Tomas Markovic
(Institute of Plasma Physics of the CAS)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
A number of tokamaks, including the largest operating one, Joint European Torus (JET), has ferromagnetic core installed in their plasma current drive system. Moreover, some auxiliary systems, such as magnetic shielding of neutral beam injection (NBI) system, or iron inserts for toroidal field ripple mitigation, consist of non-negligible amount of ferromagnetic material as well. Besides the...
Alastair Shepherd
(Culham Centre for Fusion Energy)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Neutral beam injection systems have proved themselves as the most effective form of auxiliary heating in tokamak plasmas. In positive ion based systems once the beam is neutralised there are many residual ion components which must be intercepted by suitable ion dumps. A particular challenge for ion dump design occurs when the dump must be placed close to a focus point as is the case for the...
Eva Belonohy
(JET Exploitation Unit)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The final phase of the JET Programme in Support of ITER plans to operate with 100% Tritium (TT) followed by Deuterium-Tritium (DT) operation, to help minimise risks and delays in the execution of the ITER Research Plan and the achievement of Q~10. Additional technical requirements (compared to Deuterium operation) are needed to allow operation with Tritium gas, a high DT neutron flux and...
Rosaria Villari
(EUROfusion Consortium)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Neutronics benchmark experiments are conducted at JET for validating the neutronics codes and tools used in ITER nuclear analyses to predict quantities such as the neutron flux along streaming paths and dose rates at the shutdown due to activated components. In particular, in the frame of subproject NEXP of JET-3 program, several activities are performed within EUROfusion Consortium devoted to...
Roberto Ambrosino
(Engineering department)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The European roadmap to the realisation of fusion energy has identified a number of technical challenges and defined eight different missions to face them. Mission 2 ‘Heat-exhaust systems’ addresses the challenge of reducing the heat load on the divertor targets.
Divertor Tokamak Test (DTT) facility [1]-[2] has been launched to investigate alternative power exhaust solutions for DEMO. This...
Alessandro Anemona
(ICAS)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
In the European Fusion Roadmap, one of the main challenges to be faced is the mitigation of the risk due to the impossibility of directly extrapolate to DEMO the divertor solution adopted in ITER, due to the expected very large loads. Thus a satellite experimental facility oriented toward the exploration of robust divertor solutions for power and particles exhaust and to the study of...
Gustavo Granucci
(IFP-CNR)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The proposed Divertor Test Tokamak, DTT, aims at studying power exhaust and divertor load in an integrated plasma scenario. Additional heating systems have the task to provide heating to reach a reactor relevant power flow in the SOL and guarantee the necessary PSEP/R together adequate plasma performances. About 40 MW of heating power are foreseen to have PSEP/R ≥ 15 MW/m. A mix of the three...
Giorgio Maddaluno
(FSN)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
on behalf of the EUROfusion WPDTT2 team & the DTT report contributors
Within the frame of the DTT program, included in the EuroFusion roadmap, the design of a new Tokamak dedicated to tackle the Power Exhaust problem as an integrated bulk and edge plasma problem has been developed. The main guidelines used to work out the machine parameters will be shortly illustrated.To allow the machine...
Giuseppe Di Gironimo
(Department of Industrial Engineering)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
This paper describes the activity addressed to the conceptual design of the first wall and the main containment structures of DTT device, which will be broadly presented in the invited talk "Design and definition of a Divertor TOKAMAK Test facility".
The work moved from the geometrical constraints imposed by the desired plasma shape and the configuration needed for the magnetic coils. Many...
Giuseppe Mazzitelli
(Consorzio CREATE & Seconda Università di Napoli)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The DTT (Divertor Test Tokamak) is a new facility conceived in the frame of EUROfusion roadmap with the aim to assess and possibly integrate all the relevant physics and technology divertor issues.
The general project is presented in another paper of this conference [1] and with more details in [2].
The general project includes the analysis of the site requirements from several points of view;...
Alessandro Lampasi
(Department of Fusion and Technology for Nuclear Safety and Security)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The power supplies (PSs) of the DTT proposal, as presented in the talk "Design and definition of a Divertor Tokamak Test facility" invited at this conference, have to feed:
6 central solenoid (CS) and 6 poloidal field (PF) superconducting coils, with currents up to 25 kA.
18 toroidal field (TF) superconducting coils, with a current up to 50 kA.
Some fast plasma control coils, including at...
Marco Utili
(FSN-ING)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The experimental facility THALLIUM (Test HAmmer in Lead LithIUM) was designed to experimental validate the RELAP5-3D code simulations of the pressure wave propagation in the HCLL TBM due to In-box LOCA. THALLIUM, which reproduces the geometry of the LLE loop of the HCLL TBM, was installed at the ENEA Brasimone Research Centre to support the accidental analysis of this type of test blanket...
Gianluca Barone
(Dipartimento di Ingegneria Civile e Industriale)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The 1stst Specific Grant of the Framework Partnership Agreement 372 deals with experimental activities in support of the Conceptual Design of HCLL and HCPB Test Blanket Systems. Service-2 is focused on thermal-hydraulic tests with high pressure Helium for validation and benchmarking of suitable dedicated numerical tools. In this frame, an extensive experimental campaign has been...
Alessandro Venturini
(Department of Civil and Industrial Engineering)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The experimental facility IELLLO (Integrated European Lead Lithium LOop) was designed and installed at the ENEA Brasimone Research Centre to support the design of the HCLL TBM.
This work presents the results of the experimental campaign carried out within the framework of F4E-FPA-372 and which had three main objectives. First, to produce new experimental data for flowing LLE (Lead-Lithium...
Liqin Hu
(Institute of Nuclear Energy Safety Technology)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Due to the complexity of fusion reactors on geometry and neutron physics, the Monte Carlo (MC) methods have been broadly adopted in fusion nuclear design and analysis. But for calculations that require obtaining a detailed global flux map, such as the shutdown dose rate analysis, analog MC simulations usually cost a prohibitive long run time. To make such analysis computational practicable, it...
Jing Song
(Institute of Nuclear Energy Safety Technology)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Great challenges exist in real fusion engineering projects for the current Monte Carlo (MC) methods including the calculation modeling of complex geometries, simulation of deep penetration problem, slow convergence of complex calculation, lack of experimental validation for new physical features, etc.
Several novel and advanced capabilities of the latest version of MC program SuperMC for...
Michal Kresina
(DEN)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Operation of fusion facilities using deuterium and tritium to fuel the fusion reaction will lead to generation of radioactive waste during operating and decommissioning phases. Most of these wastes are expected to be contaminated with tritium and will require a specific management strategy taking into account the physical and chemical properties of tritium. The reference management strategy...
Mercedes Medrano
(National Laboratory for Magnetic Fusion)
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The superconducting tokamak JT-60SA, aimed to support and complement the ITER experimental programme, is currently being assembled at the JAEA laboratories in Naka (Japan). Within the European contribution, Spain is responsible for providing JT-60SA cryostat.
The cryostat is a stainless steel vacuum vessel 14m diameter, 16m height which encloses the tokamak providing the vacuum environment...
Peter Lang
(Tokamak Scenario Development Division (E 1))
9/6/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
A conceptual design for a pellet injection system will be worked out, capable to support key missions of the new tokamak device JT-60SA. For exploitations in view of ITER and to resolve key physics and engineering issues for DEMO, several tasks were assigned to this system. Physics investigations aim at operation at high density in ITER and DEMO relevant plasma regime above Greenwald density,...
Nicolo Marconato
(Consorzio RFX (CNR)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The ITER Heating Neutral Beam (HNB) injectors shall be protected from stray magnetic field (several hundreds of mT) produced by the ITER PF coils and plasma current. Such stray field would hamper the production of negative ions, deflect ion trajectories in the accelerator and cause intolerable heat load on neutralizer and beam line components. In order to keep the residual magnetic field below...
Daniele Aprile
(Consorzio RFX)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
In the multi-beamlet, negative-ion based Heating Neutral Beam (HNB) Injectors presently used in fusion research, arrays of permanent magnets are embedded in the Extraction Grid (EG) for the suppression of the unwanted co-extracted electrons. These magnets cause a significant undesired deflection of the negative ion beamlets, with a typical alternate pattern, matching the orientation of the...
Stefan Hanke
(Institute for Technical Physics)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The gas cloud inside the neutralizer of MITICA (Megavolt ITER Injector and Concept Advancement), required to neutralize the negative ion beam, will be created continuously by 20 identical nozzles providing the gas needed for different operation modes. In order to validate the design, one nozzle will be characterized in detail and for a wide range of supply conditions in a dedicated experiment...
Loris Zanotto
(Consorzio RFX)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The Acceleration Grid Power Supply supplies the acceleration grids of the MITICA experiment, the full scale prototype of the ITER Neutral Beam Injector under construction in Padua (Italy) to tackle the technical challenges and prepare for the target performance objectives ahead of operation in ITER.
The AGPS is a special switching power supply with demanding requirements: high rated power (55...
Bernd Heinemann
(ITER Technology & Diagnostics)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The negative ion source test facility ELISE represents the first step in the European R&D roadmap for the neutral beam injection (NBI) systems of ITER in order to consolidate the design and to gain early experience with a large and modular Radio Frequency (RF) negative ion source. The aim of ELISE is to demonstrate the ITER requirements with respect to extracted negative hydrogen densities...
49767.
