Fusion Plasma Devices Research Articles

BENCHMARKING OF THE SERPENT 2 MONTE CARLO CODE FOR FUSION NEUTRONICS APPLICATIONS

Analyses of radiation fields resulting from a deuterium-tritium (DT) plasma in fusion devices is a critical input to the design and validation of many aspects of the reactor design, including, shielding, material lifetime and remote maintenance requirements/scheduling. Neutronics studies, which perform in-depth analysis are typically performed using radiation transport codes such as MCNP, TRIPOLI, Serpent, FLUKA and OpenMC. The Serpent 2 Monte-Carlo code, developed by VTT in Finland, is the focus of this work which seeks to benchmark the code for fusion applications. The application of Serpent 2 in fusion specific analysis requires validation of the codes performance in an energy range, and a geometrical description, which significantly differs to conventional nuclear fission analysis, for which the code was originally developed. A Serpent model of the Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) mock up experiment has been prepared and the calculated results compared against experimental data, as well as the reference Monte Carlo code MCNP. The analysis is extended to a model of DEMO with HCPB blanket concept. For this model, the flux, nuclear heating, tritium production and DPA are calculated, all of which are integral nuclear responses in fusion reactor analysis. In general, a very good agreement is demonstrated for both of the benchmarks, with any discrepancies pinpointed to different physics models implemented.

Ion temperature gradient mode driven solitons and shocks in superthermal plasma

An analysis of the ion temperature gradient driven mode solitary and shock waves in superthermal plasma is presented. Based on the observation in linear region it is shown that the phase velocity of the mode has strong dependency on superthermal particles population and on ion temperature. Linear dispersion relation is reduced to two different results by considering different limiting cases: (i) when superthermality (κ) is ignored in the system and (ii) when the fluctuations are caused by E×B convection only. In nonlinear regime we in our investigations observe that, nonlinear waves are strongly modified in superthermal plasmas. Furthermore, the different parametric analysis clearly shows that speed and the energy of these nonlinear structures driven by ion temperature gradient modify. These parametric analysis are quite constructive, which in turns may provide help in future designing of plasma fusion devices. The present investigation of this article may have its relevance to astrophysical and laboratory plasma.

Progress of advanced transport modeling of toroidal plasmas

Integrated modeling is required for predicting the behavior of burning plasma in fusion devices. Core plasma transport modeling is a key component which describes the time evolution of plasma profile. Between the simple 1D diffusive transport model and the large-scale microscopic turbulent transport simulation, advanced transport modeling which requires decent computational resource is desired. Two approaches, 1D fluid transport modeling and 3D kinetic transport modeling, are discussed.

The Plasma Characterization of Argon and Helium Gases in Inertial Electrostatic Confinement Fusion Plasma Device

The Plasma Characterization of Argon and Helium Gases in Inertial Electrostatic Confinement Fusion Plasma Device

Recent Progress on dispersion interferometers for nuclear fusion and low-temperature plasmas

A dispersion interferometer (DI) is known to be less sensitive to mechanical vibrations, which is one of the main sources of error for conventional interferometers. A simple optical configuration is also one of the advantages of the DI approach. Since the first application of a homodyne DI on a nuclear fusion plasma device in the 1990s, several interferometer techniques have improved the operation of the DI. Improvements in nonlinear crystals have also improved the performance and availability of the DI. Immunity to neutral gas density changes, which is needed for low-temperature plasma measurements, is also found. A remaining challenge is the suppression of the offset drifts, which can be significant. Recent studies confirmed that they are caused by ambient humidity changes.

Design and Operation of First Egyptian IEC Fusion Plasma Device

In this paper, Egyptian first inertial electrostatic confinement fusion (IECF) device, constructed at the Egyptian Atomic Energy Authority (EAEA-IEC), is introduced the characterization of IEC Plasma Device. It consists of 2.8 cm stainless steel cathode, 6.5 cm anode diameter with 10 cm diameter 30 cm height vacuum chamber. The discharge current and voltage of plasma discharge has been recorded using current probe and resistive voltage divider respectively. The X-ray emissions in IEC plasma device were investigated by employing time-resolved detector. The temporal distributions of detected x-rays emission are occurring during the initial 1 microsecond. The calculated rate of DD-neutron generation using the same electrode configuration about 106 – 108 neutrons/second.

