Tritium Plasma Research Articles

Plasma heating by non-linear wave-Plasma interaction

We simulate the non-linear interaction of waves with magnetized tritium plasma with the aim of determining the parameter values that characterize the response of the plasma. The wave-plasma interaction has a non-conservative Hamiltonian description. The resulting system of Hamilton's equations is integrated numerically using fourth order Runge-Kutta scheme. It is found that for wave amplitude &#945 as low as 0.01Bo the response of the plasma is remarkably different from the prediction of linear response theory. The response cannot be explained in terms of whether or not the wave frequency w is a harmonic of the ion cyclotron frequency &#937. The scaled drift velocity of the ions &#945 and the scaled phase velocity of the waves &#946 were found to be more relevant in explaining the response characteristics. For &#945 >> &#946, the plasma response is found to be chaotic while for &#945

Two dimensional simulation of ion beam-plasm interaction

Hybrid plasma simulation is a model in which different components of the plasma are treated differently. In this work the ions are treated as particles while the electrons are treated as a neutralizing background fluid through which electric signals may propagate. Deuterium ion beams incident on the tritium plasma interact with the plasma through the wake electric and magnetic fields of the beam currents. The model is described by the Vlasov-Maxwell-fluid equations with displacement currents omitted in the Ampere's equation. Euler scheme advances the ion positions and velocities while finite difference scheme approximates the spatial derivatives of field quantities. Bilinear interpolation is used in interpolating field quantities from the grid points to push particles in their respective positions. The beam parameters that determine the response of the plasma are the beam velocity Vb, the density nb and the radius rb. It was found that the sharpness of the characteristic frequencies for power absorption increased with increasing beam density while increasing the beam radius destroyed the characteristic frequencies and spread the frequency band of the continuous lines. The Joule heating of the plasma fluctuated in time and could be localized. Increasing nb delocalized the heating. The magnetic field profile did not respond sharply to beam parameters but the respond was sufficient enough to cause inhomogeneity in the direction of applied magnetic field B&#963. Journal of the Nigerian Association of Mathematical Physics Vol. 9 2005: pp. 561-572

Effects of Sputtering of and Radiation by Aluminum on Magnetized Target Fusion Plasmas

We estimate numerically the rate of radiation by aluminum impurities for parameters relevant to magnetized target fusion (MTF) plasmas. We demonstrate that the coronal equilibrium is appropriate for expected MTF plasma parameters. Using the coronal equilibrium, we estimate the power radiated per impurity ion is 0.25–0.5 × 10−16 MW for temperatures and densities relevant to present plasma parameters taken from the FRX-L experiment at Los Alamos National Laboratory and is approximately 75.0 × 10−16 MW for temperatures and densities relevant to anticipated MTF plasmas. We calculate the sputtering rate of aluminum by thermal deuterium and tritium plasma ions is a few percent assuming an impact angle of 45°. Finally, we estimate that with aluminum impurity levels of a few percent, the impurity radiation power density would be approximately 25 kW/cm3 for FRX-L conditions and 2.5 GW/cm3 for anticipated conditions in a MTF plasma. While we have assumed a sputtering model of impurity generation, the results for the power density apply for impurity levels of a few percent, regardless of the generation mechanism.

Choice of coils for a fusion reactor

In a fusion reactor a hot plasma of deuterium and tritium is confined by a strong magnetic field to produce helium ions and release energetic neutrons. The 3D geometry of a stellarator provides configurations for such a device that reduce net toroidal current that might lead to disruptions. We construct smooth coils generating an external magnetic field designed to prevent the plasma from deteriorating.

