Anna Wojcik-Gargula
(Department of Radiation Transport Physics)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Study of materials dedicated to fusion reactors is one of the most challenging tasks faced by fusion research. Unfortunately, the number of useful fast neutron sources with a proper neutron spectrum and high neutron fluence is limited. Currently, a better exploitation of the existing neutron sources, such as high flux fission research reactors or material test reactors, is necessary to develop...
Sudhirsinh Vala
(Neutron Source Up-gradation Division)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
In order to study the neutronics of fusion reactor blankets, a program is underway at the IPR using 14-MeV neutron source. An accelerator based neutron generator is under development in which 30 mA deuterium beam will be accelerated up to 300 keV energy. It will then impinge on a rotating tritium target to producing nearly isotropic 14-MeV neutrons. The expected neutron yield is 3-5 x...
Fernando Arranz
(Laboratorio Nacional de Fusion)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The LIPAc (Linear IFMIF Prototype Accelerator) is a prototype that ends in a Dump made of copper with conical shape and cooled by water moving at high speed on the outer surface.
The shape of the dump is intended for a redistribution of a very high density power of the deuteron beam to be stopped (1.12 MW) leading during normal operation to reasonable temperatures and thermal stresses well...
Gioacchino Micciche
(FSN-ING-PAN)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where fusion reactor candidate materials will be tested. The neutron flux is produced by means of a deuteron beam (250 mA, 40 MeV) that strikes a target of liquid lithium circulating in a loop. The support on which the liquid lithium flows is the most heavily exposed component to the...
Wojciech Krolas
(Institute of Nuclear Physics PAN)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
IFMIF-DONES - a powerful neutron irradiation facility for studies and certification of materials - is planned as part of the European roadmap to fusion electricity. Its main goal will be to study properties of materials under severe irradiation in a neutron field similar to the one in a fusion reactor first wall. It is a key facility to prepare for the construction of the DEMO Power Plant...
Gaetano Bongiovi
(Department of Energy)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The availability of a high flux neutron source for testing candidate materials under irradiation conditions which will be typically encountered in future fusion power reactors is a fundamental step towards the development of fusion energy. To this purpose, IFMIF (International Fusion Materials Irradiation Facility) represents the reference option to provide the fusion community with a source...
Oriol Nomen
(IREC)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The International Fusion Materials Irradiation Facility (IFMIF) aims to provide an accelerator-based, D-Li neutron source to produce high energy neutrons at sufficient intensity and irradiation volume for DEMO materials qualification. Part of the Broader Approach (BA) agreement between Japan and EURATOM, the goal of the IFMIF/EVEDA project is to work on the engineering design of IFMIF and to...
Zhiqiang Zhu
(Institute of Nuclear Energy Safety Technology (INEST))
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Because of the depletion and limitation of natural energy sources, fusion energy is the promising and irreplaceable way for energy development in the future. As the only energy conversion unit in the fusion reactor, PbLi blanket is considered as one of the important blankets for DEMO and fusion reactors, Lead Lithium (PbLi) is designed as tritium breeder, neutron multiplier and coolant. Before...
Tomas Romsy
(Faculty of Mechanical Engineering)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The liquid metal eutectic Pb-Li17 is considered as one of the possible coolants for the blanket of the fusion reactor DEMO. The main reason for usage of the eutectic Pb-Li17 is the Tritium breeding. The eutectic flow separates alloys of the structural steels and thus be the cause of them corrosion.The cold trap is a device for corrosion products removing from liquid metal.
The cold trap was...
Bernhard Ploeckl
(Max Planck Institute for Plasma Physics)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The Demonstration Fusion Power Reactor (DEMO) is supposed to be the step in between ITER and the first commercial fusion power plant. In the framework of one mission of the “Work plan for the roadmap to fusion energy 2014-2018” a work package Tritium, Fuelling and Vacuum (TFV) was launched. As part of this project, the examination of requirements for the matter injection system is ongoing...
Mikhail Gryaznevich
(Tokamak Energy Ltd)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Recent advances in the development of high temperature superconductors (HTS) [1], and encouraging results on a strong favourable dependence of electron transport on higher toroidal field (TF) in Spherical Tokamaks (ST) [2], open new prospects for a high field ST as a compact fusion reactor or a powerful neutron source [3]. The combination of the high beta (ratio of the plasma pressure to...
Carlos Otarola
(Electromechanical Engineering)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The manufacturing methods and issues found during the construction of the Stellarator of Costa Rica 1 (SCR-1) will be discussed. The SCR-1 is a small modular stellarator developed by the Instituto Tecnológico de Costa Rica (ITCR). Currently, it’s being tested for the first plasma discharge.
SCR-1 is a 2-field period small modular stellarator (Ro=0.238 m, =0.054 m, Ro/a>4.4, plasma volume...
Subrata Pradhan
(Institute for Plasma Research)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Steady State Superconducting Tokamak (SST-1) at Institute for Plasma Research is a `working’ experimental superconducting device since late 2013. SST-1has been upgraded with Plasma Facing Components and is getting prepared towards long pulse operations in both circular and elongated configurations. Initial experiments have begun in SST-1 with circular plasma configurations. SST-1 offers a...
Dennis Ronden
(Fusion physics - Remote Handling)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
This paper presents the results of a study that was performed on conceptual solutions for assembly and handling of EC components inside the EC upper and equatorial port cells. Particular topics that are discussed include the access to the waveguides and auxiliary feedthroughs of the launchers at the port plug closure plate, (dis-)assembly & alignment of the ex-vessel waveguide in the port...
Avelino Mas Sanchez
(Ecole Polytechnique Fédérale de Lausanne)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The Electron Cyclotron Upper Launcher (ECUL) is an eight beamline ITER antenna aimed to drive current locally inside the islands that may form on the q= 3/2 or 2 rational magnetic flux surfaces in order to stabilize neoclassical tearing modes (NTMs). The primary vacuum boundary at the port plug extends into the port cell region through the ex-vessel mm-wave waveguide components, defining the...
Phillip Santos Silva
(Swiss Plasma Center)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The ITER Electron Cyclotron Heating Upper Launcher (ECHUL) will be used to drive current locally inside magnetic islands located at the q=2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs). Each antenna consists of eight beam lines that are designed for the transmission of up to 1.5 MW of mm-wave power at 170 GHz. The First Confinement System (FCS) is...
Robert Bertizzolo
(EPFL-SPC (Swiss Plasma Center))
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The ITER Electron Cyclotron Heating Upper Launcher (ECHUL) will be used to drive current locally inside magnetic islands located at the q=2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs). Each antenna consists of eight beam lines that are designed for the transmission of up to 1.5 MW of mm-wave power at 170 GHz. The First Confinement System (FCS) is...
Koji Takahashi
(Department of ITER Project)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The new mirror angle detector for ITER EC launchers, applying a rotary capacitor , a RF feeder, RF circuits and several hundreds MHz RF has been developed. The rotary electrode is attached to the rotation axis of the mirror and the stationary electrode is connected to a RF feeder. The reflected RF wave at the rotary capacitor comes back to the feeder and phase of the reflected RF wave changes...
Yasuhisa Oda
(Japan Atomic Energy Agency)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The Electron Cyclotron Heating and Current Drive system developed for ITER is made of 12 sets of High Voltage Power Supplies, 24 Gyrotrons, 24 Transmission Lines and 5 Launchers, 4 UL located in upper ports and 1 EL at the equatorial level. The ITER operation requires to switch operating launcher during the plasma operation with short interval, namely mid-pulse switch operation. To change the...
Michael Bader
(Ampegon AG)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The power supply for the EC Heating system (ECPS) of ITER provides the electrical power to the 170GHz/1MW Gyrotrons. The required electrical power for the gyrotrons is not only very high but has to comply also with highest quality requirements.
This paper gives an overview of the Ampegon ECPS system procured by F4E. It describes the technical requirements of the EC Power Supply system ECPS and...
Tomasz Rzesnicki
(IHM)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The EU 1 MW, 170 GHz gyrotron with hollow cylindrical cavity has been designed within EGYC (European GYrotron Consortium) in collaboration with the industrial partner Thales Electron Devices (TED) and under the coordination of Fusion for Energy (F4E). In the frame of the EU program the short-pulse (SP) version of this tube has been designed and manufactured by KIT in collaboration with TED....
Peter Spaeh
(Institute for Applied Materials)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
ITER will be equipped with four EC (Electron Cyclotron) upper launchers of 8 MW microwave power each with the aim to counteract plasma instabilities during operation. The structural system of these launcher antennas will be installed into four upper ports of the ITER vacuum vessel.
During operation the port plug structure will be heated by nuclear heating from neutrons and photons and thermal...
Matthieu Toussaint
(Swiss Plasma Center)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The Tokamak à Configuration Variable (TCV) has been recently equipped with a 1 MW neutral beam heating (NBH) injector11. Two new stainless steel ports with rectangular aperture of 170x220mm have been manufactured and installed for this purpose. The NBH injector is connected to one of them via a stainless steel port extension. The port and its extension together form the beam duct...
