Precise determination of the plasma column position is fundamental for equilibrium maintenance and stability control in tokamak devices. This work presents a comparative study of two methodologies implemented on the Thailand Tokamak-1 (TT-1): a Toroidal Filament Model using circumferentially oriented Mirnov probes and an Optical Boundary Reconstruction (OBR) technique based on high-speed visible imaging. The filament model provides a rapid analytical solution (similar to 10-1 ms per time slice) for the plasma centroid and is adaptable, allowing many different combinations of magnetic probes to be used in the event of malfunctioning probes, while the optical method serves as a direct geometric cross-validation tool. Experimental results from Ohmic discharges demonstrate strong agreement between the two methods during stable flat-top phases, with radial and vertical displacement trends consistent within experimental margins. However, discrepancies observed during dynamic phases are analyzed in the context of magnetic noise contributions and optical emission limitations. During this dynamic phase, the OBR technique fails to calculate the plasma column position due to optical limitations, whereas the filament model remain functional throughout the discharge. Because of its speed and consistency, the study confirms the viability of the Toroidal Filament Model for future real-time feedback control systems in TT-1.

Thailand Tokamak - 1 is a compact, circular-shaped tokamak with major and minor radii of 0.65 m and 0.20 m, respectively. It successfully achieved its first plasma in 2023, generating a maximum plasma current of approximately 100 kA, in line with the magnetic field induction work sequence of the magnet system in Thailand Tokamak - 1. This paper presents the design and measurement of the magnet system in Thailand Tokamak - 1, where the vacuum magnetic field was measured using diamagnetic loops and a rack of magnetic probes inserted into a vacuum chamber. Additionally, the Ohmic discharge process and its characteristic behavior in the experiment are described and analyzed.

A novel Imaging Neutral Particle Analyzer (INPA) was newly installed in early 2024 to enhance the understanding of fast ion confinement on Large Helical Devices (LHDs). This diagnostic system, based on a magnetic spectrometer using a scintillator, provides energy-resolved radial profiles of confined fast ions by measuring charge-exchanged fast neutrals escaping from the plasma. The system utilizes a 100 nm thick carbon foil to ionize the fast neutrals, subsequently deflecting the ions toward a scintillator via the existing local magnetic field. The fast ion energy and sightline determine the position of the scintillation, while the light intensity depends on the flux of the fast ions. The INPA features two apertures, facilitating effective measurements in both clockwise and counterclockwise magnetic field directions in the LHD. This INPA was designed as a passive measurement system that measures fast ions charge exchange with background neutrals, focusing on perpendicular beam ions from 5 to 100 keV with an energy resolution of about 5.75 keV. This paper describes the details of the design, installation, and the initial results of the INPA on the LHD. This work will contribute significantly to a deeper understanding of fast ion transport due to magnetohydrodynamic instabilities.

Thailand Institute of Nuclear Technology (TINT) is developing Thailand Tokamak-1 (TT-1) from a former device HT-6 M of China. The first hydrogen plasma will be initiated in 2023. To investigate high-fl plasma and physics related to fast ions, TT-1 will be equipped with auxiliary heating systems. In this work, a feasibility study for installing a neutral beam injection (NBI) heating system in TT-1 is carried out. This work is motivated to characterize beam ion's orbits in different injection angles and to explore a condition suitable in terms of higher heating efficiency. In this work, we assume that a hydrogen beam will be launched into the TT-1 plasma with an acceleration voltage of 20 kV. The orbit simulations using the gyromotion following code LORBIT are performed in various magnetic field equilibria, i.e., different plasma current (Ip), toroidal magnetic field strength (Bt), and the magnetic axis (Rax). Furthermore, beam ions are injected in different directions, i.e., tangential co-injection and tangential counter-injection. In the case of co-injection, beam ion loss is not significant, by about 6%, whereas beam ion loss fraction is evaluated to be 26-34% in the case of counter-injection. Also, it is found that the number of lost beam ions is significantly affected by changing Ip and Rax. The results obtained in this work will directly support the experiment plan for the high-performance plasmas, design of the fast-ion diagnostic system, and systematic understanding of beam ion's confinement property and beam-ion-driven magnetohy-drodynamic (MHD) instabilities in TT-1.

