Pellet Injection Research Articles

Enhanced energy confinement after series of pellets in Wendelstein 7-X

A series of ice pellets was injected into the advanced stellarator Wendelstein 7-X (W7-X). Although the pellets were small and slow, deep and efficient particle fueling could be observed experimentally. The most striking feature appearing after the injection of the pellets, however, was a transient increase in the energy confinement time. This transient phase resembled in several aspects modes of enhanced confinement after gas-puff or pellet injection, as observed in other fusion experiments. All experimental attempts, to prolong this phase, failed. In this paper, discharges are described that show the enhanced energy confinement, and some conditions are summarized which seem to be essential in order to generate it. The focus here is on deep particle fueling by pellets, and shaping of the density profiles during and after the series of pellets. During this time, neutral gas particle re-fueling at the plasma edge is reduced, while density profile peaking and low impurity radiation losses are present.

Sandpile modelling of dual location fuelling in fusion plasmas

We modify the Chapman sandpile model (Chapman et al 2001 Phys. Rev. Lett. 86 2814) to form comparisons with pellet pacing, which is used to reduce or eliminate ELMs in a fusion plasma. We employ a variation of that model in which a pedestal with feedback is introduced (Bowie and Hole 2018 Phys. Plasmas 25 012511), which we further modify to provide for dual fuelling—sand is added both at the centre of the sandpile, and near the edge. We observe that when the additional sand is added at the top of the pedestal, mass loss events (MLEs) are largely suppressed. While this suppression comes at a cost by way of reduction in total energy confinement, that reduction is lower than the reduction in MLE size. The trade-off between MLE suppression and reduction in energy confinement depends not only on the amount of extra sand, but also on its precise location relative to the top of the pedestal. We suggest that the approach of constant dual fuelling may be equally applicable to plasmas, and may suggest a strategy for ELM suppression in fusion plasmas. We observe that when the proposed amount of extra sand is added in ‘pellets’, using frequencies and amounts based on those proposed for ELM suppression for ITER, MLEs are similarly suppressed, although MLEs are not significantly suppressed when the pellet rate does not substantially exceed the MLE frequency. This suggests that pellet injection at the top of the pedestal at small pellet size and high frequency may represent a reasonable physical proxy for our proposed scheme. However, our results suggest that it is not the synchronisation of pellets to ELM frequencies which is the key factor for ELM suppression in this regime, but rather the introduction of additional fuelling at the top of the pedestal.

Optimizing the EU-DEMO pellet fuelling scheme

• Elaborated suitable solution for the core particle fuelling of EU-DEMO fusion reactor. • Inboard pellet injection identified as suitable technique. • Conservative conventional approach, relying only on proven technical potential. • Next step: additional integration of plasma feedback control requirements. Efficient fuelling will be an essential task in the EU-DEMO. The basic requirement here is to establish the target plasma core density with a minimum particle flux by injecting mm-size solid fuel pellets. Modelling showed this requires a pellet launch from the vessel inboard. Optimization can be achieved by the pellet parameters and the injection geometry; the latter however taking into account boundary conditions resulting from system integration needs. Design activities integrating the pellet transfer system into the vessel and the breeding blanket unveiled several possible variants requiring different levels of technical efforts. Basically, all extra efforts bear the benefit of a deeper and hence more favourable particle deposition. To quantify the potential gain, a full closed loop modelling was performed calculating the required pellet particle flux for any solution considered. Results allow now to balance potential advantages against related efforts required. Furthermore, the analysis tools developed can be employed for even more refined optimization of the pellet actuator tool by e.g. taking into account the interplay of pellet fuelling with burn control requirements.

Dynamic modeling of pellet production process for pellet injection system in ITER

Abstract This article proposes a dynamic simulation model of the pellet production process for the pellet injection system (PIS) in the fusion fuel cycle. The integrated devices in the pellet production process should be designed and operated to meet safety limitations while ensuring the required quantity of fuel supply. The presented model is used to investigate the variation in fuel inventory of the system under time-varying tokamak fuel demand. The model applies to the T2 pellet production in three different operation scenarios: inductive, hybrid and non-inductive operation. The simulation results suggest the reliable storage capacity of the main devices under a wide range of changing operating conditions. Preliminary analysis of design and operation options is expected to provide useful insights into the development of PIS.

