Pellet Injection Research Articles

Simulation and developments for large pellet formation and acceleration for shattered pellet injection of the ITER DMS

Amongst the technological developments needed for ITER, the disruption mitigation system (DMS) is one of the most challenging since it has to mitigate the consequences of high-energy plasma disruptions to ensure the lifetime of ITER's structure and in-vessel components. ITER DMS is using Shattered Pellet Injection, whose technology needs to be developed for ITER specific requirements.In the framework of the “Technology Fundamental Studies for DMS” contract with ITER Organization, as part of the ITER DMS Task Force activities, and CEA, the DSBT (Low Temperature Systems Department) designed and operates a pellet injection test bench to produce and accelerate large cryogenic pellets with diameters from 10 to 28.5 mm and length/diameter ratios up to two. The objective of this activity is to study and to optimise the pellet formation and acceleration process using pure Protium, Deuterium, Neon and mixtures of these gases.Analytic and numerical studies were used to define the geometrical and cryogenic parameters for the pellet desublimation in the in-situ condensation cell of the cryostat. An analytical model has also been developed for the pellet acceleration in the single stage gas gun to validate the propellant valve design aiming at achieving velocities of up to 500 m/s for 28.5 mm protium pellets and 200 m/s for neon pellets.The complete test bench is composed of the gas propulsion system, the cryostat with its key component the pellet formation cell, the flight tube and the target chamber with viewing ports. The mechanical assembly is completed with specific instrumentation and fast data acquisition systems for the observation of the pellet formation and acceleration. The pellet integrity and trajectory are characterised by shadowgraphy with high resolution, high-speed cameras located at 1 m and 3 m after the barrel exit and with an accelerometer and a target foil in the target chamber.This document details the simulation studies, the selected test bench design and first obtained results.

Development of electromagnetic pellet injector for disruption mitigation of tokamak plasma

Disruption remains to be a serious threat to large tokamaks like the International Thermonuclear Experimental Reactor (ITER). The injection speed of disruption mitigation systems (DMS) driven by high pressure gas is limited by the sound speed of the propellant gas. When extrapolating to ITER-like tokamaks, long overall reaction duration and shallow penetration depth due to low injection speed make it stricter for plasma control system to predict the impending disruptions. Some disruptions with a short warning time may be unavoidable. Thus, a fast time response and high injection speed DMS is essential for large scale devices. The electromagnetic pellet-injection (EMPI) system is a novel massive material injection system aiming to provide rapid and effective disruption mitigation. Based on the railgun concept, EMPI can accelerate the payload to over 1000 m/s and shorten the overall reaction time to a few milliseconds. To verify the injection ability and stability of the EMPI, the prototype injector EMPI-1 has been designed and assembled. The preliminary test has been carried out using a 5.9 g armature to propel a dummy pellet and the results suggest that the EMPI configuration has a great potential to be the DMS of the large scale fusion devices.

Control-oriented core-SOL-divertor model to address integrated burn and divertor control challenges in ITER

The real-time regulation of a burning plasma’s temperature and density, or burn control, will be necessary to produce high fusion power in future tokamaks like ITER. This is made more challenging due to the plasma’s nonlinear characteristics and the interdependence between the core-plasma and edge-plasma regions. For example, a raising plasma temperature leads to increasing reactivity and therefore to more alpha-particle heating, which further increases temperature. Furthermore, a raise of the fusion power increases the heat flow through the scrape-off-layer (SOL), which can compromise the integrity of the divertor without proper safeguards. For control design, a model-based approach is attractive because it can directly incorporate the nonlinear, coupled, burning-plasma dynamics into the design. To facilitate this design approach, a control-oriented core-SOL-divertor (CSD) model is presented in this work. In this CSD model, a core-plasma model captures the nonlinear dynamics of the core’s density and temperature, and a SOL-divertor model defines the plasma conditions at the separatrix and divertor including the heat load on the target plates. The core-plasma and SOL-divertor models are coupled through the exchange of various variables. In particular, the SOL-divertor model yields the separatrix temperature and the influx of recycled particles into the core-plasma. These variables influence the power and particle balances captured by the core-plasma model. In return, the core-plasma model determines the intensity of the heat and particles fluxes across the separatrix, and this outflow strongly impacts the SOL-divertor model. Therefore, the power and density of the core-plasma, which can be readily modulated through external heating systems and pellet injection, can be viewed as control knobs for the SOL-divertor region in addition to the gas puffing. In simulations of the CSD model, it is demonstrated how external actuation can be utilized to meet burn control and divertor control objectives simultaneously.

