Bernard Bigot
(ITER Organization)
05/09/2016, 09:10
Oral
Established by the signature of the ITER Agreement in November 2006, the ITER project is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant.
Supported by...
Thomas Klinger
(Enterprise Wendelstein 7-X)
05/09/2016, 09:50
Oral
The optimized stellarator Wendelstein 7-X (W7-X) has started with the goal to demonstrate steady-state plasma operation at fusion relevant plasma parameters. This is to establish the optimized stellarator as a viable fusion power plant concept. The design of W7-X is based on the optimization of the geometric properties of the magnetic field with the aim to minimize neoclassical transport...
V. Tomarchio
(JT-60SA EU-Home Team)
05/09/2016, 11:15
Oral
JT-60SA is a superconducting tokamak developed under the Satellite Tokamak Programme of the Broader Approach Agreement between EU and Japan, and the Japanese national programme. It is designed to operate in the break-even conditions for long pulse duration (typically 100 s), with a maximum plasma current of 5.5 MA. Its scientific aim is to contribute at early realization of fusion energy, in...
Aditya Singh
(Cooling Water System)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
Fast ignition is one of the proposed ways to achieve high fusion energy gain in inertial fusion research. This scheme has an advantage that requirements of laser power and implosion process for ignition are not strict compared to that in central ignition. For a successful ignition, it is necessary to transport the energy of hot electrons to the imploded core effectively. Recently, it is found...
Andrea Zamengo
(Consorzio RFX)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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))
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Donghui Xia
(State Key Laboratory of Advanced Electromagnetic Engineering and Technology)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Yoshiteru Sakamoto
(Department of Fusion Power Systems)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Slavomir Entler
(Institute of Plasma Physics of the CAS)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Gabor Nadasi
(Plasma Physics)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Federico Binda
(Physics and Astronomy)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Anatoly Panin
(Forschungszentrum Juelich GmbH)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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")
05/09/2016, 14:20
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)
05/09/2016, 14:20
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))
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Mihails Halitovs
(University of Latvia)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Michal Poradzinski
(Department of Nuclear Fusion and Plasma Spectroscopy)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Eugenio Vallone
(Dipartimento di Energia)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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),...
Kwen-Hee Hong
(Tokamak Engineering Department)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Yury Krasikov
(Forschungszentrum Jülich GmbH)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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.)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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))
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Ladislav Vala
(Centrum výzkumu Řež)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Shin-ichi Satake
(Applied Electronics)
05/09/2016, 14:20
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))
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Jae-Hwan Kim
(Department of Blanket Systems Research)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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))
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Pietro Arena
(Dipartimento di Energia)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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,...
Jean-Charles Jaboulay
(Department of Systems and Structures Modelling)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Young-Bum Chun
(Nuclear Materials Development Division)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Jingping Xin
(Key Laboratory of Neutronics and Radiation Safety)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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)
05/09/2016, 14:20
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...
Elisabetta Carella
(National Fusion Laboratory)
05/09/2016, 16:40
Tritium behavior in a breeding blanket is a key design issue because of its impact on safety and fuel-cycle best performance. Nowadays there are only few references and any fully validated tool with predictive capabilities. Considering the difficulty in handling tritium and its fundamental role inside a fusion reactor, it is intended to prepare a simulation tool for tritium transport.In this...
Nicolai Martovetsky
(US ITER)
05/09/2016, 16:40
The ITER Central Solenoid (CS) is one of the critical elements of the machine. The CS conductor went through an intense optimization and qualification program, which included characterization of the strands, a conductor straight short sample testing in the SULTAN facility at the Swiss Plasma Center (SPC), Villigen, Switzerland, and a single-layer CS Insert coil recently tested in the Central...
Tsuyoshi Hoshino
(Breeding Functional Materials Development Group)
05/09/2016, 17:00
Any demonstration power reactor (DEMO), which applies solid breeder blankets, requires “advanced tritium breeders” with high tritium breeding ratios and increased stability at high temperatures. However, the fabrication techniques of advanced tritium breeder pebbles have yet to be established. Therefore, the R&D on the fabrication technologies of the advanced tritium breeders and the...
Yeong-Kook Oh
(KSTAR Research Center)
05/09/2016, 17:00
Extending high performance plasma discharge into long pulse steady-state operation is one of the urgent issues to be solved in preparing the ITER and fusion reactor. The KSTAR device is one of the best engineered superconducting tokamak devices which is good for exploring the science and technologies for the high performance steady-state operation due to lots of its unique features such as...
Jie Yu
(Key Laboratory of Neutronics and Radiation Safety)
05/09/2016, 17:20
China has long been active in pushing forward the fusion energy development to the demonstration of electricity generation. Two experts’ meetings were organized in 2014 by Ministry of Science and Technology (MOST) to seriously discuss the China’s fusion roadmap in particular the design and construction of magnetic confinement fusion reactor beyond ITER.
As one of the most challenging...
Simone Peruzzo
(Consorzio RFX)
05/09/2016, 17:20
After 10 years of operation since its major modification, an upgrade of the RFX-mod experiment is presently under design. The main objectives are the improvement of the control of magnetic confinement, plasma density and plasma wall interaction in both RFP and Tokamak configuration.
The main design driver requirement for the improvement of the magnetic confinement control is the enhancement of...
Alessandro Del Nevo
(ENEA CR Brasimone)
05/09/2016, 17:40
Water-Cooled Lithium-Lead Breeding Blanket (WCLL) is considered a candidate option for European DEMO reactor. Starting from previous experiences in the frame of Power Plant Conceptual Studies within EUROfusion Consortium, , ENEA and its linked third parties have proposed and are developing a multi-module blanket segment concept based on DEMO 2015 specifications. The layout of the module is...
Jose Botija
(Fusion National Laboratory)
05/09/2016, 17:40
The JT-60SA project implemented by Japan and Europe is progressing on schedule towards the first plasma in 2019. Spain (Ciemat) is in charge of the design and manufacturing of the cryostat.
The JT-60SA cryostat is a stainless steel vacuum vessel (14m diameter, 16m height) which encloses the tokamak providing the vacuum environment (10-3-3 Pa). It must withstand the external...
Nikolay Bykovsky
(Swiss Plasma Center)
05/09/2016, 17:40
Various tests performed with full-size 60 kA HTS cable prototypes for fusion magnets in EDIPO test facility demonstrated that design of HTS strand proposed at Swiss Plasma Center – stack of HTS tapes twisted and soldered between two copper profiles – is applicable for high-current fusion cables, but additional mechanical reinforcement is still needed. Based on experimentally obtained...
S. Brezinsek
(EUROfusion Consortium)
06/09/2016, 08:30
Oral
Since installation of the JET ITER-Like Wall more than 30h of plasma operation with the inertial cooled full W divertor took place. Successfully, the divertor plasma-facing components PFCs handled harsh tokamak conditions with (i) high surface temperature excursions passing the ductile-to-brittle temperature and re-crystallisation temperature multiple times, (ii) ITER-relevant steady-state and...
Jiangang Li
(for CFETR team)
06/09/2016, 09:10
Oral
The Chinese Fusion Engineering Testing Reactor (CFETR) is the next device for the Chinese magnetic confinement fusion (MCF) program which aims to bridge the gaps between the fusion experiment ITER and the demonstration reactor DEMO. CFETR will be operated in two phases: Steady-state operation and tritium self-sustainment will be the two key issues for the first phase with a modest fusion power...
Radomir Panek
(Insitute of Plasma Physics CAS)
06/09/2016, 09:50
Oral
The COMPASS tokamak with ITER-like plasma shape has been put into operation in 2009 in Institute of Plasma Physics ASCR in Prague. It has been equipped by a comprehensive set of diagnostics for edge and Scrape-Off-Layer (SOL) plasma as well as by a new a system of two Neutral Beam Injectors (NBIs), which enabled to obtain significant results in the field of edge, SOL and divertor physics.
In...
David Armstrong
(Department of Materials, Oxford University, Oxford, United Kingdom)
06/09/2016, 11:00
Tungsten is the leading candidate material for plasma facing applications in future tokamak systems, due to its high melting point, good sputtering resistance and low activity after irradiation. Despite this there has been a significant lack of study of the effect of transmutation products on the post irradiation mechanical behaviour of tungsten-based alloy systems. This will be key to...
Jean-Michel Bernard
(CEA/DRF/IRFM)
06/09/2016, 11:00
One of key missions of WEST (Tungsten (W) Environment in Steady-state Tokamak) is to pave the way towards the ITER actively cooled tungsten divertor procurement and operation. WEST PFC will operate in ITER conditions, i.e. with a heat flux on the divertor target of 10MW/m22 during 1000s and 20MW/m22 during a few tens of seconds. To achieve such heat flux levels, both...
Chang-Hoon Lee
(Korea Institute of Materials Science (KIMS))
06/09/2016, 11:20
Microstructural evolution and mechanical properties of Ti-bearing RAFM steels were investigated after aging at 550 °C for 0 ~ 1000 hr. All samples with Ti were prepared using vacuum induction melting furnace and hot rolling process, followed by heat treatment in normalizing and tempering. Microstructures including precipitates, fractured surfaces and cross-sectional microsturctures were...
Tobias Wegener
(Institut für Energie- und Klimaforschung – Plasmaphysik)
06/09/2016, 11:20
Tungsten is considered the main candidate material for the first-wall in DEMO for its high melting point, low erosion yield and low fuel retention. Nevertheless, it can cause a substantial safety issue in a loss-of-coolant accident (LOCA) in combination with air ingress into the plasma vessel, due to formation and evaporation of volatile neutron activated tungsten oxide. Self-passivating...
Atsushi Kojima
(Fusion Research and Development Directorate)
06/09/2016, 11:20
Acceleration of high-power-density negative ion beams of ~180 MW/m22 have been achieved up to 60 s for the first time. Because the achieved power density was comparable to ITER accelerator, and accelerated energy density of 10800 MJ/m22 is much higher than that for JT-60SA of 6500 MJ/m22, this achievement is one of promising results to overcome common issues...
Jiri Matejicek
(Department of Materials Engineering)
06/09/2016, 11:40
Tungsten is the main candidate material for the plasma facing components of future fusion devices. During operation, these components will be subject to severe conditions, involving both steady state and transient heat loads as well as high particle fluxes. These may lead to surface and structure modifications which influence their performance and lifetime. Therefore, it is necessary to study...
Mehdi Firdaouss
(CEA/IRFM)
06/09/2016, 11:40
The main objective of the WEST (W Environment in Steady-state Tokamak) project is to fabricate and test an ITER-like actively cooled tungsten divertor to mitigate the risks for ITER. Concerning the others Plasma Facing Components (PFC), they will also be modified and coated with W to transform Tore Supra into a fully metallic environment.
Solutions had been developed with three different...
Joseph Tooker
(General Atomics)
06/09/2016, 11:40
A new mechanism for driving current off-axis in high beta tokamaks using fast electromagnetic waves, called Helicons, will be experimentally tested for the first time in the DIII-D tokamak. This method is calculated to be more efficient than current drive using electron cyclotron waves or neutral beam injection, and it may be well suited to reactor-like configurations [1]. A low power (100 W)...
Jan Willem Coenen
(Institut für Energie- und Klimaforschung – Plasmaphysik)
06/09/2016, 12:00
Material issues pose significant challenges for future fusion reactors like DEMO. When using materials in a fusion environment a highly integrated approach is required. Cracking, oxidation and fuel management are driving issues when deciding for new materials. Neutron induced effects e.g. transmutation adding to embrittlement are crucial to material performance. Here advanced materials e.g....
Qingxi Yang
(Institute of Plasma Physics)
06/09/2016, 12:00
Lithium coating techonolgy and flowing liquid lithium limiter (Flili) have been applied on HT-7 tokamak and many significant results been obtained. A Flili for exploring lithium as potential plasma facing material was designed and manufactured for EAST tokamak, it is applied on the concept of the thin flowing flim which had been sucessfully tested in HT-7 tokamak. The Flili of EAST mainly...
Defeng Kong
(Institute of Plasma Physics)
06/09/2016, 12:00
As the next step for the fusion energy in China beyond ITER, the China Fusion Engineering Text Reactor (CFETR) aims to operate with duty time as 0.3~0.5, means that CFETR should operate at steady-state scenario. This provides a great challenge for the physical design of the heating the current driving system. In general, four different kinds of method as NBI, ECH, LHW and ICRH have been...
Tomas Markovic
(Institute of Plasma Physics of the CAS)
06/09/2016, 14:20
A number of tokamaks, including the largest operating one, Joint European Torus (JET), has ferromagnetic core installed in their plasma current drive system. Moreover, some auxiliary systems, such as magnetic shielding of neutral beam injection (NBI) system, or iron inserts for toroidal field ripple mitigation, consist of non-negligible amount of ferromagnetic material as well. Besides the...
Alastair Shepherd
(Culham Centre for Fusion Energy)
06/09/2016, 14:20
Neutral beam injection systems have proved themselves as the most effective form of auxiliary heating in tokamak plasmas. In positive ion based systems once the beam is neutralised there are many residual ion components which must be intercepted by suitable ion dumps. A particular challenge for ion dump design occurs when the dump must be placed close to a focus point as is the case for the...
Eva Belonohy
(JET Exploitation Unit)
06/09/2016, 14:20
The final phase of the JET Programme in Support of ITER plans to operate with 100% Tritium (TT) followed by Deuterium-Tritium (DT) operation, to help minimise risks and delays in the execution of the ITER Research Plan and the achievement of Q~10. Additional technical requirements (compared to Deuterium operation) are needed to allow operation with Tritium gas, a high DT neutron flux and...
Rosaria Villari
(EUROfusion Consortium)
06/09/2016, 14:20
Neutronics benchmark experiments are conducted at JET for validating the neutronics codes and tools used in ITER nuclear analyses to predict quantities such as the neutron flux along streaming paths and dose rates at the shutdown due to activated components. In particular, in the frame of subproject NEXP of JET-3 program, several activities are performed within EUROfusion Consortium devoted to...
Roberto Ambrosino
(Engineering department)
06/09/2016, 14:20
The European roadmap to the realisation of fusion energy has identified a number of technical challenges and defined eight different missions to face them. Mission 2 ‘Heat-exhaust systems’ addresses the challenge of reducing the heat load on the divertor targets.
Divertor Tokamak Test (DTT) facility [1]-[2] has been launched to investigate alternative power exhaust solutions for DEMO. This...
Alessandro Anemona
(ICAS)
06/09/2016, 14:20
In the European Fusion Roadmap, one of the main challenges to be faced is the mitigation of the risk due to the impossibility of directly extrapolate to DEMO the divertor solution adopted in ITER, due to the expected very large loads. Thus a satellite experimental facility oriented toward the exploration of robust divertor solutions for power and particles exhaust and to the study of...
Gustavo Granucci
(IFP-CNR)
06/09/2016, 14:20
The proposed Divertor Test Tokamak, DTT, aims at studying power exhaust and divertor load in an integrated plasma scenario. Additional heating systems have the task to provide heating to reach a reactor relevant power flow in the SOL and guarantee the necessary PSEP/R together adequate plasma performances. About 40 MW of heating power are foreseen to have PSEP/R ≥ 15 MW/m. A mix of the three...
Giuseppe Di Gironimo
(Department of Industrial Engineering)
06/09/2016, 14:20
This paper describes the activity addressed to the conceptual design of the first wall and the main containment structures of DTT device, which will be broadly presented in the invited talk "Design and definition of a Divertor TOKAMAK Test facility".
The work moved from the geometrical constraints imposed by the desired plasma shape and the configuration needed for the magnetic coils. Many...
Giuseppe Mazzitelli
(Consorzio CREATE & Seconda Università di Napoli)
06/09/2016, 14:20
The DTT (Divertor Test Tokamak) is a new facility conceived in the frame of EUROfusion roadmap with the aim to assess and possibly integrate all the relevant physics and technology divertor issues.
The general project is presented in another paper of this conference [1] and with more details in [2].
The general project includes the analysis of the site requirements from several points of view;...
Alessandro Lampasi
(Department of Fusion and Technology for Nuclear Safety and Security)
06/09/2016, 14:20
The power supplies (PSs) of the DTT proposal, as presented in the talk "Design and definition of a Divertor Tokamak Test facility" invited at this conference, have to feed:
6 central solenoid (CS) and 6 poloidal field (PF) superconducting coils, with currents up to 25 kA.
18 toroidal field (TF) superconducting coils, with a current up to 50 kA.
Some fast plasma control coils, including at...
Gianluca Barone
(Dipartimento di Ingegneria Civile e Industriale)
06/09/2016, 14:20
The 1stst Specific Grant of the Framework Partnership Agreement 372 deals with experimental activities in support of the Conceptual Design of HCLL and HCPB Test Blanket Systems. Service-2 is focused on thermal-hydraulic tests with high pressure Helium for validation and benchmarking of suitable dedicated numerical tools. In this frame, an extensive experimental campaign has been...
Alessandro Venturini
(Department of Civil and Industrial Engineering)
06/09/2016, 14:20
The experimental facility IELLLO (Integrated European Lead Lithium LOop) was designed and installed at the ENEA Brasimone Research Centre to support the design of the HCLL TBM.
This work presents the results of the experimental campaign carried out within the framework of F4E-FPA-372 and which had three main objectives. First, to produce new experimental data for flowing LLE (Lead-Lithium...
Liqin Hu
(Institute of Nuclear Energy Safety Technology)
06/09/2016, 14:20
Due to the complexity of fusion reactors on geometry and neutron physics, the Monte Carlo (MC) methods have been broadly adopted in fusion nuclear design and analysis. But for calculations that require obtaining a detailed global flux map, such as the shutdown dose rate analysis, analog MC simulations usually cost a prohibitive long run time. To make such analysis computational practicable, it...
Jing Song
(Institute of Nuclear Energy Safety Technology)
06/09/2016, 14:20
Great challenges exist in real fusion engineering projects for the current Monte Carlo (MC) methods including the calculation modeling of complex geometries, simulation of deep penetration problem, slow convergence of complex calculation, lack of experimental validation for new physical features, etc.
Several novel and advanced capabilities of the latest version of MC program SuperMC for...
Mercedes Medrano
(National Laboratory for Magnetic Fusion)
06/09/2016, 14:20
The superconducting tokamak JT-60SA, aimed to support and complement the ITER experimental programme, is currently being assembled at the JAEA laboratories in Naka (Japan). Within the European contribution, Spain is responsible for providing JT-60SA cryostat.
The cryostat is a stainless steel vacuum vessel 14m diameter, 16m height which encloses the tokamak providing the vacuum environment...
Peter Lang
(Tokamak Scenario Development Division (E 1))
06/09/2016, 14:20
A conceptual design for a pellet injection system will be worked out, capable to support key missions of the new tokamak device JT-60SA. For exploitations in view of ITER and to resolve key physics and engineering issues for DEMO, several tasks were assigned to this system. Physics investigations aim at operation at high density in ITER and DEMO relevant plasma regime above Greenwald density,...
Nicolo Marconato
(Consorzio RFX (CNR)
06/09/2016, 14:20
The ITER Heating Neutral Beam (HNB) injectors shall be protected from stray magnetic field (several hundreds of mT) produced by the ITER PF coils and plasma current. Such stray field would hamper the production of negative ions, deflect ion trajectories in the accelerator and cause intolerable heat load on neutralizer and beam line components. In order to keep the residual magnetic field below...
Daniele Aprile
(Consorzio RFX)
06/09/2016, 14:20
In the multi-beamlet, negative-ion based Heating Neutral Beam (HNB) Injectors presently used in fusion research, arrays of permanent magnets are embedded in the Extraction Grid (EG) for the suppression of the unwanted co-extracted electrons. These magnets cause a significant undesired deflection of the negative ion beamlets, with a typical alternate pattern, matching the orientation of the...
Stefan Hanke
(Institute for Technical Physics)
06/09/2016, 14:20
The gas cloud inside the neutralizer of MITICA (Megavolt ITER Injector and Concept Advancement), required to neutralize the negative ion beam, will be created continuously by 20 identical nozzles providing the gas needed for different operation modes. In order to validate the design, one nozzle will be characterized in detail and for a wide range of supply conditions in a dedicated experiment...
Loris Zanotto
(Consorzio RFX)
06/09/2016, 14:20
The Acceleration Grid Power Supply supplies the acceleration grids of the MITICA experiment, the full scale prototype of the ITER Neutral Beam Injector under construction in Padua (Italy) to tackle the technical challenges and prepare for the target performance objectives ahead of operation in ITER.
The AGPS is a special switching power supply with demanding requirements: high rated power (55...
Bernd Heinemann
(ITER Technology & Diagnostics)
06/09/2016, 14:20
The negative ion source test facility ELISE represents the first step in the European R&D roadmap for the neutral beam injection (NBI) systems of ITER in order to consolidate the design and to gain early experience with a large and modular Radio Frequency (RF) negative ion source. The aim of ELISE is to demonstrate the ITER requirements with respect to extracted negative hydrogen densities...
49767.
P2.025 Preparation of the ELISE test facility for long-pulse extraction of negative ion beams
Riccardo Nocentini
(ITER Technology and Diagnostics)
06/09/2016, 14:20
The test facility ELISE (Extraction from a Large Ion Source Experiment) at IPP Garching, Germany, aims to demonstrate ITER-relevant negative ion beam parameters which are required for the NBI system of ITER. ELISE is equipped with a Radio Frequency driven source and an ITER‑like extraction system with half the ITER size. An H-- or D-- beam can be extracted for 10 s every...
Chandramouli Rotti
(Diagnostic Neutral Beam)
06/09/2016, 14:20
The Beam Line Components (BLCs) for the ITER Diagnostic Neutral Beam (DNB) and Indian Test Facility (INTF) are mainly water cooled elements made from CuCrZr which are designed to absorb heat flux up to 10MW/m2 2 (e.g. Heat Transfer Element for calorimeter) according to their position in beam line. The design of these components imposes stringent requirements of having the specific...
Jaydeepkumar Joshi
(Diagnostic Neutral Beam (DNB))
06/09/2016, 14:20
The acceleration system of Beam Source(BS) of Neutral Beam(NB) system is composed of water cooled Oxygen-Free Copper multi-aperture grid systems which is designed for focusing of beamlets to a focal point located at distance>20m from the Grounded Grid. For present application in the accelerator for DNB, this focusing is obtained using a combination of segment bending and aperture offsets. In...
Hiroyuki Tobari
(Naka Fusion Institution)
06/09/2016, 14:20
Design and manufacturing of DC 1 MV components have progressed for the ITER neutral beam injector.
A multi-conductor DC 1 MV transmission line (TL) which can transmit five-different voltages of 200 kV step simultaneously has been manufactured and tested. The TL is a gas insulation tube with SF6 gas of 0.6 MPa. A layout of those conductors inside the tube was designed through electric field...
Haejin Kim
(KSTAR Research Center)
06/09/2016, 14:20
Helicon wave coupling for efficient off-axis current drive using a traveling wave antenna has been proposed. It was found that helicon wave can drive plasma current in the mid-radius of high electron beta plasmas in medium and large size tokamak due to moderate optical thickness and wave alignment nature of helicon wave in helical magnetic field. KSTAR tokamak can be a good platform to test...
Hyunho Wi
(KSTAR Research center)
06/09/2016, 14:20
Steady-state operation of a DEMO-like tokamak requires substantial off-axis current be driven by external current drive systems. Non-inductive current drive is needed to complement the bootstrap current to support the plasma current in steady state. Recently, helicon wave current drive at frequencies of 500~700 MHz is gained much attention to achieve off-axis current drive with high...
Jeehyun Kim
(Heating and current drive team)
06/09/2016, 14:20
The KSTAR LHCD system is to be upgraded for RF power up to 4 MW in 2020. The basic configuration of the system is composed of eight 5-GHz 500-kW CW klystrons, low-loss transmission line with oversized circular waveguide, and PAM launcher for the mid-plane injection. An off mid-plane injection near the upper diverter is also under consideration. A preliminary study based on a mid-plane PAM...
Taesik Seong
(Department of Physics)
06/09/2016, 14:20
The KSTAR LHCD system is using a 5-GHz, 0.5-MW c. w. klystron and oversized rectangular waveguides. The WR187 output waveguide of the klystron transmits the RF power to the LH launcher via 80-m of transmission line composed of WR284 oversized rectangular waveguide. The overall transmission loss was about 34% including 26% of Ohmic loss. In order to transfer RF power effectively from a klystron...
Brendan Crowley
(DIII-D National Fusion Facility)
06/09/2016, 14:20
The Neutral Beam system on DIII-D consists of eight ion sources. The basis of the DIII-D NB system is the Common Long Pulse Source (CLPS). The CLPS is an 80 kV high perveance, deuterium positive ion based system delivering up to 2.5 MW per source. The ion source is a filament driven magnetic bucket design and the accelerator is a slot and rail tetrode design with vertical focusing achieved...
Mirela Cengher
(General Atomics)
06/09/2016, 14:20
The gyrotron complex on DIII-D has been updated and comprises six gyrotrons installed and routinely operating reliably for injection of up to 3.6 MW into the plasma. The operational maximum of 5 s pulse length for the six gyrotrons allows up to 18 MJ total energy to be injected into the plasma. Recent system upgrades include faster launcher mirror scans and control by the plasma control...
Chiara Piron
(Consorzio RFX (CNR)
06/09/2016, 14:20
The RAPTOR - RApid Transport simulatOR code [F. Felici et al 2011 Nucl. Fusion 51 083052] is a model-based control-oriented code that predicts Tokamak plasma profile evolution in real-time. One of its key applications is in a state observer, where the real-time predictions are combined with the measurements of the available diagnostics, yielding a complete estimate of the plasma profiles.The...
Raffaele Martone
(Department of Industrial and Information Engineering, Seconda Università di Napoli, Aversa, Italy)
06/09/2016, 14:20
The Reversed Field Pinch configurations are characterized by strong asymmetries [1]; in order to prevent or mitigate possible consequent instabilities, suitable control systems are required. In RFX-mod (Padua, Italy), such a system includes a number of 192 saddle coils, independently controlled, fully covering the toroidal surface and operating in a coordinate strategy. An equal number of...
Paolo Bettini
(Consorzio RFX)
06/09/2016, 14:20
RFX-mod is equipped with an advanced active control system of MHD instabilities, which consists of 48x4 saddle coils, housed inside a stainless steel Toroidal Support Structure, and 48x4 radial field sensor loops processed in real time to drive the currents in the control coils. Thanks to the high flexibility of this system [1], RFX-mod operations in the last years have allowed to reach the...
Luca Grando
(Consorzio RFX)
06/09/2016, 14:20
RFX [1] was originally designed with a load assembly consisting of a vacuum vessel (VV) and a thick aluminum stabilizing shell, with two poloidal and two equatorial cuts (i.e. gaps). After several years of experimental campaigns, a major modification of the RFX load assembly has been introduced [2], consisting in the substitution of the aluminum shell with a thin Copper Shell (CS) and the...
Zhengping Luo
(Institute of Plasma Physics)
06/09/2016, 14:20
The Parallel plasma equilibrium reconstruction code PEFIT [1], first developed for real-time plasma shape control of the EAST tokamak (and capable of one full equilibrium reconstruction in 300ms with a calculation grid size in 65x65) is being adapted for use on MAST. PEFIT is based upon the EFIT equilibrium code algorithm, but rewritten in C using the CUDATMTM architecture in order...
Seongcheol Kim
(Department of Nuclear Engineering)
06/09/2016, 14:20
Mitigation of heat and particle fluxes reaching on divertor plates is still a critical problem even though innovative divertor concept such as super-X and snowflake divertors have been suggested. A new divertor concept for the reduction of heat and particle fluxes is to convert thermal energy to electrical energy by separating electrons from the plasma with appropriate magnetic field....
Ngoc Minh Trang Vu
(IRFM)
06/09/2016, 14:20
The control of the safety factor q and/or the electronic temperature Te profiles is a key issue to achieve advanced plasma scenarios with high repeatability. This paper will discuss the new results of such plasma internal profile control on TCV, using total plasma current Ip, and ECCD heating source. The issue is that only the ECCD heating power is controlled, since the distributed heating...
Galina Kuzmina
(National Research Centre Kurchatov Institute, Moscow, Russian Federation)
06/09/2016, 14:20
Presented work is related to the development and creation of hardware and software of Plasma Control System (PCS) platform of the modernized now tokamak T-15 [1] for the integration, configuration, testing and start-up algorithms for the calculation of electrical installation parameters, as well as for the modeling of the experiment scenario with taking into account of the real-time magnetic...
Heung-Su Kim
(National Fusion Research Institute)
06/09/2016, 14:20
Noise width (δV/V) and drift level (ΔV/Δt) in the magnetic measurements by using sensors such as magnetic field probes (MPs) and flux loops (FLs) has been fully satisfied with the requirements (δV/V < 2% and (ΔV/V)/ Δt < 2% for 60 s), for the plasma control in the KSTAR tokamak before the in-vessel control coil (IVCC) is used to control plasma shapes. From the experimental campaign of 2010,...
Jan Horacek
(Tokamak)
06/09/2016, 14:20
In order to avoid surface melting of divertor targets of big tokamak fusion reactors by localized ELM heat loads, we study a technique of spreading the flux by harmonic divertor strike point sweeping with a dedicated in-vessel twin-coil. If the sweep frequency gets above 1/tELMdecaydecay~300 Hz, local ELM plasma heat flux suppresses significantly (by...
Arkady Serikov
(Institute for Neutron Physics and Reactor Technology)
06/09/2016, 14:20
This paper presents new results of neutronics analysis performed in support for the design development of the Tritium and Deposit Monitor (TDM) to be installed inside the ITER Equatorial Port Plug (EPP) #17. This monitor is a laser based diagnostics to provide information about the tritium content in the deposited layer on the inner baffle of the ITER divertor. Neutronics analysis is performed...
