Kwang-Pyo Kim
(National Fusion Research Institute)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
To achieve the high performance plasma in the Korea Superconducing Tokamak Advanced Research (KSTAR) tokamak, Neutral Beam Injection (NBI) system has been installed and upgraded. The first NBI (NBI-1) was installed in 2010, which provides a 100 keV deuterium neutral beam of 6 MW maximum using three ion sources. The second NBI (NBI-2) with another 6 MW will complete to be constructed by 2018....
Young-Ju Lee
(Vacuum & cryogenic engineering team)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
KSTAR project has required the new helium distribution box named upgraded distribution box (DBU) for the operation of the cryogenic components such as in-vessel cryo-pump (CPI), super-sonic molecular beam injector (SMBI), and hydrogen pellet injection system (PIS). Two CPIs are inserted into the tokamak vacuum vessel and these components shall be operated at 90 K for the liquid nitrogen...
Wolfgang Biel
(Institute for Energy and Climate Research IEK-4 (Plasma Physics))
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
In the European strategy towards fusion electricity, a demonstration tokamak fusion reactor (DEMO) is foreseen as the single step between ITER and a fusion power plant. Recent studies have been focussing on the concept development for a “conservative” pulsed tokamak reactor with an electrical output power of Pel ~ 500 MW and plasma pulse duration of tpulse ~ 2 hours.
In the design process for...
Dong-Seong Park
(National Fusion Research Institute)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The nuclear fusion research is in progress for the next generation energy source in many countries. The Korea Superconducting Tokamak Advanced Research (KSTAR) in Korea, the Experimental Advanced Superconducting Tokamak (EAST) in China and the Wendelstein7-X in German are the operational superconducting fusion device in the world. The International Thermonuclear Experimental Reactor (ITER) is...
Wook Cho
(Heating and Current Drive Research Team)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
In 2015 KSTAR Campaign, the maximum injection power of the KSTAR tangential Neutral Beam Injector (KSTAR NBI-1) is 5.39MW with three ion sources. Issues in beam extraction found during the experiment were 1) a large oscillation of beam current, 2) frequent interrupts in beam extraction due to breakdown in grids, and 3) a distortion of waveform. To solve these issues, we focused on the unstable...
Soo-Hwan Park
(Advanced Technology Research Center)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
KSTAR (Korea Superconducting Tokamak Advanced Research) has used gas puffing system as main fueling method since 2008. Up to date total fueling efficiency of gas puff is less than 30 %. Pellet injection is more effective technique to control plasma density than gas puffing system and supersonic molecular beam injection. Many fusion devices such as JET, Tore Supra, ASDEX-U, HL-2A, EAST, and LHD...
Michela De Muri
(Consorzio RFX)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The Padova Research on ITER Megavolt Accelerator (PRIMA), under construction at Consorzio RFX, will host SPIDER test bed, a full-size 100 kV negative ion source, and MITICA test bed, a prototype of the whole ITER injector, aiming to develop and optimize the heating injectors to be installed in ITER.
The production of hydrogen (or deuterium) negative ions inside the sources relies mainly on the...
Mauro Pavei
(Consorzio RFX)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The heating neutral beam injectors (HNBs) at ITER are expected to deliver 33 MW of neutral beam power to the ITER plasma for the purposes of heating and current drive. This is achieved by using 2 injectors, each capable of delivering 16.5 MW of neutral beam power.
The beam source of each injector is a complex assembly composed by an RF based negative ion source having an extraction area of...
Samuele Dal Bello
(Consorzio RFX)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The ITER project requires at least two Neutral Beam Injectors, each accelerating to 1MV a 40A beam of negative deuterium ions, to deliver to the plasma a power of about 33 MW for one hour as additional heating.
Since these requirements have never been experimentally met, it was recognized necessary to build-up a test facility, named PRIMA (Padova Research on ITER Megavolt Accelerator), in...
Sergei Sytchevsky
(JSC «NIIEFA»)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Large state-of-the-art fusion devices involve extensive computations throughout the engineering design process from the concept to the commissioning. A variety of well-established software tools, such as ANSYS, OPERA, CARIDDY, TYPHOON, TORNADO has produced a range of simulation techniques and approaches for electro-magnetic (EM) simulations of principal components of tokamaks. The installation...
Boris Lyublin
(JSC "NIIEFA")
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Concrete structures of tokamak buildings are reinforced with steel rebar that produces a substantial contribution into the tokamak field both in the plasma region and in the building where the service staff and magnetically sensitive equipment will be located.
The article describes an advanced approach to modelling magnetic properties of reinforced concrete structures bearing in mind the...
Ilya Gornikel
(Alphysica GmbH)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Cryogenic systems for fusion reactors have to cope with large pulsed heat load generated during fusion experiments. The paper is focused on mitigation of pulsed heat power arriving to the cryoplant from several parallel cooling loops of tokamak superconducting magnets. A new control strategy is proposed. The pressure drop measured at the return cryoline serves as a feedback signal to...
Mahesh Vuppugalla
(Institute for Plasma Research)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Successful operation of a Neutral Beam Injector is dependent on the performance of High voltage power supply system(HVPS) for the production of ion beam. To meet the functional requirements of ion extraction, the power supplies(PS) are designed for fast output cut-off, low energy content during breakdown(BD), ability to withstand repeated BD. It is important that features of the PS are...
Jyoti Shankar Mishra
(Institute for plasma research)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Institute for Plasma Research (IPR), India has a programme of development of allied technologies with applications related to fusion reactor. A pneumatic gas gun kind Single pellet injector system (SPINS-IN) developed at IPR is successfully delivering hydrogen pellets of size 2 mm with a velocity of 700 meters/sec. It is a cryocooler based system operated at a temperature < 10 K and...
Larry Baylor
(Oak Ridge National Laboratory)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The formation and acceleration of cryogenically solidified pellets of hydrogen isotopes has long been under development for fueling fusion plasmas. Fueling with DT pellets injected from the high field side wall has been proposed for future burning plasma tokamak devices. In addition to fueling, smaller shallow penetrating pellets of deuterium injected from the low field side wall have been...
Massimo Zucchetti
(DENERG)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The compact, high field fusion experiment Ignitor aims at the demonstration, for the first time, of ignition in magnetically confined D-T plasmas, together withthe exploration of the physics of the ignition process, and of heating and control of plasmas under controlled burning conditions. The machine parameters have been established on the basis of existing knowledge of the confinement...
Dario Andres Cruz Malagon
(DENERG)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
Nuclear Fusion is a candidate as a long-term energy solution for developed countries. A fusion plasma can be fuelled by different kinds of isotopes. The advantages of Deuterium-Helium-3 (DHe) plasmas of advanced fusion reactors lie in the scarcity of neutrons (due to side DD and DT reactions), and direct conversion of the produced energy without thermal cycle.
The proposed CANDOR DHe plasma...
Sayf Elgriw
(Department of Physics and Engineering Physics)
9/8/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The interaction between resonant magnetic perturbations (RMP) and plasma is an active topic in the fusion energy research. RMP involves the use of radial magnetic fields generated by external coils installed on a tokamak device. The resonant interaction between the plasma and the RMP fields has many favorable effects such as suppression of instabilities and improvement of discharge parameters...
Pietro Vincenzi
(Consorzio RFX)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
EU DEMO studies for pulsed (DEMO1) and steady-state (DEMO2) concepts are currently in the pre-conceptual phase [1]. DEMO1 aims at producing about 2GW of fusion power with a burn time of approximately 2 hours. Within EUROfusion Power Plant Physics and Technology department, DEMO scenario modelling is carried out as part of the validation of feasibility and performance of DEMO designs. One of...
Thomas Franke
(EUROfusion Consortium)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The Heating & Current Drive (H&CD) systems in a DEMOnstration fusion power plant are one of the major energy consumers. Due to its high demand in electrical energy produced in the balance of plant (BoP) the H&CD efficiency optimization is one of the main goals of the DEMO development. The energy consumption of the H&CD sub-systems in different plant modes & states and plasma phases need to be...
Giulio Gambetta
(Consorzio RFX)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
Several novel design solutions for high performance cooling systems have been developed by Consorzio RFX, permitting to experimentally simulate the challenging heat transfer conditions foreseen in the future fusion devices. The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), aims on one hand to verify the present solution applied inside the MITICA experiment...
Alexey Dnestrovskiy
(Plasma Physics)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
Neutral Beam Current Drive (NBCD) is considered as an indispensable mechanism for a steady state regime in such contemporary projects as a tokamak based neutron source or a DEMO type thermonuclear reactor. In this report numerical calculations of NBCD with a Monte Carlo code NUBEAM are complemented by a semianalytical treatment of fast ion velocity distribution function. NBCD parameters were...
Ivan Spassovsky
(Fusion Department)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
F. Mirizzi11, M. Carpanese22, S. Ceccuzzi22, F. Ciocci22, G. Dattoli22, E. Di Palma22, A. Doria22, G.P. Gallerano22, G. Maffia22, A. Petralia22, G.L. Ravera22, E. Sabia33, I. Spassovsky22, A.A. Tuccillo22, S. Turtù22, P....
Silvio Ceccuzzi
(FSN - Fusion Physics Division)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
In the frame of the feasibility study of a Cyclotron Auto-Resonance Maser (CARM), different solutions for the distributed reflectors of the resonant cavity have been considered and compared. In detail, a 250 GHz CARM source is under design with an output power of 200 kW for pulses up to 0.2 s, representing the first milestone of a more ambitious project, aimed at achieving a CW 1 MW mm-wave...
