Small Spherical Tokamak Research Articles

Safety Analysis of Metallic Tokamak-I Against Electromagnetic and Structural Loads

During tokamak operation, the structural integrity of the vacuum vessel (VV) of Metallic Tokamak-I (MT-I), a small spherical tokamak, was evaluated. This evaluation involved simulating real experimental data of electromagnetic (EM) and structural loads using the ANSYS platform. Internal heat generation, induced currents, and inertial and pressure loads in the VV were analyzed to determine their effects on the VV. This analysis was conducted on a 180-deg sector model over a 10-ms-event period. To create multiple checkpoint events, the plasma current was assumed to be formed at variable positions of the VV, hence inducing variable current for each event. The events are divided into four cases based on the radial and vertical displacements of plasma. The response of the VV structure was calculated using coupling of EM and structural modules of ANSYS. It is observed from the numerical results that the maximum stress on the VV is in a safe range and that the temperature rise on the vessel can be reduced by natural convection only if the event is ended in 10 ms. A prolonged event can result in permanent deformation in the VV structure. A disruption event on the limiter region is also studied.

Performance of MT-I spherical tokamak with upgraded power supplies

This paper describes the upgradation of power supply systems for the improved performance of a small spherical tokamak (MT-I) with an aspect ratio of 1.67 in operation at the Pakistan Tokamak Plasma Research Institute (PTPRI), Islamabad. Three mutually independent power supplies are designed to generate current pulses of desired shape and magnitude in central solenoid (CS), toroidal field (TF) and vertical field (VF) coil systems. The main parts of a power supply are capacitor banks as energy storage devices, ignitrons and thyristors as fast switches for controlled power switching. A double capacitor bank topology with enhanced second capacitor bank voltage is found to be suitable for CS to improve the plasma startup conditions. Simulations are performed in MATLAB Simulink for confirmation of the designed power supply system. The Rogowski coil and current transformers are employed to measure plasma current and the coil’s current waveforms during the experiment in the presence of hydrogen as working gas. An ocean spectrometer and a photron high-speed camera (SA-8) are used for optical measurements and fast plasma imaging. An improvement in startup conditions is observed, enhancing the plasma current duration and light emission intensity recorded in the high-speed camera masked with a Hα filter.

Modeling of high-field-side high-density regime in the Globus-M2 tokamak

Formation of a high-field-side high-density (HFSHD) regime and the role of the high-field-side (HFS) poloidal electric field in the scrape-off layer of the spherical tokamak Globus-M2 are analyzed using SOLPS-ITER edge plasma simulations. The dependence of the HFS poloidal electric field sign and, consequently, radial drift fluxes on the discharge density is discussed. It is demonstrated that the HFS poloidal electric field is the key element in the formation of a HFSHD regime in the Globus-M2 tokamak as in ASDEX Upgrade. It is demonstrated that the physics of HFSHD formation in a small spherical tokamak is similar to that suggested by Kaveeva et al (2009 36th EPS Conf. on Plasma Physics) and is in line with experimental observations and modeling, performed later on ASDEX Upgrade.

Optimization of Wall-Conditioning Techniques on the MT-II Tokamak

The MT-II is a small spherical tokamak that is currently under construction at the Pakistan Tokamak Plasma Research Institute. Wall conditioning of the MT-II vacuum vessel (VV) is an essential step to achieve a good quality vacuum for plasma experiments. This study presents an overview of the wall-conditioning techniques implemented on the MT-II VV, including baking and glow discharge cleaning (GDC). Prior to wall conditioning, the system is checked via a helium leak test machine and residual gas analyzer (RGA) to identify and remove leaks. The VV walls are baked at ~180 °C to get rapid desorption of water vapors and other impurities. After the baking process, the partial pressure of most of the carbon- and oxygen-containing impurities is reduced. In particular, the partial pressure of water vapors is reduced by 93%. Consequently, the total leak and outgassing rate is significantly reduced. To further improve the vacuum condition in the vessel, hydrogen GDC is carried out. The fill hydrogen pressure and anode voltage are optimized to get a stable glow discharge. The RGA scan shows that GDC reduces the partial pressure of H2O, O2, and CO2 by 57%, 63%, and 51%, respectively. The results signify that baking and GDC are effective techniques for wall conditioning of the MT-II VV.

Design and Development of 0.3-T Toroidal Field Coil System for Small-Sized MT-II Spherical Tokamak

The achievement of a high toroidal magnetic field in a small spherical tokamak is challenging because of the small bore area in the central cylinder of the vacuum vessel. In this paper, we present a toroidal field coil of 0.3 T at the center of the MT-II tokamak. It has been designed, developed, and tested for installation at Pakistan Tokamak Plasma Research Institute (PTPRI). The coil is made of highly pure oxygen-free copper. It has a cross-sectional area of 10 × 15 mm2 (150 mm2) for the flow of an approximately 20-kA current to produce a 0.33 T toroidal magnetic field at the center of the tokamak. Mechanical support for the central stack of the inner legs is provided by a twisted grooved nylon cylinder to control the torque and attractive forces. The repulsive force density between the joints of the outer and inner legs is balanced by nuts and bolts along with an insulated ring of Teflon and an isolated metallic clamp from both ends. This compressive force also reduces connection resistance. The simulated currents and magnetic field are confirmed from the experimental results as well.

