The ITER Disruption Mitigation System (DMS) utilizes Shattered Pellet Injection (SPI) technology for the injection of a high amount of cryogenic material into the plasma. The ITER DMS Support Laboratory at the HUN-REN Centre for Energy Research studies pellet formation, launch and shattering using ITER-size pellet. The aim of this paper is to study the propagation of cryogenic dust and propellant gas around the pellet.Any propellant gas or debris getting into the plasma ahead of the shattered pellet might compromise the mitigation efficiency due to reduced material assimilation caused by premature cooling of the plasma.With a schlieren diagnostic, a solid hydrogen dust cloud and debris was observed to arrive before the pellet, in addition to shock waves, which is a direct evidence of propellant gas flow. A supersonic Mach-number of the gas flow was be calculated from the shockwaves.Dimension reduction on the motion picture data can depict the entire launch process in one panorama image, which can be further reduced to a single time series for each pellet launch, allowing statistical analysis on two types of pellets, produced with different freezing recipes. The time evolution of pellet launches shows a general behavior regardless of pellet velocity, material, and freezing recipe. Less debris arrives before pellets prepared with lower initial barrel pressures. The experimental proof of the gas flow, dust and debris ahead of the pellet delivered critical information for the ITER DMS, highlighting the capabilities of schlieren diagnostic.

Shattered pellet injection (SPI), currently the most effective method of disruption mitigation, is currently implemented on tokamaks worldwide for experimental purposes. Cryogenic pellets are formed and fired into an angled surface before entering the plasma. The impact with the angled surface causes the pellets to fragment into a cloud of particles with the purpose of increasing the surface area for ablation. As pellets traverse guide tubes, depending on design, there is a chance of an off-normal pellet impact. Pellet impacts are also, depending on design, possible in the plasma chamber if not fully ablated and assimilated, or if the fragment plume is not directed in the proper direction. This paper outlines a series of pellet impact tests on various tiles and components relevant to the ITER, JET, and ASDEX Upgrade SPI systems. Testing was done to assess the potential damage from pellet and fragment plume impacts through high-speed imaging and the visual inspection of components.

The shattered pellet injection (SPI) method has been chosen as the disruption mitigation system (DMS) for ITER. To protect the device from plasma disruptions that cause damaging heat and electromagnetic loads, SPI is used to inject high-Z material into the plasma. The process of SPI utilizes cryogenic cooling to form solid pellets. Pellets are accelerated down a barrel and into an angled surface, causing the pellet to shatter prior to entering the tokamak chamber. For the DMS to function reliably, the 27 separate shattered pellet injectors planned for ITER must rely on many components to provide accurate feedback data and for control functions.Each component in the DMS is exposed to an elevated background magnetic field depending on its placement and proximity to the plasma chamber. A Helmholtz coil test stand that is operated at Oak Ridge National Laboratory was utilized to test the components in relevant background field levels to assess component performance. This paper details the test design and results for in-field component operation for a variety of components. This list includes the following components: two different network switches for camera connectivity, a VAT fast shutter valve intended to reduce the flow of SPI propellant gas into the torus, a solenoid control valve intended for use in the pellet formation process, pressure/vacuum switches to be used for feedback and control, a printed circuit board piezo pressure sensor to be used to measure breech pressure, and various relays for the high-voltage pulsed power supply used to drive the SPI propellant valve.

The thermal and mechanical loads during disruptions are a major threat for large reactor-class tokamak devices. Therefore, shattered pellet injection (SPI) is selected as the baseline technology for the ITER Disruption Mitigation System (DMS). The aim of DMS Support Laboratory located at the HUN-REN Centre for Energy Research is to study the production, launch and shattering of cryogenic protium, deuterium, and neon pellets in the ITER geometry. This paper reports on the fragment analysis procedure and the first results of the fragment plume investigation performed with pellets made of the above materials and accelerated to a speed ranging between 70 and 500m/s. The experimental results show that the fragment plume consists of macroscopic and microscopic fragments. The shattering of about 500 m/s protium and deuterium pellets resulted in a most common fragment size of a few millimeters, while at lower velocities around 250m/s still the same few millimeters size fragments dominate but larger fragments have a higher proportion in the mass distribution. At low velocities (70 m/s) for neon pellets most of the pellet mass was converted into large fragments (from 10 to 20 mm ). The results are also compared with the Parks pellet fragmentation model.

A support laboratory has been set up to study pellet production, launch and shattering of cryogenic protium, deuterium, and neon pellets for the ITER disruption mitigation system, which plans to use 28.5 × 57 mm (diameter × length) protium, neon an mixture pellets in the Shattered Pellet Injectors. Such large protium pellets have not been produced and launched before, therefore the desublimation and launch process have been studied in detail in two steps. First 19 mm diameter pellets were produced, followed by the demonstration of the final pellet size. Pellet desublimation recipes were established for all pellet types, and it was found that, under certain conditions, even the large neon pellets can be launched with a propellant gas pulse, without requiring a mechanical punch device. This is attributed to cryogenic snow formation on the surface of the pellet. Conditions for the snow formation are studied and tendencies are understood using simple calculations.

