Description
Poster Session First Conference Day
The 10 MW FULGOR test facility (Fusion Long Pulse Gyrotron Laboratory) is built to meet the future needs of the gyrotron development, in particular for future fusion machines like the DEMOnstration power plant (DEMO). In the final stage it will enable KIT to test gyrotrons up to 4 MW RF output power in CW and frequencies up to 240 GHz. One of the main components of FULGOR test facility is the...
The International Thermonuclear Experimental Reactor (ITER) is currently under construction at Cadarache in southern France. The Joint European Torus (JET) is presently the largest tokamak in the world and the only one capable of using tritium. At JET, D-T fusion experiments will be conducted in 2019 (DTE2) on addressing the future ITER needs and reducing the risks of ITER operations. During...
The Water-Cooled Lithium Lead (WCLL) Breeding Blankets is one of the European blanket designs proposed for DEMO reactor. A tritium transport model inside the blankets is necessary to assess their preliminary design and safety features. Tritium transport and permeation are complex phenomena to be taken into account in the evaluation of tritium balance in order to guarantee tritium...
High field superconducting magnets are an essential technology enabling the development of magnetic confinement fusion and high energy hadron colliders. These two communities have joined efforts to design a facility for testing superconducting cables and small insert coils. We propose a large aperture Nb3Sn dipole to replace the magnet assembly of EDIPO, which was irreversibly damaged in 2016,...
The paper describes a design proposal for the ITER ELM Coil Power Supply, optimized for the simultaneous ELM mitigation and RWM stabilisation during the ITER non-inductive operation. Slow (ELM) and fast (RWM) rotating magnetic fields are generated by exciting the three sets of nine ELM-Coils at ELM-frequencies up to 5 Hz (N = 4) and RWM-frequencies up to 60 Hz (N = 1). Starting from a basic...
Injection of low-Z powders into fusion plasma has been used to improve wall conditions, similar to the standard boronization process using diborane. Powder injection has the advantage of being much simpler, non-toxic, and efficient. The W7-X stellarator is planning on utilizing powder injection in long pulse discharges; a proof-of-principle test for horizontal injection into the plasma was...
The simulation of the plasma equilibrium and its evolution is important for the study of plasma physics and for the correct design of fusion devices. For this purpose, a novel code, based on the solution of the Grad-Shafranov equation, has been fully implemented in ANSYS. It exploits the finite element method using the magnetic potential vector formulation. In this approach plasma pressure and...
The design of nuclear fusion devices like ITER requires the execution of complex multi-physics simulations, involving different analysis disciplines such as mechanical, thermal-hydraulic and neutronics. Nowadays, thanks to the novel implementation of unstructured mesh capability into MCNP6, nuclear responses can be computed over meshes conformal with the components tackling the problematic...
PROTO-SPHERA (Spherical Plasma for Helicity Relaxation Assessment) is a new concept of torus that aims to produce a Spherical Torus at closed flux surfaces and a force-free screw pinch (SP) at open flux surfaces and fed by electrodes [1]. By replacing the metal centrepost current of the spherical tokamaks with the SP plasma electrode current, the rod at the centre of the plasma, which...
In the frame of the JET3 Deuterium-Tritium (D-T) technology project within the EUROfusion Consortium program, several neutronics experiments are in preparation for the future high performance D-T campaign at the Joint European Torus (JET). The experiments will be conducted with the purpose to validate the neutronics codes and tools used in ITER, thus reducing the related uncertainties and the...
Alloying elements can possibly serve as an important technique in designing W-based plasma facing materials (PFMs) with superior comprehensive performance. To investigate the interaction between the alloying elements and point defect is one of the mainly contents to study the W material service properties under irradiation. The first-principle method based on the density function theory was...
The neutron generator (NG) is being used more increasingly in various industrial and research area such as neutron activation analysis, neutron radiography, neutron capture therapy, and so on. In such an application of neutron generator, compactness is one of the most important issue. Since neutron source is generated by deuterium-deuterium (D-D) or deuterium-tritium (D-T) fusion reaction,...
Helium, as the ash of burning D-T plasma, is an unavoidable impurity component in DEMO reactor . Its efficient removal from the burning zone of a D-T fusion reactor is most important in the path towards achievement of economic fusion power production. Edge plasma transport properties: recycling/pumping will play a key role in the problem of helium removal from reactor.
This work...
Continuous Wave (CW) gyrotrons are the key elements for electron cyclotron resonance heating and current drive in present machines and future fusion reactors. In the frame of the EUROfusion activities, a 170 GHz, 2 MW short-pulse (ms) coaxial gyrotron existing at Karlsruhe Institute of Technology (KIT) is being upgraded for operation at longer pulses (100 ms - 1 s). In the coaxial gyrotron,...
Installation of metallic plasma facing wall (ITER-like wall – ILW) [1] and replacing the previous carbon wall (JET-C) in the Joint European torus (JET) was a unique possibility to collect from the tokamak vacuum vessel the first wall erosion products (EP) – dust and flakes.
Fundamental investment about the properties of EP to comply with security reasons is given by analysing EP from other...
To restrict climate change, it is highly desirable to replace conventional power plants by emission free technologies like solar and wind power. This leads to increased shares of fluctuating sources in the power system challenging the balance between generation and demand. Moreover, as the transportation and heating sectors are expected to shift towards using more electricity the projected...
The evaluation of the neutron induced material activation plays an important role for the development of future fusion power plants for issues related to safety, engineering design and radioactive waste management. For these devices the activation codes and cross section libraries handling neutron energies up to 20 MeV are quite adequate.
Besides, in order to study the irradiation effects on...
The JT-60SA project, a superconducting tokamak developed under the Satellite Tokamak Programme of the Broader Approach Agreement between EU and Japan and of the Japan Fusion National Programme, is progressing on schedule towards the first plasma in 2020. Within the European contribution to JT-60SA, Spain is responsible for providing JT-60SA cryostat.
...
Assessment of Controllers and Scenario Control Performance for ITER First Plasma
M.L. Walker, D.A. Humphreys
General Atomics, PO Box 85608, San Diego, California 92186-5608, USA
G. Ambrosino
CREATE/Università di Napoli Federico II, Napoli, Italy
P.C. De Vries, J.A. Snipes
ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067, St. Paul-lez-Durance, France
F. Rimini
CCFE/Fusion...
Plasma current measurements will play an important role in ITER to provide real-time plasma control and machine protection. Fiber Optics Current Sensors (FOCS) with the sensing fiber installed on the external surface of the vacuum vessel is a system intended to perform this task. The FOCS signal is proportional to the current and is more suitable for the steady-state operation as compared to...
NBI Cooling Water System is used during beam operation for the removal of received heat load from vessel sub-components. During the process of shifting the NBI Vacuum Vessel to SST-1, it was needed to shift the cooling water plant. Shifting of Cooling Water Plant leads to the dis-mantling of the actual plant and designing, development and installation of a new plant. The cooling water plant is...
Finite numerical simulation on plasma generation, confinement and distribution, is lacked in design and research of linear plasma devices. As a high density plasma source, cascaded arc plasma is widely used in plasma and material interaction devices. In this paper, a high-density linear plasma device with cascaded arc source is developed, which plasma parameters and distribution is analyzed by...
Hydrogen generation reaction with water vapor of Be at high temperatures and BeO produced by this reaction that is harmful to human bodies are major drawbacks. Advanced neutron multipliers with high stability at high temperatures are desirable for fusion reactors where coolant water is extensively used. Beryllides have strong potential for use in high-temperature environments. In the framework...
The ITER Heating Neutral Beam (HNB) injector is required to deliver 16.7 MW power into the plasma from a neutralised beam of H-/D- negative ions, produced by an RF source and accelerated up to 1 MeV. To enhance the H-/D- production, the surface of the acceleration system grid facing the source (the plasma grid) will be coated with Cs to reduce its work function. Cs will be routinely evaporated...
The main purpose of the Divertor Tokamak Test (DTT) is to study solutions to mitigate the issue of power exhaust in conditions relevant for ITER and DEMO. The key feature of such a study is to equip the machine with a significant amount of auxiliary heating power (45 MW) in order to test different divertor solutions. According to the Italian project, the experiment is foreseen to operate with...
The potential of High Temperature Superconductor (HTS) for a Toroidal Field Coil (TFC) of a future fusion power plant has already be demonstrated in a conceptual design within EUROfusion [1]. One of the candidates of a high current HTS conductor for use in a fusion magnet is the so-called HTS CrossConductor (HTS CroCo) where REBCO tapes are arranged in a cross sectional optimized way.
The...
The roadmap to the realization of fusion energy describes a path towards the development of a DEMO tokamak reactor, which is supposed to provide electricity into the grid by the mid of the century [1]. The DEMO diagnostic and control (D&C) system must provide measurements with high reliability and accuracy, constrained by space restrictions in the blanket under the adverse effects induced by...
The main motivation of the current work, framed under the safety EUROfusion activities to develop DEMO, is to present the conclusions drawn from our contribution to the safety studies of the HCPB DEMO design carried out by the team tasked with AINA code development. During 2016 and 2017 a new AINA version has been built and properly validated in order to evaluate plasma evolution and in-vessel...
