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The use of spin polarized fuel could increase the deuterium-tritium (D-T) fusion cross section by a factor of 1.5 and, owing to alpha heating, increase the fusion power by an even larger factor. Issues associated with the use of polarized fuel in a reactor are identified. Theoretically, nuclei remain polarized in a hot fusion plasma. The similarity between the Lorentz force law and the Bloch equations suggests polarization can be preserved despite the rich electromagnetic spectrum present in a magnetic fusion device. The most important depolarization mechanisms can be tested in existing devices. The use of polarized deuterium and ³He in an experiment avoids the complexities of handling tritium, while encompassing the same nuclear reaction spin-physics, making it a useful proxy to study issues associated with full D-T implementation. ³He fuel with 65% polarization can be prepared by permeating optically-pumped ³He into a shell pellet. Dynamically polarized 7Li-D pellets can achieve 70% vector polarization for the deuterium. Cryogenically-frozen pellets can be injected into fusion facilities by special injectors that minimize depolarizing field gradients. Alternatively, polarized nuclei could be injected as a neutral beam. Once injected, the lifetime of the polarized fuel is monitored through measurements of escaping charged fusion products. Multiple experimental scenarios to measure the polarization lifetime in the DIII-D tokamak and other magnetic-confinement facilities are discussed, followed by outstanding issues that warrant further study. [ABSTRACT FROM AUTHOR]

The use of spin polarized fuel could increase the deuterium-tritium (D-T) fusion cross section by a factor of 1.5 and, owing to alpha heating, increase the fusion power by an even larger factor. Issues associated with the use of polarized fuel in a reactor are identified. Theoretically, nuclei remain polarized in a hot fusion plasma. The similarity between the Lorentz force law and the Bloch equations suggests polarization can be preserved despite the rich electromagnetic spectrum present in a magnetic fusion device. The most important depolarization mechanisms can be tested in existing devices. The use of polarized deuterium and 3He in an experiment avoids the complexities of handling tritium, while encompassing the same nuclear reaction spin-physics, making it a useful proxy to study issues associated with full D-T implementation. 3He fuel with 65% polarization can be prepared by permeating optically-pumped 3He into a shell pellet. Dynamically polarized 7Li-D pellets can achieve 70% vector polarization for the deuterium. Cryogenically-frozen pellets can be injected into fusion facilities by special injectors that minimize depolarizing field gradients. Alternatively, polarized nuclei could be injected as a neutral beam. Once injected, the lifetime of the polarized fuel is monitored through measurements of escaping charged fusion products. Multiple experimental scenarios to measure the polarization lifetime in the DIII-D tokamak and other magnetic-confinement facilities are discussed, followed by outstanding issues that warrant further study.

The next generation of large tokamaks, including ITER, will be equipped with a disruption mitigation system (DMS) that can be activated if a disruption is deemed to be imminent. Introducing impurities by pellet (large or shattered) or massive gas injection has been shown to be an effective mitigation mechanism on many tokamaks. The goal of the mitigation is to lessen the thermal and electromagnetic loads from the disruption without generating enough high-energy (runaway) electrons to damage the device. Variations of this mitigation process with impurity injection are presently being tested on many experiments. We have modeled one such impurity injection experiment on DIII-D using the M3D-C1 nonlinear 3D extended MHD code (Jardin et al 2012 Comput. Sci. Discovery 6 014002), The model includes an argon large pellet injection and ablation model, impurity ionization, recombination, and radiation, and runaway electron formation and subsequent evolution, including both Dreicer and avalanche sources. We obtain reasonable agreement with the experimental results for the timescale of the thermal and current quench and for the magnitude of the runaway electron plateau formed during the mitigation. This is the first 3D full MHD simulation with pellets and REs to simulate the disruption process and it also provides a partial validation of the M3D-C1 DMS model.

