Efficient plasma fuelling, impurity exhaust and power load control are essential for the successful operation of fusion reactors and have a direct impact on the fuel cycle and the achievable tritium burn-up fraction.
The physics of plasma fuelling in fusion devices and the processes determining the balance of particles in the reactor chamber are described, with emphasis on the core and edge...
The regulation of the amount of fusion power produced by future reactors will require precise control over the plasma density and temperature. Therefore, the control of the core-plasma kinetic state, usually referred to as burn control, arises as one of the most fundamental problems in nuclear fusion and will be critical to the success of burning-plasma devices like ITER. Due to the nonlinear...
Benign power exhaust in a tokamak relies on the injection of radiating impurities (plasma enhancement gases), to be controlled for the achievement of the desired power crossing the separatrix and sufficient divertor radiation. At least partial divertor detachment is required, reducing the heat flux impinging the divertor target at the separatrix below about 5 MW/m$^2$. Candidate gases for...
Deuterium (D)-Tritium (T) plasmas are considered the most promising hydrogen isotope combination for the generation of fusion energy in tokamaks. However, in contrast to electron particle transport, ion transport in mixed plasmas, notably in the presence of T, is less understood.
Differences in the electron density behavior clearly indicated in the first DT campaigns in TFTR and JET that...
In a fusion reactor plasma, impurities are present due to multiple sources. In the center, He ash is produced by fusion reactions, at the edge plasma facing components can release impurity atoms and impurities are actively seeded to reach a tolerable heat exhaust. The consequent impurity density profiles will be the result of a combination of the strength of the sources and of the transport,...
Plasma-material interaction (PMI) imposes a number of challenges on the operation of a next step fusion device or reactor associated with the lifetime of components, the sustainability of the tritium cycle, and ultimately with safety aspects. The underlying critical processes under steady-state plasma operation can be splitted into two categories: (i) erosion, transport, deposition, and dust...
Tritium (T) retention in plasma facing components (PFCs) subjected to burning plasma-material interactions (BPMI), defined here as simultaneous plasma exposure and 14 MeV neutron irradiation at reactor-relevant temperatures, will materially impact the in-vessel T inventory, achievable tritium breeding ratio (TBR), and performance limits of PFCs in fusion pilot plants (FPPs). Validated model...
A critical challenge for the long-term operation of ITER and the future U.S. fusion pilot plants will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to intense heat and neutral/ion particle fluxes under the extreme fusion nuclear environment while minimizing in-vessel inventories and ex-vessel permeation of tritium.
INL leverages a series of...
A self-sufficient fuel cycle is a significant contributor to enable commercial fusion energy. It also puts additional requirements on existing fuel cycle concepts. Driven by the need to reduce the tritium inventory in the systems to an absolute minimum, the work package TFV (Tritium – Matter Injection – Vacuum) of the European Fusion Programme has developed a three-loop fuel cycle...
A future fusion reactor is anticipated to mainly utilize pellet injection for particle fuelling. Pellets are mm-sized bodies formed from solid hydrogen fuel. Undoubtedly, delivering an adequate fuel amount with an isotope mixture adjusted to establish the optimum deuterium:tritium composition expected in the vicinity of D:T = 1:1 to the plasma core has to lay the foundation for any pellet...
The divertor system of a fusion device is always a compromise which has to meet power exhaust, particle exhaust and neutron shielding requirements at the same time. The design space of the tokamak particle exhaust function results from a number of requirements, such as geometrical parameters (for instance the divertor cassette configuration, and the position of the pumping port relative to the...
Present and planned fusion machines rely heavily on the use of neutral beam injectors, to provide plasma heating and current drive. In the case of large experiment, like ITER and beyond, the atoms injected in the plasma require a high energy (>500 keV) to penetrate the dense and large plasma and deliver the power at plasma center. This calls for the use of negative ions as precursors of the...
The deuterium/tritium fuel cycle for fusion reactors is linked to how the reactor is designed and operated. Sometimes these links are obvious such as the choice of seeding gases and burn fraction, though others are subtle or not obvious. This paper presents an introduction to the key considerations stemming from physics decisions that impact the design and operation of the fuel cycle.
Future deuterium-tritium fueled fusion power plants must breed tritium and sustain a burning plasma using a semi-closed loop fuel cycle. The DT fusion fuel cycle is an important aspect of any fusion energy configuration whose purpose is to provide fuel to the plasma, pump and separate plasma exhaust products, and recover fuel from breeding and plasma exhaust products. The current method of...