Computer simulations of 1.5D BALDUR predictive transport code which is coupled with empirically predictive boundary models for the pedestal densities and temperatures are carried out to investigate the plasma characteristics in ITER with the helium environment. In each simulation, the transports of particles and energy in the confined volume are predicted by both neoclassical and anomalous transports in which the latter is based on multi-mode model. The boundaries of the confined plasma are defined at the top of the pedestal and the width of the pedestal is assumed to be fixed. The pedestal temperatures are predicted based on the magnetic and flow shear stabilization. The pedestal densities of hydrogenic and impurity particles are empirically determined from experimental data taken from the International Pedestal Database. Based on these boundary models, the simulations show that the amount of the helium content is 9.9%. The fusion gains predicted by the model is about 3.5. However, as the helium fraction increases to 18%, the fusion gains are further reduced to 1.0 due to the fuel dilution and increasing radiative power loss. Moreover, micro-instability analysis of the ITER plasmas for different level of the helium content is also explored. The results show that, as the amount of helium content increases, the ion-temperature gradient mode is stabilized due to the decrease of the temperature gradient. The trapped-electron mode is also reduced because of increasing the electron collisionality.

JET H-mode plasma discharge 53212 simulation during the pellet fueling operation in the presence of an internal transport barrier is carried out using the 1.5D BALDUR integrated predictive modelling code. The plasma instability during ITB formation with pellet injection in a tokamak is investigated. These simulations use a neoclassical transport model and an anomalous transport model (either multimode or mixed Bohm/gyro-Bohm core transport model). The boundary condition is described at the top of the pedestal, which is calculated theoretically based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The toroidal flow calculation is based on the neoclassical viscosity toroidal velocity model. It was found that the shallower pellet does not destroy the ITB, which locating mainly between r/a = 0.8 and 0.9. Moreover, in the plasma center region (0.4<r/a<0.6) the effective electron thermal diffusivities do not change during the ablation time. However, the effective electron thermal diffusivities decrease after pellet ablation, which means a shallower pellet can improve the internal transport barrier.

The BALDUR code is used for predicting core profiles in H-mode plasma, by employing a combination of the MMM95 anomalous transport model and the NCLASS neoclassical transport model. The pedestal temperature and density of Deuteron and Carbon are taken from the experiment as a boundary conditions in the simulations. The simulated profiles, including those of the electron and ion temperature, electron, Deuteron and Carbon ion density profiles, are compared with the corresponding experimental data. The multi-parameter optimization is used for obtaining the most suitable coefficients for hydrogenic and impurity diffusion coefficient for new MMM95 transport model. This new set of coefficients for hydrogenic and impurity transport can improve the prediction of the electron and ion temperature, electron, Deuteron and Carbon ion density profiles of JET H-mode discharges within the RMS errors of 13%.

The impact of the E × B flow shear stabilization on particle transport and density peaking at JET is analyzed in the framework of integrated modelling with the CRONOS code.For that purpose,plasmas from a power scan which show a significant increasing of density peaking with the injected neutral beam injection power have been used as a modeling basis.By means of simulations with the quasilinear model GLF23 for the heat and particle transport,a strong link between the particle confinement and E × B flow shear stabilization is found.This is particularly important close to the pedestal region where the particle pinch direction becomes strongly inward for high E × B flow shear values.Such impact introduces some non-negligible deviation from the well-known collisonality dependence of the density peaking,whose general trend has been also obtained in the framework of this modelling by performing pedestal density scans.

Self-consistent simulations of sawtooth oscillations on four DEMO (demonstration fusion power plant) designs, proposed by European, Indian, Chinese, and Korean teams, are carried out using the Porcelli sawtooth triggering model and the modified version of the Kadomtsev magnetic reconnection model coupled with the BALDUR integrated code. The simulation results suggest that all sawtooth crashes are triggered by the driving force for the internal kink instability overcoming the fast ion stabilization found in all DEMOs. Conversely, different expansions of the pre-crash helical flux are found, causing different mixing radii on each design. The mixing radius of the European DEMO is found to be the largest, resulting in the largest reduction of the core plasma temperature after the sawtooth crash. It is also observed that for all DEMOs except the Chinese DEMO, regular sawtooth crashes lead to an increase in the core plasma density and radiative power density in different levels, which are due to the steepness of the pre-crash hollow density profile. Furthermore, while helium is found to be outwardly transported by sawtooth crashes for all DEMOs, the sawteeth yield a penetration of the seeded impurities, remarkably found in the European design. Interestingly, based on the European design, a decrease in the atomic mass of the plasma fuels and seeded impurities are found to yield greater flatness of the pre-crash density profile and thus a smaller impact of sawtooth oscillations on central density. Additionally, the sawteeth degrade the plasma performance in the Chinese DEMO, but demonstrate positive impacts on the performance in the European, Indian, and Korean DEMOs, which results from the increase of the central ion density caused by the sawteeth. Furthermore, variation in neutral beam injection power exhibits both positive and negative correlations of sawtooth oscillation impacts on central ion density. The sensitivity of the mixing region of DEMOs is also investigated for scans where the main plasma parameters are varied.