P2.025 Preparation of the ELISE test facility for long-pulse extraction of negative ion beams
Riccardo Nocentini
(ITER Technology and Diagnostics)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The test facility ELISE (Extraction from a Large Ion Source Experiment) at IPP Garching, Germany, aims to demonstrate ITER-relevant negative ion beam parameters which are required for the NBI system of ITER. ELISE is equipped with a Radio Frequency driven source and an ITER‑like extraction system with half the ITER size. An H-- or D-- beam can be extracted for 10 s every...
Chandramouli Rotti
(Diagnostic Neutral Beam)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The Beam Line Components (BLCs) for the ITER Diagnostic Neutral Beam (DNB) and Indian Test Facility (INTF) are mainly water cooled elements made from CuCrZr which are designed to absorb heat flux up to 10MW/m2 2 (e.g. Heat Transfer Element for calorimeter) according to their position in beam line. The design of these components imposes stringent requirements of having the specific...
Jaydeepkumar Joshi
(Diagnostic Neutral Beam (DNB))
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The acceleration system of Beam Source(BS) of Neutral Beam(NB) system is composed of water cooled Oxygen-Free Copper multi-aperture grid systems which is designed for focusing of beamlets to a focal point located at distance>20m from the Grounded Grid. For present application in the accelerator for DNB, this focusing is obtained using a combination of segment bending and aperture offsets. In...
Hiroyuki Tobari
(Naka Fusion Institution)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
Design and manufacturing of DC 1 MV components have progressed for the ITER neutral beam injector.
A multi-conductor DC 1 MV transmission line (TL) which can transmit five-different voltages of 200 kV step simultaneously has been manufactured and tested. The TL is a gas insulation tube with SF6 gas of 0.6 MPa. A layout of those conductors inside the tube was designed through electric field...
Jong-Gu Kwak
(NFRI)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The main mission of KSTAR program is exploring the physics and technologies of high performance steady state tokamak operation that are essential for future fusion reactor. Since the successful long pulse operation of 25sec at 0.5MA exceeding conventional tokamak capabilities in 2013, the duration of H-mode has been extended to over 50s which corresponds to a few times of current diffusion...
Haejin Kim
(KSTAR Research Center)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
Helicon wave coupling for efficient off-axis current drive using a traveling wave antenna has been proposed. It was found that helicon wave can drive plasma current in the mid-radius of high electron beta plasmas in medium and large size tokamak due to moderate optical thickness and wave alignment nature of helicon wave in helical magnetic field. KSTAR tokamak can be a good platform to test...
Hyunho Wi
(KSTAR Research center)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
Steady-state operation of a DEMO-like tokamak requires substantial off-axis current be driven by external current drive systems. Non-inductive current drive is needed to complement the bootstrap current to support the plasma current in steady state. Recently, helicon wave current drive at frequencies of 500~700 MHz is gained much attention to achieve off-axis current drive with high...
Jeehyun Kim
(Heating and current drive team)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The KSTAR LHCD system is to be upgraded for RF power up to 4 MW in 2020. The basic configuration of the system is composed of eight 5-GHz 500-kW CW klystrons, low-loss transmission line with oversized circular waveguide, and PAM launcher for the mid-plane injection. An off mid-plane injection near the upper diverter is also under consideration. A preliminary study based on a mid-plane PAM...
Taesik Seong
(Department of Physics)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The KSTAR LHCD system is using a 5-GHz, 0.5-MW c. w. klystron and oversized rectangular waveguides. The WR187 output waveguide of the klystron transmits the RF power to the LH launcher via 80-m of transmission line composed of WR284 oversized rectangular waveguide. The overall transmission loss was about 34% including 26% of Ohmic loss. In order to transfer RF power effectively from a klystron...
Julien Hillairet
(IRFM)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The coupling of lower hybrid (LH) range of frequencies waves to strongly magnetized plasmas is a critical issue on tokamaks as the RF power which can be transferred from the antenna to the plasma is often limited by the quality of this coupling. Development of new types of antennas aiming at improving the ability of the antenna to handle large powers in stationary conditions, as it will be...
Carl Pawley
(DIII-D)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
In the DIII-D tokamak, one of the most powerful techniques to control the density, temperature and plasma rotation is by eight independently modulated neutral beam sources with a total power of 20 MW. The rapid modulation requires a high degree of reproducibility and precise control of the ion source plasma and beam acceleration voltage. Recent changes have been made to the controls to...
Brendan Crowley
(DIII-D National Fusion Facility)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The Neutral Beam system on DIII-D consists of eight ion sources. The basis of the DIII-D NB system is the Common Long Pulse Source (CLPS). The CLPS is an 80 kV high perveance, deuterium positive ion based system delivering up to 2.5 MW per source. The ion source is a filament driven magnetic bucket design and the accelerator is a slot and rail tetrode design with vertical focusing achieved...
Mirela Cengher
(General Atomics)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The gyrotron complex on DIII-D has been updated and comprises six gyrotrons installed and routinely operating reliably for injection of up to 3.6 MW into the plasma. The operational maximum of 5 s pulse length for the six gyrotrons allows up to 18 MJ total energy to be injected into the plasma. Recent system upgrades include faster launcher mirror scans and control by the plasma control...
Chiara Piron
(Consorzio RFX (CNR)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
The RAPTOR - RApid Transport simulatOR code [F. Felici et al 2011 Nucl. Fusion 51 083052] is a model-based control-oriented code that predicts Tokamak plasma profile evolution in real-time. One of its key applications is in a state observer, where the real-time predictions are combined with the measurements of the available diagnostics, yielding a complete estimate of the plasma profiles.The...
Raffaele Martone
(Department of Industrial and Information Engineering, Seconda Università di Napoli, Aversa, Italy)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
The Reversed Field Pinch configurations are characterized by strong asymmetries [1]; in order to prevent or mitigate possible consequent instabilities, suitable control systems are required. In RFX-mod (Padua, Italy), such a system includes a number of 192 saddle coils, independently controlled, fully covering the toroidal surface and operating in a coordinate strategy. An equal number of...
Paolo Bettini
(Consorzio RFX)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
RFX-mod is equipped with an advanced active control system of MHD instabilities, which consists of 48x4 saddle coils, housed inside a stainless steel Toroidal Support Structure, and 48x4 radial field sensor loops processed in real time to drive the currents in the control coils. Thanks to the high flexibility of this system [1], RFX-mod operations in the last years have allowed to reach the...
Luca Grando
(Consorzio RFX)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
RFX [1] was originally designed with a load assembly consisting of a vacuum vessel (VV) and a thick aluminum stabilizing shell, with two poloidal and two equatorial cuts (i.e. gaps). After several years of experimental campaigns, a major modification of the RFX load assembly has been introduced [2], consisting in the substitution of the aluminum shell with a thin Copper Shell (CS) and the...
Zhengping Luo
(Institute of Plasma Physics)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
The Parallel plasma equilibrium reconstruction code PEFIT [1], first developed for real-time plasma shape control of the EAST tokamak (and capable of one full equilibrium reconstruction in 300ms with a calculation grid size in 65x65) is being adapted for use on MAST. PEFIT is based upon the EFIT equilibrium code algorithm, but rewritten in C using the CUDATMTM architecture in order...
Seongcheol Kim
(Department of Nuclear Engineering)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
Mitigation of heat and particle fluxes reaching on divertor plates is still a critical problem even though innovative divertor concept such as super-X and snowflake divertors have been suggested. A new divertor concept for the reduction of heat and particle fluxes is to convert thermal energy to electrical energy by separating electrons from the plasma with appropriate magnetic field....
Ngoc Minh Trang Vu
(IRFM)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
The control of the safety factor q and/or the electronic temperature Te profiles is a key issue to achieve advanced plasma scenarios with high repeatability. This paper will discuss the new results of such plasma internal profile control on TCV, using total plasma current Ip, and ECCD heating source. The issue is that only the ECCD heating power is controlled, since the distributed heating...
Galina Kuzmina
(National Research Centre Kurchatov Institute, Moscow, Russian Federation)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
Presented work is related to the development and creation of hardware and software of Plasma Control System (PCS) platform of the modernized now tokamak T-15 [1] for the integration, configuration, testing and start-up algorithms for the calculation of electrical installation parameters, as well as for the modeling of the experiment scenario with taking into account of the real-time magnetic...
Heung-Su Kim
(National Fusion Research Institute)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
Noise width (δV/V) and drift level (ΔV/Δt) in the magnetic measurements by using sensors such as magnetic field probes (MPs) and flux loops (FLs) has been fully satisfied with the requirements (δV/V < 2% and (ΔV/V)/ Δt < 2% for 60 s), for the plasma control in the KSTAR tokamak before the in-vessel control coil (IVCC) is used to control plasma shapes. From the experimental campaign of 2010,...
Jan Horacek
(Tokamak)
9/6/16, 2:20 PM
C. Plasma Engineering and Control
Poster
In order to avoid surface melting of divertor targets of big tokamak fusion reactors by localized ELM heat loads, we study a technique of spreading the flux by harmonic divertor strike point sweeping with a dedicated in-vessel twin-coil. If the sweep frequency gets above 1/tELMdecaydecay~300 Hz, local ELM plasma heat flux suppresses significantly (by...
Arkady Serikov
(Institute for Neutron Physics and Reactor Technology)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
This paper presents new results of neutronics analysis performed in support for the design development of the Tritium and Deposit Monitor (TDM) to be installed inside the ITER Equatorial Port Plug (EPP) #17. This monitor is a laser based diagnostics to provide information about the tritium content in the deposited layer on the inner baffle of the ITER divertor. Neutronics analysis is performed...