Presheath-sheath coupling for kinetic trajectory simulation of a magnetized plasma sheath

The coupling of presheath-sheath parameters is extended for the study of magnetized plasma sheath using the kinetic trajectory simulation (KTS) method, in which the final self-consistent states are obtained iteratively by solving the kinetic equations. In our case, it is assumed that the ion and electron velocity distribution functions are cut-off Maxwellians at the sheath entrance. The results show that the cut-off and Maxwellian maximum velocities have equal magnitudes at the sheath entrance and at wall. The presheath electron temperature has a considerable effect on the self-consistent potential profile which affects the Child sheath thickness. The latter increases from 3.8320 μm to 5.4190 μm when the presheath electron temperature increases from 10 eV to 20 eV. It is found that the number of ions reaching wall is higher than that of the electrons and hence the space charge density has its maximum value there. Furthermore, the temperature of ions in the sheath region increases with the increase in presheath ion temperature. Moreover, the cut-off distribution causes our simulation result to deviate from the theoretical result found for the Boltzmann distribution by about 3%. The coupling scheme presented here provides a basis for smooth transition of plasma parameters at the presheath-sheath interface. The proper understanding of the magnetized plasma-wall transition plays a vital role for further exploring the plasma sheath characteristics which has useful applications in fusion and industrial plasma devices.

Investigation of tungsten-coated divertor tiles at the JET tokamak with the ITER‐Like Wall

Materials migration in fusion plasma devices and fuel retention in plasma-facing components are issues of great importance for the safe operation of fusion devices. The underlying mechanisms require a good understanding in order to make predictions regarding the lifetime of wall components and to assess the amount of fuel retained in the machine mainly in co-deposited layers. To reduce fuel inventory and to investigate plasma-wall interactions a large-scale experiment at the JET (Joint European Torus) tokamak is realized: operation with the ITER-Like Wall (JET-ILW) which comprises beryllium and tungsten. The current work reports on the post-mortem analysis of W/CFC tiles retrieved after the first deuterium-deuterium campaign at JET-ILW. Specimens from different areas of the divertor have been analyzed by means of several techniques including nuclear reaction analysis and Rutherford backscattering employing a deuterium beam. In addition, X-ray fluorescence spectroscopy and scanning electron microscopy with energy dispersive X-ray analysis are used to assess the sample surface morphology and analyze the stoichiometry of the surface of the samples in order to compare with the results from the ion beam analytical techniques.

Improvement of infrared imaging video bolometer for application to deuterium experiment on the large helical device.

An infrared imaging video bolometer was improved for application to a neutron environment in fusion plasma devices, i.e., the Large Helical Device (LHD). In order to calibrate the thermal characteristics of the activated foil absorber inside the plasma vacuum vessel, the remote-controlled in situ calibration system was improved with high-surface-flatness mirrors. Furthermore, the carbon coating method was improved by introducing a vacuum evaporation technique instead of the conventional spray technique to realize the coating on both sides of the absorber with reproducibility and uniformity. The optimal thickness of the coating was also determined. Owing to these coating improvements, the reproducibility of the effective emissivity on both sides especially was improved. Finally, the variation with the neutron irradiation of the thermal characteristics of the foil absorber was investigated. It was found that the effect was not significant for the total neutron emission of 3.6 × 1018 on LHD.

Diamagnetic energy measurement during the first operational phase at the Wendelstein 7-X stellarator