Amro Bader
(Tokamak Scenario Development Department)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The use of efficient heating and current drive systems is an important research priority for DEMO. The Ion Cyclotron Resonance Heating (ICRH) is one such system justified by its inherent advantages, though in its present status (antenna situated in a port in the Vacuum Vessel (VV) is unacceptable for DEMO, where tritium self-sufficiency is to be demonstrated, and reducing the openings in the...
Bongki Jung
(Nuclear Fusion Engineering Development Division)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
A high-power pulsed arc ion source based on Marx generator has been developed at the Korea Atomic Energy Research Institute for the heating NBI system of the VEST which is a compact spherical tokamak at Seoul National University to study the reactor-relevant tokamak operating scenario[1][1]. The NBI system, with a total ion beam power of 0.8MW, was designed for the core plasma...
Sung-Ryul Huh
(Nuclear Fusion Engineering Development Division)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
Within the framework for development of the radio frequency (RF) driven positive ion source as an alternative to the conventional filament arc driven ion source for fusion applications, KAERI is currently constructing a new high power (50 kW at a frequency of 2 MHz) large area RF ion source. The ion source was designed to have a rounded rectangular geometry for covering rectangular ion...
Matteo Vallar
(Consorzio RFX)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The planned upgrade of the RFX-mod device is a good opportunity to widen the operational space of the machine, in both RFP and tokamak configurations. Installation of a power neutral beam injector (NBI) is also envisaged and a NBI system compatible with RFX-mod is already available on site. It was previously installed in TPE-RX (Tsukuba, Japan), it has a nominal power of 1.25 MW, a nominal...
Macarena Liniers
(Laboratorio Nacional de Fusión)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
Neutral Beam injection has some well-established effects on plasma behaviour, such as the power threshold observed in L to H confinement mode transitions or the fast ion excitation of Alfvén modes, whose underlying mechanisms are still under investigation.
In recent TJ-II experimental campaigns emphasis has been made in the characterisation of those Neutral Beam related effects. A study of...
Liu He
(NBI Group)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The condition of 1MW-NBI heating for toroidal experiments to increase plasma energy storage and help making H-mode discharge had been well examined on HL-2A tokomak. A new tokomak with larger size and higher parameters named HL-2M tokomak which is under construction in Southwestern Institute of Physics of China needs higher auxiliary heating power, so a new NBI beamline with maximum 5MW...
Takuya Hase
(Tokai University)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
Production of negative ions plays an essential role in Neutral Beam Injection (NBI). A negative ion beam with an energy of 1 MeV and a current of 40 A (a current density of 20 mA/cm22) is required for 3600 s to produce 16.5 MW of power. NBI predominantly uses negative hydrogen ion sources based on surface production. These negative hydrogen ion sources require cesium seeding to...
Shaofei Geng
(Department of Fusion Science)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
In order to investigated the dynamics of H-- ions and understand the extraction process inside filament-arc-driven plasmas in a Cs-seeded negative ion source, diagnostic experiments using a directional Langmuir probe combined with photodetachment measurement have been conducted. Two-dimensional flow pattern of H-- ions has been obtained as well as the profile of...
Raghuraj Singh
(IC H&CD)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
India is developing 2.5 MW RF source at VSWR 2:1 in the frequency range 35-65 MHz for ITER project. Eight such RF sources will generate total 20MW of RF power for plasma heating and current drive. A large number of high power transmission line components are required for connecting various stages of RF source. To test these passive transmission line components at high power, a 3MW test...
Manojkumar Patel
(ICH&CD)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
ITER-India is developing Ion Cyclotron Heating & Current Drive (ICH&CD) RF source in the frequency of 35 to 65 MHz. Three cascaded amplifiers along with low power RF section, AC/DC power supplies and controls will be used for getting MW level RF power from one source. In the present configuration, two tube based tuned amplifiers, i.e. driver (~150 kW) and final (1.7MW) stage amplifiers are...
Pierre Dumortier
(LPP-ERM/KMS)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The JET ICRF ITER-like Antenna (ILA) is composed of four resonant double loops (RDLs) arranged in a 2 toroidal by 2 poloidal array. Each RDL consists of two poloidally adjacent straps fed through in-vessel matching capacitors from a common Vacuum Transmission Line. Two toroidally adjacent RDLs are fed through a 3dB combiner-splitter.
The JET ILA antenna has been operating at 33, 42 and 47MHz...
Frederic Durodie`
(Laboratory for Plasma Physics)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
The ITER-like Antenna (ILA) [1] for JET is a 2 toroidal by 2 poloidal array of Resonant Double Loops (RDL). It featurs in-vessel matching capacitors feeding RF current straps in Conjugate-T (CT) manner, a low impedance quarter-wave impedance transformer and a service stub allowing hydraulic actuator and water cooling services to reach the aforementioned capacitors. A 2ndnd stage...
A. Dunaevsky
(Tri Alpha Energy)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
In the C-2 field-reversed configuration (FRC) experiment, tangential neutral beam injection (NBI), coupled with electrically-biased plasma guns at the plasma ends and advanced surface conditioning, led to dramatic reductions in turbulence-driven losses.11 Under such conditions, highly reproducible, macroscopically stable, hot FRCs with a significant fast-ion population, total plasma...
R.S. Delogu
(Consorzio RFX)
9/8/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
To study and optimize negative ion production, the SPIDER prototype (beam energy 100 keV, current 48 A) is under construction in Padova, Italy. The instrumented calorimeter STRIKE (Short-Time Retractable Instrumented Kalorimeter Experiment) has been designed with the main purpose of characterizing the SPIDER negative ion beam in terms of beam uniformity and divergence during short pulse...
Laurent Jung
(National Fusion Research Institute)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
An elaborate control of waveforms of poloidal field (PF) coils is prerequisite to ensure a reliable plasma start-up in ITER. An additional requirement in the ITER PF coil scenario development is that coil currents should be optimized to minimize quench risks during a discharge. In this paper, we use the quadratic programming method to optimize ITER PF coil currents at the initial magnetization...
Marco Cecconello
(Uppsala University)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
The ITER Radial Neutron Camera (RNC) is a diagnostic with multiple collimated inputs aiming at characterizing the neutron source. The RNC plays a primary role in the advanced control measurements and physics studies of ITER, and acts as backup for system machine protection and basic control measurements. The RNC primary design driver is the measurement of the neutron emissivity radial profile...
Gerhard Raupp
(Tokamak Scenario Development E1)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
To operate ITER and control long and finally thermonuclear discharges with very complex physics and a limited set of actuators requires a sophisticated Plasma Control System (PCS). To provide the required control functionality, the PCS will include many control loops to keep parameters within operation envelopes. These must be backed by exception handling functions, to optimize continuous...
Wolfgang Treutterer
(E1)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
Design of the ITER plasma control system is proceeding towards its next - preliminary design - stage. During the conceptual design in 2013 an overall assessment of high-level control tasks and their relationships has been conducted. The goal of the preliminary design is to show, that a reasonable implementation of the proposed concepts exists which fulfills the high-level requirements and is...
Alessandro Formisano
(Dept. of Industrial and Information Engineering)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
The magnet system in ITER is composed by three main coils groups, characterized by tight tolerances on manufacturing and assembly, to keep error fields at levels compatible with plasma operation. Additional coils correct error fields guaranteeing suitable accuracy at start of flat top [1].
Plasma initiation in ITER will be critical, since low electric field will be available, and a reduction...
Ruben Specogna
(DPIA)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
We compare three methods for the solution of eddy current problems arising in fusion technology. We first consider the Finite Element Method formulation based on the reduced magnetic vector potential [1]. This formulation provides a very sparse system matrix and is able to solve problems on meshes composed of tens of millions elements. Yet, it requires to produce the mesh for both conducting...
Luca Zabeo
(ITER Organization)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
The ITER Plasma Control System (PCS) is now approaching the second phase of development, the Preliminary Design Review (PDR). The PDR, expected at the end of 2016, is now more deeply investigating possible solutions for the different control areas aimed at operations up to 15MA with low auxiliary heating in L-mode. The entire sequence of a plasma discharge from the break-down to the...
Christopher James Rapson
(Max Planck Institute for Plasma Physics)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
Integrated control of many plasma parameters simultaneously is expected to increase the reproducibility and stability of scenarios, which are otherwise developed laboriously through trial and error. The benefits are expected to be especially important for high performance scenarios, operating near multiple stability boundaries. The two main challenges of integrated control are: firstly the...
Jorge M. Santos
(Instituto de Plasmas e Fusao Nuclear)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
On future long pulse fusion devices an extended set of diagnostics will play an increasingly important role in advanced plasma control. In particular, O-mode microwave reflectometry will be used, on ITER and foreseeably on DEMO, to complement the standard magnetic diagnostics for plasma position control. With the preliminary design of ITER’s plasma position reflectometers (PPR) presently...
Jiaxian Li
(Center for Fusion Science)
9/8/16, 2:20 PM
C. Plasma Engineering and Control
Poster
The advanced configurations (snowflake and tripod) have been designed with EFIT based on current poloidal field (PF) coils system of HL-2M to study the advanced divertor physics and support the high performance plasma operation. The characteristic parameters of the advanced configuration (the distance between two X-points, magnetic flux expansion and weak field area and so on), especially the...
