|makoto hasegawa||Last modified date：2019.12.25|
Assistant Professor / Division of Nuclear Fusion Dynamics / Research Institute for Applied Mechanics
|1.||Makoto Hasegawa, Introduction of experimental system on large plasma experimental device QUEST, さくらサイエンスプラン, 2017.12, ●Introduction of QUEST
・History of long duration discharges
●Exp. system including plasma control system on QUEST
●Usage of FPGA as software-defined technology
●Information sharing with Ethernet for coordinated operation
|2.||M. Hasegawa, K. Hanada, N. Yoshida, A. Kuzmin, H. Zushi, K. Nakamura, A. Fujisawa, H. Idei, Y. Nagashima, O. Watanabe, T. Onchi, K. Kuroda, H. Watanabe, K. Tokunaga, A. Higashijima, S. Kawasaki, and T. Nagata, Efforts toward Steady State Operation in Long Duration Discharges with the Control of Hot Wall Temperature on QUEST, 1st Asia-Pacific Conference on Plasma Physics, 2017.09, Efforts toward Steady State Operation in Long Duration Discharges
with the Control of Hot Wall Temperature on QUEST
M. Hasegawa1, K. Hanada1, N. Yoshida1, A. Kuzmin1, H. Zushi1, K. Nakamura1,
A. Fujisawa1, H. Idei1, Y. Nagashima1, O. Watanabe1, T. Onchi1, H. Watanabe1,
K. Tokunaga1, A. Higashijima1, S. Kawasaki1, and T. Nagata1
1 RIAM, Kyushu University, Japan
Achievement of steady state operation (SSO) of magnetic fusion devices is one of important issues for fusion research. Fully non-inductive plasma start-up and its maintenance up to 1h55min was successfully achieved on QUEST with a microwave of 8.2GHz, 40kW and well-controlled gas fueling and plasma-facing wall (PFW) temperature of 373K. The gas fueling is feedback controlled to keep constant in H signal, which can be an indicator of in-coming H flux to plasma facing materials (PFMs). On QUEST, the hot wall, which can be actively heated by electrical heater, was installed inside the vacuum vessel in 2014 autumn/winter (A/W) campaign, and the plasma can be sustained with high temperature PFW to investigate particle balance such as fuel recycling and wall pumping properties. Thermal insulators are installed between hot wall and vacuum vessel wall to keep the temperature of vacuum vessel wall below 423K for the protection of various diagnostics and plasma-heating devices. The function of active cooling of hot wall with cooling water channels will be installed in 2017 spring/summer (S/S) campaign.
The plasma-wall interaction (PWI) is an important subject when considering SSO, and is a wide-range issue because the matters such as material science and the plasma science are linked each other complicatedly. In these matters, especially, power balance and particle balance play important roles against SSO. The power balance in long duration discharges was sufficiently investigated in TRIAM-1M, which has the world record of plasma duration on tokamaks for more than 5h16min . During the long plasma discharge, all of the temperatures of PFMs are saturated and kept constant on TRIAM-1M. The power balance on QUEST is also investigated before 2014, in which the hot wall had been installed. Approximately 70%-90% of the injected power could be detected by calorimetric measurements of PFMs, and about half of the injected power was deposited on the vessel wall .
The total particle balance on QUEST is estimated experimentally . The time evolution of wall-pumping rate is evaluated as the difference between injected and evacuated H2 flux, which are derived from the flowmeter installed on gas fueling system and a quadrupole mass analyzer (QMS) installed on the bottom of the vessel, respectively. Absolute values of them are calibrated with consideration of the pressure and volume of gas fueling line and the relationship between flowmeter and QMS signal with the situation of no plasma. The wall-stored H can be obtained by time-integration of wall-pumping rate with setting the initial integrated value at zero. On the QUEST, the wall kept at higher temperature is rather active, and almost all stored H particles are released from the wall during the intervals of plasma discharges.
In the long duration discharges, the wall pumping occurs in the initial phase, and its rate gradually decrease. Finally, the wall-pumping rate becomes zero, and the wall saturation occurs. This tendency is likely to occur faster when its wall temperature is higher. To express this tendency, a wall model with hydrogen barrier (HB) which is formed around boundary between the deposition layer and the substrate was proposed . In this model, the time derivative of the number of H dissolved in wall (dHW/dt) is proportional to the square of HW, when the number of H trapped in defects (HT) can be negligible. The parabolic relation between dHW/dt and HW is clearly observed in low HW experimentally, and the given curves with this model is well-fitted to the experimental observation.
 H.Zushi, et al, Steady-state tokamak operation, ITB transition and sustainment and ECCD experiments in TRIAM-1M, Nuclear Fusion, 45 (2005) S142-S156
 K.Hanada, et al, Power Balance Estimation in Long Duration Discharge on QUEST, Plasma Science and Technology, 18 (2016) 1069-1075.
