Aditya Singh
(Cooling Water System)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Jinho Bae
(Tokamak Technology)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Min-Su Ha
(Tokamak Technology)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Akifumi Iwamoto
(National Institute for Fusion Science)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Chengzhi Cao
(Southwestern Institute of Physics)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
49553.
P1.006 Detailed design of ITER CCWS, CHWS and HRS: Challenges experienced and their solutions
Ajith Kumar
(Cooling Water System)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Dinesh Gupta
(Cooling water system group)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Zhiwei Xia
(Southwestern Institute of Physics)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
The function of Gas Injection System[1][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...
Michael Nagel
(Wendelstein 7-X Operation)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Chandra Prakash Dhard
(Max-Planck-Institut fuer Plasmaphysik)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
On 13thth 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...
Tamara Andreeva
(Max-Planck-Institut fuer Plasmaphysik)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Sebastien Renard
(Institute for Magnetic Fusion Research)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Paul van Eeten
(Max-Planck-Institut fur Plasmaphysik, Device Operation, Greifswald, Germany)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
David Sestak
(Institute of Plasma Physics at the Czech Academy of Science)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Minyou Ye
(School of Nuclear Sciences and Technology)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Li Liu
(School of Nuclear Sciences and Technology)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Gergo Pokol
(Institute of Nuclear Techniques)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Sulkhan Nanobashvili
(Andronikashvili Institute of Physics)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Konstantinos Kouloulias
(Department of Mechanical Engineering)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Mayuko Koga
(Graduate School of Engineering)
9/5/16, 2:20 PM
A. Experimental Fusion Devices and Supporting Facilities
Poster
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...
Andrea Zamengo
(Consorzio RFX)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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)...
Marco Boldrin
(Consorzio RFX (CNR)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Cesare Taliercio
(Consorzio RFX, Padova, Italy)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Nicola Pilan
(Consorzio RFX)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Francesco Fellin
(Consorzio RFX)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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),...
Martin Schmid
(Institute of Pulsed Power and Microwave Technology (IHM))
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Nicolas Fil
(Engineering)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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,...
Mikio Saigusa
(College of Engineering)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Andrea Bertinetti
(Politecnico di Torino)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Christos Tsironis
(Electrical and Computer Engineering)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Braj Shukla
(ECRH)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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,...
Donghui Xia
(State Key Laboratory of Advanced Electromagnetic Engineering and Technology)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Alessandro Moro
(Istituto di Fisica del Plasma "Piero Caldirola" IFP-CNR)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
49581.
P1.034 Development of an ICRH antenna system at W7-X for plasma heating and wall conditioning
Bernd Schweer
(Laboratory for Plasma Physics LPP)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Guillermo Orozco
(ITED)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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 m33) in each injector provide proper vacuum conditions (below 10-2-2 Pa) during the 10 s beam pulse with a gas feed of up to 30 Pa×m33/s. A maximum...
Yang Qing Xi
(Institute of Plasma Physics)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Christian Hopf
(Max Planck Institute for Plasma Physics)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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 ~ 3×1066 L/s for D2 do not pump helium at all, leaving only the...
Claus-Peter Kasemann
(Max Planck Institute for Plasma Physics)
9/5/16, 2:20 PM
B. Plasma Heating and Current Drive
Poster
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...
Filip Janky
(Max Planck Institute for Plasma Physics)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Ian Jenkins
(CCFE)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Yoshiteru Sakamoto
(Department of Fusion Power Systems)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Shinsuke Tokunaga
(Japan Atomic Energy Agency)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Natale Rispoli
(Istituto di Fisica del Plasma “Piero Caldirola” - IFP-CNR)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Francesco Pizzo
(Department of Industrial and Information Engineering)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Morten Lennholm
(Jet Exploitation Unit)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Kazuo Nakamura
(Nuclear Fusion Dynamics)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Peter Buxton
(Tokamak Energy Ltd)
9/5/16, 2:20 PM
C. Plasma Engineering and Control
Poster
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...
Antonio Batista
(Instituto de Plasmas e Fusão Nuclear)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
49596.
P1.049 Design development, integration and assembly of the ITER steady-state magnetic sensors
Martin Kocan
(Fircroft Engineering Services Ltd)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Ivan Duran
(Tokamak)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Slavomir Entler
(Institute of Plasma Physics of the CAS)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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....
Jorge Belo
(Instituto de Plasmas e Fusão Nuclear)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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,...
Paulo Quental
(IPFN - Instituto de Plasmas e Fusão Nuclear)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Francesco Mazzocchi
(IAM- AWP)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Juan Ayllon
(CNA)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Gabor Nadasi
(Plasma Physics)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Florian Penzel
(Max Planck Institut für Plasmaphysik)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Nancy Ageorges
(Kampf Telescope Optics)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Nicola Fonnesu
(Department of Fusion and Nuclear Safety Technology)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Marco Riva
(Fusion)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Federico Binda
(Physics and Astronomy)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Axel Klix
(Neutron Physics and Reactor Technology)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
H.J. Wang
(School of Nuclear Science and Technology)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Bo Shi
(Institute of Plasma Physics)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Jean-Marcel Travere
(CEA/IRFM)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Philippe Moreau
(Institut de Recherches sur la Fusion par confinement Magnétique)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Chen Zhang
(Cadarache Center)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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/m22 for steady state conditions and up to 20 MW/m2 2 for transient phases. As a result, surface temperature measurement of the plasma...