Alexey Gorbunov
(NRC Kurchatov Institute)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Laser-induced fluorescence (LIF) diagnostic system on ITER will be used for local measurements of helium density (nHe) and ion temperature (Ti) in the divertor region. The diagnostics is combined with divertor Thomson scattering (DTS) via common laser injection and signal collection optics. Physical aspects of the LIF method for measuring the plasma parameters and the layout of the system...
Ilya Orlovskiy
(NRC "Kurchatov Institute")
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
First mirrors (FMs) for ITER optical diagnostics induce a number of specific requirements including low sputtering rate, high neutron/gamma radiation and thermal stability to keep the optical performance in the DT plasma shots. Additionally, the FM surface must withstand the discharges by a cleaning system aimed to eliminate Be deposits. A number of experiments have shown that the mirrors made...
Konstantin Vukolov
(Kurchatov Institute)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Silica-based optical fibers have a high light transmission in visible range and so they are widely used for transmitting the light from plasma to detectors in modern thermonuclear facilities. The fiber bundle is comprised as a rule of several tens or hundreds optical fibres of 100-500 microns diameter. The lifetime of the optical fiber in ITER should be more than 15 years. Radiation resistance...
Evgeny Andreenko
(NRC “Kurchatov institute”)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The performance of ITER Main Chamber H-alpha & Visible Spectroscopy is challenged by the problem of separating the contribution of visible light emitted in the scrape-off-layer (SOL) from the background of much higher intensity, produced by the divertor stray light (DSL) reflected by the all-metal first wall (S.Kajita, et al., PPCF, 2013). A differential (bifurcated-line-of-sight) measurement...
Vincent Martin
(Bertin Technologies)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The ITER equatorial visible and infrared wide-angle viewing system is a first plasma diagnostic that will be used to image the visible plasma boundary and the in-vessel components temperatures for real-time machine protection and plasma control purposes, as well as offline physics studies. The system will be installed in four equatorial ports and will have 15 lines of sight covering most of...
Sven Gutruf
(Kampf Telescope Optics)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The H-alpha and Visible Spectroscopy Diagnostic shall measure spectrally, spatially and temporally resolved emission of hydrogen isotopes and impurities in the ITER scrape-off layer. There are four H-alpha diagnostic channels, located in 3 port plugs.
In the current design status, all main interfaces have been iterated with the Port Integrator. All major subsystems, of this complete end to...
Arnd Reutlinger
(Kampf Telescope Optics)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The H-alpha and Visible Spectroscopy Diagnostic shall measure spectrally, spatially and temporally resolved emission of hydrogen isotopes and impurities in the ITER scrape-off layer. Four H-alpha diagnostic channels are designed to observe the plasma. They are located in 3 port plugs:
- Equatorial Port #11:
TV (Top View): poloidal wide Field of View (FoV) covering the upper part of the...
Matthew Smiley
(General Atomics MFE)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
One of the diagnostic systems being provided by the US is the Upper Wide Angle Viewing System (UWAVS), which provides real-time, simultaneous visible and infrared images of the ITER divertor region via optical systems located in five upper ports. The UWAVS is designed in three main sections: in-vessel, interspace and port cell assemblies. Each assembly utilizes multiple steering and relay...
Eugene Mukhin
(Ioffe Institute)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
ITER Divertor Thomson scattering (DTS) was discussed in a number of presentations and papers. The development of diagnostic equipment for ITER DTS is under way and coming to its conclusion. Choice and justification of lasers and polychromator design as well as first mirror protection are the focus of the presentation.
Q-switched Nd:YAG laser for DTS in ITER (1.064mm, 2J, 50Hz, 3ns) is...
YoungHwa An
(National Fusion Research Institute)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The local shielding design for the detector of ITER VUV Edge Imaging Spectrometer is optimized based on the MCNP calculation using a local port cell model of ITER Upper Port #18. A back-illuminated CCD, the envisaged VUV detector for ITER VUV Edge Imaging Spectrometer will be installed at ITER Upper Port #18 port cell region, in which a harsh radiation environment is expected with neutron flux...
Bastian Weinhorst
(Institute for Neutron Physics and Reactor Technology)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The Charge Exchange Recombination Spectroscopy (CXRS) diagnostic aims to measure emission lines of impurity isotopes in the ITER plasma in order to quantify several parameters like the composition of the plasma (density of helium, deuterium or tritium), the ion temperature or rotation velocities. The core plasma CXRS shall be installed in one of the ITER Upper Port Plugs (UPP #3). Currently,...
Valentina Huber
(Forschungszentrum Jülich GmbH)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Imaging systems are an indispensable technique for successful plasma operation of fusion devices. At the JET tokamak, numerous cameras in the VIS/NIR/MWIR spectral ranges are used for plasma physics studies as well as for the real time overheating protection of the first wall and for live plasma monitoring during operation. The protection system, on the basis of the NIR imaging cameras, is...
Ivan Lupelli
(UKAEA-CCFE)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The next generation of tokamaks, e.g. ITER, will have extremely large data collection rates (~0.3PBytes per day), significantly larger than those experienced today, with consequential new challenges in data management, data analysis and modelling. With long pulse durations it is important that data be accessible during the experiment for plant monitoring in quasi real-time analysis. One of the...
Gonzalo Farias
(Escuela de Ingenieria Electrica)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Huge databases are a common situation in fusion. Physical properties of plasma are studied by thousands of signals, sampled at very high frequencies, producing enormous amount of data. A medium-size nuclear fusion device such as TJ-II can generate discharges that last around 500 milliseconds, reaching up to 100 Mbytes per one simple shot. Larger fusion devices such as JET can produce 10Gbytes...
Yi Tan
(Department of Engineering Physics)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The noises of a tokamak during operations form the "voiceprint" of a tokamak. By installing a set of microphones in several optimized positions around the tokamak machine, most noises can be detected and can be used as the “voiceprint” of the tokamak for monitoring its status. Noises of a tokamak in discharge-ready status are mainly continuous and/or cyclical noises from pumping system, water...
Igor Nedzelskiy
(Instituto de Plasmas e Fusao Nuclear)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The heavy ion beam diagnostic of the tokamak ISTTOK is operated with a 20 keV Xe++ ion beam and a multiple cell array detector to collect the secondary Xe2+2+ ions created along the primary beam path by ionizing collisions with plasma electrons. In this multichannel mode of operation, the use of standard Proca-Green 30oo parallel plate energy analyzer for the...
Matti Laan
(Institute of Physics)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Laser induced breakdown spectroscopy (LIBS) is a promising tool for remote monitoring of erosion/deposition processes at the first wall of ITER. Proper application of LIBS requires knowing the ablation rates of co-deposited layers on plasma-facing components accurately to obtain elemental depth profiles of different elements on the layers from the recorded LIBS spectra. This goal is, however,...
Gennady Sergienko
(Forschungszentrum Jülich GmbH)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Deuterium-tritium gas mixture will be used as fuel in future fusion devises like ITER. Thus it is important to monitor hydrogen isotope ratios not only in fusion plasma and in the subdivertor/exhaust gases but also retained in the plasma facing components (PFC). Residual gas analysis is traditionally used to quantify the isotope species of the PFCs in the laboratory by means of thermal...
Andrzej Wojenski
(Institute of Electronic Systems)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
This work refers to the currently being developed extended soft X-Ray plasma diagnostics system with the novel, radiation-hard generation of electronics and implemented algorithms. The system is based on the Gas Electron Multiplier detector. For the multichannel, modular systems working with very intense plasmas (e.g. laser generated plasma, plasma fluxes), the phenomenon of the coinciding...
Bernd Sebastian Schneider
(Institute for Ion Physics and Applied Physics)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The characterization of outward filamentary plasma transport in Medium-Size Tokamaks (MST) is an important objective of current fusion plasma research. We aim at improving the diagnostic of transport events in the Scrape-Off Layer (SOL) and further inside by means of various types of newly developed electrical probes combined with the associated probe measurement procedures. Presently, a New...
Tomasz Czarski
(Institute of Plasma Physics and Laser Microfusion)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The measurement system based on GEM - Gas Electron Multiplier detector is developed for X‑ray diagnostics of magnetic confinement tokamak plasmas. The multi-channel setup is designed for estimation of the energy and the position distribution of an X-ray source. The main measuring issue is the charge cluster identification by its value and position estimation. The fast and accurate mode of the...
Maryna Chernyshova
(Institute of Plasma Physics and Laser Microfusion)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Necessity to develop new diagnostics for poloidal tomography focused on the metal impurities radiation monitoring, especially tungsten emission, has become recently inevitable. Tungsten is now being used for the plasma facing material on many machines, including on the WEST project, where an actively cooled tungsten divertor is being implemented. This forced a creation of the ITER-oriented...
Evzen Losa
(Research Centre Rez)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The intended fusion reaction for ITER project is D + T → 44He (3.5 MeV) + 00n (14.1 MeV), which produces high energy neutrons. Portion of these neutrons is effectively captured in breeder blanket, however, many neutrons leak and can cause radiation damage. Monitoring of the neutron damage in ITER internals is necessary due to the aging management. 2323Na(n,2n)...
Milos Jirsa
(Institute of Physics ASCR)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Superconducting RE-BaCuO tapes of different suppliers were tested by magnetic induction (vibrating sample magnetometer, VSM) and by current transport techniques. The tests aimed at finding the best candidates for the tape utilization in a new generation of superconducting magnets for fusion reactors. The electromagnetic characteristics of the tapes as a function of temperature, magnetic field,...