Raul Luis
(Instituto de Plasmas e Fusão Nuclear)
06/09/2016, 14:20
The ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations, also known as gaps 3, 4, 5, and 6, complementing the magnetic diagnostics system. For gaps 4 and 6, the antennas are to be installed in-vessel between two blanket shield modules. The microwave signal is...
Nuno Cruz
(Instituto de Plasmas e Fusão Nuclear)
06/09/2016, 14:20
The Radial Neutron Camera (RNC) diagnostic is a neutron detection system with multiple collimators aiming at characterizing the neutron emission that will be produced by the ITER tokamak. The RNC plays a primary role for basic and advanced plasma control measurements and acts as backup for system machine protection measurements.
To achieve its goals, the RNC diagnostic needs to acquire,...
Fabio Moro
(Department of Fusion and Nuclear Safety Technology)
06/09/2016, 14:20
The ITER Radial Neutron Camera (RNC) is a multichannel detection system hosted in the Equatorial Port Plug 1 (EPP 1) designed to provide information on the neutron source total strength and emissivity profiles through the measurement of the uncollided neutron flux along a set of collimated lines of sight (LOS). Furthermore the ion temperature profile and fuel ratio (nd/nt) can be assessed by...
Anders Hjalmarsson
(Department of Physics and Astronomy)
06/09/2016, 14:20
The High Resolution Neutron Spectrometer (HRNS) system for ITER is an array of neutron spectrometers with the primary function to provide measurements of the fuel ion ratio, nT/nD, in the plasma core. Supplementary functions are to assist or provide information on fuel ion temperature and energy distributions of fuel ions and confined alpha-particles. The ITER requirement for the HRNS primary...
Mykyta Varavin
(Institute of Plasma Physics AS CR)
06/09/2016, 14:20
The COMPASS tokamak is equipped by the 2-mm microwave interferometer. This interferometer measures the electron density integrated along the central chord. Two VCO oscillators stabilized by the PLL together with multipliers generate two probing waves of the close frequency 139.3 and 140 GHz. The digital 2π-phase detector in the receiving part compares the phase between these probing waves. The...
Pavel Hacek
(Faculty of Mathematics and Physics)
06/09/2016, 14:20
Atomic beam probe (ABP) is a diagnostic tool using a detection of ions coming from an ionized part of a diagnostic beam in tokamaks. The method allows measurements of plasma density fluctuations and fast variations in the poloidal magnetic field. Therefore, it gives the possibility to follow fast changes of edge plasma current, e.g. during ELMs in H-mode.
The test detector has been installed...
Ales Havranek
(Institute of Plasma Physics of the CAS)
06/09/2016, 14:20
The COMPASS tokamak has been recently equipped with two new fast color cameras Photron FASTCAM Mini UX100 operating in visible light. A new node, including both software and hardware, was developed for these cameras to ensure automatic and reliable operation integrated to the control and data acquisition system of COMPASS. The node provides camera function control, parameter setting, data...
Petr Vondracek
(Institute of Plasma Physics of the CAS)
06/09/2016, 14:20
A new fast infrared camera Telops FAST-IR 2K was purchased on the COMPASS tokamak recently. It is equipped with a MWIR (medium wavelength infrared, 3-5 μm) InSb detector and is possible to reach framerate of 1.917 kHz in a full frame acquisition mode (320x256 px.) and up to 90 kHz in a sub-windowed acquisition (64x4 px.).
The camera allows e.g. automatic exposure control, providing autonomous...
Jaromir Zajac
(Institute of Plasma Physics AS CR)
06/09/2016, 14:20
The microwave reflectometry system on COMPASS tokamak uses the frequency modulated continuous wave (FM-CW) in K and Ka bands. The fast swept synthesizer together with the simple homodyne detection provides the complex beat frequency spectrum for the density profile reconstruction. The homodyne detection scheme limits the other applications like the Doppler reflectometry, therefore the sheme is...
Mark Szutyanyi
(Department of Mathematics and Computational Sciences)
06/09/2016, 14:20
The physics of Edge Localized Modes (ELM) is one of the most studied scientific fields in fusion research. Automatic detection of ELMs in different diagnostic signals is an important initial step during massive experimental data analysis.
This contribution contains the description of the generalized Sequential Probability Ratio Test (g-SPRT) method used for automatic ELM detection in different...
Michael Grahl
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
WENDELSTEIN 7-X and its superconducting coil system is designed for research on steady-stateoperation of stellarators. This sets high requirements on the control and data acquisition (CoDaC)system, with the archive database as one of its main components. W7-X ArchiveDB [1] is the centralstorage system for all engineering and scientific data. It stores raw data as well as processed data...
Andre Carls
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
Wendelstein 7-X (W7-X) has been finally commissioned in 2015 and is now in its first stage of operation. Due to the complex structural design and a limited life time of some components, each step of W7-X commissioning and operation is carefully monitored by a considerable amount of different sensors.
Unlike the fast machine control or the fast experiment data acquisition, the machine...
Didier Chauvin
(CEA de Cadarache DSM/IRFM)
06/09/2016, 14:20
The Wendelstein 7-X fusion device at Max-Planck-Institut für Plasma Physik (IPP) in Greifswald produced its first hydrogen plasma on 3rdrd February 2016. This marks the start of scientific operation. Wendelstein 7-X is to investigate this configuration’s suitability for use in a power plant. In order to allow for an early integral test of the main systems needed for plasma operation...
Ireneusz Ksiazek
(Institute of Physics)
06/09/2016, 14:20
The C/O monitor for W7-X will be a spectrometer of special construction with high throughput and high time resolution, suitable for controling concentration of main impurities in plasma. The spectrometer will be fixed at horizontal position and at wavelengths corresponding to Lyman a lines of H-like ions of oxygen (at 1.9 nm), nitrogen (at 2.5 nm), carbon (at 3.4 nm) and boron (at 4.9 nm). Its...
Guruparan Satheeswaran
(Forschungszentrum Jülich GmbH)
06/09/2016, 14:20
A multi-purpose manipulator (MPM) system is attached at an outer cryostat vessel port in the equatorial plane to transport electrical probes and targets to the edge of the inner vessel. From this parking position where the tip of the probe coincides with the inner vessel wall a fully controlled movement into the edge plasma for all magnetic field configurations is feasible. The distributed...
Tamas Szabolics
(Wigner Research Centre for Physics)
06/09/2016, 14:20
In the past few years a ten channel video diagnostics system was developed, built and installed for Wendestein 7-X stellarator (W7-X). The system is based on EDICAM (Event Detection and Intelligent Camera) CMOS cameras (400 fps @ 1.3 Mpixel). In the first W7-X experimental campaigh (OP1.1) the video diagnostic system is not integrated into the central control and data acquisition system of...
Tomasz Fornal
(Department of Nuclear Fusion and Plasma Spectroscopy)
06/09/2016, 14:20
Measurements of soft X-ray radiation from plasmas is a standard diagnostic which is used in many different fusion devices. Analysis of X-ray emission delivers among others, information about the electron density and temperature as well as can deliver an information about the impurity content in the plasma.
The paper describes design of the soft X-ray diagnostic, multi-foil system (MFS,) for...
Natalia Krawczyk
(Department of Nuclear Fusion and Plasma Spectroscopy)
06/09/2016, 14:20
The Wendelstein 7-X (W7-X) stellarator started its operation at the end of 2015. The first operation phase is conducted both with helium and hydrogen as working gas and has achieved first plasmas in the order of 500ms at the time this abstract has been written. The initial experiments have also been devoted to commissioning, tests and optimization of diagnostic systems.
In this paper we report...
Christian Brandt
(Max-Planck-Institute for Plasma Physics)
06/09/2016, 14:20
The quasi-steady state high power plasma experiments at Wendelstein 7-X are expected to become pioneering research benchmarking the advanced stellarator concept. The results will bring comparisons to the huge amount of experimental findings in other stellarator and tokamak devices. After the successful start of hydrogen plasmas in February 2016, the set of plasma diagnostics will be extended...
Ulrich Neuner
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
Thirteen Rogowski coils have been installed in the vacuum vessel of the stellarator Wendelstein 7-X (W 7-X). They are designed to measure the equilibrium plasma currents as Pfirsch-Schlüter current and bootstrap current. The coils will be calibrated using a conductor positioned inside the plasma vessel with an alternating current passing through it. The response of the coils is measured and...
Dirk Nicolai
(Institut für Energie- und Klimaforschung - Plasmaphysik)
06/09/2016, 14:20
The investigation of edge plasmas at W7-X requires a flexible tool for integration of a variety of different diagnostics as e. g. electrical probes, probing magnetic coils, material collection, or material exposition probes, and gas injection. A multi-purpose manipulator (MPM) system has been developed and attached to the W7-X vessel before the operational phase 1.1. The system was designed as...
Youngseok Lee
(KSTAR Research Center)
06/09/2016, 14:20
Long-pulse D-D plasma operation in the annual KSTAR plasma campaign is performed and involved Ohmic heating and auxiliary heating such as a neutral beam injection (NBI) of high power with deuterium beams. The NBI heating power reached up to 6 MW at the moment.
In addition, many energetic runaway electrons are also observed through hard-X ray (HXR) monitoring during the operation. Runaway...
Jong-ha Lee
(National Fusion Research Institute (NFRI))
06/09/2016, 14:20
To measure Zeff profile, most plasma machine equipped brehmsstrahlung measurement system like as filterscope diagnostic. In KSTAR, however, a new type brehmsstrahlung measurement system developed and tested at single point in KSTAR 10th campaign in last year.[1] In 2016 KSTAR campaign, to Zeff profile measurement, we expand this concepts of brehmsstrahlung measurement system to multi points;...
Y. Yu
(School of Nuclear Science and Technology)
06/09/2016, 14:20
Abstract：In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum...
Md Mahbub Alam
(Advanced Energy Engineering Science)
06/09/2016, 14:20
In QUEST (Q-shu University Experiments with Steady-State Spherical Tokamak), the achievement of the steady-state operation for long time discharge is one of its project objectives. For the achievement of the long time discharge, the identification of the plasma shape and position in real-time is important during the operation of the tokamak. By observing the temporal behaviours of the plasma...
Andrea Rizzolo
(Consorzio RFX)
06/09/2016, 14:20
This paper describes the final design of the Short-Time Retractable Instrumented Kalorimeter Experiment (STRIKE) for the SPIDER experiment (Source for Production of Ions of Deuterium Extracted from Radio frequency plasma) under construction at the Consorzio RFX premises. The STRIKE diagnostic will be used to characterise the SPIDER beam during short pulse operation (several seconds) to verify...
Zito Pietro
(FSN-FUSTEC-IEE)
06/09/2016, 14:20
JT-60SA is a Superconducting Tokamak in the framework of the Broader Approach Agreement between Europe and Japan. For this International Project, both the Italian National Agency for New Technology, Energy and Sustainable Economic Development (ENEA) and Comissariat à l’Energie Atomique et aux Energies alternatives (CEA) are providing ten AC/DC converters for the poloidal superconducting...
49823.
P2.082 Final tests of four switching network units procured by the European Union for JT-60SA
Miguel Pretelli
(Power Electronics)
06/09/2016, 14:20
Switching Network Units (SNUs) are inserted in the power supply circuits of modern tokamaks for plasma initiation. In the framework of the “Broader Approach” agreement, the four SNUs for the superconducting modules of the JT-60SA Central Solenoid will be procured by European Union through the Italian Agency ENEA.
The design is based on the synchronized operations of a light electromechanical...
Elena Gaio
(Power System)
06/09/2016, 14:20
Effective control of Resistive-Wall-Modes (RWM) is mandatory in JT-60SA, the satellite tokamak under construction in Naka (Japan), since one of its main objectives is to reach steady-state high-beta plasmas.
The RWM control system is based on a set of 18 in-vessel sector coils, placed on the plasma side of a conductive wall and individually fed by a dedicated fast power supply system...
Kyohei Natsume
(Tokamak System Technology)
06/09/2016, 14:20
JT-60SA is a tokamak device using superconducting coils to be built in Japan, as a joint international research and development project involving Japan and Europe. The JT-60SA helium refrigerator system (HRS) supplies supercritical or gaseous helium to cold components: superconducting coils, coil supporting structures, cryopumps, high temperature superconductor current leads (HTS CL), and...
Daniel Ciazynski
(IRFM/STEP)
06/09/2016, 14:20
The Toroidal Field system of the JT-60SA tokamak is composed of 18 NbTi superconducting coils. Half of them are provided by France within the Broader Approach Agreement. These coils are manufactured by General Electric (ex-Alstom) at Belfort, France. Each TF coil is composed of 6 cable-in-conduit conductor lengths, wound in double-pancakes, carrying a nominal current of 25.7 kA at a...
Yawei Huang
(Institute of Research into the Fundamental Laws of the Universe)
06/09/2016, 14:20
In order to check the performance of the JT-60SA Toroidal Field (TF) coils and hence mitigate their possible fabrication risks, a series of tests have been carried out in the Cold Test Facility (CTF) at CEA Saclay in nominal conditions at 5 K and 25.7 kA. One major test performed is the so called “temperature margin test" during which the inlet helium temperature of the winding pack is...
Vicente Queral
(National Fusion Laboratory)
06/09/2016, 14:20
Coil casings and coil frames for stellarators are geometrically complex components at high accuracy. A method of additive manufacturing combined with fibre-reinforced resin casting has been recently experimented [1] for the fabrication of complex coil frames. The method is named 3Dformwork and consists of additive fabrication of a hollow thin shell which is filled with resins or other...
Matthias Schneider
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
The Quench-Detection-System of the fusion experiment Wendelstein 7-X detects quench events within the superconducting magnet system constructed of 50 non-planar and 20 planar coils, 14 current leads and the bus bars. In the event of a quench the QD-System triggers the power supply of the magnetic system to shut down.
The QD-System monitors the superconducting system by 486...
Frank Fullenbach
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
The magnet system of the stellarator fusion device Wendelstein 7-X (W7-X) is composed of three different groups of coil systems. The main magnetic field is created by a superconducting magnet system that is accompanied by two sets of normal conducting coil groups, the Control Coils inside the plasma vessel and the Trim Coils (TC) positioned outside of the cryostat.
The TC system consists of...
Sheng Li
(State Key Laboratory of Electrical Insulation and Power Equipment)
06/09/2016, 14:20
The quench protection switch (QPS) is very important for ensuring the safety of the PF and TF coils of a superconductive Tokomak. The main function of a QPS is to protect the magnet as the coil quench occurs. Besides, a QPS has to withstand almost all of the coil current of some tens of kA flowing through it for a long time in the normal operation condition. This task is undertaken by the...
Qiaosen Wang
(State Key Laboratory of Electrical Insulation and Power Equipment)
06/09/2016, 14:20
Superconducting magnet is one of the most crucial components in a superconducting Tokamak. During the normal operation stage, high current of some tens of kA flows through the magnet with large inductance of ~1H. Therefore, extremely large energy (~0.1-10GJ) is stored in the magnet, which must be dissipated in the case of magnet quench in certain duration before the occurrence of local or even...
Tindaro Cicero
(Fusion for Energy)
06/09/2016, 14:20
The Normal Heat Flux (NHF) First Wall (FW) panels consist of a series of fingers, which represent the elementary plasma facing units and are designed to withstand 15,000 cycles at 2 MW/m22. The fingers are mechanically joined and supported by a back structural element called “supporting beam”. The structure of a finger is made of three different materials, stainless steel for the...
Stefano Banetta
(Fusion for Energy)
06/09/2016, 14:20
This paper describes the main activities carried out for the conclusion of the EU-DA prequalification process for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the High Heat Flux (HHF) testing of a reduced scale FW prototype (Semi-Prototype (SP)). This component is manufactured by the AREVA Company in France and has a dimension of 221 x...
Rafael Enparantza
(Design)
06/09/2016, 14:20
The Normal Heat Flux (NHF) First Wall (FW) panels are designed to withstand the heat flux from the plasma inside ITER. These components are made of beryllium tiles bonded to a copper alloy and 316L (N) stainless steel heat sink. A NHF FW panel consists of several fingers as elementary plasma facing units. This this paper presents the experimental stress and deformations measured on a...
Sergey Tomilov
(JSC “NIKIET”)
06/09/2016, 14:20
In the framework of PA realization, specialists from NIKIET and Efremov Institute are developing a design of First Wall (FW) Full Scale Prototype (FSP) in order to demonstrate its manufacturability and qualify critical technological processes. Design of FW FSP is developed based on the FW 14 type A. The semi-prototype has been manufactured in order to verify the FW design. Based upon the...
Maxim Sviridenko
(JSC NIKIET)
06/09/2016, 14:20
The JSC NIKIET is responsible for the manufacture of the First Wall (FW) beam, the fingers bodies, the mechanical attachment system and electrical connection system of the FW panel to the shield block (SB) in the framework of Procurement Arrangement 1.6.Р1А.RF.01 dated 14.02.2014.
The Electrical strap (ES) is located on the FW rear surface and used for providing current through the FW to the...
Karel Samec
(Centrum Výzkum ŘEŽ)
06/09/2016, 14:20
The heat flux on plasma-facing components in ITER, and even more so in the projected DEMO reactor will reach values in the order of several Megawatt per square meter. Evacuating this heat in a reliable manner is key to the robustness and safety of operation of any fusion reactor.
The current state-of-the-art for cooling plasma-facing components relies on cooling a high heat-resistant structure...
Phani Domalapally
(Centrum výzkumu Řež s.r.o.)
06/09/2016, 14:20
The heat loads on the First Wall (FW) of the European DEMO are not yet defined, but when extrapolated from ITER, the loads can be quite high. As the DEMO will use Eurofer 97 as the structural material and Pressurized Water Reactor (PWR) conditions at the inlet, i.e. 15.5 MPa and 285 °C, the design of the heat sink gets complicated as the thermal conductivity of the heat sink material is quite...
Pavel Zacha
(Energy Engineering)
06/09/2016, 14:20
The first wall cooling of the fusion power reactor DEMO is an important part of the fusion power plant development. A cooling ability at high heat flux conditions will affect a lifetime period of the first wall modules having a direct impact on the operating costs of the fusion power plant. According to current knowledge, the water cooling provides the largest ability to remove the high heat...
Ladislav Vesely
(Faculty of Mechanical Engineering)
06/09/2016, 14:20
Based on the requirements of F4E, an experimental device HELCZA (High Energy Load Czech Assembly) was designed for high heat flux cyclic loading of plasma-facing components of the ITER reactor, primarily for testing of the full-size first wall modules and divertor inner vertical targets.
Testing is carried out by a high power electron beam heating, and a deviation of the heat flux density at...
Radek Skoda
(Department of Energy Engineering)
06/09/2016, 14:20
The paper deals with optimal electron beam heat distribution on the HELLCZa experiment calculating the flatness of the distribution of heat input and distribution of surface temperature of various samples. A computer program has been developed for balancing the heat flux in the construction materials of the sample. The first boundary condition for this calculation were primarily functions...
Richard Jilek
(Centrum výzkumu Řež s.r.o.)
06/09/2016, 14:20
Commissioning phase of high heat flux test facility HELCZA
R. Jíleka,*a,*, J. Prokůpekaa, P. Gavilabb
aCentrum výzkumu Řež s.r.o. (CVR), Hlavní 130, 25068 Husinec-Řež, Czech Republic,
bFusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona, Spain
*Corresponding author: e-mail: Richard.Jilek@cvrez.cz, phone: +420 601 315 137
The high heat...
Andre Kunze
(Institute for Neutron Physics and Reactor Technology)
06/09/2016, 14:20
For the testing of helium cooled plasma facing components in HELOKA-HP homogeneous surface heat flux densities of up to 500 kW/m² have to be reproduced. It has been proposed to use infrared radiation heaters which consist of several quartz glass (fused silica) tubes with tungsten filaments inside to generate the heat flux. This paper presents a numerical model of the latest type of heater...
Muyuan Li
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
Maintenance of structural integrity under high-heat-flux (HHF) fatigue loads is a critical concern for assuring the reliable HHF performance of a plasma-facing divertor target component. Loss of structural integrity may lead to structural as well as functional failure of the component.
Currently, a full tungsten divertor was chosen by ITER Organization, and plenty of HHF qualification tests...
Eunnam Bang
(KSTAR research center)
06/09/2016, 14:20
This paper deals with the first commissioning of active cooling system for plasma-facing components (PFCs) and coolant removal system. During 2015 KSTAR campaign, we have achieved a 55 sec long pulse H-mode. However, some plasma shots were terminated, not because of instabilities or limitation of heating power, but because of safety limit applied to the PFC temperature: upper boundary to lock...
Jaehyun Song
(Advanced Engineering Division)
06/09/2016, 14:20
The tungsten (W) brazed flat type mock-up with swirl tube which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade. The mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 8 MW/m22 for 20 sec duration at KoHLT-EB in KAERI. In this paper, for comparison of...
Sungjin Kwon
(DEMO Technology Division)
06/09/2016, 14:20
The preliminary conceptual design study on the Korean fusion demonstration reactor (K-DEMO) tokamak consists of the vacuum vessel, the in-vessel components, and the superconducting magnet system, and so on [1]. The K-DEMO superconducting magnet system contains 16 toroidal field (TF) coils, 8 central solenoid (CS) coils and 12 poloidal field (PF) coils. The magnetic field at the plasma center...
JongSung Park
(Fusion Engineering Center)
06/09/2016, 14:20
A preliminary study on the rigorous 2-step (R2S) based shutdown dose rate calculations has been performed for the Korean fusion demonstration reactor (K-DEMO) in the vicinity of an equatorial port area using the coupled transport and activation calculation codes of MCNP6 and FISPACT. For the shutdown dose rate calculation, the equatorial port structures and port plug including shielding blocks...
Kihak Im
(DEMO Technology Division)
06/09/2016, 14:20
A pre-conceptual design study for the Korean fusion demonstration tokamak reactor (K-DEMO) has been initiated in 2012. K-DEMO is characterized by the uniqueness of high magnetic field (BT0 = 7.4 T), major and minor radii of 6.8 m and 2.1 m, and steady-state operation.
The heat load distribution by plasma radiation onto the first walls of the in-vessel components is one of the basic inputs for...
G Douglas Loesser
(Engineering)
06/09/2016, 14:20
G. Douglas Loesser1,Joris Fellinger22, Hutch Neilson11, John Mitchell11, Marc Sibilia11, Han Zhang11, P. Titus11, Irving Zatz1,1,, Arnie Lumsdaine33, Dean McGinnis33
1Princeton Plasma Physics Laboratory, James Forestall Campus, Princeton, NJ 08542, USA
2Max-Planck-Institut für Plasmaphysik,...
Joris Fellinger
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
The modular stellarator Wendelstein 7-X in Greifswald (Germany) successfully started operation in 2015 with short pulse limiter plasmas. In 2017, the next operation phase (OP) OP1.2 will start once 10 uncooled test divertor units (TDU) with graphite armor will be installed. The TDUs allow for plasma pulses of 10 s with 8 MW heating. OP2, allowing for steady state operation, is planned for 2020...
Zhongwei Wang
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
The cryopump will be installed for the high power and long pulse operation up to 30 minutes of Wendelstein 7-X (W7-X). The cryopump system plays a critical role for capturing ash particles from the plasma, including hydrogen, deuterium and even helium. In total there are 10 independent cryopumps, one cryopump for each of the 10 discrete divertor units. The cryopump is located along the pumping...
Patrick Junghanns
(Max-Planck-Institut für Plasmaphysik)
06/09/2016, 14:20
The 890 target elements of the high heat flux divertor of Wendelstein 7-X (W7-X) are made of a CuCrZr copper alloy heat sink armored with carbon reinforced carbon (CFC) NB31 tiles. Connectors with an internal diameter of 10 mm are electron beam welded to heat sink for the water inlet and outlet. They are produced by electron beam welding thicker tubes of CuCrZr and stainless steel with a...
Jean Boscary
(Max-Planck_Institut für Plasmaphysik)
06/09/2016, 14:20
The actively water-cooled target elements of the high heat flux divertor of Wendelstein 7-X (W7-X) are designed to remove a stationary heat flux of 10 MW/m² on its main area and 5 MW/m² at the end adjacent to the pumping gap. A target element is made of a CuCrZr copper alloy heat sink armored with carbon reinforced carbon (CFC) NB31 tiles. The realization of the divertor requires the...
Clara Colomer
(IDOM Nuclear Services)
06/09/2016, 14:20
The ITER Vacuum Vessel (VV) is a double wall Stainless Steel structure that surrounds the plasma. It constitutes a major safety barrier for ITER, and, because of its function, is classified as Protection Important Component (PIC). Its design and construction has to follow the RCC-MR design code rules to verify the structural integrity under electromagnetic, thermal and seismic...
49864.
P2.127 Electromagnetic Analysis for the In-Vessel Transfer Lines of Neutron Activation System
Sunil Pak
(National Fusion Research Institute)
06/09/2016, 14:20
In ITER the neutron activation system deploys several foil samples close to the plasma to measure the neutron fluence and the fusion power. These samples are transferred in a pneumatic way along the tubes installed on the vacuum vessel wall. Therefore, the tubes, namely transfer lines, get eddy current induced during plasma disruption, leading to Lorentz force by interacting the background...
Josu Eguia
(Mechanical Engineering)
06/09/2016, 14:20
The vacuum vessel of ITER is a paradigmatic example of a gargantuan system that can only be processed in-situ and from the inside. Its assembly implies performing post welding repair operations, including machining of welding seams following the internal surface of the vacuum vessel. The requirements for the machining operations are the following: accuracy +/- 0.1 mm; dynamic machining forces...
Anna Encheva
(Tokamak Department)
06/09/2016, 14:20
ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just...
Ivan Poddubnyi
(JSC "NIKIET")
06/09/2016, 14:20
In ITER blanket system, electrical connectors (“E–straps”, ES) are used to form a low impedance electrical path from shield blocks (SB) to the vacuum vessel (VV). Main functions of ES is providing current from SB to VV. ES shall withstand electromagnetic (EM) loads and Joule heating resulted from electrical current with magnitude up to 137 kA during 300 ms, accommodate cyclic relative...
Dieter Leichtle
(Fusion for Energy)
06/09/2016, 14:20
The Ion Cyclotron Heating and Current Drive system (ICH) is designed to launch RF power into the ITER plasma, and will reside in equatorial ports (EP) 13 and 15. Shutdown dose rates (SDDR) within the ICH port interspace are required to be ALARA and less than 100 μSv/h at 1066 seconds cooling, in locations where hands-on maintenance is required. The shielding performance of...
Ivan Popov
(Mechanics and Control)
06/09/2016, 14:20
In this paper the stress-strain state of the diagnostic shield modules (DSM) and the supporting frames (ISS, PCSS), located in the upper ports #2 and #8 of the tokamak ITER is investigated.
DSM is the upper port components and has two main functions: neutron radiation protection and maintenance of rigid fixation diagnostics placed in the port. DSM is operated at high temperatures, significant...
Pivkov Andrew
(Mechanics and Control)
06/09/2016, 14:20
The primary systems of future international thermonuclear experimental reactor (ITER) have to withstand major thermal, nuclear, electromagnetic and seismic loads. Therefor engineering analysis of elements of construction plays crucial role in realizing of the project as a whole. The paper describes calculations of spatial stress-strain state from major loads arising during operation upper...
Yu-Gyeong Kim
(National Fusion Research Institute)
06/09/2016, 14:20
Korea has been manufacturing two vacuum vessels of ITER and main jointing method to in-wall shield assemblies is welding. Though in-wall shield ribs holding neutron shielding blocks should sustain various design loads such as electro-magnetic forces, earthquake and their own weights, as a part of the assembly, in-service inspections are hardly possible because they are installed between...
Dong Won Lee
(Nuclear Fusion Engineering Development Department)
06/09/2016, 14:20
After the Conceptual Design Review (CDR), Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) design is being updated for the preparation of the preliminary design phase. The manufacturability is considered based on the TBM-set model of CD phase. The overall geometry of the first wall, side wall and the breeding zone was changed slightly. Thethermal-hydraulic and mechanical...
Aleksandr Nemov
(Peter the Great Saint-Petersburg Polytechnic University)
06/09/2016, 14:20
The High Field Side Reflectometry is diagnostic equipment subjected to the conditions that are severe even for ITER: magnetic field over 9T, temperatures up to 700 ºC, strongly non-uniform temperature field, specific shape of the equipment with length of in-vessel waveguides about 10m and location of waveguides close to the blanket connectors where large halo currents are expected during...
Ivan Kirienko
(Mechanics and controls)
06/09/2016, 14:20
The presentation is focused on the simulation results and approaches used for loading analyses made for DTS in-vessel equipment, including spatial stress strain state, seismic analysis, electromagnetic analysis as well as the most important load combinations.
Finite element model of the construction was updated according with updated DTS components design and separated on the following...
Jiaming Jiang
(Fusion Center for Scientist)
06/09/2016, 14:20
HL-2M RMP (Resonance Magnetic Perturbation) Coils is designed to provide a resonant perturbation magnetic field for high beta plasma operation scenarios stability control, such as Edge Localized Modes (ELMs) suppression control, Resistance Wall Model (RWM) fast control and Error magnetic field correction control, etc.
Especially, ELMs result in impulsive burst of energy deposition on to the...
Bostjan Koncar
(Jožef Stefan Institute)
06/09/2016, 14:20
Thermal radiation analysis of the DEMO tokamak based on the updated CAD design of in-vessel components and magnet system has been carried out. For the purpose of the analysis, Vacuum Vessel Thermal Shield (VVTS), Cryostat Thermal Shield (CTS) and some support structures have been created additionally (on a conceptual level) to complement the overall DEMO CAD design model. The Finite Element...
