5-9 September 2016
Prague Congress Centre
Europe/Prague timezone
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Contribution List

Displaying 834 contributions out of 834
Type: Oral Session: I1.1 B.Bigot
Established by the signature of the ITER Agreement in November 2006, the ITER project is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. Supported by impress ... More
Presented by Bernard BIGOT on 5 Sep 2016 at 09:10
Type: Oral Session: I1.2 T.Klinger
The optimized stellarator Wendelstein 7-X (W7-X) has started with the goal to demonstrate steady-state plasma operation at fusion relevant plasma parameters. This is to establish the optimized stellarator as a viable fusion power plant concept. The design of W7-X is based on the optimization of the geometric properties of the magnetic field with the aim to minimize neoclassical transport losses in ... More
Presented by Thomas KLINGER on 5 Sep 2016 at 09:50
Type: Oral Session: I1.3 V.Tomarchio
JT-60SA is a superconducting tokamak developed under the Satellite Tokamak Programme of the Broader Approach Agreement between EU and Japan, and the Japanese national programme. It is designed to operate in the break-even conditions for long pulse duration (typically 100 s), with a maximum plasma current of 5.5 MA. Its scientific aim is to contribute at early realization of fusion energy, in suppo ... More
Presented by V. TOMARCHIO on 5 Sep 2016 at 11:15
Type: Oral Session: I2.1 S. Brezinsek
Since installation of the JET ITER-Like Wall more than 30h of plasma operation with the inertial cooled full W divertor took place. Successfully, the divertor plasma-facing components PFCs handled harsh tokamak conditions with (i) high surface temperature excursions passing the ductile-to-brittle temperature and re-crystallisation temperature multiple times, (ii) ITER-relevant steady-state and pea ... More
Presented by S. BREZINSEK on 6 Sep 2016 at 08:30
Type: Oral Session: I2.2 J. Li
The Chinese Fusion Engineering Testing Reactor (CFETR) is the next device for the Chinese magnetic confinement fusion (MCF) program which aims to bridge the gaps between the fusion experiment ITER and the demonstration reactor DEMO. CFETR will be operated in two phases: Steady-state operation and tritium self-sustainment will be the two key issues for the first phase with a modest fusion power up ... More
Presented by Jiangang LI on 6 Sep 2016 at 09:10
Type: Oral Session: I2.3 R. Panek
The COMPASS tokamak with ITER-like plasma shape has been put into operation in 2009 in Institute of Plasma Physics ASCR in Prague. It has been equipped by a comprehensive set of diagnostics for edge and Scrape-Off-Layer (SOL) plasma as well as by a new a system of two Neutral Beam Injectors (NBIs), which enabled to obtain significant results in the field of edge, SOL and divertor physics. In order ... More
Presented by Radomir PANEK on 6 Sep 2016 at 09:50
Type: Oral Session: I3.1 N. Mitchell
The magnet system is one of the critical core components of the ITER magnets, defining the machine capabilities to form and drive 15MA 500MW nuclear plasmas for 100s of seconds. The magnets, the largest superconducting magnet system ever built with 50GJ of stored energy, are also technologically highly advanced components using large composite Nb3Sn 4-6K force flow cooled conductors that also, in ... More
Presented by N. MITCHELL on 7 Sep 2016 at 08:30
Type: Oral Session: I3.2 R. Heidinger
Fusion road maps defined by both Europe and Japan, Parties to the Broader Approach Agreement (BA) where the IFMIF/EVEDA project is underway, have yet again confirmed the central need of a neutron source dedicated for fusion materials qualification. In the framework of the BA, engineering design and engineering validation activities are conducted which are targeted to prepare the foundations toward ... More
Presented by R. HEIDINGER on 7 Sep 2016 at 09:10
Type: Oral Session: I3.3 U. Fischer
The European Power Plant Physics and Technology (PPPT) programme, organised within the EUROfusion Consortium, aims at developing a conceptual design of a fusion power demonstration plant (DEMO) as a central element of the roadmap to the realisation of fusion energy. Various integrated PPPT projects are being conducted to meet this goal including Breeder Blanket (BB), Safety and Environment (SAE), ... More
Presented by U. FISCHER on 7 Sep 2016 at 09:50
Type: Oral Session: I4.1 R. Vila
With start of EUROfusion Materials-WP in 2014, functional materials (FM) have been included as a new branch. Their main scopes are issues of optical and dielectric materials for DEMO applications. R&D of these materials are, in particular, essential for Diagnostics and Heating and Current Drive (H&CD) systems that  must provide critical services such as machine control, protection, performance e ... More
Presented by Rafael VILA on 8 Sep 2016 at 08:30
Type: Oral Session: I4.2 V. Toigo
The realization of the ITER Neutral Beam Test Facility (NBTF) and the start the experimental phase are important tasks of the fusion roadmap, since the target requirements of injecting to the plasma a beam of Deuterium atoms with a power up to 16.5 MW, at 1MeV of energy and with a pulse length up to 3600s have never been reached together before. The ITER NBTF, called PRIMA (Padova Research on ITER ... More
Presented by V. TOIGO on 8 Sep 2016 at 09:10
Type: Oral Session: I4.3 H. Fuenfgelder
A enhanced impurity production has often accompanied experiments using ICRF (Ion Cyclotron Range of Frequency) as heating method. Positive effects, such as the capability to deposit the power centrally even at high density and thereby reduce the central impurity accumulation, were wiped out in the all‐metal ASDEX Upgrade when the antenna limiters were also coated with W. The hypothesis that this ... More
Presented by H. FUENFGELDER on 8 Sep 2016 at 09:50
Type: Oral Session: I5.1 O. Neubauer
The mission of Wendelstein 7-X is to assess the reactor capabilities of the HELIAS stellarator line. W7-X is equipped with superconducting coils (B=2.5 T) and is sufficiently large (V=30 m<sup>3</sup>3) to potentially attain steady-state plasmas at low collisionalities and high densities at the same time. As prerequisite for long-pulse operation, W7-X will employ high power, cw microwave heating ( ... More
Presented by Olaf NEUBAUER on 9 Sep 2016 at 08:30
Type: Oral Session: I5.2 E. Tsitrone
The WEST project is targeted at minimizing risks for ITER divertor procurement and operation. It consists in implementing an actively cooled tungsten divertor for testing the ITER divertor technology under tokamak conditions in Tore Supra. The present paper gives an overview of the project status, and describes the main lines of the associated research plan. As far as the project is concerned, the ... More
Presented by E. TSITRONE on 9 Sep 2016 at 09:10
Type: Oral Session: I5.3 R. Albanese
One of the main challenges in the European roadmap toward the realisation of fusion energy with a demonstration plant DEMO [1] is to develop a heat and power exhaust system able to withstand the large loads expected in the divertor. In parallel with the programme to optimise the operation with a conventional divertor based on detached conditions to be tested on ITER, efforts are being devoted to t ... More
Presented by R. ALBANESE on 9 Sep 2016 at 09:50
Type: Oral Session: I5.4 F. Warmer
One of the high-level missions of the European Roadmap to the realisation of fusion energy is to bring the HELIAS stellarator line to maturity. The near-term focus is the scientific exploitation of the Wendelstein 7-X experiment in order to assess stellarator optimization in view of economic operation of a stellarator fusion power plant. W7-X will play a decisive role for these studies but may be ... More
Presented by F. WARMER on 9 Sep 2016 at 11:00
Type: Oral Session: I5.5 P. Bastistoni
Within the framework of the EUROfusion programme, a work-package of technology projects (WPJET3) is being carried out in conjunction with the planned DT experiment at JET with the objective of maximising the scientific and technological return of DT operations in support of ITER. To this purpose, experiments, analyses and studies are performed in the areas of neutronics, neutron induced activation ... More
Presented by P. BATISTONI on 9 Sep 2016 at 11:40
Type: Oral Session: O1A
Track: H. Fuel Cycle and Breeding Blankets
Tritium behavior in a breeding blanket is a key design issue because of its impact on safety and fuel-cycle best performance. Nowadays there are only few references and any fully validated tool with predictive capabilities. Considering the difficulty in handling tritium and its fundamental role inside a fusion reactor, it is intended to prepare a simulation tool for tritium transport.In this work ... More
Presented by Elisabetta CARELLA on 5 Sep 2016 at 16:40
Type: Oral Session: O1A
Track: H. Fuel Cycle and Breeding Blankets
Any demonstration power reactor (DEMO), which applies solid breeder blankets, requires “advanced tritium breeders” with high tritium breeding ratios and increased stability at high temperatures. However, the fabrication techniques of advanced tritium breeder pebbles have yet to be established. Therefore, the R&D on the fabrication technologies of the advanced tritium breeders and the character ... More
Presented by Tsuyoshi HOSHINO on 5 Sep 2016 at 17:00
Type: Oral Session: O1A
Track: H. Fuel Cycle and Breeding Blankets
China has long been active in pushing forward the fusion energy development to the demonstration of electricity generation. Two experts’ meetings were organized in 2014 by Ministry of Science and Technology (MOST) to seriously discuss the China’s fusion roadmap in particular the design and construction of magnetic confinement fusion reactor beyond ITER. As one of the most challenging component ... More
Presented by Jie YU on 5 Sep 2016 at 17:20
Type: Oral Session: O1A
Track: H. Fuel Cycle and Breeding Blankets
Water-Cooled Lithium-Lead Breeding Blanket (WCLL) is considered a candidate option for European DEMO reactor. Starting from previous experiences in the frame of Power Plant Conceptual Studies within EUROfusion Consortium, , ENEA and its linked third parties have proposed and are developing a multi-module blanket segment concept based on DEMO 2015 specifications. The layout of the module is based o ... More
Presented by Alessandro DEL NEVO on 5 Sep 2016 at 17:40
Type: Oral Session: O1B
Track: A. Experimental Fusion Devices and Supporting Facilities
When designing a new large experimental device, extrapolation from current knowledge and rules into unexplored design space is unavoidable, and predicting the behaviour of a new device is therefore subject to significant uncertainties. This makes it difficult to determine an optimal design. For conceptual fusion power plants, a further concern is the large possible variation in expected plasma per ... More
Presented by Hanni LUX on 5 Sep 2016 at 16:40
Type: Oral Session: O1B
Track: A. Experimental Fusion Devices and Supporting Facilities
Extending high performance plasma discharge into long pulse steady-state operation is one of the urgent issues to be solved in preparing the ITER and fusion reactor. The KSTAR device is one of the best engineered superconducting tokamak devices which is good for exploring the science and technologies for the high performance steady-state operation due to lots of its unique features such as extreme ... More
Presented by Yeong-Kook OH on 5 Sep 2016 at 17:00
Type: Oral Session: O1B
Track: A. Experimental Fusion Devices and Supporting Facilities
After 10 years of operation since its major modification, an upgrade of the RFX-mod experiment is presently under design. The main objectives are the improvement of the control of magnetic confinement, plasma density and plasma wall interaction in both RFP and Tokamak configuration. The main design driver requirement for the improvement of the magnetic confinement control is the enhancement of the ... More
Presented by Simone PERUZZO on 5 Sep 2016 at 17:20
Type: Oral Session: O1B
Track: A. Experimental Fusion Devices and Supporting Facilities
The JT-60SA project implemented by Japan and Europe is progressing on schedule towards the first plasma in 2019. Spain (Ciemat) is in charge of the design and manufacturing of the cryostat. The JT-60SA cryostat is a stainless steel vacuum vessel (14m diameter, 16m height) which encloses the tokamak providing the vacuum environment (10<sup>-3</sup>-3 Pa). It must withstand the external atmospheric ... More
Presented by Jose BOTIJA on 5 Sep 2016 at 17:40
Type: Oral Session: O1C
Track: E. Magnets and Power Supplies
The ITER Central Solenoid (CS) is one of the critical elements of the machine. The CS conductor went through an intense optimization and qualification program, which included characterization of the strands, a conductor straight short sample testing in the SULTAN facility at the Swiss Plasma Center (SPC), Villigen, Switzerland, and a single-layer CS Insert coil recently tested in the Central Solen ... More
Presented by Nicolai MARTOVETSKY on 5 Sep 2016 at 16:40
Type: Oral Session: O1C
Track: E. Magnets and Power Supplies
Pulse Step Modulation (PSM) based High Voltage Power Supply (HVPS) are widely used in applications viz. Broadcast transmitters, Particle accelerators and Neutral Beam Injectors because of inherent advantages of modular structure, high accuracy and efficiency, low ripple and fast dynamics. Typical IC RF system composed of cascaded connection of Driver stage (70 kW RF output) and End stage (1500 kW ... More
Presented by N. P. SINGH on 5 Sep 2016 at 17:00
Type: Oral Session: O1C
Track: E. Magnets and Power Supplies
JT-60SA is a fusion experiment which is jointlyconstructed by Japan and Europe and which shall contribute to the earlyrealization of fusion energy, by providing support to the operation of ITER,and by addressing key physics issues for ITER and DEMO. In order to achievethese goals, the existing JT-60U experiment will be upgraded to JT-60SA byusing superconducting coils. The 18 TF coils of the JT-60 ... More
Presented by Walid ABDEL MAKSOUD on 5 Sep 2016 at 17:20
Type: Oral Session: O1C
Track: E. Magnets and Power Supplies
Various tests performed with full-size 60 kA HTS cable prototypes for fusion magnets in EDIPO test facility demonstrated that design of HTS strand proposed at Swiss Plasma Center – stack of HTS tapes twisted and soldered between two copper profiles –  is applicable for high-current fusion cables, but additional mechanical reinforcement is still needed. Based on experimentally obtained correla ... More
Presented by Nikolay BYKOVSKY on 5 Sep 2016 at 17:40
Type: Oral Session: O2A
Track: I. Materials Technology
Tungsten is the leading candidate material for plasma facing applications in future tokamak systems, due to its high melting point, good sputtering resistance and low activity after irradiation.  Despite this there has been a significant lack of study of the effect of transmutation products on the post irradiation mechanical behaviour of tungsten-based alloy systems.  This will be key to underst ... More
Presented by David ARMSTRONG on 6 Sep 2016 at 11:00
Type: Oral Session: O2A
Track: I. Materials Technology
Microstructural evolution and mechanical properties of Ti-bearing RAFM steels were investigated after aging at 550 °C for 0 ~ 1000 hr. All samples with Ti were prepared using vacuum induction melting furnace and hot rolling process, followed by heat treatment in normalizing and tempering. Microstructures including precipitates, fractured surfaces and cross-sectional microsturctures were observed ... More
Presented by Chang-Hoon LEE on 6 Sep 2016 at 11:20
Type: Oral Session: O2A
Track: I. Materials Technology
Tungsten is the main candidate material for the plasma facing components of future fusion devices. During operation, these components will be subject to severe conditions, involving both steady state and transient heat loads as well as high particle fluxes. These may lead to surface and structure modifications which influence their performance and lifetime. Therefore, it is necessary to study thes ... More
Presented by Jiri MATEJICEK on 6 Sep 2016 at 11:40
Type: Oral Session: O2A
Track: I. Materials Technology
Material issues pose significant challenges for future fusion reactors like DEMO. When using materials in a fusion environment a highly integrated approach is required. Cracking, oxidation and fuel management are driving issues when deciding for new materials. Neutron induced effects e.g. transmutation adding to embrittlement are crucial to material performance. Here advanced materials e.g. Wf/W o ... More
Presented by Jan Willem COENEN on 6 Sep 2016 at 12:00
Type: Oral Session: O2B
Track: F. Plasma Facing Components
The conceptual design of the European DEMO power reactor is under development as part of the EUROfusion Consortium. DEMO is a high fusion power, long-pulsed, tritium self-sufficient device, and hence amongst the most critical and high-risk technologies are the divertor and main chamber plasma-facing components (PFCs). These PFCs must operate reliably under an extreme surface heat and particle flux ... More
Presented by Thomas R. BARRETT on 6 Sep 2016 at 11:00
Type: Oral Session: O2B
Track: F. Plasma Facing Components
Tungsten is considered the main candidate material for the first-wall in DEMO for its high melting point, low erosion yield and low fuel retention. Nevertheless, it can cause a substantial safety issue in a loss-of-coolant accident (LOCA) in combination with air ingress into the plasma vessel, due to formation and evaporation of volatile neutron activated tungsten oxide. Self-passivating tungsten ... More
Presented by Tobias WEGENER on 6 Sep 2016 at 11:20
Type: Oral Session: O2B
Track: F. Plasma Facing Components
The main objective of the WEST (W Environment in Steady-state Tokamak) project is to fabricate and test an ITER-like actively cooled tungsten divertor to mitigate the risks for ITER. Concerning the others Plasma Facing Components (PFC), they will also be modified and coated with W to transform Tore Supra into a fully metallic environment. Solutions had been developed with three different suppliers ... More
Presented by Mehdi FIRDAOUSS on 6 Sep 2016 at 11:40
Type: Oral Session: O2B
Track: F. Plasma Facing Components
Lithium coating techonolgy and flowing liquid lithium limiter (Flili) have been applied on HT-7 tokamak and many significant results been obtained. A Flili for exploring lithium as potential plasma facing material was designed and manufactured for EAST tokamak, it is applied on the concept of the thin flowing flim which had been sucessfully tested in HT-7 tokamak. The Flili of EAST mainly compose ... More
Presented by Qingxi YANG on 6 Sep 2016 at 12:00
Type: Oral Session: O2C
Track: B. Plasma Heating and Current Drive
One of key missions of WEST (Tungsten (W) Environment in Steady-state Tokamak) is to pave the way towards the ITER actively cooled tungsten divertor procurement and operation. WEST PFC will operate in ITER conditions, i.e. with a heat flux on the divertor target of 10MW/m<sup>2</sup>2 during 1000s and 20MW/m<sup>2</sup>2 during a few tens of seconds. To achieve such heat flux levels, both Lower Hy ... More
Presented by Jean-Michel BERNARD on 6 Sep 2016 at 11:00
Type: Oral Session: O2C
Track: B. Plasma Heating and Current Drive
Acceleration of high-power-density negative ion beams of ~180 MW/m<sup>2</sup>2 have been achieved up to 60 s for the first time. Because the achieved power density was comparable to ITER accelerator, and accelerated energy density of 10800 MJ/m<sup>2</sup>2 is much higher than that for JT-60SA of 6500 MJ/m<sup>2</sup>2, this achievement is one of promising results to overcome common issues for th ... More
Presented by Atsushi KOJIMA on 6 Sep 2016 at 11:20
Type: Oral Session: O2C
Track: B. Plasma Heating and Current Drive
A new mechanism for driving current off-axis in high beta tokamaks using fast electromagnetic waves, called Helicons, will be experimentally tested for the first time in the DIII-D tokamak. This method is calculated to be more efficient than current drive using electron cyclotron waves or neutral beam injection, and it may be well suited to reactor-like configurations [1]. A low power (100 W) 476 ... More
Presented by Joseph TOOKER on 6 Sep 2016 at 11:40
Type: Oral Session: O2C
Track: B. Plasma Heating and Current Drive
As the next step for the fusion energy in China beyond ITER, the China Fusion Engineering Text Reactor (CFETR) aims to operate with duty time as 0.3~0.5, means that CFETR should operate at steady-state scenario. This provides a great challenge for the physical design of the heating the current driving system. In general, four different kinds of method as NBI, ECH, LHW and ICRH have been developed ... More
Presented by Defeng KONG on 6 Sep 2016 at 12:00
Type: Oral Session: O3A
Track: D. Diagnostics, Data Acquisition and Remote Participation
The ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations also known as gaps 3, 4, 5, and 6, complementing the information provided by the magnetic diagnostics. For gaps 4 and 6, the antennas are to be installed in-vessel between two blanket shield modules. The microwave ... More
Presented by Hugo POLICARPO on 6 Sep 2016 at 16:40
Type: Oral Session: O3A
Track: D. Diagnostics, Data Acquisition and Remote Participation
The ITER bolometer diagnostic shall provide the measurement of the total radiation emitted from the plasma, a part of the overall energy balance. About 500 lines-of-sight (LOS) will be installed in ITER observing the whole plasma from many different angles to enable reliable measurements and tomographic reconstructions of the spatially resolved radiation profile. The LOS are bundled in up to 100 i ... More
Presented by Hans MEISTER on 6 Sep 2016 at 17:00
Type: Oral Session: O3A
Track: D. Diagnostics, Data Acquisition and Remote Participation
One of the main research lines currently investigated within the FTU programs is the possibility to adopt a technology based on liquid metals as first plasma wall. More particularly, the main attention has been devoted to the analysis of plasma performances when using a liquid lithium limiter (LLL) device. The control of the limiter surface temperature reveals to be a fundamental aspect of the LL ... More
Presented by Claudia CORRADINO on 6 Sep 2016 at 17:20
Type: Oral Session: O3A
Track: D. Diagnostics, Data Acquisition and Remote Participation
The equatorial visible infrared wide angle viewing system (WAVS) is one of the key diagnostics in ITER aiming at the machine protection and plasma control. Those two main functions are achieved by means of infrared thermography and visible observation of the main plasma facing components. The diagnostic is composed of 15 lines of sight integrated in 4 equatorial port plugs allowing coverage of abo ... More
Presented by Laurent LETELLIER on 6 Sep 2016 at 17:40
Type: Oral Session: O3B
Track: A. Experimental Fusion Devices and Supporting Facilities
Radioactive waste arisings from JET operations are projected to contain approximately 25t of non-incinerable Intermediate Level Waste (ILW) with tritium levels > 12 kBq/g. This originates primarily from plasma facing components, specifically the divertor (MKIIa) used during the JET Deuterium Tritium Experiment in 1997 (DTE1). As current UK regulations do not allow off-site disposal of ILW and rest ... More
Presented by Stephen REYNOLDS on 6 Sep 2016 at 16:40
Type: Oral Session: O3B
Track: F. Plasma Facing Components
Helium is considered as coolant in the plasma facing first wall of several blanket concepts for DEMO fusion reactors, due to the favorable properties of chemical inertness, no activation, comparatively low effort to remove tritium, no chemical corrosion and a flexible temperature range. Design analyses for the ITER Test Blanket Modules done by several design teams have shown ability to use helium ... More
Presented by Frederik ARBEITER on 6 Sep 2016 at 17:00
Type: Oral Session: O3B
Track: I. Materials Technology
Flow Channel Inserts (FCI) are key elements in a Dual Coolant Lead Lithium blanket concept for DEMO, since they provide the required thermal and electrical insulation between the He cooled structural steel and the hot liquid Pb-15.7Li flowing at around 700°C, and minimize MHD pressure loss. FCIs must be inert in contact with Pb-15.7Li and show low tritium permeability. In addition, FCIs have to e ... More
Presented by Carlota SOTO on 6 Sep 2016 at 17:20
Type: Oral Session: O3B
Track: I. Materials Technology
Iron-base alloys are the leading candidate structural material for first-wall and blanket applications in near-term fusion devices, but their long-term viability to reliably function in the harsh fusion nuclear environment remains to be established. Helium produced by transmutation reactions interacts with microstructural features such as pre-existing dislocations, martensite lath boundaries, prec ... More
Presented by Charles HENAGER on 6 Sep 2016 at 17:40
Type: Oral Session: O3C
Track: G. Vessel/In-Vessel Engineering and Remote Handling
It is recognized that ITER will be the first nuclear installation where welding and cutting of pipes are performed routinely under Remote Handling conditions. Remote pipe maintenance tooling has been developed for JET, but conditions were such that manual deployment was permitted. Ultra-high vacuum class welding and cutting are highly skilled tasks and demand the precise control of parameters suc ... More
Presented by Luke THOMSON on 6 Sep 2016 at 16:40
Type: Oral Session: O3C
Track: G. Vessel/In-Vessel Engineering and Remote Handling
Befroe join ITER project fusion technologies development in China are focus on fusion device and plasma operation related. Components on fusion device installed, removed and maintained by personnel. Robotic technologies are never applied for fusion. China joined ITER from 2004. Scientists and engineers are involved in ITER related study and technologies development. Remote handling systems are imp ... More
Presented by Damao YAO on 6 Sep 2016 at 17:00
Type: Oral Session: O3C
Track: G. Vessel/In-Vessel Engineering and Remote Handling
As part of the programme to create a viable concept design for the Eurofusion DEMO powerplant, RACE is developing a concept design for the remote maintenance system. Within the DEMO tokamak, breeding blankets will require periodic replacement. In the current DEMO design this replacement will utilize the upper vertical ports at the top of the vacuum vessel. This operation will be challenging due to ... More
Presented by Jonathan KEEP on 6 Sep 2016 at 17:20
Type: Oral Session: O3C
Track: G. Vessel/In-Vessel Engineering and Remote Handling
Nuclear power plants require periodically maintenance, including the remote handling operations of transportation performed by automated guided vehicles (AGV). The navigation system becomes a key issue given the safety constrains of the heavy load to be transported in the complex scenarios, such as the reactor building. This work presents well-known and mature navigation technologies used by AGV i ... More
Presented by Rodrigo VENTURA on 6 Sep 2016 at 17:40
Type: Oral Session: O4A
Track: D. Diagnostics, Data Acquisition and Remote Participation
In magnetic fusion devices of the next generation such as ITER, high neutron and gamma-ray yields could be detrimental to the applied diagnostic equipment such as video imaging systems as well as to electronic components of machine control systems. Semiconductors devices are particularly sensitive to the radiation, both ionizing (formation of traps at the Si/SiO2 interface with energy levels withi ... More
Presented by Alexander HUBER on 8 Sep 2016 at 11:00
Type: Oral Session: O4A
Track: D. Diagnostics, Data Acquisition and Remote Participation
This paper describes the preliminary RAMI analysis for the ITER Low Field Side Collective Thomson Scattering (LFS CTS) system based on its preliminary architecture achieved at the System Level Design. The benefits and challenges involved in a RAMI analysis since the front end of the design process of the system are discussed together with the methodology pursued. The Functional Analysis, developed ... More
Presented by Elsa HENRIQUES on 8 Sep 2016 at 11:20
Type: Oral Session: O4A
Track: D. Diagnostics, Data Acquisition and Remote Participation
Abstract: The increasingly complex Physics experiments demand innovative digital Instrumentation for critical Measurement and Control functions. Requested system capabilities are, at least: high reliability, availability, maintainability, synchronized real-time high throughput data processing and compatibility to established Standards. Some of the methods that help attaining those capabilities ar ... More
Presented by Jorge SOUSA on 8 Sep 2016 at 11:40
Type: Oral Session: O4A
Track: D. Diagnostics, Data Acquisition and Remote Participation
All optical and laser diagnostics in ITER will use mirrors to observe the plasma radiation. In the severe ITER environment mirrors may become contaminated with plasma impurities hampering the performance of corresponding diagnostics. To counteract the mirror contamination, an in-situ mirror cleaning is proposed, which relies on ion sputtering the contaminants together with affected mirror material ... More
Presented by Andrey LITNOVSKY on 8 Sep 2016 at 12:00
Type: Oral Session: O4B
Track: F. Plasma Facing Components
Plasma-facing units equipped with tungsten (W) monoblock geometry are employed at the vertical targets of the ITER divertor. This contribution discusses a statistical approach for high heat flux (HHF) tests as potential quality assessment of the ITER divertor additional to the quality assurance performed by the manufacturer during the manufacturing. The IR analysis of the local temperature evoluti ... More
Presented by Henri GREUNER on 8 Sep 2016 at 11:00
Type: Oral Session: O4B
Track: F. Plasma Facing Components
ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European International Thermonuclear Experimental Reactor (ITER) development activities for the manufacturing of the inner vertical target (IVT) plasma-facing components of the ITER divertor. During normal operation the heat flux deposited on the bottom segment of divertor is 5-10 MW/m2 but the capability to remove up to 20 M ... More
Presented by Eliseo VISCA on 8 Sep 2016 at 11:20
Type: Oral Session: O4B
Track: F. Plasma Facing Components
An overview of recent Plasma-Material Interactions (PMI) research at DIII-D tokamak using the Divertor Material Evaluation Station (DiMES) is presented. The DiMES manipulator allows exposing material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by the extensive diagnostic suite including a num ... More
Presented by Dmitry RUDAKOV on 8 Sep 2016 at 11:40
Type: Oral Session: O4B
Track: F. Plasma Facing Components
The ITER first wall panels are exposed directly to thermonuclear plasma and must extract heat loads of about 2 MW/m² (EU) to 4.7 MW/m² (RF + CN). The panels are qualified through high heat flux cyclic testing before the installation in ITER. Initially the first wall panel prototypes will undergo full-power tests, this will be followed by the pre-series panels and finally the series panels. Th ... More
Presented by Jan PROKUPEK on 8 Sep 2016 at 12:00
Type: Oral Session: O4C
Track: A. Experimental Fusion Devices and Supporting Facilities
Abstract: Fusion energy becomes essential to solve the energy problem with the increase of energy demands. Although the recent studies of fusion energy have demonstrated the feasibility of fusion power, it commonly realizes that more hard work is needed on neutronics and safety before real application of fusion energy. A high intensity D-T fusion neutron generator is keenly needed for the researc ... More
Presented by Chao LIU on 8 Sep 2016 at 11:00
Type: Oral Session: O4C
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
As part of the conceptual design studies for a European DEMO, a programme of safety studies and analyses is performed, intended to help guide the design process by assessing the safety and environmental impact of design options under consideration. They also begin to prepare for the eventual licensing of DEMO construction and operation by a European nuclear regulator. A safety approach has been ad ... More
Presented by Neill TAYLOR on 8 Sep 2016 at 11:20
Type: Oral Session: O4C
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER. Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. The F4E, Amec Foster Wheeler and INL comprehensive methodology f ... More
Presented by Andrew GRIEF on 8 Sep 2016 at 11:40
Type: Oral Session: O4C
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
The ASTEC code has been recently extended to address the analysis of the main design basis accident scenarios in fusion installations, more particularly in the ITER facility. Current efforts are focused on loss of coolant accidents (LOCA) because a strong reactivity between beryllium toxic dust and steam leading to possible formation of gaseous beryllium oxide, hydroxide and hydride during the tra ... More
Presented by Francois VIROT on 8 Sep 2016 at 12:00
Type: Oral Session: O5A
Track: H. Fuel Cycle and Breeding Blankets