Development of multifilamentary niobium–titanium and niobium–tin strands for the International Thermonuclear Experimental Reactor project

The International Thermonuclear Experimental Reactor(ITER) device should demonstrate the scientific and technological possibility of commercial fusion energy production in large scale in order to solve the worldwide energy problem in the future. The superconducting magnet system is the key part of the ITER device to supply high magnetic fields for confining the deuterium–tritium plasma. The multifilament NbTi and Nb3Sn strands with high quality have been studied to meet the specifications of superconducting strands for fabricating poloidal field coils (PF) and toroidal field coils (TF). For NbTi strands with 8306 filaments, Jc of 2910Amm−2 (4.2K, 5T, 0.1μVcm−1) has been obtained by a conventional process. The proposed process could be used for fabrication of long strands with a unit length more than 5000m. By an internal tin process the multifilamentary Nb3Sn strands with a diameter of 0.79mm and a unit length longer than 5000m have been successfully fabricated. The highest non-Cu Jcn (12T, 4.2K, 0.1μVcm−1) value of 1249Amm−2 has been obtained. The n-value of Nb3Sn strands is larger than 20 and the residual resistance ratio (RRR) value lies between 150 and 220. The formation of the Nb3Sn superconducting phase together with the evolution of microstructure has been investigated by neutron diffraction and scanning electron microscopy. The results indicate that the properties of NbTi and Nb3Sn strands have already met basically the specifications proposed by the ITER program.

Thermonuclear ringtones

The ability to confine alpha particles within a burning deuterium–tritium plasma is likely to be crucial to the future of fusion power generation. Resonant interactions between alpha particles and magnetohydrodynamic vibrations could threaten their confinement.

Status of the ITER magnets

The overall programme objective of ITER is to demonstrate the scientific and technologicalfeasibility of fusion power production for peaceful purposes. ITER would accomplish thisobjective by demonstrating controlled burn of deuterium tritium plasmas, andby demonstrating essential technologies (of a reactor) in an integrated system.The magnetic coils of ITER are all made of superconducting conductors. A comprehensiveresearch and development programme has been carried out to demonstrate the feasibility ofthe magnetic system. The major elements of this programme have been the construction ofthe toroidal field and central solenoid model coils. The testing of these model coils hasproduced all the necessary information to finalize the design of the ITER magneticsystem. In addition, a poloidal field conductor insert is being manufactured and willbe tested in 2006 at Naka (J) to confirm the design of the poloidal field coils.

Radio-frequency heating of sloshing ions in a straight field line mirror

A scenario to sustain a sloshing ion population with radio-frequency heating in a newly proposed mirror device, the straight field line mirror, is examined. The possibilities of ion cyclotron heating in two-ion species plasma have been analyzed and a scheme with longitudinal wave conversion and fundamental harmonic heating of deuterium ions in tritium plasma has been investigated. This scheme provides efficient ion heating for high deuterium “minority” concentration without substantial conversion to slow waves and heating of the electrons. Numerical calculations carried out for a reactor-scale device show that conversion of the fast magnetosonic wave to the fast Alfvén wave occurs. For reasons of strong cyclotron absorption of the fast Alfvén wave, only a small portion of the wave energy transits through the cyclotron layer and penetrates to the central part of the trap. The power deposition is peaked at the plasma core. The amount of deposited power does not depend sensitively on the parameters of the discharge. The study uses numerical three-dimensional calculations for the time-harmonic boundary problem for Maxwell’s equations. For radio-frequency heating in this scheme, a simple efficient strap antenna is proposed. It has low Q antenna and operates in the regime of global resonance overlapping.

Overview of transport, fast particle and heating and current drive physics using tritium in JET plasmas