WEST plasma control system status

This paper presents the current status of the WEST Plasma Control System (PCS). From the first campaign at the end of 2016 to the latest campaign at the end of 2022, significant upgrades have been deployed to improve several aspects of the machine performance. The vertical instability control loop has been upgraded in order to increase the operation domain (i.e. plasma elongation). This upgrade optimizes the delay at each stage of the control loop using raw magnetic data coupled to a real-time vertical displacement observer. This allows for different shape control modes that address various operation issues such as X-point formation and gap control. The control of the plasma density uses generic feedforward and/or feedback controllers to control the 21 piezo-electric valves of the gas injection system and/or the pellet injection system. The heating systems of WEST (ICRH and LHCD) are also driven by the PCS, which has the ability to control both the power and the phase of the antennas. More recently, feedback control of the loop voltage using the LHCD power has been fully implemented in order to perform plasma discharges of up to 1000 s. Also in the most recent campaigns, the entirety of the lower divertor has been actively cooled. Operating a fully actively cooled metallic machine requires real-time machine protection strategies to avoid water-leaks or other damage to the first wall. On WEST, such protections are based on spectroscopic density measurements of impurities and on the infrared monitoring of critical components. These strategies trigger either minor adjustments of the reference scenario (e.g. decreasing of the requested heating power) or a soft plasma termination.

A comparison of the influence of plasmoid-drift mechanisms on plasma fuelling by cryogenic pellets in ITER and Wendelstein 7-X

Pellet injection is the most promising technique to achieve efficient plasma core fuelling, key for attaining stationary scenarios in large magnetic confinement fusion devices. In this paper, the injection of pellets with different volumes and speeds into standard plasma scenarios in ITER (tokamak) and Wendelstein 7-X (stellarator) is studied by modeling the pellet ablation and particle deposition, focusing on the evaluation of the expected differences in pellet plasmoid drifts in tokamaks and stellarators. Since the efficiency of the damping-drift mechanisms is predicted to depend on the magnetic configuration, device-specific characteristics are expected for the temporal evolution of the plasmoid drift acceleration. For instance, plasmoid-internal Pfirsch–Schlüter currents dominate the drift damping process for stellarators, while plasmoid-external currents are more relevant for tokamaks. Also, relatively larger drifts are in principle expected for W7-X due to higher field gradients in relation to machine dimensions. However, shorter plasmoid-internal charge reconnection lengths result in the drift damping due to internal Pfirsch–Schlüter currents being more effective than in a tokamak. Therefore, the average relative drift displacement during the whole plasmoid homogenization may a priori be comparable in both magnetic configurations. Moreover, High Field Side (HFS) injection is expected to be highly advantageous to maximize pellet particle deposition in ITER, whereas it may only be beneficial in medium to high β environments in W7-X. Finally, there may be means for the optimization of pellet injection configurations in both ITER and W7-X for the considered plasma scenarios despite the sizeable differences in the relative importance of the mechanisms of plasmoid drift acceleration and deceleration in play.

Development of a Shattered Pellet Injector test bench for the ITER DMS support laboratory

In the ITER Shattered Pellet Injection system large, Ø28.5x57mm cryogenic pellets have to be created and preserved in a barrel, until the control system detects a forthcoming disruption event and triggers the propellant gas valve to launch the pellet. The high-pressure propellant gas accelerates the pellet, which shatters on an angled plate near the edge of the plasma. The ablation of the resulting pellet fragments is intended to mitigate the consequences of the disruption and to prevent the generation of runaway electrons. To support the Shattered Pellet Injector technology for the ITER Disruption Mitigation System, a new support laboratory has been built at the Centre for Energy Research. The aim of the laboratory is to develop a test bench with ITER relevant dimensions, produce d=28.5mm Hydrogen, Neon and mixture pellets, observe the pellet acceleration and shattering with automated control. This contribution describes the cryogenic pellet injector of the test bench.

