burning plasma prediction and control. This paper provides a general and streamlined overview on the advances that were mainly led by the ITPA TC joint experiments and joint activities for the last 15 years (see JEX/JA table in appendix). This paper starts with the scientific strategy and scope of the ITPA TC Topical group and overall picture of the major progress, followed by the progress of each research field: particle transport, impurity transport, ion and electron thermal turbulent transport, momentum transport, impact of 3D magnetic fields on transport, confinement mode transitions, global confinement, and reduced transport modeling. Cross references with other Topical Groups are given in order to highlight overlapped topics, such as the 3D effect on the plasma transport in the edge and L-H transition physics. The increasing overlap between the topical groups is a reflection of the progress on integrating the known physics into comprehensive models that are better and better able to reproduce the plasma transport. In recent years, such integration has become increasingly prevalent when considering transport from the SOL, through the edge pedestal, and into the plasma core. In the near future, increased collaboration also with the magneto-hydrodynamic and energetic particles community will be important as we approach burning plasma conditions in next-step fusion devices. A summary of remaining challenges and next steps for each research field is given in the Summary section.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Driving rotamak currents with minimal power dissipation

The Rotamak is a proposed thermonuclear fusion device which employs rotating magnetic fields (RMF) to generate an azimuthal current to produce a field-reversed configuration. The efficiency of the currents that produce the field reversal by RMFs was debated some 40 years ago. The debate revolved around whether the currents would incur dissipation by the conventional Spitzer perpendicular resistivity, or whether some other relation between current and dissipation would be more appropriate. By employing an electron–ion pitch-angle scattering model, we find that the dissipation is non-Spitzer in nature. However, curiously, there appears to exist a regime where the power dissipated to maintain the current becomes vanishingly small.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Reduction in RF sheath rectification with insulating antenna enclosure walls

Abstract Radiofrequency (RF) sheath rectification is one of the most prominent deleterious effects associated with ion cyclotron range of frequencies (ICRF) heating in fusion plasmas. RF sheaths and associated effects, such as impurity generation and convective cell generation, need to be mitigated to ensure that ICRF is a viable option for heating in future fusion devices. Experiments were performed on the Large Plasma Device (LAPD) at UCLA to explore the effects of using electrically-insulating antenna enclosures on RF rectified sheaths. Three different enclosure side-wall materials were used, including copper, MACOR (electrically insulating), and MACOR over copper. In the case of the MACOR–copper side walls, the non-conductive MACOR material was exposed to the bulk plasma but a layer of copper was added below to allow for image currents to flow. All three of the experiments had similar plasma density, temperature, and background magnetic field. In the case of the copper enclosure, RF rectified potentials, many times the local electron temperature, and associated formation of convective cells were observed and reported Martin M. et al (2017 Phys. Rev. Lett. 119 205002). In the experiments with MACOR and MACOR–copper enclosures, RF rectification was significantly reduced. Additionally, these latter two experiments showed no evidence of convective cell formation. Although the results from the MACOR experiment are reminiscent of the results obtained in ASDEX-U with a three-strap antenna optimized to reduce image currents on the antenna limiters Bobkov V. et al (2016 Nucl. Fusion 56 084001), the MACOR–copper experiment seems to suggest that insulating plasma facing materials have at least an equally strong impact on reducing potential rectification.

Physics↗

DOE OSTI2022

Development of Integrated Education, Training and Research Program with Interdisciplinary Applications

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Design of microwave broadband CMOS transmitter and receiver circuits for MIR and ECEI plasma diagnostics

To efficiently determine the plasma electron density fluctuations using the MIR diagnostic technique, a 55–75 GHz 65 nm-CMOS transmitter has been developed where four separate intermediate frequency (IF) signals are up-converted, amplified, and then combined to generate an 8-tone RF output; a broadband 90 nm-CMOS receiver has also been constructed, which consists of an RF-low noise amplifier (LNA), mixer, and IF amplifier. The circuits and their corresponding modules will soon be deployed on the DIII-D and NSTX-U fusion devices. A 110–140 GHz 65 nm-CMOS receiver has also been designed, which is suitable for measuring the deep-core temperature fluctuations in the DIII-D tokamak using the electron cyclotron emission imaging diagnostic system. In addition to the RF-LNA/balun, mixer, and IF amplifier, an LO balun/tripler and driving amplifier are now included in this highly integrated circuit chip. By adopting the microwave and millimeterwave system-on-chip concept in the front-end system design, this paper demonstrates that compact transmitter and receiver modules can be easily built, which, in turn, facilitates array implementation and maintenance.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