Damien Fasel
(SB-SPC)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The TCV tokamak infrastructure has been recently adapted to leave access for a neutral beam (NB) injector capable of 1MW of neutral power during 2sec into the TCV plasma. BINP has been in charge to design and to procure this equipment, taking care of the experimental constraints imposed both by the future physics objectives of TCV, as by the mechanical requirements complying with the tight...
Ugo Siravo
(Ecole Polytechnique Fédérale de Lausanne (EPFL))
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
Three RHVPSs (Regulated High Voltage Power Supplies, 84kV/80A/2s) are installed and operated at the Swiss Plasma Center for almost twenty years. Each RHVPS supplies a cluster of three gyrotrons. Two clusters are composed of diode type gyrotrons operating at the second harmonic of the TCV electron-cyclotron frequency (X2, 84GHz), whereas the third is a cluster of triode type gyrotrons operating...
Alexander N. Karpushov
(Swiss Plasma Center (SPC))
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The TCV tokamak contributes to physics understanding in fusion reactor research based with a wide experimental tool set: flexible shaping and high power electron cyclotron heating. Plasma regimes with high plasma pressure, a wide range of temperature ratios and significant populations of fast ions are now attainable by a TCV heating system upgrade. In the first stage of the TCV upgrade...
Kenji Saito
(Department of Helical Plasma Research)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The transmission line is one of the most important parts among the ion cyclotron range of frequencies (ICRF) heating devices. In the case of unwanted troubles on the line, immediate power-off is necessary for the protection of the line and for safety. In the Large Helical Device (LHD), though the causes were unclear, several troubles such as melting sometimes occurred on the line between the...
Haifeng Liu
(Institute of Fusion Science)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The heating of ions by an obliquely propagating shear Alfvén wave at frequencies a fraction of the particle cyclotron frequency is demonstrated analytically. Under consideration of the small wave amplitude, the resonance conditions in the laboratory frame are systematically derived by multi-scale expansion method. It is found that 1) the cyclotron resonance condition may occur at any wave...
Helmut Faugel
(Max Planck Institute for Plasma Physics)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
The efficiency of heating and current drive systems is the key for a successful operation of fusion demonstration power plants like DEMO. In an earlier review article, overall efficiencies of H & CD systems were estimated at 20 – 30 % [1].
In this paper we present a breakdown of the overall efficiency for ICRF (ion cyclotron range of frequencies): 1) the technical efficiencies; 2) the...
Fabrice Louche
(Plasma Physics Laboratory)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
Ion cyclotron wall conditioning (ICWC) is being developed for ITER as a baseline conditioning technique in which the ion cyclotron heating and current drive system will be employed to produce and sustain the current-less conditioning plasma. The TOMAS project (TOroidal MAgnetized System, operated at the FZ-Juelich, Germany) proposes to explore several key aspects of ICWC. This project stands...
Chun Kung
(Plasma Physics Laboratory)
9/7/16, 11:00 AM
B. Plasma Heating and Current Drive
Poster
Experimental results have shown that twelve-strap HHFW operating at 30 MHz can provide significant plasma heating for NSTX. In this case, it is important to understand the interactions between return currents on the antenna enclosure sidewalls/septa and the launched k|| spectra. CST Microwave Studio is applied to this problem with the view toward optimizing the antenna coupling to the desired...
Anett Spring
(W7-X Operation)
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
The W7-X steady state control and data acquisition system has been successfully commissioned and well established to investigate plasma break down and run the first more complex physics programs during the initial operation phase of W7-X. Already in the first weeks of plasma operation, experiment programs with up to 10 minutes containing a series of up to 20 plasma discharges have been run...
Heike Laqua
(Wendelstein 7-X Operations (OP))
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
Wendelstein 7-X (W7-X) is a superconducting stellarator undergoing the first experimental campaign after its commissioning. It’s characteristic feature is the steady state operation of the magnetic field. After an upgrade to cope with permanent heat loads of several Megawatts, W7-X will be able to run steady state discharges, too. This requires a control system that differs from the commonly...
Reinhard Vilbrandt
(Max-Planck-Institute for Plasma Physics)
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
The commissioning and final validation of the central safety system and the acceptance by the authority were very important steps immediately before the successful ignition of the first plasma in Wendelstein 7-X in December 2016.
Safety is the mandatory prerequisite for the operation of experimental devices of course to protect the personnel and the investment from hazardous situations. To...
Hexiang Wang
(Mechanical Engineering & Mechanics)
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
Ongoing work in the fusion community focuses on developing advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamic (MHD) stability, and noninductively driven plasma current. The toroidal current density profile, or alternatively the q profile, together with the normalized beta, are often used to characterize these advanced scenarios. The development of these...
Andres Pajares
(Mechanical Engineering & Mechanics)
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
Control of the plasma density and temperature to produce a certain amount of fusion power, known as burn control, is one of the key issues that need to be solved for the success of tokamak fusion reactors such as ITER. In order to reach a high fusion power to auxiliary power ratio, tokamaks must operate near temperature and density stability limits. Therefore, active control to maintain a...
Eugenio Schuster
(Mechanical Engineering & Mechanics)
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
Research on fusion plasmas in tokamaks has led to the insight that the poloidal magnetic-flux distribution within the plasma has a crucial impact on its performance. Achieving certain types of poloidal magnetic-flux profiles, or alternatively certain types of q profiles, leads to resilience against undesirable instabilities and to higher bootstrap-current fractions, which in turns favor...
Zeki Ilhan
(Mechanical Engineering & Mechanics)
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
Active control of the toroidal current density profile is among those plasma control milestones that the National Spherical Tokamak eXperiment - Upgrade (NSTX-U) program must achieve to realize its next-step operational goals characterized by the high-performance, MHD-stable plasma operation with neutral beam heating, and longer pulse durations. Motivated by the coupled, nonlinear,...
Luca Boncagni
(FSN)
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
In this work we present a new real-time acquisition and elaboration system for the two-color scanning beam interferometer installed on FTU. The real-time system provides the density informations that can be used to approximate the plasma and runaway beam radial position. Furthermore, the central chord plasma line density will be used to substitute the actual feedback signal for the fueling...
Carlo Neri
(ENEA CR Frascati)
9/7/16, 11:00 AM
C. Plasma Engineering and Control
Poster
The plasma pulse phase of Frascati Tokamak Upgrade (FTU) is driven by the dedicated system FSC (Fast Sequence Control), which has been developed in order to send all the necessary commands to the different power plants feeding the toroidal and poloidal coils during the plasma discharge, meanwhile controlling the correct outcome. In case of incorrect execution of the sequence the system is able...
Andrzej Broslawski
(Narodowe Centrum Badan Jadrowych)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The products of fusion reactions at JET are measured using different diagnostic techniques. One of the methods is based on measurements of gamma-rays, originating from reactions between fast ions and plasma impurities. During the forthcoming deuterium-tritium (DT) campaign a particular attention will be paid to 4.44 MeV gamma-rays emitted in the 99Be(α,nγ)1212C reaction....
Marian Curuia
(Institute of Atomic Physics)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The JET tangential gamma-ray spectrometer (KM6T) is undergoing an extensive upgrade in order to make it compatible with the forthcoming deuterium-tritium (DT) experiments.
The paper will present the design of the main components for the upgrade of the spectrometer beam-line: tandem collimators, gamma-ray shields, and neutron attenuators.
The existing KM6T tandem collimators will be upgraded...
Roch Kwiatkowski
(National Centre for Nuclear Research)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The diagnostic of fast ions at JET is based on the measurements of gamma-rays which are produced as a result of nuclear reactions between ions and plasma impurities. The gamma-ray spectra provide information on energetic tail of ion energy distribution.
The existent BGO detector, with a decay time of ~300 ns, is sufficient during DD campaigns. The strong neutron and gamma-ray fluxes during D-T...
Sorin Soare
(ICIT Rm. Valcea)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
A new diagnostics technique, the Lost Alpha Monitor (LAM), for the investigation of escaping alpha particles in JET has been proposed [1]. The method is based on the detection of the gamma radiation induced by the escaping particles on a target external to the plasma. For a beryllium target this reaction is 99Be(a, nγ)1212C. The implementation on JET of the LAM technique...
Marek Rubel
(Fusion Plasma Physics)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
All optical spectroscopy and imaging diagnostics in next-step fusion devices will be based on metallic mirrors. The performance of mirrors is studied in present-day tokamaks and under laboratory conditions. This work deals with comprehensive tests of mirrors: (i) exposed in JET with the ITER-Like Wall (JET-ILW); (b) irradiation by He and heavy ions to simulate the impact of neutrons under...
Jean-Marie Noterdaeme
(Applied Physics Department)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
High performance H-mode plasmas are characterized by short, repetitive edge perturbations known as edge-localized modes (ELMs). Large, unmitigated ELMs can result in significant transient heat loads released onto the plasma-facing components. Hence, characterization of ELMs and their control are crucial for avoiding a significant reduction in the divertor lifetime. This necessitates...
