|makoto hasegawa||Last modified date：2021.06.16|
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Reseacher Profiling Tool Kyushu University Pure
High energy Material science and technology studies, Plasma and Material Engineering Science, Department of Advanced Energy Engineering Science, Interdisciplinary Graduate School of Engineering Sciences Kyushu University .
Advanced Fusion Research Center, Research Institue for Applied Mechanics .
Doctor of engineering
Country of degree conferring institution (Overseas)
Field of Specialization
data-processing system, plasma heating by radio frequency, plasma control
ORCID(Open Researcher and Contributor ID)
Study on the plasma breakdown induced by micro & milli waves and static electric fields for tokamak devices, and Study and Development on data acquisition system, plasma control, and control systems for large plant.
- Development of Integrated Control Method for stedy-state operation on Tokamak Plasma
keyword : Integrated Control, Tokamak, steady-state operation, large database
- A plasma shape identification and its real-time control on steady-state operation of ST tokamak
keyword : plasma shape, real-time, control, steady state operation
- The development of integrated experimental system for large-scale plasma experiments
keyword : remote participation, data acquisition, supervision & control system, large-scale plasma experiment, IoT (Internet of Things)
2003.10Data acqusition system is required to manupulate various type, and large size data according to enlargement of experimental scale. In plasma experiments, experimenters have to decide parameters of next discharge swiftly refferencing previous various and large data. The purpose of this thema is to develop data acquisition system to manupulate and administrate various and large data in order to offer data swiftly and to encourage experimenters to execute experiment effectively..
- R&D for SNET-based Steady-state Remote Data Acquisition Method in CPD Experiments
|1.||Makoto HASEGAWA, Kazuaki HANADA, Naoaki YOSHIDA, Hiroshi IDEI, Takeshi IDO, Yoshihiko NAGASHIMA, Ryuya IKEZOE, Takumi ONCHI, Kengoh KURODA, Shoji KAWASAKI, Aki HIGASHIJIMA, Takahiro NAGATA, Shun SHIMABUKURO and Kazuo NAKAMURA, Extension of Operation Region for Steady State Operation on
QUEST by Integrated Control with Hot Walls, Plasma and Fusion Research, 10.1585/pfr.16.2402034, 16, 2402034, 2402034-1-2402034-8, 2021.01, [URL], The controllability of particle supply during long-term discharge in a high-temperature environment was investigated at the Q-shu University Experiment with steady state spherical tokamak (QUEST). QUEST has a high-temperature wall capable of active heating and cooling as a plasma-facing wall. With this hot wall, a temperature rise test was conducted with 673K as the target temperature. It was confirmed that the hot wall could maintain the temperature above 600 K. Feedback control of particle fueling was conducted to control Hα
emission, which is closely related to influx to the wall. Using this particle fueling control and setting the hot wall temperature to 473 K, it was possible to obtain a discharge of more than 6 h. In this discharge, the fueling rate of particles decreased with time, and finally became zero, losing the particle fueling controllability. However, as soon as the cooling water started to flow through the hot wall, particles could be supplied again, and controllability was restored. Thus, indicating that temperature control of the plasma first wall is important even in the high temperature environment of 473K to control particle retention of the wall..