Juan-Pablo Catalan
(Energy Engineering Department)
06/09/2016, 14:20
Shutdown dose rate (SDR) analysis plays a key role in the design of fusion facilities like ITER and DEMO. One of most used methodology to carry out SDR calculations is the rigorous-two-step (R2S) based on the coupling of transport and activation calculations. Currently, one of the most relevant lacks of this method is the possibility to propagate the effect of the uncertainties accumulated...
Heejin Shim
(Blanket Technology Team)
06/09/2016, 14:20
Molybdenum disulfide (MoS2) coating was deposited by magnetron sputtering onto the target material. The coatings of deposited MoS2 can be used in high vacuum and aerospace environments for lubrications purposes, which ultra-low friction is desirable. For these reason, the sputtered MoS2 coating method is primarily considered for ITER components and their mechanical assemblies. A common...
Piero Agostinetti
(Consorzio RFX (CNR)
06/09/2016, 14:20
A new technique, called Vacuum Tight Threaded Junction (VTTJ), has been developed and patented by Consorzio RFX, permitting to obtain low-cost and reliable non welded junctions, able to maintain vacuum tightness also in aggressive environments. The technique can be applied also if the materials to be joint are not weldable and for heterogeneous junctions (for example, between steel and copper)...
Pietro Alessandro Di Maio
(Energia)
06/09/2016, 14:20
Within the framework of EUROfusion R&D activities an intense research campaign has been carried out at the University of Palermo, in close cooperation with ENEA Brasimone, in order to investigate the thermo-mechanical performances of the DEMO Water-Cooled Lithium Lead breeding blanket (WCLL). In particular, attention has been paid to the most recent geometric configuration of the DEMO WCLL...
Antonio Froio
(NEMO Group)
06/09/2016, 14:20
The development of a system-level thermal-hydraulic model of the whole EU DEMO tokamak has been launched by the EUROfusion Project Management Unit. In order to follow the progress in the design of the tokamak components, the model should be developed in an object-oriented fashion, to ensure a high modularity. Within this framework, the first block of the model is under development at...
49885.
P2.150 CFD simulation of the magnetohydrodynamic flow inside the WCLL breeding blanket module
Alessandro Tassone
(Dipartimento di Ingegneria Astronautica)
06/09/2016, 14:20
The interaction between the molten metal and the plasma-containing magnetic field in the breeding blanket of a Tokamak fusion reactor causes the onset of a magnetohydrodynamic (MHD) flow. In order to properly design the blanket, it is important to quantify how and how much the flow features are modified compared with an ordinary hydrodynamic flow. This paper aims to characterize the evolution...
Leo Buhler
(Institute for Nuclear and Energy Technologies)
06/09/2016, 14:20
A number of liquid metal blanket designs for applications in nuclear fusion reactors is currently under development. In the water cooled lead lithium (WCLL) blanket Eurofer97 is used as structural material and liquid PbLi as breeder, neutron multiplier, and as heat transfer medium. The released heat is removed by water at a pressure of 155 bar (pressurized water reactor conditions, 285°C -...
49887.
P2.152 Structural analysis of the back supporting structure of the DEMO WCLL outboard blanket
Maria Lorena Richiusa
(Department of Energy)
06/09/2016, 14:20
Within the framework of EUROfusion R&D activities an intense research campaign has been carried out at the University of Palermo, in close cooperation with ENEA Brasimone, in order to investigate the thermo-mechanical performances of the Back-Supporting Structure (BSS) outboard segment of the DEMO Water-Cooled Lithium Lead breeding blanket (WCLL). In particular, the configuration of the BSS...
Songlin Liu
(Institute of Plasma Physics Chinese Academy of Sciences)
06/09/2016, 14:20
China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokomak reactor. Its major radius is 5.7m, minor radius is 1.6m and elongation ratio is 1.8. It is possible upgrade to R~6 m, a~2 m. CFETR mission and objectives are to bridge gaps between ITER and DEMO, and to realize fusion energy application in China. CFETR has two phases. Phase I is to demonstrate full cycle of...
Hui Bao
(School of Nuclear Science and Technology)
06/09/2016, 14:20
Square channel is widely used in the conceptual design of water cooled blanket of fusion reactor for cooling and providing appropriate inner temperature field for tritium breeding. Thermal hydraulic design of blanket directly determines the heat transfer efficiency and safety characteristics of fusion reactor. Therefore, thermal-hydraulic characteristics of square channel should be...
Kecheng Jiang
(Institute of Plasma Physics)
06/09/2016, 14:20
The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). From the security point of view, the thermal-hydraulic analysis is very essential because the blanket should remove the high heat flux radiated from the plasma and the volumetric heat generated by neutron wall loading. For the normal state of plasma burning, the jumped peak...
Xiaokang Zhang
(Institute of Plasma Physics Chinese Academy of Sciences)
06/09/2016, 14:20
The water-cooled ceramic breeder (WCCB) blanket is one of the candidates of Chinese fusion engineering test reactor (CFETR). WCCB blanket will produce radioactive waste during its operation and decommissioning processes. The radioactive characteristics of WCCB blanket, including solid structure and functional material and the liquid water coolant, are of importance for the replacement and...
Pinghui Zhao
(School of Nuclear Science and Technology)
06/09/2016, 14:20
A conceptual structural design of Water-Cooled-Solid-Breeder (WCSB) blanket, one of the breeding blanket candidates for China Fusion Engineering Test Reactor (CFETR), is now being carried on by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP). To validate the reliability of the designed blanket module, detailed thermal-hydraulic analysis is necessary. The computational fluid...
Geon-Woo Kim
(Nuclear Engineering)
06/09/2016, 14:20
Tokamak reactors like ITER or fusion DEMO reactors have serious concerns about material damages to plasma facing components (PFC) due to plasma instabilities. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. In addition, high thermal stresses due to rapid...
Angel Ibarra
(CIEMAT, Avda. Complutense 40, 28040 Madrid, Spain)
06/09/2016, 14:20
In the framework of the EUROfusion programme, Dual Coolant Lithium Lead (DCLL) breeding blanket is being investigated as a candidate for European DEMO, which is based on the use of Pb-17Li as breeder and coolant (“self-cooled breeding zone”) and high-pressure helium for cooling the structures made of reduced-activation ferritic steel (EUROFER). During the first part of the project, a...
Luis Maqueda
(Esteyco Mechanics)
06/09/2016, 14:20
General purpose finite element (FE) softwares can be readily used for the stationary analysis of breeding blankets of a nuclear fusion reactor. However, the analysis of transient effects generated during the pulsed operation mode requires transient simulations to be carried out. Nowadays, a commercial tool which can be directly used for these transient simulations with affordable computational...
Fernando Roca Urgorri
(National Fusion Laboratory)
06/09/2016, 14:20
The Dual Cooled Lithium Lead (DCLL) blanket is one of the four breeder blanket technologies under consideration within the framework of EUROfusion Consortium activities. The aim of this work is to develop a preliminary model that can track the tritium concentration along each part of the DCLL blanket and their ancillary systems at any time.
Because of tritium’s nature, the phenomena of...
Ivan Fernandez-Berceruelo
(Fusion National Laboratory)
06/09/2016, 14:20
The Dual Coolant Lead-Lithium (DCLL) is one of the breeding blanket concepts under investigation in EUROFusion. This concept is characterized by the use of self-cooled eutectic PbLi as neutron multiplier and tritium breeder and carrier, whereas supercritical helium is used to cool the first wall and other parts of the structure.
The thermal-hydraulic (TH) design of the breeding blanket, as the...
Daniel Suarez
(Department of Physics)
06/09/2016, 14:20
The conceptual design of the European Dual Coolant Lead Lithium (DCLL) breeding blanket is currently being developed in the frame of EUROfusion Project. To this aim, it is of utmost interest to estimate critical flow parameters such as: (1) pressure drop and heat transfer coefficient at both helium and lithium sides, and (2) tritium permeation ratio. Pressure drop in purely hydrodynamic flows...
Qingyun He
(School of Nuclear Science and Technology)
06/09/2016, 14:20
Liquid metal (LM) blanket concepts are designed by many countries due to its attractive features such as geometric adaptability, good thermal conductivity and heat carrying capacity, et al. However, they all have feasibility issues associated with magnetohydrodynamic (MHD) interactions under the environment of a strong control magnetic field and the flowing high electrical conductivity LM. The...
Fumito Okino
(Institute of Advanced Energy)
06/09/2016, 14:20
DCLL blanket has high energy recovery efficiency. Nevertheless by several technical issues, such as MHD pressure drop, tritium permeation and energy conversion membrane corrosion, technical readiness level(TRL) of DCLL is relatively not high. To breakthrough this situation, the authors propose a new method to recover tritium and heat from liquid lithium-lead (PbLi) droplet by non-contact in...
Andrei Khodak
(Princeton Plasma Physics Laboratory)
06/09/2016, 14:20
The analysis of Dual-Coolant Lead–Lithium (DCLL) blankets requires application of Computational Fluid Dynamics (CFD) methods for electrically conductive liquids in geometrically complex regions and in the presence of a strong magnetic field. Several general-purpose CFD codes allow modeling of the flow in complex geometric regions, with simultaneous conjugated heat transfer analysis in liquid...
Brijesh Kumar Yadav
(Institute for Plasma Research)
06/09/2016, 14:20
Indian Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) will be tested in one half of the port no#02 of ITER. In LLCB TBM, PbLi eutectic alloy is used as multiplier, breeder, and coolant for the CB zones, and Li2TiO3 ceramic breeder (CB) is used as a tritium breeding material. The LLCB TBM consists of two helium coolant circuits, one for the TBM outer box i.e. the TBM First...
Denis Obukhov
(JSC "NIIEFA" (Efremov Institute))
06/09/2016, 14:20
This paper gives an overview of the new facility for MHD and heat transfer (HT) tests of liquid metal breeder blanket mock-ups in high magnetic field. The facility named LIMITEF5 (LIquid Metal TЕst Facility, 5 T) is under construction now in JSC “NIIEFA” (D.V. Efremov Institute).
The facility includes the Lead-Lithium (LL) loop passing through the warm aperture of the superconducting...
Takuya Goto
(National Institute for Fusion Science)
06/09/2016, 14:20
Lithium molten salts (e.g., Flibe, Flinabe) have several merits as a self-cooled tritium breeding material: low reactivity, low density and low electric conductivity. On the other hand, molten salts may cause a problem of tritium migration to the structural material of the blanket due to the low hydrogen solubility. To overcome this problem, an active control of the effective hydrogen...
Igor Kupriyanov
(Beryllium Department)
06/09/2016, 14:20
The primary reasons for the selection of beryllium as an armour material for the ITER first wall are its low Z and high gettering characteristics. For this application three beryllium grades: S-65C (USA), TGP-56FW (Russia) and CN-G01 (China) have been accepted. This selection was based on the results of the ITER Qualification Program, which included characterization and testing of material...
Petra Jenus
(Department for Nanostructured Materials)
06/09/2016, 14:20
Tungsten-based composites have gained considerable attention owing to their excellent performance levels at high temperatures due to exceptional high temperature properties such as a high melting point, good thermal conductivity and a low thermal expansion coefficient. However, tungsten is also associated with a serious reduction in its strength at elevated temperatures, which is also one of...
Sasa Novak
(Department for Nanostructured Materials)
06/09/2016, 14:20
The main aim of the work has been to improve properties of the plasma-facing material for the divertor to resist high thermal loading during operation. Among the available materials we selected (carbide) particles reinforcement of tungsten, wherein the reinforcement should not chemically react with the matrix. In this respect, W2C particles offer the most attractive solution.
The paper will...
Andrei Galatanu
(National Institute of Materials Physics)
06/09/2016, 14:20
W has the highest melting point of all metals, good high temperature strength, high creep resistance and a high thermal conductivity. These properties make W a first choice for armor materials in fusion energy reactors. Unfortunately W can not be also used for structural applications, due especially to its high temperature brittle- to-ductile transition (DBT). However, when cold rolled at...
Carmen Garcia-Rosales
(Materials Department)
06/09/2016, 14:20
Tungsten is presently the main candidate material for the first wall armour of future fusion reactors. However, if a loss of coolant accident with simultaneous air ingress into the vacuum vessel occurs, the temperature of the in-vessel components would exceed 1000ºC, leading to the undesirable formation of volatile and radioactive tungsten oxides. A way to prevent this serious safety issue is...
Min Pan
(Key Laboratory of Advanced Technology of Materials (Ministry of Education))
06/09/2016, 14:20
Irradiation damage research is one of the basic issues to solve the application of first-wall materials in fusion engineering. The diffusion and recovery of the defects can greatly affect the performance of the materials in fusion. The rotation, stability, migration of the self-interstitial atoms (SIAs) in defect structures of tungsten is investigated by the first-principle method. It is found...
Vladica Nikolic
(Erich Schmid Institute of Materials Science of the Austrian Academy of Sciences)
06/09/2016, 14:20
In order to investigate possible enhancement of mechanical properties of tungsten (W) based materials by solid solutions and to examine the influence of a single alloying element on a particular property such as ductility, a versatile production method of generating a wide range of different tungsten binary alloys is presented. Magnetron sputter co – deposition was used to produce thin films...
Magdalena Galatanu
(National Institute of Materials Physics)
06/09/2016, 14:20
For DEMO fusion reactor an expected heat flux of about 10 MW/m22 should be extracted by the divertor which will have, most likely, an armour part made of W and a following heat sink part made of Cu or ODS Cu alloy. Unfortunately, for these materials the optimum operating temperature windows do not overlap. Thermal barrier materials are interface materials included in such...
Dai Hamaguchi
(Fusion Research and Development Directorate)
06/09/2016, 14:20
Copper is the candidate material for cooling components for divertor and other plasma facing components. Although CuCrZr alloy is a first choice regarding strength, toughness, and conductivities, issues related to quality control during manufacturing process and also on the possible loss of strength during brazing among fabrication of the components still remains. CuCrZr also exhibit some...
Hiroyuki Noto
(National Institute for Fusion Science)
06/09/2016, 14:20
Copper (Cu) alloy is a candidate materials for use as heat sink materials of fusion divertor because of its good thermal conductivity. In recent years a number of studies have been carried out on Cu-based materials such as Precipitation Strengthened Cu (PS-Cu).However, the material has some critical issues such as instability of microstructure at high temperature and loss of strength by...
Inigo Iturriza
(Materials and Manufacturing)
06/09/2016, 14:20
The blanket is one of the most critical component of ITER. It is directly exposed to the plasma and acts as shielding of the vacuum vessel from the neutrons and other energetic particles produced in the fusion plasma. Each of the 215 Normal Heat Flux (NHF) panels consists of a shield block and a First Wall (FW) panel. The NHF FW panels consist of a complex bimetallic structure of 316L...
Teteny Baross
(WIGNER RCP)
06/09/2016, 14:20
The actuality of the topic comes from the ITER (International Thermonuclear Experimental Reactor) fusion tokamak that is a major international experiment with the aim of demonstrating the scientific and technical feasibility of fusion as an energy source. Among others the most challenging task is to find proper materials and technology for Plasma Facing Components.
Welding by HIP (Hot...
Pinghuai Wang
(Southwestern Institute of Physics)
06/09/2016, 14:20
The CuCrZr/316L(N) explosion bonding bimetallic plates were used to make hypervapotron (HVT) cooling channel for the fingers, which is the key components of the ITER First Wall (FW). The bimetallic plates will be subjected to the same thermal cycles as the FW component, including the HIP (hot iso-static pressing) joining for bonding HVT and beryllium tiles, thus the properties of both the...
Javier de Prado
(Materials Science and Engineering Area)
06/09/2016, 14:20
Development of new materials is one of the key for the construction of the new fusion power plant (DEMO). The selected materials have to fulfill several requirements such as standing the conditions that takes place in the core (high neutron flux and temperatures close to 1200 ºC) and low activation rate.
Several techniques have been proposed to join the different parts of the first wall...
Eduard Feldbach
(Institute of Physics)
06/09/2016, 14:20
Radiation tolerant optical components of future fusion reactors have to withstand radiation of unprecedented intensity. It is widely recognized that spinel lattice of AB2O4 double oxides demonstrates enhanced resistance against neutron irradiation. Therefore, the development of spinel optical materials and understanding of their radiation damage processes is of great importance. One defect...
Jiao Peng
(Institute of Plasma Physics)
06/09/2016, 14:20
First mirror (FM) lifetime is one of critical issues for the optical diagnostic system in ITER since it greatly influences the performance of relative diagnostic. In ITER, repetitive cleaning is expected to give a positive solution to the frequent replacement of FM, thus prolonging its lifetime. Three cleaning cycles using radio frequency argon plasma were applied to the stainless steel mirror...
Richard Kembleton
(Culham Centre for Fusion Energy)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
There are a number of key design difficulties in producing an integrated demonstration fusion power plant (DEMO) design, and how these issues are resolved fundamentally affects the final overall design. Technological examples include the issue of power loading in the divertor and reducing recirculating power through efficient current drive. Additional drivers include economic considerations...
Dagui Wang
(Institute of Nuclear Energy Safety)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Breeding blanket research and development is recognized as one of the most important areas for realizing an energy-producing fusion reactor. In China, the ceramic breeder/helium coolant/ferritic steel structure is considered as the main concepts of the blanket to conduct the breeding blanket research, and on the other hand, the liquid breeder blanket is also to be investigated as the...
James Morris
(Power Plant Technology Group)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The investigation of time-dependent power requirements for a future nuclear fusion reactor is part of the systems integration task for the European Fusion Programme. All fusion power plants, whether pulsed or steady-state, will require electrical power to operate the various plant systems. Over the entire pulse cycle reactor systems will require varying levels of power over different time...
Christopher Harrington
(Culham Centre for Fusion Energy (CCFE))
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The Water-cooled Lithium-Lead (WCLL) blanket is one option under consideration for the EUROfusion DEMO programme. This blanket design must interface with the Primary Heat Transfer System, Power Conversion System, and Energy Storage System in an integrated solution to mitigate the pulsed power profile of the tokamak and deliver feasible power plant performance. The system must maintain an...
Monika Lewandowska
(Institute of Physics)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
ITER is planned to be the research type tokamak which will achieve the energy breakeven point. The next step towards the realization of fusion energy will be DEMO – the first demonstration fusion power plant producing grid electricity at the level of a few hundred MW. DEMO designers are required to maximize the conversion efficiency of the primary and secondary plant circuits. The Primary Heat...
Vaclav Dostal
(Energy engineering)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The cooling system is one of the key parts of the fusion power reactor technology. The DEMO fusion power reactor should have different heat sources (first wall, blanket, and divertor) with different temperature and power. In the current European concept of DEMO, helium and water are used as the cooling medium. However, use of Helium and water introduces some issues in terms of their properties...
Xue Zhou Jin
(Institute of Neutron Physics and Reactor Technology (INR))
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
HCPB (helium cooled pebble bed) blanket concept is one of the EU DEMO blanket concepts running for the final design selection. It is necessary to study the pressure behaviour in the blanket and the connected systems during the loss of coolant (LOCA) in a blanket module, as well as the temperature evolution in the coolant flow and the associated structures. The LOCA can be caused by...
Danilo Dongiovanni
(FSN)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Radioactive toxins confinement is a main safety function for nuclear power plants, hence the importance of confinement design parameters optimization. In this context, performing parametric assessments of thermodynamic variables thought to be relevant for confinement design can help at better framing the option design space. In the context of DEMO EUROfusion WP, FFMEA studies are going on for...
Jan Stepanek
(Department of Energy Engineering)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The first wall, blanket and divertor targets provide a physical boundary for the plasma influence and have to be intensively cooled during the operation in case of the high power fusion reactor. In the case of the LOCA accident, the released fusion power can be stopped very quickly, but the final plasma disruption may load the non-cooled components, and a large amount of heat accumulated in...
Dobromir Panayotov
(ITER Department, Fusion for Energy (F4E), Torres Diagonal Litoral B3, Barcelona, E-08019, Spain)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER. Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. The F4E, Amec Foster Wheeler and INL comprehensive...
Danna Zhou
(Institute of Nuclear Energy Safety Technology)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The helium cooled LiPb blanket concept has become a promising design for fusion reactors in the world. Considering the complex design of the blanket, it is likely that helium gas leakage into the liquid alloy may occur due to tube rupture, named in-box Loss of Coolant Accident (in-box LOCA). And corresponding shock waves likely occurred at the break position and transferred within the liquid...
Jiangtao Jia
(Key Laboratory of Neutronics and Radiation Safety)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
With China signing Test Blanket Module Arrangement (TBMA) with ITER Organization for Helium Cooled Ceramic Breeder (HCCB) Test Blanket System (TBS) in February 2014, Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS), becomes one of the leading teams undertaking its corresponding research and development, and is mainly responsible for structure material...
Marco Fabbri
(Fusion Energy Engineering Laboratory)
06/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
For almost ten years now, several safety studies of plasma-wall transients have been performed with AINA code for ITER, the European DEMO design (e.g. HCPB) and Japanese one (e.g. Water Cooled Pebbled Bed or WCPB) to establish an envelope for the worst effects of ex-vessel LOCA and overfuelling. For this purpose, for each blanket type a specific wall-model has been developed for different AINA...
Agnieszka Zaras-Szydłowska
(Institute of Plasma Physics and Laser Microfusion)
06/09/2016, 14:20
A concept and a laboratory model of the laser-driven accelerator of plasma beams for materials research is presented. The accelerator is based on the laser-induced cavity pressure acceleration (LICPA) scheme [1] and includes four parts: (1) the laser driver, (2) the plasma cavity where high-temperature plasma is created by the laser driver and a high plasma pressure is generated, (3) the...
Punit Kumar
(Department of Physics)
06/09/2016, 14:20
Interaction of high power laser fields with plasma is important for many applications including laser fusion, laser wakefield acceleration and x-ray lasers. At high laser intensities, nonlinear interactions between plasma and laser becomes significant. In the last ten years, there has been a great deal of interest on plasma systems where the quantum effects are important. Consideration of...
Koichi Nishiyama
(IFMIF/EVEDA Project Team)
06/09/2016, 14:20
IFMIF (International Fusion Material Irradiation Facility) will generate 14 MeV neutron flux for qualification and characterization of suitable structural materials of plasma exposed equipment of fusion power plants. IFMIF is an indispensable facility in the fusion roadmaps since provide neutrons with the similar characteristics as those generated in the DT fusion reactions of next steps after...
Sunao Maebara
(Rokkasho Fusion Research Institute)
06/09/2016, 14:20
For the IFMIF/EVEDA accelerator prototype RFQ linac, the operation frequency of 175MHz was selected to accelerate a large current of 125mA. The driving RF power of 1.28MW by 8 RF input couplers has to be injected into the RFQ cavity for CW operation mode. For each RF input coupler, nominal RF power of 160kW and maximum transmitted RF power of 200kW are required.
For this purpose, an RF input...
Hugo Policarpo
(IPFN - Instituto de Plasmas e Fusão Nuclear)
06/09/2016, 16:40
The ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations also known as gaps 3, 4, 5, and 6, complementing the information provided by the magnetic diagnostics. For gaps 4 and 6, the antennas are to be installed in-vessel between two blanket shield modules. The...
Stephen Reynolds
(Power and Active Operations)
06/09/2016, 16:40
Radioactive waste arisings from JET operations are projected to contain approximately 25t of non-incinerable Intermediate Level Waste (ILW) with tritium levels > 12 kBq/g. This originates primarily from plasma facing components, specifically the divertor (MKIIa) used during the JET Deuterium Tritium Experiment in 1997 (DTE1). As current UK regulations do not allow off-site disposal of ILW and...
Hans Meister
(ITER Technology & Diagnostics)
06/09/2016, 17:00
The ITER bolometer diagnostic shall provide the measurement of the total radiation emitted from the plasma, a part of the overall energy balance. About 500 lines-of-sight (LOS) will be installed in ITER observing the whole plasma from many different angles to enable reliable measurements and tomographic reconstructions of the spatially resolved radiation profile. The LOS are bundled in up to...
Frederik Arbeiter
(Institute for Neutron Physics and Reactor Technology)
06/09/2016, 17:00
Helium is considered as coolant in the plasma facing first wall of several blanket concepts for DEMO fusion reactors, due to the favorable properties of chemical inertness, no activation, comparatively low effort to remove tritium, no chemical corrosion and a flexible temperature range. Design analyses for the ITER Test Blanket Modules done by several design teams have shown ability to use...
Damao Yao
(Institute of plasma physics)
06/09/2016, 17:00
Befroe join ITER project fusion technologies development in China are focus on fusion device and plasma operation related. Components on fusion device installed, removed and maintained by personnel. Robotic technologies are never applied for fusion.
China joined ITER from 2004. Scientists and engineers are involved in ITER related study and technologies development. Remote handling systems are...
Carlota Soto
(Departament of Materials)
06/09/2016, 17:20
Flow Channel Inserts (FCI) are key elements in a Dual Coolant Lead Lithium blanket concept for DEMO, since they provide the required thermal and electrical insulation between the He cooled structural steel and the hot liquid Pb-15.7Li flowing at around 700°C, and minimize MHD pressure loss. FCIs must be inert in contact with Pb-15.7Li and show low tritium permeability. In addition, FCIs have...
Charles Henager
(Pacific Northwest National Laboratory, Richland, WA, United States)
06/09/2016, 17:40
Iron-base alloys are the leading candidate structural material for first-wall and blanket applications in near-term fusion devices, but their long-term viability to reliably function in the harsh fusion nuclear environment remains to be established. Helium produced by transmutation reactions interacts with microstructural features such as pre-existing dislocations, martensite lath boundaries,...
Rodrigo Ventura
(Institute for Systems and Robotics)
06/09/2016, 17:40
Nuclear power plants require periodically maintenance, including the remote handling operations of transportation performed by automated guided vehicles (AGV). The navigation system becomes a key issue given the safety constrains of the heavy load to be transported in the complex scenarios, such as the reactor building.
This work presents well-known and mature navigation technologies used by...
N. Mitchell
(on behalf of the ITER Central Team)
07/09/2016, 08:30
Oral
The magnet system is one of the critical core components of the ITER magnets, defining the machine capabilities to form and drive 15MA 500MW nuclear plasmas for 100s of seconds. The magnets, the largest superconducting magnet system ever built with 50GJ of stored energy, are also technologically highly advanced components using large composite Nb3Sn 4-6K force flow cooled conductors that also,...
R. Heidinger
(Fusion for Energy)
07/09/2016, 09:10
Oral
Fusion road maps defined by both Europe and Japan, Parties to the Broader Approach Agreement (BA) where the IFMIF/EVEDA project is underway, have yet again confirmed the central need of a neutron source dedicated for fusion materials qualification. In the framework of the BA, engineering design and engineering validation activities are conducted which are targeted to prepare the foundations...
U. Fischer
(Karlsruhe Institute of Technology)
07/09/2016, 09:50
Oral
The European Power Plant Physics and Technology (PPPT) programme, organised within the EUROfusion Consortium, aims at developing a conceptual design of a fusion power demonstration plant (DEMO) as a central element of the roadmap to the realisation of fusion energy.
Various integrated PPPT projects are being conducted to meet this goal including Breeder Blanket (BB), Safety and Environment...
Anna Wojcik-Gargula
(Department of Radiation Transport Physics)
07/09/2016, 11:00
Study of materials dedicated to fusion reactors is one of the most challenging tasks faced by fusion research. Unfortunately, the number of useful fast neutron sources with a proper neutron spectrum and high neutron fluence is limited. Currently, a better exploitation of the existing neutron sources, such as high flux fission research reactors or material test reactors, is necessary to develop...
Sudhirsinh Vala
(Neutron Source Up-gradation Division)
07/09/2016, 11:00
In order to study the neutronics of fusion reactor blankets, a program is underway at the IPR using 14-MeV neutron source. An accelerator based neutron generator is under development in which 30 mA deuterium beam will be accelerated up to 300 keV energy. It will then impinge on a rotating tritium target to producing nearly isotropic 14-MeV neutrons. The expected neutron yield is 3-5 x...
Fernando Arranz
(Laboratorio Nacional de Fusion)
07/09/2016, 11:00
The LIPAc (Linear IFMIF Prototype Accelerator) is a prototype that ends in a Dump made of copper with conical shape and cooled by water moving at high speed on the outer surface.
The shape of the dump is intended for a redistribution of a very high density power of the deuteron beam to be stopped (1.12 MW) leading during normal operation to reasonable temperatures and thermal stresses well...
Gioacchino Micciche
(FSN-ING-PAN)
07/09/2016, 11:00
The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where fusion reactor candidate materials will be tested. The neutron flux is produced by means of a deuteron beam (250 mA, 40 MeV) that strikes a target of liquid lithium circulating in a loop. The support on which the liquid lithium flows is the most heavily exposed component to the...
Wojciech Krolas
(Institute of Nuclear Physics PAN)
07/09/2016, 11:00
IFMIF-DONES - a powerful neutron irradiation facility for studies and certification of materials - is planned as part of the European roadmap to fusion electricity. Its main goal will be to study properties of materials under severe irradiation in a neutron field similar to the one in a fusion reactor first wall. It is a key facility to prepare for the construction of the DEMO Power Plant...
Gaetano Bongiovi
(Department of Energy)
07/09/2016, 11:00
The availability of a high flux neutron source for testing candidate materials under irradiation conditions which will be typically encountered in future fusion power reactors is a fundamental step towards the development of fusion energy. To this purpose, IFMIF (International Fusion Materials Irradiation Facility) represents the reference option to provide the fusion community with a source...