In-situ pellet growth and quality monitoring diagnostics for the ITER DMS Support Laboratory

The ITER disruption mitigation system is based on shattered pellet injection. The ITER DMS requires the use of large, 19 × 38 and 28.5 × 57 mm (diameter x length) cryogenic pellets, made of hydrogen, neon, or a mixture of those. The production of such large pellets, especially protium, is not straightforward, therefore the freezing process is monitored in real-time. A barrel view camera diagnostic was developed at the Centre of Energy Research for this purpose: the pellet is illuminated from one side of the barrel via a special vacuum adaptor featuring a high-pressure rated window, while observing it from the other side using a digital camera. It was found that the growth of the pellet can be followed easily, and issues such as bad heat contact between the cold head and the barrel can clearly be observed as well. Once the pellet is launched, its integrity and other parameters can be determined in an optically assessable vacuum chamber, using a shadowgraphy diagnostic. The pellet size can routinely be determined to sub-mm precision, and voids (or other non-solid material regions) can be identified as dark areas on the captured shadowgraphs.

Design of a shattered pellet injector and preliminary bench tests of Ne pellet formation for EAST disruption mitigation

Shattered pellet injection (SPI) technology has been demonstrated as a novel effective method for disruption mitigation in tokamaks. To further verify the superiority and reliability of this technology, in this study, a SPI test stand based on in-situ gas formation technology was designed and constructed for Experimental Advanced Superconducting Tokamak (EAST). A pellet producer, a gas feeding system, a pumping system, diagnostic units, and a shatter chamber have been developed and are served as the main components of the SPI test stand. We experimentally demonstrated that the SPI test stand developed in this study can be used for forming and accelerating impurity pellets. On the test stand, Ne pellets with a diameter of ∼5 mm were successfully formed, and their length was varied from ∼7 mm to ∼15 mm by changing the heating power of the heat sinks after the formation of the pellets. The feasibility of pellet shattering was verified using a shatter tube (inner diameter: 15.7 mm) in the shatter chamber. The results showed that Ne pellets with a velocity of ∼150 m/s could be broken into fragments without removing the propellant gas by passing them through a ∼20° bent tube. The SPI test stand developed in this study and the experimental results presented in this paper will serve as the technical basis for developing a reliable SPI system on EAST and will be a good reference for the development of such systems on other fusion devices.

The versatile fueling systems of WEST

New fueling systems have been designed for WEST, a full tungsten wall tokamak with divertor configuration. They consist of standard gas injection, supersonic molecular beam injection and pellets injection systems, each being able to fuel continuously the plasma discharge up to a duration of 1000s. A massive gas injection system is also implemented to mitigate the impact of disruptions and runaway electrons. The standard gas system consists of 4 independent lines allowing the injection of up to 4 different gas species during a plasma discharge through 11 injections areas, 4 of them being located in the divertors and in the private flux region. The pellet injection system is able to inject up to 10 pellets per second through 4 tracks: one on the low field side, and the other 3 on the high field side. The supersonic molecular beam injection system consists of 3 injectors located close to the midplane, two on the high field side and one on the low field side. On trigger, the massive gas injection system can deliver in a single puff up 800 Pa.m3 of noble gas in order to stop the plasma discharge. All the fueling systems have been characterized and can be used routinely on plasma.