In the framework of the EUROfusion DEMO Programme and its work package Tritium-Matter Injection-Vacuum (TFV), the EU is preparing the conceptual design of the inner fuel cycle of a pulsed fusion DEMO. This contribution presents the current status of the project, addresses the most demanding challenges and shows first results. The project was started in 2014. The first one and a half years were dev ... More
Presented by Christian DAY on 8 Sep 2016 at 16:40
Type: Oral Session: O5A
Track: H. Fuel Cycle and Breeding Blankets
Although the D-T reaction is the most promising for fusion and is widely promoted, the amounts of tritium necessary to provide a sustainable fuel supply do not exist naturally. Besides the tritium must be self-sufficient operating a reactor, the initial fuel loading to start up any large-scale D-T fusion reactor remains a significant issue. We have examined the feasibility of starting a reactor fr ... More
Presented by Shanliang ZHENG on 8 Sep 2016 at 17:00
Type: Oral Session: O5A
Track: H. Fuel Cycle and Breeding Blankets
TRIPOLI-4® is a 3D continuous-energy Monte-Carlo particle transport code, developed by CEA, and devoted to shielding, reactor physics, criticality safety and nuclear instrumentation. TRIPOLI-4® is currently able to simulate four kinds of particles: Neutrons from 20 MeV down to 10<sup>-5</sup>-5 eV, Photons from 50 MeV down to 1 keV, Electrons and positrons from 100 MeV down to 1 keV. The TRIPO ... More
Presented by Fadhel MALOUCH on 8 Sep 2016 at 17:20
Type: Oral Session: O5A
Track: H. Fuel Cycle and Breeding Blankets
This work was performed by CEA within the framework of one specific contract concerning the development for ITER of manufacturing procedures for the industrial ATMOSTAT (ALCEN group) and Fusion For Energy (F4E). The HCLL-TBM (Helium Cooled Lithium Lead Test Blanket Module) box assembly development implies the welding development of the following components: the Box and the Stiffening Grid (SG) mad ... More
Presented by Olivier DOYEN on 8 Sep 2016 at 17:40
Type: Oral Session: O5B
Track: F. Plasma Facing Components
To qualify new plasma facing materials (PFM) and to evaluate the high heat flux performance under ITER or DEMO relevant loading conditions, extensive High Heat Flux (HHF) testing is indispensable. This includes performance tests under cyclic stationary thermal loads and screening of different material candidates under relevant transients such as Edge Localized Modes (ELMs) with high pulse numbers. ... More
Presented by Jochen LINKE on 8 Sep 2016 at 16:40
Type: Oral Session: O5B
Track: B. Plasma Heating and Current Drive
This paper presents the Radio-Frequency (RF) design of a new type of slow-wave Lower Hybrid Current Drive (LHCD) launcher, based on the Slotted Waveguide Antenna (SWA) concept, which is particularly attractive for the use in future magnetic fusion reactors. When compared to conventional LHCD slow-wave launchers, SWA are less obstructive, allow an “off-port” extension of the launcher and are pa ... More
Presented by Walid HELOU on 8 Sep 2016 at 17:00
Type: Oral Session: O5B
Track: B. Plasma Heating and Current Drive
Long term options for a steady state DEMO may require the availability of gyrotrons with an operating frequency above 200 GHz together with an RF output power of significantly more than 1 MW and a total gyrotron efficiency higher than 60 %. Fast frequency tuning in steps of around 2-3 GHz will be needed for control of plasma stability. Multi-purpose operation at leaps of about 30 – 40 GHz (e. ... More
Presented by John JELONNEK on 8 Sep 2016 at 17:20
Type: Oral Session: O5B
Track: B. Plasma Heating and Current Drive
The main advantages of Ion Cyclotron Resonance Heating and Current Drive (ICRH&CD) are its ability to achieve power deposition in the centre of the plasma column without any density limit along with direct heating of plasma ions. The challenge is then to couple large amount of power through the plasma boundary, where an evanesence layer has to be crossed, without exceeding the voltage standoff at ... More
Presented by Riccardo RAGONA on 8 Sep 2016 at 17:40
Type: Oral Session: O5C
Track: E. Magnets and Power Supplies
The Wendelstein 7-X stellarator (W7-X), one of the largest stellarator fusion experiments, is presently in the first operational phase at the Max Planck Institute for Plasma Physics in Greifswald, Germany. The W7-X shall prove the reactor relevance of the optimized stellarator concept. To confine 30m<sup>3</sup>3 plasma the W7-X machine has a superconducting magnet system with 50 non-planar and 20 ... More
Presented by Konrad RISSE on 8 Sep 2016 at 16:40
Type: Oral Session: O5C
Track: C. Plasma Engineering and Control
The Wendelstein 7-X safety control system is one of the main central control entities and ensures personal safety and investment protection. Its proper definition and setup has been a major precondition for the operation permit by the authorities and was inspected by external reviewers several times. The safety control systems has a distributed architecture comprising of the central safety system ... More
Presented by Andreas WERNER on 8 Sep 2016 at 17:00
Type: Oral Session: O5C
Track: C. Plasma Engineering and Control
ASDEX Upgrade came into operation in 1991. It was designed as a tokamak with reactor relevant shaping. The coil and control system allows to operate in lower single null (LSN), double null (DN) or upper single null (USN) with up to 1.6 MA plasma current and an initially open divertor configuration. Divertor enhancements were concentrated on the lower divertor that was finally transferred to a soli ... More
Presented by Albrecht HERRMANN on 8 Sep 2016 at 17:20
Type: Oral Session: O5C
Track: C. Plasma Engineering and Control
To collect meaningful experimental data, it is necessary to maintain consistent operating conditions in the tokamak plasma across repeated discharges. Presently, the desired plasma formation conditions, such as the shape of the plasma current profile, are achieved in a trial and error fashion, which can be a lengthy, wasteful process. In this work, model-based control techniques including optimal ... More
Presented by William WEHNER on 8 Sep 2016 at 17:40
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 1
A tee or an elbow behaves very differently from a straight pipe in resisting bending moment. When a straight pipe is bent, its cross section remains circular and the stresses increase linearly with distance from the neutral axis. However, when an elbow or a tee is bent, its cross section gets deformed into an oval shape. This geometrical deformity results in increased stresses, which are accounted ... More
Presented by Aditya SINGH on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 2
The purpose of the Upending Tool (UT) is to upend the vacuum vessel (VV) 40-degree sectors and the toroidal field coils (TFC) from horizontal delivery orientations to vertical assembly orientations. According to the ITER assembly procedure, this upending operation is carried out by four hooks of the tokamak crane. And the VV and TFC which are upended with UT are transfer from the UT to sector sub- ... More
Presented by Jinho BAE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 3
The Sector Sub-assembly Tool is a special tool for assembly of ITER Tokamak and is used to sub-assemble the 40° Tokamak sector which consists of vacuum vessel sector, vacuum vessel thermal shield sector and two toroidal field coils. The sector assembled in the assembly building is a basic and fundamental unit for the construction of the ITER Tokamak. Therefore, the design and structural integrity ... More
Presented by Min-Su HA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 4
A 600 W He refrigerator/liquefier with variable temperature supplies was constructed in National Institute for Fusion Science (NIFS) and its operation is started. Several cool-downs of large sized superconductors and magnets, such as a conductor of ITER TF coil and a JT-60SA superconducting coil, will be performed. The cooling performance is confirmed to meet its specifications. Two dummy heat loa ... More
Presented by Akifumi IWAMOTO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 5
This paper describes the analysis performed for the final design review of the ITER Gas Distribution System (GDS) manifolds to verify the system structural integrity. The GDS manifolds, which consist of Gas Fuelling (GF) manifold and Neutral Beam (NB) manifold, are complex combination pipes, of which gas supply lines and evacuation line are enclosed in a guard pipe. Based on the loading conditions ... More
Presented by Chengzhi CAO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 6
While the decisive feat of any concept is ‘successful implementable design’, the process of converting the concept into practically executable design is critical and challenging. It is usual to initiate any design on the basis of challenges visible during the conceptualization, as no project can really be a repeat of another. However, during conceptual design phase, it may not be possible to i ... More
Presented by Ajith KUMAR on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 7
ITER is an experimental fusion reactor being constructed in south of France which will demonstrate the scientific and technological capability in the direction of future commercial fusion power plant. The enormous amount of heat generated from the experimental reactor (mainly from the In-vessel components of Tokamak and its auxiliary systems) shall be removed by the Primary, Secondary and Tertiary ... More
Presented by Dinesh GUPTA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 8
The function of Gas Injection System<sup>[1]</sup>[1] (GIS), in ITER machine, is to deliver the fuelling and impurity gases into the torus. As an important sub-system of GIS, Fusion Power Shut-down System (FPSS) provides the function of emergency shut down for torus safety. The assessment of magnetic field in Tokamak building shows that a high stray field will exist in port cells during burning pl ... More
Presented by Zhiwei XIA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 9
The first cool down of the stellarator fusion experiment Wendelstein 7-X was achieved within 4 weeks in March 2015. A helium refrigerator with a cooling power of 7 kW at 4.5 K was used to cool down 456 tons of cold mass. The Outer Vessel (OV) of the cryostat contains 70 superconducting coils that are threaded over the twisted Plasma Vessel (PV). These coils are attached to a massive support struct ... More
Presented by Michael NAGEL on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 10
On 13<sup>th</sup>th February 2015 began the cool-down of about 450 tons cold mass of Wendelstein 7-X i.e. 70 superconducting magnets, 14 currents leads, massive support structure and the thermal shield, enclosed within a vacuum vessel of about 15.4 m outer diameter. After a smooth cool-down, the temperatures around 5 K, within the so called Short Standby Mode with the thermal shield return temper ... More
Presented by Chandra Prakash DHARD on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 11
Wendelstein 7-X (W7-X), went into operation in December 2015 at the Max-Planck-Institut für Plasmaphysik in Greifswald, Germany, is a modular advanced stellarator with a magnetic field optimized for good plasma confinement and stability [1]. The magnet system of W7-X consists of 70 superconducting coils - ten non-planar and four planar in each out of five modules of the machine. Preliminary simul ... More
Presented by Tamara ANDREEVA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 12
Wendelstein 7-X (W7-X) is a fusion device of the stellarator type with optimized magnetic field geometry and superconducting coils. The scientific goals of W7-X are to confirm the predicted improvement of the plasma confinement and to demonstrate the technical suitability of such a device as a fusion reactor. It is undergoing its first operation phase at the Max Planck Institute for Plasma Physics ... More
Presented by Sebastien RENARD on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 13
The Wendelstein 7-X stellarator started its first operational phase in October 2015 at the Max-Planck-Institute for Plasma Physics in Greifswald with the goal to verify that a stellarator magnetic confinement concept is a viable option for a fusion power plant. The main components of the W7-X cryostat system are the plasma vessel (PV), outer vessel (OV), 254 ports, thermal insulation, vessel suppo ... More
Presented by Paul VAN EETEN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 14
This contribution describes the electromagnetic and structural analysis of the new structural design of the COMPASS-U tokamak. The electromagnetic calculations solve force effects on tokamak coils using ANSYS Maxwell 3D code. The calculations were performed for three different combinations of excited coils and for two different plasma positions. The structural analysis was performed then using ANS ... More
Presented by David SESTAK on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 15
The design of the Chinese Fusion Engineering Test Reactor(CFETR) must integrate a great number of working documents and data from many groups, and distribute these materials to everyone in time, therefore, the parallel design work in different places could be properly managed, and the schedule, as well as the cost, could be ensured. An integration design platform has been built with this demand; ... More
Presented by Minyou YE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 16
The ramp up scenario design, which considers of both physics and engineering constrains, plays an important part in fusion device design. The Tokamak Simulation Code (TSC), coupling with some auxiliary heating codes, has been implemented in the CFETR system code to construct the workflow of the CFETR ramp up scenario designs. In this workflow, the CFETR geometric construction design and some preli ... More
Presented by Li LIU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 17
The HESEL code has been used to simulate scrape-off-layer (SOL) electrostatic interchange-driven low-frequency turbulence in various EAST tokamak discharges [1]. The recently installed Lithium Beam Emission Spectroscopy (LiBES) diagnostic system on EAST provides well resolved non-intrusive 2D measurements of SOL turbulence [2]. This paper presents results of comparison of statistical properties of ... More
Presented by Gergo POKOL on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 18
Various ways of filling the open magnetic trap with plasma are used in different experiments on study of plasma in order to develop methods of plasma heating and confinement, to study the interaction of electromagnetic waves with magnetoactive plasma etc. Among all existing methods the ultra high frequency (UHF) contactless methods are used frequently. We have proposed the method of filling the op ... More
Presented by Sulkhan NANOBASHVILI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 19
Increased cooling performance is eagerly required by the cutting edge engineering and industrial technology. Nanofluids have attracted considerable interest due to their potential to enhance the thermal performance of conventional heat transfer fluids. However, heat transfer in nanofluids is a controversial research theme as there is yet no conclusive answer to explain the underlying heat transfer ... More
Presented by Konstantinos KOULOULIAS on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 20
Fast ignition is one of the proposed ways to achieve high fusion energy gain in inertial fusion research. This scheme has an advantage that requirements of laser power and implosion process for ignition are not strict compared to that in central ignition. For a successful ignition, it is necessary to transport the energy of hot electrons to the imploded core effectively. Recently, it is found that ... More
Presented by Mayuko KOGA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 21
SPIDER experiment, currently under construction at the Neutral Beam Test Facility (NBTF) in Padua, Italy, is a full-size prototype of the ion source for the ITER Neutral Beam (NB) injectors part of the ITER project. The Ion Source and Extraction Power Supplies (ISEPS) for SPIDER are supplied by OCEM Energy Technology s.r.l. (OCEM) under a procurement contract with Fusion for Energy (F4E) covering ... More
Presented by Andrea ZAMENGO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 22
SPIDER (Source for the Production of Ions of Deuterium Extracted from RF plasma) is the 100keV Ion Source Test facility (presently under construction in the Neutral Beam Test Facility at Consorzio RFX premises, in Padua, Italy) representing the full scale prototype of the Ion Source (IS) for the ITER 1 MeV Neutral Beam Injector (NBI).  SPIDER Ion Source, polarized at -100kVdc Power Supply, is me ... More
Presented by Marco BOLDRIN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 23
The SPIDER Central Interlock is a centralized electronic system to coordinate the protection functions within the SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma), i.e. the full-ion source prototype of the ITER Neutral Beam Injector. Due to the system time requirements, the SPIDER Central Interlock has been implemented by using PLCs for the slow functi ... More
Presented by Cesare TALIERCIO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 24
The ITER project requires at least two Neutral Beam Injectors (NBIs), 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. A full-size negative ion source (SPIDER - Source for Production of Ion of Deuterium Extracted from RF plasma) and a prototype of the whole 1 MV ITER injector (MITICA - Megavolt ITER ... More
Presented by Nicola PILAN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 25
The ITER project requires at least two Neutral Beam Injectors (NBIs), 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 It ... More
Presented by Francesco FELLIN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 26
The construction of the new FULGOR test facility (Fusion Long Pulse Gyrotron Laboratory) at KIT is in full swing. This will significantly expand the experimental capabilities at KIT  to CW tests of high power gyrotrons of up to 4 MW ouput power at operating frequencies up to 240 GHz. Thus, this facility will be a significant platform for the verification of the performance of current CW gyrotro ... More
Presented by Martin SCHMID on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 27
The power handling of RF components can be limited   by a resonant process known as Multipactor effect. Multipactor can be fatal   to microwave systems in space communication payloads or in experimental fusion devices. Multipactor simulations   are used to predict voltage thresholds but the results highly depends on the electron emission properties of the RF components materials. Moreover, b ... More
Presented by Nicolas FIL on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 28
A neoclassical tearing mode (NTM) can be controlled by electron cyclotron current drive (ECCD). Up to now, ECCD with pulse modulated gyrotron operation at a duty of 50% have been done to drive current into only O-point of magnetic island of NTM. The fast directional switch have been developed for improving a stabilizing efficiency of NTM [1]. It makes the duty of ECCD system to 100% by switching b ... More
Presented by Mikio SAIGUSA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 29
During operation, the resonance cavity of a high power gyrotron experiences a very large heat load (>15 MW/m2), localized on a very short ( < 1 cm) length, where any thermal deformation should be carefully controlled to guarantee the gyrotron performance. Different strategies can be considered for the removal of the heat there, among which we focus here on the use of mini-channels drilled in the a ... More
Presented by Andrea BERTINETTI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 30
The stabilization of appearing MHD modes (NTMs, RWMs) is a key factor in optimizing tokamak operation towards fusion power production. In NTM control, the primary actuator is a confluence of focused electromagnetic wave beams, which are generated by high-power millimetre-wave sources (gyrotrons), transferred through waveguides and injected into the plasma by a controlled electromechanical launcher ... More
Presented by Christos TSIRONIS on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 31
SST-1 Tokamak employs Electron Cyclotron Resonance (ECR) assisted pre-ionization as an effective support towards low loop-voltage plasma start-up at fundamental (O-mode) and second harmonic (X-mode). A 42GHz 500KW 500ms ECR source is used for this purpose. In recent experimental campaigns in SST-1, several experiments have been carried out on ECR assisted pre-ionization, plasma start-up, possible ... More
Presented by Braj SHUKLA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 32
To carry out research related to electron cyclotron waves, 6 MW ECH systems including four 105 GHz/1 MW/2 s and two 140 GHz/1 MW/3 s units will be developed on the HL-2M tokamak being built in the first stage. Dual-frequency transmission lines with same components for the 105 GHz and 140 GHz systems are designed to make the fabrication easier. The corrugated waveguides are used to ensure the bandw ... More
Presented by Donghui XIA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 33
The JT-60SA tokamak is scheduled to start operations in 2019 to support the ITER experimental programme and to provide key information for the design of DEMO scenarios. The device will count on ECRH and NBI as auxiliary heating and EC operations are foreseen for EC assisted startup, EC Wall Cleaning (ECWC), bulk heating and current drive and MHD control, for example. 7 MW of total injected EC powe ... More
Presented by Alessandro MORO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 34
An ICRH antenna system is developed and will be attached to W7-X for the operational phase 1.2. An antenna box with two straps with surfaces adapted to the 3d LCFS in standard magnetic configuration (m/n=5/5), is located at the low field side in the equatorial plane. The antenna system is optimised for plasma heating and wall conditioning in presence of magnetic field. Each strap is connected via ... More
Presented by Bernd SCHWEER on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 35
The experimental devices ASDEX Upgrade (AUG) and Wendelstein‑7X (W‑7X) are both equipped with two neutral beam injectors each for plasma heating (up to 20 MW). Four large titanium sublimation pumps (TSPs) (4×1.5×0.2 m<sup>3</sup>3) in each injector provide proper vacuum conditions (below 10<sup>-2</sup>-2 Pa) during the 10 s beam pulse with a gas feed of up to 30 Pa×m<sup>3</sup>3/s. A ... More
Presented by Guillermo OROZCO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 36
Abstract: Wave heating in the Ion cyclotron range of Frequencies (ICRF) has been a method of choice for plasma heating in fusion research because of its flexibility, cost effectiveness and plug-to-power efficiency. A new three-strap ICRF antenna, designed for ASDEX Upgrade, and aiming to lower RF sheath by preventing undesirable currents induced in the antenna frame,  demonstrated experimentally ... More
Presented by Yang QING XI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 37
ASDEX Upgrade’s (AUG) neutral beam injection (NBI) is primarily designed for deuterium injection and delivers 20 MW heating power from two injectors with four beams each at 60 and 93 keV, respectively. As opposed to the cryosorption pumps of the JET NBI, the Ti getter pumps of the AUG NBI with a pumping speed of ~ 3×10<sup>6</sup>6 L/s for D2 do not pump helium at all, leaving only the conven ... More
Presented by Christian HOPF on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: B. Plasma Heating and Current Drive Board #: 38
One of the biggest challenges for a fusion reactor with magnetic confinement is the controlled removal of the heating power. ASDEX Upgrade (AUG) is the leading experiment in this area and investigates integrated solutions that combine high heating power and wall materials suitable for reactors. A measure of the challenge to remove the power in the divertor region is given by the normalized output ... More
Presented by Claus-Peter KASEMANN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 39
DEMO is aproposed demonstration fusion power plant which is under design. Fusion power, Pfus, has to be controlled at certain level to produce sufficient net electricity. However, this increases power through separatrix, Psep, and thus can produce excessive heat flux to the divertor which can lead to damage. Due to neutron radiation, the materials are even more susceptible to damage for a given h ... More
Presented by Filip JANKY on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 40
A project on the scale of DEMO requires a formal systems engineering approach. Mapping the interfaces, dependencies and relationships between subsystems permits an understanding of a conceptual design from a set of complementary and consistent perspectives. It also helps to prevent clashes and incompatibility between subsystems at a later stage of engineering design. The first stage of this work h ... More
Presented by Ian JENKINS on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 41
Recent DEMO physics study has focused on several issues raised from the JA Model 2014 concept. The concept is characterized by a fusion power of ~1.5 GW and a major radius of 8.5 m based on the technical assessments of divertor heat removal capability, overall tritium breeding ratio TBR > 1.05, full inductive ramp-up of plasma current, and so on. A problem is compatibility between divertor detachm ... More
Presented by Yoshiteru SAKAMOTO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 42
Controllability of output power is one of the essential requirements for DEMO. Fuel control is expected as primary knob for the fusion power control. Pellet injection is considered as primary fueling technique in DEMO as with the ITER. Difference of requirement for fueling system in DEMO compared to ITER comes from demand of larger output. It consequences requirement of more fueling efficiency to ... More
Presented by Shinsuke TOKUNAGA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 43
Tokamak plasmas, in low safety factor scenarios, are prone to magnetohydrodynamic (MHD) low m,n instabilities which may affect the energy and particle confinement time and possibly lead to disruptive plasma termination. In presently operating tokamaks high space resolution (~2cm) and high time resolution (0.01-0.1ms) Electron Cyclotron Emission (ECE) diagnostics are embedded in the control loop fi ... More
Presented by Natale RISPOLI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 44
The Magnet and Power Supplies system in JET includes a ferromagnetic core able to increase the transformer effect by improving the magnetic coupling with the plasma. The iron configuration is based on an inner cylindrical core and eight returning limbs; the ferromagnetic circuit is designed in such a way that the inner column saturates during standard operations [1]. The modelling of the magnetic ... More
Presented by Francesco PIZZO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 45
Robust high performance plasma scenarios are being developed to exploit the unique capability of JET to operate with Tritium and Deuterium. In this context, real time control schemes are used to guide the plasma into the desired state and maintain it there. Other real time schemes detect undesirable behaviour and trigger appropriate actions to assure the best experimental results without unnecessa ... More
Presented by Morten LENNHOLM on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 46
In the present RF-driven (ECCD) steady-state plasma on QUEST (Bt = 0.25 T, R = 0.68 m, a = 0.40 m), plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to plasma current fitting (PCF) method. The current in the open magnetic surface seems due to orbit-driven current by high-energy particles in RF-driven plasma.  So bas ... More
Presented by Kazuo NAKAMURA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: C. Plasma Engineering and Control Board #: 47
Merging compression startup, pioneered on START, is a successful and robust method for plasma breakdown and plasma current startup which does not involve a solenoid. Tokamak Energy is currently constructing a relatively small (R~0.4m) high toroidal field (BT>2T) spherical tokamak (aspect ratio ~ 1.8) called ST40 which will have ~2MA of plasma current. A consequence of the ambitiously high toroidal ... More
Presented by Peter BUXTON on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 48
The main objective of this work is to demonstrate that a digital integrator based on the chopper modulation concept is capable of meeting the ITER requirements. The ITER magnetics diagnostic requires a maximum drift of 500 uV.s/hour, among other specifications, for the respective signal integrators. As of today, known COTS integrator modules do not fully comply simultaneously with all ITER require ... More
Presented by Antonio BATISTA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 49
The final design of the steady-state sensor diagnostic, developed collaboratively by ITER Organization and IPP Prague, is presented. The steady-state sensors – a subsystem of the ITER magnetic diagnostics – will contribute to the measurement of the plasma current, plasma-wall clearance, and local perturbations of the magnetic flux surfaces near the wall. The diagnostic consists of an array of ... More
Presented by Martin KOCAN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 50
Hall sensors with their small dimensions, simple principle of operation, and large dynamic range offer an attractive non-inductive method of magnetic field measurements for future fusion reactors operating in steady state regime. The applicability of commercially available Hall sensors, which are based on semiconductor sensing layer, is strongly limited by insufficient range of operational tempera ... More
Presented by Ivan DURAN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 51
A prototype electronics for the ITER ex-vessel steady state magnetic field metallic Hall sensors based on the analog lock-in signal processing with dynamic quadrature offset cancelation was developed and tested. Testing was carried out on Bismuth Hall sensors placed in the SAMM test assembly. The magnetic coils are used for measuring the magnetic field of the fusion reactor conventionally. However ... More
Presented by Slavomir ENTLER on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 52
The Plasma Position Reflectometry (PPR) diagnostic will be used in ITER to measure the plasma position/shape in order to provide a reference for the magnetic diagnostics during very long (>1000s) pulse operation, where the position deduced from the magnetics is known to be subject to substantial error. It consists of five reflectometers distributed at four locations, known as gaps 3-6, operating i ... More
Presented by Jorge BELO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 53
ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations also known as gaps 3, 4, 5, and 6, complementing the information provided by magnetic diagnostics. For gaps 4 and 6, the antennas are to be installed in-vessel between two blanket shield modules. The microwave signal ... More
Presented by Paulo QUENTAL on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 54
The future nuclear fusion power plants will require Electron Cyclotron Heating and Current Drive (ECH&CD) systems to heat up and stabilize the plasma inside the vacuum vessel. One of the key components of such systems is the Chemical Vapor Deposition (CVD) diamond window. The purpose of this device is to act as vacuum and tritium boundary while providing a high microwave transparency with minimal ... More
Presented by Francesco MAZZOCCHI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 55
Scintillator based fast-ion loss detectors (FILD) are used in virtually all major tokamaks and stellarators to study the fast-ion losses induced by magnetohydrodynamic (MHD) fluctuations. FILD systems provide velocity-space measurements of fast-ion losses with alfvenic temporal resolution. This information is crucial to identify the MHD fluctuations responsible for the actual fast-ion losses and t ... More
Presented by Juan AYLLON on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 56
As part of ITER's fusion diagnostic systems, metal foil – miniaturised metal resistor type bolometer cameras are envisaged to provide the measurement of the total plasma radiation. For this kind of bolometer sensor the temperature of a measurement and a reference absorber is realised by metallic meanders on their back side, which are combined in an electrical configuration of a Wheatstone bridge ... More
Presented by Gabor NADASI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 57
The ITER bolometer diagnostic will have to provide accurate measurements of the plasma radiation in a varying thermal environment of up to 250°C. Current fusion experiments perform regular in-situ calibration of the detector properties, assuming stable calibration parameters within short discharge times, e.g. 10 s on ASDEX Upgrade. For long-pulse fusion experiments, e.g. W7-X, the diagnostic is o ... More
