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Masaharu Fukuyama (Department of Advanced Energy Engineering Science IGSES Kyushu University)19/09/2018, 11:00
A transmission line has been newly developed in QUEST spherical tokamak (ST) for the highly efficient electron cyclotron heating and current drive (ECHCD) experiments with a 28 GHz gyrotron. Waves in plasmas can be described in terms of two eigenmodes, so-called extra-ordinary/ordinary modes, coming from anisotropic property of the electron motion on the magnetic-field direction. The modes...
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Mikhail Gryaznevich (ST40 Tokamak Energy Ltd)19/09/2018, 11:00
Spherical Tokamak (ST) path to Fusion has been proposed in [1] and experiments on STs have demonstrated feasibility of this approach. Advances in High Temperature Superconductor technology [2] allows significant increase in the Toroidal field which was found to improve confinement in STs. The combination of the high beta, which has been achieved in STs [3], and high TF that can be produced by...
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Denis Ivanov (NRC Kurchatov Institute)19/09/2018, 11:00
The next stage of the upgrade of T-15MD machine (R=1.5m, a=0.67m, B≤2T, Ipl=2MA) to the superconducting one (excluding the limits of pulse duration) with basically the same geometry is suggested. The estimations show the possibility to make a toroidal magnet with aspect ratio A=2.2, magnetic field on the axis Bo=5T, maximum magnetic field Bm= 12.5T. Such increase in Bo provides the possibility...
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Dr Albrecht Herrmann (ASDEX Upgrade Max-Planck-Institut fŸr Plasmaphysik)19/09/2018, 11:00
ASDEX Upgrade is a tokamak that can be operated with the strike lines in the upper and/or the lower divertor. In 2016 a project was started to develop and install a new upper divertor with internal coils and an in-vessel cryo pump. The aim is to investigate advanced magnetic configurations that may facilitate the access to detachment via an enhanced flux tube expansion and/or connection...
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Paula Sirén (VTT Technical Research Centre of Finland)19/09/2018, 11:00
New development steps of ASCOT and AFSI (ASCOT Fusion Source Integrator) based synthetic neutron diagnostics and validation at JET are reported in this contribution. Synthetic neutron diagnostics are important not only in existing tokamaks, where they are used to interpret experimental data, but also in the design of future reactors including ITER, DEMO and beyond, where neutrons are one of...
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JEAN-CLAUDE HATCHRESSIAN (IRFM CEA)19/09/2018, 11:00
Actively Cooled Plasma Facing Components (ACPFC) are required to allow for long plasma discharges in magnetic fusion devices. Prior to their installation, the integrity of ACPFC has to be checked under relevant experimental conditions in order to prevent serious water leaks in the vacuum vessel.
Since 1990, the French Magnetic Fusion Research Institute (CEA/IRFM) has developed specific leak...
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Roberto Ambrosino (DIETI University of Naples Federico II - Consorzio CREATE)19/09/2018, 11:00
The European roadmap to the realisation of fusion energy has identified a number of technical challenges and defined eight different missions to face them. Mission 2 ‘Heat- exhaust systems’ addresses the challenge of reducing the heat load on the divertor targets. Divertor Tokamak Test (DTT) facility [1]-[2] has been launched to investigate alternative power exhaust solutions for DEMO. DTT...
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Dr Zhengping Luo (Institute of Plasma Physics Chinese Academy of Sciences)19/09/2018, 11:00
Advanced magnetic divertor configuration is one of the attractive methods to spread the heat fluxes over divertor targets in tokamak because of enhanced scrape-off layer transport and an increased plasma wetted area on divertor target. Exact snowflake (SF) for EAST is only possible at very low plasma current due to poloidal coil system limitation. EAST can be operated in quite flexible plasma...
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Dr Shigetoshi Nakamura19/09/2018, 11:00
In order to reduce thermal radiation to cryogenic components, vacuum vessel (VV), ports and cryostat are covered with thermal shield in JT-60SA. The design requirements are follows;
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- Surface temperature shall be < 100K during operation and < 140K during VV baking.
- It should have integrity against electro-magnetic force during plasma disruptions and seismic load (0.6G in horizontal and 0.4G... -
Dr Matteo Vallar (Consorzio RFX ENEA)19/09/2018, 11:00
The TCV tokamak contributes to physics understanding in fusion reactor research by a wide set of experimental tools, like flexible shaping and high power ECRH. Plasma regimes with high pressure, a wider range of temperature ratios and significant fast-ion population are now attainable with the TCV heating system upgrade. A 1 MW, 25 keV deuterium heating neutral beam (NB) has been installed in...
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Paolo Tinti (Consorzio RFX)19/09/2018, 11:00
The first ever built full-scale prototype of the ITER heating neutral beam injector is the MITICA experiment at PRIMA Neutral Beam Test Facility, under realization in Padua, Italy.
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This experiment consists of several in-vessel components: most of them are actively cooled by a large cooling plant, still under construction. Coolant is deionized water produced by a Chemical Control System... -
Dr Stefano Alberti (Swiss Plasma Center EPFL)19/09/2018, 11:00
The upgrade of the EC-system of the TCV tokamak has entered in the realization phase as part of a broader upgrade of TCV[1]. The first of the two MW-class, dual-frequency gyrotrons (84 or 126GHz/2s/1MW) has been delivered by Thales Electron Devices and the full commissioning and characterization is expected to be completed during the first half of 2018. The design of this gyrotron includes new...
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Jean-Michel BERNARD (IRFM Commissariat ˆ l'Žnergie atomique et aux Žnergies alternatives)19/09/2018, 11:00
Before their installation and commissioning in the tokamak, WEST ICRF launchers undergo two categories of pre-qualifications tests. These tests aim at accelerating the commissioning of the launchers in the tokamak.
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The first category of tests is milliwatt-range radio-frequency (RF) experiments which allow checking the launchers coupling capabilities, impedance matching and their... -
1062. P3.015 Numerical analyses and tests for optimized and enhanced heat transfer solutions in DEMOGiulio Gambetta (Dipartimento Ingegneria Industriale universitˆ degli Studi di Padova)19/09/2018, 11:00
A set of novel design solutions for high performance cooling systems have been developed and tested by Consorzio RFX, achieving, with experimental tests, the challenging heat transfer conditions foreseen for Heating and Current Drive Systems of present and future nuclear fusion devices.
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The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), has the triple... -
Humberto Torreblanca (Energy and Advanced Concepts General Atomics)19/09/2018, 11:00
A comb-line antenna to demonstrate efficient off-axis non-inductive current drive from the absorption of toroidally directed very high harmonic fast waves is being designed and built for DIII-D [1].
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The antenna consists of a toroidal array of 30 modules, each 5 cm wide by 21 cm tall, so that the array spans 1.5 m on outer wall just above the tokamak midplane.
This antenna will be fed with 1... -
Silvio Ceccuzzi (ENEA)19/09/2018, 11:00
An Italian Divertor Tokamak Test (DTT) facility has been proposed to tackle a major mission of the European roadmap to fusion electricity, i.e., the problem of power exhaust. DTT is conceived to accomplish very different magnetic configurations and to reproduce edge conditions as close as possible to DEMO, allowing a reactor-relevant exploration of alternative power exhaust solutions in an...
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Donghui Xia (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics Huazhong University of Science and Technology)19/09/2018, 11:00
To meet requirements of heating and physics experiments on J-TEXT, we are developing a 105GHz/500kW/1s electron cyclotron resonance heating (ECRH) system. With the toroidal field of about 2T for normal discharges on J-TEXT, this system will mainly work at the second extraordinary mode. The ECRH system consists of a Gycom gyrotron with a superconducting magnet and related power supplies, a 30m...
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Braj Shukla (ECRH Institute for Plasma Research)19/09/2018, 11:00
The Aditya tokamak has been upgraded to Aditya-U by changing it’s vacuum vessel from rectangular to circular to accommodate the diverter coil for shaped plasma. The tokamak has been commissioned and now operating with routine plasma experiments.
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The 42GHz ECRH (Electron Cyclotron Resonance Heating) system has been integrated with the tokamak. The system is capable to deliver 500kW power for... -
Riccardo Nocentini (ITER Technology and Diagnostics Max Planck Institute for Plasma Physics)19/09/2018, 11:00
Within the framework of the ion source development for the ITER and DEMO Neutral Beam Injection (NBI) systems, IPP Garching has recently upgraded the radiofrequency-driven negative ion source testbed BATMAN. One of the requirements for the ITER NBI system is to produce a beam power density homogeneity above 90% over its large extraction area of about 0.2 m2. This requirement is going to be...
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Niek den Harder (ITER Technology Diagnostics Max-Planck-Institut fŸr Plasmaphysik)19/09/2018, 11:00
Neutral Beam Injection is the main heating system on a variety of fusion devices, and will be the main heating system of ITER. Especially during high-power operation in long pulse devices, it is important that the losses in the beamline are well quantified. There are several types of beamline losses, such as geometrical losses, due to scraping of the beam on apertures, and reionisation losses,...
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Takayuki Kobayashi (National Institutes for Quantum and Radiological Science and Technology)19/09/2018, 11:00
Wide range poloidal (~60°) and toroidal (~30°) beam steering capability and the reliability for high-power (0.8 MW/waveguide), long-pulse (100 s) operation are required for the launcher of the Electron Cyclotron Heating and Current Drive (ECH/CD) system in JT-60SA (Super Advanced). The two directional beam steering launcher has been designed by a linear motion antenna concept, which has an...
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Yikun Jin (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics Huazhong University of Science and Technology)19/09/2018, 11:00
Based on the Gycom gyrotron of a diode type with a single-stage depressed collector, a 105GHz/500kW/1s electron cyclotron resonrance heating sytem is being developed on J-TEXT tokamak. To modulate output power of the gyrotron, we designed a 33kV/1A anode power supply based on the pulse step modulation technology. The power supply consists of 40 modules with output voltage of 800 V and 10...
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Gustavo Granucci (IFP-CNR)19/09/2018, 11:00
The purpose of the italian Divertor Tokamak Test (DTT, 6 T, 5.5 MA, R0 = 2.08 m, a = 0.65 m for ~ 100 s) is to study power exhaust and divertor load in an integrated plasma scenario. To accomplish this mission DTT will be equipped with 45 MW of additional heating power to fulfill a PSEP/R ≥ 15 MW/m studing alternative divertor configurations in view of ITER operations and DEMO design. The...
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Bernd Heinemann (ITED Max-Planck-Institut fŸr Plasmaphysik)19/09/2018, 11:00
Neutral Beam Injection (NBI) for ITER shall deliver in total 33 MW heating power to the plasma with two injectors at a beam energy of 1 MV. Taking neutralisation efficiency and all losses along the beam path into account a negative ion current density of 329 A/m2 (H- for 1000s) and 286 A/m2 (D- for 3600s) has to be extracted from each ion source (size 1 × 2 m2) with a beam non-uniformity below...
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Luca Grando (Consorzio RFX)19/09/2018, 11:00
The ITER project requires at least two Heating Neutral Beam Injectors (NBIs), each accelerating up to 1MV a 40A beam of negative H-/D-.ions, to deliver to the plasma a total power of about 33 MW for one hour.
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Since these requirements have never been experimentally met, it was recognized necessary to build-up a test facility, named PRIMA including both a full-size negative ion source (SPIDER -... -
Mi Joung (National Fusion Research Institute)19/09/2018, 11:00
Since KSTAR first plasma operation, ECH played a key role in obtaining various experimental results such as ECH preionization, ECH-assisted startup, plasma rotation study, impurity transport study, high poloidal beta operation, and long pulse operation. The main heating systems in KSTAR are NBI and ECH which are planed to provide 12 MW NBI by 2019 and 6 MW ECH by 2020 to prepare long pulse,...
