The challenge of tokamak auxiliary heating method ICRH is to couple large amount of power through the plasma boundary, where an evanescence layer has to be crossed, without exceeding the voltage standoff at the antenna. Travelling Wave Array antenna has been proposed for ICRH of future fusion reactor such as DEMO in view to decrease the antenna power density. However, so far the voltage standoff of

In future tokamaks, the control of burning plasmas will require careful regulation of the plasma density and temperature. Along with the design of effective burn-control systems, understanding how the fusion power varies in the density-temperature space is vital for the operation of fusion power plants. In this work, the steady-state operational space of ITER is studied using a control-oriented core-plasma

The SMall Aspect Ratio Tokamak (SMART) is a new spherical device that is currently being designed at the University of Seville. SMART is a compact machine with a plasma major radius (R) greater than 0.4 m, plasma minor radius (a) greater than 0.2 m, an aspect ratio (A) over than 1.7 and an elongation (k) of more than 2. It will be equipped with 4 poloidal field coils, 4 divertor field coils, 12 toroidal

The COMPASS-U is a new tokamak in the design phase at Institute of Plasma Physics, Prague, Czech Republic. It is a medium-sized tokamak with high magnetic field (Bt = 5 T) and high temperature vessel (< 500 °C). Hence, large electromagnetic forces are expected on the Passive Stabilizing Plates (PSP) and Vacuum Vessel (VV) during plasma disruption events. A non-selfconsistent approach is selected for

Tungsten (W) is one of candidate surface materials for a plasma facing component (PFC) in a magnetically confinement fusion reactor such as a divertor from the perspective of its high melting point, high thermal conductivity, low tritium retention and low sputtering yield. In the fusion reactor in operating, the divertor surface is exposed by the transient heat load generated by plasma disruption or

The port stubs are larger welding parts of stainless steel plate, featuring with complicated structure and big overall dimensions contain complex local flanging shape and demand high forming quality. It is very important to forecast the welding stress and control the welding residual deformation exactly to ensure the safe reliability of the structure. In this paper, the welding distortions of port

Due to the higher hydrogen solubility and diffusivity, gadolinium sesquioxide (Gd2O3) could be expected as a proton conductive material. In this study, monoclinic Gd2O3 was used to evaluate the possible proton conductivity in Ar (σAr), Ar+4% H2 (σH2), Ar+4% D2 (σD2), Ar+H2O (σH2O), Ar+D2O (σD2O), and O2 (σO2) atmospheres in the temperature range from 500°C to 1000°C, to understand its possible application

The edge coherent mode (ECM) is a promising operational regime in the Experimental Advanced Superconducting Tokamak (EAST). In order to understand the physical mechanisms of ECM thoroughly, the automatic recognition of ECM is necessary. In this work, we define an adaptive denoising inequality based on the variance, and propose a dual-channel convolution feature extraction model to obtain a visualized

The constitutive equation of the 155/157Gd areal density and thermal neutron shielding rate was established by Monte Carlo N Particle Transport Code (MCNP) and used to design the novel Gd/316L stainless steel thermal neutron shielding materials. When the 155/157Gd areal density is greater than 0.01 g/cm2, the thermal neutron absorption rate of the Gd/316L composite reaches approximately 99%. The novel

For successful tokamak operation with adequate mode selection of the electron cyclotron wave, launching wave polarization should be accurately controlled. In this study, a new method is proposed to efficiently identify the response of polarizers using a geometric approach in the Stokes space, focusing on the fact that polarization can be represented as a vector in Stokes space and the response of a

To solve the problem of voltage overshoot during switching-off transient in H-bridge, an active lossless H-bridge snubber composing of auxiliary switch and snubber capacitor is discussed, which aims to enable the H-bridge to realize a higher voltage and current. When the main switch is turned off, the capacitor absorbs the overshoot energy and the energy is returned to DC side through the auxiliary

Future tokamaks will require robust technologies for the mitigation of heat exhaust onto the plasma-facing components. As a first step towards this development, a system has been developed at DIII-D that estimates and controls in real-time the heat flux to the PFCs. Real-time estimation of the peak power flux from this model-based approach is validated with off-line infra-red measurements for various