The ITER lower port is designed to support divertor remote handling and vacum pumping. To meet the purpose, it will be assembling to each main vessel on the vacuum vessel manufacturing site. Before delivering to the sector shop, a series of fuctional and mechanical test, which is so-called factory acceptance test (FAT) should be performed by the manufacturer. The ITER FAT should be complying...
The four ITER EC (Electron Cyclotron) Upper Launchers inject up to 8 MW microwave power each with the aim to counteract plasma instabilities during plasma operations. The structural system of these launcher antennas will be installed into four upper ports of the ITER vacuum vessel.
The structural part of the Upper Launchers which forms the plasma facing component is called the Blanket Shield...
The EU fusion roadmap defines as a goal the development of a DEMO, which achieves a high plasma operation time and demonstrates Tritium self-sufficiency and net electricity output. A number of design issues have been identified as critical, either because the solution chosen in ITER is not suitable in DEMO or because it is a DEMO-specific issue not present in ITER. All of these will affect in...
Radioactive dust will accumulate in the vacuum vessel (VV) of ITER after plasma operations. Thus, the ITER Blanket Remote Handling System (BRHS) will be installed in the VV to handle the blanket modules, which can weigh up to 4.5 ton and be larger than 1.5 m, stably and with a high degree of positioning accuracy. The BRHS itself also needs to undergo regular maintenance in the Hot Cell...
In our previous work, the joint between oxide dispersion strengthened copper alloy (ODS-Cu) and tungsten (W) demonstrated superior fracture strength (~200 MPa). In the present study, deformation and fracture behavior of the bonding layer and its vicinity after the three-point bending test was investigated. Consequently, it was found that the crack initiation site was dominantly in the tungsten...
DEMO represents DEMOnstration Power station, which is a nuclear fusion power station and it is proposed to be built after ITER experimental nuclear fusion reactor. It is impossible to do any change or repair work during the nuclear operation by human directly due to the high radiation and extreme temperature in the fusion reactor. And the solution to solve these issues is adopt the remote...
A liquid-lithium (Li) free-surface stream flowing under a high vacuum serves as a Li target for the planned International Fusion Materials Irradiation Facility (IFMIF). As the primary Japanese activity for the Li target system of the IFMIF/EVEDA (i.e. Engineering Validation and Engineering Design Activities) project, implemented under the Broader Approach (BA) Agreement, cavitation-like...
A mechanical support structure (a.k.a. “platform”) has been designed to provide mechanical support and thermal conductance for the inductive magnetic sensors installed on the inner shell of ITER vacuum vessel (VV) for equilibrium and high-frequency magnetic field measurement. The platform design is modular so as to simplify the on-site installation process. It consists of a permanent and a...
A mechanical support structure (a.k.a. “platform”) has been designed to provide mechanical support and thermal conductance for the inductive magnetic sensors installed on the inner shell of ITER vacuum vessel (VV) for equilibrium and high-frequency magnetic field measurement. The platform design is modular so as to simplify the on-site installation process. It consists of a permanent and a...
Design parameters of the ITER Plasma Position Reflectometer (PPR) in-port-plug antennas are determined and then their measurement performance is assessed using 2D full wave analysis.
Two ITER scenarios were selected when considering the optimum antenna position and orientation, namely the baseline scenario (15 MA D-T) and the low density one planned for the initial non-active phase at 7.5 MA....
The ITER Heating Neutral Beam (HNB) injector is required to deliver 16.7 MW power into the plasma from a neutralised beam of H-/D- negative ions, produced by an ICP RF source and accelerated up to 1 MeV. To enhance the H-/D- production, the surface of the acceleration system grid facing the source (the plasma grid) will be coated with Cs because of its low work function. Cs will be routinely...
The ITER Heating Neutral Beam (HNB) injector is required to deliver 16.7 MW power into the plasma from a neutralised beam of H-/D- negative ions, produced by an ICP RF source and accelerated up to 1 MeV. To enhance the H-/D- production, the surface of the acceleration system grid facing the source (the plasma grid) will be coated with Cs because of its low work function. Cs will be routinely...
Future designs of fusion devices are going to make use of a tritium production systems, several of which are considered using PbLi alloy as a breeder. Apart from tritium, other volatile and non-volatile species are being formed, either as products of the neutron irradiation or as corrosion products. The volatile impurities must be eliminated based on safety concerns (e.g. polonium) or blanket...
This paper designs a cascaded four-quadrant 24-pulse converter fed by a 6-phase pulsed motor-generator (M-G) aiming at the requirements of HL-2M tokamak, which is mainly used for the control of vertical instability of plasma. The optimally designed four-quadrant 24-pulse converter cascaded by four-quadrant converters is able to balance the loads of the double Y of M-G. Owing to the fact that...
Each of the 4 ITER Electron Cyclotron Heating Upper Launcher (ECHUL) features 8 transmission lines (TLs) used to inject 170 GHz microwave power into the plasma at a level of up to 1.31 MW (at the TL diamond window) per line. The millimetre waves are guided through a quasi-optical section consisting of three fixed mirror sets (M1, M2 and M3) and one front steering mirror set (M4).
The M2...
The aim of this paper is to identify the design criteria of the Electrical Power Supply of the Lithium Systems of DEMO-Oriented NEutron Source (DONES) power plant. This facility is planned as a simplified IFMIF-like plant to provide, in a reduced time scale and with a reduced cost, information on materials damage due to neutron irradiation. In particular, a general overview of the current...
The torus window unit is a very particular component of the ITER EC H&CD upper launcher aiming to provide the vacuum and confinement primary boundary between the vacuum vessel and the transmission lines (TLs). The high power 170 GHz millimeter-wave beams generated by the gyrotrons travel along the TLs and pass through the window units, before being quasi-optically guided into the plasma via...
The DC equipment of ITER poloidal field converter will be interconnected by DC busbar with water cooled aluminum busbars with cross section 200×60 mm, whose segments will be connected by aluminum flexible links in order to compensate that of thermal expansion. Because the contact surface between DC busbar and flexible links are small and the high current up to 30 kA flowed through, the power...
The vacuum systems of neutral beam injectors have very demanding requirements in terms of gas type, pumping speed and throughput. Due to its high affinity to hydrogenic species, non-evaporable getter (NEG) is in principle a good pumping technology candidate for the deployment in neutral beams, which require the injection of a serious amount of hydrogen in order to operate. Getter materials...
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. These launcher antennas will be installed into four upper ports of the ITER vacuum vessel.
Beside their functional purpose the port plugs which are the structural system of the launchers have to provide as much shielding as...
In the fusion reactor, tungsten will be exposed to high heat flux, neutrons, ash and fuel plasma of fusion reaction including tritium. The irradiation defects generated by neutrons will dynamically migrate, which results in the accumulation and annealing of irradiation defects. The irradiation defects in tungsten will act as potential trapping sites for hydrogen isotopes and, therefore,...
The structural materials in fusion reactors as DEMO and future power plants are under strong irradiation and will suffer from radiation damages. The knowledge of the radiation induced degradation is planned to be investigated in IFMIF-DONES, a facility in which fast neutrons are produced by a reaction of a D-beam with a liquid lithium target. The operation of such a system requires the control...
A new experimental device has been designed, manufactured and tested for quasi-2 dimensional turbulence studies in magnetized electrolyte system. This experiment can provide significant information about the interaction of large scale shear flows (zonal flows) and smaller scale turbulent vortices. This physics problem has a relevance in different scientific areas such as the turbulent...
Institute for Plasma Research is developing an Experimental Helium Cooling Loop (EHCL) as a part of R&D activities in fusion blanket technologies. This helium cooling system is designed for testing various nuclear fusion components such as tritium breeding blanket, helium-cooled divertor, and any other components which can be operated within EHCL operating window. The cooling channels of...
A medium sized water-cooled Divertor DOME has been manufactured at Institute for Plasma Research (IPR), India. Divertor Plasma Facing Components (PFCs) such as DOME and Reflector Plate has multi-layered joints which are made of various materials such as Tungsten (W), OFHC Copper (Cu), Copper alloy (CuCrZr) and SS316L etc. Joining of such multi-layered joints is known to be problematic as being...
Machine learning has garnered increasing attention within the fusion community in recent years, with much of the focus going toward implementation of disruption predictors. However, disruption detection is but one possible area in which the large body of fusion experimental data, accrued over decades, can be put to use. In particular, this data can be utilized to assist in the implementation...
In the Large Helical Device (LHD), the development of high-power and long-pulse ICRF system is ongoing. Frequency was fixed at 38.47 MHz for the optimization of devices. At this frequency, plasma is heated with the minority ion heating of hydrogen and the second harmonic heating of deuterium. Field-Aligned-Impedance-Transforming (FAIT) antenna has the potential performance of high-power...
The high neutron flux inherent in fusion reactors creates high heat loads in the components surrounding the plasma. These heat load needs to be managed through active cooling. These components also become highly activated so require remote maintenance, hence the connection and disconnection of these cooling systems becomes an important functionality of these maintenance activities. The...
The Poloidal Field(PF) coils are one of the main sub-system of ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences(ASIPP) as per the Poloidal Field coils cooperation agreement between ASIPP and Fusion for Energy(F4E).