|2.||Makoto Hasegawa, Kazuaki Hanada, Hiroshi Idei, Shoji Kawasaki, Takahiro Nagata, Ryuya Ikezoe, Takumi Onchi, Kengoh Kuroda, Aki Higashijima , Predictive maintenance and safety operation by device integration on the QUEST large experimental device, Heliyon, accepted, 2020.06, As technology has improved in recent years, it has become possible to create new valuable functions by combining various devices and sensors in a network. This concept is referred to as the Internet of Things (IoT), and predictive maintenance is a new valuable function associated with the IoT. In large-scale experimental facilities with many researchers, it is not desirable that experiments cannot be performed due to sudden failure of equipment. For this reason, it is important to predict the failure in advance based on the measurement results of sensors and to perform repairs in a planned manner. On the Q-shu University experiment with steady-state spherical tokamak (QUEST) large experimental device, it is necessary to drive a large current of 50 kA, and the diagnosis of its power line deterioration is well performed as predictive maintenance through the evaluation of its contact resistances of several micro ohms order on the network. In addition, as an example of the IoT, mechanisms to assist safe operation, such as a sound alarm system and an entrance management system, are built by sharing experimental information between devices via the network..|
|3.||Makoto Hasegawa, Kazuo Nakamura, Kazuaki Hanada, Shoji Kawasaki, Arseniy Kuzmin, Hiroshi Idei, Kazutoshi Tokunaga, Yoshihiko Nagashima, Takumi Onchi, Kengoh Kuroda, Osamu Watanabe, Aki Higashijima, Takahiro Nagata, Modification of plasma control system and hot-wall temperature control system for long-duration plasma sustainment in QUEST, Fusion Engineering and Design, 10.1016/j.fusengdes.2018.02.069, 129, 202-206, 2018.04, In tokamaks, the temperature of the plasma-facing wall is an important parameter for achieving particle balanceand therefore steady-state operation. QUEST, which is a middle-sized spherical tokamak, has hot walls that act asplasma-facing walls. They can be actively heated with sheath heaters and actively cooled with water. To controlthe wall temperature, heating and cooling systems have been developed. These systems adjust the power of thesheath heaters and the motor valves of the cooling system, respectively. The two systems communicate viaEthernet through UDP and control the hot-wall temperature cooperatively. The plasma control system (PCS) inQUEST has also been modified, especially with respect to gas fueling, in order to enable long-duration plasmasustainment. A feedback controller has been installed in the PCS, together with a mass flow controller, allowingHα emission from the plasma which is used as a reference signal, to be well controlled. Plasma density calculationsusing a field-programmable gate array are proposed for the feedback control system..|
|4.||Makoto Hasegawa, Kazuo Nakamura, Hideki Zushi, K.Hanada, Fujisawa Akihide, K. Tokunaga, Hiroshi Idei, Nagashima Yoshihiko, Shoji Kawasaki, Hisatoshi Nakashima, Aki Higashijima, Current status and prospect of plasma control system for steady-stateoperation on QUEST, Fusion Engineering and Design, http://dx.doi.org/10.1016/j.fusengdes.2016.04.016, 112, 699-702, Vol. 112, pp. 699-702, 2016.12, The plasma control system (PCS) of QUEST is developed according to the progress of QUEST project. Sinceone of the critical goals of the project is to achieve the steady-state operation with high temperaturevacuum 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 ofreflective memories (RFMs). The PCS controls the plasma edge position with the real-time identificationof 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 H signal filtered by a digital low-pass filterare proposed and suitable for the steady-state operation on QUEST. The present status and prospect ofthe PCS are presented with recent topics..|
|5.||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 a high-performance control system bydecentralization with reflective memory on QUEST, Fusion Engineering and Design, 96-97, 629-632, 96-97 (2015) 629-632, 2015.07, The plasma control system (PCS) of QUEST was a centralized system, which lost its scalability because ofthe overload imposed on its central processing unit (CPU) of the PCS, making it impossible to add newfunctions. Thus, the PCS is distributed into a main workstation (WS) and subsystem (SS) with a reflectivememory (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 WSwith a maximum latency of 250 s. The WS and the SS each have quad-core CPUs, and tasks are executedin parallel. Although one of the four cores is intermittently occupied by up to 90% by this transaction, theoccupation is normally 60%. A time correction procedure is used to map the recorded data sets on theWS and the SS to a common time base by referring to the time difference between two systems..|
