A real time Wall Monitoring System (WMS) is used on the WEST tokamak during the C4 experimental campaign. The WMS uses the wall surface temperatures from 6 fields of view of the Infrared viewing system. It extracts the raw digital data from selected areas, converts it to temperatures using the calibration and write it on the shared memory network being used by the Plasma Control System (PCS). The PCS

We have studied pulse waveform dependence of tungsten mass erosion under disruption-like heat loads. In the experiment, tungsten samples were placed in a vacuum chamber and irradiated by a disruption-like heat load using a 1064 nm Nd: YAG laser. Mass erosion of tungsten samples by laser irradiation was measured by a microbalance. Laser intensity (up to 6.2 GW/m2), pulse duration (4 ms–5 ms), and pulsed

The ADITYA-U tokamak (R0 = 0.75 m, a = 0.25 m) is designed to shape plasma column in both single and double null diverter configurations. It is quite well known that sustaining a shaped plasma in tokamak requires very good plasma column position control, both horizontal and vertical. An FPGA-based proportional–integral–derivative (PID) control system has been designed and operated to achieve horizontal

A new divertor concept REVOLVER-D2 that uses metal pebbles as a divertor target has been proposed for the LHD-type helical fusion reactor FFHR-c1. Demonstration of sufficient plasma shielding rate is one of the important issues to show the feasibility of the REVOLVER-D2. For this purpose, an experiment on the plasma irradiation to the metal pebble flow has been conducted in the sheet plasma device

Institute for Plasma Research (IPR) is establishing an intense neutron generator facility. It is an accelerator-based 14 MeV neutron generator that produces 1012 neutrons per second. The neutron generator facility has various components that are made of different kinds of materials. It gets radio-activated due to the direct interaction of neutrons and secondary radiations. The radio activation of components

A bulk ultrafine-grained (UFG) W-ZrC-Re alloy is fabricated via a developed bottom-up powder metallurgy process. The detailed fabrication processes including the mechanical alloying and spark plasma sintering have been introduced. The relative density of this UFG W-ZrC-Re is as high as 99.5 %, and the average grain size is about 0.4 μm and uniform dispersion particles with the average particle size

In order to achieve high power generation efficiency, facility compactness, high safety, and high coexistence with fuel tritium in the 3 G Wth class FFHR(Force Free Helical Reactor) power plant, optimization, performance and operation scenario of the power generation system of the power plant model using the axial flow type uniaxial supercritical CO2 gas turbine power generation system were examined

The ECRH (Electron Cyclotron Resonance Heating) system was used in the EAST (Experimental Advanced Superconducting Tokamak) experiments for several years, and some good experimental results have been obtained. However, the performance of the gyrotrons is degraded and the ability to stabilize output RF (Radio Frequency) power is reduced. Two single-frequency 140 GHz gyrotrons are planned to be repaired

Within the framework of the European fusion strategy, two tritium Breeder Blankets concepts are developed to be tested in ITER as Test Blanket Modules (TBM): the Water-Cooled Lithium-Lead (WCLL) which uses the liquid Pb-16Li as a breeder, neutron multiplier and tritium carrier, and the Helium-Cooled Pebble-Bed (HCPB) with lithiated ceramic and beryllium pebbles as breeder and neutron multiplier, respectively

In addition to a reduced number of feeders, which would simplify their integration in the cryostat of the tokamak, the present study considers the possibility to reduce the voltage during safety discharge of the toroidal field (TF) coils. A lowered discharge voltage would ease the manufacture of the TF coils and reduce the operational risks. Both aspects closely related to the discharge time constant

Being the Vacuum Vessel Pressure Suppression System (VVPSS) one of the most important passive safety systems to be foreseen in DEMO plant, design and integration challenges have to be faced to ensure that best performance within safety requirements are always achieved. In this framework, parametric safety analyses have been performed to support VVPSS design activities; in particular to determine the