Fragment plume diagnostics for cryogenic pellet shattering studies: Development and first experimental results

Shattered pellet injection is the baseline technology for the ITER Disruption Mitigation System (DMS) [1]. ITER DMS requires the use of large, 28.5 × 57 mm (diameter x length) cryogenic pellets, made of hydrogen, neon, or a mixture of those gasses and accelerated to several hundred m/s velocity. Beside the production, acceleration and transfer of such ITER-size pellets, the characterization of the shattering is one of the main task of the ITER DMS Support Laboratory at center of Energy Research as part of the ITER DMS Task Force program to validate the design choices of the DMS [2]. For this purpose, a fragment plume diagnostic chamber and two optical diagnostics using fast framing cameras, namely the laser curtain at two distances from the shattering head and a shattering observation were designed and implemented. Preliminary experiments for different pellet material and velocities at 20⁰ and 30⁰ shattering angle showed a strong pellets sublimation during shattering process. The fragments seem to slide along the shattering plate, no significant total reflection was detected. No qualitative difference between 20⁰ and 30⁰ shattering angle was observed, but clear differences for different materials and or pellet speeds.

Self-consistent investigation of density fueling needs on ITER and CFETR utilizing the new Pellet Ablation Module

Self-consistent modeling using the stability, transport, equilibrium, and pedestal (STEP) workflow in the OMFIT integrated modeling framework (predicting pedestal with EPED, core profiles with TGYRO, current profile with ONETWO, and EFIT for equilibrium) suggests ITER and future devices such as China Fusion Engineering Test Reactor (CFETR) Zhuang et al (2019 Nucl. Fusion 59 112010) will benefit from high-density operation (Greenwald limit fraction 0.7−1.3). Regimes with an operational density near the Greenwald limit will likely need peaked density profiles so that the pedestal density remains below the Greenwald limit. Peaked density profiles can be achieved with the help of pellet injection. A flexible Pellet Ablation Module (PAM), which predicts the density source based on a comprehensive analytical pellet ablation model, has been developed for predicting pellet fueling for transport studies, and has been incorporated into the STEP workflow for predictive modeling. This workflow is applied to DIII-D and finds good agreement with experiments. On ITER the effect of pellet fueling is examined in an advanced inductive scenario, where a fusion gain of up to Q = 9 is predicted with strong central pellet fueling. On CFETR, with a mid-radius density source, an average of electrons s−1 are required to achieve the density and temperature profiles necessary for the 1000 MW advanced scenario with a tritium burn-up fraction of .

Mixed-Integer MPC Strategies for Fueling and Density Control in Fusion Tokamaks

Model predictive control (MPC) is promising for fueling and core density feedback control in nuclear fusion tokamaks, where the primary actuators, frozen hydrogen fuel pellets fired into the plasma, are discrete. Previous density feedback control approaches have only approximated pellet injection as a continuous input due to the complexity that it introduces. In this letter, we model plasma density and pellet injection as a hybrid system and propose two MPC strategies for density control: mixed-integer (MI) MPC using a conventional mixed-integer programming (MIP) solver and MPC utilizing our novel modification of the penalty term homotopy (PTH) algorithm. By relaxing the integer requirements, the PTH algorithm transforms the MIP problem into a series of continuous optimization problems, reducing computational complexity. Our novel modification to the PTH algorithm ensures that it can handle path constraints, making it viable for constrained hybrid MPC in general. Both strategies perform well with regards to reference tracking without violating path constraints and satisfy the computation time limit for real-time control of the pellet injection system. However, the computation time of the PTH-based MPC strategy consistently outpaces the conventional MI-MPC strategy.

ECH and ECCD modelling studies for DTT

In this work the Electron Cyclotron (EC) physics performances of the EC system foreseen for the new Divertor Tokamak Test facility (DTT) are investigated using the beam tracing code GRAY on the flat top phase of the most recent DTT full power scenario. The whole core plasma region can be reached by EC beams with complete absorption, assuring bulk heating and core current drive (CD) for profile tailoring, and NTM mitigation in correspondence of the rational surfaces. A detailed analysis regarding modifications of the EC propagation, absorption and CD location due to density fluctuations caused by pellet injection is performed. The compatibility between the EC system and the pellet injection system is verified: the density variations due to pellet injection are foreseen to negligibly influence the EC performances, allowing the EC beams to reach the plasma central region for bulk heating and to drive current on the rational surfaces for NTM mitigation. Finally, the polarization variations originated by the angle steering foreseen for the operational and physics tasks accomplishment during the flat top phase of the discharge are assessed. Negligible power losses have been found keeping fixed polarization during the needed steering.