The ITER Diagnostic Residual Gas Analyzer (DRGA) will measure the distribution of gas species, i.e., deuterium (D), tritium (T), and impurities, in the divertor exhaust stream and in the plasma periphery, with time resolution relevant to fusion plasma–wall particle dynamics. The uniqueness of the DRGA, over previous implementations of plasma dynamics residual gas analysis, is an integrated approach, combining mass and low-temperature plasma-activated optical spectroscopy, in a differentially pumped analysis station. A further unique feature of the ITER divertor-specific DRGA is an ~8-m separation of the analysis station from the sampled pumping duct, while still maintaining a ~1-s response time for hydrogen isotopic concentrations. ITER DRGA final design activities are strongly benefiting from testing of prototypical DRGA components and methods on present fusion devices, most currently on JET and W7-X. DRGA systems are implemented on both these devices and include sensors (and pumping methods) that are directly relevant to the ITER DRGA design. The recent JET-DTE2 campaign has provided the first experience on operating the combined ITER DRGA sensors with D-T plasmas. While enhancing system design for ITER, this experience on operating devices has also revealed additional engineering challenges, which further guide the continuing final design project. Meanwhile, the recent determination that the ITER DRGA, with slight optimization, will resolve the helium isotopes well enough to support an ITER pre-DT, He-3-based heating scheme, has greatly increased ITER Research Program interest in the DRGA and its implementation well ahead of the DT phase.

Shattered pellet injection (SPI) is the technology chosen for the ITER Disruption Mitigation System and is explored at several fusion research devices, like DIII-D and JET and J-TEXT. The ITER disruption mitigation strategy relies on multiple injections to achieve RE (runaway electron) avoidance with optimum TQ (thermal quench), CQ (current quench) durations to adequately reduce wall loads. To demonstrate the feasibility of the multiple injection and to extrapolate to ITER, experiments with two identical injectors toroidally opposite to each other are needed urgently. KSTAR (Korea Superconducting Tokamak Advanced Research) can be a unique testbed to study the plasma disruption mitigation for ITER. KSTAR has installed two identical injectors in 180 degrees of toroidal opposite positions in 2019. For this system, ORNL (Oak Ridge National Laboratory) provided the two injectors, the shatter tubes, and auxiliary systems. NFRI (National Fusion Research Institute) provided the infrastructure of a vacuum pumping system, control & data acquisition system, and installed additional diagnostic systems for SPI in collaboration with the ITER Organization. This paper describes the engineering achievements during installation on KSTAR and the initial results of single and multiple SPI experiments in the 2019 campaign.

ITER is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. Rapid progress is being made in project construction, and the facility is now taking shape at St-Paul-lez-Durance in southern France. In the course of designing and manufacturing of the systems making up the ITER tokamak and the ITER facility, extensive ground-breaking R&D has been implemented by the ITER partners across a wide range of technology and science areas which underpin the achievement of the project’s engineering and fusion plasma performance requirements. Significant developments have been made in the production of high performance Nb 3 Sn superconducting strand and in magnet technologies supporting the construction of the largest superconducting magnets produced to date. High heat flux plasma facing components have been fabricated which are capable of sustaining quasi-stationary heat loads of up to 10 MW m −2 and transient loads of up to 20 MW m −2 . Fusion nuclear technologies such as remote maintenance and tritium breeding have received specific emphasis within the ITER R&D program, since extensive deployment of these technologies is foreseen. Diagnostic systems face particular challenges in the ITER environment, and wide-ranging R&D activities have been implemented to develop novel solutions to ensure an adequate measurement capability in ITER DT operation. Routine and reliable operation in ITER will require a highly effective capability for the detection, avoidance and mitigation of disruptions, and significant science and technology R&D is underway to establish this capability. The overall integration of the control requirements for the ITER plasma and facility, in particular during burning plasma operation, has presented new challenges for fusion control systems, including the need for robust safety and hardware (investment) protection. These challenges are being addressed via the implementation of the most extensive and ambitious control system to date. The paper introduces the ITER project and its major goals in relation to the development of fusion energy and provides an overview of key innovations which have been made in these areas of fusion technology and science in support of ITER construction.

During the ITER plasma operation state, in order to prevent explosion or contamination risk, leaks of the gas supply pipes in the GIS gas fueling manifold especially the hydrogen isotope pipes shall be detected. To detect the leaks of the gas supply pipes, methodology of pressure monitoring and gas composition monitoring are employed, basing on the design of pressure cascade and interspace venting. By comparing the pressure signal of the gas supply pipe and the interspace, leaks of the gas supply pipe could be identified. By setting reasonable flow rate for interspace venting, at the same time monitoring the hydrogen isotope concentration in the venting flow, leaks of the hydrogen isotope pipes could be identified. Pressure alarm and H2/T2 alarm are set respectively, and leak of different dimension is identified.

