Fusion Experimental Device Research Articles

Development of FAIR conductor and HTS coil for fusion experimental device

This study is aimed at the development of high-temperature superconducting (HTS) magnets for application in a fusion experimental device next to the Large Helical Device (LHD). By applying the features of an HTS, high current density and high stability can be balanced. As a candidate conductor, REBCO tapes and pure aluminum sheets are laminated and placed in the groove of an aluminum alloy jacket with a circular cross-section, after joining a lid to the jacket using friction stir welding, and twisting the conductor to homogenize its electrical and mechanical properties. The FAIR conductor derives its name from the processes and materials used in its development: Friction stir welding, an Aluminum alloy jacket, Indirect cooling, and REBCO tapes. Initially, the degradation of the critical current of the FAIR conductor is observed, which was eventually resolved. The development status of the FAIR conductor has been reported.

Charge exchange recombination spectroscopy at Wendelstein 7-X.

The Charge Exchange Recombination Spectroscopy (CXRS) diagnostic has become a routine diagnostic on almost all major high temperature fusion experimental devices. For the optimized stellarator Wendelstein 7-X (W7-X), a highly flexible and extensive CXRS diagnostic has been built to provide high-resolution local measurements of several important plasma parameters using the recently commissioned neutral beam heating. This paper outlines the design specifics of the W7-X CXRS system and gives examples of the initial results obtained, including typical ion temperature profiles for several common heating scenarios, toroidal flow and radial electric field derived from velocity measurements, beam attenuation via beam emission spectra, and normalized impurity density profiles under some typical plasma conditions.

Modelling reflectometry diagnostics: finite-difference time-domain simulation of reflectometry in fusion plasmas

Reflectometry simulations are particularly important since they allow to assess the measuring capabilities in experimental fusion devices and to predict the performance of future ones. We present a brief overview of reflectometry and introduce the family of REFMUL* codes for time-dependent reflectometry simulation. REFMUL* codes are Finite Difference Time Domain (FDTD) that allow to set up synthetic diagnostics to assess the behaviour of reflectometry diagnostics. This is illustrated in the current manuscript using the example of the Plasma Position Reflectometers of DEMO.

Wendelstein 7-X Magnets: Experiences Gained During the First Years of Operation

The Wendelstein 7-X (W7-X) experimental fusion device went into operation in 2015 after intensive commissioning. Meanwhile, the third plasma operation phase started and ran until October 2018. W7-X has three magnet systems. The superconducting magnet system creates the main magnetic field of W7-X. It consists of 70 superconducting coils, divided into seven individual circuits with ten coils each. Seven equal power supplies provide the electrical current to power the magnets. Seven magnet protection systems are also part of the system. A magnet protection system allows fast discharge of the magnets in case of severe failures, e.g., a quench that means a sudden transition from the superconducting to the normal conducting state. A special sensor system, the quench detection system, checks the status of the magnets continuously. During each of the operation phases, the superconducting magnet system is kept under cryogenic conditions at about 4 K. For that, a helium refrigerator with total power of 7 kW at 4.5 K runs steady state 24/7. The second magnet system is the trim coil system, a set of five copper coils, placed at the outer side of the machine cryostat. The coils are powered by five identical power supplies. The third magnet system is the control coil system, a set of ten copper coils, placed inside of the plasma vessel behind the divertor targets. Ten 4-quadrant power supplies power each coil separately. The power supplies can deliver bidirectional direct currents and, as per request by the experimental program, an alternating current with adjustable frequencies between 1 and 20 Hz. An operation phase of W7-X comprises about 20 weeks. During the phase, the magnet systems are normally operated 2 or 3 days per week. The superconducting magnet system is usually switched on in the morning, kept energized during the day, and ramped down in the evening. This paper analyzes the operation phases, reports on the issues during the operation, and names countermeasures and improvements performed during the breaks between the operation phases.