The fuel cycle of the European DEMO reactor comprises three loops, where the first two – the Direct Internal Recycling (DIRL) and the Inner Tritium Plant (INTL) Loop – are directly coupled to the reactor. These loops include components that act as actuators on the plasma, as vacuum pumps, pellet injectors or gas injection valves that can and must be controlled on a given timescale.
This...
The minimization of the Fuel Cycle inventory in a pulsed DEMO-like power reactor can be provided in the Direct Internal Recycling (DIR) concept by adding an additional short-cut between the pumped torus exhaust gas and the fuelling systems [1]. The Fuel Cycle which includes plasma fuelling and exhaust, as well as several exhaust processing and isotope separation processes, is one of the key...
Recently, as many countries are developing their demonstration fusion plant, Korea has also begun developing the Korean-style demonstration fusion plant. For the fuel cycle design, the handling of a large amount of tritium is an essential problem to be solved. Several activities related to process modeling and simulation are in progress for process design optimization and tritium inventory...
Significant production of radioactive metal dust, mainly due to erosion of plasma facing components, is expected to be present in the vacuum vessel of DEMO while it is operating. A relevant tritium content is expected to be present in this dust. Therefore, the removal of tritium and the management of this radioactive dust for safety reasons and for eventual remanufacturing or appropriate...
When a plasma experiment using deuterium (D) gas is conducted in a large fusion test device, a small amount of tritium is produced in the plasma. The produced tritium can be used to evaluate tritium behavior and inventory in fusion systems as a tracer because of its small amount. As one of the large fusion test devices, the deuterium plasma experiment with the Large Helical Device (LHD) have...
DTT (Divertor Tokamak Test Facility) is a new facility, currently under build, in which various scaled experiments for testing different magnetic configurations and alternative solutions for the power exhaust system of DEMO will be performed. Although the divertor system is not finalized yet, the machine and port geometry set limitations on the divertor pumping system operational space. In the...
The Diagnostic Residual Gas Analyzer (DRGA), an integrated, multi-sensor diagnostic system, will access and sample the ITER sub-divertor region, in the ducts of the cryogenic pumps, out-of-site of the main plasma chamber. It will deliver time resolved neutral gas composition measurements directly related to fuel cycle processes in the core plasma, in plasma-wall equilibration timescales [1,...
A future fusion reactor based on the tokamak concept in particular will need to employ methods to mitigate both edge localized modes (ELMs) and disruptions. Both of these unwanted plasma events can lead to high heat fluxes that can damage internal plasma facing components. The mitigation of these events in the plasma relies on the injection of material into the plasma to trigger a plasma...
There is a strong relationship between fusion machine physics characteristics and the associated fuel cycle systems. As these grow in size and features, so may the facility hazards and the need for hazard mitigation. This talk will introduce the consideration of the need for hazard mitigation alongside the physics/fuel cycle relationship.
The fuel cycle of future demonstration and fusion power plants is a complex and highly dynamic system by nature, resulting from the pulsed operation of the tokamak as well as from a number of cyclic operations employed within its processing systems. The fuel cycle nevertheless has to guarantee the availability of fuel in the right quantities and composition to the plasma fueling systems, while...
Mechanisms underlying tritium retention in chamber materials can be roughly divided into two groups: trapping in deposition layers and that in bulk of materials. The contribution of trapping in the bulk to the total tritium retention could be larger in DEMO than that in existing fusion devices due to far longer discharge pulse that allows diffusion of tritium into deeper region of the...
In fusion DEMOs, tritium (T) decontamination scenario before maintenance begins is a key issue. Hence, it is important that T decontamination under vacuum conditions before opening the plasma vacuum vessels. Currently, JA-DEMO team has not yet determined the allowable value of residual T in the vacuum vessel, but it is necessary to indicate a candidate T decontamination technique. Furthermore,...
JET is the largest tokamak in use and currently the only one capable of handling radioactive tritium (T). It operates since 2011 with the ITER-like wall (ILW), which consists of a tungsten (W) divertor and a beryllium (Be) main chamber. Following preparatory campaigns in deuterium (D), hydrogen (H) then T, JET has operated the second Deuterium Tritium Experimental campaign (DTE2, after DTE1 in...
The tritium self-sufficiency is one of the most challenging feasibility and attractiveness issues in the development of fusion systems, which is also the main science mission of CFETR. The tritium burning rate is the key which can be improved by the increased particle confinement time and central fueling. Our study has been engaged in the research of central fueling technology and particle...
