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.     

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.     

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.     

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.     

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.     

Thermal Evolution of the Proton Irradiated Structure in Tungsten–5 wt% Tantalum

We have monitored the thermal evolution of the proton irradiated structure of W–5 wt% Ta alloy by in-situ annealing in a transmission electron microscope at fusion reactor temperatures of 500–1300 °C. The interstitial-type a/2<111> dislocation loops emit self-interstitial atoms and glide to the free sample surface during the early stages of annealing. The resultant vacancy excess in the matrix originates vacancy-type a/2<111> dislocation loops that grow by loop and vacancy absorption in the temperature range of 600–900 °C. Voids form at 1000 °C, by either vacancy absorption or loop collapse, and grow progressively up to 1300 °C. Tantalum delays void formation by a vacancy-solute trapping mechanism.     

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.     

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.     

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.     

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.     

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.     

D-T experiments on TFTR

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of 1 and yielded a maximum fusion power of 9.2 MW. The fusion power density in the core of the plasma was 1.8 MW m^–3 approximating that expected in a D-T fusion reactor. In other experiments TFTR has produced 6.4 MJ of fusion energy in one pulse satisfying the original 1976 goal of producing 1 to 10 MJ of fusion energy per pulse. A TFTR plasma with T/D density ratio of 1 was found to have 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of _E. The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfvén Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels.Work supported by U.S. Department of Energy Contract No. DE-AC02-76-CHO-3073.     

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.     

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.     

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.     

Effect of Internal Breeding of Tritium and Helium-3 on the Ignition of an ICF Fuel Pellet

Self-heating condition and following ignition in an Inertial Confinement Fusion (ICF) fuel pellet is evaluated by calculating the power equations, dynamically. In fact, the self-heating condition is a criterion that determines the minimum parameters of a fuel (such as temperature, density and areal density) that can be ignited. Deuterium is the main component of ICF fuels as large amounts of it are naturally available. In addition, the use of deuterium as a fuel in ICF causes the production of tritium and helium-3. However, pure deuterium has a high ignition temperature (\(\hbox {T}\ge 40\,\hbox {keV}\)) which makes it inefficient. In this paper, the power equations are solved, dynamically, and it has been indicated that internal tritium and helium-3 production at early evolution of compressed deuterium fuel causes ignition at lower predicted temperatures.     

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49–50:689–695, 2000; Tillack et al. in Fusion Eng Des 65:215–261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794–1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3–23, 2006).     

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.