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Yizhe Tian (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics Huazhong University of Science and Technology)19/09/2018, 11:00
To match the electron cyclotron wave with the plasma efficiently, we design two polarizers including a linear polarizer and an elliptical polarizer for the 105 GHz electron cycltron resonance heating system on J-TEXT. The linear polarizer is mainly used to change the rotation angle of the wave, while the ellipticity of the wave is regulated by the ellptical polarizer. The sinusoidal grooves...
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Gian Luca Ravera (Fusion and Nuclear Safety Departement ENEA Frascati)19/09/2018, 11:00
Electron Cyclotron Resonance Frequency (ECRF) systems in future fusion devices, like the DEMO-nstration reactor, foresee an operational frequency in the range 230-280 GHz to match the plasma characteristics. The Cyclotron Auto Resonance Masers (CARM) characterized by a high value of a frequency Doppler up-shift, could represent an alternative to gyrotrons and the design of a 250 GHz, 0.5 MW...
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Mauro Cappelli (FSN ENEA)19/09/2018, 11:00
Purpose of this work is to describe the DONES Central Instrumentation and Control System (CICS). A functional definition of the main systems will be given, together with a general overview of the current status of the CICS and the differences with respect to the corresponding system developed during the IFMIF-EVEDA phase. The overall architecture of the Control System, the definition,...
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Jian Liu (southwestern institute of physics)19/09/2018, 11:00
Due to the high precision requirements of HL-2M tokamak sub-components assembly, the survey control network with high precision should be established. With high accuracy distance measurement of laser tracker system and distance intersection method, the local coordinates of reference points(or the relative locations of the reference points) in the survey control network are calculated. There...
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Kazuo Nakamura (Research Institute for Applied Mechanics Kyushu University)19/09/2018, 11:00
In the EC-driven (8.2 GHz) steady-state plasma on QUEST, plasma current seems to flow in the open magnetic surface in the outside of the closed magnetic surface in the low-field region according to plasma current fitting (PCF) method. First, plasma equilibrium solution was fitted assuming all plasma current is flowing in the inside of the LCFS. It was solved within isotropic pressure profile...
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Manabu takechi (Naka Fusion Institute Fusion Energy Research and Development Directorate National Institutes for Quantum and Radiological Science and Technology (QST))19/09/2018, 11:00
Disruption simulations with DINA code are performed for JT-60SA design. The simulation results have been applied to design of many components, not only for the vacuum vessel and in-vessel components, but also for peripheral components. For instance, for design of in-vessel coils, the stabilizing plate and magnetic sensors, EM force induced by hallo current and eddy current at disruption were...
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Rodrigo Castro (Laboratorio Nacional de Energ’a de Fusion CIEMAT)19/09/2018, 11:00
ITER’s CODAC archiving system manages currently three different sets of data: DAN, SDN and PON, that correspond with the data that is transmitted by different networks: Data Archiving Network (data produced by data acquisition, diagnostics and data analysis), Synchronous Data Network (real time control network), and Plan Operational Network (control data). In this sense, ITER’s CODAC data...
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Xueqing Zhao (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics Huazhong University of Science and Technology)19/09/2018, 11:00
On tokamaks, there are many diagnostics, which need real-time data acquisition and processing to provide useful information for plasma control. Some of the diagnostics required fast processing of multiple very high sampling rate signals. It is often difficult to achieve even with modern multi-core CPUs. This is due to moving large amount of data from the digitizers to the system ram would hurt...
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Jesús Vega (Laboratorio Nacional de Fusi—n CIEMAT)19/09/2018, 11:00
Recent research on disruption prediction shows that predictors based on analysing the amplitude evolution of magnetic signals outperforms the results obtained by using simple thresholds. To accomplish this, the disruptive and non-disruptive information of discharges can be compressed into two centroids in a particular parameter space (PS). During a running discharge, points in the PS are...
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Wei Zheng (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics Huazhong University of Science and Technology)19/09/2018, 11:00
ITER CODAC is the most sophisticated tokamak control and data acquisition system. The core of ITER CODAC is built around the EPICS toolkit. EPICS is very mature in accelerator community. However, there are still works trying to improve existing control system software like tango and EPICS 7 mainly driven by the needs of more flexible system and development of computer technology. This paper...
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Marcello Cinque (Electrical Engineering and Information Technology Federico II University of Naples)19/09/2018, 11:00
The Plasma Control System (PCS) is one of the main ITER systems. It is in charge of running the plasma discharge, by receiving data from the real-time diagnostics, and by computing the commands to be processed by various plant systems to act on the plasma (e.g., the power supplies of the poloidal field coil circuits and the additional heating systems).
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To this aim, the PCS will implement... -
Gerardo D'Elia (DTE-ICT-HPC ENEA)19/09/2018, 11:00
The four-barrel, two-stage gun Ignitor Pellet Injector (IPI) was developed in collaboration between ENEA and ORNL to provide cryogenic Deuterium pellets of different mass and speed to be launched into tokamak plasmas with arbitrary timing. The prototype injector is presently located at Oak Ridge (TN, USA), and is normally operated locally through a control and data acquisition system developed...
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Sangwon YUN (NFRI)19/09/2018, 11:00
The control systems for the Korea Superconducting Tokamak Advanced Research (KSTAR) have been implemented based on the Experimental Physics and Industrial Control System (EPICS) which is a framework for control systems widely used on accelerators and fusion devices including the ITER, an international fusion experiment.
In terms of operation and maintenance of KSTAR control systems, there was...
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Jari Varje (Department of Applied Physics Aalto University)19/09/2018, 11:00
The limits for the heat loads on the DEMO first wall are significantly stricter compared to those of ITER due to cooling and breeding blanket requirements. In addition to the thermal particle and radiation loads, fast particles in the form of fusion alphas and NBI ions with high energies can escape the confinement due to various magnetic perturbations and produce a significant heat load on the...
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Marco Ariola (Dipartimento di Ingegneria universitˆ di Napoli Parthenope - CREATE)19/09/2018, 11:00
This paper describes the preliminary design of a position, current and shape control for DEMO tokamak. This preliminary design relies on the availability of magnetic sensor measurements for the vertical position and for the plasma-wall gaps. The controller is designed basing on the CREATE-L model of the DEMO 2017 Single-Null (SN) configuration, and then is tested using the nonlinear evolution...
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Dr Giuliana Sias (Department of Electrical and Electronic Engineering University of Cagliari)19/09/2018, 11:00
Plasma behavior in the SOL of tokamaks is driven by turbulence in the edge region where density and temperature gradients are large. This generates intermittent structures of increased density and temperature known as filaments, which extend along the magnetic field lines. The protection of plasma facing components in the next step devices is a primary concern. In this context, the...
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Alexander Huber (Institut fŸr Energie- und Klimaforschung Plasmaphysik Forschungszentrum JŸlich GmbH)19/09/2018, 11:00
The risk of damaging the metallic PFCs on JET-ILW by beryllium melting or cracking of tungsten owing to thermal fatigue requires a reliable active protection system: it shall avoid damage to the plasma-facing components (PFCs). To address this issue, a real-time protection system comprising newly installed imaging diagnostics, real-time algorithms for hot spot detection and alarm handling...
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ANDREY ALEKSEEV (Fusion Centre)19/09/2018, 11:00
Single crystal Molybdenum is one of the most promising materials for the First Mirror (FM) for ITER optical diagnostics due to high resistance to erosion under the neutral atom bombardment. Other advantages are: low CTE, high thermal conductivity, good mechanical properties at elevated temperatures. The FMs are normally located in the front-end of ITER port plugs, being subject to the...
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Dr Tamás Szabolics (Plasma Physics Wigner Research Centre for Physics Hungarian Academy of Sciences)19/09/2018, 11:00
In the past two campaigns of Wendelstein 7-X stellarator the overview video diagnostics played an important role in the daily experiments. The current software implementations went through numerous improvements and changes according to the continuously changing requirements. However, while the control software could handle all the needs, the changes reached a point where the redesign and...
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Dr Santiago Cabrera (Laboratorio Nacional de Fusi—n CIEMAT)19/09/2018, 11:00
The Plasma Position Reflectometry (PPR) diagnostic system (PBS 55F3) is planned to provide information related to the edge electron density profile and plasma position at four defined locations distributed both poloidally and toroidally in the ITER vacuum vessel.
The sections of the ex-vessel transmission lines (TL) in the Gallery, between the two secondary confinement barriers, are...
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Roch Kwiatkowski (Department of Nuclear Techniques Equipment National Centre for Nuclear Research)19/09/2018, 11:00
The paper reports on measurements of neutron flux emitted from a 14 MeV DT neutron generator. Such devices are widely used in material sciences, industry, medicine, etc. The used neutron generator (NSD-35) provides a controllable emission of a stabilized neutron flux, up to about 2∙E8 neutrons per second in 4pi angle. According to manufacturer, more than 90% of the neutrons emitted are 14-MeV...
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Andrey Ushakov (TNO P.O. Box 155 NL-2600 AD)19/09/2018, 11:00
One of the important aspects of the plasma cleaning of the front-end mirrors (FM) in ITER UWAVS diagnostics is to understand surface roughness after multiple cleaning runs and to minimize possible contamination due to unwanted sputtering of the mirror surface and neighboring walls. The capacitive coupled RF 30-60 MHz is a candidate for the UWAVS FM cleaning. It generates ion fluxes of tens of...
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Alexey Sushkov (Kurchatov Complex of Fusion Power and Plasma Technology National Research Centre Kurchatov Institute)19/09/2018, 11:00
Magnetic diagnostics plays an important role in tokamak operation. Magnetic data are used for real-time control of plasma current, shape and position and for post-discharge analysis of magnetohydrodynamic plasma instabilities and equilibrium reconstruction. The magnetic diagnostic of the T-15MD will consist of more than 500 inductive sensors of various types: poloidal flux loops, saddle loops,...
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Gonzalo Farias (Ingenier’a Electrica Pontificia Universidad Catolica de Valparaiso)19/09/2018, 11:00
Over the years, humanity has needed energy constantly due increase both population and the technology. That’s why conventional methods of energy production are not enough to cover new demands especially in environmental area because of pollution generated.
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Energy generated by nuclear fusion solve both problems so achieve full control over it internal process, this implies an analysis over... -
Xi Jiang (FZJ)19/09/2018, 11:00
One of the keen interests of plasma-facing material diagnostics is the in situ measurement of fuel content on the material surface and monitoring the outgassing process in the early stage after plasma exposure. The W samples are exposed to D plasma (with a typical fluence of 5×1025 D/m2) on the linear device PSI-2. Laser induced breakdown spectroscopy (LIBS) is employed as the detecting...
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Gábor Anda (Plasma Physics HAS WIGNER RCP)19/09/2018, 11:00
A 60 keV neutral Alkali beam system was designed, built and installed for beam emission spectroscopy measurement of edge plasma on W7-X.
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The injector consists of three parts: a recently developed thermionic (lithium or sodium) ion source (j≥2.5mA/cm2), a high focusing efficiency ion optic (~50% of the extracted current can be found in the plasma) and a newly developed recirculating... -
Shifeng Mao (School of Physical Sciences University of Science and Technology of China)19/09/2018, 11:00
Charge eXchange Recombination Spectroscopic (CXRS) diagnostic system was successfully applied on EAST campaign. The CXRS located at D port was designed to focus on the tangential neutral injection beam from Port A on EAST. However, the tangential beam and the perpendicular beam are always injected into the plasma at the same time for the better heating and current driving. Therefore, spectrum...