Presented by Florian PENZEL on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 58
In ITER, like in any fusion reactor, the plasma-wall interaction is unavoidable. It leads to material erosion and potential re-deposition or other surface morphology changes, as well as dust formation and tritium retention. The decision to start ITER operations with a full-W divertor has significantly reduced the expected erosion of the divertor target making observation of the target during disch ... More
Presented by Nancy AGEORGES on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 60
The assessment of the Shutdown Dose Rate (SDR) due to neutron activation is a major safety issue for fusion devices and in the last decade several benchmark experiments have been conducted at JET during Deuterium-Deuterium shutdown for the validation of the numerical tools used in ITER nuclear analyses. The future Deuterium-Tritium campaign at JET (DTE-2) will provide a unique opportunity to valid ... More
Presented by Nicola FONNESU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 61
The Neutron Camera is a Joint European Torus (JET) diagnostic with the main function of measuring the 2.5 MeV (DD) and 14 MeV (DT) neutron emissivity profile over a poloidal plasma cross-section using line-integrated measurements along a number of collimated channels (lines-of-sight, LOS).  Measurements are performed using two detectors: NE213 liquid scintillators (DD, low power DT) and BC418 pl ... More
Presented by Marco RIVA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 64
The signal of a neutron detector can be divided into an unscattered and a scattered component. In fusion, the unscattered, direct component reaches the detector directly from the fusion plasma. The scattered neutrons, on the other hand, reach the detector after interacting with some of the materials in the fusion device. More specifically, the backscatter component is defined as the signal from ne ... More
Presented by Federico BINDA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 66
The second experimental deuterium-tritium (DT2) campaign is planned at JET in 2019. Acalibration of the JET neutron emission monitoring system, consisting of fission chambers (KN1) and of an activation system (KN2), will be carried out with a compact deuterium-tritium neutron generator (NG) with suitable intensity (≈5x10 8 n/s). The accuracy goal for this calibration is <10% uncertainty at 14 Me ... More
Presented by Axel KLIX on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 67
Abstract:Beam Emission Spectroscopy (BES) diagnostic based on neutral beam injection (NBI) has recently been developed in EAST tokamak. A 128-channel Hamamatsu S8550 APD detector array is chosen as the core device. Three cavity interference filter with a center frequency of 659.33nm and a bandwidth of 1.59nm is used to eliminate the interference Dα signal and carbon impurities radiation. This B ... More
Presented by H.J. WANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 68
H-mode is the main operation mode in the future fusion reactor and L-H transition is one of the concerning issue of H-mode research[1]. Much effort has been made on the research of L-H transition, however, the detail characters of the L-H transition need more research to afford reference for the optimization of H-mode plasma discharge [2-4]. An infrared(IR)/visible endoscope system was built on th ... More
Presented by Bo SHI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 69
The WEST (W Environment for Steady-state Tokamak) [1] project consists in a major upgrade of the superconducting medium size tokamak Tore Supra to minimize risks for the ITER divertor procurement in terms of cost, delays and performance. This modification consists in changing the present circular magnetic configuration to a divertor configuration and implementing an ITER like actively cooled Tu ... More
Presented by Jean-Marcel TRAVERE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 70
The WEST project consists in a major upgrade of the superconducting medium size tokamak Tore Supra to minimize risks for ITER divertor procurement and operation. This modification consists in changing the present circular magnetic configuration to a divertor configuration and implementing an ITER like actively cooled Tungsten divertor. Heat load on divertor target will range from a few MW/m<sup>2< ... More
Presented by Philippe MOREAU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 71
For the long-pulse high-confinement discharges in future tokamaks, the equilibrium of plasma requires an interaction and energy exchange with the first wall materials. The heat flux resulting from this interaction is of the order of 10 MW/m<sup>2</sup>2 for steady state conditions and up to 20 MW/m<sup>2 </sup>2 for transient phases. As a result, surface temperature measurement of the plasma facin ... More
Presented by Chen ZHANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 72
JT-60SA Thomson scattering system will measure electron temperature and density profile. A YAG laser will be toroidally injected to the JT-60SA on its equatorial plane. If the beam profile changes from flat-top to peaked profile, the laser beam breaks the vacuum window. Thus, we designed beam transfer optics as long as ~50 m using a relay image technique. The beam transfer optics designed for the ... More
Presented by Hiroshi TOJO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 73
JT-60SA, which has fully super conducting coils, is designed and now being constructed for demonstrate and develop steady-state high beta operation in order to supplement ITER toward DEMO.  In order to obtain the information for the control and the physics research on JT-60SA plasma, we developed the many types of magnetic sensors.  Compared to JT-60U, JT-60SA needs larger magnetic sensors and l ... More
Presented by Manabu TAKECHI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 74
The JT-60SA superconducting tokamak is proposed to be equipped with a Lithium Beam Emission Spectroscopy (LiBES) and Deuterium Beam Emission Spectroscopy (DBES) diagnostic systems. The purpose of the LiBES system is SOL and plasma edge density profile measurements and density fluctuation measurements in the SOL and outer edge regions, whereas the DBES system on the heating beams would have the cap ... More
Presented by Ors ASZTALOS on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 75
In order to extend the operational space of RFX-mod in both RFP and Tokamak configurations, a major refurbishment of the load assembly is under study. It includes the removal of the vacuum vessel to increase the plasma-shell proximity and modifications of the support structure to obtain a new vacuum-tight chamber. This entails the design of a new electromagnetic measure system, taking into account ... More
Presented by Giuseppe MARCHIORI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 77
Optical emission spectroscopy with inversion process is used to obtain local emission spectrum from line integrated spectra. Tomographic inversion techniques are widely used with complicated noise reduction and sufficient viewing line of sights. On the other hand, optical probe has advantage of direct measurement although it may lead to plasma perturbation. An optical probe with outer diameter of ... More
Presented by Jae-young JANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 78
Helium transport study is essential in burning plasma to prevent fuel dilution from the helium ash accumulation. Charge exchange spectroscopy (CES) is widely used to measure impurity density as well as toroidal rotation and ion temperature. Single-handed CES system have a low accuracy in impurity density measurement due to the large errors in absolute intensity calibration and neutral beam modelli ... More
Presented by YooSung KIM on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 79
A Thomson scattering(TS) system is developed and commissioned for measuring and analyzing spatial profiles of electron temperature(Te) and density(Ne) of Versatile Experiment Spherical Torus(VEST). Since the estimated Ne of VEST plasma is ~5x10<sup>18</sup>18m<sup>-3</sup>-3 which is lower than typical Ne in other tokamaks, each part of the system is carefully designed to maximize the number of co ... More
Presented by Young-Gi KIM on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 80
The combined system of Charge Exchange Spectroscopy (CES) and Beam Emission Spectroscopy (BES) will be developed in Versatile Experimental Spherical Torus(VEST).  to measure ion temperature and rotation velocity by not using impurity but fuel hydrogen ion emission line directly. In order to use this system, Diagnostic Neutral Beam Injection (DNBI) system is necessary to supply high energy neutral ... More
Presented by Kihyun LEE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 81
The degradation of transport current property by the high mechanical strain on the practical Nb3Sn wire is serious problem to apply for the future fusion magnet operated under higher electromagnetic force environment beyond ITER. Recently, we approached to the solid solution ternary Cu-Sn (Cu-Sn-X) matrices for the development of the high mechanical strength bronze processed Nb3Sn wires. Generally ... More
Presented by Yoshimitsu HISHINUMA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 82
It is accepted that plasma exhaust is a major challenge for DEMO and future power plants and the reference approach is to use a design similar to JET and ITER. There is not yet full confidence this will extrapolate successfully and be compatible with a maximum power flux of 5-10 MWm<sup>-2</sup>-2 on the Plasma Facing Components. Detachment provides an attractive solution to the power exhaust prob ... More
Presented by Simon MCINTOSH on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 83
Current models used for thermal–hydraulic analyses of forced-flow superconducting cables used in fusion technology, such as e.g. Cable-in-Conduit Conductors, are typically 1-D and they require reliable predictive correlations for the transverse mass-, momentum- and energy transport processes occurring between the different cable components in order to reliably assess any fusion magnet design in ... More
Presented by Aleksandra DEMBKOWSKA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 84
In the European path towards the tokamak reactor DEMO, led by the EUROfusion consortium with the aim of demonstrating electricity production by fusion energy by 2050, the Toroidal Field Coils are under conceptual design. Three different winding pack (WP) options have been proposed by different European parties. In this paper, we consider the ENEA proposal, featuring a layer-wound WP with graded su ... More
Presented by Alberto BRIGHENTI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 85
Since the year 2013, the Swiss Plasma Center (SPC) has proposed a Toroidal Field (TF) layout for the DEMO- EUROFusion tokamak, based on a graded winding pack made of layers of Nb3Sn (react-and-wind) and NbTi conductors. In summer 2015, a new reference baseline is issued for the DEMO- EUROFusion tokamak, leading to an update of the TF coil requirements, e.g. the operating current has been reduced f ... More
Presented by Boris STEPANOV on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 86
A reliable and realistic cost estimate is of paramount importance for the management of large projects, to assist the budget and planning phases. In the case of DEMO, the cost estimate helps driving the selection among competing design options. The achievement of a target construction price < 2 B€ for a 500 MWe fusion power plant is a necessary condition in order to sell electricity to the mar ... More
Presented by Pierluigi BRUZZONE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 87
Three alternative designs of the toroidal field (TF) coil were proposed for the European DEMO being developed under the Eurofusion Consortium. The most ambitious TF coil winding pack in terms of technological deviation from the ITER TF coil design and consequent potential cost saving, the so-called WP1, is based on the react&wind technology of Nb3Sn layer-wound flat multistage conductors. We prese ... More
Presented by Kamil SEDLAK on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 88
In the framework of the EUROfusion DEMO project, studies are conducted in several European institutions for designing the tokamak magnet systems. In order to generate the high magnetic fields required for the plasma confinement and control, the reactor should be equipped with superconducting magnets, the reference design being based on Cable-In-Conduit Conductors cooled at cryogenic temperatures b ... More
Presented by Quentin LE COZ on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 89
The present study aims to minimise the outer radius of the CS coil of European DEMO in order to reduce the size and the cost of the whole tokamak. In a previous study, it has been demonstrated that the outer radius of the CS coil can be reduced maintaining the generated magnetic flux at 320 Vs using high-temperature superconductors (HTS). This first study was based on a uniform current density in ... More
Presented by Rainer WESCHE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 90
Successful operation of Demonstration Reactors is a key step in the fusion development. The structural integrity of the superconducting magnets producing high magnetic fields that are crucial for optimization of a fusion reactor performance must be ensured. Combinations of calculation approaches, reasonable modelling simplifications and clever prioritization at each analysis phase facilitate desig ... More
Presented by Anatoly PANIN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 91
Tokamak toroidal field coils (TFCs) characterized by a tilting in the azimuthal direction lead to several potential advantages, most notably the relieving of the stresses in the most critical area at the inboard side. As a consequence, much of the heavy steel structures needed to withstand the huge electromagnetic forces in conventional magnets can be reduced. Mechanically unloading the TFCs makes ... More
Presented by Renato GATTO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 92
SST-2 is a medium size fusion reactor machine under design at Institute for Plasma Research, India. It is being planned to operate between 100-300 MW of fusion power with main objectives of breeding of Tritium, Tritium handling studies and as a test bed for materials and components. SST-2 physics requirements of toroidal field Bt = 5.42 T at plasma major radius R = 4.42 m and the maximum allowable ... More
Presented by Aashoo SHARMA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 94
Presently, the Tokamak T-15MD (T-15U) is being built. All elements of the magnet system have been manufactured by the end of 2015. The magnet system of the Tokamak T-15MD will obtain and confine the hot plasma in the divertor configuration. The tokamak T-15MD magnet system includes the toroidal winding, the poloidal magnet system and supporting structures. The toroidal winding consists of 16 D-sha ... More
Presented by Petr KHVOSTENKO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 95
Spherical Tokamaks used in magnetic fusion have a small centre stack by design.  This causes a very high field on the conductor.  ST40 is a 3 Tesla spherical tokamak with a major radius of R=40cm and minor radius of a=26cm being built by Tokamak Energy. The high toroidal field (TF) requirement requires a wire current of 250kA flowing in each of the 24 limbs totalling 6 MA in the centre stack. Jo ... More
Presented by Bill HUANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: E. Magnets and Power Supplies Board #: 96
High-Temperature Superconductor (HTS) material REBCO has high critical currents even in high magnetic fields. The use of such material for future fusion magnets was already proposed in 2004, but the aspect ratio of REBCO, which is available as thin tapes only, made the realization of a high current cable in the current range of several 10 kA at magnetic fields around 12 T difficult. In the last ye ... More
Presented by Walter H. FIETZ on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 101
In the last years, W and W-Ti and W-V alloys, with grain sizes of hundreds of nanometers and densification very close to 100%, have been produced following a powder metallurgy route that consists of mechanical alloying and consolidation by hot isostatic pressing (HIP). In spite of the submicron-grained microstructure, and the dispersion of second phase nanoparticles, these alloys do not exhibit a ... More
Presented by Angel MUNOZ on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 103
Tungsten and Cu-alloys are currently proposed as reference candidate material for ITER first wall and divertor. Tungsten is proposed for its high fusion temperature and Cu-Cr-Zr alloys for their high thermal conductivity together good mechanical properties.  However its behavior under the extreme irradiation conditions as expected in ITER or DEMO is still unknown. Due to the determinant role of H ... More
Presented by Fernando MOTA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 104
The exhaust of power and particles is regarded as a major challenge in view of the design of a nuclear fusion demonstration power plant (DEMO). In such a reactor, highly loaded plasma facing components (PFCs), like the divertor targets, have to withstand both severe high heat flux (HHF) loads and considerable neutron irradiation. Existing divertor target designs, as e.g. the ITER-like monoblock co ... More
Presented by Alexander VON MULLER on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 105
Joining of armor material tungsten to other alloys and especially to copper components which will act as heat sinks in divertor application showed lacks due to the restricted miscibility of tungsten and copper. This negative behavior leads to bad or missing metallurgical W – Cu reactions with the consequence of reduced mechanical stability or high risks of cracking if any joining was realized. I ... More
Presented by Wolfgang KRAUSS on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 106
Although high room temperature strength (300-1000 MPa) and conductivity (200-360 W/m-K) have been achieved in Cu alloys, these alloys suffer significant thermal creep deformation at temperatures above 300-400oC. Deformation analysis indicates dislocation creep and grain boundary sliding are occurring. Design requirements for improved high-performance copper alloys are: 1) thermally stable microstr ... More
Presented by Steven ZINKLE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 107
The ITER operation program, as well as the DEMO operational, foresees for the vertical targets strike point region high steady state thermal fluxes that can be sustained only by components designed and manufactured accordingly. Their life-time is limited mainly by thermal fatigue caused by cyclic thermal loads inducing high mechanical stresses.The Plasma Facing components of the ITER divertor are ... More
Presented by Selanna ROCCELLA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 108
Fusion device materials have been modified over the years for the main aim of using optimal materials in ITER fusion device. Post-mortem analysis of materials used in JET provides valuable information for further material development and improvements required. One of key fusion device elements is the divertor. It minimizes plasma contamination and draws a big part of thermal and neutron load in th ... More
Presented by Mihails HALITOVS on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 109
Application of liquid lithium as a plasma facing material has some features proved by a lot of experiments with lithium devices in plasma accelerators KSPU, MK-200UG and “Plasma focus” facility. Then, the experiments carried out in operating tokamaks and stellarator (NSTX, FTU, T11-M, EAST, TJ-II) using liquid lithium and lithium CPS as intrachamber devices have shown the advisability of lithi ... More
Presented by Timur KULSARTOV on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 110
The use of bare Reduced Activation Ferritic Martensitic (RAFM) steels has been proposed for the first wall in a reactor [1]. Thus, it is necessary to understand the performance of RAFM steels under fusion-relevant condition. To date, the effects of simultaneous irradiation of hydrogen isotopes and He in F82H haven’t been examined in detail. We previously examined hydrogen retention properties, a ... More
Presented by Koki YAKUSIJI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 111
ASDEX Upgrade (AUG) is the only tokamak in Europe to have low activation ferritic steel in the inner vessel wall. The project is a first step towards the extensive use of ferritic steel in future fusion reactors. The ‘ad hoc’ ferritic steel built with low activation capability is the so called Eurofer. As the low activation property is not a requirement for AUG, the material selected for the p ... More
Presented by Irene ZAMMUTO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 112
The design of the demonstration fusion reactor DEMO presents challenges beyond those faced by the ITER project and may require the implementation of different solutions. One of the biggest challenges is managing the heat flux to the main chamber wall. The presently predicted total heating power in DEMO is more than 3 times that predicted for ITER value, while the major radius is only 1.5 times gre ... More
Presented by Francesco MAVIGLIA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 113
Yu. Igitkhanov, R. Fetzer and B. Bazylev Karlsruhe Institute of Technology (KIT), Karlsruhe, Germany juri.igitkhanov@partner.kit.edu The first assessments has shown that the edge localized modes (ELM) in the fusion power plant DEMO will pose a severe tread to the plasma facing components (PFC) by causing a surface melting and erosion [1]. In this work we estimate the degree of the ELM mitigation ... More
Presented by Yuri IGITKHANOV on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 114
The DEMO device is expected to operate in H-mode. On the other hand it is postulated that the divertor power load cannot exceed 5MW/m<sup>2 </sup>2 . In case of liquid divertor, vaporizing additionally enhances the plate material flux into the bulk. Impurities with large atomic number (Z) dilute the plasma core less, however, they radiate more in the core than those with smaller Z. Liquid tin (Sn) ... More
Presented by Michal PORADZINSKI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 115
Handling of the huge power exhausting from the core region to the SOL/divertor region is one of the crucial issues for a DEMO reactor design. In previous study for JA compact DEMO concept, SlimCS (a major radius of 5.5m), numerical simulation by an integrated divertor codes SONIC showed the divertor target heat load of < 10 MW/m<sup>2</sup>2 for the fusion power of < 1.5 GW and the large impurity ... More
Presented by Kazuo HOSHINO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 116
After the preliminary exploring phases for devising initial design concepts and performing design studies, the divertor project (WPDIV) of the EUROfusion consortium is currently entering into the final stage of the first half R&D round which is planned to be completed by the end of 2016. The core missions of WPDIV are to deliver feasible pre-conceptual design solutions for the divertor of an early ... More
Presented by Jeong-Ha YOU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 117
DEMO development is currently in the Pre-Conceptual Design Activity and the Divertor that is in charge of power exhaust and removal of impurity particles represents the key in-vessel component, with its Plasma Facing Units (PFU) exposed to the plasma and hence subjected to very high heat loads. During 2015 the integrated R&D project launched in the EUROfusion Consortium  studied how to approach a ... More
Presented by Fabio CRESCENZI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 118
The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. The operational reliability of the divertor target relies essentially on the structural integrity of the component, in particular, at material interfaces, where thermal stresses tend to be concentrated and thus cracks are most likely to initiate. In this context, the qualit ... More
Presented by Franklin GALLAY on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 119
In the framework of the work package "Divertor" of the EUROfusion action, a research campaign has been jointly carried out for the subproject "Cassette design and integration" by ENEA and University of Palermo to investigate the thermal-hydraulic performance of the DEMO divertor cassette body cooling system. A comparative evaluation study has been performed considering the different options of div ... More
Presented by Eugenio VALLONE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 120
In the framework of the work package "Divertor" of the EUROfusion action, a research campaign has been jointly carried out for the subproject "Cassette design and integration" by ENEA and University of Palermo to investigate the thermal-hydraulic performance of the DEMO divertor cassette cooling system. A comparative evaluation study has been performed considering three different options of coolin ... More
Presented by Silvia GARITTA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 121
East Advenced Superconducting Toakmak (EAST) is a superconducting magnet toakmak and its goal is to achieve the magnetic confinement fusion. The major plasma disruption(MD) or the vertical displacement event(VDE) all will produce toroidal eddy current in the vacuum vessel(VV) with plasma facing components(PFCs) and cause mechanical forces, which represent one of the most vital loads for tokamak. T ... More
Presented by Sumei LIU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 122
A medium sized Tokamak HL-2M is being designed and constructed in Southwestern Institute of Physics of China. This device can be operated with high plasma current 2.5 MA and toroidal magnetic field 3 T. Advanced divertor configurations with snowflake, tripod etc. are envisaged to study the divertor physics under high heating power and high core plasma performance operation. To accommodate the vari ... More
Presented by Lijun CAI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 123
The China Fusion Engineering Testing Reactor (CFETR) aims at bridging the gap between ITER and DEMO. Its scientific mission is to produce fusion power of 200 MW with tritium self-sustention and duty cycle of 0.3-0.5. The big fusion power and the auxiliary heating power of 100-140 MW, makes the design of CFETR divertor challenging. Previous work focuses on the plasma configuration and the first rou ... More
Presented by Xuebing PENG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: F. Plasma Facing Components Board #: 124
The Chinese Fusion Engineering Test Reactor (CFETR) is under design. Divertor is the most pivotal PFC to manage power and He ash exhaust. Based on the main goal of CFETR, it has a similar P/R~14 MW/m to ITER. Impurity seeding has been considered a promising means to enhance the radiation from the plasma edge and hence to reduce the target heat load, especially on carbon-free wall conditions. We ha ... More
Presented by Xiaoju LIU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 125
ITER (Nuclear Facility INB-174) Vacuum Vessel is divided into 9 similar sectors where In-Vessel Diagnostics and Operational Instrumentation are located and which require the provision of Electrical Services. The electrical Services are connected through Feed-outs at the primary vacuum interface and distributed in the vacuum vessel by cable looms ( up to 12 per sector). A cable tail will be routed ... More
Presented by Jorge GONZALEZ on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 127
Thermal shield (TS) is one of the components in the ITER tokamak to minimize radiation heat load from vacuum vessel and cryostat to magnet structure that operates at 4.5 K. The TS main components (TSMC) are vacuum vessel TS (VVTS), cryostat TS (CTS) and support TS (STS). The TSMC are cooled by 80 K helium gas, which is supplied from the cryoplant via manifold pipes. The surface emissivity of the T ... More
Presented by Dong Kwon KANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 128
The ITER In-Vessel Viewing System (IVVS) consists of six identical units located at the B1 level of the Tokamak complex, at lower ports 3, 5, 9, 11, 15 and 17. They can be deployed to perform in-vessel inspections between plasma pulses or during a shutdown. When not in use, each unit is housed inside a dedicated port extending from the Vacuum Vessel (VV) outer wall to the port cell (PC), locked by ... More
Presented by Davide FLAMMINI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 129
Nuclear heating of the vacuum vessel (VV) is an important issue for the design and the safe operation of ITER. The heating distribution must be known with high accuracy to identify hot spots which may be crucial for the reliable operation. The VV is heated by neutrons passing through the blanket shield modules and gaps, and photons generated in the VV structure. The heating distribution is thus st ... More
Presented by Anton TRAVLEEV on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 130
ITER vacuum vessel (VV) is composed of 9 sectors, and each sector is completed through an assembly of 4 segments which are independently fabricated. Compared with Upper, Equatorial and Lower segment which have relatively large curvature in a 3 dimensional configuration, Inboard segment is the most difficult in aspect of a welding distortion control although it seems to be simply in fabrication due ... More
Presented by Kwen-Hee HONG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 131
The ITER vacuum system will be one of the largest, most complex vacuum systems ever to be built and includes a number of large volume systems such as the Cryostat (~ 8500 m<sup>3</sup>3), Torus (~1330 m<sup>3</sup>3), and the Neutral Beams (~180 m<sup>3</sup>3 each). The vacuum system comprises of custom and commercially available components and adapted commercial vacuum technology. For a componen ... More
Presented by Liam WORTH on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 132
ITER Thermal shield (TS) is a thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2K. TS supports are designed to endure a dead weight, seismic load, electro-magnetic load and thermal loads. In the design and analysis of the TS supports, deterministic values of the geometry or dimension of ... More
Presented by Chang Hyun NOH on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 133
The first mirrors of ITER diagnostic systems are the most vulnerable ones since they are directed to the plasma and are subjected to erosion and intensive impurity deposition. In order to prolong the lifetime of the first mirror and to keep its high optical performance and maintainability, single crystalline molybdenum and rhodium have been considered as mirror materials, subject to intensive inve ... More
Presented by Yury KRASIKOV on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 134
ITER Diagnostic Port Plugs will operate with water at high pressures and temperatures. Because of these conditions of operation, the diagnostic Port Plugs are under the French Regulation on Pressure Equipment / Nuclear Pressure Equipment. This paper focuses on the assessments performed in order to substantiate application of Article 2 paragraph II of French decree 99-1046 relieving diagnostic port ... More
Presented by Thibaud GIACOMIN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 135
The cryogenic superconducting joint box is an important part of ITER HTS current leads, which is made of Copper-316L bi-metallic explosion bonded plate. The bimetal interface of the joint has the direct effect on the mechanical properties of the joint and testing performance at low temperature. This paper describes work on the development of water immersion ultrasonic testing technology, and its a ... More
Presented by Jiang BEIYAN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 136
The French Tore Supra tokamak is upgraded in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test actively cooled tungsten Plasma Facing Units (PFU) under long plasma discharge. As the existing cooling loop B30 cannot ensure the cooling of th ... More
Presented by Stephane GAZZOTTI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 137
In order to fully validate ‘’ITER-like’’ actively water cooled tungsten plasma facing units, addressing the issues of long plasma discharges, an axisymmetric divertor structure has been studied and manufactured for the implementation in the WEST (Tungsten (W) Environment in Steady state Tokamak) tokamak platform. This assembly, called divertor structure and coils (4m diameter, 20 tonnes), ... More
Presented by Louis DOCEUL on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 138
