ITER reached in November 2017 completion of 50% of the work required to achieve First Plasma. Progress is most visible in the completion of many key buildings, such as the tokamak assembly building, the cryogenic plant, and the magnet power supply building have been completed. The tokamak building will be ready for equipment in 2020 and the bioshield is already to full height. Key systems...
The European Roadmap to the realisation of fusion energy breaks the quest for fusion energy into eight missions. For each mission, it reviews the current status of research, identifies open issues, proposes a research and development programme and estimates the required resources. It points out the needs to intensify industrial involvement, to educate the fusion scientists and engineers of the...
In the European road map towards the realisation of fusion energy, one of the challenges is the power exhaust for DEMO. If the ITER baseline strategy can’t be extrapolated to DEMO, tens of years will delay the realization of a fusion plant. So in parallel to ITER exploitation, it is mandatory to test alternative solutions for the heat loads on the divertor as risk mitigation for DEMO.
In the...
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...
Traditionally, remote maintenance in fusion and other nuclear plants has made use of man-in-the-loop telemanipulator devices in order to deal with the relatively unpredictable nature of tasks, and complex environments. Future fusion devices will require maintenance orders of magnitude more complex than at present, however it is infeasible to scale remote maintenance operations teams linearly...
The authors have pointed out that initial tritium needed for starting operation of fusion reactor can be made by DD and low T discharges with self sufficient blankets. Practical commissioning plan of Japanese DEMO was recently planned as a part of DEMO design activity. The early campaigns will require longer than a year of repeated low power pulses for operational purposes as the “power...
Advanced magnetic divertor configuration is one of the attractive methods to spread the heat fluxes over divertor targets in tokamak because of enhanced scrape-off layer transport and an increased plasma wetted area on divertor target. Exact snowflake (SF) for EAST is only possible at very low plasma current due to poloidal coil system limitation. However, we found an alternative way to...
JET presents some unique capabilities: the reactor fuel, ITER wall materials and the capability to confine the alphas. JET next T-T and D-T experimental campaigns can therefore address major physics and technological gaps for the development of fusion energy: the isotopic effects on confinement, the access the H mode and ELM behaviour. The total yield of the final D-T phase is expected to be...
The first of nine ITER Torus and Cryostat Cryopumps has been successfully manufactured and delivered to ITER in summer 2017. This Pre- Production Cryopump is the first of the ITER cryopumps and may be used for the first pump down of the vacuum vessel or the cryostat.
The pump has a 1.8 m diameter and a length of about 3 m and contains cryogenic pressure equipment with a charcoal coated...
The containment vessel of the Joint European Torus is a huge, complicated assembly with a myriad of components, all of which are important for plasma operation. As a research device, JET has been operated over many years and has been extensively rebuilt. During each maintenance shutdown, inspections and measurements of the Vacuum Vessel are carried out by means of dual-camera Stereo surveys,...
There is proposed a new upper divertor for the ASDEX Upgrade tokamak experiment [1]. It is planned to be equipped with internal coils for investigation of advanced magnetic configurations like e.g. „snowflake“. Due to the close vicinity of the coils to the plasma, high induced and very stiff voltages are expected during disruption events. Because only very vague analytical estimates of...
The divertor configuration defines the power exhaust capabilities of DEMO as one of the major key design parameters and sets a number of requirements on the tokamak layout, including port sizes, PF coil positions, and size of TF coils. It also requires a corresponding configuration of plasma-facing components and a remote handling scheme to be able to handle the cassettes and associated...
Within the Early Neutron Source (ENS) project of EUROfusion the design of the accelerator based irradiation facility IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) is under development. The main mission of IFMIF- DONES is to provide the irradiation data needed for the construction of DEMO, a fusion power demonstration reactor developed in the...
Tritium needed as a fuel for fusion reactors is produced via neutron capture by lithium-6 (6Li). However, natural Li contains only about 7.8% 6Li, and enrichment of 6Li up to 90% is required for adequate tritium breeding in fusion reactors. In Japan, lithium isotope enrichment methods have been developed to avoid the environmental hazards of using mercury. However, the isotope separation...
There is an increasing need for integrating individual plasma-control algorithms with the ultimate goal of simultaneously regulating more than one plasma property. Some of these integrated-control solutions should have the capability of arbitrating the authority of the individual plasma-control algorithms over the available actuators within the tokamak. Such decision-making process must run in...
The ITER machine consists of a large number of highly integrated and complex systems, with critical functional positional requirements (e.g. accurate positioning of magnets to minimize error fields and location of plasma facing components with respect to magnetic axis) and reduced design clearances to maximize Tokamak performances and limit costs. Deviations from specified part tolerances and...
The Institute of Plasma Physics of the CAS in Prague has recently started construction of new COMPASS-U tokamak. It will be a compact, medium-size (R = 0,85 m, a = 0,3 m), high-magnetic-field (5 T) device. COMPASS-U will be equipped by a flexible set of poloidal field coils and capable to operate with plasma current up to 2 MA and, therefore, high plasma density (~ 10^20 m^-3). The device is...
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 realization of a DEMOnstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle by 2050, is viewed by Europe as the remaining crucial step towards the exploitation of fusion power. The EUROfusion Consortium, in the frame of the European Horizon 2020 Program, is assessing four different...
Fusion energy becomes essential to solve the problem of increasing energy demands. A high intensity D-T fusion neutron generator is keenly needed for the research and development (R&D) of fusion technology, especially for fusion materials research.
The Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS) has launched the High Intensity D-T Fusion Neutron...
Unique gas retention and transport characteristics of group V elements (V, Nb, Ta) have long attracted a significant interest, in particular among the nuclear fusion community. The nominally high hydrogen isotope permeability and diffusion at the expected operational temperatures, together with the negative activation energy for the solubility present these materials as a promising choice for...
Monitoring the fuel content of plasma-facing components is a key challenge for fusion devices like Wendelstein 7-X (W7-X) [1], equipped with graphite PFCs or ITER with beryllium/tungsten components. In the case of ITER, it is essential to limit the tritium content in the first wall to comply with safety regulations and to sustain the tritium cycle. In W7-X the measurement and control of the...
Tungsten is the main candidate material for the first wall (FW) armour of future fusion reactors. However, a loss of coolant accident with simultaneous air ingress into the vacuum vessel would lead to temperatures of the in-vessel components exceeding 1000ºC, resulting in the formation of volatile and radioactive tungsten oxides. A way to prevent this important safety concern is the addition...
Several breeding blanket concepts for the DEMO reactor employ the eutectic Pb–16Li as breeder material, namely Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL). These three concepts share, with different incidences, three major technological challenges: tritium containment, steel corrosion and magnetohydrodynamic drag. Here, we describe...
The experiments that are planned over the next few years at the Joint European Torus (JET), notably including a deuterium-tritium (DT) experimental phase, are expected to produce large neutron yields, up to 1.7E21 neutrons. The scientific objectives of the experiments are linked with a technology programme, WPJET3, to deliver the maximum scientific and technological return from those...
Europe has elaborated a Roadmap to the realisation of fusion energy in which ‘ITER is the key facility and its success is the most important overarching objective of the programme’. EUROfusion has seized the unique opportunity to develop an integrated programme on devices of different sizes, i.e. on EU Medium-Size Tokamaks (MSTs), and, on JET in order to provide a step-ladder approach for...
JT-60SA is a highly-shaped large superconducting Tokamak under construction by EU and Japan. The mission of JT-60SA is to support ITER and to complement ITER towards DEMO by resolving key physics and engineering issues. Fabrication and installation of components of JT-60SA by EU-Japan Integrated Project Team are progressing on schedule towards the first plasma in Sep. 2020. On the Cryostat...
Wendelstein 7-X (W7-X), a fivefold symmetric stellarator located at the Max-Planck-Institute for Plasma Physics in Greifswald, Germany, was successfully taken in operation with short pulse limiter plasmas in 2015. Hereafter, ten symmetrically positioned un-cooled graphite divertors were installed, the plasma facing wall was refurbished with graphite tiles and various auxiliary systems and...
The European DEMO design will potentially use single phase water cooling in various components that require protection against corrosion damage. Coolant conditions will be similar to fission PWRs but with additional considerations arising from materials choices (Eurofer-97, CuCrZr), 14 MeV neutron irradiation, the presence of tritium, and strong magnetic fields. Presently, many aspects of...
Shaped Tokamak discharges with an insertable polarized electrode have been executed in RFX-mod to achieve H-mode regime. This was aimed at reproducing successful experiments of stable operation at q<2 by feedback stabilization of m=2, n=1 mode already performed with low and high-beta circular discharges. Equilibrium magnetic configurations with a wide range of plasma shapes have been...
In the upcoming operational phase OP1.2b of the Wendelstein 7-X stellarator in 2018 it is planned to have the Neutral Beam Injection (NBI) Heating System operational. Any un-absorbed heating power is dumped on the NBI beam dump graphite tiles that are cooled using CuCrZr-cooling structures. The Heat Shield Thermography (HST) system is present to prevent damage and overheating of the graphite...
The key mission of the new tokamak JT-60SA is to conduct exploitations in view of ITER and to resolve key physics and engineering issues for DEMO. Its pellet launching system was designed to cover according requirements by providing a powerful and flexible tool for the control of density profile and ELM frequency. Therefore, the systems lay out had to be adapted for pellet injection via a...
The reduction of heat loads of divertor target is one of the main challenges addressed by the European roadmap to the realisation of fusion energy. In particular, eight different missions have been identified overall, of which Mission 2 ‘Heat-exhaust systems’ is specifically devoted to this goal. Recently, the Divertor Tokamak Test (DTT) facility [1] has been proposed with the aim of...
Controlling the plasma density in a future fusion reactor will be mainly attributed to pellet injection using a control algorithm based on a rather difficult density measurement. The underlying technology to capacitate the Pellet Launching System (PLS) for the requirements is challenging. The ASDEX Upgrade (AUG) PLS was retrofitted for this task, intensifying the integration into the Discharge...
During disruptions runaway electrons(REs) often drift from high field side to low field side in J-TEXT. It may damage plasma facing components when REs strike the first wall with high energies. In order to mitigate the damage, a novel approach called magnetic energy transfer(MET) based on the principle of electromagnetic coupling is presented in this paper. A set of extra coils with a high...
The project of tokamak Ignitor is one of the main themes of long-term scientific cooperation between the Russian Federation and the Italian Republic. Currently, negotiations on the development of technical design tokamak Ignitor with placement on the site of TRINITI (Moscow, Troitsk, Russia). The discussion on preparing of the Russian-Italian Inter-government agreement on realization of...
In order to operate a large research facility, one needs software tools assisting the organization and the operation follow-up. In the past, separated software tools were used on Tore Supra but for West, an integrated approach was chosen. The West Operation management Software Suite (WOSS) allows a streamline management of information from the planned program up to the realized experiments and...
The EC-system of the TCV tokamak is progressively being upgraded with the addition of two MW-class dual-frequency gyrotrons (84 and 126GHz/2s/1MW) being manufactured by Thales Electron Devices with the first gyrotron delivered to SPC at the end of 2017. In order to connect the two gyrotrons to the existing low field side and top launchers, new waveguide routing from gyrotron hall to TCV...
DIII-D plays a vital role in the development of the physics basis for fusion energy and the ITER design. Designed in the 1970’s and built in the early 1980’s, the system started operations in 1986 and has provided a reliable platform for fusion experiments for over 30 years. A hallmark of DIII-D operations has been its ability to adapt to the changing needs of the fusion research community...
Neutral Beam Injection (NBI) is a robust, established heating and current drive method in fusion experiments. Among its strengths is high current drive efficiency that may pave the path for steady state operation of a tokamak reactor with an economically viable recirculating power fraction. For large tokamaks like ITER and DEMO the use of negative ions is mandatory due to the vanishing...
The future EC systems will consist of several gyrotrons sources providing MW-level millimeter wave power at a frequency around or above 170 GHz. The development of matched loads is necessary to test the new sources, the components for the transmission lines and the launchers, and must ensure high qualification for compatibility with the nuclear environment. The load low reflectivity and high...
Achieving the plasma temperature expected for nuclear fusion requires external heating systems, such as dedicated Radio-Frequency antennas. Dimensions, power level and manufacturing cost which are at stake make it impossible to build scale-one mock-up during design and prototyping phases. For that reason, modelling the electromagnetic interactions between magnetized plasmas and Radio-Frequency...
High energy (800 keV) Neutral Beam Injection (NBI) is one of the methods being considered to heat EU DEMO plasma [1]. A major issue of present NBI systems is the limited efficiency of the gas neutralizer (for ITER NBI ~55%), which impacts on the overall system efficiency. An attractive method, but still undemonstrated at full performances, is the photo-neutralization of the negative D-ion...
This paper presents a milliwatt-range testbed that has been recently designed and manufactured for the RF characterization of the WEST ICRF launchers. The low power testbed is integrated into the TITAN test facility. This extends the capabilities of TITAN from testing at high voltage/high current parts of the launchers in vacuum, to the characterization of the launchers coupling capabilities,...
To decrease the power density and associated high voltage a distributed antenna system is proposed as ICRH system for the reactor. Among the different solutions a layout made from a set of TWA sections is considered as the most promising [1, 2]. It optimizes coupling to the plasma, is load resilient and avoids large values for the VSWR in the feeding lines. The total radiated power scales as...
Overmoded corrugated waveguide is used in high power microwave applications such as Electron Cyclotron Heating systems, where it is necessary to transmit high power at very low loss. In the primary propagating HE11 mode, corrugated waveguide is effective over a large frequency bandwidth. This operational flexibility becomes important in multi-frequency systems. For 50-mm diameter aluminium...
The accurate determination of the emitted radiation is an important element in the interpretation of Tokamak performance and in the design of experiments. The spatial distribution of the total emitted radiation is typically determined with quite sophisticated tomographic techniques. On JET, a new tomographic inversion method, based on the Maximum Likelihood, has been very recently developed...
The construction of SPIDER, the experimental device of the ITER Neutral Beam Test Facility (NBTF) devoted to the study of the ion source, is completed and SPIDER will soon start operation. The construction of MITICA, the full-size prototype of the ITER HNBI, is in progress. SPIDER and MITICA operation presents many possible hazards including radiation (D.Lgs.230/95–Cat.A), electrical,...
The integration of heating current drive (HCD) systems in the EU DEMO tokamak must address a number of issues, namely space constraints in the tokamak building, remote handling requirements, breeding blanket penetration, neutron and photon radiation shielding, compliance of penetrations of the primary vacuum with safety and vacuum criteria, and a large number of loading conditions, in...
The low aspect ratio (a/R=2.2) D-shaped tokamak T-15MD with toroidal field of 2T on axis is currently under construction in the Kurchatov Institute [1, 2]. Ion-cyclotron resonance heating (ICRH) is considered as an important heating method for this device [1]. In addition, ion cyclotron current drive (ICCD) could contribute to sustain the non-inductive plasma current for long-pulse operation....
Neutral Beam Injection (NBI) is used for non-inductive heating, current drive, fueling and diagnostics in most major magnetic confinement fusion devices. The DIII-D device comprises eight NBI ion sources based on the US Common Long Pulse Source (CLPS), with a total output power of 20 MW.
Here we report on efforts to improve performance and longevity of the NBI system by initiating a R&D...
The test facility BATMAN was dedicated since its start in 1996 to the development of radio frequency driven negative hydrogen ion sources for ITER NBI with focus on formation and extraction of negative ions, technological developments and improved concepts. During 2017 the test facility has been upgraded in order to replace the former extraction system (which was derived from a positive ion...
Within the Power Plant Physics and Technology (PPPT) programme in the EUROfusion Consortium design activities are currently in progress for the development of a DEMOnstration Fusion Power Plant (DEMO). In this framework, the design of the machine and the integration of in-vessel components require neutronics analyses fundamental to verify the tritium self-sufficiency, the shielding...
A high-powered “comb-line” helicon antenna for use within the DIII-D Tokamak is currently in design and fabrication at General Atomics. The antenna will drive current in high beta discharges using electromagnetic helicon waves. The high powered helicon antenna (HPHA) is expected to couple up to one MW of power into DIII-D plasmas at a frequency of 476 MHz. The antenna design includes 30...
Arc detection is an essential protection system for high power RF systems. It is commonly realised by monitoring the Voltage Standing Wave Ratio (VSWR) in the transmission lines. The JET ILA is a load tolerant ICRF antenna composed of 8 short straps grouped in 4 Resonant Double Loops (RDLs). In this type of antenna, there is a low impedance section in which the standard VSWR protection is...
The WEST tokamak is aiming at testing ITER like divertor component. This requires to address new control challenges like X-point configuration magnetic control or heat loads control in metallic environment and event handling challenges to sustain long duration H-mode that are in line with ITER needs.
To address these requirements, a new Plasma Control System (PCS) has been built using a...
The future AUG control research is expected to cope with a large number of control tasks using a limited number of actuators in high performance regime. Essential part of a control system for future tokamaks is an intelligent actuator management that will be responsible for allocating the most convenient actuators to the control tasks of the highest importance.
Activities in this field have...
High-speed long pulse archiving systems are critically sensitive to the latencies produced by hardware and software along full archiving chain. Therefore, detailed studying of this phenomenon, estimating its impact on archiving process, correct selecting and commissioning of the hardware for archiving purpose, optimizing of archiving system configuration are indispensable steps to achieve...
Wendelstein 7-X (W7-X) stellerator has been designed to support a long-term and continuous operation. In that concern, corresponded scientists have access on the archived data (signals) anytime-anywhere, where archived signals can be referenced via project-specific unique identifiers, referred to as signal-addresses.
At the same time, different projects in the fusion research such as W7-X and...
Wendelstein 7-X (W7-X) completed its second operation phase (OP1.2a) in December 2017. A large number of diagnostics were operated in nearly 1000 experiment programs by an international research team. For the documentation of W7-X experiment programs, a new electronic logbook software was developed and eventually used for the first time in OP1.2a. The software was designed for the needs of...
Identifying the plasma equilibrium operating space in terms of e.g. plasma current Ip, internal inductance li(3) and magnetic flux state ψ is a central task in the design of future tokamaks. The operating space is typically limited (for a given plasma shape) by constraints on the Poloidal Field (PF) system such as maximal allowable currents, fields and forces in PF coils. The typical tool to...
Discharge scenarios and control schemes in ASDEX Upgrade (AUG) are evolving more and more complex. Especially in physics investigations for ITER and DEMO sophisticated scenarios exploit the operational space. This increases the probability of design flaws or human errors in the pulse configuration, but also aggravates the potential damage in the failure case.
The ASDEX Upgrade Flight...
The construction of MITICA, the full-size prototype of the ITER Heating Neutral Beam Injectors (HNBs) is in progress at the Neutral Beam Test Facility (NBTF) located in Padova, Italy.
The design of the central control (CODAS) and interlock (CIS) systems is progressing taking into account the requirements coming from the MITICA plant units, in terms of number and type of interface signals, data...
Adequate avoidance and mitigation of disruptions must be ensured if ITER is to meet its objective of high performance burning plasma operation. A comprehensive disruption mitigation system (DMS) is being designed to ensure that thermal, electromagnetic and runaway electron (RE) loads are reduced to tolerable levels. The strategy relies on the injection of impurities/fuel using an array of...
The ISTTOK, a large aspect ratio fully ohmic tokamak operated at IPFN-IST is presently scientifically exploring an AC) regime, aiming to extend much longer pulses, up to one second plasma and around 40 current inversions.
The control of the earlier single-pulse plasma formation and sustaining was essentially deterministic using pre-programed delays on a set of timing channels generated within...
The ITER Neutral Beam Test Facility (NBTF) is currently under construction at Consorzio RFX, Padova, Italy. The NBTF includes two experimental devices: SPIDER for the study of the ion source and MITICA for the development of the full-size HNB prototype. Even if the two experiments share many architectural aspects in their Control and Data Acquisition System (CODAS), there are nevertheless some...
It is possible to decompose an event prediction problem into a hazard function (event rate model) and a phase space trajectory (dynamical system evolution). In contrast to typical event prediction approaches (such as those attempted in present tokamak disruption systems) the hazard function has two significant advantages. First, it has a time localized and quantitative interpretation (events...
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...
Bolometer cameras in ITER will be mounted in Port Plugs, on Divertor Cassettes and on the Vacuum Vessel (VV) wall behind Blanket Modules (BMs). For the first assembly phase the platform (Cable fixations and the lower part of the internal signal chain) of VV cameras has to be delivered to fix the signal cables and protect their termination. During First Plasma, the as-built magnetic axis and...
We report on the selection, implementation and successful demonstration of a new automated laser control system for JET’s Far Infrared Interferometer, a diagnostic essential for machine protection. The new control system allows all laser subsystems and sensors to be interlocked and operated remotely in a precise and preprogramed manner, functionalities that are essential for reliable operation...
For the future JET deuterium-tritium (DT) campaigns different gamma diagnostics, in particular the JET Gamma-ray Camera (GC) and the JET Gamma-ray Spectrometer (GS), have been upgraded in the last few years. The main demands for new detectors to be used during DT campaigns are connected with expected high count rates of 0.5 Mcps and with a required energy resolution equal or better than 5% at...
Portable neutron generators (NGs) are widely used in many applications e.g. medicine, materials analysis and plasma diagnostics calibrations. The NGs based on DT reaction provide a controllable 14-MeV neutron emission of high-intensity flux. The knowledge of the total neutron yield is important, and in some of the applications energy distribution of produced particles are also essential....
H-alpha and Visible Spectroscopy is one of the ITER first-plasma diagnostics providing full poloidal coverage of plasma scrape-off layer near the first wall. There are two poloidal-view channels in EPP11, one tangential-view channel in EPP12, and one divertor-view channel in UPP02. At the moment, the final design phase is ongoing, requiring proper testing of design solutions to identify the...
Thomson scattering (TS) system is developed to measure the electron temperature and density of Versatile Experiment Spherical Torus (VEST). Since it is the key diagnostics for measuring the local electron properties of the core plasma, each part of the system is carefully designed to provide a reliable measurement result. Besides, as additional heating devices such as neutral beam injection...
ITER and DEMO will use a optimum 50%:50% deuterium-tritium gas mixture as fuel. The fusion reaction rates depend on the hydrogen isotope ratios, therefore, these ratios are important to monitor both in the confined fusion plasma itself and in the pump ducts. Penning gauge spectroscopy of Balmer-α lines of hydrogen isotopes is widely used in present-day experiments to determine the hydrogen...
On Wendelstein 7-X a Sodium beam emission spectroscopy (BES) diagnostic system has been installed in 2017 in order to measure plasma edge density and turbulence. The diagnostic setup consists of two parts: an alkali beam injector and an observation system through which we can observe the light emission by the alkali beam.
The observation system consists of two parts, which operate in parallel:...
Within EUROfusion WPJET3 programme, the unique 14 MeV neutron yields produced in the scheduled JET DT campaign will be exploited to validate codes, models, procedures and data currently used in ITER design in order to reduce the related uncertainties and the associated risks in the machine operation.
One relevant experiment selected for DT is the irradiation of the HCPB-TBM mock-up of ITER...
LD pumped Nd:YAG lasers developed for ITER Divertor Thomson Scattering (DTS) diagnostic can operate with high power 3ns/(1–2)J/(50–100)Hz at wavelengths 1064nm and 946nm.
1064nm Nd:YAG laser technology is well established and, when creating the laser, we are mainly focused on quality of the laser radiation. The laser beam quality is usually affected by thermally induced lens and birefringence...
The first mirrors of optical diagnostics in ITER are exposed to high radiation and fluxes of particles which escape the plasma, in the order of 10^20 m-2 s-1. They are thus the most vulnerable optical component in the optics chain inside port plugs, being subject to erosion, especially by fast charge-exchange neutrals, or to deposition of impurities at flux rates which can reach 0.1nm/s. The...
An important goal of tokamak plasmas is the control of magneto-hydrodynamic (MHD) instabilities with low m, n (poloidal and toroidal mode numbers), which can influence the confinement time of energy and particles and possibly lead to plasma disruption. These instabilities, which appear as rotating magnetic islands, can be reduced or completely suppressed by a current driven by electron...
The Charge Exchange Recombination Spectroscopy diagnostic system on the ITER plasma core (CXRS core) will provide spatially resolved measurement of ITER plasma parameters. The optical front-end is located in upper port 3 and the collected light in the wavelength range of 460 nm to 665 nm is routed to spectrometers housed in the tritium building.
Continuing the efforts described in [1] of...
A calibration procedure is proposed for the entire lifetime of the ITER Radial Neutron Camera (RNC) diagnostic. The proposed calibration is divided in different phases: pre-delivery calibration, post delivery calibration and periodic calibrations at different time intervals. The RNC calibration relies exclusively on radiation sources available at the domestic agency in the pre-delivery phase...
A new Infrared diagnostic has been developed by IRFM and installed in the WEST tokamak to measure surface temperature of the actively cooled W-monoblocs components as foreseen for the ITER Divertor units, with a very high spatial resolution of 100µm.
The goals are to investigate the effects of the shaping of these components on the heat load deposition pattern, the evolution of damages...
To determine the power produced in a fusion device, an accurate estimation of the neutron yield is necessary. It is a fundamental operational quantity and, being linked to plasma performance parameters, it is an important measure of fusion success. The neutron yield is also needed to support the operational safety case and is the prime input to operational and maintenance doses.
A second...
COMPASS-U [Panek et al. Fus. Eng. Des. 123 (2017) 11-16], a high magnetic field tokamak with hot walls, will be designed and built at IPP Prague replacing the currently operated COMPASS tokamak. Unique features of this new device bring noticeable constraints and requirements, which make the development of necessary plasma diagnostics highly demanding. In the contribution, main expected...
The main objective of WEST is to study the behavior of the ITER like Plasma Facing Components (PFCs) and to test the resistance and ageing of these components under high heat loads. To achieve these objectives, two independent thermal diagnostics have been developed and installed in the lower divertor of the WEST tokamak. The first one is based on 20 thermocouples embedded at 7.5 mm from the...
The presentation is focused on approaches and results of simulations and used for loading analyses made for new design of the Divertor Thomson Scattering (DTS) in-vessel equipment, including spatial stress strain state, seismic analysis, electromagnetic analysis as well as the most important load combinations.
The ITER Divertor Thomson Scattering system is designed to provide an instrument...
Tangentially viewed images of plasma from high-speed cameras are intuitive and reliable data which could be used to reconstruct the plasma boundary. A new image acquisition and processing system has been developed for optical boundary reconstruction on the Experimental Advanced Superconducting Tokamak (EAST). The head section of the optical system is an imaging lens, followed by fiber optic...
In many nuclear applications, sensors are widely used in order to detect high energy particles; one of the available technologies is the scintillator, which is generally coupled with a photomultiplier and pulse amplifier. The different particles incident on the scintillator produce electrical pulses having different shape; moreover the amplitude of these signals is related to the particle...