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Wei Tao (Department of Engineering and Applied Physics University of Science and Technology of China)19/09/2018, 11:00
The beam emission spectrometer that shares the same collection optics with the existing core charge exchange recombination spectroscopy (cCXRS) on EAST has recently been upgraded. The enhanced system allows the simultaneous measurements of red- and blue-shifted parts of the Doppler spectrum as well as the active charge exchange line (Dα n = 3-2 656.1nm) from the main ions. One curved strip on...
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Hiroshi Tojo (National Institutes for Quantum and Radiological Science and Technology)19/09/2018, 11:00
A Thomson scattering system providing electron temperature and density profile requires a high energetic laser. YAG lasers amplified by flash lamps (~50-100 Hz of repetition frequency with a few Joules output) sometimes suffer from wavefront distortion and peaked beam profile. The wavefront distortion deteriorates beam profile in the far field. Since a focusing lens is used toward the plasma...
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Bruno Santos (Instituto de Plasmas e Fusˆo Nuclear Instituto Superior TÖcnico Universidade de Lisboa)19/09/2018, 11:00
The ITER Radial Neutron Camera (RNC) Data Acquisition (DAQ) prototype is based on the PCIe protocol as the interface to be used between the I/O unit and the host PC, enabling for the scalability of the final RNC DAQ system and allowing a sustainable 2 MHz peak event to cope with the long plasma discharges, up to half an hour.
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The high performance computer receives the acquired data through the... -
Dr Petr Vondracek (Institute of Plasma Physics of the CAS)19/09/2018, 11:00
A new fast divertor infra-red (IR) thermography system was put into operation at COMPASS. It provides full radial coverage of the bottom open divertor with pixel resolution ~ 0.6-1.1 mm/px. on the target surface (0.04-0.12 mm/px. mapped to the outer midplane) and time resolution better than 20 µs. This setup provides unique capabilities for heat flux profile measurements simultaneously in the...
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Yixuan Zhou (Department of Engineering and Applied Physics University of Science and Technology of China)19/09/2018, 11:00
A high throughput spectrometer has been developed for the measurement of the plasma ion temperature fluctuations on EAST tokamak. The designed spectrometer operates at the spectral range of 527.5±5 nm, then the emission lines from CVI at 529.1nm and NeX at 524.9nm can be observed simultaneously. The collimated and focus lenses are specially developed in order to realize the maximize...
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Nancy Ageorges (Kampf Telescope Optics GmbH)19/09/2018, 11:00
H-alpha and Visible Spectroscopy is one of the ITER first-plasma optical diagnostics providing full poloidal coverage of plasma scrape-off layer (SOL)by two poloidal-view channels in EP11, one tangential-view channel in EP12, and one divertor-view channel in UP02. The diagnostics is composed of several optical sub-units, which transfer the SOL image to the narrow-band filtered cameras located...
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Minghui Xia (State Key Laboratory of Advanced Electronmagnetic Engineering and Technology Huazhong University of Science Technology)19/09/2018, 11:00
Visible light high-speed imaging systems (VLHIS) are widely used in imaging and diagnosing plasmas in tokamak, e.g., plasma boundary position, structures, fast fluctuations, for its visibility and easy operation. To monitor the discharge process on J-TEXT tokamak in a high temporal resolution and study of the visible light emissivity distribution, the plasma boundary shape, a new VLHIS has...
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Johan Oosterbeek (Stellarator-Heizung und -Optimierung Max-Planck-Institut fŸr Plasmaphysik W7-X)19/09/2018, 11:00
The Wendelstein 7-X (W7-X) experiment is equipped with an Electron Cyclotron Resonance Heating installation consisting of 10 gyrotrons capable of delivering upto 7.5 MW of Electron Cyclotron Wave power at the 140 GHz resonance in the plasma. Normally, the gyrotron power is delivered in a very narrow band of several 100 MHz around the gyrotron frequency and the gyrotrons are optimized to...
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Sebastien VIVES (IRFM CEA)19/09/2018, 11:00
The equatorial visible and infrared Wide Angle Viewing System (WAVS) for ITER is one of the key diagnostics for machine protection, plasma control and physics analysis. To achieve these objectives, the WAVS will monitor the surface temperature of the Plasma Facing Components (PFCs) by infrared (IR) thermography (3-5µm range) and will image the edge plasma emission in the visible range. It will...
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Rita Pereira (IPFN IST)19/09/2018, 11:00
The ITER project aims at building an experimental fusion device, twice the size of the largest current device in operation, JET, to demonstrate the scientific and technical feasibility of fusion power. Presently, ITER is being equipped with a set of diagnostics to provide accurate measurements of plasma behavior and performance, as the neutron diagnostics. In particular, the Radial Neutron...
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Mykyta Varavin (Tokamak COMPASS Institute of Plasma Physics of the Czech Academy of Sciences)19/09/2018, 11:00
The present COMPASS tokamak at the Institute of Plasma Physics in Prague is equipped with the 2-mm interferometer, which gives a possibility to measure line average electron densities up to 1.2x1020 m-3 (the critical density for the interferometer probing waves is 2.43x1020 m-3). A high magnetic field tokamak, COMPASS-U [Panek et al., Fus. Eng.des. 123 (2017) 11-16], will be designed and built...
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Nicolò Marconato (Consorzio RFX)19/09/2018, 11:00
A major modification of the RFX-mod toroidal load assembly has been decided in order to improve passive MHD control and to minimise the braking torque on the plasma, thus extending the operational space in both RFP and Tokamak configurations. With the removal of the vacuum vessel, the support structure will be modified in order to obtain a new vacuum-tight chamber and the first wall tiles will...
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Dr Klaus-Peter Weiss (Institute for Technical Physics Karlsruhe Institute of Technology)19/09/2018, 11:00
The ITER Poloidal-Field (PF) magnet system is composed of six circular coils consisting of superconducting winding packs made up from a stack of Double Pancakes. Due to the large coil sizes the coils PF2, PF3, PF4 and PF5 are to be fabricated adjacent to the ITER site in a dedicated PF Coil fabrication building. The cold testing of the full coils PF2 – PF6 will be carried out in the same...
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Dr Stuart Dowson (CCFE/UKAEA)19/09/2018, 11:00
The Alfvén Eigenmode Active Diagnostic system (AEAD) has undergone a major upgrade and redesign to provide a state of the art excitation and real-time detection system for JET.
The new system consists of individual 4kW amplifiers for each of the six antennas, allowing for increased current, separate excitation and real time control of relative phasing between antenna currents. The amplifiers...
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Slavomir Entler (Tokamak Institute of Plasma Physics of the CAS)19/09/2018, 11:00
The magnetic diagnostic is essential for today’s tokamaks to determine the plasma position, stability, energy content and additional parameters critical for safe operation of these devices. Conventional sensors, such as the inductive sensors, have to be supplemented by steady-state magnetic sensors in devices with long pulse capability, as is planned for the ITER reactor. In ITER, the set of...
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Fabio Moro (Fusion and Technology for Nuclear Safety and Security Department ENEA)19/09/2018, 11:00
The ITER Radial Neutron Camera (RNC) is a multichannel detection system hosted in the Equatorial Port Plug 1 (EPP 1). It is designed to measure the uncollided neutron flux from the plasma, providing information on the neutron emissivity profile and total neutron strength. The RNC structure consists of two sub-systems based on fan-shaped arrays of cylindrical collimators: the ex-port system,...
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Keeman Kim (National Fusion Research Institute)19/09/2018, 11:00
Based on the Korean Fusion Energy Development Promotion Law was enacted in 2007, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. One of the key components of the K-DEMO, the superconducting magnet system consists of 16 TF (Toroidal Field), 8 CS (Central Solenoid) and 12 PF (Poloidal Field) coils. All of the TF, CS and PF coil...
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Sylvie Nicollet (IRFM CEA)19/09/2018, 11:00
The Toroidal Field (TF) system of the Tore Supra/WEST tokamak comprises 18 NbTi superconducting coils, cooled by a static superfluid helium bath at 1.8 K and carrying a nominal current of 1255 A. The 19th December 2017, at the end of plasma run #52205, the current Fast Safety Discharge (FSD) was triggered after a quench of TFC-09.
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A numerical model has been developed with SuperMagnet... -
Chiarasole Fiamozzi Zignani (FSN ENEA)19/09/2018, 11:00
Cable-in-Conduit Conductors (CICCs) are complex systems whose behaviour is not directly predictable studying their single components, and the explanation of their observed properties is not straightforward. The knowledge of the strain (Eps_th) distribution of Nb3Sn filaments in a CICC cross-section is a key parameter in understanding the performance and its evolution when the cable undergoes...
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Xiaowu Yu (Institute of Plasma Physics Chinese Academy of Sciences)19/09/2018, 11:00
The Poloidal Field(PF) coils are one of the main sub-system of ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences(ASIPP) as per the Poloidal Field coils cooperation agreement between ASIPP and Fusion for Energy(F4E).
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The PF6 coil winding is constructed from nine double pancakes (DPs) which are plane cylindrical solenoids of about... -
Yoshimitsu Hishinuma (Helical plasma research National Institute for Fusion Science)19/09/2018, 11:00
The degradation of transport current property by the high mechanical strain on the practical Nb3Sn wire is serious problem to apply for the future fusion magnet operated under higher electromagnetic force environment. Therefore, increase of the mechanical strength on Nb3Sn wire is the most important research subject. Recently, we approached to the solid solution strength process on the ternary...
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Aleksandra Dembkowska (Faculty of Mechanical Engineering and Mechatronics West Pomeranian University of Technology in Szczecin)19/09/2018, 11:00
The Central Solenoid (CS) coil of the European DEMO tokamak will consist of five modules, namely CSU3, CSU2, CS1, CSL2 and CSL3, located vertically one above the other. The central CS1 module will be subjected to the most demanding operating conditions (the highest magnetic field and mechanical loads). The design concept of the CS1 winding pack with superconductor and stainless steel grading...
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Francesco Fellin (consorzio RFX)19/09/2018, 11:00
The first ever built full-scale prototype of the ITER heating neutral beam injector is the MITICA experiment at PRIMA-NBTF, under realization in Padua, Italy.
The MITICA experiment includes many auxiliary plants; this paper is focused on the integration of the High Voltage Power Supply (-1 MV), hosted in a Faraday cage (HVD, High Voltage Deck) inside a dedicated Building at PRIMA-NBTF.
The...
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Andrea Zappatore (Dipartimento Energia Politecnico di Torino)19/09/2018, 11:00
The HTS CrossConductor (HTS CroCo) was recently proposed by Karlsruhe Institute of Technology as novel concept for the winding pack of future fusion magnets.
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The conductor concept is based on a cable-in-conduit configuration (CICC), in which 6 HTS CroCo macro-strands are twisted around a round copper core, jacketed in a stainless steel conduit and cooled by forced-flow supercritical helium at... -
Fabien Ferlay (CEA)19/09/2018, 11:00
Assembling of ITER components is a major challenge especially because of the large size and weight of its components, and the high accuracy that has to be reached. As part of the Magnets Infrastructures Facility for Iter (MIFI) agreement between CEA and IO, CEA works on the assembly sequence of ITER TF coils OIS (Outer Interface Structure) composed of shear pins and bolts. These pins and bolt...
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Dr Huan Wu (Institute of Plasma Physics, Chinese Academy of Sciences)19/09/2018, 11:00
The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. The Fusion for energy (F4E) is in charge of supplying 5 Poloidal field coils (PF2-PF6) as in-kind contributions to ITER project. In 2013, F4E commissioned the task of PF6 coil fabrication to Institute of Plasma Physics Chinese Academy of Sciences (ASIPP). The PF6 coil consists of 9 double pancakes (DPs). Before...
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Song Zhou (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics ( IFPP ) Huazhong University of Science and Technology)19/09/2018, 11:00
The high heat load on divertor target plate is one of the essential issues for future fusion reactors. In stellarator, the island divertor configuration has a long magnetic field line connection length. It is beneficial to increase the equivalent radial transport and the power decay length, and consequently reduce the peaking heat load on the divertor target plate. Therefore, it is significant...