ITER PF6 winding pack is composed by stacking of 9 double pancakes. Series double pancakes are being wound in ASIPP...
ITER-RH system is used to exchange the divertor’s 54 cassette assemblies in the vessel. Water hydraulics and servo valves are currently used in the task requiring high accuracy tracking and the use of de-mineralized water. The main concern has been robustness of the technology. Only few suitable commercial water servo valves exist and problems e.g. with jamming and wear been encountered. A...
The activated/toxic dust resuspension inside the vacuum vessel of future fusion devices as ITER or DEMO is a safety issue of main concern. In case of a LOVA or a LOCA, dusts produced during the normal and off-normal conditions can be released inside the tokamak building or towards the external environment. These accidents are not expected during the whole lifetime of the ITER machine, though...
FFHR depicts the conceptual design of an LHD-type helical fusion reactor that is being developed by the National Institute for Fusion Science. Several design mechanisms for FFHR have been investigated. For instance, FFHR-d1A is a self-ignition demonstration reactor that operates at a magnetic field intensity of 4.7 T and has a major radius of 15.6 m. FFHR-c1 is a compact-type sub-ignition...
The screening of a neutral gas by plasma from the top of the private flux region (PFR) of the latest DEMO divertor configuration without the dome structure is analysed. The effect of the neutral gas compression in the PFR is assessed by using the direct simulation Monte Carlo method (DSMC) and the ambipolar approximation for the simulation of neutral molecule dissociation and ionization as a...
Plasma facing component such as breeding blanket and divertor in fusion reactors are supposed to be assembled by welding and joining of parts made of reduced-activation ferritic-martensitic (RAFM) steel, and accordingly the structural integrity is significantly affected by the properties of the joint. Conventional fusion welding results in a wide heat-affected zone (HAZ) where a well-known...
The qualification Full Scale Prototype (FSP) of the Enhanced Heat Flux (EHF) First Wall (FW) for the ITER blanket will be manufactured by JSC NIKIET and JSC NIIEFA as part of the Procurement Arrangement between the ITER Organization and the Russian Federation Domestic Agency. The FSP design is based on the FW panel #14 type A and includes: the supporting structure (FW beam), plasma facing...
The design study of a DEMOnstration (DEMO) Fusion Plant is one of the main points of the European Roadmap to Fusion Electricity [F. Romanelli, http://www.efda.org/wpcms/wp-content/uploads/2013/01/JG12.356-web.pdf]. The pre-conceptual design phase of DEMO is presently used to explore a flexible range of the main machine geometrical design parameters, including machine magnetic configurations,...
The ITER vacuum vessel (VV) contains transducers for vessel and blanket instrumentation and for the measurement of plasma performance. The electrical and optical signals to and from these transducers are managed through the electrical services infrastructure project 55.NE.V0. This system is responsible for transmitting electrical signals from the VV inner skin to the outside of the vacuum...
The stellarator Wendelstein 7-X has been prepared for long pulse operation in the first operational campaign. Forschungszentrum Jülich has contributed diagnostics for investigation of plasma wall interaction processes in presence of an island divertor and steady state plasma at high density and low temperature.
A versatile optical observation system has been developed for local...
Lithium rich advanced ceramic breeder pebbles composed of lithium orthosilicate with a strengthening phase of lithium metatitanate are intended as tritium breeders for future fusion reactors. The EU breeding blanket being designed for trial in ITER will feature the pebbles in the form of pebble beds in the wall of the reactor. Upon irradiation with neutrons, the lithium will decay into tritium...
Wendelstein 7-X (W7-X) is a modular advanced stellarator, which successfully went into operation in December 2015 at the Max-Planck-Institut für Plasmaphysik in Greifswald, Germany and continued to thrive at the experimental campaign OP1.2a (August-December 2017). The term modular stellarator refers to a generalized stellarator configuration with nested magnetic surfaces created by a system of...
The European fusion electricity roadmap sets out a strategy for a collaboration to achieve the goal of generating fusion electricity by 2050. It has been developed based on a goal-oriented approach with eight different missions including development of Heat-Exhaust systems which must be capable of withstanding the large heat/particle fluxes of a fusion power plant. This paper summarises the...
As the first funding period of EUROfusion Consortium is nearing its end, the work package “DEMO divertor” (WPDIV) is entering the final project period concluding its preconceptual design activities. The primary mission of WPDIV is to deliver holistic design solutions of the divertor targets as well as the cassette and to assure the availability of required technologies at least with a...
The ITER Magnet System will be the largest and most challenging integrated superconducting magnet system ever built. For the Central Solenoid (CS), cable – in – conduit - conductors (CICCs) of nearly one kilometre length are produced, but still, it will be necessary to connect several lengths together to wind the gigantic 110 tonnes coils. The creation of these superconducting joints is one of...
The ITER Magnet System will be the largest and most challenging integrated superconducting magnet system ever built. For the Central Solenoid (CS), cable – in – conduit - conductors (CICCs) of nearly one kilometre length are produced, but still, it will be necessary to connect several lengths together to wind the gigantic 110 tonnes coils. The creation of these superconducting joints is one of...
This paper mainly introduces the experimental analysis of the dummy load prototype, whose functions are to verify the capability of the ITER magnetic power supply systems to operate at their rated power levels without energizing the superconducting coils. The rated inductance of dummy load is 6.73 mH and the pulse test currents are 45 kA, 55 kA and 68 kA. To meet the requirements of the large...
A pressure, volume, temperature, and concentration (PVT-c) method, which is widely used to measure the amount of tritium owing to its effectiveness, requires a desorption process from uranium tritides and a transfer process to a measurement tank in the tritium storage and delivery system (SDS) for a tokamak-type nuclear fusion reactor. In addition, the repeated processes for the PVT-c,...
This paper reports on an experimental evaluation of wall shear stress in a double contraction nozzle to produce a liquid lithium (Li) target being intended for use as a beam target for the intense fusion neutron sources such as the International Fusion Materials Irradiation Facility (IFMIF), the Advanced Fusion Neutron Source (A-FNS), and the DEMO Oriented Neutron Source (DONES). The current...
The quench protection switch (QPS) is indispensable to protect the magnet coils from the damage of a quench in a superconducting Tokamak. In this paper, a QPS based on the artificial current zero is involved. The vacuum circuit breaker (VCB), which is driven by a high-speed electromagnetic repulsion mechanism, is used as the main circuit breaker (MCB). Two kinds of commercial vacuum...
JSC NIKIET is a main supplier of ITER in-vessel components and its responsibility includes the manufacture of the FW beam, finger bodies, and the mechanical attachment system of the Enhanced Heat Flux (EHF) First Wall (FW) Panels in the framework of Procurement Arrangement 1.6.Р1А.RF.01 dated 14.02.2014. The mechanical attachment system comprises the central bolt, threaded barrel and system...
Progress in technological fields such as High Temperature Superconductors, Additive manufacturing, new diagnostics, and innovative materials, has led to new scenarios and to a second generation of Fusion Reactor designs. A new Affordable Robust Compact (ARC) Fusion Reactor, which meets its goal in a cheaper, smaller but even more powerful, faster way to achieve Fusion Energy, has been designed...
An exploratory risk analysis of ITER Cask & Plug Remote Handling System (CPRHS) has been performed under a system engineering approach considering the CPRHS in various operational states with the associated loads.
A Functional Breakdown Structure was developed from the 4 main functions fulfilled by the CPRHS: to dock, to handle, to transport and to confine. During Tokamak maintenance...
Neutral gas pressures in the vacuum vessel of ITER will be measured by hot cathode ionization gauges. The design is based on the ASDEX pressure gauge which is operated successfully in many fusion experiments worldwide. Further development is needed to fulfill superior requirements: the upper measuring limit has to be at least 20 Pa in hydrogen at a magnetic flux density of up to 8 T. The...
Neutron multipliers with lower swelling and higher stability at elevated temperatures are desired for the pebble bed blankets of designed DEMO rector. Among beryllium-based intermetallic alloys, vanadium beryllide Be12V is considered to be an attractive material from the point of view of its potential use as an advanced neutron multiplier of the breeding blanket. Preliminary assessment of its...
The fabrication of the modules for the ITER Central Solenoid (CS) is in progress at General Atomics (GA) in Poway, California, USA. This purpose built facility has been established with the requisite tools and machines to fabricate the seven 110 tonne CS modules (six required plus one spare). The current schedule has the first module’s fabrication completing in 2018 followed by electrical and...
The power supply set for the EU EC Heating system (ECPS) of ITER provides up to 6 MVA electrical power to two 170GHz/1MW Gyrotrons. The required electrical power for the gyrotrons is both very high and has to comply also with highest quality requirements. These performance indicators were proven with full voltage modulation at rates up to 5kHz.
Ampegon’s newly developed power supply topology...
High heat load test were performed by using 1) E-beam for tungsten blocks and divertor mock-up, and 2) Long pulse H-mode plasmas in KSTAR for tungsten blocks mounted on stainless steel base.
Tungsten blocks are exposed to a heat flux of 13 MW/m2 from the top with a beam spot size around 11.5 mm in diameter, 100 kV and 12.5 mA, while the divertor mock-up is exposed to much higher heat flux up...