 K. Hanada, et al, Investigation of hydrogen recycling property and its control with hot wall in long duration discharges on QUEST, Nuclear Fusion, (2017) to be published.
 K. Hanada, et al, Particle balance in long duration RF driven plasmas on QUEST, Journal of Nuclear Materials, 463 (2015) 1084-1086..
|3.||Makoto Hasegawa, and QUEST group, Modifications of Plasma Control System and Central Control System for Integrated Control of Long Plasma Sustainment on QUEST, 11th IAEA Technical Meeting (TM) on the Control, Data acquisition and Remote Participation for Fusion Research, 2017.05, Modifications of Plasma Control System and Central Control System for Integrated Control of Long Plasma Sustainment on QUEST
Makoto Hasegawa1, Kazuo Nakamura1, Kazuaki Hanada1, Shoji Kawasaki1, Arseniy Kuzmin1, Hiroshi Idei1, Kazutoshi Tokunaga1, Yoshihiko Nagashima1, Takumi Onchi1, Kengoh Kuroda1,
Osamu Watanabe1, Aki Higashijima1, and Takahiro Nagata1
1Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka, Japan
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Achievement of steady state operation (SSO) is one of important issues for future magnetic fusion devices. The world record of plasma duration on tokamaks for more than 5h16min was achieved in TRIAM-1M , where particle balance and power balance are investigated. On QUEST, which is a middle sized spherical tokamak installed on the same place after the closing of TRIAM-1M experiments, these issues are also vigorously investigated, and the fully non-inductive plasma start-up and its maintenance up to 1h55min was successfully achieved  with a microwave of 8.2 GHz, 40 kW and well-controlled gas fueling and plasma facing wall (PFW) temperature of 373 K.
On QUEST, the hot wall which can be actively heated by electrical heaters was installed inside vacuum vessel in 2014, and the plasma discharge is sustained with high temperature PFW to investigate particle balance such as wall pumping properties. The function of active cooling for hot wall with the cooling water will be installed in 2017 spring. These controls of heating with electrical heaters and cooling with cooling water will be managed by the central control system and its peripheral subsystems with the coordination of them. On the other hand, the gas fueling during plasma discharge is feed-back controlled with referring to the signal level of H which is an indicator of in-coming H flux to PFWs. This control is managed by a Proportional-Integral-Differential (PID) control on the plasma control system using a mass-flow controller. The modifications and coordination of these control systems for long discharges are introduced.
1. H.Zushi, et al, “Steady-state tokamak operation, ITB transition and sustainment and ECCD experiments in TRIAM-1M”, Nuclear Fusion, 45 (2005) S142-S156.
2. K. Hanada, et al, Investigation of hydrogen recycling property and its control with hot wall in long duration discharges on QUEST, Nuclear Fusion, (2017) to be published..
|4.||Makoto Hasegawa, Integrated control system on spherical tokamaks, 4th A3 Foresight Summer School and Workshop on Spherical Torus (ST), 2016.08.|
|5.||Makoto Hasegawa, Kazuo Nakamura, hideki zushi, kazuaki hanada, Akihide Fujisawa, KAZUTOSHI TOKUNAGA, Hiroshi Idei, Yoshihiko Nagashima, Aki Higashijima, Shoji Kawasaki, Hisatoshi Nakashima, Aleksandrovich Arseniy Kuzmin, Takumi Onchi, Osamu Watanabe, Kishore Mishra, Real-time identification of plasma current and its position with hall sensors for long-pulse operation on QUEST, 8th IAEA Technical Meeting on "Steady State Operation of Magnetic Fusion Devices", 2015.05, [URL], For long-pulse operation, the control of plasma current and its position is important to manage the heat loads, particle flux, and so on. Thus, the plasma current and its position have to be identified correctly during a long plasma discharge in real time. The plasma current and its position are usually calculated with signals of a rogowski coil sensor, magnetic flux loop sensors, and magnetic pick-up coil sensors. In order to calculate with these signals, the time-integrations of raw signals with electrical circuits or numerical calculations are required. However, these time-integrations cause drift errors which become larger according to the duration of the plasma discharge. And, this disturbs the correct identification and the control of a long-pulse plasma discharge.