Integration of coupled modeling of the core and divertor plasmas into “FC-FNS” code and application to DEMO-FNS project

Abstract The “FC-FNS” (Fusion Cycle for Fusion Neutron Source) fuel cycle model is used to calculate the hydrogen isotope flows in the fuel systems of a tokamak-based fusion neutron source (FNS). The present development incorporates into the FNS model the coupled modeling of the core and divertor plasmas that provides the relation between the plasma parameters and particle fluxes through the vacuum chamber in steady state operation. The upgraded “FC-FNS” code is applied to the analysis of the DEMO-FNS project based on a tokamak with the parameters R/a = 3.2 m/1 m, B = 5 T, Ipl = 4–5 MА, PNBI =30 MW, PECR =6 MW and DT fusion power Pfus =40 MW. The possibility of reducing the hydrogen fluxes in the pumping and injection systems when using an admixture of Ne in a diverter is substantiated. The operation mode of the DEMO-FNS fueling systems depending on the main plasma parameters is analyzed and the gas flows for different D2/T2/DT pellet injection modes are evaluated.

Deployment of multiple shattered pellet injection systems in KSTAR

Abstract Shattered pellet injection is of the most attractive way to mitigate the plasma disruption in fusion research facilities up to now. DIII-D has already utilized it and achieved very positive results. ITER has decided to adopt this technology for the DMS (disruption mitigation system) for PFPO-1 (Pre-Fusion Power Operation phase 1). The validation between simulation code and experiment, and continuous engineering development need to be carried out to meet the DMS’s requirement and reliability. KSTAR (Korea Superconducting Tokamak Advanced Research) is a possible candidate to test the urgent issues of plasma disruption for ITER. KSTAR can install two injectors in toroidal opposite positions. For this work, ORNL (Oak Ridge National Laboratory) will provide the two injectors, the shatter tubes and auxiliary systems. NFRI (National Fusion Research Institute) is preparing the infrastructure of a pumping system, control and data acquisition system, and arranging the location of diagnostic and the heating systems. This presentation describes the basic requirements and the engineering challenges to be solved for successful deployment and operation of multiple SPI injectors in 2019.

Shattered pellet penetration in low and high energy plasmas on DIII-D

Shattered pellet injection (SPI) has been adopted as the baseline disruption mitigation system for ITER, as the radiative payload penetration into DIII-D plasmas from SPI is superior to those using the massive gas injection (MGI) method. Because of the substantial differences in the energy content of ITER plasma and those in present experiments, reliable 3D MHD modeling, benchmarked against present experiments is needed to project to ITER plasmas. In support of these needs, the depth of SPI fragment penetration in DIII-D plasmas was investigated by injecting SPI into two discharges with vastly different energy content and pedestal height. 400 Torr-L pure Ne fragmented pellets at a velocity of about 200 m s−1 were injected into a 0.2 MJ L-mode discharge and a 2 MJ super H-mode discharge. Results show deep penetration of SPI fragments into low-energy plasmas in DIII-D. SPI fragment penetration is reduced as the plasma energy content increases, with some discharges exhibiting penetration that is confined to the outer regions of the plasma. The injected SPI fragments are also spread out over a distance of about 20 cm, which results in some fragments arriving near the end of or after the thermal quench is over.

Numerical modelling of solid hydrogen flow through an extruder die

Harnessing energy by nuclear fusion has progressed into technological developments as a result of sufficient progress made in the theoretical investigations. Fueling a tokamak by pellet injection is reported to be most efficient among other fueling techniques. It is thus important to produce a solid hydrogen filament by extrusion process using an extruder-die setup. It is required to arrive at an optimum configuration of the die as the pressure drop in the die decides the load on the extruder. The present analysis focuses on developing a CFD model of a die taking into account the complex rheological behavior of solid hydrogen. A systematic parametric analysis is performed to study the influence of die geometry on the flow behavior. The un-yielded regions are identified from the simulation results. It has been found that the un-yielded region decreases as the contraction ratio and contraction angle is decreased. The un-yielded zone is predominantly present in the region before contraction. However, at lower flow rates, the un-yielded zone appears even after the contraction.