In vessel Mirnov coils are an essential diagnostic in present day tokamaks. Their use in ITER and future Fusion reactors presents some disadvantages linked to the high radiation environment. Furthermore large Electro Magnetic forces can be experienced by the coil, due to the pulsed operation of the tokamak device [1], and disruptions [2].
Since the operation with the ITER-like wall, JET has...
In the European DEMOnstration nuclear fusion power plant (DEMO), the desired toroidal magnetic field is produced by a magnet system composed of 16 Toroidal Field (TF) coils, according to the last 2017 reference baseline. The total stored energy of about 140 GJ, more than three times that of ITER TF coils, has to be quickly dissipated in case of quench by a suitable Quench Protection (QP)...
The species-selective (or Optical) Penning gauge approach to the measurement H2/D2/T2 fuel isotopic composition [1] and He/D2 concentration [2] in the neutralized particle exhaust of fusion devices is almost universally used nowadays across all fusion facilities. Although recent studies have shown that, through spectroscopic detection optimization, He/D2 detection is feasible down to at...
This paper is focused on the analysis of existing industrial-scale process for recycling of DEMO steel components (Eurofer, AISI 316L) and Lithium orthosilicates breeder. The aim is the assessment of their practical feasibility and the individuation of preparatory activities to be performed for facilitating and improving the recycling.
In detail, the thermodynamic analysis of recovering 14C...
A new tritium facility to study the interaction of tritium with fusion relevant materials, and its retention and release, has been produced. Tritium retention is a major issue for fusion power devices. The new facility allows implanting of a range of gases into samples, including tritium. This facility is currently used for the UKAEA led Tritium retention in Controlled and Evolving...
Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) has been intensively studied for the EU DEMO. However, several feasibility issues remain for a HCPB-class DEMO reactor, namely the large diameter of the Primary Heat Transfer System pipework, the resulting large coolant inventory and large expansion volumes required after an ex-vessel loss of coolant accident, the limited operational...
The central safety system (cSS) of W7-X consists of two parts. The safety related PLC with its corresponding periphery, such as sensors and actors, fulfills the requirements of occupational safety and ensures basic investment protection. The reaction time from the signalization of dangerous faults to the initiation of protection measures like W7-X emergency stop or media shut-off is in the...
Two sets of upgrades are being implemented on the TCV tokamak. The first set involves the installation of neutral beam injection (NBI) and new Electron Cyclotron (EC) power sources, to heat the ions and vary the electron to ion temperature ratio, for ITER relevant β values. A tangential 15-30keV, 1MW, 2s NBI is operational on TCV since 2015. A second 1MW, ~50keV beam, also tangential but...
The complex power and particle wall loading conditions in fusion devices lead to various surface modifications of plasma-facing components (PFCs). To assess the consequences of these modifications on power handling capability and lifetime of PFCs, detailed microscopic studies of the surface and internal structure are required. Essential are analyses of the same area before and after plasma...
The work presents the results of high temperature brazing of tungsten with EK-181 steel by rapidly quenched into ribbon filler alloys based on copper. Compositions of the filler alloys were chosen with consideration to the requirement of reduced activation that is necessary for DEMO reactor. All the joints were manufactured at 1100oC in a vacuum furnace. To analyse microstructure and...
The ITER magnet system will be the largest superconducting magnet system ever built. The system, all inside a cryostat, is mainly composed by a central solenoid (CS) split in 6 modules, a set of 18 toroidal field (TF) D-shaped coils and 6 poloidal field (PF) coils. Each of these coils use variable type of cable-in-conduit-conductors (CICC) actively cooled by supercritical helium forced flow....
Following the decommissioning of JET, and other future fusion reactors, there will be large amounts of tritiated waste requiring disposal. An appropriate containment strategy is required for storage of this waste. Studies have so far demonstrated that stainless steel appears to be the most promising containment material, but little is known about the permeation of hydrogen isotopes through...
Tokamak-based fusion neutron source (FNS) [Kuteev B.V. et al 2010 Plasma Phys. Rep. 36 281, Kuteev B.V.et al Nucl. Fusion 55 (2015) 073035] is the centerpiece of the fusion-fission hybrid reactor (combining nuclear and thermonuclear technologies). In Russia, for the demonstration of stationary and hybrid technologies, the DEMO-FNS project has been developed, which should operate at least 5000...
A laser-induced fluorescence (LIF) diagnostic has been designed for measuring helium density (nHe) and ion temperature (Ti) in the outer leg of the ITER divertor. The LIF diagnostic is integrated with the divertor Thomson scattering (DTS) diagnostics via common injection and collection optics. Optimisation of previously proposed spectroscopic schemes, and lasers suitable for nHe and Ti...
The stellarator Wendelstein 7-X (W7-X) is a fusion device designed for steady state operation. It is a
complex technical system. To cope with the complexity a modular, component-based control and
data acquisition system has been developed.
During operation phases of W7-X components steadily evolve. For instance measurement devices for
diagnostics get improved, technical processes are...
In large Neutral Beam Injectors for fusion applications, the efficiency of ion beam neutralization and transport to the tokamak plasma strongly depends on the divergence and the deflection angle of each single beamlet with respect to its ideal trajectory. In fact, a very narrow window is available for the particle beam to pass through the neutralizer panels and the duct reaching the tokamak...
Transient analysis in a water-cooled fusion DEMO reactor have been performed to support the WCLL (Water-Cooled Lithium Lead) breeding blanket design. In this framework, the Design Basis Accident analysis of an in-box LOCA has been carried out.
The WCLL breeding blanket concept relies on Lithium Lead (LiPb) as breeder, neutron multiplier and tritium carrier, which is cooled by water at 15.5 MPa...
Low pressure plasma spraying (LPS) and spark plasma sintering (SPS) are attractive techniques to prepare W armor layers on substrate materials. The properties of LPS-W and SPS-W depend on fabrication conditions. In this study, LPS-W and SPS-W layers were prepared on graphite and carbon fiber reinforced carbon composite (CFC) substrates at different temperatures, and D retention after plasma...
The behaviour of the SOL plasma of the Italian projected DTT is analysed for the standard divertor configuration by means of the integrated COREDIV code simulations when either Lithium or Tin are used as liquid target materials.
The DTT tokamak is expected to operate in H-mode, which requires the value of power to scrape-off layer above the L-H threshold. On the other hand it is postulated...
Operation of a future demonstration fusion reactor (DEMO) requires the handling of a significant power flux that crosses the separatrix and enters the scrape-off layer. A considerable amount of energy has to be dissipated before the heat flux reaches divertor plates. The divertor may be exposed to high heat fluxes causing high temperature gradients and material fatigue. Such challenging...
In this paper we present the analysis of System Requirements and Interfaces of the Heating and Current Drive (HCD) system of the Demonstration Fusion Power Reactor DEMO.
The work was performed applying Model-Based Systems Engineering (MBSE) refining the HCD System Architecture for assessing the system functions, its interdependencies and its overall integration into DEMO. Two concepts for DEMO...
The ITER Vaccum Vessel (VV) is supported by the nine VV gravity supports (VVGS) located on the cryostat toroidal pedestal. The VVGS is dual hinge type that fastened by dowel on the hinge-block hole. The primary hinge restrains a vertical and toroidal movement of the VV system against fast displacements by the seismic events or fast transients. The secondary hinge restrains steady vertical...
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 operations. STRIKE is made of 16 1D Carbon Fibre Composite (CFC) tiles, intercepting the whole beam and observed on the rear side by infrared (IR)...
The divertor, being the main power exhaust of a tokamak, is exposed to high heat
fluxes and therefore must be precisely aligned to prevent leading edges. Since the
transition from carbon to tungsten tiles in ASDEX Upgrade it was found that a specific
assembly in the divertor was misaligned up to 1.5 mm after the experimental
campaigns. This lead to...
Within the EU, the current grade of advanced ceramic breeder pebbles is composed of a mixture of Li4SiO4 (LOS) and Li2TiO3 (LMT). These pebbles are fabricated at KIT by the melt-based process “KALOS”. The addition of LMT is beneficial for two aspects: the mechanical strength of the pebbles is considerably increased and the long-term stability at high temperatures is improved. Nevertheless, the...
Large-scale isotope separation in a DEMO tritium plant poses significant challenges. Alternatives to distillation and palladium-based adsorption (used in the tritium fuel cycle for JET) remain elusive, despite the disadvantages: Cryodistillation is energy intensive and lacks inherent safety due to the high tritium inventories in the liquid phase, that inevitably expand to vapour in the event...
ITER in-vessel magnetic sensors play a key role for ITER plasma operation. Each of these sensors is accommodated in a platform mounted on the inner surface of ITER vacuum vessel and behind the blanket.
A full set of engineering analysis has been performed on the platform to assess the feasibility of the design configuration.
Electromagnetic (EM) Sub-Modelling technique has been used for very...
The ITER Upper Visible/Infrared Wide Angle Viewing System diagnostic will provide key measurements for machine protection and plasma control. The system, installed in five upper port plugs, will monitor the ITER divertor but also part of the ITER first wall using both high definition infrared and visible cameras typically running at 100 Hz. The plant system I&C will process about 40 Gb/s of...