Shielding configuration studies for the fusion plasma experiment of the Thailand tokamak upgrade are presented in this work. The neutron rate of 1014 n/s of DT fusion is determined to assess the next step operation in phase II of the Thailand tokamak upgrade in the future. In order to optimize the materials and dimensions of the shielding, a series of MCNP simulation was performed to assess the neutron streaming in order to enhance understanding of neutron and gamma transport in the plasma torus, hall and the pathways shielding. Effect of three commonly available concrete composition (ordinary concrete, barite concrete and boron frits-baryte concrete) have been investigated as a torus wall shielding and entrance maze. Four orders of magnitude of fast neutron flux reduction was observed at outside torus hall. The thermal neutron was significantly reduced using the labyrinth structure. The conceptual design of the biological radiation shielding is assessed and presented. This study provides support for the future neutron and gamma radiation safety at the Thailand tokamak upgrade facility. (C) 2019 The Japan Society of Plasma Science and Nuclear Fusion Research

Simulations of future Thailand tokamak plasmas are carried out using a CRONOS integrated predictive modelling code. The design of the reactor is based on nominal parameters of HT-6M tokamak. The code consists of a 1D transport solver with general 2D magnetic equilibria, and includes several heat, particle and impurities transport models as well as heat, particle and momentum sources. In this work, a combination of a mixed Bohm/gyro-Bohm anomalous transport model and an NCLASS neoclassical transport model are used to calculate plasma core diffusivities. The boundary condition of the simulations is taken to be at the top of the pedestal which is calculated based on an international multi-tokamak scaling. Sensitivity analyses on plasma performance of the future Thailand tokamak are investigated by varying plasma current, toroidal magnetic field and external heating schemes. It is found that the performance in H-mode plasmas such as transport barrier at plasma edge and central temperatures are found to be sensitive to heating schemes and their magnitudes. Additionally, ICRH and LH methods appear to be the most effective scheme of heating for ion and electron temperatures, respectively. Central ion temperature in the range of 120 - 750 eV and central electron temperature in the range of 1,100 - 2,750 eV with heating are expected.

This study investigates the plasma performance in HT-6M tokamak using 1.5D integrated predictive modeling code BALDUR. The simulations are carried out under the designed plasma conditions, including R = 65 cm, a = 20 cm, B-T = 1.5 T, n(e) = 1 x 10(19) m(-3) and I-p = 40 - 150 kA without external heating. In these simulations, a combination of turbulence and neoclassical transports is used for predicting thermal and particle transport. Thus, the plasma evolution for plasma current, temperature, and density can be predicted under a designed condition. In addition, the influence of current rampup for the plasma performanceis investigated. The scenario study for the tokamak is also carried out by varying plasma current. To summarize the results yield the electron temperature at the center T-e(0) = 477 - 1,551 eV (MMM95) and 328 - 1,384 eV (Mixed B/gB), the ion temperature at the center T-i(0) = 26 - 50 eV (MMM95) and 18 - 42 eV (Mixed B/gB) the electron densit y = 6.4 x 10(18) - 1.4 x 10(19) m(-3) in both Mixed B/gB and MMM95 simulations. Using the obtained plasma parameters, the radiated power of the carbon impurity is assessed. (C) 2018 The Japan Society of Plasma Science and Nuclear Fusion Research

The criteria for the L-H transition and ITB transition in fusion plasmas is studied based on the bifurcation concept. Three transport equations including thermal, particle and toroidal momentum density are solved simultaneously, resulting in the prediction of plasma pressure, plasma density and toroidal velocity profiles at steady state. The thermal and particle transport include both neoclassical and anomalous effects with the velocity shear dependent suppression effect. The results show that the flux (thermal/particle/momentum) versus gradient (pressure/density/velocity) space of each field independently exhibits s-curve bifurcation nature in which a forward L-H and ITB transitions require higher flux than that of the respective backward transitions, hence hysteresis behaviors. In addition, it is found that there exist certain regimes where the transitions are possible. In particular, the ratios of anomalous over neoclassical transport must exceed certain thresholds.