Using Monte Carlo (MC) transport codes for fusion device shielding calculation is very challenging due to the complexity and heavy shielding of a fusion reactor. An effective on-the-fly (OTF) global variance reduction method was developed and verified before. The OTF method is designed to speed up the general-purpose MC global total neutron distribution by generating and updating the global weight

The superconducting Central Solenoid (CS) of the EU-DEMO tokamak coil will comprise five modules, labeled CSU3, CSU2, CS1, CSL2 and CSL3, situated vertically one above the other. The central CS1 module will be operated under the most severe conditions, i.e. the largest mechanical loads and magnetic field. Two alternative designs of the CS1 module are being developed by the CEA and EPFL-SPC teams. The

One of the main issues in view of the realization of a DEMOnstration fusion reactor is the availability of a sufficient external supply of tritium (T) to start operation. T is an unstable nuclide, which is almost absent in nature and is currently available as by-product in e.g. CANDU, whose operation in the next decades (both in terms of life extension of existing reactors and construction of new ones)

ASDEX Upgrade (AUG) is the experimental tokamak based in Germany that since many years, 1991, explores technical solutions to address physic aspects. This time, the aim is to prove that alternative divertor configurations (X-divertor or Snowflake divertor) can mitigate the power exhaust problem, in a machine with high heating power like AUG. To realize the required magnetic configurations [2], two

The nuclear fusion reactor ITER (International Thermonuclear Experimental Reactor) foresees a Pressure Suppression System (PSS) in order to manage a Loss of Coolant Accident (LOCA) in the Vacuum Vessel (VV). A LOCA scenario is postulated to occur in the cooling systems of plasma facing components. The Vacuum Vessel Pressure Suppression System (VVPSS) has a key safety function because a large internal

Several welding methods are used in the manufacturing of the CFETR vacuum vessel. These welding methods will inevitably lead to the problems of welding residual stress and welding distortion. Relieving welding residual stress is one of the important works in CFETR vacuum vessel welding and vibration stress relief (VSR) is one of the methods to relief the residual stress in the welding of the CFETR

The ITER Neutral Beam Test Facility includes development, testing, and optimization of the full prototype of the ITER Heating Neutral Beam injectors (HNBs), named MITICA. A 40 MW precursor D−/H− beam will be produced and then neutralised and filtered along the beamline, aiming to obtain 18 MW D0/H0 beam at the calorimeter and 16.5 MW during operation into the plasma. A gas neutraliser and an electrostatic

The 4.6GHz LHCD system is an important part of the auxiliary heating system on EAST tokamak. The cathode high voltage power supply(HVPS) is based on pulse step modulation (PSM) technology, in which 64 modules are used to output 50kVDC in series. The modulation frequency of single PSM module is 50Hz, it switch on and off every 20ms, so the regulation speed of system is limited. In the face of the interference

There is proposed a new upper divertor for the ASDEX Upgrade (AUG) tokamak experiment [1]. It is planned to be equipped with internal coils for investigation of advanced magnetic configurations like e.g. „snowflake“ [2]. Due to the close vicinity of the coils to the plasma, high induced and very stiff voltages are expected during disruption events, as shown in [3,4],. This voltage is a high risk for

This paper introduces an optimal estimation methodology to estimate the boundary shape of plasma in Damavand tokamak (DT). The first step of the introduced methodology is determining an initial guess of the plasma center. An accurate initial guess of the plasma center can improve the performance of the boundary shape estimation. A methodology to find the best arrangement of filaments around the plasma

A hard X-ray measurement system is developed to observe the activities of runaway electrons in versatile experiment spherical torus (VEST) at Seoul National University. The detection system provides information on the generation of runaway electrons by detecting hard X-rays from its bremsstrahlung radiation. To acquire such information, the X-ray spectrum at several tens to hundreds of keV at the maximum