Thomson scattering diagnostic system consists of a laser, a collection system, spectroscopy and a digitizer section. Recently, KSTAR Thomson scattering system has found some problems in collection system. The first problem is that the light transmission of the lens glass drops to less than 10% due to the browning of the lens due to the neutron, and when the plasma disruption occurs, the impact...
The first mirrors of all optical diagnostics will be exposed to the fluxes of neutrals, mainly D, T and Be, from the plasma in ITER. This can lead to formation of erosion and deposition zones on the mirror surface. The location of the zones will depend on D,T/Be flux ratio and geometry of the cutouts in the diagnostic shield module (DSM). In H-alpha diagnostics, the first mirrors in equatorial...
The reliability of the optical diagnostics in ITER critically depends on radiation resistance of the fiber optics for a transmission of plasma light to remote detectors. The design of H-alpha diagnostics includes fiber bundles about 60 m long between the port cell and the diagnostic room. The first 10 m of the bundles run through the gamma-neutron fields. This part of the bundle will...
The detection of retained nuclear fuel in plasma facing components (PFCs) is currently one of the critical issues for ITER because of the impact tritium can have on the machine operation and safety. Laser Induced Breakdown Spectroscopy (LIBS) is a promising technique providing both qualitative and quantitative composition of the chemical elements retained in PFCs: it does not require sample...
Unidirectional carbon fiber-carbon matrix (CFC) composite tiles can be used to diagnose the main features of a particle beam such as its power, its divergence and uniformity. The diagnostic calorimeter STRIKE will be used with such aim for the negative hydrogen ion beam produced by the ITER ion source prototype SPIDER, starting operation at Consorzio RFX (Padova, Italy) in 2018. By exposing...
The ITER bolometer diagnostic shall provide the measurement of the total radiation emitted from the plasma, a part of the overall energy balance. Up to 550 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 performance of the...
Abstract- As an important component of ITER PF6 coil, the superconducting joint is used for the mechanical and electrical connection between inner conductors and also between the coil and the outer feeder. PF6 coil plays the role in plasma current drive, position and shape control. Both the working current and the rate of working current change are high, and it works in the complex magnetic...
The helium inlet is one of the most important components of ITER Poloidal Field (PF) coils. The insulation structure of helium inlet is critical to provide sufficient electrical and mechanical properties in practical application. In this paper, an ITER PF6 coil double pancake helium inlet trial mock-up was designed and manufactured by simulating the actual manufacturing process. A thermal...
The Swiss Plasma Center (SPC) has developed a Toroidal Field (TF) layout for the EUROfusion DEMO tokamak, based on a reference baseline of 2015. Each TF coil winding pack consists of 12 single layers wound with Nb3Sn graded conductors, connected in series by inter-layer joints, which are embedded in the winding pack. The react-and-wind (R&W) manufacturing technique is foreseen for the TF coil...
The coupling currents loss for fusion conductors is frequently assessed applying a sinusoidal field sweep of fixed, small amplitude and variable frequency. From the initial slope of the loss curve, the coupling loss time constant is derived and applied in the loss calculation over the whole range of field transient. In this work, the traditional AC loss assessment is compared to an...
The high current dc disconnector is the significant switch in ITER poloidal field converter power supply system. The high temperature rise in the contact area of disconnector is the key factor which limit the capacity of current carrying. In order to reduce the contact resistance and improve the capacity of current sharing and carrying, two new contact structures of circular contact and ring...
Nb thin films on different substrates have been prepared by DC magnetron sputtering for capacitor application. The influences of substrates on film morphology, crystallographic structure and characters have been investigated. Superconducting transition temperatures of the films have been measured and relationship between the transition temperatures and deposition conditions have been studied....
One of the biggest challenges for a fusion reactor with magnetic confinement is the controlled removal of the heating power. ASDEX Upgrade (AUG) is one of the leading experiments in this area and investigates integrated solutions that combine high heating power and wall materials suitable for reactors. To increase the normalized output power towards the values intended for ITER and DEMO, the...
To qualify the insulation design to be applied on the ITER PF6 coil, two short beam-shaped mock-up’s taking the form of the conductors combined with insulation in the coil have been tested in simulated Tokamak operation environment. All the results of the mechanical and electrical tests, including compressive fatigue, push-out, thermal cycling, DC and AC high-voltage withstanding, AC partial...
The International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) AC/DC in series converters are composed by three converter units in series to supply megawatt energy to PF coil. With the characteristic that high power, complex operating modes, large amount of snubber capacitors and stray inductances, any inappropriate starting mechanism could introduce over-voltage and then...
The International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) AC/DC. converter is composed by thyristor-based phase controlled converter modules. As the core component of ITER PF AC/DC. converter, the thyristor is very sensitive to over-voltage and could be broken down in microseconds, therefore, the transient over-voltage protection strategy is desperately essential to...
The ITER magnet system comprises 30 main superconductive coils, which will create steady-state and slowly varying magnetic fields with a total energy up to 50 GJ. The fast protective discharge of this energy in case of a quench (superconductor-to-normal transition) is provided by closing the coil current circuits with the help of protective make switches (PMS) and inserting discharge resistors...
Due to safety and reliability concerns, high voltage fast switch is required for Tokamak power system. It can cut off the high voltage power supply within a little time usually in microsecond, when fault occurs, system tests and load switches. IGBT is an ideal choice as the basic component of the switch to satisfy the fast response requirement. Obviously, connecting multiple IGBTs in series...
PF5 and 6 coil assembly tool is used to transfer, support and align the PF 5 & 6 coil. The tools are comprised of PF5 lifting adapters, PF5 temporary support and align units, PF6 lifting adapters, and PF6 temporary support and align unit. PF 5 and 6 coil will be lifted from assembly hall to the tokamak pit using PF 5 and 6 lifting adapter. Then, PF5 and 6 will be temporarily placed on their...
Cryogenic circuit for cooling of a superconducting magnet like tokamak has a lot of branch and has to be designed efficiently considering the conductance for cooling. The KSTAR PF cryogenic circuit has one hundred eight cooling paths for fourteen superconducting magnets and CS structure. The five cryogenic valves has been installed to provide the same mass flow rate to cooling channel of...
Tearing Mode (TM) creates magnetic islands in the tokamak, which will cause mode locking and major disruption. A new method for applying modulated magnetic perturbation is explored to suppress magnetic island and accelerate island rotation by using external resonant perturbation (RMP) coils in the J-TEXT tokamak. The phase difference between TM and external RMP is denoted by Φ. RMP has a...
China has started the research and development of the negative-ion-based neutral beam injection (N-NBI) prototype for Chinese Fusion Engineering Testing Reactor (CFETR). The prototype needs an acceleration grid power supply (AGPS) rated at 200kV/25A/3600s. A single stage inverter-type high voltage power supply is applied as AGPS of the prototype. The AGPS consists of phase-controlled...
In order to form an independent capacity of design and engineering construction for negative-ion-based neutral beam injection (N-NBI) system, and lay the foundation for the construction of Chinese Fusion Engineering Testing Reactor (CFETR). China has started the research and development of the N-NBI prototype for CFETR. The prototype is designed to accelerate hydrogen negative ions up to...
The Linear IFMIF (International Fusion Materials Irradiation Facility) Prototype Accelerator (LIPAc) injector consists of a 140 mA proton/deuteron source, its associated low energy beam transport line (LEBT) as well as ancillaries such as water cooling skid, vacuum groups, High Voltage Power Supplies (HVPS), etc . A specific element, the beam “Chopper”, was included in the LEBT to generate...
The JET tokamak is connected to the United Kingdom 400kV National Grid by three Super Grid Transformers (SGTs) through a 36kV power network. The 36kV system supplies power to the toroidal (TF) and poloidal (PF) field coils, heating systems and their auxiliaries. The voltage drop on the system is limited. Dropping below 30kV will first trip the heating systems followed by a total loss of power...
In order to realize the 200keV negative-ion-based neutral beam injector (N-NBI) prototype of China Fusion Engineering Test Reactor (CFETR), an acceleration grid power supply (AGPS) is under development. The AGPS adopts single stage inverter type topology and is rated at 200kV/25A/3600s. Here, the paper presents a control system that can fulfill the high requirements in both steady state and...
The ITER Magnets are in the procurement and assembly phases.
During these phases, critical components need to be tested and assembly procedures need to be developed and qualified on mock-ups.
To this end, ITER Organization (IO) and CEA have built up a support structure with space, expertise and equipment: MIFI – Magnet Infrastructure Facilities for ITER.
In this framework, IO and CEA perform...
The KSTAR has seven pairs of ring-shaped poloidal field (PF) superconducting coils with rectangular cross section. The central solenoid (CS) is a vertical stack of four pairs of PF coils compressed axially by preloading structures. The axial compression (remaining preload) on the CS coils is monitored by a strain measurement, one of the important monitoring parameters for safe operation of...
The Wendelstein 7-X stellarator (W7-X), one of the largest stellarator fusion experiments, will start the third plasma operation campaign mid of 2018 at the Max Planck Institute for Plasma Physics in Greifswald, Germany. The main objective of the experiment is to prove the reactor relevance of the stellarator design.
The W7-X experiment has a superconducting magnet system with 50 non-planar...
ITER (IO) magnet coil power supply system is the world largest AC/DC conversion system which is jointly contributed by China, Korea including PF, CS, VS, TF and CC AC/DC converters with total capacity 2853 MVA and the associated Fast Discharge Unit, Switch Network Unit and DC Busbar from Russia Federation. All ITER coil power supply AC/DC converters will be installed in the magnet power...
Segment-fabrication of high-temperature superconducting (HTS) magnet is an attractive concept to solve engineering issues of helical fusion reactor having huge and complex superconducting helical coils. There are two designs as the segment-fabrication: 1) Remountable (demountable) coil option; the entire half-pitch helical coil segments are connected with demountable multi-conductor joints, 2)...
In ITER, DC busbars will be used to connect the AC/DC converters to the superconducting coils of the magnet system and will run from buildings B32, B33 through building B74 to TOKAMAK building B11; the total length will exceed 5 km. The busbars will be interconnected by flexible links for thermal expansion compensation.
The busbars used in the TF, PF/CS and CC coil power supply systems are...
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 synergistic effects of transient heat loads in conjunction with stationary plasma were investigated. All tests were executed in the linear plasma device PSI-2 at a base temperature of around 800 °C, which was achieved by the plasma exposure and an ohmic heater attached to the sample holder. Moreover, to simulate the ELM-like transient thermal events, a Nd:YAG laser with a wavelength of...
The Plasma Facing Components (PFCs) of the National Spherical Torus Experiment Upgrade (NSTX-U) protect the vacuum chamber wall from high plasma heat fluxes which are mostly carried by energetic particles that flow along magnetic field lines. The magnetic field lines will have very shallow impingement angles to the PFC surfaces (as small as 1o), a consequence of flux expansion at the divertor....
In the Stellarator Wendelstein 7-X with its twisted 3D magnetic field geometry, studies of material migration with respect to first wall components becomes very important in view of the envisioned long-pulse operation. A variety of erosion/deposition probes were installed on graphite plasma-facing components exposed at three different nominal heat load levels between 0.1 and 10 MW/m2. At the...
ITER and DEMO plasma facing components (PFCs) should remove the extreme heat flux up to 10 MW/m^2 and the various type of PFCs have been developed for enhancing the heat transfer performance such as hypervapotron and twisted tape insertion. For the limitation of complexity to mechanical machining, three-dimensional (3D) metal printing technology by direct energy deposition is used to fabricate...
The total area of the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER) is 650 m2, 40% of which is the responsibility of the RF DA. The NIIEFA should manufacture and test 179 first wall panels (FWP), which requires 7000 high-heat-flux units and 305 000 beryllium tiles. As is known, beryllium is a toxic material, therefore stringent requirements are imposed on the...
In-vessel plasma facing components such as first-wall, blanket and divertor modules should withstand harsh design conditions. In particular, since the divertor module undergoes extreme thermal loads, several tests for mono- and multi-block mock-ups as well as lots of stress analyses for the mock-ups and module themselves have been carried out. However, there were a little fracture mechanics...
The First Plasma Protection Components (FPPC) are a set of structures temporarily installed inside the vacuum vessel specifically for first plasma operation. In addition to aiding plasma generation, these components protect the unshielded vacuum vessel and ancillary systems from the plasma. Upon completion of first plasma operation, these components will be removed.
A significant design and...
The thermal property of the jet stabilized by an internal flow resistance has been investigated in order to apply the liquid metal divertor consisting of molten tin shower jets named the REVOLVER-D, to the helical fusion reactor, FFHR. The allowable heat load on the REVOLVER-D is higher than that of conventional solid divertors. The droplet formation of the jet can be avoided by inserting an...
As part of an ongoing divertor upgrade of the TCV tokamak [1] it is planned to add gas baffles on the inner and outer wall of the vacuum vessel to form a divertor chamber of variable closure. The baffles promise to increase the compression of neutral particles in the divertor and, thereby extend the divertor research on TCV towards more reactor relevant, highly dissipative divertor regimes....
Recent efforts dedicated to the mitigation of tungsten (W) brittleness have demonstrated that tungsten fiber-reinforced tungsten composites show toughness even at room temperature. This is caused by extrinsic mechanisms induced by the incorporated tungsten wire used as reinforcing element. High temperature operation and manufacturing of the fiber-reinforced composites might result in a change...
Partly insulated fin structure has been proposed to mitigate the temperature stratification in the flowing-type liquid metal divertor. This structure consists of partly insulated fins which are infused in the flowing liquid metal. Our previous study observed the generation of a wavy flow by a checker-board like arrangement of insulated parts experimentally and numerically. Moreover, magnitude...
The elucidation of hydrogen recycling in plasma facing materials is one of key issues to sustain the steady state plasma during fusion operation. QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) is operated by only hydrogen plasma with all metal plasma facing wall under higher wall temperature of 473 K. However, a mixed material deposition layers contained with carbon,...
Plasma must be heated by external heating for ignition in the future fusion reactor, ICRF heating is a favorable high-density plasma heating method since the fast wave launched from ICRF antenna can be transmitted to plasma core even in high-density plasma. With the heating power larger, the exposed antenna surface enduring heat becoming higher, Faraday shied (FS), as one of the key...
Within the framework of the Work Package DIV 1 - “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the thermal-hydraulic performances of the DEMO divertor cassette cooling system. The research activity has been focussed onto the most recent design of the Cassette Body (CB)...
The on-line measurement, removal and recovery of hydrogen isotopes in plasma-facing materials are important issues for Tokamak during long-time discharge operations. The laser induced desorption system (LIDS) was designed and built from the laser induced breakdown spectroscopy (LIBS) system with a quadrupole mass spectrometer (QMS). As one part of the comprehensive ECR plasma system, LIDS can...
Within the framework of the Work Package DIV 1 - “Divertor Cassette Design and Integration” of the EUROfusion action, a research campaign has been jointly carried out by University of Palermo and ENEA to investigate the steady state thermal-hydraulic behaviour of the DEMO divertor cassette cooling system, focussing the attention on its Plasma Facing Components (PFCs). The research campaign has...
Adhesion plays a pivotal role in numerous aspects of tokamak-generated dust such as in-situ removal techniques, post-mortem collection activities, resuspension during loss-of-vacuum accidents and in-plasma remobilization. Due to insurmountable difficulties in the theoretical treatment of the interaction between technical (rough, polycrystalline, adsorbate covered) surfaces, adhesive or...
High-performance cooling is of vital importance for the cutting-edge technology of today, from nanoelectronic devices to nuclear reactors. For fusion reactors, subcooled boiling heat transfer is expected to play a critical role for the safe and efficient operation of components exposed to high heat flux. Recent advances in nanotechnology have allowed the development of a new category of...
The Chinese Domestic Agency (DA) is procuring ITER Enhanced Heat Flux (EHF) First Wall (FW) panels, representing 12% of the total number of ITER FW panels being procured. The EHF FW panel shall withstand a surface heat load up to 4.7 MW/m2 during ITER operation. Prior to the implementation of the ITER Procurement Arrangement (PA), several key technologies in manufacturing the EHF FW panel have...
Future fusion power plants require the development of a first wall armor material withstanding extreme particle and heat loads. Considering safety, the formation of long-lived radioactive isotopes when irradiated with neutrons and a tritium inventory has to be prevented. As tungsten (W) meets these safety requirements, has a low erosion rate, high melting point, and high thermal conductivity,...
The next Deuterium-Tritium campaign will push JET ITER-like wall to high divertor power and energy levels. During the 2016 campaign, the Strike Point (SP) sweeping technique enabled us to run relevant H-mode scenarios without exceeding the temperature limits imposed by JET Operation Instructions (JOIs). In the subsequent shutdown, six outer divertor tungsten-coated 2D carbon fibre composite...
High Magnetic field Helicon experiment (HMHX) is a linear helicon wave plasma (HWP) source with high axial magnetic field (B0<6300 G), which address fuel retention in first wall materials. High flux Ar/D2 plasmas are produced using an inner half helical antenna with RF power source operating at a frequency of 13.56 MHz at power levels up to 5 kW. Langmuir probe, OES and Hiden EQP...
Tungsten is foreseen as plasma facing material for ITER. The tungsten testing under high heat loads is very important for ITER operation prediction. In a real tokamak conditions combination of the heat and particle fluxes could enhance tungsten destruction and erosion. Some mechanisms of plasma-surface interaction can lead to a concentration of heat flux onto the small zones of the divertor...
In order to realize a commercial feasible fusion reactor, the life time of plasma facing materials (PFM) and components (PFC) is one of the key issues. Steady state loads in the range of 10 MW/m² and in addition millions of transient events with 0.6 - 3.5 GW/m² represent a huge challenge and lead to severe damages. As first wall, tungsten armor on a low activating structural steel is planned....
The cooling performance of surface structuring for enhancing heat transfer in cooling channels of helium-gas cooled First Wall applications and their prospects of success in efficiency and effectiveness were investigated for several thermal-hydraulic conditions and structure designs in the last years. Cooling channel structured by upstream and downstream directed, truncated 60° V-shaped ribs,...
The first neutral beam injector (NBI) experiments of the Wendelstein 7-X stellarator will start in summer 2018. The modelling of the fast ion production and slowing down processes [1,2] predicts losses of the NBI fast ions to the first wall on the order of 15%. One location receiving a high load (possibly peaking at several MW/m2) is the immersion tube for optical and infrared monitoring of...
Each of the neutral beam injectors in the experimental devices ASDEX Upgrade (AUG) and Wendelstein-7X (W-7X) can be equipped with up to four positive ion sources with a neutral beam output power of 2.5 MW each. For the conditioning of the system, a movable calorimeter is placed in the path of the neutral beam to dump the heat load. The core of the calorimeter consists of a set of so called...
The ITER maintenance is done by means of Remote Handling (RH) systems. During maintenance operations, the RH operator is intended to utilize user interfaces for commanding, monitoring and controlling the RH Equipment. The user interfaces are, for example, GUIs, haptic and joystick devices, Virtual Reality (VR) systems and camera views on the RH Equipment and its environment. Many RH tasks...
If a remote-handling system were to become stuck and unrecoverable from the vacuum vessel, the fusion reactor would be forced to cease all operations. Proven recovery technology must be established for remote-handling systems of fusion reactors to ensure the system is recoverable from expected failures. The recovery technology for the ITER Blanket Remote Handling System is in the form of...
A novel endoscope has been developed for the inspection of long and complex-shaped cooling pipe of ITER Thermal Shield (TS). The mechanical design has been improved and its endoscopic images are clarified for various pipe surface conditions. Main break-through is to reduce the cable friction against metal pipe inner surface as well as to maintain the elastic rigidity of the cable for insertion...
In magnetically confined fusion devices the plasma operation takes place in a hermetically sealed vacuum vessel (VV) of unconventional size and shape that enables the crucial high-vacuum environment.
Apart from this basic purpose the DEMO VV has to fulfil several additional requirements.
It has to provide support to the in-vessel components (IVCs) in all operational conditions in particular...
The Russian Federation is responsible for manufacturing and delivery 54 main and 4 spare Dome, that is the part of ITER divertor cassette.
The main elements of the Dome are: Umbrella manifold, Outer and Inner manifolds made of 316L(N)-IG steel, the tubes of 19,05 mm and 141.3 mm diameter with 1.65 mm and 9.53 mm of thick walls respectively made of 316L steel and 34 plasma-facing units of...
A number of blanket arrangements and maintenance options are being investigated within various fusion DEMO studies. The defined segmentation, general arrangement and maintenance approach for the vacuum vessel blankets has a major impact on the overall device configuration. The K-DEMO blanket arrangement is centered on an approach to minimize the number of blanket segments that are accessed...
In the European DEMO program, the design development of the demonstration power plant is currently in its pre-conceptional phase. This work includes also the design development of the vacuum vessel, where lower ports are important appendices that house the Metal Foil Pumps and the Linear Diffusions Pumps as major components of the vacuum pumping and fuel processing systems (the so-called...
The EAST superconducting tokamak upper divertor had been updated to tungsten divertor. Based on the tungsten divertor operation 10MW/m^2 heat load can be exhausted. Long pules (100s) H mode plasma was obtained. The lower divertor of EAST is still carbon plasma facing material. Upgrade the divertor to tungten the EAST will be full metal first wall with tungten for upper and lower divertor and...
The ITER vacuum vessel (VV) is a torus-shaped, double-wall structure with shielding and cooling water between the shells. Low distortion welding techniques are chosen in order to manufacture the 4 poloidal segments (PS) composing each sector, weld them together and then assemble on site the nine toroidal sectors to form the complete torus.
Control of the distortions during the welding process...
Tailor-welded blanks are used in the manufacture of CFETR (China Fusion Engineering Test Reactor) port hub.In order to obtain high manufacturing precision, CFETR port hub is welded into a whole by EBW which has features of higher precision ，smaller deformation then melt welding methods.According to the inherent strain theory, the inherent strain of EBW welding is calculated ,then the result...
Wendelstein 7-X (W7-X) is a fivefold optimized stellarator in operation in Greifswald, Germany. W7-X Plasma Vessel (PV) consists of five modules made with 17mm thick steel and having 254 openings for ports, necessary for cooling, heating and diagnostic purposes. Both ports number and their structures are different in each PV module.
During rare plasma disruption, the plasma bootstrap current...
The ITER Vacuum Vessel is a torus-shaped, double shell stainless steel structure made up of nine welded sectors, with five being manufactured by the European Domestic Agency of the ITER Project (Fusion for Energy). Despite the large sector dimensions (around 11 x 7 x 6 metres) and the considerable weld lengths (in excess of 1.7 kilometres per sector) each sector must achieve very tight...
In the ITER or the future DEMO fusion reactors, 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 manipulators. In order to adapt to the complex environment, the accuracy of the manipulators is necessary to be improved. The kinematic...
Chinese Fusion Engineering Testing Reactor(CFETR) is a super conducting magnet Tokamak, and the key component begun to be studied in advance. 1/8 full size vacuum vessel(VV) as a research project, which purpose is to fully grasp the key technology of molding, welding , non-destructive testing and measurement in the aspect of building large-scale vacuum, and accumulate experience for the formal...
Remote welding of cooling water pipes is one of the technological challenges for maintenance of nuclear fusion reactor. In ITER, more than 1,000 in-vessel welds are performed for the installation of First Wall (FW) and Shield Block (SB), of which failures during D-T operation require complete remote handling of these pipes due to irradiation environment in the vacuum vessel. The welds in FW...
This paper describes the manufacturing study of ITER Lower Cryostat Thermal Shield (LCTS) cylinder components, which were delivered to ITER site. Fabrication of LCTS cylinder had been proceeded according to the following processes: 1) plate cutting, 2) shell to flange welding, 3) cooling pipe welding, 4) flange final machining, 5) pre-assembly of 60 degree sector, 6) silver coating, 7) final...
ITER Remote Handling equipment controllers provide measurement and diagnostics data about the remote handling equipment and devices they control, about themselves and their operating environment. This information is aimed for the RH operators to reduce downtime of the Remote Handling systems by anticipating maintenance needs and failure conditions.
In this paper, the development of the...
The realization of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle is viewed by Europe and many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. The DEMO machine has three main entrance levels to the...
Thermal Shield (TS) in ITER tokamak reduces heat loads from vacuum vessel and cryostat to superconducting magnet structure. Its delivery is scheduled to begin from September 2018 including TS Main Components (TSMC) and manifold pipes. Unlike to other components in ITER tokamak, most of TSMC have slender structure with panel thickness of 20 mm. Due to its structural uniqueness, TSMC cannot be...
The main components of the ITER tokamak are assembled from the nine sub-assemblies of the 40° sectors. During the sub-assembly, the Toroidal Field coils (TFCs) are installed outside of the Vacuum Vessel Thermal Shield (VVTS) by rotation. The clearance between TFCs and VVTS is very small, in relation to the cooling tube end points, which connect to the thermal shield manifolds. The dimensional...
This paper mainly analyzes the process of the 1/16 sector vacuum vessel (VV) welding from two 1/32 VV sectors in the 1/8 VV sector research and development project of China Fusion Engineering Test Reactor (CFETR). In the numerical simulation, the inherent strain method was applied to analyze the welding deformation and shrinkage of 1/16 VV sector with and without welding tools respectively....
Due to the limited irradiation lifetime of the structural material used for in-vessel components in DEMO, and subsequent future fusion power plants, it will be necessary to replace all breeding blankets within the given planned maintenance window in order to meet DEMO availability targets. It is assumed that failure of in-vessel components cannot be excluded, whilst in-situ repair is...
The Multi-Purpose-Manipulator (MPM) has been operated, as a versatile carrier system for probes, since the first campaign OP. 1.1 in 2015 at Wendelstein 7X (W-7X). The combined probe, a combination of Langmuir, Mach and magnetic probes, was used. For the second campaign OP. 1.2a in 2017, with an island divertor, an upgraded combined probe, a fluctuation probe, a retarding field analyzer (RFA),...
This presentation shows results of calculations of the Upper port plug (UPP) and ex-vessel components of ITER upper ports №2 and №8. Detailed finite-element models of modernized UPP construction were developed taking into account nonlinear contact interaction between the diagnostic shield module and the UPP structure. In the structural analysis under design loads the stress-strain state (SSS)...
A significant analysis effort was undertaken to address the challenging ITER Blanket Manifold design requirements resulting from electromagnetic (EM) major disruptions and vertical displacements events in a very demanding neutronic environment.
The effort was focused on maintaining the structural integrity of the component itself and minimizing the loads transferred to the Vacuum Vessel to...
In the framework of the European “HORIZON 2020” research program, the EUROfusion Consortium develops a design of a fusion demonstrator (DEMO). CEA-Saclay, with the support of Wigner-CR and IPP-CR, is in charge of one of the four Breeding Blanket (BB) concepts investigated in Europe for DEMO: the Helium Cooled Lithium Lead (HCLL) BB. The BB directly surrounding the plasma is a major component...
