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Recent progress of HT-7 superconducting Tokamak PFPS control system
Fusion Engineering and Design 60 (2002) 421– 426 www.elsevier.com/locate/fusengdes
Recent progress of HT-7 superconducting Tokamak PFPS control system L. Wang *, J.R. Luo, F.X. Yin, B. Shen, P.J. Qin, H.Z. Wang, P. Fu Institute of Plasma Physics, Academia Sinica, P.O. Box 1126, Hefei, Anhui 230031, People’s Republic of China
Abstract The recent progress of the poloidal field control system of the HT-7 super-conducting Tokamak is presented. The distributed control system (DCS) was implemented on the poloidal field power supplies (PFPS) in 1998. The successful development of the new PFPS trigger system allowed the control cycle of the PF power supplies to reach 1–1.2 ms. The response speed of the control system was improved by ten times with reference to the former system. Secondly, based on software de-coupling of plasma equilibrium parameters, a multivariable feedback control scheme was implemented. Thirdly, the efficiency of HT-7 experiment operation was greatly improved by realizing the discharge experiment database of HT-7. © 2002 Elsevier Science B.V. All rights reserved. Keywords: Tokamak; Poloidal field; Multivariable feedback; Thyristor trigger; Discharge experiment database
1. Introduction HT-7 was reconstructed from the former T-7 Tokamak originally built in Kurchatov Institute in Moscow as shown in Fig. 1. Its main objectives are to investigate reactor relevant issues, such as advanced operation modes and plasma-wall interaction in the near steady state condition. The former PFPS control system of HT-7 was built at the beginning of 1990s. The former PFPS control system adopted * Corresponding author. Tel.: + 86-551-5591-333; fax: + 86-551-5591-310. E-mail address: [email protected] (L. Wang).
various independent control systems each based on single variable control schemes, with low response velocity and long delay time. As HT-7 is a Tokamak equipped with iron core and outer and inner copper shell screens cooled by liquid nitrogen, there is a strong coupling between the vertical and the ohmic heating magnetic fields [1]. Due to the shielding effect of the copper shells and the magnetic coupling, an accurate control of the plasma position at high plasma current was very difficult and the repeatability of the plasma discharges resulted is poor. Hence, to improve the time response of the active feedback control and to solve the coupling problems between vertical and ohmic heating magnetic fields, a new PFPS control system was built in 1998.
0920-3796/02/$ - see front matter © 2002 Elsevier Science B.V. All rights reserved. PII: S 0 9 2 0 - 3 7 9 6 ( 0 2 ) 0 0 0 4 2 - X
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L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
2. DCS system configuration
2.1. PFPS control components and work theory The poloidal field power supply is one of the most important parts of the plasma discharge circuit. The PFPS system generates the ohmicheating field, the vertical field, the biased field, and the horizontal and vertical correcting field. A 120 MW AC flywheel generator provides the energy for the ohmic-heating and vertical fields through one transformer and a set of thyristor converters (64 MW). The power supply for the biased and for the horizontal and vertical-correcting fields is provided by a normally controlled rectifier system. It controls more than 200 switches and the interlocks in the switching sequence for the total of five types of fields above mentioned, and the thyristor converter angle. The PFPS control computer produces control commands and presets the experimental parameters. Time order signals from one programmable logic controller (PLC) are used to control the switches of PFPS and the time process of the feedback loops. The PFPS system controls the PF magnetic field by modifying the PFPS output voltage.
Fig. 1. Schematic diagram of HT-7 Tokamak, 1, 2, 3, 4, 5, 6, 7, 8, 9 poloidal field coil; ten plasma current and vessel; 11 iron core; 12 inner copper shell; 13 outer copper shell.
2.2.3. Operability Due to the limited accessibility by operators to the experiment during the operating periods, a friendly Human Machine Interface (HMI) is required for the PFPS control system in order to control and monitor remotely the experiment. 2.2.4. Reliability and safety As the stored energy in the plasma and in the poloidal coils is extremely high, unexpected events such as plasma disruptions and exceeding of current limits in the coils may cause fatal damage to the components of vacuum vessel (VV) and magnetic coils. 2.3. PFPS control system configuration
2.2. Design of the PFPS control system The characteristic configurations of the superconduction Tokamak dictate that control system design meet the following requirements:
As shown in Fig. 2, the poloidal control system is based on a distributed design. The control system is composed of one host computer at an upper hierarchical level, and of a set of computers at a lower level. These include: the multivariable
2.2.1. Real-time The response time of the PFPS control and its delay should be kept as small as possible in order to control rapidly the plasma dynamics, especially the horizontal position of the plasma [2]. 2.2.2. Software de-coupling By analyzing the coupling relationship between the vertical field, the ohmic-heating field and the plasma current, the mathematical model to handle the power supply voltage, plasma current and plasma position was established.