We propose the use of hall sensors for the long-pulse operation. A hall sensor does not require time-integration, and does not cause the drift error. On QUEST, several hall sensors are installed on the outside of the vacuum vessel. Although the quick behavior of plasma cannot be sensed with hall sensors because of eddy current effects, this enables high-accuracy measurements with the static environment of no plasma, no RF, and no vacuum. This also enables the repair or replacement of hall sensors without a vacuum purge. The plasma current and its position are identified in real time with these hall sensor signals. The plasma position and current are calculated by the evaluation of intensity ratios and intensities itself, respectively.
|6.||Makoto Hasegawa, Kazuo Nakamura, hideki zushi, kazuaki hanada, Akihide Fujisawa, KAZUTOSHI TOKUNAGA, Hiroshi Idei, Yoshihiko Nagashima, Shoji Kawasaki, Hisatoshi Nakashima, Aki Higashijima, Current Status and Prospect of Plasma Control System for Steady-state Operation on QUEST, 10th IAEA Technical Meeting on Control, Data Acquisition and Remote Participation for Fusion Research, 2015.04, Plasma control system (PCS) on QUEST has been developed for the achievement of the steady-state sustainment of tokamak plasma. QUEST is a spherical tokamak , on which high temperature all metal vessel wall up to 500 K is planned for the steady-state operation under unity recycling ratio. Achievement of steady-state operation in tokamak plasma is one of a key issue to realize cost-effective fusion power plants. In the aim of this, many kind of controls are required such as plasma position and its shape control, particle balance control, and heat load control. Current status and prospect of PCS for steady-state operation on QUEST are described.
For the control of plasma position and its shape, these parameters have to be identified in real time and steadily. Though magnetic sensors of rogowski coils, pick-up coils, and flux loops are usually used for this identification, these sensors are not suitable for the long time measurements because drift error induced by time integration occurs. On QUEST, in addition to these sensors, hall sensors are used, which are suitable for the long time measurement because of no drift errors. Furthermore, hall sensors can be expected to have an ease of maintenance and high accuracy because these are located on the outside of vacuum vessel wall where is less noisy environment compared to the inside one. The plasma current and position are calculated with just hall sensor signals, assuming the plasma as a filament current located on the inside of vacuum vessel. In this procedure, the plasma position and plasma current are evaluated with ratios and intensities of hall sensor signals, respectively. In addition to this, plasma shape is also evaluated in real time with a shape identification method . These procedures are applicable to the control of plasma position and its shape for steady-state operation.
For the control of particle balance, a fueling feed-back control is implemented, which is referring Ha signals instead of plasma density. The fueling gas is puffed when an actual Ha signal intensity is lower the target intensity, and the actual signal gradually comes close to the target signal with a setting of the pulse prohibited duration. The actual Ha signal is well controlled with this method on over 10 minutes plasma discharge. Other several approaches such as a distributing system for steady-state operation will be discussed.
1. K. Hanada, K. Sato, H. Zushi, K. Nakamura, M. Sakamoto, H. Idei, et al., “Steady-State Operation Scenario and the First Experimental Result on QUEST”, Plasma and Fusion Research, 5, S1007 (2010).
2. M. Hasegawa, K. Nakamura, H. Zushi, K. Hanada, a. Fujisawa, K. Matsuoka, et al., “Development of plasma control system for divertor configuration on QUEST”, Fusion Engineering and Design, 88, 1074–1077 (2013).
|7.||Makoto Hasegawa, Kazuo Nakamura, hideki zushi, kazuaki hanada, Akihide Fujisawa, Osamu Mitarai, KAZUTOSHI TOKUNAGA, Hiroshi Idei, Yoshihiko Nagashima, Shoji Kawasaki, Hisatoshi Nakashima, Aki Higashijima, Development of high performance control system by decentralization with reflective memory on QUEST, 28th Symposium on Fusion Technology, 2014.10, Plasma control systems for tokamak plasmas are required to make control signals in real-time with simultaneously acquiring various data and calculating meaningful physical quantities. Since the physical quantities and the control signals have relationship with each other, a centralized control system is principally desirable for the grasp of these parameters. However, the computational loads on the CPU of plasma control workstation (WS) become too large to build a highly integrated control system, because it makes difficult to execute in real-time. In actual, the CPU utilization of the WS for the spherical tokamak QUEST becomes almost full.
We propose to develop a decentralized control system. In this system, each control system has a reflective memory connected to each other with optical fibers, and shares various data via reflective memory. The good point of this system is to increase the CPU resource. Furthermore, the electrical insulation is ensured spontaneously. On the other hand, the synchronization accuracy between each system may become worse.