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Dr Mattia Dan (Consorzio RFX)19/09/2018, 11:00
The Acceleration Grid Power Supply (AGPS) is a system devoted to supply the acceleration grids of the MITICA experiment, the full scale prototype of the ITER Neutral Beam Injector (NBI). The AGPS is a special switching power supply with demanding requirements: high rated power (about 55 MW), extremely high output voltage (-1MV dc), long duration pulses. The procurement of the AGPS is split in...
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Ilia Ivashov (IEK-4 Forschungszentrum JŸlich GmbH)19/09/2018, 11:00
The toroidal field (TF) coil system is one of the most mechanically stressed system in a tokamak. Structural integrity of the system must be maintained on the global and on the local scale, where the stress state in each conductor jacket as well as in insulation is to be within structural allowables. Solving this task head-on leads to very high computational demands. In this work a methodology...
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Gherardo Romanelli (ENEA)19/09/2018, 11:00
The Divertor Tokamak Test (DTT) facility is a satellite experiment in the same research and development framework of the European DEMOnstrating fusion power reactor. It shall evaluate different divertor solutions for power and particles exhaust, and shall investigate the plasma-material interaction scaled to long pulse operation. It is part of the European Fusion Roadmap and shall be...
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Lorenzo Giannini (FSN ENEA)19/09/2018, 11:00
The “Divertor Tokamak Test” facility, DTT, is a project for an experimental tokamak reactor developed in Italy, in the framework of the European Fusion Roadmap.
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In this design phase of the machine it is necessary to ensure the structural integrity of the superconducting magnets.
This work focuses on the analysis of the stresses that are generated in the central solenoid of the tokamak, the CS... -
Yong YANG (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics State Key Laboratory of Advanced Electromagnetic Engineering and Technology School of Electrical and Electronic Engineering Huazhong University of Science and Tec)19/09/2018, 11:00
Due to the high intensity stray magnetic field around the tokamak device, static magnetic field immunity test is an essential procedure to verify the reliable operation of electrical and electronic equipment nearby. For the safety and reliability concerns for the Chinese Fusion Engineering Test Reactor (CFETR) and future tokamak devices, a large-scale high-intensity static magnetic field...
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Kun Wang (Institute of Plasma Physics Chinese Academy of Sciences University of Science and Technology of China)19/09/2018, 11:00
When the quench occurrence in the operating course of the large fusion device, huge energy stored inside the magnetic load and the maximum current flow from the superconducting load can reach 100kA with maximum inductance value to be 2H that will lead an irreversible damage on the device. By dissipating the energy by means of the fast discharge resistor(FDR) system connected in series to the...
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Alessandro Anemona (Superconductivity Lab Enea)19/09/2018, 11:00
DTT is the acronym of “Divertor Tokamak Test” facility, a project for a compact but flexible tokamak reactor which has been conceived in the framework of the European Fusion Roadmap. It will be built in Italy and shall act as a satellite experimental facility to integrate the extrapolation of the ITER results to the EU-DEMO machine. It is thus mainly aimed at the exploration of different...
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Dr Carmela Ciummo (PPL Princeton University )19/09/2018, 11:00
The National Spherical Torus eXperiment Upgrade (NSTX-U) is an experimental device funded by the U.S. Department of Energy (DOE) at the Princeton Plasma Physics Laboratory (PPPL). NSTX-U (http://nstx-u.pppl.gov/home) is an upgrade of the original NSTX device that operated successfully for more than 10 years as a proof-of-principle demonstration of the ST concept.
During early phases of...
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Paolo Rossi (Department of Fusion and Technology for Nuclear Safety and Security ENEA)19/09/2018, 11:00
In the framework of the Broader Approach program, ENEA supplied the Toroidal Field (TF) coil casings for JT-60SA tokamak.
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ENEA commissioned the manufacture of the full set of eighteen casings for the integration of the TF coils plus two additional spare casings to the company Walter Tosto (Chieti, Italy).
The casing is segmented in one outboard straight leg, an outboard curved leg and three... -
Olga Ogorodnikova (Plasma Physics National Research Nuclear University ÈMEPHI)19/09/2018, 11:00
In the process of burning fusion plasmas, plasma-facing materials such as tungsten-based materials (W) will be exposed to energetic particles of hydrogen isotopes and helium (He), high heat flux, and neutrons. In this regard, a study of accumulation of hydrogen isotopes and He in W under normal operation conditions and transit events appears necessary for assessment of safety of fusion reactor...
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Dr Mehdi Firdaouss (IRFM CEA)19/09/2018, 11:00
Actively cooled plasma facing components (PFC) are often made of CuCrZr, whereas the cooling pipes are made of stainless steel. Both materials are not easily joined, and a common solution is electron beam welding, using a ring made of Inconel or Ni as intermediate.
This paper reveals the potential of using explosive welding as an alternative joining technique for multi-material transitions of...
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Masashi Shimada (Fusion Safety Program Idaho National Laboratory)19/09/2018, 11:00
Nuclear fusion promises to deliver an abundant, carbon-free and clean energy source for the future. Before the realization of nuclear fusion energy, the fusion community must solve immense technological safety challenges related to tritium permeation in materials under an extreme fusion nuclear environment. Tritium behavior in materials determines two crucial safety evaluation source terms:...
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Valentina Huber (JŸlich Supercomputing Center Forschungszentrum JŸlich GmbH)19/09/2018, 11:00
Plasma Facing Components (PFC) in JET with metal ITER-like wall are subjected to high heat fluxes which can lead to damages such as beryllium melting or thermal fatigue of tungsten. The hot spots formation at the re-ionization zones due to impact of the re-ionised neutrals injected by the heating system as well as due to RF-induced fast ion losses is recognized as a big threat due to quick...
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Giacomo Dose (Department of Industrial Engineering University of Rome Tor Vergata)19/09/2018, 11:00
In a fusion reactor, heat exhaust is one of the most challenging engineering issues, due to the high heat flux (HHF) expected on the divertor targets. The tungsten (W) monoblock design represents one of the most suitable technological solution for plasma facing components, since it has already met the ITER requirements. However, further research is required to investigate improved solutions to...
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Jan Prokůpek (Energetics and Fusion Technologies Research Centre ñe_)19/09/2018, 11:00
The ITER first wall panels are exposed directly to thermonuclear plasma and must extract heat loads of about 2 MW/m² (Normal Heat Flux) to 4.7 MW/m² (Enhanced Heat Flux). The manufacturer of the normal heat flux first wall panels shall be qualified through deep high heat flux cyclic testing campaign counting thousands of cycles within the heat flux range up to 2.5 MW/m². To ensure correct...
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Romain Monier (Centre technique Framatome)19/09/2018, 11:00
This poster describes the main steps realized for the manufacturing of a full scale First Wall panel to ITER. This full scale prototype (FSP) is foreseen to be delivered in 2019 to F4E in order to perform high heat flux tests. The dimensions of this prototype are 1360 mm x 850 mm x 500 mm. It consists of a bi-metallic support structure made from 15-25 mm thick CuCrZr alloy embedded with...
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Pavel Zacha (Department of Energy Engineering Faculty of Mechanical Engineering Czech Technical University in Prague)19/09/2018, 11:00
The water-cooled lithium lead (WCLL) blanket is considered as one of the possible candidates for the EU DEMO blanket in the present EU fusion roadmap. One of the critical points of the first wall design is the maximal allowed thermal load of the Eurofer97 steel within the limiting temperature of 550 °C. Therefore, the initial reactor geometrical concept of the WCLL blanket allows a heat flux...
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Yury Gasparyan (Plasma Physics National Research Nuclear University MEPhI)19/09/2018, 11:00
Hydrogen co-deposition with sputtered particles is one of the main channels of hydrogen isotope accumulation in today’s tokamaks. According to experiments in tokamaks and in labaratory conditions hydrogen concentration in co-deposits show that can be very high (up to tens of atomic percents) for various materials at low deposition temperature even in the case of low hydrogen solubility. This...
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Tomáš Melichar (Research centre ñe_)19/09/2018, 11:00
Helium cooled First wall (FW) is being developed within the breeding blanket (BB) workpackage of the EUROfusion project as one of FW options for the European DEMO. The helium cooling system has to be adapted to high thermal loads and at the same time to achieve reasonable hydraulic parameters. Moreover, different values of the heat fluxes are expected at the DEMO FW depending on the position...
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Arnold Lumsdaine (Oak Ridge National Laboratory)19/09/2018, 11:00
Plasma-facing components based on so-called monoblocks are planned for use in the divertor region of long-pulse plasma devices such as ITER and JT-60SA due to their capacity to handle high heat fluxes with active water cooling. The plasma-facing materials that are preferred for these monoblocks are tungsten for ITER or carbon-carbon fiber composite (CFC) for JT-60SA. The requirements for the...
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selanna roccella (Department of Fusion and Technology for Nuclear Safety and Security ENEA)19/09/2018, 11:00
The Plasma Facing Units are the components of the ITER's divertor target exposed to the plasma. PFUs are cooling pipes made of copper covered by tungsten monoblocks as armour.
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The non-destructive ultrasonic control is the simplest and most economical test for PFU control. It has also proved to be extremely reliable and accurate in identifying and sizing defects.
ENEA has been working on... -
Jiliang Wu (Institute of Materials China Academy of Engineering Physics)19/09/2018, 11:00
Tritium takes the most cost of fusion project when it has been regularly operated.In order to give a proper fuel combustion rate and recycling efficiency,it is necessary to assess the amount of hydrogen isotopes(accompanied with helium)retent in the plasma facing materials(PFM).
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A comprehensive ECR plasma system for tritium (named CEPT)is designed and built for the assessment of tritium... -
Dmitry Rudakov (Center for Energy Research University of California San Diego)19/09/2018, 11:00
The Metal Rings Campaign in DIII-D allowed for studies of tungsten sourcing and transport from poloidally localized, isotopically distinct surfaces in a low-Z background. Two 5 cm wide toroidal rings of W-coated tile inserts were installed in the lower divertor of DIII-D. The outboard (shelf) ring was coated with isotopically enriched W-182; the inboard (floor) ring used a natural W coating....
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Alexandr Vasilyev (Budker Institute of Nuclear Physics SB RAS)19/09/2018, 11:00
Transient heat fluxes up to 1 MJ·m-2 on divertor area are expected during operation of ITER. They can lead to severe erosion of plasma-facing components. Studies on tungsten damaging under thermal shocks are widespread, but they are mainly concentrated on postmortem analysis of the exposed samples. Main feature of the experiments conducting on electron beam based test facility called BETA is...
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Tatsuya Hayashi (Tokai University Hiratsuka-shi)19/09/2018, 11:00
Divertors are responsible for removing the exhaust helium ash generated after fusion in a magnetic fusion reactor. Tungsten (W) was selected as the plasma facing material in the ITER divertor region because of its high melting temperature and thermal conductivity and low sputtering erosion yield. Therefore, it is crucial to understand the behavior of hydrogen isotopes in W contained in the...
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Toshikio Takimoto (Tokai University Kitakaname)19/09/2018, 11:00
Steady-state fusion reactors and DEMO reactors will have much higher heat flux from the core than that from ITER, which itself exhibits heat flux that is several times larger than that available in the current fusion reactors. The detached plasma is effective for reducing heat load. However, since the generation of detached plasma requires to introduce a large quantity of gas, there is...
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Dr Eliseo Visca (FSN ENEA)19/09/2018, 11:00
The scope of contract F4E-OPE-138 Lot 1, assigned to Ansaldo Nucleare S.p.A (ANN) by Fusion for Energy, the EU-Domestic Agency, is the fabrication and qualification of a representative full scale prototype of the International Thermonuclear Experimental Reactor (ITER) divertor inner vertical target which procurement falls under the EU responsibility.