Raul Luis
(Instituto de Plasmas e Fusão Nuclear)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations, also known as gaps 3, 4, 5, and 6, complementing the magnetic diagnostics system. For gaps 4 and 6, the antennas are to be installed in-vessel between two blanket shield modules. The microwave signal is...
Nuno Cruz
(Instituto de Plasmas e Fusão Nuclear)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Radial Neutron Camera (RNC) diagnostic is a neutron detection system with multiple collimators aiming at characterizing the neutron emission that will be produced by the ITER tokamak. The RNC plays a primary role for basic and advanced plasma control measurements and acts as backup for system machine protection measurements.
To achieve its goals, the RNC diagnostic needs to acquire,...
Fabio Moro
(Department of Fusion and Nuclear Safety Technology)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The ITER Radial Neutron Camera (RNC) is a multichannel detection system hosted in the Equatorial Port Plug 1 (EPP 1) designed to provide information on the neutron source total strength and emissivity profiles through the measurement of the uncollided neutron flux along a set of collimated lines of sight (LOS). Furthermore the ion temperature profile and fuel ratio (nd/nt) can be assessed by...
Anders Hjalmarsson
(Department of Physics and Astronomy)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The High Resolution Neutron Spectrometer (HRNS) system for ITER is an array of neutron spectrometers with the primary function to provide measurements of the fuel ion ratio, nT/nD, in the plasma core. Supplementary functions are to assist or provide information on fuel ion temperature and energy distributions of fuel ions and confined alpha-particles. The ITER requirement for the HRNS primary...
Mykyta Varavin
(Institute of Plasma Physics AS CR)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The COMPASS tokamak is equipped by the 2-mm microwave interferometer. This interferometer measures the electron density integrated along the central chord. Two VCO oscillators stabilized by the PLL together with multipliers generate two probing waves of the close frequency 139.3 and 140 GHz. The digital 2π-phase detector in the receiving part compares the phase between these probing waves. The...
Pavel Hacek
(Faculty of Mathematics and Physics)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Atomic beam probe (ABP) is a diagnostic tool using a detection of ions coming from an ionized part of a diagnostic beam in tokamaks. The method allows measurements of plasma density fluctuations and fast variations in the poloidal magnetic field. Therefore, it gives the possibility to follow fast changes of edge plasma current, e.g. during ELMs in H-mode.
The test detector has been installed...
Ales Havranek
(Institute of Plasma Physics of the CAS)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The COMPASS tokamak has been recently equipped with two new fast color cameras Photron FASTCAM Mini UX100 operating in visible light. A new node, including both software and hardware, was developed for these cameras to ensure automatic and reliable operation integrated to the control and data acquisition system of COMPASS. The node provides camera function control, parameter setting, data...
Petr Vondracek
(Institute of Plasma Physics of the CAS)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
A new fast infrared camera Telops FAST-IR 2K was purchased on the COMPASS tokamak recently. It is equipped with a MWIR (medium wavelength infrared, 3-5 μm) InSb detector and is possible to reach framerate of 1.917 kHz in a full frame acquisition mode (320x256 px.) and up to 90 kHz in a sub-windowed acquisition (64x4 px.).
The camera allows e.g. automatic exposure control, providing autonomous...
Jaromir Zajac
(Institute of Plasma Physics AS CR)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The microwave reflectometry system on COMPASS tokamak uses the frequency modulated continuous wave (FM-CW) in K and Ka bands. The fast swept synthesizer together with the simple homodyne detection provides the complex beat frequency spectrum for the density profile reconstruction. The homodyne detection scheme limits the other applications like the Doppler reflectometry, therefore the sheme is...
Mark Szutyanyi
(Department of Mathematics and Computational Sciences)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The physics of Edge Localized Modes (ELM) is one of the most studied scientific fields in fusion research. Automatic detection of ELMs in different diagnostic signals is an important initial step during massive experimental data analysis.
This contribution contains the description of the generalized Sequential Probability Ratio Test (g-SPRT) method used for automatic ELM detection in different...
Michael Grahl
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
WENDELSTEIN 7-X and its superconducting coil system is designed for research on steady-stateoperation of stellarators. This sets high requirements on the control and data acquisition (CoDaC)system, with the archive database as one of its main components. W7-X ArchiveDB [1] is the centralstorage system for all engineering and scientific data. It stores raw data as well as processed data...
Andre Carls
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Wendelstein 7-X (W7-X) has been finally commissioned in 2015 and is now in its first stage of operation. Due to the complex structural design and a limited life time of some components, each step of W7-X commissioning and operation is carefully monitored by a considerable amount of different sensors.
Unlike the fast machine control or the fast experiment data acquisition, the machine...
Dirk Pilopp
(E4)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
WENDELSTEIN 7-X (W7-X) is a superconducting helical advanced stellarator which is currently in operation phase 1.1 at the Max-Planck-Institut für Plasmaphysik in Greifswald.
During this startup period five uncooled inboard poloidal limiter structures made from fine corn graphite protect the plasma vessel wall, since the divertor, heat shields and carbon tiles are not installed yet. At 10 ports...
Didier Chauvin
(CEA de Cadarache DSM/IRFM)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Wendelstein 7-X fusion device at Max-Planck-Institut für Plasma Physik (IPP) in Greifswald produced its first hydrogen plasma on 3rdrd February 2016. This marks the start of scientific operation. Wendelstein 7-X is to investigate this configuration’s suitability for use in a power plant. In order to allow for an early integral test of the main systems needed for plasma operation...
Ireneusz Ksiazek
(Institute of Physics)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The C/O monitor for W7-X will be a spectrometer of special construction with high throughput and high time resolution, suitable for controling concentration of main impurities in plasma. The spectrometer will be fixed at horizontal position and at wavelengths corresponding to Lyman a lines of H-like ions of oxygen (at 1.9 nm), nitrogen (at 2.5 nm), carbon (at 3.4 nm) and boron (at 4.9 nm). Its...
Guruparan Satheeswaran
(Forschungszentrum Jülich GmbH)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
A multi-purpose manipulator (MPM) system is attached at an outer cryostat vessel port in the equatorial plane to transport electrical probes and targets to the edge of the inner vessel. From this parking position where the tip of the probe coincides with the inner vessel wall a fully controlled movement into the edge plasma for all magnetic field configurations is feasible. The distributed...
Tamas Szabolics
(Wigner Research Centre for Physics)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
In the past few years a ten channel video diagnostics system was developed, built and installed for Wendestein 7-X stellarator (W7-X). The system is based on EDICAM (Event Detection and Intelligent Camera) CMOS cameras (400 fps @ 1.3 Mpixel). In the first W7-X experimental campaigh (OP1.1) the video diagnostic system is not integrated into the central control and data acquisition system of...
Tomasz Fornal
(Department of Nuclear Fusion and Plasma Spectroscopy)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Measurements of soft X-ray radiation from plasmas is a standard diagnostic which is used in many different fusion devices. Analysis of X-ray emission delivers among others, information about the electron density and temperature as well as can deliver an information about the impurity content in the plasma.
The paper describes design of the soft X-ray diagnostic, multi-foil system (MFS,) for...
Natalia Krawczyk
(Department of Nuclear Fusion and Plasma Spectroscopy)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Wendelstein 7-X (W7-X) stellarator started its operation at the end of 2015. The first operation phase is conducted both with helium and hydrogen as working gas and has achieved first plasmas in the order of 500ms at the time this abstract has been written. The initial experiments have also been devoted to commissioning, tests and optimization of diagnostic systems.
In this paper we report...
Christian Brandt
(Max-Planck-Institute for Plasma Physics)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The quasi-steady state high power plasma experiments at Wendelstein 7-X are expected to become pioneering research benchmarking the advanced stellarator concept. The results will bring comparisons to the huge amount of experimental findings in other stellarator and tokamak devices. After the successful start of hydrogen plasmas in February 2016, the set of plasma diagnostics will be extended...
Ulrich Neuner
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Thirteen Rogowski coils have been installed in the vacuum vessel of the stellarator Wendelstein 7-X (W 7-X). They are designed to measure the equilibrium plasma currents as Pfirsch-Schlüter current and bootstrap current. The coils will be calibrated using a conductor positioned inside the plasma vessel with an alternating current passing through it. The response of the coils is measured and...
Dirk Nicolai
(Institut für Energie- und Klimaforschung - Plasmaphysik)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The investigation of edge plasmas at W7-X requires a flexible tool for integration of a variety of different diagnostics as e. g. electrical probes, probing magnetic coils, material collection, or material exposition probes, and gas injection. A multi-purpose manipulator (MPM) system has been developed and attached to the W7-X vessel before the operational phase 1.1. The system was designed as...
Youngseok Lee
(KSTAR Research Center)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Long-pulse D-D plasma operation in the annual KSTAR plasma campaign is performed and involved Ohmic heating and auxiliary heating such as a neutral beam injection (NBI) of high power with deuterium beams. The NBI heating power reached up to 6 MW at the moment.
In addition, many energetic runaway electrons are also observed through hard-X ray (HXR) monitoring during the operation. Runaway...
Jong-ha Lee
(National Fusion Research Institute (NFRI))
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
To measure Zeff profile, most plasma machine equipped brehmsstrahlung measurement system like as filterscope diagnostic. In KSTAR, however, a new type brehmsstrahlung measurement system developed and tested at single point in KSTAR 10th campaign in last year.[1] In 2016 KSTAR campaign, to Zeff profile measurement, we expand this concepts of brehmsstrahlung measurement system to multi points;...