Xinsheng Yang
(Southwest Jiaotong University)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
As the only high-temperature superconductors (HTS) that can be made into round wires without anisotropy, Bi-2212 has significant potential applications as CICC (cable in conduit conductor) for large-scaled superconducting magnets in fusion reactors. However, Bi-2212 is brittle and sensitive to strain which leads to a low mechanical performance. The effort on studying the impact of strain on...
Jonathan Hollocombe
(Theory and Modelling)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The SAGE 2 2 European Horizon 2020 project (grant agreement 671500), led by Seagate with 10 partners, is investigating the needs of future exascale storage systems for data intensive applications. CCFE is one of the partners and SPECTRE (SPECtral Research Engine) is one of the tools being developed to take advantage of the improved data I/O and throughput capability of the SAGE...
Alexey Arkhipov
(Max-Planck Institute for Plasma Physics)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The ITER Diagnostic Pressure Gauges (DPG) shall provide the measurement of the neutral gas pressure, which is an important parameter for basic control of the operation of ITER machine as well as for input to physics models of the plasma boundary. The reference sensor is a hot cathode ionization gauge, which is able to operate in an environment with strong magnetic fields (up to 8 Tesla),...
Sebastian Friese
(Institut für Energie- und Klimaforschung)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The shutter mechanical concept for the ITER core plasma CXRS Fast Shutter is based on elastic bending of a deformable arm structure (length ≈ 1.8 m) which blocks or opens the path of plasma emitted light aiming at the diagnostics first mirror. Bending of the shutter arms is induced by an actuator and will be restrained using the limiting bumpers, where, although the arms are preloaded against...
Andrey Ushakov
(TNO)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The PBS55 Upper-port Wide Area Viewing System (UWAVS) provides real-time, simultaneous visible and IR images of the ITER diverter region via optical systems located in the upper port plugs of the ITER vacuum vessel. Wall temperature and radiance measurements are performed based on the IR-images. Due to mirror contamination with reactor material deposits the optical performance will deteriorate...
Mathias Dibon
(Max-Planck-Institute for Plasmaphysics)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
A disruption is a major plasma instability that follows a sudden loss of plasma energy. During such an event, large electromagnetic forces and high heat loads occur, as well as electrons at relativistic speed. These effects can cause damage to the plasma facing components and thus have to be mitigated. For this purpose high speed gas valves are used to inject a strong pulse of noble gas onto...
Nikola Jaksic
(Max Planck Institute for Plasma Physics)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
“This project has received funding from the Euratom research and training programme 2014-2018”
In plasma fusion research the neutral gas density is usually measured using hot cathode ionisation gauges which are modified for the application in high magnetic fields and for a measurement range between 10-3-3 Pa and 20 Pa. For obtaining sufficient electron emission, high temperatures in...
Fang Liu
(Institute of Plasma Physics)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Bi2Sr2CaCu2Ox is a potential material for the superconducting magnets of the next generation of Fusion reactor. A R&D activity based on Bi2212 wire is running at ASIPP for the feasibility demonstration of CICC. One sub-size conductor cabled with 42 wires was designed and manufactured. A test method was designed and performed to measure the joints resistance and critical current of the Bi2212...
Pascal de Marne
(Max-Planck-Institut fuer Plasmaphysik)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
Manipulators are an important tool to position diagnostics or samples near to the plasma without breaking the vacuum of fusion devices. They can be used for different purposes like measuring plasma parameters with electrical or magnetic probes near to the core plasma or to investigate plasma-wall interaction by exposing dedicated samples. ASDEX Upgrade is operating a set of manipulators, the...
Qin Zeng
(School of Nuclear Science and Technology)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Chinese Fusion Engineering Testing Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plant and to demonstrate generation of fusion power in China. In order to select the most suitable blanket proposal for CFETR, the three blanket concepts (i.e. the helium cooled solid breeder blanket, the liquid LiPb blanket, and the water cooled ceramic breeder...
Weibin Xi
(Tokamak Design Division)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
The original EAST magnet feeders have been operated for over 7 years since 2006. With the improvement of experimental parameters, a new magnet feeder system has been designed for the upgrade project of the EAST. It consists of 13 pairs of superconducting bus-lines with total length over 900 m and 13 pairs high temperature superconducting current leads. Each original bus-line connecting new...
Diogo Eloi Aguiam
(Instituto de Plasmas e Fusão Nuclear)
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The new multichannel X-mode reflectometer installed on ASDEX Upgrade measures the plasma density profile evolution at different positions in front of the ICRF antenna. The reflectometer operates in the extended U-band (40–68 GHz) microwave region, measuring density profiles up to 101919 m-3-3 with magnetic fields between 1.5 T and 2.7 T. In this heterodyne reflectometer...
Fabio Pollastrone
(FSN (Nuclear Fusion and Fission and Related Technologies Department))
9/8/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
The electrical pattern recognition can be useful in several applications, generally it is used to detect particular events or anomalies in the signal under analysis or to identify precursors, especially in electrophysiology. Each application requires customized algorithms and appropriate signal processing capabilities. In this paper we present an application of pattern recognition to real-time...
Hee-Jae Ahn
(NFRI)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
The central solenoid (CS) of the KSTAR consists of four pairs of superconducting coils compressed axially by preloading structures. The axial pre-compression was designed to 15 MN at 5 K, which could suppress the maximum repulsive force of the coils based on reference operation scenarios. Tolerances in-between insulations, buffers, wedges, blocks and shells have been precisely controlled...
Markus Teschke
(E1 - tokamak scenario development)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
BUSSARD is a new inverter system at the nuclear fusion experiment ASDEX Upgrade for mitigation of ELMs and execution of other, physics related experiments. The concept and first results were presented in detail [1]. Four-phase operation was routinely done during shot campaign 2015/16 and many experience in operation was gained. Now, the completion of BUSSARD is almost finished and many...
Nils Arden
(Max-Planck Institute for Plasma Physics)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Recently an inverter system (called BUSSARD) was assembled to individually feed the 16 in-vessel saddle coils of the fusion experiment ASDEX Upgrade (AUG).The new inverter system consists of 16 inverters, each with an output current of up to 1.3 kA and a bandwidth of up to 500 Hz in arbitrary waveforms. Currently, the system is in operation with 4 inverters feeding four in serial connected...
Wang HaiBing
(Center for Fusion Science)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
Study on 300MVA pulse generator starting system
HaiBing Wang, WeiMin Xuan, JianFei Peng, HuaJun Li, LiRong Xu, HaoTian Hu, li Kang
Southwestern Institute of Physics, Chengdu, Sichuan, China
For supplying power for HL-2M Tokamak, a new 300MVA pulse generator has been developed. The new generator with 400 tons of rotor to stored energy will be driven by an 8500kW asynchronous motor. The...
Jianfei Peng
(Tokamak Power Supply Division)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
A new motor generator (MG) system is building mainly for the poloidal field power supply system of the HL-2M Tokamak. This MG system will be capable of providing a peak capacity of 300 MVA and delivering up to 1350 MJ per pulse at 15 min intervals. The system consists of a 300 MVA MG and its auxiliary systems. The MG adopts the semi umbrella vertical shaft type and consists of an 8500kW...
Shouzhi Wang
(Department of Engineering Physics)
9/8/16, 2:20 PM
E. Magnets and Power Supplies
Poster
A high voltage power supply (HVPS) used for the ECRH system on the SUNIST tokamak is introduced. It is able to output a 50 ms pulse of -40 kV / 15 A in every 5 minutes. The voltage drop for the whole flat top is less than 2%. In each arcing events, the maximum energy delivered to the load is less than 15 Joules.
The HVPS is based on Marx Generator and PSM technologies using fast switch...
G. Pintsuk
(Forschungszentrum Jülich GmbH)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
The WEST (W -for tungsten- Environment in Steady-state Tokamak) project is based on an upgrade of Tore Supra tokamak. ITER-like actively cooled tungsten targets (monoblocks) will be integrated in the lower divertor and a new set of actively cooled tungsten coated plasma facing components will cover a part of the vessel to provide a fully metallic environment.
In preparation of the production...
Youngjae Park
(Department of Nuclear Engineering)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Development of reliable high heat flux removal techniques is an important issue to design plasma facing components in a fusion reactor. The ITER-like divertor cooling design based on water-subcooled flow boiling is one of the well-developed divertor cooling schemes. To withstand such a high heat flux in the vertical target of the ITER divertor, a twisted tape is inserted into a CuCrZr tube...
Kyung-Min Kim
(National Fusion Research Institute)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
It is so important that the bonding technology between tungsten and dissimilar metals for the PFC of ITER and DEMO. The development of tungsten brazing technology was first launched for the KSTAR PFC.
Flat type tungsten block was brazed on CuCrZr in vacuum at a temperature of 980 °C for 30 minutes using silver free brazing alloy. A OFHC-copper was used as an interlayer between tungsten and...
Dong Jun Kim
(Korea Atomic Energy Research Institute)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Tungsten coated mock-ups for developing the Plasma facing component (PFC) werefabricated and tested in the plasma torch and high heat flux test facility with electron beam,which can be used in the repair of the damaged PFCs. For evaluating the life-time of the tungsten coated mock-up, the erosion rate was measured and thermal-lifetime analyses were performed with the fabricated mock-up. And...