The charged particle heat load expected for the DEMO Single Module Segment first wall (FW) during current off-normal plasma scenarios indicates that protection is needed for avoiding/reducing damage to the breeding blanket FW due to the deposition of a huge amount of energy in a small timescale [1]. Within the EUROfusion framework of heat load analysis and design of DEMO wall and FW protections during

The morphological and structural changes of two different dual-phase tungsten alloy (either 3% or 5% of an added mixture of Fe and Ni) are investigated after deuterium plasma irradiation. Heat flux factor of deuterium plasma pulses is comparable to transient events in ITER. This study compares the changes in tungsten alloy materials between two different regimes with a different number of plasma pulses

A robust safety case for a fusion power plant for electricity generation must demonstrate that the radiological risk to workers and the public under any credible accident scenario is as low as reasonably practicable (ALARP). Whilst the hazard potential of a fusion power plant is significantly less than that of a fission power plant, a fusion power plant will still contain radiological inventories.

Liquid metal Lead-Lithium (PbLi) flow channels in Indian LLCB (lead lithium cooled ceramic breeder) Test Blanket Module consist of a sequence of parallel channels which are fed from a common inlet header. Thin radial-toroidal vertical plates known as partition plates at the central plane perpendicular to the applied magnetic field are primarily envisaged to enhance the mechanical strength of high aspect

The grating polarizer is a critical component in the millimeter-wave transmission line in an electron cyclotron resonance heating (ECRH) system, the main function of which is to change the polarization characteristics of the millimeter-wave. This paper presents an optimal polarization strategy for the millimeter-wave transmission line in an ECRH system. With two identical sinusoidal-groove elliptical

This paper presents the high flux neutron shielding design and extensive neutronics calculations of GDT based fusion neutron source ALIANCE. Neutron distribution of ALIANCE is strongly inhomogeneous along the axis: significant portion of the neutron flux is generated near the two mirrors, while the rest of it is spread over the remaining central volume of plasma. The shielding design includes 40 cm

Friction stir processing (FSP) is a solid-state material processing technique to enhance the surface properties of metallic materials. In this study, Cu-1.0%Cr alloy is friction stir processed with an intention to improve its surface hardness and wear resistance. A single-pass FSP was performed by varying the tool traverse speed from 50 to 200 mm/min in steps of 50 mm/min and the tool rotational speed

This article presents development of remote handling tool for FW central bolt tightening and its demonstration of the functional requirements in prototype testing. A compliance mechanism was newly designed and successfully demonstrated for the capacity to handle the misalignment of the thread part and to withstand 10 kN m torque. Maraging steel wrench with out-of-standard Torx bit was designed and

The qualification of materials to be used in the future DEMOnstration power plant (DEMO) is the main task of the DEMO Oriented Neutron Source facility (DONES). Over the next decades, experiments will be executed to analyze the reaction of in-vessel materials for DEMO which are exposed to neutrons of high energy. An efficient execution of experiments requires a steady maintenance and replacement of

Heat pipes (HP) are considered being used in the divertor target because of their high thermal conductivity and their capability to substantially enlarge the heat transfer area to the cooling circuit. Here, a divertor target concept based on a heat pipe design is introduced being able to dissipate heat fluxes of up to 20 MW/m2. It consists of a 230 mm long water-based heat pipe with a capillary structure

Shattered pellet injection (SPI) is the technology chosen for the ITER Disruption Mitigation System and is explored at several fusion research devices, like DIII-D and JET and J-TEXT. The ITER disruption mitigation strategy relies on multiple injections to achieve RE (runaway electron) avoidance with optimum TQ (thermal quench), CQ (current quench) durations to adequately reduce wall loads. To demonstrate

A new in-pile experimental loop has been established in CMRR for ceramic tritium breeder irradiation. The most typical of this loop is the capability of online reloading of irradiated samples. The irradiation temperature in the pellets bed can be controlled within the range of 300～750 °C, with a spatial inhomogeneity of ±10 °C. Up to now, a series of irradiation experiments have been carried out on