Y. Yu
(School of Nuclear Science and Technology)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Abstract：In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum...
Md Mahbub Alam
(Advanced Energy Engineering Science)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
In QUEST (Q-shu University Experiments with Steady-State Spherical Tokamak), the achievement of the steady-state operation for long time discharge is one of its project objectives. For the achievement of the long time discharge, the identification of the plasma shape and position in real-time is important during the operation of the tokamak. By observing the temporal behaviours of the plasma...
Andrea Rizzolo
(Consorzio RFX)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
This paper describes the final design of the Short-Time Retractable Instrumented Kalorimeter Experiment (STRIKE) for the SPIDER experiment (Source for Production of Ions of Deuterium Extracted from Radio frequency plasma) under construction at the Consorzio RFX premises. The STRIKE diagnostic will be used to characterise the SPIDER beam during short pulse operation (several seconds) to verify...
Gabriele Croci
(Physics)
9/6/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Neutron measurements are proposed for the SPIDER/MITICA Neutral Beam Injection (NBI) prototypes in Padua. Neutron emission is here due to reactions between the beam and the adsorbed deuterons in the target and thus depends on the deuteron absorption level in the beam calorimeter.
We have investigated such process at the “half size” ITER NBI ELISE facility of the Max-Planck Institut. A first...
Zito Pietro
(FSN-FUSTEC-IEE)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
JT-60SA is a Superconducting Tokamak in the framework of the Broader Approach Agreement between Europe and Japan. For this International Project, both the Italian National Agency for New Technology, Energy and Sustainable Economic Development (ENEA) and Comissariat à l’Energie Atomique et aux Energies alternatives (CEA) are providing ten AC/DC converters for the poloidal superconducting...
49823.
P2.082 Final tests of four switching network units procured by the European Union for JT-60SA
Miguel Pretelli
(Power Electronics)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Switching Network Units (SNUs) are inserted in the power supply circuits of modern tokamaks for plasma initiation. In the framework of the “Broader Approach” agreement, the four SNUs for the superconducting modules of the JT-60SA Central Solenoid will be procured by European Union through the Italian Agency ENEA.
The design is based on the synchronized operations of a light electromechanical...
Elena Gaio
(Power System)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Effective control of Resistive-Wall-Modes (RWM) is mandatory in JT-60SA, the satellite tokamak under construction in Naka (Japan), since one of its main objectives is to reach steady-state high-beta plasmas.
The RWM control system is based on a set of 18 in-vessel sector coils, placed on the plasma side of a conductive wall and individually fed by a dedicated fast power supply system...
Kyohei Natsume
(Tokamak System Technology)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
JT-60SA is a tokamak device using superconducting coils to be built in Japan, as a joint international research and development project involving Japan and Europe. The JT-60SA helium refrigerator system (HRS) supplies supercritical or gaseous helium to cold components: superconducting coils, coil supporting structures, cryopumps, high temperature superconductor current leads (HTS CL), and...
Katsuhiko Tsuchiya
(QST)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
The programme of constructing JT-60SA device is progressing as a satellite tokamak of the Broader Approach project. JT-60SA has superconducting poloidal field (PF) coil system which is procured by JAEA, and 18 D-shaped toroidal field (TF) coils of which Europe has been in charged. PF coil system consists of a central solenoid (CS) with four solenoid modules and six circular coils which are...
Paolo Rossi
(ENEA)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
In the framework of the Broader Approach program, ENEA is in charge of the in-kind supply of 18 Toroidal Field (TF) coil casings for the superconducting tokamak JT-60SA being assembled in Naka site, Japan.
ENEA commissioned the company Walter Tosto (Chieti, Italy) the fabrication of two sets of 9 casings to be delivered to ASG Superconductors (Genoa, Italy) and GE (Belfort, France), in charge...
Sylvie Nicollet
(IRFM)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
The Toroidal Field system of the JT-60SA tokamak comprises 18 NbTi superconducting coils. In each TF coil (TFC), 6 Cable-In-Conduit Conductor (CICC) lengths are wound in 6 double-pancakes (DP) and carry a nominal current of 25.7 kA at a temperature of 5 K. These coils are tested in the Cold Test Facility (CTF, CEA Saclay), the test program including a quench for each of the first coils of the...
Gian Mario Polli
(FSN)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
ENEA, in the framework of Broader Approach program for the early realization of fusion with the construction of JT-60SA tokamak, has committed to procure 9 of the 18 TF coils of JT-60SA magnet system. Within 2016 six coils will be completed and delivered to the cold test facility in Saclay, France, for the final acceptance tests before their shipment to Naka site for the...
Daniel Ciazynski
(IRFM/STEP)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
The Toroidal Field system of the JT-60SA tokamak is composed of 18 NbTi superconducting coils. Half of them are provided by France within the Broader Approach Agreement. These coils are manufactured by General Electric (ex-Alstom) at Belfort, France. Each TF coil is composed of 6 cable-in-conduit conductor lengths, wound in double-pancakes, carrying a nominal current of 25.7 kA at a...
Yawei Huang
(Institute of Research into the Fundamental Laws of the Universe)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
In order to check the performance of the JT-60SA Toroidal Field (TF) coils and hence mitigate their possible fabrication risks, a series of tests have been carried out in the Cold Test Facility (CTF) at CEA Saclay in nominal conditions at 5 K and 25.7 kA. One major test performed is the so called “temperature margin test" during which the inlet helium temperature of the winding pack is...
Patrick Decool
(CEA)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
In the frame of the Broader Approach, CEA provides 9 + 1 spare TF coils for the JT-60SA tokamak. Mid 2011, a manufacturing contract was awarded to Alstom, Belfort, now General Electric. The first years were dedicated to the manufacturing process definition, the critical phases qualification through a set of 12 mockups, the manufacturing QA definition and the procurement and commissionning of...
Vicente Queral
(National Fusion Laboratory)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Coil casings and coil frames for stellarators are geometrically complex components at high accuracy. A method of additive manufacturing combined with fibre-reinforced resin casting has been recently experimented [1] for the fabrication of complex coil frames. The method is named 3Dformwork and consists of additive fabrication of a hollow thin shell which is filled with resins or other...
Matthias Schneider
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
The Quench-Detection-System of the fusion experiment Wendelstein 7-X detects quench events within the superconducting magnet system constructed of 50 non-planar and 20 planar coils, 14 current leads and the bus bars. In the event of a quench the QD-System triggers the power supply of the magnetic system to shut down.
The QD-System monitors the superconducting system by 486...
Frank Fullenbach
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
The magnet system of the stellarator fusion device Wendelstein 7-X (W7-X) is composed of three different groups of coil systems. The main magnetic field is created by a superconducting magnet system that is accompanied by two sets of normal conducting coil groups, the Control Coils inside the plasma vessel and the Trim Coils (TC) positioned outside of the cryostat.
The TC system consists of...
Sheng Li
(State Key Laboratory of Electrical Insulation and Power Equipment)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
The quench protection switch (QPS) is very important for ensuring the safety of the PF and TF coils of a superconductive Tokomak. The main function of a QPS is to protect the magnet as the coil quench occurs. Besides, a QPS has to withstand almost all of the coil current of some tens of kA flowing through it for a long time in the normal operation condition. This task is undertaken by the...
Qiaosen Wang
(State Key Laboratory of Electrical Insulation and Power Equipment)
9/6/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Superconducting magnet is one of the most crucial components in a superconducting Tokamak. During the normal operation stage, high current of some tens of kA flows through the magnet with large inductance of ~1H. Therefore, extremely large energy (~0.1-10GJ) is stored in the magnet, which must be dissipated in the case of magnet quench in certain duration before the occurrence of local or even...
Tindaro Cicero
(Fusion for Energy)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The Normal Heat Flux (NHF) First Wall (FW) panels consist of a series of fingers, which represent the elementary plasma facing units and are designed to withstand 15,000 cycles at 2 MW/m22. The fingers are mechanically joined and supported by a back structural element called “supporting beam”. The structure of a finger is made of three different materials, stainless steel for the...
Stefano Banetta
(Fusion for Energy)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
This paper describes the main activities carried out for the conclusion of the EU-DA prequalification process for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the High Heat Flux (HHF) testing of a reduced scale FW prototype (Semi-Prototype (SP)). This component is manufactured by the AREVA Company in France and has a dimension of 221 x...
Rafael Enparantza
(Design)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The Normal Heat Flux (NHF) First Wall (FW) panels are designed to withstand the heat flux from the plasma inside ITER. These components are made of beryllium tiles bonded to a copper alloy and 316L (N) stainless steel heat sink. A NHF FW panel consists of several fingers as elementary plasma facing units. This this paper presents the experimental stress and deformations measured on a...
Sergey Tomilov
(JSC “NIKIET”)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
In the framework of PA realization, specialists from NIKIET and Efremov Institute are developing a design of First Wall (FW) Full Scale Prototype (FSP) in order to demonstrate its manufacturability and qualify critical technological processes. Design of FW FSP is developed based on the FW 14 type A. The semi-prototype has been manufactured in order to verify the FW design. Based upon the...
Maxim Sviridenko
(JSC NIKIET)
9/6/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The JSC NIKIET is responsible for the manufacture of the First Wall (FW) beam, the fingers bodies, the mechanical attachment system and electrical connection system of the FW panel to the shield block (SB) in the framework of Procurement Arrangement 1.6.Р1А.RF.01 dated 14.02.2014.