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ENEA, as major partner of the contract... -
Dr Marianne Richou (IRFM CEA)19/09/2018, 11:00
Operational reliability of the divertor target relies essentially on the structural integrity of the component, in particular, of the material interfaces, where thermal stresses tend to be concentrated. To improve bonding quality, a concept developed in the frame of the EUROfusion project WPDIV for the DEMO divertor, consists in the use of functionally graded material (FGM) as interlayer...
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Rodrigo Mateus (IPFN Instituto de Plasmas e Fusˆo Nuclear)19/09/2018, 11:00
Erosion, co-deposition of impurities and heat load effects leads to compound formation in PFC enhancing delamination mechanisms with re-emission of dust particles detrimental to the plasma stability of fusion devices. Beryllium (Be) and tungsten (W) PFC were used in the first wall and divertor in JET, and carbon (C) has achieved new relevance as impurity in the same reactor. Earlier...
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Dr Petra Jenus (Department for Nanostructured Materials Jozef Stefan Institute)19/09/2018, 11:00
Lately, tungsten has gained considerable attention of the fusion scientific community due to its performance at high temperatures. On the other hand, tungsten is affected with a serious reduction of strength at elevated temperatures, latter being one of the main drawbacks of its usefulness as a plasma facing material in fusion reactors.1 Therefore, the main aim of this work has been to improve...
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Juan Du (Institute of Energy and Climate Research IEK-2 Forschungszentrum JŸlich GmbH FZJ)19/09/2018, 11:00
The ITER first wall (FW) panels consist of plasma facing Be tiles, the CuCrZr alloy as heat sink material, and the stainless steel as structural material. A copper layer of 1~2 mm is used between the Be tile and the CuCrZr for stress compensation. Cyclic high heat flux tests employing the electron beam facility results indicate that the failure/weak spot usually occurs at the joint corners...
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Torsten Braeuer (Max-Planck-Institute for Plasma Physics)19/09/2018, 11:00
Wendelstein 7-X (W7-X) is equipped with ten symmetric arranged divertor units consisting of horizontal and vertical targets each. In the current completion phase, Scraper Elements (SE) have been installed in front of two out of ten divertor units to protect the gap between the horizontal and vertical targets (pumping gap) from thermal overload out of the plasma. During the next plasma...
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Chengzhi Cao (Southwestern Institute of Physics)19/09/2018, 11:00
With the aims of high performance plasma toward ITER and even a fusion reactor, heat exhaust would be a serious problem for HL-2M. In this work, impurity seeding is considered to solve heat exhaust problem by radiative divertor. SOLPS-ITER simulations are performed for Ne and Ar impurities from three seeding locations (lower dome, inner target and outer target) with the standard lower single...
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Kieran Flinders (Central Engineering Culham Centre for Fusion Energy (UKAEA))19/09/2018, 11:00
High heat flux testing is a vital part of engineering component validation for fusion technology. The Heat by Induction to Verify Extremes (HIVE) facility is designed to improve the practicalities of this aspect of component testing. It provides a faster turnaround for smaller concepts and a more cost-effective approach by utilising induction heating within a small vacuum vessel.
Due to the...
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Antonio Castaldo (Dipartimento di Ingegneria Elettrica e delle Tecnologie dell'Informazione University of Naples Federico II)19/09/2018, 11:00
The European roadmap to the realisation of fusion energy has identified a number of technical challenges and defined eight different missions to face them. Mission 2 ‘Heat-exhaust systems’ addresses the challenge of reducing the heat load on the divertor targets. Part of this mission is an assessment of several alternatives to the conventional divertor configuration, including ‘Advanced...
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Shin-ichi Satake (Applied Electronics Tokyo University of Science)19/09/2018, 11:00
The simulation plays an important role to estimate characteristics of cooling in plasma facing components such as blanket and divertor. An objective of this study is to perform large -scale direct numerical simulation (DNS) on heat transfer of turbulent flow on coolant water flow. The coolant flow conditions in plasma facing components are assumed to be Reynolds number of a higher order. To...
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Kyoungo Nam (Tokamak Engineering Department National Fusion Research Institute(NFRI))19/09/2018, 11:00
An ITER Tokamak machine with a torus shape is composed of nine units of 40◦ sectors. The sector sub-assembly tool (SSAT) is dedicated assembly tool to integrate vacuum vessel (VV) sector, VV thermal shield (VVTS) segments, VVTS port shrouds, two toroidal field coils (TFC) and various intercoil structures into 40◦ sector.
For the sub-assembly of 40◦ sector, SSAT shall have sufficient strength...
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YoungHwa An (National Fusion Research Institute)19/09/2018, 11:00
The nuclear heat and the shut-down dose rate (SDDR) in the ITER upper port 18 (UP18) was estimated to provide the nuclear heat load for the structural analysis of UP18 and to provide the basis for the further SDDR mitigation strategy of UP18. The UP18 MCNP model has been developed based on the actual CAD model, which was integrated into C-model, the global MCNP model for ITER. While ITER UP18...
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Dr Chulkyu Park (Tokamak Engineering National Fusion Research Institute)19/09/2018, 11:00
Vacuum vessel which is the first confinement barrier of Tokamak fusion reactor should have numerous interfaces such as Blanket, In-vessel-coils, etc. Those interface components should be assembled by fastening of special shape of bolts to the threaded holes in the Vacuum vessel with threaded inserts and to be disassembled for maintenance during Tokamak operation. Threaded connection between...
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Sungjin Kwon (DEMO Technology Division Nationa Fusion Research Institute Yuseong-Gu)19/09/2018, 11:00
The variation of plasma current and magnetic fields generated by superconducting magnet coils causes electromagnetic (EM) loads especially during the abrupt plasma current changes such as major disruption, the vertical displacement event (VDE) of plasma, and the fast discharge. The EM loads are one of the most important external loads for in-vessel components like blanket and divertor modules....
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Christian Vorpahl (Power Plant Physics Technology EUROfusion)19/09/2018, 11:00
In the current pre-concept phase of the European DEMO, integration studies of the systems in the Upper Port area are being carried out. In DEMO, the Upper Port of the Vacuum Vessel is extraordinarily large to allow for the vertical extraction of the Breeding Blanket segments. This requires a number of components inside and outside the port to be integrated with tight space constraints: The...
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Dr Viktor Modestov (Peter the Great St.Petersburg Polytechnic University)19/09/2018, 11:00
The presentation is focused on approaches and results of simulations and used for loading analyses made for Upper Vertical Neutron Camera (UVNC), including spatial stress strain state, seismic analysis, electromagnetic analysis as well as the most important load combinations.
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The Vertical Neutron Camera is a multichannel neutron collimator intended to measure the time resolved neutron... -
Ronan Kelly (RACE UK Atomic Energy Authority)19/09/2018, 11:00
The operation of nuclear fusion facilities must be carefully planned and monitored due to the potential damage to equipment or personnel caused by radiation fields. A method for visualising such three-dimensional (3D) radiation fields in real-time is presented. An interactive volumetric representation is achieved using view-dependent ray casting of a scalar field in three...
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ROCCO MOZZILLO (Department of Industrial Engineering CREATE UNIVERSITY OF NAPLES FEDERICO II)19/09/2018, 11:00
One of the most critical components in the design of DEMO Power Plant is the Breeding Blanket (BB). Currently, four candidates are investigated as options for DEMO. One of these is the Water Coolant Lithium Lead (WCLL) Breeding Blanket (BB). During the previous years a conceptual design of WCLL BB has been developed. At the current state some open issues related to the manufacturability and...
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Dr Alfredo Portone (CCFE UKAEA-CCFE Culham Science Centre Abingdon Oxon)19/09/2018, 11:00
In magnetic confinement fusion ITER represents the most challenging projects conceived ever. The assessment of ITER structural behavior is not trivial since it requires the application of loads coming from different types of analysis (electromagnetic (EMAG), thermal, dynamic, etc.), which are usually run using different software and Finite Element (FE) models, onto mechanical (MECH) models...
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Dr Martin Mittwollen (Institute of Materials Handling and Logistics Karlsruhe Institute of Technology (KIT))19/09/2018, 11:00
The DEMO Oriented Neutrons Source (DONES) is the dedicated facility for testing and enabling of the qualification for different materials to be utilized in the future fusion reactor DEMO. The neutron irradiation damages not only the material samples to be tested but also impacts the plant hardware in and around the Test Cell.
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For that reason, preventive and predictive maintenance activities... -
Daigo Tsuru (Department of Tokamak System Technology Naka Fusion Institute National Institutes for Quantum and Radiological Science and Technology (QST))19/09/2018, 11:00
This article introduces overview of inboard first wall of the JT-60SA device, especially for the initial operation phase including the first plasma. The objective of the inboard first wall is to protect magnetic sensors from plasma. There is no cooling water for in-vessel components in the initial operation phase of JT-60SA, and it will be installed in the later phase. Graphite armour tiles...
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Takao Hayashi (National Institutes for Quantum and Radiological Science and Technology (QST))19/09/2018, 11:00
This article introduces remote handling tools for hydraulic connection of Divertor Cassette in JT-60SA, especially for cutting and aligning tools for re-welding accessing from inside of the cooling pipe. Remote handling system is necessary for the maintenance and repair of the divertor cassette in JT-60SA. Because the space around the cooling pipe connected with the divertor cassette is very...
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Jinho Bae (National Fusion Research Institute Yuseong-gu)19/09/2018, 11:00
Sector Lifting Tool (SLT) are purpose-built tool for the lifting and transferring ITER components. SLT consists of the Sector Lifting Tool (SLT) with the lifting attachments. The purpose of the SLT is to lift and transfer Vacuum Vessel (VV) and Toroidal Field Coil (TFC) from Upending Tool to Sector Sub-assembly Tool (SSAT). After the sub-assembly at SSAT in assembly hall, 40° Sector which is...
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Stuart Budden (DEMO RACE UKAEA)19/09/2018, 11:00
RACE has been developing a concept design for the remote maintenance system for the EUROfusion DEMO powerplant. Within the DEMO tokamak, tritium breeding blankets will require periodic replacement which is currently designed to utilize the upper vertical ports at the top of the vacuum vessel. This operation will be challenging due to the scale of the blankets (~10m tall, up to 80 tonnes). The...
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Catarina Vidal (Instituto de Plasmas e Fusˆo Nuclear Universidade de Lisboa - Instituto Superior TÖcnico)19/09/2018, 11:00
The Plasma Position Reflectometry (PPR) diagnostic systems, to be installed in the International Thermonuclear Experimental Reactor (ITER), will measure the edge electron density profile of the plasma, providing real-time supplementary contribution to the magnetic measurements of the plasma-wall distance. Some of the diagnostic components will be placed inside the vacuum vessel (VV) and...
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Ludovic Allegretti (CEA / DRF / IRFM Institute for Research on Magnetic Fusion CEA - Commissariat ˆ l'Žnergie atomique et aux Žnergies alternatives)19/09/2018, 11:00
Tokamaks, as complex technical devices, need regular maintenances to insure optimal operational conditions. The major 2012-2016 shutdown, dedicated to the upgrade of Tore Supra, was the opportunity to engage important maintenance actions, preparing the restart and insuring the optimal sustainability of the future subsystems of the WEST tokamak. An overall maintenance plan, based on a risk...
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Dr Gunnar Ehrke (E4-Stellerator Edge and Divertor Physics Max-Planck-Institute for Plasma Physics)19/09/2018, 11:00
The cryo-vacuum pump (CVP) system, consisting of 10 units distributed symmetrically inside the Wendelstein 7-X plasma vessel, will be installed together with the 10 units of the actively cooled high heat flux divertor. One pump each is located below the corresponding divertor, and positioned as close as possible to the flux line strike points in order to allow efficient control of plasma...