Radiation Protection Design and Licensing for an Experimental Fusion Facility: The Italian and European Approaches

An experimental nuclear fusion device could be seen as a step toward the development of the future nuclear fusion power plant. If compared with other possible solutions to the energy problem, nuclear fusion has advantages that ensure sustainability and security. In particular, considering the radioactivity and the radioactive waste produced in a nuclear fusion plant, the component materials for the plant could be selected in order to limit the decay period, making recycling possible in a new reactor after about 100 yr from the beginning of decommissioning. To achieve this and other pertinent goals, many experimental machines have been developed and operated worldwide in the last decades, underlining that radiation protection and worker exposure are critical aspects of these facilities due to the high-flux, high-energy neutrons produced in the fusion reactions. Direct radiation, material activation, tritium diffusion, and other related issues pose a real challenge to demonstrating that these devices are safer than nuclear fission facilities. In Italy, for the past 30 yr, a limited number of fusion facilities have been constructed and operated, mainly at the ENEA Frascati Center, where a new one, the Italian Divertor Tokamak Test Facility (DTT), is now under development. The radiation protection approach, addressed by national licensing requirements, shows that respecting the constraints for worker exposure to ionizing radiation is not always straightforward. In the current analysis the main radiation protection issues encountered in the Italian fusion facilities are considered and discussed, and the technical and legal requirements are described. The licensing process for this kind of device is outlined and compared with that of other European countries.The following aspects are considered throughout the current study: description of the installation, plant, and systems; suitability of the area; buildings and structures; radioprotection structures and organization; exposure of personnel; accident analysis and relevant radiological consequences; and radioactive waste assessment and management.In conclusion, the analysis points out the need for special attention to the radiological exposure of workers in order to demonstrate at least the same level of safety as that reached at nuclear fission facilities.

Method for determining the shape and size of a knock-on tail using the Doppler-broadened γ-ray emission spectrum

Analysis and experimental validation of energetic non-Maxwellian components formed in ion velocity distribution functions enhance our understanding of important physical phenomena in burning plasmas. As one such energetic component, knock-on tails appear due to nuclear elastic scattering (NES). It is important to experimentally validate the NES effects and analysis model of knock-on tails. For this purpose, we propose a method for determining the shape and size of a knock-on tail from the Doppler-broadened γ-ray emission spectrum in a proton-beam-injected deuterium plasma confined in an ITER-like experimental fusion device. In the proposed method, we adopt a two-temperature Maxwellian model and utilize the γ-ray-generating 6Li(d,nγ)7Be and 6Li(d,pγ)7Li nuclear reactions. We discuss both the validity of the proposed method and the expected accuracy of the shape and size of the measured knock-on tail.

Study on Butt Weld Joint of Thick Plate Superconducting Coil Structure to Reduce Welding Residual Deformation

In a large fusion experimental device, the magnetic field in the coil is rather high to create higher magnetic field at the plasma center, and the coil current is very large resulting in generation of huge electromagnetic force. To sustain such huge electromagnetic force, thick section components are required for the coil structure of which material is generally an austenitic stainless steel to avoid magnetization. On the process of the welding design,the welding residual deformation must be taken into account to keep the position of the current center of the coil and to avoid the degradation of the superconducting conductor by the excessive welding residual strain. In this paper, a partial welding which was used for the LHD construction will be focused and the image of the conceptual welding design of ITER TF coil butt joints will be discussed for an example of the fusion application.

Design optimization of a fast-neutron detector with scintillating fibers for triton burnup experiments at fusion experimental devices.

Time-resolved triton burnup studies have been carried out to estimate the behavior of alpha particles in DD fusion experimental devices. In those studies, 14 MeV neutrons emitted through DT reactions in DD plasmas should be measured selectively in the backgrounds of DD neutrons and gamma rays. For this purpose, a scintillating-fiber (Sci-Fi) based fast-neutron detector has been adapted because of its advantages such as fast response, design flexibility in detection efficiency by changing the number of Sci-Fi, and discrimination property against 2.4 MeV neutrons produced through DD reactions and gamma rays. However, its length had conventionally been set to around 10 cm without an optimization study of its design parameters to meet the requirements as 14 MeV neutron detector. In the present study, we tested three types of Sci-Fi detectors with three different lengths and compared with the simulated results of energy deposition, through which we tried to understand the phenomena in the detection process of fast neutrons. From the results, it has been shown that, due to the self-shielding of neutrons by Sci-Fi and the attenuation of scintillation photons during the transmission process to the photomultiplier tube, the optimal length of Sci-Fi is concluded to be about 6 cm.

Design of the soft-start resistance in PSM high voltage power supply for ECRH

The soft-start resistance plays an important role in high voltage power supply (HVPS) based on pulse step modulation (PSM) for ECRH on EAST, which is fusion experimental device for total superconducting tokamak in china, and has a great influence on the transient process of system operation. Reasonable selection of the soft-start resistance can guarantee the stability of the circuit and ensure that the device withstands maximum voltage and current. So the response process is analysed detailedly by proposing DC equivalent circuit model, and the process analysis and performance comparison are given under the conditions of different soft-start resistance parameters. At last the theoretical analysis is proved by simplorer simulations. The soft-start resistance is designed for the HVPS of 140 GHz ECRH system for a smooth charging without overshoot of the capacitance, and the final experimental results show that it is in agreement with the theoretical analysis and is stable and reliable to the power devices.