The Electrical strap (ES) is located on the FW rear surface and used for providing current through the FW to the...
Karel Samec
(Centrum Výzkum ŘEŽ)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The heat flux on plasma-facing components in ITER, and even more so in the projected DEMO reactor will reach values in the order of several Megawatt per square meter. Evacuating this heat in a reliable manner is key to the robustness and safety of operation of any fusion reactor.
The current state-of-the-art for cooling plasma-facing components relies on cooling a high heat-resistant structure...
Phani Domalapally
(Centrum výzkumu Řež s.r.o.)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The heat loads on the First Wall (FW) of the European DEMO are not yet defined, but when extrapolated from ITER, the loads can be quite high. As the DEMO will use Eurofer 97 as the structural material and Pressurized Water Reactor (PWR) conditions at the inlet, i.e. 15.5 MPa and 285 °C, the design of the heat sink gets complicated as the thermal conductivity of the heat sink material is quite...
Pavel Zacha
(Energy Engineering)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The first wall cooling of the fusion power reactor DEMO is an important part of the fusion power plant development. A cooling ability at high heat flux conditions will affect a lifetime period of the first wall modules having a direct impact on the operating costs of the fusion power plant. According to current knowledge, the water cooling provides the largest ability to remove the high heat...
Ladislav Vesely
(Faculty of Mechanical Engineering)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
Based on the requirements of F4E, an experimental device HELCZA (High Energy Load Czech Assembly) was designed for high heat flux cyclic loading of plasma-facing components of the ITER reactor, primarily for testing of the full-size first wall modules and divertor inner vertical targets.
Testing is carried out by a high power electron beam heating, and a deviation of the heat flux density at...
Radek Skoda
(Department of Energy Engineering)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The paper deals with optimal electron beam heat distribution on the HELLCZa experiment calculating the flatness of the distribution of heat input and distribution of surface temperature of various samples. A computer program has been developed for balancing the heat flux in the construction materials of the sample. The first boundary condition for this calculation were primarily functions...
Richard Jilek
(Centrum výzkumu Řež s.r.o.)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
Commissioning phase of high heat flux test facility HELCZA
R. Jíleka,*a,*, J. Prokůpekaa, P. Gavilabb
aCentrum výzkumu Řež s.r.o. (CVR), Hlavní 130, 25068 Husinec-Řež, Czech Republic,
bFusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona, Spain
*Corresponding author: e-mail: Richard.Jilek@cvrez.cz, phone: +420 601 315 137
The high heat...
Andre Kunze
(Institute for Neutron Physics and Reactor Technology)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
For the testing of helium cooled plasma facing components in HELOKA-HP homogeneous surface heat flux densities of up to 500 kW/m² have to be reproduced. It has been proposed to use infrared radiation heaters which consist of several quartz glass (fused silica) tubes with tungsten filaments inside to generate the heat flux. This paper presents a numerical model of the latest type of heater...
Muyuan Li
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
Maintenance of structural integrity under high-heat-flux (HHF) fatigue loads is a critical concern for assuring the reliable HHF performance of a plasma-facing divertor target component. Loss of structural integrity may lead to structural as well as functional failure of the component.
Currently, a full tungsten divertor was chosen by ITER Organization, and plenty of HHF qualification tests...
Sergey Pestchanyi
(INR)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
Transient heat fluxes onto the tungsten divertor targets during disruptions in ITER may cause severe melting, leading to intolerable damage. However, for sufficiently energetic transients, tungsten vaporized from the target in the initial stage of the heat pulse will generate a protective plasma shield in front of the target, greatly reducing the incoming heat flux. This vapour shielding is a...
Eunnam Bang
(KSTAR research center)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
This paper deals with the first commissioning of active cooling system for plasma-facing components (PFCs) and coolant removal system. During 2015 KSTAR campaign, we have achieved a 55 sec long pulse H-mode. However, some plasma shots were terminated, not because of instabilities or limitation of heating power, but because of safety limit applied to the PFC temperature: upper boundary to lock...
Jaehyun Song
(Advanced Engineering Division)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The tungsten (W) brazed flat type mock-up with swirl tube which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade. The mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 8 MW/m22 for 20 sec duration at KoHLT-EB in KAERI. In this paper, for comparison of...
Sungjin Kwon
(DEMO Technology Division)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The preliminary conceptual design study on the Korean fusion demonstration reactor (K-DEMO) tokamak consists of the vacuum vessel, the in-vessel components, and the superconducting magnet system, and so on [1]. The K-DEMO superconducting magnet system contains 16 toroidal field (TF) coils, 8 central solenoid (CS) coils and 12 poloidal field (PF) coils. The magnetic field at the plasma center...
JongSung Park
(Fusion Engineering Center)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
A preliminary study on the rigorous 2-step (R2S) based shutdown dose rate calculations has been performed for the Korean fusion demonstration reactor (K-DEMO) in the vicinity of an equatorial port area using the coupled transport and activation calculation codes of MCNP6 and FISPACT. For the shutdown dose rate calculation, the equatorial port structures and port plug including shielding blocks...
Kihak Im
(DEMO Technology Division)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
A pre-conceptual design study for the Korean fusion demonstration tokamak reactor (K-DEMO) has been initiated in 2012. K-DEMO is characterized by the uniqueness of high magnetic field (BT0 = 7.4 T), major and minor radii of 6.8 m and 2.1 m, and steady-state operation.
The heat load distribution by plasma radiation onto the first walls of the in-vessel components is one of the basic inputs for...
G Douglas Loesser
(Engineering)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
G. Douglas Loesser1,Joris Fellinger22, Hutch Neilson11, John Mitchell11, Marc Sibilia11, Han Zhang11, P. Titus11, Irving Zatz1,1,, Arnie Lumsdaine33, Dean McGinnis33
1Princeton Plasma Physics Laboratory, James Forestall Campus, Princeton, NJ 08542, USA
2Max-Planck-Institut für Plasmaphysik,...
Joris Fellinger
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The modular stellarator Wendelstein 7-X in Greifswald (Germany) successfully started operation in 2015 with short pulse limiter plasmas. In 2017, the next operation phase (OP) OP1.2 will start once 10 uncooled test divertor units (TDU) with graphite armor will be installed. The TDUs allow for plasma pulses of 10 s with 8 MW heating. OP2, allowing for steady state operation, is planned for 2020...
Zhongwei Wang
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The cryopump will be installed for the high power and long pulse operation up to 30 minutes of Wendelstein 7-X (W7-X). The cryopump system plays a critical role for capturing ash particles from the plasma, including hydrogen, deuterium and even helium. In total there are 10 independent cryopumps, one cryopump for each of the 10 discrete divertor units. The cryopump is located along the pumping...
Patrick Junghanns
(Max-Planck-Institut für Plasmaphysik)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The 890 target elements of the high heat flux divertor of Wendelstein 7-X (W7-X) are made of a CuCrZr copper alloy heat sink armored with carbon reinforced carbon (CFC) NB31 tiles. Connectors with an internal diameter of 10 mm are electron beam welded to heat sink for the water inlet and outlet. They are produced by electron beam welding thicker tubes of CuCrZr and stainless steel with a...
Jean Boscary
(Max-Planck_Institut für Plasmaphysik)
9/6/16, 2:20 PM
F. Plasma Facing Components
Poster
The actively water-cooled target elements of the high heat flux divertor of Wendelstein 7-X (W7-X) are designed to remove a stationary heat flux of 10 MW/m² on its main area and 5 MW/m² at the end adjacent to the pumping gap. A target element is made of a CuCrZr copper alloy heat sink armored with carbon reinforced carbon (CFC) NB31 tiles. The realization of the divertor requires the...
Clara Colomer
(IDOM Nuclear Services)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The ITER Vacuum Vessel (VV) is a double wall Stainless Steel structure that surrounds the plasma. It constitutes a major safety barrier for ITER, and, because of its function, is classified as Protection Important Component (PIC). Its design and construction has to follow the RCC-MR design code rules to verify the structural integrity under electromagnetic, thermal and seismic...
49864.
P2.127 Electromagnetic Analysis for the In-Vessel Transfer Lines of Neutron Activation System
Sunil Pak
(National Fusion Research Institute)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
In ITER the neutron activation system deploys several foil samples close to the plasma to measure the neutron fluence and the fusion power. These samples are transferred in a pneumatic way along the tubes installed on the vacuum vessel wall. Therefore, the tubes, namely transfer lines, get eddy current induced during plasma disruption, leading to Lorentz force by interacting the background...
Josu Eguia
(Mechanical Engineering)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The vacuum vessel of ITER is a paradigmatic example of a gargantuan system that can only be processed in-situ and from the inside. Its assembly implies performing post welding repair operations, including machining of welding seams following the internal surface of the vacuum vessel. The requirements for the machining operations are the following: accuracy +/- 0.1 mm; dynamic machining forces...
Anna Encheva
(Tokamak Department)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just...
Ivan Poddubnyi
(JSC "NIKIET")
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
In ITER blanket system, electrical connectors (“E–straps”, ES) are used to form a low impedance electrical path from shield blocks (SB) to the vacuum vessel (VV). Main functions of ES is providing current from SB to VV. ES shall withstand electromagnetic (EM) loads and Joule heating resulted from electrical current with magnitude up to 137 kA during 300 ms, accommodate cyclic relative...