Zhiqiang Zhu
(Institute of Nuclear Energy Safety Technology (INEST))
07/09/2016, 11:00
Because of the depletion and limitation of natural energy sources, fusion energy is the promising and irreplaceable way for energy development in the future. As the only energy conversion unit in the fusion reactor, PbLi blanket is considered as one of the important blankets for DEMO and fusion reactors, Lead Lithium (PbLi) is designed as tritium breeder, neutron multiplier and coolant. Before...
Tomas Romsy
(Faculty of Mechanical Engineering)
07/09/2016, 11:00
The liquid metal eutectic Pb-Li17 is considered as one of the possible coolants for the blanket of the fusion reactor DEMO. The main reason for usage of the eutectic Pb-Li17 is the Tritium breeding. The eutectic flow separates alloys of the structural steels and thus be the cause of them corrosion.The cold trap is a device for corrosion products removing from liquid metal.
The cold trap was...
Bernhard Ploeckl
(Max Planck Institute for Plasma Physics)
07/09/2016, 11:00
The Demonstration Fusion Power Reactor (DEMO) is supposed to be the step in between ITER and the first commercial fusion power plant. In the framework of one mission of the “Work plan for the roadmap to fusion energy 2014-2018” a work package Tritium, Fuelling and Vacuum (TFV) was launched. As part of this project, the examination of requirements for the matter injection system is ongoing...
Mikhail Gryaznevich
(Tokamak Energy Ltd)
07/09/2016, 11:00
Recent advances in the development of high temperature superconductors (HTS) [1], and encouraging results on a strong favourable dependence of electron transport on higher toroidal field (TF) in Spherical Tokamaks (ST) [2], open new prospects for a high field ST as a compact fusion reactor or a powerful neutron source [3]. The combination of the high beta (ratio of the plasma pressure to...
Carlos Otarola
(Electromechanical Engineering)
07/09/2016, 11:00
The manufacturing methods and issues found during the construction of the Stellarator of Costa Rica 1 (SCR-1) will be discussed. The SCR-1 is a small modular stellarator developed by the Instituto Tecnológico de Costa Rica (ITCR). Currently, it’s being tested for the first plasma discharge.
SCR-1 is a 2-field period small modular stellarator (Ro=0.238 m, =0.054 m, Ro/a>4.4, plasma volume...
Subrata Pradhan
(Institute for Plasma Research)
07/09/2016, 11:00
Steady State Superconducting Tokamak (SST-1) at Institute for Plasma Research is a `working’ experimental superconducting device since late 2013. SST-1has been upgraded with Plasma Facing Components and is getting prepared towards long pulse operations in both circular and elongated configurations. Initial experiments have begun in SST-1 with circular plasma configurations. SST-1 offers a...
Dennis Ronden
(Fusion physics - Remote Handling)
07/09/2016, 11:00
This paper presents the results of a study that was performed on conceptual solutions for assembly and handling of EC components inside the EC upper and equatorial port cells. Particular topics that are discussed include the access to the waveguides and auxiliary feedthroughs of the launchers at the port plug closure plate, (dis-)assembly & alignment of the ex-vessel waveguide in the port...
Avelino Mas Sanchez
(Ecole Polytechnique Fédérale de Lausanne)
07/09/2016, 11:00
The Electron Cyclotron Upper Launcher (ECUL) is an eight beamline ITER antenna aimed to drive current locally inside the islands that may form on the q= 3/2 or 2 rational magnetic flux surfaces in order to stabilize neoclassical tearing modes (NTMs). The primary vacuum boundary at the port plug extends into the port cell region through the ex-vessel mm-wave waveguide components, defining the...
Phillip Santos Silva
(Swiss Plasma Center)
07/09/2016, 11:00
The ITER Electron Cyclotron Heating Upper Launcher (ECHUL) will be used to drive current locally inside magnetic islands located at the q=2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs). Each antenna consists of eight beam lines that are designed for the transmission of up to 1.5 MW of mm-wave power at 170 GHz. The First Confinement System (FCS) is...
Robert Bertizzolo
(EPFL-SPC (Swiss Plasma Center))
07/09/2016, 11:00
The ITER Electron Cyclotron Heating Upper Launcher (ECHUL) will be used to drive current locally inside magnetic islands located at the q=2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs). Each antenna consists of eight beam lines that are designed for the transmission of up to 1.5 MW of mm-wave power at 170 GHz. The First Confinement System (FCS) is...
Koji Takahashi
(Department of ITER Project)
07/09/2016, 11:00
The new mirror angle detector for ITER EC launchers, applying a rotary capacitor , a RF feeder, RF circuits and several hundreds MHz RF has been developed. The rotary electrode is attached to the rotation axis of the mirror and the stationary electrode is connected to a RF feeder. The reflected RF wave at the rotary capacitor comes back to the feeder and phase of the reflected RF wave changes...
Yasuhisa Oda
(Japan Atomic Energy Agency)
07/09/2016, 11:00
The Electron Cyclotron Heating and Current Drive system developed for ITER is made of 12 sets of High Voltage Power Supplies, 24 Gyrotrons, 24 Transmission Lines and 5 Launchers, 4 UL located in upper ports and 1 EL at the equatorial level. The ITER operation requires to switch operating launcher during the plasma operation with short interval, namely mid-pulse switch operation. To change the...
Michael Bader
(Ampegon AG)
07/09/2016, 11:00
The power supply for the EC Heating system (ECPS) of ITER provides the electrical power to the 170GHz/1MW Gyrotrons. The required electrical power for the gyrotrons is not only very high but has to comply also with highest quality requirements.
This paper gives an overview of the Ampegon ECPS system procured by F4E. It describes the technical requirements of the EC Power Supply system ECPS and...
Peter Spaeh
(Institute for Applied Materials)
07/09/2016, 11:00
ITER will be equipped with four EC (Electron Cyclotron) upper launchers of 8 MW microwave power each with the aim to counteract plasma instabilities during operation. The structural system of these launcher antennas will be installed into four upper ports of the ITER vacuum vessel.
During operation the port plug structure will be heated by nuclear heating from neutrons and photons and thermal...
Matthieu Toussaint
(Swiss Plasma Center)
07/09/2016, 11:00
The Tokamak à Configuration Variable (TCV) has been recently equipped with a 1 MW neutral beam heating (NBH) injector11. Two new stainless steel ports with rectangular aperture of 170x220mm have been manufactured and installed for this purpose. The NBH injector is connected to one of them via a stainless steel port extension. The port and its extension together form the beam duct...
Ugo Siravo
(Ecole Polytechnique Fédérale de Lausanne (EPFL))
07/09/2016, 11:00
Three RHVPSs (Regulated High Voltage Power Supplies, 84kV/80A/2s) are installed and operated at the Swiss Plasma Center for almost twenty years. Each RHVPS supplies a cluster of three gyrotrons. Two clusters are composed of diode type gyrotrons operating at the second harmonic of the TCV electron-cyclotron frequency (X2, 84GHz), whereas the third is a cluster of triode type gyrotrons operating...
Alexander N. Karpushov
(Swiss Plasma Center (SPC))
07/09/2016, 11:00
The TCV tokamak contributes to physics understanding in fusion reactor research based with a wide experimental tool set: flexible shaping and high power electron cyclotron heating. Plasma regimes with high plasma pressure, a wide range of temperature ratios and significant populations of fast ions are now attainable by a TCV heating system upgrade. In the first stage of the TCV upgrade...
Kenji Saito
(Department of Helical Plasma Research)
07/09/2016, 11:00
The transmission line is one of the most important parts among the ion cyclotron range of frequencies (ICRF) heating devices. In the case of unwanted troubles on the line, immediate power-off is necessary for the protection of the line and for safety. In the Large Helical Device (LHD), though the causes were unclear, several troubles such as melting sometimes occurred on the line between the...
Haifeng Liu
(Institute of Fusion Science)
07/09/2016, 11:00
The heating of ions by an obliquely propagating shear Alfvén wave at frequencies a fraction of the particle cyclotron frequency is demonstrated analytically. Under consideration of the small wave amplitude, the resonance conditions in the laboratory frame are systematically derived by multi-scale expansion method. It is found that 1) the cyclotron resonance condition may occur at any wave...
Helmut Faugel
(Max Planck Institute for Plasma Physics)
07/09/2016, 11:00
The efficiency of heating and current drive systems is the key for a successful operation of fusion demonstration power plants like DEMO. In an earlier review article, overall efficiencies of H & CD systems were estimated at 20 – 30 % [1].
In this paper we present a breakdown of the overall efficiency for ICRF (ion cyclotron range of frequencies): 1) the technical efficiencies; 2) the...
Fabrice Louche
(Plasma Physics Laboratory)
07/09/2016, 11:00
Ion cyclotron wall conditioning (ICWC) is being developed for ITER as a baseline conditioning technique in which the ion cyclotron heating and current drive system will be employed to produce and sustain the current-less conditioning plasma. The TOMAS project (TOroidal MAgnetized System, operated at the FZ-Juelich, Germany) proposes to explore several key aspects of ICWC. This project stands...
Chun Kung
(Plasma Physics Laboratory)
07/09/2016, 11:00
Experimental results have shown that twelve-strap HHFW operating at 30 MHz can provide significant plasma heating for NSTX. In this case, it is important to understand the interactions between return currents on the antenna enclosure sidewalls/septa and the launched k|| spectra. CST Microwave Studio is applied to this problem with the view toward optimizing the antenna coupling to the desired...
Anett Spring
(W7-X Operation)
07/09/2016, 11:00
The W7-X steady state control and data acquisition system has been successfully commissioned and well established to investigate plasma break down and run the first more complex physics programs during the initial operation phase of W7-X. Already in the first weeks of plasma operation, experiment programs with up to 10 minutes containing a series of up to 20 plasma discharges have been run...
Heike Laqua
(Wendelstein 7-X Operations (OP))
07/09/2016, 11:00
Wendelstein 7-X (W7-X) is a superconducting stellarator undergoing the first experimental campaign after its commissioning. It’s characteristic feature is the steady state operation of the magnetic field. After an upgrade to cope with permanent heat loads of several Megawatts, W7-X will be able to run steady state discharges, too. This requires a control system that differs from the commonly...
Reinhard Vilbrandt
(Max-Planck-Institute for Plasma Physics)
07/09/2016, 11:00
The commissioning and final validation of the central safety system and the acceptance by the authority were very important steps immediately before the successful ignition of the first plasma in Wendelstein 7-X in December 2016.
Safety is the mandatory prerequisite for the operation of experimental devices of course to protect the personnel and the investment from hazardous situations. To...
Hexiang Wang
(Mechanical Engineering & Mechanics)
07/09/2016, 11:00
Ongoing work in the fusion community focuses on developing advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamic (MHD) stability, and noninductively driven plasma current. The toroidal current density profile, or alternatively the q profile, together with the normalized beta, are often used to characterize these advanced scenarios. The development of these...
Andres Pajares
(Mechanical Engineering & Mechanics)
07/09/2016, 11:00
Control of the plasma density and temperature to produce a certain amount of fusion power, known as burn control, is one of the key issues that need to be solved for the success of tokamak fusion reactors such as ITER. In order to reach a high fusion power to auxiliary power ratio, tokamaks must operate near temperature and density stability limits. Therefore, active control to maintain a...
Eugenio Schuster
(Mechanical Engineering & Mechanics)
07/09/2016, 11:00
Research on fusion plasmas in tokamaks has led to the insight that the poloidal magnetic-flux distribution within the plasma has a crucial impact on its performance. Achieving certain types of poloidal magnetic-flux profiles, or alternatively certain types of q profiles, leads to resilience against undesirable instabilities and to higher bootstrap-current fractions, which in turns favor...
Zeki Ilhan
(Mechanical Engineering & Mechanics)
07/09/2016, 11:00
Active control of the toroidal current density profile is among those plasma control milestones that the National Spherical Tokamak eXperiment - Upgrade (NSTX-U) program must achieve to realize its next-step operational goals characterized by the high-performance, MHD-stable plasma operation with neutral beam heating, and longer pulse durations. Motivated by the coupled, nonlinear,...
Carlo Neri
(ENEA CR Frascati)
07/09/2016, 11:00
The plasma pulse phase of Frascati Tokamak Upgrade (FTU) is driven by the dedicated system FSC (Fast Sequence Control), which has been developed in order to send all the necessary commands to the different power plants feeding the toroidal and poloidal coils during the plasma discharge, meanwhile controlling the correct outcome. In case of incorrect execution of the sequence the system is able...
Andrzej Broslawski
(Narodowe Centrum Badan Jadrowych)
07/09/2016, 11:00
The products of fusion reactions at JET are measured using different diagnostic techniques. One of the methods is based on measurements of gamma-rays, originating from reactions between fast ions and plasma impurities. During the forthcoming deuterium-tritium (DT) campaign a particular attention will be paid to 4.44 MeV gamma-rays emitted in the 99Be(α,nγ)1212C reaction....
Marian Curuia
(Institute of Atomic Physics)
07/09/2016, 11:00
The JET tangential gamma-ray spectrometer (KM6T) is undergoing an extensive upgrade in order to make it compatible with the forthcoming deuterium-tritium (DT) experiments.
The paper will present the design of the main components for the upgrade of the spectrometer beam-line: tandem collimators, gamma-ray shields, and neutron attenuators.
The existing KM6T tandem collimators will be upgraded...
Roch Kwiatkowski
(National Centre for Nuclear Research)
07/09/2016, 11:00
The diagnostic of fast ions at JET is based on the measurements of gamma-rays which are produced as a result of nuclear reactions between ions and plasma impurities. The gamma-ray spectra provide information on energetic tail of ion energy distribution.
The existent BGO detector, with a decay time of ~300 ns, is sufficient during DD campaigns. The strong neutron and gamma-ray fluxes during D-T...
Sorin Soare
(ICIT Rm. Valcea)
07/09/2016, 11:00
A new diagnostics technique, the Lost Alpha Monitor (LAM), for the investigation of escaping alpha particles in JET has been proposed [1]. The method is based on the detection of the gamma radiation induced by the escaping particles on a target external to the plasma. For a beryllium target this reaction is 99Be(a, nγ)1212C. The implementation on JET of the LAM technique...
Marek Rubel
(Fusion Plasma Physics)
07/09/2016, 11:00
All optical spectroscopy and imaging diagnostics in next-step fusion devices will be based on metallic mirrors. The performance of mirrors is studied in present-day tokamaks and under laboratory conditions. This work deals with comprehensive tests of mirrors: (i) exposed in JET with the ITER-Like Wall (JET-ILW); (b) irradiation by He and heavy ions to simulate the impact of neutrons under...
Jean-Marie Noterdaeme
(Applied Physics Department)
07/09/2016, 11:00
High performance H-mode plasmas are characterized by short, repetitive edge perturbations known as edge-localized modes (ELMs). Large, unmitigated ELMs can result in significant transient heat loads released onto the plasma-facing components. Hence, characterization of ELMs and their control are crucial for avoiding a significant reduction in the divertor lifetime. This necessitates...
Janne Lyytinen
(Smart Industry and Energy Systems, VTT Technical Research Centre of Finland Ltd, Tampere, Finland)
07/09/2016, 11:00
ITER fusion reactor is a very complex machine which has several different subsystems. It is still a research reactor and the testing results will be implemented in the next generation reactors. In the testing phase of the reactor there will be several sensors and instruments assembled inside the vessel for diagnostics purposes. One of the key diagnostics areas will be the divertor...
Miklos Palankai
(Plasma Physics Department)
07/09/2016, 11:00
Electrical Services provide the electrical infrastructure to serve the diagnostics installed on the ITER Tokamak. The components of the Diagnostics are located all over on the inner and outer shell of the vacuum vessel, in the ports, on the Divertor Cassettes and in the Cryostat as well. Sensors require electrical transmission lines to transmit both of the diagnostic and control signals across...
Christian Vorpahl
(Port Plugs & Diagnostics Integration Division)
07/09/2016, 11:00
Numerous plasma-near mirrors of optical diagnostics of ITER require protection from erosion and deposition caused by impinging energetic particles. This is achieved by approximately 60 individual Diagnostic Shutters, rather simple mechanical devices which obstruct the mirror’s sight towards the plasma when the diagnostic is not in use. If a shutter fails to operate, so does the respective...
Vladislav Kotov
(Institut für Energie- und Klimaforschung – Plasmaphysik)
07/09/2016, 11:00
First mirrors are plasma facing components which redirect light to the protected optical diagnostics. Initial investigations [A. Litnovsky et al. Nuclear Fusion 49 (2009) 075015, V. Kotov et al. Fusion Eng. Des. 89 (2011) 1583] showed that deposition of impurities (Be, Fe etc.) may cause drastic degradation of the mirror reflectivity and thus severely restrict the diagnostic performance. Very...
Laura Garcia-Ruesgas
(Department of Engineering Graphics)
07/09/2016, 11:00
During the final design review of Diagnostic Port Plugs, it has been highlighted that the current system of fixation, based on gaps, while it is not harmful for the port plug, it throws large uncertainties over the alignment of the optical systems placed inside the DSMs at the same time that the real mechanical behaviour of the assembly is clearly unknown. Due to the fact that the DSM is not...
Jean-Marc Drevon
(Bertin Systèmes Instrumentation)
07/09/2016, 11:00
ITER will have a set of 45 diagnostics to ensure controlled operation. Many of them are integrated in the ITER ports. Housed in generic structures, this modular integration is designed to help diagnostics withstanding the plasma loads whilst complying with the French regulations. Now that the Domestic Agencies and ITER Organization are developing the preliminary or even final designs of the...
Antonio Carpeno
(Telematics and Electronics Department)
07/09/2016, 11:00
The iRIO-3DLab platform has been devised to enhance the learning process and reduce the development time for engineers in charge of designing intelligent DAQ systems based on PXIe technology and distributed control systems such as EPICS. iRIO-3DLab consists of an Opensim-based virtual world that aims to promote the understanding of how such a kind of DAQ system works, and how the EPICS IOC...
Hiteshkumar Dhola
(Power Supply Group)
07/09/2016, 11:00
A Dual output (27kV & 15kV), 3MW High Voltage Power Supply (ICHVPS) has been installed and integrated with a Diacrode based RF source to be used for ICRF system. The ICHVPS Controller is based on LabVIEW Real-time PXI controller, which supports all control and monitoring operations of the PSM based power supply. The controller supports all essential features like, fast dynamics, low ripple and...
Bruno Santos
(Instituto de Plasmas e Fusão Nuclear)
07/09/2016, 11:00
The Advanced Telecommunications Computing Architecture (ATCA) standard defines a high performance technical solution that meets the requirements for fast controllers on large-scale physics experiments like ITER. This platform provides high throughput, scalability and features for high availability such as redundancy and intelligent platform management which are essential for steady state...
Antonio Rodrigues
(Instituto de Plasmas e Fusão Nuclear)
07/09/2016, 11:00
Control and Data Acquisition (CDAQ) systems applied to large physics experiments like ITER, are designed, among other features, for High-Availability (HA). A CDAQ system based on the PCI Industrial Computer Manufacturers Group (PICMG) 3.x AdvancedTCA Base Specification and Intelligent Platform Management Interface (IPMI) standards grants these features. One of the key functions of the HA is...
Paulo Carvalho
(IPFN/IST)
07/09/2016, 11:00
The Advanced Telecommunications Computing Architecture (ATCA) specification implements important key features such as high reliability, high availability, redundancy and serviceability for control and data acquisition instrumentation fault condition, hardware malfunction, firmware updates and hardware reconfiguration.
Red Hat Enterprise Linux and corresponding kernels already have built-in...
Rita C. Pereira
(Instituto Superior Técnio)
07/09/2016, 11:00
The Radial Neutron Camera (RNC) and the Radial Gamma-Ray Spectrometer (RGRS) are two ITER diagnostics, devoted, respectively, to the real-time measurement of the neutron emissivity profile (to be used for plasma control purposes) and to the measurement of the confined alpha profile and runaway electrons. The two systems are closely related as they share the same equatorial port plug and part...
Jeremie Dubray
(Ecole Polytechnique Fédérale de Lausanne)
07/09/2016, 11:00
The Swiss Plasma Center (SPC) is involved in the development and the operation of gyrotrons for fusion application (TCV tokamak, W7-X, ITER) and for medical application as well (spectroscopy DNP/NMR). In this framework, embedded control systems based on National Instrument (NI) compact Reconfigurable Input Output (cRIO) and compact Data AcQuisition (cDAQ) offer versatile solutions for...
Karishma Qureshi
(Institute for Plasma Research)
07/09/2016, 11:00
Cryogenic Instrumentation is a unique and vast field and requires an in-depth understanding of the process and instrumentation. 26 channels Data Acquisition System is required for the 6 nos. of Cryogenics Pumps LN2 cool down experiment. The data acquisition system measures 22 nos. of temperature signals, 2 nos. of level signals of the buffers and 2 nos. of Nitrogen Dewar Signals (Pressure and...
Adriano Francesco Luchetta
(Consorzio RFX)
07/09/2016, 11:00
The Control and Data Acquisition System (CODAS) of SPIDER, the first experiment of the Neutral Beam Test Facility, is under installation and undergoing the commissioning and first operation phases.
The system hardware is nearly compliant with the ITER CODAC catalog for slow and fast plant systems. The system software is based on a combination of software frameworks that altogether collaborate...
Eduardo Rodriguez
(Department of Construction and Manufacturing Engineering)
07/09/2016, 11:00
This paper is focused on the computation of EM loads induced by plasma current disruptions on the Diagnostics positioned inside the Equatorial Port Plugs, and more explicitly, on the creation of a detailed set of tools (Finite Element ‘FE’ models and routines) which allow the automatic characterization of the EM phenomena (DINA) as well as they provide versatility for the adding/removing of...
Takeo Nishitani
(National Institute for Fusion Science)
07/09/2016, 11:00
The Large Helical Device (LHD) plans to start the deuterium experiment in March of 2017, where a maximum neutron yield of 2.1x101616 neutrons/3 sec is expected. For the deuterium experiment, neutron flux monitors, a neutron profile monitor, a neutron activation system and other neutron detectors have been prepared. The characteristics of those neutron diagnostics, such as the...
Gabor Veres
(Department of Plasma Physics)
07/09/2016, 11:00
Devices that are capable of measuring the total plasma radiation in fusion reactor experiments are indispensable for safe and reliable plasma operation. One of the most widespread type of these kind of devices are metal absorber–metal resistor bolometers where the radiation is absorbed by a metallic layer and the change of the layer’s temperature is measured by metal resistors. Based on the...
Rafał Krawczyk
(Institute of Electronic Systems)
07/09/2016, 11:00
The development of GEM detector based acquisition systems resulted in the increase of throughput and resolution in the new revision of the system. The FPGA-based electronics is used to acquire, diagnose and to preliminarily analyze the data of soft X-ray emitted by hot plasma in Tokamak. Moreover, the development of electronics allowed to implement algorithms, so far performed offline after...
B. Bieg
(Institute of Physics)
07/09/2016, 11:00
On the basis of the angle variables technique (AVT) changes of polarimetry state of electromagnetic wave passing through the thermonuclear plasma in the poloidal plane have been analyzed. The first section analyzes the changes in polarization state depending on the angle at which the test beam was sent, for the same plasma parameters.
Subsequently, for a given geometry, using numerical...
Jose Martinez-Fernandez
(Laboratorio Nacional de Fusión (LNF))
07/09/2016, 11:00
This work describes the preliminary assessment of the different waveguide technologies for the ex-vessel transmission lines of the Plasma Position Reflectometer (PPR) in ITER.
Initially, both oversized rectangular and circular corrugated waveguides were considered for the study; the former due to reduced costs and ease of procurement and the latter due to better performance in terms of...
Chuan Li
(State Key Laboratory of Advanced Electromagnetic Engineering and Technology)
07/09/2016, 11:00
This paper mainly introduces the seismic analysis of the high-power dc reactor prototype, whose functions are to limit the ripple current and the increasing rate of fault current in the ITER poloidal field (PF) converter. The stacked reactors with the assembly dimension (L×W×H) of 2955 mm×1639 mm×3296 mm and weight about 5 tons are fixed to the steel base by five support components. In order...
Hideki Kajitani
(ITER department)
07/09/2016, 11:00
Japan Atomic Energy Agency (JAEA), as Japan Domestic Agency, has responsibility to procure 9 ITER Toroidal Field (TF) coils. JAEA completed proto double-pancake (DP) trials aiming at qualification and optimization of manufacturing procedure of TF coil in 2015. Series production of DPs then started and winding of 14 DPs, heat treatment of 11 DPs, fabrication of 9 radial plates (RP), transfer of...
Roberto Bonifetto
(Energy Department)
07/09/2016, 11:00
The ITER Central Solenoid Model Coil (CSMC) is a superconducting solenoid operated at the JAEA centre of Naka, Japan, since 2000 to test the performance of insert coils in its bore, where it produces a magnetic field of 13 T representative of the ITER CS operating conditions.
In 2015, the ITER Central Solenoid Insert (CSI), whose Nb3Sn cable-in-conduit conductor (CICC) will be adopted for the...
Kurt Schaubel
(ITER CS Project)
07/09/2016, 11:00
General Atomics (GA) is currently manufacturing the ITER Central Solenoid Modules (CSM) under contract to US ITER at Oak Ridge National Laboratory, under the sponsorship of the Department of Energy Office of Science. The contract includes the design and qualification of manufacturing processes and tooling necessary to fabricate seven CSM (6 + 1 spare) that constitute the ITER Central Solenoid....
Alberto Ferro
(Consorzio RFX)
07/09/2016, 11:00
The Residual Ion Dump Power Supply (RIDPS) is part of the Ground Related Power Supplies, to be manufactured by OCEM Energy Technology s.r.l. (OCEM) for the MITICA experiment and for the two ITER Heating Neutral Beam Injectors (HNBI). MITICA is the full-scale prototype of the HNBI, under construction in the PRIMA Neutral Beam Test Facility in Padua, Italy.
The RIDPS is devoted to feed the...
Vanni Toigo
(Consorzio RFX)
07/09/2016, 11:00
The Neutral Beam Injector (NBI) is required to inject in ITER plasma Deuteron particles which, once generated in the Ion Source (IS) polarized at -1MV, are accelerated at ground potential and then neutralized. This voltage level is very demanding for the power supply system, requiring several non-standard components. This paper describes the design status of two main NBI components: High...
Francesca Cau
(Fusion for Energy)
07/09/2016, 11:00
The winding pack of the ITER Toroidal Field (TF) coils is composed of 134 turns of Nb3Sn Cable in Conduit Conductor (CICCs) wound in 7 double pancakes and cooled by supercritical helium (He) at cryogenic temperature. The cooling of the Stainless Steel (SS) case supporting the winding pack is guaranteed by He circulation in 74 parallel channels. A 2D approach to compute the temperature...
Rustam Enikeev
(Efremov Institute)
07/09/2016, 11:00
The superconductive coils of ITER magnet system will be energized by ac/dc converters. Before each plasma pulse the magnet system will be pre-charged with energy (8GJ) to be used for generating the toroidal loop voltage required for the gas mixture breakdown and plasma formation. This will be realized by inserting energy dissipating resistors in series with the central solenoid (CS) modules...
Maksim Manzuk
(Joint Stock Company "D.V. Efremov Institute of Electrophysical Apparatus")
07/09/2016, 11:00
High current DC switches play a very important role in the ITER coil power supply system (CPSS) being key components of its two major parts: switching network units (SNU) for plasma initiation and fast discharge units (FDU) for superconducting coils energy extraction in case of quench. For both functions, circuit-breakers rated up to 70 kA steady-state current and 10 kV voltage are required...
50027.
P3.088 On optimization of air cooling system of FDR dissipating energy from ITER magnet coils
Victor Tanchuk
(JSC "NIIEFA")
07/09/2016, 11:00
The Fast Discharge Resistors (FDR) under development at NIIEFA are intended together with switching equipment to dissipate energy released in case of quench of the ITER superconducting coils, thereby protecting them against failure.
FDRs are made of sections consisting of resistive elements enclosed in boxes. Two-four sections stacked vertically form a separate module.
During energy release...
Neway Atnafu
(Engineering)
07/09/2016, 11:00
NSTX-U COILS BUS BARS DESIGN AND CONSTRUCTION**
Neway D. Atnafu, L. Dudek, A. Khodak, S. Gerhardt, S. Ramakrishnan, M. Smith, P. Titus
Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451
natnafu@pppl.gov
The construction of the NSTX upgrade project was completed in the fall of 2015. The multi-year capital project was budgeted at $94 Million. The reactor will used to run...
Katsunao Uenishi
(Osaka University)
07/09/2016, 11:00
Sputtering properties of tungsten (W) should be evaluated correctly for lifetime estimation of divertor components. Especially, at elevated temperatures, recrystallization would cause grain structure reconstruction, which would influence sputtering properties and surface morphology changes. However, the detailed studies haven’t been performed.
Actually, the temperature of divertor could...
Takeru Maeji
(Osaka University)
07/09/2016, 11:00
Currently, In regard to the plasma facing material, Tungsten (W) is a major candidate at ITER. A recent study has been reported indicating that the transient thermal load such as ELM or disruption causes metal surface melting or evaporation of W. However, the property and behavior of the W above the melting point has not yet been sufficiently known, and many of the previous studies are...
Daisuke Inoue
(Osaka University)
07/09/2016, 11:00
Tungsten (W) is a primary candidate of plasma-facing materials for fusion reactors. But erosion due to melting and evaporation of W caused by transient heat loads are concerned. A pulsed laser simulating the transient heat loads was irradiated to three tungsten materials and the behavior of the molten layer was investigated. In addition, aluminum (Al) and tin (Sn) was deposited on W and the...
