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...