First predictive simulations for deuterium shattered pellet injection in ASDEX Upgrade

First simulations of deuterium shattered pellet injection into an ASDEX Upgrade H-Mode plasma with the JOREK MHD code are presented. Resistivity is increased by one order of magnitude in most simulations to reduce computational costs and allow for extensive parameter scans. The effect of various physical parameters on MHD activity and thermal quench (TQ) dynamics is studied and MHD influence on ablation is shown. TQs are obtained quickly after injection in most simulations with a typical duration of 100 microseconds, which slows down at lower resistivity. Although the n = 1 magnetic perturbation dominates in the simulations, toroidal harmonics up to n = 10 contribute to stochastization and stochastic transport in the plasma core. The post-TQ density profile remains hollow for a few hundred microseconds. However, when flux surfaces re-form around the magnetic axis, the density becomes monotonic, again, suggesting beneficial behavior for runaway electron avoidance/mitigation. With 1021 atoms injected, TQ is typically incomplete and triggered when the shards reach the q = 2 rational surface. At a larger number of injected atoms, TQ can set in even before the shards reach this surface. For low field side injection considered here, repeated formation of outward convection cells is observed in the ablation region reducing material assimilation. This is due to sudden rise of pressure in the high density cloud when the stochastic region expands further releasing heat from the hot core. After TQ, strong sheared poloidal rotation is created by Maxwell stress, which contributes to re-formation of flux surfaces.

Construction and experimental tests of the high frequency pellet injection system on EAST tokamak

A high frequency deuterium pellet injection system based on extrusion and pneumatic acceleration has been recently developed and built for EAST tokamak, which is one of fully superconducting Tokamaks with ITER-like plasma configuration. It is mainly used for plasma edge localized mode control and can achieve 1–50 Hz injection rate. It has been completely installed on EAST up to now. A new differential pumping system and pellet transmission lines have been designed according to the actual condition of EAST. Recently we have tested the system systematically and injected multiple pellets into type-I ELMy H-mode plasma on EAST for the first time. The test and experimental results show that the system can inject continuous pellets into plasma stably, and ELMs can be triggered by deuterium pellets in H-mode plasma with negligible impact on density and stored energy. And it was also found that ELM trigger efficiency is less than 100 % sometimes. This lays an important foundation for further experimental study of ELM modulation using deuterium pellet injection on EAST.

Transition between Isotope-Mixing and Nonmixing States in Hydrogen-Deuterium Mixture Plasmas

The transition between isotope-mixing and nonmixing states in hydrogen-deuterium mixture plasmas is observed in the isotope (hydrogen and deuterium) mixture plasma in the Large Helical Device. In the nonmixing state, the isotope density ratio profile is nonuniform when the beam fueling isotope species differs from the recycling isotope species and the profile varies significantly depending on the ratio of the recycling isotope species, although the electron density profile shape is unchanged. The fast transition from nonmixing state to isotope-mixing state (nearly uniform profile of isotope ion density ratio) is observed associated with the change of electron density profile from peaked to hollow profile by the pellet injection near the plasma periphery. The transition from nonmixing to isotope-mixing state strongly correlates with the increase of turbulence measurements and the transition of turbulence state from TEM to ion temperature gradient is predicted by gyrokinetic simulation.

Experimental Pellet Shatter Thresholds and Analysis of Shatter Tube Ejecta for Disruption Mitigation Cryogenic Pellets

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.

Fuel pellet injection into heavy-ion inertial fusion reactor

In inertial confinement fusion (ICF), the scientific issues include the generation and transport of driver energy, the pellet design, the uniform target implosion physics, the realistic nuclear fusion reactor design, etc. In this paper, we present a pellet injection into a power reactor in heavy ion inertial fusion. We employ a magnetic correction method to reduce the pellet alignment error in an ICF 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. A fuel pellet would be covered by Pb and is cooled down by a liquid He. The cryo-pellet coated by Pb becomes a superconductor, when its temperature is below 7.2K. The magnetic field repels the superconductor due to the Meissner effect. We employ this effect to correct the fuel pellet injection velocity. The results of simulations demonstrate that the magnetic correction system is effective and robust to reduce the pellet alignment error.

Non-linear magnetohydrodynamic simulations of pellet triggered edge-localized modes in JET

Non-linear magnetohydrodynamic simulations of pellet-triggered edge-localized modes (ELMs) in JET plasma have been carried out with the JOREK code. The pellet particle fuelling efficiency and the power flux at the divertor target during the pellet-triggered ELM have been studied. The pellet injection in unstable plasma delivers the particle fuelling but the pellet fuelling rate is smaller than the rate of particle loss during the pellet triggered ELM. The JOREK simulations estimate the power flux at the divertor target and found good agreement with the experimental observation. The energy deposition of the pellet triggered ELM shows a toroidally asymmetric profile. However, and due to this toroidal asymmetry, this effect cannot be captured by the existing layout of the divertor infra-red (IR) cameras available in JET. This work highlights the benefit of having a larger number of IR cameras to analyse the heat flux for the experiments which are assumed to be toroidally asymmetric, such as the pellet and/or gas injection experiments.