Suk-Kwon Kim
(Nuclear Fusion Engineering Development Division)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
The Developments of plasma facing components (PFCs) are the key items for the nuclear fusion reactors. The most components for the tokamak PFCs are the blanket first wall, divertor, heating ports, and diagnostics ports. These PFCs are composed of the armour materials, the heat sink for the cooling, and the structural materials. Be, W, C-composites, and advanced materials were selected for...
Shenghong Huang
(Modern mechanics)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
After years of exploration and development, research of magnetic confinement nuclear fusion is progressed into stage of experimental fusion reactor construction and test. As a key plasma-facing component, the anti-fatigue performance of first wall of fusion reactor receives widely concerns. Due to the fact of enduring both periodic loads of pulse operating mode and shock loads of transient...
Rajamannar Swamy Kidambi
(Divertor & First Wall Technology Development Division)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
This paper deals with the design of High Pressure High Temperature Water Circulation System (HPHT-WCS) for High Heat Flux Test Facility (HHFTF) of IPR and its related thermal hydraulic experiments. HHFTF has been established at IPR, India for testing performance of plasma facing components under intense heat loads expected in plasma fusion devices. Plasma facing components of the present day...
Kohei Hamaguchi
(Division of Sustainable Energy and Environmental Engneering)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
It is desirable to develop tungsten (W) diverter in Tokamak-type nuclear fusion reactor including the International Thermonuclear Experimental Reactor (ITER). W has the highest melting point in all metals and thus is a promising material of the diverter. Since the diverter will repetitively undergo high heat flux of 100MW/m2 2 at least in a few tens of millisecond or less when...
Ryuji Ohsone
(Japan Atomic Energy Agency)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
A hot isostatic pressing(HIP) method is one of the candidate process to fabricate the fusion blanket the first wall with built in cooling channels. Thin plates and rectangular tubes made of reduced activation ferritic/martensitic (RAFM) steel, such as F82H, are consolidated by the HIP method. The first wall quality therefore depends on the integrity of the formed HIP joint. In laboratory scale...
Toshikio Takimoto
(Tokai University)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
In the magnetic confinement fusion reactor for high power and long pulse operation, enormous heat flux (exceeding 10 MW/m22) is expected to flow onto divertor plates from core plasma. In order to reduce this heat load, the divertor geometry on stationary detached plasma formation must be realized. In addition, the neutral particle flowback into the core plasma is necessary to...
Arnold Lumsdaine
(Oak Ridge National Laboratory)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
One of the critical challenges for the development of next generation fusion facilities, such as a Fusion Nuclear Science Facility (FNSF) or DEMO, is the understanding of plasma material interactions (PMI). The field of PMI occurs at the intersection of plasma physics, materials science, and engineering, and requires expertise and research and development in each of these fields. Making...
Keith Smith
(Materion Beryllium and Composites Elmore, OH, United States)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
In its current design, the ITER fusion machine will use tens of thousands of beryllium tiles as plasma-facing components in its First Wall. S-65 is one of three grades of beryllium which has been accepted by the ITER International Organization for use in the reactor. The beryllium material for ITER has to pass through many machining and manufacturing processes after being consolidated by...
Mizuki Noguchi
(Advanced Energy Engineering Science)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
It is important to understand tritium (T) desorption behavior from plasma-facing materials of a fusion reactor in order to discuss tritium recovery method from in-vessel components. Tungsten (W) is a candidate material for plasma-facing components. Although a sputtering rate of W by hydrogen isotopes is low, a certain amount of W deposition layer will be formed on plasma-facing wall. In this...
Irina Tazhibayeva
(Insitute of Atomic Energy NNC RK)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Tritium is a prospect fuel material for future fusion power reactors, thus tritium breeding in these reactors is one of the design challenges, which can be solved by using the lithium-containing materials for contrstruction of the reactors’ blankets. Also of great interest is use of lithium as a plasma-facing material, for example, in the form of lithium-capillary porous systems (CPS). Such...
Alexey Popkov
(Plasma Physics Department)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Lithium is considered as a promising material for plasma-facing components (PFC) in future fusion devices. A number of experiments have already demonstrated positive effects of lithization and using of Li based PFCs on plasma operation. During operation of the machine, lithium is deposited on the surrounding walls and in shadowed areas. One can expect a high concentration of hydrogen isotopes...
Fumitaka Ishikawa
(Tokai University)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Tungsten is important candidates for plasma-facing component applications on the development of magnetic fusion reactors. Particularly, it is important to understand the behavior of hydrogen isotopes in tungsten of the diverter wall material. In this study, we have performed the irradiation experiments using deuterium and helium mixed plasma in order to investigate the deuterium retention and...
Daniel Iglesias
(UKAEA-CCFE)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Virtual prototyping enhances traditional engineering analysis workflow when a quick evaluation of complex load cases is required. During design, commissioning or operating phases, components can be virtually tested in realistic conditions by using previously validated numerical models and experimental databases.
Three complementary applications have been developed under this approach for the...
Masayuki Tokitani
(Department of Helical Plasma Research)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
The study is focused on modification of surfaces of the tungsten-coated divertor tiles used in the first campaign (2011-2012) of the JET tokamak with the ITER-lLike Wall (JET-ILW). The analyses by means of several material research techniques have been carried out at International Fusion Energy Research Centre (IFERC), JAEA Rokkasho.
Samples, in the form of disks (17 mm in diameter), extracted...
Aleksander Drenik
(EUROfusion Consortium)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
After the transition to full metal wall configurations at AUG and subsequently at JET, impurity seeding became necessary to maintain the divertor heat loads below material limits in H-mode discharges. Among the studied impurities, nitrogen (N) was found to be the most favourable option. However, it was also found that N2-seeding leads to formation of ammonia (NH3). Nitrogen and NH3 retained in...
Thomas Hartl
(E1)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Glow discharge cleaning (GDC) and coating of the plasma facing components (PFC) is still crucial for fusion research machines to reach demands on plasma cleanliness for elaborate investigations. To correspond with latest experimental findings the GDC-system of ASDEX Upgrade (AUG) has been remodeled entirely.After transition to tungsten PFCs it becomes evident that Helium implanted during GDC...
Rudolf Neu
(Plasmarand und Wand)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Since 2014 ASDEX Upgrade (AUG) is using bulk tungsten tiles at the outer divertor strike-point. In two experimental campaigns more than 2000 plasma discharges with up to 10 s duration and 100 MJ plasma heating were successfully conducted, without impairment by the W tiles. However, an inspection after the campaigns revealed that a large number of tiles suffered from deep cracking, mostly...
Johan Oosterbeek
(Eindhoven University of Technology)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
Diagnostic systems are essential for the development of ITER discharges and to reach the ITER goals. Many of these diagnostics require a line of sight to relay signals from the plasma to the diagnostic, typically located outside the torus shall. Such diagnostics then require vacuum windows that isolate the torus vacuum and crucially ensure tritium containment. While such windows are routine in...
Hun-Chea Jung
(ITER Korea)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
The ITER blanket shield block (SB) is one of the in-vessel components, which is designed to provide nuclear shielding and to supply the cooling water to vacuum vessel and external component. The ITER SB is classified the VQC 1A as vacuum classification and its manufacturing process and cleaning procedure shall comply with the ultra-high vacuum conditions necessary for machine operation and...
Paul Edwards
(Tokamak Engineering Department)
9/8/16, 2:20 PM
F. Plasma Facing Components
Poster
The Final Design Review for the Blanket Manifold (BM) was successfully held in December 2015. Since the Conceptual Design Review, a concerted effort has been necessary on finalisation of the multi-pipe design, verification by analysis and practical validation to address challenging design requirements, and installation/maintenance processes.
During normal operating conditions the BM provide...
Yongbo Wang
(Lappeenranta University of Technology)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
For ITER or the future DEMO remote maintenance system (WPRM), several types of special tailored automatic manipulators are needed for vacuum vessel (VV) component transportation, inspection, and removal from and replacement to the VV wall. These tailored manipulators, such as Multi-purpose Deployer, Articulated Inspection Arm (AIA), Diverter Cassette Mover etc., should be calibrated with very...
Takahito Maruyama
(Department of ITER Project)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
How to recover from failures of components in radiation environment is an important issue of the ITER remote handling systems. Recovery operations of the remote handling systems must be performed remotely due to limitation of human access. For the ITER Blanket Remote Handling system, failure modes have been analysed, and the analysis has concluded that electrical failures of actuators, which...
Yuto Noguchi
(Fusion Research and Development Directorate)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The ITER blanket module has hydraulic connections to the cooling water manifold. The connections are designed to be cut and re-welded remotely in the vacuum vessel during blanket maintenance due to irradiation of in-vessel components after D-T experiment. In course of the R&D activities for in-vessel pipe welding, a study [1] demonstrated that good weld quality can be achieved by correcting...
Naveen Rastogi
(Remote Handling Division)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
An integrated control system architecture has been defined for the implementation of ITER Remote Handling (RH) equipment systems. The RH Core System (RHCS) is a standard software platform used for the development of ITER RH equipment controller applications to facilitate the integration with this system. It installs on top of the CODAC core system and provides a uniform platform for the...
Jean-Pierre Friconneau
(ITER)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
ITER is a large scale fusion device designed to study the high temperature fusion reaction between tritium and deuterium. The success of a tokamak-type fusion reactor will depend to a great extent on developing reliable and safe methods of carrying out routine maintenance and repairs remotely.
