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...
