A cryopump can be utilized as an impurity removal component of a direct internal recirculation (DIR) system for the fusion fuel cycle. The DIR facilitates a low fuel inventory by continuously pumping unburnt fuel while removing impurities from the fusion exhaust stream. A cryopump can target multiple impurity species by maintaining a temperature lower than the gas triple-point temperature that promotes desublimation. The desublimation/condensation of gases in cryopumps can be characterized by the sticking coefficient, which is defined as the probability for a gas particle to stick to a (cryo-)surface upon collision. The sticking coefficient is one of the important design/operation parameters for cryopumps, and it depends on a variety of surface and gas properties.In this study, molecular dynamics simulations were utilized to estimate the sticking coefficients of typical fusion gas impurity species N2, CO2, and CH4 over a Cu surface for a range of gas temperatures and surface coverages. The molecular dynamics study showed that the sticking coefficients for gases decrease with an increase in gas temperature. The presence of a single full monolayer of condensate on the metallic surface showed an adverse effect on the sticking of gases; however the sticking improved with two full monolayers of condensate on the surface. The sticking of gases over the mixed condensate on a surface was more favorable than the condensate of the same species for N2 and CH4, with an exception for CO2, which showed a decrease in sticking over the mixed condensate.

Cryopump-based direct internal recycling (DIR) of fusion fuel is an attractive prospect because it provides pumping of helium ash from a reactor along with separation of helium and other impurities from the fuel. Previous studies have demonstrated a continuously regenerating cryopump, referred to as the Snail pump, that separates helium ash from fusion fuel for reactor-relevant flow rates. In this study, a conceptual cryopump was designed as a proof-of-principle study to target other impurities, besides helium ash, from the fusion exhaust that would complement the Snail pump. The impurity-removal cryopump consists of four sets of chevron fins, two sets operating at 80 K and the other two at 30 K. A computational study using direct simulation Monte Carlo (DSMC) was performed with a gas mixture of 96.5% D2, 2.0% He, and 0.5% each of CO2, N2, and CH4. In this configuration, 80 K chevron fin sets are capable of capturing CO2 and 30 K sets capable of capturing CO2, N2, and CH4. The computational study showed that the cryopump is capable of reducing the impurity content by more than two orders of magnitude from the flow.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

Studies have been performed on the release mechanism for large pellets using high pressure gas in a shattered pellet injector. Typically, pellets are dislodged from the cryogenic surface and accelerated down a barrel using high pressure gas delivered by a fast-acting propellant valve. The pellets impact an angled surface which shatters the pellet into many small fragments before entering the plasma. This technique was initially demonstrated on DIII-D (Commaux et al 2016 Nucl. Fusion 56 046007) and is now deployed on JET, KSTAR, ASDEX-Upgrade, and other tokamaks around the world in support of ITER’s disruption mitigation system design and physics basis. The large hydrogen, 28.5 mm diameter, 2 length-to-diameter ratio, pellets foreseen for ITER SPI operation have low material strength and low heat of sublimation, which cause the pellets to be fragile and highly reactive to the impact of warm propellant gas. Due to the size of the pellets, significantly more propellant gas is required to dislodge and accelerate them. This creates a potentially significant propellant gas removal issue as 2–6 bar-L of gas is expected to be required for release and speed control. The research presented in this paper is an in-depth exploration of the parameters that are keys to reliable pellet release and speed control. Computational fluid dynamics (CFD) modeling of propellant flows through various breech designs was conducted to determine the force generated on the back surface of a pellet. These simulations assumed the use of the ORNL designed flyer plate valve. CFD modeling combined with experimental measurements provide adequate insight to determine a path to an optimal valve and breech design for ITER SPI pellet release and speed control while minimizing propellant gas usage.

Shattered pellet injection (SPI) experiments on Joint European Torus (JET) are an important element in determining the physics basis for mitigating disruptions in ITER. The initial design of the JET SPI system included three barrels to produce pellets with diameters of 4.5, 8.1, and 12.5 mm. The variability of the pellet speed by operating with and without a mechanical punch was limited and led to poor pellet integrity, so the mechanical punch was removed. Fragment size distribution is a function of pellet speed and the desire to change the resulting fragment size distribution was not originally a requirement. After the first set of SPI experiments on JET, different pellet sizes were considered to enable dual injection experiments with identical pellet diameters. It was also determined that speed control is necessary to improve experimental repeatability and to determine how the fragment size distribution impacts mitigation performance. Two barrels were fabricated to form pellets of 10 mm diameter and a third to form an 8.1 mm diameter pellet. New propellant valves were also fabricated and characterized to improve repeatability and overall performance. To have fine control of pellet speed, inserts to reduce the breech volume were fabricated and installed. Laboratory testing was conducted to ensure pellet release and provide a comparison of propellant gas delivered versus pellet speeds for a range of pellet types and mixtures. This article will also discuss how lessons learned from the JET SPI modifications can be applied to other SPI systems, such as the ITER SPI system, as controlling pellet release with the least amount of propellant gas is essential for optimal SPI effectiveness.