First field test of a novel optical gas analyser in the exhaust of Wendelstein 7-X

A novel optical gas analyser, designed for isotope-resolved exhaust composition measurement, was field-tested at Wendelstein 7-X (W7-X) to validate its laboratory-proven concept under operational fusion experiment conditions. The system, Optix, comprises a cold cathode remote plasma generator and a high-resolution Fabry–Perot spectrometer and was deployed in the exhaust line of W7-X during the OP2.3 campaign. The injection of 3 He and 4 He for minority ion-cyclotron heating provided a test case for helium isotope discrimination. Despite limitations due to background gas and low partial pressures of the target species, isotope-resolved spectral signatures were successfully observed, demonstrating the fundamental viability of the Optix approach. Additionally, the spectrometer was evaluated for plasma emission measurements from both core and edge sightlines. While helium line emission was detectable, interpretation was hindered by complex background signals, highlighting the benefits of controlled remote plasma generators for spectroscopy. This first deployment provides critical insight into pressure requirements, spectral resolution, and operational constraints, informing future applications of optical exhaust diagnostics in fusion devices.

magnetic confinement fusion↗

DOE OSTI2025

Amorphization and siliconization of silicon carbide as a first wall material

The understanding and prediction of silicon carbide (SiC) material evolution exposed to SOL plasma conditions is of prime interest because SiC represents a promising main chamber wall plasma-facing material for next-step fusion devices (low hydrogenic diffusion, good mechanical and thermal properties under neutron irradiation). Gross and net Si erosion rates from SiC surfaces in contact with a well-diagnosed L-mode plasma in the DIII-D tokamak have been simulated and the surface concentrations of impurities have been tracked as a function of time. Coupled simulation of surface model and impurity transport demonstrates amorphization of crystalline SiC exposed to L-mode plasma due to the accumulation of displacement damages under ion irradiation. This affects the lifetime of SiC plasma facing components. Surface evolution is tightly coupled to impurity transport in the plasma and therefore needs to be integrated with impurity transport simulations to effectively predict Si erosion rates and sub-surface concentrations as a function of time. The simulation workflow couples a semi-analytical surface model to the impurity transport code GITR. The surface model is a homogeneous mixed-material model that tracks physical & chemical sputtering and reflection of impurities. Gross erosion is primarily influenced by the background plasma parameters and redeposition patterns are mainly influenced by the prompt redeposition due to the gyro-orbits of impurity ions. Although crystalline form of SiC is preferable for fusion wall applications because of resistance to neutron irradiation, this work indicates that crystalline SiC will undergo amorphization under D plasma contact with implications of higher sputtering and fuel retention. These results direct us to explore the effects of amorphization on crystalline SiC and further the physics basis of SiC usage as first wall material for fusion environments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Kinetic Plasma Simulation Capabilities in the MOOSE Framework: Verification of Particle-Particle Collisions

High-fidelity simulations of complex plasma systems allow researchers to gain key insights into and understanding of these systems. To facilitate massively parallel high-fidelity plasma simulations, finite-element-based particle-in-cell capabilities are being developed within the open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) based framework called Software for Advanced Large-scale Analysis of MAgnetic confinement for Numerical Design, Engineering & Research (SALAMANDER). While SALAMANDER’s primary objective is modeling edge plasmas and plasma-facing components in fusion devices, the particle-in-cell capabilities being developed are general and will support modeling low-temperature plasmas as well. Previously, collisionless magnetostatic simulation capabilities have been verified with the two-stream and Dorey-Guest-Harris instabilities, and single particle motion. Collisions were implemented using the direct simulation Monte Carlo method, and verification of this capability will be presented here several verification problems: relaxation of a randomly initialized gas to a Maxwellian distribution, Fourier heat flow, and comparison of reaction rates to both analytic calculations and those calculated using a multi-term Boltzmann solver.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Implementation of a High-Speed Multichannel Data Acquisition System for Magnetic Diagnostics and Plasma Centroid Position Control in ISTTOK

In tokamak and other fusion devices, magnetic control is the main tool that allows to regulate the plasma current, position and shape; it is in charge of actuating the desired plasma current waveform, steering the plasma position to a given set point and maintain the plasma shape close to a prescribed plasma equilibrium. This work describes the application of several physics concepts and computational tools in order to obtain a novel optimal controller for the plasma centroid position, which has been implemented and tested in the real-time plasma control system at the ISTTOK tokamak. A key point for the development of the new control system was the installation of a recently upgraded hardware, that numerically integrates in real-time the magnetic probes signals.

poloidal field coils↗