Tokamak Operation Research Articles

Methods of theoretical calculations and of experimental researches of the system atomic hydrogen – metal

All the main directions of energy development suggest or already implement the use of hydrogen. In addition, the interaction of low-energy hydrogen atoms with metals is also of considerable interest, both from the point of view of fundamental research and in connection with the operation of large tokamaks and thermonuclear reactors. The paper presents a literature review of the features of the interaction of hydrogen with metals.It is shown that metal-hydrogen reactions, which lead to the formation of metal hydrides, are considered as a special type of such interaction.Modern methods of experimental study of heterogeneous reactions, topochemistry of metal - hydrogen reactions, dependences of the rate of interaction on pressure and temperature are considered, models of surface processes occurring during the interaction of hydrogen with a metal are discussed.A kinetic method for studying the mechanism of interaction of atomic hydrogen with hydride-forming metals is proposed.

Upgrade of the cathode HVPS system for 4.6 GHz LHCD on EAST using high speed PSM

The 4.6GHz LHCD system is an important part of the auxiliary heating system on EAST tokamak. The cathode high voltage power supply(HVPS) is based on pulse step modulation (PSM) technology, in which 64 modules are used to output 50kVDC in series. The modulation frequency of single PSM module is 50Hz, it switch on and off every 20ms, so the regulation speed of system is limited. In the face of the interference caused by the voltage fluctuation of the grid side during the tokamak operation, the HVPS cannot make more rapid response. It means that a large fluctuation of the output voltage will be produced and output performance of microwave will be affected. In order to improve the regulation speed and anti-interference capability of system, the high frequency PSM module with 1kHz modulation frequency capability was designed to replace some of the original low frequency modules. The fluctuation of output voltage was reduced by the rapid regulation capability of the high frequency module. The experimental results show that the output voltage fluctuation rate of the upgraded power supply is reduced by 50%, which better meets the control requirements of klystron for voltage accuracy and stability, and the reliability of the system is guaranteed.

First temperature database achieved with Fiber Bragg Grating sensors in uncooled plasma facing components of the WEST lower divertor

Plasma Facing Components (PFCs) temperature measurement is required to ensure safe high power for long pulse tokamak operation and for physics studies. A set of twenty thermocouples (TCs) and four optical fiber temperature sensing probes, each of them including eleven wavelength-multiplexed fiber Bragg gratings (FBGs) written with UV radiation, have been integrated and deployed in the WEST lower divertor. To avoid any thermal erasure occurring above 400 °C, the gratings have been regenerated using a high temperature annealing process. The FBG probes are embedded in tungsten-coated graphite components with no active cooling. The diagnostic was working correctly from the first plasma breakdown achieved in WEST (in 2017) until the first dismantling of the lower divertor three years later (2020). Collected data are sorted by temperature range to investigate the ageing of both PFCs and diagnostics. Measurements obtained with FBGs and TCs are consistent, no long-term drift is reported during the overall period of use. The maximum temperature reported by FBGs in the divertor is 830 °C obtained with 90 MJ of cumulated plasma energy in one pulse. A short but reversible collapse of the Bragg peak (from 5 down to 2 dB) has been observed under high temperature due to the strong temperature gradients measured along the fiber (about 20 °C/mm, for an FBG size of 3 mm). This feedback demonstrates that regenerated FBGs are suitable for long-term temperature monitoring in tokamak environment with high heat load, high temperature and gradient, strong magnetic field and vacuum constraint.

Progress of Divertor Heat and Particle Flux Control in EAST for Advanced Steady-State Operation in the Last 10 Years

Active control of the excessively high heat and particle fluxes on the divertor target plates is of fundamental importance to the steady state operation of tokamaks, especially for fusion reactors. A series of experiments have been carried out on this critical issue to relieve the plasma-wall interactions in the experimental advanced superconducting tokamak (EAST) in the last ten years, not only contributing to the long pulse operation of EAST itself, but also providing physical understandings and potential techniques to the next-generation devices like ITER. We have characterized the power deposition pattern and broadened the divertor footprint width effectively. The plasma-wetted area is actively handled using either 3-dimentional edge magnetic topology or advanced plasma equilibrium, thereby peak heat flux around the strike point is reduced. Active control of detachment or radiation compatible with core plasma performance has progressed significantly in very recent years, with a series of active feedback control modules developed and utilized successfully, based on the divertor physics advances with both experiments and simulation. The upper divertor of EAST was upgraded from graphite to active water-cooling ITER-like tungsten in 2014, exhibiting much enhanced heat removal capability. As for the particle exhaust including both fueling and impurity particles, in addition to wall conditioning and impurity source control, the efficiency of particle flux exhaust is optimized by making full use of the divertor closure and the plasma drifts in both scrape-off layer and divertor volume. These heat and particle exhaust advances contribute greatly to a series of EAST achievements like H-mode operation over 100 s. A brief near-term plan on the integrated control of divertor plasma-wall interactions in long-time scale will also be introduced, aiming to provide favorable divertor operation solution for ITER and CFETR.

