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A conceptual design for a pellet injection system will be worked out, capable to support key missions of the new tokamak device JT-60SA. For exploitations in view of ITER and to resolve key physics and engineering issues for DEMO, several tasks were assigned to this system. Physics investigations aim at operation at high density in ITER and DEMO relevant plasma regime above Greenwald density, power exhaust techniques with radiation layers, particle balance studies, and ELM control and mitigation. The postulated engineering requirement is to quantify pellet actuation on electron density for application within the advanced real-time control scheme by controlling density gradients. Our intended pellet system comprises three major components: pellet source, accelerator, and guiding system. The guiding system must be installed inside the torus vessel already under construction; hence still possible launch geometry was pursued first. Three different options have been identified: inboard, outboard, and top launch. The first one is most promising with respect to fuelling performance but will impose pellet speed restrictions to about 470 m/s for adequate pellet sizes. Both others offer headroom for significantly higher injection speed but under less favourable physics boundary conditions. In order to evaluate expected performances for all relevant plasma scenarios, detailed modelling efforts for every launch geometry option have been made. For a suitable pellet source covering all requirements, several options are at hand including commercial providers. For the accelerator, the high speed option up to about 4000 m/s could be covered by a two-stage gas gun. Single-stage gas guns and centrifuges can cover the speed range up to about 1000 m/s for the basic work load since both fulfil the requirements for pellet size and speed. Due to a higher speed precision resulting in less timing jitter, a centrifuge would be better suited for control requirements. [ABSTRACT FROM AUTHOR]

The envisaged Pellet Engineering Testbed (PET) in the framework of Direct Internal Recycling Integrated Development Platform Karlsruhe (DIPAK) will provide unique opportunities to advance pellet injection technology and its implementation in the fuel cycle as well as into plasma control of EU-DEMO. This work analyses the parameters for Pellet Launching System (PLS) control according to requirements from plasma control. Main parameter is the particle flux and the related accuracy with respect to amount of flux and arrival time of pellets. Others are addressed as well like dynamic response to setpoint changes and adaptivity of isotope composition of the hydrogen ice. The performance of the PLS is limited by the mass loss in the guiding tubes, both by erosion (applies to any pellet) and loss of an entire pellet, that may apply to only a few pellets, provided a suitable guiding tube geometry is selected depending on the required injection speed. DIPAK-PET has to provide data points in order to enable proper guiding tube design activities based on modelling. A short description of the proposed technical solution for the Pellet Launching System is provided and setup for the Programmable Logic Control (PLC) is suggested comprising a MasterPLC and a dedicated Human Machine Interface (operators' desk). A non-exhaustive proposal of experiments to be performed in DIPAK-PET is listed in order to verify the estimated performance values for the relevant PLS parameters. [ABSTRACT FROM AUTHOR]

• Core fuelling of EU-DEMO tokamak requires pellet injection from the HFS at ≳1 km/s. • Two complementary approaches are being pursued to achieve this goal. • Guide tubes with bend radii ≥6 m to try delivering intact pellets at ∼1 km/s. • High-speed (∼3 km/s) pellet injection along DLS paths from the HFS or vertical port. • Scatter cone of free-flight high-speed pellets measured and neutron flux estimated. Pellet injection represents, to date, the most promising option for core fuelling of the EU-DEMO tokamak. Simulations with the HPI2 pellet ablation/deposition code indicate, however, that sufficiently deep fuel deposition requires injection from the High Field Side (HFS) at velocities ≳1 km/s. Two complementary inboard injection schemes are being explored: one makes use of guide tubes with curvature radii ≥6 m in the attempt of preserving pellet integrity at speeds of ˜1 km/s, the other is investigating the feasibility of injecting high-speed (˜3 km/s) pellets along "direct line of sight" (DLS) trajectories, from either the HFS or a vertical port. Options using quasi-vertical DLS paths routed across the upper vertical port have been explored first, as they can be more easily integrated, Unfortunately, the radial position of the available vertical access (≳9 m from the machine axis) turns out to be unfavorable; further simulations with the HPI2 code predict indeed that vertical injection may be effective only if pellets trajectories are well inboard the magnetic axis. High-speed injection through oblique inboard "DLS" paths, not interfering with the Central Solenoid (CS), are instead predicted to yield good performance, provided that the injection location is ≲2.5 m from the equatorial mid-plane. The angular spread of high-speed free-flight pellets, recently measured using an existing facility, turns out to be enclosed within ˜ 0.7°. This scatter cone may require significant cut off volume of the Breeding Blanket (BB). Moreover, DLS in-vessel conical penetrations may increase the neutron flux outside of the bio-shield, and also result in a significant heat load in the cryogenic pellet source. These issues are being investigated, to identify suitable shielding strategies; preliminary results are reported. The suitability of straight guide tubes to reduce the scatter cone, and hence the corresponding open cross section on BB penetration and the neutron streaming, will be explored as a further step. [ABSTRACT FROM AUTHOR]

