Simulation of Radial Electric Field and Internal Transport Barrier Formation in Small Size Divertor Tokamak Plasma Edge

The simulation of the radial electric field shear, which is responsible for L-H transition by means B2SOLPS0.5.2D transport code, gives the dependence of this shear on plasma parameters. Also, as result of uni-directional neutral beam heating, internal transport barrier is formed and ion radial heat flux q _ir starts to decrease. Furthermore, the dependence of radial electric field shear on ion temperature gradient ITG has also investigated.     

Triple-probe Diagnostic Measurements in Plasma of GLAST Spherical Tokamak

Triple-probe has been developed and operated successfully to characterize ECRH-assisted argon as well as hydrogen microwave plasmas in GLAST Spherical Tokamak. This technique enables to determine transient plasma parameters such as floating potential, electron temperature and electron number density in rapidly time-varying plasmas. An effective electron heating mechanism is applied to produce microwave plasma by injecting radiofrequency (RF) radiation at a frequency of 2.45 GHz in the presence of resonant toroidal magnetic field. Plasma parameters and corresponding fluctuations are measured as a function of time in different gas fill pressures for various applied magnetic fields. The results demonstrate the dependence of plasma parameters such as V _f , T _e , n _e and their fluctuations on gas fill pressure during the pre-ionization phase of the GLAST operation. Plasma behavior is observed to be closely depending on the coupling of RF power during microwave discharge. Additionally, the hydrogen plasma shows pronounced fluctuations in comparison with argon plasma with some decrease in electron temperature and densities.     

Modeling of Heat Fluxes in Edge Plasma of Small Size Divertor Tokamak

The heat flows out from the tokamak core region are collected on the divertor plates and external wall. Control of heat flux exhaust in the SOL and divertor plates regions is one of the important issues in tokamak physics. There are important phenomena affecting heat flows were simulated. The simulation is based on the B2SOLPS5.0 2D multifluid code. It is demonstrated that, the following results: (1) The simulation shows that, the operation of small size divertor tokamak, the divertor plate with/without impurities influence on profiles of electron, ion temperatures, and heat loads significantly. (2) Under normal direction of parallel (toroidal) magnetic field and different values of edge plasma density, strong “SOL” heat flow exists directed towards the LFS (outer) plate. (3) The simulation results show that, the increasing of the plasma density strong influence on the ion and electron poloidal heat fluxes profile significantly. The ion and electron polodial heat flux increase by factor “~8” and “2.4” times. (4) The simulation results show that the in–out asymmetry of heat fluxes was reversed when switching on/off E × B drifts in the edge plasma of this tokamak. (5) The simulation results show correlation between the in–out asymmetry divertor heat fluxes and E × B drift velocity. (6) The observed heat loads asymmetry between HFS and LFS plates can be explained with the radial electric field in SOL. (7) Also the simulation results performed result in, the in–out asymmetry strong influence on the characteristic length of ion poloidal heat flux.     

Simulation of the Radial Electric Field in the Small Size Divertor Tokamak for Discharge with Neutral Beam Injection (NBI)

The radial electric field in the edge plasma of small size divertor tokamak can be simulated by B2SOLPS0.5.2D fluid transport code. The simulation provides the follow results: (1) Switching on and off the part of the parallel plasma viscosity driven by parallel ion diamagnetic heat flux (Bekheit in J. Fusion Energ 27(4), 338–345, 2008; Schneider et al. in Nucl. Fusion 41:387, 2001) and Counter-NBI plasma heating change profile of radial electric field significantly. (2) Switching on and off the parallel plasma viscosity driven by parallel ion diamagnetic heat flux leads to the radial electric field is toroidal magnetic field dependence (3) For the case of counter-NBI plasma heating, the switching on and off the current driven by part parallel plasma viscosity depends on the ion diamagnetic heat flux leads to the ion poloidal velocity is toroidal magnetic field B_T dependence. (4) The profile of the radial electric field in edge plasma of small size divertor tokamak is consistent with poloidal rotation velocity.     

Measurements and Simulations of Neutron Emission Versus Deuterium Filling Pressure in Plasma Focus Device PF-24

A series of experiments were carried out using the middle energy (dense) plasma focus device PF-24 with deuterium as a working gas under pressure in a range between 2 and 5 mbar for 17 kV of charging voltage. The relationship between these operating deuterium pressures in reference to the total neutron yield (Y_n) was estimated. The 5-phase Lee code was used to simulate the measured discharge current and neutron yield (Y_n) using a phenomenological beam-target neutron generating mechanism, which was incorporated in the model. Comparison of the Y_n versus pressure using fitted model parameters was made at each point of pressure. The good agreement between measured and computed Y_n values was achieved for discharges with lower neutron emission. The measured Y_n (below 2.6?×?10⁹ n/discharge—the median value) has been reproducible by the model for 73% of simulated discharges, while above the median value its prediction were incorrect. The kinetic plasma parameters which were measured and computed using the Lee code for different pressures are: the time to a current sheath collapse (t_c), the average axial current sheath velocity (v_z) and the so called velocity factor (RF). Good agreement was found in the whole range of deuterium pressures between the computed and measured results for these kinematic quantities. Presented findings in this work suggest that the character of neutron emission is more complex than it would seem from classical interpretation of neutron production based on a beam-target model.     