Vladimir Khripunov
(Fusion Reactor Department)
07/09/2016, 11:00
Primary radiation damage (atomic displacements) and Helium and Hydrogen production rates in plasma facing components (PFCs) of a fusion system are usually determined by the high energy parts of neutron spectra formed in plasma chamber from the initial fusion neutron source. According to presented estimates, the energetic alphas and protons, appearing in PFC materials in the (n,a) and (n,p)...
Dmitry Terentyev
(SCK-CEN)
07/09/2016, 11:00
Recent theoretical and subsequent experimental studies suggest that the uptake and release of deuterium (D) in tungsten (W) under high flux plasma exposure (i.e. under ITER-relevant conditions) is controlled by dislocation microstructure induced by the plasma itself. A comprehensive mechanism for the nucleation and growth of D bubbles on dislocation network under high flux low-energy plasma...
Bong Guen Hong
(Chonbuk National University)
07/09/2016, 11:00
We investigate the ablation characteristics of plasma facing materials (PFM) using thermal plasma facilities. A high enthalpy, 400 kW plasma testing facility which uses an enhanced segmented arc torch as a plasma source and 55 kW vacuum plasma spraying system produce particle flux greater than 102424/(m22sec) and heat flux greater than 10 MW/m22, levels that...
Samuel A. Humphry-Baker
(Department of Materials)
07/09/2016, 11:00
High-field spherical tokamaks may be a viable technology for relatively compact fusion power devices (Costley et al Nucl. Fus. 2015). However, such reactors leave little space for shielding of the central column, which must protect the inner superconducting magnets from high energy neutrons. Tungsten carbide cermets are promising candidate materials for such shields: They have high thermal...
Valentina Marascu
(National Institute for Laser)
07/09/2016, 11:00
Controlled fusion research represents an important step for sustainable energy production once with the development of the International Thermonuclear Experimental Reactor (ITER). ITER proposes a deuterium - tritium fusion reaction for hot plasma creation. During plasma- wall interactions, small tungsten particles, from nm to microns will be produced in the tokamak chamber. These particles can...
Richard E. Nygren
(Sandia National Laboratories)
07/09/2016, 11:00
Power exhaust is perhaps foremost among the issues for ITER and post-ITER devices, as well as for existing large confinement devices as they increase power. A related concern is the alignment of plasma facing components to avoid protruding (leading) edges that would intercept field lines and incur very high loads and high erosion. This concern prompted the transient melt experiment in JET,...
Rodrigo Mateus
(Instituto de Plasmas e Fusão Nuclear)
07/09/2016, 11:00
Migration of impurities during ITER plasma discharges will result in the formation of co-deposited mixed materials on the surface of plasma facing components (PFC) with properties distinct from those of the original PFC. These issues have motivated the fusion community to investigate Be-W coatings, in particular their fuel retention behaviour, since in ITER the deposits will present a...
Liga Avotina
(Institute of Chemical Physics)
07/09/2016, 11:00
Tungsten covered carbon materials due to good thermal conductivity of carbon based materials (up to ~250 Wm-1-1K-1 -1 for carbon fiber composites [1]) are suitable for use in fusion devices, like ITER (International Thermonuclear Experimental Reactor) [2], as divertor materials. However, during the plasma wall interactions, erosion and re-deposition, as well as formation...
Hanns Gietl
(Max-Planck-Institut für Plasmaphysik (IPP))
07/09/2016, 11:00
Tungsten is a promising plasma facing material for future fusion reactors due to its unique property combination such as low sputter yield, high melting point and low activation. The main drawbacks for the use of pure tungsten are the brittleness below the ductile-to-brittle transition temperature and the embrittlement during operation e.g. by overheating and neutron irradiation. This...
Cristian Ruset
(Plasma Phisics and Nuclear Fusion)
07/09/2016, 11:00
Tungsten coatings deposited on carbon materials such as carbon fibre composite (CFC) or fine grain graphite (FGG) are currently used in fusion devices as armour for plasma facing components (PFC). About 1800 CFC tiles were W-coated for the ITER-like Wall at JET and more than 1300 FGG tiles were coated for the ASDEX Upgrade tokamak. At present the W coating production is on going for the first...
Keisuke Azuma
(Graduate School of Science)
07/09/2016, 11:00
Tungsten (W) is a candidate for plasma facing materials in D-T fusion reactors due to its higher melting point and lower sputtering yield. During the plasma operation, W will be exposed to energetic particles including hydrogen isotopes, neutrons, and impurities like carbon (C). It is well known that hydrogen isotopes are trapped in the defects produced by the energetic particle irradiation....
Yuya Miyoshi
(Japan Atomic Energy Agency)
07/09/2016, 11:00
Understanding of the heat load profile on the first wall (1stst wall) is one of the key issues to establish the DEMO blanket concept, because the thermal stress on the each blanket module depends on its surface heat load, and it will vary with the 1stst wall shape, the toroidal/poloidal position and the plasma equilibrium. Thus, the 1stst wall surface of the...
Sebastian Ruck
(Institute of Neutron Physics and Reactor Technology)
07/09/2016, 11:00
Rib-roughening the helium-gas cooled channels in plasma-facing components of DEMO (First Wall (FW), limiters or the divertor) enhances heat transfer and reduces structural material operation temperatures. The rib-elements induce a three-dimensional, unsteady flow field and heat transfer is augmented by mixing the fluid in the near wall regions and boundary layers. Whereas the overall heat...
Ali Abou-Sena
(Institute of Neutron Physics and Reactor Technology)
07/09/2016, 11:00
The First Wall (FW) of the EU Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) faces the fusion plasma and experiences high heat fluxes; therefore its cooling channels design is a key R&D task for qualifying the HCPB TBM for the fusion reactors ITER and DEMO. Within the manufacturing and qualification activities performed in KIT for the HCPB TBM, a First Wall Mock-up (FWM) was...
Tomas Melichar
(Research Centre Rez)
07/09/2016, 11:00
Dual Coolant Lithium Lead (DCLL) is one of the four breeding blanket concepts being developed within the EUROfusion project as candidates for the European DEMO. One of the most challenging components of breeding blanket in terms of thermal-hydraulic is a first wall. In order to handle the high thermal loads that the DCLL first wall is facing a proper design of a helium cooling system is...
50048.
P3.120 Development of force reconstruction method on EU ITER TBM based on strain measurements
Christian Zeile
(Karlsruhe Institute of Technology (KIT))
07/09/2016, 11:00
The EU ITER Test Blanket Module (TBM) sets, which consist of TBM box and shield, will be located inside the equatorial port #16 of ITER. One of the important objectives of the TBM program, starting from the first H-H phase, is the validation of the theoretical predictions of the structural behavior of the TBM set under thermal, mechanical and electromagnetic loads. High electromagnetic forces...
Taishi Sugiyama
(Graduate school of energy science)
07/09/2016, 11:00
DEMO reactor must achieve total TBR >1 with high level of accuracy and confidence in the design process. However there is no relevant neutron sources before ITER /TBM, and even in ITER, neutron field is considerably different due to the shield blankets surrounding TBMs. This study proposes verification technique to experimentally simulate reactor neutron field and evaluates its expected...
Ivan Alessio Maione
(Institute for Neutron Physics and Reactor Technology)
07/09/2016, 11:00
Off-normal operations in Tokamak reactors result in the induction of eddy currents that, coupled with the large magnetic field, impose strong electromagnetic forces (Lorentz’s forces) to fusion reactor components. In addition the presence of ferromagnetic material induces Maxwell’s forces as interaction between the magnetized material and the external magnetic field that are thus present also...
Sergey Grashin
(NRC "Kurchatov Institute")
07/09/2016, 11:00
In 2015 the graphite limiter was replaced by the tungsten one on the T-10 tokamak. The limiter was made in “Efremov Institute” from the ITER-grade “POLEMA” tungsten used for ITER divertor plates manufacturing. “POLEMA” tungsten doesn’t contain any impurities and has a high thermal conductivity and heat capacity. Tungsten has a polycrystalline structure with a grain size about 30µm. The...
Aleksey Arakcheev
(Budker Institute of Nuclear Physics)
07/09/2016, 11:00
The residual mechanical deformation and stress were measured in the preliminary experiments carried out at synchrotron radiation (SR) scattering stations on VEPP-3 in the Siberian Center of Synchrotron and Terahertz Radiation. Significant changes in the SR diffraction are found as the result of material recrystallization or irradiation of the material by plasma or high energy ions. It implies...
Vladimir Weinzettl
(Institute of Plasma Physics of The Czech Academy of Sciences)
07/09/2016, 11:00
Dust transport is among important issues for ITER and DEMO, where material erosion will be significant. One of possible mechanisms how material is eroded from plasma facing surfaces is the remobilization of dust particles linked to their lifetime there and to the formation of dust accumulation sites. On the COMPASS tokamak, dust remobilization experiments have been performed using a tungsten...
Christian Bachmann
(Power Plant Physics and Technology)
07/09/2016, 11:00
An essential goal of the EU fusion roadmap is the development of design and technology of a Demonstration Fusion Power Reactor (DEMO) to follow ITER. A pragmatic approach is advocated considering a pulsed tokamak based on mature technologies and reliable regimes of operation, extrapolated as far as possible from the ITER experience. The EUROfusion Power Plant Physics and Technology Department...
Fabio Cismondi
(Eurofusion-PPPT)
07/09/2016, 11:00
In the framework of the EUROfusion DEMO Programme, the Programme Management Unit (PMU) is assuming the role of the plant and tokamak design integration. It is recognized, in part thanks to the ITER experience, that due to the large number of complex systems assembled into the tokamak vessel for integration it is of vital importance to address the in-vessel integration at an early stage in the...
Gandolfo Alessandro Spagnuolo
(Institute for Neutron Physics and Reactor Technology (INR))
07/09/2016, 11:00
The development of the fusion technology reliability involves, among other issues, the improvement of simulation tools to be used for the design of reactor key components, such as the Breeding Blanket (BB), where the engineering requirements and constraints are of nuclear, material and safety kind. For this reason, advanced simulation tools are needed. In the European DEMO project, several...
Giuseppe Mazzone
(Unità Tecnica Fusione)
07/09/2016, 11:00
Among the design activities of the DEMO divertor cassette carried out in the frame of EUROfusion an important parameter is the operating temperature of the divertor cassette. As for the DEMO breeding blanket Eurofer has been chosen as structural material of the divertor cassette due to its low long-term activation, low creep and swelling behavior under neutron fluence. The choice of the...
Domenico Marzullo
(Department of Industrial Engineering)
07/09/2016, 11:00
This paper presents the pre-conceptual design activities conducted for the European DEMO divertor, focusing on cassette design and Plasma Facing Components (PFC) integration. Following the systems engineering principles for the conceptual stage, high level design requirements are collected and conceptual 3D model of divertor’s cassette is presented. The work moved from the geometrical and...
Youji Someya
(Sector of fusion research and development)
07/09/2016, 11:00
Periodical replacement of in-vessel components is required for DEMO. The surface dose rate of in-vessel components for DEMO with fusion power of 1.5 GW is higher than that of shielding blanket in ITER by double digits. In addition, DEMO requires five-year cooling time for decreasing its dose rate to the level of ITER. Therefore, it is difficult to adopt the in-vessel maintenance scheme as ITER...
Peter Titus
(Analysis Branch Mechanical Engineering Division)
07/09/2016, 11:00
The Korean fusion demonstration reactor (K-DEMO) is in the early stages of conceptual design. Ceramic breeder blanket modules are being investigated. These have had extensive nuclear and thermal evaluations. Structural assessments are in process. This paper presents stress analyses performed at PPPL in support of the blanket design. Disruption loading, including the effects of ferromagnetic...
Rocco Mozzillo
(Industrial Engineering)
07/09/2016, 11:00
One of the most critical components in the design of DEMO Power Plant is the Breeding Blanket (BB). Currently, four candidates are investigated as options for DEMO. One of these is the Water Coolant Lithium Lead (WCLL) Breeding Blanket (BB). A new WCLL BB concept design has been proposed and investigated, starting from DEMO 2015 reference configuration. The first activity driving the BB design...
Hiroyasu Utoh
(Japan Atomic Energy Agency)
07/09/2016, 11:00
Maintenance is one of the critical issues in the DEMO design. Several maintenance schemes has been comparatively evaluated from the viewpoint of plasma positional control, in-vessel transferring mechanism of blanket segment, and pipe connection in order to establish a feasible reactor maintenance scheme on the DEMO reactor. Two options has been selected as likely remote maintenance schemes on...
Ming Li
(Mechanical Engineering)
07/09/2016, 11:00
In the inside engineering of DEMO, the robotic machines or manipulators are foreseeable to be widely employed, which often have to deal with the demanding working conditions. The construction of the dynamic model of the robotic machine or manipulator can not only benefit the performance evaluation of the manipulator in the early design stage, but also can be incorporated into the control...
Alberto Vale
(Instituto de Plasmas e Fusao Nuclear)
07/09/2016, 11:00
In DEMO, the ex-vessel Remote Maintenance Systems (RMS) are responsible for the replacement and transportation of the plasma facing components. The ex-vessel operations of transportation are performed by cranes or by means of cask transfer systems (CTS) moved by trolleys.
The main loads of transportation are the blankets and divertors. The blankets are extracted and transported vertically by...
Romain Sibois
(Remote Operation and Virtual Reality)
07/09/2016, 11:00
The next European fusion reactor after ITER is called DEMO. The development implementing ITER experiences has taken place within EUROfusion Programme. One of the reactor maintenance system development tasks has been focused on Divertor Maintenance system. The maintenance of DEMO involving handling hazardous components shall be carried out remotely such as the installation and removal of the...
Kumarpalsinh Jadeja
(Institute for plasma Research)
07/09/2016, 11:00
The First Indian tokamak, ADITYA had successfully completed 25 years of operation of limiter plasma at the Institute for Plasma Research (IPR). After achieving the targeted plasma and successfully carrying out many major tokamak experiments, the up-gradation of ADITYA tokamak with diverter configuration was planed. The upgradation includes the replacement of rectangular cross section vacuum...
Hitoshi Tamura
(Department of Helical Plasma Research)
07/09/2016, 11:00
The design activity of a conceptual design of a helical fusion reactor FFHR-d1 is progressing at the National Institute for Fusion Science. The superconducting magnet system of FFHR-d1 comprises one pair of helical coils, two sets of vertical field coils, and the coil support structure. The major and the minor radii of the helical coil are 5.6 m and 3.774 m, respectively. The magnetic field at...
Axel von der Weth
(INR-MET)
07/09/2016, 11:00
The hydrogen isotopes Tritium and Deuterium will be the fuel of future fusion power plants. These isotopes will be in contact with components of the reactor, as well as with auxiliary systems. For safety studies and the overall Tritium budget, hydrogen transport parameters are necessary to perform according analyses. Reduced Activating Ferritic Martensitic (RAFM) steels at operation conditions...
Marta Malo
(Fundación UNED-Ciemat)
07/09/2016, 11:00
Tritium permeation through containment structures is an important factor for safety and design analysis of fusion energy systems. This process controls several key aspects of the system performance, including the amount of radioactive tritium released to environment, the requirements on tritium breeding ratio, the tritium recycling from the first wall, and it influences the selection of...
Maribel C Gazquez
(Fusion National Laboratory)
07/09/2016, 11:00
Al-based coatings are proposed as anti-permeation and anti-corrosion barrier in Pb-Li breeding blankets -Water Cooled Lithium-Lead (WCLL), Helium Cooled Lithium-Lead (HCLL) and Dual Coolant Lithium-Lead (DCLL). In this work, Al2O3 coatings have been prepared by Pulsed Laser Deposition (PLD) at Istituto Italiano di Tecnologia (IIT) and they have been qualified in Pb-Li to evaluate its...
Takumi Chikada
(Graduate School of Integrated Science and Technology)
07/09/2016, 11:00
Tritium permeation through structural materials in fusion blankets is one of the most important issues in terms of a fuel loss and radiological hazard. Tritium permeation barriers (TPBs) have been developed for several decades, and erbium oxide (Er2O3) coatings have recently been intensively studied as TPBs. However, irradiation effects in TPB coatings on hydrogen isotope permeation have not...
Belit Garcinuno
(Fusion Technology Division)
07/09/2016, 11:00
Tritium recovery is one of the major issues of a future DEMO reactor, in order to accomplish with the requirement of tritium self-sufficiency. Different techniques have been proposed over the years for the extraction of tritium, depending on the Breeding Blanket technology. After a preliminary selection, the EUROfusion Programme has considered the Permeation Against Vacuum (PAV) technique as...
Arthur Brooks
(Engineering Analysis)
07/09/2016, 11:00
Modeling Blanket Ferromagnetic Loading using Edge Potential Elements
Arthur W Brooks 1 1, Han Zhang11
1Princeton Plasma Physics Laboratory abrooks@pppl.gov
Future fusion experiments and reactors will require first wall materials that can survive the thermal and nuclear radiation environment without structural degradation. Candidate materials that are under consideration...
Kenzo Ibano
(Graduate School of Engineering)
07/09/2016, 11:00
For the fusion reactor operations, the tritium (T) retention and permeation in the reactor walls are important for points of views of safety and fuel cycle. It is known that T retention in tungsten (W) is less severe compared with carbon (C). However, recent experimental studies revealed that the neutron irradiated damage, surface recrystallization, and fuzz formation by He ion irradiation...
Matthias Kolb
(Institute for Applied Materials)
07/09/2016, 11:00
Advanced ceramic breeder pebbles composed of a mixture of Li4SiO4 (LOS) and Li2TiO3 (LMT) are fabricated and developed at KIT by a melt-based process (KALOS). The produced pebbles are easily characterized for their non-nuclear properties. Yet, as the main properties of a tritium breeder material are the generation and release of tritium, these characteristics also have to be examined.
Neutron...
Zhenxing Liu
(Department of Reactor Engineering Research & Design)
07/09/2016, 11:00
Abstract: This paper presents the results of experimental study of the columns packed with Palladium deposited on kieselguhr (Pd/k). The characteristic of pressure resistance and separation of hydrogen isotopes of the Pd/k column was investigated. The corresponding relationships among pressure resistance characteristics of Pd/k separation column and Pd/k material physicochemical properties,...
Fred Thomas
(York Plasma Institute)
07/09/2016, 11:00
Tritium self-sufficiency is a fundamental requirement for future DT fusion demonstration and commercial power plants. Hence, prior to the construction of expensive, complex fusion breeder blanket assemblies there should be a concerted effort to quantify and ultimately reduce the uncertainties associated with various nuclear observables. This will enable tritium self-sufficient blankets to be...
Jonathan Klabacha
(Nuclear Engineering)
07/09/2016, 11:00
Looking towards the future of fusion devices, detailed understanding of the underlying working properties is desired knowledge. Even though there are many fusion devices available and extensive operating data is being collected, computational analysis is an underlying requirement to fully understand how a fusion device will operate. Due to the extensive complexity of fusion devices a...
Jonathan Shimwell
(Culham Centre for Fusion Energy)
07/09/2016, 11:00
The Helium Cooled Pebble Bed (HCPB) breeder blanket is being developed as part of the European Fusion Programme. Part of the programme is to investigate blanket modules relevant for future demonstration fusion power plants. This paper presents fluid dynamic, thermomechanical and neutronic analyses of the helium cooled pebble bed with an alternative neutron multiplier, Be12Ti which is...
Julia M. Heuser
(Institute for Applied Materials (IAM))
07/09/2016, 11:00
The investigation of Ceramic Breeders (CB) is of great concern for the development of the solid breeder concept for ITER and DEMO. To ensure an adequate tritium production of the breeder material several requirements like a high lithium density, good tritium release behaviour, and a high resistance against irradiation as well as thermomechanical stresses have to be fulfilled. Lithium...
Lida Magielsen
(Research and Innovation)
07/09/2016, 11:00
In the frame of the European Tritium Breeder blanket development for DEMO, two high dose irradiations of beryllium and beryllides, to be used as neutron multiplier, have been performedin the High Flux Reactor Petten (NL). From one irradiation, to 3000 appm He, the post irradiation results have been published in previous proceedings.
In the second High Dose Beryllium irradiation (HIDOBE-02),...
Huaqin Kou
(Institute of Materials)
07/09/2016, 11:00
Fast and efficient activation of ZrCo is beneficial to promote its application to hydrogen isotopes storage in the fusion energy field. To obtain the optimum activation procedures, the influences of temperature and hydrogen pressure on the activation behavior of ZrCo were systematically investigated. Experimental results showed that fast and efficient activation of ZrCo could be achieved by...
Wei Li
(School of Nuclear Science and Technology)
07/09/2016, 11:00
Chinese Fusion Engineering Test Reactor（CFETR）is a necessary and feasible engineering test reactor which aims at developing the fusion energy while the helium cooled solid breeder blanket (HCSB) is one of the most significant component of it. During the reactor operation stage, the blanket will be activated to produce highly radioactive substances by high energy neutrons irradiation. In order...
Hyoseong Gwon
(Department of Blanket System Research)
07/09/2016, 11:00
Decay heat produced by neutron irradiation can lead to temperature rise in blanket even after plasma shutdown. The excessive temperature increase of blanket structure would be concerned with increase of decay heat when assuming loss of cooling capability for blanket even though vacuum vessel is assumed to be normally cooled with a safety function. The neutron wall loading is designed to be...
Yi-Hyun Park
(TBM Technology Team)
07/09/2016, 11:00
Lithium-containing ceramics (Li-ceramics) are considering as tritium breeding material in pebble-bed form for solid-type breeding blanket in fusion reactor. The tritium breeding material requires small particle size to reduce diffusion distance of generated tritium in the intercrystalline. In addition, the essential resource, especially enriched Li-6, has to recover from the used tritium...
Masaru Nakamichi
(Sector of Fusion Research and Development)
07/09/2016, 11:00
Hydrogen generation via an oxidation reaction of beryllium as an existing neutron multiplier with steam at high temperatures should be reduced on safety hazard for a fusion reactor. Therefore, advanced neutron multipliers with high stability at high temperatures are desirable for the fusion reactor in which water coolant is extensively used. Beryllium intermetallic compounds (beryllides) are...
Ryoutarou Yamamoto
(Advanced of Energy Engineering science)
07/09/2016, 11:00
Understanding of Li evaporation property is important because Li mass transfer decreases tritium breeding ratio and influences tritium behavior. In JAEA, the development of Li2TiO3with excess Li has been performed as an advanced tritium breeder. The present authors revealed in previous works that a layer existing on the pebble surface includes Li2CO3 and it contributes Li mass loss. Recently,...
Yu Otani
(Prime Mover Engineering)
07/09/2016, 11:00
Lithium metatitanate (Li2TiO3) is one of the candidate materials among the solid tritium breeders proposed because of its good tritium release property and high chemical stability [1]. Lithium metatitanate with excess Li (Li2+xTiO3+y) has been recognized as an another prominent candidate material owing to its higher Li density [2]. Demonstration power plant (DEMO) reactors require tritium...
Kiyoto Shin-mura
(Course of Mechanical Engineering)
07/09/2016, 11:00
Lithium metatitanate (Li2TiO3) is one of the candidate materials for solid tritium breeder proposed because of its good tritium release property and high chemical stability [1], and Lithium metatitanate with excess Li (Li2+xTiO3+y) has been recognized as a prominent candidate material owing to its higher Li density [2]. However, demonstration power plant (DEMO) reactors require tritium...
Arturs Zarins
(Institute of Chemical Physics)
07/09/2016, 11:00
Modified lithium orthosilicate pebbles with additions of titanium dioxide are suggested as an alternative tritium breeding ceramic for the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM). The tritium breeding ceramic in the HCPB TBM will be under the action of harsh operation conditions. Radiolysis can take place as a result, and unstable radiation-induced defects (RD) and radiolysis...
Marigrazia Moscardini
(Institute for Applied Materials)
07/09/2016, 11:00
Five ITER project members are actively involved in the fabrication of tritium breeding ceramics pebbles. Different fabrication processes developed by these members strongly influence the characteristics of pebbles produced. One of the main characteristics is the sphericity of pebbles. The spherical shape is the one desired; however the manufacture of perfect round particles is not simple. For...
Fernando Sanchez
(National Fusion Laboratory (LNF))
07/09/2016, 11:00
SiC is a primary candidate for flow channel inserts in blankets due to their excellent thermo-mechanical properties. During reactor operation SiC will be exposed to tritium in a hostile radiation environment. Absorption, diffusion, and desorption will occur, and are expected to depend on the neutron and ionizing radiation conditions. We present work to assess the effect of displacement damage...
Yasuhisa Oya
(College of Science)
07/09/2016, 11:00
Silicon carbide (SiC) is considered to be used for blanket modules for high temperature gas–cooling system in D-T fusion reactors, as SiC/SiC composites. During D-T fusion operation, SiC will be exposed to heavy radiation conditions by neutron and/or gamma-ray. These radiation induces the formation of various damages by a collision process and an electron excitation process, leading to the...
Changho Park
(Japan Agency for Quantum and Radiological Science and Technology)
07/09/2016, 11:00
Lead−lithium (Pb−Li) alloy are considered as a coolant and a tritium breeder for fusion reactor blanket systems. One of the critical requirements for the realization of this systems is the compatibility of silicon carbide (SiC) and its composites as structural and/or functional materials. The authors investigated that inclusions, possibly Li−oxides in Pb−Li may have certain impacts on...
Enrique Ascasibar
(National Fusion Laboratory)
07/09/2016, 11:00
During ITER and DEMO reactor operation the proposed Li-Pb blanket flow channel inserts made of SiC ceramic material will be exposed to both radiation and tritium. Absorption, diffusion, and desorption of tritium is expected to occur and these processes will strongly depend on the irradiation conditions, neutron flux, and purely ionizing radiation. Previous results have shown that marked...
Saerom Kwon
(Japan Atomic Energy Agency)
07/09/2016, 11:00
In our previous copper benchmark experiment we had pointed out that the elastic scattering and capture reaction data of the copper had included some problems in the resonance region, which had caused a large underestimation of reaction rates of non-threshold reactions. In order to corroborate this issue, we carried out a new benchmark experiment on copper in the neutron field with more low...
Masayuki Ohta
(Japan Atomic Energy Agency)
07/09/2016, 11:00
In our previous benchmark experiment on molybdenum at JAEA/FNS, we found problems of the (n,2n) and (n,γ) cross sections in Mo of JENDL-4.0. However, the Mo data only above a few hundred eV were investigated, because there were few neutrons with lower energy in the Mo assembly in the previous experiment. We perform a new benchmark experiment on Mo in order to validate the Mo data in the lower...
Snejana Bakardjieva
(Institute of Inorganic Chemistry AS CR)
07/09/2016, 11:00
Materials from the group of layered Mn+1AXn phases are new type of nanolaminates which can be used in many technical applications, especially as viable candidates for high-radiation structural application in fusion technology.
It has been proposed that the novel physical properties of MAX phases arise from their atomic structure, combining “ceramic” MX6 octahedra layers with a single...
Jan Engels
(Institut für Energie- und Klimaforschung – Plasmaphysik)
07/09/2016, 11:00
In fusion power plants a tritium permeation barrier is required in order to prevent the loss of the fuel inventory. Moreover, the tritium permeation barrier is necessary to avoid that the radioactive tritium accumulates in the first wall, the cooling system, and other parts of the power plant. Oxide thin films, e.g. Er2O3 and Y2O3, are promising candidates as tritium permeation barrier layers....
Monika Vilemova
(Materials Engineering)
07/09/2016, 11:00
Pure tungsten is considered as the most suitable plasma facing material for the reactor first wall. However, number of studies points out serious drawbacks related to tungsten mechanical properties that negatively affect lifetime of first wall components. Serious risk for the divertor comes from abnormal events, such as disruptions, vertical displacement events (VDEs) and edge localized modes...
Sven-Erik Wulf
(Institute for Applied Materials)
07/09/2016, 11:00
Different breeding blanket designs for a future fusion power plant (DEMO) consider Eurofer steel as a main structural material. Nevertheless, RAFM steels suffer from severe corrosion attack in Pb-15.7Li, which acts as breeding material in the liquid breeder blanket designs, e.g. HCLL, WCLL and DCLL. The resulting corrosion products may cause safety risks e.g. concerning tube plugging due to...
Shuhei Nogami
(Department of Quantum Science and Energy Engineering)
07/09/2016, 11:00
Tungsten (W) is a primary candidate for fusion reactor divertor because of its high melting point, thermal conductivity and sputtering resistance. To improve its structural reliability, improvement of mechanical properties and suppression of recrystallization of the W materials are necessary. It is well known that the grain refining, work hardening, solid solution strengthening, and dispersion...
Anatoli Popov
(Institute of Solid State Physics)
07/09/2016, 11:00
The radiation-resistant insulators (MgO, Al2O3, MgAl2O4, BeO etc) are important key materials for fusion reactors. It is very important to predict/simulate not only the kinetics of diffusion-controlled defect accumulation under neutron irradiation, but also a long-time defect structure evolution including thermal defect annealing. Here we developed and applied the advanced theoretical approach...
Teruya Tanaka
(National Institute for Fusion Science)
07/09/2016, 11:00
In our previous study, a Cr2O3 layer was formed on a reduced activation ferritic/martenstic (RAFM) steel substrates by heat treatment under a reduced atmosphere and it could suppress hydrogen permeation by ~2 orders at 550-650 ooC. Since the Cr2O3 layer was stable at high temperatures in air, it was also a preferable underlayer for multi-layer ceramic coating with the metal organic...
Jumpei Mochizuki
(Shizuoka University)
07/09/2016, 11:00
Tritium permeation barrier (TPB) has been investigated for the establishment of an efficient fuel cycle and radiological safety in fusion power plants. One of critical issues for TPB is degradation caused by introduction of cracks and pores. Even if a microscopic crack is introduced, tritium permeation is drastically increased. The development of self-healing coating is one of techniques for...