Zsolt Vizvary
(CCFE)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The ITER first mirrors are the components of optical diagnostic systems closest to the plasma. Deposition may build up on the surfaces of the mirror affecting their ability to fulfil their function. However, physics modelling of this layer growth is fraught with uncertainty. A new experiment is underway on JET, under contract to ITER, with primary objective to test if, under realistic plasma...
Janne Lyytinen
(Smart Industry and Energy Systems, VTT Technical Research Centre of Finland Ltd, Tampere, Finland)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
ITER fusion reactor is a very complex machine which has several different subsystems. It is still a research reactor and the testing results will be implemented in the next generation reactors. In the testing phase of the reactor there will be several sensors and instruments assembled inside the vessel for diagnostics purposes. One of the key diagnostics areas will be the divertor...
Miklos Palankai
(Plasma Physics Department)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Electrical Services provide the electrical infrastructure to serve the diagnostics installed on the ITER Tokamak. The components of the Diagnostics are located all over on the inner and outer shell of the vacuum vessel, in the ports, on the Divertor Cassettes and in the Cryostat as well. Sensors require electrical transmission lines to transmit both of the diagnostic and control signals across...
Christian Vorpahl
(Port Plugs & Diagnostics Integration Division)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Numerous plasma-near mirrors of optical diagnostics of ITER require protection from erosion and deposition caused by impinging energetic particles. This is achieved by approximately 60 individual Diagnostic Shutters, rather simple mechanical devices which obstruct the mirror’s sight towards the plasma when the diagnostic is not in use. If a shutter fails to operate, so does the respective...
Vladislav Kotov
(Institut für Energie- und Klimaforschung – Plasmaphysik)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
First mirrors are plasma facing components which redirect light to the protected optical diagnostics. Initial investigations [A. Litnovsky et al. Nuclear Fusion 49 (2009) 075015, V. Kotov et al. Fusion Eng. Des. 89 (2011) 1583] showed that deposition of impurities (Be, Fe etc.) may cause drastic degradation of the mirror reflectivity and thus severely restrict the diagnostic performance. Very...
Laura Garcia-Ruesgas
(Department of Engineering Graphics)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
During the final design review of Diagnostic Port Plugs, it has been highlighted that the current system of fixation, based on gaps, while it is not harmful for the port plug, it throws large uncertainties over the alignment of the optical systems placed inside the DSMs at the same time that the real mechanical behaviour of the assembly is clearly unknown. Due to the fact that the DSM is not...
Jean-Marc Drevon
(Bertin Systèmes Instrumentation)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
ITER will have a set of 45 diagnostics to ensure controlled operation. Many of them are integrated in the ITER ports. Housed in generic structures, this modular integration is designed to help diagnostics withstanding the plasma loads whilst complying with the French regulations. Now that the Domestic Agencies and ITER Organization are developing the preliminary or even final designs of the...
Yuhu Zhai
(Engineering)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
ITER is the world’s largest fusion device currently under construction in the South of France with over 60 diagnostic systems to be installed inside the port plugs, the interspace or the port cell region of various diagnostic ports. The plasma facing Diagnostic First Wall (DFW) and its supporting Diagnostic Shielding Modules (DSM) are designed to protect front-end diagnostics from plasma...
Kwon Giil
(Control team)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
To achieve the real time controllability of plasma, real-time network is required in fusion experiments place. KSTAR Plasma control system(PCS) adopted the reflective memory (RFM) as a real time network. Since RFM based network has low latency and low jitter. However, KSTAR is also adopted Synchronous Data bus Network (SDN) as real time network to provide real time network to fueling system....
Antonio Carpeno
(Telematics and Electronics Department)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The iRIO-3DLab platform has been devised to enhance the learning process and reduce the development time for engineers in charge of designing intelligent DAQ systems based on PXIe technology and distributed control systems such as EPICS. iRIO-3DLab consists of an Opensim-based virtual world that aims to promote the understanding of how such a kind of DAQ system works, and how the EPICS IOC...
Hiteshkumar Dhola
(Power Supply Group)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
A Dual output (27kV & 15kV), 3MW High Voltage Power Supply (ICHVPS) has been installed and integrated with a Diacrode based RF source to be used for ICRF system. The ICHVPS Controller is based on LabVIEW Real-time PXI controller, which supports all control and monitoring operations of the PSM based power supply. The controller supports all essential features like, fast dynamics, low ripple and...
Bruno Santos
(Instituto de Plasmas e Fusão Nuclear)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Advanced Telecommunications Computing Architecture (ATCA) standard defines a high performance technical solution that meets the requirements for fast controllers on large-scale physics experiments like ITER. This platform provides high throughput, scalability and features for high availability such as redundancy and intelligent platform management which are essential for steady state...
Antonio Rodrigues
(Instituto de Plasmas e Fusão Nuclear)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Control and Data Acquisition (CDAQ) systems applied to large physics experiments like ITER, are designed, among other features, for High-Availability (HA). A CDAQ system based on the PCI Industrial Computer Manufacturers Group (PICMG) 3.x AdvancedTCA Base Specification and Intelligent Platform Management Interface (IPMI) standards grants these features. One of the key functions of the HA is...
Paulo Carvalho
(IPFN/IST)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Advanced Telecommunications Computing Architecture (ATCA) specification implements important key features such as high reliability, high availability, redundancy and serviceability for control and data acquisition instrumentation fault condition, hardware malfunction, firmware updates and hardware reconfiguration.
Red Hat Enterprise Linux and corresponding kernels already have built-in...
Rita C. Pereira
(Instituto Superior Técnio)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Radial Neutron Camera (RNC) and the Radial Gamma-Ray Spectrometer (RGRS) are two ITER diagnostics, devoted, respectively, to the real-time measurement of the neutron emissivity profile (to be used for plasma control purposes) and to the measurement of the confined alpha profile and runaway electrons. The two systems are closely related as they share the same equatorial port plug and part...
Jeremie Dubray
(Ecole Polytechnique Fédérale de Lausanne)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Swiss Plasma Center (SPC) is involved in the development and the operation of gyrotrons for fusion application (TCV tokamak, W7-X, ITER) and for medical application as well (spectroscopy DNP/NMR). In this framework, embedded control systems based on National Instrument (NI) compact Reconfigurable Input Output (cRIO) and compact Data AcQuisition (cDAQ) offer versatile solutions for...
Karishma Qureshi
(Institute for Plasma Research)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Cryogenic Instrumentation is a unique and vast field and requires an in-depth understanding of the process and instrumentation. 26 channels Data Acquisition System is required for the 6 nos. of Cryogenics Pumps LN2 cool down experiment. The data acquisition system measures 22 nos. of temperature signals, 2 nos. of level signals of the buffers and 2 nos. of Nitrogen Dewar Signals (Pressure and...
Adriano Francesco Luchetta
(Consorzio RFX)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Control and Data Acquisition System (CODAS) of SPIDER, the first experiment of the Neutral Beam Test Facility, is under installation and undergoing the commissioning and first operation phases.
The system hardware is nearly compliant with the ITER CODAC catalog for slow and fast plant systems. The system software is based on a combination of software frameworks that altogether collaborate...
Eduardo Rodriguez
(Department of Construction and Manufacturing Engineering)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
This paper is focused on the computation of EM loads induced by plasma current disruptions on the Diagnostics positioned inside the Equatorial Port Plugs, and more explicitly, on the creation of a detailed set of tools (Finite Element ‘FE’ models and routines) which allow the automatic characterization of the EM phenomena (DINA) as well as they provide versatility for the adding/removing of...
Takeo Nishitani
(National Institute for Fusion Science)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Large Helical Device (LHD) plans to start the deuterium experiment in March of 2017, where a maximum neutron yield of 2.1x101616 neutrons/3 sec is expected. For the deuterium experiment, neutron flux monitors, a neutron profile monitor, a neutron activation system and other neutron detectors have been prepared. The characteristics of those neutron diagnostics, such as the...
Gabor Veres
(Department of Plasma Physics)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Devices that are capable of measuring the total plasma radiation in fusion reactor experiments are indispensable for safe and reliable plasma operation. One of the most widespread type of these kind of devices are metal absorber–metal resistor bolometers where the radiation is absorbed by a metallic layer and the change of the layer’s temperature is measured by metal resistors. Based on the...
Rafał Krawczyk
(Institute of Electronic Systems)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The development of GEM detector based acquisition systems resulted in the increase of throughput and resolution in the new revision of the system. The FPGA-based electronics is used to acquire, diagnose and to preliminarily analyze the data of soft X-ray emitted by hot plasma in Tokamak. Moreover, the development of electronics allowed to implement algorithms, so far performed offline after...
B. Bieg
(Institute of Physics)
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
On the basis of the angle variables technique (AVT) changes of polarimetry state of electromagnetic wave passing through the thermonuclear plasma in the poloidal plane have been analyzed. The first section analyzes the changes in polarization state depending on the angle at which the test beam was sent, for the same plasma parameters.
Subsequently, for a given geometry, using numerical...