A set of coupled particle and thermal transport equations is used to study a formation and sustainability of an edge transport barrier (ETB) in tokamak plasmas based on two-field bifurcation. The two transport equations are numerically solved for spatio-temporal profiles of plasma pressure and density. The plasma core transport includes both neoclassical and turbulent effects, where the latter can be suppressed by flow shear mechanism. The flow shear, approximated from the force balance equation, is proportional to the product of pressure and density gradients, resulting in non-linearity behaviors in this calculation. The main thermal and particle sources are assumed to be localized near plasma center and edge, respectively. It is found that the fluxes versus gradients regime illustrates bifurcation nature of the plasma. This picture of the plasma implies hysteresis properties in fluxes versus gradients space. Hence, near marginal point, the perturbation in thermal or particle sources can trigger an L-H transition. Due to hysteresis, the triggered H-mode can be sustained and the central plasma pressure and density can be enhanced.

In this work, the impurity accumulations and their performance in the presence of both ITB and ETB in ITER and DEMO plasmas are investigated using a BALDUR integrated predictive modelling code. In these simulations, a combination of a neoclassical transport model NCLASS and an anomalous transport model Mixed Bohm/gyro-Bohm is used. The boundary condition is described at the top of the pedestal, which is calculated theoretically based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The toroidal flow is calculated based on the NTV (neoclassical toroidal viscosity) toroidal velocity model. The time evolution of plasma temperature and density profiles of ITER and DEMO (Korean K-DEMO and Japanese DEMO models A, B and C) plasmas are simulated in H-mode scenario with and without ITB formation. It is found that Japanese DEMO model C yields highest plasma temperature; while Korean DEMO yields the best plasma performance among those designs considered. Impurity accumulation is found to be highest in Japanese DEMO model B.

This research studies plasma performance in fusion Tokamak system by investigating parameters such as plasma pressure in the presence of an edge transport barrier (ETB) and an internal transport barrier (ITB) as the off-axis driven current position is varied. The plasma is modeled based on the bifurcation concept using a suppression function that can result in formation of transport barriers. In this model, thermal and particle transport equations, including both neoclassical and anomalous effects, are solved simultaneously in slab geometry. The neoclassical coefficients are assumed to be constant while the anomalous coefficients depend on gradients of local pressure and density. The suppression function, depending on flow shear and magnetic shear, is assumed to affect only on the anomalous channel. The flow shear can be calculated from the force balance equation, while the magnetic shear is calculated from the given plasma current. It is found that as the position of driven current peak is moved outwards from the plasma center, the central pressure is increased. But at some point it stars to decline, mostly when the driven current peak has reached the outer half of the plasma. The higher pressure value results from the combination of ETB and ITB formations. The drop in central pressure occurs because ITB stats to disappear.

The Thailand Plasma Focus II (TPF-II) is a 3.3 kJ dense plasma focus that was developed at Walailak University, Thailand. The aim of the device is to study the production of ion beams in the keV energy range and their applications for the color modification of gemstones. A high-energy ion beam is produced by heating and acceleration in the pinch phase of the plasma focus. The heating process is determined by the maximum electrical current, which can be optimized by variation of the system’s inductance. Lee model code was implemented to optimize the configuration of the electrodes. The current waveforms for the different initial conditions were used to obtain the system’s inductance, which was verified by a short circuit test. It was found that the inductance and resistance were about 153 nH and 12 mΩ, respectively.

The improvement of water quality using by atmospheric plasma produced from a dielectric barrier discharge (DBD) was studied. An experiment was set-up with a 4 mm diameter pipe, which contains 2 electrodes and has an air flow with the rate of 15 liters per minute. Surface water, domestic wastewater and DI water were treated with the DBD plasma for some period of time. Electricity was supplied at 3.5 kV with the frequency of 5.5 kHz. Some key parameters of water quality includes the level of chemical oxygen demand (COD), total suspended solid (TSS), color, and odor are measured before and after. The result showed that strong acid with pH below 2 was observed after 60 minutes plasma treatment for the DI water, while the surface water and wastewater needs about 120 minutes to pH below 2 even though the pH value are about the same at the beginning. Moreover, It was formed that the COD, TSS microorganism was noticeably decreased, therefore the increasing of transparency level. This result confirms that atmospheric DBD plasma generated acidity in water as reduce amount of organic and suspended solid in water.

B. Chatthong1, W. Kanjanaput2, J. Promping3, R. Picha3 and T. Onjun2 1School of Manufacturing Systems and Mechanical Engineering, Sirindhorn International Institute of Technology, Thammasat University, Pathum Thani, Thailand 2School of Manufacturing Systems and Mechanical Engineering, Sirindhorn International Institute of Technology, Thammasat University, Pathum Thani, Thailand 3Thailand Institute of Nuclear Technology, Bangkok, Thailand