Seira Horikoshi
(Graduate School of Integrated Science and Technology)
07/09/2016, 11:00
To establish liquid lithium-lead blanket concepts, the development of a functional coating as a tritium permeation barrier with corrosion resistance is required. In our previous study, erbium oxide (erbia)-iron two-layer coatings showed a better compatibility than erbia single-layer coatings with keeping a high permeation reduction factor (PRF). In this study, hydrogen isotope migration...
Hynek Hadraba
(Institute of Physics of Materials)
07/09/2016, 11:00
The structural components used for construction of future generation of fission reactors and fusion reactors will undergo demanding service conditions as high neutron doses, high temperature and extremely corrosive environment. The nano-structured oxide dispersion steels (ODS) containing small amounts of homogeneously dispersed nano-size yttria particles were developed as structural material...
Filip Siska
(Institute of Physics of Materials)
07/09/2016, 11:00
ODS steels are candidates for the structural material in the future fusion power plants. Their main advantage is high strength and creep resistance at high temperatures. Such high performance is achieved by the presence of the oxide particles in the microstructure. Nowadays, the best ODS steels contain particles of Y2O3 which are stable at high temperatures. However, yttrium is expensive and...
Simon Heuer
(Forschungszentrum Jülich GmbH)
07/09/2016, 11:00
Future fusion reactors may exhibit first walls composed of a tungsten (W) armor, that is attached to a subjacent stainless steel (SS) structure. Joining these materials for the application at hand is challenging because the pulsed operation of TOKAMAK reactors induces thermo-mechanical stresses and strains at the W/SS interface due to differing materials properties. These cyclic loads will...
Ming Sun
(Institute of Nuclear Energy Safety Technology)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Fusion reactor is one of new type reactors being developed , and it is cleaner and more efficient than the fission reactor. Each SSCs (Structures, Systems, Components) has different safety importance to fusion reactors. So it is necessary to classify the SSCs of fusion reactors. And the safety classification of SSCs for fusion reactor is the important basis of reactor design and construction....
Miao Nie
(Key Laboratory of Neutronics and Radiation Safety)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The high reliability and availability of Tritium extraction system (TES) will be needed is necessary for safety operation of circulation and processing of tritium purge gas. Reliability, availability, maintainability, inspectability (RAMI) analysis of the TES should be performed during the design and operation phase. Since there is no TES failure rate data available from fusion operating...
Shijun Qin
(Tokamak Design Division)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the EAST in-vessel components cooling system based on currently available design is presented. The following sub-systems were considered in the analysis: the EAST PFCs heat-sink cooling system, two water pumps system, cooling loop including cycle feed pipe and cycle return pipe lines, secondary...
Paul-Martin Steffen
(Institute of Energy and Climate Research – Nuclear Waste Management and Reactor Safety (IEK-6))
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
In case of a severe accident inside the ITER fusion facility, there exist several scenarios in which hydrogen may be produced and released into the suppression tank. Assuming the accidental ingress of air, the formation of flammable gas mixtures may lead to explosions and severe component failure. One option to mitigate such hypothetical scenarios is the installation of passive auto-catalytic...
Tonio Pinna
(Nuclear Fusion and Safety Technologies Department (FSN-FUSTEC-TEN))
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Safety studies are performed in the frame of the conceptual design studies for the European DEMO reactor to assess the safety and environmental impact of design options. An exhaustive set of reference accident sequences are defined in order to evaluate plant response in the most challenging events and compliance with safety requirements. The identification of a comprehensive set of accident...
Richard Brown
(PMU)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The generation and investigation of alternative design solutions and their benchmarking against criteria that are traceable to high level objectives is a fundamental facet of a holistic systems engineering approach. During the pre-conceptual design phase of DEMO, characterisation studies for multiple plant concepts are being conducted in parallel to explore the design space and evaluate the...
Fabrizio Franza
(Institute for Neutron Physics and Reactor Technology)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
A fusion power plant is characterized by many subsystems operating under extreme thermal and nuclear conditions, thus compelling to be designed according to physics and engineering constraints. For such an operation, dedicated tools called systems codes are currently used. At Karlsruhe Institute of Technology (KIT), a dedicated modelling campaign has been recently launched aiming to study the...
Lei Lu
(Neutronics and Nuclear Data Group)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
McCad is a geometry conversion tool developed at the Karlsruhe Institute of Technology (KIT) for the automatic conversion of CAD models into the constructive solid geometry (CSG) representation. The resulting geometry models can then be used in Monte Carlo (MC) particle transport simulations applied in design analyses of fusion reactors like the DEMO tokamak developed within the European Power...
Xiaoman Cheng
(Institute of Plasma Physics)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). In this work, the Primary Heat Transfer System (PHTS) of the WCCB blanket was designed based on the configuration of the blanket sectors, employing two identical loops at this stage. And each loop consists of a steam generator, a pressurizer and a main pump,...
Taehyun Tak
(KSTAR Control Research Team)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The ITER Central Interlock System (CIS) architecture is composed of four categories of hardware: fast architecture, slow PLC based architecture, hardwired architecture and servers.
The CIS fast architecture receives interlock events from various local plant systems of ITER and communicates the corresponding actions to any other local plant systems in order to avoid or mitigate the damage to...
Rafael Juarez
(Departamento de Ingeniería Energética)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
ITER is a prominent facility in the development of the nuclear fusion. It presents 44 ports providing access to the Vacuum Vessel at three different heights: Lower, Equatorial and Upper ports. Out of them, 22 ports, correspond to Diagnostics ports. They host a diversity of diagnostics systems, designed by the different ITER Domestics Agencies (DAs). They are later integrated into the different...
Jia Li
(School of Nuclear Science and Technology)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
In order to control the global sample frequency, GVR method is deemed to be a practical way. But it is common that GVR method needs too many steps of weigh window iteration and it may fall into a long-history problem. We introduce a novel approach that is GVR method combined with reduced density in model, which could improve the calculation efficiency of GVR method in the following two...
Shengpeng Yu
(Institute of Nuclear Energy Safety Technology)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The advantages of CAD based Automatic Modeling make it possible to efficiently describe and verify complex nuclear system, such as ITER, for Nuclear Analysis. SuperMC/MCAM, the most widely applied CAD based Automatic Modeling tool for Monte Carlo, is currently focusing on modeling for Monte Carlo partile transport programs.
Being more and more detailed, the radiation shielding modeling of...
Miguel Correia
(Instituto de Plasmas e Fusão Nuclear)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
High availability (HA) is a key element in the specification of next generation Fusion devices, targeting steady-state operation. HA is especially required on mission-critical systems, as is the case of experimental Fusion devices and future Fusion power plants, where safety of people, environment and the infrastructure/investment is a primordial priority.
IPFN developed control and data...
Qi Yang
(Key Laboratory of Neutronics and Radiation Safety)
07/09/2016, 11:00
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Activation study is very important for fusion reactors, from the view of component maintenance, occupational radiation exposure, and radioactive waste management. SuperMC is a multi-functional, intelligent, accurate and user-friendly simulation software system with comprehensive functions of transport simulation, material activation and transmutation, radiation source term and dose, etc.
The...
Rafael Vila
(Fusion Materials Unit)
08/09/2016, 08:30
Oral
With start of EUROfusion Materials-WP in 2014, functional materials (FM) have been included as a new branch. Their main scopes are issues of optical and dielectric materials for DEMO applications. R&D of these materials are, in particular, essential for Diagnostics and Heating and Current Drive (H&CD) systems that must provide critical services such as machine control, protection, performance...
V. Toigo
(Consorzio RFX)
08/09/2016, 09:10
Oral
The realization of the ITER Neutral Beam Test Facility (NBTF) and the start the experimental phase are important tasks of the fusion roadmap, since the target requirements of injecting to the plasma a beam of Deuterium atoms with a power up to 16.5 MW, at 1MeV of energy and with a pulse length up to 3600s have never been reached together before.
The ITER NBTF, called PRIMA (Padova Research on...
H. Fuenfgelder
(Max-Planck-Institut für Plasmaphysik)
08/09/2016, 09:50
Oral
A enhanced impurity production has often accompanied experiments using ICRF (Ion Cyclotron Range of Frequency) as heating method. Positive effects, such as the capability to deposit the power centrally even at high density and thereby reduce the central impurity accumulation, were wiped out in the all‐metal ASDEX Upgrade when the antenna limiters were also coated with W. The hypothesis that...
Alexander Huber
(Forschungszentrum Jülich GmbH)
08/09/2016, 11:00
In magnetic fusion devices of the next generation such as ITER, high neutron and gamma-ray yields could be detrimental to the applied diagnostic equipment such as video imaging systems as well as to electronic components of machine control systems. Semiconductors devices are particularly sensitive to the radiation, both ionizing (formation of traps at the Si/SiO2 interface with energy levels...
Henri Greuner
(Max-Planck-Institut für Plasmaphysik)
08/09/2016, 11:00
Plasma-facing units equipped with tungsten (W) monoblock geometry are employed at the vertical targets of the ITER divertor. This contribution discusses a statistical approach for high heat flux (HHF) tests as potential quality assessment of the ITER divertor additional to the quality assurance performed by the manufacturer during the manufacturing.
The IR analysis of the local temperature...
Chao Liu
(Key Laboratory of Neutronics and Radiation Safety)
08/09/2016, 11:00
Abstract:
Fusion energy becomes essential to solve the energy problem with the increase of energy demands. Although the recent studies of fusion energy have demonstrated the feasibility of fusion power, it commonly realizes that more hard work is needed on neutronics and safety before real application of fusion energy. A high intensity D-T fusion neutron generator is keenly needed for the...
Elsa Henriques
(LAETA, IDMEC, Instituto Superior Tecnico, Universidade de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa, Portugal)
08/09/2016, 11:20
This paper describes the preliminary RAMI analysis for the ITER Low Field Side Collective Thomson Scattering (LFS CTS) system based on its preliminary architecture achieved at the System Level Design. The benefits and challenges involved in a RAMI analysis since the front end of the design process of the system are discussed together with the methodology pursued. The Functional Analysis,...
Eliseo Visca
(Department of Fusion and Technology for Nuclear Safety and Security)
08/09/2016, 11:20
ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European International Thermonuclear Experimental Reactor (ITER) development activities for the manufacturing of the inner vertical target (IVT) plasma-facing components of the ITER divertor.
During normal operation the heat flux deposited on the bottom segment of divertor is 5-10 MW/m2 but the capability to remove up to...
Neill Taylor
(Culham Centre for Fusion Energy)
08/09/2016, 11:20
As part of the conceptual design studies for a European DEMO, a programme of safety studies and analyses is performed, intended to help guide the design process by assessing the safety and environmental impact of design options under consideration. They also begin to prepare for the eventual licensing of DEMO construction and operation by a European nuclear regulator. A safety approach has...
Jorge Sousa
(Instituto de Plasmas e Fusão Nuclear)
08/09/2016, 11:40
Abstract:
The increasingly complex Physics experiments demand innovative digital Instrumentation for critical Measurement and Control functions. Requested system capabilities are, at least: high reliability, availability, maintainability, synchronized real-time high throughput data processing and compatibility to established Standards.
Some of the methods that help attaining those capabilities...
Dmitry Rudakov
(Center for Energy Research)
08/09/2016, 11:40
An overview of recent Plasma-Material Interactions (PMI) research at DIII-D tokamak using the Divertor Material Evaluation Station (DiMES) is presented. The DiMES manipulator allows exposing material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by the extensive diagnostic suite including a...
Andrew Grief
(Amec Foster Wheeler)
08/09/2016, 11:40
'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER. Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. The F4E, Amec Foster Wheeler and INL comprehensive...
Andrey Litnovsky
(Forschungszentrum Jülich)
08/09/2016, 12:00
All optical and laser diagnostics in ITER will use mirrors to observe the plasma radiation. In the severe ITER environment mirrors may become contaminated with plasma impurities hampering the performance of corresponding diagnostics. To counteract the mirror contamination, an in-situ mirror cleaning is proposed, which relies on ion sputtering the contaminants together with affected mirror...
Jan Prokupek
(Technological Experimental Loops)
08/09/2016, 12:00
The ITER first wall panels are exposed directly to thermonuclear plasma and must extract heat loads of about 2 MW/m² (EU) to 4.7 MW/m² (RF + CN). The panels are qualified through high heat flux cyclic testing before the installation in ITER. Initially the first wall panel prototypes will undergo full-power tests, this will be followed by the pre-series panels and finally the series panels.
The...
Kwang-Pyo Kim
(National Fusion Research Institute)
08/09/2016, 14:20
To achieve the high performance plasma in the Korea Superconducing Tokamak Advanced Research (KSTAR) tokamak, Neutral Beam Injection (NBI) system has been installed and upgraded. The first NBI (NBI-1) was installed in 2010, which provides a 100 keV deuterium neutral beam of 6 MW maximum using three ion sources. The second NBI (NBI-2) with another 6 MW will complete to be constructed by 2018....
Young-Ju Lee
(Vacuum & cryogenic engineering team)
08/09/2016, 14:20
KSTAR project has required the new helium distribution box named upgraded distribution box (DBU) for the operation of the cryogenic components such as in-vessel cryo-pump (CPI), super-sonic molecular beam injector (SMBI), and hydrogen pellet injection system (PIS). Two CPIs are inserted into the tokamak vacuum vessel and these components shall be operated at 90 K for the liquid nitrogen...
Wolfgang Biel
(Institute for Energy and Climate Research IEK-4 (Plasma Physics))
08/09/2016, 14:20
In the European strategy towards fusion electricity, a demonstration tokamak fusion reactor (DEMO) is foreseen as the single step between ITER and a fusion power plant. Recent studies have been focussing on the concept development for a “conservative” pulsed tokamak reactor with an electrical output power of Pel ~ 500 MW and plasma pulse duration of tpulse ~ 2 hours.
In the design process for...
Dong-Seong Park
(National Fusion Research Institute)
08/09/2016, 14:20
The nuclear fusion research is in progress for the next generation energy source in many countries. The Korea Superconducting Tokamak Advanced Research (KSTAR) in Korea, the Experimental Advanced Superconducting Tokamak (EAST) in China and the Wendelstein7-X in German are the operational superconducting fusion device in the world. The International Thermonuclear Experimental Reactor (ITER) is...
Wook Cho
(Heating and Current Drive Research Team)
08/09/2016, 14:20
In 2015 KSTAR Campaign, the maximum injection power of the KSTAR tangential Neutral Beam Injector (KSTAR NBI-1) is 5.39MW with three ion sources. Issues in beam extraction found during the experiment were 1) a large oscillation of beam current, 2) frequent interrupts in beam extraction due to breakdown in grids, and 3) a distortion of waveform. To solve these issues, we focused on the unstable...
Soo-Hwan Park
(Advanced Technology Research Center)
08/09/2016, 14:20
KSTAR (Korea Superconducting Tokamak Advanced Research) has used gas puffing system as main fueling method since 2008. Up to date total fueling efficiency of gas puff is less than 30 %. Pellet injection is more effective technique to control plasma density than gas puffing system and supersonic molecular beam injection. Many fusion devices such as JET, Tore Supra, ASDEX-U, HL-2A, EAST, and LHD...
Michela De Muri
(Consorzio RFX)
08/09/2016, 14:20
The Padova Research on ITER Megavolt Accelerator (PRIMA), under construction at Consorzio RFX, will host SPIDER test bed, a full-size 100 kV negative ion source, and MITICA test bed, a prototype of the whole ITER injector, aiming to develop and optimize the heating injectors to be installed in ITER.
The production of hydrogen (or deuterium) negative ions inside the sources relies mainly on the...
Mauro Pavei
(Consorzio RFX)
08/09/2016, 14:20
The heating neutral beam injectors (HNBs) at ITER are expected to deliver 33 MW of neutral beam power to the ITER plasma for the purposes of heating and current drive. This is achieved by using 2 injectors, each capable of delivering 16.5 MW of neutral beam power.
The beam source of each injector is a complex assembly composed by an RF based negative ion source having an extraction area of...
Samuele Dal Bello
(Consorzio RFX)
08/09/2016, 14:20
The ITER project requires at least two Neutral Beam Injectors, each accelerating to 1MV a 40A beam of negative deuterium ions, to deliver to the plasma a power of about 33 MW for one hour as additional heating.
Since these requirements have never been experimentally met, it was recognized necessary to build-up a test facility, named PRIMA (Padova Research on ITER Megavolt Accelerator), in...
Sergei Sytchevsky
(JSC «NIIEFA»)
08/09/2016, 14:20
Large state-of-the-art fusion devices involve extensive computations throughout the engineering design process from the concept to the commissioning. A variety of well-established software tools, such as ANSYS, OPERA, CARIDDY, TYPHOON, TORNADO has produced a range of simulation techniques and approaches for electro-magnetic (EM) simulations of principal components of tokamaks. The installation...
Boris Lyublin
(JSC "NIIEFA")
08/09/2016, 14:20
Concrete structures of tokamak buildings are reinforced with steel rebar that produces a substantial contribution into the tokamak field both in the plasma region and in the building where the service staff and magnetically sensitive equipment will be located.
The article describes an advanced approach to modelling magnetic properties of reinforced concrete structures bearing in mind the...
Ilya Gornikel
(Alphysica GmbH)
08/09/2016, 14:20
Cryogenic systems for fusion reactors have to cope with large pulsed heat load generated during fusion experiments. The paper is focused on mitigation of pulsed heat power arriving to the cryoplant from several parallel cooling loops of tokamak superconducting magnets. A new control strategy is proposed. The pressure drop measured at the return cryoline serves as a feedback signal to...
Mahesh Vuppugalla
(Institute for Plasma Research)
08/09/2016, 14:20
Successful operation of a Neutral Beam Injector is dependent on the performance of High voltage power supply system(HVPS) for the production of ion beam. To meet the functional requirements of ion extraction, the power supplies(PS) are designed for fast output cut-off, low energy content during breakdown(BD), ability to withstand repeated BD. It is important that features of the PS are...
Jyoti Shankar Mishra
(Institute for plasma research)
08/09/2016, 14:20
Institute for Plasma Research (IPR), India has a programme of development of allied technologies with applications related to fusion reactor. A pneumatic gas gun kind Single pellet injector system (SPINS-IN) developed at IPR is successfully delivering hydrogen pellets of size 2 mm with a velocity of 700 meters/sec. It is a cryocooler based system operated at a temperature < 10 K and...
Larry Baylor
(Oak Ridge National Laboratory)
08/09/2016, 14:20
The formation and acceleration of cryogenically solidified pellets of hydrogen isotopes has long been under development for fueling fusion plasmas. Fueling with DT pellets injected from the high field side wall has been proposed for future burning plasma tokamak devices. In addition to fueling, smaller shallow penetrating pellets of deuterium injected from the low field side wall have been...
Massimo Zucchetti
(DENERG)
08/09/2016, 14:20
The compact, high field fusion experiment Ignitor aims at the demonstration, for the first time, of ignition in magnetically confined D-T plasmas, together withthe exploration of the physics of the ignition process, and of heating and control of plasmas under controlled burning conditions. The machine parameters have been established on the basis of existing knowledge of the confinement...
Dario Andres Cruz Malagon
(DENERG)
08/09/2016, 14:20
Nuclear Fusion is a candidate as a long-term energy solution for developed countries. A fusion plasma can be fuelled by different kinds of isotopes. The advantages of Deuterium-Helium-3 (DHe) plasmas of advanced fusion reactors lie in the scarcity of neutrons (due to side DD and DT reactions), and direct conversion of the produced energy without thermal cycle.
The proposed CANDOR DHe plasma...
Sayf Elgriw
(Department of Physics and Engineering Physics)
08/09/2016, 14:20
The interaction between resonant magnetic perturbations (RMP) and plasma is an active topic in the fusion energy research. RMP involves the use of radial magnetic fields generated by external coils installed on a tokamak device. The resonant interaction between the plasma and the RMP fields has many favorable effects such as suppression of instabilities and improvement of discharge parameters...
Pietro Vincenzi
(Consorzio RFX)
08/09/2016, 14:20
EU DEMO studies for pulsed (DEMO1) and steady-state (DEMO2) concepts are currently in the pre-conceptual phase [1]. DEMO1 aims at producing about 2GW of fusion power with a burn time of approximately 2 hours. Within EUROfusion Power Plant Physics and Technology department, DEMO scenario modelling is carried out as part of the validation of feasibility and performance of DEMO designs. One of...
Thomas Franke
(EUROfusion Consortium)
08/09/2016, 14:20
The Heating & Current Drive (H&CD) systems in a DEMOnstration fusion power plant are one of the major energy consumers. Due to its high demand in electrical energy produced in the balance of plant (BoP) the H&CD efficiency optimization is one of the main goals of the DEMO development. The energy consumption of the H&CD sub-systems in different plant modes & states and plasma phases need to be...
Giulio Gambetta
(Consorzio RFX)
08/09/2016, 14:20
Several novel design solutions for high performance cooling systems have been developed by Consorzio RFX, permitting to experimentally simulate the challenging heat transfer conditions foreseen in the future fusion devices. The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), aims on one hand to verify the present solution applied inside the MITICA experiment...
Alexey Dnestrovskiy
(Plasma Physics)
08/09/2016, 14:20
Neutral Beam Current Drive (NBCD) is considered as an indispensable mechanism for a steady state regime in such contemporary projects as a tokamak based neutron source or a DEMO type thermonuclear reactor. In this report numerical calculations of NBCD with a Monte Carlo code NUBEAM are complemented by a semianalytical treatment of fast ion velocity distribution function. NBCD parameters were...
Ivan Spassovsky
(Fusion Department)
08/09/2016, 14:20
F. Mirizzi11, M. Carpanese22, S. Ceccuzzi22, F. Ciocci22, G. Dattoli22, E. Di Palma22, A. Doria22, G.P. Gallerano22, G. Maffia22, A. Petralia22, G.L. Ravera22, E. Sabia33, I. Spassovsky22, A.A. Tuccillo22, S. Turtù22, P....
Silvio Ceccuzzi
(FSN - Fusion Physics Division)
08/09/2016, 14:20
In the frame of the feasibility study of a Cyclotron Auto-Resonance Maser (CARM), different solutions for the distributed reflectors of the resonant cavity have been considered and compared. In detail, a 250 GHz CARM source is under design with an output power of 200 kW for pulses up to 0.2 s, representing the first milestone of a more ambitious project, aimed at achieving a CW 1 MW mm-wave...
Amro Bader
(Tokamak Scenario Development Department)
08/09/2016, 14:20
The use of efficient heating and current drive systems is an important research priority for DEMO. The Ion Cyclotron Resonance Heating (ICRH) is one such system justified by its inherent advantages, though in its present status (antenna situated in a port in the Vacuum Vessel (VV) is unacceptable for DEMO, where tritium self-sufficiency is to be demonstrated, and reducing the openings in the...
Bongki Jung
(Nuclear Fusion Engineering Development Division)
08/09/2016, 14:20
A high-power pulsed arc ion source based on Marx generator has been developed at the Korea Atomic Energy Research Institute for the heating NBI system of the VEST which is a compact spherical tokamak at Seoul National University to study the reactor-relevant tokamak operating scenario[1][1]. The NBI system, with a total ion beam power of 0.8MW, was designed for the core plasma...
Sung-Ryul Huh
(Nuclear Fusion Engineering Development Division)
08/09/2016, 14:20
Within the framework for development of the radio frequency (RF) driven positive ion source as an alternative to the conventional filament arc driven ion source for fusion applications, KAERI is currently constructing a new high power (50 kW at a frequency of 2 MHz) large area RF ion source. The ion source was designed to have a rounded rectangular geometry for covering rectangular ion...
Matteo Vallar
(Consorzio RFX)
08/09/2016, 14:20
The planned upgrade of the RFX-mod device is a good opportunity to widen the operational space of the machine, in both RFP and tokamak configurations. Installation of a power neutral beam injector (NBI) is also envisaged and a NBI system compatible with RFX-mod is already available on site. It was previously installed in TPE-RX (Tsukuba, Japan), it has a nominal power of 1.25 MW, a nominal...
Macarena Liniers
(Laboratorio Nacional de Fusión)
08/09/2016, 14:20
Neutral Beam injection has some well-established effects on plasma behaviour, such as the power threshold observed in L to H confinement mode transitions or the fast ion excitation of Alfvén modes, whose underlying mechanisms are still under investigation.
In recent TJ-II experimental campaigns emphasis has been made in the characterisation of those Neutral Beam related effects. A study of...
Takuya Hase
(Tokai University)
08/09/2016, 14:20
Production of negative ions plays an essential role in Neutral Beam Injection (NBI). A negative ion beam with an energy of 1 MeV and a current of 40 A (a current density of 20 mA/cm22) is required for 3600 s to produce 16.5 MW of power. NBI predominantly uses negative hydrogen ion sources based on surface production. These negative hydrogen ion sources require cesium seeding to...
Shaofei Geng
(Department of Fusion Science)
08/09/2016, 14:20
In order to investigated the dynamics of H-- ions and understand the extraction process inside filament-arc-driven plasmas in a Cs-seeded negative ion source, diagnostic experiments using a directional Langmuir probe combined with photodetachment measurement have been conducted. Two-dimensional flow pattern of H-- ions has been obtained as well as the profile of...
Manojkumar Patel
(ICH&CD)
08/09/2016, 14:20
ITER-India is developing Ion Cyclotron Heating & Current Drive (ICH&CD) RF source in the frequency of 35 to 65 MHz. Three cascaded amplifiers along with low power RF section, AC/DC power supplies and controls will be used for getting MW level RF power from one source. In the present configuration, two tube based tuned amplifiers, i.e. driver (~150 kW) and final (1.7MW) stage amplifiers are...
Pierre Dumortier
(LPP-ERM/KMS)
08/09/2016, 14:20
The JET ICRF ITER-like Antenna (ILA) is composed of four resonant double loops (RDLs) arranged in a 2 toroidal by 2 poloidal array. Each RDL consists of two poloidally adjacent straps fed through in-vessel matching capacitors from a common Vacuum Transmission Line. Two toroidally adjacent RDLs are fed through a 3dB combiner-splitter.
The JET ILA antenna has been operating at 33, 42 and 47MHz...
Frederic Durodie`
(Laboratory for Plasma Physics)
08/09/2016, 14:20
The ITER-like Antenna (ILA) [1] for JET is a 2 toroidal by 2 poloidal array of Resonant Double Loops (RDL). It featurs in-vessel matching capacitors feeding RF current straps in Conjugate-T (CT) manner, a low impedance quarter-wave impedance transformer and a service stub allowing hydraulic actuator and water cooling services to reach the aforementioned capacitors. A 2ndnd stage...
A. Dunaevsky
(Tri Alpha Energy)
08/09/2016, 14:20
In the C-2 field-reversed configuration (FRC) experiment, tangential neutral beam injection (NBI), coupled with electrically-biased plasma guns at the plasma ends and advanced surface conditioning, led to dramatic reductions in turbulence-driven losses.11 Under such conditions, highly reproducible, macroscopically stable, hot FRCs with a significant fast-ion population, total plasma...
R.S. Delogu
(Consorzio RFX)
08/09/2016, 14:20
To study and optimize negative ion production, the SPIDER prototype (beam energy 100 keV, current 48 A) is under construction in Padova, Italy. The instrumented calorimeter STRIKE (Short-Time Retractable Instrumented Kalorimeter Experiment) has been designed with the main purpose of characterizing the SPIDER negative ion beam in terms of beam uniformity and divergence during short pulse...
Laurent Jung
(National Fusion Research Institute)
08/09/2016, 14:20
An elaborate control of waveforms of poloidal field (PF) coils is prerequisite to ensure a reliable plasma start-up in ITER. An additional requirement in the ITER PF coil scenario development is that coil currents should be optimized to minimize quench risks during a discharge. In this paper, we use the quadratic programming method to optimize ITER PF coil currents at the initial magnetization...
Marco Cecconello
(Uppsala University)
08/09/2016, 14:20
The ITER Radial Neutron Camera (RNC) is a diagnostic with multiple collimated inputs aiming at characterizing the neutron source. The RNC plays a primary role in the advanced control measurements and physics studies of ITER, and acts as backup for system machine protection and basic control measurements. The RNC primary design driver is the measurement of the neutron emissivity radial profile...
Gerhard Raupp
(Tokamak Scenario Development E1)
08/09/2016, 14:20
To operate ITER and control long and finally thermonuclear discharges with very complex physics and a limited set of actuators requires a sophisticated Plasma Control System (PCS). To provide the required control functionality, the PCS will include many control loops to keep parameters within operation envelopes. These must be backed by exception handling functions, to optimize continuous...
Alessandro Formisano
(Dept. of Industrial and Information Engineering)
08/09/2016, 14:20
The magnet system in ITER is composed by three main coils groups, characterized by tight tolerances on manufacturing and assembly, to keep error fields at levels compatible with plasma operation. Additional coils correct error fields guaranteeing suitable accuracy at start of flat top [1].
Plasma initiation in ITER will be critical, since low electric field will be available, and a reduction...
Luca Zabeo
(ITER Organization)
08/09/2016, 14:20
The ITER Plasma Control System (PCS) is now approaching the second phase of development, the Preliminary Design Review (PDR). The PDR, expected at the end of 2016, is now more deeply investigating possible solutions for the different control areas aimed at operations up to 15MA with low auxiliary heating in L-mode. The entire sequence of a plasma discharge from the break-down to the...
Christopher James Rapson
(Max Planck Institute for Plasma Physics)
08/09/2016, 14:20
Integrated control of many plasma parameters simultaneously is expected to increase the reproducibility and stability of scenarios, which are otherwise developed laboriously through trial and error. The benefits are expected to be especially important for high performance scenarios, operating near multiple stability boundaries. The two main challenges of integrated control are: firstly the...