The dual shattered pellet injection (SPI) systems installed at the KSTAR (Korean Superconducting Tokamak Advanced Research) have been operational since 2019 to conduct disruption mitigation studies in support of ITER DMS (Disruption Mitigation System) design activities. The system has the advantage of firing pellets simultaneously from the two identical pellet injectors installed at 180-degree opposite positions on KSTAR. Each injector is capable of fabricating and firing three pellets of different sizes (diameter, length). The characteristics of the initial pellet (geometry, velocity, etc.) can be verified using a microwave cavity (MWC) system and an imaging optical pellet diagnostics (OPD) system.Each SPI system must achieve specific conditions related to pellet fabrication and firing prior to plasma experiments. We realized that we didn't have the sufficient time and conditions to do off-line testing in 2022, so we have developed a suitable solution. First, we have to disconnect the SPI system from the KSTAR vacuum vessel and build a separate test chamber in the injection line. Inside the new test chamber, we have to install the same shatter tube used in the current experiments and equip it with measuring devices to observe the fragmented pellets. We also have to switch the fabrication process of the pellets from manual to semi-automatic using a feedback pressure controller on the barrel pressure. In this paper, we will describe the new experimental environment for independent SPI testing and the future plan.

The mitigation of plasma disruptions in tokamaks has become a very important topic in magnetic fusion research, motived by the potential challenges that may occur in ITER disruptions due to the high magnetic field and high plasma current. Such disruptions can have a deleterious effect on the internal components due to the fast dissipation of the plasma thermal energy and the magnetic stored energy leading to large forces, as well as the possible formation of several megaamperes of energetic runaway electrons during the current quench. Oak Ridge National Laboratory has been developing and deploying technology to inject material into the plasma to rapidly radiate the thermal energy and start a fast plasma current ramp down to dissipate the magnetic stored energy. The choice of materials to inject and the injection technology have evolved over the past decades to arrive at the present systems planned for ITER based on cryogenic pellets of hydrogen-neon mixtures for thermal mitigation and hydrogen pellets for runaway electron mitigation. This scheme injects shattered cryogenic material into the plasma from pellets formed in situ in a pipe gun and fired onto angled metal surfaces at the end of the injection line just before entering the plasma.In this paper, we describe the evolution of schemes and technologies that have been employed for disruption mitigation and runaway electron prevention and dissipation, discuss how they have performed in present-day experiments, and give the outlook for the use of this technology in a burning plasma and how it may continue to evolve in the future.

Dual shattered pellet injection (SPI) systems are installed in KSTAR (Korea Superconducting Tokamak Advanced Research) to study disruption mitigation in support of the ITER DMS (Disruption Mitigation System) design activities. KSTAR has two identical pellet injectors that can shoot pellets simultaneously from opposite toroidal locations. Each injector can shoot up to of three different diameter pellets. The pellet integrity and velocity after formation have been measured using microwave cavity (MWC) systems since 2019. However, this diagnostic provides only limited information on the pellet integrity, especially for partially broken pellets. Therefore, an optical pellet diagnostic (OPD) system monitoring the pellet has been installed in the first stage viewport of the differential pumping system and is operational since 2021. The KSTAR SPI systems offer a wide parameter range that must be characterized before the experimental campaign. To satisfy the experimental requirements, a proper procedure for pellet formation and acceleration must be commissioned. We carry out the commissioning by varying pellet parameters such as formation duration, temperature, heating, propellant gas pressure, and pulse length. The resulting pellet velocity, integrity, and tilt can be measured with the OPD system. These tests can be performed without plasma operation because no additional information is needed from the plasma response to the injection. This paper describes the results of the characteristic tests of single and multiple injections of dual SPI.