Jose Martinez-Fernandez
(Laboratorio Nacional de Fusión (LNF))
9/7/16, 11:00 AM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
This work describes the preliminary assessment of the different waveguide technologies for the ex-vessel transmission lines of the Plasma Position Reflectometer (PPR) in ITER.
Initially, both oversized rectangular and circular corrugated waveguides were considered for the study; the former due to reduced costs and ease of procurement and the latter due to better performance in terms of...
Chuan Li
(State Key Laboratory of Advanced Electromagnetic Engineering and Technology)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
This paper mainly introduces the seismic analysis of the high-power dc reactor prototype, whose functions are to limit the ripple current and the increasing rate of fault current in the ITER poloidal field (PF) converter. The stacked reactors with the assembly dimension (L×W×H) of 2955 mm×1639 mm×3296 mm and weight about 5 tons are fixed to the steel base by five support components. In order...
Andrew Ash
(CCFE)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
It is conceivable that electrical arcs can occur during the failure of a large superconducting magnet following an unmitigated quench accident. To assess such accidents, it is important to employ appropriate arc models to calculate the voltage current characteristics and heat dissipation as a function of conditions such as pressure and arc length. Although electrical arcs have been studied for...
Hideki Kajitani
(ITER department)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
Japan Atomic Energy Agency (JAEA), as Japan Domestic Agency, has responsibility to procure 9 ITER Toroidal Field (TF) coils. JAEA completed proto double-pancake (DP) trials aiming at qualification and optimization of manufacturing procedure of TF coil in 2015. Series production of DPs then started and winding of 14 DPs, heat treatment of 11 DPs, fabrication of 9 radial plates (RP), transfer of...
Roberto Bonifetto
(Energy Department)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
The ITER Central Solenoid Model Coil (CSMC) is a superconducting solenoid operated at the JAEA centre of Naka, Japan, since 2000 to test the performance of insert coils in its bore, where it produces a magnetic field of 13 T representative of the ITER CS operating conditions.
In 2015, the ITER Central Solenoid Insert (CSI), whose Nb3Sn cable-in-conduit conductor (CICC) will be adopted for the...
Kurt Schaubel
(ITER CS Project)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
General Atomics (GA) is currently manufacturing the ITER Central Solenoid Modules (CSM) under contract to US ITER at Oak Ridge National Laboratory, under the sponsorship of the Department of Energy Office of Science. The contract includes the design and qualification of manufacturing processes and tooling necessary to fabricate seven CSM (6 + 1 spare) that constitute the ITER Central Solenoid....
Alberto Ferro
(Consorzio RFX)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
The Residual Ion Dump Power Supply (RIDPS) is part of the Ground Related Power Supplies, to be manufactured by OCEM Energy Technology s.r.l. (OCEM) for the MITICA experiment and for the two ITER Heating Neutral Beam Injectors (HNBI). MITICA is the full-scale prototype of the HNBI, under construction in the PRIMA Neutral Beam Test Facility in Padua, Italy.
The RIDPS is devoted to feed the...
Vanni Toigo
(Consorzio RFX)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
The Neutral Beam Injector (NBI) is required to inject in ITER plasma Deuteron particles which, once generated in the Ion Source (IS) polarized at -1MV, are accelerated at ground potential and then neutralized. This voltage level is very demanding for the power supply system, requiring several non-standard components. This paper describes the design status of two main NBI components: High...
Francesca Cau
(Fusion for Energy)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
The winding pack of the ITER Toroidal Field (TF) coils is composed of 134 turns of Nb3Sn Cable in Conduit Conductor (CICCs) wound in 7 double pancakes and cooled by supercritical helium (He) at cryogenic temperature. The cooling of the Stainless Steel (SS) case supporting the winding pack is guaranteed by He circulation in 74 parallel channels. A 2D approach to compute the temperature...
Rustam Enikeev
(Efremov Institute)
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
The superconductive coils of ITER magnet system will be energized by ac/dc converters. Before each plasma pulse the magnet system will be pre-charged with energy (8GJ) to be used for generating the toroidal loop voltage required for the gas mixture breakdown and plasma formation. This will be realized by inserting energy dissipating resistors in series with the central solenoid (CS) modules...
Maksim Manzuk
(Joint Stock Company "D.V. Efremov Institute of Electrophysical Apparatus")
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
High current DC switches play a very important role in the ITER coil power supply system (CPSS) being key components of its two major parts: switching network units (SNU) for plasma initiation and fast discharge units (FDU) for superconducting coils energy extraction in case of quench. For both functions, circuit-breakers rated up to 70 kA steady-state current and 10 kV voltage are required...
50027.
P3.088 On optimization of air cooling system of FDR dissipating energy from ITER magnet coils
Victor Tanchuk
(JSC "NIIEFA")
9/7/16, 11:00 AM
E. Magnets and Power Supplies
Poster
The Fast Discharge Resistors (FDR) under development at NIIEFA are intended together with switching equipment to dissipate energy released in case of quench of the ITER superconducting coils, thereby protecting them against failure.
FDRs are made of sections consisting of resistive elements enclosed in boxes. Two-four sections stacked vertically form a separate module.
During energy release...
Neway Atnafu
(Engineering)
9/7/16, 11:00 AM
A. Experimental Fusion Devices and Supporting Facilities
Poster
NSTX-U COILS BUS BARS DESIGN AND CONSTRUCTION**
Neway D. Atnafu, L. Dudek, A. Khodak, S. Gerhardt, S. Ramakrishnan, M. Smith, P. Titus
Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451
natnafu@pppl.gov
The construction of the NSTX upgrade project was completed in the fall of 2015. The multi-year capital project was budgeted at $94 Million. The reactor will used to run...
Katsunao Uenishi
(Osaka University)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Sputtering properties of tungsten (W) should be evaluated correctly for lifetime estimation of divertor components. Especially, at elevated temperatures, recrystallization would cause grain structure reconstruction, which would influence sputtering properties and surface morphology changes. However, the detailed studies haven’t been performed.
Actually, the temperature of divertor could...
Takeru Maeji
(Osaka University)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Currently, In regard to the plasma facing material, Tungsten (W) is a major candidate at ITER. A recent study has been reported indicating that the transient thermal load such as ELM or disruption causes metal surface melting or evaporation of W. However, the property and behavior of the W above the melting point has not yet been sufficiently known, and many of the previous studies are...
Daisuke Inoue
(Osaka University)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Tungsten (W) is a primary candidate of plasma-facing materials for fusion reactors. But erosion due to melting and evaporation of W caused by transient heat loads are concerned. A pulsed laser simulating the transient heat loads was irradiated to three tungsten materials and the behavior of the molten layer was investigated. In addition, aluminum (Al) and tin (Sn) was deposited on W and the...
Vladimir Khripunov
(Fusion Reactor Department)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Primary radiation damage (atomic displacements) and Helium and Hydrogen production rates in plasma facing components (PFCs) of a fusion system are usually determined by the high energy parts of neutron spectra formed in plasma chamber from the initial fusion neutron source. According to presented estimates, the energetic alphas and protons, appearing in PFC materials in the (n,a) and (n,p)...
Dmitry Terentyev
(SCK-CEN)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Recent theoretical and subsequent experimental studies suggest that the uptake and release of deuterium (D) in tungsten (W) under high flux plasma exposure (i.e. under ITER-relevant conditions) is controlled by dislocation microstructure induced by the plasma itself. A comprehensive mechanism for the nucleation and growth of D bubbles on dislocation network under high flux low-energy plasma...
Bong Guen Hong
(Chonbuk National University)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
We investigate the ablation characteristics of plasma facing materials (PFM) using thermal plasma facilities. A high enthalpy, 400 kW plasma testing facility which uses an enhanced segmented arc torch as a plasma source and 55 kW vacuum plasma spraying system produce particle flux greater than 102424/(m22sec) and heat flux greater than 10 MW/m22, levels that...
Samuel A. Humphry-Baker
(Department of Materials)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
High-field spherical tokamaks may be a viable technology for relatively compact fusion power devices (Costley et al Nucl. Fus. 2015). However, such reactors leave little space for shielding of the central column, which must protect the inner superconducting magnets from high energy neutrons. Tungsten carbide cermets are promising candidate materials for such shields: They have high thermal...
Valentina Marascu
(National Institute for Laser)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Controlled fusion research represents an important step for sustainable energy production once with the development of the International Thermonuclear Experimental Reactor (ITER). ITER proposes a deuterium - tritium fusion reaction for hot plasma creation. During plasma- wall interactions, small tungsten particles, from nm to microns will be produced in the tokamak chamber. These particles can...
Richard E. Nygren
(Sandia National Laboratories)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Power exhaust is perhaps foremost among the issues for ITER and post-ITER devices, as well as for existing large confinement devices as they increase power. A related concern is the alignment of plasma facing components to avoid protruding (leading) edges that would intercept field lines and incur very high loads and high erosion. This concern prompted the transient melt experiment in JET,...
