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Abstract: Construction has just begun in the JT-60 Super Advanced (JT-60SA) tokamak project, which is part of the Japanese domestic program and the Japan – EU international program “ITER Broader Approach”. All plasma facing components (PFCs) will be actively cooled because they are exposed to high heat flux for long duration. A lower single-null closed divertor with a vertical target (VT) will be installed when experiments begin. A divertor cassette integrated with coolant pipe connections will be used for remote handling maintenance. Each cassette covers a 10° sector in the toroidal direction. Modularized PFCs such as VTs, baffles, and domes will be assembled on divertor cassettes. Static structural analysis shows that the displacement and stress in the divertor module are generally small for determining the dead weight, coolant pressure, and electromagnetic forces (EMFs). However, the support structure of the outer baffle requires improvement because eddy currents generate concentrated overturning EMFs. [ABSTRACT FROM AUTHOR]

Abstract: The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program “Broader Approach”. The maximum heat flux to JT-60SA divertor is estimated to ∼15MW/m2 for 100s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R&D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R&D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15MW/m2 as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor. [Copyright &y& Elsevier]

Abstract: In TRIAM-1M, plasma discharge can be sustained for over five hours [H. Zushi, et al., Steady-state tokamak operation, ITB transition and sustainment and ECCD experiments in TRIAM-1M, Nucl. Fusion 45 (2005) S142–S156]. In order to avoid sitting in front of one console for the purpose of monitoring the plasma discharge, it is recommended that the experimental information be accessible from any location at any time. In addition, simple services to access experimental information are required in order to promote the participation of multiple researchers in the TRIAM-1M experiment. Thus, A WEB-based integrated data processing system that provides management for experiment planning, an experimental log, numerical data, and plasma supervision has been installed in the TRIAM-1M. These services are composed primarily of an Apache WEB server, a Tomcat JSP/Servlet container, and a MySQL relational database. This system is constructed using the object-oriented Java language, which is easy to maintain and develop because of the intrinsic characteristics of the Java language. When participating in experiments, researchers are required only to prepare a WEB browser on any platform and are no longer required to memorize complex operations because all services are provided with a uniform user interface through a WEB browser. Furthermore, with the integration of these services, the required information and numerical data can be provided promptly by tracing HTML links that are created dynamically by server applications. [Copyright &y& Elsevier]

As part of study to establish in-vessel tritium reduction methods in carbon plasma facing fusion devices such as ITER, measurement of tritium removal characteristics of glow discharge methods, usually used for wall conditioning, have been applied to and examined in the vacuum vessel of JT-60U. Release rates of hydrogen isotopes (tritium and deuterium) as well as hydrocarbons from the JT-60U vacuum vessel induced by glow discharge cleaning (GDC) with He and H2 were measured. Release characteristics of hydrogen isotopes could be classified into three release processes, and each process was well described by a simple exponential decrease with time. It was found that H2 GDC showed the superior hydrogen isotope release characteristics to the He GDC, probably because of chemical processes, such as isotope exchanges assisted by the chemical sputtering process between discharged hydrogen and hydrogen isotopes in the plasma facing carbon tiles were enhanced by the H2 glow discharge. Based on the release kinetics observed in the present work, it is estimated that it takes several days to reduce tritium (deuterium) inventory within the removal depth by the H2 GDC in JT-60U to a half by continuous H2 GDC at 573 K. [Copyright &y& Elsevier]