A series of experiments have been executed at JET to assess the efficacy of the newly installed shattered pellet injection (SPI) system in mitigating the effects of disruptions. Issues, important for the ITER disruption mitigation system, such as thermal load mitigation, avoidance of runaway electron (RE) formation, radiation asymmetries during thermal quench mitigation, electromagnetic load control and RE energy dissipation have been addressed over a large parameter range. The efficiency of the mitigation has been examined for the various SPI injection strategies. The paper summarises the results from these JET SPI experiments and discusses their implications for the ITER disruption mitigation scheme.

Runaway electrons (REs) created during tokamak disruptions pose a threat to the reliable operation of future larger machines. Experiments using shattered pellet injection (SPI) have been carried out at the JET tokamak to investigate ways to prevent their generation or suppress them if avoidance is not sufficient. Avoidance is possible if the SPI contains a sufficiently low fraction of high-Z material, or if it is fired early in advance of a disruption prone to runaway generation. These results are consistent with previous similar findings obtained with Massive Gas Injection. Suppression of an already accelerated beam is not efficient using High-Z material, but deuterium leads to harmless terminations without heat loads. This effect is due to the combination of a large magnetohydrodynamic instability scattering REs on a large area and the absence of runaway regeneration during the subsequent current collapse thanks to the flushing of high-Z impurities from the runaway companion plasma. This effect also works in situations where the runaway beam moves upwards and undergoes scraping-off on the wall.

A high-voltage pulsed power supply (HVPPS) has been designed, prototyped, and tested for driving an eddy current actuated propellant valve for the International Thermonuclear Experimental Reactor (ITER) disruption mitigation system. The high-voltage (HV) dc supply output voltage is software programmable, and the energy storage capacitor bank can be readily reconfigured as 200, 400, 600, and 800 $\mu \text{F}$ , enabling testing and optimization of both the valve drive and valve systems. Multiple system parameters are monitored before, during, and after each firing of the valve. The system parameters are both displayed and stored for further analysis. Control of the setup, firing sequence, and data collection is automated using a LabVIEW-based control program. The programmability and reconfigurability of this system collectively provide a flexible and robust platform for system refinement and optimization. In this article, a summary of the system will be provided including operational sequences, HV switching and associated triggering methods and circuits, and results measured while firing a solid frozen pellet. In addition, planned refinement activities toward meeting all requirements for ITER integration will be discussed.

To fulfill the increasing needs of diagnostic support for researchers in plasma technology, a portable diagnostic package (PDP) equipped for both laser Thomson scattering (TS) and optical emission spectroscopy has been designed and constructed at Oak Ridge National Laboratory (ORNL), aiming to measure the temperature and number density of electrons and temperatures of ions in plasma devices. The PDP has been initially implemented on a high density and low temperature electrothermal arc source (ET-arc) at ORNL to test its TS capability. TS from the plasmas in the ET-arc has been obtained using the PDP. The electron temperature and number density were determined from TS spectra. These results were then compared to measurements from previous studies on the ET-arc. The TS diagnostic measured 0.8 ± 0.1, 1.3 ± 0.2, and 0.7 ± 0.1 eV and (4.4 ± 0.5) × 1021, (5.9 ± 0.7) × 1021, and (4.3 ± 0.5) x 1021 m-3, respectively, from three lines of sight that transect the plasma column.

Reliably mitigating disruptions is essential for ITER to meet its long-term operational research plan without damage to the in-vessel components. Currently, the shattered pellet injection (SPI) technique is the most effective radiator of thermal energy and has been chosen for the baseline disruption mitigation system (DMS) for ITER. The SPI process uses cryogenic temperatures to desublimate material into the barrel of a pipe gun forming a solid cylindrical pellet. Pellets for ITER will initially be hydrogen and hydrogen–neon mixtures. Once formed, pellets are dislodged and accelerated using high-pressure gas (40–60 bar) delivered by a fast-opening valve. The solenoid valves currently used for SPI experiments will not operate in an ITER environment due to the large background magnetic field. An ITER prototype fast-opening valve, called a flyer plate valve (FPV), has been designed and has undergone a wide range of testing. The FPV operates by pulsing current through a pancake coil that is closely coupled with a “flyer plate.” The flyer plate is an aluminum plate in which eddy currents are generated creating a repulsive force from the pancake coil. The force generated in the flyer plate rapidly lifts the valve tip off the seat and delivers a pulse of gas to the rear of the pellet, breaking it free from the barrel and accelerating the pellet downstream to its intended target. The design of the valve has been iterated on over the lifetime of this project, as the DMS for ITER shifted from massive gas injection (MGI) to SPI. The most recent design has been tested, and operational ranges have been mapped. The valve must survive 3000+ cycles in an ITER-like magnetic field. The principal functional requirement of this valve is to reliably dislodge and accelerate hydrogen (or H–Ne mixture) pellets into ITER. The valve was mated with an ITER SPI test stand and has been shown to be capable of launching pellets reliably. The valve and power supply design will be discussed in this article, along with the various testing setups used to determine the feasibility of this valve for use on ITER.