Rodrigo Mateus
(Instituto de Plasmas e Fusão Nuclear)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Migration of impurities during ITER plasma discharges will result in the formation of co-deposited mixed materials on the surface of plasma facing components (PFC) with properties distinct from those of the original PFC. These issues have motivated the fusion community to investigate Be-W coatings, in particular their fuel retention behaviour, since in ITER the deposits will present a...
Liga Avotina
(Institute of Chemical Physics)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Tungsten covered carbon materials due to good thermal conductivity of carbon based materials (up to ~250 Wm-1-1K-1 -1 for carbon fiber composites [1]) are suitable for use in fusion devices, like ITER (International Thermonuclear Experimental Reactor) [2], as divertor materials. However, during the plasma wall interactions, erosion and re-deposition, as well as formation...
Hanns Gietl
(Max-Planck-Institut für Plasmaphysik (IPP))
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Tungsten is a promising plasma facing material for future fusion reactors due to its unique property combination such as low sputter yield, high melting point and low activation. The main drawbacks for the use of pure tungsten are the brittleness below the ductile-to-brittle transition temperature and the embrittlement during operation e.g. by overheating and neutron irradiation. This...
Cristian Ruset
(Plasma Phisics and Nuclear Fusion)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Tungsten coatings deposited on carbon materials such as carbon fibre composite (CFC) or fine grain graphite (FGG) are currently used in fusion devices as armour for plasma facing components (PFC). About 1800 CFC tiles were W-coated for the ITER-like Wall at JET and more than 1300 FGG tiles were coated for the ASDEX Upgrade tokamak. At present the W coating production is on going for the first...
Keisuke Azuma
(Graduate School of Science)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Tungsten (W) is a candidate for plasma facing materials in D-T fusion reactors due to its higher melting point and lower sputtering yield. During the plasma operation, W will be exposed to energetic particles including hydrogen isotopes, neutrons, and impurities like carbon (C). It is well known that hydrogen isotopes are trapped in the defects produced by the energetic particle irradiation....
Yuya Miyoshi
(Japan Atomic Energy Agency)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Understanding of the heat load profile on the first wall (1stst wall) is one of the key issues to establish the DEMO blanket concept, because the thermal stress on the each blanket module depends on its surface heat load, and it will vary with the 1stst wall shape, the toroidal/poloidal position and the plasma equilibrium. Thus, the 1stst wall surface of the...
Sebastian Ruck
(Institute of Neutron Physics and Reactor Technology)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Rib-roughening the helium-gas cooled channels in plasma-facing components of DEMO (First Wall (FW), limiters or the divertor) enhances heat transfer and reduces structural material operation temperatures. The rib-elements induce a three-dimensional, unsteady flow field and heat transfer is augmented by mixing the fluid in the near wall regions and boundary layers. Whereas the overall heat...
Julien Aubert
(DEN)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
The EUROfusion Consortium develops a design of a fusion power demonstrator plant (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. Among the 4 candidates for...
Ali Abou-Sena
(Institute of Neutron Physics and Reactor Technology)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
The First Wall (FW) of the EU Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) faces the fusion plasma and experiences high heat fluxes; therefore its cooling channels design is a key R&D task for qualifying the HCPB TBM for the fusion reactors ITER and DEMO. Within the manufacturing and qualification activities performed in KIT for the HCPB TBM, a First Wall Mock-up (FWM) was...
Tomas Melichar
(Research Centre Rez)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Dual Coolant Lithium Lead (DCLL) is one of the four breeding blanket concepts being developed within the EUROfusion project as candidates for the European DEMO. One of the most challenging components of breeding blanket in terms of thermal-hydraulic is a first wall. In order to handle the high thermal loads that the DCLL first wall is facing a proper design of a helium cooling system is...
50048.
P3.120 Development of force reconstruction method on EU ITER TBM based on strain measurements
Christian Zeile
(Karlsruhe Institute of Technology (KIT))
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
The EU ITER Test Blanket Module (TBM) sets, which consist of TBM box and shield, will be located inside the equatorial port #16 of ITER. One of the important objectives of the TBM program, starting from the first H-H phase, is the validation of the theoretical predictions of the structural behavior of the TBM set under thermal, mechanical and electromagnetic loads. High electromagnetic forces...
Taishi Sugiyama
(Graduate school of energy science)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
DEMO reactor must achieve total TBR >1 with high level of accuracy and confidence in the design process. However there is no relevant neutron sources before ITER /TBM, and even in ITER, neutron field is considerably different due to the shield blankets surrounding TBMs. This study proposes verification technique to experimentally simulate reactor neutron field and evaluates its expected...
Ivan Alessio Maione
(Institute for Neutron Physics and Reactor Technology)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Off-normal operations in Tokamak reactors result in the induction of eddy currents that, coupled with the large magnetic field, impose strong electromagnetic forces (Lorentz’s forces) to fusion reactor components. In addition the presence of ferromagnetic material induces Maxwell’s forces as interaction between the magnetized material and the external magnetic field that are thus present also...
Tristan Batal
(IRFM)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test W monoblock Plasma Facing Units (PFU) under long plasma discharge (up to 1000s), with thermal loads of the same magnitude as those...
Sergey Grashin
(NRC "Kurchatov Institute")
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
In 2015 the graphite limiter was replaced by the tungsten one on the T-10 tokamak. The limiter was made in “Efremov Institute” from the ITER-grade “POLEMA” tungsten used for ITER divertor plates manufacturing. “POLEMA” tungsten doesn’t contain any impurities and has a high thermal conductivity and heat capacity. Tungsten has a polycrystalline structure with a grain size about 30µm. The...
Aleksey Arakcheev
(Budker Institute of Nuclear Physics)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
The residual mechanical deformation and stress were measured in the preliminary experiments carried out at synchrotron radiation (SR) scattering stations on VEPP-3 in the Siberian Center of Synchrotron and Terahertz Radiation. Significant changes in the SR diffraction are found as the result of material recrystallization or irradiation of the material by plasma or high energy ions. It implies...
Vladimir Weinzettl
(Institute of Plasma Physics of The Czech Academy of Sciences)
9/7/16, 11:00 AM
F. Plasma Facing Components
Poster
Dust transport is among important issues for ITER and DEMO, where material erosion will be significant. One of possible mechanisms how material is eroded from plasma facing surfaces is the remobilization of dust particles linked to their lifetime there and to the formation of dust accumulation sites. On the COMPASS tokamak, dust remobilization experiments have been performed using a tungsten...
Christian Bachmann
(Power Plant Physics and Technology)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
An essential goal of the EU fusion roadmap is the development of design and technology of a Demonstration Fusion Power Reactor (DEMO) to follow ITER. A pragmatic approach is advocated considering a pulsed tokamak based on mature technologies and reliable regimes of operation, extrapolated as far as possible from the ITER experience. The EUROfusion Power Plant Physics and Technology Department...
Fabio Cismondi
(Eurofusion-PPPT)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
In the framework of the EUROfusion DEMO Programme, the Programme Management Unit (PMU) is assuming the role of the plant and tokamak design integration. It is recognized, in part thanks to the ITER experience, that due to the large number of complex systems assembled into the tokamak vessel for integration it is of vital importance to address the in-vessel integration at an early stage in the...
Gandolfo Alessandro Spagnuolo
(Institute for Neutron Physics and Reactor Technology (INR))
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The development of the fusion technology reliability involves, among other issues, the improvement of simulation tools to be used for the design of reactor key components, such as the Breeding Blanket (BB), where the engineering requirements and constraints are of nuclear, material and safety kind. For this reason, advanced simulation tools are needed. In the European DEMO project, several...
Giuseppe Mazzone
(Unità Tecnica Fusione)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Among the design activities of the DEMO divertor cassette carried out in the frame of EUROfusion an important parameter is the operating temperature of the divertor cassette. As for the DEMO breeding blanket Eurofer has been chosen as structural material of the divertor cassette due to its low long-term activation, low creep and swelling behavior under neutron fluence. The choice of the...
Domenico Marzullo
(Department of Industrial Engineering)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
This paper presents the pre-conceptual design activities conducted for the European DEMO divertor, focusing on cassette design and Plasma Facing Components (PFC) integration. Following the systems engineering principles for the conceptual stage, high level design requirements are collected and conceptual 3D model of divertor’s cassette is presented. The work moved from the geometrical and...
Youji Someya
(Sector of fusion research and development)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Periodical replacement of in-vessel components is required for DEMO. The surface dose rate of in-vessel components for DEMO with fusion power of 1.5 GW is higher than that of shielding blanket in ITER by double digits. In addition, DEMO requires five-year cooling time for decreasing its dose rate to the level of ITER. Therefore, it is difficult to adopt the in-vessel maintenance scheme as ITER...
Peter Titus
(Analysis Branch Mechanical Engineering Division)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The Korean fusion demonstration reactor (K-DEMO) is in the early stages of conceptual design. Ceramic breeder blanket modules are being investigated. These have had extensive nuclear and thermal evaluations. Structural assessments are in process. This paper presents stress analyses performed at PPPL in support of the blanket design. Disruption loading, including the effects of ferromagnetic...
