Titanium beryllide is considered as an advanced neutron multiplier in the Helium Cooled Pebble Bed (HCPB) blanket of DEMO. Neutron irradiation of titanium beryllide together with beryllium as a reference material in material testing nuclear reactors can give essential data for the DEMO blanket design. Be-7at.%Ti (Be-7Ti) as well as Be were irradiated in the HFR, Petten, the Netherlands, at four temperatures

For the first time, the TiBe12 blocks have been manufactured by powder metallurgy using industrial equipment at the Ulba Metallurgical Plant. Such blocks will be used for neutron multiplication in the helium-cooled breeding blanket for the EU DEMO fusion reactor. The vacuum annealing of a cold-pressed composite of Be and Ti powders at 600–1275 °C approved the conditions for TiBe12 synthesis. After

This work describes integrated numerical modelling applied to DEMO discharges with tungsten wall and divertor, using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multi-fluid transport in the SOL. The model is self-consistent with respect to both the effects of impurities on the α-power level and the interaction between

Within the framework of the EUROfusion activities regarding the EU-DEMO Breeding Blanket (BB) concept, the University of Palermo is long-time involved, in close cooperation with ENEA, in the design of the Water Cooled Lithium Lead (WCLL) BB, that is one of the two concepts under consideration for the DEMO reactor. It is mainly characterized by a liquid lithium-lead eutectic alloy acting as breeder

The acceleration grid power supply (AGPS) is developed to provide accelerator power for China Fusion Experimental Rector (CFETR) Negative-ion Neutral Beam Injection (N-NBI) system. The rated output parameters of this AGPS are -1000kV/80A/80MW. The preliminary design of the AGPS is proposed in this paper. To improve the steady-state and dynamic performance of the AGPS, a new CFETR AGPS conversion system

For upcoming nuclear fusion energy reactors, like the China Fusion Engineering Test Reactor (CFETR) and EU-DEMO, the superconducting Cable-In-Conduit Conductors (CICC) are in the design phase, and the operating conditions like electromagnetic forces can be higher than in previous devices like ITER. The prototype conductors for the Central Solenoid (CS) coils in the CFETR, for example, are designed

Within the framework of the Work Package DIV 1 - “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the thermal-hydraulic behaviour of the DEMO divertor cassette cooling system, focussing the attention on the 2018 configuration of the Plasma Facing Components (PFCs) circuit consistent

Generally, DEMO requires larger toroidal field (TF) coils than ITER, resulting in one of the major difficulties, the tolerance in TF coil fabrication. This paper presents the possible solutions based on the design study on Japan’s DEMO (JA DEMO). It was confirmed that, in the case of adopting a mitigated tolerance by a factor of 2.5–5 compared with that of ITER, the resulting error field of TF coils

Super-X test facility can simulate the electromagnetic performances of a superconducting conductor under fusion reactor operating conditions, in which DC magnets are a core assemble. The DC test magnet provides 15 T central magnetic field. The field homogeneity of test area is more than or equal to 98 %. The magnet comprises three pairs of Nb3Sn coils (i.e., high-field, mid-field and low-field.). Included

Open issues for fusion reactor technology are related to the breeder material to be used for the nuclear fusion reaction. Despite in the last decades several studies were carried out, the selection, characterization and reliability of breeder material are still open issues. In this study, the attention is focused on the production of lithium orthosilicate (Li4SiO4) by means of drip casting method.

MARTe is a framework for real-time control that has been used in several fusion experiments. Recently, a new version named MARTe2 has been developed adhering to software quality standards. MDSplus is a data system widely adopted in the fusion community. MDSplus provides fast data acquisition and access to pulse files and is intended to provide a complete interface both for the configuration of the

Within the framework of EUROfusion R&D activity, a research campaign has been carried out at the University of Palermo, in close cooperation with ENEA labs, in order to preliminary design the top cap foreseen for the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket segments. Due to the high heat and pressure loads acting on such component, its design results particularly demanding and a specific