The JT-60SA Tokamak is provided with a cryogenic system with a refrigeration capacity of 9KW (eqv.) at 4.5 K. Before commissioning and during occasional warm-up periods the total 3.6 t helium inventory is stored in six pressure vessels, which have been procured by Europe. Each vessel is 22 m long, has a diameter of 4 m, a 250 m<sup>3</sup>3 volume, and weighs about 73 t. As the vessels will store ... More
Presented by Antonino CARDELLA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 140
Important challenges for fusion technology deal with the design of safety systems designed to protect the Vacuum Vessel (VV) in the case of pressurizing accidents like the LOCA (Loss Of Coolant Accident). This accident is caused by the failure of a number of elements of the Tokamak Water Cooling System and may result in relevant consequences for the integrity of the reactor. To prevent or to mitig ... More
Presented by D. MAZED on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 141
The flexible in-vessel inspection system (FIVIS) for EAST is a unique 10-degree-of-freedom manipulator for its serial structure of arcuate deployed Big Arm and its planar Small Arm (end effector):the Big Arm takes the Small Arm to all positions of the toroidal vacuum vessel (VV) along its equatorial plane,achieving a full coverage of VV’s first wall. In the in-vessel inspection process, the Big ... More
Presented by Weijun ZHANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 142
The remote handling in-vessel inspection manipulator specially developed for EAST superconducting tokamak has proven its kinematics feasibility in scale one toroidal vessel and its survivability under 120 °C high temperature. To adapt this manipulator for real in-vessel operation, most of its joint components, such as motors and reducers, must be isolated in sealed spaces to prevent possible cont ... More
Presented by Liang DU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 143
EAMA (EAST Articulated Maintenance Arm) is an articulated serial robot arm working in experimental advanced superconductor tokamak for inspection and maintenance. Redundant flexible structure of EAMA increases reach capability, however, it reduces accuracy and speed due to the compliance introduced into each joint. This deteriorates EAMA into oscillation and produces undesirable disturbance. In t ... More
Presented by Jing WU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 144
EAST Articulated Maintenance Arm (EAMA) is a highly redundant serial robot system with 11 degree of freedoms (DOFs) in total. It will allow remote inspection and simple repair of plasma facing components (PFCs) in EAST vacuum vessel (VV) without breaking down the ultra-high vacuum condition during physical experiments. Due to its long-reach mechanisms with a weight more than 100 kg, the gravity ef ... More
Presented by Shanshuang SHI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 145
The design requirements for the DEMO Blanket Primary Heat Transfer System, both for the water and helium concepts have been defined. The plasma facing components cooling circuits have to fulfill several requirements dictated by safety and operational criteria. In particular, the Blanket PHTS of a fusion reactor shall transfer the heat load coming from the plasma to the secondary side to allow powe ... More
Presented by Dario CARLONI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 146
Europe is developing two reference tritium Breeder Blankets concepts that will be tested in ITER under the form of Test Blanket Modules (TBMs): i) Helium-Cooled Lithium-Lead (HCLL) which uses liquid Pb-16Li as both breeder and neutron multiplier, ii) Helium-Cooled Pebble-Bed (HCPB) with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Both concepts are using the EU ... More
Presented by Milan ZMITKO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 147
The Test Blanket Module (TBM) and its associated ancillary systems (including cooling systems, tritium extraction system, coolant purification, PbLi loop, I&C) form the Test Blanket System (TBS). The TBSs will be fully integrated in the ITER machine and buildings. Therefore, testing of the TBS integration and maintenance in ITER port cell prior to its installation and operation in the ITER machine ... More
Presented by Ladislav VALA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 148
Europe is developing two reference tritium breeder blankets concepts that will be tested in ITER under form of Test Blanket Systems (TBSs): (i) the helium-cooled lithium-lead (HCLL) which uses liquid Pb16Li as both breeder and neutron multiplier, (ii) the helium-cooled pebble-bed (HCPB) with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. One of core documents to ... More
Presented by Jose GALABERT on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 149
It is widely believed that fusion DEMO reactor will need significant amount of tritium at the beginning of its operation. However, the authors have pointed out that steady deuterium operation can produce sufficient tritium in a reasonable period of DD operation by DD reaction followed by exponential breeding in the blanket. The present study further suggests that realistic Power Ascension Tests (P ... More
Presented by Satoshi KONISHI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 150
The basis of a thermonuclear fusion reactor is neutron source (FNS) based on the tokamak [1]. FNS should provide steady flow of fusion neutrons with a capacity of 10-50 MW, which reached close to the pulse values of existing installations JET and JT-60U. Fuel cycle technologies (FC) is one of the key elements for the FNS. FC systems should provide treatment and storage of deuterium and tritium, a ... More
Presented by Sergey ANANYEV on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 151
Thermal hydraulic and accident analysis codes such as RELAP5-3D and MELCOR rely on an equation of state to specify all the thermodynamic properties of fusion-relevant working fluids such as PbLi.  The existing liquid metal fluid properties in both RELAP5-3D and MELCOR are based on a five parameter "soft sphere" equation of state for which parameter sets that approximately reproduce experiment dat ... More
Presented by Paul HUMRICKHOUSE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 152
The Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) is one of the four BB concepts being investigated in the EU for their possible implementation in DEMO. During 2011-2013 initial HCPB BB conceptual studies were performed based on a design extrapolation from the ITER’s HCPB Test Blanket Module, leading to the so called “beer-box” BB concept. During 2014 the “beer-box” BB concept su ... More
Presented by Francisco A. HERNANDEZ GONZALEZ on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 153
Within the Power Plant Physics and Technology (PPPT) programme of EUROfusion, a major development effort is devoted to the conceptual design of a DEMO reactor which has the capability to breed Tritium for self-sufficiency. This DEMO is assumed to be suitable for the accommodation of any blanket type out of the existing concepts. For the neutronics analyses, a generic DEMO model is thus set-up wh ... More
Presented by Pavel PERESLAVTSEV on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 154
Lithium density and tritium release behaviour are key properties in the design and synthesis of Li-containing solid breeders for the helium cooled pebble blanket (HCBP) concept. Radiation and high temperature may give rise to changes in both material composition and microstructure, hence important aspects including chemical compatibility and tritium production/extraction effectiveness may be stron ... More
Presented by Alejandro MORONO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 155
The tritium release behaviour of candidate ceramic materials for the HCPB breeder concept is still an issue. High experimental costs, long experimental periods, and handling difficulties for activated materials after being tested in experimental fission reactors have motivated the validation of alternative methods for testing the gas desorption behaviour of tritium breeder materials. In the framew ... More
Presented by Maria GONZALEZ on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 156
The simulation plays an important role to estimate characteristics of cooling in a blanket for such high heating plasma in ITER-BA. An objective of this study is to perform large -scale direct numerical simulation (DNS) on heat transfer of turbulent flow on coolant materials assumed gas flow.  The coolant flow conditions in ITER-BA are assumed to be Reynolds number of a higher order. To investiga ... More
Presented by Shin-ichi SATAKE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 157
All solid breeder concepts, considered to be tested in ITER, make use of lithium-based ceramics in the form of pebble-packed beds as tritium breeder. A thorough understanding of the effective thermal conductivity of the ceramic breeding pebble beds in fusion relevant conditions is essential for the design of the breeder blanket modules of the future fusion reactors. An experimental set-up for the ... More
Presented by Simone PUPESCHI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 158
Solid blanket is a core candidate of blanket structure for CFETR (Chinese Fusion Engineering Testing Reactor), and the effective thermal conductivity of ceramic pebble beds is a very significant parameter for the thermo-mechanical design of solid blankets. In order to obtain the effective thermal conductivity, theoretical calculation and experimental measurement are two common methods. Compared wi ... More
Presented by Shuang WANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 159
Tritium breeder pebble bed plays a vital role in tritium breeding for fusion solid blanket. And thermo-physical properties of it affect the thermo-mechanical and structural design of solid blanket directly. Theoretical and experimental study on effective thermal conductivity of ceramic pebble beds have been carried out in this paper. Firstly, a new theoretical model, coupling the contact a ... More
Presented by Hongli CHEN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 160
Thermal transport efficiency of a tritium breeding pebble bed can strongly affect tritium self-sufficiency of the magnetic confinement fusion solid breeding blanket system.  The effective thermal conductivity of the pebble bed is related not only to its configuration, such as dimensions, pebble size, and pebble material porosity, but also to its environment, such as helium temperature, flow veloc ... More
Presented by Yuanjie LI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 161
As a water-cooled solid breeder blanket of a fusion reactor, safety concern has become one of the most critical issues. In specific, Be pebbles as a multiplier have been well-known to generate hydrogen and exothermally react while reacted with water vapor at high temperature. In contrary to these Be pebbles, Beryllium intermetallic compounds (beryllides) are one of promising materials because of i ... More
Presented by Jae-Hwan KIM on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 162
The development of system code for CFETR (China Fusion Engineering Test Reactor) is in progress for the optimization of the CFETR design in both core physics and engineering. As one of the key modules, the neutronics interface module has been implemented within the engineering framework of CFETR system code. The neutronics interface module, which is designed to work in conjunction with the general ... More
Presented by Kun XU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 163
Chinese Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion device that was proposed to achieve 200 MW fusion power, 30-50% duty time factor, and tritium self-sufficiency. As a candidate blanket concept for CFETR, a helium cooled solid breeder (HCSB) blanket was designed following the specific requirements. The helium cooling system (HCS) is an important ancillary system of HCSB blanket ... More
Presented by Shuai WANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 164
After the Conceptual Design Review (CDR), Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) is in progress of the preliminary design phase. The detained design work was performed on the connecting supports which are connected between the TBM and the TBM-shield. The geometric design of the connecting supports are referred from the connection design of the blanket first wall. The othe ... More
Presented by Seong Dae PARK on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 165
Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be operated at elevated temperature with high pressure helium coolant during normal operation in ITER. One of the main ancillary systems of HCCR-TBS is Helium Cooling System (HCS) which play an important role to extract heat from HCCR Test Blanket Module (TBM) by the helium coolant to keep the operational temperature and ... More
Presented by Mu-Young AHN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 166
A helium circulator, to provide up to 1.5 kg/s of helium flow with pressure of 8 MPa, has been developed for the HCCR-TBS. To overcome the pressure drop of the helium cooling system of the HCCR TBS, the circulator is designed maximum speed of 70,000 RPM with electric power of 150 kWe to meet compression ratio of 1.1. One of the major design features of the circulator is that the impeller and the s ... More
Presented by Eo Hwak LEE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 167
Within the framework of EUROfusion R&D activities CEA-Saclay has carried out an investigation of the thermal and mechanical performances of alternative designs intended to enhance the Tritium Breeding Ratio (TBR) of the Helium-Cooled Lithium Lead (HCLL) blanket for DEMO. Neutronic calculations performed on the 2014 DEMO HCLL layout have indeed predicted a value of TBR equal to 1.07, lower than the ... More
Presented by Pietro ARENA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 168
In 2008-2009 experiments have been performed to investigate liquid metal magnetohydrodynamic (MHD) flows in a scaled mock-up of a helium cooled lead lithium (HCLL) blanket. In order to improve the mechanical stiffness of the blanket module the design of the stiffening plate between two hydraulically connected breeder units (BUs) has been later modified. In the former design the liquid metal passed ... More
Presented by Chiara MISTRANGELO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 169
Research Centre Rez (CVR) is actively involved in research and development of a purification technique of the liquid lithium-lead eutectic alloy based on use of a cold trap. The first activities linked to this field are dated since 2003. They are carried out within the major European fusion projects (F4E, EFDA and EUROfusion) and the Czech national CANUT project. For the cold trap development, the ... More
Presented by Otakar FRYBORT on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 170
In a prospect of future fusion power plants construction, diferent concepts of tritium breeding blankets are being developed within the EUROfusion breeding blanket work package. Three main concepts using Pb-17Li as breeder, the HCLL (Helium Cooled Lithium Lead), WCLL (Water Cooled Lithium Lead) nad DCLL (Dual Coolant Lithium Lead) are developped as candidate technologies for european DEMO facility ... More
Presented by Michal KORDAC on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 171
The EUROfusion Consortium aims at developing a conceptual design of a fusion power demonstrator (DEMO). The breeding blanket facing the plasma is one of the key components of DEMO. It must ensure tritium self-sufficiency and heat removal functions. The Helium Cooled Lithium Lead (HCLL) blanket concept is one the four breeding blanket concepts investigated for DEMO. It uses the liquid lithium lead ... More
Presented by Jean-Charles JABOULAY on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 172
The dual functional lead lithium (DFLL) test blanket module (TBM) concept has been proposed by FDS team to demonstrate the techniques basis of DEMO liquid blanket concepts, including quasi-statistic lead lithium (SLL) breeder blanket and the dual-cooling lead lithium (DLL) blanket. In recent years, series R&D work for DFLL-TBM carried out are mainly on five topics: 1) Structural materials (i.e. CL ... More
Presented by Qunying HUANG on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 173
Former Investigations clearly had revealed that embrittlement and hardening of RAFM steel after 15 - 70 dpa neutron irradiation damage remarkably can be reduced by short time post-irradiation annealing (PIA) at 550 °C [1, 2]. The purpose of this study is to demonstrate the repeatability of the damage- and recovery-mechanisms to RAFM 7-10% CrWVTa, ODS EUROFER, Boron doped heats of the prior 33 ... More
Presented by Hans-Christian SCHNEIDER on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 174
Oxide dispersion strengthened ferritic steels (ODS FS) are candidate structural materials for future fusion reactors thanks to their high temperature strength, high creep resistance, and good resistance to neutron radiation. Their outstanding behavior is a direct consequence of their extremely fine microstructure and the presence of highly stable and finely distributed nanometric oxide precipitate ... More
Presented by Nerea ORDAS on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 175
F82H is the reduced activation ferritc/martensitic (RAFM) steel which has been developed in Japan. Its chemical composition was designed based on the composition of high Cr heat resistant steel, Mod9Cr-1Mo, reducing activity level by replacing Mo to W, Nb to Ta, and reduce N level to suppress 14C formation. In order to prove its potential as the structural materials, it is critical to provide data ... More
Presented by Hiroyasu TANIGAWA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 176
The box structure of water-cooled solid breeding (WCSB) blanket fabricated by F82H is being developed in Japan for the DEMO reactor. In the DEMO operation, the structural materials in the region of first wall (FW) will be exposed to severe fusion neutron irradiation. One of the issues is the loss of ductility for the structural materials due to severe fusion neutron irradiation. In the case of in- ... More
Presented by Takeshi MIYAZAWA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 177
The hot isostatic pressing (HIP) is the key technology to fabricate the first wall of the fusion blanket system. Generally, the Charpy impact test is applied to evaluate the failure behavior of the HIP joint however there is a drawback that this cannot be applied to the practical thin-walled first wall component since the Charpy impact test requires a long bar specimen. Alternatively the authors r ... More
Presented by Takashi NOZAWA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 178
Reduced activation ferritic/martensitic steel, as typified by F82H, is a promising candidate for structural material of DEMO fusion reactors. To prevent plasma sputtering, tungsten (W) coating was essentially required. Vacuum plasma spray (VPS) is one of candidate coating processes, but the key issues are the degraded mechanical and thermal properties due to its relatively higher porosity and smal ... More
Presented by Kazumi OZAWA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 179
Connection between blanket and out-vessel component is essential to fusion reactors. In the present study, electron beam welding was carried out to fabricate a dissimilar-metals joint between a blanket structural material, F82H steel, and an out-vessel component material, 316L steel. Impact properties and deformation behavior of the joint were analyzed after neutron irradiation. Two types of Charp ... More
Presented by Haiying FU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 180
Heavy ion irradiation technique has been used for simulating fusion neutron irradiation on materials. However mechanical testing technologies were limited due to the thin irradiated layer only up to several um in depth. Nanoindentation hardness were often used for evaluating irradiation hardening behaviro of ion-irradiated subsurface. This study investigates micro-pillar compression behavior of io ... More
Presented by Ryuta KASADA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 181
The small punch (SP) test method is a one of the small specimen test techniques (SSTT). This method has several advantages: it requires only a small specimen, its test method is simple, and it is able to evaluate various mechanical properties. For these reasons, the SP method is commonly used in post-irradiation testing (PIE) of nuclear materials and as a damage evaluation technique for actual str ... More
Presented by Toshiya NAKATA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 182
The R&D of high performance fuel cladding materials has been considered to be essential for the realization of fusion and Gen IV fission energy systems. The 9Cr oxide dispersion strengthened (ODS) martensitic steels was developed for applying as cladding materials of sodium-cooled fast breeder reactors (FBRs). The steels exhibited good compatibility with sodium, while the corrosion resistance was ... More
Presented by Noriyuki Y. IWATA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 183
A reduced-activation ferritic steel, F82H steel, is the primary candidate structural material for fusion blanket. It has been clarified that long term aging degrades both strength and ductility due to precipitation of Laves phase (Fe2W) and other changes in microstructure. In order to evaluate the degradation and to clarify its mechanisms, the present study analyzed the tensile properties of F82H ... More
Presented by Takuya NAGASAKA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 184
ITER Vacuum Vessel (VV) is made of double walls connected by ribs structure and flexible housings, space between these walls is filled up with In Wall Shielding (IWS) blocks to (1) reduce neutrons streaming out of plasma and (2) reduce toroidal magnetic field ripple. These blocks will be connected to the VV through a supporting structure of Support Rib (SR) and Lower Bracket (LB) assembly. SR and ... More
Presented by Yatinkumar SARVAIYA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 185
Measurement and calculations of long-lived radionuclide activity forming in the 14 MeV neutron field, in <sup>6</sup>6Li-D converter were done, in some steel composites of ITER. The activation was conducted in September, 2014 in the thermal-to-14MeV neutron converter constructed in National Centre for Nuclear Research in Poland. This irradiation facility was placed in the core of MARIA research fi ... More
Presented by Władysław POHORECKI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 186
In wall Shielding blocks will be inserted between inner and outer shell on ITER Vacuum Vessel (VV) and will fill up about 60% of volume between two shells. IWS blocks comprise of number of plates stacked together with fasteners. There are two types of IWS blocks, (i) Primary IWS blocks made of Austenitic stainless steels (SS304B4 and B7) to provide neutron shielding to all components inside cryost ... More
Presented by Abha MAHESHWARI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 187
The ITER Correction Coils (CCs) consist of three sets of six coils, Bottom (BCC), Side (SCC) and Top Correction Coils (TCC), respectively, located in between the toroidal (TF) and poloidal field (PF) magnets. The CCs rely on 10 kA NbTi Cable-in-Conduit Conductor (CICC). Each CC winding pack is enclosed inside a 20 mm thick stainless steel case, providing structural reinforcement against the electr ... More
Presented by Stefano SGOBBA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 188
Reduced activation ferritic-martensitic (RAFM) steel is considered a primary candidate for the structural material in a fusion reactor. The operational design window for a blanket is limited by the high-temperature creep and low-temperature irradiation embrittlement of the structural material, and it is therefore essential to develop RAFM steel which can withstand high temperatures and high energy ... More
Presented by Young-Bum CHUN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 189
Chemical compatibility between Korean reduced activation ferritic-martensitic alloy (ARAA) and lithium meta-titanate breeder was investigated under operation conditions; high temperature and helium purge gas including low concentration of hydrogen. ARAA specimens were embedded inside lithium meta-titanate powder and compacted under the load of 200 MPa to form block-shaped samples. The samples were ... More
Presented by Seungyon CHO on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 190
Reheating cracking susceptibility in the weld heat-affected zone (HAZ) of reduced activation ferritic-martensitic (RAFM) steels was explored by evaluating stress-rupture parameters (SRP), which depends on rupture strength and ductility. The HAZs simulation and stress-rupture experiments were carried out using a Gleeble simulator at various temperatures, corresponding to post-weld heat treatment (P ... More
Presented by Joonoh MOON on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 191
The effect of addition of Ti on microstructures and mechanical properties in RAFM steels were investigated. Ti-bearing RAFM steels, designed based on the thermodynamic calculation, were fabricated by vacuum induction melting and hot-rolling process. All samples were heat treated by normalizing and tempering, resulting in tempered martensite with M23C6 carbides and MX precipitates. The microstructu ... More
Presented by Jun Young PARK on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 192
The property of functional material for the design of the breeding blanket is very essential. Since the stress due to the thermal load on breeding blanket structure is one of the main design driver, the thermal property of the material is very important for thermal-structural and thermo-hydraulic analysis. In particular, the thermal conductivity is one of necessary input data for these analyses pe ... More
Presented by Youngmin LEE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 194
China low activation martensitic (CLAM) steel, one of the three main reduced activation ferritic/martensitic steels (RAFMs) under development in the world, has been selected as the primary structural material of ITER testing blanket material (TBM) in China. It is important to understand the neutron irradiation effects of CLAM steel, especially in an environment with high energy and high dose neutr ... More
Presented by Jingping XIN on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 195
China low activation martensitic (CLAM) steel has been selected as the primary structure material of FDS series PbLi blankets for fusion reactors, CN helium cooled ceramic breeder (HCCB) test blanket module (TBM) for ITER and the blanket of other future fusion reactors. Tantalum (Ta) is the essential element for reduced activation ferritic/martensitic (RAFM) steels, and the effect of Ta content ... More
Presented by Shaojun LIU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: I. Materials Technology Board #: 196
Activities under the EUROfusion work package (WP) JET3 programme have been established to enable the technological exploitation of the planned JET experiments over the next few years, which culminates in a D-T experimental campaign, DTE-2. In the areas of nuclear technology and nuclear safety the programme offers a unique opportunity to provide experimental data that is relevant to ITER. The key p ... More
Presented by Lee PACKER on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 197
The preliminary safety and operating design requirements are being defined aiming at obtaining the license for construction with a relatively large operational domain to assure an easy control and adequate availability of DEMO. The DEMO design approach is being organized, by taking into account the Nuclear Power Plant experience and the lessons learnt from ITER and GEN IV. Outstanding challenges r ... More
Presented by Sergio CIATTAGLIA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 198
Abstract : A fusion DEMO reactor, like other advanced nuclear energy systems, must satisfy a range of goals including a high level of public and worker safety, low environmental impact, high availability, a closed fuel cycle, and the potential to be economically competitive. It is well known that the experience of the ITER project will facilitate DEMO programs in developing a safety approach and s ... More
Presented by Yican WU on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 199
Environment assessment of large inventory tritium for fusion devices is an important issue before fusion energy commercially used. Different with other radioactive substance, tritium has particular processes of atmosphere dispersion, dry & wet deposition, oxidation in air & soil, reemission, transfer among the soil, plants, animals and human beings. In our previous work, a virtual point source met ... More
Presented by Muyi NI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 200
In large fusion machines, as the foreseen DEMO, the high energy neutrons produced will cause the transmutation of the interacting materials which become a source of radioactive waste. One of the main presuppositions for the global interest in nuclear fusion is that it should be cleaner and safer comparing with traditional nuclear technology. This implies, among other considerations, that the radio ... More
Presented by Raquel GARCIA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 201
Demonstrating tritium self-sufficiency is an important goal of the European tokamak DEMOnstration reactor developed within the Power Plant Physics and Technology (PPPT) EUROfusion programme. Currently four breeder blanket concepts are being considered; the Helium Cooled Pebble Bed (HCPB), Helium Cooled Lithium-Lead (HCLL), Dual Cooled Lithium-Lead (DCLL) and Water Cooled Lithium-Lead (WCLL). The d ... More
Presented by Tim EADE on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 202
The pre-conceptual design concept on the Korean fusion demonstration reactor (K-DEMO) has been studied in Korea since 2012. In the fusion reactor, neutrons produced from fusion reactions cause activation of fusion reactor devices. For the safety of fusion devices and workers during operation and maintenance, it is important to calculate activation and to evaluate shutdown dose rate (SDR). In this ... More
Presented by Jae Hyun KIM on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 203
Coolant activation is important concern for nuclear fusion devices, where water is being used in heat transfer systems. Production of nitrogen-16 isotope is one of the main hazards in such systems and should be taken with care. In this work, the examination of the neutron activation in water cooling systems, that might be used in future fusion devices, was carried out. Primary heat transfer system ... More
Presented by Andrius TIDIKAS on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 204
The problem of Source Term qualification is one of the most important topics in order to predict possible releases of the Activation Products (APs) and tritium from the DEMO Fusion reactor. The prevention of any possible consequence, which can affect the environment and the population, is the mission of Fusion technology. In the frame of the EUROfusion Work Package of Safety Analyses and Environme ... More
Presented by Guido MAZZINI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 205
In frame of project Eurofusion, WPSAE (safety and environment) were reviewed existing detritiation technique for different material types and identified techniques for further development for short –term reuse, long – term reuse, recycling and disposal. Moreover criteria for assessment were proposed and technique were described. The most efficient treatment technique for different group of mat ... More
Presented by Lucie KARASKOVA NENADALOVA on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 206
After the tritium handling operation, it is an important issues to take an appropriate disposal method of tritium handling facility contaminated with tritium. In Kyushu University, according to the relocation program to the new campus, decommissioning operation of tritium handling facility located in the former campus had been performed. This handling facility made of concrete was used for acceler ... More
Presented by Toshiharu TAKEISHI on 5 Sep 2016 at 14:20
Type: Poster Session: P1 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 207
In previous studies, the authors proposed a novel nuclear fusion biomass gasification plant concept as an alternative to conventional nuclear fusion power plants. This gasification plant concept utilizes the heat from fusion blanket to convert biomass into synthetic gas (H2 + CO), and then convert it into liquid fuels, e.g. methanol or diesel. Through this nuclear fusion gasification plant concept ... More
Presented by Shutaro TAKEDA on 5 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 1