Rocco Mozzillo
(Industrial Engineering)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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). A new WCLL BB concept design has been proposed and investigated, starting from DEMO 2015 reference configuration. The first activity driving the BB design...
Hiroyasu Utoh
(Japan Atomic Energy Agency)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Maintenance is one of the critical issues in the DEMO design. Several maintenance schemes has been comparatively evaluated from the viewpoint of plasma positional control, in-vessel transferring mechanism of blanket segment, and pipe connection in order to establish a feasible reactor maintenance scheme on the DEMO reactor. Two options has been selected as likely remote maintenance schemes on...
Ming Li
(Mechanical Engineering)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
In the inside engineering of DEMO, the robotic machines or manipulators are foreseeable to be widely employed, which often have to deal with the demanding working conditions. The construction of the dynamic model of the robotic machine or manipulator can not only benefit the performance evaluation of the manipulator in the early design stage, but also can be incorporated into the control...
Alberto Vale
(Instituto de Plasmas e Fusao Nuclear)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
In DEMO, the ex-vessel Remote Maintenance Systems (RMS) are responsible for the replacement and transportation of the plasma facing components. The ex-vessel operations of transportation are performed by cranes or by means of cask transfer systems (CTS) moved by trolleys.
The main loads of transportation are the blankets and divertors. The blankets are extracted and transported vertically by...
Dan Wolff
(RACE)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
As part of the conceptual design studies for a European DEMO, a range of Tokamak geometries are being considered. As identified in the EFDA Roadmap to the realisation of Fusion Energy: “The integration of the Remote Maintenance system within the DEMO plant is an essential task within the DEMO CDA phase. This will involve establishing requirements, functions and interfaces with many other...
Romain Sibois
(Remote Operation and Virtual Reality)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The next European fusion reactor after ITER is called DEMO. The development implementing ITER experiences has taken place within EUROfusion Programme. One of the reactor maintenance system development tasks has been focused on Divertor Maintenance system. The maintenance of DEMO involving handling hazardous components shall be carried out remotely such as the installation and removal of the...
Kumarpalsinh Jadeja
(Institute for plasma Research)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The First Indian tokamak, ADITYA had successfully completed 25 years of operation of limiter plasma at the Institute for Plasma Research (IPR). After achieving the targeted plasma and successfully carrying out many major tokamak experiments, the up-gradation of ADITYA tokamak with diverter configuration was planed. The upgradation includes the replacement of rectangular cross section vacuum...
Hitoshi Tamura
(Department of Helical Plasma Research)
9/7/16, 11:00 AM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The design activity of a conceptual design of a helical fusion reactor FFHR-d1 is progressing at the National Institute for Fusion Science. The superconducting magnet system of FFHR-d1 comprises one pair of helical coils, two sets of vertical field coils, and the coil support structure. The major and the minor radii of the helical coil are 5.6 m and 3.774 m, respectively. The magnetic field at...
Axel von der Weth
(INR-MET)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
The hydrogen isotopes Tritium and Deuterium will be the fuel of future fusion power plants. These isotopes will be in contact with components of the reactor, as well as with auxiliary systems. For safety studies and the overall Tritium budget, hydrogen transport parameters are necessary to perform according analyses. Reduced Activating Ferritic Martensitic (RAFM) steels at operation conditions...
Marta Malo
(Fundación UNED-Ciemat)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Tritium permeation through containment structures is an important factor for safety and design analysis of fusion energy systems. This process controls several key aspects of the system performance, including the amount of radioactive tritium released to environment, the requirements on tritium breeding ratio, the tritium recycling from the first wall, and it influences the selection of...
Maribel C Gazquez
(Fusion National Laboratory)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Al-based coatings are proposed as anti-permeation and anti-corrosion barrier in Pb-Li breeding blankets -Water Cooled Lithium-Lead (WCLL), Helium Cooled Lithium-Lead (HCLL) and Dual Coolant Lithium-Lead (DCLL). In this work, Al2O3 coatings have been prepared by Pulsed Laser Deposition (PLD) at Istituto Italiano di Tecnologia (IIT) and they have been qualified in Pb-Li to evaluate its...
Takumi Chikada
(Graduate School of Integrated Science and Technology)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Tritium permeation through structural materials in fusion blankets is one of the most important issues in terms of a fuel loss and radiological hazard. Tritium permeation barriers (TPBs) have been developed for several decades, and erbium oxide (Er2O3) coatings have recently been intensively studied as TPBs. However, irradiation effects in TPB coatings on hydrogen isotope permeation have not...
Keisuke Mukai
(IAM-KWT)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
In a helium cooled pebble bed (HCPB) DEMO reactor, ceramic breeder pebbles are packed in EUROFER structural steel blanket and generate tritium as a consequence of the reactions between lithium and neutrons. As breeder pebbles and EUROFER are contacted at a high temperature for a long period during the operation, corrosive attack to EUROFER could occur even with the low activities of ceramic...
Jae Sung Yoon
(KAERI)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Korea has designed a helium cooled ceramic reflector (HCCR) test blanket module (TBM), including a TBM shield, called a TBM set, that will be tested in ITER. The HCCR TBM is composed of four sub-modules and a back manifold. In addition, each sub-module is composed of a first wall (FW), a breeding box with seven-layer breeding zone (BZ), and side walls with the cooling path. Korean RAFM steel...
Belit Garcinuno
(Fusion Technology Division)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Tritium recovery is one of the major issues of a future DEMO reactor, in order to accomplish with the requirement of tritium self-sufficiency. Different techniques have been proposed over the years for the extraction of tritium, depending on the Breeding Blanket technology. After a preliminary selection, the EUROfusion Programme has considered the Permeation Against Vacuum (PAV) technique as...
Arthur Brooks
(Engineering Analysis)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Modeling Blanket Ferromagnetic Loading using Edge Potential Elements
Arthur W Brooks 1 1, Han Zhang11
1Princeton Plasma Physics Laboratory abrooks@pppl.gov
Future fusion experiments and reactors will require first wall materials that can survive the thermal and nuclear radiation environment without structural degradation. Candidate materials that are under consideration...
Kenzo Ibano
(Graduate School of Engineering)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
For the fusion reactor operations, the tritium (T) retention and permeation in the reactor walls are important for points of views of safety and fuel cycle. It is known that T retention in tungsten (W) is less severe compared with carbon (C). However, recent experimental studies revealed that the neutron irradiated damage, surface recrystallization, and fuzz formation by He ion irradiation...
Matthias Kolb
(Institute for Applied Materials)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Advanced ceramic breeder pebbles composed of a mixture of Li4SiO4 (LOS) and Li2TiO3 (LMT) are fabricated and developed at KIT by a melt-based process (KALOS). The produced pebbles are easily characterized for their non-nuclear properties. Yet, as the main properties of a tritium breeder material are the generation and release of tritium, these characteristics also have to be examined.
Neutron...
Zhenxing Liu
(Department of Reactor Engineering Research & Design)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Abstract: This paper presents the results of experimental study of the columns packed with Palladium deposited on kieselguhr (Pd/k). The characteristic of pressure resistance and separation of hydrogen isotopes of the Pd/k column was investigated. The corresponding relationships among pressure resistance characteristics of Pd/k separation column and Pd/k material physicochemical properties,...
Fred Thomas
(York Plasma Institute)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Tritium self-sufficiency is a fundamental requirement for future DT fusion demonstration and commercial power plants. Hence, prior to the construction of expensive, complex fusion breeder blanket assemblies there should be a concerted effort to quantify and ultimately reduce the uncertainties associated with various nuclear observables. This will enable tritium self-sufficient blankets to be...
Richard Walker
(CCFE)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
In preparation for the design of a future tritium-handling plant for the DEMO fusion reactor, a study was undertaken to consider the activation of gases, in addition to those used as fuel, which are to be injected into DEMO for the purpose of reducing damage to the divertor.
Likely impurity gases were identified as nitrogen, neon, argon, krypton and xenon, with no clear consensus as to which...
Jonathan Klabacha
(Nuclear Engineering)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Looking towards the future of fusion devices, detailed understanding of the underlying working properties is desired knowledge. Even though there are many fusion devices available and extensive operating data is being collected, computational analysis is an underlying requirement to fully understand how a fusion device will operate. Due to the extensive complexity of fusion devices a...
Jonathan Shimwell
(Culham Centre for Fusion Energy)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
The Helium Cooled Pebble Bed (HCPB) breeder blanket is being developed as part of the European Fusion Programme. Part of the programme is to investigate blanket modules relevant for future demonstration fusion power plants. This paper presents fluid dynamic, thermomechanical and neutronic analyses of the helium cooled pebble bed with an alternative neutron multiplier, Be12Ti which is...
Julia M. Heuser
(Institute for Applied Materials (IAM))
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
The investigation of Ceramic Breeders (CB) is of great concern for the development of the solid breeder concept for ITER and DEMO. To ensure an adequate tritium production of the breeder material several requirements like a high lithium density, good tritium release behaviour, and a high resistance against irradiation as well as thermomechanical stresses have to be fulfilled. Lithium...