The magnetic diagnostic system at the Wendelstein 7-X stellarator includes three diamagnetic loops to measure magnetic flux changes in the plasma. Their signals are directly related to the plasma energy. The diagnostic design with respect to materials, component cooling and data acquisition is built to be fully steady-state capable within the harsh environment of a fusion plasma device. During the first operational phase, two diamagnetic loops have been put into operation, each of them close to one of the up-down symmetric main planes of the plasma column with a bean-shaped and triangular-shaped cross-section, respectively. Both loops measured reliable energies in accordance to theoretical expectations. The triangular-shaped diamagnetic loop is equipped with four compensation coils. They are used to compensate errors during the energy measurement due to small fluctuations of externally driven currents in the main superconducting magnetic field coils and eddy currents in the adjacent vacuum vessel and thereby increase the time-resolution allowing to measure fast changes of the plasma energy. The diamagnetic flux measurements agree well with corresponding estimations of diamagnetic signals using three-dimensional Biot–Savart calculations. A consistency check for the diamagnetic energy is performed by a reconstruction of the associated Pfirsch–Schlüter current distribution and a comparison of predicted signals with measurements of an arrangement of eight plasma encircling Rogowski coil segments. Additionally, the measured diamagnetic energy is compared to kinetic energy calculations based on density and temperature measurements performed by the Thomson scattering diagnostic and the x-ray imaging crystal spectrometer diagnostic. The resulting energy confinement times are similar to predictions of empirical scaling laws, like ISS04. For upcoming operational periods of Wendelstein 7-X, the diamagnetic energy measurement will be used to generate an interlock signal, which will turn off the main plasma heating systems in case of a sudden, unwanted plasma collapse.

Detection of Delayed Neutrons from Fissionable Samples: Monte Carlo Modelling and Physical Assumptions for a Design of the DET-12 Device

An activation of fissionable materials with neutrons has been considered as a possible neutron diagnostic of D–D and D–T fusion plasma. Fission reaction caused by fusion neutrons leads up to emission of secondary neutrons: prompt and delayed. Physical assumptions have been outlined to design a new device (DET-12) for measurements of delayed neutrons emitted from samples of fissionable materials activated with neutrons at big fusion-plasma devices. The aim is to support a classic neutron activation method used as one of plasma diagnostics at tokamaks or stellarators. An interpretation of the time decay of delayed neutrons enables an assessment of the primary neutron flux which induced fission reaction. Monte Carlo calculations have been carried out in order to elaborate the method considered. Nuclides like: pure 235U, 238U and 232Th, have been selected as possible materials to be irradiated. Physical fundamentals of generation of the delayed neutrons are mentioned and a resulting concept of the DET-12 device, built in the Institute of Nuclear Physics, Poland, is presented. A general size and dimensions of particular constituent material layers, and a number and placement of neutron detectors are optimized by means of Monte Carlo modelling. Recommendations for a technical design of the measuring chamber were formulated. Detection efficiency of DET-12 has been also estimated.

Physics of increased edge electron temperature and density turbulence during ELM-free QH-mode operation on DIII-D

For the first time, we report increased edge electron temperature and density turbulence levels (T̃e and ñe) in Edge Localized Mode free Quiescent H-mode (ELM-free QH-mode) plasmas as compared to the ELMing time period. ELMs can severely damage plasma facing components in fusion plasma devices due to their large transient energy transport, making ELM-free operation a highly sought after goal. The QH-mode is a candidate for this goal as it is ELM-free for times limited only by hardware constraints. It is found that the driving gradients decrease during the QH-mode compared to the ELMing phase, however, a significant decrease in the ExB shearing rate is also observed that taken together is consistent with the increased turbulence. These results are significant as the prediction and control of ELM-free H-mode regimes are crucial for the operation of future fusion devices such as ITER. The changes in the linear growth rates calculated by CGYRO [Candy et al., J. Comput. Phys. 324, 73 (2016)] and the measured ExB shearing rate between ELMing and QH-mode phases are qualitatively consistent with these turbulence changes. Comparison with ELMing and 3D fields ELM suppressed H-mode finds a similar increase in T̃e and ñe, however, with distinctly different origins, the increased driving gradients rather than the changes in the ExB shearing rate in 3D fields ELM suppressed the H-mode. However, linear gyrokinetic calculation results are generally consistent with the increased turbulence in both ELM-controlled discharges.

A dynamic monitoring approach for the surface morphology evolution measurement of plasma facing components by means of speckle interferometry.