Hiroshi Tojo
(Japan Atomic Energy Agency)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Manabu Takechi
(Japan Atomic Energy Agency)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
49619.
P1.074 Feasibility study on the JT-60SA tokamak beam emission spectroscopy diagnostic systems
Ors Asztalos
(Institute of Nuclear Techniques)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Giuseppe Marchiori
(Consorzio RFX)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Jae-young Jang
(Department of Nuclear Engineering)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
YooSung Kim
(Department of Nuclear Enigneering)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Young-Gi Kim
(Department of Nuclear Engineering)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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 ~5x101818m-3-3 which is lower than typical Ne in other tokamaks, each part of the system is carefully designed to maximize the number...
Kihyun Lee
(Department of Engineering)
9/5/16, 2:20 PM
D. Diagnostics, Data Acquisition and Remote Participation
Poster
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...
Yoshimitsu Hishinuma
(National Institute for Fusion Science)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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....
Simon McIntosh
(Culham Centre for Fusion Energy)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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-2-2 on the Plasma Facing Components.
Detachment provides an attractive solution to the power exhaust...
Aleksandra Dembkowska
(Faculty of Mechanical Engineering and Mechatronics)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Alberto Brighenti
(Energy Department)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Boris Stepanov
(EPFL-SPC)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Pierluigi Bruzzone
(Swiss Plasma Center)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Kamil Sedlak
(Swiss Plasma Center)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
49632.
P1.088 Towards a multi-physic platform for fusion magnet design – Application to DEMO TF coil
Quentin Le Coz
(IRFM)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Rainer Wesche
(SPC)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Anatoly Panin
(Forschungszentrum Juelich GmbH)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Renato Gatto
(Department of Astronautical)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Aashoo Sharma
(Institute for Plasma Research)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Petr Khvostenko
(NRC"Kurchatov Institute")
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Bill Huang
(Tokamak Energy)
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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....
Walter H. Fietz
(Karlsruhe Institute of Technology (KIT))
9/5/16, 2:20 PM
E. Magnets and Power Supplies
Poster
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...
Angel Munoz
(Departamento de Física)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Fernando Mota
(Laboratorio Nacional de Fusion)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Alexander von Muller
(Max-Planck-Institut für Plasmaphysik)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Wolfgang Krauss
(Institute for Applied Materials)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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....
Steven Zinkle
(University of Tennessee)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Selanna Roccella
(FSN)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Mihails Halitovs
(University of Latvia)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Timur Kulsartov
(Institute of Atomic Energy of National Nuclear Center of the Republic of Kazakhstan)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Koki Yakusiji
(Osaka university)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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,...
Irene Zammuto
(Max Planck Institut für Plasmaphysik)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Francesco Maviglia
(Power Plant Physics & Technology Department)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Yuri Igitkhanov
(ITEP)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Michal Poradzinski
(Department of Nuclear Fusion and Plasma Spectroscopy)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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/m2 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...
Kazuo Hoshino
(Japan Atomic Energy Agency)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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/m22 for the fusion power of < 1.5 GW and the large...
Jeong-Ha You
(Max Planck Institute for Plasma Physics)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Fabio Crescenzi
(Fusion and Technology for Nuclear Safety and Security)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Franklin Gallay
(CEA)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Eugenio Vallone
(Dipartimento di Energia)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Silvia Garitta
(Dipartimento di Energia)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Sumei Liu
(School of Engineering)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Lijun Cai
(Southwestern Institute of Physics)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Xuebing Peng
(Insititute of Plasma Physics)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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...
Xiaoju Liu
(Institute of plasma physics chinese academy of sciences)
9/5/16, 2:20 PM
F. Plasma Facing Components
Poster
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....
Jorge Gonzalez
(RÜECKER LYPSA)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Dong Kwon Kang
(ITER Korea)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Davide Flammini
(Department of Fusion and Nuclear Safety Technology)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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),...
Anton Travleev
(INR-NK)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Kwen-Hee Hong
(Tokamak Engineering Department)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Liam Worth
(ITER Organization)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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 m33), Torus (~1330 m33), and the Neutral Beams (~180 m33 each).
The vacuum system comprises of custom and commercially available components and adapted commercial vacuum technology. For a...
Chang Hyun Noh
(NFRI)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Yury Krasikov
(Forschungszentrum Jülich GmbH)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Thibaud Giacomin
(Port Plugs & Diagnostics Integration Division)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Jiang Beiyan
(Hefei Juneng Electro Physics High-tech Development Co.)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Stephane Gazzotti
(CEA IRFM)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Louis Doceul
(CEA Cadarache)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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), is...