Lida Magielsen
(Research and Innovation)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
In the frame of the European Tritium Breeder blanket development for DEMO, two high dose irradiations of beryllium and beryllides, to be used as neutron multiplier, have been performedin the High Flux Reactor Petten (NL). From one irradiation, to 3000 appm He, the post irradiation results have been published in previous proceedings.
In the second High Dose Beryllium irradiation (HIDOBE-02),...
Huaqin Kou
(Institute of Materials)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Fast and efficient activation of ZrCo is beneficial to promote its application to hydrogen isotopes storage in the fusion energy field. To obtain the optimum activation procedures, the influences of temperature and hydrogen pressure on the activation behavior of ZrCo were systematically investigated. Experimental results showed that fast and efficient activation of ZrCo could be achieved by...
Wei Li
(School of Nuclear Science and Technology)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Chinese Fusion Engineering Test Reactor（CFETR）is a necessary and feasible engineering test reactor which aims at developing the fusion energy while the helium cooled solid breeder blanket (HCSB) is one of the most significant component of it. During the reactor operation stage, the blanket will be activated to produce highly radioactive substances by high energy neutrons irradiation. In order...
Hyoseong Gwon
(Department of Blanket System Research)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Decay heat produced by neutron irradiation can lead to temperature rise in blanket even after plasma shutdown. The excessive temperature increase of blanket structure would be concerned with increase of decay heat when assuming loss of cooling capability for blanket even though vacuum vessel is assumed to be normally cooled with a safety function. The neutron wall loading is designed to be...
Yi-Hyun Park
(TBM Technology Team)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Lithium-containing ceramics (Li-ceramics) are considering as tritium breeding material in pebble-bed form for solid-type breeding blanket in fusion reactor. The tritium breeding material requires small particle size to reduce diffusion distance of generated tritium in the intercrystalline. In addition, the essential resource, especially enriched Li-6, has to recover from the used tritium...
Masaru Nakamichi
(Sector of Fusion Research and Development)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Hydrogen generation via an oxidation reaction of beryllium as an existing neutron multiplier with steam at high temperatures should be reduced on safety hazard for a fusion reactor. Therefore, advanced neutron multipliers with high stability at high temperatures are desirable for the fusion reactor in which water coolant is extensively used. Beryllium intermetallic compounds (beryllides) are...
Ryoutarou Yamamoto
(Advanced of Energy Engineering science)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Understanding of Li evaporation property is important because Li mass transfer decreases tritium breeding ratio and influences tritium behavior. In JAEA, the development of Li2TiO3with excess Li has been performed as an advanced tritium breeder. The present authors revealed in previous works that a layer existing on the pebble surface includes Li2CO3 and it contributes Li mass loss. Recently,...
Yu Otani
(Prime Mover Engineering)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Lithium metatitanate (Li2TiO3) is one of the candidate materials among the solid tritium breeders proposed because of its good tritium release property and high chemical stability [1]. Lithium metatitanate with excess Li (Li2+xTiO3+y) has been recognized as an another prominent candidate material owing to its higher Li density [2]. Demonstration power plant (DEMO) reactors require tritium...
Kiyoto Shin-mura
(Course of Mechanical Engineering)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Lithium metatitanate (Li2TiO3) is one of the candidate materials for solid tritium breeder proposed because of its good tritium release property and high chemical stability [1], and Lithium metatitanate with excess Li (Li2+xTiO3+y) has been recognized as a prominent candidate material owing to its higher Li density [2]. However, demonstration power plant (DEMO) reactors require tritium...
Arturs Zarins
(Institute of Chemical Physics)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Modified lithium orthosilicate pebbles with additions of titanium dioxide are suggested as an alternative tritium breeding ceramic for the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM). The tritium breeding ceramic in the HCPB TBM will be under the action of harsh operation conditions. Radiolysis can take place as a result, and unstable radiation-induced defects (RD) and radiolysis...
Marigrazia Moscardini
(Institute for Applied Materials)
9/7/16, 11:00 AM
H. Fuel Cycle and Breeding Blankets
Poster
Five ITER project members are actively involved in the fabrication of tritium breeding ceramics pebbles. Different fabrication processes developed by these members strongly influence the characteristics of pebbles produced. One of the main characteristics is the sphericity of pebbles. The spherical shape is the one desired; however the manufacture of perfect round particles is not simple. For...
Fernando Sanchez
(National Fusion Laboratory (LNF))
9/7/16, 11:00 AM
I. Materials Technology
Poster
SiC is a primary candidate for flow channel inserts in blankets due to their excellent thermo-mechanical properties. During reactor operation SiC will be exposed to tritium in a hostile radiation environment. Absorption, diffusion, and desorption will occur, and are expected to depend on the neutron and ionizing radiation conditions. We present work to assess the effect of displacement damage...
Yasuhisa Oya
(College of Science)
9/7/16, 11:00 AM
I. Materials Technology
Poster
Silicon carbide (SiC) is considered to be used for blanket modules for high temperature gas–cooling system in D-T fusion reactors, as SiC/SiC composites. During D-T fusion operation, SiC will be exposed to heavy radiation conditions by neutron and/or gamma-ray. These radiation induces the formation of various damages by a collision process and an electron excitation process, leading to the...
Changho Park
(Japan Agency for Quantum and Radiological Science and Technology)
9/7/16, 11:00 AM
I. Materials Technology
Poster
Lead−lithium (Pb−Li) alloy are considered as a coolant and a tritium breeder for fusion reactor blanket systems. One of the critical requirements for the realization of this systems is the compatibility of silicon carbide (SiC) and its composites as structural and/or functional materials. The authors investigated that inclusions, possibly Li−oxides in Pb−Li may have certain impacts on...
Enrique Ascasibar
(National Fusion Laboratory)
9/7/16, 11:00 AM
I. Materials Technology
Poster
During ITER and DEMO reactor operation the proposed Li-Pb blanket flow channel inserts made of SiC ceramic material will be exposed to both radiation and tritium. Absorption, diffusion, and desorption of tritium is expected to occur and these processes will strongly depend on the irradiation conditions, neutron flux, and purely ionizing radiation. Previous results have shown that marked...
Saerom Kwon
(Japan Atomic Energy Agency)
9/7/16, 11:00 AM
I. Materials Technology
Poster
In our previous copper benchmark experiment we had pointed out that the elastic scattering and capture reaction data of the copper had included some problems in the resonance region, which had caused a large underestimation of reaction rates of non-threshold reactions. In order to corroborate this issue, we carried out a new benchmark experiment on copper in the neutron field with more low...
Masayuki Ohta
(Japan Atomic Energy Agency)
9/7/16, 11:00 AM
I. Materials Technology
Poster
In our previous benchmark experiment on molybdenum at JAEA/FNS, we found problems of the (n,2n) and (n,γ) cross sections in Mo of JENDL-4.0. However, the Mo data only above a few hundred eV were investigated, because there were few neutrons with lower energy in the Mo assembly in the previous experiment. We perform a new benchmark experiment on Mo in order to validate the Mo data in the lower...
Snejana Bakardjieva
(Institute of Inorganic Chemistry AS CR)
9/7/16, 11:00 AM
I. Materials Technology
Poster
Materials from the group of layered Mn+1AXn phases are new type of nanolaminates which can be used in many technical applications, especially as viable candidates for high-radiation structural application in fusion technology.
It has been proposed that the novel physical properties of MAX phases arise from their atomic structure, combining “ceramic” MX6 octahedra layers with a single...
Jan Engels
(Institut für Energie- und Klimaforschung – Plasmaphysik)
9/7/16, 11:00 AM
I. Materials Technology
Poster
In fusion power plants a tritium permeation barrier is required in order to prevent the loss of the fuel inventory. Moreover, the tritium permeation barrier is necessary to avoid that the radioactive tritium accumulates in the first wall, the cooling system, and other parts of the power plant. Oxide thin films, e.g. Er2O3 and Y2O3, are promising candidates as tritium permeation barrier layers....
Monika Vilemova
(Materials Engineering)
9/7/16, 11:00 AM
I. Materials Technology
Poster
Pure tungsten is considered as the most suitable plasma facing material for the reactor first wall. However, number of studies points out serious drawbacks related to tungsten mechanical properties that negatively affect lifetime of first wall components. Serious risk for the divertor comes from abnormal events, such as disruptions, vertical displacement events (VDEs) and edge localized modes...
Sven-Erik Wulf
(Institute for Applied Materials)
9/7/16, 11:00 AM
I. Materials Technology
Poster
Different breeding blanket designs for a future fusion power plant (DEMO) consider Eurofer steel as a main structural material. Nevertheless, RAFM steels suffer from severe corrosion attack in Pb-15.7Li, which acts as breeding material in the liquid breeder blanket designs, e.g. HCLL, WCLL and DCLL. The resulting corrosion products may cause safety risks e.g. concerning tube plugging due to...