Remote Handling System (RHS) are used to perform remotely the maintenance of the vacuum vessel. They...
50252.
P4.133 Irradiation tests of radiation hard components for ITER blanket remote handling system
Makiko Saito
(Naka Fusion Institute)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The ITER Blanket Remote Handling System (BRHS) will handle the blanket modules (BMs), which can weigh up to 4.5 ton and be larger than 1.5 m, stably and with a high degree of positioning accuracy. When the ITER has stopped plasma operations for maintenance, the BRHS will be installed in the vacuum vessel, whose components are radioactive, to remove and install the BMs. Therefore, the BRHS will...
Bingyan Mao
(Laboratory of Intelligent Machines)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
In the ITER or the future DEMO reactor systems, due to the neutron activation, the remote handling tasks such as inspection, repair and/or maintenance of in-vessel and ex-vessel components must be carried out using a wide variety of special tailored automatic manipulators. The structure of these manipulators can be designed as a pure serial articulated arm or a pure parallel mechanism, but for...
Paulo Carvalho
(Instituto de Plasmas e Fusao Nuclear)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Experimental fusion reactors aim at the exploration of the nuclear fusion as a viable energy resource. Remote Handling Systems (RHS) are specially designed for regular operations of inspection and maintenance inside the reactors, such as the In-Vessel Transporter, an extendable robotic arm deployed in the equatorial level of ITER. The reactor is shutdown during the installation and operation...
Huapeng Wu
(Lappeenranta university of technology)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The EAMA robot is a long slender arm for tokamak inspection and maintenance. In such conditions, grasp techniques ignoring or trying to avoid contact with the components of the vacuum chamber brings bottlenecks on the system control. During the grasping and releasing objects the contact with vacuum chamber is a critical condition for providing robust and achievable solutions of robot control....
Cephise Louison
(IRFM)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
The development of fusion plants is more and more challenging. Compared to previous fusion experimental devices, integration constraints, maintenance and safety requirements are key parameters in the ITER project. Components are designed in parallel and we must consider integration, assembly and maintenance issues, which might have an impact on the overall design. That also implies to consider...
Tom H. Owen
(Remote Applications in Challenging Environments (RACE))
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Mascot is a two-armed dexterous master-slave telemanipulator device linked by force-reflecting servomechanisms, giving the operator a tactile sensation of doing the work. Mascot version 4.5 is currently in use at the Joint European Torus (JET) experimental nuclear fusion facility. Its role is to maintain the inside of the reactor vessel without the need for manned entry. The slave is...
Wang Rui
(Institute of Plasma Physics Chinese Academy of Sciences)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Full penetration welding and 100% volumetric examination of weld joints are strictly required for all welds of pressure retaining parts of the CFETR Vacuum Vessel (VV) according to the design manual. However not every welding joint can be tested using RT method due to component structure and welding position. Therefore, the ultrasonic testing (UT) has been selected as an alternative...
Jianguo Ma
(Institute of Plasma Physics)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
With the development of CFETR engineering design, a full-scale sector prototype of vacuum vessel has been carried out as one of the major R&D projects. The welding structure between vacuum vessel sectors in field assembly is modeled in this prototype, and NG-TIG is taken for an applicable welding strategy with small welding deformation, high-quality welds and excellent adaptability to the...
Zhihong Liu
(instititue of plasma physics chinese academy of sciences)
9/8/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
Chinese Fusion Engineering Testing Reactor (CFETR) is a superconducting magnet Tokamak, it has the equivalent scale with complementary function to International Thermonuclear Experimental Reactor (ITER). The vacuum vessel (VV) which has a double-layer structure，Cooling water circulating through the double-layer structure will remove the heat generated during operation. The VV will provides a...
Kanetsugu Isobe
(Japan Atomic Energy Agency)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
In the one of Broader Approach (BA) activities aiming to the development for a DEMO fusion reactor, the R&D of tritium technology has been carried from 2007. The period consists of Phase 1 (2007-2010) and Phase 2 (2010-2016). International Fusion Energy Research Center (IFERC) including DEMO R&D building was constructed in Rokkasho BA site of Japan. The R&D building is a facility to handle...
Juro Yagi
(Department of Helical Plasma Research)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
One of the major concerns for molten salt breeding blanket system is the low tritium solubility, high equilibrium tritium pressure in other words, of the molten salts including FLiBe, FLiNaBe and FLiNaK. Owing to this, vanadium alloy (V-4Cr-4Ti) has been thought to be inappropriate as a structure material in molten salt breeding blanket because of its high tritium solubility.
The concept of...
Youhua Chen
(University of Science and Technology of China)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The neutron multiplier and the tritium breeder materials are made into millimeter-sized particles and arranged in the solid breeder blanket. Helium (mixed with 0.1% content of H2) is used as the purge gas to sweep tritium out when it flows through the pebble beds. Previous research shows that binary pebble beds present a better performance in tritium breeding than unitary pebble beds. Since...
Benedikt J. Peters
(Institute for Technical Physics)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The effect of superpermeability is capable of separating hydrogen and its isotopes out of gas mixtures at low pressures even against a pressure gradient. This process allows strongly enhanced permeation. It relies on metal membranes that are exposed to atomic hydrogen. If the surface inhibits the chemisorption on its surface, the atomic hydrogen can still enter the bulk, but hydrogen...
Karine Liger
(CEA Cadarache)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
In the framework of tritium recovery from tritiated water, efficiency of packed bed membrane reactors have been successfully demonstrated. Thanks to protium isotope swamping, tritium can be recovered from tritiated water under the valuable Q2 form (Q = H, D or T) by means of isotope exchange reactions occurring on catalyst surface. The use of permselective Pd-based membrane allows withdrawal...
Alessia Santucci
(ENEA)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The blanket concepts investigated under the EUROfusion program rely on water or helium as the primary coolant medium; the main duty of the coolant is to recover the thermal power from the first wall and the blanket units and drive it into the Primary Heat Transfer System (PHTS).
The coolant path goes through three different systems: the breeder, the tritium plant and the PHTS. In the breeding...
Silvano Tosti
(ENEA)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Fusion plasma exhaust is generally composed of unburned fuel (deuterium and tritium), helium and few impurities. However for a metal wall machine (like DEMO) that reaches elevated powers, a certain amount of plasma enhancement gas (nitrogen, Ar, Ne, etc.) could be used as seeding for enhancing the radiative power and decreasing the power load over the plasma facing components. The recovery of...
Marco Incelli
(DEIM)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Pd-based membrane reactors are well-known technologies in the fuel cycle of the next fusion plants. In this work the application of Pd-Ag membranes have been studied in order to recover tritium in both molecular (Q2) and, especially, oxidised (Q2O) form in the tritium extraction system (TES) of the HCPB blanket.
The membrane reactor is made up of a Pd-Ag membrane tube filled with a catalyst....
Laetitia Frances
(ITEP)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Tritium self-sufficiency and management in nuclear fusion power plants is still challenging. Advanced technologies to extract tritium from lead lithium (Pb-16Li) as possible breeder material are required. The Vacuum Sieve Tray (VST) method consists in pushing Pb-16Li through a tray of submillimeter scaled nozzles towards a chamber maintained under dynamic vacuum. At the exit of each nozzle, an...
Yasunori Iwai
(Department of Blanket Systems Research)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Effect of halogenated gas on detritiation efficiency of the detritiation system was investigated. In order to accelerate tritium safety of the Japanese DEMO reactor, the detritiation system should be designed taking possible off normal events such as fire carefully into consideration. In an event of fire in a tritium processing room, halogenated gases such as hydrogen chloride, halogenated...
Kwangjin Jung
(University of Science and Technology (UST))
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The hydrogen isotope storage and delivery system (SDS) is a complex system that includes many individual components. One of the most important parts of the SDS is a metal hydride bed, which stores and delivers the hydrogen isotopes and pure gases required for a nuclear fusion reactor. We have been developing a metal hydride bed using depleted uranium (DU). The hydrogen delivery performance of...
Yeanjin Kim
(quantum energy chemical engineering)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The hydrogen isotope storage and delivery system (SDS) is a part of a nuclear fuel cycle. It is a complex system that is composed of numerous components such as a metal hydride bed, measuring tank, and other essential components. Depleted uranium (DU) was chosen as a hydrogen isotope storage material because of its rapid reactivity. We designed and manufactured the DU hydride bed to store the...
Alina Niculescu
(National Institute for Cryogenics and Isotopes Technologies - ICSI)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Cryogenic distillation (CD) process is being employed, among other applications, in tritium separation technologies and in case of ITER is one of the key proceses in the fuel cycle. The ITER Isotope Separation System has to process by cryogenic distillation various mixtures of H-D-T depending from the various torus operation scenarious.
Cryogenic distillation has also been employed to separate...
George Ana
(National Institute for Cryogenics and Isotopes Technologies - ICSI)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
During normal operation of a CANDU reactor, large amounts of tritiated heavy water is being produced as result of neutron absorption by the heavy water used as moderator and cooling agent. Tritium in the heavy water, being radioactive, brings a significant contribution to the personal doses and also represents an environmental hazard if a waterspill occurs.
The Pilot Plant for T2 and D2...
Tao Jiang
(Center for Fusion Science)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Being part of the ITER fuelling system, the primary functions of the Gas Injection System (GIS) include providing gases for plasma discharge, wall conditioning, and neutral beam injectors. The Gas Distribution System(GDS) is a key sub-system of the GIS, which shall distribute gases obtained from the Tritium Plant, to the Gas Valve Boxes for the Pellet Injection System, Gas Fuelling System,...