The precompression system of the ITER Central Solenoid (CS) consists of nine identical units designed to provide an axial precompression force of 210 MN at room temperature to the stack of the six independently energised modules of the CS. This load allows the modules to be maintained in contact with each other in the vertical direction during operation. Precompression is achieved by using tensioning

OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. It supports geometry modeling using constructive solid geometry, and is capable of performing fixed source particle transport simulation based on continuous-energy and group-wise nuclear cross-section data. The lack of variance reduction technique is one of the main drawbacks of this code, which prevent

In order to reach commercially-relevant availability, future fusion power plants must minimise the duration of maintenance shutdowns. However, as radiation levels increase, so too will the number of maintenance tasks that must be performed without human access. To meet these conflicting constraints, remote maintenance systems must therefore significantly increase their capabilities, performing tasks

For the purpose of enhancing thresholds of nucleate boiling and ensuring working safeties for DEMO reactor, numerical analyses of the influences of twisted tapes and fluid thermal-hydraulic parameters on DNB-type CHF and pressure drops have been carried out on plain tubes and tubes with twisted tape inserts. The research campaign has been conducted following the Eulerian multiphase model and the critical

The ITER project demanding operating conditions, as well as other enabling experiments in relevant fusion projects, present new challenges to the diagnostics, control and instrumentation. The most critical requirements relate to the high acquisition rates (up to some gigasample per second), high physical event rate (up to several megaevents per second), need to handle enormous quantities of data in

One of the unique features of ASDEX Upgrade is the heating power with respect to the size of the plasma. The power at the separatrix divided by the major plasma radius Psep/R representing the normalized heat flux at the divertor is 10 MW/m for 3 s. This is 2/3 of the power expected for a fusion power plant. In order to achieve this goal, the additional heating systems have been strengthened by four

Two models are currently taken as reference for the breeding blanket (BB) of DEMO, the Helium Cooled Pebble Bed (HCPB) and the Water Cooled Lithium Lead (WCLL). For both the models two Balance of Plant (BOP) configurations are currently investigated. They are all based on four independent primary heat transfer systems (PHTSs). The largest PHTS is devoted to remove the BB thermal power, two PHTSs for

For a strict control of tritium migration in fusion reactor fuel systems, tritium permeation barrier coatings have been developed for several decades. In liquid blanket concepts, corrosion of the coatings by liquid tritium breeders is a serious concern in addition to tritium permeation. Furthermore, the coatings are exposed to high-energy neutrons in an actual reactor, which would bring a synergy of

In this study, we investigated the material properties of tungsten mono-blocks undergone high heat flux test. The ITER-like tungsten mono-block specimens were fabricated by two kinds of joining methods. The tungsten-Cu interlayer joint was made by gas pressure casting and Cu interlayer-CuCrZr cooling tube were joined by hot radial pressing (HRP) or brazing method. To evaluate the integrity of the fabricated

In the Experimental Advanced Superconducting Tokamak (EAST), the water in the cooling channels of the upper divertor is difficult to be pushed out by the current drainage system resulting in the possible divertor damage. Therefore, studies on the draining process using the gas pushing for this divertor is essential to improve the current drainage system. In this paper, the hydraulic characteristics

Plasma Facing Components (PFCs) temperature measurement is required to ensure safe high power for long pulse tokamak operation and for physics studies. A set of twenty thermocouples (TCs) and four optical fiber temperature sensing probes, each of them including eleven wavelength-multiplexed fiber Bragg gratings (FBGs) written with UV radiation, have been integrated and deployed in the WEST lower divertor

The Water-Cooled Lithium Lead (WCLL) is a blanket concept pursued in the framework of Test Blanket Module (TBM) campaign in ITER. Even if the liquid metal is circulated slowly in the component, magnetohydrodynamic (MHD) pressure losses are still expected to be significant. The aim of this paper is to assess the MHD pressure losses in the TBM frontal part, also called Breeding Zone (BZ). There, important

Protective aluminum-based coatings represent a promising anti-permeation and anti-corrosion barrier for breeding blanket systems developed for European DEMO fusion reactor. Following the prior in-depth characterizations, selected coating candidates were subjected to a combined test consisting of contact with liquid Pb-16Li, its repeated in-situ solidification and re-melting, tritium permeation during