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Cyra Neugebauer (Karlsruher Institute of Technology (KIT))19/09/2018, 11:00
An important goal for DEMO is the tritium inventory reduction in the fuel cycle. For that, the residence time must be minimized and the tritium content in the individual fuel cycle sub-systems must be reduced. One activity foresees the implementation of an isotope rebalancing and protium removal unit - requires less recycling, has lower hold-up and has a lower residence time than cryogenic...
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Michael Hubeny (IEK-4 Forschungszentrum JŸlich)19/09/2018, 11:00
Steady-state and long pulse exposure of plasma-facing materials in reactor-relevant conditions are an integral step towards the qualification of next-step materials with respect to erosion, fuel retention and morphology changes in view of reactor applications.
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W7-X will allow plasma operation of up to 30 minutes in its second operation phase (OP2) and thus provides an ideal framework for the... -
Beom Seok Kim (DEMO Technology Division National Fusion Research Institute (NFRI))19/09/2018, 11:00
It is foreseen from the decay heat analysis that the total decay heat from the blankets reaches up to 55.6 MW immediately after two years of the full power operation of K-DEMO with the fusion power of 2.2 GW. Especially, the estimation shows that the decay heat from an outboard blanket made of Reduced Activation Ferritic Martensitic (RAFM) steel and tungsten first wall would be tens of...
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Vincent Bruno (IRFM CEA)19/09/2018, 11:00
Currently on fusion devices, diagnostics are mainly aiming at plasma analysis and control. However, operational and programmatic needs have appeared for regular in-vessel components monitoring during plasma campaign. Light robotics systems could meet this requirement and may be a way as well to replace human interventions to fix damaged in vessel components. To minimize the impact on machine...
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Aljaz Kolsek (Universidad Nacional de Educaci—n a Distancia)19/09/2018, 11:00
The ITER Equatorial Port #12 is a first plasma port, which has undergone the Preliminary Design Review (PDR) in November 2017. In support of the PDR, the following nuclear analyses have been conducted: i) the nuclear heat has been calculated in the port plug, as one of the principal thermal loads considered in the design, and ii) the shutdown dose rates (SDDR) have been estimated in the port...
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Laurent Gargiulo (IRFM CEA)19/09/2018, 11:00
Since 2013, CEA has carried out an in-depth modification of the Tore Supra tokamak to build the WEST platform, targeted at supporting the ITER tungsten divertor detailed design, manufacturing and operation. The changes included the modification of the magnetic configuration with new in-vessel coils, the replacement of all carbon Plasma Facing Components (PFCs) by new tungsten elements and the...
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1203. P3.156 Mechanical design of the hydraulic lift and rotation drive system for a DIII-D beamlineMatthew Hansink (General Atomics)19/09/2018, 11:00
Neutral beams are one of the methods to inject power into a tokamak for plasma heating. The DIII-D tokamak has four neutral beam injectors with two ion sources each, located at toroidal angles of 30º, 150º, 210º, and 330º. As originally installed, each could inject up to 5 MW of neutral beam power in the co-injection orientation (nearly parallel to the plasma current). One of the systems, the...
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Keisuke Mukai (Institute of Advanced Energy Kyoto University Uji)19/09/2018, 11:00
This research aims to develop a two-dimensional analysis of neutron flux within the blanket modules by using a compact discharge device as neutron source and imaging plates for detector. Neutron detectable imaging plate is composed photostimulated luminescence (PSL) material and converter such as gadolinium, allowing a high spatial resolution neutron radiography in a wide dynamic range of 10^5...
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Dr Ladislav Vala (Fusion Research Centre Rez)19/09/2018, 11:00
An initial conceptual study of integration of reflectometry diagnostics in the European DEMO has been carried out in the previous years within the EUROfusion project. This study considered antennas and waveguides incorporated in a full poloidal section attached to the Helium-cooled Lithium Lead (HCLL) breeding blanket segments. However, this concept of a diagnostics slim cassette would reduce...
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Kecheng Jiang (Fusion reactor materials science technology division Institute of Plasma Physics Chinese Academy of Sciences)19/09/2018, 11:00
The breeding blanket First Wall is the first boundary separating the fusion plasma and its energetic particles from the rest of the Tokamak. In DEMO reactor, the First Wall integrated in the blanket is in charge of 1) removing the surface heat load connected with the charged particles and the volumetric power density arising from plasma; 2) ensuring the structural integrity of the blanket,...
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ALICE YING (Mechanical and Aerospace Engineering Department University of California Los Angeles)19/09/2018, 11:00
Ceramic breeder pebble beds undergo complex thermomechanical interactions during blanket operation due to stress build-up and relaxation under the effects of confined thermal expansion, thermal cycling, and creep. Understanding the evolution of such processes can aid in guiding blanket design, breeder materials developments, predicting performance and possible failure modes. This study...
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Raffaella Testoni (Dipartimento Energia Politecnico di Torino)19/09/2018, 11:00
The Helium-Cooled Lithium Lead (HCLL) breeding blanket is one of the European blanket designs proposed for DEMO reactor. A tritium transport model is fundamental for the correct assessment of both design and safety, in order to guarantee tritium self-sufficiency and to characterise tritium con-centrations, inventories and losses. The present 2D transport model takes into account a single...
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Geon-Woo Kim (Nuclear Engineering Seoul National University)19/09/2018, 11:00
The blanket system of Korean fusion demonstration reactor (K-DEMO) has a cooling channel through which pressurized water flows to cool down the heat from nucleate heating and plasma radiation. In order to evaluate the cooling performance of blanket, a computational fluid dynamics (CFD) code has been widely used as well as used in commercial heat exchangers. However, CFD can show a large...
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SooHwan Park (KSTAR Research Center National Fusion Research Institute)19/09/2018, 11:00
Pellet injection system of 20 Hz has been operated in KSTAR (Korea Superconducting Tokamak Advanced Research) since 2016. The pellet can be injected to the plasma with different size, velocity and frequency during plasma experiments. The pellet trajectory is interesting topic in KSTAR so the related investigation is carried out outside of tokamak at first. We introduce the preliminary result...
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Jisoo Kim (University of Science and Technology)19/09/2018, 11:00
Uranium (U), which has three allotropic crystal modifications (alpha-, beta-, and gamma-U), is a strong candidate medium for storing and delivering hydrogen or hydrogen isotopes. Alpha-, beta-, and gamma-U are stable at a temperature of up to 668°C, from 668°C to 775°C, and above 775°C, respectively. Because the temperature of the uranium hydride (UHx) formation is limited at room temperature...
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WOO-CHAN JUNG (RESEARCH INSTITUTE DAESUNG INDUSTRIAL GASES)19/09/2018, 11:00
There are various gas components in the exhaust gas of the D-T fusion reaction. All of the hydrogen isotopes are recovered and reused as fuel, and the remaining components are released to the environment. Before releasing to the environment, all substances containing trace amounts of Q2 and Q (such as CH4) must be recovered. An oxidation / adsorption process can be used for this purpose. By...
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Tirui Xia (CNEC High Temperature Reactor Holdings Co.Ltd. Haidian District)19/09/2018, 11:00
This paper is aimed at addressing critical issues related to tritium separation in fusion reactors. One of the effective tritium separation technology is using high temperature proton conducting materials as hydrogen isotope separation membranes. When a direct current is applied to the electrochemical hydrogen pump, hydrogen and its isotopes in the anode side can be electrochemically extracted...
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Mingzhun LEI (Institute of Plasma Physics Chinese Academy of Sciences)19/09/2018, 11:00
China Fusion Engineering Test Reactor (CFETR), the next-step fusion device of China, is proposed to design and operate in two phases. The physical parameters and machine sizes of CFETR have been updated in 2018. It is required that one blanket design can cover two operation phases of CFETR. The water cooled ceramic breeder (WCCB) blanket for CFETR phase II, one candidate CFETR blanket option,...
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Alina Elena Niculescu (Cryogenic Pilot Plant National Institute for Research and Development for Cryogenic and Isotopic Technologies ICSI Rm.Valcea)19/09/2018, 11:00
Nuclear reactors whether they are based on fusion (JET, ITER, DEMO), fission (e.g. CANDU type), or cooled using molted salts (MSR’s) generate significant amounts of wastes in the form of low level tritiated light water or heavy water, for which there is an increasing interest to process and recover tritium (in gas form) and deuterium (as heavy water). Current water treatment systems allow the...
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Vladimir Khripunov (Thermonuclear Reactor Department National Research Center Kurchatov Institute)19/09/2018, 11:00
While the future fusion power reactors will consume and reproduce tritium for their operations, essential amounts of tritium will be required from external sources for their initial start-ups in the commissioning periods. Up-to-date evaluations of the start-up inventories are comparable with or even exceed the available commercial tritium resources in the world nuclear industry. At present the...
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Antonio Frattolillo (Fusion and Technology for Nuclear Safety and Security Department ENEA Italian National Agency for New Technologies Energy and Sustainable Economic Development)19/09/2018, 11:00
Core fuelling of the EU-DEMO tokamak is under investigation within the EUROfusion Work Package “Tritium, Fuelling and Vacuum”. Pellet injection still represents the most promising option. Modelling of pellet penetration and fuel deposition profiles for different injection locations, assuming specific DEMO plasma scenarios and the ITER reference pellet mass (6×1021 atoms), indicates that...
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Nicolae Bidica (National Research and Development Institute for Cryogenic ans Isotopic Technologies)19/09/2018, 11:00
Tritium permeation into the structural materials and further in the coolant of the fusion devices is one of the most important safety issues. Various mathematical model and experiments have been carried out to estimate the amount of tritium permeated in the key components of the fusion devices. However, some issues related to the permeation of hydrogen isotopes through metals, like those...
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Elena Nunnenmann (Institute for Neutron Physics and Reactor Technology Karlsruhe Institute of Technology)19/09/2018, 11:00
The high-energy particle physics Monte Carlo code toolkit GEANT4 has been expanded for fusion energy-range neutron transport simulations based on evaluated nuclear cross-section libraries. Verification and Validation (V&V) analyses were conducted with nuclear data from the ENDF/B-VII.0 and the JEFF-3.1 library to show the suitability for fusion applications. Two computational benchmarks with a...
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Anisia Mihaela Bornea (Pilot Plant National RD Institute for Cryogenics and Isotopic Technologies-ICSI)19/09/2018, 11:00
The research activity for development of catalytic package that equips water-hydrogen catalytic isotopic exchange columns was of permanent interest for the Institute’s research team, mainly motivated by the integration of the Liquid Phase Catalytic Exchange (LPCE) process in most of the detritiation technologies for tritiated water generated from nuclear reactors.
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In recent years, our... -
Ivan Alessio Maione (Institute for Neutron Physics and Reactor Technology Karlsruhe Institute of Technology)19/09/2018, 11:00
Within the EUROFusion consortium, a big effort is made in order to analyze the electromagnetic loads that act on the in-vessel components during normal and off-normal operations, being an important input for their structural assessment. With regard to the Breeding Blanket (BB) project, a global DEMO EM model, feasible to account for different blankets design, has been developed last year with...
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Alexander Spitsyn (NRC Kurchatov Institute)19/09/2018, 11:00
Fusion reactors materials (FRM) will be exposed to 14 MeV fusion neutrons and damaged up to 15 dpa/year. The investigation of neutron irradiated materials is possible only in special conditions in a hot cell. The MeV-range energy ions can be used to simulate the effect of neutron-induced damages in FRM. Such simulation experiments can be used to study the effect of displacements on the...