Francesca Bombarda
(FSN Department)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The injection of cryogenic pellets from the low field side (LFS) has long been in use for core fueling of fusion devices. However, with higher plasma temperatures and bigger sizes, this technique becomes increasingly inadequate to ensure effective core particle deposition; injection from the high field side (HFS) has shown better results, despite the severe limitations imposed to the pellet...
Igor Vinyar
(PELIN)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
High frequency pellet injectors have been developed for edge localized mode mitigation and plasma fuelling of the EAST and KSTAR tokamaks. Each pellet injector is able to inject solid deuterium or hydrogen pellets at steady state mode. Both injectors consist of a continuous ice generator based on a screw extruder cooled by liquid helium and pneumatic punches for pellet fabrication and...
Dimitris Valougeorgis
(Mechanical Engineering)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Recently, an integrated software algorithm for modeling gas distribution systems operating under vacuum conditions has been developed [1]. It has been successfully applied to model the 2012 ITER divertor pumping system and results have been provided for the flow patterns in the cassettes and the divertor ring, as well as for the throughputs in the burn and dwell phases. In all cases the input...
Antonio Frattolillo
(ENEA C.R. Frascati)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Core fuelling of DEMO fusion reactor is under investigation within the EUROfusion Work Package "Tritium, Fuelling and Vacuum". An extensive analysis of fuelling requirements and technologies, suggests that pellet injection still represents, to date, the most realistic option. Modelling of both pellet penetration and fuel deposition profiles for different injection locations, assuming a...
Silvio Giors
(Plant Engineering Department)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The ITER vacuum system, one of the largest and most complex vacuum systems ever to be built, will use first of a kind cryopumps to provide high vacuum conditions to the torus vessel, cryostat vessel, and neutral beam injectors. In order to evacuate the high gas flows required by the plasma scenarios, the cryopumps will need sequential regenerations with unprecedented high frequencies.
The...
Ranjana Gangradey
(Development of cryopump and pellet injector system)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Indigenous cryoadsorption cryopump with large pumping speeds gases like hydrogen and helium is developed and a set of experiments performed at the Institute for Plasma Research (IPR). India. Towards its successful realization, technological bottlenecks were identified, studied and resolved. Hydroformed cryopanels were developed from concept leading to the design and product realization with...
Thomas Giegerich
(Institute for Technical Physics)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
The reduction of tritium inventories is a key challenge for DEMO and future fusion power plants. As large amounts of tritium have to be processed in the inner fuel cycle, an inventory-optimized vacuum pumping process – the KALPUREX process – has been developed at KIT. Here, continuously working and non-cryogenic vacuum pump trains will be used in order to keep the tritium residence times and...
Jordi Abella
(Analytical and Applied Chemistry)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Accurate and reliable tritium management is of basic importance for the correct operation conditions of the blanket tritium cycle. The determination of the hydrogen isotopes concentration in liquid metal is of high interest for the blanket correct design and operation. Sensors based on solid state electrolytes can be used to that purpose. It is worth mentioning that these type of sensors offer...
Luigi Candido
(Department of Energy)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
A crucial issue for the design of HCLL (Helium Cooled Lead Lithium) Test Blanket Module of ITER and HCLL, WCLL, DCLL Breeder Blanket of DEMO is to efficiently characterise the tritium inventory inside the blanket and the permeation of tritium into the coolant in order to reduce as much as possible the radiological hazard towards the external environment. A fast and reliable sensor is required...
Yuki Edao
(Department of Blanket Systems Research)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Various methods of tritium measurement have been applied depending on a chemical formof tritium. A method combined oxidation catalyst and water bubblers has been used as one of the most quantitative analysis methods for gaseous tritium. We previously developed a quantitative analysis system to measure gaseous tritium in a high accuracy using by an organic-based hydrophobic platinum catalyst....
David Wilson
(CCFE)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
In support of ITER, two experimental campaigns are foreseen to take place at JET, the first with tritium only and a second with deuterium plus tritium to explore the machine fusion potential. To support the tritium operation, a total of five Tritium Introduction Modules (TIMs) are expected to be installed at JET, one on top of the machine, another in the mid-plane and three in the divertor...
Ivo Carvalho
(Instituto de Plasmas e Fusão Nuclear)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
As part of the JET Programme in Support for ITER, campaigns with pure Tritium-Tritium (TT) fuel and Deuterium-Tritium (DT) mixture are planned at JET. Unlike the previous DT campaign at JET, these campaigns require a much higher tritium flow rate, particularly, the TT campaign can require up to 3.7 grams of tritium on a single pulse. Five tritium introduction modules (TIMs) fed from the Active...
Oliver Leys
(Institute for Applied Materials)
9/8/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
Advanced tritium breeder pebbles, composed of lithium orthosilicate with additions of lithium metatitanate as a secondary strengthening phase, are produced using a melt-based process. Synthesis powders are heated to high temperatures in a platinum alloy crucible, forming a melt, which is then ejected through a nozzle to form a laminar jet. Longitudinal surface instabilities cause the...
Kuo Tian
(Karlsruhe Institute of Technology)
9/8/16, 2:20 PM
I. Materials Technology
Poster
As the complementary work of IFMIF-EVEDA (International Fusion Material Irradiation Facility- Engineering Validation and Engineering Design Activities) project, WPENS (Work Package Early Neutron Source) project in the framework of EUROfusion activities is committed to the engineering design of an IFMIF-DONES (Demo Oriented Neutron Source) facility, which is an accelerator based intense...
Sachiko Yoshihashi
(Department of Applied Nuclear Technology)
9/8/16, 2:20 PM
I. Materials Technology
Poster
In the international fusion materials irradiation facility (IFMIF), 14 MeV neutrons are generated by 40 MeV deuteron beam injection into a high-speed liquid lithium (Li) plane jet, flowing along a vertical concave wall in vacuum. Measurement of a free surface flow and fluctuation of the thickness are required to produce a stable neutron field and maintain the safety of Li target system.In...
Sergej Gordeev
(Institute for Neutronic Physics and Reactor Technology)
9/8/16, 2:20 PM
I. Materials Technology
Poster
The configuration of the Early Neutron Source (ENS) is the IFMIF-DONES (DEMO Oriented Neutron Source) approach, based on an IFMIF-type neutron source. It aims providing an intense fusion-like neutron spectrum with the objective to qualify on an accelerated time scale structural materials to be used in the future DEMO fusion reactor. IFMIF-DONES is based on the interaction of single 40MeV 125mA...
Hiroo Kondo
(Japan Atomic Energy Agency)
9/8/16, 2:20 PM
I. Materials Technology
Poster
A liquid-Li free-surface stream flowing at 15 m/s under a high vacuum of 1E−3 Pa is to serve as a beam target (Li target) for the planned International Fusion Materials Irradiation Facility (IFMIF). The Engineering Validation and Engineering Design Activities (EVEDA) for the IFMIF are implemented under the Broader Approach. As a major activity of the Li target facility, the EVEDA Li test loop...
Eiji Hoashi
(Osaka University)
9/8/16, 2:20 PM
I. Materials Technology
Poster
A high-speed liquid metal lithium jet (Li jet) with a free surface is planned as a target irradiated by two deuteron beams to generate a neutron field in an accelerator based neutron source, such as that in the international fusion materials irradiation facility (IFMIF). In the IFMIF, it is desirable to stabilize the Li jet for the efficiency of the neutron generation and the safety of...
Takafumi Okita
(Division of Sustainable Energy and Environmental Engineering)
9/8/16, 2:20 PM
I. Materials Technology
Poster
Liquid metal flow has been expected to be applied in various fields. For example, sodium and lithium (Li) are applied as a coolant in the fast-breeder reactor and space nuclear reactor, Li jet as a beam target in the International Fusion Materials Facility (IFMIF) and as a charge stripper in Radioactive Isotope Beam Facility (RIBF) at RIKEN, lithium-lead (Li-Pb) as a liquid metal blanket in a...
Georg Schlindwein
(Institute for Neutron Physics and Reactor Technology (INR))
9/8/16, 2:20 PM
I. Materials Technology
Poster
The so called High Flux Test Module (HFTM) represents the component of IFMIF (International Fusion Irradiation Facility) in which material specimens are being placed that accumulate the highest neutron induced damage rates (≥20 dpa/fpy). Damage rates of this magnitude are limited to a volume of ~500 cm³ (attenuation in beam direction) behind a beam footprint of 20x5 cm. The high flux region of...
Christine Klein
(INR)
9/8/16, 2:20 PM
I. Materials Technology
Poster
During the EVEDA phase of the International Fusion Materials Irradiation Facility (IFMIF), the High Flux Test Module (HFTM) was developed as dedicated irradiation device for Small Specimen Test Technique . In the intensive IFMIF neutron radiation field the specimens are contained in temperature controlled irradiation rigs. Since one of the requirements for the HFTM is to provide a uniform...
Shotaro Matsuda
(Division of Sustainable Energy and Environmental Engineering)
9/8/16, 2:20 PM
I. Materials Technology
Poster
International Fusion Material Irradiation Facility (IFMIF) is the facility generating the high flux and high energy neutron to develop a material for a nuclear fusion reactor. In the IFMIF, high-speed liquid lithium (Li) jet is used as the target irradiated by two deuteron beams. Since the Li jet must flow with high velocity for the heat removal, it is important to research on the...