After the successful short pulse operation phase 1 (OP1) of the stellarator Wendelstein 7-X with maximal plasma energies of 200 MJ, the upcoming long pulse OP2 aims at stepwise higher energies up to 18 GJ. With the knowledge of partly unexpected experimental heat load distributions on plasma facing components (PFC), their allowable loads were re-evaluated which consequently lead to some adaptation

As part of ITER research to devise suitable facilities withstanding extreme plasma, divertor modules have been designed in a way equipping tungsten armor blocks with optimized shapes at target regions. However, in spite of comprehensive considerations for the material and geometry, thermally damaged cracks were observed from recent limited mock-up tests. Objectives of the present numerical study are

The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium production, shield the Vacuum Vessel and remove the heat generated by plasma thermal radiation and nuclear reactions. It relies on PbLi eutectic alloy adopted as breeder and neutron multiplier and refrigerate by subcooled pressurized water. The last function is fulfilled by two independent cooling

Absolute extreme ultraviolet (AXUV) diode arrays are used to measure the plasma radiation for a wide spectral range in the fusion devices, providing valuable information about plasma radiated power and transport phenomena, especially during fast plasma events (such as plasma disruption and edge-localized mode). Design and optimization of the AXUV diode arrays are performed for the HL-2 M tokamak. 16-channel

Tungsten “Metal wall” experiments are planned at JT-60SA to contribute to ITER. Because tungsten coated CFC materials are attractive for the “metal wall” experiments, their fabrication techniques need to be developed. Aiming to the application of divertor, a sinter bonding method is being developed to join thick tungsten plates and CFC blocks. Present research reports about the fabrication of W/CFC

Fueling a plasma reactor by pellet injection is the most efficient method. Production of a solid hydrogen filament using a Twin screw extruder (TSE) is reported to be most stable among various extrusion techniques. The non-Newtonian behavior of solid hydrogen and non-linearities arising due to temperature dependent thermophysical properties in addition to rotation of two intermeshing screws make the

Highly deformed tungsten (W) thin foils could show excellent low temperature ductility, even at room temperature. A laminated composite using pure W foils has been proposed for the application to the plasma-facing material of divertors, which is a composite material consisting of stacked pure W thin foils and interlayer materials. However, the degradation of low temperature ductility attributed to

Recent advances in tritium transport modeling of helium-cooled ceramic breeding blankets systems has shined light into some tritium management issues. A detailed component model accounting for multi-physics, design, and operational features is necessary to provide accurate estimations of tritium permeation rates to the building/environment- a safety and licensing concern for a fusion nuclear reactor

The blanket system is a key component to fulfill tritium breeding and nuclear energy conversion in the fusion reactor. This paper reports and presents the design and analysis of an equatorial inboard water-cooled ceramic breeder (WCCB) blanket module for Chinese Fusion Engineering Test Reactor (CFETR). The detailed structure design and the schematics of coolant flow scheme optimization has been presented

The design activity on the CFQS quasi-axisymmetric stellarator which is conducted as a joint international project is ongoing by National Institute for Fusion Science (NIFS) in Japan and Southwest Jiaotong University (SWJTU) in China. The CFQS magnetic field generating coils consist of sixteen modular coils (MCs) in total with four different types, four poloidal field coils (PFCs), and twelve toroidal

Divertor Tokamak Test (DTT) facility, the Italian fusion machine, foresees the Electron Cyclotron Resonance Heating (ECRH) system to provide additional heating power to plasma ohmic-heating. Such system will be composed of an initial set of 8 mm-wave 1-MW sources at 170 GHz, which will be raised to 16 in a second stage of installation. The power of each source will be launched into the plasma by a

For the EU DEMO Tokamak, Electron Cyclotron (EC) launching systems for plasma heating and stabilization are under development. Various concepts for the optical system are currently studied of which the Mid Steering Antenna (MSA) with a steering mirror at a recessed position behind the breeding blanket (BB), the Open Ended Waveguide (OEWG) concept with quasi-optical beam propagation by fixed mirrors