Fig. 2. Hardware configuration in the PFPS control system.
L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
Fig. 3. Synchronous signal waveform interference.
controller, the PLC, the power supply trigger (PS trigger) and the vertical position feedback controller (VP controller), the experiment data display unit, the parameter presetting unit, a local network and some additional devices [3].
3. PFPS control system main progress
423
trigger pluses are issued. There are 12 firing commands of the of thyristor converter in a PFPS cycle. So at present the control system cycle and its delay time are less than 1–1.2 and 2.2 ms, respectively [4], a good result if compared with the values of the cycle time of the former system (12–14 ms). The control system response time has been greatly improved. This PFPS trigger plays also an important role on ohmic-heating field and vertical field access time control. It was shown that the trigger has an important effect to ensure plasma discharge waveform repeatability during HT-7 experiment in 2000.
3.2. Multi6ariable feedback control system Due to the iron core transformer, the inner and outer copper shells, and the coupling between the ohmic-heating and the vertical coils, all electromagnetic parameters in HT-7 PFPS are nonlinear and time-variant. The presence of the copper shells introduces a delay for the externally generated magnetic field to penetrate into the plasma region [5].
3.1. PS trigger system The power supply trigger consists of the synchronous signal pretreatment circuit, 200 MHz IPC-based equipped with four timer/count digital cards and one pulse —modulating amplifier. Due to the operation of the large power thyristor converters, the voltage provided by the freewheel generator (the frequency of which is in the range 70–90 Hz) is seriously distorted, as shown in Fig. 3. For this reason, it is necessary to pre-treat the synchronous signal to satisfy the control requirements. A type of high-pass and low-pass filter with high Q value was successfully used in the electronic circuit for the synchronous signal treatment. The IPC acquires the value of the h angle through the parallel interface from the multivariable feedback controller while the CPU detects in real time the PS voltage period from the AC flywheel generator. Trigger pulse timer values are sent to the timer card and successively 12 phase
3.2.1. Multi6ariable feedback controller theory The following approximations have been introduced. The Tokamak iron core transformer is considered as infinitely long. The equivalent model of the copper shells is expressed by four coils. By using the linearized approach described above and the equivalent model of the copper shells, the plasma multivariable feedback control model can be obtained. The equation of the poloidal field circuit and of the plasma equilibrium can be expressed as follows:
n
d[L] d[I] V [I]+ [L] + = [V] dt dt I Based on the above mentioned theoretical analysis and related electromagnetic calculation, the transformation function can be written as follows:
L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
424
n
Æ 7.1×103(s − 154)(s + 9.4) à Y (s +61.4)(s + 7.1) =à 5.0× 102(s −685.8)(s +4.1) Ip Ã È (s+ 61.4)(s + 7.1)(s +4.5)
n
4.1 × 102(s+ 790)(s + 5.1)(s +12) Ç Ã U1 (s + 61.4)(s + 4.5)(s +7.1) Ã g1 3 5.6 × 10 (s+7.4)(s + 6.0) Ã U0 (s +61.4)(s +7.1)(s + 4.5) É
The multivariable feedback control system design is based on this with some simplification.
3.2.2. Multi6ariable feedback controller According to the measured plasma current, plasma horizontal position, a set of others magnetic signals coming from the electromagnetic measurements and the preset signals, the results of the de-coupling calculation are transferred to the PS trigger controller as a angle every 1 ms. The plasma current and the horizontal position are controlled by feedback control. 3.2.3. The 6ertical position feedback controller The error values between the collected real time signals and the preset reference is computed. The CPU of the vertical position feedback controller implements a PID regulator and produces the control reference of the feedback voltage for the vertical position. A D/A converter provides the output analog vertical voltage reference which modifies the horizontal field coil current in order to control the vertical position.
Fig. 4. The plasma horizontal position experiment.
The performance of the control system, validated by the experiment discharges, is as follows.