Design of a Continuous Pellet Fueling System for Wendelstein 7-X

A continuous pellet fueling system (CPFS) is currently being designed at the Oak Ridge National Laboratory (ORNL) for the long pulse operation of the Wendelstein 7-X (W7-X) stellarator. The purpose of the CPFS is to provide deep continuous fueling for feedback-controlled high-density operation and mitigation of predicted hollow density profiles. The system will provide the capability to inject cylindrical pellets of solid hydrogen or deuterium into the plasma core, with flexibility to vary the pellet size, velocity, and injection frequency. Pellets are nominally of 3 mm in diameter and have a length between 1 and 4 mm. The heart of the CPFS is a vertically oriented, twin-screw extruder, cooled by three Gifford-McMahon cryocoolers in parallel, designed to form a continuous filament of hydrogen or deuterium. The filament width, which determines the pellet length, can be adjusted by means of a variable nozzle driven by a linear actuator at the base of the extruder. Coupled to the nozzle is a solenoid-operated gas gun and cutter assembly. A pneumatic propellant valve pulses a burst of ~60 bar helium to accelerate the cut pellet into W7-X. Three gaps in the guide tubes provide pumping locations to remove the helium propellant before it reaches the plasma. The maximum velocity of the pellet is limited by its ability to survive navigating the guide tube trajectory intact. A microwave cavity located within the guide tube provides the capability to measure the pellet size and velocity. The mechanical and thermal designs of the W7-X extruder, adjustable nozzle, and gun and cutter assembly design are described. A guide tube design and experimental pellet survivability test results are presented.

Integration of the TESPEL injection system at W7-X

Since impurities determine the plasma performance in magnetically-confined fusion plasmas, analyzing their behavior and controlling their confinement is significantly important for a stable and steady-state operation of fusion devices. A comparison of the impurity transport in the core region of large stellarators, such as Wendelstein 7-X and Large Helical Device (LHD) is quite useful to gain a better understanding of the impurity transport in reactor-relevant stellarator plasmas. Therefore, a Tracer-Encapsulated Solid Pellet (TESPEL) injection system has been newly installed for the operational phase OP1.2b of W7-X. The injector was designed and manufactured by NIFS to achieve a similar TESPEL deposition location as in LHD. Technical guidelines and functional specifications, addressing the specific conditions at W7-X had been implemented in close collaboration between NIFS and IPP. This resulted in a compact 3-stage differential pumping system considering spatial restrictions due to existing and future diagnostics, material requirements and limitations due to the magnetic stray field in the vicinity of the turbo pumps. This contribution reports on the design and the integration of the new TESPEL injection system to W7-X and first achievements during the commissioning of the system.

Spontaneous drop of the Li-containing flake as a trigger of ITB-event in plasmas of the T-10 tokamak during simultaneous co+counter injection of EC waves

Various triggers of ITB-events (ITB-event is an abrupt and non-local reduction of the heat flux within a radius of ∼50% of the minor radius) have been previously found at the JT-60U, the T-10 tokamaks and the LHD stellarator. This short paper describes a new trigger of ITB-events discovered in the T-10 plasmas in experiments on simultaneous co+counter axial EC current drive performed using the X-wave with a frequency equal to the second ECR harmonic. In these experiments, the tungsten limiter and lithium wall coating were used. The ionization of atoms in the Li-containing flake, which occurs at the plasma edge, generally causes the rise of the electron temperature and density in the core of the plasma column. The electron heat flux reduces abruptly within the radial range of 0.2 < r/a < 0.8, just as it was at the LHD in experiments on the injection of small C8H8 pellets. The energy confinement time abruptly increases by ∼10%. After the transient phase, the ITB appears, which is located between the radii corresponding to the safety factors of q = 1 and q ≈ 1.4. Typically, the ITB lifetime is equal to the plasma energy lifetime. As far as we know, the triggering of ITB-events by the drop of Li-containing flake has been never reported in tokamaks and stellarators.