A number of tokamaks, including the largest operating one, Joint European Torus (JET), has ferromagnetic core installed in their plasma current drive system. Moreover, some auxiliary systems, such as magnetic shielding of neutral beam injection (NBI) system, or iron inserts for toroidal field ripple mitigation, consist of non-negligible amount of ferromagnetic material as well. Besides the intende ... More
Presented by Tomas MARKOVIC on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 2
Neutral beam injection systems have proved themselves as the most effective form of auxiliary heating in tokamak plasmas. In positive ion based systems once the beam is neutralised there are many residual ion components which must be intercepted by suitable ion dumps. A particular challenge for ion dump design occurs when the dump must be placed close to a focus point as is the case for the curved ... More
Presented by Alastair SHEPHERD on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 3
The final phase of the JET Programme in Support of ITER plans to operate with 100% Tritium (TT) followed by Deuterium-Tritium (DT) operation, to help minimise risks and delays in the execution of the ITER Research Plan and the achievement of Q~10. Additional technical requirements (compared to Deuterium operation) are needed to allow operation with Tritium gas, a high DT neutron flux and neutron a ... More
Presented by Eva BELONOHY on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 4
Neutronics benchmark experiments are conducted at JET for validating the neutronics codes and tools used in ITER nuclear analyses to predict quantities such as the neutron flux along streaming paths and dose rates at the shutdown due to activated components. In particular, in the frame of subproject NEXP of JET-3 program, several activities are performed within EUROfusion Consortium devoted to the ... More
Presented by Rosaria VILLARI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 5
The European roadmap to the realisation of fusion energy has identified a number of technical challenges and defined eight different missions to face them. Mission 2 ‘Heat-exhaust systems’ addresses the challenge of reducing the heat load on the divertor targets. Divertor Tokamak Test (DTT) facility [1]-[2] has been launched to investigate alternative power exhaust solutions for DEMO. This tok ... More
Presented by Roberto AMBROSINO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 6
In the European Fusion Roadmap, one of the main challenges to be faced is the mitigation of the risk due to the impossibility of directly extrapolate to DEMO the divertor solution adopted in ITER, due to the expected very large loads. Thus a satellite experimental facility oriented toward the exploration of robust divertor solutions for power and particles exhaust and to the study of plasma-materi ... More
Presented by Alessandro ANEMONA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 7
The proposed Divertor Test Tokamak, DTT, aims at studying power exhaust and divertor load in an integrated plasma scenario. Additional heating systems have the task to provide heating to reach a reactor relevant power flow in the SOL and guarantee the necessary PSEP/R together adequate plasma performances. About 40 MW of heating power are foreseen to have PSEP/R ≥ 15 MW/m. A mix of the three ... More
Presented by Gustavo GRANUCCI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 8
on behalf of the EUROfusion WPDTT2 team & the DTT report contributors Within the frame of the DTT program, included in the EuroFusion roadmap, the design of a new Tokamak dedicated to tackle the Power Exhaust problem as an integrated bulk and edge plasma problem has been developed. The main guidelines used to work out the machine parameters will be shortly illustrated.To allow the machine flexibil ... More
Presented by Giorgio MADDALUNO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 9
This paper describes the activity addressed to the conceptual design of the first wall and the main containment structures of DTT device, which will be broadly presented in the invited talk "Design and definition of a Divertor TOKAMAK Test facility". The work moved from the geometrical constraints imposed by the desired plasma shape and the configuration needed for the magnetic coils.  Many other ... More
Presented by Giuseppe DI GIRONIMO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 10
The DTT (Divertor Test Tokamak) is a new facility conceived in the frame of EUROfusion roadmap with the aim to assess and possibly integrate all the relevant physics and technology divertor issues. The general project is presented in another paper of this conference [1] and with more details in [2]. The general project includes the analysis of the site requirements from several points of view; amo ... More
Presented by Giuseppe MAZZITELLI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 11
The power supplies (PSs) of the DTT proposal, as presented in the talk "Design and definition of a Divertor Tokamak Test facility" invited at this conference, have to feed:   6 central solenoid (CS) and 6 poloidal field (PF) superconducting coils, with currents up to 25 kA. 18 toroidal field (TF) superconducting coils, with a current up to 50 kA. Some fast plasma control coils, including at leas ... More
Presented by Alessandro LAMPASI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 12
The experimental facility THALLIUM (Test HAmmer in Lead LithIUM) was designed to experimental validate the RELAP5-3D code simulations of the pressure wave propagation in the HCLL TBM due to In-box LOCA. THALLIUM, which reproduces the geometry of the LLE loop of the HCLL TBM, was installed at the ENEA Brasimone Research Centre to support the accidental analysis of this type of test blanket module. ... More
Presented by Marco UTILI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 13
The 1<sup>st</sup>st Specific Grant of the Framework Partnership Agreement 372 deals with experimental activities in support of the Conceptual Design of HCLL and HCPB Test Blanket Systems. Service-2 is focused on thermal-hydraulic tests with high pressure Helium for validation and benchmarking of suitable dedicated numerical tools. In this frame, an extensive experimental campaign has been execut ... More
Presented by Gianluca BARONE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 14
The experimental facility IELLLO (Integrated European Lead Lithium LOop) was designed and installed at the ENEA Brasimone Research Centre to support the design of the HCLL TBM. This work presents the results of the experimental campaign carried out within the framework of F4E-FPA-372 and which had three main objectives. First, to produce new experimental data for flowing LLE (Lead-Lithium Eutectic ... More
Presented by Alessandro VENTURINI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 15
Due to the complexity of fusion reactors on geometry and neutron physics, the Monte Carlo (MC) methods have been broadly adopted in fusion nuclear design and analysis. But for calculations that require obtaining a detailed global flux map, such as the shutdown dose rate analysis, analog MC simulations usually cost a prohibitive long run time. To make such analysis computational practicable, it is ... More
Presented by Liqin HU on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 16
Great challenges exist in real fusion engineering projects for the current Monte Carlo (MC) methods including the calculation modeling of complex geometries, simulation of deep penetration problem, slow convergence of complex calculation, lack of experimental validation for new physical features, etc. Several novel and advanced capabilities of the latest version of MC program SuperMC for fusion ap ... More
Presented by Jing SONG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 17
Operation of fusion facilities using deuterium and tritium to fuel the fusion reaction will lead to generation of radioactive waste during operating and decommissioning phases. Most of these wastes are expected to be contaminated with tritium and will require a specific management strategy taking into account the physical and chemical properties of tritium. The reference management strategy for tr ... More
Presented by Michal KRESINA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 18
The superconducting tokamak JT-60SA, aimed to support and complement the ITER experimental programme, is currently being assembled at the JAEA laboratories in Naka (Japan). Within the European contribution, Spain is responsible for providing JT-60SA cryostat. The cryostat is a stainless steel vacuum vessel 14m diameter, 16m height which encloses the tokamak providing the vacuum environment (10-3 P ... More
Presented by Mercedes MEDRANO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 19
A conceptual design for a pellet injection system will be worked out, capable to support key missions of the new tokamak device JT-60SA. For exploitations in view of ITER and to resolve key physics and engineering issues for DEMO, several tasks were assigned to this system. Physics investigations aim at operation at high density in ITER and DEMO relevant plasma regime above Greenwald density, powe ... More
Presented by Peter LANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 20
The ITER Heating Neutral Beam (HNB) injectors shall be protected from stray magnetic field (several hundreds of mT) produced by the ITER PF coils and plasma current. Such stray field would hamper the production of negative ions, deflect ion trajectories in the accelerator and cause intolerable heat load on neutralizer and beam line components. In order to keep the residual magnetic field below acc ... More
Presented by Nicolo MARCONATO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 21
In the multi-beamlet, negative-ion based Heating Neutral Beam (HNB) Injectors presently used in fusion research, arrays of permanent magnets are embedded in the Extraction Grid (EG) for the suppression of the unwanted co-extracted electrons. These magnets cause a significant undesired deflection of the negative ion beamlets, with a typical alternate pattern, matching the orientation of the magnet ... More
Presented by Daniele APRILE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 22
The gas cloud inside the neutralizer of MITICA (Megavolt ITER Injector and Concept Advancement), required to neutralize the negative ion beam, will be created continuously by 20 identical nozzles providing the gas needed for different operation modes. In order to validate the design, one nozzle will be characterized in detail and for a wide range of supply conditions in a dedicated experiment at K ... More
Presented by Stefan HANKE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 23
The Acceleration Grid Power Supply supplies the acceleration grids of the MITICA experiment, the full scale prototype of the ITER Neutral Beam Injector under construction in Padua (Italy) to tackle the technical challenges and prepare for the target performance objectives ahead of operation in ITER. The AGPS is a special switching power supply with demanding requirements: high rated power (55 MW), ... More
Presented by Loris ZANOTTO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 24
The negative ion source test facility ELISE represents the first step in the European R&D roadmap for the neutral beam injection (NBI) systems of ITER in order to consolidate the design and to gain early experience with a large and modular Radio Frequency (RF) negative ion source. The aim of ELISE is to demonstrate the ITER requirements with respect to extracted negative hydrogen densities (329 A/ ... More
Presented by Bernd HEINEMANN on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 25
The test facility ELISE (Extraction from a Large Ion Source Experiment) at IPP Garching, Germany, aims to demonstrate ITER-relevant negative ion beam parameters which are required for the NBI system of ITER. ELISE is equipped with a Radio Frequency driven source and an ITER‑like extraction system with half the ITER size. An H<sup>-</sup>- or D<sup>-</sup>- beam can be extracted for 10 s every 3 ... More
Presented by Riccardo NOCENTINI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 26
The Beam Line Components (BLCs) for the ITER Diagnostic Neutral Beam (DNB) and Indian Test Facility (INTF) are mainly water cooled elements made from CuCrZr which are designed to absorb heat flux up to 10MW/m<sup>2 </sup>2 (e.g. Heat Transfer Element for calorimeter) according to their position in beam line. The design of these components imposes stringent requirements of having the specific chemi ... More
Presented by Chandramouli ROTTI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 27
The acceleration system of Beam Source(BS) of Neutral Beam(NB) system is composed of water cooled Oxygen-Free Copper multi-aperture grid systems which is designed for focusing of beamlets to a focal point located at distance>20m from the Grounded Grid. For present application in the accelerator for DNB, this focusing is obtained using a combination of segment bending and aperture offsets. In verti ... More
Presented by Jaydeepkumar JOSHI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 28
Design and manufacturing of DC 1 MV components have progressed for the ITER neutral beam injector. A multi-conductor DC 1 MV transmission line (TL) which can transmit five-different voltages of 200 kV step simultaneously has been manufactured and tested. The TL is a gas insulation tube with SF6 gas of 0.6 MPa. A layout of those conductors inside the tube was designed through electric field analys ... More
Presented by Hiroyuki TOBARI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 29
The main mission of KSTAR program is exploring the physics and technologies of high performance steady state tokamak operation that are essential for future fusion reactor. Since the successful long pulse operation of 25sec at 0.5MA exceeding conventional tokamak capabilities in 2013, the duration of H-mode has been extended to over 50s which corresponds to a few times of current diffusion time. I ... More
Presented by Jong-Gu KWAK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 30
Helicon wave coupling for efficient off-axis current drive using a traveling wave antenna has been proposed. It was found that helicon wave can drive plasma current in the mid-radius of high electron beta plasmas in medium and large size tokamak due to moderate optical thickness and wave alignment nature of helicon wave in helical magnetic field. KSTAR tokamak can be a good platform to test this c ... More
Presented by Haejin KIM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 31
Steady-state operation of a DEMO-like tokamak requires substantial off-axis current be driven by external current drive systems. Non-inductive current drive is needed to complement the bootstrap current to support the plasma current in steady state. Recently, helicon wave current drive at frequencies of 500~700 MHz is gained much attention to achieve off-axis current drive with high efficiency. He ... More
Presented by Hyunho WI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 32
The KSTAR LHCD system is to be upgraded for RF power up to 4 MW in 2020. The basic configuration of the system is composed of eight 5-GHz 500-kW CW klystrons, low-loss transmission line with oversized circular waveguide, and PAM launcher for the mid-plane injection. An off mid-plane injection near the upper diverter is also under consideration. A preliminary study based on a mid-plane PAM launch ... More
Presented by Jeehyun KIM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 33
The KSTAR LHCD system is using a 5-GHz, 0.5-MW c. w. klystron and oversized rectangular waveguides. The WR187 output waveguide of the klystron transmits the RF power to the LH launcher via 80-m of transmission line composed of WR284 oversized rectangular waveguide. The overall transmission loss was about 34% including 26% of Ohmic loss. In order to transfer RF power effectively from a klystron to ... More
Presented by Taesik SEONG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 34
The coupling of lower hybrid (LH) range of frequencies waves to strongly magnetized plasmas is a critical issue on tokamaks as the RF power which can be transferred from the antenna to the plasma is often limited by the quality of this coupling. Development of new types of antennas aiming at improving the ability of the antenna to handle large powers in stationary conditions, as it will be req ... More
Presented by Julien HILLAIRET on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 35
In the DIII-D tokamak, one of the most powerful techniques to control the density, temperature and plasma rotation is by eight independently modulated neutral beam sources with a total power of 20 MW. The rapid modulation requires a high degree of reproducibility and precise control of the ion source plasma and beam acceleration voltage.  Recent changes have been made to the controls to provide a ... More
Presented by Carl PAWLEY on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 36
The Neutral Beam system on DIII-D consists of eight ion sources. The basis of the DIII-D NB system is the Common Long Pulse Source (CLPS). The CLPS is an 80 kV high perveance, deuterium positive ion based system delivering up to 2.5 MW per source. The ion source is a filament driven magnetic bucket design and the accelerator is a slot and rail tetrode design with vertical focusing achieved through ... More
Presented by Brendan CROWLEY on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 37
The gyrotron complex on DIII-D has been updated and comprises six gyrotrons installed and routinely operating reliably for injection of up to 3.6 MW into the plasma. The operational maximum of 5 s pulse length for the six gyrotrons allows up to 18 MJ total energy to be injected into the plasma. Recent system upgrades include faster launcher mirror scans and control by the plasma control system ... More
Presented by Mirela CENGHER on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 38
The RAPTOR - RApid Transport simulatOR code [F. Felici et al 2011 Nucl. Fusion 51 083052] is a model-based control-oriented code that predicts Tokamak plasma profile evolution in real-time. One of its key applications is in a state observer, where the real-time predictions are combined with the measurements of the available diagnostics, yielding a complete estimate of the plasma profiles.The state ... More
Presented by Chiara PIRON on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 39
The Reversed Field Pinch configurations are characterized by strong asymmetries [1]; in order to prevent or mitigate possible consequent instabilities, suitable control systems are required. In RFX-mod (Padua, Italy), such a system includes a number of 192 saddle coils, independently controlled, fully covering the toroidal surface and operating in a coordinate strategy. An equal number of saddle p ... More
Presented by Raffaele MARTONE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 40
RFX-mod is equipped with an advanced active control system of MHD instabilities, which consists of 48x4 saddle coils, housed inside a stainless steel Toroidal Support Structure, and 48x4 radial field sensor loops processed in real time to drive the currents in the control coils. Thanks to the high flexibility of this system [1], RFX-mod operations in the last years have allowed to reach the design ... More
Presented by Paolo BETTINI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 41
RFX [1] was originally designed with a load assembly consisting of a vacuum vessel (VV) and a thick aluminum stabilizing shell, with two poloidal and two equatorial cuts (i.e. gaps). After several years of experimental campaigns, a major modification of the RFX load assembly has been introduced [2], consisting in the substitution of the aluminum shell with a thin Copper Shell (CS) and the installa ... More
Presented by Luca GRANDO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 42
The Parallel plasma equilibrium reconstruction code PEFIT [1], first developed for real-time plasma shape control of the EAST tokamak (and capable of one full equilibrium reconstruction in 300ms with a calculation grid size in 65x65) is being adapted for use on MAST. PEFIT is based upon the EFIT equilibrium code algorithm, but rewritten in C using the CUDA<sup>TM</sup>TM architecture in order to t ... More
Presented by Zhengping LUO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 43
Mitigation of heat and particle fluxes reaching on divertor plates is still a critical problem even though innovative divertor concept such as super-X and snowflake divertors have been suggested. A new divertor concept for the reduction of heat and particle fluxes is to convert thermal energy to electrical energy by separating electrons from the plasma with appropriate magnetic field. Feasibility ... More
Presented by Seongcheol KIM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 44
The control of the safety factor q and/or the electronic temperature Te profiles is a key issue to achieve advanced plasma scenarios with high repeatability. This paper will discuss the new results of such plasma internal profile control on TCV, using total plasma current Ip, and ECCD heating source. The issue is that only the ECCD heating power is controlled, since the distributed heating profile ... More
Presented by Ngoc Minh Trang VU on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 45
Presented work is related to the development and creation of hardware and software of Plasma Control System (PCS) platform of the modernized now tokamak T-15 [1] for the integration, configuration, testing and start-up algorithms for the calculation of electrical installation parameters, as well as for the modeling of the experiment scenario with taking into account of the real-time magnetic plasm ... More
Presented by Galina KUZMINA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 46
Noise width (δV/V) and drift level (ΔV/Δt) in the magnetic measurements by using sensors such as magnetic field probes (MPs) and flux loops (FLs) has been fully satisfied with the requirements (δV/V < 2% and (ΔV/V)/ Δt < 2% for 60 s), for the plasma control in the KSTAR tokamak before the in-vessel control coil (IVCC) is used to control plasma shapes. From the experimental campaign of 2010, ... More
Presented by Heung-Su KIM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: C. Plasma Engineering and Control Board #: 47
In order to avoid surface melting of divertor targets of big tokamak fusion reactors by localized ELM heat loads, we study a technique of spreading the flux by harmonic divertor strike point sweeping with a dedicated in-vessel twin-coil. If the sweep frequency gets above 1/t<sub>ELM</sub><sup>decay</sup>decay~300 Hz, local ELM plasma heat flux suppresses significantly (by factor=1+2λsweep/λdivet ... More
Presented by Jan HORACEK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 48
This paper presents new results of neutronics analysis performed in support for the design development of the Tritium and Deposit Monitor (TDM) to be installed inside the ITER Equatorial Port Plug (EPP) #17. This monitor is a laser based diagnostics to provide information about the tritium content in the deposited layer on the inner baffle of the ITER divertor. Neutronics analysis is performed wit ... More
Presented by Arkady SERIKOV on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 49
The ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations, also known as gaps 3, 4, 5, and 6, complementing the magnetic diagnostics system. For gaps 4 and 6, the antennas are to be installed in-vessel between two blanket shield modules. The microwave signal is routed to ... More
Presented by Raul LUIS on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 50
The Radial Neutron Camera (RNC) diagnostic is a neutron detection system with multiple collimators aiming at characterizing the neutron emission that will be produced by the ITER tokamak. The RNC plays a primary role for basic and advanced plasma control measurements and acts as backup for system machine protection measurements. To achieve its goals, the RNC diagnostic needs to acquire, process an ... More
Presented by Nuno CRUZ on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 51
The ITER Radial Neutron Camera (RNC) is a multichannel detection system hosted in the Equatorial Port Plug 1 (EPP 1) designed to provide information on the neutron source total strength and emissivity profiles through the measurement of the uncollided neutron flux along a set of collimated lines of sight (LOS). Furthermore the ion temperature profile and fuel ratio (nd/nt) can be assessed by means ... More
Presented by Fabio MORO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 53
The High Resolution Neutron Spectrometer (HRNS) system for ITER is an array of neutron spectrometers with the primary function to provide measurements of the fuel ion ratio, nT/nD, in the plasma core. Supplementary functions are to assist or provide information on fuel ion temperature and energy distributions of fuel ions and confined alpha-particles. The ITER requirement for the HRNS primary func ... More
Presented by Anders HJALMARSSON on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 56
The COMPASS tokamak is equipped by the 2-mm microwave interferometer. This interferometer measures the electron density integrated along the central chord. Two VCO oscillators stabilized by the PLL together with multipliers generate two probing waves of the close frequency 139.3 and 140 GHz. The digital 2π-phase detector in the receiving part compares the phase between these probing waves. The re ... More
Presented by Mykyta VARAVIN on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 57
Atomic beam probe (ABP) is a diagnostic tool using a detection of ions coming from an ionized part of a diagnostic beam in tokamaks. The method allows measurements of plasma density fluctuations and fast variations in the poloidal magnetic field. Therefore, it gives the possibility to follow fast changes of edge plasma current, e.g. during ELMs in H-mode. The test detector has been installed on th ... More
Presented by Pavel HACEK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 58
The COMPASS tokamak has been recently equipped with two new fast color cameras Photron FASTCAM Mini UX100 operating in visible light. A new node, including both software and hardware, was developed for these cameras to ensure automatic and reliable operation integrated to the control and data acquisition system of COMPASS. The node provides camera function control, parameter setting, data transfer ... More
Presented by Ales HAVRANEK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 59
A new fast infrared camera Telops FAST-IR 2K was purchased on the COMPASS tokamak recently. It is equipped with a MWIR (medium wavelength infrared, 3-5 μm) InSb detector and is possible to reach framerate of 1.917 kHz in a full frame acquisition mode (320x256 px.) and up to 90 kHz in a sub-windowed acquisition (64x4 px.). The camera allows e.g. automatic exposure control, providing autonomou ... More
Presented by Petr VONDRACEK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 60
The microwave reflectometry system on COMPASS tokamak uses the frequency modulated continuous wave (FM-CW) in K and Ka bands. The fast swept synthesizer together with the simple homodyne detection provides the complex beat frequency spectrum for the density profile reconstruction. The homodyne detection scheme limits the other applications like the Doppler reflectometry, therefore the sheme is reb ... More
Presented by Jaromir ZAJAC on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 61
The physics of Edge Localized Modes (ELM) is one of the most studied scientific fields in fusion research. Automatic detection of ELMs in different diagnostic signals is an important initial step during massive experimental data analysis. This contribution contains the description of the generalized Sequential Probability Ratio Test (g-SPRT) method used for automatic ELM detection in different dia ... More
Presented by Mark SZUTYANYI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 62
WENDELSTEIN 7-X and its superconducting coil system is designed for research on steady-stateoperation of stellarators. This sets high requirements on the control and data acquisition (CoDaC)system, with the archive database as one of its main components. W7-X ArchiveDB [1] is the centralstorage system for all engineering and scientific data. It stores raw data as well as processed data andprovides ... More
Presented by Michael GRAHL on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 63
Wendelstein 7-X (W7-X) has been finally commissioned in 2015 and is now in its first stage of operation. Due to the complex structural design and a limited life time of some components, each step of W7-X commissioning and operation is carefully monitored by a considerable amount of different sensors. Unlike the fast machine control or the fast experiment data acquisition, the machine instrumentati ... More
Presented by Andre CARLS on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 64
WENDELSTEIN 7-X (W7-X) is a superconducting helical advanced stellarator which is currently in operation phase 1.1 at the Max-Planck-Institut für Plasmaphysik in Greifswald. During this startup period five uncooled inboard poloidal limiter structures made from fine corn graphite protect the plasma vessel wall, since the divertor, heat shields and carbon tiles are not installed yet. At 10 ports im ... More
Presented by Dirk PILOPP on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 65
The Wendelstein 7-X fusion device at Max-Planck-Institut für Plasma Physik (IPP) in Greifswald produced its first hydrogen plasma on 3<sup>rd</sup>rd February 2016. This marks the start of scientific operation. Wendelstein 7-X is to investigate this configuration’s suitability for use in a power plant. In order to allow for an early integral test of the main systems needed for plasma operation ... More
Presented by Didier CHAUVIN on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 66
The C/O monitor for W7-X will be a spectrometer of special construction with high throughput and high time resolution, suitable for controling concentration of main impurities in plasma. The spectrometer will be fixed at horizontal position and at wavelengths corresponding to Lyman a lines of H-like ions of oxygen (at 1.9 nm), nitrogen (at 2.5 nm), carbon (at 3.4 nm) and boron (at 4.9 nm). Its pur ... More
Presented by Ireneusz KSIAZEK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 67
A multi-purpose manipulator (MPM) system is attached at an outer cryostat vessel port in the equato­rial plane to transport electrical probes and targets to the edge of the inner vessel. From this parking position where the tip of the probe coincides with the inner vessel wall a fully controlled movement into the edge plasma for all magnetic field configurations is feasible. The distributed contr ... More
Presented by Guruparan SATHEESWARAN on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 68
In the past few years a ten channel video diagnostics system was developed, built and installed for Wendestein 7-X stellarator (W7-X). The system is based on EDICAM (Event Detection and Intelligent Camera) CMOS cameras (400 fps @ 1.3 Mpixel).  In the first W7-X experimental campaigh (OP1.1) the video diagnostic  system is not integrated into the central control and data acquisition system of W7- ... More
Presented by Tamas SZABOLICS on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 69
Measurements of soft X-ray radiation from plasmas is a standard diagnostic which is used in many different fusion devices. Analysis of X-ray emission delivers among others, information about the electron density and temperature as well as can deliver an information about the impurity content in the plasma. The paper describes design of the soft X-ray diagnostic, multi-foil system (MFS,) for the st ... More
Presented by Tomasz FORNAL on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 70
The Wendelstein 7-X (W7-X) stellarator started its operation at the end of 2015. The first operation phase is conducted both with helium and hydrogen as working gas and has achieved first plasmas in the order of 500ms at the time this abstract has been written. The initial experiments have also been devoted to commissioning, tests and optimization of diagnostic systems. In this paper we report on ... More
Presented by Natalia KRAWCZYK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 71
The quasi-steady state high power plasma experiments at Wendelstein 7-X are expected to become pioneering research benchmarking the advanced stellarator concept. The results will bring comparisons to the huge amount of experimental findings in other stellarator and tokamak devices. After the successful start of hydrogen plasmas in February 2016, the set of plasma diagnostics will be extended durin ... More
Presented by Christian BRANDT on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 72