A review of the joints for the ITER CS CICC is given. More detailed discussion of the design and performance of the ITER CS joints is presented including successes and the revealed problems. ITER CS has three types of joints: 1) sintered joints to connect conductor lengths in the CS module; 2) coaxial joints to connect the CS module terminations to the superconducting buses; 3) twin box...
Detailed understanding of mechanisms underlying DNA damages by low energy beta-rays from tritium is important for evaluation of impact of tritium release from fusion devices to the environment. In this study, the rate of double strand breaks (DSBs) of DNA in tritiated water was measured using a single molecule observation method.
Genome size linear double strand DNA molecules of bacteriophage...
This paper represents the tokamak in-vessel image sequence classification method that used to automatically infer plasma status. Fast framing standard CCD cameras are installed on KSTAR (Korea Superconducting Tokamak Advanced Research) to monitor plasma shape, plasma motion and plasma status. The images generated by the CCD cameras were used for plasma start-up studies and plasma disruption...
Due to the fact that during Tokamak operation，Plasma Facing Materials（PFM）are able to trap part of the fuel（particularly Tritium）, these resident fuel have to be measured and removed. LIBS（Laser induced breakdown spectroscopy） and LIDS （Laser induced desorption spectrometry）are two of the most promising techniques to solve these issues which allow to achieve an on-line and ultra-sensitive...
One of the main difficulties of designing fusion reactor is the development of plasma-facing materials that have to be resilient to the proximity of plasma. Pure tungsten is a primary candidate for this material but has to be strengthened either with particles or fibers to improve its’ brittleness at moderate temperatures and inhibit recrystallization as well as grain growth at higher ones....
The main functions of ITER Gas injection system(GIS) are providing gas fueling(H2,D2, T2, 4He/3He, N2/Ne, Ar) for plasma, wall conditioning operation and neutral beam injectors. If there is leak on the gas supply lines during ITER plasma operation state, abnormal gas composition will affect or have potential to affect operation. Furthermore, Out-leak of Hydrogen or Tritium from gas supply...
The Demo-Oriented NEutron Source (DONES) is an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. An intense flux of highly energetic neutrons is generated by the nuclear reactions of a 125mA beam of deuterons at 40MeV striking a liquid lithium target. The neutron flux achieves a damage rate of 8-10 dpa/fpy in a volume of about 0.3 l...
Magnetic interaction between a tokamak reactor and its iron reinforced-concrete basement has been studied using the analytical model and ANSYS electromagnetic code. When the magnetic material is used for tokamak building, the leakage magnetic field from the tokamak is enhanced due to the normal angle incidence of the magnetic field line to the magnetic material wall. As this study is...
W/CuCrZr PFCs will be used in ITER divertor and are strong candidate for the use in high heat flux regions of the upgraded KSTAR and K-DEMO. Development of hot isostatic pressing (HIP) bonding technology is in progress for the fabrication and qualification of tungsten divertor. We manufactured the first W/CuCrZr flat type small size mock-ups by HIP technology using PVD for interlayer...
Various types of multilayer laser mirrors and piezoelements underwent radiation tests to assess the influence of neutron and gamma-ray fluxes similar to those expected in diagnostic ports of ITER divertor. The optical and thermal performance of laser mirrors and the piezoelectric coefficient of the piezo-elements were under investigation. The test was performed in the RIAR irradiation facility...
High-energy, high-intensity neutrons emitted from the fusion plasma present a stringent environment for the structural materials present in the fusion device. This has significant life-limiting effects on the reactor components. The neutrons interact with the material initiating nuclear reaction leading to the production of radioactive isotopes, gas molecules and material defects. These gases,...
The present work is performed within the framework of the EUROfusion DEMO project. Previously, it was demonstrated that for a maintained magnetic flux the use of HTS conductors at highest magnetic field in a layer-wound CS coil would allow the reduction of its outer radius by around 0.5 m as compared to the DEMO reference design using only Nb3Sn conductors. Alternatively, the superior high...
The neutron fluence is an important normalization parameter for the material specimens to be
irradiated in the Early Neutron Source (ENS). The activation foil method appears suitable for
this purpose considering cost, low technical requirements and invasivness.
Small packages of thin activation foils can be placed in several locations: on the outer surface
of the HFTM, on the outside of...
Korea has designed a helium cooled ceramic reflector (HCCR) breeding blanket for developing the Korean DEMO and fusion reactor, including the development of the reduced activation alloy, ARRA (Advanced Reduced Activation Alloy). From the lesson of the developing procedure of the HCCR test blanket module (TBM) for ITER, it is known that the various fabrication methods, such as electron beam...
Due to its main function as provider of a thermal radiation opaque barrier to the superconducting magnets, the ITER Thermal Shield (TS from now) design guarantees an appropriate thermal behaviour during operation. All the methods and strategies implemented with this purpose on the design, manufacturing and assembly of the TS, constitute the so called TS Thermal Integrity Management. The scope...
The ITER project requires at least two Neutral Beam Injectors, each accelerating up to 1MV a 40A beam of negative deuterium ions, so as to deliver to the plasma a power of about 33 MW for one hour.
Since these requirements have never been experimentally met, it was recognized necessary to build-up a test facility, named PRIMA, that is in an advanced state of realization and which includes both...
In the Water-Cooled Lithium Lead (WCLL) blanket, a critical problem faced by the design is to ensure that the breeding zone (BZ) is properly cooled by the refrigeration system, thus to keep the structural materials under the maximum allowed temperature. For this purpose, CFD simulations are carried over using ANSYS CFX to investigate how the cooling system performances are affected by the...
Over the last few years, new magnetic control algorithms have been developed and tested on the EAST tokamak. The aim is to improve the overall plasma performances and to open the way to the control of advanced plasma magnetic configurations [1]. In order to achieve such an objective, an architecture based on a MIMO plasma shape controller was proposed in [2].
This architecture relies on...
An Ingress of Coolant Event (ICE) is postulated to occur in the ITER Vacuum Vessel (VV) due to a breach on the first-wall cooling channels. The pressure raise in the VV is limited by means of a Vacuum Vessel Pressure Suppression System (VVPSS), consisting of relief lines connected to the VV and discharging the steam to four Vapor Suppression Tanks (VST) partially filled with water: one...
While most of previous numerical analyses have been carried out under thermal and electromagnetic loads due to their significance, severe dynamic loads may also threat its structural integrity. The present study is to investigate resistance of complex ITER divertor module against typical seismic loads. Two kinds of huge finite element models, which consists of cassette body, inner and outer...
Due to the unique combination of excellent thermal properties, low sputter yield, hydrogen retention and activation, tungsten is the main candidate for the first wall material in future fusion devices. However, its intrinsic brittleness and its susceptibility to operational embrittlement is a major concern. To overcome this drawback, tungsten fiber reinforced tungsten composites featuring...
Neutral beam injection is one of the primary auxiliary heating systems for tokamak plasmas. Once the neutral beam leaves the neutraliser collisions with background neutral particles in the beamline and tokamak vessel re-ionises part of the neutral beam. These particles can be deflected by the tokamak magnetic field, potentially damaging unshielded components.
The first stage of the Mega Amp...
The quasi-symmetric fivefold modular Wendelstein 7-X (W7-X) stellarator consists of three groups of coil systems, i.e. superconducting magnet, trim coil and control coil systems. The control coil system contains ten identical 3D shaped control coils (CC) situated behind the baffle plates of corresponding divertor unit, and is designated to rectify the error field and to sweep hot spots on the...
Main goals of breeding blanket development in Korea are to develop and verify the integrated blanket design tools; to develop the core technologies such as blanket materials, blanket cooling, and tritium fuel cycle technologies; and to develop and evaluate fabrication and joining technologies. Several breeding concepts are considered as candidates for the Korean DEMO blanket concept. As a...
On the way towards a comprehensive design of DEMO, step by step all the systems and components must be introduced as their definition or refinement progresses, in order to demonstrate the viability of a design on larger scale, i.e. leaving fewer margins to undetermined questions.
Among the EUROfusion Programme, new aspects have been recently fixed or furtherly developed as the Divertor, the...
The goals of this work are the neutronic modeling of the ITER Upper Port (UP) environment according to the updates of ITER CAD model, the assessment of neutronic effects caused by that update and proposing improvements of the radiation conditions. The update has been applied to the ITER-reference neutronics simulation model called “C-Model” which includes the standard components and generic...
In future fusion power plants, such as DEMO, D-T neutron emission is predicted to exceed 1e21 neutrons/second. Accurately monitoring neutron energies and intensities will be the primary method for estimating fusion power, and calculating key parameters, including the tritium breeding ratio and nuclear heating. The Novel Neutron Detector for Fusion (VERDI) project, implemented under the...
Within the Power Plant Physics and Technology (PPPT) programme of EUROfusion, an intensive development effort is devoted to the detailed design of a solid breeder blanket for a demonstration fusion reactor (DEMO) with the inherent capability of a highly efficient tritium breeding. A novel design of the Helium Cooled Pebble Bed (HCPB) breeding blanket based on a Single Module Segment (SMS) and...
Helium flows at low pressure (0.3 MPa) are used to cool the specimen capsules and the structure of the neutron irradiated High Flux Test Module (HFTM) of the DEMO-Oriented Neutron Source (DONES). The flow path includes inlet and outlet ducts with large cross sections, but also mini-channels with 1 mm gap width, where a high velocity low Reynolds number laminar to turbulent transitional heated...