Yuefeng Qiu
(Karlsruhe Institute of Technology)
9/8/16, 2:20 PM
I. Materials Technology
Poster
The location of the lithium quench tank (QT) is an important safety related issue in the design of the test cell (TC) of the IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented Neutron Source). In the current reference design, the QT is situated outside the TC and is connected to the target assembly through a long lithium outlet channel penetrating the TC floor....
Florian Schwab
(Institute for Neutron Physics and Reactor Technology (INR))
9/8/16, 2:20 PM
I. Materials Technology
Poster
The High Flux Test Module (HFTM) of the International Fusion Materials Irradiation Facility (IFMIF) is a device to enable irradiation of Small Scale Testing Technique (SSTT) specimens by neutrons up to a structural damage of 50 displacements per atom (dpa) in an irradiation campaign of one year. The IFMIF source generates neutrons with a D-T-fusion-relevant energy spectrum and a flux to...
Giuseppe Pruneri
(Fusion department)
9/8/16, 2:20 PM
I. Materials Technology
Poster
The Conventional Facilities of the Linear IFMIF Prototype Accelerator (LIPAc)
Authors
G.Pruneri, P.Cara, R.Heidinger, A. Kasugai, J. Knaster, S. Ohira, Y.Okumura, K.Sakamoto, and the LIPAc Integrated Project Team.
The International Fusion Material Irradiation Facility (IFMIF) aims at qualifying and characterising materials capable to withstand the intense neutron flux originated in the D-T...
Pedro Ortego
(Neutronic Calculations)
9/8/16, 2:20 PM
I. Materials Technology
Poster
In the conceptual design of the beam dump shielding for the foreseen fusion-relevant irradiation facility IFMIF, an inner lead cylinder performs the shielding of the highly activated copper cone undergoing the deuteron beam bombardment and low-alloy steel is used for front shielding. In order to reduce the residual dose around the beam dump at beam-off conditions and dose at hands-on...
Yuki Iwama
(Department of Sustainable Energy and Environmental Engineering)
9/8/16, 2:20 PM
I. Materials Technology
Poster
It is desirable to develop liquid lithium-lead (Li-Pb) blanket for helical-type fusion reactor because of its high cooling and tritium-recovering abilities. Since heat transport under a strong magnetic field in a fusion reactor determines the performance of liquid metal blanket (LMB), it is important to clarify the mechanism of the interaction between Li-Pb flow and the magnetic field. On the...
Petr Stupka
(TEO)
9/8/16, 2:20 PM
I. Materials Technology
Poster
Envisioned fusion facilities for energy production are currently under development within EUROfusion program. In these devices, a D-T plasma is used as energy source. While deuterium is abundant, tritium has to be produced on-site. Tritium, as one of the hydrogen isotopes, easily diffuses through metallic walls of its confinements. Such ‘tritium leakage’ can be limited by developing an...
Masatoshi Kondo
(Research Laboratory for Nuclear Reactors)
9/8/16, 2:20 PM
I. Materials Technology
Poster
The development of functional layers such as the tritium permeation barrier and the anti-corrosion barrier is one of the important issues for the development of liquid breeder blanket. The functional layers with the self-healing function have been developed based on the mechanism of the oxide layer formation. The oxides of yttria (Y2O3) and zirconia (ZrO2) have an excellent chemical stability....
Daniele Martelli
(Department of Civil and Industrial Engineering)
9/8/16, 2:20 PM
I. Materials Technology
Poster
The use of PbLi and RAFM steels in blanket applications requires a better understanding of material compatibility related to physical/chemical corrosion phenomena in the 450-550°C temperature range. The impact of corrosion includes deterioration of the mechanical integrity of the blanket structure due to the wall thinning. Furthermore, serious concerns are associated with the transport of...
Gorka Alberro
(Nuclear Engineering and Fluid Mechanics)
9/8/16, 2:20 PM
I. Materials Technology
Poster
The importance of the hydrogen isotopes transport parameters of Sieverts’ constant and diffusivity in the eutectic lead lithium alloy is well known, as long as it is vital for the determination of tritium management strategies at liquid-metal breeding blanket systems [Helium Cooled Lithium Lead (HCLL), or Dual-Coolant Lead-lithium (DCLL)].
Tritium transport parameters as solubility and...
Sergi Colominas
(Analytical Chemistry)
9/8/16, 2:20 PM
I. Materials Technology
Poster
Lithium 6 is the substance required to generate in-situ tritium in fusion reactors. Because of that, lithium monitoring in lithium-lead eutectic (Pb-15.7Li) is of great importance for the performance of the liquid blanket. Lithium measurements will be required in order to proof tritium self-sufficiency in liquid metal breeding systems. On-line lithium sensors must be designed and tested in...
Jiang Haiyan
(School of Materials Science and Engineering)
9/8/16, 2:20 PM
I. Materials Technology
Poster
In this study, rotating experimental devices were built to investigate the compatibility of the fusion reactor materials RAFM steel, 316L(N) steel，CuCrZr alloy with the Al2O3–water nanofluids. Based on the ITER water-cooling program，the experimental condition parameters were fluid velocity of 1.13 and 3.71m/s，fluid temperature of 70±1◦◦C，testing duration of 2136h，nanofluid mass...
Fabio Tieri
(Fusion Nuclear Tecnologies)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The ASTEC code is a lumped parameter code originally designed to perform safety analysis in fission nuclear power plants. Recently some modules of ASTEC have been modified by IRSN to be applicable for the safety analysis in the nuclear fusion plants. In particular the CPA module ( for the thermal-hydraulics of the containment) and the SOPHAEROS module (to model the physical phenomena...
Jean-Francois Ciparisse
(Industrial Engineering)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
One of the main concerns in Tokamak operation is the dust resuspension and fallout in case of LOVA (Loss Of Vacuum Accident) and LOCA (Loss Of Coolant Accident), as the metallic powders contained in the vessel are radioactive and therefore harmful. Furthermore, they can react explosively with the incoming oxygen if the local composition falls inside the flammability interval and if a hot point...
Luigi Antonio Poggi
(Industrial Engineering)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
STARDUST-U facility is an experimental facility voted to help the scientific community to better understand the problem of dust re-suspension and mobilization in case of Loss Of Vacuum Accidents (LOVAs) or Loss Of Coolant Accidents (LOCAs) inside the next generation fusion reactors like the International Thermonuclear Reactor (ITER) or the Demonstration Power Plant (DEMO).In this work the...
Andrea Malizia
(Industrial Engineering)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The future nuclear plants like ITER, DEMO or PROTO are interested by the problems of dust creation and resuspension. Radioactive dust, if resuspended by accidents in the vacuum vessel, can be dangerous because of its toxicity and capacity to explode under certain conditions. The authors have been working since 2006 on dust resuspension problems through the STARDUST facility before and the...
Jonathan Naish
(Technology)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Effective data visualisation is a key part of the scientific process with complex geometric datasets. It is the bridge between the quantitative content of the data and human intuition. Immersion in virtual reality (VR) provides benefits beyond the traditional “desktop” visualization tools and it leads to a demonstrably better perception of dataspace geometry, more intuitive data...
Zaixin Li
(Center For Fusion Science)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Chinese Fusion Engineering Testing Reactor (CFETR) is aimed to obtain the technologies to fill the gaps between ITER and DEMO. The helium cooled ceramic breeder (HCCB) blanket is one of the candidates for CFETR. Ceramics Li4SiO4, beryllium and helium of 8 MPa were selected as tritium breeding material, neutron multiplication and coolant, respectively. CLF steel developed in SWIP, one of...
Mikhail Subbotin
(CERN)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
In the framework of the joint Russian – Italian collaboration on the development of the IGNITOR project some preliminary estimates of the risk factors that may be occurring during the realization of the project were recently carried out.
A distinctive feature of the IGNITOR project is the fact that it contains some innovative solutions in the areas of research, engineering and technology,...
Andre Haußler
(Institute for Neutron Physics and Reactor Technology (INR))
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The Helical-Axis Advanced Stellarator (HELIAS) is the leading stellarator concept in Europe. Its prime example, Wendelstein 7-X, successfully started operation in 2015. Based on the 5-field-period symmetry, the HELIAS 5-B engineering design study emerged which is a stellarator power reactor concept designed for 3000MW fusion power.
The stellarator confines the hot plasma by external field...
Chiara Bustreo
(Consorzio RFX)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The cost of the electricity (COE) generated by a fusion power plant is a key driver for the technology future energy market deployment. Hence, the ongoing researches on the pulsed DEMO design optimization, taking into account the physical and technical constraints, are putting priorities on the minimization of the DEMO direct costs that indeed greatly influence the COE.
Also the duty cycle of...
Hyun Soo Tho
(Strategy Division)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
This paper is focused on the analysis of spillover benefits of the ongoing R&D program on nuclear fusion in Korea. The spillover effects are understood here as positive externalitiesof publicly funded R&D activities that may be revealed at the companies’ level in the form of newly created knowledge stock; development of innovative products/ processes with broader market applications;...
Alexander Rydzy
(FSN-FUSTEC-TES)
9/8/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Ever since the ENEA Fusion Department has been involved in the technology transfer of its knowledge in the field of nuclear fusion from the R&D scope to the execution of large projects together with industry, it has been outlined the importance of working by a quality management system (QMS) and of applying the principles of the Project Management. The head of the ENEA Fusion Department took...