Plasma Facing Components (PFCs) in a magnetically confined fusion plasma device will be exposed to high heat load and particle fluxes, and it would cause PFCs' surface morphology to change due to material erosion and redeposition from plasma wall interactions. The state of PFCs' surface condition will seriously affect the performance of long-pulse or steady state plasma discharge in a tokamak; it will even constitute an enormous threat to the operation and the safety of fusion plasma devices. The PFCs' surface morphology evolution measurement could provide important information about PFCs' real-time status or damage situation and it would help to a better understanding of the plasma wall interaction process and mechanism. Meanwhile through monitoring the distribution of dust deposition in a tokamak and providing an upper limit on the amount of loose dust, the PFCs' surface morphology measurement could indirectly contribute to keep fusion operational limits and fusion device safety. Aiming at in situ dynamic monitoring PFCs' surface morphology evolution, a laboratory experimental platform DUT-SIEP (Dalian University of Technology-speckle interferometry experimental platform) based on the speckle interferometry technique has been constructed at Dalian University of Technology (DUT) in China. With directional specific designing and focusing on the real detection condition of EAST (Experimental Advanced Superconducting Tokamak), the DUT-SIEP could realize a variable measurement range, widely increased from 0.1 μm to 300 μm, with high spatial resolution (<1 mm) and ultra-high time resolution (<2 s for EAST measuring conditions). Three main components of the DUT-SIEP are all integrated and synchronized by a time schedule control and data acquisition terminal and coupled with a three-dimensional phase unwrapping algorithm, the surface morphology information of target samples can be obtained and reconstructed in real-time. A local surface morphology of the real divertor tiles adopted from EAST has been measured, and the feasibility and reliability of this new experimental platform have been demonstrated.

Quasi-3D electron cyclotron emission imaging on J-TEXT

Electron cyclotron emission imaging (ECEI) can provide measurements of 2D electron temperature fluctuation with high temporal and spatial resolution in magnetic fusion plasma devices. Two ECEI systems located in different toroidal ports with 67.5 degree separation have been implemented on J-TEXT to study the 3D structure of magnetohydrodynamic (MHD) instabilities. Each system consists of 12 (vertical) × 16 (horizontal) = 192 channels and the image of the 2nd harmonic X-mode electron cyclotron emission can be captured continuously in the core plasma region. The field curvature adjustment lens concept is developed to control the imaging plane for receiving optics of the ECEI systems. Field curvature of the image can be controlled to match the emission layer. Consequently, a quasi-3D image of the MHD instability in the core of the plasma has been achieved.

The role of rectified currents in far-field RF sheaths and in SOL losses of HHFW power on NSTX

Radio-frequency (RF) rectification is an important sheath phenomenon for wave heating of plasma in fusion devices and is proposed to be the mechanism responsible for converting high-harmonic fast-wave (HHFW) power in the National Spherical Torus eXperiment (NSTX) into a heat flux to the divertor. RF rectification has two aspects: current rectification and voltage rectification, and, while the latter is emphasized in many application, we demonstrate the importance of current rectification in analysis of the NSTX divertor during HHFW heating. When rectified currents are accounted for in first-principle models for the heat flux to the tiles, we predict a sizeable enhancement for the heat flux in the presence of an RF field: for one case studied, the predicted heat flux increases from 0.103 MW/m2 to 0.209 MW/m2. We also demonstrate how this rectification scales with injected HHFW power by tracking probe characteristics during a HHFW power ramp; the rectified current may be clamped at a certain level. This work is important for minimizing SOL losses of HHFW power in NSTX-U but may also have implications for near-field studies of ICRF antennae: ignoring rectified current may lead to underestimated heat fluxes and overestimated rectified voltages.

Design and development of high pressure high temperature water circulation system for HHFTF

High Heat Flux Test Facility (HHFTF) at Institute for Plasma Research in Bhat in India is established for testing thermal performance of plasma facing materials and components under intense heat loads expected in plasma fusion devices. It mainly consists of the following sub-systems like vacuum system, high power electron beam system, diagnostic and calibration system, data acquisition & control system and high pressure high temperature water circulation system (HPHT-WCS). In order to test the water cooled test mock-ups or components at ITER like relevant parameters in HHFTF, HPHT-WCS is established that can supply water at desired pressures, temperatures and flow rates. HPHT-WCS, is capable of providing demineralised water up to 160°C temperature, 60bar pressure and 4.85kg/s mass flow rate. The loop is designed on a skid base frame which has piping connected to various process components and the system is operated remotely through a data acquisition and control system. The main requirements and characteristics, design and development of the HPHT-WCS are presented.