Advances in physics understanding of high poloidal beta regime toward steady-state operation of CFETR

Experimental and modeling investigations of high βp scenarios on DIII-D and EAST tokamaks show advantages in high energy confinement, avoidance of n = 1 MHD, and core-edge integration with reduced heat flux, making this scenario an attractive option for China Fusion Engineering Test Reactor steady-state operation. Experiments show that plasmas with high confinement and high density can be achieved with neutral beam injection on DIII-D (βp ∼ 2.2, βN ∼ 3.5, fBS ∼ 50%, fGw ∼ 1.0, and H98y2 ∼ 1.5) and pure RF power on EAST (βP ∼ 2.0, βN ∼ 1.6, fBS ∼ 50%, fGw ∼ 0.8, and H98y2 > 1.3). By tailoring the current density profile, a q-profile with local (off-axis) negative shear is achieved, which yields improved confinement and MHD stability. Transport analysis and simulation suggest that the combination of a high density gradient and high Shafranov shift allows turbulence stabilization and higher confinement. Using on-axis Electron Cyclotron Heating injection, tungsten accumulation is avoided on EAST, and this is reproduced in modeling. Reduced heat flux (by > 40%) and maintenance of high core confinement is achieved with active feedback control of the radiated divertor, an important result for long pulse operation in tokamaks. The improved physics understanding and validated modeling tools are used to design a 1 GW steady-state scenario for CFETR.

Measuring the electron temperature and identifying plasma detachment using machine learning and spectroscopy.

A machine learning approach has been implemented to measure the electron temperature directly from the emission spectra of a tokamak plasma. This approach utilized a neural network (NN) trained on a dataset of 1865 time slices from operation of the DIII-D tokamak using extreme ultraviolet/vacuum ultraviolet emission spectroscopy matched with high-accuracy divertor Thomson scattering measurements of the electron temperature, Te. This NN is shown to be particularly good at predicting Te at low temperatures (Te < 10 eV) where the NN demonstrated a mean average error of less than 1eV. Trained to detect plasma detachment in the tokamak divertor, a NN classifier was able to correctly identify detached states (Te < 5 eV) with a 99% accuracy (an F1 score of 0.96) at an acquisition rate 10× faster than the Thomson scattering measurement. The performance of the model is understood by examining a set of 4800 theoretical spectra generated using collisional radiative modeling that was also used to predict the performance of a low-cost spectrometer viewing nitrogen emission in the visible wavelengths. These results provide a proof-of-principle that low-cost spectrometers leveraged with machine learning can be used to boost the performance of more expensive diagnostics on fusion devices and be used independently as a fast and accurate Te measurement and detachment classifier.

Analysis of a Possibility to Use Parallel Non-Twisted Stacks of HTS Tapes as Cable in High Current Conductor of Tokamak Toroidal Field Coils

Development of high current cable-in-conduit conductors (CICC) based on HTS tapes aimed for fusion magnets has started recently. The majority of the suggested CICC use as a HTS cable either, twisted stacks of HTS tapes (stacked technology), or some layers of tapes wrapped around a cylindrical core (CORC technology). An idea to use parallel non-twisted stacks of HTS tapes in a cable design is received skeptically by the fusion community, although such a cable design for tokamak toroidal field (TF) coils has some advantages over twisted ones. The twofold benefits is determined by parallel (deviation- θ <; 5°) orientation of toroidal field vector to surface of HTS tape and mutual orientation of background field and operating current vectors along conductor. Due to such orientation the critical current of tape-Ic(B;T) is considerably (in 5-6 times) higher, due to an anisotropy of HTS tape electro-physical properties. In addition a compressive force is oriented in the most beneficial direction. The main arguments against the non-twisted stacks in the cable design is a concern that the favorable condition described above are not the same in every location of TF coil, or that they will not be favorable in all modes of tokamak operation and the time of TF magnet charging is unreasonably long because of the screening current. The analysis carried out shows that the negative effect of screening currents decreases significantly due to a wavelike magnetic field distribution along conductor and that suggested design of HTS CICC is expected to be successful at any tokamak operation scenario.