A pilot plant for tritium removal from tritiated water is in pre-operational stage at ICSI Ramnicu Valcea and is based on catalytic isotopic exchange (LPCE) between tritiated heavy water and hydrogen/deuterium followed by cryogenic distillation (CD) aiming to recover tritium. As any detritiation plant or tritium processing plant/laboratory, also the Expriemntal Pilot Plant for D-T separation...
The Coolant Purification Systems, together with the Tritium Extraction System (TES), the Tritium Removal System (TRS) and the two Helium Cooling Systems (HCSs) belong to the ancillary systems of Helium Cooled Lead Lithium (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Modules (TBMs ) which are currently in the preliminary design phase in view of their installation and operation in...
Nowadays, Fusion Energy is one of the most important sources under study. During the last years, different designs of fusion reactor were considered. At the MIT, an innovative design was created: ARC, the Affordable Robust Compact reactor. It takes advantage of the innovative aspect of recent progress in fusion technology, such as High Temperature Superconductors, that permit to decrease the...
Within the framework of the pre-conceptual design of the EU-DEMO Breeding Blanket (BB) supported by EUROfusion action, the University of Palermo is involved, as ENEA linked third-party, in the development of the Water Cooled Lithium Lead (WCLL) BB concept.
Results of the research activities carried out have highlighted that changes in the proposed WCLL BB design have to be considered,...
In order to achieve tritium self-sufficiency of fusion reactors, tritium will be generated in breeding blankets by neutron bombardment of lithium and then will be extracted to refuel the plasma. The Vacuum Sieve Tray (VST) was proposed to ensure tritium extraction from liquid breeding blankets, composed of lead-lithium. It consists in letting the liquid metal fall through submillimeter...
Pd/Ag membranes are one of the reference technologies for the fuel cycle of deuterium-tritium fusion machines. This technology is proposed to be implemented in tritium recovery systems, due to their exclusive selectivity towards molecular hydrogen isotopes. For instance, these membranes are proposed to process and separate Q2 (Q = H, D, T) species from impurities (e.g., inert gases) coming...
Tritium permeation loss in the fusion reactor is an important issue. Silicon carbide (SiC) is considered as an important material for Tritium permeation barriers due to its excellent properties (including low diffusivity). Steady-state and high-flux helicon-wave excited Ar/CH4/SiH4 plasma were used to synthesis SiC film onto 316L stainless steel. The surface profile and the thickness of the...
Long-lived fission products (LLFPs) are one major factor of the radioactivity and decay heat of high level wastes produced by light water reactors. Transmutation of LLFPs into non-radioactive or short-lived nuclides is an efficient way to reduce the amount of high level waste. Earlier studies have proposed to use a part of the breeding blanket of a fusion reactor to transmute LLFPs. The...
In the European Helium-Cooled Pebble Bed (HCPB) concept of the DEMO blanket, beryllium pebbles with a diameter of 1 mm are planned to be used as neutron multiplier. A study pebble bed mock-up behavior under high-dose neutron irradiation at the HCPB relevant temperatures in a material testing nuclear reactor should provide an essential database for blanket designers.
Beryllium pebbles with...
Within the framework of the EUROfusion project activities concerning the EU-DEMO Breeding Blanket (BB), University of Palermo is long-time involved, in close cooperation with ENEA, on the design of the Water-Cooled Lithium Lead (WCLL) BB, which is currently under consideration to be adopted in the EU-DEMO reactor.
The WCLL BB concept foresees liquid Pb-15.7Li eutectic alloy as breeder and...
To be accepted in the future energy landscape, fusion reactors must be inherently safe by design. An unresolved safety issue is the undesired production of highly radiotoxic Po-210 in the liquid Pb-Li eutectic used in many breeding blanket concepts. Po-210 is the end product of consequent neutron captures and beta decays, initiated by a neutron capture by Pb-208.
Po-210 is an intense alpha...
In fusion devices, the retention of the fusion fuel deuterium (D) and tritium (T) in plasma-facing components (PFCs) is a major concern. Measurement of the hydrogen isotope content in PFCs and test samples gives insight into the retention physics.
In FREDIS, Thermal Desorption Spectrometry (TDS) is performed in an evacuated (p < 1E-8 hPa) quartz tube (Ø52 mm), where samples are heated by 6...
Nowadays the Systems Engineering (SE) methodology is strongly applied in several fields of engineering such as Chemical and Process Industries, Civil and Enterprise application as well as Service and Healthcare systems. Furthermore, the SE represents a powerful interdisciplinary mean to enable the realisation of complex systems taking into account the customer and Stakeholder´s needs by...
Tritium permeation through structure materials in fusion blanket systems is a critical issue from the perspectives of fuel loss and radiological hazard. In the previous studies, detailed hydrogen isotope permeation behaviors in reduced activation ferritic/martensitic (RAFM) steels have been investigated; however, it is supposed that the surface of the RAFM steel will be oxidized under an...
Tritium permeation through structural materials in a fusion reactor fuel system causes fuel inefficiency and tritium leakage to the environment. Tritium permeation barrier (TPB) has been intensively developed using ceramic coatings to establish liquid blanket concepts for several decades. In a TPB coating, not only tritium permeation reduction but also tolerance to high dose radiation is...
The characteristic identification of the functional material is important in the performance prediction of a breeding blanket, which is one of the main components of the fusion power plant. The functional material of the solid type ceramic breeding blanket is mainly used in the form of a pebble bed, which is an aggregate group of pebbles. Various experimental methods such as laser flash, hot...
Tritium breeder pebble beds are multiphase materials where ceramic pebbles and purge gas coexist. Therefore, the heat transfer in the bed is influenced by both solid particles and helium gas. Furthermore, due to their discrete nature, pebble beds show a complex fully coupled thermo-mechanical behaviour. Simulations carried out with the Discrete Element Method (DEM) allow evaluating the...
Project of IGNITOR tokamak is one of main directions of scientific collaboration between Russia and Italy. Project is entering stage of technical design for location of the machine on TRINITI site in Troitsk, Moscow region.
The IGNITOR machine differs considerably from other machines based on tokamak concept by using a super strong magnetic field (13 Tesla) and plasma current (11 MA). It will...
In the framework of gloveboxes tritiated gaseous effluent treatment, efficiency of packed bed membrane reactors has been successfully demonstrated under lab scale. In such an intensified process, tritium from tritiated water can be recovered under the valuable Q2 form (Q = H, D or T) thanks to isotope exchange reactions on catalyst surface. In the meanwhile, the use of permselective Pd-based...
In the framework of the European “HORIZON 2020” innovation and research program, the EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO). One of the key components in the fusion reactor is the Breeding Blanket (BB) surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. CEA-Saclay, with the...
The experimental facility THALLIUM was designed and installed at ENEA C.R. Brasimone to investigate the consequence of a HCLL-TBS (Helium Cooled Lithium Lead) In box LOCA, ensuring a good level of geometrical relevance with HCLL-TBM. Within the framework of the contractual activities agreed with Fusion for Energy, a first experimental campaign was carried out in HCLL-TBS relevant conditions....
Practical plan of “continuous tritium recovery by PbLi droplets in vacuum” campaign is introduced. This campaign aims to verify the viability of tritium recovery method by PbLi droplets in vacuum as a prototype design level. Following verifications are to be performed. 1) To verify the steady state extraction efficiency of tritium from a vacuum sieve tray for 1.1 mm diameter droplet. Predicted...
To start up an initial fusion reactor and for technical tests for tritium circulation and blanket system, it is necessary to provide sufficient amount of tritium from an outside device. Tritium production using a high-temperature gas-cooled reactor has been proposed. [1]. It was reported that 500–800 g of tritium could be produced during one year of operation using a 600 MW thermal output...
The fuel cycle of the tritium plant has to safely handle the fuel gases including tritium and provide those gases to the fusion reactor. Given a required amount of tritium for fuelling scenarios considering ramp-up, flat-top, and ramp-down, a scheduling model is developed base on the state-task-network representation to provide the optimal operation plan for DT plasma operation including...
China has long been active in pushing forward the fusion energy development to the demonstration of electricity generation. As one of the most challenging components in DEMO, great efforts have been put on the development of breeder blanket and three blanket schemes were studied in China for fusion engineering test complementary with ITER (International Thermonuclear Experimental Reactor). In...
Recent works have shown that low grain sizes are favorable to improve ductility and machinability of tungsten, as well as the resistance to ablation and spallation, which are key properties for the use of this material in thermonuclear fusion environment [Reiser et al. Int. Journal of Refractory Metals and Hard Materials, 64 (2017) 261]. However, current production routes are not suitable for...
Tungsten is to be used as plasma facing material of divertor target for ITER. Though the present ITER specification for divertor target requires pure tungsten as a plasma facing materials, much research effort is being devoted to improve the material properties of tungsten for the application to severe conditions predicted in DEMO reactor. Among the material properties of tungsten, thermal...
The Dual-beam ion irradiation facility for FUsion materials (DiFU) is under development at the Ruđer Bošković Institute in Zagreb, Croatia, allowing irradiation of fusion-related samples by one or two ion beams. Two ion beams come to the DiFU chamber at an angle of 170 between them, from 6 MV HVE Tandem VDG and 1 MV HVE Tandetron accelerator. Ion beam handling and scanning systems enable fast...
Liquid tin (Sn) is a promising coolant of liquid divertor systems due to its low vapor pressure. However, the material compatibility with structural materials is important issue for the development of the liquid divertor system. The purpose of the present study is to explore corrosion resistant materials in the liquid Sn. The corrosion tests were performed in a static Sn at 773K with various...
Radiation induced lattice defects strongly affect functionality of optical components, which will play a substantial role in various diagnostic systems of future fusion reactors. It is widely recognized that spinel lattice of double oxides (e.g. MgAl2O4) demonstrates enhanced radiation tolerance. One can expect a higher radiation tolerance of single cation spinels because in this case...
The constitutive behavior of Reduced Activation Ferritic/Martensitic (RAFM) steel for the potential blanket material of fusion reactor was modeled and implemented into a crystal plasticity finite element method (CPFEM). In the developed constitutive model, the plasticity was formulated by the slip system activation in the body centered cubic single crystal and the stress of polycrystal was...
The present paper summarizes the current status of the Secondary Heat Removal System (SHRS) of the IFMIF-DONES (International Fusion Material Irradiation Facility- Demo Oriented Neutron Source). As part of the Lithium systems (LS) in IFMIF-DONES, SHRS is the responsible sub-system for the removal of the heat which develops in the LS-Test Assembly (TA) during the Li - DT reaction. In this way,...
Nuclear fusion is a promising way to fulfill the current and future energy needs in a cleaner way and many challenges must be overcome to be able to achieve a sustained nuclear fusion reaction. Tungsten with a high melting point, high sputtering threshold and low tritium inventory is the choice for the plasma facing material and CuCrZr alloy, with high conductivity and strength, for the heat...
Tungsten is the leading choice as plasma-facing material in fusion reactors. Its brittleness can be alleviated by alloying with few-% rhenium. The earliest nuclear application of the W-Re system was as thermocouples in experimental breeder reactors. A drop in efficiency of the thermocouples was noticed and attributed to radiation-induced precipitation. Many studies followed with varied...
IFMIF-DONES (International Fusion Materials Irradiation Facility — DEMO-Oriented Neutron Source) will be built as a powerful neutron source to test suitable materials planned for the construction of future tokamaks like DEMO (Demonstration Fusion Power Plant). In the commissioning phase of IFMIF-DONES it is foreseen that a Start-Up Monitoring Module (STUMM) will be used for the...
Fusion first wall materials are required to withstand large neutron damage during their lifetime. This damage comprises of knock on induced displacement damage and a material composition change through transmutation reactions. Protons of up to 5 MeV and heavy ions from accelerators are capable of reproducing displacement damage similar to that from neutrons in a fusion reactor. As the...
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 as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. DONES facility is being developed within the EUROfusion workpackage WPENS which main objective is...
Several blanket concepts (e.g., HCLL, WCLL, DCLL) are based on the application of the liquid breeder Pb-15.7Li, which is in direct contact with the structural components. Compatibility testing has shown that the structural materials (e.g., Eurofer) always suffer from corrosion attack which mainly depends on the operation temperature and flow velocity of the liquid breeder. The governing...
One of the functions of ITER diagnostic port-plugs is neutron protection of the equipment installed in the port, as well as reducing the radiation background in the area of reactor elements requiring access for maintenance personnel. Engineering restrictions on the full weight of the port plug and the amount of water in the reactor do not allow the use traditional iron-water protection. Boron...
The divertor component is subject to some of the most extreme loading conditions in a fusion environment and the safety design window is relatively narrow. Variation in mechanical properties of the same material throughout a component is common place in practice and can be caused by both local effects of processing and joining, and local variations in thermal history, neutron flux and other...
The functional materials of solid-type breeding blanket concepts for fusion reactor are used in a pebble bed form. In order to verify the performance and safety of the breeding blanket, the thermal conductivity of pebble bed is required. The hot wire, hot disk, guarded hot plate and laser flash method are considered as measurement technique for the thermal conductivity of pebble bed. This...
Interactions of 14.1 MeV energy neutrons with fusion reactor materials will result into the production of energetic recoils (primary knock on atoms) that lead to displacement damage in the reactor materials. These neutrons will yield different recoils atoms species of different energy and mass based on different reaction channels. Prediction of displacement per atom (dpa) requires energy...
Due to the low activation and excellent neutron irradiation resistance, the Reduced Activation Ferritic/Martensitic (RAFM) steel has been considered as the primary candidate structural material for the blanket of the first fusion reactor plant. The rigorous serving condition of a fusion reactor lead to to multiaxial stress-strain condition of the blanket. With the multiaxial cyclic loading,...
A reduced-activation ferritic steel, F82H steel, is the primary candidate structural material for fusion blanket. Neutron irradiation properties are estimated by using miniature specimens. Since the thickness of the gauge section of the miniature tensile specimens and of wall thickness for creep tubes is less than 1 mm, deformation volume is much smaller than that of standard size specimens....
We are carrying on design activities of an advanced fusion neutron source (A-FNS) in Japan. A large amount of neutrons are produced by Li(d,n) reaction bombarding a 40 MeV deuteron beam of 125 mA with a liquid Li target at the A-FNS. In the Li(d,n) reaction, there are reaction processes with strong angular dependence such as proton stripping and ones with weak dependence such as evaporation....
Reduced activation ferritic/martensitic (RAFM) steel, e.g., F82H, is the leading candidate structural material for fusion blanket. Of many blanket concepts, the water-cooled ceramic breeder blanket is an attractive concept because of its compactness and its compatibility with the technologies in conventional light water reactor. For tritium breeding, it is necessary to manage the corrosion of...
In the frame of the activities promoted and encouraged by the EURO-fusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently posed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among...
Fusion reactors represent a future evolution of the nuclear technology improving the world-wide energy portfolio. The experimental fusion reactor under construction (ITER) and the planned industrial fusion reactors (DEMO) are large and complex facilities. For their operation it is necessary to ensure safety and solve several technical issues. The limitation of the radiological and mobilizable...
Safety assessment is a key issue for the licensing of DONES facility, the DEMO-Oriented Neutron Source. A first phase of the safety assessment include Failure Mode Analyses of systems to identify postulated initiating events (PIEs), while deterministic accident analyses are lately performed to estimate source terms in radiological hazards. In addition, the deterministic analyses are also the...
In fusion plants the overnight cost (capital cost minus financing) is expected to be the main contribution to the cost of electricity. The overnight costs (euro/kWe) for ITER and DEMO are several times higher than the ones of present commercial sources (wind, photovoltaic, fission, coal, gas, …). It is shown that cost reduction from high learning rates in future commercial plants should not be...
The Chinese Fusion Engineering Test Reactor (CFETR) bridges the gap between ITER and a demonstration fusion power plant (DEMO). The primary objectives of CFETR are: demonstrate tritium self-sufficiency, ~ 1 GW fusion power, operate in steady-state and have a duty cycle of 30-50 %. CFETR is in the pre-conceptual design phase and is currently envisaged to be a two-phase machine (phase I ~ 200...
The Helical-Axis Advanced Stellarator (HELIAS) is the leading stellarator concept in Europe and developed at the Max-Planck-Institute for Plasma Physics (IPP). Based on the 5-field-period symmetry, the HELIAS-5B engineering design study emerged which aims at a stellarator power reactor designed for 3000 MW fusion power.
The stellarator confines hot plasma only by external superconducting field...
CFETR is now in engineering design phase (EDP) and will hopefully complete this phase around 2020. It is essential to evaluate the public impact due to the radioactivity release in this stage. In this work, the radioactive inventory and property of source terms were evaluated, discussed and compared with ITER. Then, under normal operation radioactivity release limit was researched...
The world’s largest tokamak fusion device-ITER is under construction in Cadarache, France, and first plasma will be officially identified in 2025. Building on the work of ITER, various countries are planning the steps needed for the fusion demonstration reactor (DEMO), and many conceptual designs for the fusion power plants (FPP) have been developed, such as PPCS of the European Union, ARIES...
The current conceptual design of the Primary Heat Transfer System (PHTS) of the water-cooled EU DEMO foresees two independent cooling circuits, the breeding zone PHTS and the first wall PHTS. During the pulse (120 minutes) the first delivers thermal power to the turbine, the latter delivers thermal power to the Intermediate Heat Transfer System (IHTS) equipped by an Energy Storage System...
The new experimental facility LIFUS5/Mod3 has been designed, manufactured and installed to investigate the phenomena connected with the thermodynamic and chemical interaction between lithium-lead and water in case of in-box LOCA (Loss of Coolant Accident) of the WCLL breeding blanket concept and to validate the chemical model implemented in SIMMER code for fusion application. In order to...
Selection of technologies that are considered for detritiation and recycling of DEMO waste materials has been made. To study treatment of waste technological process at DEMO facility where it has to be accounted groups of specific materials as composites, metals, oxides and others. From them the DEMO facility will be constructed and which are considered for further modification of this unit....
With the advance of fusion research both in physics and engineering and the start of the conceptual design development of respective fusion plants it becomes important to study how fusion will interact with the energy system. However, in order to study the interface between a fusion plant and the energy grid, the details of the Balance of Plant (BoP) must be known and modeled.
Recently, first...
One of the four breeding blanket candidates as options for European DEMO nuclear fusion reactor is the Water Coolant Lithium Lead Breeding Blanket (WCLL BB). The WCLL in-box LOCA (Loss of Coolant Accident) is a major safety concern of this component, therefore transient behavior will be investigated to support the design, to evaluate the consequences and to adopt mitigating countermeasures....
Safety studies for fusion facilities are commonly conducted using codes originally developed for fission reactor accident analysis and adapted to model the fusion-relevant phenomena. Nevertheless there are many “fission developed” methods still not considered in fusion safety assessment which could offer significant advantages in the fusion power commercialization. Along with solving the...
This paper aims to make a further investigation on economic analysis of fusion-biomass hybrid model based on previously reported concept. This system postulates to gasify cellulose and lignin which are compositions of biomass derived from agricultural and forestry sectors to produce synthetic gas for artificial diesel and hydrogen production. Gasification process (C6H10O5 + H2O → 6H2+ 6CO)...
The production of dust inside the nuclear fusion power plants is one of the safety issues of this technology. Dust is generated because of plasma-material interactions and deposits in the bottom regions of the TOKAMAK. In case of a Loss Of Vacuum Accident (LOVA), the dust may be resuspended, threatening the functioning and the safety of these reactors. A deep study of this phenomenon is...
As an import component of the amplifier of inertia confined fusion (ICF), the flash lamp pumping efficiency have a lot to do with the amplifier efficiency and. In this paper, a kind of flash lamp with a new design has been manufactured to acquire high performance and it has been proved to be able to acquire high performance than the tradition flash lamps. Comparisons between the new and the...
The implementation of Machine Learning (ML) techniques has considerably improved the prediction of disruptions. However, they usually provide outcomes difficult to understand from a physics point of view due to their mathematical formulation. The objective of this work is to attain an interpretable equation of an accurate ML disruption predictor. The equation could be used for real-time...
An Ion Cyclotron Range of Frequency (ICRF) system is one of the options to provide power on DEMO for a number of tasks, which have all been experimentally verified on present machines: breakdown assist, heating the plasma, controlling sawteeth, removing central impurities, current ramp down assist and wall conditioning. ICRF has a number of advantages for a machine like DEMO. The system has a...
The Chinese Fusion Engineering Testing Reactor (CFETR) will be operated in two phases. Phase I focuses on Pfusion=200 MW, Qplasma=1–5, TBR>1.0. Phase II emphasizes DEMO validation, which means Qplasma>10, Pfusion>1 GW (e.g., 1.5GW). CFETR has updated its core parameters in 2018. The major /minor radius have been changed from R=5.7m/a=1.8m to R=7.2m/a=2.2m. It is also required that one blanket...
Innovative high-frequency magnetic sensors have been designed and manufactured in-house for installation on the Tokamak à Configuration Variable (TCV), and are currently routinely operational during the TCV experimental campaigns. These sensors combines the Low Temperature Co-fired Ceramic (LTCC) and the classical thick-film technologies, and are in various aspects similar to the large...
The Water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European DEMO nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant and EUROFER as structural material. Current design is based on DEMO 2017 specifications. Two separate water systems are in...
The construction of SPIDER, the prototype of the full-size, radio frequency, negative ion source of the ITER
Heating Neutral Beam Injectors, has been completed at the Neutral Beam Test Facility in Padova (Italy). All
SPIDER components have been delivered and have successfully undergone the site acceptance tests.
The mechanical components (vessel, ion source, accelerator, and calorimeter),...
The Helium Cooled Pebble Bed (HCPB) breeding blanket is the reference blanket proposed for the EU-DEMO. A concept for the HCPB based on a cooling plate (CP) “sandwich” architecture built in Multi-Module Segments was developed for the EU-DEMO 2015 tokamak baseline. This architecture significantly improved the tritium breeding performance with respect to the former “beer-box”-like concept and...
The WEST platform aims at testing ITER like W divertor targets in an integrated tokamak environment. To operate long plasma discharge, the IR thermography is required to monitor the main plasma facing components by means of real time surface temperature measurements, while providing essential data for various physics studies.
To monitor the new divertor targets, the WEST IR thermography...
Electron-Cyclotron-Resonance Heating and Current Drive (ECRH&CD) is one of the preferable auxiliary heating systems for magnetically confined nuclear fusion plasmas. ECRH&CD is widely used in tokamaks and helical machines for start-up and plasma control.
Since many years KIT is strongly involved in the development of high power gyrotrons for use in ECRH. KIT is pursuing two development lines:...
A ten-channel overview video diagnostic system was installed and commissioned at
Wendelstein 7-X (W7-X) optimised stellarator. The cameras serve both for surveillance of the
first wall (ensuring safe device operation) and allow for physics studies. The wide range of
applications is ensured by a highly flexible data acquisition and on-board data processing.
Combination with magnetic field line...
Cryogenic hydrogen adsorption using molecular sieve beds is considered to be one of the main candidate processes for recovery of produced tritium from purge gas in breeding blankets and it has been chosen for separation of hydrogen isotopes in Tritium Extraction System (TES) of Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS). Various absorbents and their performance...
The Neutral Beam Injection System (NBI) on the DIII-D tokamak includes eight ion sources operating nominally at 75-80 kV, each capable of injecting up to 2.5 MW for plasma heating and current drive. As DIII-D physics experiments evolve to explore new regimes in fusion energy research, the capabilities of the NBI systems are being improved to help provide the necessary tools.
One of the NBI...
Indian Test Facility (INTF) is designed for “Full characterization of the Diagnostic Neutral Beam (DNB)” for ITER, to unveil the possible challenges in production, neutralization and transportation of neutral beam over the path length of ~20.67m. This facility consist of a vacuum vessel (with volume >180m3) which has been designed and manufactured as per the rules of ASME Sec.VIII Div. 1, to...
The main expertise of the Membrane Laboratory of ENEA Frascati is related to the study and development of Pd-based membrane technologies (both permeators and catalytic reactors), which are one of the reference processes in the fuel cycle of nuclear fusion reactors. Principal characteristics of Pd-based membranes are infinitive hydrogen selectivity, elevated hydrogen permeability, modularity,...
The Diagnostic Shielding Modules (DSM) are secondary containers of diagnostic Port Plugs where shielding and diagnostics components have to be integrated. For the development of equatorial DSMs several key requirements have to be met. The total dry Plug weight shall not exceed the allowable maximum (45 t) in other to guarantee the consistency of the design with the specification of other...
A magnetically insulated baffled (MIB) probe offers the advantages of direct measurements of the plasma parameters (including plasma potential, electron temperature and ion temperature etc.), while being non-emitting and electrically floating[1].The MIB probe was constructed by retracting the conducting plug of a classical Langmuir probe inside an insulating tube placed perpendicular to the...
The ITER neutral beam test facility under construction in Padova will host two experimental devices: SPIDER, a 100 kV negative H/D RF source, and MITICA, a full scale, 1MeV deuterium beam injector. A detection system called Close-contact Neutron Emission Surface Mapping (CNESM) is under development with the aim to resolve the horizontal beam intensity profile in MITICA and one of the eight...
The validation of the key technologies relevant for a DEMO Breeding Blanket is one of the main objectives of the design and operation of the Test Blanket Systems (TBS) in ITER. . In compliance with the main features and technical requirements of the parent breeding blanket concepts, the European TBM Project is developing the HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble...
Pulsed power supply systems are employed in many fusion projects and in other scientific applications. A novel power supply was specifically developed to feed pulsed loads (low duty cycle), as the resistive or superconducting coils used to produce high magnetic fields for some seconds or longer.
Thanks to the integrated energy storage devices, it is not necessary to draw directly from the...
Three of the four breeder blanket concepts for a DEMO Reactor use the eutectic Pb–16Li enriched at 90% in 6Li as breeder material: Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL). Moreover the WCLL is one of the blanket concepts that will be qualified in the ITER reactor, therefore the development and design of lead lithium loops and...
For future fusion plants, such as DEMO, there is a great need for detectors capable to accurately monitor neutrons under the harsh conditions imposed by fusion environment. In particular, detectors required in Test Blanket Modules must be capable to accurately measure neutron fluence under high and variable neutron count rates, high gamma background, high temperature and high and variable...
To reach fusion conditions and control plasma configuration in ITER, the next step towards establishing nuclear fusion as viable energy source, suitable combination of additional heating and current drive systems is necessary. Among them, two Neutral Beam Injectors (NBI) will provide 33MW hydrogen/deuterium particles electrostatically accelerated to 1MeV; efficient gas-cell neutralisation at...
In the framework of the EUROfusion DEMO Programme the EU has elaborated a completely novel and most innovative fuel cycle architecture, driven by the need to reduce the tritium inventory to an absolute minimum.
To achieve this goal, batchwise processes used in the fusion fuel cycle so far were replaced by continuous processes wherever possible. This includes the change from discontinuous...