• The data from the mass spectrometer was clustered by Euclidean distance to evaluate the presence or absence of vacuum leaks. • The leaked data and the no-leak data were classified into different clusters in the dendrogram, and this clustering was effective. • In this clustering, it is necessary to use data that is not affected by plasma discharge or fluctuation due to vacuum pumping over a long period. In order to generate high-performance plasma for future fusion power generation, it is desirable to keep high quality vacuum during experiment. Mass spectrometer is commonly used to monitor the vacuum quality and to record the amount of atoms and molecules in the vacuum vessel. Leak is the most serious accident to avoid that can nullify an experiment and even harm researchers. Detecting leaks are ever more important since it can be easily overlooked, e.g., when the deterioration in the vacuum degree is modest. This forces the researcher to carefully observe the vacuum and mass spectrometer data. This article presents a way to suggest potential leaks in the vacuum vessel by analyzing mass spectrometer data. This is done by utilizing the Euclidean distance between composition ratios at different times for the clustering using the daily composition ratio. We show that our cluster analysis is an effective way of separating these two cases, which results in a semi-automatic determination of leaks is more efficient than the current norm, which is to check many measures to find a small abnormality in the data manually. We plan further model improvements for long-term evaluation. [ABSTRACT FROM AUTHOR]

Abstract: On CPD [1] (Compact PWI experimental Device) experiment, information of electronic logbook and sequence status are distributed by Web services to prepare future experimental environment such as steady state operation and remote participation. Hence, all the researchers can acquire information with a Web browser installed on a personal computer if they are connected to the Internet. However, to carry a notebook computer all the time is a burden to researchers. Furthermore, the researchers may not be always connected to the Internet. Mobile phones are superior in portability compared to notebook computers, and are easy to connect with Internet through the wireless network of the telecom carriers. Moreover, since recent mobile phones have full browsing function, their affinities to the Web services are becoming high. On this account, Web services for mobile phones are developed to access experimental information. For sequence monitoring, a mobile application MIDlet that utilizes special functions of mobile phone such as sound and vibration is also developed to draw attentions of researchers to sequence status. [Copyright &y& Elsevier]

• Plasma current seems to flow in the open magnetic surface in the outside of the closed magnetic surface in the low-field region. • Considering the toroidal rotation, the equilibrium is fitted within nested magnetic surfaces by SU-EFIT. • Sum of contributions from toroidal rotation and pressure gradient to plasma current density is constant. In the EC-driven (8.2 GHz) steady-state plasma on QUEST, plasma current seems to flow in the open magnetic surface in the outside of the closed magnetic surface in the low-field region according to plasma current fitting method without taking equilibrium into account. In our previous work, plasma equilibrium solution was fitted assuming all plasma current is flowing in the inside of the Last Closed Flux Surface (LCFS). It was solved within isotropic pressure profile by EFIT code. Opposite-polarity current density region appeared in the high-field region. Here in this article, considering the toroidal rotation, the equilibrium is fitted within nested magnetic surfaces by SU-EFIT. Though the plasma magnetic axis shifts outward due to the centrifugal force, the opposite polarity current does not disappear in the high-field region. And relation between the toroidal rotation speed and the poloidal beta value will be discussed. [ABSTRACT FROM AUTHOR]

• New high voltage power supply (PS) for a klystron has been developed. • This PS is used for the electron cyclotron heating (ECH). • 8.56 GHz-RF with the PS will be a promising ECH system in QUEST spherical tokamak. • An IGBT and a reactor can shut down the PS quickly to protect the klystron. • AC switching has been set up for the overload and overheat of the PS. A high voltage DC power supply for the cathode of the 8.56 GHz CW klystron has been set up for electron cyclotron heating (ECH) in steady state tokamak operation on QUEST spherical tokamak. The power supply is equipped with an IGBT array and a reactor for fast shutoff of the voltage in 10 μs, where the influx of electric energy at the short circuit is limited to 5 J. AC switches also have been installed in the three-phase power lines. High voltage relays are useful to save electric energy consumption. Fast three-phase AC switching by IGBT-stack is applicable to reduce the electric load of the components of the klystron power supply. [ABSTRACT FROM AUTHOR]

In tokamaks, the temperature of the plasma-facing wall is an important parameter for achieving particle balance and therefore steady-state operation. QUEST, which is a middle-sized spherical tokamak, has hot walls that act as plasma-facing walls. They can be actively heated with sheath heaters and actively cooled with water. To control the wall temperature, heating and cooling systems have been developed. These systems adjust the power of the sheath heaters and the motor valves of the cooling system, respectively. The two systems communicate via Ethernet through UDP and control the hot-wall temperature cooperatively. The plasma control system (PCS) in QUEST has also been modified, especially with respect to gas fueling, in order to enable long-duration plasma sustainment. A feedback controller has been installed in the PCS, together with a mass flow controller, allowing Hα emission from the plasma which is used as a reference signal, to be well controlled. Plasma density calculations using a field-programmable gate array are proposed for the feedback control system. [ABSTRACT FROM AUTHOR]