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Richard Walker (CCFE)19/09/2018, 11:00
A DEMO fusion reactor will need to be self-sufficient in tritium fuel, with breeding planned within blankets surrounding the vessel. However, at the high temperatures within the breeding blanket region, tritium will readily permeate into the coolant through most materials of construction, causing a loss of this valuable commodity and contamination of the coolant stream. Special coatings have...
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Daniele Martelli (Dipartimento di Ingegneria Civile e Industriale universitˆ di Pisa)19/09/2018, 11:00
The In-Box Loss Of Coolant (LOCA) postulated accident is considered as a major concern for the safety involving the development of EU-DEMO fusion reactor. Related to the renewed interest in the Water Cooled Lithium Lead (WCLL) blanket concept, a unique and innovative experimental campaign is under development at ENEA Brasimone research center aiming at investigating consequences of an In-Box...
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Tomas Romsy (Department of Energy Engineering Faculty of Mechanical Engineering Czech Technical University in Prague)19/09/2018, 11:00
The eutectic liquid metal LiPb is considered as one of the tritium breeders of the first fusion power reactors. The flowing liquid metal dissolves alloying elements of the structural steels and thus causes their corrosion. The proposed type of the cold trap is a device providing extraction of corrosion products from liquid metal by gravity separation, which occurs at lower temperatures than...
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Cheol-Woo Lee (Korea Atomic Energy Research Institute)19/09/2018, 11:00
The high peak value of nuclear heat distribution in the fusion breeding blanket is expected to make cooling system design difficult for DEMO. The maximum peak value of about 10 W/cm3 is assumed in the Test Blanket Module with the maximum operational power of 700 MW in ITER. The peak value of nuclear heat distribution in the blanket of DEMO will be increased in proportion to the operational...
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Seong Dae Park (Nuclear Fusion Technology Development Division Korea Atomic Energy Research Institute)19/09/2018, 11:00
The contact resistance effect in the interface between pebble beds was studied with CFD analysis. The lithium ceramics is used as breeder with the form of sphere-shaped pebbles for the extraction of the tritium in some Test Blanket Module (TBM) candidates of ITER. The flow of the gas is essential for the extraction of the tritium. The effects of the gas flow was considered. The effect of fluid...
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Michiko AHN FURUDATE (Dept. of Mechatronics Engineering Chungnum National University)19/09/2018, 11:00
Tritium recovery rate is one of most important parameters to design highly efficient fuel cycle in fusion reactors. To estimate the tritium recovery rate accurately, chemical reactions in the tritium recovery process must to be studied in detail. In solid type breeding blankets, tritium is expected to be released from the breeder pebbles in the form of HTO into purge gas surrounding the...
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Dr Luis Mora Vallejo (F4E)19/09/2018, 11:00
The ITER project will require large cryopumps of flat-geometry to pump the Heating Neutral Beam Injectors (NBI), and similar cryopumps to pump the diagnostic Neutral Beam (DNB). The cryogenic supply uses supercritical Helium for the cryopanels and gaseous Helium for the thermal shields of the cryopumps. The cryogenic fluids will be produced by a large cryogenic plant, and then distributed by...
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wei mao (The Institute of Engineering Innovation The University of Tokyo)19/09/2018, 11:00
To limit hydrogen leakages in a breeding blanket of fusion reactor, a hydrogen permeation barrier can be used. Erbium oxide was selected as a promising candidate with a low hydrogen diffusion. Thus, the purpose of this study is to understand the irradiation effect of helium ions, originating from fission of lithium exposed to fusion-induced neutrons in the blanket, on the hydrogen diffusivity...
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Lazar Alin (Experimental Pilot Plant for Tritium and Deuterium Separation National Institute for Cryogenics and Isotope Separation)19/09/2018, 11:00
In cryogenic distillation columns complex phenomena appear, some of them are neededand othersmust be avoided, such the non-uniform cooling of the distillation column or the impossibility of transfer of the cooling power to the gases mixture with major changes in the separation dynamics. The loss of separation capacity or the inability to reach optimal operating parameters are caused by...
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Dr Hongli Chen19/09/2018, 11:00
As an integral part of the solid blanket, ceramic tritium breeder pebble bed plays a vital role in tritium breeding during the operation of the fusion reactor. The packing structure of the pebble bed has an impact on its thermomechanical behavior and tritium exaction in the solid blanket, which is actually affected by packing methods, particle size, container size, particle smoothness,...
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Sergi Colominas (Analytical and Applied Chemistry Department ETS Institut Qu’mic de Sarriˆ Universitat Ramon Llull)19/09/2018, 11:00
Tritium management is one of the main challenges that future nuclear fusion energy has to achieve. Accurate tritium monitoring is a basic task in order to have relying fusion reactors. High temperature sensors have to be developed to make this monitoring a reality. Hydrogen sensors based on solid-state electrolytes can be a reliable option to perform this monitoring. These types of sensors...
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Marc Nel-lo (Analytical and Applied Chemistry Department ETS Institut Qu’mic de Sarriˆ Universitat Ramon Llull)19/09/2018, 11:00
Lithium 6 is the isotope required to generate in-situ tritium in fusion reactors. Because of that, lithium monitoring in lithium-lead eutectic (Pb-15.7Li) is of great importance for the performance of the liquid blanket. Lithium measurements will be required in order to proof tritium self-sufficiency in liquid metal breeding systems. On-line lithium sensors must be designed and tested in order...
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Vijulie Mihai (Experimental Pilot Plant for Tritium and Deuterium Separation National Institute for Cryogenics and Isotope Separation)19/09/2018, 11:00
The hydrogen isotopes separation plants have special requests related to safety operation and avoidance of radiological fluid leakage and explosion conditions. For the LPCE, part of the ICSI Rm.Valcea “Experimental Pilot Plant for Tritium and Deuterium Separation”, the process transformation from a laboratory setup into a semi-industrial plant, as well as migration from a local control to an...
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Pierluigi Chiovaro (Department of Energy Information Engineering and Mathematical Models University of Palermo)19/09/2018, 11:00
On the end of 2017, in the framework of EUROfusion R&D activities, a close collaboration between EU and China has started aiming at elaborating joint strategies for the development of the Water Cooled Lithium Lead (WCLL) and the Water Cooled Ceramic Breeder (WCCB) Breeding Blanket (BB) concepts. In this framework, an intense research campaign has been carried out at the University of Palermo,...
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Jingjie Shen (Nuclear Professional School The University of Tokyo)19/09/2018, 11:00
Oxide-dispersion-strengthened (ODS) steels have been developed as one of prospective candidate materials for fast reactor cladding as well as fusion reactor blanket applications. The anisotropy in microstructure and tensile properties in the range from room temperature (RT) to 973 K of the 12Cr ODS steel with the nominal composition of Fe-12Cr-2W-0.3Ti-0.25Y2O3 (in wt.%) was investigated by...
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Rafael Vila (FUSION CIEMAT)19/09/2018, 11:00
Functional materials have diverse applications in fusion reactors and it is clear that insulators are among the most versatile groups. They are the base of all the electric and radiofrequency systems in diagnostics and heating systems from DC to very high frequencies (RF, H&CD, NBI…). Additionally, insulators are subjected to quite different conditions (voltage, temperature, frequency...)...
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Vaughan Thompson (Central Engineering UKAEA)19/09/2018, 11:00
Remote maintenance in fusion machines such as JET and ITER relies on sliding interfaces such as bolted joints. Experience in JET, where removal torques much higher than installation values with uncoated bolts is commonplace, led to the installation of experimental bolted assemblies in 2015: the first of its kind in JET. These assemblies included some 660B stainless steel ITER Blanket-specific...
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Simon Heuer (Forschungszentrum JŸlich GmbH Institut fŸr Energie und Klimaforschung - Plasmaphysik Partner of the Trilateral Euregio Cluster (TEC) Forschungszentrum JŸlich)19/09/2018, 11:00
For the European demonstration power plant, four types of breeding blankets are under consideration. All designs agree in the basic materials selection, that is Eurofer used as structural material and tungsten used as armour material. Detailed thermo-mechanical finite element analyses show that a direct joint of these materials will not last over the anticipated lifetime of the blankets due to...
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Ivan-Aleksander Kodeli (Jozef Stefan Institute)19/09/2018, 11:00
Shielding Integral Benchmark Archive and Database (SINBAD) project started in the early 1990’s at the Organization for Economic Cooperation and Development’s Nuclear Energy Agency Data Bank (OECD/NEADB) and the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL) with the goal to preserve and make available the information on the performed radiation...
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Xiaodong Mao (Department of Nuclear Structural and Shielding Materials Insititute of Nuclear Energy Safety Technology Chinese Academy of Sciences)19/09/2018, 11:00
Oxide dispersion strengthened (ODS) steel is one of the most promising candidate structural materials for fusion nuclear systems. It is widely recognized that to design and to control macroscopic materials properties of ODS steel successfully, a fundamental understanding of the atomic-scale structure and chemistry of oxide/matrix interfaces is necessary, owing to the fact that oxide/matrix...
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1246. P3.199 Development of W-monoblock divertor components with included thermal barrier interfacesMagdalena Galatanu (National Institute of Materials Physics)19/09/2018, 11:00
In the case of DEMO fusion reactor, the divertor should be able to extract a steady heat flux of about 10 MW/m2. A promising concept is the W-monoblock. which should be connected to a CuCrZr or an advanced Cu ODS alloy pipe passing through the W component. Taking into account the optimum operating temperature windows for W and existing Cu-based alloys and the thermal expansion coefficients...
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Isabel Garcia-Cortes (Laboratorio Nacional de Fusion CIEMAT)19/09/2018, 11:00
To date the research on structural materials for future fusion reactors has been focused on the evolution of mechanical properties with irradiation dose, energy, temperature, etc. However, the performance of materials irradiated under the presence of magnetic fields remains unclear. This aspect becomes critical, as structural materials in fusion reactors will need to withstand intense and...
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Kuo Tian (Institute for Neutron Physics and Reactor Technology (INR) Karlsruhe Insitute of Technology)19/09/2018, 11:00
The IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) is planned to deliver an intensive neutron source (5×10^16 n/s) for irradiation experiments that are confined and shielded by the Test Cell (TC). During the operation of the facility, unexpected degradation (by irradiation or corrosion) or damage (by handling etc.) of the TC leak tight liner,...
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Angel Muñoz (F’sica Aplicada Universidad Carlos III de Madrid)19/09/2018, 11:00
It has been paid a great attention to the production of Tungsten/Copper (W/Cu) composites, as they appear promising materials to form part of the cooling system of the divertor of the future fusion reactors. However, further assessments of the microstructure and mechanical characteristics of these composites are required for the designs of the divertor. In this study, the mechanical behavior...
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Monika Vilémová (Materials engineering Institute of Plasma Physics of the CAS)19/09/2018, 11:00
Tungsten has many advantageous features; however, it is rather susceptible to oxidation at temperatures above 500 °C. By the addition of various oxide-forming elements to tungsten, self-passivation is induced. During exposure of the alloy to air, a passivation layer is formed on its surface, thereby preventing further tungsten oxidation, material degradation and related radiation spreading....
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Dehong Chen (Key Laboratory of Neutronics and Radiation Safety Institute of Nuclear Energy Safety Technology Chinese Academy of Sciences)19/09/2018, 11:00
Gas Dynamic Trap (GDT) is very attractive as a kind of fusion neutron source for testing fusion materials and components as well as driving fusion-fission hybrid reactor due to its linear and compact structure, low physics and technology requirement, relatively low cost and tritium consumption. These years, the conceptual designs of GDT-based neuron source for above two purposes, named...