High-temperature superconductors (HTS) have the potential to enable the operation of a future fusion reactor at higher magnetic fields (> 14 T) or at higher temperatures compared to conventional low- temperature superconductors. In particular, the operation at high magnetic fields with good temperature margin is perceived to be an important advantage of HTS in a fusion power plant.
Fusion...
A transmission line has been newly developed in QUEST spherical tokamak (ST) for the highly efficient electron cyclotron heating and current drive (ECHCD) experiments with a 28 GHz gyrotron. Waves in plasmas can be described in terms of two eigenmodes, so-called extra-ordinary/ordinary modes, coming from anisotropic property of the electron motion on the magnetic-field direction. The modes...
Spherical Tokamak (ST) path to Fusion has been proposed in [1] and experiments on STs have demonstrated feasibility of this approach. Advances in High Temperature Superconductor technology [2] allows significant increase in the Toroidal field which was found to improve confinement in STs. The combination of the high beta, which has been achieved in STs [3], and high TF that can be produced by...
The next stage of the upgrade of T-15MD machine (R=1.5m, a=0.67m, B≤2T, Ipl=2MA) to the superconducting one (excluding the limits of pulse duration) with basically the same geometry is suggested. The estimations show the possibility to make a toroidal magnet with aspect ratio A=2.2, magnetic field on the axis Bo=5T, maximum magnetic field Bm= 12.5T. Such increase in Bo provides the possibility...
ASDEX Upgrade is a tokamak that can be operated with the strike lines in the upper and/or the lower divertor. In 2016 a project was started to develop and install a new upper divertor with internal coils and an in-vessel cryo pump. The aim is to investigate advanced magnetic configurations that may facilitate the access to detachment via an enhanced flux tube expansion and/or connection...
New development steps of ASCOT and AFSI (ASCOT Fusion Source Integrator) based synthetic neutron diagnostics and validation at JET are reported in this contribution. Synthetic neutron diagnostics are important not only in existing tokamaks, where they are used to interpret experimental data, but also in the design of future reactors including ITER, DEMO and beyond, where neutrons are one of...
Actively Cooled Plasma Facing Components (ACPFC) are required to allow for long plasma discharges in magnetic fusion devices. Prior to their installation, the integrity of ACPFC has to be checked under relevant experimental conditions in order to prevent serious water leaks in the vacuum vessel.
Since 1990, the French Magnetic Fusion Research Institute (CEA/IRFM) has developed specific leak...
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. DTT...
Advanced magnetic divertor configuration is one of the attractive methods to spread the heat fluxes over divertor targets in tokamak because of enhanced scrape-off layer transport and an increased plasma wetted area on divertor target. Exact snowflake (SF) for EAST is only possible at very low plasma current due to poloidal coil system limitation. EAST can be operated in quite flexible plasma...
The TCV tokamak contributes to physics understanding in fusion reactor research by a wide set of experimental tools, like flexible shaping and high power ECRH. Plasma regimes with high pressure, a wider range of temperature ratios and significant fast-ion population are now attainable with the TCV heating system upgrade. A 1 MW, 25 keV deuterium heating neutral beam (NB) has been installed in...
The first ever built full-scale prototype of the ITER heating neutral beam injector is the MITICA experiment at PRIMA Neutral Beam Test Facility, under realization in Padua, Italy.
This experiment consists of several in-vessel components: most of them are actively cooled by a large cooling plant, still under construction. Coolant is deionized water produced by a Chemical Control System...
The upgrade of the EC-system of the TCV tokamak has entered in the realization phase as part of a broader upgrade of TCV[1]. The first of the two MW-class, dual-frequency gyrotrons (84 or 126GHz/2s/1MW) has been delivered by Thales Electron Devices and the full commissioning and characterization is expected to be completed during the first half of 2018. The design of this gyrotron includes new...
Before their installation and commissioning in the tokamak, WEST ICRF launchers undergo two categories of pre-qualifications tests. These tests aim at accelerating the commissioning of the launchers in the tokamak.
The first category of tests is milliwatt-range radio-frequency (RF) experiments which allow checking the launchers coupling capabilities, impedance matching and their...
A set of novel design solutions for high performance cooling systems have been developed and tested by Consorzio RFX, achieving, with experimental tests, the challenging heat transfer conditions foreseen for Heating and Current Drive Systems of present and future nuclear fusion devices.
The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), has the triple...
A comb-line antenna to demonstrate efficient off-axis non-inductive current drive from the absorption of toroidally directed very high harmonic fast waves is being designed and built for DIII-D [1].
The antenna consists of a toroidal array of 30 modules, each 5 cm wide by 21 cm tall, so that the array spans 1.5 m on outer wall just above the tokamak midplane.
This antenna will be fed with 1...
To meet requirements of heating and physics experiments on J-TEXT, we are developing a 105GHz/500kW/1s electron cyclotron resonance heating (ECRH) system. With the toroidal field of about 2T for normal discharges on J-TEXT, this system will mainly work at the second extraordinary mode. The ECRH system consists of a Gycom gyrotron with a superconducting magnet and related power supplies, a 30m...
The Aditya tokamak has been upgraded to Aditya-U by changing it’s vacuum vessel from rectangular to circular to accommodate the diverter coil for shaped plasma. The tokamak has been commissioned and now operating with routine plasma experiments.
The 42GHz ECRH (Electron Cyclotron Resonance Heating) system has been integrated with the tokamak. The system is capable to deliver 500kW power for...
Within the framework of the ion source development for the ITER and DEMO Neutral Beam Injection (NBI) systems, IPP Garching has recently upgraded the radiofrequency-driven negative ion source testbed BATMAN. One of the requirements for the ITER NBI system is to produce a beam power density homogeneity above 90% over its large extraction area of about 0.2 m2. This requirement is going to be...
Neutral Beam Injection is the main heating system on a variety of fusion devices, and will be the main heating system of ITER. Especially during high-power operation in long pulse devices, it is important that the losses in the beamline are well quantified. There are several types of beamline losses, such as geometrical losses, due to scraping of the beam on apertures, and reionisation losses,...
Wide range poloidal (~60°) and toroidal (~30°) beam steering capability and the reliability for high-power (0.8 MW/waveguide), long-pulse (100 s) operation are required for the launcher of the Electron Cyclotron Heating and Current Drive (ECH/CD) system in JT-60SA (Super Advanced). The two directional beam steering launcher has been designed by a linear motion antenna concept, which has an...
Based on the Gycom gyrotron of a diode type with a single-stage depressed collector, a 105GHz/500kW/1s electron cyclotron resonrance heating sytem is being developed on J-TEXT tokamak. To modulate output power of the gyrotron, we designed a 33kV/1A anode power supply based on the pulse step modulation technology. The power supply consists of 40 modules with output voltage of 800 V and 10...
The purpose of the italian Divertor Tokamak Test (DTT, 6 T, 5.5 MA, R0 = 2.08 m, a = 0.65 m for ~ 100 s) is to study power exhaust and divertor load in an integrated plasma scenario. To accomplish this mission DTT will be equipped with 45 MW of additional heating power to fulfill a PSEP/R ≥ 15 MW/m studing alternative divertor configurations in view of ITER operations and DEMO design. The...
Neutral Beam Injection (NBI) for ITER shall deliver in total 33 MW heating power to the plasma with two injectors at a beam energy of 1 MV. Taking neutralisation efficiency and all losses along the beam path into account a negative ion current density of 329 A/m2 (H- for 1000s) and 286 A/m2 (D- for 3600s) has to be extracted from each ion source (size 1 × 2 m2) with a beam non-uniformity below...
The ITER project requires at least two Heating Neutral Beam Injectors (NBIs), each accelerating up to 1MV a 40A beam of negative H-/D-.ions, to deliver to the plasma a total 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 including both a full-size negative ion source (SPIDER -...
Since KSTAR first plasma operation, ECH played a key role in obtaining various experimental results such as ECH preionization, ECH-assisted startup, plasma rotation study, impurity transport study, high poloidal beta operation, and long pulse operation. The main heating systems in KSTAR are NBI and ECH which are planed to provide 12 MW NBI by 2019 and 6 MW ECH by 2020 to prepare long pulse,...
To match the electron cyclotron wave with the plasma efficiently, we design two polarizers including a linear polarizer and an elliptical polarizer for the 105 GHz electron cycltron resonance heating system on J-TEXT. The linear polarizer is mainly used to change the rotation angle of the wave, while the ellipticity of the wave is regulated by the ellptical polarizer. The sinusoidal grooves...
Electron Cyclotron Resonance Frequency (ECRF) systems in future fusion devices, like the DEMO-nstration reactor, foresee an operational frequency in the range 230-280 GHz to match the plasma characteristics. The Cyclotron Auto Resonance Masers (CARM) characterized by a high value of a frequency Doppler up-shift, could represent an alternative to gyrotrons and the design of a 250 GHz, 0.5 MW...
Purpose of this work is to describe the DONES Central Instrumentation and Control System (CICS). A functional definition of the main systems will be given, together with a general overview of the current status of the CICS and the differences with respect to the corresponding system developed during the IFMIF-EVEDA phase. The overall architecture of the Control System, the definition,...
Due to the high precision requirements of HL-2M tokamak sub-components assembly, the survey control network with high precision should be established. With high accuracy distance measurement of laser tracker system and distance intersection method, the local coordinates of reference points(or the relative locations of the reference points) in the survey control network are calculated. There...
In the EC-driven (8.2 GHz) steady-state plasma on QUEST, plasma current seems to flow in the open magnetic surface in the outside of the closed magnetic surface in the low-field region according to plasma current fitting (PCF) method. First, plasma equilibrium solution was fitted assuming all plasma current is flowing in the inside of the LCFS. It was solved within isotropic pressure profile...
Disruption simulations with DINA code are performed for JT-60SA design. The simulation results have been applied to design of many components, not only for the vacuum vessel and in-vessel components, but also for peripheral components. For instance, for design of in-vessel coils, the stabilizing plate and magnetic sensors, EM force induced by hallo current and eddy current at disruption were...
ITER’s CODAC archiving system manages currently three different sets of data: DAN, SDN and PON, that correspond with the data that is transmitted by different networks: Data Archiving Network (data produced by data acquisition, diagnostics and data analysis), Synchronous Data Network (real time control network), and Plan Operational Network (control data). In this sense, ITER’s CODAC data...
On tokamaks, there are many diagnostics, which need real-time data acquisition and processing to provide useful information for plasma control. Some of the diagnostics required fast processing of multiple very high sampling rate signals. It is often difficult to achieve even with modern multi-core CPUs. This is due to moving large amount of data from the digitizers to the system ram would hurt...
Recent research on disruption prediction shows that predictors based on analysing the amplitude evolution of magnetic signals outperforms the results obtained by using simple thresholds. To accomplish this, the disruptive and non-disruptive information of discharges can be compressed into two centroids in a particular parameter space (PS). During a running discharge, points in the PS are...
ITER CODAC is the most sophisticated tokamak control and data acquisition system. The core of ITER CODAC is built around the EPICS toolkit. EPICS is very mature in accelerator community. However, there are still works trying to improve existing control system software like tango and EPICS 7 mainly driven by the needs of more flexible system and development of computer technology. This paper...
The Plasma Control System (PCS) is one of the main ITER systems. It is in charge of running the plasma discharge, by receiving data from the real-time diagnostics, and by computing the commands to be processed by various plant systems to act on the plasma (e.g., the power supplies of the poloidal field coil circuits and the additional heating systems).
To this aim, the PCS will implement...
The four-barrel, two-stage gun Ignitor Pellet Injector (IPI) was developed in collaboration between ENEA and ORNL to provide cryogenic Deuterium pellets of different mass and speed to be launched into tokamak plasmas with arbitrary timing. The prototype injector is presently located at Oak Ridge (TN, USA), and is normally operated locally through a control and data acquisition system developed...
The limits for the heat loads on the DEMO first wall are significantly stricter compared to those of ITER due to cooling and breeding blanket requirements. In addition to the thermal particle and radiation loads, fast particles in the form of fusion alphas and NBI ions with high energies can escape the confinement due to various magnetic perturbations and produce a significant heat load on the...
This paper describes the preliminary design of a position, current and shape control for DEMO tokamak. This preliminary design relies on the availability of magnetic sensor measurements for the vertical position and for the plasma-wall gaps. The controller is designed basing on the CREATE-L model of the DEMO 2017 Single-Null (SN) configuration, and then is tested using the nonlinear evolution...
Plasma behavior in the SOL of tokamaks is driven by turbulence in the edge region where density and temperature gradients are large. This generates intermittent structures of increased density and temperature known as filaments, which extend along the magnetic field lines. The protection of plasma facing components in the next step devices is a primary concern. In this context, the...
The risk of damaging the metallic PFCs on JET-ILW by beryllium melting or cracking of tungsten owing to thermal fatigue requires a reliable active protection system: it shall avoid damage to the plasma-facing components (PFCs). To address this issue, a real-time protection system comprising newly installed imaging diagnostics, real-time algorithms for hot spot detection and alarm handling...
Single crystal Molybdenum is one of the most promising materials for the First Mirror (FM) for ITER optical diagnostics due to high resistance to erosion under the neutral atom bombardment. Other advantages are: low CTE, high thermal conductivity, good mechanical properties at elevated temperatures. The FMs are normally located in the front-end of ITER port plugs, being subject to the...
In the past two campaigns of Wendelstein 7-X stellarator the overview video diagnostics played an important role in the daily experiments. The current software implementations went through numerous improvements and changes according to the continuously changing requirements. However, while the control software could handle all the needs, the changes reached a point where the redesign and...
The Plasma Position Reflectometry (PPR) diagnostic system (PBS 55F3) is planned to provide information related to the edge electron density profile and plasma position at four defined locations distributed both poloidally and toroidally in the ITER vacuum vessel.
The sections of the ex-vessel transmission lines (TL) in the Gallery, between the two secondary confinement barriers, are...
The paper reports on measurements of neutron flux emitted from a 14 MeV DT neutron generator. Such devices are widely used in material sciences, industry, medicine, etc. The used neutron generator (NSD-35) provides a controllable emission of a stabilized neutron flux, up to about 2∙E8 neutrons per second in 4pi angle. According to manufacturer, more than 90% of the neutrons emitted are 14-MeV...
One of the important aspects of the plasma cleaning of the front-end mirrors (FM) in ITER UWAVS diagnostics is to understand surface roughness after multiple cleaning runs and to minimize possible contamination due to unwanted sputtering of the mirror surface and neighboring walls. The capacitive coupled RF 30-60 MHz is a candidate for the UWAVS FM cleaning. It generates ion fluxes of tens of...
Magnetic diagnostics plays an important role in tokamak operation. Magnetic data are used for real-time control of plasma current, shape and position and for post-discharge analysis of magnetohydrodynamic plasma instabilities and equilibrium reconstruction. The magnetic diagnostic of the T-15MD will consist of more than 500 inductive sensors of various types: poloidal flux loops, saddle loops,...
Over the years, humanity has needed energy constantly due increase both population and the technology. That’s why conventional methods of energy production are not enough to cover new demands especially in environmental area because of pollution generated.
Energy generated by nuclear fusion solve both problems so achieve full control over it internal process, this implies an analysis over...
A 60 keV neutral Alkali beam system was designed, built and installed for beam emission spectroscopy measurement of edge plasma on W7-X.
The injector consists of three parts: a recently developed thermionic (lithium or sodium) ion source (j≥2.5mA/cm2), a high focusing efficiency ion optic (~50% of the extracted current can be found in the plasma) and a newly developed recirculating...
Charge eXchange Recombination Spectroscopic (CXRS) diagnostic system was successfully applied on EAST campaign. The CXRS located at D port was designed to focus on the tangential neutral injection beam from Port A on EAST. However, the tangential beam and the perpendicular beam are always injected into the plasma at the same time for the better heating and current driving. Therefore, spectrum...
The beam emission spectrometer that shares the same collection optics with the existing core charge exchange recombination spectroscopy (cCXRS) on EAST has recently been upgraded. The enhanced system allows the simultaneous measurements of red- and blue-shifted parts of the Doppler spectrum as well as the active charge exchange line (Dα n = 3-2 656.1nm) from the main ions. One curved strip on...
A Thomson scattering system providing electron temperature and density profile requires a high energetic laser. YAG lasers amplified by flash lamps (~50-100 Hz of repetition frequency with a few Joules output) sometimes suffer from wavefront distortion and peaked beam profile. The wavefront distortion deteriorates beam profile in the far field. Since a focusing lens is used toward the plasma...
The ITER Radial Neutron Camera (RNC) Data Acquisition (DAQ) prototype is based on the PCIe protocol as the interface to be used between the I/O unit and the host PC, enabling for the scalability of the final RNC DAQ system and allowing a sustainable 2 MHz peak event to cope with the long plasma discharges, up to half an hour.
The high performance computer receives the acquired data through the...
A new fast divertor infra-red (IR) thermography system was put into operation at COMPASS. It provides full radial coverage of the bottom open divertor with pixel resolution ~ 0.6-1.1 mm/px. on the target surface (0.04-0.12 mm/px. mapped to the outer midplane) and time resolution better than 20 µs. This setup provides unique capabilities for heat flux profile measurements simultaneously in the...
A high throughput spectrometer has been developed for the measurement of the plasma ion temperature fluctuations on EAST tokamak. The designed spectrometer operates at the spectral range of 527.5±5 nm, then the emission lines from CVI at 529.1nm and NeX at 524.9nm can be observed simultaneously. The collimated and focus lenses are specially developed in order to realize the maximize...
H-alpha and Visible Spectroscopy is one of the ITER first-plasma optical diagnostics providing full poloidal coverage of plasma scrape-off layer (SOL)by two poloidal-view channels in EP11, one tangential-view channel in EP12, and one divertor-view channel in UP02. The diagnostics is composed of several optical sub-units, which transfer the SOL image to the narrow-band filtered cameras located...
Visible light high-speed imaging systems (VLHIS) are widely used in imaging and diagnosing plasmas in tokamak, e.g., plasma boundary position, structures, fast fluctuations, for its visibility and easy operation. To monitor the discharge process on J-TEXT tokamak in a high temporal resolution and study of the visible light emissivity distribution, the plasma boundary shape, a new VLHIS has...
The Wendelstein 7-X (W7-X) experiment is equipped with an Electron Cyclotron Resonance Heating installation consisting of 10 gyrotrons capable of delivering upto 7.5 MW of Electron Cyclotron Wave power at the 140 GHz resonance in the plasma. Normally, the gyrotron power is delivered in a very narrow band of several 100 MHz around the gyrotron frequency and the gyrotrons are optimized to...
The equatorial visible and infrared Wide Angle Viewing System (WAVS) for ITER is one of the key diagnostics for machine protection, plasma control and physics analysis. To achieve these objectives, the WAVS will monitor the surface temperature of the Plasma Facing Components (PFCs) by infrared (IR) thermography (3-5µm range) and will image the edge plasma emission in the visible range. It will...
The present COMPASS tokamak at the Institute of Plasma Physics in Prague is equipped with the 2-mm interferometer, which gives a possibility to measure line average electron densities up to 1.2x1020 m-3 (the critical density for the interferometer probing waves is 2.43x1020 m-3). A high magnetic field tokamak, COMPASS-U [Panek et al., Fus. Eng.des. 123 (2017) 11-16], will be designed and built...
A major modification of the RFX-mod toroidal load assembly has been decided in order to improve passive MHD control and to minimise the braking torque on the plasma, thus extending the operational space in both RFP and Tokamak configurations. With the removal of the vacuum vessel, the support structure will be modified in order to obtain a new vacuum-tight chamber and the first wall tiles will...
The ITER Poloidal-Field (PF) magnet system is composed of six circular coils consisting of superconducting winding packs made up from a stack of Double Pancakes. Due to the large coil sizes the coils PF2, PF3, PF4 and PF5 are to be fabricated adjacent to the ITER site in a dedicated PF Coil fabrication building. The cold testing of the full coils PF2 – PF6 will be carried out in the same...
The Alfvén Eigenmode Active Diagnostic system (AEAD) has undergone a major upgrade and redesign to provide a state of the art excitation and real-time detection system for JET.
The new system consists of individual 4kW amplifiers for each of the six antennas, allowing for increased current, separate excitation and real time control of relative phasing between antenna currents. The amplifiers...
The magnetic diagnostic is essential for today’s tokamaks to determine the plasma position, stability, energy content and additional parameters critical for safe operation of these devices. Conventional sensors, such as the inductive sensors, have to be supplemented by steady-state magnetic sensors in devices with long pulse capability, as is planned for the ITER reactor. In ITER, the set of...
The ITER Radial Neutron Camera (RNC) is a multichannel detection system hosted in the Equatorial Port Plug 1 (EPP 1). It is designed to measure the uncollided neutron flux from the plasma, providing information on the neutron emissivity profile and total neutron strength. The RNC structure consists of two sub-systems based on fan-shaped arrays of cylindrical collimators: the ex-port system,...
Based on the Korean Fusion Energy Development Promotion Law was enacted in 2007, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. One of the key components of the K-DEMO, the superconducting magnet system consists of 16 TF (Toroidal Field), 8 CS (Central Solenoid) and 12 PF (Poloidal Field) coils. All of the TF, CS and PF coil...
The Toroidal Field (TF) system of the Tore Supra/WEST tokamak comprises 18 NbTi superconducting coils, cooled by a static superfluid helium bath at 1.8 K and carrying a nominal current of 1255 A. The 19th December 2017, at the end of plasma run #52205, the current Fast Safety Discharge (FSD) was triggered after a quench of TFC-09.
A numerical model has been developed with SuperMagnet...
Cable-in-Conduit Conductors (CICCs) are complex systems whose behaviour is not directly predictable studying their single components, and the explanation of their observed properties is not straightforward. The knowledge of the strain (Eps_th) distribution of Nb3Sn filaments in a CICC cross-section is a key parameter in understanding the performance and its evolution when the cable undergoes...
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).
The PF6 coil winding is constructed from nine double pancakes (DPs) which are plane cylindrical solenoids of about...
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. Therefore, increase of the mechanical strength on Nb3Sn wire is the most important research subject. Recently, we approached to the solid solution strength process on the ternary...
The Central Solenoid (CS) coil of the European DEMO tokamak will consist of five modules, namely CSU3, CSU2, CS1, CSL2 and CSL3, located vertically one above the other. The central CS1 module will be subjected to the most demanding operating conditions (the highest magnetic field and mechanical loads). The design concept of the CS1 winding pack with superconductor and stainless steel grading...
The first ever built full-scale prototype of the ITER heating neutral beam injector is the MITICA experiment at PRIMA-NBTF, under realization in Padua, Italy.
The MITICA experiment includes many auxiliary plants; this paper is focused on the integration of the High Voltage Power Supply (-1 MV), hosted in a Faraday cage (HVD, High Voltage Deck) inside a dedicated Building at PRIMA-NBTF.
The...
The HTS CrossConductor (HTS CroCo) was recently proposed by Karlsruhe Institute of Technology as novel concept for the winding pack of future fusion magnets.
The conductor concept is based on a cable-in-conduit configuration (CICC), in which 6 HTS CroCo macro-strands are twisted around a round copper core, jacketed in a stainless steel conduit and cooled by forced-flow supercritical helium at...
The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. The Fusion for energy (F4E) is in charge of supplying 5 Poloidal field coils (PF2-PF6) as in-kind contributions to ITER project. In 2013, F4E commissioned the task of PF6 coil fabrication to Institute of Plasma Physics Chinese Academy of Sciences (ASIPP). The PF6 coil consists of 9 double pancakes (DPs). Before...
The high heat load on divertor target plate is one of the essential issues for future fusion reactors. In stellarator, the island divertor configuration has a long magnetic field line connection length. It is beneficial to increase the equivalent radial transport and the power decay length, and consequently reduce the peaking heat load on the divertor target plate. Therefore, it is significant...
The Acceleration Grid Power Supply (AGPS) is a system devoted to supply the acceleration grids of the MITICA experiment, the full scale prototype of the ITER Neutral Beam Injector (NBI). The AGPS is a special switching power supply with demanding requirements: high rated power (about 55 MW), extremely high output voltage (-1MV dc), long duration pulses. The procurement of the AGPS is split in...
The toroidal field (TF) coil system is one of the most mechanically stressed system in a tokamak. Structural integrity of the system must be maintained on the global and on the local scale, where the stress state in each conductor jacket as well as in insulation is to be within structural allowables. Solving this task head-on leads to very high computational demands. In this work a methodology...
The Divertor Tokamak Test (DTT) facility is a satellite experiment in the same research and development framework of the European DEMOnstrating fusion power reactor. It shall evaluate different divertor solutions for power and particles exhaust, and shall investigate the plasma-material interaction scaled to long pulse operation. It is part of the European Fusion Roadmap and shall be...
The “Divertor Tokamak Test” facility, DTT, is a project for an experimental tokamak reactor developed in Italy, in the framework of the European Fusion Roadmap.
In this design phase of the machine it is necessary to ensure the structural integrity of the superconducting magnets.
This work focuses on the analysis of the stresses that are generated in the central solenoid of the tokamak, the CS...
Due to the high intensity stray magnetic field around the tokamak device, static magnetic field immunity test is an essential procedure to verify the reliable operation of electrical and electronic equipment nearby. For the safety and reliability concerns for the Chinese Fusion Engineering Test Reactor (CFETR) and future tokamak devices, a large-scale high-intensity static magnetic field...
When the quench occurrence in the operating course of the large fusion device, huge energy stored inside the magnetic load and the maximum current flow from the superconducting load can reach 100kA with maximum inductance value to be 2H that will lead an irreversible damage on the device. By dissipating the energy by means of the fast discharge resistor(FDR) system connected in series to the...
DTT is the acronym of “Divertor Tokamak Test” facility, a project for a compact but flexible tokamak reactor which has been conceived in the framework of the European Fusion Roadmap. It will be built in Italy and shall act as a satellite experimental facility to integrate the extrapolation of the ITER results to the EU-DEMO machine. It is thus mainly aimed at the exploration of different...
The National Spherical Torus eXperiment Upgrade (NSTX-U) is an experimental device funded by the U.S. Department of Energy (DOE) at the Princeton Plasma Physics Laboratory (PPPL). NSTX-U (http://nstx-u.pppl.gov/home) is an upgrade of the original NSTX device that operated successfully for more than 10 years as a proof-of-principle demonstration of the ST concept.
During early phases of...
In the framework of the Broader Approach program, ENEA supplied the Toroidal Field (TF) coil casings for JT-60SA tokamak.
ENEA commissioned the manufacture of the full set of eighteen casings for the integration of the TF coils plus two additional spare casings to the company Walter Tosto (Chieti, Italy).
The casing is segmented in one outboard straight leg, an outboard curved leg and three...
In the process of burning fusion plasmas, plasma-facing materials such as tungsten-based materials (W) will be exposed to energetic particles of hydrogen isotopes and helium (He), high heat flux, and neutrons. In this regard, a study of accumulation of hydrogen isotopes and He in W under normal operation conditions and transit events appears necessary for assessment of safety of fusion reactor...