Thirteen Rogowski coils have been installed in the vacuum vessel of the stellarator Wendelstein 7-X (W 7-X). They are designed to measure the equilibrium plasma currents as Pfirsch-Schlüter current and bootstrap current. The coils will be calibrated using a conductor positioned inside the plasma vessel with an alternating current passing through it. The response of the coils is measured and compa ... More
Presented by Ulrich NEUNER on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 73
The investigation of edge plasmas at W7-X requires a flexible tool for integration of a variety of different diagnostics as e. g. electrical probes, probing magnetic coils, material collection, or material exposition probes, and gas injection. A multi-purpose manipulator (MPM) system has been developed and attached to the W7-X vessel before the operational phase 1.1. The system was designed as use ... More
Presented by Dirk NICOLAI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 74
Long-pulse D-D plasma operation in the annual KSTAR plasma campaign is performed and involved Ohmic heating and auxiliary heating such as a neutral beam injection (NBI) of high power with deuterium beams. The NBI heating power reached up to 6 MW at the moment. In addition, many energetic runaway electrons are also observed through hard-X ray (HXR) monitoring during the operation. Runaway electrons ... More
Presented by Youngseok LEE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 75
To measure Zeff profile, most plasma machine equipped brehmsstrahlung measurement system like as filterscope diagnostic. In KSTAR, however, a new type brehmsstrahlung measurement system developed and tested at single point in KSTAR 10th campaign in last year.[1] In 2016 KSTAR campaign, to Zeff profile measurement, we expand this concepts of brehmsstrahlung measurement system to multi points; two f ... More
Presented by Jong-ha LEE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 76
Abstract:In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum wa ... More
Presented by Y. YU on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 78
In QUEST (Q-shu University Experiments with Steady-State Spherical Tokamak), the achievement of the steady-state operation for long time discharge is one of its project objectives. For the achievement of the long time discharge, the identification of the plasma shape and position in real-time is important during the operation of the tokamak. By observing the temporal behaviours of the plasma shape ... More
Presented by Md Mahbub ALAM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 79
This paper describes the final design of the Short-Time Retractable Instrumented Kalorimeter Experiment (STRIKE) for the SPIDER experiment (Source for Production of Ions of Deuterium Extracted from Radio frequency plasma) under construction at the Consorzio RFX premises. The STRIKE diagnostic will be used to characterise the SPIDER beam during short pulse operation (several seconds) to verify the ... More
Presented by Andrea RIZZOLO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 80
Neutron measurements are proposed for the SPIDER/MITICA Neutral Beam Injection (NBI) prototypes in Padua. Neutron emission is here due to reactions between the beam and the adsorbed deuterons in the target and thus depends on the deuteron absorption level in the beam calorimeter. We have investigated such process at the “half size” ITER NBI ELISE facility of the Max-Planck Institut. A first me ... More
Presented by Gabriele CROCI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 81
JT-60SA is a Superconducting Tokamak in the framework of the Broader Approach Agreement between Europe and Japan. For this International Project, both the Italian National Agency for New Technology, Energy and Sustainable Economic Development (ENEA) and Comissariat à l’Energie Atomique et aux Energies alternatives (CEA) are providing ten AC/DC converters for the poloidal superconducting magnets ... More
Presented by Zito PIETRO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 82
Switching Network Units (SNUs) are inserted in the power supply circuits of modern tokamaks for plasma initiation. In the framework of the “Broader Approach” agreement, the four SNUs for the superconducting modules of the JT-60SA Central Solenoid will be procured by European Union through the Italian Agency ENEA. The design is based on the synchronized operations of a light electromechanical c ... More
Presented by Miguel PRETELLI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 83
Effective control of Resistive-Wall-Modes (RWM) is mandatory in JT-60SA, the satellite tokamak under construction in Naka (Japan), since one of its main objectives is to reach steady-state high-beta plasmas. The RWM control system is based on a set of 18 in-vessel sector coils, placed on the plasma side of a conductive wall and individually fed by a dedicated fast power supply system (RWM-PS). For ... More
Presented by Elena GAIO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 84
JT-60SA is a tokamak device using superconducting coils to be built in Japan, as a joint international research and development project involving Japan and Europe. The JT-60SA helium refrigerator system (HRS) supplies supercritical or gaseous helium to cold components: superconducting coils, coil supporting structures, cryopumps, high temperature superconductor current leads (HTS CL), and thermal ... More
Presented by Kyohei NATSUME on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 85
The programme of constructing JT-60SA device is progressing as a satellite tokamak of the Broader Approach project. JT-60SA has superconducting poloidal field (PF) coil system which is procured by JAEA, and 18 D-shaped toroidal field (TF) coils of which Europe has been in charged. PF coil system consists of a central solenoid (CS) with four solenoid modules and six circular coils which are utilize ... More
Presented by Katsuhiko TSUCHIYA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 86
In the framework of the Broader Approach program, ENEA is in charge of the in-kind supply of 18 Toroidal Field (TF) coil casings for the superconducting tokamak JT-60SA being assembled in Naka site, Japan. ENEA commissioned the company Walter Tosto (Chieti, Italy) the fabrication of two sets of 9 casings to be delivered to ASG Superconductors (Genoa, Italy) and GE (Belfort, France), in charge the ... More
Presented by Paolo ROSSI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 87
The Toroidal Field system of the JT-60SA tokamak comprises 18 NbTi superconducting coils. In each TF coil (TFC), 6 Cable-In-Conduit Conductor (CICC) lengths are wound in 6 double-pancakes (DP) and carry a nominal current of 25.7 kA at a temperature of 5 K. These coils are tested in the Cold Test Facility (CTF, CEA Saclay), the test program including a quench for each of the first coils of the two ... More
Presented by Sylvie NICOLLET on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 88
ENEA, in the framework of Broader Approach program for the early realization of fusion with the construction of JT-60SA tokamak, has committed to procure 9 of the 18 TF coils of JT-60SA magnet system. Within 2016 six coils will be completed and delivered to the cold test facility in Saclay, France, for the final acceptance tests before their shipment to Naka site for the assembly. Manufacturing ha ... More
Presented by Gian Mario POLLI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 89
The Toroidal Field system of the JT-60SA tokamak is composed of 18 NbTi superconducting coils. Half of them are provided by France within the Broader Approach Agreement. These coils are manufactured by General Electric (ex-Alstom) at Belfort, France. Each TF coil is composed of 6 cable-in-conduit conductor lengths, wound in double-pancakes, carrying a nominal current of 25.7 kA at a temperature of ... More
Presented by Daniel CIAZYNSKI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 90
In order to check the performance of the JT-60SA Toroidal Field (TF) coils and hence mitigate their possible fabrication risks, a series of tests have been carried out in the Cold Test Facility (CTF) at CEA Saclay in nominal conditions at 5 K and 25.7 kA. One major test performed is the so called “temperature margin test" during which the inlet helium temperature of the winding pack is increased ... More
Presented by Yawei HUANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 91
In the frame of the Broader Approach, CEA provides 9 + 1 spare TF coils for the JT-60SA tokamak. Mid 2011, a manufacturing contract was awarded to Alstom, Belfort, now General Electric. The first years were dedicated to the manufacturing process definition, the critical phases qualification through a set of 12 mockups, the manufacturing QA definition and the procurement and commissionning of the t ... More
Presented by Patrick DECOOL on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 92
Coil casings and coil frames for stellarators are geometrically complex components at high accuracy. A method of additive manufacturing combined with fibre-reinforced resin casting has been recently experimented [1] for the fabrication of complex coil frames. The method is named 3Dformwork and consists of additive fabrication of a hollow thin shell which is filled with resins or other appropriate ... More
Presented by Vicente QUERAL on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 93
The Quench-Detection-System of the fusion experiment Wendelstein 7-X detects quench events within the superconducting magnet system constructed of 50 non-planar and 20 planar coils, 14 current leads and the bus bars. In the event of a quench the QD-System triggers the power supply of the magnetic system to shut down. The QD-System monitors the superconducting system by 486 Quench-Detections-U ... More
Presented by Matthias SCHNEIDER on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 94
The magnet system of the stellarator fusion device Wendelstein 7-X (W7-X) is composed of three different groups of coil systems. The main magnetic field is created by a superconducting magnet system that is accompanied by two sets of normal conducting coil groups, the Control Coils inside the plasma vessel and the Trim Coils (TC) positioned outside of the cryostat. The TC system consists of five c ... More
Presented by Frank FULLENBACH on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 95
The quench protection switch (QPS) is very important for ensuring the safety of the PF and TF coils of a superconductive Tokomak. The main function of a QPS is to protect the magnet as the coil quench occurs. Besides, a QPS has to withstand almost all of the coil current of some tens of kA flowing through it for a long time in the normal operation condition. This task is undertaken by the by-pass ... More
Presented by Sheng LI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: E. Magnets and Power Supplies Board #: 96
Superconducting magnet is one of the most crucial components in a superconducting Tokamak. During the normal operation stage, high current of some tens of kA flows through the magnet with large inductance of ~1H. Therefore, extremely large energy (~0.1-10GJ) is stored in the magnet, which must be dissipated in the case of magnet quench in certain duration before the occurrence of local or even ove ... More
Presented by Qiaosen WANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 101
The Normal Heat Flux (NHF) First Wall (FW) panels consist of a series of fingers, which represent the elementary plasma facing units and are designed to withstand 15,000 cycles at 2 MW/m<sup>2</sup>2. The fingers are mechanically joined and supported by a back structural element called “supporting beam”. The structure of a finger is made of three different materials, stainless steel for the su ... More
Presented by Tindaro CICERO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 102
This paper describes the main activities carried out for the conclusion of the EU-DA prequalification process for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the High Heat Flux (HHF) testing of a reduced scale FW prototype (Semi-Prototype (SP)). This component is manufactured by the AREVA Company in France and has a dimension of 221 x 665 ... More
Presented by Stefano BANETTA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 103
The Normal Heat Flux (NHF) First Wall (FW) panels are designed to withstand the heat flux from the plasma inside ITER. These components are made of beryllium tiles bonded to a copper alloy and 316L (N) stainless steel heat sink. A NHF FW panel consists of several fingers as elementary plasma facing units. This this paper presents the experimental stress and deformations measured on a 10-fingers mo ... More
Presented by Rafael ENPARANTZA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 104
In the framework of PA realization, specialists from NIKIET and Efremov Institute are developing a design of First Wall (FW) Full Scale Prototype (FSP) in order to demonstrate its manufacturability and qualify critical technological processes. Design of FW FSP is developed based on the FW 14 type A. The semi-prototype has been manufactured in order to verify the FW design. Based upon the manufactu ... More
Presented by Sergey TOMILOV on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: B. Plasma Heating and Current Drive Board #: 106
The JSC NIKIET is responsible for the manufacture of the First Wall (FW) beam, the fingers bodies, the mechanical attachment system and electrical connection system of the FW panel to the shield block (SB) in the framework of Procurement Arrangement 1.6.Р1А.RF.01 dated 14.02.2014. The Electrical strap (ES) is located on the FW rear surface and used for providing current through the FW to the SB ... More
Presented by Maxim SVIRIDENKO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 107
The heat flux on plasma-facing components in ITER, and even more so in the projected DEMO reactor will reach values in the order of several Megawatt per square meter. Evacuating this heat in a reliable manner is key to the robustness and safety of operation of any fusion reactor. The current state-of-the-art for cooling plasma-facing components relies on cooling a high heat-resistant structure us ... More
Presented by Karel SAMEC on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 108
The heat loads on the First Wall (FW) of the European DEMO are not yet defined, but when extrapolated from ITER, the loads can be quite high. As the DEMO will use Eurofer 97 as the structural material and Pressurized Water Reactor (PWR) conditions at the inlet, i.e. 15.5 MPa and 285 °C, the design of the heat sink gets complicated as the thermal conductivity of the heat sink material is quite low ... More
Presented by Phani DOMALAPALLY on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 109
The first wall cooling of the fusion power reactor DEMO is an important part of the fusion power plant development. A cooling ability at high heat flux conditions will affect a lifetime period of the first wall modules having a direct impact on the operating costs of the fusion power plant. According to current knowledge, the water cooling provides the largest ability to remove the high heat flux ... More
Presented by Pavel ZACHA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 110
Based on the requirements of F4E, an experimental device HELCZA (High Energy Load Czech Assembly) was designed for high heat flux cyclic loading of plasma-facing components of the ITER reactor, primarily for testing of the full-size first wall modules and divertor inner vertical targets. Testing is carried out by a high power electron beam heating, and a deviation of the heat flux density at any ... More
Presented by Ladislav VESELY on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 111
The paper deals with optimal electron beam heat distribution on the HELLCZa experiment calculating the flatness of the distribution of heat input and distribution of surface temperature of various samples. A computer program has been developed for balancing the heat flux in the construction materials of the sample. The first boundary condition for this calculation were primarily functions describi ... More
Presented by Radek SKODA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 112
Commissioning phase of high heat flux test facility HELCZA R. Jílek<sup>a,*</sup>a,*, J. Prokůpek<sup>a</sup>a, P. Gavila<sup>b</sup>b aCentrum výzkumu Řež s.r.o. (CVR), Hlavní 130, 25068 Husinec-Řež, Czech Republic, bFusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona, Spain *Corresponding author: e-mail: Richard.Jilek@cvrez.cz, phone: +420 601 315 137 The hig ... More
Presented by Richard JILEK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 113
For the testing of helium cooled plasma facing components in HELOKA-HP homogeneous surface heat flux densities of up to 500 kW/m² have to be reproduced. It has been proposed to use infrared radiation heaters which consist of several quartz glass (fused silica) tubes with tungsten filaments inside to generate the heat flux. This paper presents a numerical model of the latest type of heater which h ... More
Presented by Andre KUNZE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 114
Maintenance of structural integrity under high-heat-flux (HHF) fatigue loads is a critical concern for assuring the reliable HHF performance of a plasma-facing divertor target component. Loss of structural integrity may lead to structural as well as functional failure of the component. Currently, a full tungsten divertor was chosen by ITER Organization, and plenty of HHF qualification tests have b ... More
Presented by Muyuan LI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 115
Transient heat fluxes onto the tungsten divertor targets during disruptions in ITER may cause severe melting, leading to intolerable damage. However, for sufficiently energetic transients, tungsten vaporized from the target in the initial stage of the heat pulse will generate a protective plasma shield in front of the target, greatly reducing the incoming heat flux. This vapour shielding is a comp ... More
Presented by Sergey PESTCHANYI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 116
This paper deals with the first commissioning of active cooling system for plasma-facing components (PFCs) and coolant removal system. During 2015 KSTAR campaign, we have achieved a 55 sec long pulse H-mode. However, some plasma shots were terminated, not because of instabilities or limitation of heating power, but because of safety limit applied to the PFC temperature: upper boundary to lock the ... More
Presented by Eunnam BANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 117
The tungsten (W) brazed flat type mock-up with swirl tube which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade. The mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 8 MW/m<sup>2</sup>2 for 20 sec duration at KoHLT-EB in KAERI. In this paper, for comparison of two ... More
Presented by Jaehyun SONG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 118
The preliminary conceptual design study on the Korean fusion demonstration reactor (K-DEMO) tokamak consists of the vacuum vessel, the in-vessel components, and the superconducting magnet system, and so on [1]. The K-DEMO superconducting magnet system contains 16 toroidal field (TF) coils, 8 central solenoid (CS) coils and 12 poloidal field (PF) coils. The magnetic field at the plasma center is ab ... More
Presented by Sungjin KWON on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 119
A preliminary study on the rigorous 2-step (R2S) based shutdown dose rate calculations has been performed for the Korean fusion demonstration reactor (K-DEMO) in the vicinity of an equatorial port area using the coupled transport and activation calculation codes of MCNP6 and FISPACT. For the shutdown dose rate calculation, the equatorial port structures and port plug including shielding blocks wer ... More
Presented by JongSung PARK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 120
A pre-conceptual design study for the Korean fusion demonstration tokamak reactor (K-DEMO) has been initiated in 2012. K-DEMO is characterized by the uniqueness of high magnetic field (BT0 = 7.4 T), major and minor radii of 6.8 m and 2.1 m, and steady-state operation. The heat load distribution by plasma radiation onto the first walls of the in-vessel components is one of the basic inputs for the ... More
Presented by Kihak IM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 121
G. Douglas Loesser1,Joris Fellinger<sup>2</sup>2, Hutch Neilson<sup>1</sup>1, John Mitchell<sup>1</sup>1, Marc Sibilia<sup>1</sup>1, Han Zhang<sup>1</sup>1, P. Titus<sup>1</sup>1, Irving Zatz<sup>1,</sup>1,, Arnie Lumsdaine<sup>3</sup>3, Dean McGinnis<sup>3</sup>3 1Princeton Plasma Physics Laboratory, James Forestall Campus, Princeton, NJ 08542, USA 2Max-Planck-Institut für Plasmaphysik, Teilinst ... More
Presented by G Douglas LOESSER on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 122
The modular stellarator Wendelstein 7-X in Greifswald (Germany) successfully started operation in 2015 with short pulse limiter plasmas. In 2017, the next operation phase (OP) OP1.2 will start once 10 uncooled test divertor units (TDU) with graphite armor will be installed. The TDUs allow for plasma pulses of 10 s with 8 MW heating. OP2, allowing for steady state operation, is planned for 2020 aft ... More
Presented by Joris FELLINGER on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 123
The cryopump will be installed for the high power and long pulse operation up to 30 minutes of Wendelstein 7-X (W7-X). The cryopump system plays a critical role for capturing ash particles from the plasma, including hydrogen, deuterium and even helium. In total there are 10 independent cryopumps, one cryopump for each of the 10 discrete divertor units. The cryopump is located along the pumping gap ... More
Presented by Zhongwei WANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 124
The 890 target elements of the high heat flux divertor of Wendelstein 7-X (W7-X) are made of a CuCrZr copper alloy heat sink armored with carbon reinforced carbon (CFC) NB31 tiles. Connectors with an internal diameter of 10 mm are electron beam welded to heat sink for the water inlet and outlet. They are produced by electron beam welding thicker tubes of CuCrZr and stainless steel with a Nickel 27 ... More
Presented by Patrick JUNGHANNS on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: F. Plasma Facing Components Board #: 125
The actively water-cooled target elements of the high heat flux divertor of Wendelstein 7-X (W7-X) are designed to remove a stationary heat flux of 10 MW/m² on its main area and 5 MW/m² at the end adjacent to the pumping gap. A target element is made of a CuCrZr copper alloy heat sink armored with carbon reinforced carbon (CFC) NB31 tiles. The realization of the divertor requires the production ... More
Presented by Jean BOSCARY on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 126
The ITER Vacuum Vessel (VV) is a double wall Stainless Steel structure that surrounds the plasma. It constitutes a major safety barrier for ITER, and, because of its function, is classified as Protection Important Component (PIC). Its design and construction has to follow the RCC-MR design code rules to verify the structural integrity under electromagnetic, thermal and seismic loads. Computation F ... More
Presented by Clara COLOMER on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 127
In ITER the neutron activation system deploys several foil samples close to the plasma to measure the neutron fluence and the fusion power. These samples are transferred in a pneumatic way along the tubes installed on the vacuum vessel wall. Therefore, the tubes, namely transfer lines, get eddy current induced during plasma disruption, leading to Lorentz force by interacting the background magneti ... More
Presented by Sunil PAK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 128
The vacuum vessel of ITER is a paradigmatic example of a gargantuan system that can only be processed in-situ and from the inside. Its assembly implies performing post welding repair operations, including machining of welding seams following the internal surface of the vacuum vessel. The requirements for the machining operations are the following: accuracy +/- 0.1 mm; dynamic machining forces 3 kN ... More
Presented by Josu EGUIA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 129
ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just behin ... More
Presented by Anna ENCHEVA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 130
In ITER blanket system, electrical connectors (“E–straps”, ES) are used to form a low impedance electrical path from shield blocks (SB) to the vacuum vessel (VV). Main functions of ES is providing current from SB to VV. ES shall withstand electromagnetic (EM) loads and Joule heating resulted from electrical current with magnitude up to 137 kA during 300 ms, accommodate cyclic relative displ ... More
Presented by Ivan PODDUBNYI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 131
The Ion Cyclotron Heating and Current Drive system (ICH) is designed to launch RF power into the ITER plasma, and will reside in equatorial ports (EP) 13 and 15. Shutdown dose rates (SDDR) within the ICH port interspace are required to be ALARA and less than 100 μSv/h at 10<sup>6</sup>6 seconds cooling, in locations where hands-on maintenance is required. The shielding performance of in-vessel, v ... More
Presented by Dieter LEICHTLE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 132
In this paper the stress-strain state of the diagnostic shield modules (DSM) and the supporting frames (ISS, PCSS), located in the upper ports #2 and #8 of the tokamak ITER is investigated. DSM is the upper port components and has two main functions: neutron radiation protection and maintenance of rigid fixation diagnostics placed in the port. DSM is operated at high temperatures, significant ele ... More
Presented by Ivan POPOV on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 133
The primary systems of future international thermonuclear experimental reactor (ITER) have to withstand major thermal, nuclear, electromagnetic and seismic loads. Therefor engineering analysis of elements of construction plays crucial role in realizing of the project as a whole. The paper describes calculations of spatial stress-strain state from major loads arising during operation upper vertical ... More
Presented by Pivkov ANDREW on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 134
Korea has been manufacturing two vacuum vessels of ITER and main jointing method to in-wall shield assemblies is welding. Though in-wall shield ribs holding neutron shielding blocks should sustain various design loads such as electro-magnetic forces, earthquake and their own weights, as a part of the assembly, in-service inspections are hardly possible because they are installed between double-wal ... More
Presented by Yu-Gyeong KIM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 135
After the Conceptual Design Review (CDR), Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) design is being updated for the preparation of the preliminary design phase. The manufacturability is considered based on the TBM-set model of CD phase. The overall geometry of the first wall, side wall and the breeding zone was changed slightly. Thethermal-hydraulic and mechanical analysis a ... More
Presented by Dong Won LEE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 136
The High Field Side Reflectometry is diagnostic equipment subjected to the conditions that are severe even for ITER: magnetic field over 9T, temperatures up to 700 ºC, strongly non-uniform temperature field, specific shape of the equipment with length of in-vessel waveguides about 10m and location of waveguides close to the blanket connectors where large halo currents are expected during disrupti ... More
Presented by Aleksandr NEMOV on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 137
The presentation is focused on the simulation results and approaches used for loading analyses made for DTS in-vessel equipment, including spatial stress strain state, seismic analysis, electromagnetic analysis as well as the most important load combinations. Finite element model of the construction was updated according with updated DTS components design and separated on the following constructio ... More
Presented by Ivan KIRIENKO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 138
HL-2M RMP (Resonance Magnetic Perturbation) Coils is designed to provide a resonant perturbation magnetic field for high beta plasma operation scenarios stability control, such as Edge Localized Modes (ELMs) suppression control, Resistance Wall Model (RWM) fast control and Error magnetic field correction control, etc.  Especially, ELMs result in impulsive burst of energy deposition on to the “P ... More
Presented by Jiaming JIANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 140
Thermal radiation analysis of the DEMO tokamak based on the updated CAD design of in-vessel components and magnet system has been carried out. For the purpose of the analysis, Vacuum Vessel Thermal Shield (VVTS), Cryostat Thermal Shield (CTS) and some support structures have been created additionally (on a conceptual level) to complement the overall DEMO CAD design model. The Finite Element (F ... More
Presented by Bostjan KONCAR on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 141
This study is a part of the structural activity being conducted in the framework of the structural design of a DEMO Divertor. The thermal and structural analysis has already been started since a year and the first results has been partly published in a previous paper. The Cassette Body is being analyzed considering the most critical types of loads (e.g. coolant pressure, volumetric neutron heatin ... More
Presented by Paolo FROSI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 142
Shutdown dose rate (SDR) analysis plays a key role in the design of fusion facilities like ITER and DEMO. One of most used methodology to carry out SDR calculations is the rigorous-two-step (R2S) based on the coupling of transport and activation calculations. Currently, one of the most relevant lacks of this method is the possibility to propagate the effect of the uncertainties accumulated along t ... More
Presented by Juan-Pablo CATALAN on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 143
Molybdenum disulfide (MoS2) coating was deposited by magnetron sputtering onto the target material. The coatings of deposited MoS2 can be used in high vacuum and aerospace environments for lubrications purposes, which ultra-low friction is desirable. For these reason, the sputtered MoS2 coating method is primarily considered for ITER components and their mechanical assemblies. A common deposition ... More
Presented by Heejin SHIM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: G. Vessel/In-Vessel Engineering and Remote Handling Board #: 144
A new technique, called Vacuum Tight Threaded Junction (VTTJ), has been developed and patented by Consorzio RFX, permitting to obtain low-cost and reliable non welded junctions, able to maintain vacuum tightness also in aggressive environments. The technique can be applied also if the materials to be joint are not weldable and for heterogeneous junctions (for example, between steel and copper) and ... More
Presented by Piero AGOSTINETTI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 146
Within the framework of EUROfusion R&D activities an intense research campaign has been carried out at the University of Palermo, in close cooperation with ENEA Brasimone, in order to investigate the thermo-mechanical performances of the DEMO Water-Cooled Lithium Lead breeding blanket (WCLL). In particular, attention has been paid to the most recent geometric configuration of the DEMO WCLL outboar ... More
Presented by Pietro Alessandro DI MAIO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 147
The development of a system-level thermal-hydraulic model of the whole EU DEMO tokamak has been launched by the EUROfusion Project Management Unit. In order to follow the progress in the design of the tokamak components, the model should be developed in an object-oriented fashion, to ensure a high modularity. Within this framework, the first block of the model is under development at Politecnico d ... More
Presented by Antonio FROIO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 148
The Breeding Blanket is a key component in a fusion power plant in charge of ensuring tritium breeding, neutron shielding and energy extraction. Water-Cooled Lithium-Lead Breeding Blanket (WCLL) is considered a candidate option in view of the risk mitigation strategy for the realization of DEMO. Indeed, this design might benefit of efficient cooling performances of water as coolant, as well as of ... More
Presented by Nicola FORGIONE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 149