|6.||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 a plasma control system for steady-state operation on QUEST, JOURNAL OF THE KOREAN PHYSICAL SOCIETY, 10.3938/jkps.65.1191, 65, 8, 1191-1195, 2014.10, A drift error correction technique with machine vision and a real-time equilibrium calculation code have been developed on the QUEST (Q-shu university experiment with the steady-state spherical tokamak) for steady-state operation. The drift error caused by the long time-integration of magnetic raw signals has to be removed. With a captured image of the plasma’s cross section, the plasma’s position is identified by use of image filters. The measured magnetic flux values are corrected to the calculated flux values estimated by using this plasma position. The correction with the captured image work as expected in the preliminary result using a flashlight instead of a plasma..|
|7.||Hasegawa, M.; Nakamura, K.; Zushi, H.; Hanada, K.; Fujisawa, A.; Matsuoka, K.; Mitarai, O.; Idei, H.; Nagashima, Y.; Tokunaga, K.; Kawasaki, S.; Nakashima, H.; Higashijima, A., Development of plasma control system for divertor configuration on QUEST, FUSION ENGINEERING AND DESIGN, 10.1016/j.fusengdes.2013.03.035, 88.0, 6.0, 1074.0-1077.0, 0.0, 2013.10, 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 100 kHz using FPGA (Field-Programmable Gate Array) modules and transferred to a main calculation loop at 4 kHz. With these signals, plasma shapes are identified in real time at 2 kHz 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. (C) 2013 Elsevier B.V. All rights reserved..|
|8.||Makoto Hasegawa, Kazuo Nakamura, Kazutoshi Tokunaga, Hideki Zushi, Kazuaki Hanada, Akihide Fujisawa, Hiroshi Idei, Shoji Kawasaki, Hisatoshi Nakashima, Aki Higashijima, A Plasma Shape Identification with Magnetic Analysis for the Real-time Control on QUEST, IEEJ Transactions on Fundamentals and Materials, 132.0, 7.0, 477.0-484.0, 2012.07, In order to identify plasma shape, there is a way to represent the plasma current profile with several parameters, and adjust these parameters with least-square technique in order for calculated magnetic values to accord with measured ones. Here, the plasma shape parameters such as minor radius, elongation, and triangularity are chosen as the fitting parameters to represent plasma shape more directly, and the applicability to the control of the plasma shape are described by evaluating its calculation time. In order to find minimum of an objective function with least-square technique, two methods are compared, namely a linear approximation method and a downhill simplex method. While high accuracies of the measured magnetic signals are required, the good reproducibility is obtained, and the plasma shape identification can be done within several milliseconds in both methods..|
|9.||Hasegawa, M.; Higashijima, A.; Nakamura, K.; Hanada, K.; Sato, K. N.; Sakamoto, M.; Idei, H.; Kawasaki, S.; Nakashima, H., A WEB-based integrated data processing system for the TRIAM-1M, FUSION ENGINEERING AND DESIGN, 10.1016/j.fusengdes.2007.12.012, 83.0, 4.0, 588.0-593.0, 2008.05, 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. (C) 2007 Elsevier B.V. All rights reserved..|
|10.||Hasegawa, M.; Nakamura, K.; Higashijima, A.; Kawasaki, S.; Nakashima, H.; Sato, K. N.; Zushi, H.; Hanada, K.; Sakamoto, A.; Idei, H., High accessible experimental information on CPD experiment, FUSION ENGINEERING AND DESIGN, 10.1016/j.fusengdes.2007.10.014, 83.0, 2.0, 402.0-405.0, 2008.04, On CPD  (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. (C) 2007 Elsevier B.V. All rights reserved..|
|11.||Hasegawa, Makoto; Hanada, Kazuaki; Sato, Kohnosuke; Nakamura, Kazuo; Zushi, Hideki; Sakamoto, Mizuki; Idei, Hiroshi; Kawasaki, Shoji; Nakashima, Hisatoshi; Higashijima, Aki, Initial plasma production by Townsend avalanche breakdown on QUEST tokamak, JAPANESE JOURNAL OF APPLIED PHYSICS, 10.1143/JJAP.47.287, 47.0, 1.0, 287.0-292.0, 2008.01, On tokamak devices, an induction electric field induced by poloidal field (PF) coils plays a role to produce initial plasma. On a DIII-D tokamak, the required induction electric field for plasma breakdown agrees well with theoretical predictions based on the Townsend avalanche theory. According to the Townsend avalanche theory, the minimum induction electric field for plasma breakdown depends on neutral gas pressure and connection length. For stable plasma breakdown, a sufficiently large induction electric field is required. However, in the case of spherical tokamaks without electric insulation in the toroidal direction, the effect of eddy currents flowing in the toroidal direction should be considered in evaluating a feasible induction electric field because this effect suppresses the generation of an induction electric field. On a QUEST spherical tokamak, the possibility of Townsend avalanche breakdown is studied by evaluating the connection length and achievable induction electric field. The connection length is greater than 100 in in the case where a null point is set to be R = 0.56 in with a CS coil current of 2.0 kA and a PF26 coil current of 0.4 kA. Moreover, the induction electric field is about 1.5 V at this point including the effect of eddy currents. With these values, the initial plasma production by the induction electric field is sufficiently possible on QUEST..|