Actively cooled plasma facing components (PFC) are often made of CuCrZr, whereas the cooling pipes are made of stainless steel. Both materials are not easily joined, and a common solution is electron beam welding, using a ring made of Inconel or Ni as intermediate.
This paper reveals the potential of using explosive welding as an alternative joining technique for multi-material transitions of...
Nuclear fusion promises to deliver an abundant, carbon-free and clean energy source for the future. Before the realization of nuclear fusion energy, the fusion community must solve immense technological safety challenges related to tritium permeation in materials under an extreme fusion nuclear environment. Tritium behavior in materials determines two crucial safety evaluation source terms:...
Plasma Facing Components (PFC) in JET with metal ITER-like wall are subjected to high heat fluxes which can lead to damages such as beryllium melting or thermal fatigue of tungsten. The hot spots formation at the re-ionization zones due to impact of the re-ionised neutrals injected by the heating system as well as due to RF-induced fast ion losses is recognized as a big threat due to quick...
In a fusion reactor, heat exhaust is one of the most challenging engineering issues, due to the high heat flux (HHF) expected on the divertor targets. The tungsten (W) monoblock design represents one of the most suitable technological solution for plasma facing components, since it has already met the ITER requirements. However, further research is required to investigate improved solutions to...
The ITER first wall panels are exposed directly to thermonuclear plasma and must extract heat loads of about 2 MW/m² (Normal Heat Flux) to 4.7 MW/m² (Enhanced Heat Flux). The manufacturer of the normal heat flux first wall panels shall be qualified through deep high heat flux cyclic testing campaign counting thousands of cycles within the heat flux range up to 2.5 MW/m². To ensure correct...
This poster describes the main steps realized for the manufacturing of a full scale First Wall panel to ITER. This full scale prototype (FSP) is foreseen to be delivered in 2019 to F4E in order to perform high heat flux tests. The dimensions of this prototype are 1360 mm x 850 mm x 500 mm. It consists of a bi-metallic support structure made from 15-25 mm thick CuCrZr alloy embedded with...
The water-cooled lithium lead (WCLL) blanket is considered as one of the possible candidates for the EU DEMO blanket in the present EU fusion roadmap. One of the critical points of the first wall design is the maximal allowed thermal load of the Eurofer97 steel within the limiting temperature of 550 °C. Therefore, the initial reactor geometrical concept of the WCLL blanket allows a heat flux...
Hydrogen co-deposition with sputtered particles is one of the main channels of hydrogen isotope accumulation in today’s tokamaks. According to experiments in tokamaks and in labaratory conditions hydrogen concentration in co-deposits show that can be very high (up to tens of atomic percents) for various materials at low deposition temperature even in the case of low hydrogen solubility. This...
Helium cooled First wall (FW) is being developed within the breeding blanket (BB) workpackage of the EUROfusion project as one of FW options for the European DEMO. The helium cooling system has to be adapted to high thermal loads and at the same time to achieve reasonable hydraulic parameters. Moreover, different values of the heat fluxes are expected at the DEMO FW depending on the position...
Plasma-facing components based on so-called monoblocks are planned for use in the divertor region of long-pulse plasma devices such as ITER and JT-60SA due to their capacity to handle high heat fluxes with active water cooling. The plasma-facing materials that are preferred for these monoblocks are tungsten for ITER or carbon-carbon fiber composite (CFC) for JT-60SA. The requirements for the...
The Plasma Facing Units are the components of the ITER's divertor target exposed to the plasma. PFUs are cooling pipes made of copper covered by tungsten monoblocks as armour.
The non-destructive ultrasonic control is the simplest and most economical test for PFU control. It has also proved to be extremely reliable and accurate in identifying and sizing defects.
ENEA has been working on...
Tritium takes the most cost of fusion project when it has been regularly operated.In order to give a proper fuel combustion rate and recycling efficiency,it is necessary to assess the amount of hydrogen isotopes(accompanied with helium)retent in the plasma facing materials(PFM).
A comprehensive ECR plasma system for tritium (named CEPT)is designed and built for the assessment of tritium...
The Metal Rings Campaign in DIII-D allowed for studies of tungsten sourcing and transport from poloidally localized, isotopically distinct surfaces in a low-Z background. Two 5 cm wide toroidal rings of W-coated tile inserts were installed in the lower divertor of DIII-D. The outboard (shelf) ring was coated with isotopically enriched W-182; the inboard (floor) ring used a natural W coating....
Transient heat fluxes up to 1 MJ·m-2 on divertor area are expected during operation of ITER. They can lead to severe erosion of plasma-facing components. Studies on tungsten damaging under thermal shocks are widespread, but they are mainly concentrated on postmortem analysis of the exposed samples. Main feature of the experiments conducting on electron beam based test facility called BETA is...
Divertors are responsible for removing the exhaust helium ash generated after fusion in a magnetic fusion reactor. Tungsten (W) was selected as the plasma facing material in the ITER divertor region because of its high melting temperature and thermal conductivity and low sputtering erosion yield. Therefore, it is crucial to understand the behavior of hydrogen isotopes in W contained in the...
Steady-state fusion reactors and DEMO reactors will have much higher heat flux from the core than that from ITER, which itself exhibits heat flux that is several times larger than that available in the current fusion reactors. The detached plasma is effective for reducing heat load. However, since the generation of detached plasma requires to introduce a large quantity of gas, there is...
The scope of contract F4E-OPE-138 Lot 1, assigned to Ansaldo Nucleare S.p.A (ANN) by Fusion for Energy, the EU-Domestic Agency, is the fabrication and qualification of a representative full scale prototype of the International Thermonuclear Experimental Reactor (ITER) divertor inner vertical target which procurement falls under the EU responsibility.
ENEA, as major partner of the contract...
Operational reliability of the divertor target relies essentially on the structural integrity of the component, in particular, of the material interfaces, where thermal stresses tend to be concentrated. To improve bonding quality, a concept developed in the frame of the EUROfusion project WPDIV for the DEMO divertor, consists in the use of functionally graded material (FGM) as interlayer...
Erosion, co-deposition of impurities and heat load effects leads to compound formation in PFC enhancing delamination mechanisms with re-emission of dust particles detrimental to the plasma stability of fusion devices. Beryllium (Be) and tungsten (W) PFC were used in the first wall and divertor in JET, and carbon (C) has achieved new relevance as impurity in the same reactor. Earlier...
Lately, tungsten has gained considerable attention of the fusion scientific community due to its performance at high temperatures. On the other hand, tungsten is affected with a serious reduction of strength at elevated temperatures, latter being one of the main drawbacks of its usefulness as a plasma facing material in fusion reactors.1 Therefore, the main aim of this work has been to improve...
The ITER first wall (FW) panels consist of plasma facing Be tiles, the CuCrZr alloy as heat sink material, and the stainless steel as structural material. A copper layer of 1~2 mm is used between the Be tile and the CuCrZr for stress compensation. Cyclic high heat flux tests employing the electron beam facility results indicate that the failure/weak spot usually occurs at the joint corners...
Wendelstein 7-X (W7-X) is equipped with ten symmetric arranged divertor units consisting of horizontal and vertical targets each. In the current completion phase, Scraper Elements (SE) have been installed in front of two out of ten divertor units to protect the gap between the horizontal and vertical targets (pumping gap) from thermal overload out of the plasma. During the next plasma...
With the aims of high performance plasma toward ITER and even a fusion reactor, heat exhaust would be a serious problem for HL-2M. In this work, impurity seeding is considered to solve heat exhaust problem by radiative divertor. SOLPS-ITER simulations are performed for Ne and Ar impurities from three seeding locations (lower dome, inner target and outer target) with the standard lower single...
High heat flux testing is a vital part of engineering component validation for fusion technology. The Heat by Induction to Verify Extremes (HIVE) facility is designed to improve the practicalities of this aspect of component testing. It provides a faster turnaround for smaller concepts and a more cost-effective approach by utilising induction heating within a small vacuum vessel.
Due to the...
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. Part of this mission is an assessment of several alternatives to the conventional divertor configuration, including ‘Advanced...
The simulation plays an important role to estimate characteristics of cooling in plasma facing components such as blanket and divertor. An objective of this study is to perform large -scale direct numerical simulation (DNS) on heat transfer of turbulent flow on coolant water flow. The coolant flow conditions in plasma facing components are assumed to be Reynolds number of a higher order. To...
An ITER Tokamak machine with a torus shape is composed of nine units of 40◦ sectors. The sector sub-assembly tool (SSAT) is dedicated assembly tool to integrate vacuum vessel (VV) sector, VV thermal shield (VVTS) segments, VVTS port shrouds, two toroidal field coils (TFC) and various intercoil structures into 40◦ sector.
For the sub-assembly of 40◦ sector, SSAT shall have sufficient strength...
The nuclear heat and the shut-down dose rate (SDDR) in the ITER upper port 18 (UP18) was estimated to provide the nuclear heat load for the structural analysis of UP18 and to provide the basis for the further SDDR mitigation strategy of UP18. The UP18 MCNP model has been developed based on the actual CAD model, which was integrated into C-model, the global MCNP model for ITER. While ITER UP18...
Vacuum vessel which is the first confinement barrier of Tokamak fusion reactor should have numerous interfaces such as Blanket, In-vessel-coils, etc. Those interface components should be assembled by fastening of special shape of bolts to the threaded holes in the Vacuum vessel with threaded inserts and to be disassembled for maintenance during Tokamak operation. Threaded connection between...
The variation of plasma current and magnetic fields generated by superconducting magnet coils causes electromagnetic (EM) loads especially during the abrupt plasma current changes such as major disruption, the vertical displacement event (VDE) of plasma, and the fast discharge. The EM loads are one of the most important external loads for in-vessel components like blanket and divertor modules....
In the current pre-concept phase of the European DEMO, integration studies of the systems in the Upper Port area are being carried out. In DEMO, the Upper Port of the Vacuum Vessel is extraordinarily large to allow for the vertical extraction of the Breeding Blanket segments. This requires a number of components inside and outside the port to be integrated with tight space constraints: The...
The presentation is focused on approaches and results of simulations and used for loading analyses made for Upper Vertical Neutron Camera (UVNC), including spatial stress strain state, seismic analysis, electromagnetic analysis as well as the most important load combinations.
The Vertical Neutron Camera is a multichannel neutron collimator intended to measure the time resolved neutron...
The operation of nuclear fusion facilities must be carefully planned and monitored due to the potential damage to equipment or personnel caused by radiation fields. A method for visualising such three-dimensional (3D) radiation fields in real-time is presented. An interactive volumetric representation is achieved using view-dependent ray casting of a scalar field in three...
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). During the previous years a conceptual design of WCLL BB has been developed. At the current state some open issues related to the manufacturability and...
In magnetic confinement fusion ITER represents the most challenging projects conceived ever. The assessment of ITER structural behavior is not trivial since it requires the application of loads coming from different types of analysis (electromagnetic (EMAG), thermal, dynamic, etc.), which are usually run using different software and Finite Element (FE) models, onto mechanical (MECH) models...
The DEMO Oriented Neutrons Source (DONES) is the dedicated facility for testing and enabling of the qualification for different materials to be utilized in the future fusion reactor DEMO. The neutron irradiation damages not only the material samples to be tested but also impacts the plant hardware in and around the Test Cell.
For that reason, preventive and predictive maintenance activities...
This article introduces overview of inboard first wall of the JT-60SA device, especially for the initial operation phase including the first plasma. The objective of the inboard first wall is to protect magnetic sensors from plasma. There is no cooling water for in-vessel components in the initial operation phase of JT-60SA, and it will be installed in the later phase. Graphite armour tiles...
This article introduces remote handling tools for hydraulic connection of Divertor Cassette in JT-60SA, especially for cutting and aligning tools for re-welding accessing from inside of the cooling pipe. Remote handling system is necessary for the maintenance and repair of the divertor cassette in JT-60SA. Because the space around the cooling pipe connected with the divertor cassette is very...
Sector Lifting Tool (SLT) are purpose-built tool for the lifting and transferring ITER components. SLT consists of the Sector Lifting Tool (SLT) with the lifting attachments. The purpose of the SLT is to lift and transfer Vacuum Vessel (VV) and Toroidal Field Coil (TFC) from Upending Tool to Sector Sub-assembly Tool (SSAT). After the sub-assembly at SSAT in assembly hall, 40° Sector which is...
RACE has been developing a concept design for the remote maintenance system for the EUROfusion DEMO powerplant. Within the DEMO tokamak, tritium breeding blankets will require periodic replacement which is currently designed to utilize the upper vertical ports at the top of the vacuum vessel. This operation will be challenging due to the scale of the blankets (~10m tall, up to 80 tonnes). The...
The Plasma Position Reflectometry (PPR) diagnostic systems, to be installed in the International Thermonuclear Experimental Reactor (ITER), will measure the edge electron density profile of the plasma, providing real-time supplementary contribution to the magnetic measurements of the plasma-wall distance. Some of the diagnostic components will be placed inside the vacuum vessel (VV) and...
Tokamaks, as complex technical devices, need regular maintenances to insure optimal operational conditions. The major 2012-2016 shutdown, dedicated to the upgrade of Tore Supra, was the opportunity to engage important maintenance actions, preparing the restart and insuring the optimal sustainability of the future subsystems of the WEST tokamak. An overall maintenance plan, based on a risk...
The cryo-vacuum pump (CVP) system, consisting of 10 units distributed symmetrically inside the Wendelstein 7-X plasma vessel, will be installed together with the 10 units of the actively cooled high heat flux divertor. One pump each is located below the corresponding divertor, and positioned as close as possible to the flux line strike points in order to allow efficient control of plasma...
An important goal for DEMO is the tritium inventory reduction in the fuel cycle. For that, the residence time must be minimized and the tritium content in the individual fuel cycle sub-systems must be reduced. One activity foresees the implementation of an isotope rebalancing and protium removal unit - requires less recycling, has lower hold-up and has a lower residence time than cryogenic...
Steady-state and long pulse exposure of plasma-facing materials in reactor-relevant conditions are an integral step towards the qualification of next-step materials with respect to erosion, fuel retention and morphology changes in view of reactor applications.
W7-X will allow plasma operation of up to 30 minutes in its second operation phase (OP2) and thus provides an ideal framework for the...
It is foreseen from the decay heat analysis that the total decay heat from the blankets reaches up to 55.6 MW immediately after two years of the full power operation of K-DEMO with the fusion power of 2.2 GW. Especially, the estimation shows that the decay heat from an outboard blanket made of Reduced Activation Ferritic Martensitic (RAFM) steel and tungsten first wall would be tens of...
The ITER Equatorial Port #12 is a first plasma port, which has undergone the Preliminary Design Review (PDR) in November 2017. In support of the PDR, the following nuclear analyses have been conducted: i) the nuclear heat has been calculated in the port plug, as one of the principal thermal loads considered in the design, and ii) the shutdown dose rates (SDDR) have been estimated in the port...
Since 2013, CEA has carried out an in-depth modification of the Tore Supra tokamak to build the WEST platform, targeted at supporting the ITER tungsten divertor detailed design, manufacturing and operation. The changes included the modification of the magnetic configuration with new in-vessel coils, the replacement of all carbon Plasma Facing Components (PFCs) by new tungsten elements and the...
Neutral beams are one of the methods to inject power into a tokamak for plasma heating. The DIII-D tokamak has four neutral beam injectors with two ion sources each, located at toroidal angles of 30º, 150º, 210º, and 330º. As originally installed, each could inject up to 5 MW of neutral beam power in the co-injection orientation (nearly parallel to the plasma current). One of the systems, the...
This research aims to develop a two-dimensional analysis of neutron flux within the blanket modules by using a compact discharge device as neutron source and imaging plates for detector. Neutron detectable imaging plate is composed photostimulated luminescence (PSL) material and converter such as gadolinium, allowing a high spatial resolution neutron radiography in a wide dynamic range of 10^5...
An initial conceptual study of integration of reflectometry diagnostics in the European DEMO has been carried out in the previous years within the EUROfusion project. This study considered antennas and waveguides incorporated in a full poloidal section attached to the Helium-cooled Lithium Lead (HCLL) breeding blanket segments. However, this concept of a diagnostics slim cassette would reduce...
The breeding blanket First Wall is the first boundary separating the fusion plasma and its energetic particles from the rest of the Tokamak. In DEMO reactor, the First Wall integrated in the blanket is in charge of 1) removing the surface heat load connected with the charged particles and the volumetric power density arising from plasma; 2) ensuring the structural integrity of the blanket,...
Ceramic breeder pebble beds undergo complex thermomechanical interactions during blanket operation due to stress build-up and relaxation under the effects of confined thermal expansion, thermal cycling, and creep. Understanding the evolution of such processes can aid in guiding blanket design, breeder materials developments, predicting performance and possible failure modes. This study...
The Helium-Cooled Lithium Lead (HCLL) breeding blanket is one of the European blanket designs proposed for DEMO reactor. A tritium transport model is fundamental for the correct assessment of both design and safety, in order to guarantee tritium self-sufficiency and to characterise tritium con-centrations, inventories and losses. The present 2D transport model takes into account a single...
The blanket system of Korean fusion demonstration reactor (K-DEMO) has a cooling channel through which pressurized water flows to cool down the heat from nucleate heating and plasma radiation. In order to evaluate the cooling performance of blanket, a computational fluid dynamics (CFD) code has been widely used as well as used in commercial heat exchangers. However, CFD can show a large...
Pellet injection system of 20 Hz has been operated in KSTAR (Korea Superconducting Tokamak Advanced Research) since 2016. The pellet can be injected to the plasma with different size, velocity and frequency during plasma experiments. The pellet trajectory is interesting topic in KSTAR so the related investigation is carried out outside of tokamak at first. We introduce the preliminary result...
Uranium (U), which has three allotropic crystal modifications (alpha-, beta-, and gamma-U), is a strong candidate medium for storing and delivering hydrogen or hydrogen isotopes. Alpha-, beta-, and gamma-U are stable at a temperature of up to 668°C, from 668°C to 775°C, and above 775°C, respectively. Because the temperature of the uranium hydride (UHx) formation is limited at room temperature...
There are various gas components in the exhaust gas of the D-T fusion reaction. All of the hydrogen isotopes are recovered and reused as fuel, and the remaining components are released to the environment. Before releasing to the environment, all substances containing trace amounts of Q2 and Q (such as CH4) must be recovered. An oxidation / adsorption process can be used for this purpose. By...
This paper is aimed at addressing critical issues related to tritium separation in fusion reactors. One of the effective tritium separation technology is using high temperature proton conducting materials as hydrogen isotope separation membranes. When a direct current is applied to the electrochemical hydrogen pump, hydrogen and its isotopes in the anode side can be electrochemically extracted...
China Fusion Engineering Test Reactor (CFETR), the next-step fusion device of China, is proposed to design and operate in two phases. The physical parameters and machine sizes of CFETR have been updated in 2018. It is required that one blanket design can cover two operation phases of CFETR. The water cooled ceramic breeder (WCCB) blanket for CFETR phase II, one candidate CFETR blanket option,...
Nuclear reactors whether they are based on fusion (JET, ITER, DEMO), fission (e.g. CANDU type), or cooled using molted salts (MSR’s) generate significant amounts of wastes in the form of low level tritiated light water or heavy water, for which there is an increasing interest to process and recover tritium (in gas form) and deuterium (as heavy water). Current water treatment systems allow the...
While the future fusion power reactors will consume and reproduce tritium for their operations, essential amounts of tritium will be required from external sources for their initial start-ups in the commissioning periods. Up-to-date evaluations of the start-up inventories are comparable with or even exceed the available commercial tritium resources in the world nuclear industry. At present the...
Core fuelling of the EU-DEMO tokamak is under investigation within the EUROfusion Work Package “Tritium, Fuelling and Vacuum”. Pellet injection still represents the most promising option. Modelling of pellet penetration and fuel deposition profiles for different injection locations, assuming specific DEMO plasma scenarios and the ITER reference pellet mass (6×1021 atoms), indicates that...
Tritium permeation into the structural materials and further in the coolant of the fusion devices is one of the most important safety issues. Various mathematical model and experiments have been carried out to estimate the amount of tritium permeated in the key components of the fusion devices. However, some issues related to the permeation of hydrogen isotopes through metals, like those...
The high-energy particle physics Monte Carlo code toolkit GEANT4 has been expanded for fusion energy-range neutron transport simulations based on evaluated nuclear cross-section libraries. Verification and Validation (V&V) analyses were conducted with nuclear data from the ENDF/B-VII.0 and the JEFF-3.1 library to show the suitability for fusion applications. Two computational benchmarks with a...
The research activity for development of catalytic package that equips water-hydrogen catalytic isotopic exchange columns was of permanent interest for the Institute’s research team, mainly motivated by the integration of the Liquid Phase Catalytic Exchange (LPCE) process in most of the detritiation technologies for tritiated water generated from nuclear reactors.
In recent years, our...
Within the EUROFusion consortium, a big effort is made in order to analyze the electromagnetic loads that act on the in-vessel components during normal and off-normal operations, being an important input for their structural assessment. With regard to the Breeding Blanket (BB) project, a global DEMO EM model, feasible to account for different blankets design, has been developed last year with...
Fusion reactors materials (FRM) will be exposed to 14 MeV fusion neutrons and damaged up to 15 dpa/year. The investigation of neutron irradiated materials is possible only in special conditions in a hot cell. The MeV-range energy ions can be used to simulate the effect of neutron-induced damages in FRM. Such simulation experiments can be used to study the effect of displacements on the...
The In-Box Loss Of Coolant (LOCA) postulated accident is considered as a major concern for the safety involving the development of EU-DEMO fusion reactor. Related to the renewed interest in the Water Cooled Lithium Lead (WCLL) blanket concept, a unique and innovative experimental campaign is under development at ENEA Brasimone research center aiming at investigating consequences of an In-Box...
The eutectic liquid metal LiPb is considered as one of the tritium breeders of the first fusion power reactors. The flowing liquid metal dissolves alloying elements of the structural steels and thus causes their corrosion. The proposed type of the cold trap is a device providing extraction of corrosion products from liquid metal by gravity separation, which occurs at lower temperatures than...
The high peak value of nuclear heat distribution in the fusion breeding blanket is expected to make cooling system design difficult for DEMO. The maximum peak value of about 10 W/cm3 is assumed in the Test Blanket Module with the maximum operational power of 700 MW in ITER. The peak value of nuclear heat distribution in the blanket of DEMO will be increased in proportion to the operational...
The contact resistance effect in the interface between pebble beds was studied with CFD analysis. The lithium ceramics is used as breeder with the form of sphere-shaped pebbles for the extraction of the tritium in some Test Blanket Module (TBM) candidates of ITER. The flow of the gas is essential for the extraction of the tritium. The effects of the gas flow was considered. The effect of fluid...
Tritium recovery rate is one of most important parameters to design highly efficient fuel cycle in fusion reactors. To estimate the tritium recovery rate accurately, chemical reactions in the tritium recovery process must to be studied in detail. In solid type breeding blankets, tritium is expected to be released from the breeder pebbles in the form of HTO into purge gas surrounding the...
The ITER project will require large cryopumps of flat-geometry to pump the Heating Neutral Beam Injectors (NBI), and similar cryopumps to pump the diagnostic Neutral Beam (DNB). The cryogenic supply uses supercritical Helium for the cryopanels and gaseous Helium for the thermal shields of the cryopumps. The cryogenic fluids will be produced by a large cryogenic plant, and then distributed by...
To limit hydrogen leakages in a breeding blanket of fusion reactor, a hydrogen permeation barrier can be used. Erbium oxide was selected as a promising candidate with a low hydrogen diffusion. Thus, the purpose of this study is to understand the irradiation effect of helium ions, originating from fission of lithium exposed to fusion-induced neutrons in the blanket, on the hydrogen diffusivity...
In cryogenic distillation columns complex phenomena appear, some of them are neededand othersmust be avoided, such the non-uniform cooling of the distillation column or the impossibility of transfer of the cooling power to the gases mixture with major changes in the separation dynamics. The loss of separation capacity or the inability to reach optimal operating parameters are caused by...
Tritium management is one of the main challenges that future nuclear fusion energy has to achieve. Accurate tritium monitoring is a basic task in order to have relying fusion reactors. High temperature sensors have to be developed to make this monitoring a reality. Hydrogen sensors based on solid-state electrolytes can be a reliable option to perform this monitoring. These types of sensors...
Lithium 6 is the isotope 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 order...
The hydrogen isotopes separation plants have special requests related to safety operation and avoidance of radiological fluid leakage and explosion conditions. For the LPCE, part of the ICSI Rm.Valcea “Experimental Pilot Plant for Tritium and Deuterium Separation”, the process transformation from a laboratory setup into a semi-industrial plant, as well as migration from a local control to an...
On the end of 2017, in the framework of EUROfusion R&D activities, a close collaboration between EU and China has started aiming at elaborating joint strategies for the development of the Water Cooled Lithium Lead (WCLL) and the Water Cooled Ceramic Breeder (WCCB) Breeding Blanket (BB) concepts. In this framework, an intense research campaign has been carried out at the University of Palermo,...
Oxide-dispersion-strengthened (ODS) steels have been developed as one of prospective candidate materials for fast reactor cladding as well as fusion reactor blanket applications. The anisotropy in microstructure and tensile properties in the range from room temperature (RT) to 973 K of the 12Cr ODS steel with the nominal composition of Fe-12Cr-2W-0.3Ti-0.25Y2O3 (in wt.%) was investigated by...
Functional materials have diverse applications in fusion reactors and it is clear that insulators are among the most versatile groups. They are the base of all the electric and radiofrequency systems in diagnostics and heating systems from DC to very high frequencies (RF, H&CD, NBI…). Additionally, insulators are subjected to quite different conditions (voltage, temperature, frequency...)...
Remote maintenance in fusion machines such as JET and ITER relies on sliding interfaces such as bolted joints. Experience in JET, where removal torques much higher than installation values with uncoated bolts is commonplace, led to the installation of experimental bolted assemblies in 2015: the first of its kind in JET. These assemblies included some 660B stainless steel ITER Blanket-specific...
For the European demonstration power plant, four types of breeding blankets are under consideration. All designs agree in the basic materials selection, that is Eurofer used as structural material and tungsten used as armour material. Detailed thermo-mechanical finite element analyses show that a direct joint of these materials will not last over the anticipated lifetime of the blankets due to...
Shielding Integral Benchmark Archive and Database (SINBAD) project started in the early 1990’s at the Organization for Economic Cooperation and Development’s Nuclear Energy Agency Data Bank (OECD/NEADB) and the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL) with the goal to preserve and make available the information on the performed radiation...
Oxide dispersion strengthened (ODS) steel is one of the most promising candidate structural materials for fusion nuclear systems. It is widely recognized that to design and to control macroscopic materials properties of ODS steel successfully, a fundamental understanding of the atomic-scale structure and chemistry of oxide/matrix interfaces is necessary, owing to the fact that oxide/matrix...