EUV spectral shape variation of tungsten unresolved transition arrays in electron temperature range of 2–4 keV observed in the Large Helical Device

Spectroscopic studies of emissions released from tungsten ions combined with a pellet injection technique have been conducted in the Large Helical Device. The tungsten Unresolved Transition Array (UTA) spectrum was observed in the wavelength ranges of extreme ultraviolet (EUV) 6–60 Å and 130–340 Å, and the electron temperature dependence of the UTA spectral shape was investigated in the electron temperature region < 4.3 keV. The UTAs of W24+–W33+ at 20–33 Å, W37+–W42+ at 45–47 Å, W27+–W29+ at 48–55 Å, and W7+– W27+ at 170–210 Å were observed. Unidentified UTAs were also found at 230–270 Å and 280–320 Å. As the electron temperature increased further above 4 keV, the W37+–W42+ UTA at 45–47 Å was maintained, while the other UTAs became less intense.

Numerical Simulation Analysis of Flow Field on Repetitive Pellet Injector with Single-Screw Extruder in CFETR

Pellet injection (PI) is the preferred fueling method in the future fusion reactor. It is particularly important to study the flow field characteristics of the frozen fuel extrusion process for the future steady operation of the pellet injector. In order to study the influence of groove depth on extrusion flux and conveying capacity, the flow field characteristics of a repetitive pellet injector with a single-screw extruder in the China Fusion Engineering Test Reactor (CFETR) was numerically simulated with POLYFLOW software. Thus, information about pressure field, viscous heating, and velocity field distribution was obtained. The results indicate that to a certain extent, increasing the groove depth (while maintaining the gaps between the screw and extrusion cylinder) is beneficial for the conveying capacity and pressure building capacity. The results of the numerical simulations show that at a screw speed of 120 rpm, screw outer diameter of 20 mm, screw length of 230 mm, screw groove depth of 6 mm, and screw prism gap of 0.3 mm, solid hydrogen can be stably extruded, and the velocity of the extruded ice at the nozzle is 0.15 m/s, which meets the design requirement of the CFETR PI system. These results also provide good references for structure design and performance optimization of the CFETR pellet injector.

Enhanced Material Assimilation in a Toroidal Plasma Using Mixed H_{2}+Ne Pellet Injection and Implications to ITER.

The ablation and assimilation of cryogenic pure H_{2} and mixed H_{2}+Ne pellets, which are foreseen to be used by the ITER tokamak for mitigating thermal and electromagnetic loads of major disruptions, are observed by spatially and temporally resolved measurements. It is experimentally demonstrated that a small fraction (here ≈5%) of neon added to hydrogenic pellets enhances the core density assimilation with reduced outward transport for the low magnetic-field side injection. This is consistent with theoretical expectations that line radiation increased by doped neon in dense plasmoids suppresses the plasmoid pressure and reduces the E[over →]×B[over →] transport of the ablated material.

Overview of runaway current suppression and dissipation on J-TEXT tokamak

The main works on disruption mitigation including suppression and mitigation of runaway current on the J-TEXT tokamak are summarized in this paper. Two strategies for the mitigation of runaway electron (RE) beams are applied in experiments. The first strategy enables the REs to be completely suppressed by means of supersonic molecular beam injection and resonant magnetic perturbation which can enhance RE loss, magnetic energy transfer which can reduce the electric field, and secondary massive gas injection (MGI) which can increase the collisional damping. For the second strategy, the runaway current is allowed to form but should be dissipated or soft landed within tolerance. It is observed that the runaway current can be significantly dissipated by MGI, and the dissipation rate increases with the injected impurity particle number and eventually stabilizes at 28 MA s−1. The dissipation rate of the runaway current can be up to 3 MA s−1 by ohmic field. Shattered pellet injection has been chosen as the main disruption mitigation method, which has the capability of injecting material deeper into the plasma for higher density assimilation when compared to MGI. Moreover, simulation works show that the RE seeds in the plasma are strongly influenced under different phases and sizes of 2/1 mode locked islands during thermal quench. The robust runaway suppression and runaway current dissipation provide an important insight on the disruption mitigation for future large tokamaks.