Within the framework of EUROfusion Power Plant Physics & Technology Work Programme, the Water Cooled Lithium Lead (WCLL) is one of the four breeding blanket (BB) concepts considered as possible candidate for the realization of DEMO fusion power plant. ENEA CR Brasimone has developed during 2015 a new design of the outboard module based on horizontal (i.e radial-toroidal) water cooling tubes in the ... More
Presented by Emanuela MARTELLI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 150
The interaction between the molten metal and the plasma-containing magnetic field in the breeding blanket of a Tokamak fusion reactor causes the onset of a magnetohydrodynamic (MHD) flow. In order to properly design the blanket, it is important to quantify how and how much the flow features are modified compared with an ordinary hydrodynamic flow. This paper aims to characterize the evolution of t ... More
Presented by Alessandro TASSONE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 151
A number of liquid metal blanket designs for applications in nuclear fusion reactors is currently under development. In the water cooled lead lithium (WCLL) blanket Eurofer97 is used as structural material and liquid PbLi as breeder, neutron multiplier, and as heat transfer medium. The released heat is removed by water at a pressure of 155 bar (pressurized water reactor conditions, 285°C - 325°C ... More
Presented by Leo BUHLER on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 152
Within the framework of EUROfusion R&D activities an intense research campaign has been carried out at the University of Palermo, in close cooperation with ENEA Brasimone, in order to investigate the thermo-mechanical performances of the Back-Supporting Structure (BSS) outboard segment of the DEMO Water-Cooled Lithium Lead breeding blanket (WCLL). In particular, the configuration of the BSS outboa ... More
Presented by Maria Lorena RICHIUSA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 153
The interaction between lithium-lead and water is a major concern of Water Coolant Lithium Lead (WCLL) breeding blanket design concept, therefore deterministic safety analysis of the in-box LOCA postulated accident is of primary importance. In this framework, a past experimental campaign was carried out in LIFUS5 to investigate the evolution and the consequences of the interaction. Then, these dat ... More
Presented by Marica EBOLI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 154
China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokomak reactor. Its major radius is 5.7m, minor radius is 1.6m and elongation ratio is 1.8.  It is possible upgrade to R~6 m, a~2 m. CFETR mission and objectives are to bridge gaps between ITER and DEMO, and to realize fusion energy application in China. CFETR has two phases. Phase I is to demonstrate full cycle of fus ... More
Presented by Songlin LIU on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 155
Square channel is widely used in the conceptual design of water cooled blanket of fusion reactor for cooling and providing appropriate inner temperature field for tritium breeding. Thermal hydraulic design of blanket directly determines the heat transfer efficiency and safety characteristics of fusion reactor. Therefore, thermal-hydraulic characteristics of square channel should be investigated. T ... More
Presented by Hui BAO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 156
The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). From the security point of view, the thermal-hydraulic analysis is very essential because the blanket should remove the high heat flux radiated from the plasma and the volumetric heat generated by neutron wall loading. For the normal state of plasma burning, the jumped peak heat ... More
Presented by Kecheng JIANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 157
The water-cooled ceramic breeder (WCCB) blanket is one of the candidates of  Chinese fusion engineering test reactor (CFETR). WCCB blanket will produce radioactive waste during its operation and decommissioning processes. The radioactive characteristics of WCCB blanket, including solid structure and functional material and the liquid water coolant, are of importance for the replacement and manage ... More
Presented by Xiaokang ZHANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 158
A conceptual structural design of Water-Cooled-Solid-Breeder (WCSB) blanket, one of the breeding blanket candidates for China Fusion Engineering Test Reactor (CFETR), is now being carried on by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP). To validate the reliability of the designed blanket module, detailed thermal-hydraulic analysis is necessary. The computational fluid dynamic ... More
Presented by Pinghui ZHAO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 159
Tokamak reactors like ITER or fusion DEMO reactors have serious concerns about material damages to plasma facing components (PFC) due to plasma instabilities. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. In addition, high thermal stresses due to rapid changes o ... More
Presented by Geon-Woo KIM on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 160
In the framework of the EUROfusion programme, Dual Coolant Lithium Lead (DCLL) breeding blanket is being investigated as a candidate for European DEMO, which is based on the use of Pb-17Li as breeder and coolant (“self-cooled breeding zone”) and high-pressure helium for cooling the structures made of reduced-activation ferritic steel (EUROFER). During the first part of the project, a conceptua ... More
Presented by Angel IBARRA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 161
General purpose finite element (FE) softwares can be readily used for the stationary analysis of breeding blankets of a nuclear fusion reactor. However, the analysis of transient effects generated during the pulsed operation mode requires transient simulations to be carried out. Nowadays, a commercial tool which can be directly used for these transient simulations with affordable computational tim ... More
Presented by Luis MAQUEDA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 162
The Dual Cooled Lithium Lead (DCLL) blanket is one of the four breeder blanket technologies under consideration within the framework of EUROfusion Consortium activities. The aim of this work is to develop a preliminary model that can track the tritium concentration along each part of the DCLL blanket and their ancillary systems at any time. Because of tritium’s nature, the phenomena of diffusion ... More
Presented by Fernando ROCA URGORRI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 163
The Dual Coolant Lead-Lithium (DCLL) is one of the breeding blanket concepts under investigation in EUROFusion. This concept is characterized by the use of self-cooled eutectic PbLi as neutron multiplier and tritium breeder and carrier, whereas supercritical helium is used to cool the first wall and other parts of the structure. The thermal-hydraulic (TH) design of the breeding blanket, as the mai ... More
Presented by Ivan FERNANDEZ-BERCERUELO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 164
The conceptual design of the European Dual Coolant Lead Lithium (DCLL) breeding blanket is currently being developed in the frame of EUROfusion Project. To this aim, it is of utmost interest to estimate critical flow parameters such as: (1) pressure drop and heat transfer coefficient at both helium and lithium sides, and (2) tritium permeation ratio. Pressure drop in purely hydrodynamic flows (suc ... More
Presented by Daniel SUAREZ on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 165
Liquid metal (LM) blanket concepts are designed by many countries due to its attractive features such as geometric adaptability, good thermal conductivity and heat carrying capacity, et al. However, they all have feasibility issues associated with magnetohydrodynamic (MHD) interactions under the environment of a strong control magnetic field and the flowing high electrical conductivity LM. The MHD ... More
Presented by Qingyun HE on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 166
DCLL blanket has high energy recovery efficiency. Nevertheless by several technical issues, such as MHD pressure drop, tritium permeation and energy conversion membrane corrosion, technical readiness level(TRL) of DCLL is relatively not high. To breakthrough this situation, the authors propose a new method to recover tritium and heat from liquid lithium-lead (PbLi) droplet by non-contact in vacuum ... More
Presented by Fumito OKINO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 167
The analysis of Dual-Coolant Lead–Lithium (DCLL) blankets requires application of Computational Fluid Dynamics (CFD) methods for electrically conductive liquids in geometrically complex regions and in the presence of a strong magnetic field. Several general-purpose CFD codes allow modeling of the flow in complex geometric regions, with simultaneous conjugated heat transfer analysis in liquid and ... More
Presented by Andrei KHODAK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 170
Indian Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) will be tested in one half of the port no#02 of ITER. In LLCB TBM, PbLi eutectic alloy is used as multiplier, breeder, and coolant for the CB zones, and Li2TiO3 ceramic breeder (CB) is used as a tritium breeding material. The LLCB TBM consists of two helium coolant circuits, one for the TBM outer box i.e. the TBM First Wal ... More
Presented by Brijesh Kumar YADAV on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 171
This paper gives an overview of the new facility for MHD and heat transfer (HT) tests of liquid metal breeder blanket mock-ups in high magnetic field. The facility named LIMITEF5 (LIquid Metal TЕst Facility, 5 T) is under construction now in JSC “NIIEFA” (D.V. Efremov Institute). The facility includes the Lead-Lithium (LL) loop passing through the warm aperture of the superconducting magnet. ... More
Presented by Denis OBUKHOV on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: H. Fuel Cycle and Breeding Blankets Board #: 172
Lithium molten salts (e.g., Flibe, Flinabe) have several merits as a self-cooled tritium breeding material: low reactivity, low density and low electric conductivity. On the other hand, molten salts may cause a problem of tritium migration to the structural material of the blanket due to the low hydrogen solubility. To overcome this problem, an active control of the effective hydrogen solubility o ... More
Presented by Takuya GOTO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 173
The primary reasons for the selection of beryllium as an armour material for the ITER first wall are its low Z and high gettering characteristics. For this application three beryllium grades: S-65C (USA), TGP-56FW (Russia) and CN-G01 (China) have been accepted. This selection was based on the results of the ITER Qualification Program, which included characterization and testing of material perform ... More
Presented by Igor KUPRIYANOV on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 174
Tungsten-based composites have gained considerable attention owing to their excellent performance levels at high temperatures due to exceptional high temperature properties such as a high melting point, good thermal conductivity and a low thermal expansion coefficient.  However, tungsten is also associated with a serious reduction in its strength at elevated temperatures, which is also one of the ... More
Presented by Petra JENUS on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 175
The main aim of the work has been to improve properties of the plasma-facing material for the divertor to resist high thermal loading during operation. Among the available materials we selected (carbide) particles reinforcement of tungsten, wherein the reinforcement should not chemically react with the matrix. In this respect, W2C particles offer the most attractive solution. The paper will presen ... More
Presented by Sasa NOVAK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 176
W has the highest melting point of all metals, good high temperature strength, high creep resistance and a high thermal conductivity. These properties make W a first choice for armor materials in fusion energy reactors. Unfortunately W can not be also used for structural applications, due especially to its high temperature brittle- to-ductile transition (DBT). However, when cold rolled at about 40 ... More
Presented by Andrei GALATANU on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 177
Tungsten is presently the main candidate material for the first wall armour of future fusion reactors. However, if a loss of coolant accident with simultaneous air ingress into the vacuum vessel occurs, the temperature of the in-vessel components would exceed 1000ºC, leading to the undesirable formation of volatile and radioactive tungsten oxides. A way to prevent this serious safety issue is the ... More
Presented by Carmen GARCIA-ROSALES on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 179
Irradiation damage research is one of the basic issues to solve the application of first-wall materials in fusion engineering. The diffusion and recovery of the defects can greatly affect the performance of the materials in fusion. The rotation, stability, migration of the self-interstitial atoms (SIAs) in defect structures of tungsten is investigated by the first-principle method. It is found th ... More
Presented by Min PAN on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 180
In order to investigate possible enhancement of mechanical properties of tungsten (W) based materials by solid solutions and to examine the influence of a single alloying element on a particular property such as ductility, a versatile production method of generating a wide range of different tungsten binary alloys is presented. Magnetron sputter co – deposition was used to produce thin films of ... More
Presented by Vladica NIKOLIC on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 181
For DEMO fusion reactor an expected heat flux of about 10 MW/m<sup>2</sup>2 should be extracted by the divertor which will have, most likely, an armour part made of W and a following heat sink part made of Cu or ODS Cu alloy. Unfortunately, for these materials the optimum operating temperature windows do not overlap. Thermal barrier materials are interface materials included in such components, ai ... More
Presented by Magdalena GALATANU on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 182
Copper-based materials are considered the most promising candidates for water-cooled components of the heat sink systems of future fusion reactors. Although pure copper is the material with the higher thermal conductivity, the detriment of its mechanical strength on increasing temperature restricts its use at high temperature. In the last years, ODS Cu-Y2O3 and Cu-Y alloys have been produced follo ... More
Presented by Gabriel CARRO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 183
Copper is the candidate material for cooling components for divertor and other plasma facing components. Although CuCrZr alloy is a first choice regarding strength, toughness, and conductivities, issues related to quality control during manufacturing process and also on the possible loss of strength during brazing among fabrication of the components still remains. CuCrZr also exhibit some weakness ... More
Presented by Dai HAMAGUCHI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 184
 Copper (Cu) alloy is a candidate materials for use as heat sink materials of fusion divertor because of its good thermal conductivity. In recent years a number of studies have been carried out on Cu-based materials such as Precipitation Strengthened Cu (PS-Cu).However, the material has some critical issues such as instability of microstructure at high temperature and loss of strength by irradi ... More
Presented by Hiroyuki NOTO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 185
The blanket is one of the most critical component of ITER. It is directly exposed to the plasma and acts as shielding of the vacuum vessel from the neutrons and other energetic particles produced in the fusion plasma. Each of the 215 Normal Heat Flux (NHF) panels consists of a shield block and a First Wall (FW) panel. The NHF FW panels consist of a complex bimetallic structure of 316L stainless st ... More
Presented by Inigo ITURRIZA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 186
The actuality of the topic comes from the ITER (International Thermonuclear Experimental Reactor) fusion tokamak that is a major international experiment with the aim of demonstrating the scientific and technical feasibility of fusion as an energy source. Among others the most challenging task is to find proper materials and technology for Plasma Facing Components. Welding by HIP (Hot Isostatic Pr ... More
Presented by Teteny BAROSS on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 187
The CuCrZr/316L(N) explosion bonding bimetallic plates were used to make hypervapotron (HVT) cooling channel for the fingers, which is the key components of the ITER First Wall (FW). The bimetallic plates will be subjected to the same thermal cycles as the FW component, including the HIP (hot iso-static pressing) joining for bonding HVT and beryllium tiles, thus the properties of both the CuCrZr / ... More
Presented by Pinghuai WANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 188
Development of new materials is one of the key for the construction of the new fusion power plant (DEMO). The selected materials have to fulfill several requirements such as standing the conditions that takes place in the core (high neutron flux and temperatures close to 1200 ºC) and low activation rate. Several techniques have been proposed to join the different parts of the first wall componen ... More
Presented by Javier DE PRADO on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 190
Radiation tolerant optical components of future fusion reactors have to withstand radiation of unprecedented intensity. It is widely recognized that spinel lattice of AB2O4 double oxides demonstrates enhanced resistance against neutron irradiation. Therefore, the development of spinel optical materials and understanding of their radiation damage processes is of great importance. One defect type of ... More
Presented by Eduard FELDBACH on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: I. Materials Technology Board #: 191
First mirror (FM) lifetime is one of critical issues for the optical diagnostic system in ITER since it greatly influences the performance of relative diagnostic. In ITER, repetitive cleaning is expected to give a positive solution to the frequent replacement of FM, thus prolonging its lifetime. Three cleaning cycles using radio frequency argon plasma were applied to the stainless steel mirror wit ... More
Presented by Jiao PENG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 192
There are a number of key design difficulties in producing an integrated demonstration fusion power plant (DEMO) design, and how these issues are resolved fundamentally affects the final overall design. Technological examples include the issue of power loading in the divertor and reducing recirculating power through efficient current drive. Additional drivers include economic considerations such a ... More
Presented by Richard KEMBLETON on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 193
Breeding blanket research and development is recognized as one of the most important areas for realizing an energy-producing fusion reactor. In China, the ceramic breeder/helium coolant/ferritic steel structure is considered as the main concepts of the blanket to conduct the breeding blanket research, and on the other hand, the liquid breeder blanket is also to be investigated as the alternative o ... More
Presented by Dagui WANG on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 194
The investigation of time-dependent power requirements for a future nuclear fusion reactor is part of the systems integration task for the European Fusion Programme. All fusion power plants, whether pulsed or steady-state, will require electrical power to operate the various plant systems. Over the entire pulse cycle reactor systems will require varying levels of power over different time periods. ... More
Presented by James MORRIS on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 195
The Water-cooled Lithium-Lead (WCLL) blanket is one option under consideration for the EUROfusion DEMO programme. This blanket design must interface with the Primary Heat Transfer System, Power Conversion System, and Energy Storage System in an integrated solution to mitigate the pulsed power profile of the tokamak and deliver feasible power plant performance. The system must maintain an acceptabl ... More
Presented by Christopher HARRINGTON on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 196
ITER is planned to be the research type tokamak which will achieve the energy breakeven point. The next step towards the realization of fusion energy will be DEMO – the first demonstration fusion power plant producing grid electricity at the level of a few hundred MW. DEMO designers are required to maximize the conversion efficiency of the primary and secondary plant circuits. The Primary Heat T ... More
Presented by Monika LEWANDOWSKA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 197
The cooling system is one of the key parts of the fusion power reactor technology. The DEMO fusion power reactor should have different heat sources (first wall, blanket, and divertor) with different temperature and power. In the current European concept of DEMO, helium and water are used as the cooling medium. However, use of Helium and water introduces some issues in terms of their properties and ... More
Presented by Vaclav DOSTAL on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 198
HCPB (helium cooled pebble bed) blanket concept is one of the EU DEMO blanket concepts running for the final design selection. It is necessary to study the pressure behaviour in the blanket and the connected systems during the loss of coolant (LOCA) in a blanket module, as well as the temperature evolution in the coolant flow and the associated structures. The LOCA can be caused by rupture/leak of ... More
Presented by Xue Zhou JIN on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 199
Radioactive toxins confinement is a main safety function for nuclear power plants, hence the importance of confinement design parameters optimization. In this context, performing parametric assessments of thermodynamic variables thought to be relevant for confinement design can help at better framing the option design space. In the context of DEMO EUROfusion WP, FFMEA studies are going on for the ... More
Presented by Danilo DONGIOVANNI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 200
The first wall, blanket and divertor targets provide a physical boundary for the plasma influence and have to be intensively cooled during the operation in case of the high power fusion reactor. In the case of the LOCA accident, the released fusion power can be stopped very quickly, but the final plasma disruption may load the non-cooled components, and a large amount of heat accumulated in the co ... More
Presented by Jan STEPANEK on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 201
'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER. Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. The F4E, Amec Foster Wheeler and INL comprehensive methodology f ... More
Presented by Dobromir PANAYOTOV on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 202
The helium cooled LiPb blanket concept has become a promising design for fusion reactors in the world. Considering the complex design of the blanket, it is likely that helium gas leakage into the liquid alloy may occur due to tube rupture, named in-box Loss of Coolant Accident (in-box LOCA). And corresponding shock waves likely occurred at the break position and transferred within the liquid metal ... More
Presented by Danna ZHOU on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 203
With China signing Test Blanket Module Arrangement (TBMA) with ITER Organization for Helium Cooled Ceramic Breeder (HCCB) Test Blanket System (TBS) in February 2014, Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS), becomes one of the leading teams undertaking its corresponding research and development, and is mainly responsible for structure material develo ... More
Presented by Jiangtao JIA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer Board #: 204
For almost ten years now, several safety studies of plasma-wall transients have been performed with AINA code for ITER, the European DEMO design (e.g. HCPB) and Japanese one (e.g. Water Cooled Pebbled Bed or WCPB) to establish an envelope for the worst effects of ex-vessel LOCA and overfuelling. For this purpose, for each blanket type a specific wall-model has been developed for different AINA cod ... More
Presented by Marco FABBRI on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: K. Laser and Accelerator Technologies Board #: 206
A concept and a laboratory model of the laser-driven accelerator of plasma beams for materials research is presented. The accelerator is based on the laser-induced cavity pressure acceleration (LICPA) scheme [1] and includes four parts: (1) the laser driver, (2) the plasma cavity where high-temperature plasma is created by the laser driver  and a high plasma pressure is generated, (3) the acceler ... More
Presented by Agnieszka ZARAS-SZYDŁOWSKA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: K. Laser and Accelerator Technologies Board #: 207
Interaction of high power laser fields with plasma is important for many applications including laser fusion, laser wakefield acceleration and x-ray lasers. At high laser intensities, nonlinear interactions between plasma and laser becomes significant. In the last ten years, there has been a great deal of interest on plasma systems where the quantum effects are important. Consideration of quantum ... More
Presented by Punit KUMAR on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: K. Laser and Accelerator Technologies Board #: 208
IFMIF (International Fusion Material Irradiation Facility) will generate 14 MeV neutron flux for qualification and characterization of suitable structural materials of plasma exposed equipment of fusion power plants. IFMIF is an indispensable facility in the fusion roadmaps since provide neutrons with the similar characteristics as those generated in the DT fusion reactions of next steps after ITE ... More
Presented by Koichi NISHIYAMA on 6 Sep 2016 at 14:20
Type: Poster Session: P2 Poster session
Track: K. Laser and Accelerator Technologies Board #: 209
For the IFMIF/EVEDA accelerator prototype RFQ linac, the operation frequency of 175MHz was selected to accelerate a large current of 125mA. The driving RF power of 1.28MW by 8 RF input couplers has to be injected into the RFQ cavity for CW operation mode. For each RF input coupler, nominal RF power of 160kW and maximum transmitted RF power of 200kW are required. For this purpose, an RF input coupl ... More
Presented by Sunao MAEBARA on 6 Sep 2016 at 14:20
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 1
Study of materials dedicated to fusion reactors is one of the most challenging tasks faced by fusion research. Unfortunately, the number of useful fast neutron sources with a proper neutron spectrum and high neutron fluence is limited. Currently, a better exploitation of the existing neutron sources, such as high flux fission research reactors or material test reactors, is necessary to develop fur ... More
Presented by Anna WOJCIK-GARGULA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 2
In order to study the neutronics of fusion reactor blankets, a program is underway at the IPR using 14-MeV neutron source. An accelerator based neutron generator is under development in which 30 mA deuterium beam will be accelerated up to 300 keV energy. It will then impinge on a rotating tritium target to producing nearly isotropic 14-MeV neutrons. The expected neutron yield is 3-5 x 10<sup>12</s ... More
Presented by Sudhirsinh VALA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 3
The LIPAc (Linear IFMIF Prototype Accelerator) is a prototype that ends in a Dump made of copper with conical shape and cooled by water moving at high speed on the outer surface. The shape of the dump is intended for a redistribution of a very high density power of the deuteron beam to be stopped (1.12 MW) leading during normal operation to reasonable temperatures and thermal stresses well below t ... More
Presented by Fernando ARRANZ on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 5
The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where fusion reactor candidate materials will be tested. The neutron flux is produced by means of a deuteron beam (250 mA, 40 MeV) that strikes a target of liquid lithium circulating in a loop. The support on which the liquid lithium flows is the most heavily exposed component to the neu ... More
Presented by Gioacchino MICCICHE on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 6
IFMIF-DONES - a powerful neutron irradiation facility for studies and certification of materials - is planned as part of the European roadmap to fusion electricity. Its main goal will be to study properties of materials under severe irradiation in a neutron field similar to the one in a fusion  reactor first wall. It is a key facility to prepare for the construction of the DEMO Power Plant envisa ... More
Presented by Wojciech KROLAS on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 7
The availability of a high flux neutron source for testing candidate materials under irradiation conditions which will be typically encountered in future fusion power reactors is a fundamental step towards the development of fusion energy. To this purpose, IFMIF (International Fusion Materials Irradiation Facility) represents the reference option to provide the fusion community with a source capab ... More
Presented by Gaetano BONGIOVI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 8
The International Fusion Materials Irradiation Facility (IFMIF) aims to provide an accelerator-based, D-Li neutron source to produce high energy neutrons at sufficient intensity and irradiation volume for DEMO materials qualification. Part of the Broader Approach (BA) agreement between Japan and EURATOM, the goal of the IFMIF/EVEDA project is to work on the engineering design of IFMIF and to valid ... More
Presented by Oriol NOMEN on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 9
Because of the depletion and limitation of natural energy sources, fusion energy is the promising and irreplaceable way for energy development in the future. As the only energy conversion unit in the fusion reactor, PbLi blanket is considered as one of the important blankets for DEMO and fusion reactors, Lead Lithium (PbLi) is designed as tritium breeder, neutron multiplier and coolant. Before the ... More
Presented by Zhiqiang ZHU on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 11
The liquid metal eutectic Pb-Li17 is considered as one of the possible coolants for the blanket of the fusion reactor DEMO. The main reason for usage of the eutectic Pb-Li17 is the Tritium breeding. The eutectic flow separates alloys of the structural steels and thus be the cause of them corrosion.The cold trap is a device for corrosion products removing from liquid metal. The cold trap was develo ... More
Presented by Tomas ROMSY on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 12
The Demonstration Fusion Power Reactor (DEMO) is supposed to be the step in between ITER and the first commercial fusion power plant. In the framework of one mission of the “Work plan for the roadmap to fusion energy 2014-2018” a work package Tritium, Fuelling and Vacuum (TFV) was launched. As part of this project, the examination of requirements for the matter injection system is ongoing cove ... More
Presented by Bernhard PLOECKL on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 13
Recent advances in the development of high temperature superconductors (HTS) [1], and encouraging results on a strong favourable dependence of electron transport on higher toroidal field (TF) in Spherical Tokamaks (ST) [2], open new prospects for a high field ST as a compact fusion reactor or a powerful neutron source [3]. The combination of the high beta (ratio of the plasma pressure to magnetic ... More
Presented by Mikhail GRYAZNEVICH on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 14
The manufacturing methods and issues found during the construction of the Stellarator of Costa Rica 1 (SCR-1) will be discussed. The SCR-1 is a small modular stellarator developed by the Instituto Tecnológico de Costa Rica (ITCR). Currently, it’s being tested for the first plasma discharge. SCR-1 is a 2-field period small modular stellarator (Ro=0.238 m, <a>=0.054 m, Ro/a>4.4, plasma volume =0. ... More
Presented by Carlos OTAROLA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: A. Experimental Fusion Devices and Supporting Facilities Board #: 15
Steady State Superconducting Tokamak (SST-1) at Institute for Plasma Research is a `working’ experimental superconducting device since late 2013. SST-1has been upgraded with Plasma Facing Components and is getting prepared towards long pulse operations in both circular and elongated configurations. Initial experiments have begun in SST-1 with circular plasma configurations. SST-1 offers a unique ... More