Conceptual Neutronics Study of a Hybrid Fusion Neutron Source FNS-C with Aqueous Blanket

The modern challenges of nuclear energy are the replenishment of dwindling reserves of nuclear fuel and the development of a closed nuclear fuel cycle while complying with strict radiation safety requirements. A fusion neutron source has unique capabilities to solve these problems. The preliminary results of a neutronic analysis of the FNS-C fusion-fission hybrid neutron source with a thorium-uranium aqueous blanket by the Monte Carlo method computer simulation, using the MCNP-4 code with the ENDF/B-VII cross-section library, gives satisfactory results for the study of the possibility of creating a compact source of fusion neutrons based on a small spherical tokamak for commercial use. The obtained results show that the FNS-C hybrid blanket generates enough tritium to fully ensure the uninterrupted operation of the FNS-C throughout the year. The reproduction coefficient of 233U is 1.027 at a consumption of 1304 kg/year of the fissile material in the aqueous blanket containing 232Th enriched to 1.47% 233U. The FNS-C is operated with an effective neutron multiplication factor keff ~ 0.99 with reactivity ρ = –0.006249 in the presence of delayed neutrons, which corresponds to the safest state of the core of thermal neutron fission reactors. The thermal power of the FNS-C at keff ~ 0.99 is ~3 GW, which is comparable to the thermal power of fission reactors. This indicates the potential possibility of creating a safe thorium-uranium breeder power reactor based on a fusion neutron source. The results of the study were obtained for the simplified approximate geometrical FNS-C model. To confirm the preliminary results, it is necessary to develop a more accurate calculation model of the FNS-C machine.

Vacuum Chamber of the MEPhIST-1 Tokamak

One of the most important structural elements of a tokamak is a vacuum discharge chamber, which satisfies specific requirements concerning the induction method of plasma generation and, accordingly, the presence of alternating magnetic fields additionally acting on the chamber. Moreover, the chamber should not only be able to create and confine a current plasma but also be able to diagnose its parameters through appropriate pipes. The small spherical tokamak being created at MEPhI for both educational and research purposes has an extensive program of such studies; therefore, the discharge chamber should not only allow fast evacuation and withstand alternating magnetic fields but also be convenient for measuring its parameters, using various diagnostics, as well as allow quick replacement of internal elements and introduction of additional power into the chamber. The paper describes the design of the vacuum chamber of the MEPhIST-1 tokamak, which meets these requirements, and presents calculations to justify the choice of material and the thickness of the characteristic elements of the chamber, and also presents the test results of the finished chamber.

Development of magnetic diagnostics for Glass Spherical Tokamak (GLAST)

GLAST is a small spherical tokamak with an insulating vacuum vessel of pyrex. It has major and minor radii of 20 and 10 cm, respectively. Magnetic measurements play a fundamental role in plasma breakdown and start-up studies. This paper presents the development, fabrication, calibration, and installation of magnetic diagnostics on GLAST III Tokamak. Magnetic diagnostics consist of the flux loops, magnetic field pickup sensors, and Rogowski coil. The signals obtained from various coils are integrated with passive analog integrators. The experimental results obtained from these magnetic diagnostics are presented and discussed.

Integrated data acquisition, storage and retrieval for glass spherical tokamak (GLAST)

GLAST is a small spherical tokamak with an insulating vacuum vessel of pyrex. The data acquisition (DAQ) system of GLAST is in evolving process. It integrates different data sources as an assortment of data acquisition hardware and software. This paper presents the evolution of the hardware setup and software implementations of the DAQ system for GLAST tokamak. It includes all the sub-systems required for the successful acquisition of a signal from the transducer to the DAQ hardware and then to the database. The hardware for the DAQ system comprises of National Instruments USB based DAQ cards, high resolution Tektronix oscilloscopes and several onsite developed signal conditioning (SC) modules. The software layer for front and back end handling is developed using LabVIEW and MATLAB for acquisition, storage, data retrieval, and post-processing purposes. The main features of this system include system configuration, shot implementation, data saving and data sharing between the connected users. In addition to the evolution of the DAQ system, future directions about the control system of GLAST tokamak experiments have also been highlighted.