Au-Ni and Rh Electroplated Coatings Evaluation for ITER Ion Cyclotron Resonance Heating Radio-Frequency Sliding Contacts

Within the large scale fusion experimental device ITER, Ion Cyclotron Resonance Heating (ICRH) system is one of the three heating systems which will supply total heating power of 20 MW (40-55 MHz) up to one hour operation. Radio-Frequency (RF) contacts are integrated within the antennas for assembly and operation considerations, which will face extremely harsh service conditions, including neutron irradiation, heavy electrical loads (RF current reaches up to 2 kA with a linear current density of 4.8 kA/m), high thermal loads and also long-duration vacuum baking at 250°C before each experimental plasma campaign. CuCrZr and 316L steel have been shown to be proper base candidate materials for ITER RF contact louvers and conductors respectively. However, in order to limit the wear and the diffusion phenomena at the RF contact as well as to reduce the contact resistance, functional protective layers should be developed. The aim of this work is to investigate Au-Ni and Rh functional layers, electroplated on CuCrZr and 316L respectively. The efficiency of the Au-Ni/Rh coated pairs was evaluated through thermal ageing diffusion tests, using EDS cross-section mapping and XRD techniques. Wear and electrical contact performances of the Au-Ni/Rh original and thermally aged pairs have also been deeply studied on a dedicated tribometer operated at ITER relevant conditions.

Implementing DevOps practices at the control and data acquisition system of an experimental fusion device

The stellarator Wendelstein 7-X (W7-X) is a fusion device designed for steady state operation. It is a complex technical system. To cope with the complexity a modular, component-based control and data acquisition system has been developed.During operation phases of W7-X components steadily evolve. For instance, measurement devices for diagnostics are improved, technical processes are optimized, experienced limits of the machine have to be taken into account or simple “bug fixing” is done. This requires continuous further development of the components while operating them at W7-X – a typical use case for DevOps practices.DevOps is a software engineering practice. The term is a compound of development and operations. It aims at shorter development cycles, while the quality of the changed system must stay at a high level. This is achieved by using a highly automated tool chain.

Safety Valve and Helium Recovery System for JT-60SA Superconducting Coils

JT-60SA is a fusion experimental device using superconducting coils and a joint international research and development project involving Japan and Europe. It is to be built in Naka, Japan using infrastructure of the existing JT-60 Upgrade experiment. The missions of the design of JT-60SA safety valve and helium recovery system are to ensure safety for the excess pressure in the superconducting coil system, and to fully recover a large amount of helium gas to storage vessels during the normal transition (quench) or the fast discharge of the coils. In this paper, we designed the JT-60SA safety valve system. Installed places of pressure relief valves, flow diagrams, a concept with a selection of valve type, the detailed design and the operation scheme, and manufacturing of safety valve units are reported.

Characterizations of thermal stability and electrical performance of Au-Ni coating on CuCrZr substrate for high vacuum radio-frequency contact application

Radio-frequency (RF) contacts—which are an example of electrical contacts—are commonly employed on accelerators and nuclear fusion experimental devices. RF contacts with a current load of 2 kA for steady-state operation were designed for application to the International Thermonuclear Experimental Reactor (ITER) device. In contrast to the typical working conditions of general commercial electrical contacts, those of RF contacts employed on fusion devices include high vacuum, high temperature, and neutron radiation. CuCrZr is currently of interest as a base material for the manufacture of louvers of RF contacts, which has excellent thermal and electrical properties and has low creep rate at 250 °C. In this study, a hard Au coating (Au-Ni) was electroplated on CuCrZr samples and the samples were then subjected to thermal aging treatment at 250 °C for 500 h in order to simulate the vacuum-commissioning process of the ITER. The effects of thermal aging on the hardness, elastic modulus, crystallite size, and compositions of the coating were investigated via microstructural and mechanical characterizations of the coating material. Metal atom migration in different coating layers during thermal aging was characterized and evaluated via scanning electron microscopy/energy dispersive X-ray spectroscopy observations of the cross-sectional surfaces, and the obtained results could be used to directly select the coating thickness for the final RF contact component. The contact resistance—an important parameter of the RF contact—was measured in a dedicated testbed built to simulate fusion reactor conditions between CuCrZr pins and stainless steel plates coated with Au-Ni and Rh, respectively.