Controlled neoclassical tearing mode (NTM) healing by fueling pellets and its impact on electron cyclotron current drive requirements for complete NTM stabilization

Controlled partial stabilization of core m/n = 2/1 neoclassical tearing modes (NTMs) by fueling deuterium pellets is reported in DIII-D and KSTAR H-mode plasmas (m/n are the poloidal/toroidal mode numbers). Analyses of DIII-D data exploring possible physics origins show that an explanation is offered by NTM-turbulence multi-scale interaction, triggered by a sudden increase of local gradients near q = 2 caused by the pellet. Pellet injection from the high-field side allows deep fueling which reaches the island region. In turn, low-k turbulent density fluctuations () increase by in the island region. This can drive transport across the island separatrices, reducing the pressure flat spot at the O-point and diminishing the NTM drive. The Mirnov probe array detects the reduction of the 2/1 magnetic amplitude by up to . Causality between elevated gradients outside of the island, turbulence spreading into the island and reduced NTM drive is qualitatively supported by non-linear gyrokinetic turbulence simulations. These show increased penetration of ion-scale from the background plasma to the O-point region when the background gradient is increased. This interaction has potentially far reaching consequences as it can lead to a reduction of the required electron cyclotron current density () for NTM suppression by 70%, as predicted by the modified Rutherford equation. This beneficial effect of fueling pellets can be important as is the anticipated active NTM control technique for ITER, but its efficiency will be lowered by third harmonic absorption in Pre-Fusion Power Operation-1 (PFPO-1) at half magnetic field.

Predictions of Plasma Behavior Due to Pellet Injection for Future Thailand Tokamak

Predictions of Plasma Behavior Due to Pellet Injection for Future Thailand Tokamak

Time-dependent runaway electron simulations: Ampere–Faraday equations implemented in CQL3D

The runaway electron (RE) distributions driven by a large toroidal electric field induced by the drop in the temperature profile due to disruption or pellets are comprehensively simulated by the 3D Fokker–Planck (FP) solver CQL3D (Harvey and McCoy 1992 Proc. of IAEA TCM), recently coupled to the Ampere–Faraday (AF) equations. The evolution of the toroidal current in a plasma occurs on a resistive time scale, τres = 4πa2/(c2η), which is typically of the order of seconds in present tokamaks. Here, a and η are respectively plasma radius or radial extent of a current density perturbation, and Ohmic resistivity. From the Faraday EM equation, the toroidal electric field is proportional to the time derivative of the poloidal magnetic field, which, from the Ampere equation, is proportional to the toroidal current. Thus, the toroidal electric field rapidly increases due to an abrupt temperature drop decrease in conductivity, to prevent change in the toroidal current faster than τres. This is a example of Lenz’s law. For example, in simulations with KPRAD (Whyte et al 2003 J. Nucl. Mater. 313–6 1239) of neon pellet injection into a DIII-D shot, Te drops from 2 keV to 10 eV in 0.1 ms and Zeff increases 1–4, giving that the electric field increases 3500× to 0.8 V cm−1. As described in Harvey et al (2000 PoP 7 4590), this places much of the tail electron distribution beyond the Dreicer runaway velocity, giving so-called ‘hot-tail runaways’ which for a time are the dominant source of runaways, more so than the knockon source. In this prior calculation, performed for a single flux surface, the toroidal current density is held constant, on the basis that τres is large. Most of the initial current can be converted to runaway current, which is then dangerous, particularly for ITER. A more comprehensive A–F model recently implemented in CQL3D, taking into account the time-development of the full-plasma-width toroidal electric field on time-scales of order τres applies an iterative technique for the toroidal field previously developed for a different application (Kupfer et al 1996 PoP 3 3644), maintaining the implicit-in-time evolution of CQL3D. The degree of runaway current formation is reduced in AF augmented CQL3D, but the basic mechanism of ‘hot-tail runaways’ remains a dominant contribution to the REs at early times after the Te drop in these simulations. On the other hand, a NIMROD (Sovinec et al 2004 J. Comput. Phys. 195 355) simulation of shattered-pellet shutdown of DIII-D plasma (Kim 2018 APS/DPP Meeting), gives a slower thermal quench; when the plasma profiles and electric field are coupled one-way to CQL3D, the ‘hot-tail’ REs are much less, and growth of RE is dominated by the knockon process.