In the case of DEMO fusion reactor, the divertor should be able to extract a steady heat flux of about 10 MW/m2. A promising concept is the W-monoblock. which should be connected to a CuCrZr or an advanced Cu ODS alloy pipe passing through the W component. Taking into account the optimum operating temperature windows for W and existing Cu-based alloys and the thermal expansion coefficients...
To date the research on structural materials for future fusion reactors has been focused on the evolution of mechanical properties with irradiation dose, energy, temperature, etc. However, the performance of materials irradiated under the presence of magnetic fields remains unclear. This aspect becomes critical, as structural materials in fusion reactors will need to withstand intense and...
The IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) is planned to deliver an intensive neutron source (5×10^16 n/s) for irradiation experiments that are confined and shielded by the Test Cell (TC). During the operation of the facility, unexpected degradation (by irradiation or corrosion) or damage (by handling etc.) of the TC leak tight liner,...
It has been paid a great attention to the production of Tungsten/Copper (W/Cu) composites, as they appear promising materials to form part of the cooling system of the divertor of the future fusion reactors. However, further assessments of the microstructure and mechanical characteristics of these composites are required for the designs of the divertor. In this study, the mechanical behavior...
Tungsten has many advantageous features; however, it is rather susceptible to oxidation at temperatures above 500 °C. By the addition of various oxide-forming elements to tungsten, self-passivation is induced. During exposure of the alloy to air, a passivation layer is formed on its surface, thereby preventing further tungsten oxidation, material degradation and related radiation spreading....
Gas Dynamic Trap (GDT) is very attractive as a kind of fusion neutron source for testing fusion materials and components as well as driving fusion-fission hybrid reactor due to its linear and compact structure, low physics and technology requirement, relatively low cost and tritium consumption. These years, the conceptual designs of GDT-based neuron source for above two purposes, named...
The use of non-evaporable getter (NEG) pumps is common in many UHV applications including surface science, analytical instruments and very large vacuum systems for high energy physics. In the past years, getter solutions based on the new sintered alloy ZAO® have been developed enabling operation in the HV regime, i.e. 10-6 Pa and above. The properties of this NEG material make it appealing for...
The high mechanical strength of ODS FS, and their resistance to creep and neutron radiation damage up to 750 ºC are attributed to extremely fine microstructures with high density of very stable nanometric precipitates, generally Y-Ti-O oxides. The STARS route (Surface Treatment of gas Atomized powder followed by Reactive Synthesis) proposed by Ceit avoids mechanical alloying to introduce...
The ability to estimate the in-service performance and lifespan of components is key to realising a commercially viable fusion energy device. The finite element method (FEM) is used to estimate performance of a component design with computational simulations. Image-based FEM (IBFEM) converts 3D images (e.g. X-ray tomography) into high-resolution models for part-specific simulations that...
The capsules of the IFMIF-DONES High Flux Test Module (HFTM) are packed densely with Eurofer specimens. A filling material (previously NaK-78 and presently sodium) is needed to fill any empty volume to improve the heat conduction and obtain uniform temperature distribution. Sodium is replacing NaK-78 because potassium generates argon isotopes leading to a pressure increase and formation of...
Electrochemical techniques such as electroplating of metals, electrochemical machining (ECM), electroforming, anodizing and electropolishing of metal surfaces have been established successfully in a variety of industrial processes. A wide range of applications are available such as the electrodeposition of decorative metal coatings on plastics and metals, corrosion protection of mass products...
A copper (Cu) alloy, having a high thermal conductivity, is a promised material for heat sink of diverters in a force free helical reactor (FFHR). Recently, Hishinuma’s group succeeded in fabrication of oxide dispersion strengthened (ODS) Cu alloys using mechanical alloying (MA) and hot isostatic pressing (HIP) process. ODS is expected to bring about high-temperature strength and irradiation...
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 workpackage WPENS as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. The objective of this workpackage is to be ready for...
W-laminates are multi layered composites realized from alternately stacked W and a second metal foils. Such materials are promising candidates for W-based structural materials for fusion reactors like DEMO or beyond concepts, due to the fact that cold-rolled ultrafine-grained thin W foils show exceptional properties in terms of ductility, toughness and ductile to brittle transition (DBT), in...
Er2O3 coatings with different structures were deposited on type 316 stainless steel substrates by magnetron sputtering and corroded by liquid lithium for corrosion resistance study. The microstructure of the Er2O3 coatings was controlled by using two different methods, one the Er metal layer was deposited and oxidized successively, and the other directly by sputtering with Er2O3 deposition....
Eurofer97 is one of the leading candidates of reduced activation ferritic martensitic (RAFM) steels for first wall structural materials of early demonstration fusion power plants. During fusion plant operation, intense neutron irradiation damage on first wall materials can cause significant irradiation embrittlement and greatly reduce the fracture toughness of RAFM steels. Therefore, it is...
Fusion systems codes are essential computational tools aimed to simulate the physics and the engineering features of a fusion power station. The main objective of a system code is to find one (or more) reactor configurations, which simultaneously comply with the physics operational limits, the engineering constraints and the net electric output requirements.
As such simulation tools need to...
This paper describes recent progress at the Idaho National Laboratory (INL) in developing the MELCOR-TMAP computer code for fusion. The MELCOR-TMAP for fusion computer code is being developed by the INL’s Fusion Safety Program (FSP) [1] by modifying the US Nuclear Regulatory Commission’s (NRC’s) MELCOR [2] computer code for fission reactor severe accident analyses. The MELCOR code was chosen...
DEMO is planned to be a prototype fusion power plant capable of demonstrating production of electricity at the level of a few hundred MW. DEMO is considered to be an intermediate step between the ITER experimental reactor and a commercial power plant. Design and assessment studies on the European (EU) DEMO are carried out by the EUROfusion consortium. The Primary Heat Transfer System (PHTS)...
Externalities are defined as a cost that arises when the social or economic activities of one group of persons have an impact on another group and when that impact is not fully accounted, or compensated for, by the first group (ExternE project). External costs are not usually considered in the total cost of electricity causing market failures. To fairly compare the different electricity...
ENEA through the Department of Fusion and Technologies for Nuclear Safety (FSN) actively participates, playing a fundamental role, in the realization of ITER, contributing with the industry to the design and construction of many components ranging from diagnostic, power supply systems, superconducting magnets and Test Blanket Module auxiliary systems. The French Nuclear Safety Authority (ASN),...
A hybrid fusion-fission (HFF) reactor based on a Reversed Field Pinch (RFP) configuration looks as an attractive option from both a technical (simple design, easy machine assembly and maintenance) as well as economic perspective (low investment costs due the absence of large Heating and Current Drive systems and superconductive toroidal field coils).
The hybrid reactor studied here has a RFP...
In the second half of this century, the European energy mix will be very likely completely decarbonized. Two main options are available to generate carbon free electricity: either to rely on renewable energy sources or to further differentiate the energy mix by including nuclear power.
In the former case a large storage capacity and/or back-up dispatchable generation are required to compensate...
The paper focuses on the design of appropriate power cycles for fusion power reactor, two S-CO2 Brayton cycles, and its positive and negative aspects. The goal of the paper is to propose a suitable power cycle and its optimization for the European fusion power plant DEMO2. Comparison of cycles in terms of using more heat resources at once is depicted. The study gives a principal preview of...
Helium-3 is a rare isotope of helium (1.37 ppm as fraction of total helium – natural abundance), with applications in medicine, industry, security, and science. Due to its high request, the world is experiencing nowadays a shortage of helium-3.
The most common source of helium-3 is the disintegration of tritium. Tritium is an unstable isotope of hydrogen, with a half-life of 12.3 years, and is...
The IFMIF (International Fusion Materials Irradiation Facility) project aiming at material tests for a future fusion power plant is now in the Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach Agreement between Japan and EU. As part of the activities the construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress at Rokkasho,...
Detritiation system (DS) is the key system to ensure safety of a fusion reactor. The DS must be designed to make sure of detritiation when an extraordinary event such as fire happens. Assuming that an accidental release of tritium and production of hydrocarbons by combustion of cables in case of fire occurs simultaneously, tritiated methane will be generated by the reaction between tritium and...
The exposure by shutdown dose and production of radioactive waste predicted from the activation analysis are interesting issues of fusion reactor facility design in the view of radiation safety. Impurities of the irradiated material, such as cobalt in the structural material, are occasionally an important factor in the evaluation of the induced activity.
Concrete is used as the neutron shield...
Large-scale R&D projects are experiencing frequent delays due to high development uncertainties. Schedule issues are creating a series of problems that are causing delays in the entire projects by increasing the cost of projects and thereby reducing the reliability resulting in delays in timely tasks such as building the R&D facilities. In this study, considering the fact that the technology...
In the last decade, it has been intensively studied to heat a compressed DT fuel to an igniting temperatures of about 5 keV by using picosecond laser pulses. In the present work, we have investigated to create high temperature (> 10 keV) plasma at relatively high densities, by using a femtosecond laser pulse combined with a specially structured micron-sized target. The structured target is...
Current models used for thermal–hydraulic analyses of forced-flow superconducting cables, used in the fusion technology, such as, e.g. Cable-in-Conduit Conductors (CICCs), are typically 1-D. They require reliable predictive expressions for the transverse mass-, momentum- and energy transfer processes between different cable components, in order to reliably simulate the behavior of any...
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 fusion power plant. CFETR detail engineering design and R&D project have been approved by Chinese government. The activities have been started in the end of last year. CFETR new design focus...
This paper reviews the material strategy of the EU fusion roadmap and the recent progress of activities within the EUROfusion materials research program. It highlights, both, the characterization and validation of in-vessel components baseline materials, i.e. EUROFER97, CuCrZr and tungsten as well as the development and characterization of advanced structural and high heat flux materials for...
The need of a high-intensity, 14 MeV-peaked neutron source for the qualification of materials under fusion-relevant conditions has been recognized in the European (EU) fusion programme as an essential step towards the design and licensing of DEMO and future commercial fusion power plants. This need has pushed the EU to support the development of a Li(d,nx) neutron source called IFMIF-DONES...
ITER is probably the most ambitious energy project in the world today, whose main objective is demonstrating the scientific and technical feasibility of nuclear fusion as an energy source. 35 nations are collaborating to build the world's largest tokamak fusion reactor, a magnetic fusion device that has been designed to prove the feasibility of fusion as a large-scale and carbon-free source of...
The Linear IFMIF (International Fusion Materials Irradiation Facility) Prototype Accelerator (LIPAc) is a key activity to demonstrate the validity of the low energy section of an IFMIF deuteron accelerator up to 9 MeV with a beam current of 125 mA in CW. For the successful acceleration of the high power beam, the input beam to the 5 MeV LIPAc RFQ should be fully characterized and controlled to...
The divertor which is used to discharge reaction energy is the core component of the EAST Tokamak. The divertor is cooled by cooling water when the EAST machine is in operation. In order to prevent the cooling water from corroding the components, and to avoid the uneven baking temperature caused by the cooling water residue during the next baking, the cooling water inside the divertor pipe...
A new Deuterium-Tritium campaign (DTE2) is planned at JET in 2019/20, with a proposed 14 MeV neutron budget nearly an order of magnitude higher than any previous DT campaigns. With this proposed budget, the achievable neutron fluence on the first wall of JET will be up to about 10E20 n/m2, comparable to that occurring in ITER at the end of life in the rear part of the port plug, where several...
An optimization approach that incorporates the predictive transport code TRANSP is proposed for tokamak scenario development. Optimization methods are often employed to develop open-loop control strategies to aid access to high performance tokamak scenarios [Fusion Eng. Des. 123 (2017) 513–517]. In general, the optimization approaches use control-oriented models, i.e. models that are...
The original DTT (Divertor Tokamak Test facility) proposal presented in 2015 [1] in the challenging area of plasma exhaust has been described in detail in [2] with a critical review of several aspects.
Afterwards, according to the conclusions of the DTT Workshop held at Frascati in June 2017, various points have been examined to improve the proposal in the design review phase.
One issue is the...
Using helium as a working gas in COMPASS tokamak is a vital part of the research program. Unfortunately plasma breakdown in helium is quite difficult owing to heliums high ionization potential. In the absence of impurities with lower ionization energies helium plasma breakdown is very sensitive to start-up parameters and thus unreliable. To mitigate this issue several methods have been...
A new high magnetic field tokamak, COMPASS-U [Panek et al., Fusion Eng. Des. 123 (2017) 11 – 16], replacing the currently operating COMPASS tokamak at IPP Prague is being designed and constructed nowadays. As a result of high magnetic field in the machine (B0 = 5 T) large forces (up to 6.5 MN) acting especially on the TF coils are expected. In order to keep the loading of the coils, which will...
An upgrade of the RFX-mod experiment is presently in the final design phase, with the main objectives of improving the control of magnetic confinement, plasma density and plasma wall interaction in both RFP and Tokamak configuration.
The achievement of these aims implies a major change and reconfiguration of the internal components of the machine assembly: the present toroidal support...
In 2017 the Wendelstein 7-X stellarator (W7-X) has performed the second experimental phase (OP 1.2a), introducing divertor operation with a inertially cooled test divertor unit (TDU), and graphite-covered first wall.
The divertor operation resulted in a substantial improvement of the plasma parameters and the inertially cooled TDU and the graphite walls also allowed much longer discharges with...
Neutral Beam Injectors (NBI) for DEMO-like reactors will need deuterium neutrals at a high energy (>0.8MeV) and a fair injector overall efficiency (>50%) for plasma heating and current drive. The neutralization efficiency of positive ions drops for energy higher than 100keV/nucleon and so NBIs based on negative ions are required. A conceptual design of injectors (so called Siphore at the IRFM...
In ITER, each Heating Neutral Beam injector (HNB) will deliver about 16.5 MW heating power by accelerating a negative ion beam up to the energy of 1 MeV for a duration of 1 hour.
To this purpose, a large RF-driven 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). SPIDER...
In arc discharge plasma, an ion species ratio depends on plasma density and ion confinement time. To increase an atomic ion species ratio in an NBI (neutral beam injector) ion source, various ion sources have been designed to have higher ratio of the plasma volume to the effective ion loss area so that the ion confinement time could be longer. For that reason, most of NBI ion sources have a...
A 3.7GHz lower hybrid current drive (LHCD) system was built on HL-2M in 2017. A Full-Active Multijunction (PAM) launcher is installed. The peak parallel refractive index is 2.25 with a range from 2.1 to 2.4. ALOHA code predicts a low Reflection Coefficient (RC) within a large range of the HL-2M plasma density. TE10 to TE30 mode converters are designed for the antenna to divide the power in...
Electron cyclotron heating and current drive (ECHCD) system with a 28 GHz gyrotron has been prepared for non-inductive electron cyclotron (EC) plasma ramp-up in the QUEST spherical tokamak (ST). Non-inductive plasma start-up using the EC waves is a key issue for advanced tokamak reactor concepts as well as for the ST concept. There are two important aspects of conducting the present ECHCD...
The KSTAR's NBI heating system achieves an output of about 5.46 MW, 8.7s through the 2017 KSTAR Campaign. The second NBI system (NBI2) has being installed to increase the output of the NBI heating system. The NBI2 system consists of one on-axis and two off-axis ion sources, and the maximum output is designed to be 6 MW based on the neutral beam output. The ion source of NBI2 is in the same...
Radio Frequency (RF) waves in the Ion Cyclotron Range of Frequency (ICRF) are successfully used to heat fusion plasmas. For a better understanding of the ICRF wave propagation, absorption and other processes in the plasma good in-vessel diagnostics are essential.
In recent years, a number of high frequency B-field probes have been installed in the ASDEX Upgrade tokamak. These probes, arranged...
A short-pulse and high power neutral beam injection (NBI) system is developed for the Versatile Experiment Spherical Torus (VEST) as a main plasma heating device. The NBI system is designed to inject above 0.5MW neutral beam heating power at the hydrogen ion beam energy of 20 keV in the pulse length of 10 ms. In addition, a design feature of VEST NBI system is changing the beam injection...
The small scale prototype negative ion source has been designedfor KSTAR negative ion source. The target performance of the ion source is to extract 0.5 A of 200 keV D-. The aperture geometries of the accelerator gridsare based on the ITER HNBI reference design and optimized for the small scale prototype ion source.The accelerator consists of plasma grid (PG), extraction grid (EG), and ground...
To achieve the high performance plasma in the Korea Superconducing Tokamak Advanced Research (KSTAR) tokamak, Neutral Beam Injection (NBI) system has been installed. The first NBI (NBI-1) was installed in 2010, which provides a 100 keV deuterium neutral beam of 5.5 MW maximum using three ion sources. The second NBI (NBI-2) with another 6.0 MW will be constructed until 2019. In this process,...
Production of negative ion plays an essential role in Neutral Beam Injection (NBI). Research on a cesium-free negative ion source using sheet plasma has being carried out. The sheet plasma is suitable to produce negative ions because the electron temperature in the central region of the plasma is as high as 10 -15 eV, whereas in the periphery of the plasma, a low temperature of a few eV is...
The ITER Heating Neutral Beam injector will be equipped with a beam source that will provide a negative beam of 40A (H2 or D2). The R&D activities undertaken in Europe to pursue this challenging goal comprise three experiments:
ELISE the half size ion source experiment operating in IPP Garching
SPIDER the full size ion source experiment at Neutral Beam Test Facility (NBTF) site in Padua...
The ITER ECRH system consists of 24 gyrotrons and their associated power supplies providing up to 24 MW millimeter wave heating power at a frequency of 170GHz, a set of transmission lines connecting the gyrotrons with the equatorial and the four upper launchers. With its high frequency this heating system provides the unique capability of driving locally current due to the small beam focus of...
The paper presents the ICR-heating system of the TSP TRINITI complex: purpose; features of implementation; characteristics; power supply system; physical condition; and also discusses the possibility of upgrading the system of ICR-heating of the TSP TRINITI complex for the Ignitor project.
According to the project of tokamak Ignitor to accelerate the plasma ignition and facilitate the access...
Negative ions play an essential role for Neutral Beam Injection (NBI) system of steady state magnetic nuclear fusion. We have development of negative ion sounrce in a cesium-free discharge by the magnetized sheet plasma device, TPD-Sheet IV [1]. Negative ions are formed by volume-production, that is, the dissociative attachment of low energy electrons (Te = 1-2 eV) to highly vibrationally...
A Test Facility (TF) has been designed and installed at SPC to allow for the commissioning of the EU gyrotrons developed in view of their integration to the ITER EC system. The first phase of operation of this TF was dedicated to the development of the EU 2MW coaxial cavity gyrotron [1,2]. The EU gyrotron development for ITER has been reoriented since then and is presently advancing a...
Fenix, the ASDEX Upgrade (AUG) flight simulator under development, is based on the Plasma Control System Simulation Platform (PCSSP) developed for ITER and the ASTRA transport code. Fenix will give a session leader the possibility to check whether the discharge will meet experimental goals prior to execution. It reads the AUG discharge program and checks if all the parameters and reference...
The electron cyclotron heating system in DIII-D comprises four 110-GHz gyrotrons and one 117.5-GHz gyrotron able to inject more than 3 MW into the plasma for an administrative limit of 5 s during the 2018 experimental campaign. In a major controls upgrade, the gyrotron high voltage reference waveform is no longer generated by an obsolete pre-programmed waveform generator but by a newly...
The development of a conceptual design for a demonstration fusion power plant (DEMO) is a key priority of the recent European fusion program. The DEMO design faces an even higher challenge taking into account that, compared to ITER, the European DEMO design has a fusion power that is four times higher and a major radius that is only 1.5 times larger than ITER. From a first review of the wall...
Plasma control design increasingly depends on fast simulations able to connect to operational plasma control system (PCS) software for iterative development. GSevolve is a nonlinear free-boundary simulation that evolves the Grad-Shafranov equilibrium including current and pressure profiles, and can connect to all versions of the DIII-D PCS operational on devices around the world. Its ability...
A careful control of poloidal field (PF) coil currents is indispensable to assure a successful plasma initiation in ITER. This requires the development of an accurate modeling tool which can evaluate PF coil current and voltage waveforms leading to a satisfactory breakdown condition. In particular, it is of most importance to provide the necessary loop voltage with a sufficiently large field...
The Southern Europe Thomson Backscattering Source for Applied Research (STAR) is a compact hard X-ray source designed by INFN for advanced applied materials-science research and founded by Progetto
MaTeRiA. MaTeRiA has been realized thanks to a partnership between the University of Calabria and CNISM (Consorzio Nazionale Interuniversitario per le Scienze fisiche della Materia).
The machine...
The magnetic diagnostic system in RFX-mod [1] includes local field probes, flux loops and saddle loops. It plays an important role in real time plasma control and in several off line studies of the plasma configurations and behaviours. The failure or malfunction of one or more components can cause negative consequences in several applications, including the plasma equilibrium and stability...
The reconstruction of the Plasma Boundary (PB) in fusion reactors is a critical task in several applications, including plasma control and several off-line studies of the plasma configurations.
A widely shared definition for PB is the Last Closed Magnetic Surface (LCMS) within the vacuum vessel. In the limits of plasma equilibrium conditions, the PB reconstruction should imply the solution of...
Integrated control of the toroidal current density profile, or alternatively the q-profile, and plasma stored energy is essential to achieve advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamics stability, and noninductively driven plasma current. The q-profile evolution is closely related to the evolution of the poloidal magnetic flux profile, whose dynamics...
The JET Real-Time Protection Sequencer (RTPS) co-ordinates responses for magnetic and kinetic actuators to protect the ITER-Like Wall from possible melting events and other undesirable scenarios. It allows programmable stop responses per pulse, based on alarms raised by other systems.
The architecture combines a modular run-time application developed using MARTe (Multithreaded Application for...
The Satellite Tokamak Programme (STP) is the main project within the Broader Approach agreement. The STP includes the construction of the JT-60SA superconductive tokamak and its exploitation as an ITER “satellite” facility. In view of JT-60SA operations, Japanese and European scientists are developing different tools to support preliminary studies. In this context, a set of tools for the...
Mitigation of the intense heat flux to the divertor is a crucial issue for safety operation of ITER and next-step devices. The divertor heat fluxes can be significantly reduced by operating detached plasmas, where a huge amount of energy carried by plasma particles is converted into isotropic radiation. In COMPASS-U the power decay length is expected to be small due to high magnetic field,...
ST40 is a new high field spherical tokamak built by Tokamak Energy Ltd. (TE). When operating at full power, the main parameters of ST40 will be: R0=0.4–0.6m, A=1.7–2.0, Ipl=2MA, Bt=3T, κ=2.5, and it will have up to 2MW of auxiliary heating power, and a pulse length of 1-2s. The first operational phase of ST40, at limited performance, took place in early 2018, and its results are presented by...
The Real Time Protection Sequencer allows the Session Leader to program magnetic and kinetic actuators in response to alarms. We will deal with the State Diagram, actuator conflicts and Jump to Termination, which drove the changes to the software and architecture in [1].
The State Diagram determines what happens when one response type follows another, e.g a main chamber hot spot response...
The runaway electron generation during plasma disruptions presents a danger to the vacuum vessel and associated instrumentation. The presented work concerns application of semiconductor detectors for study and characterization of runaway electrons events. The recent advances in the field of semiconductor detectors allow for the development of new diagnostic methods, utilizing their particle...
C/O monitor system is a specially dedicated spectrometer for Wendelstein 7-X, which is planned to be installed before the second Operational Phase (OP 2). It will be a high throughput and high time resolution system which will be able to monitor the main low-Z impurities in the plasma. It will be fixed at nearly horizontal position for the observation of Lyman-alpha of H-like ions of carbon...
Neutron spectrometer based on diamond detector has proposed for the D-D fusion neutron measurements at the KSTAR tokamak in future. Neutron flux monitoring and neutron spectrometry at the KSTAR tokamak as well as at fusion reactors are one of important diagnostics. Diamond has excellent properties in environments such as radiation harsh, high temperature, small size, and so on.
The aim of...
KSTAR plasma experiments have been carried out without baking the inner wall (called as cold wall condition) until 2016. The drift (ΔVsen/Vsen) in magnetic sensor signal was able to satisfy with the requirement for the plasma control by adjusting the offset level of the input in the integrator for several the vacuum shots. Here, the required value of (ΔVsen/Vsen) should be below 2 % during 60...
In this paper the method of complex geometrical optics (CGO) is applied to the analysis of polarymetric laser beam propagation and diffraction in tokamak plasma. The paraxial CGO reduces the problem of beam diffraction in inhomogeneous medium to the set of ordinary differential equations, which significantly simplifies calculation as compared with full-wave approach. Numerical simulations for...
Shutter systems for various diagnostics at the Wendelstein 7-X stellarator
K. Höchel, R. Laube, C. Brandt, M. Otte, M. Schülke and the W7-X Team
Max-Planck-Institut für Plasmaphysik, Greifswald, Germany
The stellarator experiment Wendelstein 7-X (W7-X) has completed the first two operational phases. During the last operation phase OP1.2a in 2017 a maximum energy limit of 80 MJ within a 24 s...
The High Field Side Reflectometry (HFSR) is one of the ITER diagnostics, which provides information about plasma state by measuring the electron density profile. HFSR diagnostics undergoes a strong action of different physical nature loads.
Five various designs have been developed and studied since 2014 as a result of interaction of Peter the Great St. Petersburg Polytechnic University, NRC...
Since impurities determine the plasma performance in magnetically-confined fusion plasmas, analyzing their behavior and controlling their confinement is significantly important for a stable and steady-state operation of fusion devices. A comparison of the impurity transport in the core region of large stellarators, such as Wendelstein 7-X and Large Helical Device (LHD) is quite useful to gain...
First mirrors (FMs) of optical diagnostics in ITER will operate in a harsh environmental conditions. Deposition of plasma-facing materials on the mirror surface will cause a degradation of the nominal diagnostic performance. During ITER lifetime the mirror optical characteristics are intended to be periodically recovered by an appropriate cleaning technique. Cleaning system based on the RF...
In a tokamak device magnetic diagnostics play a key role in the understanding of plasma physics, for control and safe operation. JET tokamak has hundreds of magnetic sensors distributed over the torus, designed to withstand neutron fluence. By the end of 2016 experimental campaign C36B, JET lost several pick-up coils used both for equilibrium reconstruction (slow coils) and MHD analysis (fast...
Paper describes recent advances in Heavy Ion Beam Probing (HIBP), a unique diagnostics to measure the core plasma potential in the T-10 tokamak (R = 1.5 m, a = 0.3 m, B_tor = 1.5 - 2.5 T). Fine focused (< 1 cm) and intense (<130 mkA) Tl+ beams with energy up to E = 330 keV, equipped by advanced control and data acquisition system provides the measurements in the wide density interval n_e =...