Highlights • Advanced technology modules were added to the Sycomore fusion reactor system code. • Increased maximum toroidal field coil steel stresses yield large reactor size reduction. • Increase of the tritium burn-up requires large reactors. • Water-cooled and helium-cooled divertors are equivalent for DEMO1-class pulsed designs. Abstract The next step for fusion energy after the ITER tokamak is the demonstration power plant DEMO. In this framework, system codes are used to address high-level key design issues for the DEMO pre-conceptual phase. They aim at capturing the interactions between the subsystems of a fusion reactor. SYCOMORE is a modular system code which includes physics and technology models coupled to an optimizer in order to explore a large design parameter space. In the present paper, trade-off studies focused on technology modules are reported including the influence of some design-driving assumptions on the reactor performances and size, starting from a European DEMO1-like design (more than 500 MW net electric power and 2 h burn duration). The increase of the mechanical stress limits in TF and CS magnets can help reducing the reactor size, slightly more when high temperature superconductors are used in the TF coil. The tritium breeding ratio can be improved to more than 1.10 by a moderate increase of the size, but the tritium burn-up ratio needs one additional meter of major radius for every percent increase. Divertor coolant options are also compared, showing some differences between helium, hot and cold water scenarios at various incident divertor heat fluxes. [ABSTRACT FROM AUTHOR]

Core fuelling of DEMO tokamak fusion reactor is under investigation within the EUROfusion Work Package “Tritium, Fuelling and Vacuum”. An extensive analysis of fuelling requirements and technologies, suggests that pellet injection still represents, to date, the most realistic option. Modelling of both pellet penetration and fuel deposition profiles for different injection locations, assuming a specific plasma reference scenario and the ITER reference pellet mass (6 × 10 21 atoms), indicates that: 1) Low Field Side (LFS) injection is inadequate; 2) Vertical injection may be effective only provided that pellets are injected at ∼ 10 km/s from a radial position ≤∼8 m; 3) effective core fuelling can be achieved launching pellets from the High Field Side (HFS) at ∼1 km/s. HFS injection was therefore selected as the reference scheme, though scenarios featuring less steep density and temperature gradients at the plasma edge could induce to reconsider vertical injection at speeds in the range of 4–5 km/s. To deliver intact pellets at 1 km/s from the HFS, the use of guide tubes with a bend radius ≥6 m is envisaged. The results of above simulations rely on the hypothesis that pellets are delivered at the plasma edge with the desired mass and speed. However, mass erosion and fracturing of pellets inside the guide tube (severely limiting the transfer speed), as well as pressure build up and speed losses at relevant injection rates, might hamper the use of curved guide tubes. An additional innovative approach, aimed at identifying inboard straight “free flight” injection paths, to inject pellets from the HFS at significantly higher speeds, is proposed and discussed as a backup solution. Outboard high-speed injection is still being considered, instead, for JT-60SA. [ABSTRACT FROM AUTHOR]

DEMO will be a burning fusion device, most likely a tokamak type and thus, the plasma needs to be re-fuelled. A project was launched in the European DEMO Programme to develop concepts for the matter injection. This approach and the actual status of the project are presented. In a first step, requirements for actuators on plasma core density were elaborated. Related modelling activities indicated that only sufficiently deep fuel deposition can achieve target operational parameters. Hence, suitable techniques had to be identified and evaluated with respect to their availability and capability. Several techniques for pellet injection have been benchmarked in view of the defined requirements. Finally, cryogenic pellet injection was chosen as the main actuator on plasma core density. From further modelling activities, assuming for the pellet mass the ITER reference value, required launching speeds were derived, with respect to different injection geometries. Gas puffing and the respective technical system are necessary for pre-fill, ramp-up and plasma confinement enhancement. The ITER gas injection system (GIS) is assessed in view of suitability for DEMO. The piping system and the manifold concept can be adopted. The Gas Valve Box (GVB) is considered not to be an optimum solution for DEMO. Instead of this GVB, a pressure based RUN/VENT flow regulation and injection system is proposed in order to meet DEMO requirements. The principle of this system is described as well as some considerations about injection locations; further orienting gas flux numbers are provided. [ABSTRACT FROM AUTHOR]

The peak power flux densities from plasma ion bombardment on tokamak armor surfaces can deviate from geometric-optical projection when the ion thermal Larmor radius becomes comparable to the characteristic size of the surface relief (gaps, misalignments, shaping). We present a fast and reliable tool to estimate the effect of finite ion Larmor radius on heat load deposition for complex 3D elements. The two possible geometrical design options foreseen for WEST tungsten monoblocks are asymmetric chamfer shaping and no shaping on top of monoblocks. This paper presents an evaluation of peak power flux densities for a WEST ITER-like tungsten monoblock plasma facing component which is vertically misaligned with respect to the other components of a maximum expected value of +0.3 mm, for steady state convected plasma power with ion temperature ranging from 10 to 500 eV. With the +0.5 mm asymmetric chamfer, leading edges between plasma facing components are shadowed and there is no significant difference between the optical calculations and the full ion orbit model. For an unshaped and +0.3 mm misaligned monoblock, when considering the full ion orbit model, the heat load on the exposed edge is reduced by a factor of 3–5 compared to the geometric-optical projection. Energy is found to be deposited over several millimetres on top of the monoblock just after the leading edge which prevents the tungsten armor from melting. [ABSTRACT FROM AUTHOR]