The plasma control system (PCS) of QUEST is developed according to the progress of QUEST project. Since one of the critical goals of the project is to achieve the steady-state operation with high temperature vacuum vessel wall, the PCS is also required to have the capability to control the plasma for a long period. For the increase of the loads to processing power of the PCS, the PCS is decentralized with the use of reflective memories (RFMs). The PCS controls the plasma edge position with the real-time identification of plasma current and its position. This identification is done with not only flux loops but also hall sensors. The gas fueling method by piezo valve with monitoring the Hα signal filtered by a digital low-pass filter are proposed and suitable for the steady-state operation on QUEST. The present status and prospect of the PCS are presented with recent topics. [ABSTRACT FROM AUTHOR]

In the present RF-driven divertor plasma of QUEST, it has been observed that orbit-driven current flows in the open magnetic surfaces outside of the closed magnetic surfaces. To observe this phenomenon and the characteristics of the orbit-driven current, current density profiles have been calculated on two different equilibrium conditions. We calculated current density profiles from particle guiding center orbits both for the fundamental and the second harmonic resonances for the 8.2 GHz electron cyclotron current drive. From this calculation, hollow current density profiles have been obtained with significant characteristics on both conditions. Only positive current distribution has been observed in the open magnetic surfaces outside of the closed magnetic surfaces. [ABSTRACT FROM AUTHOR]

As QUEST has a small central solenoid (CS), a larger Ohmic discharge current has been obtained when the plasma shifts to the inboard side. This tendency restricts a divertor operation to the smaller plasma current regime. As the inner divertor coil has a smaller mutual inductance, it would be expected that its utilization seems to be better for easier plasma current ramp-up for a divertor operation. In this work, we made comparative studies on the plasma current ramp-up for two divertor coils. It is found that while the inner divertor coil with smaller mutual inductance needs a larger coil current, the outer divertor coil with larger mutual inductance needs a smaller coil current for divertor operation. Thus we have found that the plasma current ramp-up characteristics are almost similar for both configurations. We also propose a new fueling method for spherical tokamak (ST) using the coaxial helicity injection (CHI). The main plasma current would be generated at first, and then the CHI plasma current is created between bottom two electrode plates and merged into the main plasma current for fueling. [ABSTRACT FROM AUTHOR]

In the present RF-driven (ECCD) steady-state plasma on QUEST ( B t = 0.25 T, R = 0.68 m, a = 0.40 m), plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to plasma current fitting (PCF) method. We consider that the current in the open magnetic surface is due to orbit-driven current by high-energy particles in RF-driven plasma. So based on the analysis of current density profile based on the orbit-driven current, plasma equilibrium is to be calculated. We calculated high energy particles guiding center orbits as a contour plot of conserved variable in Hamiltonian formulation and considered particles initial position with different levels of energy and pitch angles that satisfy resonance condition. Then the profile of orbit-driven current is estimated by multiplying the particle density on the resonance surface and the velocity on the orbits. This analysis shows negative current near the magnetic axis and hollow current profile is expected even if pressure driven current is considered. Considering the hollow current profile shifted toward the low-field region, the equilibrium is fitted by J-EFIT coded by MATLAB. [ABSTRACT FROM AUTHOR]

The plasma control system (PCS) of QUEST was a centralized system, which lost its scalability because of the overload imposed on its central processing unit (CPU) of the PCS, making it impossible to add new functions. Thus, the PCS is distributed into a main workstation (WS) and subsystem (SS) with a reflective memory (RFM) in order to share data between these systems so as to mitigate the load on each system. As a result, 128 double-precision floating-point numbers (DBLs) can be transferred from the SS to the WS with a maximum latency of 250 μs. The WS and the SS each have quad-core CPUs, and tasks are executed in parallel. Although one of the four cores is intermittently occupied by up to 90% by this transaction, the occupation is normally 60%. A time correction procedure is used to map the recorded data sets on the WS and the SS to a common time base by referring to the time difference between two systems. [ABSTRACT FROM AUTHOR]