Prospect Toward Steady-State Helical Fusion Reactor Based on Progress of LHD Project Entering the Deuterium Experiment Phase

Large Helical Device (LHD) is one of the world largest superconducting fusion experiment devices, having demonstrated its inherent advantage for steady-state operation since the start of experiments in 1998. LHD has also demonstrated reliable operation of the large-scale superconducting magnet system for almost two decades. Development of the challenging heating systems, such as negative-ion-based neutral beam injection (NBI), high-power and high-frequency electron cyclotron heating, and steady-state ion cyclotron heating, have led to wide-ranging physics and engineering achievements. LHD has progressed to the next stage, that is, the deuterium experiment starting in March 2017, which should further extend plasma parameters toward reactor-relevant regime. For establishing firm basis for designing steady-state helical fusion reactor, advanced physics research, such as on isotope effect, energetic particle confinement, and plasma–wall interaction, will be intensively performed in the deuterium experiments. In an engineering aspect, the upgrade of NBI system has been carried out in preparation to the deuterium experiment, and it should contribute to future NBI development for fusion reactors including ITER. For enhancement of the particle control, the closed divertor system has been installed with pumping capability. Diagnostics for neutron measurements are newly developed and installed for the deuterium experiment. Aligned with all the progress of LHD project in terms of engineering and physics aspects, the conceptual design activity of the LHD-type helical fusion reactor, FFHR-d1, has been programmatically conducted. In parallel to the design study, engineering research and development for the component development have been performed, including those based on employing challenging ideas such as high-temperature superconductor, liquid metal ergodic divertor, and molten-salt breeder blanket. The present status of LHD project entering the deuterium experiment phase is overviewed with putting emphasis on the engineering aspects, and then the engineering research and development activities toward steady-state helical fusion reactor are described.

Neutronics Analysis of Helium Cooled Ceramic Breeder Blanket with S-shaped Lithium Zone and Cooling Plate for CFETR

China fusion engineering test reactor (CFETR) is a tokamak fusion experimental device under design to bridge the research and development gaps between international thermonuclear experimental reactor and demonstration. Helium cooled ceramic breeder (HCCB) blanket is the candidate blanket concepts for CFETR. A blanket with S-shaped lithium zone and cooling pipes reduces the space of helium manifold and is beneficial to tritium breeding. The neutronics analyses of the HCCB blanket have been performed with the Monte Carlo code monte carlo N particle transport code and nuclear cross section data from the fusion evaluated nuclear data library-3.1b data library. The tritium breeding ratio without port is 1.22 and meets the requirements of tritium self-sufficiency, as well as, the nuclear heating power of 214 MW is achieved and the global energy multiplication factor is 1.33. The shielding performance was shown to meet the design limit. The aim of the analysis is to provide the reference for the design and optimization of CFETR blanket system.

Estimation of X-Mode Reflectometry First Fringe Frequency Using Neural Networks

One of the main challenges in X-mode reflectometry is the correct determination of the group delay measurement used for density profile reconstruction. The X-mode upper cutoff group delay measurement can be used to reconstruct the electron density profiles from the near zero density. However, due to the broad operational conditions of experimental fusion devices, the start of the upper cutoff region can occur at any probing frequency. The first fringe (FF) of the interference signal measured by reflectometry that corresponds to the start of the upper cutoff reflection is used together with the magnetic field profile to determine vacuum distance between the reflectometer antenna and the start of the plasma. An incorrect estimation of the FF probing frequency not only introduces a radial error but also a group delay error, affecting the shape of the resulting density profile. In this paper, we present the new developments in the automatic FF estimation required for the reliable reconstruction of density profiles, used in the multichannel X-mode density profile reflectometry diagnostic recently installed on ASDEX Upgrade. An improved algorithm to estimate and track the frequency of the FF along a discharge is introduced. Tests show that it is able to correctly determine the FF for most discharges. However, for a number of unanticipated cases, the algorithm provides jitter and imprecise results, introducing errors in the reconstructed density profiles. We also present a novel neural network (NN) approach for the first time for the estimation of the FF frequency. A comprehensive training set was carefully selected by experienced reflectometry diagnosticians and used to train the NN model using the open source software libraries TensorFlow and Keras. The resulting NN is able to provide more precise FF estimations than the previous algorithm. The reconstructed density profiles, using both algorithms, are presented and compared.