3.2.4. Accuracy 9 5 KA (Ip= 150 KA flat-top phase); 91 cm (horizontal plasma position control); 91 cm (vertical plasma position control).
3.2.5. Stability The feedback control system showed a good interference rejection, as demonstrated in many experiment sessions where injected ICRH power and LHCD current drive were used extensively [6].
3.2.6. De-coupling The de-coupling results in some experiments to test the ability of the multivariable feedback control system are shown in Fig. 4.
Fig. 5. The waveform of six different pulses.
L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
425
Table 1 Poloidal field power supply protection setup Protection level
Limit values
C level
IPlasmaB30 kA and Rectifier current reverse Tdischarge pulse\70 ms Iohmic field \2.8 kA Ivertical field\5.0 kA Iohmic field\3.5 kA Rectifier current reverse. Lock rectifier trigger pulse. AC generator excitation current reverse Ivertical field\8.0 kA Iohmic field\5.0 kA Rectifier current reverse. Lock rectifier trigger pulse. AC generator excitation switch off Ivertical field\9.0 kA Rectifier current reverse. Lock rectifier trigger pulse. AC generator excitation switch off Ivertical field\9.5 kA Power supply main circuit switch off
B level
A level AA level
Protection active
3.2.7. Repeatability Fig. 5 illustrates six different discharge waveforms, which show very good repeatability. 3.3. Friendly HMI and discharge experiment database Windows 98 and DOS were used as operating systems for the HT-7 PFPS control system. Control programs were developed in VISUAL BASIC and C language. PFPS control system communication method is as follows: the huge amounts of experiment data are transferred by file sharing in the network. The control commands are transferred by a serial line. Remote and local controls are based on network Win-sock and standard Sockets ActiveX program of the VISUAL BASIC language. In order to make efficient use of HT-7 experimental data; (for example capacitor charge voltage, discharge trigger pulse length, and plasma current and position by operator setup) it was saved on one large disk in the host computer as a file named by the shot-number. This data file will be loaded when the same experiment conditions will be proposed. With the introduction of the HT-7 experiment database in 2000, considerable HT-7 operation time was saved. It is now easy to load plasma discharge waveforms for pulses with the plasma current in the range from
Protection device Multivariable controller
Multivariable controller, PLC
Multivariable controller PLC Multivariable controller PLC. Over-time relay Over-current relay
200 to 150 KA simply by recalling the pulses by a shot number. The host computer manages all the lower level computers of the PFPS control system, including control commands, communication between network and computers, preset of experiment parameters, their reception, saving and displaying. The PFPS control system interfaces have the following features. Monitoring of the operation status of the real time control system. Presetting of the experimental condition parameters. Presetting of the plasma discharge waveform and display of the result. Easy exchange of the experimental model. Use of the discharge database for the experiment. Simple modification of remote and local control. Simple and convenient operation.
3.4. PFPS control system protection function There are four protection levels in the PFPS control system as shown in Table 1. The HT-7 Tokamak components of the VV and the magnetic coils have been protected successfully since the construction of the PFPS protection system was accomplished in 1999.
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L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
4. Conclusions
References
The PFPS control system realizes equilibrium control of the plasma by improving the response time of the control system. Multivariable realtime control technique, advanced network communication and digital technologies were applied to HT-7 Tokamak. With the recent progress of the PFPS control system, high performance and long pulse operation in HT-7 Tokamak were obtained. The new PFPS control system provided good reproducibility of the HT-7 experiment pulses. Many new physics experiment results were achieved in 2000.
[1] J.K. Xie, et al., HT-7 Super-conducting Tokamak and its Operation, Asipp/157, 1999. [2] T. Kimura, et al., JT-60U Plasma Control System, Fusion Technol. 32 (1997) 1. [3] L. Wang, et al., Upgraded poloidal field control system for HT-7 Tokamak, Plasma Sci. Technol. 2 (3) (2000) 311. [4] L. Wang, et al., IPC-based Thyristors Trigger System Following Variable Frequencies, ASIPP/CH9804, in Chinese. [5] P. Fu, et al., Multivariable feedback control and analysis of horizontal position and current of plasma in HT-7 supper-conducting Tokamak, Acta Phys. Sin. 48 (4) (1999) 685 in Chinese. [6] J.R. Luo, et al., Feedback control system on HT-7 Tokamak, Plasma Sci. Technol. 2 (1) (2000) 1.