Cross Section Calculations of (n,2n) and (n,p) Nuclear Reactions on Germanium Isotopes at 14–15 MeV

Neutron incident reaction cross sections of Germanium isotopes (^70,72,74,76Ge) were investigated for the (n,2n) and (n,p) reactions around 14–15 MeV. Cross section calculations have been presented for ⁷⁰Ge(n,2n)⁶⁹Ge, ⁷²Ge(n,2n)⁷¹Ge, ⁷⁴Ge(n,2n)⁷³Ge, ⁷⁶Ge(n,2n)⁷⁵Ge, ⁷⁰Ge(n,p)⁷⁰Ga, ⁷²Ge(n,p)⁷²Ga, ⁷⁴Ge(n,p)⁷⁴Ga, and ⁷⁶Ge(n,p)⁷⁶Ga reactions. Theoretical calculations were performed with four different computer codes: ALICE/ASH for the Geometry Dependent Hybrid model, TALYS 1.6 for two component Exciton model, EMPIRE 3.2 Malta for Exciton model and PCROSS for Full Exciton model with the incident neutron energy up to 20 MeV. The (n,2n) and (n,p) reaction cross section calculations were compared with empirical formulas derived by several researchers and compared with the experimental data obtained from EXFOR database as well as with evaluated Nuclear data files (ENDF/B-VII.1: USA 2014). Results show good agreement between the theoretical calculations having a major importance in nuclear data evaluation calculations and the experimental data from literature.     

The Effect of Plasma Profiles on the Critical Value of $$n\uptau_{E}$$ for Ignition

The effect of plasma profiles for ignition condition in a stationary D–T plasma is investigated using the energy conservation equations for ions and electrons, assuming that steady state fusion power is produced with no external power. The alpha power heating is sufficiently large to sustain the plasma and to balance the combined Bremsstarhlung and thermal conduction losses. The space dependent Lawson criteria is derived and critical condition is identified. As a result of this analysis we have shown that the optimum temperature might be \(\bar{T} \approx 26\,{\text{keV}}\) and that the peaked profiles with \(n\sim\left( {1 - \frac{{r^{2} }}{{a^{2} }}} \right)^{{v_{n} }}\), ν _n  = 1, and \(T\sim\left( {1 - \frac{{r^{2} }}{{a^{2} }}} \right)^{{v_{T} }} ,\,v_{T} = 2\) are good to minimizing \(\bar{n}\uptau_{E}\) for ignition. The results for these profiles show the critical value of \((\bar{n}\uptau_{E} )_{min} = 0.08 \times 10^{20 } \,{\text{m}}^{ - 3} \,{\text{s}}\) showing the reduction by 1/3 from the reference value limit ν _n  = ν _T  = 0. For a 26 keV plasma with an energy confinement time of 1 s, a pressure of about 6.24 atm is required for the plasma to be ignited; that is, it is sustained purely by the self-heating of the fusion alpha particles.     

Study of Effects of the Effective Edge Safety Factor on the Energy confinement Time in IR-T1 Tokamak

We present study of the effects of effective edge safety factor on the energy confinement time in IR-T1 tokamak. For this purpose, four magnetic pickup coils were designed, constructed, and installed on the outer surface of the IR-T1, and then Shafranov parameter is obtained from them. Therefore, the effective edge safety factor obtained. Also, energy confinement time is obtained using diamagnetic loop. Experimental results on IR-T1 show that the maximum energy confinement time relate to the low values of effective edge safety factor (2.5 < q _eff (a) < 2.8). This is agreement with theoretical approach.     