Dieter Leichtle
(Fusion for Energy)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The Ion Cyclotron Heating and Current Drive system (ICH) is designed to launch RF power into the ITER plasma, and will reside in equatorial ports (EP) 13 and 15. Shutdown dose rates (SDDR) within the ICH port interspace are required to be ALARA and less than 100 μSv/h at 1066 seconds cooling, in locations where hands-on maintenance is required. The shielding performance of...
Ivan Popov
(Mechanics and Control)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
In this paper the stress-strain state of the diagnostic shield modules (DSM) and the supporting frames (ISS, PCSS), located in the upper ports #2 and #8 of the tokamak ITER is investigated.
DSM is the upper port components and has two main functions: neutron radiation protection and maintenance of rigid fixation diagnostics placed in the port. DSM is operated at high temperatures, significant...
Pivkov Andrew
(Mechanics and Control)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The primary systems of future international thermonuclear experimental reactor (ITER) have to withstand major thermal, nuclear, electromagnetic and seismic loads. Therefor engineering analysis of elements of construction plays crucial role in realizing of the project as a whole. The paper describes calculations of spatial stress-strain state from major loads arising during operation upper...
Yu-Gyeong Kim
(National Fusion Research Institute)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Korea has been manufacturing two vacuum vessels of ITER and main jointing method to in-wall shield assemblies is welding. Though in-wall shield ribs holding neutron shielding blocks should sustain various design loads such as electro-magnetic forces, earthquake and their own weights, as a part of the assembly, in-service inspections are hardly possible because they are installed between...
Dong Won Lee
(Nuclear Fusion Engineering Development Department)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
After the Conceptual Design Review (CDR), Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) design is being updated for the preparation of the preliminary design phase. The manufacturability is considered based on the TBM-set model of CD phase. The overall geometry of the first wall, side wall and the breeding zone was changed slightly. Thethermal-hydraulic and mechanical...
Aleksandr Nemov
(Peter the Great Saint-Petersburg Polytechnic University)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The High Field Side Reflectometry is diagnostic equipment subjected to the conditions that are severe even for ITER: magnetic field over 9T, temperatures up to 700 ºC, strongly non-uniform temperature field, specific shape of the equipment with length of in-vessel waveguides about 10m and location of waveguides close to the blanket connectors where large halo currents are expected during...
Ivan Kirienko
(Mechanics and controls)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The presentation is focused on the simulation results and approaches used for loading analyses made for DTS in-vessel equipment, including spatial stress strain state, seismic analysis, electromagnetic analysis as well as the most important load combinations.
Finite element model of the construction was updated according with updated DTS components design and separated on the following...
Jiaming Jiang
(Fusion Center for Scientist)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
HL-2M RMP (Resonance Magnetic Perturbation) Coils is designed to provide a resonant perturbation magnetic field for high beta plasma operation scenarios stability control, such as Edge Localized Modes (ELMs) suppression control, Resistance Wall Model (RWM) fast control and Error magnetic field correction control, etc.
Especially, ELMs result in impulsive burst of energy deposition on to the...
Bostjan Koncar
(Jožef Stefan Institute)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Thermal radiation analysis of the DEMO tokamak based on the updated CAD design of in-vessel components and magnet system has been carried out. For the purpose of the analysis, Vacuum Vessel Thermal Shield (VVTS), Cryostat Thermal Shield (CTS) and some support structures have been created additionally (on a conceptual level) to complement the overall DEMO CAD design model. The Finite Element...
Paolo Frosi
(Fusion)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
This study is a part of the structural activity being conducted in the framework of the structural design of a DEMO Divertor. The thermal and structural analysis has already been started since a year and the first results has been partly published in a previous paper. The Cassette Body is being analyzed considering the most critical types of loads (e.g. coolant pressure, volumetric neutron...
Juan-Pablo Catalan
(Energy Engineering Department)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Shutdown dose rate (SDR) analysis plays a key role in the design of fusion facilities like ITER and DEMO. One of most used methodology to carry out SDR calculations is the rigorous-two-step (R2S) based on the coupling of transport and activation calculations. Currently, one of the most relevant lacks of this method is the possibility to propagate the effect of the uncertainties accumulated...
Heejin Shim
(Blanket Technology Team)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Molybdenum disulfide (MoS2) coating was deposited by magnetron sputtering onto the target material. The coatings of deposited MoS2 can be used in high vacuum and aerospace environments for lubrications purposes, which ultra-low friction is desirable. For these reason, the sputtered MoS2 coating method is primarily considered for ITER components and their mechanical assemblies. A common...
Piero Agostinetti
(Consorzio RFX (CNR)
9/6/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
A new technique, called Vacuum Tight Threaded Junction (VTTJ), has been developed and patented by Consorzio RFX, permitting to obtain low-cost and reliable non welded junctions, able to maintain vacuum tightness also in aggressive environments. The technique can be applied also if the materials to be joint are not weldable and for heterogeneous junctions (for example, between steel and copper)...
Pietro Alessandro Di Maio
(Energia)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Within the framework of EUROfusion R&D activities an intense research campaign has been carried out at the University of Palermo, in close cooperation with ENEA Brasimone, in order to investigate the thermo-mechanical performances of the DEMO Water-Cooled Lithium Lead breeding blanket (WCLL). In particular, attention has been paid to the most recent geometric configuration of the DEMO WCLL...
Antonio Froio
(NEMO Group)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The development of a system-level thermal-hydraulic model of the whole EU DEMO tokamak has been launched by the EUROfusion Project Management Unit. In order to follow the progress in the design of the tokamak components, the model should be developed in an object-oriented fashion, to ensure a high modularity. Within this framework, the first block of the model is under development at...
Nicola Forgione
(DICI)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The Breeding Blanket is a key component in a fusion power plant in charge of ensuring tritium breeding, neutron shielding and energy extraction. Water-Cooled Lithium-Lead Breeding Blanket (WCLL) is considered a candidate option in view of the risk mitigation strategy for the realization of DEMO. Indeed, this design might benefit of efficient cooling performances of water as coolant, as well as...
Emanuela Martelli
(DIAEE)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Within the framework of EUROfusion Power Plant Physics & Technology Work Programme, the Water Cooled Lithium Lead (WCLL) is one of the four breeding blanket (BB) concepts considered as possible candidate for the realization of DEMO fusion power plant. ENEA CR Brasimone has developed during 2015 a new design of the outboard module based on horizontal (i.e radial-toroidal) water cooling tubes in...
49885.
P2.150 CFD simulation of the magnetohydrodynamic flow inside the WCLL breeding blanket module
Alessandro Tassone
(Dipartimento di Ingegneria Astronautica)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The interaction between the molten metal and the plasma-containing magnetic field in the breeding blanket of a Tokamak fusion reactor causes the onset of a magnetohydrodynamic (MHD) flow. In order to properly design the blanket, it is important to quantify how and how much the flow features are modified compared with an ordinary hydrodynamic flow. This paper aims to characterize the evolution...
Leo Buhler
(Institute for Nuclear and Energy Technologies)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
A number of liquid metal blanket designs for applications in nuclear fusion reactors is currently under development. In the water cooled lead lithium (WCLL) blanket Eurofer97 is used as structural material and liquid PbLi as breeder, neutron multiplier, and as heat transfer medium. The released heat is removed by water at a pressure of 155 bar (pressurized water reactor conditions, 285°C -...
49887.
P2.152 Structural analysis of the back supporting structure of the DEMO WCLL outboard blanket
Maria Lorena Richiusa
(Department of Energy)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Within the framework of EUROfusion R&D activities an intense research campaign has been carried out at the University of Palermo, in close cooperation with ENEA Brasimone, in order to investigate the thermo-mechanical performances of the Back-Supporting Structure (BSS) outboard segment of the DEMO Water-Cooled Lithium Lead breeding blanket (WCLL). In particular, the configuration of the BSS...
Marica Eboli
(DICI)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The interaction between lithium-lead and water is a major concern of Water Coolant Lithium Lead (WCLL) breeding blanket design concept, therefore deterministic safety analysis of the in-box LOCA postulated accident is of primary importance. In this framework, a past experimental campaign was carried out in LIFUS5 to investigate the evolution and the consequences of the interaction. Then, these...
Songlin Liu
(Institute of Plasma Physics Chinese Academy of Sciences)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokomak reactor. Its major radius is 5.7m, minor radius is 1.6m and elongation ratio is 1.8. It is possible upgrade to R~6 m, a~2 m. CFETR mission and objectives are to bridge gaps between ITER and DEMO, and to realize fusion energy application in China. CFETR has two phases. Phase I is to demonstrate full cycle of...
Hui Bao
(School of Nuclear Science and Technology)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Square channel is widely used in the conceptual design of water cooled blanket of fusion reactor for cooling and providing appropriate inner temperature field for tritium breeding. Thermal hydraulic design of blanket directly determines the heat transfer efficiency and safety characteristics of fusion reactor. Therefore, thermal-hydraulic characteristics of square channel should be...
Kecheng Jiang
(Institute of Plasma Physics)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). From the security point of view, the thermal-hydraulic analysis is very essential because the blanket should remove the high heat flux radiated from the plasma and the volumetric heat generated by neutron wall loading. For the normal state of plasma burning, the jumped peak...
Xiaokang Zhang
(Institute of Plasma Physics Chinese Academy of Sciences)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The water-cooled ceramic breeder (WCCB) blanket is one of the candidates of Chinese fusion engineering test reactor (CFETR). WCCB blanket will produce radioactive waste during its operation and decommissioning processes. The radioactive characteristics of WCCB blanket, including solid structure and functional material and the liquid water coolant, are of importance for the replacement and...