Recent progress of HT-7 superconducting Tokamak PFPS control system L. Wang *, J.R. Luo, F.X. Yin, B. Shen, P.J. Qin, H.Z. Wang, P. Fu Institute of Plasma Physics, Academia Sinica, P.O. Box 1126, Hefei, Anhui 230031, People’s Republic of China
Abstract The recent progress of the poloidal field control system of the HT-7 super-conducting Tokamak is presented. The distributed control system (DCS) was implemented on the poloidal field power supplies (PFPS) in 1998. The successful development of the new PFPS trigger system allowed the control cycle of the PF power supplies to reach 1–1.2 ms. The response speed of the control system was improved by ten times with reference to the former system. Secondly, based on software de-coupling of plasma equilibrium parameters, a multivariable feedback control scheme was implemented. Thirdly, the efficiency of HT-7 experiment operation was greatly improved by realizing the discharge experiment database of HT-7. © 2002 Elsevier Science B.V. All rights reserved. Keywords: Tokamak; Poloidal field; Multivariable feedback; Thyristor trigger; Discharge experiment database
1. Introduction HT-7 was reconstructed from the former T-7 Tokamak originally built in Kurchatov Institute in Moscow as shown in Fig. 1. Its main objectives are to investigate reactor relevant issues, such as advanced operation modes and plasma-wall interaction in the near steady state condition. The former PFPS control system of HT-7 was built at the beginning of 1990s. The former PFPS control system adopted * Corresponding author. Tel.: + 86-551-5591-333; fax: + 86-551-5591-310. E-mail address: [email protected] (L. Wang).
various independent control systems each based on single variable control schemes, with low response velocity and long delay time. As HT-7 is a Tokamak equipped with iron core and outer and inner copper shell screens cooled by liquid nitrogen, there is a strong coupling between the vertical and the ohmic heating magnetic fields [1]. Due to the shielding effect of the copper shells and the magnetic coupling, an accurate control of the plasma position at high plasma current was very difficult and the repeatability of the plasma discharges resulted is poor. Hence, to improve the time response of the active feedback control and to solve the coupling problems between vertical and ohmic heating magnetic fields, a new PFPS control system was built in 1998.
0920-3796/02/$ - see front matter © 2002 Elsevier Science B.V. All rights reserved. PII: S 0 9 2 0 - 3 7 9 6 ( 0 2 ) 0 0 0 4 2 - X
422
L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
2. DCS system configuration
2.1. PFPS control components and work theory The poloidal field power supply is one of the most important parts of the plasma discharge circuit. The PFPS system generates the ohmicheating field, the vertical field, the biased field, and the horizontal and vertical correcting field. A 120 MW AC flywheel generator provides the energy for the ohmic-heating and vertical fields through one transformer and a set of thyristor converters (64 MW). The power supply for the biased and for the horizontal and vertical-correcting fields is provided by a normally controlled rectifier system. It controls more than 200 switches and the interlocks in the switching sequence for the total of five types of fields above mentioned, and the thyristor converter angle. The PFPS control computer produces control commands and presets the experimental parameters. Time order signals from one programmable logic controller (PLC) are used to control the switches of PFPS and the time process of the feedback loops. The PFPS system controls the PF magnetic field by modifying the PFPS output voltage.
Fig. 1. Schematic diagram of HT-7 Tokamak, 1, 2, 3, 4, 5, 6, 7, 8, 9 poloidal field coil; ten plasma current and vessel; 11 iron core; 12 inner copper shell; 13 outer copper shell.
2.2.3. Operability Due to the limited accessibility by operators to the experiment during the operating periods, a friendly Human Machine Interface (HMI) is required for the PFPS control system in order to control and monitor remotely the experiment. 2.2.4. Reliability and safety As the stored energy in the plasma and in the poloidal coils is extremely high, unexpected events such as plasma disruptions and exceeding of current limits in the coils may cause fatal damage to the components of vacuum vessel (VV) and magnetic coils. 2.3. PFPS control system configuration
2.2. Design of the PFPS control system The characteristic configurations of the superconduction Tokamak dictate that control system design meet the following requirements:
As shown in Fig. 2, the poloidal control system is based on a distributed design. The control system is composed of one host computer at an upper hierarchical level, and of a set of computers at a lower level. These include: the multivariable
2.2.1. Real-time The response time of the PFPS control and its delay should be kept as small as possible in order to control rapidly the plasma dynamics, especially the horizontal position of the plasma [2]. 2.2.2. Software de-coupling By analyzing the coupling relationship between the vertical field, the ohmic-heating field and the plasma current, the mathematical model to handle the power supply voltage, plasma current and plasma position was established.