Particle Trajectories and Energy Distribution from a New IEC Fusion Device: A Many-Body Approach

This paper reports the explorations on the particle dynamics, ion distribution, energy spectra and temperature in a new-designed inertial electrostatic confinement fusion device in case of low azimuthal magnetic field. The proposed design has six bar-sized cathodes at the vicinity of the central region and a central DC current-carrying bar injects a homogeneous azimuthal magnetic flux on the particles. The cylindrical device is simulated in the fully ionized Deuterium media. Following the 3D design of the chamber, the real-time simulations have been performed by the time integration of the electrical and magnetic forces. The model uses the many-body approach with the particle–particle and particle–chamber interactions. To implement the particle–chamber interaction, the finite difference method has been applied. Besides, the model includes reflection effects of particles from the electrically grounded chamber wall. According to the simulations, the particle trajectories exhibit complex fluctuations in the central region and nearby the chamber walls. The ion temperature has been calculated around T _i  = 35 keV for the source potential V = ?150 kV. In addition, the ion distribution indicates that 68 % of ions can be collected in the central region. According to the velocity distribution, there exists a double Gaussian distribution with a low velocity peak. In addition, nearly 61 % of ions stay in the energy scale between 2 keV ≤ E ≤ 39 keV. The averaged neutron rate is estimated as 5.96 × 10⁵ n/s.     

Power and Sizes of Tokamak Fusion Neutron Sources with NBI-Enhanced Reaction Rate

Steady state tokamak with deuterium–tritium plasma is considered as a basis for fusion neutron source for a hybrid fusion–fission reactor. Prototypes of such a system can be developed on the basis of the present day tokamaks as the plasma power gain factor Q ~ 1 is required for hybrid applications. Significant population of fast ions can be supported by a powerful neutral beam injection heating in regimes with Q ~ 1. The reaction rate for fast ions greatly exceeds the rate for thermal Maxwellian ions. The possible ranges of parameters are discussed for medium size tokamaks with minor plasma radius a = 0.5–1 m. Power and sizes of the neutron source are determined by the value of the injection energy. Power gain Q ≈ 1 can be achieved with injection energy of deuterium about 130 keV and tritium energy about 200 keV. Neutron power of 30–40 MW can be realized with a ≈ 1 m, and about of few megawatts with a ≈ 0.5 m.     

Measurement of Plasma Energy Confinement Time in Presence of Resonant Helical Field in IR-T1 Tokamak

Plasma energy confinement time is one of the main parameters of tokamak plasma and Lawson criterion. In this paper we present an experimental method especially based on diamagnetic loop (toroidal flux loop) for measurement of this parameter in presence of resonance helical field (RHF) in IR-T1 tokamak. For this purpose a diamagnetic loop with its compensation coil constructed and installed on outer surface of the IR-T1. Also in this work we measured the plasma current and plasma voltage from the Rogowski coil and poloidal flux loop measurements. Measurement results of plasma energy confinement time with and without RHF (L = 2, L = 3, L = 2 & 3) show that the addition of a relatively small amount of RHF could be effective for improving the quality of tokamak plasma discharge by flatting the plasma current and increasing the energy confinement time.     

Effects of Resonant Helical Field on Plasma Internal Inductance in IR-T1 Tokamak

Measurement of plasma internal inductance is important in tokamak plasma experiments (plasma internal inductance relates to the plasma current profile). In this paper we present an experimental investigation of effects of Resonant Helical Field (RHF) on the plasma internal inductance in IR-T1 tokamak. For this purpose, four magnetic probes and also a diamagnetic loop with its compensation coil were constructed and installed on outer surface of the IR-T1 tokamak, and Shafranov parameter, poloidal Beta, and then the internal inductance determined. In order to investigate the effects of RHF on internal inductance, we measured it in presence and also in absence of different modes of the RHF (L = 2, L = 3, L = 2&3). Experimental results show that L = 3 mode can flat the plasma current and increase the plasma internal inductance.     

TF Ripple Effects on Plasma Energy Confinement Time in IR-T1 Tokamak

In this paper we present analysis of the effects of Toroidal Field ripple (TF ripple) on the plasma energy confinement time in IR-T1 Tokamak. For this purpose, a diamagnetic loop with its compensation coil were designed and installed on outer surface of the IR-T1 tokamak. Amplitude of the TF ripple is obtained 0.01, and also the effects of TF ripple on the plasma energy confinement time discussed. In presence of the TF ripple and in low field side of the IR-T1 tokamak chamber (θ = 0), the local value of energy confinement time increased, whereas in the high field side (θ = 180), the energy confinement time decreased.     

RHF Effect on Shafranov Parameter and Shafranov Shift in IR-T1 Tokamak

In this paper we present an experimental study of effects of Resonant Helical Field (RHF) on Shafranov parameter and Shafranov shift in IR-T1 tokamak. For this purpose a four magnetic pickup coils were designed, constructed, and installed on outer surface of the IR-T1 tokamak chamber, and then the Shafranov parameter and Shafranov shift obtained. On the other hand, the external RHF applied on tokamak plasma and its effects on results measured. Experimental results of measurements with and without RHF (L = 2, L = 3, L = 2 & 3) show that the addition of a relatively small amount of RHF especially L = 3 mode could be effective for improving the quality of tokamak plasma discharge by flatting the plasma current and reducing the Shafranov parameter and Shafranov shift.     