Mathematical modeling using batch scheduling approach for optimal tritium inventory in the ISS of the ITER fuel cycle

An optimal model to schedule fuel regeneration is developed for the Tokamak Cryopumps and Cryogenic Viscous Compressors (CVCs) of the fuel cycle of ITER. The objective is to minimize the tritium inventory inside the Isotope Separation Systems (ISS) while supplying the tokamak’s required amount of fuel considering fueling stages: ramp-up, flat-top, and ramp-down for each tokamak operation mode (Inductive, Hybrid, and Non-Inductive). Case studies of CVC regeneration times are presented to illustrate how CVC regeneration affects the ISS tritium working inventory. Results showed that nine iterations of CVC regeneration are enough to maintain ISS working tritium inventory management during Non-Inductive tokamak operation mode. The developed model uses mixed integer linear programming of the state-task network.

Plasma confinement mode classification using a sequence-to-sequence neural network with attention

In a typical fusion experiment, the plasma can have several possible confinement modes. At the tokamak à configuration variable, aside from the low (L) and high (H) confinement modes, an additional mode, dithering (D), is frequently observed. Developing methods that automatically detect these modes is considered to be important for future tokamak operation. Previous work (Matos et al 2020 Nucl. Fusion 60 036022) with deep learning methods, particularly convolutional long short-term memory networks (conv-LSTMs), indicates that they are a suitable approach. Nevertheless, those models are sensitive to noise in the temporal alignment of labels, and that model in particular is limited to making individual decisions taking into account only the input data at a given timestep and the past data, represented in its hidden state. In this work, we propose an architecture for a sequence-to-sequence neural network model with attention which solves both of those issues. Using a carefully calibrated dataset, we compare the performance of a conv-LSTM with that of our proposed sequence-to-sequence model, and show two results: one, that the conv-LSTM can be improved upon with new data; two, that the sequence-to-sequence model can improve the results even further, achieving excellent scores on both train and test data.

Active control of electron cyclotron emission radiometer channel frequencies for improved electron temperature measurements.

As advanced scenarios are developed for tokamak operations, the demand for flexibility of the electron cyclotron emission (ECE) channels' locations has increased. The tunable feature of yttrium iron garnet (YIG) filters provides this spatial flexibility. Here, we present a method of performing ECE measurements on fixed flux surfaces instead of fixed frequencies. This is achieved by adjusting YIG filters utilized in the intermediate frequency section to frequencies associated with flux surfaces in regions of interest during the discharge. The key components are the application of tunable YIG filters and a control program that calculates the filter settings using flux information from real-time reconstruction equilibria (EFIT). This fast procedure facilitates Te measurements in regions of interest to investigate plasma dynamic behaviors.

First measurements of aluminum erosion by neutral particles on EAST tokamak

To evaluate the erosion effects of neutral particles on the first wall materials during tokamak operation, aluminum (Al) coated samples were exposed to a sequence of plasma discharges in 2018 and 2019 EAST campaigns. In order to avoid the incidence of charged particles, the samples were installed in a molybdenum (Mo) cylindrical sleeve and positioned far from the main plasma. The erosion thicknesses of Al layer were measured by Rutherford Backscattering Spectrometry (RBS) before and after the plasma exposure. The average net erosion rates of the samples in the two campaigns are around 1.18 × 1013 atoms cm−2 s−1 and 4.64 × 1013 atoms cm−2 s−1, respectively. The different erosion rate may mainly result from different plasma density and heating power. The theoretical calculated erosion rates using the neutral distribution under different density conditions are in fair agreement with the experimental results.

SHS Synthesis, SPS Densification and Mechanical Properties of Nanometric Tungsten

Recent studies have shown that low grain sizes are favorable to improve ductility and machinability in tungsten, as well as a resistance to ablation and spallation, which are key properties for the use of this material in a thermonuclear fusion environment (Tokamaks such as ITER). However, as one of the possible incidents during Tokamak operation is the leakage of air or water from the cooling system inside the chamber, resulting in the so-called loss of vacuum accident (LOVA), extensive oxidation may arise on tungsten components, and the use of an alloy with improved oxidation resistance is therefore highly desirable. As current production routes are not suitable for the fabrication of bulk nanostructured tungsten or tungsten alloys samples, we have proposed a new methodology based on powder metallurgy, including the powder synthesis, the densification procedure, and preliminary mechanical testing, which was successfully applied to pure tungsten. A similar study is hereby presented on tungsten-chromium alloys with up to 6 wt.% Cr. Results show that full tungsten densification may be obtained by SPS at a temperature lower than 1600 °C. The resulting morphology strongly depends on the amount of the alloying element, presenting a possible second phase of chromium oxide, but always keeps a partial nanostructure inherited from the synthesized powders. Such microstructure had previously been identified as being favorable to the use of these materials in fusion environments and for improved mechanical properties, including hardness, yield strength and ductility, all of which is confirmed by the present study.