Fig. 2. Hardware configuration in the PFPS control system.
L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
Fig. 3. Synchronous signal waveform interference.
controller, the PLC, the power supply trigger (PS trigger) and the vertical position feedback controller (VP controller), the experiment data display unit, the parameter presetting unit, a local network and some additional devices [3].
3. PFPS control system main progress
423
trigger pluses are issued. There are 12 firing commands of the of thyristor converter in a PFPS cycle. So at present the control system cycle and its delay time are less than 1–1.2 and 2.2 ms, respectively [4], a good result if compared with the values of the cycle time of the former system (12–14 ms). The control system response time has been greatly improved. This PFPS trigger plays also an important role on ohmic-heating field and vertical field access time control. It was shown that the trigger has an important effect to ensure plasma discharge waveform repeatability during HT-7 experiment in 2000.
3.2. Multi6ariable feedback control system Due to the iron core transformer, the inner and outer copper shells, and the coupling between the ohmic-heating and the vertical coils, all electromagnetic parameters in HT-7 PFPS are nonlinear and time-variant. The presence of the copper shells introduces a delay for the externally generated magnetic field to penetrate into the plasma region [5].
3.1. PS trigger system The power supply trigger consists of the synchronous signal pretreatment circuit, 200 MHz IPC-based equipped with four timer/count digital cards and one pulse —modulating amplifier. Due to the operation of the large power thyristor converters, the voltage provided by the freewheel generator (the frequency of which is in the range 70–90 Hz) is seriously distorted, as shown in Fig. 3. For this reason, it is necessary to pre-treat the synchronous signal to satisfy the control requirements. A type of high-pass and low-pass filter with high Q value was successfully used in the electronic circuit for the synchronous signal treatment. The IPC acquires the value of the h angle through the parallel interface from the multivariable feedback controller while the CPU detects in real time the PS voltage period from the AC flywheel generator. Trigger pulse timer values are sent to the timer card and successively 12 phase
3.2.1. Multi6ariable feedback controller theory The following approximations have been introduced. The Tokamak iron core transformer is considered as infinitely long. The equivalent model of the copper shells is expressed by four coils. By using the linearized approach described above and the equivalent model of the copper shells, the plasma multivariable feedback control model can be obtained. The equation of the poloidal field circuit and of the plasma equilibrium can be expressed as follows:
n
d[L] d[I] V [I]+ [L] + = [V] dt dt I Based on the above mentioned theoretical analysis and related electromagnetic calculation, the transformation function can be written as follows:
L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
424
n
Æ 7.1×103(s − 154)(s + 9.4) à Y (s +61.4)(s + 7.1) =à 5.0× 102(s −685.8)(s +4.1) Ip Ã È (s+ 61.4)(s + 7.1)(s +4.5)
n
4.1 × 102(s+ 790)(s + 5.1)(s +12) Ç Ã U1 (s + 61.4)(s + 4.5)(s +7.1) Ã g1 3 5.6 × 10 (s+7.4)(s + 6.0) Ã U0 (s +61.4)(s +7.1)(s + 4.5) É
The multivariable feedback control system design is based on this with some simplification.
3.2.2. Multi6ariable feedback controller According to the measured plasma current, plasma horizontal position, a set of others magnetic signals coming from the electromagnetic measurements and the preset signals, the results of the de-coupling calculation are transferred to the PS trigger controller as a angle every 1 ms. The plasma current and the horizontal position are controlled by feedback control. 3.2.3. The 6ertical position feedback controller The error values between the collected real time signals and the preset reference is computed. The CPU of the vertical position feedback controller implements a PID regulator and produces the control reference of the feedback voltage for the vertical position. A D/A converter provides the output analog vertical voltage reference which modifies the horizontal field coil current in order to control the vertical position.
Fig. 4. The plasma horizontal position experiment.
The performance of the control system, validated by the experiment discharges, is as follows.
3.2.4. Accuracy 9 5 KA (Ip= 150 KA flat-top phase); 91 cm (horizontal plasma position control); 91 cm (vertical plasma position control).