Ion-acoustic solitary waves in a positron beam plasma with electron trapping and nonextensivity effects

Ion-acoustic solitary waves (IASWs) are investigated in a plasma having a cold positron beam fluid, electrons following a vortex-like distribution with entropic index q, and dynamic ions. Using a standard procedure, a pseudo-potential energy equation is derived. The presence of nonextensive q- distributed trapped electrons and cold positron beam has been shown to influence the small amplitude soliton structure quite significantly. From the analysis of our results, it is shown that compressive IASWs are supported in this plasma model. As the real plasma situations are observed with plasma species having a relative flow, our present analysis should be beneficial for comprehending the electrostatic solitary structures observed in fusion plasma devices and positron winds observed in astrophysical plasmas.

First Breakthrough for Future Air-Breathing Magneto-Plasma Propulsion Systems

A new breakthrough in jet propulsion technology since the invention of the jet engine is achieved. The first critical tests for future air-breathing magneto-plasma propulsion systems have been successfully completed. In this regard, it is also the first time that a pinching dense plasma focus discharge could be ignited at one atmosphere and driven in pulse mode using very fast, nanosecond electrostatic excitations to induce self-organized plasma channels for ignition of the propulsive main discharge. Depending on the capacitor voltage (200-600 V) the energy input at one atmosphere varies from 52-320 J/pulse corresponding to impulse bits from 1.2-8.0 mNs. Such a new pulsed plasma propulsion system driven with one thousand pulses per second would already have thrust-to-area ratios (50-150 kN/m²) of modern jet engines. An array of thrusters could enable future aircrafts and airships to start from ground and reach altitudes up to 50km and beyond. The needed high power could be provided by future compact plasma fusion reactors already in development by aerospace companies. The magneto-plasma compressor itself was originally developed by Russian scientists as plasma fusion device and was later miniaturized for supersonic flow control applications. So the first breakthrough is based on a spin-off plasma fusion technology.

From Electrojet to ITER: India's Journey in Experimental Plasma Physics

India has an international presence in Plasma Physics and its diverse applications such as thermonuclear fusion, material processing, strategic and environmental applications and plasma devices. From a modest start in the early 1970s, we have made great strides in the field of experimental plasma physics. Capacity building in techniques relevant to plasma production, manipulation and parameter control, pulsed power, creation of magnetic fields of complex geometries, clean vacuum and pumping systems, development and deployment of diagnostics to enable understanding of fundamental processes in plasmas and computer simulation to model plasma phenomena have been truly remarkable. Parallel to this, a community of physicists, engineers and computer experts has grown and matured. Funding mechanisms and financial support essential to broad base the research and development activity by drawing in Universities and education institutes have been nucleated. It is through these activities that the human resource and technology development essential to sustain India's ambitious forays into magnetic confinement fusion and industrial and strategic plasma applications has taken place. This paper is an attempt to give a historical perspective to this journey, which started at the Physical Research Laboratory, Ahmedabad and later, involved the Institute for Plasma Research at Gandhinagar, many DAE Institutions, IITs and Universities.

Calculation of levels, transition rates, and lifetimes for the arsenic isoelectronic sequence Sn XVIII-Ba XXIV, W XLII

Multi-configuration Dirac–Fock (MCDF) calculations of energy levels, wavelengths, oscillator strengths, lifetimes, and electric dipole (E1), magnetic dipole (M1), electric quadrupole (E2), magnetic quadrupole (M2) transition rates are reported for the arsenic isoelectronic sequence Sn XVIII-Ba XXIV, W XLII. Results are presented among the 86 levels of the 4s24p3, 4s4p4, 4p5, 4s24p24d, and 4s4p34d configurations in each ion. The relativistic atomic structure package GRASP2K is adopted for the calculations, in which the contributions from the correlations within the n≤7 complexes, Breit interaction (BI) and quantum electrodynamics (QED) effects are taking into account. The many-body perturbation theory (MBPT) method is also employed as an independent calculation for comparison purposes, taking W XLII as an example. Calculated results are compared with data from other calculations and the observed values from the Atomic Spectra Database (ASD) of the National Institute of Standards and Technology (NIST). Good agreements are obtained. i.e, the accuracy of our energy levels is assessed to be better than 0.6%. These accurate theoretical data should be useful for diagnostics of hot plasmas in fusion devices.