Non-evaporable getter pump operations in the TCV tokamak

A non-evaporable getter pump using the ZAO alloy was installed on TCV and operated during the 2019 experimental campaign. The pump performance, determined from in-situ pumping speed measurements, indicates that current getter pump technologies are well suited to tokamak operations with high power exhaust plasmas, and provide an interesting alternative for applications in fusion experiments.

Real-time feedback control system for ADITYA-U horizontal plasma position stabilisation

The ADITYA-U tokamak (R0 = 0.75 m, a = 0.25 m) is designed to shape plasma column in both single and double null diverter configurations. It is quite well known that sustaining a shaped plasma in tokamak requires very good plasma column position control, both horizontal and vertical. An FPGA-based proportional–integral–derivative (PID) control system has been designed and operated to achieve horizontal plasma position control in ADITYA-U tokamak. The complete system has been rigorously tested with sample signals before implementing to the ADITYA-U plasma discharges. The control system is integrated and time-synchronized with the plasma discharge operation of ADITYA –U. Furthermore, the system has been trained to take appropriate actions during the disruption or plasma failure in the tokamak operation. Detailed experimental results have been obtained by the operation of the digital PID controller. The complete design, installation, operation, tuning of the system along with all the relevant testing and operating experience of the digital PID controller for real-time horizontal plasma position control is presented in the paper.

On Monitoring Tearing Modes Stability in Toroidally Rotating Tokamak Equilibria

In tokamak operation, the control of dangerous MHD instabilities, possibly in real-time scenarios, must rely on prompt robust diagnostics of the state and stability of the system. The set of magnetic signals measured on the outside of the plasma boundary, based on the Zakharov–Shafranov, Shkarowsky, Wootton (ZSSW) current moments, has been always used for reliable monitoring of key characteristics of the instantaneous equilibrium condition, such as the quantities $\Delta _{\mathrm{ Sh}}$ , the Shafranov centroid shift, $\beta _{p} $ related to the thermal energy content, and $l_{i}$ related to the current profile peakedness. In addition, the fast pickup coils monitor the external magnetic field fluctuations due to internal MHD activity, however, without the possibility of radial localization of the source. Here, we explore the potential usefulness of more complete use of ZSSW moments in association with the information from fast B perturbation signals to detect tearing stability conditions. For clarity, we set up an analysis of the measurable response to tearing perturbations based on an exact equilibrium model, which is an extension of the Solov’ev case with the addition of an equilibrium, nonuniform plasma rotation $\Omega (\psi)$ . The relation of the selected (externally measurable) ZSSW moments to the calculated stability index is mapped for different rotation values. The footprint of the stability condition $\Delta ' on some current moments on the outer surface can then identify stability boundaries for different rotation conditions. This first discussion on an idealized exact model is proposed for testing the concept for application to realistic equilibria since it relies on a few externally monitorable quantities and very basic assumptions on the tearing modes physics.

Design study and preliminary feasibility analysis of divertor with ITER-Like PFU for advanced tokamak

Abstract The divertor is one of the key plasma facing components (PFCs) in fusion device and shall exhaust high heat flux coming from the plasma. The paper presents an engineering design solution accommodating the physical requirements and the removal of 10 MW/m2 steady heat flux. The proposed divertor design employs reliable PFCs structure which can be manufactured by proven mature technologies. There are effective measures to ensure assembly and installation accuracy in this design concept. The convenient maintenance is guaranteed by small modular design. Preliminary feasibility analysis has been carried out with respect to thermal and hydraulic, which verifies that the design successfully meets the requirements for the plasma operation of advanced tokamak.