Measurements of the Thermal Neutron Macroscopic Absorption Cross Section for Neutron Absorbing Layers

Objective of this study is measuring the macroscopic cross section of a neutron absorbing layer for thermal neutrons. For this purpose a neutron source and BF ₃ detector have been applied. For measuring macroscopic cross section of thermal neutrons by the \( I = BI_{0} e^{{ - \sum\nolimits_{tot} t}} \) Formula, it is necessary to provide suitable geometric conditions in order to assume the production and build-up coefficient to be the unit value (=1). To fulfill required conditions for this assumption, surface of the detector is covered with a 2 mm thick layer of cadmium. Radiation window of the detector has a 3 cm diameter, situated directly in front of the source. By placing the cadmium cover over the detector, variation of \( Ln{\frac{{I_{0}^{{}} }}{I}} \) values verses thickness of absorbent layer, renders linear function behavior, making it possible to measure the macroscopic cross section. The next stage is applying the MCNP code by simulating F1 tally and cosine-cards for calculating Total Macroscopic Cross-Section. Validation of this study is achieved through comparison of simulation by the MCNP code and results rendered by experiment measurements.     

Measurement of Plasma Poloidal Beta in Presence of Resonant Helical Field in IR-T1 Tokamak

Precise measurements of poloidal beta and internal inductance are essential for tokamak plasma experiments. In this paper we present an experimental investigation of effects of Resonant Helical Field (RHF) on the poloidal beta in IR-T1 tokamak. For this purpose, a diamagnetic loop with its compensation coil were constructed and installed on outer surface of the IR-T1 tokamak, and then poloidal beta measured. In order to investigate the effects of RHF on the poloidal beta, we measured it with and without introducing of different modes of the RHF (L = 2, L = 3, L = 2 & 3). Experimental results discussed.     

Synthesis and adsorption properties of modified SDB carrier

Styrene–divinylbenzene copolymer (SDB) is a key material for preparing Pt/SDB hydrophobic catalyst, which could be used for the treatment and purification of tritiated water. In this paper, a modified SDB (MSDB) carrier based on tert-butyl styrene (t-Bu-St), methyl methacrylate (MMA), styrene (St) and divinylbenzene (DVB) was prepared by aqueous suspension polymerization. Static adsorption experiment shows that the MSDB has the best adsorption performance at the molar ratio of St, DVB, t-Bu-St and MMA of 1:0.85:0.3:0.3 with n-heptane as porogen. The adsorption behavior of MSDB is analyzed by theoretical formulas. Data show that the adsorption process is in accordance with the Lagrange pseudo-first-order kinetics and Freundlich isotherm (1/n?<?1), which is exothermic and entropy decrease (ΔH⁰?<?0, ΔS⁰?<?0) in a temperature range of 293.15–333.15 K.     

The Effect of Magnetic Field of Multicusp and Wall Material on Electron Trajectories

In this paper, the effect of electron confinement with magnetic fields in the multicusp ion source has been investigated. That is, electron confinement with magnetic fields plays a very important role for the generation of negative ions at plasma. Three-dimensional spatial distributions of electrons production are obtained for a multicusp ion source. The electron confinement of magnetic fields from various surface materials (such as Al₂O₃, Al, Au, Cu, w and stainless-steel) have been compared in a multicusp plasma source. The electron confinement effect becomes stronger with increasing N (the number of rows of permanent magnets) and using Al for plasma chamber wall material. The results of investigations have demonstrated good correspondence with experimental data, and therefore the adequacy of the developed approach and the possibility to build more effective source on this basis.     

Modelling of Small Size Divertor Tokamak Plasma Edge in the Higher Confinement Regime

A version of the B2SOLPS0.5.2D fluid transport code is the new version of B2SOLPS fluid transport code, which is suited technique to simulate the edge plasma of small size divertor tokamak in the H- regime. The results of simulation provide the following: (1) the radial electric field inside the transport barrier is consistent with the neoclassical nature of the radial electric field. (2) The absolute value of the radial electric field shear at inner side of internal transport barrier is small and consistent with the value of shear before the L–H transition, while the value of shear at barrier is significantly large. (3) As a result of strong radial electric field shear and strong barrier formation the diffusion coefficient reduced by factor ~3 with respect to L-mode while ion heat conductivity reduced by factor ~22 with respect to L-mode inside the barrier. (4) The toroidal (Parallel) flux is directed along co-current direction as L-mode but at inner side of barrier is significantly large in absolute value. (5) The radial profile of toroidal rotation in vicinity of transition layer is determined by the parameter δ (width of the transition layer) depending on the collisionality and anomalous diffusion coefficient.