Tritium release into the primary coolant of the JMTR and the JRR-3 M had been studied, and it is found that tritium recoil release from the chain reaction of beryllium neutron reflectors is dominant. To prevent the tritium recoil release, Al, Ti, V, Ni and Zr are selected as the candidate tritium recoil barrier materials in this feasibility study. It is clear that 20∼40 μm thickness is required depending

This paper describes ITER PF converter system architecture in details. According to its real-time and reliability characteristics, ITER PF converter system uses an industrial Ethernet fieldbus based on EtherCAT protocol with high real-time communication performance to develop its remote monitoring system. This system follows ITER design specification, it realize the real-time monitoring of thousands

The cellular structure is one of the reasons for the improvement of 316 L performance which fabricated by additive manufacturing technology. However, the cellular structure tends to disappear at high temperatures. In this study, oxide nanoparticles were introduced into the 316 L matrix to improve the thermal stability of the cellular structure, this material was successfully fabricated by mechanical

A new laboratory-scale linear plasma device, PSIEC (Plasma-Surface Interaction system under Extreme Conditions), has been designed and constructed at Hefei University of Technology, China, to study plasma-material interactions for fusion reactor application. The PSIEC facility can generate continuous plasmas of hydrogen, deuterium, helium, argon and nitrogen. The electron density of these plasmas ranges

A stable uniaxial tensile testing method has been proposed for small specimens that can obtain stress-strain relations and strength of metallic materials. Based on energy equivalent assumptions, a semi-analytical model with five parameters has been put forward. Those parameters can be determined by a small amount of finite element analysis (FEA), and more extensive numerical validation is carried out

The exhaust detritiation system (EDS) has been operating for the deuterium plasma experiment in large fusion test device since 2016. The EDS consists of two systems: the molecular sieves (MS) type for plasma exhaust gas, the polymer membrane (PM) type for vacuum vessel purge gas during the maintenance activity. The tritium removal performance of the EDS was evaluated for four years. As the operation

The divertor is one of the key plasma facing components (PFCs) in fusion device and shall exhaust high heat flux coming from the plasma. The paper presents an engineering design solution accommodating the physical requirements and the removal of 10 MW/m2 steady heat flux. The proposed divertor design employs reliable PFCs structure which can be manufactured by proven mature technologies. There are

To evaluate the capability of electron cyclotron emission (ECE) diagnostic in the Chinese Fusion Engineering Testing Reactor (CFETR), ECE spectra have been simulated using the code SPECE. The results indicate that measurements of the 2nd harmonic X-mode and the 1st harmonic O-mode ECE spectra combined are capable of providing the information of local electron temperature with fairly good spatial resolution

We report about the soft X-ray (SXR) tomography system which is being developed on Keda Torus eXperiment (KTX). The tomographic system aims at studying three dimensional effects in reversed field pinch with high plasma current, particularly in quasi-single-helicity state (QSH). The view distribution of the probes has been carefully designed to reflect the experimental constraints. Bayesian experimental

The superconducting DTT magnetic system needs a quench detection systems (QDSs) fast enough to trigger the dumping of the magnetic energy in case of quench and avoid irreversible damage of the cable systems. With this aim, a primary system based on the detection of the resistive voltage associated with the quench offers the best quench detection guarantees. The tokamak environment is affected by several

This report summarizes the status of several R&D tasks devoted to characterization of the basic behavior and definition of some features of the RF discharge mirror cleaning systems for ITER spectroscopy diagnostics. First results of mirror cleaning system engineering development and its implementation on ITER are described. Key requirements and specifications for such mirror cleaning systems for ITER