The Project of MEPhIST Tokamak

The concept of the small spherical tokamak project is described, the main objectives of the project are to increase competencies at the University in the training of personnel in the field of physics and technologies of controlled thermonuclear fusion, as well as attracting young people to this area. The implementation of the project also implies giving impetus and new opportunities for improving the methods of plasma diagnostics developed at the University, studies on the plasma surface interactions and computer simulation of processes in the plasma and on the plasma facing surfaces. The installation has a large radius of 25 cm, an aspect ratio of less than 2, and a vertical elongation of ∼ 3, which allows, in principle, for small sizes and costs of the installation, taking into account the subsequent increase in the toroidal field to ∼ 2 T, to carry out important studies for progress in plasma performance in tokamaks. Namely, research on physics of plasma confinement, current drive with RF power and plasma interaction with materials. The project includes a phased implementation with a multiple increase in the magnetic field and plasma current, as well as the possibility of quick and convenient access to the internal elements of the discharge chamber and the simultaneous use of a large number of diagnostics.

Concept of DT Fuel Cycle for a Fusion Neutron Source

A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 1018 neutrons per second is under designe in Russia. Such a FNS is of interest for many applications: (i) basic and applied research (neutron scattering, etc); (ii) testing the structural materials for fusion reactors; (iii) control of sub-critical nuclear systems and (iv) nuclear waste processing (including transmutation of minor actinides). This paper describes of fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ~0.5 and ~0.3m, magnetic field ~1.5 T, heating power less than 15MW and plasma current 1–2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems.

Estimation of Electron Temperature on Glass Spherical Tokamak (GLAST)

Glass Spherical Tokamak (GLAST) is a small spherical tokamak indigenously developed in Pakistan with an insulating vacuum vessel. A commercially available 2.45 GHz magnetron is used as pre-ionization source for plasma current startup. Different diagnostic systems like Rogowski coils, magnetic probes, flux loops, Langmuir probe, fast imaging and emission spectroscopy are installed on the device. The plasma temperature inside of GLAST, at the time of maxima of plasma current, is estimated by taking into account the Spitzer resistivity calculations with some experimentally determined plasma parameters. The plasma resistance is calculated by using Ohm's law with plasma current and loop voltage as experimentally determined inputs. The plasma resistivity is then determined by using length and area of the plasma column. Finally, the average plasma electron temperature is predicted to be 12.65eV for taking neon (Ne) as a working gas.

Steady-state operation in compact tokamaks with copper coils

This paper considers a fast track to non-energy applications of nuclear fusion that is associated with the ‘fusion for neutrons’ (F4N) paradigm. Being a useful product accompanying energy, fusion neutrons are more valuable than the energy released in DT reactions and they are urgently needed for research purposes and to develop and validate modern technologies. In the near future neutron yield in fusion devices will become significantly larger than that of fission and accelerator sources. This paper describes a compact tokamak fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ∼0.5 and ∼0.3 m with magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1–2 MA. The production rate of DT neutrons of (3–10) × 1017 n s−1 and their flux at the first wall of 0.2 MW m−2 ensure that the device is capable of fusion–fission demonstration experiments. The problems of major concern are discharge initiation, current drive, plasma—fast ion beam stability and high first wall and divertor loads. The conceptual design provides solutions to these problems and suggests the feasibility of the FNS-ST.

Active particle control experiments and critical particle flux discriminating between the wall pumping and fuelling in the compact plasma wall interaction device CPD spherical tokamak

Two approaches associated with wall recycling have been performed in a small spherical tokamak device CPD (compact plasma wall interaction experimental device), that is, (1) demonstration of active particle recycling control, namely, ‘active wall pumping’ using a rotating poloidal limiter whose surface is continuously gettered by lithium and (2) a basic study of the key parameters which discriminates between ‘wall pumping and fuelling’. For the former, active control of ‘wall pumping’ has been demonstrated during 50 kW RF current drive discharges whose pulse length is typically ∼300 ms. Although the rotating limiter is located at the outer board, as soon as the rotating drum is gettered with lithium, hydrogen recycling measured with Hα spectroscopy decreases by about a factor of 3 not only near the limiter but also in the centre stack region. Also, the oxygen impurity level measured with O II spectroscopy is reduced by about a factor of 3. As a consequence of the reduced recycling and impurity level, RF driven current has nearly doubled at the same vertical magnetic field. For the latter, global plasma wall interaction with plasma facing components in the vessel is studied in a simple torus produced by electron cyclotron waves with Ip < 1 kA. A static gas balance (pressure measurement) without external pumping systems has been performed to investigate the role of particle flux on a transition of ‘wall fuelling’ to ‘wall pumping’. It is found that a critical particle flux exists to discriminate between them. Beyond the critical value, a large fraction (∼80%) of pressure drop (‘wall pumping’) is found, suggesting that almost all injected particles are retained in the wall. Below it, a significant pressure rise (‘wall fuelling’) is found, which indicates that particles are fuelled from the wall during/just after the discharge. Shot history effects (integrated particle recycling behaviour from the plasma facing surfaces) are seen on that the critical particle flux is reducing from 0.8 × 10−4 to ∼0.1 × 10−4 Torr during the experimental campaign (∼3000 shots). In the wall pumping pressure range the wall pumping fraction is reduced with increasing surface temperature up to 150 °C.

Study on wall recycling behaviour in CPD spherical tokamak

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