A BOUT ++ three-field magnetohydrodynamic model is employed to study the triggering and evolution of edge localized mode (ELM) by Li pellets injected along the outer mid-plane in the EAST configuration. The linear simulation shows that compared with a large deposition on the pedestal top (scenario I), a smaller deposition within the steep-gradient pedestal region (scenario II) can stimulate much larger linear growth rates of all-n peeling-ballooning modes (PBMs). The nonlinear simulation shows that there exists a pellet size threshold for ELM triggering for two deposition locations; the threshold for scenario I predicted in the present study matches the EAST observation well. Comparison of the two scenarios reveals that a smaller deposition is sufficient to trigger an ELM in a much shorter time in scenario II, whose ELM size is comparable to that in scenario I. This conclusion confirms previous DIII-D and ASDEX-Upgrade observations, suggesting that the steep-gradient pedestal region is a favorable deposition location for ELM triggering with minimum pellet size. Simulation analyses also find that the positive radial gradient of the hump-like pressure profile in the outer mid-plane induced by the pellet deposition plays a different role in the two scenarios. In scenario I, the force resulting from the gradient hinders the outflow of core plasmas and in return, the perturbation is suppressed from spreading inwards after ELM crashes. In scenario II, with a sizable deposition, the gradient results in another competitive perturbation growth region during the linear phase, thus dispersing the free energy and reducing the efficiency of destabilizing PBMs by pellet injection. The suppressing effect of saturated zonal flow on other modes, the short ELM fast crash phase, and the restricting transport effect of the positive radial pressure gradient work together to constrain the pedestal energy loss, especially when the pellet deposition amount is high.

For a commercial laser inertial fusion energy reactor, a highly repetitive operation in which many fuel pellets must be illuminated by laser spots. One of the most efficient ways to achieve this is to control the laser pointing by following fluctuations of the target position. The paper shows the precise control of laser pointing in the horizontal direction with a repetition rate of 10 Hz. Free-falling test pellets of 1 mm in diameter have been illuminated by a laser of 1.6 mm in diameter with cancellation of horizontal fluctuation over 4 mm. The difference in centroids between the laser spot and the illuminated test pellets is 86 μ m (standard deviation). This corresponds to a 92% engagement within a 0.15 mm difference, which is a condition for successful nuclear fusions at the Hamamatsu facility. This is the proof-of-principle demonstration of the target-supply tracking and homing laser at a repetition rate of 10 Hz for the actualization of a commercial reactor.

• A DAC system has been developed using the NI-PXIe hardware and LabVIEW application for pellet injection system. • Functions like pellet freezing, launching and data acquisition are performed from a single controller and GUI. • The GUI allows switching between slow and fast data acquisition modes with a single click. • Pellet freezing and launching experiments successfully conducted using it. • The system will be integrated with the tokamak for pellet-plasma interaction study. A data acquisition and control (DAC) system has been developed for the solid hydrogen pellet injection system. This injector is a gas gun type injector, where solid hydrogen pellet ice is formed using a closed cycle cryocooler, and high-pressure helium gas or a pneumatic punch is used to dislodge the pellet. The DAC system is based on National Instrument's embedded controller NI PXIe-8133, along with associated I/O DAC cards and the LabVIEW application. A graphical user interface (GUI) developed using LabVIEW software allows users to remotely control the pellet formation and injection process. The data is stored locally on a desktop computer or in cloud storage for further analysis. The images of the injected pellet were obtained by using a Phantom V-1210 high-speed camera at 100-Killo FPS, which reveals the size and speed of the pellet. The developed DAC system provides the flexibility needed to operate the injector remotely during the plasma discharge in a tokamak environment. The injector setup system has been successfully tested in a test bench operation, and it will be integrated with the ADITYA-U tokamak. [ABSTRACT FROM AUTHOR]

As a result of self-consistent simulations using the ASTRA + SOLPS and FC-FNS codes, the range of parameters was found, within which the required fraction of tritium in the core plasma, = 0.5, can be maintained by the injection of pellets with different isotope composition. For the DEMO-FNS tokamak, in the operating window of parameters, the interdependence of the fuel component fluxes used for plasma fueling is analyzed. The effects of the pellet injection from the high-magnetic-field and low-magnetic-field sides, used for the core plasma fueling and stimulating ELMs, on the fluxes in the DEMO-FNS fuel cycle systems are considered. It is shown that in the regimes with driven, convective ELMs, it is necessary to increase by one order of magnitude the capacities of the systems for pellet injection and separation of hydrogen isotopes, as compared to the estimates made previously without taking ELMs into account. The dependence of the frequency of the fuel pellet injection from the high-field side on their size and the particle confinement time in the plasma is found. It is shown that in the operating window of parameters found, it is necessary to inject pellets from the low-field side to stimulate ELMs. It is demonstrated that the additional stimulation of ELMs affects the requirements for the core plasma fueling. [ABSTRACT FROM AUTHOR]

Pellet injection is regarded as the only realistic actuator for core density control in future reactors such as ITER and DEMO. However, a control strategy that can reliably regulate the plasma close to operational limits using multiple pellet injectors is not yet available. In this paper, we present the first integrated model control simulations where a dedicated model-predictive controller is included in JINTRAC. We show that, when continuous actuators are considered, a simple transport model with a steady-state disturbance rejection paradigm is capable of capturing the particle transport dynamics for multiple transport models and scenarios. This in turn allows the model-predictive controller to deal with the uncertainty and minimize the control error given the limited actuation space. Furthermore, we show that for ITER and DEMO relevant pellet sizes, the discrete, nonlinear dynamics of pellet injection will limit the control performance and jeopardize the constraints if not accounted for by the controller. Hence, we conclude that for high-performance control on future reactors, controllers will have to be developed that explicitly deal with the discrete pellet dynamics.