The GE cPCI-5565PIORC of National Instruments Corporation is used as the reflective memory, which has 256 Mbytes memory and 170Mbyte/sec transfer rate. The most popular data type to share is double-precision real type (DBL) which needs 8 bytes to represent. The actual data read or write time is measured. Especially, within the period of 4 kHz which is the period of WS, more than 1000 to 2000 DBLs can be read or write. This means about 50 Mbytes/sec transfer rate for the one directional data sharing. For the bidirectional data sharing, each system has to repeat the read-write procedure. This would take more time. In the presentation, we will introduce the actual implementation of the reflective memory to the decentralized control system and its performance..
|8.||Makoto Hasegawa, Kazuo Nakamura, Hideki Zushi, Kazuaki Hanada, Akihide Fujisawa, Keisuke Matsuoka, Hiroshi Idei, Yoshihiko Nagashima, Kazutoshi Tokunaga, Shoji Kawasaki, Hisatoshi Nakashima, Aki Higashijima, Development of plasma control system for steady state operation on QUEST, 9th Asia Plasma and Fusion Association Conference, 2013.11, [URL], A long time plasma sustainment is an important issue for the future nuclear fusion plasma. In QUEST (Q-shu university experiment with steady-state spherical tokamak), a steady state operation is also one of project objectives. Thus, the long time identification and its control of the plasma position and its shape are important for the steady state operation. However, the long time identification is difficult, as long as the integrated magnetic signals such as magnetic fluxes or magnetic fields are used because the integration errors, namely, drift errors occur and prevent the accurate identification.
The WS is composed of PXI systems of the National Instruments Corporation, which contains a controller module (2.26 GHz Intel Core 2 Quad processor, memory: 2 GBytes) based on a real-time operating system, one DIO module (16-channel digital input and output), and six FPGA modules (eight-channel analog input and output in each module). In the WS, the several tasks can be performed in parallel because a multi-quad-core processor is used in the controller module. One task is for the control of a DIO module and FPGA modules. Another task, referred to as a main loop, is for the calculation of control signals by the acquired data. These two tasks are performed at 4 kHz. In addition, the real-time equilibrium calculation and the plasma image analysis are executed in parallel on other cores, respectively. The calculation period of the image analysis will be several seconds. That is sufficient to correct the drifts of magnetic fluxes. In this presentation, the development status of this control system will be introduced.
|9.||長谷川 真, Development of real-time equilibrium control system on QUEST, Workshop on QUEST and Related ST RF Startup and Sustainment Plasma Research, 2013.02.|
|10.||Makoto Hasegawa, Kazuo Nakamura, hideki zushi, kazuaki hanada, Akihide Fujisawa, Keisuke Matsuoka, Osamu Mitarai, Hiroshi Idei, Yoshihiko Nagashima, Kazutoshi Tokunaga, Shoji Kawasaki, Hisatoshi Nakashima, Aki Higashijima, Development of plasma control system for divertor configuration on QUEST, 27th Symposium on Fusion Technology (SOFT 2012), 2012.09, A plasma control system in order to sustain divertor configurations is developed on QUEST (Q-shu university experiment with steady-state spherical tokamak). Magnetic fluxes are numerically integrated by 100 kHz frequency with usage of FPGA (Field-Programmable Gate Array) modules, and transferred to a main calculation loop with 4 kHz. With these signals, plasma shapes are identified in real time with 2 kHz frequency 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 position controls are tested using PID (proportional–integral–derivative) control loops for target positions. Whereas the outside edge position can not be controlled by outer PF coil current, the inside edge position can be controlled by inner PF coil current..|
|11.||Makoto Hasegawa, Kazuo Nakamura, KAZUTOSHI TOKUNAGA, hideki zushi, kazuaki hanada, Akihide Fujisawa, Hiroshi Idei, Shoji Kawasaki, Hisatoshi Nakashima, Aki Higashijima, Development of Control System for Divertor Configuration on QUEST, 16th International Workshop on Spherical Torus (ISTW2011), 2011.09, In a similar way to other spherical tokamaks, the achievement of steady state operation with divertor configuration is one of important issues for the QUEST project. The control system for this has been developed in the QUEST. The identification of plasma position and its configuration is required for the control. One of the methods adopted in this control system is to adjust plasma shape parameters such as elongation and triangularity directly so that the calculated magnetic signals become the same values as measured ones. Although this method cannot be adopted if the plasma shape is complicated, one can expect that the time to calculate become short because there is no need to calculate values such as flux values at inside area of vacuum vessel but just installed positions of magnetic sensors. This calculation method has been installed into the control system of the QUEST which is composed of 4 CPU cores and Real Time-OS and operates main control loop with 4 kHz period. And, this calculation with 22 flux loop signals is finished within 1msec by using parallel processing technology.
The horizontal and vertical plasma positions are controlled by active coils called HCUL coils and PF26 coils, respectively with simple PID control method. The current of PF26 coils changes not only vertical magnetic field but n-index. Furthermore, the n-index also varies gradually in the process that plasma configuration changes from limiter configuration to divertor configuration. This change affects vertical control, and the appropriate PID gain values differ by each magnetic configuration. For this, the mechanism to regulate each gain values automatically according to the magnetic configuration will be also installed into the control system.
|12.||Reports of utilization of SNET on QUEST experiment, [URL].|
|13.||Divertor design study using SOLDOR divertor simulation code on QUEST.|