In the present RF-driven (ECCD) steady-state plasma on QUEST, the plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to the plasma current fitting (PCF) method. The current in the open magnetic surface is due to the orbit-driven current by high-energy particles in the RF-driven plasma. High-energy particles guiding center orbits are calculated as contour plots of conserved variables in the Hamiltonian formulation considering particles in the initial position with different energies and pitch angles satisfying the resonance condition. A negative current appears near the magnetic axis, and a hollow current profile is expected even if the pressure driven current is considered. The equilibrium is fitted within nested magnetic surfaces by J-EFIT coded by MATLAB using the hollow current profile shift toward the low-field region. Although the plasma boundary shape reflects the plasma current density profile, the equilibrium shape fitted by J-EFIT does not coincide with the orbit-driven current profile. However, introducing an extension of the current profile without assuming nested contours into the J-EFIT code appropriately fits the plasma shape with the hollow current profile to the measured magnetic data. [ABSTRACT FROM AUTHOR]

Abstract: A plasma control system to sustain divertor configurations is developed on QUEST (Q-shu university experiment with steady-state spherical tokamak). Magnetic fluxes are numerically integrated at 100kHz using FPGA (Field-Programmable Gate Array) modules and transferred to a main calculation loop at 4kHz. With these signals, plasma shapes are identified in real time at 2kHz under the assumption that the plasma current can be represented as one filament current. This calculation is done in another calculation loop in parallel by taking advantage of a multi-core processor of the plasma control system. The inside and outside plasma edge positions are controlled to their target positions using PID (proportional-integral-derivative) control loops. Whereas the outside edge position can not be controlled by the outer PF coil current, the inside edge position can be controlled by the inner PF coil current. [Copyright &y& Elsevier]

Abstract: CCS (Cauchy Condition Surface) method is a numerical approach to reproduce plasma shape, which has good precision in conventional tokamaks. In order to apply it in plasma shape reproduction of ST (Spherical Tokamak), the calculation precision of the CCS method in a spherical tokamak CPD (Compact PWI experimental Device) (B t =0.25T, R =0.3m, a =0.2m) has been analyzed. The precision was confirmed also in ST and decided to be applied to a spherical tokamak QUEST (B t =0.25T, R =0.68m, a =0.40m).. In present stage from the magnetic measurement, it is known that the eddy current effect is large in QUEST experiment, and there are no special magnetic measurements for eddy current now, so some proper model should be selected to evaluate the eddy current effect. The eddy current density by not only CS (Center Solenoid) coil but also plasma current is calculated using EDDYCAL (JAEA). The eddy current magnitudes are taken as unknown variables and solved together with plasma shape reconstruction in ohmic discharge and ECCD (Electron Cyclotron Current Drive) discharge. [Copyright &y& Elsevier]

Abstract: Particle retention and recycling in plasma fusion devices are generally associated with the diffusion of atomic hydrogen into the materials. The resulted permeation of atomic hydrogen is known as plasma driven permeation (PDP). This permeation may also be significant, even in the walls, which are not directly exposed to the plasma. Under similar conditions, the permeation flux (Γ perm) of hydrogen through a 30μm thick Ni membrane heated at 412–575K has been measured in the spherical tokamak QUEST. Γ perm is being measured during the scans of different operating parameters like RF power (P RF), chamber pressure (P chamber), discharge widths (τ dis) and vertical magnetic field (B Z). Simultaneously edge plasma density and spectral intensities of atomic (Balmer) lines and molecular (Fulcher) bands have been compared with the permeation measurements. A linear relationship has been established between the time integrated Γ perm i.e. permeation fluence (Q perm) and the time integrated H α intensity i.e. H α fluence (Q α). Q perm also shows a strong relationship with the edge plasma density and various spectral fluences. The obtained results are discussed for exploring the applicability of the permeation probes in measuring the atomic flux near the first walls. [ABSTRACT FROM AUTHOR]