MITICA, the full scale prototype of ITER Heating Neutral Beam Injector required to heat up ITER plasma with 16.5 MW injected power, is under realization at the Neutral Beam Test Facility (NBTF) in Padova (Italy) with the contributions of JApanese and EUropean Domestic Agencies (JADA and EUDA, respectively). The objective of MITICA is to produce a 16.5 MW neutral beam, obtained by accelerating negative

The use of real-time tomography at JET opens up new possibilities for monitoring the plasma radiation profile and for taking preventive or mitigating actions against impending disruptions. By monitoring the radiated power in different plasma regions, such as core, edge and divertor, it is possible to set up multiple alarms for the radiative phenomena that usually precede major disruptions. The approach

Understanding the physics of fast ions in a fusion plasma is widely considered as one of the crucial tasks for the reliable operation of fusion tokamak reactors. Measurements on tokamaks have shown that gamma rays are produced when fast ions react with either plasma fuel ions or with the plasma impurities such as beryllium, carbon and oxygen. The spectroscopy of these gamma rays can be used for measurement

As a critical part of a breeding blanket of fusion reactor or DEMO, it is a common method to design and optimize the first wall separately before the mechanical evaluation of the fully integrated blanket. To understand the difference between the simplified first wall models and the fully integrated first wall/blanket model, the differences of the mechanical behavior between the simplified and integrated

In this paper, computational fluid dynamics (CFD) software ANSYS Fluent was used to study the influence of Laser Entrance Holes (LEHs) film transmissivity on the indirect-drive cryogenic inertial confinement fusion (ICF) targets after the shield is removed. The results show that when LEHs film transmissivity (tLEHs) increases from 0.01 to 1 under the steady-state calculation, the maximum temperature

A tomographic reconstruction method is developed for the visible light imaging diagnostic system of Experimental Advanced Superconducting Tokamak (EAST) to provide the emissive profile on the poloidal cross section of the divertor region. Weight matrix is calculated by a ray tracing method according to the optical path of the visible light imaging diagnostic system in the delicate three-dimensional

Precipitate phases in a reduced activation ferritic/martensitic steel CLAM (HEAT-0912) were observed and analyzed using transmission electron microscopy and energy dispersive spectroscopy. The steel samples were normalized at 980 °C for 30 min and then tempered at 760 °C for 90 min followed by an air cooling. Through the selected-area electron diffraction pattern analyses and energy dispersive X-ray

Chinese Fusion Engineering Test Reactor (CFETR) is an ITER-like test superconducting TOKAMAK fusion reactor. In CFETR, blankets face core plasma directly and are in charge of tritium breeding and neutron shielding, while the condition of superconducting coils needs to be maintained in a low temperature for providing a steady magnetic field to torus. Thus, doing neutronic analyses for blankets and superconducting

The neutron spectrum is one of the essential parameters in the design of advanced nuclear energy systems. Neutron spectra of different nuclear energy systems have wide differences so that accurate regulation of a specific neutron spectrum is of great significance to its development. In this research, an adaptive optimization reverse regulation method for neutron spectrum based on standardized response

The dissimilar welds joint of 2.25Cr-1Mo steel and lean duplex LDX2101 steel was prepared using the Gas Tungsten Arc Welding (GTAW) process and employing the Inconel 617 filler. The microstructural stability of the dissimilar welds joint (DWJ) was studied using the optical microscope and field emission scanning electron microscope (FESEM) for as-welded (AW) and post-weld heat treatment (PWHT) condition

In this work the capabilities of RELAP5-3D are tested against the experimental data gathered in the first experimental campaign on THALLIUM (Test HAmmer in Lead LIthiUM). THALLIUM, a LiPb facility, was built in 2015 at ENEA R.C. Brasimone with the aim to study the In-box LOCA by reproducing the LiPb loop of the HCLL TBS and the pipe forest at 1:1 scale. HCLL TBS (Helium Cooled Lithium-Lead Test Blanket