By combining the X-mode polarized lower and upper cut-off reflections obtained from the density profile reflectometer, we have successfully attained a comprehensive density profile spanning from the edge to the core region in pellet injection (PI) experiments on EAST. During the homogenization process after PI on EAST, an innovative method was introduced to quantify the local density increase. This approach employed the distinctive ‘dual-reflection’ phenomenon observed in the EAST microwave reflectometer, encompassing measurements of both the X-mode lower and upper cut-off frequencies. Furthermore, experimental investigations were carried out on EAST to comprehensively explore the parallel and poloidal expansion of the high-density pellet cloud. Notably, this study marks the first instance of measuring expansion velocities of pellet materials in both parallel and poloidal directions on EAST. A comparative analysis was performed initially between these experimental measurements and simulation results obtained from the HPI2 code, marking a pivotal stride towards enhancing its applicability in EAST.

Single carbon pellet disruption mitigation simulations using M3D- C 1 were conducted for an NSTX-U-like plasma to support the electromagnetic pellet injection concept. A carbon ablation model has been implemented in M3D- C 1 and tested with available data. 2D simulations were conducted in order to estimate the amount of carbon needed to quench the plasma, finding that the content in a 1 mm radius vitreous carbon pellet (∼3.2 Ă— 1020 atoms) would be enough if it is entirely ablated. 3D simulations were performed, scanning over pellet velocity and parallel thermal conductivity, as well as different injection directions and pellet concepts (solid pellets and shell pellets). The sensitivity of the thermal quench and other related quantities to these parameters has been evaluated. A 1 mm radius solid pellet only partially ablates at velocities of 300 m sâˆ'1 or higher, thus being unable to fully quench the plasma. To further enhance the ablation, approximations to an array of pellets and the shell pellet concept were also explored. 3D field line stochastization plays an important role in both quenching the center of the plasma and in heat flux losses, thus lowering the amount of carbon needed to mitigate the plasma when compared to the 2D case. This study constitutes an important step forward in ‘predict-first’ simulations for disruption mitigation in NSTX-U and other devices, such as ITER. [ABSTRACT FROM AUTHOR]

The aim of the article is a description of the first realization of the pellet injection experiment at the COMPASS tokamak. In order to study the interaction of the solid state materials with runaway electrons in tokamaks, the room temperature solid state pellet injector (RTSP) was borrowed from the ASDEX Upgrade tokamak and reinstalled at COMPASS. Firstly, RTSP was characterized at the test bed at Max Planck Institute in Garching and important quantities for successful reinstallation such as pellet divergence and propellant gas throughput were measured. Obtained results were taken into account during design and assembly of the dedicated vacuum system needed for the RTSP operation. The RTSP system was operated in the runaway electron campaigns at the COMPASS tokamak and reliable delivery of graphite pellets (2 × 1. 5 mm) was demonstrated. The preliminary analysis of the pellet injection impact proving successful pellet firing into a plasma, is given. [Display omitted] • Successful reinstallation of the room temperature solid state pellet injector from the ASDEX Upgrade tokamak at the COMPASS tokamak. • Realization of the first pellet injection experiment at the COMPASS tokamak. • Usage of solid state pellets for investigation of runaway electrons. • Observation of graphite pellet explosion caused by fast electron volumetric heating in great detail. [ABSTRACT FROM AUTHOR]

A BOUT++ three-field MHD model is applied to study the effect of the pedestal pressure profile modified by the pellet injection from the low field side on the nonlinear behaviors of ELM. The linear simulation shows that the growth rates of all toroidal modes can be enhanced regardless of the deposition location and are more affected by the pellet atoms depositing in the steep pedestal region. From the nonlinear simulation, a threshold behavior for influencing the natural ELM by the ablated pellet in terms of the rate of the modified pressure over the unperturbed one at the most affected position (Rp=Pmod,MAP/P0,MAP) is demonstrated when the pellet deposits its atoms on the pedestal top. In contrast, when the pellet deposits its atoms within the steep pedestal region, the ELM size is increased at low and high Rp but decreased in an intermediate range. This result can perhaps explain the probabilistic behavior of ELM triggering by small pellet injection on EAST observed in previous experimental results.