The Fast Discharge Resistors (FDRs) under development at NIIEFA are intended together with switching equipment to dissipate energy released in case of a quench of the ITER superconducting coils, thereby protecting them against failure. FDRs are made of sections consisting of steel resistive elements enclosed in boxes. Two-four sections stacked vertically form a separate module. During energy...
The ITER bolometer provides an absolutely calibrated measurement of the radiation emitted by the plasma which is a part of the total energy balance. The development is especially challenging because of the extreme environmental conditions within the vacuum vessel (VV) during plasma operation. The bolometer has to guarantee reliable measurements within an environment characterized by high...
The current design baseline for the EU DEMO implements the KALPUREX process for the fusion fuel cycle. This process aims to reduce the tritium inventory by separating hydrogen from other gases within the tokamak building and feeding it back to the matter injection system. The best candidate for the hydrogen separation unit close to the torus is a metal foil pump that relies on the effect of...
System parameters and the optimal radial build of a tokamak fusion system with a normal aspect ratio were found through the coupled analysis of a tokamak system and neutron transport. Neutron impact on shielding and tritium breeding capability are self-consistently incorporated together with plasma physics and engineering constraints in determining the radial builds. The plasma physics and...
As to the ultrasonic testing of argon arc seam of 50mm austenitic stainless steel China Fusion Engineering Test Reactor(CFETR) vacuum vessel mock-ups, there are some limitations if we adopt the traditional ultrasonic probe or linear array phased array probe. In this paper, we designed a Dual Matrix Array(DMA) probe based on the CIVA, and then analyze the optimal principle of the probe...
Insulated pads are used on ITER blanket module connectors and the first wall; their main insulating function is to break any current loop between the shield block and vacuum vessel and/or between the first wall and shield block. The design of the pads consists of a cylindrical or prismatic body manufactured from NiAl-bronze, a ceramic insulating coating (Al2O3 or MgAl2O4) which is applied on...
The proposed work refers to the development of gaseous detectors for application at tokamak plasma radiation monitoring. Soft−semi hard X-ray region radiation measurement of magnetic fusion plasmas is a standard way of accessing valuable information on particle transport and magnetic configuration.
In this work, Gas Electron Multiplier (GEM) based imaging technique is proposed to perform...
Upgrade of the DIII-D neutral beams leads to enhanced heat loads on many components, such as pole shields, calorimeter and collimator. Higher power is now desired for the neutral beams, increasing from 2.6 MW to 3.2 MW per source leading to a normal heat flux loads of up to 55 MW/m2 for the calorimeter. Original designs experienced local melting and fatigue cracks during operation at 2.6 MW....
The design of a major refurbishment of the toroidal complex of the RFX-mod experiment is going to be finalized before starting the realization phase. The Inconel vacuum vessel will be removed and the stainless steel supporting structure will be modified so as to become vacuum tight. The plasma facing graphite tiles will be mounted onto the inner surface of the copper shell, allowing an...
Diagnostic mirrors are planned to be used as plasma-viewing optical elements in all optical and laser-based diagnostics in ITER. Degradation of mirrors due to e.g. deposition of plasma impurities will hamper the entire performance of affected diagnostics. In situ mirror cleaning by plasma sputtering is presently envisaged for the recovery of optical reflectivity of contaminated...
The JET tokamak has been in operation since 1983, producing ~92500 pulses so far. For the period 2000 to 2016 (not including DTE1 in 1997), information on every shutdown, commissioning phase and experimental campaign has been logged, providing unprecedented operation reliability statistics and a model for studying reliability, availability, maintainability and inspectability (RAMI) in fusion...
The Divertor Tokamak Test (DTT) machine has been proposed by ENEA, in collaboration with other Italian institutions, to investigate power exhaust solutions with an experiment integrating all DEMO relevant physics and technology issues. The DTT machine will be able to host, in different phases of its life-time, advanced divertor magnetic configurations (snowflake, super-X, double null) and...
General Atomics is currently fabricating superconducting magnet modules for ITER Central Solenoid in its Poway, CA facility. A critical step during final testing of the modules is high voltage checks of the insulation in Paschen conditions. A qualification coil was fabricated using the same techniques and equipment as the CS Modules. The qualification coil insulation was tested at voltages...
The ex-vessel Remote Maintenance Systems in the DEMOnstration Power Station (DEMO) are responsible for the replacement and transportation of the plasma facing components. The ex-vessel operations of transportation are performed by overhead systems or ground vehicles. The time duration of the transportation operations has to be taken into account for the reactor shutdown. The space required to...
According to the National Fusion Energy Program in Korea, Volumetric Fusion Neutron Source (temporarily called, V-FNS) has been planned and Compact Fusion Neutron Source (temporarily called, C-FNS) development was started at KAERI, which can be used in the fusion and also the fission/industrial applications such as radiotracing isotope production, radiography, and so on, in which the various...
The ITER Heating Neutral Beam (HNB) injector RF plasma source is required to generate a 40A D- or 46A H- ion current, with low electron/ion ratio (<1) and high uniformity over the extraction area (800 mm x 1600 mm). The source prototype SPIDER in the Neutral Beam Test Facility at Consorzio RFX has been developed to demonstrate these performances and it is now under final installation and...
The catalytic separation of hydrogen isotopes is of particular interest for nuclear industry from the point of view of tritium recovery and its use in fusion reactors. Isotopic exchange may take place in the homogeneous (gaseous) phase or in the heterogeneous phase (hydrogen or gaseous deuterium and water or liquid heavy water). Catalysts are necessary both for the homogeneous phase reaction...
The absolute calibration of the detection efficiency for the total neutron yield in the whole plasma is one of the most important issues in the neutron diagnostics such a neutron flux monitor (NFM). In many magnetic confinement devises, those neutron detectors are calibrated by moving or rotating a neutron source such as a Cf-252 radioactive source or a compact neutron generator on the...
In the ITER Magnet System, ten thousand tonnes of superconducting cable – in – conduit - conductor (CICC) are cooled down by a forced flow of supercritical helium, which is supplied from helium inlets. For the ITER Central Solenoid (CS), consisting of six independent pancake wound modules, the He inlets consist of three overlapping holes covered by an oblong shaped boss, welded to the CS...
The authors exposed a radiatively cooled, ~195-mm-long, lithium-filled tantalum heat pipe (HP) to a hydrogen plasma in DIFFER’s linear plasma source Magnum PSI continuously for ~2 hours. We kept the overall heat load on the inclined HP constant, varied the tilt and peak heat flux to ~2.5 MWm2. The HP operated at ~1000-1100 C. Diagnostics included near infra-red thermography from two...
The ITER project is being undertaken at Cadarache, France, to construct and operate an experimental nuclear fusion facility. The aim of this paper is the description of the implementation of the French Order of February 7, 2012, concerning Basic Nuclear Installation (also called “INB”) within the European Union Domestic Agency (EU-DA), specifically on the Electron Cyclotron Upper Launcher (EC...
Presently, the Tokamak T-15MD is being built in the NRC “Kurchatov Institute”. Vacuum vessel was manufactured and passed the preliminary vacuum tests at the plant in St. Petersburg (Efremov Institute) in 2016. Vacuum vessel consists of toroidal shell made of a 321stainless steel of 5 mm and 8 mm thick, horizontal and vertical ports (152 in total), in-vessel elements. The chamber volume is 47...
Based on the reference design HCPB2016 (helium cooled pebble bed) in the pre-conceptual design studies for the European DEMO, the primary heat transfer system (PHTS) for DEMO baseline 2015, and current parameter study for the plasma disruption conditions and the affected FW surface areas, ex-vessel LOCA (loss of coolant) with a double-ended guillotine break of a main pipe in the PHTS has been...
Among the eight core missions towards the realization of nuclear fusion, a future reactor must ensure efficient and safe power exhaust through the divertor and First Wall (FW). The greatest challenges arise from the occurrence of plasma transients. A simulation of a DEMO-like FW Plasma Facing Component (PFC) was carried out assuming Vertical Displacement Event (VDE) and ramp-up limiter...
A RAMI (Reliability, Availability, Maintainability and Inspectability) assessment performed on the ITER Test Blanket Module ancillary systems is presented. The assessment is aimed at evaluating design criticalities possibly jeopardizing the achievement of the overall 75% availability requirements for the considered ITER plant. The Ancillary systems of the European Test Blanket Systems for ITER...
The European Roadmap to the realisation of fusion energy, carried out by the EUROfusion
consortium, considers the stellarator concept as a possible long-term alternative to a tokamak fusion
power plant. To this purpose a pivotal issue is the design of a helical-axis advanced stellarator
(HELIAS) machine equipped with a tritium breeding blanket (BB), considering the achievements
and the design...
The Electron Cyclotron diamond window which is located inside the port cell serves, together with an isolation valve, as primary vacuum boundary between the ITER vacuum vessel, the transmission lines and the atmospheric environment and it functions as confinement barrier. The window consists of an ultra-low loss Chemical Vapor Deposition (CVD) diamond disk brazed into a metallic housing and it...