Nitrogen seeding, necessary for divertor heat-load mitigation in ITER, has been shown to lead to ammonia formation which would be a severe operational and safety issue in ITER. Predictions of ammonia production in ITER are based on data from present day fusion devices. Ammonia is mainly detected by residual gas analysis (RGA). Detection of ammonia is impeded by the presence of water and methane which, in a mixed H-D system, leave signatures in the same range of the mass spectra. A statistical model is used to ascribe an average isotope ratio to each gaseous species. The model is tested with simulated RGA recordings with varying concentration of ammonia to evaluate the sensitivity to fitting parameter boundaries, noise in the recordings and mis-matching cracking patterns. The analysis shows that the fitting procedure may in some occasions substitute species among each other, resulting in faulty concentrations. Nevertheless, the right choice of parameter boundaries ensures correct fitting results. Finally, the fitting procedure is applied to experimental data from nitrogen-seeeded discharges at AUG and JET. (C) 2017 Published by Elsevier B.V.

Disruptions, the fast accidental losses of plasma current and stored energy in tokamaks, represent a significant risk to the mechanical structure as well as the plasma facing components of reactor-scale fusion facilities like ITER. At JET, the tokamak experiment closest to ITER in terms of operating parameters and size, massive gas injection has been established as a disruption mitigation method. As a "last resort" measure it reduces thermal and electromagnetic loads during disruptions which can potentially have a serious impact on the beryllium and tungsten plasma-facing materials of the main chamber and divertor. For the planned deuterium-tritium experiments, a new Disruption Mitigation System (DMS) has been designed and installed and is presented in this article. The new DMS at JET consists of an all metal gate valve compatible with gas injections, a fast high pressure eddy current driven valve, a high voltage power supply and a gas handling system providing six supply lines for pure and mixed noble and flammable gases (Ar, Ne, Kr, D-2, etc.). The valve throughput varies with the injection pressure and gas type (efficiency - injected/charged gas 50-97%); the maximum injected amount of gas is approximately 4.6 kPa m(3) (at maximum system pressure of 5.0 MPa). (C) 2015 Published by Elsevier B.V.

A new Disruption Mitigation System (DMS) based on Massive Gas Injections (MGI) has been installed at the JET-tokamak. The key component of this system is a fast eddy current driven valve, which is capable of injecting up to 4.6 x 10(-3) MPa m(3) in less than 5 ms. Along with this valve a new gas handling system has been installed, whose control had to be integrated into the JET-operation. The operation of the DMS requires interaction with several other systems. Although Massive Gas Injections are used to ameliorate potentially severe damage to the tokamak plant and plasma facing components caused by disruptions, they introduce a high risk for example to auxiliary heating systems or diagnostics, which could be damaged by high vacuum pressures. In addition to this, the presence of high pressure (of noble and flammable gases) in combination with high voltages represents a risk not only to the actual DMS plant itself (in case of a failure) but also to personnel in the vicinity. These varieties of risks have been addressed and are described in this article. (C) 2015 Elsevier B.V. All rights reserved.

Disruptions are a major concern for next-generation tokamaks, including ITER. Heat loads, electromagnetic forces and runaway electrons generated by disruptions have to be mitigated for a reliable operation of future machines. Massive gas injection is one of the methods proposed for disruption mitigation. This article reports the first use of massive gas injection as an active disruption protection system at JET. During the 2011–2012 campaigns, 67 disruptions have been mitigated by the disruption mitigation valve (DMV) following a detection by mode lock amplitude and loop voltage changes. Most of disruptions where the valve was intended to be used were successfully mitigated by the DMV, although at different stages of the typical slow disruptions of the ITER-like wall. The fraction of magnetic and thermal energy radiated during the disruption was found to be increased by the action of the DMV. Vertical forces dispersion was also reduced. No non-sustained breakdown was observed following pulses terminated by the disruption mitigation valve.

With installation of the ITER-like wall in JET a major diagnostic upgrade to measure the neutral gas pressure and composition in the sub-divertor region has been completed, to characterise retention and outgassing of the new metallic first wall. The upgrade includes two new magnetically shielded systems consisting of sensitive capacitance manometers and residual gas analysers, both capable of providing data during plasma operation. These enable absolute pressure and gas composition measurements (pressure range: 10(-5)-10(-1) mbar, mass range: 1-200 amu, respectively) and have been used to characterise the neutral gas behaviour under various plasma conditions.

Disruptions are a critical issue for large scale tokamaks due to the potential damage to plasma facing components. Massive Gas Injection (MGI) is considered as a ‘last resort’ method for disruption mitigation. A MGI system based on the Disruption Mitigation Valve (DMV) has been brought into operation at JET. Injections of neon, argon and its mixtures with deuterium show distinct effects on the machine condition during and after MGI-induced disruptions. MGI with pure argon shows a continuous accumulation in consecutive pulses. Neon on the contrary shows a fast saturation due to trapping in carbon PFCs.