Retention and release of hydrogen isotopes in carbon materials priorly charged in gas phase

Thermal desorption of deuterium from isotropic graphites and a CFC material priorly charged with deuterium gas has been investigated to obtain quantitative information on hydrogen recycling and tritium inventory in fusion experimental devices and future reactors. The spectra of thermal desorption spectrometry (TDS) appear to have five peaks, which are defined as peaks 1–5 in order of the increase of temperature. Diffusion coefficients of deuterium for the absorption process are evaluated for the samples charged with deuterium gas for various absorption times. Absorption pressure dependence of TDS spectra shows a unique change in the temperatures of peak 2. As for the absorption temperature dependence of TDS spectra, peak 4 has a drastic change in the shape of spectra and also in the peak temperatures. The mechanism of desorption for peak 5 will be associated to a detrapping process with a dispersion of energies in the range of 4.0–4.9 eV.

Interfacing ATCA Hot-Swap with PCIe Hot-Plug for high-availability instrumentation in critical systems

High availability (HA) is a key requirement of next generation Fusion devices, targeting steady-state operation. HA is demanded in mission-critical systems, as is the case of experimental Fusion devices and future Fusion power plants, where safety of people, environment, infrastructure and investment is primordial. IPFN (Institute of Plasma and Nuclear Fusion, Lisbon − Portugal) developed control and data acquisition instrumentation based on the Advanced Telecommunications Computing Architecture (ATCA), composed of digitizing (nodes) and switching (hub) boards, which establish a PCIe data network through ATCA’s backplane Fabric Interface, connecting to an external host computer. The ATCA standard was selected to achieve a high level of availability, as it is a modular platform of Field Replaceable Units (FRU) used with several redundancy resources. Its mandatory hardware platform management, includes a “Hot Swap” mechanism, which allows FRU insertion and extraction without having to power off the system. From the host/software perspective, ATCA nodes and hubs correspond to PCIe endpoint and switch devices. The PCIe standard specifies a Hot-Plug procedure for device add and removal for adapter cards only, whereas for other form-factors the Hot-Plug mechanism is implementation-dependent and should be defined by the form-factor itself. Although ATCA has an extension for PCIe on Fabric Interface, it does not define Hot-Plug procedures. From an initial Hot-Plug solution developed for IPFN’s platform, the paper attempts to establish a relationship between these specifications, which could help to standardize PCIe Hot-Plug implementation for the ATCA form-factor, fostering the development of instrumentation with HA requirements.

Electron emission under uniform magnetic field of materials for fusion and space applications

The power handling capacity of RF components can be limited by a resonant process known as Multipactor effect. Multipactor-induced breakdowns can be critical to microwave systems in space communication payloads or in experimental fusion devices. Multipactor simulations can be used to predict power thresholds but the results highly depend on the electron emission properties of the RF component materials. Moreover, in both space and fusion applications, the RF devices can be subjected to DC magnetic fields. This magnetic field may affect the electron emission properties. In order to improve understanding and predictability of the multipactor effect, the assessment of DC magnetic field on the electron emission properties is required; an experimental setup has been developed to measure the Total Electron Emission Yield (TEEY) under DC magnetic field. The aim of this paper is to describe the newly developed experimental setup and the associated TEEY measurements techniques. The effect of DC magnetic field on the TEEY of copper is investigated.

Operators’ accessibility studies for assembly and maintenance scenarios using virtual reality

The development of fusion plants is more and more challenging. Compared to previous fusion experimental devices, integration constraints, maintenance and safety requirements are key parameters in the ITER project. Components are designed in parallel and we must consider integration, assembly and maintenance issues, which might have a huge impact on the overall design. That also implies to consider the operator’s feedback to assess the feasibility of accessibility or maintenance processes.Virtual reality (VR) provides tools to optimize such integration. In 2010, the CEA IRFM decided to upgrade its design tools, by using VR during the life cycle (from design to operations) of a fusion component. The VR platform is intensively used in the design and assembly studies of WEST components. In particular, feasibility of the assembly scenario is assessed by the operators involving in the real assembly work.To study this aspect, the use of static manikins is quite frequent in the industry. However, more complex studies, like the feasibility of assembly and maintenance tasks in complex and very confined environments, require enhanced features such as dynamic and biomechanically realistic virtual humans. We also study the contribution of tactile feedback to improve physical presence and interaction in the virtual environment (VE), which is very important for the validation of a given task’s feasibility and the ergonomic evaluation of the posture and gesture of the operator. In particular, we show that adapted behavior in respect to physical elements of the VE can be obtained using a dynamic co-localized representation of the subject’s body and a pseudo-haptic tactile feedback.In this paper, we present integration studies involving operators and recent advances in the assessment of maintenance feasibility.