Neutral gas pressure is one of the main parameters for basic control of ITER operation. Diagnostic Pressure Gauges shall provide pressure measurements in the range from 10-4 Pa to 20 Pa with an accuracy of 20 % and a time resolution of 50 ms. In total 52 DPG sensor heads will be installed in 4 lower ports, 4 divertor cassettes and 2 equatorial ports. The overall DPG system has 15 interfaces...
In the framework of the DEMO divertor project of EUROfusion an extensive R&D program has been carried out to develop advanced design concepts for hot water cooled divertor targets. These plasma-facing components made of W blocks as plasma facing material and CuCrZr tubes as cooling tubes should allow a reliable DEMO operation for 2 h long pulses and maximum heat fluxes up to 20 MW/m². Compared...
This paper presents the recent progress in the pre-conceptual design activities for the DEMO divertor Cassette Body, performed in the framework of the work package “Divertor” of the EUROfusion Power Plant Physics & Technology (PPPT) program. According to Systems Engineering Principles, the divertor CAD model is reviewed, considering the updates in the DEMO configuration model presented by the...
The realization of the 19.6 m² highly heat loaded surface of the actively water-cooled divertor of Wendelstein 7-X (W7-X) requires the installation of 100 target modules distributed in ten discrete similar divertor units. A target module is made of target elements mounted onto rails joined by a stiffening plate forming a frame with an attachment system to the plasma vessel. The target modules...
F4E undertook the qualification of so-called “Additional Suppliers” in order to enhance competition among the potential bidders and secure the procurement of the ITER Divertor Inner Vertical Target.
In order to assess the performances of W armoured Plasma Facing Components under the conditions expected in the divertor target strike point region, a total of 36 W monoblock mock-ups were...
An upgrade to the lower divertor is currently being planned for EAST superconducting tokamak, aiming at >400s long-pulse H-mode operations with a full metal wall and a divertor heat load of ~10MW/m2. A new divertor concept for EAST, “Tightly Baffled Divertor”, suited to water-cooled W/Cu plasma face components (PFCs) with minimized divertor volume, has been proposed to achieve Te,target<5eV...
Within the framework of EUROfusion activities, an alternative Helium-Cooled Molten Lead Ceramic Breeder (HC-MLCB) solid breeding blanket is being also developed at KIT for European DEMO. This concept is proposed as an alternative near-term breeding blanket and it is based on a fission-like “fuel-breeder pin” assembly configuration. Molten lead is used here as the neutron multiplier, Li4SiO4 in...
Stray radiation at 60GHz and 170GHz is an engineering challenge for the integrity of various window assemblies in ITER. Their protection and long term performance preservation are essential for both the operational safety of the device and its scientific exploitation. This contribution focuses on the assessment of Electron Cyclotron Resonance Heating (ECRH) and Collective Thomson Scattering...
This contribution provides summary of two purification experiments of liquid metal breeder Pb-16Li by a cold trap. The behavior of artificially added impurities were studied in non-isothermal ferritic loop Meliloo v1. During these experiments a Mn concentrations followed the solubility curve as published by Barker. More advanced trap design was tested in austenitic loop Meliloo v2. This trap...
To proceed the solid breeder concept for ITER and DEMO it is essential to investigate Ceramic Breeder (CB) materials’ properties. 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 neutron irradiation as well as thermomechanical stresses have to be fulfilled. Lithium...
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 a loss of coolant accident (LOCA) occurred. In contrary to these Be pebbles, Beryllium intermetallic compounds (beryllides) are one of promising materials because of...
The European Gyrotron Consortium (EGYC) is developing the EU 1 MW, 170 GHz CW industrial prototype gyrotron for ITER in cooperation with the industrial partner Thales Electron Devices (TED) and under the coordination of Fusion for Energy (F4E). This hollow cylindrical cavity gyrotron is based on the 1 MW, 170 GHz short-pulse (SP) modular gyrotron that has been designed and manufactured by KIT...
An ongoing study about the influence of neutron irradiation on the mechanical properties of the first wall’s structure materials is presented in this work. EUROFER97 and an Oxide Dispersion Strengthened EUROFER steel were irradiated in the Petten High Flux Reactor up to a nominal dose of 15 displacements per atom at temperatures between 250 and 450°C and investigated by an advanced method of...
China Fusion Engineering Testing Reactor (CFETR) will be built to test and verify the feasibility of engineering and technology in practice for the future fusion reactor. Long pulse and steady-state operation will be demonstrated with duty cycle time not less than 30~50%.During plasma operation, the in-vessel components of the fusion reactor will be activated and contaminated with tritium....
The highly loaded surface of the actively water-cooled divertor of Wendelstein 7-X (W7-X) is made of 100 individual target modules. In each target module, a set of target elements is water-cooled in parallel and fed by manifolds. A target element is made of a CuCrZr copper alloy heat sink, armored with CFC NB31 tiles. Due to the width of the target elements, CFC tiles had to be successively...
In Chinese Fusion Engineering Test Rector (CFETR), blanket is a key component, responsible for producing and transporting tritium, energy conversion and output, so its safety is of particular concern. The water-cooled ceramic breeder blanket (WCCB) is one of three candidate blankets for CFETR. To confirm safety of WCCB, sufficient data are required to estimate the thermal-hydraulic state and...
Four Electron Cyclotron Upper Launchers (EC UL) will be used at ITER to counteract magneto-hydrodynamic plasma instabilities by aiming up to 20 MW of mm-wave power at 170 GHz. This mm-wave power will be injected through eight ex-vessel waveguide assemblies for each EC UL to the in-vessel waveguides. The power exiting the in-vessel waveguides located inside the Port Plug will be directed by...
The First Wall (FW) of DEMO or following fusion power reactors will be exposed to high heat fluxes by thermal radiation and energetic particles from the plasma. During steady state, values of over 1 MW/m² are expected for the EU DEMO concept. The function of the FW therefore relies on (1) good thermal conduction from the plasma facing surface through the channel material, and (2) good heat...
ITER Divertor maintenance equipment work under considerable ambient temperature and radiation load. The heavy components are moved with equipment powered with water hydraulics, with demineralised water as a pressure medium. None of this has yet been tested in ITER-relevant environmental conditions and over projected duty cycles and loading. Hence, a project was undertaken to ascertain the...
The ITER plasma-facing components (PFC) are now fully designed and procurement is underway. A key utility in such design is field line tracing for different magnetic equilibria which allows the definition of component front surface shaping. On ITER, this design phase has deployed both analytic theory [1] and the tracing codes CASTEM and PFCFLUX [2]. Attention is now turning towards the...
Design studies on the helical fusion reactor FFHR-c1 has been progressed. The main goal of the FFHR-c1 is to demonstrate one-year steady-state sustainment of the fusion plasma with self-produced electricity and tritium. The major radius of the plasma, R, is ~10 m and the magnetic field strength at the plasma center, B, is ~8 T. High-temperature superconductor (HTS) magnet coils are adopted in...
The Lower Mirror one (LM1) is part of the in-vessel quasi-optical beam propagation system for the ITER Electron Cyclotron (EC) Upper Launcher (UL), in which each of eight beams of mm-waves are reflected from four mirrors during passage to the plasma. 60000 thermal cycles are foreseen at frequencies lower than 3Hz and power levels up to 1.31MW per beam.
This paper reports the means used to...
The WCLL (Water Cooled Lithium Lead) is a European option of the breeder blanket dedicated for DEMO fusion power reactor as being developed in the frame of EUROfusion’s Power Plant and Technology (PPPT) programme. The intense neutron radiation produced results in a strong activation of the breeder blanket structural elements. The activation and decay heat generation of the WCLL components need...
The validation and testing of tritium breeding blankets concepts, which are relevant for a future commercial reactor, is one of the goals of the ITER project. To achieve these objectives, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are tested in three ITER equatorial ports. Each TBM and its associated shield form a TBM-set that is mechanically attached to a steel frame....
Tritium permeation through structural materials is a significant issue for the Japan’s DEMO reactor blanket concept. Reduced activation ferritic steel F82H is a prime candidate for the blanket structural material. The previous study showed a thin chromium oxide layer formed on a steel substrate worked as tritium permeation barrier; however, heat treatment parameters at atmospheric pressure for...
A key feature of the developed T-15MD tokamak Plasma Control System (PCS) is its ability to rapidly design, test and deploy real-time shot scenario algorithms. PCS platform consist of two levels:
1. High application-specific level: model development and linear approximation, calculation of the experiment scenario, controllers design and experiment simulation (Mathlab Simulink RT...
The control, data access and communication system (CODAC) designed to solve the tasks of planning, preparing and conducting the experiment, collecting, processing and complex analysis of the experimental results at the IGNITOR tokamak fusion project.
It is proposed to build CODAC based on modern failsafe dual-redundant industrial equipment manufactured by National Instruments, Schneider...
Pure tungsten is a potential candidate for armor material of fusion reactors as it possesses superior thermal properties and radiation resistance. Application at the desired operation temperatures for longer times will result in a loss of strength accompanied by embrittlement due to thermal activated changes in the microstructure, in particular due to recrystallization, undermining tungstens...
The Plasma Exhausts Gases (PEGs) proposed to reduce the power load over the plasma facing components are separated by the Plasma Exhaust Processing System of DEMO.