In the fuel cycle of ITER and the future DEMO fusion reactors, which has many processes of tritium provision/ consumption/recovery/loss, such as, fueling, DT burning, vacuum pumping, purification, isotope separation and storage, detritiation, etc., tritium accountancy control will be required in its operation and safety management. In contrast to ITER, which will have frequent intervals between operations

The energy balance of lithium recovery by electrodialysis using the solid electrolyte La0.57Li0.29TiO3 was investigated using AC impedance spectroscopy and overpotential measurements. It was found that most input energy is consumed in the electrolysis of the water associated with electrodialysis. Despite the use of low-performing model electrodes and electrolyte plates with low Li+ conductivity, most

Detached divertor scenarios at ASDEX Upgrade (AUG) need high gas fluxes, which are mainly pumped with the gas binding in-vessel cryopump (CP). To cope with the large quantities of hydrogen released from the CP during regeneration, a hydrogen compatible pumping system (HCPS), which is built to comply with safety regulations for explosion protection, has been built up in parallel to the already existing

Power Extraction System (PES) is the major port of outputting fusion energy in the Chinese Fusion Engineering Testing Reactor (CFETR). The large amount of thermal power is transferred by Power Extraction System Primary Heat Transfer System (PES PHTS) to achieve the effect of cooling plasma facing components (PFCs) and power generating. In addition, PES PHTS has the function of baking the PFCs. Therefore

To support Chinese Inertial Confined Fusion research, a thermal cycling absorption process (TCAP) system was previously developed for protium removal from D–T fuels. However, to monitor the product quality in each TCAP cycle (10 min), an on-line analysis method is required. At ITER, the Analytical System is an individual system used, which meets the static analysis requirement via interconnecting lines

The estimation of Activated Corrosion Products (ACPs) in fusion technology is an important input to safety assessments (e.g. occupational radiation exposure, accidental analyses). The present work aims to calculate ACPs for the First Wall (FW) and Breeder Zone (BZ) water cooling loops of the DEMO Water Cooled Lithium Lead (WCLL) fusion reactor. The code used is PACTITER v2.1. Geometrical and thermal-hydraulic

Tritium permeation from breeder material to the Water Coolant System (WCS) in Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) is one of the technological issues to be solved in the design of the European DEMO. Since the tritium extraction from the Water Coolant System is more challenging and expensive than the extraction from the eutectic alloy PbLi, it is mandatory to use of a protective coating

In this study, the hydrogen retention properties of solid tin exposed to DC glow plasma were investigated using thermal desorption spectroscopy. The measurements were performed by varying the plasma exposure time from 0 (no exposure) to 40 min. The resulting retained hydrogen from the ion fluence of the plasma was quantified as ~10−3. The retained hydrogen increased with the exposure time.

In the framework of ITER diagnostics, the visible and infrared equatorial Wide Angle Viewing System (WAVS, referenced as 55.G1.C0 in ITER Plant Breakdown Structure) plays a key role. Indeed, this system is devoted to monitor the surface temperature of the plasma facing components by infrared thermography and to image the edge plasma emission in the visible range. As a consequence, it is mandatory to

The Karlsruhe Tritium Neutrino (KATRIN) experiment aims to determine the effective mass of the electron antineutrino by investigating the tritium β-spectrum close to the energetic endpoint. To achieve this, there are stringent and challenging requirements on the stability of the gaseous tritium source. The tritium loop system has the task to provide the <0.1 % stabilized flow rate of tritium gas into

On ITER, long pulse gyrotrons are required as a power source for electron cyclotron heating (ECH) and current drive (CD). The microwaves are guided from the gyrotrons, which are placed far from the Tokamak, into the plasma by transmission lines (TLs) and a launching antenna (launcher). Each of the four ECH Upper launchers features eight waveguide (WG) TLs, with at least 95% of the power from the gyrotrons