DEMO requires a novel fuel cycle architecture with Direct Internal Recycling in order to drastically minimize the tritium inventory. One central element of this architecture is the separation (and compression) of hydrogenic gas from the exhaust gas stream, close to the divertor, which can immediately be recycled to feed the matter injection systems. The hydrogen-selective separation will be achieved

The JET control and data acquisition system (CODAS) is an integrated system that provides all the pulse based and continuous data acquisition, real time and slow control and control room interfaces for JET. It has a long history, dating back to the beginning of JET in 1980. It utilises both commercially available hardware along with many in-house modules. The software has grown up and evolved largely

In 2013 Tore Supra limiter tokamak at CEA Cadarache went through major changes to become the WEST tokamak (Tungsten [W] Environment in Steady state Tokamak). The tokamak was upgraded into an X point divertor device. Taking advantage of its long discharge capability, WEST's goal is to minimize technological and operational risks of the actively cooled tungsten divertor for ITER. At the same time, Control

This paper gives guidelines for assessing the structural integrity of plasma facing components (PFC)when irradiated to the levels expected in DEMO after two full power years. The paper is part II of a 3-part paper describing the EuroFusion DEMO Divertor group (WPDIV) Inelastic Analysis procedure (IAP), created to improve the assessment of PFCs, and specifically those constructed from tungsten armour

Active feedback stabilization of the dominant resistive wall mode (RWM) for an ITER H-mode scenario at high plasma pressure using infinite-horizon model predictive control (MPC) is presented. The MPC approach is closely-related to linear-quadratic-Gaussian (LQG) control, improving the performance in the vicinity of constraints. The control-oriented model for MPC is obtained with model reduction from

Integrated detritiation system (IDS) would play a key role in controlling the tritium release and atmosphere cleanup under accident scenarios in fusion reactors, and thus it is very significant to evaluate and ensure its reliability. Up to now, many conceptual designs for IDS have been developed worldwide, yet few are focused on its reliability studies. Due to complex redundant designs and functional

Tritium permeation through structural materials in fusion reactor blanket systems is a critical issue from the viewpoints of fuel loss and radiological safety. Ceramic coatings have been investigated to prevent tritium permeation; however, corrosion of the coatings by blanket materials, especially corrosive liquid tritium breeders such as lithium-lead is serious. In our previous study, the improvement

Superconducting cable-in-conduct-conductors (CICC) are multi-stage structures with complex geometries. We study the contact behavior and tensile stiffness for a new design of CICC. The geometrical configurations of the new design of CICC is first formulated. Based on the superposition principle, a simple theoretical model is proposed to study the contact behavior of the new design of CICC. A wide range

The ITER Fibre Optic Current Sensor (FOCS) is a dc coupled current measurement system that will provide back-up ITER plasma current measurements using a sensing fibre located on the outer surface of the vacuum vessel. To address the impact of a harsh tokamak environment on the sensor performance we installed a FOCS system on JET and performed current measurements in various machine operating scenarios

Recent advances in selective laser melting (SLM) 3D printing technology allows additive manufacture of lower hybrid current drive (LHCD) RF launchers from a new material, Glenn Research Copper 84 (GRCop-84), a Cr2Nb (8 at. % Cr, 4 at. % Nb) precipitation hardened alloy, in configurations unachievable with conventional machining. Rough surfaces in additive manufactured components are a limiting factor

Researchers began exploring methods for selecting tolerances at design level to ensure competitive products since the late 1980. Reliability of mechanical product depends greatly on different performance parameters among which tolerance optimization design is one of the most important activities. Application of the tolerance design to mechanical assemblies has benefits for obtaining valuable information

In the Water-Cooled Lithium Lead (WCLL) blanket, the eutectic alloy lithium-lead (PbLi) is used as tritium breeder and carrier, neutron multiplier and heat transfer medium. The liquid metal is distributed to and collected from the breeding zone through a compact poloidal manifold composed of two co-axial rectangular channels. The external channel, tasked with distribution, and the internal one, assigned