|12.||Hasegawa, Makoto; Hanada, Kazuaki; Sato, Kohnosuke; Nakamura, Kazuo; Zushi, Hideki; Sakamoto, Mizuki; Idei, Hiroshi; Kawasaki, Shoji; Nakashima, Hisatoshi; Higashijima, Aki, Model of inductive plasma production assisted by radio-frequency wave in tokamaks, JOURNAL OF NUCLEAR MATERIALS, 10.1143/JPSJ.76.084501, 76.0, 8.0, 0.0-0.0, 084501, 2007.08, For initial plasma production, an induction electric field generated by applying voltage to a poloidal field (PF) coil system is used to produce a Townsend avalanche breakdown. When the avalanche margins are small, as for the International Thermonuclear Experimental Reactor (ITER) in which the induction electric field is about 0.3 V/m, the assistance of radio-frequency waves (RF) is provided to reduce the induction electric field required for reliable breakdown. However, the conditions of RF-assisted breakdown are not clear. Here, the effects of both RF and induction electric field on the RF-assisted breakdown are evaluated considering the electron loss. When traveling loss is the dominant loss, a simple model of an extended Townsend avalanche is proposed. In this model, the induction electric field required for RF-assisted breakdown can be decreased to half that required for induction breakdown..|
|13.||Makoto HASEGAWA, Kazuaki HANADA, Kohnosuke SATO, Kazuo NAKAMURA, Hideki ZUSHI, Mizuki SAKAMOTO, Hiroshi IDEI, Shoji KAWASAKI, Hisatoshi NAKASHIMA and Aki HIGASHIJIMA , Townsend Avalanche Breakdown Assisted by Radio Frequency Wave in Tokamaks, Plasma and Fusion Research, 2.0, RC, Article No. 007, 2007.03.|
|14.||Nakamura, K; Ji, ZS; Shen, B; Qin, PJ; Itoh, S; Hanada, K; Sakamoto, M; Jotaki, E; Hasegawa, M; Iyomasa, A; Kawasaki, S; Nakashima, H, Magnetic sensorless sensing of plasma position in the superconducting tokamak HT-7, PLASMA SCIENCE & TECHNOLOGY, 10.1088/1009-0630/6/5/006, 6.0, 5.0, 2459.0-2462.0, 0.0, 2004.10, Magnetic sensorless sensing experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The horizontal position is calculated from the vertical field coil current and voltage without using signals of magnetic probes placed nearby a plasma. The calculations are focused on the ripple frequency component of the power supply. There is no drift problem with the time integration of magnetic probe signals. The error of the derived plasma position is lower than 2% of the plasma minor radius..|
|15.||M. Hasegawa, K. Hanada, S. Itoh, K. Nakamura, H. Zushi, M. Sakamoto, E. Jotaki, S. V. Kulkarni, A. Iyomasa, S. Kawasaki, H. Nakashima, K. Nagasaki, Plasma Experiments Using a New 170GHz EC System and a Simple Model for Plasma Production, JOURNAL OF NUCLEAR MATERIALS, Vol.80, No.1 pp.53-58, 2004.01.|
|16.||Nakamura, K; Ji, ZS; Shun, B; Qin, PQ; Itoh, S; Hanada, K; Sakamoto, M; Jotaki, E; Hasegawa, M; Iyomasa, A; Kawasaki, S; Nakashima, H, Sensorless sensing of plasma horizontal position on HT-7, FUSION ENGINEERING AND DESIGN, 10.1016/S0920-3796(03)00308-9, #
Works, Software and Database
|8.||Monitoring system of experiment hall entering personnel
Many personnel participate in the QUEST experiment. Because the high RF power is used and hard x-ray is generated during plasma sustain, the experiment operator is required to confirm that nobody stays in the experimental hall before the start of plasma discharge. Thus, the instrument which enables personnel to switch on the lamps of his name in the experimental hall is installed at the entrance of this hall, and this can communicate with the central control system to send the information of entering personnel. If any one stays in the hall, the central control system stops the experiment sequence according to the information of this system. This contributes for the safety operation of the QUEST.
|10.||Central Control System of QUEST.|
|11.||We built and released the site which enables simple numerical calcuration such as magnetic configuration and electron trajectory on QUEST device.
|12.||We built up and released the community site to all the collaborators for interactively sharing information such as technical, conference presentation, and research papers to promote studies and experiments of QUEST project.
|1.||M. Hasegawa, K. Hanada, N. Yoshida, H. Idei, T. Ido, Y. Nagashima, R. Ikezoe, T. Onchi, K. Kuroda, S. Kawasaki, A. Higashi, T. Nagata, S. Shimabukuro, K. Nakamura, Extension of Operation Area for Steady State Operation on QUEST by Integrated Control with Hot Walls, 9th International Workshop, RIAM 2021, 2021.01, [URL], Steady state operation of magnetically confined plasmas has been studied on the Q-shu University Experiment with steady state spherical tokamak (QUEST). In future power plants, to increase power generation efficiency by heat exchange, its operation temperature will be around 773 K, and it is desirable to realize steady state operation. However, the uncontrollable increase in plasma density, which tends to occur with high temperature walls, prevents the realization of them. They are definitely relating to plasma-wall interaction (PWI) and particle balance, and more detailed studies are required.