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Dr Fabrizio Siviero (Solutions for Vacuum Systems Development Lab SAES Getters S.p.A.)19/09/2018, 11:00
The use of non-evaporable getter (NEG) pumps is common in many UHV applications including surface science, analytical instruments and very large vacuum systems for high energy physics. In the past years, getter solutions based on the new sintered alloy ZAO® have been developed enabling operation in the HV regime, i.e. 10-6 Pa and above. The properties of this NEG material make it appealing for...
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Dr Nerea Ordas (Materials and Manufacturing Ceit)19/09/2018, 11:00
The high mechanical strength of ODS FS, and their resistance to creep and neutron radiation damage up to 750 ºC are attributed to extremely fine microstructures with high density of very stable nanometric precipitates, generally Y-Ti-O oxides. The STARS route (Surface Treatment of gas Atomized powder followed by Reactive Synthesis) proposed by Ceit avoids mechanical alloying to introduce...
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Dr Alexander von Mueller (IPP)19/09/2018, 11:00
The ability to estimate the in-service performance and lifespan of components is key to realising a commercially viable fusion energy device. The finite element method (FEM) is used to estimate performance of a component design with computational simulations. Image-based FEM (IBFEM) converts 3D images (e.g. X-ray tomography) into high-resolution models for part-specific simulations that...
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Dr Elisa Sal Broco19/09/2018, 11:00
Flow Channel Inserts (FCIs) are key elements in the Dual Coolant Lead Lihitum (DCLL) blanket since they decouple electrically the flowing PbLi and the blanket steel structure, minimizing the MHD pressure drop. Furthermore, in the high-temperature version of the DCLL (where the PbLi may reach temperatures up to 700 °C), FCIs also protect the steel structure from the hot liquid metal.
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The... -
Dr Ali Abou-Sena (Institute of Neutron Physics and Reactor Technology (INR) Karlsruhe Institute of Technology (KIT))19/09/2018, 11:00
The capsules of the IFMIF-DONES High Flux Test Module (HFTM) are packed densely with Eurofer specimens. A filling material (previously NaK-78 and presently sodium) is needed to fill any empty volume to improve the heat conduction and obtain uniform temperature distribution. Sodium is replacing NaK-78 because potassium generates argon isotopes leading to a pressure increase and formation of...
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Julia Lorenz (Institute for Applied Materials Karlsruhe Institute of Technology)19/09/2018, 11:00
Electrochemical techniques such as electroplating of metals, electrochemical machining (ECM), electroforming, anodizing and electropolishing of metal surfaces have been established successfully in a variety of industrial processes. A wide range of applications are available such as the electrodeposition of decorative metal coatings on plastics and metals, corrosion protection of mass products...
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Yusuke Shimada (Institute for materials research Tohoku University Sendai)19/09/2018, 11:00
A copper (Cu) alloy, having a high thermal conductivity, is a promised material for heat sink of diverters in a force free helical reactor (FFHR). Recently, Hishinuma’s group succeeded in fabrication of oxide dispersion strengthened (ODS) Cu alloys using mechanical alloying (MA) and hot isostatic pressing (HIP) process. ODS is expected to bring about high-temperature strength and irradiation...
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JAVIER BARCENA RISUEÑO (Mechanical Empresarios Agrupados Internacional EAI)19/09/2018, 11:00
In the framework of the EU fusion roadmap implementing activities, an accelerator-based Li(d,n) neutron source called DONES (Demo-Oriented early NEutron Source) is being designed within the EUROfusion workpackage WPENS as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. The objective of this workpackage is to be ready for...
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Andrei Galatanu (National Institute of Materials Physics)19/09/2018, 11:00
W-laminates are multi layered composites realized from alternately stacked W and a second metal foils. Such materials are promising candidates for W-based structural materials for fusion reactors like DEMO or beyond concepts, due to the fact that cold-rolled ultrafine-grained thin W foils show exceptional properties in terms of ductility, toughness and ductile to brittle transition (DBT), in...
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Yong Zhao (Superconductivity and New Energy Research and Development Center Southwest Jiaotong University)19/09/2018, 11:00
Er2O3 coatings with different structures were deposited on type 316 stainless steel substrates by magnetron sputtering and corroded by liquid lithium for corrosion resistance study. The microstructure of the Er2O3 coatings was controlled by using two different methods, one the Er metal layer was deposited and oxidized successively, and the other directly by sputtering with Er2O3 deposition....
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Michal Kresina (DSN CEA)19/09/2018, 11:00
The Materials Detritiation Facility (MDF) has been designed to thermally treat solid non-combustible radioactive waste produced during operations of the Joint European Torus (JET) that is classified as Intermediate Level Waste (ILW) in the UK due to its tritium inventory (>12kBq/g). The wastes primarily comprise Inconel, steels, aluminium alloys, copper and carbon-based composites. These...
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Xiang Chen (Nuclear Materials Science and Technology Group Oak Ridge National Laboratory)19/09/2018, 11:00
Eurofer97 is one of the leading candidates of reduced activation ferritic martensitic (RAFM) steels for first wall structural materials of early demonstration fusion power plants. During fusion plant operation, intense neutron irradiation damage on first wall materials can cause significant irradiation embrittlement and greatly reduce the fracture toughness of RAFM steels. Therefore, it is...
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Fabrizio Franza (Institute for Neutron Physics and Reactor Technology Karlsruhe Institute of Technology)19/09/2018, 11:00
Fusion systems codes are essential computational tools aimed to simulate the physics and the engineering features of a fusion power station. The main objective of a system code is to find one (or more) reactor configurations, which simultaneously comply with the physics operational limits, the engineering constraints and the net electric output requirements.
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As such simulation tools need to... -
Brad Merrill (Idaho National Laboratory)19/09/2018, 11:00
This paper describes recent progress at the Idaho National Laboratory (INL) in developing the MELCOR-TMAP computer code for fusion. The MELCOR-TMAP for fusion computer code is being developed by the INL’s Fusion Safety Program (FSP) [1] by modifying the US Nuclear Regulatory Commission’s (NRC’s) MELCOR [2] computer code for fission reactor severe accident analyses. The MELCOR code was chosen...
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Monika Lewandowska (Faculty of Maritime Technology and Transport West Pomeranian University of Technology Szczecin)19/09/2018, 11:00
DEMO is planned to be a prototype fusion power plant capable of demonstrating production of electricity at the level of a few hundred MW. DEMO is considered to be an intermediate step between the ITER experimental reactor and a commercial power plant. Design and assessment studies on the European (EU) DEMO are carried out by the EUROfusion consortium. The Primary Heat Transfer System (PHTS)...
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Francesco Gracceva (ENEA)19/09/2018, 11:00
Externalities are defined as a cost that arises when the social or economic activities of one group of persons have an impact on another group and when that impact is not fully accounted, or compensated for, by the first group (ExternE project). External costs are not usually considered in the total cost of electricity causing market failures. To fairly compare the different electricity...
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Konstantina Voukelatou (FSN ENEA)19/09/2018, 11:00
ENEA through the Department of Fusion and Technologies for Nuclear Safety (FSN) actively participates, playing a fundamental role, in the realization of ITER, contributing with the industry to the design and construction of many components ranging from diagnostic, power supply systems, superconducting magnets and Test Blanket Module auxiliary systems. The French Nuclear Safety Authority (ASN),...
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Chiara Bustreo (Consorzio RFX)19/09/2018, 11:00
A hybrid fusion-fission (HFF) reactor based on a Reversed Field Pinch (RFP) configuration looks as an attractive option from both a technical (simple design, easy machine assembly and maintenance) as well as economic perspective (low investment costs due the absence of large Heating and Current Drive systems and superconductive toroidal field coils).
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The hybrid reactor studied here has a RFP... -
Giuseppe Zollino (Consorzio RFX)19/09/2018, 11:00
In the second half of this century, the European energy mix will be very likely completely decarbonized. Two main options are available to generate carbon free electricity: either to rely on renewable energy sources or to further differentiate the energy mix by including nuclear power.
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In the former case a large storage capacity and/or back-up dispatchable generation are required to compensate... -
Jan Syblík (Department of Energy Engineering Czech Technical University in Prague)19/09/2018, 11:00
The paper focuses on the design of appropriate power cycles for fusion power reactor, two S-CO2 Brayton cycles, and its positive and negative aspects. The goal of the paper is to propose a suitable power cycle and its optimization for the European fusion power plant DEMO2. Comparison of cycles in terms of using more heat resources at once is depicted. The study gives a principal preview of...
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Gheorghe Bulubasa (National Research and Development for Cryogenics and Isotopic Technologies - ICSI Rm. Valcea)19/09/2018, 11:00
Helium-3 is a rare isotope of helium (1.37 ppm as fraction of total helium – natural abundance), with applications in medicine, industry, security, and science. Due to its high request, the world is experiencing nowadays a shortage of helium-3.
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The most common source of helium-3 is the disintegration of tritium. Tritium is an unstable isotope of hydrogen, with a half-life of 12.3 years, and is... -
Gianluca Barone (Dipartimento di Ingegneria Civile e Industriale (DICI) Univeristy of Pisa)19/09/2018, 11:00
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Flavio Crisanti19/09/2018, 11:00
The Divertor Tokamak Test (DTT) facility is a project proposed by the Italian Consortium aimed to test the physics and technology of various alternative divertor concepts in order to design a heat and power exhaust system able to withstand the large loads expected in the divertor of a DEMO fusion power plant.
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Even though DTT is a machine operating without tritium, a significant 2.5 MeV neutron... -
Keitaro Kondo (Department of Fusion Reactor Materials Research National Institutes for Quantum and Radiological Science and Technology (QST))19/09/2018, 11:00
The IFMIF (International Fusion Materials Irradiation Facility) project aiming at material tests for a future fusion power plant is now in the Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach Agreement between Japan and EU. As part of the activities the construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress at Rokkasho,...
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Yuki Edao (National Institutes for Quantum and Radiological Science and Technology)19/09/2018, 11:00
Detritiation system (DS) is the key system to ensure safety of a fusion reactor. The DS must be designed to make sure of detritiation when an extraordinary event such as fire happens. Assuming that an accidental release of tritium and production of hydrocarbons by combustion of cables in case of fire occurs simultaneously, tritiated methane will be generated by the reaction between tritium and...
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Jae Hyun Kim (Nuclear Engineering Department Hanyang University)19/09/2018, 11:00
The exposure by shutdown dose and production of radioactive waste predicted from the activation analysis are interesting issues of fusion reactor facility design in the view of radiation safety. Impurities of the irradiated material, such as cobalt in the structural material, are occasionally an important factor in the evaluation of the induced activity.
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Concrete is used as the neutron shield... -
Jinwoo Park (Department of Energy System Engineering Seoul National University)19/09/2018, 11:00
Large-scale R&D projects are experiencing frequent delays due to high development uncertainties. Schedule issues are creating a series of problems that are causing delays in the entire projects by increasing the cost of projects and thereby reducing the reliability resulting in delays in timely tasks such as building the R&D facilities. In this study, considering the fact that the technology...
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Masakatsu Murakami (Institute of Laser Engineering Osaka University)19/09/2018, 11:00
In the last decade, it has been intensively studied to heat a compressed DT fuel to an igniting temperatures of about 5 keV by using picosecond laser pulses. In the present work, we have investigated to create high temperature (> 10 keV) plasma at relatively high densities, by using a femtosecond laser pulse combined with a specially structured micron-sized target. The structured target is...
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Paweł Herbin (Institute of Physics Faculty of Mechanical Engineering and Mechatronics West Pomeranian University of Technology)19/09/2018, 11:00
Current models used for thermal–hydraulic analyses of forced-flow superconducting cables, used in the fusion technology, such as, e.g. Cable-in-Conduit Conductors (CICCs), are typically 1-D. They require reliable predictive expressions for the transverse mass-, momentum- and energy transfer processes between different cable components, in order to reliably simulate the behavior of any...
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