Mitigating disruptions is essential for the longevity of future large tokamak experimental devices and reactors. Currently, shattered pellet injection (SPI) technique is the most effective mitigation technique found thus far, and has been chosen for the baseline disruption mitigation (DM) system for ITER. To optimally design SPI systems, the survivability of the pellet throughout the pre-shatter flight and the resulting shatter spray must be better understood. Experimental tests of low-angle single strike impacts of neon and argon pellets were conducted to determine the minimum normal kinetic impact energy that pellets can withstand throughout guide tube travel. Knowing the maximum normal kinetic energy that pellets of relevant materials and temperatures can withstand during flight will allow for an optimal SPI system design. Characterization of the downstream shatter spray was performed for various shatter tube geometries using pure argon and neon pellets. The experimentally measured post-shatter fragment size distribution is compared to theoretical models. The most accurate model found from this comparison is a statistics-based model for brittle material shattering that is correlated with the relevant physical parameters of SPI pellets. Extrapolation of the model to ITER size pellets is presented. [ABSTRACT FROM AUTHOR]

The features of a flow generated during the injection of a supersonic small body (pellet) from the channel of a spherically blunted cylinder into a supersonic flow are studied. The cylinder's diameter is 35 times larger than the diameter of the injected body. The numerical simulation is carried out using multilevel Cartesian grids with the local adaptation based on the wavelet analysis. The body's motion is simulated by the free boundary method. The dynamics of the moving body's interaction with the bow shock from the cylinder, the formation of the reverse flow region between the bodies, its evolution and disappearance, and the subsequent establishment of a stationary flow are studied. The decrease of the main body's drag to 20% of the initial value is demonstrated. [ABSTRACT FROM AUTHOR]

• Type-III ELMy H-mode plasmas have been fueled by pellets on EAST tokamak. • A string of small ELMs were induced after pellet injection from LFS at a low velocity. • The changed edge plasma parameters were proposed as the reason for the ELM modification. Type-III ELMy H-mode plasmas are a valuable operating scenario, as they may simultaneously achieve fusion power gain of Q = 10 and acceptable steady-state and peak power fluxes to the plasma-facing components. It is important to investigate cryogenic deuterium pellet injection fueling in this kind of H-mode plasma since it has been chosen as a main plasma fueling technique for ITER. Recently, pellet fueling experiments have been successfully carried out on EAST, an ITER-like tokamak. A maximum electron density of 0.73 × n GW has been obtained. It is found that ELM behavior was modified immediately and a string of small ELMs were induced when pellets were injected at a low velocity from ˜40 cm above the mid-plane on the low field side, which is different from the fact that pellets trigger signal ELMs in Type-I ELMy H-mode plasmas in other tokamaks. The decreased edge temperature caused by the cryogenic pellet was proposed as the reason for this ELM modification. These results will provide a useful reference for the pellet fueling in a Type-III ELMy H-mode plasma on ITER in the future. [ABSTRACT FROM AUTHOR]

In inertial confinement fusion, the scientific issues include the generation and transport of driver energy, the pellet design, the uniform target implosion physics, and the realistic nuclear fusion reactor design. In this paper, we present a pellet injection into a power reactor in heavy-ion inertial fusion (HIF). We employ a magnetic correction method to reduce the pellet alignment error in HIF reactor chamber, including the gravity, the reactor gas drag force, and the injection errors. We found that the magnetic correction device proposed in this paper is effective to construct a robust pellet injection system with a sufficiently small pellet alignment error. [ABSTRACT FROM AUTHOR]

Highlights • The smallest pellet (1.5 mm of length, 2 mm of diameter) injection is successful. • Slower pellet injection less than 150 m/s is achieved. • Pellet trajectory is closer to plasma by straight extension part. Abstract 20 Hz pellet injection system was installed for improving a fueling system of KSTAR (Korea Superconducting Tokamak Advanced Research) in 2016. Several things of hardware such as a liquid helium transfer line, an extruder nozzle and an extension part of pellet guide tube are changed for 2017 campaign. For plasma experiments, the pellet injection system uses the smallest pellet (1.5 mm of length, 2 mm of diameter). The pellet velocity can be variable from 150 m/s to over 200 m/s by controlling of propellant gas pressure. The clear increase of plasma density observed during KSTAR experiments is also presented in this paper. [ABSTRACT FROM AUTHOR]