Magnetic measurements at long pulse magnetic confinement fusion devices require implementation of the true steady state magnetic field sensors in order to achieve required precision of plasma position measurement. Inductive sensors can suffer from a range of temperature gradient and radiation induced offsets which together with the intrinsic offsets of analogue integrators can lead to unwanted...
For an ITER optical diagnostic, the components located within the port interspace area must be equipped with the proper remotely controlled positioning/alignment mechanisms meeting the variety of functional, environmental and load requirements. For the H-alpha diagnostics the major requirements are: 1) high structural stiffness and positioning stability under the ~100kg weight of Long Focus...
For an ITER optical diagnostic, the components located within the port interspace area must be equipped with the proper remotely controlled positioning/alignment mechanisms meeting the variety of functional, environmental and load requirements. For the H-alpha diagnostics the major requirements are: 1) high structural stiffness and positioning stability under the ~100kg weight of Long Focus...
The WEST thermal fusion reactor is currently an upgraded evaluation environment for the diagnostics to-be-deployed in the constructed ITER tokamak. In particular, fast metallic impurities diagnostics is tested, that in future can increase reaction efficiency and provide safe mode of operation with the divertor. A diagnostic system has been provided in the framework of the collaboration and the...
A Neutron Activation System (NAS) is a highly useful and reliable tool for neutron flux and spectrum measurements in fusion devices such as ITER or ENS. The underlying process involves neutron irradiation of chosen material probes followed by their gamma spectroscopy. The gamma spectra and the further data analyses produce quantitative information on the neutron field characteristics in the...
Understanding of resonant interaction between particles and Alfven eigenmodes is one of the most important issues for fusion plasma research. Destabilization of modes by energetic ion population and energetic ion redistribution by unstable modes have been studied by many researchers. However, the interaction of bulk ions with modes has almost not been investigated. Because some bulk ions also...
Understanding core MHD activities is of particular interest on ITER. The fast measurement of core density fluctuations might be a viable avenue for the exploration of plasma core activities in some of the ITER operational scenarios. Fluctuation beam emission spectroscopy (BES) would utilize the diagnostic neutral beam shot into the plasma, where beam atoms get to excited states due to...
In magnetically confining plasma experiments, measurement of ion dynamics is of great
importance to study the impurity transport in the scrape-off-layer (SOL) and divertor.
Impurities from the divertor areas may degrade the core energy confinement, if transported too
far upstream in the SOL. The Doppler coherence imaging spectroscopy (CIS) is a relatively
new...
The Collective Thomson Scattering (CTS) will be the ITER diagnostic responsible for measuring the alpha-particle velocity distribution. Using mirrors, a 1 MW microwave beam is directed into the plasma via an opening in the plasma-facing wall. The microwaves will scatter off fluctuations in the plasma, and the scattered signal is recorded after transmission through a series of mirrors and...
Due to the successful program of ASDEX Upgrade (AUG) the high current converters come to their limits. At the same time the risk of failure increases because of ageing. At the moment all converters are in use and thus the area of operations is fixed. Low power, high density discharges of AUG are limited by the OH transformer flux. Therefore a new thyristor converter group, called “Group 7”, is...
Within the framework of the ITER project, CEA/SBT is in charge of the design, manufacturing and delivery of 277 Venturi tube flowmeters for the control of the superconducting magnets. Six types of flowmeters were developed for either the control of the supercritical helium flow in the magnets at cryogenic temperature (4.5 K) or the operation of the current leads at room temperature. The last...
The Experimental Advanced Superconducting Tokamak (EAST) poloidal Field Power Supply has recently implemented the upgrade of the control system. Inspired by the ITER Control Data Access and Communication (CODAC) system, experimental physics and industrial control system (EPICS) has been chosen for the control system. The Data Acquisition (DAQ) system is an important subsystem of the overall...
In the Poloidal Field (PF) Power-supply System of International Thermonuclear Experimental Reactor (ITER), the phase-control thyristor converter is used to supply power to the PF superconducting coil. A precise zero scale is provided for the thyristor trigger by synchronization technology. The method of extracting the synchronization signal on the rectifier transformer primary side is widely...
This paper describes the development and testing of a 15 kA Solid Circuit Breaker (SCB) applied to the Switch Network Unit (SNU) of the Experimental Advanced Superconducting Tokamak (EAST). The circuit scheme composed of parallel connected integrated gate-commutated thyristors (IGCT) and the diode bridge to realize bidirectional breaker ability is presented firstly. In this paper, the stray...
Current start-up experiment and Steady-State Tokamak Operation (SSTO) are performed by Electron Cyclotron Heating (ECH) in QUEST spherical tokamak. SSTO discharge can be maintained longer than 2 hours using continuous 8.2 GHz ECH. For a new ECH for SSTO, 8.56 GHz high power klystron system is in preparation. The klystron can work continuously, and its incident RF power is 250 kW maximum....
The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. Fusion for energy (F4E) is responsible for supply five of them(PF2-PF6) as in-kind contributions to ITER project. ITER PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the Poloidal Field coils cooperation agreement signed between ASIPP and F4E.
The...
The Quench Protection Circuit (QPC) can provide a fast and reliable protection for superconducting magnet. According to the conceptual design of China Fusion Engineering Test Reactor (CFETR), a high power QPC with mechanical-static hybrid switch is proposed, which is designed for the superconducting Toroidal Field (TF) magnets. The rated currents and maximum reapplied interruption voltages for...
The external bypass, as an important components of the international thermonuclear experimental reactor (ITER) poloidal field converter unit (PFCU), will provide a freewheeling loop for the load current to protect the magnets and PF converter modules from being damaged by over-current and over-voltage under fault conditions. The triggering bypass protection test is used to verify that the...
The SPIDER experiment, currently starting to operate at the Neutral Beam Test Facility (NBTF) in Padova, is the full-size prototype of the radiofrequency (RF) ion source for ITER Neutral Beam Injector (NBI). It features a 1MHz RF system including 4 generators, each one rated for delivering up to 200kW of RF active power. A single generator is composed of 2 tetrodes connected with the push-pull...
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High-temperature superconductor (HTS) CrossConductors (HTS CroCo) are twisted stacked strands built from HTS tapes, optimized for high engineering current density and easy long length production [1,2].
The production for a 35 kA DC cable demonstrator required increased amounts of HTS CroCos for which the production of HTS CroCos was extended to 8 m length. This milestone shows clearly that...
The electron cyclotron resonance heating group installs four new gyrotrons to enlarge the heating power at ASDEX Upgrade (Axially Symmetric Divertor Experiment). Gyrotrons are microwave oscillators for additional plasma heating. Therefor two new direct current power supplies are needed. One power supply is able to feed two gyrotrons.
The engineering data for both power supplies are:
- dc...
The ELM triggering with H-mode will cause energy loss of plasma in Tokamak device. The phenomenon results in the timely response of control system and motivates the rapid variation of current in PF coils which affects the AC loss and stability of superconducting magnet. The cryogenic parameters of the superconducting magnet will be analyzed according to the EAST experimental data for recent...
The Power Supply system for Resistive Wall Modes (RWM) control in the JT-60SA experiment is an Italian in-kind contribution to JT-60SA within the Broader Approach Agreement.
A very efficient control of RWM is necessary to access plasma currents with high βN values (3-5) sustained for a hundred seconds, as foreseen in the JT-60SA research plan.
The development of the whole RWM control system...
Magnet power supply (PS) system of JT-60SA is being implemented within the Broader Approach Agreement aiming to achieve first plasma in 2020. The system consists of several components such as DC power supply, switching network unit, quench protection circuit and motor-generator. All the components have their own internal controller (LCC) for stand-alone operation so the individual...
The plasma breakdown and ramp up in fusion experiments require high peaks of power, increasing with the size of the machines and plasma current value. In ITER, the power peak reaches 600 MW and in DEMO the present estimation gives values higher than 1 GW, far beyond the limits generally fixed by the power grid operators, even for special plants.
Experience from existing fusion experiments,...
The European DEMO, which will follow ITER according to the Roadmap of the European fusion program, is presently under Pre-Conceptual Design. The DEMO Plant Electrical System (PES) will include the Power Supply (PS) and the Electrical Power Generation systems. In all the present tokamaks, the main magnets, which constitute the major loads, are powered by thyristor rectifiers; in DEMO they...
In large superconducting magnet system, the release of inductive energy stored inside the superconducting coils provides a considerable potential of dc-arc hazards in case of an accident such as unmitigated quench. Safety analysis and numerical calculation have raised great concerns. This paper will present an electrical arc simulation method using Matlab/Simulink software to simulate the...
The pre-conceptual design of the European DEMO fusion tokamak is currently being developed under the coordination of the EUROfusion Consortium. This paper reports the mechanical analysis of the central solenoid (CS), which comes right after the phase of definition of the winding pack proposed by the CEA. An analytical model is firstly developed in order to approximate the required axial...
The ITER PF6 coil is composed by 9 double pancake(DP).The vacuum pressure impregnation（VPI）technology is an important process for double pancake manufacturing.The double pancake winding is contained within the VPI mould consisting of vacuum chamber and hard mould.The vacuum chamber provide the vacuum and pressure conditions required by VPI process,The hard mould provide the heating and shape...
The low-temperature superconducting (LTS) joint box is an important part of ITER HTS current leads. Enabling to provide the required functionality, the LTS joint boxes are made out of Copper-316L bi-metallic explosion plates. The bimetal interface of the joint has the direct effect on the mechanical properties of the joint and testing performance at low temperature. For this reason, the...
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. One design object of JT-60SA is maximization of the plasma volume and the instrumentation port availability in the condition that several buildings, heating instruments, and diagnostics of JT-60U are reused. The cryogenic pipe...
Tokamak operations requires high-current coils. The systems typically used to supply these coils (self-commutated converters, resistive switching network units) present several drawbacks: high reactive power, harmonics, discontinuous loads, dead times, energy dissipated as heat during breakdown. Some traditional solutions involve large flywheels and power factor compensators. The achievable...
The thermal hydraulic analysis of the DEMO cryo-magnetic system has the main objective to minimize the refrigeration power. The cryo-magnetic system includes the superconducting magnets cooled by forced flow supercritical helium at about 4.4 K, the cryo-distribution lines and valve boxes, and the cryogenic system with several cold boxes. The DEMO cryogenic system is at the pre-conceptional...
The long plasma pulse duration and the large thermal loads expected in the EU DEMO reactor, currently under design, represent a challenge for the power exhaust and ask for a new, robust design of the divertor. For this reason, the design of a satellite fusion experiment, the Divertor Tokamak Test (DTT) facility, is being pursued in Italy. This fully superconductive compact tokamak, which must...
The construction of the full-superconducting tokamak JT-60 Super Advanced (JT-60SA) is in progress under the JA-EU broader approach agreement. During cool down, nominal operation and warm-up the thermal shields, the superconducting magnets and their structures, the high temperature superconductor current leads (HTS CL) and the divertor cryo pumps have to be supplied with helium at specific...
The Reduced Activation Martensitic/Ferritic steel (RAFM steel) is known as one of the most important materials for future fusion reactor blanket and the welding process is unavoidable for blanket structure, so it is extremely necessary to research his welding performance. Some simulations and experiments about anti-fatigue of RAFM steel welding specimens have been done in this paper. The...
For a recent Japanese (JA) DEMO reactor design (Rp: 8 m size), the exhausted power to the SOL (Psep) is expected to be 200-300 MW, where large power handling (Psep/Rp = 24-35 MW/m) is required in the SOL and divertor. The conventional divertor design with the divertor leg length of 1.7 m was proposed, where the peak heat load at the divertor target was simulated to be less than 10 MW/m2 with...
Due to the higher melting point and lower sputtering yield, tungsten (W) is considered as the candidate for plasma facing materials (PFMs) in the future fusion reactors. Under working condition, W will be exposed to 14 MeV neutrons produced by D-T fusion reaction, as well as energetic particles such as hydrogen isotope, helium ion. The damages introduced by charge-exchanged particles are...
Tungsten (W) is a refractory metal foreseen as plasma-facing material in future magnetic confinement fusion devices. Furthermore, W containing metallic composite materials, such as W particle- or fibre-reinforced composites, are currently regarded as promising advanced materials to enhance the performance and integrity of highly heat loaded plasma-facing components (PFCs). In principle, W is...
The use of a liquid metal in the plasma-facing divertor components of the DEMO project has attractive properties such as self-healing and neutron resilience and a liquid metal handling system is considered as a candidate technology. The linear plasma device Magnum-PSI is capable of producing DEMO-relevant plasma fluxes which well replicate expected divertor conditions. The present research is...
In this paper, effects of initial condition on seismic analysis for the HCCR TBM-set are evaluated using finite element analysis. Because of difficulty to predict when an earthquake occurs during operation, various scenarios are considered in the structural integrity assessment in ITER. To perform the simplified analysis, it is important to understand the effects of initial conditions on the...
Understanding of hydrogen isotope behaviors in plasma-facing wall is important from viewpoints of fuel control and tritium safety. Tungsten (W) is a candidate material of plasma-facing components. Although a sputtering rate of W is low, a certain amount of W deposition layer would be formed on the plasma-facing wall during a long time operation of a fusion reactor. However, hydrogen isotope...
The high heat flux test facility HELCZA move forward in commissioning phase to the high heat flux qualification tests. The purpose of the qualification tests is the demonstration of facility’s readiness for the testing of plasma facing components primarily first wall panels at high heat loads. For the qualification tests the flat cooper FW mock-up with representative size for electron beam...
A massive tungsten divertor, Div-III, was installed into ASDEX Upgrade (AUG) in 2014. Div-III is an adiabatically loaded component and consists of massive tungsten tiles clamped into their supporting structures. Before installing the new component, extensive studies, including Finite Element Modeling and high heat flux tests in the test facility GLADIS, were carried out. After the first...
Tungsten (W) is considered as a primary candidate material for plasma facing components, which endures high fluence plasma in future fusion reactor. Extremely insoluble helium(He) introduced into W exposed to high fluence He plasma tend to agglomerate into bubbles, which play a significant role in the radiation-induced micro-structural evolution and properties degradation. Therefore, to...
The GDOES (Glow Discharge Optical Emission Spectrometry) technique has been already used for investigation of erosion/re-deposition processes of W coated CFC tiles from JET. The challenge for GDOES was to measure the depth profile of deuterium together with the other elements from the W coatings exposed to JET plasmas. A tentative investigation using GDOES for deuterium analysis was performed...
Plasma facing components (PFC) within a fusion device are subjected to a harsh operating environment such as high heat fluxes and exposure to high flux of hydrogen isotopes. This exposure can lead to a high fuel retention that can raise serious concern from safety point of view. One of the reason for the use of W as a material for construction of the first wall is aimed to reduce the fuel...
At present, the study of structural and functional materials’ properties of fusion reactors is carried out as a part of implementation of ITER and DEMO projects.
The research of liquid metals application possibility as a plasma facing material (PFM) is one of the most important area. Studies carried out on this subject have shown that lithium is a good material for use as a PFM in a fusion...
In future fusion reactor, divertor is a key component where large heat flux from the confined plasma need to be exhausted. In CFETR Integration Design Platform, where the unified environment for both physics and engineering design is provided, a workflow is developed for the design of the geometry of divertor targets. According to the 0-D design and equilibrium of the designed plasma shape,...
In 1998 Makhankov [1] described the concept of modular exchangeable plasma facing components (PFCs) based on liquid metal heat pipes which are radiatively cooled. Here we present results from recent experiments with a lithium filled tubular heat pipe owned by Sandia National Laboratories. The tantalum envelope (~20mm diameter by ~200mm long) was heated on its side wall using a hydrogen plasma...
The European roadmap to fusion electricity attributes critical importance to the development of reliable plasma-facing component (PFC) technology. In the EUROfusion divertor project, a range of PFC concepts are explored, and two “Phases” of small-scale mock-up design, manufacture and high heat flux testing are underway. The rationale is to understand shortcomings revealed by the first phase in...
Erosion, deposition, and fuel retention on different plasma-facing components (PFC) are critical issues for next-step fusion devices such as ITER and DEMO. Since 2011, JET has been operated with the ITER-like wall (ILW): tungsten (W) in the divertor and beryllium (Be) in the main chamber. So far there have been three experimental campaigns in 2011-2012 (ILW-1), 2013-2014 (ILW-2) and 2015-2016...
The development of the future fusion reactor is highly associated to the improvement of the current materials, specially tungsten materials, to withstand the extreme conditions giving inside the reactor vessel during service life. Tungsten has extraordinary physical characteristics as plasma facing materials (high thermal conductivity, sputtering resistance and melting point). However, it has...
Several European First Wall (FW)/blanket concepts for DEMO are high-pressure (8 MPa) Helium-cooled systems. Typical radiative steady state loads delivered from the fusion plasma are predicted up to 0.45MW/m², but peak values could reach and excess 1MW/m². At such high heat fluxes, it is a major engineering challenge to dimension the cooling for moderate temperatures and moderate stresses,...
The first wall panels of the ITER main chamber will be completely armored with beryllium. The primary reasons for the selection of beryllium as an armor material for the ITER first wall are its low Z, high oxygen gettering characteristics and also high thermal conductivity. During plasma operation in the ITER, beryllium besides low cyclic heat loads (normal events) will be suffered by high...
Fusion fuel retention is a major issue for fusion devices. A set of divertor HFGC tile samples from first 2011-12 JET ITER-like wall (ILW-1) campaign were analysed in the research to complement the data obtained by other analysis methods [1].
Experiments were carried out using Hositrad MGT 6-300 Multi Gas Analyser with Thermal Desorption. The device operates with two Quadrupole Mass...
For the use in a fusion reactor, tungsten has unique properties such as a high melting point, low sputter yield and hydrogen retention as well as moderate activation. The brittleness below the ductile-to-brittle transition temperature and the embrittlement during operation are the main drawbacks for the use of pure tungsten in plasma facing components. Tungsten fibre-reinforced tungsten...
China Fusion Engineering Test Reactor (CFETR) is a tokamak reactor under design. Due to the maintenance, assembly of the blanket and heating, diagnostics of the plasma, ports must be opened on the vacuum vessel and the blanket module of CFETR. However, the ports affects the radiation shielding performance of CFETR, especially the equatorial ports where the neutron wall loading appears a...
The world’s largest modular stellarator, Wendelstein 7-X (W7-X), is in operation since 2015. Stepwise increase of operation parameters has its final goal in the demonstration of steady-state operation capabilities with pulses up to 30 minutes. Such pulses require a constant heating of the plasma due to losses by plasma-wall interactions, particle drift and radiation losses. The latter are...
Remote maintenance (RM) of highly activated components in the test cell (TC) is a key issue of design of accelerator-driven (IFMIF-type) fusion neutron sources. We newly propose a sophisticated RM method for the target assembly (TA) of an IFMIF-type Advanced Fusion Neutron Source (A-FNS), aiming at improving maintainability of the in-TC components in this study. Basic ideas of the proposed RM...
Remote Handling (RH) maintenance in hazardous environments is a challenging task to be performed on systems and components to ensure that they work as per design and that the requirements on the plant availability are fulfilled. According to this, systems and components have to be designed for easy assembly. This approach leads to an improvement and efficient maintenance process, reducing...
The China Fusion Engineering Test Reactor (CFETR) is the testing fusion device which would be the prototype for future commercial reactor. However, the traditional maintenance way is mainly remote handling which is time-wasting and low efficiency. To meet the demands for more complicated maintenance, it is no hesitation to start the more intelligent devices for the fusion device. The dual arm...
The ITER Cryostat, the largest SS vacuum pressure chamber ever built which provides the vacuum confinement to components operating in ITER ranging from 4.5 k to 80 k. Cryostat Design Model was qualified by ITER. As a Safety Important Class system, Design qualification at every change in its development and installation phases is mandatory. The Cryostat system is currently at manufacturing...
The ITER cryostat—the largest stainless steel vacuum pressure chamber ever built which provides the vacuum environment for components operating in the range from 4.5 k to 80 k like ITER vacuum vessel and the superconducting magnets. The Cryostat is divided into four section, of which, Base section is most complex because of its web shaped structure sandwiched between two 60mm thick plates with...
It is foreseen the most challenging tasks currently in remote maintenance of DEMO are the remote handlings of multi module blanket segment and divertor. Both of them are large and heavy, and must manoeuvre through precise trajectories in a limited space by remote manipulators. The manipulator deformation, due to the heavy payload, will deviate the handled component from the desired...
The full exchange of the ITER divertor is performed with the Divertor Remote Handling System during ITER long-term maintenance campaigns. Access to the vessel is possible through the lower maintenance port. The size of the port is highly constrained, therefore, high power density actuation systems are required to lift and transport the 10-tonne cassette assemblies in and out the vessel. It has...
Remote Handling (RH) equipment are deployed to exchange the ITER Divertor, segmented in 54 cassettes. The RH equipment are powered and controlled with water hydraulics, using self-supplied demineralised water as a pressure medium. Water hydraulics servo control technology has been successfully proven at Divertor Test Platform 2 (DTP2) with a full-scale prototype, namely Cassette...
The conceptual design activity of the Demonstration Fusion Power Reactor (DEMO) is in progress in the Power Plant Physics and Technology (PPPT) programme within the EUROfusion Consortium. In this work neutronics studies, fundamental for the nuclear design of DEMO, are presented for the horizontal lower port, an optional concept that is currently under investigation. Two possible configurations...
The fusion reactor produces a large amount of tritium dust in the vacuum vessel due to the plasma unstable events, decontamination is an indispensability dispose in the IVC maintenance and decommissioning, it also plays an extremely important role in reducing the radioactive pollutants diffusion, cumulative radiation dose and staff occupational exposure level, controlling radioactive effluent...
COMPASS Upgrade tokamak is a medium-size high-magnetic-field device currently in the conceptual design phase [Panek et al. Fus. Eng. Des. 123 (2017) 11-16]. Due to the high plasma current (up to 2 MA) and the strong magnetic field (up to 5 T), large electromagnetic forces on conducting structures surrounding plasma are expected during disruptions. To address this issue, electromagnetic loads...
One of the major challenge in the commercialisation of magnetic confinement fusion is maintaining the powerplant reactor in a sufficiently short period of time to achieve commercial levels of plant availability. To inform the development of an appropriate remote maintenance strategy for EU DEMO, a simplified, parametric model, called the Maintenance Duration Estimator (MDE), has been created...
The numerical analysis of robotic mechanisms for remote maintenance and inspection inside nuclear fusion reactors has to face several issues. Indeed, these robots are subject to large deformations, which are either induced by their own mechanical structure or by the heavy payloads which they usually handle. In many applications, robotic systems are usually modeled with rigid elements, although...
In this work the authors present the latest progresses in the conceptual design of the first wall and the main containment structures of DTT device. The previous DTT baseline design was reviewed in terms of both materials and plasma shape. This in turn led to a new all-welded double-wall vacuum vessel structure, made of AISI 316L(N) stainless steel. While the basic design has still 18 sectors,...
The banana-shaped segment transport using all vertical maintenance ports (BSAV scheme) was selected as the primary maintenance option on Japanese DEMO. Among various engineering issues on the BSAV scheme, recent progress on remote maintenance (RM) focuses on in-vessel transferring mechanism of the segment, support structure of the segment and pipe connection. In the BSAV scheme, cooperative...
In Early 2018 a new challenging project of building unique Tokamak device Compass-U has begun. Apart from others, the technical attributes will include elevated operating temperatures of vacuum vessel and plasma facing components, active cooling of temperature sensitive diagnostics, cooling of the magnetic coils to LN temperatures and etc. The listed requirements of operating parameters pose...
The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at inspecting plasma facing surfaces of in-vessel components for both damage and erosion, both of which are related to the amount of mobilised dust present in the Vacuum Vessel.
Key design improvements from the on-going IVVS preliminary design have recently been incorporated into a...
One of the principal means of controlling welding distortion of metallic welded components is the use of assembly jigs. The knowledge about the effect of jigs on weld distortion is based in the qualitative, generally neither systematic nor documented experience of each manufacturer. Such knowledge needs to be complemented by specific research for components with the tight tolerances, intricate...
During welding of large, massive and complex assemblies, such as the ITER Vacuum Vessel PS3, the accumulated shrinkage of parallel welds may cause a cumulative distortion effect which significantly varies the dimensions and geometry of the welded portion while still unfinished. In case of the PS3, as consequence of such distortion, the space left to fit the outer shell plates into to the...
EAST Tokamak is a complex fusion device which requires high-quality first wall condition for the long pulse and H-mode plasma operation. During plasma phases, some of the first wall components are damaged due to high thermal stress and electromagnetic force and control is necessary in case of doubt about their condition. Detection and locating the damaged PFCs are the precondition to determine...
Tritium breeders is required to have a high lithium atom density from the viewpoint of tritium breeding ratio. The first candidate material being studied in Japan at the present time is Li2+xTiO3 which is Li2TiO3 containing excess Lithium [1], however development of a material having even higher Lithium atom density is still under way. Li8ZrO6 has the highest lithium atom density except for...
The development of new manufacturing methods for the production of key components for nuclear fusion reactors by selective laser manufacturing (SLM) is currently under investigation at Karlsruhe Institute of Technology. SLM offers great potential compared with conventional manufacturing methods, especially for fabrication of thin- and double-walled structures like sandwich-type flow channel...
Several manufacturing routines were developed in KIT for First Wall components of the Helium Cooled Pebble Bed concept: for the Test Blanket Module of ITER in 2010-2015 and the Breeder Blanket for DEMO since 2014. The overall fabrication strategy consists of two main steps: 1) the manufacturing of a semi-finished plate penetrated by channels, and 2) the forming of the plate with channels into...
Reduced Activation Ferritic Martensitic (RAFM) steel is currently under intense consideration as a structural material for the blanket applications in fusion reactor. The concept of blanket module utilizes both solid i.e. Li2TiO3 and liquid breeder material, i.e. Eutectic Pb-17Li operating at 320-480C. The critical issues like liquid metal corrosion of RAFM steel, tritium permeation into RAFM...
In a detritiation system, the function of recombiner is oxidation of tritium components. Detritiation system of a nuclear fusion facility should maintain its detritiation performance even in an event of accidents of the facility such as fire. The major technical background on recombiner design are detailed kinetics data, impact of gaseous impurities on kinetics, characteristics on pressure...
In dual-coolant lead lithium blankets, foreseen in fusion power plants, the liquid metal PbLi flows at sufficiently large velocity to guarantee a suitable removal of the volumetric heat generated in the fluid. The moving electrically conducting fluid under the influence of the plasma-confining magnetic field induces currents that create strong electromagnetic Lorentz forces and a high...
Water cooled ceramic breeder (WCCB) blanket has been considered as the primary design in Japan. There are many thin cooling pipes in the blanket and internal pressure would be applied to the blanket by the pipe rupture. A number of efforts have been made for keeping pressure resistance of the blanket with box structure. Pressure resistance of the box structure could be reinforced with internal...