Jorge M. Santos
(Instituto de Plasmas e Fusao Nuclear)
08/09/2016, 14:20
On future long pulse fusion devices an extended set of diagnostics will play an increasingly important role in advanced plasma control. In particular, O-mode microwave reflectometry will be used, on ITER and foreseeably on DEMO, to complement the standard magnetic diagnostics for plasma position control. With the preliminary design of ITER’s plasma position reflectometers (PPR) presently...
Jiaxian Li
(Center for Fusion Science)
08/09/2016, 14:20
The advanced configurations (snowflake and tripod) have been designed with EFIT based on current poloidal field (PF) coils system of HL-2M to study the advanced divertor physics and support the high performance plasma operation. The characteristic parameters of the advanced configuration (the distance between two X-points, magnetic flux expansion and weak field area and so on), especially the...
Alexey Gorbunov
(NRC Kurchatov Institute)
08/09/2016, 14:20
Laser-induced fluorescence (LIF) diagnostic system on ITER will be used for local measurements of helium density (nHe) and ion temperature (Ti) in the divertor region. The diagnostics is combined with divertor Thomson scattering (DTS) via common laser injection and signal collection optics. Physical aspects of the LIF method for measuring the plasma parameters and the layout of the system...
Ilya Orlovskiy
(NRC "Kurchatov Institute")
08/09/2016, 14:20
First mirrors (FMs) for ITER optical diagnostics induce a number of specific requirements including low sputtering rate, high neutron/gamma radiation and thermal stability to keep the optical performance in the DT plasma shots. Additionally, the FM surface must withstand the discharges by a cleaning system aimed to eliminate Be deposits. A number of experiments have shown that the mirrors made...
Konstantin Vukolov
(Kurchatov Institute)
08/09/2016, 14:20
Silica-based optical fibers have a high light transmission in visible range and so they are widely used for transmitting the light from plasma to detectors in modern thermonuclear facilities. The fiber bundle is comprised as a rule of several tens or hundreds optical fibres of 100-500 microns diameter. The lifetime of the optical fiber in ITER should be more than 15 years. Radiation resistance...
Evgeny Andreenko
(NRC “Kurchatov institute”)
08/09/2016, 14:20
The performance of ITER Main Chamber H-alpha & Visible Spectroscopy is challenged by the problem of separating the contribution of visible light emitted in the scrape-off-layer (SOL) from the background of much higher intensity, produced by the divertor stray light (DSL) reflected by the all-metal first wall (S.Kajita, et al., PPCF, 2013). A differential (bifurcated-line-of-sight) measurement...
Vincent Martin
(Bertin Technologies)
08/09/2016, 14:20
The ITER equatorial visible and infrared wide-angle viewing system is a first plasma diagnostic that will be used to image the visible plasma boundary and the in-vessel components temperatures for real-time machine protection and plasma control purposes, as well as offline physics studies. The system will be installed in four equatorial ports and will have 15 lines of sight covering most of...
Sven Gutruf
(Kampf Telescope Optics)
08/09/2016, 14:20
The H-alpha and Visible Spectroscopy Diagnostic shall measure spectrally, spatially and temporally resolved emission of hydrogen isotopes and impurities in the ITER scrape-off layer. There are four H-alpha diagnostic channels, located in 3 port plugs.
In the current design status, all main interfaces have been iterated with the Port Integrator. All major subsystems, of this complete end to...
Arnd Reutlinger
(Kampf Telescope Optics)
08/09/2016, 14:20
The H-alpha and Visible Spectroscopy Diagnostic shall measure spectrally, spatially and temporally resolved emission of hydrogen isotopes and impurities in the ITER scrape-off layer. Four H-alpha diagnostic channels are designed to observe the plasma. They are located in 3 port plugs:
- Equatorial Port #11:
TV (Top View): poloidal wide Field of View (FoV) covering the upper part of the...
Matthew Smiley
(General Atomics MFE)
08/09/2016, 14:20
One of the diagnostic systems being provided by the US is the Upper Wide Angle Viewing System (UWAVS), which provides real-time, simultaneous visible and infrared images of the ITER divertor region via optical systems located in five upper ports. The UWAVS is designed in three main sections: in-vessel, interspace and port cell assemblies. Each assembly utilizes multiple steering and relay...
Eugene Mukhin
(Ioffe Institute)
08/09/2016, 14:20
ITER Divertor Thomson scattering (DTS) was discussed in a number of presentations and papers. The development of diagnostic equipment for ITER DTS is under way and coming to its conclusion. Choice and justification of lasers and polychromator design as well as first mirror protection are the focus of the presentation.
Q-switched Nd:YAG laser for DTS in ITER (1.064mm, 2J, 50Hz, 3ns) is...
YoungHwa An
(National Fusion Research Institute)
08/09/2016, 14:20
The local shielding design for the detector of ITER VUV Edge Imaging Spectrometer is optimized based on the MCNP calculation using a local port cell model of ITER Upper Port #18. A back-illuminated CCD, the envisaged VUV detector for ITER VUV Edge Imaging Spectrometer will be installed at ITER Upper Port #18 port cell region, in which a harsh radiation environment is expected with neutron flux...
Bastian Weinhorst
(Institute for Neutron Physics and Reactor Technology)
08/09/2016, 14:20
The Charge Exchange Recombination Spectroscopy (CXRS) diagnostic aims to measure emission lines of impurity isotopes in the ITER plasma in order to quantify several parameters like the composition of the plasma (density of helium, deuterium or tritium), the ion temperature or rotation velocities. The core plasma CXRS shall be installed in one of the ITER Upper Port Plugs (UPP #3). Currently,...
Valentina Huber
(Forschungszentrum Jülich GmbH)
08/09/2016, 14:20
Imaging systems are an indispensable technique for successful plasma operation of fusion devices. At the JET tokamak, numerous cameras in the VIS/NIR/MWIR spectral ranges are used for plasma physics studies as well as for the real time overheating protection of the first wall and for live plasma monitoring during operation. The protection system, on the basis of the NIR imaging cameras, is...
Gonzalo Farias
(Escuela de Ingenieria Electrica)
08/09/2016, 14:20
Huge databases are a common situation in fusion. Physical properties of plasma are studied by thousands of signals, sampled at very high frequencies, producing enormous amount of data. A medium-size nuclear fusion device such as TJ-II can generate discharges that last around 500 milliseconds, reaching up to 100 Mbytes per one simple shot. Larger fusion devices such as JET can produce 10Gbytes...
Yi Tan
(Department of Engineering Physics)
08/09/2016, 14:20
The noises of a tokamak during operations form the "voiceprint" of a tokamak. By installing a set of microphones in several optimized positions around the tokamak machine, most noises can be detected and can be used as the “voiceprint” of the tokamak for monitoring its status. Noises of a tokamak in discharge-ready status are mainly continuous and/or cyclical noises from pumping system, water...
Igor Nedzelskiy
(Instituto de Plasmas e Fusao Nuclear)
08/09/2016, 14:20
The heavy ion beam diagnostic of the tokamak ISTTOK is operated with a 20 keV Xe++ ion beam and a multiple cell array detector to collect the secondary Xe2+2+ ions created along the primary beam path by ionizing collisions with plasma electrons. In this multichannel mode of operation, the use of standard Proca-Green 30oo parallel plate energy analyzer for the...
Matti Laan
(Institute of Physics)
08/09/2016, 14:20
Laser induced breakdown spectroscopy (LIBS) is a promising tool for remote monitoring of erosion/deposition processes at the first wall of ITER. Proper application of LIBS requires knowing the ablation rates of co-deposited layers on plasma-facing components accurately to obtain elemental depth profiles of different elements on the layers from the recorded LIBS spectra. This goal is, however,...
Gennady Sergienko
(Forschungszentrum Jülich GmbH)
08/09/2016, 14:20
Deuterium-tritium gas mixture will be used as fuel in future fusion devises like ITER. Thus it is important to monitor hydrogen isotope ratios not only in fusion plasma and in the subdivertor/exhaust gases but also retained in the plasma facing components (PFC). Residual gas analysis is traditionally used to quantify the isotope species of the PFCs in the laboratory by means of thermal...
Andrzej Wojenski
(Institute of Electronic Systems)
08/09/2016, 14:20
This work refers to the currently being developed extended soft X-Ray plasma diagnostics system with the novel, radiation-hard generation of electronics and implemented algorithms. The system is based on the Gas Electron Multiplier detector. For the multichannel, modular systems working with very intense plasmas (e.g. laser generated plasma, plasma fluxes), the phenomenon of the coinciding...
Bernd Sebastian Schneider
(Institute for Ion Physics and Applied Physics)
08/09/2016, 14:20
The characterization of outward filamentary plasma transport in Medium-Size Tokamaks (MST) is an important objective of current fusion plasma research. We aim at improving the diagnostic of transport events in the Scrape-Off Layer (SOL) and further inside by means of various types of newly developed electrical probes combined with the associated probe measurement procedures. Presently, a New...
Tomasz Czarski
(Institute of Plasma Physics and Laser Microfusion)
08/09/2016, 14:20
The measurement system based on GEM - Gas Electron Multiplier detector is developed for X‑ray diagnostics of magnetic confinement tokamak plasmas. The multi-channel setup is designed for estimation of the energy and the position distribution of an X-ray source. The main measuring issue is the charge cluster identification by its value and position estimation. The fast and accurate mode of the...
Maryna Chernyshova
(Institute of Plasma Physics and Laser Microfusion)
08/09/2016, 14:20
Necessity to develop new diagnostics for poloidal tomography focused on the metal impurities radiation monitoring, especially tungsten emission, has become recently inevitable. Tungsten is now being used for the plasma facing material on many machines, including on the WEST project, where an actively cooled tungsten divertor is being implemented. This forced a creation of the ITER-oriented...
Evzen Losa
(Research Centre Rez)
08/09/2016, 14:20
The intended fusion reaction for ITER project is D + T → 44He (3.5 MeV) + 00n (14.1 MeV), which produces high energy neutrons. Portion of these neutrons is effectively captured in breeder blanket, however, many neutrons leak and can cause radiation damage. Monitoring of the neutron damage in ITER internals is necessary due to the aging management. 2323Na(n,2n)...
Milos Jirsa
(Institute of Physics ASCR)
08/09/2016, 14:20
Superconducting RE-BaCuO tapes of different suppliers were tested by magnetic induction (vibrating sample magnetometer, VSM) and by current transport techniques. The tests aimed at finding the best candidates for the tape utilization in a new generation of superconducting magnets for fusion reactors. The electromagnetic characteristics of the tapes as a function of temperature, magnetic field,...
Xinsheng Yang
(Southwest Jiaotong University)
08/09/2016, 14:20
As the only high-temperature superconductors (HTS) that can be made into round wires without anisotropy, Bi-2212 has significant potential applications as CICC (cable in conduit conductor) for large-scaled superconducting magnets in fusion reactors. However, Bi-2212 is brittle and sensitive to strain which leads to a low mechanical performance. The effort on studying the impact of strain on...
Jonathan Hollocombe
(Theory and Modelling)
08/09/2016, 14:20
The SAGE 2 2 European Horizon 2020 project (grant agreement 671500), led by Seagate with 10 partners, is investigating the needs of future exascale storage systems for data intensive applications. CCFE is one of the partners and SPECTRE (SPECtral Research Engine) is one of the tools being developed to take advantage of the improved data I/O and throughput capability of the SAGE...
Alexey Arkhipov
(Max-Planck Institute for Plasma Physics)
08/09/2016, 14:20
The ITER Diagnostic Pressure Gauges (DPG) shall provide the measurement of the neutral gas pressure, which is an important parameter for basic control of the operation of ITER machine as well as for input to physics models of the plasma boundary. The reference sensor is a hot cathode ionization gauge, which is able to operate in an environment with strong magnetic fields (up to 8 Tesla),...
Sebastian Friese
(Institut für Energie- und Klimaforschung)
08/09/2016, 14:20
The shutter mechanical concept for the ITER core plasma CXRS Fast Shutter is based on elastic bending of a deformable arm structure (length ≈ 1.8 m) which blocks or opens the path of plasma emitted light aiming at the diagnostics first mirror. Bending of the shutter arms is induced by an actuator and will be restrained using the limiting bumpers, where, although the arms are preloaded against...
Mathias Dibon
(Max-Planck-Institute for Plasmaphysics)
08/09/2016, 14:20
A disruption is a major plasma instability that follows a sudden loss of plasma energy. During such an event, large electromagnetic forces and high heat loads occur, as well as electrons at relativistic speed. These effects can cause damage to the plasma facing components and thus have to be mitigated. For this purpose high speed gas valves are used to inject a strong pulse of noble gas onto...
Nikola Jaksic
(Max Planck Institute for Plasma Physics)
08/09/2016, 14:20
“This project has received funding from the Euratom research and training programme 2014-2018”
In plasma fusion research the neutral gas density is usually measured using hot cathode ionisation gauges which are modified for the application in high magnetic fields and for a measurement range between 10-3-3 Pa and 20 Pa. For obtaining sufficient electron emission, high temperatures in...
Fang Liu
(Institute of Plasma Physics)
08/09/2016, 14:20
Bi2Sr2CaCu2Ox is a potential material for the superconducting magnets of the next generation of Fusion reactor. A R&D activity based on Bi2212 wire is running at ASIPP for the feasibility demonstration of CICC. One sub-size conductor cabled with 42 wires was designed and manufactured. A test method was designed and performed to measure the joints resistance and critical current of the Bi2212...
Pascal de Marne
(Max-Planck-Institut fuer Plasmaphysik)
08/09/2016, 14:20
Manipulators are an important tool to position diagnostics or samples near to the plasma without breaking the vacuum of fusion devices. They can be used for different purposes like measuring plasma parameters with electrical or magnetic probes near to the core plasma or to investigate plasma-wall interaction by exposing dedicated samples. ASDEX Upgrade is operating a set of manipulators, the...
Qin Zeng
(School of Nuclear Science and Technology)
08/09/2016, 14:20
Chinese Fusion Engineering Testing Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plant and to demonstrate generation of fusion power in China. In order to select the most suitable blanket proposal for CFETR, the three blanket concepts (i.e. the helium cooled solid breeder blanket, the liquid LiPb blanket, and the water cooled ceramic breeder...
Weibin Xi
(Tokamak Design Division)
08/09/2016, 14:20
The original EAST magnet feeders have been operated for over 7 years since 2006. With the improvement of experimental parameters, a new magnet feeder system has been designed for the upgrade project of the EAST. It consists of 13 pairs of superconducting bus-lines with total length over 900 m and 13 pairs high temperature superconducting current leads. Each original bus-line connecting new...
Diogo Eloi Aguiam
(Instituto de Plasmas e Fusão Nuclear)
08/09/2016, 14:20
The new multichannel X-mode reflectometer installed on ASDEX Upgrade measures the plasma density profile evolution at different positions in front of the ICRF antenna. The reflectometer operates in the extended U-band (40–68 GHz) microwave region, measuring density profiles up to 101919 m-3-3 with magnetic fields between 1.5 T and 2.7 T. In this heterodyne reflectometer...
Fabio Pollastrone
(FSN (Nuclear Fusion and Fission and Related Technologies Department))
08/09/2016, 14:20
The electrical pattern recognition can be useful in several applications, generally it is used to detect particular events or anomalies in the signal under analysis or to identify precursors, especially in electrophysiology. Each application requires customized algorithms and appropriate signal processing capabilities. In this paper we present an application of pattern recognition to real-time...
Markus Teschke
(E1 - tokamak scenario development)
08/09/2016, 14:20
BUSSARD is a new inverter system at the nuclear fusion experiment ASDEX Upgrade for mitigation of ELMs and execution of other, physics related experiments. The concept and first results were presented in detail [1]. Four-phase operation was routinely done during shot campaign 2015/16 and many experience in operation was gained. Now, the completion of BUSSARD is almost finished and many...
Nils Arden
(Max-Planck Institute for Plasma Physics)
08/09/2016, 14:20
Recently an inverter system (called BUSSARD) was assembled to individually feed the 16 in-vessel saddle coils of the fusion experiment ASDEX Upgrade (AUG).The new inverter system consists of 16 inverters, each with an output current of up to 1.3 kA and a bandwidth of up to 500 Hz in arbitrary waveforms. Currently, the system is in operation with 4 inverters feeding four in serial connected...
Wang HaiBing
(Center for Fusion Science)
08/09/2016, 14:20
Study on 300MVA pulse generator starting system
HaiBing Wang, WeiMin Xuan, JianFei Peng, HuaJun Li, LiRong Xu, HaoTian Hu, li Kang
Southwestern Institute of Physics, Chengdu, Sichuan, China
For supplying power for HL-2M Tokamak, a new 300MVA pulse generator has been developed. The new generator with 400 tons of rotor to stored energy will be driven by an 8500kW asynchronous motor. The...
Jianfei Peng
(Tokamak Power Supply Division)
08/09/2016, 14:20
A new motor generator (MG) system is building mainly for the poloidal field power supply system of the HL-2M Tokamak. This MG system will be capable of providing a peak capacity of 300 MVA and delivering up to 1350 MJ per pulse at 15 min intervals. The system consists of a 300 MVA MG and its auxiliary systems. The MG adopts the semi umbrella vertical shaft type and consists of an 8500kW...
Shouzhi Wang
(Department of Engineering Physics)
08/09/2016, 14:20
A high voltage power supply (HVPS) used for the ECRH system on the SUNIST tokamak is introduced. It is able to output a 50 ms pulse of -40 kV / 15 A in every 5 minutes. The voltage drop for the whole flat top is less than 2%. In each arcing events, the maximum energy delivered to the load is less than 15 Joules.
The HVPS is based on Marx Generator and PSM technologies using fast switch...
G. Pintsuk
(Forschungszentrum Jülich GmbH)
08/09/2016, 14:20
The WEST (W -for tungsten- Environment in Steady-state Tokamak) project is based on an upgrade of Tore Supra tokamak. ITER-like actively cooled tungsten targets (monoblocks) will be integrated in the lower divertor and a new set of actively cooled tungsten coated plasma facing components will cover a part of the vessel to provide a fully metallic environment.
In preparation of the production...
Youngjae Park
(Department of Nuclear Engineering)
08/09/2016, 14:20
Development of reliable high heat flux removal techniques is an important issue to design plasma facing components in a fusion reactor. The ITER-like divertor cooling design based on water-subcooled flow boiling is one of the well-developed divertor cooling schemes. To withstand such a high heat flux in the vertical target of the ITER divertor, a twisted tape is inserted into a CuCrZr tube...
Kyung-Min Kim
(National Fusion Research Institute)
08/09/2016, 14:20
It is so important that the bonding technology between tungsten and dissimilar metals for the PFC of ITER and DEMO. The development of tungsten brazing technology was first launched for the KSTAR PFC.
Flat type tungsten block was brazed on CuCrZr in vacuum at a temperature of 980 °C for 30 minutes using silver free brazing alloy. A OFHC-copper was used as an interlayer between tungsten and...
Dong Jun Kim
(Korea Atomic Energy Research Institute)
08/09/2016, 14:20
Tungsten coated mock-ups for developing the Plasma facing component (PFC) werefabricated and tested in the plasma torch and high heat flux test facility with electron beam,which can be used in the repair of the damaged PFCs. For evaluating the life-time of the tungsten coated mock-up, the erosion rate was measured and thermal-lifetime analyses were performed with the fabricated mock-up. And...
Suk-Kwon Kim
(Nuclear Fusion Engineering Development Division)
08/09/2016, 14:20
The Developments of plasma facing components (PFCs) are the key items for the nuclear fusion reactors. The most components for the tokamak PFCs are the blanket first wall, divertor, heating ports, and diagnostics ports. These PFCs are composed of the armour materials, the heat sink for the cooling, and the structural materials. Be, W, C-composites, and advanced materials were selected for...
Shenghong Huang
(Modern mechanics)
08/09/2016, 14:20
After years of exploration and development, research of magnetic confinement nuclear fusion is progressed into stage of experimental fusion reactor construction and test. As a key plasma-facing component, the anti-fatigue performance of first wall of fusion reactor receives widely concerns. Due to the fact of enduring both periodic loads of pulse operating mode and shock loads of transient...
Rajamannar Swamy Kidambi
(Divertor & First Wall Technology Development Division)
08/09/2016, 14:20
This paper deals with the design of High Pressure High Temperature Water Circulation System (HPHT-WCS) for High Heat Flux Test Facility (HHFTF) of IPR and its related thermal hydraulic experiments. HHFTF has been established at IPR, India for testing performance of plasma facing components under intense heat loads expected in plasma fusion devices. Plasma facing components of the present day...
Kohei Hamaguchi
(Division of Sustainable Energy and Environmental Engneering)
08/09/2016, 14:20
It is desirable to develop tungsten (W) diverter in Tokamak-type nuclear fusion reactor including the International Thermonuclear Experimental Reactor (ITER). W has the highest melting point in all metals and thus is a promising material of the diverter. Since the diverter will repetitively undergo high heat flux of 100MW/m2 2 at least in a few tens of millisecond or less when...
Ryuji Ohsone
(Japan Atomic Energy Agency)
08/09/2016, 14:20
A hot isostatic pressing(HIP) method is one of the candidate process to fabricate the fusion blanket the first wall with built in cooling channels. Thin plates and rectangular tubes made of reduced activation ferritic/martensitic (RAFM) steel, such as F82H, are consolidated by the HIP method. The first wall quality therefore depends on the integrity of the formed HIP joint. In laboratory scale...
Toshikio Takimoto
(Tokai University)
08/09/2016, 14:20
In the magnetic confinement fusion reactor for high power and long pulse operation, enormous heat flux (exceeding 10 MW/m22) is expected to flow onto divertor plates from core plasma. In order to reduce this heat load, the divertor geometry on stationary detached plasma formation must be realized. In addition, the neutral particle flowback into the core plasma is necessary to...
Arnold Lumsdaine
(Oak Ridge National Laboratory)
08/09/2016, 14:20
One of the critical challenges for the development of next generation fusion facilities, such as a Fusion Nuclear Science Facility (FNSF) or DEMO, is the understanding of plasma material interactions (PMI). The field of PMI occurs at the intersection of plasma physics, materials science, and engineering, and requires expertise and research and development in each of these fields. Making...
Keith Smith
(Materion Beryllium and Composites Elmore, OH, United States)
08/09/2016, 14:20
In its current design, the ITER fusion machine will use tens of thousands of beryllium tiles as plasma-facing components in its First Wall. S-65 is one of three grades of beryllium which has been accepted by the ITER International Organization for use in the reactor. The beryllium material for ITER has to pass through many machining and manufacturing processes after being consolidated by...
Mizuki Noguchi
(Advanced Energy Engineering Science)
08/09/2016, 14:20
It is important to understand tritium (T) desorption behavior from plasma-facing materials of a fusion reactor in order to discuss tritium recovery method from in-vessel components. Tungsten (W) is a candidate material for plasma-facing components. Although a sputtering rate of W by hydrogen isotopes is low, a certain amount of W deposition layer will be formed on plasma-facing wall. In this...
Irina Tazhibayeva
(Insitute of Atomic Energy NNC RK)
08/09/2016, 14:20
Tritium is a prospect fuel material for future fusion power reactors, thus tritium breeding in these reactors is one of the design challenges, which can be solved by using the lithium-containing materials for contrstruction of the reactors’ blankets. Also of great interest is use of lithium as a plasma-facing material, for example, in the form of lithium-capillary porous systems (CPS). Such...
Alexey Popkov
(Plasma Physics Department)
08/09/2016, 14:20
Lithium is considered as a promising material for plasma-facing components (PFC) in future fusion devices. A number of experiments have already demonstrated positive effects of lithization and using of Li based PFCs on plasma operation. During operation of the machine, lithium is deposited on the surrounding walls and in shadowed areas. One can expect a high concentration of hydrogen isotopes...
Fumitaka Ishikawa
(Tokai University)
08/09/2016, 14:20
Tungsten is important candidates for plasma-facing component applications on the development of magnetic fusion reactors. Particularly, it is important to understand the behavior of hydrogen isotopes in tungsten of the diverter wall material. In this study, we have performed the irradiation experiments using deuterium and helium mixed plasma in order to investigate the deuterium retention and...
Daniel Iglesias
(UKAEA-CCFE)
08/09/2016, 14:20
Virtual prototyping enhances traditional engineering analysis workflow when a quick evaluation of complex load cases is required. During design, commissioning or operating phases, components can be virtually tested in realistic conditions by using previously validated numerical models and experimental databases.
Three complementary applications have been developed under this approach for the...
Masayuki Tokitani
(Department of Helical Plasma Research)
08/09/2016, 14:20
The study is focused on modification of surfaces of the tungsten-coated divertor tiles used in the first campaign (2011-2012) of the JET tokamak with the ITER-lLike Wall (JET-ILW). The analyses by means of several material research techniques have been carried out at International Fusion Energy Research Centre (IFERC), JAEA Rokkasho.
Samples, in the form of disks (17 mm in diameter), extracted...
Aleksander Drenik
(EUROfusion Consortium)
08/09/2016, 14:20
After the transition to full metal wall configurations at AUG and subsequently at JET, impurity seeding became necessary to maintain the divertor heat loads below material limits in H-mode discharges. Among the studied impurities, nitrogen (N) was found to be the most favourable option. However, it was also found that N2-seeding leads to formation of ammonia (NH3). Nitrogen and NH3 retained in...
Rudolf Neu
(Plasmarand und Wand)
08/09/2016, 14:20
Since 2014 ASDEX Upgrade (AUG) is using bulk tungsten tiles at the outer divertor strike-point. In two experimental campaigns more than 2000 plasma discharges with up to 10 s duration and 100 MJ plasma heating were successfully conducted, without impairment by the W tiles. However, an inspection after the campaigns revealed that a large number of tiles suffered from deep cracking, mostly...
Johan Oosterbeek
(Eindhoven University of Technology)
08/09/2016, 14:20
Diagnostic systems are essential for the development of ITER discharges and to reach the ITER goals. Many of these diagnostics require a line of sight to relay signals from the plasma to the diagnostic, typically located outside the torus shall. Such diagnostics then require vacuum windows that isolate the torus vacuum and crucially ensure tritium containment. While such windows are routine in...
Hun-Chea Jung
(ITER Korea)
08/09/2016, 14:20
The ITER blanket shield block (SB) is one of the in-vessel components, which is designed to provide nuclear shielding and to supply the cooling water to vacuum vessel and external component. The ITER SB is classified the VQC 1A as vacuum classification and its manufacturing process and cleaning procedure shall comply with the ultra-high vacuum conditions necessary for machine operation and...
Paul Edwards
(Tokamak Engineering Department)
08/09/2016, 14:20
The Final Design Review for the Blanket Manifold (BM) was successfully held in December 2015. Since the Conceptual Design Review, a concerted effort has been necessary on finalisation of the multi-pipe design, verification by analysis and practical validation to address challenging design requirements, and installation/maintenance processes.
During normal operating conditions the BM provide...
Yongbo Wang
(Lappeenranta University of Technology)
08/09/2016, 14:20
For ITER or the future DEMO remote maintenance system (WPRM), several types of special tailored automatic manipulators are needed for vacuum vessel (VV) component transportation, inspection, and removal from and replacement to the VV wall. These tailored manipulators, such as Multi-purpose Deployer, Articulated Inspection Arm (AIA), Diverter Cassette Mover etc., should be calibrated with very...
Takahito Maruyama
(Department of ITER Project)
08/09/2016, 14:20
How to recover from failures of components in radiation environment is an important issue of the ITER remote handling systems. Recovery operations of the remote handling systems must be performed remotely due to limitation of human access. For the ITER Blanket Remote Handling system, failure modes have been analysed, and the analysis has concluded that electrical failures of actuators, which...
Yuto Noguchi
(Fusion Research and Development Directorate)
08/09/2016, 14:20
The ITER blanket module has hydraulic connections to the cooling water manifold. The connections are designed to be cut and re-welded remotely in the vacuum vessel during blanket maintenance due to irradiation of in-vessel components after D-T experiment. In course of the R&D activities for in-vessel pipe welding, a study [1] demonstrated that good weld quality can be achieved by correcting...
Naveen Rastogi
(Remote Handling Division)
08/09/2016, 14:20
An integrated control system architecture has been defined for the implementation of ITER Remote Handling (RH) equipment systems. The RH Core System (RHCS) is a standard software platform used for the development of ITER RH equipment controller applications to facilitate the integration with this system. It installs on top of the CODAC core system and provides a uniform platform for the...
Jean-Pierre Friconneau
(ITER)
08/09/2016, 14:20
ITER is a large scale fusion device designed to study the high temperature fusion reaction between tritium and deuterium. The success of a tokamak-type fusion reactor will depend to a great extent on developing reliable and safe methods of carrying out routine maintenance and repairs remotely.
Remote Handling System (RHS) are used to perform remotely the maintenance of the vacuum vessel. They...
50252.
P4.133 Irradiation tests of radiation hard components for ITER blanket remote handling system
Makiko Saito
(Naka Fusion Institute)
08/09/2016, 14:20
The ITER Blanket Remote Handling System (BRHS) will handle the blanket modules (BMs), which can weigh up to 4.5 ton and be larger than 1.5 m, stably and with a high degree of positioning accuracy. When the ITER has stopped plasma operations for maintenance, the BRHS will be installed in the vacuum vessel, whose components are radioactive, to remove and install the BMs. Therefore, the BRHS will...
Bingyan Mao
(Laboratory of Intelligent Machines)
08/09/2016, 14:20
In the ITER or the future DEMO reactor systems, due to the neutron activation, the remote handling tasks such as inspection, repair and/or maintenance of in-vessel and ex-vessel components must be carried out using a wide variety of special tailored automatic manipulators. The structure of these manipulators can be designed as a pure serial articulated arm or a pure parallel mechanism, but for...