In solid tritium breeder blanket, the packing structures of the tritium breeder pebble bed and the neutron multiplier pebble bed are very important to analyze the tritium breeder ratio and the heat and mass transfer process in the pebble bed. In this study, the numerous simulations were conducted to investigated the effects of bed dimension, friction coefficient between pebbles, pebble size distribution

Tritium permeation barrier (TPB) coating technology is one of the most effective methods to solve the problem of permeation and leakage of tritium through the structural materials to the coolant, environment, and other functional components in tritium breeder blanket modules (TBMs). Alumina coatings prepared by in-situ oxidation of aluminized steel shows a high permeation reduced factor (PRF) and self-healing

The in-box Loss of Coolant Accident is a major safety concern for the Water Cooled Lithium Lead Breeding Blanket design. SIMMER-III code has been modified by University of Pisa and ENEA C.R. Brasimone to perform deterministic safety investigation of such accidental scenario. Thanks these modifications, this version of the code has unique features for dealing with the PbLi water reaction related phenomena

In order to study the adaptation of the semiconductors for the special power supply application, a capacitive pulsed power supply is designed by using the pulse thyristors. The modular integrated method is adopted to redesign the main circuit of the power supply. The synchronous triggering of multiple sets of thyristors enables high current output of the pulsed power supply, which can avoid excessive

A pyro-breaker composed of insulating and switching sections was designed in this work. The switch was actuated by the detonation of an explosive, which serves as the backup switch in quenching protection system (QPS) of comprehensive research facility for fusion technology (CRAFT). Computational investigation on the underwater explosively actuated switching was carried out in this study. The pressure

COntrol, Data Acquisition and Communication (CODAC) real-time software codes are key elements for the operation of a fusion device both for the machine protection and for the optimization of the experiments. In 2013, following the WEST (W -for tungsten- Environment Steady-state Tokamak) upgrade of Tore-Supra, the whole legacy acquisition system has been re-factored. The WEST CODAC framework which inherited

For the fusion reactor ITER, a mandatory monitor of the fusion device and plasma will be performed with optical diagnostic systems. For the metallic first mirrors, the recovery of the reflectivity losses due to dust deposition is proposed to be carried out for 14 different optical diagnostic systems by the plasma cleaning technique. In this work, we studied the influence of the electrode area on the

EAST Articulated Maintenance Arm (EAMA) is a 8-DOF redundant articulated serial manipulator utilized to conduct maintenance tasks in EAST (Experimental Advanced Superconducting Tokamak). Due to the redundancy and total length (8.7395 m) of manipulator and the narrow space in the CASK and the curved vacuum vessel(VV), it is difficult to perform online collision-free end effector trajectory tracking

A neural network model of the effects of neutral beam injection on DIII-D has been developed. The training and testing data used by the model have been generated by the NUBEAM module of TRANSP for experimental discharges from the 2018 DIII-D campaign. Using a principle component analysis to reduce the dimensionality of profile data, the model has been shown to reproduce the results of the Monte Carlo

This paper introduces key technologies and main advances in the development of the 1/8 vacuum vessel (VV) sector mockup of China fusion engineering test reactor (CFETR). The contour accuracy of each VV shell is controlled within ±2 mm through hot forming, solution treatment, and correction of each hyperboloid shell, and cold forming of the single-curved shell. The welding structure between the 1/8

The breeding blanket will be the most important functional component generating tritium as fuel for a nuclear fusion reactor to maintain the D-T reactions. It is very significant to investigate the tritium generation rate and its distribution in a blanket system to achieve an effective blanket concept before any engineering activities startup. A neutronics probe ball method is proposed to estimate

One of the most crucial issues in the design of the ITER machine is the electromagnetic (EM) loads associated with eddy currents induced in the conducting structures during plasma disruptions. The ITER database contains tens of calculation scenarios for possible plasma disruption events. The duration of these scenarios is usually limited by the end of the plasma current quench when the toroidal plasma