The thermal-hydraulic characteristics of helium as a purge gas for the removal of the tritium from a spherical packed rectangular pebble bed reactor are studied in this paper using the DEM-CFDand porous medium approaches. Numerical models are developed for the two approaches considering radiation heat transfer, which are validated with the published experimental results. Further, a comparison is made

The vacuum vessel is very important component for China Fusion Engineering Test Reactor (CFETR), it is a double-layer shell structure with complex contour. The vacuum vessel is mainly formed by welding, which requires high precision of forming dimension. Welding distortion brings difficulties to the forming dimension control of vacuum vessel, so it is necessary to control the welding distortion in

Understanding phenomena associated with the multiple effects/interactions of the fusion nuclear environment on liquid metal flow is required to correctly design liquid metal (LM) blankets for fusion facilities. These effects are investigated in the present work by numerically simulating 3D LM MHD flow. The simulated geometry consists of a straight, vertical duct which runs perpendicular to a strong

In the design of solid breeder blanket, normally the homogenous model is used for the neutronics analysis to assess the performance of blanket, but the adaptability of homogenous model is neglected. Referring to the design of CFETR (China Fusion Engineering Test Reactor) blanket, impact assessments of homogeneous model with different coolants, helium and water, are investigated with 3D high-fidelity

The in-box Loss of Coolant Accident (LOCA) scenario is considered as one of the most affecting safety concerns for the Water-Cooled Lead Lithium Breeding Blanket (WCLL-BB) modules of the DEMOnstration (DEMO) reactor, which is sequentially followed by a multi-phase multi-component physical and chemical interaction. Therefore, the transient behavior of such accidents has to be carefully investigated

The role of the ITER thermal shield (TS) is to minimize the radiation heat load transferred from the warm components to the superconducting magnet operating at 4.5 K. Fabrication of the TS components is ongoing. Manufacturing of LCTS (Lower Cryostat TS) cylinder and VVTS (Vacuum Vessel TS) sector #6 was completed. The leak tightness of the cooling tubes, the on-site installation feasibility and the

Diffusion bonding methods are one of the joining methods for Plasma Facing Components in fusion reactors like ITER or in later decades the DEMO reactor. Diffusion bonding of 316 L specimens were investigated on the Gleeble 3800 thermomechanical physical simulator at University of Dunaújváros at 1000–1050 °C, with 20−30 MPa, through 40−80 min under high vacuum conditions. A new method was found that

In the Water-Cooled Lithium Lead (WCLL) blanket, the eutectic alloy lithium-lead (PbLi) is used as tritium breeder and carrier, neutron multiplier, and heat transfer medium. The PbLi hydraulic loop section in the range of the reactor field coils, which includes the blanket and a non-negligible length of the connection pipes, is affected by intense magnetic fields which cause the transition to a MHD

This paper reports the gamma-ray irradiation effects on optical coatings and polarizers to be used in the ITER edge Thomson scattering system (ETS). Outside a diagnostic port-plug, in the Interspace, a total dose of the order of 1 MGy is expected through 20 years of ITER operation. In this study, gamma-ray irradiation experiments were performed up to 10 MGy. A high peak power laser beam, i.e. a pulse

Deuterium (D) plasma driven permeation (PDP) experiments for tungsten (W) samples were conducted by a linear radio frequency (RF) plasma device. In the PDP experiment, the W sample surface is perpendicular to the grain elongation direction. The D ion flux is in the order of 1021 m−2 s−1 determined by a double Langmuir probe. The results show that bias had a limited effect on the D plasma driven permeation

Flow and stock of fuels in a fusion reactor has been investigated from the aspect of system-dynamics. In particular, operational scenario of the start-up only by deuterium has been discussed with an integrated model of burning plasmas including effects of dilution due to helium ash. The start-up only from deuterium has been attracting interests since initial loading of tritium in a fusion reactor is

Contextual metadata becomes an important factor for the work with research data, which is created in large amounts at fusion experiments. In its first operational phases, the stellarator experiment Wendelstein 7-X (W7-X) produces about 500 TB of experimental data in only a few weeks of scientific campaign. A new central logbook software was implemented for the processing of contextual metadata at W7-X