Antonino Cardella
(Broader Fusion Development)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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 m33 volume, and weighs about 73 t. As the vessels will...
D. Mazed
(Department of Civil and Industrial Engineering (DICI))
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Weijun Zhang
(Robotics Institute)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Liang Du
(Robotics Institute)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Jing Wu
(Institute of Plasma Physics Chinese Academy of Sciences)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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....
Shanshuang Shi
(Lab of Intelligent Machines)
9/5/16, 2:20 PM
G. Vessel/In-Vessel Engineering and Remote Handling
Poster
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...
Dario Carloni
(KIT)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Milan Zmitko
(ITER)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Ladislav Vala
(Centrum výzkumu Řež)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Jose Galabert
(Fusion for Energy)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Satoshi Konishi
(Institute of Advanced Energy)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Sergey Ananyev
(Complex physical and chemical technologies)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Paul Humrickhouse
(Fusion Safety Program)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Francisco A. Hernandez Gonzalez
(Institute of Neutron Physics and Reactor Technology)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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 suffered...
Pavel Pereslavtsev
(Karlsruhe Institute for Technology)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Alejandro Morono
(National Fusion Laboratory)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Maria Gonzalez
(LNF-DTF)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Shin-ichi Satake
(Applied Electronics)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Simone Pupeschi
(Institute for Applied Materials (IAM))
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Shuang Wang
(School of Nuclear Science and Technology)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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....
Hongli Chen
(School of Nuclear Science and Technology)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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 areas...
Yuanjie Li
(USTC)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Jae-Hwan Kim
(Department of Blanket Systems Research)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Kun Xu
(School of Nuclear Sciences and Technology)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Shuai Wang
(School of Nuclear Science and Technology)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Seong Dae Park
(Korea Atomic Energy Research Institute (KAERI))
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Mu-Young Ahn
(National Fusion Research Institute)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Eo Hwak Lee
(KAERI)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Pietro Arena
(Dipartimento di Energia)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Chiara Mistrangelo
(Institute for Nuclear and Energy Technologies)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Otakar Frybort
(Technical calculations department)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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,...
Michal Kordac
(TEO)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Jean-Charles Jaboulay
(Department of Systems and Structures Modelling)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
49708.
P1.172 R&D activities and latest progress of dual functional lead lithium test blanket module
Qunying Huang
(Key Laboratory of Neutronics and Radiation Safety)
9/5/16, 2:20 PM
H. Fuel Cycle and Breeding Blankets
Poster
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...
Hans-Christian Schneider
(Institute for Applied Materials)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Nerea Ordas
(Materials and Manufacturing)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Hiroyasu Tanigawa
(Department of Fusion Reactor Materials Research)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Takeshi Miyazawa
(Japan Atomic Energy Agency)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Takashi Nozawa
(Japan Atomic Enegy Agency)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Kazumi Ozawa
(Fusion Research and Development Directorate)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Haiying Fu
(Fusion System)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Ryuta Kasada
(Institute of Advanced Energy)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Toshiya Nakata
(Division of Industrial Innovation Sciences)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Noriyuki Y. Iwata
(National Institute of Technology)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Takuya Nagasaka
(National Institute for Fusion Science)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Yatinkumar Sarvaiya
(Quality Assurance)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Władysław Pohorecki
(Faculty of Energy and Fuels)
9/5/16, 2:20 PM
I. Materials Technology
Poster
Measurement and calculations of long-lived radionuclide activity forming in the 14 MeV neutron field, in 66Li-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...
Abha Maheshwari
(In Wall Shielding)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Stefano Sgobba
(CERN)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Young-Bum Chun
(Nuclear Materials Development Division)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Seungyon Cho
(National Fusion Research Institute)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Joonoh Moon
(Ferrous Alloy Department)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Jun Young Park
(Korea Institute of Materials Science)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Youngmin Lee
(NFRI)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Jingping Xin
(Key Laboratory of Neutronics and Radiation Safety)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Shaojun Liu
(Key Laboratory of Neutronics and Radiation Safety)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Lee Packer
(Nuclear Technology Department)
9/5/16, 2:20 PM
I. Materials Technology
Poster
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...
Sergio Ciattaglia
(Power Plant Physics & Technology Department)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Yican Wu
(Key Laboratory of Neutronics and Radiation Safety)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Muyi Ni
(Institute of Nuclear Energy Safety Technology)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Raquel Garcia
(Power Engineering Department)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Tim Eade
(Culham Centre for Fusion Energy)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Jae Hyun Kim
(Nuclear Engineering)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Andrius Tidikas
(Laboratory of Nuclear Installation Safety)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Guido Mazzini
(Nuclear Safety Research Section)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Lucie Karaskova Nenadalova
(Nuclear Fuel Cycle)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Toshiharu Takeishi
(Applied Quantum Physics and Nuclear Engineering)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
Shutaro Takeda
(Institute of Advanced Energy)
9/5/16, 2:20 PM
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
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...