Fiber bundles will be subjected to gamma-neutron irradiation in ITER Port Interspace & Cell areas, and the highest radiation dose will be accumulated in their head parts (collectors). Radiation tests have to be done to prove their radiation hardness. The paper is devoted to the radiation test of the fiber-optic collector prototypes at Co-60 gamma source. The collectors were manufactured using Epo-Tek

The SPIDER in-vessel actively cooled components shall be drained in case of major maintenance to limit, as much as possible, atmospheric corrosion inside the circuits, to prevent water spreading, and finally to allow the execution of vacuum leak tests of the components before re-installation of the SPIDER Beam Source (SBS) inside the Vacuum Vessel. Draining of hydraulic circuits is first foreseen by

The ITER project is the largest international collaboration in the scientific field ever set up. It forms the heart of a unique collaborative agreement to build the first experimental nuclear fusion device designed to prove the scientific and technological feasibility of sustained fusion power generation. Among hundreds of systems, the In-Vessel Coils System plays a crucial role in ITER: ensuring and

A few types of flint glasses intended for use in achromat lenses of optical diagnostics in ITER were tested recently in a nuclear reactor and demonstrated sufficient resistance to fast neutron fluences from 1012 to 1016 n/cm2. However, there is an evidence that some flint grades may have significant transient radiation-induced absorbance whereas post-irradiation measurements do not reveal any degradation

The low-frequency resonance suppressor effectively reduce the low-order harmonics and resonance amplification generated by the EAST poloidal field power system during the operation. This article takes the control parameters of the three-level neutral-point clamp rectifier in the low-frequency resonance suppression device as the research object to reduce the steady-state delay of the system and to ensure

Based on the latest optical design of ITER ECH&CD equatorial launcher, the power absorption at the plasma breakdown is evaluated with considering reflections by the 3-dimensional first wall structure. The beam diffusion is simulated by a bunch of spreading rays and the reflections on the first wall are analyzed for each ray. The power absorption is evaluated by the optical thickness when the rays pass

Deuterated-tritiated polymers are a type of promising target materials for inertial confinement fusion (ICF). The preparation of a deuterated-tritiated polymer was performed from a deuterated polybutadiene under tritium through solid state catalytic tritiation with 5 wt% Pd/BaSO4 as a catalyst. After reaction, the gas mixture in the vessel and the raw T2/Ar mixture in the flask were analyzed with a

Lower Vertical Neutron Camera (LVNC) is a diagnostic system installed into the Diagnostic Rack of the ITER Lower Port #14. LVNC includes water cooling system, six detectors units, electrical feedthroughs and cables. Finite element model and finite volume model created with ANSYS software were used to obtain temperature and stress-strain state during Normal Operation and Baking modes.

The hydriding reaction with hydride materials such as ZrCo, which has low hydrogen equilibrium pressure at room temperature, can be used in the helium-3 purification process to remove tritium impurity from a mixture of helium-3 and tritium. However, the helium interferes with the hydriding reaction, so it is necessary to confirm whether the helium affects the reduction of the tritium partial pressure

Erbium oxide (Er2O3) coating has gained much attention as an effective measure to assist in solving some destructive problems in Tokamak blanket system. In this paper, Er2O3 coatings were deposited by sol-gel method on the SUS 304 stainless steel substrate, and then immersed the coating specimens in liquid lithium at 500℃ lasting for 600 h for corrosion resistance study. The coatings prepared at different

An optimal model to schedule fuel regeneration is developed for the Tokamak Cryopumps and Cryogenic Viscous Compressors (CVCs) of the fuel cycle of ITER. The objective is to minimize the tritium inventory inside the Isotope Separation Systems (ISS) while supplying the tokamak’s required amount of fuel considering fueling stages: ramp-up, flat-top, and ramp-down for each tokamak operation mode (Inductive

OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It is capable of performing fixed source particle transport simulation based on continuous-energy nuclear cross-section data. In this work, OpenMC has been benchmarked and verified for the application to accelerator-based neutron sources, e.g. the International Fusion Material Irradiation Facility- DEMO Oriented