The isotherms of hydrogen isotope, absorbed or desorbed by palladium particles of various sizes were investigated at ∼100 kPa of various absorption-desorption cycles at room temperature (25 ℃). The maximum hydrogen (D) content of these samples, the hydrogen (D) absorption capactiy, was determined using the Langmuir monolayer coverage equation from the isotherms after several cycles. A ratio of the

Within the framework of EUROfusion DEMO First Wall and limiter design activities, the protection of the First Wall against power deposition peaks is being considered. During steady-state operation, the radiative power from the plasma could be considered more uniformly spread on plasma-facing components than the charged particle power deposition. The presence of openings (i.e. gaps between segments

In a complicated fusion device, the robot is at risk of colliding with its surroundings when it moves. In this article, a hybrid collision detection perceptron is studied to ensure the safety for both the fusion device and the robot. Primary component analysis is adopted as a preprocessor for extracting the features of the data. The stochastic gradient descent, Adam and Adagrad are used as classifiers

Huazhong University of Science and Technology (HUST) plans to develop a radio frequency (RF) ion source experimental setup for a high intensity neutron source which is at the conceptual design stage. The positive deuterium ion beam with the current of ∼14.5 A, the energy of 20 keV and the pulse length of ∼2 s is required. This paper introduces the RF driven deuterium ion source, and focuses on the

The purpose of the ITER Radial Neutron Camera (RNC) is the measurement of the plasma neutron emissivity profile [neutrons∙s−1 m-3] for burn control purposes. The present RNC design consists of 22 collimated detector systems providing a set of line-integrated neutron measurements [neutrons∙s−1 m-2] with full coverage of the plasma poloidal cross-section; the neutron emissivity can be recovered from

Improvements in strike point control enabled tests with the new small angle slot (SAS) divertor installed at DIII-D with precision as tight as ±1 mm, barring brief perturbations from ELMs. Aside from being narrower than previous baffle structures in DIII-D, simulations indicate that performance of the SAS divertor should be very sensitive to strike point position, making the improved strike point control

A novel 9Cr-ODS martensitic steel was designed for application in the advanced fusion reactors by reducing carbon content to an extremely low level to avoid the precipitation of M23C6 carbides and increasing manganese content to promote the martensite formation. The preliminary investigation was carried out by means of thermodynamics calculation, phase transformation measurement, microstructural observation

The super conducting stellarator Wendelstein 7-X (Bosch et al., 2017) started its first operational phase in October 2015 at the Max-Planck-Institute for Plasma Physics in Greifswald with the goal to verify that a stellarator magnetic confinement concept is a viable option for a fusion power plant, i.e. showing confinement comparable to tokamaks and running in steady state operation. Between 2015 and

It is conceived that tritium can be extracted from breeding liquid metals by using vacuum permeator. The properties of hydrogen isotopes under various operating conditions in the permeation system are basic condition of the optimal design of Permeator Against Vacuum (PAV). The capabilities of two different numerical models that can be used for predicting the permeation process of hydrogen isotope controlled

The nearly spherical shaped ceramic pebble beds in the breeder blanket of the future fusion reactor are purged by low-pressure gas to channelize the produced tritium fuel into the tritium extraction system. The required pumping power for the flowing gas in pebble beds can be estimated using the pressure drop across the pebble beds. The aim of this work is to measure the gas pressure drop experimentally

In-vessel lighting system is being developed for ITER. The diagnostic system is for inspecting and detecting eventual damages and defects of in-vessel components between discharges by illuminating the vacuum vessel with light sources. In this work, feasibility study of the in-vessel lighting system is performed using a ray tracing simulation. We installed sources and detectors in a ray tracing model

A novel multistage cascade membrane chemical exchange (MCE)process was introduced to separate lithium isotope using solvent resistant polyethylene vinyl alcohol (EVAL) hollow fiber membranes. A MCE pilot with 10 pieces 4-inch membrane modules was designed and commissioned using an azo chelating agents (HA system) as the organic phase. 7Li was enriched from a natural abundance of 92.47 %–92.92 %. The