Presented by Subrata PRADHAN on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 18
This paper presents the results of a study that was performed on conceptual solutions for assembly and handling of EC components inside the EC upper and equatorial port cells. Particular topics that are discussed include the access to the waveguides and auxiliary feedthroughs of the launchers at the port plug closure plate, (dis-)assembly & alignment of the ex-vessel waveguide in the port interspa ... More
Presented by Dennis RONDEN on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 19
The Electron Cyclotron Upper Launcher (ECUL) is an eight beamline ITER antenna aimed to drive current locally inside the islands that may form on the q= 3/2 or 2 rational magnetic flux surfaces in order to stabilize neoclassical tearing modes (NTMs). The primary vacuum boundary at the port plug extends into the port cell region through the ex-vessel mm-wave waveguide components, defining the so-ca ... More
Presented by Avelino MAS SANCHEZ on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 20
The ITER Electron Cyclotron Heating Upper Launcher (ECHUL) will be used to drive current locally inside magnetic islands located at the q=2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs). Each antenna consists of eight beam lines that are designed for the transmission of up to 1.5 MW of mm-wave power at 170 GHz. The First Confinement System (FCS) is formed b ... More
Presented by Phillip SANTOS SILVA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 21
The ITER Electron Cyclotron Heating Upper Launcher (ECHUL) will be used to drive current locally inside magnetic islands located at the q=2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs). Each antenna consists of eight beam lines that are designed for the transmission of up to 1.5 MW of mm-wave power at 170 GHz. The First Confinement System (FCS) is formed b ... More
Presented by Robert BERTIZZOLO on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 22
The new mirror angle detector for ITER EC launchers, applying a rotary capacitor , a RF feeder, RF circuits and several hundreds MHz RF has been developed. The rotary electrode is attached to the rotation axis of the mirror and the stationary electrode is connected to a RF feeder. The reflected RF wave at the rotary capacitor comes back to the feeder and phase of the reflected RF wave changes depe ... More
Presented by Koji TAKAHASHI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 23
The Electron Cyclotron Heating and Current Drive system developed for ITER is made of 12 sets of High Voltage Power Supplies, 24 Gyrotrons, 24 Transmission Lines and 5 Launchers, 4 UL located in upper ports and 1 EL at the equatorial level. The ITER operation requires to switch operating launcher during the plasma operation with short interval, namely mid-pulse switch operation. To change the wave ... More
Presented by Yasuhisa ODA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 24
The power supply for the EC Heating system (ECPS) of ITER provides the electrical power to the 170GHz/1MW Gyrotrons. The required electrical power for the gyrotrons is not only very high but has to comply also with highest quality requirements. This paper gives an overview of the Ampegon ECPS system procured by F4E. It describes the technical requirements of the EC Power Supply system ECPS and exp ... More
Presented by Michael BADER on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 25
The EU 1 MW, 170 GHz gyrotron with hollow cylindrical cavity has been designed within EGYC (European GYrotron Consortium) in collaboration with the industrial partner Thales Electron Devices (TED) and under the coordination of Fusion for Energy (F4E). In the frame of the EU program the short-pulse (SP) version of this tube has been designed and manufactured by KIT in collaboration with TED. The ... More
Presented by Tomasz RZESNICKI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 26
ITER will be equipped with four EC (Electron Cyclotron) upper launchers of 8 MW microwave power each with the aim to counteract plasma instabilities during operation. The structural system of these launcher antennas will be installed into four upper ports of the ITER vacuum vessel. During operation the port plug structure will be heated by nuclear heating from neutrons and photons and thermal radi ... More
Presented by Peter SPAEH on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 27
The Tokamak à Configuration Variable (TCV) has been recently equipped with a 1 MW neutral beam heating (NBH) injector<sup>1</sup>1. Two new stainless steel ports with rectangular aperture of 170x220mm have been manufactured and installed for this purpose. The NBH injector is connected to one of them via a stainless steel port extension. The port and its extension together form the beam duct betwe ... More
Presented by Matthieu TOUSSAINT on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 28
The TCV tokamak infrastructure has been recently adapted to leave access for a neutral beam (NB) injector capable of 1MW of neutral power during 2sec into the TCV plasma. BINP has been in charge to design and to procure this equipment, taking care of the experimental constraints imposed both by the future physics objectives of TCV, as by the mechanical requirements complying with the tight space ... More
Presented by Damien FASEL on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 29
Three RHVPSs (Regulated High Voltage Power Supplies, 84kV/80A/2s) are installed and operated at the Swiss Plasma Center for almost twenty years. Each RHVPS supplies a cluster of three gyrotrons. Two clusters are composed of diode type gyrotrons operating at the second harmonic of the TCV electron-cyclotron frequency (X2, 84GHz), whereas the third is a cluster of triode type gyrotrons operating at ... More
Presented by Ugo SIRAVO on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 30
The TCV tokamak contributes to physics understanding in fusion reactor research based with a wide experimental tool set: flexible shaping and high power electron cyclotron heating. Plasma regimes with high plasma pressure, a wide range of temperature ratios and significant populations of fast ions are now attainable by a TCV heating system upgrade. In the first stage of the TCV upgrade program, a ... More
Presented by Alexander N. KARPUSHOV on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 31
The transmission line is one of the most important parts among the ion cyclotron range of frequencies (ICRF) heating devices. In the case of unwanted troubles on the line, immediate power-off is necessary for the protection of the line and for safety. In the Large Helical Device (LHD), though the causes were unclear, several troubles such as melting sometimes occurred on the line between the Final ... More
Presented by Kenji SAITO on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 32
The heating of ions by an obliquely propagating shear Alfvén wave at frequencies a fraction of the particle cyclotron frequency is demonstrated analytically. Under consideration of the small wave amplitude, the resonance conditions in the laboratory frame are systematically derived by multi-scale expansion method. It is found that 1) the cyclotron resonance condition may occur at any wave frequen ... More
Presented by Haifeng LIU on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 33
The efficiency of heating and current drive systems is the key for a successful operation of fusion demonstration power plants like DEMO. In an earlier review article, overall efficiencies of H & CD systems were estimated at 20 – 30 % [1]. In this paper we present a breakdown of the overall efficiency for ICRF (ion cyclotron range of frequencies): 1) the technical efficiencies; 2) the interface ... More
Presented by Helmut FAUGEL on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 34
Ion cyclotron wall conditioning (ICWC) is being developed for ITER as a baseline conditioning technique in which the ion cyclotron heating and current drive system will be employed to produce and sustain the current-less conditioning plasma. The TOMAS project (TOroidal MAgnetized System, operated at the FZ-Juelich, Germany) proposes to explore several key aspects of ICWC. This project stands on tw ... More
Presented by Fabrice LOUCHE on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: B. Plasma Heating and Current Drive Board #: 35
Experimental results have shown that twelve-strap HHFW operating at 30 MHz can provide significant plasma heating for NSTX. In this case, it is important to understand the interactions between return currents on the antenna enclosure sidewalls/septa and the launched k|| spectra. CST Microwave Studio is applied to this problem with the view toward optimizing the antenna coupling to the desired spe ... More
Presented by Chun KUNG on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 36
The W7-X steady state control and data acquisition system has been successfully commissioned and well established to investigate plasma break down and run the first more complex physics programs during the initial operation phase of W7-X. Already in the first weeks of plasma operation, experiment programs with up to 10 minutes containing a series of up to 20 plasma discharges have been run routine ... More
Presented by Anett SPRING on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 37
Wendelstein 7-X (W7-X) is a superconducting stellarator undergoing the first experimental campaign after its commissioning. It’s characteristic feature is the steady state operation of the magnetic field. After an upgrade to cope with permanent heat loads of several Megawatts, W7-X will be able to run steady state discharges, too. This requires a control system that differs from the commonly use ... More
Presented by Heike LAQUA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 38
The commissioning and final validation of the central safety system and the acceptance by the authority were very important steps immediately before the successful ignition of the first plasma in Wendelstein 7-X in December 2016. Safety is the mandatory prerequisite for the operation of experimental devices of course to protect the personnel and the investment from hazardous situations. To fulfill ... More
Presented by Reinhard VILBRANDT on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 39
Ongoing work in the fusion community focuses on developing advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamic (MHD) stability, and noninductively driven plasma current. The toroidal current density profile, or alternatively the q profile, together with the normalized beta, are often used to characterize these advanced scenarios. The development of these advance ... More
Presented by Hexiang WANG on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 40
Control of the plasma density and temperature to produce a certain amount of fusion power, known as burn control, is one of the key issues that need to be solved for the success of tokamak fusion reactors such as ITER. In order to reach a high fusion power to auxiliary power ratio, tokamaks must operate near temperature and density stability limits. Therefore, active control to maintain a desired ... More
Presented by Andres PAJARES on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 41
Research on fusion plasmas in tokamaks has led to the insight that the poloidal magnetic-flux distribution within the plasma has a crucial impact on its performance. Achieving certain types of poloidal magnetic-flux profiles, or alternatively certain types of q profiles, leads to resilience against undesirable instabilities and to higher bootstrap-current fractions, which in turns favor steady-sta ... More
Presented by Eugenio SCHUSTER on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 42
Active control of the toroidal current density profile is among those plasma control milestones that the National Spherical Tokamak eXperiment - Upgrade (NSTX-U) program must achieve to realize its next-step operational goals characterized by the high-performance, MHD-stable plasma operation with neutral beam heating, and longer pulse durations. Motivated by the coupled, nonlinear, multivariable, ... More
Presented by Zeki ILHAN on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 43
In this work we present a new real-time acquisition and elaboration system for the two-color scanning beam interferometer installed on FTU. The real-time system provides the density informations that can be used to approximate the plasma and runaway beam radial position. Furthermore, the central chord plasma line density will be used to substitute the actual feedback signal for the fueling control ... More
Presented by Luca BONCAGNI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: C. Plasma Engineering and Control Board #: 44
The plasma pulse phase of Frascati Tokamak Upgrade (FTU) is driven by the dedicated system FSC (Fast Sequence Control), which has been developed in order to send all the necessary commands to the different power plants feeding the toroidal and poloidal coils during the plasma discharge, meanwhile controlling the correct outcome. In case of incorrect execution of the sequence the system is able to ... More
Presented by Carlo NERI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 45
The products of fusion reactions at JET are measured using different diagnostic techniques. One of the methods is based on measurements of gamma-rays, originating from reactions between fast ions and plasma impurities. During the forthcoming deuterium-tritium (DT) campaign a particular attention will be paid to 4.44 MeV gamma-rays emitted in the <sup>9</sup>9Be(α,nγ)<sup>12</sup>12C reaction. Ga ... More
Presented by Andrzej BROSLAWSKI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 46
The JET tangential gamma-ray spectrometer (KM6T) is undergoing an extensive upgrade in order to make it compatible with the forthcoming deuterium-tritium (DT) experiments. The paper will present the design of the main components for the upgrade of the spectrometer beam-line: tandem collimators, gamma-ray shields, and neutron attenuators. The existing KM6T tandem collimators  will be upgraded by i ... More
Presented by Marian CURUIA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 47
The diagnostic of fast ions at JET is based on the measurements of gamma-rays which are produced as a result of nuclear reactions between ions and plasma impurities. The gamma-ray spectra provide information on energetic tail of ion energy distribution. The existent BGO detector, with a decay time of ~300 ns, is sufficient during DD campaigns. The strong neutron and gamma-ray fluxes during D-T ex ... More
Presented by Roch KWIATKOWSKI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 48
A new diagnostics technique, the Lost Alpha Monitor (LAM), for the investigation of escaping alpha particles in JET has been proposed [1]. The method is based on the detection of the gamma radiation induced by the escaping particles on a target external to the plasma. For a beryllium target this reaction is <sup>9</sup>9Be(a, nγ)<sup>12</sup>12C. The implementation on JET of the LAM technique wou ... More
Presented by Sorin SOARE on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 49
All optical spectroscopy and imaging diagnostics in next-step fusion devices will be based on metallic mirrors. The performance of mirrors is studied in present-day tokamaks and under laboratory conditions. This work deals with comprehensive tests of mirrors: (i) exposed in JET with the ITER-Like Wall (JET-ILW); (b) irradiation by He and heavy ions to simulate the impact of neutrons under reactor ... More
Presented by Marek RUBEL on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 50
High performance H-mode plasmas are characterized by short, repetitive edge perturbations known as edge-localized modes (ELMs). Large, unmitigated ELMs can result in significant transient heat loads released onto the plasma-facing components. Hence, characterization of ELMs and their control are crucial for avoiding a significant reduction in the divertor lifetime. This necessitates discrimination ... More
Presented by Jean-Marie NOTERDAEME on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 51
The ITER first mirrors are the components of optical diagnostic systems closest to the plasma. Deposition may build up on the surfaces of the mirror affecting their ability to fulfil their function. However, physics modelling of this layer growth is fraught with uncertainty. A new experiment is underway on JET, under contract to ITER, with primary objective to test if, under realistic plasma and w ... More
Presented by Zsolt VIZVARY on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 52
ITER fusion reactor is a very complex machine which has several different subsystems. It is still a research reactor and the testing results will be implemented in the next generation reactors. In the testing phase of the reactor there will be several sensors and instruments assembled inside the vessel for diagnostics purposes. One of the key diagnostics areas will be the divertor environment. Du ... More
Presented by Janne LYYTINEN on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 53
Electrical Services provide the electrical infrastructure to serve the diagnostics installed on the ITER Tokamak. The components of the Diagnostics are located all over on the inner and outer shell of the vacuum vessel, in the ports, on the Divertor Cassettes and in the Cryostat as well. Sensors require electrical transmission lines to transmit both of the diagnostic and control signals across the ... More
Presented by Miklos PALANKAI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 54
Numerous plasma-near mirrors of optical diagnostics of ITER require protection from erosion and deposition caused by impinging energetic particles. This is achieved by approximately 60 individual Diagnostic Shutters, rather simple mechanical devices which obstruct the mirror’s sight towards the plasma when the diagnostic is not in use. If a shutter fails to operate, so does the respective diagno ... More
Presented by Christian VORPAHL on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 55
First mirrors are plasma facing components which redirect light to the protected optical diagnostics. Initial investigations [A. Litnovsky et al. Nuclear Fusion 49 (2009) 075015, V. Kotov et al. Fusion Eng. Des. 89 (2011) 1583] showed that deposition of impurities (Be, Fe etc.) may cause drastic degradation of the mirror reflectivity and thus severely restrict the diagnostic performance. Very mode ... More
Presented by Vladislav KOTOV on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 56
During the final design review of Diagnostic Port Plugs, it has been highlighted that the current system of fixation, based on gaps, while it is not harmful for the port plug, it throws large uncertainties over the alignment of the optical systems placed inside the DSMs at the same time that the real mechanical behaviour of the assembly is clearly unknown. Due to the fact that the DSM is not rigid ... More
Presented by Laura GARCIA-RUESGAS on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 57
ITER will have a set of 45 diagnostics to ensure controlled operation. Many of them are integrated in the ITER ports. Housed in generic structures, this modular integration is designed to help diagnostics withstanding the plasma loads whilst complying with the French regulations. Now that the Domestic Agencies and ITER Organization are developing the preliminary or even final designs of the diagno ... More
Presented by Jean-Marc DREVON on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 58
ITER is the world’s largest fusion device currently under construction in the South of France with over 60 diagnostic systems to be installed inside the port plugs, the interspace or the port cell region of various diagnostic ports. The plasma facing Diagnostic First Wall (DFW) and its supporting Diagnostic Shielding Modules (DSM) are designed to protect front-end diagnostics from plasma neutron ... More
Presented by Yuhu ZHAI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 59
To achieve the real time controllability of plasma, real-time network is required in fusion experiments place. KSTAR Plasma control system(PCS) adopted the reflective memory (RFM) as a real time network. Since RFM based network has low latency and low jitter. However, KSTAR is also adopted Synchronous Data bus Network (SDN) as real time network to provide real time network to fueling system. Since ... More
Presented by Kwon GIIL on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 60
The iRIO-3DLab platform has been devised to enhance the learning process and reduce the development time for engineers in charge of designing intelligent DAQ systems based on PXIe technology and distributed control systems such as EPICS. iRIO-3DLab consists of an Opensim-based virtual world that aims to promote the understanding of how such a kind of DAQ system works, and how the EPICS IOC should ... More
Presented by Antonio CARPENO on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 62
A Dual output (27kV & 15kV), 3MW High Voltage Power Supply (ICHVPS) has been installed and integrated with a Diacrode based RF source to be used for ICRF system. The ICHVPS Controller is based on LabVIEW Real-time PXI controller, which supports all control and monitoring operations of the PSM based power supply. The controller supports all essential features like, fast dynamics, low ripple and pro ... More
Presented by Hiteshkumar DHOLA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 63
The Advanced Telecommunications Computing Architecture (ATCA) standard defines a high performance technical solution that meets the requirements for fast controllers on large-scale physics experiments like ITER. This platform provides high throughput, scalability and features for high availability such as redundancy and intelligent platform management which are essential for steady state experimen ... More
Presented by Bruno SANTOS on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 64
Control and Data Acquisition (CDAQ) systems applied to large physics experiments like ITER, are designed, among other features, for High-Availability (HA). A CDAQ system based on the PCI Industrial Computer Manufacturers Group (PICMG) 3.x AdvancedTCA Base Specification and Intelligent Platform Management Interface (IPMI) standards grants these features. One of the key functions of the HA is the ho ... More
Presented by Antonio RODRIGUES on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 65
The Advanced Telecommunications Computing Architecture (ATCA) specification implements important key features such as high reliability, high availability, redundancy and serviceability for control and data acquisition instrumentation fault condition, hardware malfunction, firmware updates and hardware reconfiguration. Red Hat Enterprise Linux and corresponding kernels already have built-in mechani ... More
Presented by Paulo CARVALHO on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 66
The Radial Neutron Camera (RNC) and the Radial Gamma-Ray Spectrometer (RGRS) are two ITER diagnostics, devoted, respectively, to the real-time measurement of the neutron emissivity profile (to be used for plasma control purposes) and to the measurement of the confined alpha profile and runaway electrons. The two systems are closely related as they share the same equatorial port plug and part of th ... More
Presented by Rita C. PEREIRA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 67
The Swiss Plasma Center (SPC) is involved in the development and the operation of gyrotrons for fusion application (TCV tokamak, W7-X, ITER) and for medical application as well (spectroscopy DNP/NMR). In this framework, embedded control systems based on National Instrument (NI) compact Reconfigurable Input Output (cRIO) and compact Data AcQuisition (cDAQ) offer versatile solutions for dedicated ap ... More
Presented by Jeremie DUBRAY on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 68
Cryogenic Instrumentation is a unique and vast field and requires an in-depth understanding of the process and instrumentation. 26 channels Data Acquisition System is required for the 6 nos. of Cryogenics Pumps LN2 cool down experiment. The data acquisition system measures 22 nos. of temperature signals, 2 nos. of level signals of the buffers and 2 nos. of Nitrogen Dewar Signals (Pressure and Leve ... More
Presented by Karishma QURESHI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 69
The Control and Data Acquisition System (CODAS) of SPIDER, the first experiment of the Neutral Beam Test Facility, is under installation and undergoing the commissioning and first operation phases. The system hardware is nearly compliant with the ITER CODAC catalog for slow and fast plant systems. The system software is based on a combination of software frameworks that altogether collaborate to p ... More
Presented by Adriano Francesco LUCHETTA on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 70
This paper is focused on the computation of EM loads induced by plasma current disruptions on the Diagnostics positioned inside the Equatorial Port Plugs, and more explicitly, on the creation of a detailed set of tools (Finite Element ‘FE’ models and routines) which allow the automatic characterization of the EM phenomena (DINA) as well as they provide versatility for the adding/removing of th ... More
Presented by Eduardo RODRIGUEZ on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 71
The Large Helical Device (LHD) plans to start the deuterium experiment in March of 2017, where a maximum neutron yield of 2.1x10<sup>16</sup>16 neutrons/3 sec is expected.  For the deuterium experiment, neutron flux monitors, a neutron profile monitor, a neutron activation system and other neutron detectors have been prepared.  The characteristics of those neutron diagnostics, such as the det ... More
Presented by Takeo NISHITANI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 72
Devices that are capable of measuring the total plasma radiation in fusion reactor experiments are indispensable for safe and reliable plasma operation. One of the most widespread type of these kind of devices are metal absorber–metal resistor bolometers where the radiation is absorbed by a metallic layer and the change of the layer’s temperature is measured by metal resistors. Based on the me ... More
Presented by Gabor VERES on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 73
The development of GEM detector based acquisition systems resulted in the increase of throughput and resolution in the new revision of the system. The FPGA-based electronics is used to acquire, diagnose and to preliminarily analyze the data of soft X-ray emitted by hot plasma in Tokamak. Moreover, the development of electronics allowed to implement algorithms, so far performed offline after the ex ... More
Presented by Rafał KRAWCZYK on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 74
On the basis of the angle variables technique (AVT) changes of polarimetry state of electromagnetic wave passing through the thermonuclear plasma in the poloidal plane have been analyzed. The first section analyzes the changes in polarization state depending on the angle at which the test beam was sent, for the same plasma parameters. Subsequently, for a given geometry, using numerical calculation ... More
Presented by B. BIEG on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: D. Diagnostics, Data Acquisition and Remote Participation Board #: 75
This work describes the preliminary assessment of the different waveguide technologies for the ex-vessel transmission lines of the Plasma Position Reflectometer (PPR) in ITER. Initially, both oversized rectangular and circular corrugated waveguides were considered for the study; the former due to reduced costs and ease of procurement and the latter due to better performance in terms of attenuation ... More
Presented by Jose MARTINEZ-FERNANDEZ on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 76
This paper mainly introduces the seismic analysis of the high-power dc reactor prototype, whose functions are to limit the ripple current and the increasing rate of fault current in the ITER poloidal field (PF) converter. The stacked reactors with the assembly dimension (L×W×H) of 2955 mm×1639 mm×3296 mm and weight about 5 tons are fixed to the steel base by five support components. In order t ... More
Presented by Chuan LI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 77
It is conceivable that electrical arcs can occur during the failure of a large superconducting magnet following an unmitigated quench accident. To assess such accidents, it is important to employ appropriate arc models to calculate the voltage current characteristics and heat dissipation as a function of conditions such as pressure and arc length. Although electrical arcs have been studied for man ... More
Presented by Andrew ASH on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 78
Japan Atomic Energy Agency (JAEA), as Japan Domestic Agency, has responsibility to procure 9 ITER Toroidal Field (TF) coils. JAEA completed proto double-pancake (DP) trials aiming at qualification and optimization of manufacturing procedure of TF coil in 2015. Series production of DPs then started and winding of 14 DPs, heat treatment of 11 DPs, fabrication of 9 radial plates (RP), transfer of 7 D ... More
Presented by Hideki KAJITANI on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 80
The ITER Central Solenoid Model Coil (CSMC) is a superconducting solenoid operated at the JAEA centre of Naka, Japan, since 2000 to test the performance of insert coils in its bore, where it produces a magnetic field of 13 T representative of the ITER CS operating conditions. In 2015, the ITER Central Solenoid Insert (CSI), whose Nb3Sn cable-in-conduit conductor (CICC) will be adopted for the 3L m ... More
Presented by Roberto BONIFETTO on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 81
General Atomics (GA) is currently manufacturing the ITER Central Solenoid Modules (CSM) under contract to US ITER at Oak Ridge National Laboratory, under the sponsorship of the Department of Energy Office of Science. The contract includes the design and qualification of manufacturing processes and tooling necessary to fabricate seven CSM (6 + 1 spare) that constitute the ITER Central Solenoid. The ... More
Presented by Kurt SCHAUBEL on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 82
The Residual Ion Dump Power Supply (RIDPS) is part of the Ground Related Power Supplies, to be manufactured by OCEM Energy Technology s.r.l. (OCEM) for the MITICA experiment and for the two ITER Heating Neutral Beam Injectors (HNBI). MITICA is the full-scale prototype of the HNBI, under construction in the PRIMA Neutral Beam Test Facility in Padua, Italy. The RIDPS is devoted to feed the plates of ... More
Presented by Alberto FERRO on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 83
The Neutral Beam Injector (NBI) is required to inject in ITER plasma Deuteron particles which, once generated in the Ion Source (IS) polarized at -1MV, are accelerated at ground potential and then neutralized. This voltage level is very demanding for the power supply system, requiring several non-standard components. This paper describes the design status of two main NBI components: High Voltage D ... More
Presented by Vanni TOIGO on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 85
The winding pack of the ITER Toroidal Field (TF) coils is composed of 134 turns of Nb3Sn Cable in Conduit Conductor (CICCs) wound in 7 double pancakes and cooled by supercritical helium (He) at cryogenic temperature. The cooling of the Stainless Steel (SS) case supporting the winding pack is guaranteed by He circulation in 74 parallel channels. A 2D approach to compute the temperature distributio ... More
Presented by Francesca CAU on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 86
The superconductive coils of ITER magnet system will be energized by ac/dc converters. Before each plasma pulse the magnet system will be pre-charged with energy (8GJ) to be used for generating the toroidal loop voltage required for the gas mixture breakdown and plasma formation. This will be realized by inserting energy dissipating resistors in series with the central solenoid (CS) modules and tw ... More
Presented by Rustam ENIKEEV on 7 Sep 2016 at 11:00
Type: Poster Session: P3 Poster session
Track: E. Magnets and Power Supplies Board #: 87
High current DC switches play a very important role in the ITER coil power supply system (CPSS) being key components of its two major parts: switching network units (SNU) for plasma initiation and fast discharge units (FDU) for superconducting coils energy extraction in case of quench.  For both functions, circuit-breakers rated up to 70 kA steady-state current and 10 kV voltage are required to t ... More
Presented by Maksim MANZUK on 7 Sep 2016 at 11:00