Results are presented from the JET Trace Tritium Experimental (TTE) campaign using minority tritium (T) plasmas (nT/nD < 3%). Thermal tritium particle transport coefficients (DT, vT) are found to exceed neo-classical values in all regimes, except in ELMy H-modes at high densities and in the region of internal transport barriers (ITBs) in reversed shear plasmas. In ELMy H-mode dimensionless parameter scans, at q95 ∼ 2.8 and triangularity δ = 0.2, the T particle transport scales in a gyro-Bohm manner in the inner plasma (r/a < 0.4), whilst the outer plasma particle transport scaling is more Bohm-like. Dimensionless parameter scans show contrasting behaviour for the trace particle confinement (increases with collisionality, ν* and β) and bulk energy confinement (decreases with ν* and is independent of β). In an extended ELMy H-mode data set, with ρ*, ν*, β and q varied but with neo-classical tearing modes (NTMs) either absent or limited to weak, benign core modes (4/3 or above), the multiparameter fit to the normalized diffusion coefficient in the outer plasma (0.65 < r/a < 0.8) gives DT/Bϕ ∼ ρ*2.46ν*−0.23β−1.01q2.03. In hybrid scenarios (qmin ∼ 1, low positive shear, no sawteeth), the T particle confinement is found to scale with increasing triangularity and plasma current. Comparing regimes (ELMy H-mode, ITB plasma and hybrid scenarios) in the outer plasma region, a correlation of high values of DT with high values of vT is seen. The normalized diffusion coefficients for the hybrid and ITB scenarios do not fit the scaling derived for ELMy H-modes. The normalized tritium diffusion scales with normalized poloidal Larmor radius in a manner close to gyro-Bohm , with an added inverse β dependence. The effects of ELMs, sawteeth and NTMs on the T particle transport are described. Fast-ion confinement in current-hole (CH) plasmas was tested in TTE by tritium neutral beam injection into JET CH plasmas. γ-rays from the reactions of fusion alpha and beryllium impurities (9Be(α, nγ)12C) characterized the fast fusion-alpha population evolution. The γ-decay times are consistent with classical alpha plus parent fast triton slowing down times (τTs + ταs) for high plasma currents (Ip > 2 MA) and monotonic q-profiles. In CH discharges the γ-ray emission decay times are much lower than classical (τTs+ταs), indicating alpha confinement degradation, due to the orbit losses and particle orbit drift predicted by a 3-D Fokker–Planck numerical code and modelled using TRANSP.

ICRF heating in volumetric neutron source JUST-T for transmutation of minor actinides

Plasma heating by waves in the cyclotron frequency range (ICRF) in parallel with neutral beams is proposed for the volumetric neutron source based on the JUST-T tokamak project. Calculations of the transverse refractive index ( ) were performed for slow waves (SW) and fast magnetosonic waves (FW) versus the plasma radius. The coefficients of SW and FW decay for ICRH at the harmonics of the ion-cyclotron frequency and for ions and electrons were calculated. The calculations were performed for three DT mixtures: (a) deuterium minority in tritium plasmas; (b) n D∼ n T; and (c) tritium minority in deuterium plasmas. The calculations were also performed for 3He minority in DT plasmas with comparable D and T concentrations. The effective formation of travelling fast waves in VNS JUST-T by phasing several loop antennas located some distance away in the toroidal direction to provide current drive is investigated.

Simulation of multi-atomic interactions in H–O–W system with the MD code CADAC

For future tokamak reactors, chemical erosion of tungsten armour surfaces under impact of hot deuterium–tritium plasma that contains impurities, for instance oxygen, is an important issue. Oxygen can form volatile molecular complexes O x W y at the surface, and the retained H-atoms form the volatile complexes H x O y , which mitigates the erosion (H states for hydrogen isotopes). The plasma impact can substantially destroy the complexes. To describe this H–O–W system, the molecular dynamics (MD) code CADAC was earlier developed using only pair–atomic interactions. Now CADAC is extended for multi-body forces to simulate molecular organization of atoms near the tungsten surface. The approach uses the Abell's model of empirical bond-order potentials in addition combined, for the first time, with a valence concept. CADAC simulates chemical features using atomic valences and the Morse potentials. The new model is introduced and model parameters are estimated.

Radioactive waste management and safety for the Ignitor fusion experiment

Ignitor is a proposed compact high-magnetic field tokamak aimed at studying plasma burning conditions in Deuterium–Tritium plasmas. The determination to carry out this project in Italy has driven to taking into account several candidate sites. Recently, due to several technical reasons, it was identified a new site, the ENEL–TERNA plant in Rondissone (TO). This paper deals with two main aspects of the Ignitor safety: accident analysis with dose assessment to personnel and public, and radioactive waste management evaluations. Concerning the first point, accidents with a frequency higher than 10 −7 per year bring to a maximum dose for the most exposed individual (MEI) below 1 mSv. We have applied the Italian waste management regulations to the Ignitor experiment radioactive materials: none of them should be classified in the high level waste (HLW) category and most materials are classified as low level waste (LLW). If European proposed waste management strategies were applied, all Ignitor radioactive materials could be recycled.