Shuhei Nogami
(Department of Quantum Science and Energy Engineering)
9/7/16, 11:00 AM
I. Materials Technology
Poster
Tungsten (W) is a primary candidate for fusion reactor divertor because of its high melting point, thermal conductivity and sputtering resistance. To improve its structural reliability, improvement of mechanical properties and suppression of recrystallization of the W materials are necessary. It is well known that the grain refining, work hardening, solid solution strengthening, and dispersion...
Anatoli Popov
(Institute of Solid State Physics)
9/7/16, 11:00 AM
I. Materials Technology
Poster
The radiation-resistant insulators (MgO, Al2O3, MgAl2O4, BeO etc) are important key materials for fusion reactors. It is very important to predict/simulate not only the kinetics of diffusion-controlled defect accumulation under neutron irradiation, but also a long-time defect structure evolution including thermal defect annealing. Here we developed and applied the advanced theoretical approach...
Teruya Tanaka
(National Institute for Fusion Science)
9/7/16, 11:00 AM
I. Materials Technology
Poster
In our previous study, a Cr2O3 layer was formed on a reduced activation ferritic/martenstic (RAFM) steel substrates by heat treatment under a reduced atmosphere and it could suppress hydrogen permeation by ~2 orders at 550-650 ooC. Since the Cr2O3 layer was stable at high temperatures in air, it was also a preferable underlayer for multi-layer ceramic coating with the metal organic...
Jumpei Mochizuki
(Shizuoka University)
9/7/16, 11:00 AM
I. Materials Technology
Poster
Tritium permeation barrier (TPB) has been investigated for the establishment of an efficient fuel cycle and radiological safety in fusion power plants. One of critical issues for TPB is degradation caused by introduction of cracks and pores. Even if a microscopic crack is introduced, tritium permeation is drastically increased. The development of self-healing coating is one of techniques for...
Seira Horikoshi
(Graduate School of Integrated Science and Technology)
9/7/16, 11:00 AM
I. Materials Technology
Poster
To establish liquid lithium-lead blanket concepts, the development of a functional coating as a tritium permeation barrier with corrosion resistance is required. In our previous study, erbium oxide (erbia)-iron two-layer coatings showed a better compatibility than erbia single-layer coatings with keeping a high permeation reduction factor (PRF). In this study, hydrogen isotope migration...
Hynek Hadraba
(Institute of Physics of Materials)
9/7/16, 11:00 AM
I. Materials Technology
Poster
The structural components used for construction of future generation of fission reactors and fusion reactors will undergo demanding service conditions as high neutron doses, high temperature and extremely corrosive environment. The nano-structured oxide dispersion steels (ODS) containing small amounts of homogeneously dispersed nano-size yttria particles were developed as structural material...
Filip Siska
(Institute of Physics of Materials)
9/7/16, 11:00 AM
I. Materials Technology
Poster
ODS steels are candidates for the structural material in the future fusion power plants. Their main advantage is high strength and creep resistance at high temperatures. Such high performance is achieved by the presence of the oxide particles in the microstructure. Nowadays, the best ODS steels contain particles of Y2O3 which are stable at high temperatures. However, yttrium is expensive and...
Simon Heuer
(Forschungszentrum Jülich GmbH)
9/7/16, 11:00 AM
I. Materials Technology
Poster
Future fusion reactors may exhibit first walls composed of a tungsten (W) armor, that is attached to a subjacent stainless steel (SS) structure. Joining these materials for the application at hand is challenging because the pulsed operation of TOKAMAK reactors induces thermo-mechanical stresses and strains at the W/SS interface due to differing materials properties. These cyclic loads will...
Ming Sun
(Institute of Nuclear Energy Safety Technology)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Fusion reactor is one of new type reactors being developed , and it is cleaner and more efficient than the fission reactor. Each SSCs (Structures, Systems, Components) has different safety importance to fusion reactors. So it is necessary to classify the SSCs of fusion reactors. And the safety classification of SSCs for fusion reactor is the important basis of reactor design and construction....
Miao Nie
(Key Laboratory of Neutronics and Radiation Safety)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The high reliability and availability of Tritium extraction system (TES) will be needed is necessary for safety operation of circulation and processing of tritium purge gas. Reliability, availability, maintainability, inspectability (RAMI) analysis of the TES should be performed during the design and operation phase. Since there is no TES failure rate data available from fusion operating...
Shijun Qin
(Tokamak Design Division)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the EAST in-vessel components cooling system based on currently available design is presented. The following sub-systems were considered in the analysis: the EAST PFCs heat-sink cooling system, two water pumps system, cooling loop including cycle feed pipe and cycle return pipe lines, secondary...
Paul-Martin Steffen
(Institute of Energy and Climate Research – Nuclear Waste Management and Reactor Safety (IEK-6))
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
In case of a severe accident inside the ITER fusion facility, there exist several scenarios in which hydrogen may be produced and released into the suppression tank. Assuming the accidental ingress of air, the formation of flammable gas mixtures may lead to explosions and severe component failure. One option to mitigate such hypothetical scenarios is the installation of passive auto-catalytic...
Tonio Pinna
(Nuclear Fusion and Safety Technologies Department (FSN-FUSTEC-TEN))
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Safety studies are performed in the frame of the conceptual design studies for the European DEMO reactor to assess the safety and environmental impact of design options. An exhaustive set of reference accident sequences are defined in order to evaluate plant response in the most challenging events and compliance with safety requirements. The identification of a comprehensive set of accident...
Richard Brown
(PMU)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The generation and investigation of alternative design solutions and their benchmarking against criteria that are traceable to high level objectives is a fundamental facet of a holistic systems engineering approach. During the pre-conceptual design phase of DEMO, characterisation studies for multiple plant concepts are being conducted in parallel to explore the design space and evaluate the...
Fabrizio Franza
(Institute for Neutron Physics and Reactor Technology)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
A fusion power plant is characterized by many subsystems operating under extreme thermal and nuclear conditions, thus compelling to be designed according to physics and engineering constraints. For such an operation, dedicated tools called systems codes are currently used. At Karlsruhe Institute of Technology (KIT), a dedicated modelling campaign has been recently launched aiming to study the...
Lei Lu
(Neutronics and Nuclear Data Group)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
McCad is a geometry conversion tool developed at the Karlsruhe Institute of Technology (KIT) for the automatic conversion of CAD models into the constructive solid geometry (CSG) representation. The resulting geometry models can then be used in Monte Carlo (MC) particle transport simulations applied in design analyses of fusion reactors like the DEMO tokamak developed within the European Power...
Xiaoman Cheng
(Institute of Plasma Physics)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). In this work, the Primary Heat Transfer System (PHTS) of the WCCB blanket was designed based on the configuration of the blanket sectors, employing two identical loops at this stage. And each loop consists of a steam generator, a pressurizer and a main pump,...
Taehyun Tak
(KSTAR Control Research Team)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The ITER Central Interlock System (CIS) architecture is composed of four categories of hardware: fast architecture, slow PLC based architecture, hardwired architecture and servers.
The CIS fast architecture receives interlock events from various local plant systems of ITER and communicates the corresponding actions to any other local plant systems in order to avoid or mitigate the damage to...
Rafael Juarez
(Departamento de Ingeniería Energética)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
ITER is a prominent facility in the development of the nuclear fusion. It presents 44 ports providing access to the Vacuum Vessel at three different heights: Lower, Equatorial and Upper ports. Out of them, 22 ports, correspond to Diagnostics ports. They host a diversity of diagnostics systems, designed by the different ITER Domestics Agencies (DAs). They are later integrated into the different...
Jia Li
(School of Nuclear Science and Technology)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
In order to control the global sample frequency, GVR method is deemed to be a practical way. But it is common that GVR method needs too many steps of weigh window iteration and it may fall into a long-history problem. We introduce a novel approach that is GVR method combined with reduced density in model, which could improve the calculation efficiency of GVR method in the following two...
Shengpeng Yu
(Institute of Nuclear Energy Safety Technology)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The advantages of CAD based Automatic Modeling make it possible to efficiently describe and verify complex nuclear system, such as ITER, for Nuclear Analysis. SuperMC/MCAM, the most widely applied CAD based Automatic Modeling tool for Monte Carlo, is currently focusing on modeling for Monte Carlo partile transport programs.
Being more and more detailed, the radiation shielding modeling of...
Miguel Correia
(Instituto de Plasmas e Fusão Nuclear)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
High availability (HA) is a key element in the specification of next generation Fusion devices, targeting steady-state operation. HA is especially required on mission-critical systems, as is the case of experimental Fusion devices and future Fusion power plants, where safety of people, environment and the infrastructure/investment is a primordial priority.
IPFN developed control and data...
Qi Yang
(Key Laboratory of Neutronics and Radiation Safety)
9/7/16, 11:00 AM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Activation study is very important for fusion reactors, from the view of component maintenance, occupational radiation exposure, and radioactive waste management. SuperMC is a multi-functional, intelligent, accurate and user-friendly simulation software system with comprehensive functions of transport simulation, material activation and transmutation, radiation source term and dose, etc.
The...