This work concerns the conceptual design of the pipework and the main equipment of the Primary Heat Transfer System (PHTS) of EU-DEMO fusion power plant. EU-DEMO is considered to be the nearest-term reactor design to follow ITER; it shall be capable of demonstrating production of electricity, operation with a closed fuel-cycle and to be a facilitating machine between ITER and a commercial...
To achieve the validation and testing of tritium breeding blanket concepts, mock-ups of breeding blankets, called Test-Blanket-Modules (TBMs), are tested in three equatorial ports of the ITER tokamak. Each TBM and its associated shield form a TBM-Set that is mechanically attached to a TBM Frame. A TBM Frame and two TBM-Sets form a TBM port plug (TBM-PP). Actually different TBM versions will be...
In liquid metal blanket concepts for nuclear fusion reactors which are currently under development, the liquid metal PbLi serves as neutron multiplier, breeder material and shield against high neutron radiation. In helium-cooled or water-cooled blanket designs, the liquid metal may flow only at very small velocity since the entire heat released in the liquid metal is removed by water or...
The Water-Cooled Lithium-Lead Breeding Blanket is a candidate option for the European DEMO nuclear fusion reactor. The blanket is a key component in charge of ensuring Tritium self-sufficiency, shielding the Vacuum Vessel and removing the heat generated in the tokamak plasma. The last function is fulfilled by the First Wall and Breeding Zone independent cooling systems.
Several layouts of the...
The UK Government has invested ~€100M to create two new UKAEA centres for fusion research – Hydrogen-3 Advanced Technology (H3AT) and the Fusion Technology Facilities (FTF) both opening in 2020-21. FTF and H3AT will foster close cooperation with industry, academia and other international laboratories to develop and transfer knowledge between partners, offering opportunities to undertake R&D to...
A new facility, CLIPPER, is being constructed at CIEMAT to investigate tritium extraction from PbLi. It consists in a forced circulation loop with the main objective of validating the technique of permeation against vacuum. Originally, CLIPPER was designed with two zones operating at different temperatures and connected through a recuperator, which gives most of the required thermal jump. In...
In thermonuclear fusion reactor, tritium generated by nuclear reaction of lithium isotope with mass number six (6Li) and neutron is used as fuel. To maintain the nuclear reaction in the reactor, it is necessary to concentrate the 6Li isotope, which exists at only about 7.8mol% naturally, to 40–90wt%. The mercury amalgam method is the only practical method, but its environmental burden is large...
Plasma enhancement gases (PEGs) (such as: nitrogen, neon, argon and other inert gases) are injected into the plasma of several tokamaks in order to reduce the power load over the plasma facing component.
The exhaust gas in DEMO reactor consists of more than 90% of unburned fuel gas (D and T) and the remaining part will be He and impurities.
In DEMO reactor it is foreseen to recover the...
Cryogenic distillation (CD) of hydrogen in combination with Liquid Phase Catalytic Exchange (LPCE) or Combined Electrolytic Catalytic Exchange (CECE) process is used for tritium removal/recovery from tritiated water both for ITER/DEMO and for fission rectors like CANDU.
Tritiated water is being obtained after long time operation of CANDU reactors, or in case of ITER mainly by the...
The He-cooled Lithium Lead breeder blanket could be developed with the utilization of relatively mature material technology, which is used by Reduced Activation Ferritic / Martensitic (RAFM) steel as the structural material in China. It is necessary to analyse the first wall structure heated because of the thermal stress problems from two coolants in the blanket. The deviation from the thermal...
In future fusion DEMO reactors, coolant and breeding materials will flow along loops experiencing very different neutron irradiation environments, from those corresponding to in-vessel systems like blankets and divertor, up to regions without significant neutron radiation farther from the reactor. Moreover, in these loops there are material extractions due to the functioning of the tritium...
The research focuses on testing carbon molecular sieve membrane (CMS) for separation of the exhaust gas from fusion reactors.
The exhaust gas in tokamak demonstration reactor (DEMO) consists of more than 90% of unburned fuel gas (D and T) and the remaining part will be He, plasma enhancement gases (PEGs) and impurities. Plasma enhancement gases (PEGs) (such as: nitrogen, neon, argon and...
Lithium metatitanate (Li2TiO3) is one of the leading candidates for application as a breeder blanket material for the helium cooled pebble bed (HCPB) concept.
During operation, transmutation of lithium contained within the blanket material as a result of neutron capture produces both tritium and helium. For efficient tritium production, high ceramic density is desirable in order to increase...
Water Detritiation Systems (WDS) are used for CANDU (to remove tritium from heavy water) and also ITER. A water detritiaton facility was developed at ICSI Rm. Valcea based on Liquid Phase Catalytic Exchange (LPCE) combined with Cryogenic Distillation (CD). Initial application was for developing Cernavoda Tritium Removal Facility, but later these combined technologies were considered for fusion...
Gas adsorption processes are widely used in industrial applications including vacuum pumping in fusion reactors (getters and cryogenic pumps). In gas adsorption flows mass transfer occurs coupled with heat transfer. Modeling of such flows must be based on kinetic theory by applying the Boltzmann equation (BE) or kinetic model equations or alternatively the Direct Simulation Monte Carlo Method...
One of the main challenges for DEMO is to overcome the Power Conversion System (PCS) issues for a pulsed reactor. PCS is a complex system where the secondary circuits, connected to the Primary Heat Transport Systems (PHTS), should be integrated into an industrial and reliable system based, as much as possible, on proven technology. PCS should be designed to remove heat from PHTSs, during pulse...
Within the framework of the EU-DEMO Breeding Blanket (BB) research activities, the Water-Cooled Lithium Lead (WCLL) BB concept is the only one which adopts pressurized sub-cooled water as coolant and a Heavy Liquid Metal (HLM), namely Pb-15.7Li alloy, as breeder and neutron multiplier.
Cooling water, characterized by operative conditions typical of PWR fission reactors (temperature in the...
At the moment, the lead-lithium eutectic was chosen as the material of the reactor blankets of ITER and DEMO. During the reactor operation radioactive tritium will be produced in the blanket, but the interaction parameters of hydrogen isotopes with lead-lithium eutectic are poorly studied. The most significant processes affecting the release of tritium from the lead-lithium eutectic is its...
In the pathway towards the achievement of a demonstration nuclear fusion reactor DEMO, the construction of a neutron irradiation plant is a priority. Within the European framework, the construction of a facility that is capable of producing a significant amount of irradiation damage in materials as soon as possible was decided. The design for this Early Neutron Source (ENS) is DONES...
The Laboratory for Laser Energetics (LLE) at the University of Rochester supports a 35 kJ, direct-drive, laser to compress DT ice to study inertial confinement fusion physics. One millimeter diameter, hollow, 6 to 10 micron-thick wall, plastic spheres are charged with deuterium-tritium (DT) gas mixtures up to several hundred atmospheres. These targets are cooled to cryogenic temperatures in...
China Fusion Engineering Test Reactor (CFETR) is proposed to be the next generation fusion device of China. The conceptual design of CFETR has been completed and the engineering design is about to start. This work gives an overview of neutronics design analyses of helium cooled ceramic breeder blanket at institute of plasma physics Chinese academy of sciences (ASIPP), for CFETR phase 1 with...
The realisation of the breeding blanket system represents one of the crucial points in the design of future generation fusion reactors. At the time being, problems regarding materials compatibility persist. To protect structural materials from the harsh environment of breeding blankets, different materials have been studied as corrosion-resistant, anti-permeation coatings. Among them, Yttrium...
A conceptual design of breeding blanket module with pressure tightness (as 17.2 MPa) against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. The cooling water for DEMO is operated at the PWR water conditions of 15.5 MPa and 290 ºC-325 ºC. Since the pressure loss of the cooling water system was 1.2MPa, the design pressure of the coolant is set to be...
The liquid lead-lithium (PbLi) blanket concept has become a promising design for fusion DEMO and power plant reactors. To promote the successful application of fusion energy, some R&D activities on the PbLi blanket have been performed, such as structure material corrosion, thermal hydraulics, magnetic-hydrodynamic (MHD) effect, coolant impurities technology and LOCA/LOFA, etc.. Therefore, it...
Beryllium pebbles produced by the rotating electrode method were selected as a reference neutron multiplier material for the DEMO fusion reactor blanket. Recently they were characterized for this application after neutron irradiation at relevant temperatures. Under neutron irradiation at elevated temperatures, beryllium suffers from significant volumetric swelling stimulated by transmutation...
Conceptual design activity for Advanced Fusion Neutron Source (A-FNS) is being carried out for neutron irradiation test of fusion DEMO reactor materials. We are to apply multi activation foils for A-FNS neutron monitor system in order to measure the neutron spectrum using an activation method with an unfolding code. It is important to evaluate the dosimetry cross section data above 20 MeV...
Plasma Laboratory for Fusion Energy and Applications at Instituto Tecnológico de Costa Rica (ITCR) plans the design of SCR-2; a Quasi-Toroidally Symmetric (QAS) two-field period modular Stellarator, aspect ratio ~5 formed by 24 modular magnetic coils. SCR-2 coils design is based on ESTELL QAS configuration (project cancelled) [1], supplied by Max Planck Institute for Plasma Physics,...
Liquid lithium can cause serious corrosion on the surface of metal structural materials that used in the blanket and first wall of fusion device. Fast and accurate compositional depth profile measurement for the boundary layer of the corroded specimen will reveal the clues for the understanding and evaluation of the liquid lithium corrosion process as well as the involved corrosion mechanism....
The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. Fusion for energy (F4E) is responsible for supply five of them(PF2-PF6) as in-kind contributions to ITER project. ITER PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the Poloidal Field coils cooperation agreement signed between ASIPP and F4E.
An...
A plasma-facing components (PFC) such as a first wall or a divertor are exposed to high heat flux due to high thermal radiation and high energy particles emitted from a core plasma. Various heat removal techniques had been proposed and developed to remove the high heat load of the PFC. We newly developed a high efficient cooling technique by inducing a turbulent flow in a cooling channel with...
Tungsten is a refractory metal with very good thermal properties i.e. high thermal conductivity, high melting point, good high-temperature fatigue strength etc. Therefore, tungsten is of great interest in thermonuclear fusion research, mainly acting as a heat sink and an erosion protection on the first wall panels. However, mechanical properties and oxidation resistance of tungsten are rather...
Diffusion bonding methods as a candidate solution for Plasma Facing Components in fusion reactors involved significant investigations over the last decades. For diffusion bonding methods the Gleeble 3800 thermomechanical simulator provides a different method for the diffusion welding process: instead of having a furnace with radiation heating and axial forces in a vacuum chamber – the heating...
Tungsten (W) is the leading solid material for fusion plasma facing component (PFC) applications because it has many desirable properties. Future fusion power systems PFCs must tolerate an extremely hostile environment that includes severe heat loads, neutron damage, and surface modifications driven by energetic particle impingement. However, W and most W-alloys exhibit low fracture toughness...
The accelerator-driven fast neutron sources of broad- and quasi-monoenergetic spectra are operated at the NPI Rez Fast Neutron Facility utilizing the Be(thick) and 7Li(C) target stations and the variable energy proton beam (up to 37 MeV) from the isochronous cyclotron U-120M. The beryllium target station standardly operated with 35 MeV proton beam is used for production of IFMIF-like...
The CoCrFeNiMn high entropy alloy (HEA) and its low-activation variants strengthened by dispersion of nano-oxides prepared via mechanical alloying were investigated. The two ways of oxide dispersion creation were verified: direct adding of oxides and internal oxidation of oxidizable elements by adding the gaseous oxygen to the alloyed powder. The grain refinement of the one phase FCC alloy by...
Control of tritium permeation through structural materials is important in terms of fuel economy and radiological safety in fusion reactors; therefore, ceramic coatings have been researched as tritium permeation barrier for decades. Microstructural changes of the coatings, such as crystallization and deterioration, influence hydrogen isotope permeation behavior under reactor operation. In...
Fatigue behaviour of the oxide dispersion strengthened CoCrFeNiMn high entropy alloy was characterised at the first time. The initial powder was prepared via mechanical alloying and the material densification was done by the SPS technique. A microstructure, as well as basic mechanical characteristics, were obtained from the tensile test. These data allowed to set parameters for fatigue...
W-V alloys have already been considered for their use in the first wall armor components, but no information about their oxidation resistance is available. To guaranty passive safety of the future fusion reactors during a loss of coolant accident and the ingress of air in the fusion vessel, a full characterization of the oxidation behavior of new tungsten based materials is required. After a...
In the framework of future generation nuclear reactors, structural materials will face environmental conditions even more challenging with the highest radiation damage levels. To deal with this, oxide nanoceramics have been proposed. Oxide nanoceramics combine the enhanced radiation tolerance of nanocrystalline materials with the chemical inertness of oxides. In this work, the properties of...
Innovation in materials technology goes in parallel with advancements in material characterization techniques. Recent years have seen a large increase in use of transmission Kikuchi diffraction (tKD) to solving complex materials science problems, including nuclear materials and irradiation damage. A lack of high statistics in transmission electron microscopy (TEM) characterization of...
The development of plasma-facing components with a high laser-induced damage threshold (LIDT) is an important part of R&D program for laser-aided diagnostics in ITER. A number of papers have been published studying LIDT of ITER materials. Most of them, however, are the investigations of integrity of first wall materials using laser radiation for simulation of pulsed plasma impact during...
A purification system for lithium loop is one of critical issues to design the accelerator-driven (IFMIF-type) fusion neutron source. It is written in IFMIF Intermediate Engineering Design Report that the levels of nonmetallic impurities such as nitrogen, oxygen, carbon, hydrogen isotopes and tritium should be less than 10, 10, 10, 10 and 1 wt.ppm, respectively, although there is little basis...
Together with the outstanding high-temperature mechanical and physical properties, the highest melting point of all the metals and thermal stability against recrystallization make tungsten (W) one of the main armour and heat sink materials candidate. Nevertheless, its applicability as a high-performance structural material is somewhat limited due to its typically brittle character at low...
IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) is an IFMIF-based neutron irradiation facility which aims at providing the irradiation data required for the construction of a DEMO fusion power plant. Comparing with IFMIF, DONES consists of only one deuteron accelerators (40 MeV and 125 mA), and utilizes only the High Flux Test Module (HFTM) for...
In order to facilitate the modeling of thermal hydraulic transients and accident scenarios in fusion reactors, a number of additional fluids were added alongside water in the original ATHENA code, which was based on RELAP5. The same libraries were later ported to both RELAP5-3D and MELCOR 1.8.6 for Fusion, both in use today. The libraries included a number of liquid metals of potential...
In Large Helical Device (LHD) at National Institute for Fusion Science, deuterium plasma experiment with d (d, n) 3He reaction was performed from March to July 2017. The neutron generated in plasma is a direct evidence of this reaction. In addition, the neutron spectrum measurement will be useful in fusion engineering to be able to estimate the activation quantity of the fusion reactor...
The Helium Cooled Pebble Bed (HCPB) blanket concept is one of four EU DEMO (Demonstration Power Plant) blanket concepts currently under development. As part of the general strategy for the qualification of the design in view of licensing and operation, several experiments have been foreseen in to be carried out in the HELOKA facility at KIT.
This work presents the experimental results of a...
Knowledge retention and transfer strategies are crucial for the success of every project. Especially on complex innovation projects, within high-tech sectors such as the aerospace industry or fusion, inter-organizational knowledge management needs to be promoted during the entire project life cycle. Project-created knowledge is based on the learning from the day-to-day activities, planning...
Within the frame of European Fusion Technology Program, two neutronics mock-up (HCLL and HCPB) experiments have been performed using 14-MeV neutron generator (NG) to validate neutronics computational tools and nuclear data library. SuperMC, which is a nuclear design and safety evaluation software system developed by the FDS Team, was verified with HCLL mock-up experiment based on the latest...
Accurate calculation of the dose rate level around nuclear facilities after shutdown can provide important reference for the operation and maintenance of nuclear device, and also has important significance for the design of radiation shielding system and the disposal of nuclide waste. In this paper, based on the advanced neutron/photon transport calculation and activation calculation of...
This paper presents application and optimization of the method for the quantification of beryllium dust particles in experimental complex HELCZA (High Energy Load CZech Assembly) by selected available analytical instruments mainly by portable fluorometry as a procedure ensuing from the NIOSH (National Institute for Occupational Safety and Health) method. The experimental complex HELCZA, a high...
Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation, SuperMC, developed by FDS Team in China, is a large-scale integrated software system. Taking neutron transport calculation as the core, SuperMC supports the whole process neutronics calculation containing depletion, radiation source term/dose/biohazard, material activation and transmutation. Besides, SuperMC...
Radiation transport models for fusion neutronics analysis are becoming increasingly complex, further exacerbating problems in the creation and integration of neutronics models found in traditional analysis methods using MCNP. Serpent 2, an alternative radiation transport code developed at VTT Technical Research Centre of Finland, is considered as a potential method for neutronics analysis in...
The neutron-induced activation of materials is an important issue for fusion facilities. Photons emitted by the activated material result in a photon field, whose spatial distribution must be taken into account when planning maintenance during shutdown and for decommissioning. One of the approaches to calculate the shut-down dose rate (SDDR) is the rigorous 2-step method (R2S) which is based...
The occupational radiation exposure (ORE) assessment is one of the key aspects of the licensing process for International Tokamak Experimental Reactor (ITER) currently under construction in Cadarache (France). As this machine is the first of its kind, the maintenance activities for the replacement and repair of components are foreseen to be frequent and complex. In this context the remote...
Rigorous-Two-Steps (R2S) is one of the most important methodologies to estimate Shutdown Dose Rate (SDR) in relevant fusion facilities. This method is based on three calculations: first, a neutron transport calculation is performed to estimate the neutron flux in the facility during the irradiation phase; then, this neutron flux is used as input data for activation calculations, primary to...
The EU DEMO reactor, under pre-conceptual design within the EUROfusion Consortium, should produce several MW electrical power from nuclear fusion by the 2050s. DEMO shall be equipped with a Primary Heat Transfer System (PHTS) to remove the thermal power deposited in the plasma-facing components and convert it into electricity, and the associated safety-related components and subsystems require...
The identification of structures, systems and components (SSCs) performing safety functions is of a paramount importance for an EU DEMO development consistent with their implications in each phase of the project: design, fabrication, commissioning, operation, maintenance, inspections and tests. Moreover, this activity has to be performed at the early design stage for the correct definition of...
The DEMO preliminary safety and operating design requirements are being defined aiming at obtaining the license with a relatively large operational domain.
The DEMO design approach is being organized, by taking into account the Nuclear Power Plant ITER and Generation IV lesson learnt. Outstanding challenges remain in areas exhibiting large gaps beyond ITER. Those require a pragmatic approach,...
ICSI has completed in 2015 the conceptual design of the Cernavoda Tritium Removal Facility (CTRF). CTRF is located at CNE Cernavoda, a NPP subsidiary of SNN Bucharest, and is sized to process heavy water from 2 CANDU reactors, treating 40 kg/h heavy water over 40 years with a detritiation factor of 100. CTRF removes tritium using liquid phase catalytic exchange (LPCE) paired with cryogenic...
ITER will operate with a full-tungsten (W) divertor made of 54 water-cooled cassettes equipped with W armoured plasma-facing units. The units that protect the vertical targets are made of W monoblocks bonded to poloidal copper-chromium-zirconium (CuCrZr) cooling tubes.
The maximum design heat flux on W monoblocks located in the vertical target part is specified as 20 MW/m2 for 10 seconds...
JT-60SA will be the world’s largest superconducting tokamak when it is assembled in 2020 in Naka, Japan (R=3m, a=1.2m). It is being constructed jointly by institutions in the EU and Japan under the Broader Approach agreement. The assembly of its 400-tonne toroidal field (TF) magnet, designed for an on-axis field of 2.25T, will be completed in summer 2018.
After cryogenic testing in Saclay,...
The Mega Amp Spherical Tokamak (MAST) is currently being extensively upgraded to provide a system that will be able to add to the knowledge base for ITER as well as testing innovative reactor systems such as the Super-X divertor. MAST-U will have increased coil system and power supply capability. Therefore, to ensure operation of MAST-U within safe engineering limits, the machine protection...
Within the Broader Approach Program between Europe and Japan for the early realization of fusion with the construction of the JT-60SA tokamak, ENEA, the Italian Agency for New Technologies, Energy and Sustainable Economic Development, was in charge, among others, to supply ten superconducting Toroidal Field (TF) coils for the JT-60SA magnet system. The related procurement started in 2011 and...
Thermo-mechanical stability, oxidation and fuel management are driving issues behind the development of new plasma-facing materials for fusion. In recent years significant progress has been made in developing new material types with enhanced toughness (fibre reinforced tungsten (W) – Wf/W) compared to bulk tungsten, Smart-W, with suppressed oxidation and also new advanced copper based...
Active control of the toroidal current density profile is one of the plasma control milestones that the National Spherical Tokamak eXperiment - Upgrade (NSTX-U) program must achieve to realize and sustain high-performance, MHD-stable plasma operation. As a first step towards the realization of this goal, a nonlinear, control-oriented, physics-based model describing the temporal evolution of...
Plasma Position Reflectometry (PPR) presents an alternative to the employment of magnetic based diagnostics in the determination of the plasma separatrix position. For future controlled nuclear fusion devices, where harsh radiation environment may induce drifts and even damage magnetic probes, PPR can play a major role as a diagnostic for plasma position control during machine operation.
PPR...
Processes of material migration, fuel retention and dust generation are key elements in studies of plasma-facing components (PFC) in the JET tokamak with ITER-Like Wall with beryllium limiters and tungsten divertor. Detailed determination of quantity, location, morphology and size of dust are carried out at JET to respond to the ITER needs for safety assessment and to provide input for...
A new concept of the liquid metal limiter/divertor, REVLOVER-D, which uses multiple free-falling jets of liquid tin as a target, has been proposed. This concept can accommodate the high heat load of several tens of MW per square meter, whereas formation of the stable continuous flow is one of key issues. To avoid the transformation of the jets into droplets due to surface tension instability,...
In the future of fusion activities, the mastering of tritium release is a challenge which needs tritium R&D as well as better understanding of tritium impacts on health and environment. We have been working for years on such topics and the purpose of this presentation is to highlight some of the outcomes of these activities.
First, tritium inventory (TI) of different fusion relevant materials...
DTT is the acronym of “Divertor Tokamak Test” facility, a project for a compact but flexible tokamak reactor which has been conceived in the framework of the European Fusion Roadmap. It will be built in Italy and shall act as a satellite experimental facility to integrate the extrapolation of the ITER results to the EU-DEMO machine. It is thus mainly aimed at the exploration of different...
The Central Solenoid (CS), a key component of the ITER Magnet system, using a 45 kA Nb3Sn conductor, includes six identical coils, called modules, to form a solenoid, enclosed inside a structure providing vertical pre-compression and mechanical support. Procurement of the components of the ITER CS is the responsibility of US ITER, the ITER Domestic Agency of the USA, while the assembly of...
The work described in this paper was performed in the frame of the European Fusion Programme (EUROfusion), Power Plant Physics and Technology (PPPT) section, Balance of Plant (BoP) work package for EU DEMO Fusion Power Plant (FPP). DEMO BoP mainly consists of the Primary Heat Transfer System (PHTS), the Intermediate Heat Transfer System (IHTS) that uses HITEC salt as coolant and is equipped...
Neutron diffraction measurements have been carried out for non-destructive characterization of the residual stress fields in a mock-up of the ITER-like divertor target plasma-facing component which consists of 4 tungsten blocks joined to a copper alloy (CuCrZr) cooling pipe via a thick soft copper interlayer. The mock-up was manufactured by the hot radial pressing technique at ENEA-Frascati in...
The design of China Fusion Engineering Test Reactor (CFETR) involves complex system structure and intricate constrains, and the design work will be performed by a great number of geographically distributed groups. To support the design work consistently and effectively, the CFETR Integration Design Platform (CIDP) is being developed [IEEE Trans. Plasma Sci. 45 (2017) 512; Fusion Eng. Des. 123...
ITER high temperature superconducting current lead is a critical component for the magnet system, which has the benefit of reduction in the heat load of the cryogenic system compare with the conventional current lead. The current lead is located in the coil terminal box and dry box in the ITER feeder system. As a warm to cold transition section, the current lead fed the huge current from the...
The WEST (W -for tungsten- Environment in Steady-state Tokamak) tokamak, is based on an upgrade of the Tore Supra machine [1]. It consists in implementing an actively cooled tungsten divertor for testing high heat flux technology. Beside the presently tested ITER divertor technology (i.e. W-monoblock), the WEST team has also devoted time, in collaboration with ASIPP (China), into the...
The development of a solution for the removal of heat and particles from the reactor is a key area of present-day fusion research as it determines the performance, lifetime and safety of future fusion power plants. The linear plasma facility Magnum-PSI is capable of exposing materials to steady-state plasma conditions similar to those foreseen in ITER and DEMO. In addition, the machine is...
The design of a superconducting magnet system of a fusion reactor candidate is usually based on the Cable-in Conduit Conductor (CICC) concept. CICC consist of complex structures with several hundreds of highly packed, multistage twisted superconductor and copper strands, cooling structures – all wrapped in thin steel foils and jacketed in relatively thick stainless-steel pipes. As recently...
Pre-conceptual design studies for a European Demonstration Fusion Power Plant (DEMO) have been in progress since 2014. At this stage, while a range of design options are being considered, it is essential that assessments are carried out of the safety and environmental impact of these options. This is not only to ensure that the DEMO plant is optimised for safety performance, but also that it...
As an important part of the Roadmap to Fusion Electricity, Europe is conducting a pre-conceptual design study of a DEMO Plant to come in operation around the middle of this century with the main aims to demonstrate the production of few hundred MWs of net electricity and to demonstrate feasibility of operation with a closed-tritium fuel cycle.
This paper provides and overview of the newly...
On the road toward fusion energy, ITER is the first fusion installation which will have enough radioactive inventory to be potentially dangerous for the public and the environment. As such, ITER has a licensed nuclear facility status and ITER Organization, the operator, has to follow a licensing process through which it has to demonstrate to the regulator that the installation is safe at all...
Reliable neutronic assessments are essential for the design and the safe operation of high performance fusion facilities. For the under construction ITER device, accurate and complete evaluations of the nuclear responses from the various radiation sources are mandatory to optimize the shielding design, guarantee a sufficient protection of critical components and minimize the occupational...
The development of the plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) [1] faces significant challenges [2]. These comprise the required reliability of operation, the high accuracy to which the plasma parameters are to be controlled, and the robustness of components and methods against any adverse effects or disturbances.
The ongoing...
In modern innovation ecosystems, academic and practitioners’ studies point out the importance of contracts and relationships between innovation actors. Among others, science-based partners, namely universities and Public Research Organizations (PROs), are valuable external sources of knowledge and fundamental pillars of the innovation ecosystems. Although consensus exists on the fact that...