3.2.5. Stability The feedback control system showed a good interference rejection, as demonstrated in many experiment sessions where injected ICRH power and LHCD current drive were used extensively [6].
3.2.6. De-coupling The de-coupling results in some experiments to test the ability of the multivariable feedback control system are shown in Fig. 4.
Fig. 5. The waveform of six different pulses.
L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
425
Table 1 Poloidal field power supply protection setup Protection level
Limit values
C level
IPlasmaB30 kA and Rectifier current reverse Tdischarge pulse\70 ms Iohmic field \2.8 kA Ivertical field\5.0 kA Iohmic field\3.5 kA Rectifier current reverse. Lock rectifier trigger pulse. AC generator excitation current reverse Ivertical field\8.0 kA Iohmic field\5.0 kA Rectifier current reverse. Lock rectifier trigger pulse. AC generator excitation switch off Ivertical field\9.0 kA Rectifier current reverse. Lock rectifier trigger pulse. AC generator excitation switch off Ivertical field\9.5 kA Power supply main circuit switch off
B level
A level AA level
Protection active
3.2.7. Repeatability Fig. 5 illustrates six different discharge waveforms, which show very good repeatability. 3.3. Friendly HMI and discharge experiment database Windows 98 and DOS were used as operating systems for the HT-7 PFPS control system. Control programs were developed in VISUAL BASIC and C language. PFPS control system communication method is as follows: the huge amounts of experiment data are transferred by file sharing in the network. The control commands are transferred by a serial line. Remote and local controls are based on network Win-sock and standard Sockets ActiveX program of the VISUAL BASIC language. In order to make efficient use of HT-7 experimental data; (for example capacitor charge voltage, discharge trigger pulse length, and plasma current and position by operator setup) it was saved on one large disk in the host computer as a file named by the shot-number. This data file will be loaded when the same experiment conditions will be proposed. With the introduction of the HT-7 experiment database in 2000, considerable HT-7 operation time was saved. It is now easy to load plasma discharge waveforms for pulses with the plasma current in the range from
Protection device Multivariable controller
Multivariable controller, PLC
Multivariable controller PLC Multivariable controller PLC. Over-time relay Over-current relay
200 to 150 KA simply by recalling the pulses by a shot number. The host computer manages all the lower level computers of the PFPS control system, including control commands, communication between network and computers, preset of experiment parameters, their reception, saving and displaying. The PFPS control system interfaces have the following features. Monitoring of the operation status of the real time control system. Presetting of the experimental condition parameters. Presetting of the plasma discharge waveform and display of the result. Easy exchange of the experimental model. Use of the discharge database for the experiment. Simple modification of remote and local control. Simple and convenient operation.
3.4. PFPS control system protection function There are four protection levels in the PFPS control system as shown in Table 1. The HT-7 Tokamak components of the VV and the magnetic coils have been protected successfully since the construction of the PFPS protection system was accomplished in 1999.
426
L. Wang et al. / Fusion Engineering and Design 60 (2002) 421–426
4. Conclusions
References
The PFPS control system realizes equilibrium control of the plasma by improving the response time of the control system. Multivariable realtime control technique, advanced network communication and digital technologies were applied to HT-7 Tokamak. With the recent progress of the PFPS control system, high performance and long pulse operation in HT-7 Tokamak were obtained. The new PFPS control system provided good reproducibility of the HT-7 experiment pulses. Many new physics experiment results were achieved in 2000.
[1] J.K. Xie, et al., HT-7 Super-conducting Tokamak and its Operation, Asipp/157, 1999. [2] T. Kimura, et al., JT-60U Plasma Control System, Fusion Technol. 32 (1997) 1. [3] L. Wang, et al., Upgraded poloidal field control system for HT-7 Tokamak, Plasma Sci. Technol. 2 (3) (2000) 311. [4] L. Wang, et al., IPC-based Thyristors Trigger System Following Variable Frequencies, ASIPP/CH9804, in Chinese. [5] P. Fu, et al., Multivariable feedback control and analysis of horizontal position and current of plasma in HT-7 supper-conducting Tokamak, Acta Phys. Sin. 48 (4) (1999) 685 in Chinese. [6] J.R. Luo, et al., Feedback control system on HT-7 Tokamak, Plasma Sci. Technol. 2 (1) (2000) 1.