Two kinds of ceramic porous membranes (with top layer of pore size 0.2 m and 3-4 nm, respectively) used commercially for the filtration of liquids have been tested in order to verify their application for the PEG separation. The...
The ENEA Fusion Department (FSN) operates in the field of nuclear fusion under a Quality Management System (QMS) according to ISO 9001 since 2011. At the beginning this methodology applied in R&D activities of a Research Institution such as ENEA seemed to be far from the industrial reality according to an internal and external perspective. But now that the construction of ITER reactor became a...
In the framework of the EU fusion roadmap implementing activities, an accelerator-based Li(d,n) neutron source called DONES (Demo-Oriented early NEutron Source) is being designed within the EUROfusion work-package WPENS as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. DONES will employ a high speed liquid lithium jet struck...
Beryllium was selected as the plasma facing material for the ITER First Wall. Realization of the advantages of beryllium as a plasma facing material depends on the reliability of the critical beryllium joint with the heat sink made from CuCrZr alloy. This paper considers the method of induction brazing as the technology for this critical joint.
To prevent the formation of brittle...
In few last decades, great attention was paid to development in the field of fusion technology. Currently, the International Thermonuclear Experimental Reactor (ITER) is under construction followed by Demonstration Power Station (DEMO) which should be first nuclear fusion power plant in the world. Both of these facilities have one point in common – high power density and thus great demands to...
MITICA is the full scale prototype of ITER Heating Neutral Beam (HNB), designed to deliver 16.5MW of heating power to ITER plasma, currently under construction at the Neutral Beam Test Facility in Padova (Italy). In ITER HNB, negative ions (H-/D-) are produced in the Ion Source (IS) polarized to ground at -1012kV, then extracted by 12kV extraction voltage, accelerated to ground at 1MeV energy...
A design update of the ITER In-Vessel Coils (IVCs) has been launched after the prototype coil manufacture in 2014 revealed some major issues in particular related to brazing and joints inside the coils. In parallel a review and update of the plasma operating scenarios and requirements of the IVCs system has been done and a refined set of plasma pulses and corresponding load scenarios of the...
The in-vessel pressure gauge refers to a vacuum gauge installed inside a vacuum vessel of tokamak. The inside of the vacuum vessel in which the fusion reaction occurs have to discharge the impurities and ashes those generated as a byproduct of the fusion reaction to sustain efficient state. Also the impurities and ashes of the plasma impinging on the divertor plate along the magnetic field are...
The DEMO Oriented Neutrons Source (DONES) consists of complex systems and massive components that need to be on site assembled and maintained. For several of them it is required to perform maintenance, inspection and monitoring tasks over many years in a hostile environment and in efficient, safe and reliable manner. The maintenance of DONES’ systems and components, located mainly in the Test...
In the solid Breeder Blanket (BB) concepts both tritium release and heat recovery depend on the thermal performances of the breeding zone. Within the R&D activities of the Helium Cooled Pebble Bed (HCPB) breeding blanket, the knowledge of the thermal diffusivity of the breeder beds is of fundamental importance to model the transient heat transfer during the power pulses of the fusion machine....
The components of the ITER Diagnostics are located all over on the inner and outer shell of the Vacuum Vessel, in the Ports, on the Divertor Cassettes and in the Cryostat as well. Sensors require electrical transmission lines to transmit both of the diagnostic and control signals across the vacuum boundaries. To transmit the signals, Mineral Insulated cables will be used.
During the last 2...
Heat removal from liquid metal film flow has been widely studied for liquid divertor concepts of fusion reactor. In this study, thermal mixing characteristics of the liquid metal film-flow with locally heated on the surface under the vertical magnetic field was experimentally investigated by using various types of obstacle as a vortex generator. The temperature distributions on the bottom wall...
The International Fusion Materials Irradiation Facility – DEMO-Oriented Neutron Source (IFMIF-DONES) is planned to generate a high flux of 5E16 neutrons/s with First Wall relevant energy spectrum. The High Flux Test Module (HFTM), is the dedicated assembly to bring the material specimens into the high flux region of the neutron source and maintain the specified irradiation conditions.
Based...
In the ITER Heating Neutral Beam Injector (HNB) the remaining charged particles after the neutralization process will be removed by an Electrostatic Residual Ion Dump (ERID) where electrostatic fields are used to deflect the ions that are so dumped on to five panels, each one composed of 18 separate CuCrZr Beam Stopping Elements (BSEs).
The thermal loads applied on panels were calculated for...
The Blanket System provides a physical boundary for the plasma transients and contributes to the thermal and nuclear shielding of the vacuum vessel (VV). It consists of modular shielding elements, the blanket modules (BM), which are attached to the VV. Each BM consists of two major components: a plasma-facing first wall panel (FW) and a shield block (SB). They are connected by means of a...
High environmental constraints are applied on the ITER magnets and therefore on their cryogenics thermometric chains. Accurate and reliable temperature measurements of ITER magnets and their cooling circuits is of fundamental importance to make sure they operate under well controlled and reliable conditions. Therefore, thermometric chains shall reach a high operation reliability. In this...
At the present stage of the Demonstration Fusion Neutron Source (DEMO-FNS) design the actual problem is a development and use of the three-dimensional model of this device to the solution of various neutronics problems for the integration of the basic technological systems of tokamak. The radiation safety and the development of the radiation shield are crucial problems which significantly...
At the present time, in the NRC Kurchatov Institute within the Federal Target Program “Nuclear energy-technologies of new generation for period 2010 - 2015 and to the prospect until 2020” the tokamak T-15MD with supporting facilities is being built. The preassembly of the tokamak T-15MD magnet system together with vacuum chamber was completed at a plant in Bryansk. All elements of the magnet...
Determining tritium concentration within plasma facing components (PFC) of a thermonuclear reactor is crucial in terms of safety. As an example, tritium implantation can be high at the material surface (10-2 % at in W) and low in the bulk of the PFC (10s of ppm). In addition, a simultaneous implantation of tritium, deuterium and helium takes place. An in situ technique used to measure...
Tungsten coatings have received a great deal of attention as a technical solution for plasma facing components (PFC) in present-day tokamaks owing to their advantages over bulk tungsten, such as lower cost and weight. Nevertheless, tungsten (W) coatings are hard and fragile. Their lifetime is mainly limited by two degradation mechanisms occurring during the operation of the tokamak: erosion...
In ITER, each circuits of the central solenoid as well as poloidal field coils PF1 and PF6 is provided with a system for plasma initiation, called the Switching Network Unit (SNU), able to provide up to 8.5 kV for the coils. This will be obtained by inserting resistors in series with the pre-energized coils with the help of a DC current commutation unit (CCU) composed of connected-in-parallel...
In ITER, each circuits of the central solenoid as well as poloidal field coils PF1 and PF6 is provided with a system for plasma initiation, called the Switching Network Unit (SNU), able to provide up to 8.5 kV for the coils. This will be obtained by inserting resistors in series with the pre-energized coils with the help of a DC current commutation unit (CCU) composed of connected-in-parallel...
The Phase-Contrast Imaging (PCI) system is used to measure plasma density fluctuations in the W7-X stellarator at the Max-Planck-Institut für Plasmaphysik (IPP) in Greifswald, Germany. For this purpose, an expanded CO2 laser beam with a wavelength of 10.6m passes through the plasma and the scattered laser beam components yield information on plasma density fluctuations. The laser beam is...
Systems codes are a powerful tool for designing the next generation of nuclear fusion reactors. By exploring a large design space in a single calculation, they can obtain highly optimised solutions. However, while a single design is informative, it does not give the whole picture. Often new designs will push boundaries, whether that involves scaling to new physical regimes or applying new...
Upgrade of the Thomson scattering (TS) system in Versatile Experiment Spherical Torus (VEST) is planned for measuring the electron temperature and density with higher reliability and higher time resolution. The existing TS system has difficulties on measuring single plasma discharge, since it uses a laser with energy of 0.65 J and repetition rate of 10 Hz, while the pulse duration of the...
This paper presents the design activities and test of a vertical target mock-up, developed under the pre-conceptual design phase for DEMO Work Package DIV-1 “Divertor Cassette Design and Integration” under EUROfusion Power Plant Physics & Technology (PPPT) program.
The activities about the Divertor Outboard Vertical Target cooling mock-up are presented in term of CAD model (CATIA),...
COntrol, Data Acquisition and Communication (CODAC) real-time software codes are key elements for the operation of a fusion device as they can play a key role both for the machine protection and for the optimization of the experiments. The updating or upgrading of these software codes may be needed quite frequently in order to either correct bugs or include new functionalities, while these...
The ITER Plasma Control System (PCS) is an essential component for ITER operations. It will include multiple controls loops as well as a number of support functions dedicated to providing input control parameters and distributing commands to actuators. In addition, a supervisory system within the PCS architecture will manage the orchestration of the PCS control loops during the discharge as...
Advanced lithium orthosilicate (OSi) pebbles with additions of lithium metatitanate (MTi) as a secondary phase have attracted international attention as an alternative candidate for the tritium breeding in nuclear fusion reactors. In this research, the formation of radiation-induced defects (RD) in the OSi pebbles with various contents of MTi was analysed using X-ray induced luminescence...