Cryogenic pellets are used for injection into fusion plasmas to add fuel to build up density and replace the ions lost from fusion reactions and imperfect confinement in the plasma. These pellets are formed at cryogenic temperatures with pure hydrogenic isotopes or mixtures of the isotopes. Technology to make these pellets and inject them into plasmas has been under development for many years, and various methods using freezing or desublimation have been shown to produce high-quality solid pellets suitable for injection. The throughput needed and possible impurity content from the necessary recirculation of fusion exhaust gases are two of the key issues to overcome for fusion pellet fueling systems in long-pulse burning plasmas. Here, we describe the technical challenges associated with these issues and the capability of pellet formation extruders to overcome them. Cryogenic pellets of deuterium, neon, and argon are also used in fusion tokamak devices for disruption mitigation in the form of large pellets that can be injected on demand to quickly dissipate the plasma thermal energy through radiation and add significant density in order to prevent runaway electron formation. Here, the issue is not throughput as with the fueling pellets but rather is the time it takes to form pellets of the size needed and the ability to dislodge them immediately on demand when needed to mitigate a disruption. The method used to make these pellets by desublimation is described, and examples related to how pellet size and input gas parameters affect the formation time are provided.

Digital holography has been proposed to fulfill a need for an imaging diagnostic capable of in situ monitoring of surface erosion caused by plasma-material interaction in nuclear fusion devices. A digital holography diagnostic for 3D surface erosion measurement has been developed at Oak Ridge National Laboratory with the goal of deployment on a plasma device. A proof-of-concept in situ demonstration is planned which would involve measurement of plasma erosion on targets exposed to an electrothermal arc source. This work presents the results of an ex situ characterization of the capability and limitations of holographic imaging of targets exposed to the arc source. Targets were designed to provide a fiducial for comparison of deformed and unaffected areas. The results indicated that the average net erosion was ∼150 nm/plasma exposure, which is expected to be within the diagnostic's measurement capacity. Surface roughness averages determined by holographic image analysis showed good agreement with measurements taken with a profilometer. The limit of the holography diagnostic's x-y spatial resolution was characterized by comparison with scanning electron microscope imaging.

One technique for mitigating disruptions in a tokamak is shattered pellet injection (SPI). SPI is a process in which a large solid pellet consisting of deuterium, neon, or argon is desublimated in a pipe gun barsrel and launched downstream. Pellets are shattered just before entering the plasma by an impact with an angled tube. Injection of these materials into the plasma radiates stored thermal energy, limits current decay rates, suppresses the generation of runaway electrons, and dissipates runaway electrons if necessary. A critical element of the SPI system is a fast-acting valve that releases high-pressure gas to dislodge and accelerate pellets directly, or indirectly via a mechanical punch. A prototype valve sized for the ITER SPI system has been designed and fabricated. A pulsed high-voltage power supply energizes the valve's internal magnetic coil, which induces eddy currents in the adjacent flyer plate resulting in a repulsive force between the flyer plate and the coil. The flyer plate action lifts a valve seat, allowing high-pressure gas to flow from the valve plenum to the downstream (breech) location of the pellet or mechanical punch. All of the valve's internal components are designed to operate in ITER-level static background magnetic fields. A study was conducted to optimize the downstream pressure response for a range of valve sizes and operating pressures. In particular, the study analyzes the breech pressure response associated with varying plenum pressures as well as varying breech volumes. A computational fluid dynamics simulation was built in STAR-CCM+ and validated against data from laboratory experiments. The resulting simulation outputs, in the form of downstream responses for a variety of initial plenum pressures and breech volumes, will be used as a complement to experimental data to ensure the pressure pulse is suitable for pellet survivability. These data, combined with studies on pellet shear strength and shock response, will be applied to optimization of overall operating parameters of the ITER SPI system.

