FAQs

Physicists have been exploring the properties of plasmas within tokamak devices since the 1960s. It is well known that beyond certain operational boundary conditions—for example, when plasma current, pressure or density rises too high for a given magnetic field—the plasma can become unstable.

A disruption is an instability that may develop within the tokamak plasma. Disruptions lead to the degradation or loss of the magnetic confinement of the plasma, and because of the high amount of energy contained within the plasma, the loss of confinement during a disruption can cause a significant thermal loading of in-vessel components together with high mechanical strains on the in-vessel components, the vacuum vessel and the coils in the tokamak.

In some cases, because of the large electric fields created during the disruptions, a relativistic electron beam (containing "runaway electrons") forms that can penetrate several millimetres into the in-vessel components when it is eventually lost from the plasma.

Unless mitigating action is taken, plasma-facing components can suffer local damage due to the thermal loads and to the deposition of runaway electrons during disruptions. In addition, in extreme cases, the mechanical strains on the components during disruptions may cause some deformation.

Disruptions are not triggered randomly; they only occur when well-defined limits are exceeded. Disruptions have been observed, avoided and mitigated in most operating tokamaks. One of ITER's objectives is to perfect a stable operating scenario through experimentation so that disruptions become a relatively rare event. During the first years of operation, ITER operators will most likely deliberately provoke disruptive events. Their aim will be to analyze, and to learn to control, these events at reduced plasma parameters and low plasma energy so that disruptions cannot cause damage to the ITER components in experiments at the highest plasma current and energy. 

By "pushing" the machine toward disruptions at modest plasma parameters, ITER operators will find its stability boundaries. Once these stability limits have been identified, there is no reason for plasmas in the ITER Tokamak to become disruptive spontaneously as the plasma current and plasma energy is increased, provided that this is done within the stability region identified.

There is abundant literature on the subject of disruptions (see, in particular, Nuclear Fusion) and on the operational strategies to avoid disruptions and to mitigate their effects when they cannot be avoided.

Disruptions are an integral part of the official (and public) physics basis for ITER, which has been extensively refereed by the scientific community ("ITER Physics Basis," Nuclear Fusion, 47; 2007 complemented the initial 1999 report). Disruptions represent an active field of research in the fusion community in order to perfect the avoidance and mitigation schemes being developed for ITER.

The European tokamak JET and the French tokamak Tore Supra, as well as many others in the world, have been operated in a completely safe and satisfactory manner since 1983 and 1988 respectively. When exploring new plasma regimes, or during dedicated experiments to study disruptions and their mitigation, disruptions can occur several times a day in these two machines and others, but they have never led to the destruction or rupture of their vacuum vessels.

Because disruptions are expected in ITER, they have been planned for. The ITER vacuum vessel and in-vessel components have been designed to withstand the forces produced by about 3,000 disruptions at full plasma performance over the course of their lifetime. ITER's resistance to disruptions is based on scaling laws ("engineering laws") that have determined the values chosen for ITER; these values have been validated by experiments on other tokamaks.

It is important to understand that disruptions are not a safety-class issue for ITER: there is absolutely no risk for the integrity of the vacuum vessel. But as the high energy loads during disruptions can, over time, damage the surface of plasma-facing components such as divertor targets and first wall panels, these components may need—and have been designed—to be replaced. This takes time and reduces the availability of ITER for experiments. It is therefore important to develop disruption mitigation techniques that reduce the forces and the energy loads on ITER's components so that the time between interventions to replace these components is as long as possible, thereby optimizing the scientific exploitation of ITER.  

During the progressive commissioning of ITER, the machine will be tested with plasma currents and plasma energies lower than the nominal values required for fusion energy production. In this way, the potential degradation of ITER's components by disruptions during this initial learning phase will be minimized. We will begin with low current and low-energy plasmas to learn how to avoid and mitigate the effects of disruptions on ITER before moving on to more advanced operational scenarios with higher currents and higher energies (thus larger forces and energy loads on components).

The ITER strategy is not radically different from that already followed in the operation of the largest existing tokamak JET, which achieved plasma currents of 6-7 MA (as compared to the 15 MA nominal plasma current planned in ITER).

In summary, the ITER engineering design allows for disruptions to occur in approximately 10 percent of plasma pulses. The early, low-energy/low-plasma-current phase of ITER will permit physicists to characterize disruptions on ITER without risks to the machine. Disruption mitigation is one of the specific scientific missions of ITER, with direct relevance to the future development of fusion power plants based on the tokamak concept.

ITER's Disruption Mitigation System (DMS) passed its final design review in March 2024 and has been greenlit to move forward to manufacturing. 

Two promising methods were investigated for ITER scenarios. Following years of development work, shattered pellet injection—in which massive amounts (up to 500 g) of particles are introduced into the plasma within 10 milliseconds—has been chosen as the baseline technique to disperse the energy of a disruption before it can concentrate its load on the wall of the containment vessel. In determining the best method, or combination of methods, for disruption mitigation, the ITER Organization took performance, reliability, flexibility, and cost into account. 

An international task force established in 2018 under the leadership of the ITER Organization carried out an extensive program to refine system design specifications and to perform engineering work for the industrialization of the technology. One sub-group, Experiments, focused on validating the ITER shattered pellet injection scheme on operating tokamaks such as ASDEX Upgrade (Germany), KSTAR (Korea), Tore Supra (France), DIII-D (USA), and JET (EU). Another, Technology, focused on improving the technique itself and making it fit for the demanding environment in ITER. Finally, the Theory & Modelling sub-group worked to interpret the experiments and extrapolation to ITER.

The Disruption Mitigation System in ITER will function automatically, triggered as disruptions occur during plasma pulses by dedicated sensors and algorithms that can evaluate the likelihood of an impending disruption. With at least 10 pulses planned per day during operational phases, and disruptions expected in approximately 10 percent of these, it is accurate to say that the Disruption Mitigation System will operate routinely—probably daily—during operation, at least during the initial phases as the ITER operational scenarios are being developed.