Pinghui Zhao
(School of Nuclear Science and Technology)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
A conceptual structural design of Water-Cooled-Solid-Breeder (WCSB) blanket, one of the breeding blanket candidates for China Fusion Engineering Test Reactor (CFETR), is now being carried on by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP). To validate the reliability of the designed blanket module, detailed thermal-hydraulic analysis is necessary. The computational fluid...
Geon-Woo Kim
(Nuclear Engineering)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Tokamak reactors like ITER or fusion DEMO reactors have serious concerns about material damages to plasma facing components (PFC) due to plasma instabilities. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. In addition, high thermal stresses due to rapid...
Angel Ibarra
(CIEMAT, Avda. Complutense 40, 28040 Madrid, Spain)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
In the framework of the EUROfusion programme, Dual Coolant Lithium Lead (DCLL) breeding blanket is being investigated as a candidate for European DEMO, which is based on the use of Pb-17Li as breeder and coolant (“self-cooled breeding zone”) and high-pressure helium for cooling the structures made of reduced-activation ferritic steel (EUROFER). During the first part of the project, a...
Luis Maqueda
(Esteyco Mechanics)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
General purpose finite element (FE) softwares can be readily used for the stationary analysis of breeding blankets of a nuclear fusion reactor. However, the analysis of transient effects generated during the pulsed operation mode requires transient simulations to be carried out. Nowadays, a commercial tool which can be directly used for these transient simulations with affordable computational...
Fernando Roca Urgorri
(National Fusion Laboratory)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The Dual Cooled Lithium Lead (DCLL) blanket is one of the four breeder blanket technologies under consideration within the framework of EUROfusion Consortium activities. The aim of this work is to develop a preliminary model that can track the tritium concentration along each part of the DCLL blanket and their ancillary systems at any time.
Because of tritium’s nature, the phenomena of...
Ivan Fernandez-Berceruelo
(Fusion National Laboratory)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The Dual Coolant Lead-Lithium (DCLL) is one of the breeding blanket concepts under investigation in EUROFusion. This concept is characterized by the use of self-cooled eutectic PbLi as neutron multiplier and tritium breeder and carrier, whereas supercritical helium is used to cool the first wall and other parts of the structure.
The thermal-hydraulic (TH) design of the breeding blanket, as the...
Daniel Suarez
(Department of Physics)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The conceptual design of the European Dual Coolant Lead Lithium (DCLL) breeding blanket is currently being developed in the frame of EUROfusion Project. To this aim, it is of utmost interest to estimate critical flow parameters such as: (1) pressure drop and heat transfer coefficient at both helium and lithium sides, and (2) tritium permeation ratio. Pressure drop in purely hydrodynamic flows...
Qingyun He
(School of Nuclear Science and Technology)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Liquid metal (LM) blanket concepts are designed by many countries due to its attractive features such as geometric adaptability, good thermal conductivity and heat carrying capacity, et al. However, they all have feasibility issues associated with magnetohydrodynamic (MHD) interactions under the environment of a strong control magnetic field and the flowing high electrical conductivity LM. The...
Fumito Okino
(Institute of Advanced Energy)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
DCLL blanket has high energy recovery efficiency. Nevertheless by several technical issues, such as MHD pressure drop, tritium permeation and energy conversion membrane corrosion, technical readiness level(TRL) of DCLL is relatively not high. To breakthrough this situation, the authors propose a new method to recover tritium and heat from liquid lithium-lead (PbLi) droplet by non-contact in...
Andrei Khodak
(Princeton Plasma Physics Laboratory)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The analysis of Dual-Coolant Lead–Lithium (DCLL) blankets requires application of Computational Fluid Dynamics (CFD) methods for electrically conductive liquids in geometrically complex regions and in the presence of a strong magnetic field. Several general-purpose CFD codes allow modeling of the flow in complex geometric regions, with simultaneous conjugated heat transfer analysis in liquid...
Brijesh Kumar Yadav
(Institute for Plasma Research)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Indian Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) will be tested in one half of the port no#02 of ITER. In LLCB TBM, PbLi eutectic alloy is used as multiplier, breeder, and coolant for the CB zones, and Li2TiO3 ceramic breeder (CB) is used as a tritium breeding material. The LLCB TBM consists of two helium coolant circuits, one for the TBM outer box i.e. the TBM First...
Denis Obukhov
(JSC "NIIEFA" (Efremov Institute))
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
This paper gives an overview of the new facility for MHD and heat transfer (HT) tests of liquid metal breeder blanket mock-ups in high magnetic field. The facility named LIMITEF5 (LIquid Metal TЕst Facility, 5 T) is under construction now in JSC “NIIEFA” (D.V. Efremov Institute).
The facility includes the Lead-Lithium (LL) loop passing through the warm aperture of the superconducting...
Takuya Goto
(National Institute for Fusion Science)
9/6/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Lithium molten salts (e.g., Flibe, Flinabe) have several merits as a self-cooled tritium breeding material: low reactivity, low density and low electric conductivity. On the other hand, molten salts may cause a problem of tritium migration to the structural material of the blanket due to the low hydrogen solubility. To overcome this problem, an active control of the effective hydrogen...
Igor Kupriyanov
(Beryllium Department)
9/6/16, 2:20 PM
I. Materials Technology
Poster
The primary reasons for the selection of beryllium as an armour material for the ITER first wall are its low Z and high gettering characteristics. For this application three beryllium grades: S-65C (USA), TGP-56FW (Russia) and CN-G01 (China) have been accepted. This selection was based on the results of the ITER Qualification Program, which included characterization and testing of material...
Petra Jenus
(Department for Nanostructured Materials)
9/6/16, 2:20 PM
I. Materials Technology
Poster
Tungsten-based composites have gained considerable attention owing to their excellent performance levels at high temperatures due to exceptional high temperature properties such as a high melting point, good thermal conductivity and a low thermal expansion coefficient. However, tungsten is also associated with a serious reduction in its strength at elevated temperatures, which is also one of...
Sasa Novak
(Department for Nanostructured Materials)
9/6/16, 2:20 PM
I. Materials Technology
Poster
The main aim of the work has been to improve properties of the plasma-facing material for the divertor to resist high thermal loading during operation. Among the available materials we selected (carbide) particles reinforcement of tungsten, wherein the reinforcement should not chemically react with the matrix. In this respect, W2C particles offer the most attractive solution.
The paper will...
Andrei Galatanu
(National Institute of Materials Physics)
9/6/16, 2:20 PM
I. Materials Technology
Poster
W has the highest melting point of all metals, good high temperature strength, high creep resistance and a high thermal conductivity. These properties make W a first choice for armor materials in fusion energy reactors. Unfortunately W can not be also used for structural applications, due especially to its high temperature brittle- to-ductile transition (DBT). However, when cold rolled at...
Carmen Garcia-Rosales
(Materials Department)
9/6/16, 2:20 PM
I. Materials Technology
Poster
Tungsten is presently the main candidate material for the first wall armour of future fusion reactors. However, if a loss of coolant accident with simultaneous air ingress into the vacuum vessel occurs, the temperature of the in-vessel components would exceed 1000ºC, leading to the undesirable formation of volatile and radioactive tungsten oxides. A way to prevent this serious safety issue is...
Min Pan
(Key Laboratory of Advanced Technology of Materials (Ministry of Education))
9/6/16, 2:20 PM
I. Materials Technology
Poster
Irradiation damage research is one of the basic issues to solve the application of first-wall materials in fusion engineering. The diffusion and recovery of the defects can greatly affect the performance of the materials in fusion. The rotation, stability, migration of the self-interstitial atoms (SIAs) in defect structures of tungsten is investigated by the first-principle method. It is found...
Vladica Nikolic
(Erich Schmid Institute of Materials Science of the Austrian Academy of Sciences)
9/6/16, 2:20 PM
I. Materials Technology
Poster
In order to investigate possible enhancement of mechanical properties of tungsten (W) based materials by solid solutions and to examine the influence of a single alloying element on a particular property such as ductility, a versatile production method of generating a wide range of different tungsten binary alloys is presented. Magnetron sputter co – deposition was used to produce thin films...
Magdalena Galatanu
(National Institute of Materials Physics)
9/6/16, 2:20 PM
I. Materials Technology
Poster
For DEMO fusion reactor an expected heat flux of about 10 MW/m22 should be extracted by the divertor which will have, most likely, an armour part made of W and a following heat sink part made of Cu or ODS Cu alloy. Unfortunately, for these materials the optimum operating temperature windows do not overlap. Thermal barrier materials are interface materials included in such...
Gabriel Carro
(Physics)
9/6/16, 2:20 PM
I. Materials Technology
Poster
Copper-based materials are considered the most promising candidates for water-cooled components of the heat sink systems of future fusion reactors. Although pure copper is the material with the higher thermal conductivity, the detriment of its mechanical strength on increasing temperature restricts its use at high temperature.
In the last years, ODS Cu-Y2O3 and Cu-Y alloys have been produced...
Dai Hamaguchi
(Fusion Research and Development Directorate)
9/6/16, 2:20 PM
I. Materials Technology
Poster
Copper is the candidate material for cooling components for divertor and other plasma facing components. Although CuCrZr alloy is a first choice regarding strength, toughness, and conductivities, issues related to quality control during manufacturing process and also on the possible loss of strength during brazing among fabrication of the components still remains. CuCrZr also exhibit some...