Analyses of deuterium retention in tungsten and graphite first wall materials by laser-induced ablation spectroscopy on EAST

• Combined LIBS and LIAS measurements were applied to determine the dynamic deuterium retention in C and W during EAST discharges. • LIAS measurements show more than three times higher deuterium retention in graphite than in tungsten during EAST discharges. • The deuterium retention on the graphite PFC is saturated within 0.2 s plasma exposure for limited magnetic configuration when the plasma actively touches the first wall. • The deuterium retention in the start-up phase is significantly larger than during the discharge for diverted magnetic configuration. Plasma-wall interactions at the first wall in fusion experimental devices are critical for the life time of plasma-facing components and tokamak operation. In-situ measurements of these plasma-wall interaction processes are required in long pulse devices in order to understand better the underlying mechanisms as well as monitoring of the tritium content of the first wall as an immediate safety application. In this work we present measurements of the deuterium retention in the first wall of EAST using laser-induced ablation spectroscopy. The diagnostic is applied during plasma operation and the dynamic retention for graphite and tungsten as plasma-facing materials are compared, showing more than three times higher short-term retention in graphite than in tungsten. Laser-based analysis is applied during a shift of the vertical positions of the plasma column in limited as well as in the start-up phase of discharges in diverted magnetic configuration. Plasma core densities of n e = (1.5–3.2) × 10 19 /m 3 are chosen to vary the ion flux to the first wall and study saturation effects of deuterium on the surface first wall materials during tokamak plasma operation. A linear dependence of the retention from ion fluence is observed for graphite as PFC up to 3 × 10 18 D + /cm 2 , followed by a constant retention up to fluences of 3.5 × 10 19 D + /cm 2 for nearby first wall temperatures of T PFC ≈ 350 K.

Stability nature of ohmic discharges with hollow current profiles on the HL-2A tokamak

The equilibrium configurations with hollow current density profiles can effectively improve plasma confinement. The stability of the ohmic discharges with hollow current density profiles on the HL-2A tokamak was studied for various parameters, such as plasma poloidal beta βp, the axial and edge safety factors q0 and qs, as well as the minimum safety factor . The results suggest that the restriction to the parameters of discharges with hollow current density profiles is stricter than that of conventional discharges with peaked current density profiles. In the case of q profile with reverse shear, the instability of the configuration is significantly improved with higher βp. As a result, the configuration with a hollow toroidal current profile is stable for high βp operations. Only if q0, qs and satisfy the specific limitations, i.e. on HL-2A, only if both and hold, together with and , may the configurations with the reverse shears be stable for hollow-current-type Ohmic discharge. The results may give hints for the role of the discharges with hollow toroidal current densities in the coming advanced tokamak operations.

An overview of the numerical methods for tokamak plasma equilibrium computation implemented in the NICE code

The code NICE (Newton direct and Inverse Computation for Equilibrium) enables to solve numerically several problems of plasma free-boundary equilibrium computations in a tokamak: plasma free-boundary only reconstruction and magnetic measurements interpolation, full free-boundary equilibrium reconstruction from magnetic measurements and possibly internal measurements (interferometry, classical linear approximation polarimetry or Stokes model polarimetry, Motional Stark Effect and pressure), direct and inverse, static and quasi-static free-boundary equilibrium computations.NICE unifies and upgrades 3 former codes VacTH [1], EQUINOX [2] and CEDRES++ [3]. The strength of NICE is to gather in a single finite element framework different equilibrium computation modes. It makes intensive use of Newton method and Sequential Quadratic Programming method to solve non linear problems.NICE is used routinely for WEST tokamak operation. It is also adapted to the IMAS (ITER Modelling and Analysis Suite) format which makes it usable on many different fusion tokamak reactors.In this document we give a general overview of the numerical methods implemented in NICE as well as a number of computation examples.

Investigation of turbulent transport regimes in the tokamak edge by using two-fluid simulations

The results of flux-driven, two-fluid simulations in single-null configurations are used to investigate the processes determining the turbulent transport in the tokamak edge. Three turbulent transport regimes are identified: (i) a developed transport regime with turbulence driven by an interchange instability, which shares a number of features with the standard L-mode of tokamak operation; (ii) a suppressed transport regime, characterized by a higher value of the energy confinement time, low-amplitude relative fluctuations driven by a Kelvin–Helmholtz instability, a strong $\boldsymbol {E}\times \boldsymbol {B}$ sheared flow and the formation of a transport barrier, which recalls the H-mode; and (iii) a degraded confinement regime, characterized by a catastrophically large interchange-driven turbulent transport, which recalls the crossing of the Greenwald density limit. We derive an analytical expression of the pressure gradient length in the three regimes. The transition from the developed transport regime to the suppressed transport regime is obtained by increasing the heat source or decreasing the collisionality and vice versa for the transition from the developed transport regime to the degraded confinement regime. An analytical expression of the power threshold to access the suppressed transport regime, linked to the power threshold for H-mode access, as well as the maximum density achievable before entering the degraded confinement regime, related to the Greenwald density, are also derived. The experimental dependencies of the power threshold for H-mode access on density, tokamak major radius and isotope mass are retrieved. The analytical estimate of the density limit contains the correct dependence on the plasma current and on the tokamak minor radius.