Investigation of the compatibility of pellet fueling with ELM-free H-mode plasmas in EAST tokamak

Experiments on pellet fueling have been carried out in edge localized mode (ELM)-free high-confinement mode (H-mode) plasmas with q 95 ∼ 6 in the EAST tokamak. Cryogenic deuterium pellets were injected into the ELM-free plasmas at a frequency of 10 or 5 Hz from ∼45 cm above the mid-plane on the low-field side. It is found that the ELM-free H-mode plasmas are still sustained even if both the edge and core plasma are impacted by the pellet injections (PIs). Several small ELMs would appear and the edge coherent mode accompanying the ELM-free phase fades or even disappears just after the pellet events, but the plasma would rapidly recover to the ELM-free state. Although the ELMing phase is very short, it may be an issue that still needs to be resolved in the future. Meanwhile, the 3/2 tearing mode often appearing in ELM-free discharges would be stabilized by the PIs, and the high-Z impurity concentration would be reduced during the PIs. All these results will be meaningful to International Thermonuclear Experimental Reactor and future fusion reactors.

Plasma Instability During ITBs Formation with Pellet Injection in Tokamak

JET H-mode plasma discharge 53212 simulation during the pellet fueling operation in the presence of an internal transport barrier is carried out using the 1.5D BALDUR integrated predictive modelling code. The plasma instability during ITB formation with pellet injection in a tokamak is investigated. These simulations use a neoclassical transport model and an anomalous transport model (either multimode or mixed Bohm/gyro-Bohm core transport model). The boundary condition is described at the top of the pedestal, which is calculated theoretically based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The toroidal flow calculation is based on the neoclassical viscosity toroidal velocity model. It was found that the shallower pellet does not destroy the ITB, which locating mainly between r/a = 0.8 and 0.9. Moreover, in the plasma center region (0.4&lt;r/a&lt;0.6) the effective electron thermal diffusivities do not change during the ablation time. However, the effective electron thermal diffusivities decrease after pellet ablation, which means a shallower pellet can improve the internal transport barrier.

Reducing the L-H transition power threshold in ITER-similar-shape DIII-D hydrogen plasmas

Recent dedicated DIII-D experiments in low-torque, ITER-similar-shape (ISS) hydrogen plasmas (at a plasma current I p ∼ 1.5 MA and ITER-similar edge safety factor q 95 ∼ 3.6) show that the L-H transition power threshold P LH can be reduced substantially (∼30%) with L-mode helium admixtures n He/n e ⩽ 25%. In the ensuing H-mode, helium ion fractions n He/n H remain below 25%. H-mode normalized pressure and confinement quality are only slightly affected by helium seeding, and Z eff ⩽ 2.15 (including helium and carbon content). The plasmas investigated here are electron-heat dominated, with temperatures T e(0)/T i(0) ⩾ 1 and edge heat flux ratio Q e/Q i(ρ = 0.95) ∼ 1.2–1.5. Without mitigation, P LH is higher by a factor of 2–3 in comparison to similar ISS deuterium plasmas. ISS hydrogen plasmas with lower plasma current I p ∼ 1 MA (increased edge safety factor q 95 ∼ 5.1) exhibit a substantially lower power threshold. This plasma current dependence, also observed previously on ASDEX-U and in JET, is not accounted for by the commonly used 2008 ITPA multi-machine threshold scaling, but could potentially allow H-mode access at marginal heating power during the initial plasma current ramp-up. Attempts to reduce P LH with low-field- and high-field-side hydrogen pellet injection, using 1.7 mm diameter pellets, have not demonstrated a robust threshold reduction, in contrast to successful earlier experiments with larger 2.7 mm pellets. Techniques for reducing P LH are very important for ITER, in particular for accessing H-mode in hydrogen plasmas during the Pre-Fusion Power Operation-1 (PFPO-1) campaign with marginal auxiliary heating power (20–30 MW of ECH).