Hiroyuki Noto
(National Institute for Fusion Science)
9/6/16, 2:20 PM
I. Materials Technology
Poster
Copper (Cu) alloy is a candidate materials for use as heat sink materials of fusion divertor because of its good thermal conductivity. In recent years a number of studies have been carried out on Cu-based materials such as Precipitation Strengthened Cu (PS-Cu).However, the material has some critical issues such as instability of microstructure at high temperature and loss of strength by...
Inigo Iturriza
(Materials and Manufacturing)
9/6/16, 2:20 PM
I. Materials Technology
Poster
The blanket is one of the most critical component of ITER. It is directly exposed to the plasma and acts as shielding of the vacuum vessel from the neutrons and other energetic particles produced in the fusion plasma. Each of the 215 Normal Heat Flux (NHF) panels consists of a shield block and a First Wall (FW) panel. The NHF FW panels consist of a complex bimetallic structure of 316L...
Teteny Baross
(WIGNER RCP)
9/6/16, 2:20 PM
I. Materials Technology
Poster
The actuality of the topic comes from the ITER (International Thermonuclear Experimental Reactor) fusion tokamak that is a major international experiment with the aim of demonstrating the scientific and technical feasibility of fusion as an energy source. Among others the most challenging task is to find proper materials and technology for Plasma Facing Components.
Welding by HIP (Hot...
Pinghuai Wang
(Southwestern Institute of Physics)
9/6/16, 2:20 PM
I. Materials Technology
Poster
The CuCrZr/316L(N) explosion bonding bimetallic plates were used to make hypervapotron (HVT) cooling channel for the fingers, which is the key components of the ITER First Wall (FW). The bimetallic plates will be subjected to the same thermal cycles as the FW component, including the HIP (hot iso-static pressing) joining for bonding HVT and beryllium tiles, thus the properties of both the...
Javier de Prado
(Materials Science and Engineering Area)
9/6/16, 2:20 PM
I. Materials Technology
Poster
Development of new materials is one of the key for the construction of the new fusion power plant (DEMO). The selected materials have to fulfill several requirements such as standing the conditions that takes place in the core (high neutron flux and temperatures close to 1200 ºC) and low activation rate.
Several techniques have been proposed to join the different parts of the first wall...
Eduard Feldbach
(Institute of Physics)
9/6/16, 2:20 PM
I. Materials Technology
Poster
Radiation tolerant optical components of future fusion reactors have to withstand radiation of unprecedented intensity. It is widely recognized that spinel lattice of AB2O4 double oxides demonstrates enhanced resistance against neutron irradiation. Therefore, the development of spinel optical materials and understanding of their radiation damage processes is of great importance. One defect...
Jiao Peng
(Institute of Plasma Physics)
9/6/16, 2:20 PM
I. Materials Technology
Poster
First mirror (FM) lifetime is one of critical issues for the optical diagnostic system in ITER since it greatly influences the performance of relative diagnostic. In ITER, repetitive cleaning is expected to give a positive solution to the frequent replacement of FM, thus prolonging its lifetime. Three cleaning cycles using radio frequency argon plasma were applied to the stainless steel mirror...
Richard Kembleton
(Culham Centre for Fusion Energy)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
There are a number of key design difficulties in producing an integrated demonstration fusion power plant (DEMO) design, and how these issues are resolved fundamentally affects the final overall design. Technological examples include the issue of power loading in the divertor and reducing recirculating power through efficient current drive. Additional drivers include economic considerations...
Dagui Wang
(Institute of Nuclear Energy Safety)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Breeding blanket research and development is recognized as one of the most important areas for realizing an energy-producing fusion reactor. In China, the ceramic breeder/helium coolant/ferritic steel structure is considered as the main concepts of the blanket to conduct the breeding blanket research, and on the other hand, the liquid breeder blanket is also to be investigated as the...
James Morris
(Power Plant Technology Group)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The investigation of time-dependent power requirements for a future nuclear fusion reactor is part of the systems integration task for the European Fusion Programme. All fusion power plants, whether pulsed or steady-state, will require electrical power to operate the various plant systems. Over the entire pulse cycle reactor systems will require varying levels of power over different time...
Christopher Harrington
(Culham Centre for Fusion Energy (CCFE))
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The Water-cooled Lithium-Lead (WCLL) blanket is one option under consideration for the EUROfusion DEMO programme. This blanket design must interface with the Primary Heat Transfer System, Power Conversion System, and Energy Storage System in an integrated solution to mitigate the pulsed power profile of the tokamak and deliver feasible power plant performance. The system must maintain an...
Monika Lewandowska
(Institute of Physics)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
ITER is planned to be the research type tokamak which will achieve the energy breakeven point. The next step towards the realization of fusion energy will be DEMO – the first demonstration fusion power plant producing grid electricity at the level of a few hundred MW. DEMO designers are required to maximize the conversion efficiency of the primary and secondary plant circuits. The Primary Heat...
Vaclav Dostal
(Energy engineering)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The cooling system is one of the key parts of the fusion power reactor technology. The DEMO fusion power reactor should have different heat sources (first wall, blanket, and divertor) with different temperature and power. In the current European concept of DEMO, helium and water are used as the cooling medium. However, use of Helium and water introduces some issues in terms of their properties...
Xue Zhou Jin
(Institute of Neutron Physics and Reactor Technology (INR))
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
HCPB (helium cooled pebble bed) blanket concept is one of the EU DEMO blanket concepts running for the final design selection. It is necessary to study the pressure behaviour in the blanket and the connected systems during the loss of coolant (LOCA) in a blanket module, as well as the temperature evolution in the coolant flow and the associated structures. The LOCA can be caused by...
Danilo Dongiovanni
(FSN)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Radioactive toxins confinement is a main safety function for nuclear power plants, hence the importance of confinement design parameters optimization. In this context, performing parametric assessments of thermodynamic variables thought to be relevant for confinement design can help at better framing the option design space. In the context of DEMO EUROfusion WP, FFMEA studies are going on for...
Jan Stepanek
(Department of Energy Engineering)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The first wall, blanket and divertor targets provide a physical boundary for the plasma influence and have to be intensively cooled during the operation in case of the high power fusion reactor. In the case of the LOCA accident, the released fusion power can be stopped very quickly, but the final plasma disruption may load the non-cooled components, and a large amount of heat accumulated in...
Dobromir Panayotov
(ITER Department, Fusion for Energy (F4E), Torres Diagonal Litoral B3, Barcelona, E-08019, Spain)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER. Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. The F4E, Amec Foster Wheeler and INL comprehensive...
Danna Zhou
(Institute of Nuclear Energy Safety Technology)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The helium cooled LiPb blanket concept has become a promising design for fusion reactors in the world. Considering the complex design of the blanket, it is likely that helium gas leakage into the liquid alloy may occur due to tube rupture, named in-box Loss of Coolant Accident (in-box LOCA). And corresponding shock waves likely occurred at the break position and transferred within the liquid...
Jiangtao Jia
(Key Laboratory of Neutronics and Radiation Safety)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
With China signing Test Blanket Module Arrangement (TBMA) with ITER Organization for Helium Cooled Ceramic Breeder (HCCB) Test Blanket System (TBS) in February 2014, Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS), becomes one of the leading teams undertaking its corresponding research and development, and is mainly responsible for structure material...
Marco Fabbri
(Fusion Energy Engineering Laboratory)
9/6/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
For almost ten years now, several safety studies of plasma-wall transients have been performed with AINA code for ITER, the European DEMO design (e.g. HCPB) and Japanese one (e.g. Water Cooled Pebbled Bed or WCPB) to establish an envelope for the worst effects of ex-vessel LOCA and overfuelling. For this purpose, for each blanket type a specific wall-model has been developed for different AINA...
Agnieszka Zaras-Szydłowska
(Institute of Plasma Physics and Laser Microfusion)
9/6/16, 2:20 PM
K. Laser and Accelerator Technologies
Poster
A concept and a laboratory model of the laser-driven accelerator of plasma beams for materials research is presented. The accelerator is based on the laser-induced cavity pressure acceleration (LICPA) scheme [1] and includes four parts: (1) the laser driver, (2) the plasma cavity where high-temperature plasma is created by the laser driver and a high plasma pressure is generated, (3) the...
Punit Kumar
(Department of Physics)
9/6/16, 2:20 PM
K. Laser and Accelerator Technologies
Poster
Interaction of high power laser fields with plasma is important for many applications including laser fusion, laser wakefield acceleration and x-ray lasers. At high laser intensities, nonlinear interactions between plasma and laser becomes significant. In the last ten years, there has been a great deal of interest on plasma systems where the quantum effects are important. Consideration of...
Koichi Nishiyama
(IFMIF/EVEDA Project Team)
9/6/16, 2:20 PM
K. Laser and Accelerator Technologies
Poster
IFMIF (International Fusion Material Irradiation Facility) will generate 14 MeV neutron flux for qualification and characterization of suitable structural materials of plasma exposed equipment of fusion power plants. IFMIF is an indispensable facility in the fusion roadmaps since provide neutrons with the similar characteristics as those generated in the DT fusion reactions of next steps after...
Sunao Maebara
(Rokkasho Fusion Research Institute)
9/6/16, 2:20 PM
K. Laser and Accelerator Technologies
Poster
For the IFMIF/EVEDA accelerator prototype RFQ linac, the operation frequency of 175MHz was selected to accelerate a large current of 125mA. The driving RF power of 1.28MW by 8 RF input couplers has to be injected into the RFQ cavity for CW operation mode. For each RF input coupler, nominal RF power of 160kW and maximum transmitted RF power of 200kW are required.
For this purpose, an RF input...