QUEST has all-metal plasma facing walls (PFWs) and its temperature can be controlled with resistive heaters and cooling water. The typical operating temperature range on the PFWs is between room temperature and less than 773 K. To control the temperature on the PFWs, hot walls and radiation shields have been installed inside the vacuum vessel on QUEST. The radiation shields prevent the outflow of heat so that the hot wall, namely PFW, can be maintained at high temperatures.
Recently, tokamak plasma discharges longer than 6 h were achieved with full non-inductive electron cyclotron current drive (ECCD) using a RF source of 8.2 GHz. The input RF power was in the range of 20 kW, and the wall temperature was higher than 423 K. In these shots, particle fueling were properly feedback-controlled to keep plasma influx constant by referring Ha emission. However, in typical long discharge, the fueling gradually decreased and finally stopped due to a so-called “wall saturation”. This indicates the unknown particle sources is significantly affecting in the particle balance. On the other hand, when the cooling water for the hot walls was circulated under these situations, the particle fueling was observed to restart spontaneously. This also indicates that the integrated control including particle fueling and temperature of PFWs is important and this realization contributes expansion of operation area toward steady state operations..
|2.||M. Hasegawa, K. Hanada, N. Yoshida, H. Idei, T. Ido, Y. Nagashima, R. Ikezoe, T. Onchi, K. Kuroda, S. Kawasaki, A. Higashi, T. Nagata, S. Shimabukuro, K. Nakamura
, Extension of Operation Area for Steady State Operation on QUEST by Integrated Control with Hot Walls
, The 29th International Toki Conference on Plasma and Fusion Research (ITC29), 2021.01, [URL], Steady state operation of magnetically confined plasmas has been studied on the Q-shu University Experiment with steady state spherical tokamak (QUEST). In future power plants, to increase power generation efficiency by heat exchange, its operation temperature will be around 773 K, and it is desirable to realize steady state operation. However, the uncontrollable increase in plasma density, which tends to occur with high temperature walls, prevents the realization of them. They are definitely relating to plasma-wall interaction (PWI) and particle balance, and more detailed studies are required.
QUEST has all-metal plasma facing walls (PFWs) and its temperature can be controlled with resistive heaters and cooling water. The typical operating temperature range on the PFWs is between room temperature and less than 773 K. To control the temperature on the PFWs, hot walls and radiation shields have been installed inside the vacuum vessel on QUEST. The radiation shields prevent the outflow of heat so that the hot wall, namely PFW, can be maintained at high temperatures.
Recently, tokamak plasma discharges longer than 6 h were achieved with full non-inductive electron cyclotron current drive (ECCD) using a RF source of 8.2 GHz. The input RF power was in the range of 20 kW, and the wall temperature was higher than 423 K. In these shots, particle fueling were properly feedback-controlled to keep plasma influx constant by referring Ha emission. However, in typical long discharge, the fueling gradually decreased and finally stopped due to a so-called “wall saturation”. This indicates the unknown particle sources is significantly affecting in the particle balance. On the other hand, when the cooling water for the hot walls was circulated under these situations, the particle fueling was observed to restart spontaneously. This also indicates that the integrated control including particle fueling and temperature of PFWs is important and this realization contributes expansion of operation area toward steady state operations.
|3.||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
|4.||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..
|5.||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..
|6.||Makoto Hasegawa, Integrated control system on spherical tokamaks, 4th A3 Foresight Summer School and Workshop on Spherical Torus (ST), 2016.08.|
|7.||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.
|8.||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).
|9.||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..
|10.||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.
|11.||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.|
|12.||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..|
|13.||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.
|14.||Reports of utilization of SNET on QUEST experiment, [URL].|
|15.||Divertor design study using SOLDOR divertor simulation code on QUEST.|