Paulo Carvalho
(Instituto de Plasmas e Fusao Nuclear)
08/09/2016, 14:20
Experimental fusion reactors aim at the exploration of the nuclear fusion as a viable energy resource. Remote Handling Systems (RHS) are specially designed for regular operations of inspection and maintenance inside the reactors, such as the In-Vessel Transporter, an extendable robotic arm deployed in the equatorial level of ITER. The reactor is shutdown during the installation and operation...
Huapeng Wu
(Lappeenranta university of technology)
08/09/2016, 14:20
The EAMA robot is a long slender arm for tokamak inspection and maintenance. In such conditions, grasp techniques ignoring or trying to avoid contact with the components of the vacuum chamber brings bottlenecks on the system control. During the grasping and releasing objects the contact with vacuum chamber is a critical condition for providing robust and achievable solutions of robot control....
Tom H. Owen
(Remote Applications in Challenging Environments (RACE))
08/09/2016, 14:20
Mascot is a two-armed dexterous master-slave telemanipulator device linked by force-reflecting servomechanisms, giving the operator a tactile sensation of doing the work. Mascot version 4.5 is currently in use at the Joint European Torus (JET) experimental nuclear fusion facility. Its role is to maintain the inside of the reactor vessel without the need for manned entry. The slave is...
Wang Rui
(Institute of Plasma Physics Chinese Academy of Sciences)
08/09/2016, 14:20
Full penetration welding and 100% volumetric examination of weld joints are strictly required for all welds of pressure retaining parts of the CFETR Vacuum Vessel (VV) according to the design manual. However not every welding joint can be tested using RT method due to component structure and welding position. Therefore, the ultrasonic testing (UT) has been selected as an alternative...
Jianguo Ma
(Institute of Plasma Physics)
08/09/2016, 14:20
With the development of CFETR engineering design, a full-scale sector prototype of vacuum vessel has been carried out as one of the major R&D projects. The welding structure between vacuum vessel sectors in field assembly is modeled in this prototype, and NG-TIG is taken for an applicable welding strategy with small welding deformation, high-quality welds and excellent adaptability to the...
Zhihong Liu
(instititue of plasma physics chinese academy of sciences)
08/09/2016, 14:20
Chinese Fusion Engineering Testing Reactor (CFETR) is a superconducting magnet Tokamak, it has the equivalent scale with complementary function to International Thermonuclear Experimental Reactor (ITER). The vacuum vessel (VV) which has a double-layer structure，Cooling water circulating through the double-layer structure will remove the heat generated during operation. The VV will provides a...
Kanetsugu Isobe
(Japan Atomic Energy Agency)
08/09/2016, 14:20
In the one of Broader Approach (BA) activities aiming to the development for a DEMO fusion reactor, the R&D of tritium technology has been carried from 2007. The period consists of Phase 1 (2007-2010) and Phase 2 (2010-2016). International Fusion Energy Research Center (IFERC) including DEMO R&D building was constructed in Rokkasho BA site of Japan. The R&D building is a facility to handle...
Juro Yagi
(Department of Helical Plasma Research)
08/09/2016, 14:20
One of the major concerns for molten salt breeding blanket system is the low tritium solubility, high equilibrium tritium pressure in other words, of the molten salts including FLiBe, FLiNaBe and FLiNaK. Owing to this, vanadium alloy (V-4Cr-4Ti) has been thought to be inappropriate as a structure material in molten salt breeding blanket because of its high tritium solubility.
The concept of...
Youhua Chen
(University of Science and Technology of China)
08/09/2016, 14:20
The neutron multiplier and the tritium breeder materials are made into millimeter-sized particles and arranged in the solid breeder blanket. Helium (mixed with 0.1% content of H2) is used as the purge gas to sweep tritium out when it flows through the pebble beds. Previous research shows that binary pebble beds present a better performance in tritium breeding than unitary pebble beds. Since...
Benedikt J. Peters
(Institute for Technical Physics)
08/09/2016, 14:20
The effect of superpermeability is capable of separating hydrogen and its isotopes out of gas mixtures at low pressures even against a pressure gradient. This process allows strongly enhanced permeation. It relies on metal membranes that are exposed to atomic hydrogen. If the surface inhibits the chemisorption on its surface, the atomic hydrogen can still enter the bulk, but hydrogen...
Karine Liger
(CEA Cadarache)
08/09/2016, 14:20
In the framework of tritium recovery from tritiated water, efficiency of packed bed membrane reactors have been successfully demonstrated. Thanks to protium isotope swamping, tritium can be recovered from tritiated water under the valuable Q2 form (Q = H, D or T) by means of isotope exchange reactions occurring on catalyst surface. The use of permselective Pd-based membrane allows withdrawal...
Yasunori Iwai
(Department of Blanket Systems Research)
08/09/2016, 14:20
Effect of halogenated gas on detritiation efficiency of the detritiation system was investigated. In order to accelerate tritium safety of the Japanese DEMO reactor, the detritiation system should be designed taking possible off normal events such as fire carefully into consideration. In an event of fire in a tritium processing room, halogenated gases such as hydrogen chloride, halogenated...
Kwangjin Jung
(University of Science and Technology (UST))
08/09/2016, 14:20
The hydrogen isotope storage and delivery system (SDS) is a complex system that includes many individual components. One of the most important parts of the SDS is a metal hydride bed, which stores and delivers the hydrogen isotopes and pure gases required for a nuclear fusion reactor. We have been developing a metal hydride bed using depleted uranium (DU). The hydrogen delivery performance of...
Yeanjin Kim
(quantum energy chemical engineering)
08/09/2016, 14:20
The hydrogen isotope storage and delivery system (SDS) is a part of a nuclear fuel cycle. It is a complex system that is composed of numerous components such as a metal hydride bed, measuring tank, and other essential components. Depleted uranium (DU) was chosen as a hydrogen isotope storage material because of its rapid reactivity. We designed and manufactured the DU hydride bed to store the...
Alina Niculescu
(National Institute for Cryogenics and Isotopes Technologies - ICSI)
08/09/2016, 14:20
Cryogenic distillation (CD) process is being employed, among other applications, in tritium separation technologies and in case of ITER is one of the key proceses in the fuel cycle. The ITER Isotope Separation System has to process by cryogenic distillation various mixtures of H-D-T depending from the various torus operation scenarious.
Cryogenic distillation has also been employed to separate...
George Ana
(National Institute for Cryogenics and Isotopes Technologies - ICSI)
08/09/2016, 14:20
During normal operation of a CANDU reactor, large amounts of tritiated heavy water is being produced as result of neutron absorption by the heavy water used as moderator and cooling agent. Tritium in the heavy water, being radioactive, brings a significant contribution to the personal doses and also represents an environmental hazard if a waterspill occurs.
The Pilot Plant for T2 and D2...
Tao Jiang
(Center for Fusion Science)
08/09/2016, 14:20
Being part of the ITER fuelling system, the primary functions of the Gas Injection System (GIS) include providing gases for plasma discharge, wall conditioning, and neutral beam injectors. The Gas Distribution System(GDS) is a key sub-system of the GIS, which shall distribute gases obtained from the Tritium Plant, to the Gas Valve Boxes for the Pellet Injection System, Gas Fuelling System,...
Francesca Bombarda
(FSN Department)
08/09/2016, 14:20
The injection of cryogenic pellets from the low field side (LFS) has long been in use for core fueling of fusion devices. However, with higher plasma temperatures and bigger sizes, this technique becomes increasingly inadequate to ensure effective core particle deposition; injection from the high field side (HFS) has shown better results, despite the severe limitations imposed to the pellet...
Dimitris Valougeorgis
(Mechanical Engineering)
08/09/2016, 14:20
Recently, an integrated software algorithm for modeling gas distribution systems operating under vacuum conditions has been developed [1]. It has been successfully applied to model the 2012 ITER divertor pumping system and results have been provided for the flow patterns in the cassettes and the divertor ring, as well as for the throughputs in the burn and dwell phases. In all cases the input...
Antonio Frattolillo
(ENEA C.R. Frascati)
08/09/2016, 14:20
Core fuelling of DEMO fusion reactor is under investigation within the EUROfusion Work Package "Tritium, Fuelling and Vacuum". An extensive analysis of fuelling requirements and technologies, suggests that pellet injection still represents, to date, the most realistic option. Modelling of both pellet penetration and fuel deposition profiles for different injection locations, assuming a...
Silvio Giors
(Plant Engineering Department)
08/09/2016, 14:20
The ITER vacuum system, one of the largest and most complex vacuum systems ever to be built, will use first of a kind cryopumps to provide high vacuum conditions to the torus vessel, cryostat vessel, and neutral beam injectors. In order to evacuate the high gas flows required by the plasma scenarios, the cryopumps will need sequential regenerations with unprecedented high frequencies.
The...
Ranjana Gangradey
(Development of cryopump and pellet injector system)
08/09/2016, 14:20
Indigenous cryoadsorption cryopump with large pumping speeds gases like hydrogen and helium is developed and a set of experiments performed at the Institute for Plasma Research (IPR). India. Towards its successful realization, technological bottlenecks were identified, studied and resolved. Hydroformed cryopanels were developed from concept leading to the design and product realization with...
Thomas Giegerich
(Institute for Technical Physics)
08/09/2016, 14:20
The reduction of tritium inventories is a key challenge for DEMO and future fusion power plants. As large amounts of tritium have to be processed in the inner fuel cycle, an inventory-optimized vacuum pumping process – the KALPUREX process – has been developed at KIT. Here, continuously working and non-cryogenic vacuum pump trains will be used in order to keep the tritium residence times and...
Jordi Abella
(Analytical and Applied Chemistry)
08/09/2016, 14:20
Accurate and reliable tritium management is of basic importance for the correct operation conditions of the blanket tritium cycle. The determination of the hydrogen isotopes concentration in liquid metal is of high interest for the blanket correct design and operation. Sensors based on solid state electrolytes can be used to that purpose. It is worth mentioning that these type of sensors offer...
Luigi Candido
(Department of Energy)
08/09/2016, 14:20
A crucial issue for the design of HCLL (Helium Cooled Lead Lithium) Test Blanket Module of ITER and HCLL, WCLL, DCLL Breeder Blanket of DEMO is to efficiently characterise the tritium inventory inside the blanket and the permeation of tritium into the coolant in order to reduce as much as possible the radiological hazard towards the external environment. A fast and reliable sensor is required...
Yuki Edao
(Department of Blanket Systems Research)
08/09/2016, 14:20
Various methods of tritium measurement have been applied depending on a chemical formof tritium. A method combined oxidation catalyst and water bubblers has been used as one of the most quantitative analysis methods for gaseous tritium. We previously developed a quantitative analysis system to measure gaseous tritium in a high accuracy using by an organic-based hydrophobic platinum catalyst....
Ivo Carvalho
(Instituto de Plasmas e Fusão Nuclear)
08/09/2016, 14:20
As part of the JET Programme in Support for ITER, campaigns with pure Tritium-Tritium (TT) fuel and Deuterium-Tritium (DT) mixture are planned at JET. Unlike the previous DT campaign at JET, these campaigns require a much higher tritium flow rate, particularly, the TT campaign can require up to 3.7 grams of tritium on a single pulse. Five tritium introduction modules (TIMs) fed from the Active...
Oliver Leys
(Institute for Applied Materials)
08/09/2016, 14:20
Advanced tritium breeder pebbles, composed of lithium orthosilicate with additions of lithium metatitanate as a secondary strengthening phase, are produced using a melt-based process. Synthesis powders are heated to high temperatures in a platinum alloy crucible, forming a melt, which is then ejected through a nozzle to form a laminar jet. Longitudinal surface instabilities cause the...
Kuo Tian
(Karlsruhe Institute of Technology)
08/09/2016, 14:20
As the complementary work of IFMIF-EVEDA (International Fusion Material Irradiation Facility- Engineering Validation and Engineering Design Activities) project, WPENS (Work Package Early Neutron Source) project in the framework of EUROfusion activities is committed to the engineering design of an IFMIF-DONES (Demo Oriented Neutron Source) facility, which is an accelerator based intense...
Sachiko Yoshihashi
(Department of Applied Nuclear Technology)
08/09/2016, 14:20
In the international fusion materials irradiation facility (IFMIF), 14 MeV neutrons are generated by 40 MeV deuteron beam injection into a high-speed liquid lithium (Li) plane jet, flowing along a vertical concave wall in vacuum. Measurement of a free surface flow and fluctuation of the thickness are required to produce a stable neutron field and maintain the safety of Li target system.In...
Sergej Gordeev
(Institute for Neutronic Physics and Reactor Technology)
08/09/2016, 14:20
The configuration of the Early Neutron Source (ENS) is the IFMIF-DONES (DEMO Oriented Neutron Source) approach, based on an IFMIF-type neutron source. It aims providing an intense fusion-like neutron spectrum with the objective to qualify on an accelerated time scale structural materials to be used in the future DEMO fusion reactor. IFMIF-DONES is based on the interaction of single 40MeV 125mA...
Hiroo Kondo
(Japan Atomic Energy Agency)
08/09/2016, 14:20
A liquid-Li free-surface stream flowing at 15 m/s under a high vacuum of 1E−3 Pa is to serve as a beam target (Li target) for the planned International Fusion Materials Irradiation Facility (IFMIF). The Engineering Validation and Engineering Design Activities (EVEDA) for the IFMIF are implemented under the Broader Approach. As a major activity of the Li target facility, the EVEDA Li test loop...
Eiji Hoashi
(Osaka University)
08/09/2016, 14:20
A high-speed liquid metal lithium jet (Li jet) with a free surface is planned as a target irradiated by two deuteron beams to generate a neutron field in an accelerator based neutron source, such as that in the international fusion materials irradiation facility (IFMIF). In the IFMIF, it is desirable to stabilize the Li jet for the efficiency of the neutron generation and the safety of...
Takafumi Okita
(Division of Sustainable Energy and Environmental Engineering)
08/09/2016, 14:20
Liquid metal flow has been expected to be applied in various fields. For example, sodium and lithium (Li) are applied as a coolant in the fast-breeder reactor and space nuclear reactor, Li jet as a beam target in the International Fusion Materials Facility (IFMIF) and as a charge stripper in Radioactive Isotope Beam Facility (RIBF) at RIKEN, lithium-lead (Li-Pb) as a liquid metal blanket in a...
Georg Schlindwein
(Institute for Neutron Physics and Reactor Technology (INR))
08/09/2016, 14:20
The so called High Flux Test Module (HFTM) represents the component of IFMIF (International Fusion Irradiation Facility) in which material specimens are being placed that accumulate the highest neutron induced damage rates (≥20 dpa/fpy). Damage rates of this magnitude are limited to a volume of ~500 cm³ (attenuation in beam direction) behind a beam footprint of 20x5 cm. The high flux region of...
Shotaro Matsuda
(Division of Sustainable Energy and Environmental Engineering)
08/09/2016, 14:20
International Fusion Material Irradiation Facility (IFMIF) is the facility generating the high flux and high energy neutron to develop a material for a nuclear fusion reactor. In the IFMIF, high-speed liquid lithium (Li) jet is used as the target irradiated by two deuteron beams. Since the Li jet must flow with high velocity for the heat removal, it is important to research on the...
Yuefeng Qiu
(Karlsruhe Institute of Technology)
08/09/2016, 14:20
The location of the lithium quench tank (QT) is an important safety related issue in the design of the test cell (TC) of the IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented Neutron Source). In the current reference design, the QT is situated outside the TC and is connected to the target assembly through a long lithium outlet channel penetrating the TC floor....
Florian Schwab
(Institute for Neutron Physics and Reactor Technology (INR))
08/09/2016, 14:20
The High Flux Test Module (HFTM) of the International Fusion Materials Irradiation Facility (IFMIF) is a device to enable irradiation of Small Scale Testing Technique (SSTT) specimens by neutrons up to a structural damage of 50 displacements per atom (dpa) in an irradiation campaign of one year. The IFMIF source generates neutrons with a D-T-fusion-relevant energy spectrum and a flux to...
Giuseppe Pruneri
(Fusion department)
08/09/2016, 14:20
The Conventional Facilities of the Linear IFMIF Prototype Accelerator (LIPAc)
Authors
G.Pruneri, P.Cara, R.Heidinger, A. Kasugai, J. Knaster, S. Ohira, Y.Okumura, K.Sakamoto, and the LIPAc Integrated Project Team.
The International Fusion Material Irradiation Facility (IFMIF) aims at qualifying and characterising materials capable to withstand the intense neutron flux originated in the D-T...
Pedro Ortego
(Neutronic Calculations)
08/09/2016, 14:20
In the conceptual design of the beam dump shielding for the foreseen fusion-relevant irradiation facility IFMIF, an inner lead cylinder performs the shielding of the highly activated copper cone undergoing the deuteron beam bombardment and low-alloy steel is used for front shielding. In order to reduce the residual dose around the beam dump at beam-off conditions and dose at hands-on...
Yuki Iwama
(Department of Sustainable Energy and Environmental Engineering)
08/09/2016, 14:20
It is desirable to develop liquid lithium-lead (Li-Pb) blanket for helical-type fusion reactor because of its high cooling and tritium-recovering abilities. Since heat transport under a strong magnetic field in a fusion reactor determines the performance of liquid metal blanket (LMB), it is important to clarify the mechanism of the interaction between Li-Pb flow and the magnetic field. On the...
Masatoshi Kondo
(Research Laboratory for Nuclear Reactors)
08/09/2016, 14:20
The development of functional layers such as the tritium permeation barrier and the anti-corrosion barrier is one of the important issues for the development of liquid breeder blanket. The functional layers with the self-healing function have been developed based on the mechanism of the oxide layer formation. The oxides of yttria (Y2O3) and zirconia (ZrO2) have an excellent chemical stability....
Daniele Martelli
(Department of Civil and Industrial Engineering)
08/09/2016, 14:20
The use of PbLi and RAFM steels in blanket applications requires a better understanding of material compatibility related to physical/chemical corrosion phenomena in the 450-550°C temperature range. The impact of corrosion includes deterioration of the mechanical integrity of the blanket structure due to the wall thinning. Furthermore, serious concerns are associated with the transport of...
Gorka Alberro
(Nuclear Engineering and Fluid Mechanics)
08/09/2016, 14:20
The importance of the hydrogen isotopes transport parameters of Sieverts’ constant and diffusivity in the eutectic lead lithium alloy is well known, as long as it is vital for the determination of tritium management strategies at liquid-metal breeding blanket systems [Helium Cooled Lithium Lead (HCLL), or Dual-Coolant Lead-lithium (DCLL)].
Tritium transport parameters as solubility and...
Sergi Colominas
(Analytical Chemistry)
08/09/2016, 14:20
Lithium 6 is the substance required to generate in-situ tritium in fusion reactors. Because of that, lithium monitoring in lithium-lead eutectic (Pb-15.7Li) is of great importance for the performance of the liquid blanket. Lithium measurements will be required in order to proof tritium self-sufficiency in liquid metal breeding systems. On-line lithium sensors must be designed and tested in...
Jiang Haiyan
(School of Materials Science and Engineering)
08/09/2016, 14:20
In this study, rotating experimental devices were built to investigate the compatibility of the fusion reactor materials RAFM steel, 316L(N) steel，CuCrZr alloy with the Al2O3–water nanofluids. Based on the ITER water-cooling program，the experimental condition parameters were fluid velocity of 1.13 and 3.71m/s，fluid temperature of 70±1◦◦C，testing duration of 2136h，nanofluid mass...
Fabio Tieri
(Fusion Nuclear Tecnologies)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The ASTEC code is a lumped parameter code originally designed to perform safety analysis in fission nuclear power plants. Recently some modules of ASTEC have been modified by IRSN to be applicable for the safety analysis in the nuclear fusion plants. In particular the CPA module ( for the thermal-hydraulics of the containment) and the SOPHAEROS module (to model the physical phenomena...
Jean-Francois Ciparisse
(Industrial Engineering)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
One of the main concerns in Tokamak operation is the dust resuspension and fallout in case of LOVA (Loss Of Vacuum Accident) and LOCA (Loss Of Coolant Accident), as the metallic powders contained in the vessel are radioactive and therefore harmful. Furthermore, they can react explosively with the incoming oxygen if the local composition falls inside the flammability interval and if a hot point...
Luigi Antonio Poggi
(Industrial Engineering)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
STARDUST-U facility is an experimental facility voted to help the scientific community to better understand the problem of dust re-suspension and mobilization in case of Loss Of Vacuum Accidents (LOVAs) or Loss Of Coolant Accidents (LOCAs) inside the next generation fusion reactors like the International Thermonuclear Reactor (ITER) or the Demonstration Power Plant (DEMO).In this work the...
Andrea Malizia
(Industrial Engineering)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The future nuclear plants like ITER, DEMO or PROTO are interested by the problems of dust creation and resuspension. Radioactive dust, if resuspended by accidents in the vacuum vessel, can be dangerous because of its toxicity and capacity to explode under certain conditions. The authors have been working since 2006 on dust resuspension problems through the STARDUST facility before and the...
Jonathan Naish
(Technology)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Effective data visualisation is a key part of the scientific process with complex geometric datasets. It is the bridge between the quantitative content of the data and human intuition. Immersion in virtual reality (VR) provides benefits beyond the traditional “desktop” visualization tools and it leads to a demonstrably better perception of dataspace geometry, more intuitive data...
Zaixin Li
(Center For Fusion Science)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Chinese Fusion Engineering Testing Reactor (CFETR) is aimed to obtain the technologies to fill the gaps between ITER and DEMO. The helium cooled ceramic breeder (HCCB) blanket is one of the candidates for CFETR. Ceramics Li4SiO4, beryllium and helium of 8 MPa were selected as tritium breeding material, neutron multiplication and coolant, respectively. CLF steel developed in SWIP, one of...
Mikhail Subbotin
(CERN)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
In the framework of the joint Russian – Italian collaboration on the development of the IGNITOR project some preliminary estimates of the risk factors that may be occurring during the realization of the project were recently carried out.
A distinctive feature of the IGNITOR project is the fact that it contains some innovative solutions in the areas of research, engineering and technology,...
Andre Haußler
(Institute for Neutron Physics and Reactor Technology (INR))
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The Helical-Axis Advanced Stellarator (HELIAS) is the leading stellarator concept in Europe. Its prime example, Wendelstein 7-X, successfully started operation in 2015. Based on the 5-field-period symmetry, the HELIAS 5-B engineering design study emerged which is a stellarator power reactor concept designed for 3000MW fusion power.
The stellarator confines the hot plasma by external field...
Chiara Bustreo
(Consorzio RFX)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
The cost of the electricity (COE) generated by a fusion power plant is a key driver for the technology future energy market deployment. Hence, the ongoing researches on the pulsed DEMO design optimization, taking into account the physical and technical constraints, are putting priorities on the minimization of the DEMO direct costs that indeed greatly influence the COE.
Also the duty cycle of...
Hyun Soo Tho
(Strategy Division)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
This paper is focused on the analysis of spillover benefits of the ongoing R&D program on nuclear fusion in Korea. The spillover effects are understood here as positive externalitiesof publicly funded R&D activities that may be revealed at the companies’ level in the form of newly created knowledge stock; development of innovative products/ processes with broader market applications;...
Alexander Rydzy
(FSN-FUSTEC-TES)
08/09/2016, 14:20
J. Power Plants Safety and Environment, Socio-Economics and Technology Transfer
Poster
Ever since the ENEA Fusion Department has been involved in the technology transfer of its knowledge in the field of nuclear fusion from the R&D scope to the execution of large projects together with industry, it has been outlined the importance of working by a quality management system (QMS) and of applying the principles of the Project Management. The head of the ENEA Fusion Department took...
Christian Day
(Karlsruhe Institute of Technology (KIT))
08/09/2016, 16:40
In the framework of the EUROfusion DEMO Programme and its work package Tritium-Matter Injection-Vacuum (TFV), the EU is preparing the conceptual design of the inner fuel cycle of a pulsed fusion DEMO. This contribution presents the current status of the project, addresses the most demanding challenges and shows first results.
The project was started in 2014. The first one and a half years were...
Jochen Linke
(Forschungszentrum Jülich GmbH)
08/09/2016, 16:40
To qualify new plasma facing materials (PFM) and to evaluate the high heat flux performance under ITER or DEMO relevant loading conditions, extensive High Heat Flux (HHF) testing is indispensable. This includes performance tests under cyclic stationary thermal loads and screening of different material candidates under relevant transients such as Edge Localized Modes (ELMs) with high pulse...
Konrad Risse
(W7-X Operation)
08/09/2016, 16:40
The Wendelstein 7-X stellarator (W7-X), one of the largest stellarator fusion experiments, is presently in the first operational phase at the Max Planck Institute for Plasma Physics in Greifswald, Germany. The W7-X shall prove the reactor relevance of the optimized stellarator concept. To confine 30m33 plasma the W7-X machine has a superconducting magnet system with 50 non-planar...
Andreas Werner
(Wendelstein 7-X Operation CoDaC)
08/09/2016, 17:00
The Wendelstein 7-X safety control system is one of the main central control entities and ensures personal safety and investment protection. Its proper definition and setup has been a major precondition for the operation permit by the authorities and was inspected by external reviewers several times. The safety control systems has a distributed architecture comprising of the central safety...
Fadhel Malouch
(Den-Service d’études des réacteurs et de mathématiques appliquées (SERMA))
08/09/2016, 17:20
TRIPOLI-4® is a 3D continuous-energy Monte-Carlo particle transport code, developed by CEA, and devoted to shielding, reactor physics, criticality safety and nuclear instrumentation. TRIPOLI-4® is currently able to simulate four kinds of particles:
Neutrons from 20 MeV down to 10-5-5 eV,
Photons from 50 MeV down to 1 keV,
Electrons and positrons from 100 MeV down to 1 keV.
The...
John Jelonnek
(Institute for Pulsed Power and Microwave Technology (IHM))
08/09/2016, 17:20
Long term options for a steady state DEMO may require the availability of gyrotrons with an operating frequency above 200 GHz together with an RF output power of significantly more than 1 MW and a total gyrotron efficiency higher than 60 %. Fast frequency tuning in steps of around 2-3 GHz will be needed for control of plasma stability. Multi-purpose operation at leaps of about 30 – 40 GHz...
Albrecht Herrmann
(MPI für Plasmaphysik)
08/09/2016, 17:20
ASDEX Upgrade came into operation in 1991. It was designed as a tokamak with reactor relevant shaping. The coil and control system allows to operate in lower single null (LSN), double null (DN) or upper single null (USN) with up to 1.6 MA plasma current and an initially open divertor configuration. Divertor enhancements were concentrated on the lower divertor that was finally transferred to a...
Olivier Doyen
(Laboratoire des Technologies d’Assemblage)
08/09/2016, 17:40
This work was performed by CEA within the framework of one specific contract concerning the development for ITER of manufacturing procedures for the industrial ATMOSTAT (ALCEN group) and Fusion For Energy (F4E). The HCLL-TBM (Helium Cooled Lithium Lead Test Blanket Module) box assembly development implies the welding development of the following components: the Box and the Stiffening Grid (SG)...
Riccardo Ragona
(Laboratory for Plasma Physics)
08/09/2016, 17:40
The main advantages of Ion Cyclotron Resonance Heating and Current Drive (ICRH&CD) are its ability to achieve power deposition in the centre of the plasma column without any density limit along with direct heating of plasma ions. The challenge is then to couple large amount of power through the plasma boundary, where an evanesence layer has to be crossed, without exceeding the voltage standoff...
William Wehner
(Mechanical Engineering & Mechanics)
08/09/2016, 17:40
To collect meaningful experimental data, it is necessary to maintain consistent operating conditions in the tokamak plasma across repeated discharges. Presently, the desired plasma formation conditions, such as the shape of the plasma current profile, are achieved in a trial and error fashion, which can be a lengthy, wasteful process. In this work, model-based control techniques including...
Olaf Neubauer
(Forschungszentrum Jülich GmbH)
09/09/2016, 08:30
Oral
The mission of Wendelstein 7-X is to assess the reactor capabilities of the HELIAS stellarator line. W7-X is equipped with superconducting coils (B=2.5 T) and is sufficiently large (V=30 m33) to potentially attain steady-state plasmas at low collisionalities and high densities at the same time. As prerequisite for long-pulse operation, W7-X will employ high power, cw microwave...
R. Albanese
(on behalf of the EUROfusion WPDTT2 team & the DTT report contributors)
09/09/2016, 09:50
Oral
One of the main challenges in the European roadmap toward the realisation of fusion energy with a demonstration plant DEMO [1] is to develop a heat and power exhaust system able to withstand the large loads expected in the divertor. In parallel with the programme to optimise the operation with a conventional divertor based on detached conditions to be tested on ITER, efforts are being devoted...
F. Warmer
(Max Planck Institute for Plasma Physics)
09/09/2016, 11:00
Oral
One of the high-level missions of the European Roadmap to the realisation of fusion energy is to bring the HELIAS stellarator line to maturity. The near-term focus is the scientific exploitation of the Wendelstein 7-X experiment in order to assess stellarator optimization in view of economic operation of a stellarator fusion power plant. W7-X will play a decisive role for these studies but may...
P. Batistoni
(EUROfusion Consortium)
09/09/2016, 11:40
Oral
Within the framework of the EUROfusion programme, a work-package of technology projects (WPJET3) is being carried out in conjunction with the planned DT experiment at JET with the objective of maximising the scientific and technological return of DT operations in support of ITER. To this purpose, experiments, analyses and studies are performed in the areas of neutronics, neutron induced...