Detritiation of the JET Carbon Tiles by Flame Heating

Operation with tritium plasma led to contamination of the JET in-vessel components with tritium at a level exceeding 12kBq/g, which is the Low Level Waste (LLW) threshold in the UK. Carbon tiles used at JET for protecting the pumped divertor and inner wall against heat and neutron flux create one of the Intermediate Level Waste (ILW) streams to deal with during the JET decommissioning in the future. To reduce quantity and cost of ILW disposal from JET, the study has been initiated for development of detritiation techniques. This paper presents a brief description of the study of the JET carbon tiles detritiation using flame heating.

Analysis and optimization of the impact of ferromagnetic inserts on the toroidal field ripple in the International Thermonuclear Experimental Reactor

Perturbations of the axial symmetry of a toroidal magnetic field, or the toroidal field (TF) ripple, in tokamaks can potentially lead to losses of high-energy particles from the plasma. These losses can be significant and are one of the major concerns with respect to both the energy blance of the deuterium–tritium plasma and the heat load limits in the International Thermonuclear Experimental Reactor (ITER). In the ITER design, ferromagnetic inserts are used to reduce the TF ripple. In this work the impact of the ferromagnetic inserts on the TF ripple in the ITER is analysed to optimize their filling factor within the design constraints. The analysis of the high-mode spectrum of the amplitudes of ripple high modes is also carried out.

Ion temperature and plasma rotation profile effects in the neutron emission spectrum

The instrumental factors and measuring conditions affecting neutron emission spectrometry measurements of tokamak plasmas are described and analyzed. The measured energy broadening and shift of the neutron emission is used to deduce ion temperature (Ti) and toroidal plasma rotation velocity (Vt) representing average (effective) values for the nonuniform plasma volume viewed. Analytical expressions are derived for the relationship between the line-volume integrated effective temperature (Teff) and the radial profile Ti(r) for the case of thermal plasmas with isotropic neutron emission; effects on Teff due to spectral broadening from the radial dependence Vt(r) were also considered. The analysis method presented here is applied to high quality data obtained with the magnetic proton recoil neutron spectrometer installed at Joint European Torus for measurements of deuterium–tritium plasmas. Similarly, cases of anisotropic neutron emission were quantitatively assessed.

The TOFOR spectrometer for 2.5 MeV neutron measurements at JET

Neutron emission spectroscopy has been demonstrated to be a powerful plasma diagnostic at tokamaks. This was shown with the magnetic proton recoil spectrometer developed for measurement of the 14 MeV neutron emission from deuterium–tritium (DT) plasmas at the Joint European Torus (JET). For diagnosis of D plasmas, a 2.5 MeV spectrometer is needed with a factor of 100 higher efficiency to maintain the count rate because of the lower neutron emission rate. The time-of-flight (TOF) technique has the right attributes for these measurements. However, previous instruments have not achieved the full potential of the technique, especially, with respect to count rate capability. A TOF spectrometer for optimized rate (TOFOR) has been conceptually defined and is now under design and development for construction and use at JET. The TOFOR design goal is a count rate capability of about 300 kHz which should be attainable in high power D discharges at JET. The TOFOR project is now in a research and development stage where the instrument is being characterized through neutron transport calculations and test experiments. The tests aim to determine the response of the plastic scintillator elements from which TOFOR will be constructed. This contribution will present the TOFOR design principles, the results of the simulations and test measurements, and the steps taken to reach the optimized design.