Abstract: Experiments to study wall recycling behaviour have been performed in the small spherical tokamak compact plasma–wall interaction experimental device (CPD) from the viewpoint of global as well as local plasma wall interaction condition. Electron cyclotron resonance (ECR) plasma of typically ∼50 to 400ms duration is produced using ∼40 to 80kW RF power. In order to study the global wall recycling behaviour, pressure measurements are carried out just before and after the ECR plasma in the absence of any external pumping. The recycling behaviour is found to change from release to pumping beyond a certain level of pressure value which is again found to be a function of shot history. The real-time local wall behaviour is studied in similar RF plasma using a rotating tungsten limiter, actively coated with lithium. Measurement of Hα light intensity in front of the rotating surface has indicated a clear reduction (∼10%) in the steady-state hydrogen recycling with continuous Li gettering of several minutes. [Copyright &y& Elsevier]

Abstract: JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broader Approach Agreement, and it aims to contribute to the complement of the ITER experiments and to the DEMO reactor design by the study of steady-state high-beta plasma experiments. The conceptual design of the JT-60SA tokamak and the peripheral systems has been carried out in close collaboration of EU and Japan from middle of 2006, and the JT-60SA Conceptual Design Report which involves an analysis of plasma physics and engineering design is summarized in May 2007. This paper intends to present an overall view of the JT-60SA project, and the latest design status of the key engineering issues such as superconducting magnets, divertor target, plasma heating devices, and magnet power supplies. [Copyright &y& Elsevier]

Abstract: Superconducting tokamak TRIAM-1M has just ceased its operation in December 2005. However, in order to develop the systematic study on plasma–wall interaction in long duration discharges in the TRIAM-1M tokamak, spherical tokamak QUEST is under construction. To fulfill the mission of QUEST under a small group of the university in collaboration with other group, control and interlock system must be designed appropriately. Hence, a spherical tokamak CPD was constructed and the experiment is carried out for preparation of QUEST. The control and interlock system for QUEST is tested and developed during the experiment in the CPD. [Copyright &y& Elsevier]

Abstract: A discharge longer than 5h was successfully achieved on TRIAM-1M by fully non-inductive lower hybrid current drive (LHCD). The heat load distribution into the plasma facing components (PFCs) during the 5h discharge was investigated using calorimetric measurements, which estimated that the injected radio frequency (RF) power coincided with the total heat load amount to the PFCs. The power balance, including the portion of direct loss power of the fast electrons and the heat flux due to the charge exchange (CX) process, was also investigated. [Copyright &y& Elsevier]

Abstract: Magnetic sensorless control experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. Previously the horizontal position was calculated from the vertical field coil current and voltage without using signals of magnetic sensors like magnetic coils and flux loops placed near the plasma. The calculations are made focusing on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem of integrator of magnetic sensors. Two kinds of experiments were carried out, to keep the position constant and swing the position in a triangular waveform. [Copyright &y& Elsevier]

Abstract: Characterization of Tokamak exhaust gas, gas species and their concentrations, is important for the design of the tritium fuel processing system. The exhaust gas from JT-60U during the experimental campaign has been investigated with micro gas chromatography (MGC). Gas species, such as hydrogen isotopes, CH4, C2H2 +C2H4, C2H6 and CO2, in the exhaust gas from a plasma discharge could be identified successfully. In this campaign, the total ratio of hydrogen isotopes and impurity species in exhaust gas was estimated to be 39 to 1. From the comparison between both amount of gas species and some plasma parameters, there was a tendency observed indicating that the amount of hydrogen isotopes and carbon compounds increased with maximum electron density. This relation suggests that the amount of hydrogen isotopes and carbon compounds in the exhaust gas is increases with the extension of plasma performance. [1] Both contributed equally as first authors. [Copyright &y& Elsevier]