Like for any power plant, refuelling will be one key issue in a future fusion reactor. It has to be safe and efficient; due to the operational principle, it also needs to occur in steady state. From previous investigations, it became clear that injection of pellets - mm-size solid bodies produced from frozen fuel ? is the only possible and technically realistic solution for this task. Beyond their capability for fuelling, pellets have been found potentially useful as well for other purposes, e.g. to control edge localised modes. Therefore, efforts are required to develop a pellet launching system (PLS) capable to meet all basic reactor requirements. To foster this kind of technology evolution and to enhance the PLS applied at ASDEX Upgrade (AUG), a collaborative investigation between IPP and ORNL has been initiated. This aims to replace the existing pellet source able to deliver to a centrifuge injector up to 130 stored pellets produced in a batch process with repetition rates up to 80 Hz by a steady state source. The new source is based on the ORNL extruder types and capable of delivering either in true steady state or at least from a reservoir with sufficient solid fuel to produce up to 2000 pellets. Experiments extruding ice with different square cross sections showed a maximum throughput of about 300 mm3/s for Protium (H) and 215 mm3/s for Deuterium (D) with a 1.9 mm square cross section. Ice temperatures of 9 K for H and 16 K for D were found to be optimum. For D, extrusion rates correspond to a 1.9 mm pellet delivery rate of about 25 Hz. Doping the ice with up to 2% Neon was achieved at the expense of moderately reduced extrusion speeds. A new scheme has been designed to inject pellets in parallel from multiple sources for further enhancing the delivery rate.

Shattered pellet injection (SPI) has been chosen as the baseline disruption mitigation system on ITER due to its ability to rapidly inject material deep into the plasma to greatly increase the plasma density and radiate the thermal energy. SPI utilizes a mechanical punch or high-pressure gas to release and accelerate a pellet that has been cryogenically desublimated in the barrel of a pipe gun. Various material injection combinations could possibly be implemented during different phases of a disruption event to radiate plasma energy, reduce electromagnetic loads on machine components, avoid the formation of runaway electrons, or to dissipate runaway electrons that form. Each injection phase could possibly utilize combinations of deuterium, neon, or argon. In this paper we outline experimental measurements of pellet material shear strength at SPI operating temperatures to understand the force needed to release SPI pellets. Deuterium, neon, argon, and deuterium-neon mixture pellets with diameters of 8.5, 12.5, and 15.7 mm are formed at a range of relevant gas pressures and temperatures and dislodged from the cold zone with a slow-moving piston driven by a motor. The slow-moving piston is kept above the triple point temperature of the material while the pellet is forming, then cooled to below the triple point temperature before contacting the pellet to minimize any thermal conduction to the pellet. The piston incorporates a load cell to measure the force applied when the pellet breaks away from the cold zone in the barrel. The ability of the gas and punch methods to exceed the shear strength of the studied pellet materials for release has been analyzed. High-pressure gas delivered by fast-opening valves produce pressure shock to the pellet due to supersonic expansion of the propellant gas. Pressure (and therefore, force) oscillations are present due to transverse density propagation throughout the breech volume. Mechanical punches deliver an impact force through a high-kinetic energy impact. The effect of the mechanical shock on the pellet has been explored and is presented in this paper. Scaling to larger ITER-size SPI pellets will be described.

Extensive investigations, within the EUROfusion Work Package "Tritium, Fuelling and Vacuum", indicate that sufficiently deep fuel deposition inside H-mode plasmas of the EU-DEMO tokamak requires injection of fuel pellets from the High Field Side (HFS) at speeds greater than or similar to 1 km/s. To implement this, two different approaches are being pursued: one makes use of "conventional" curved guide tubes, featuring large bend radii (greater than or similar to 6 m), to transport 1 km/s pellets to the HFS while trying to preserve their mass and integrity; the other explores the feasibility of injecting high-speed (greater than or similar to 2 km/s) pellets from the HFS, along "Direct-Line-of-Sight" (DLS) paths. This paper focuses on the latter approach. Recent tests with an existing ENEA-ORNL high-speed injector have confirmed that the trajectories of free-flight pellets, travelling under vacuum at speeds up to 2.4 km/s, spread within an angle less than or similar to 0.68 degrees. Despite their small scatter cone, free-flight pellets may require too much cut off volume of the Breeding Blanket (BB), due to the large distance between the injector and the plasma. The introduction of a straight DLS guiding tube transporting the high-speed pellets, to avoid significant loss of BB material, has been investigated. The existing ENEA-ORNL injector has been modified to accommodate a 10 mm i.d. DLS guide tube, and intact pellets have been consistently delivered downstream of the guide at speeds up to 2.6 km/s, with remarkably reduced scatter cone, thus showing the viability of this innovative approach.