Tritium recovery from co-deposited layers using 193-nm laser

Recovery of tritium from co-deposited layers formed in deuterium–tritium plasma operations of the TFTR (Tokamak Fusion Test Reactor) was investigated by the use of an ArF excimer laser operating at the wavelength of 193 nm. At the laser energy density of 0.1 J/cm2, a transient spike of the tritium-release rate was observed at initial irradiation. Hydrogen isotopes were released in the form of hydrogen-isotope molecules during the laser irradiation in vacuum, suggesting that tritium can be recovered readily from the released gases. In a second experiment, hydrogen (tritium) recovery from the co-deposited layers on JT-60 tiles that had experienced hydrogen-plasma operations was investigated by laser ablation with a focused beam of the excimer laser. The removal rate of the co-deposited layers was quite low when the laser energy density was smaller than the ablation threshold (1.0 J/cm2), but reached 1.1 μm/pulse at the laser energy density of 7.6 J/cm2. The effective absorption coefficient in the co-deposited layers at the laser wavelength was determined to be 1.9 μm-1. The temperature of the surface during the irradiation at the laser energy density of 0.5 J/cm2 was measured on the basis of Planck’s law of radiation, and the maximum temperature during the irradiation decreased from 3570 K at the initial irradiation to 2550 K at the 1000th pulse of the irradiation.

Prediction/modelling of the neutron emission from JET discharges

The neutron emission from the JET tokamak is investigated using an extensive set of diagnostics, permitting the instantaneous neutron yield, the radial profile of the neutron emission and neutron energy spectra to be studied. Apart from their importance as an immediate indication of plasma fusion performance, the customary use for neutron measurements is as a test of the internal consistency of the non-neutron diagnostic data, from which the expected neutron production can be predicted. However, because contours of equal neutron emissivity are not necessarily coincident with magnetic flux surfaces, a fully satisfactory numerical analysis requires the application of highly complex transport codes such as TRANSP. In this paper, a far simpler approach is adopted wherein the neutron emission spatial profiles are used to define the plasma geometry. A two-volume model is used, with a core volume that encompasses about (2/3) of the neutron emission and the peripheral volume the remainder. The overall approach provides an interpretation of the measured neutron data, for both deuterium and deuterium–tritium (D–T) plasma discharges, that are as accurate as the basic non-nuclear plasma data warrant. The model includes the empirical assumption that particles, along with their energies and momenta, are transported macroscopically in accordance with classical conservation laws. This first-order estimate of cross-field transport (which, for D–T plasmas, determines the D : T fuel concentration ratio in the plasma core) is fine-tuned to reproduce the experimental ion and electron temperature data. The success of this model demonstrates that the observed plasma rotation rates, temperatures and the resulting neutron emission can be broadly explained in terms of macroscopic transport.

Dedicated Facility and Glovebox for Tritium Recovery from Carbon Flakes

ABSTRACTThe facility dedicated to full tritium recovery from carbon flakes and dust generated in the JET fusion reactor during operation with tritium plasma has been designed and is being constructed at the JET Active Gas Handling plant. This facility will also be used for the study of a correlation between the tritium inventory in carbon flakes/dust and tiles exposed to tritium plasma and the rate of tritium out-gassing from them.

Neutron spectrometry for D–T plasmas in JET, using a tandem annular-radiator proton-recoil spectrometer

A selection of the 14-MeV neutron spectra obtained at the JET Joint Undertaking tokamak during the deuterium–tritium operating campaign in 1997 are presented and analyzed. While several neutron spectrometers were operational during this campaign, the present paper is concerned solely with one: the tandem annular-radiator proton-recoil spectrometer (or proton recoil telescope, for brevity). During neutral beam heating with combined d- and t-beams, analysis of the spectra can define the core fuel composition (D:T) ratio. The spectra are sensitive to the population balance of the fast ions streaming in directions parallel and opposite to that of the injected beams. During ICRF heating of minority deuterium in bulk tritium plasmas, the spectra provide measurements of the effective temperature of the fast-deuteron energy tail and of its relative strength, which vary with the deuterium concentration. This information contributes to the overall understanding of the fusion performance of the various operating scenarios.