Tritium Plasma Research Articles

Reactor relevant ICRF heating in JET DT plasmas

Ion cyclotron resonance heating experiments have been carried out in reactor relevant scenarios during the deuterium-tritium (DTE1) campaign in JET. For the first time it was possible to assess the scheme with a deuterium minority in tritium plasmas: this produced a steady state DT fusion power of up to 1.66 MW for input ICRF power of 6 MW. Strong ion heating, with a core ion temperature of up to 13 keV, was observed when using 3He minority heating in both 50:50 DT and tritium dominated H mode plasmas. Second harmonic tritium heating was also studied and, as expected for JET plasma conditions, produced mainly electron heating. Finally, the `inverted' scenario of tritium minority in deuterium was successfully demonstrated.

ITER: an integrated approcah to ignited plasmas

The programmatic objective of the International Thermonuclear Experimental Reactor (ITER) calls for it to demonstrate controlled ignition and extended burn of deuterium–tritium plasmas, to demonstrate essential fusion reactor technologies in an integrated system, and to perform integrated testing of high–heat flux and nuclear components. The ITER design, as embodied in the Final Design Report, and its physics basis are presented and the projections of plasma performance are summarized together with the assessment of ITER9s engineering feasibility and of the progress in validating technology R and D. The future prospects for ITER and its potential central role as an integrated physics and engineering experiment in the overall development of controlled fusion as a source of useful energy are discussed.

Bulk ion heating with ICRH in JET DT plasmas

Reactor relevant ICRH scenarios have been assessed during DT experiments on the JETtokamak using H mode divertor discharges with ITER-like shapes and safety factors. Deuteriumminority heating in tritium plasmas was demonstrated for the first time. For 9% deuterium, anICRH power of 6 MW gave 1.66 MW of fusion power from reactions between suprathermaldeuterons and thermal tritons. The Q value of the steady state discharge reached 0.22 for thelength of the RF flat-top (2.7 s), corresponding to three plasma energy replacement times. TheDoppler broadened neutron spectrum showed a deuteron energy of 125 keV, which was optimumfor fusion and close to the critical energy. Thus, strong bulk ion heating was obtained at the sametime as high fusion efficiency. Deuterium fractions around 20% produced the strongest ionheating together with a strong reduction of the suprathermal deuteron tail. The ELMs had low amplitude and high frequency and each ELM transported less plasmaenergy content than the 1% required by ITER. The energy confinement time, on the ITERH97-Pscale, was 0.90, which is sufficient for ignition in ITER. 3He minority heating, in approximately50:50 D:T plasmas with up to 10% 3He, also demonstrated strong bulk ion heating. Central iontemperatures up to 13 keV were achieved, together with central electron temperatures up to 12 keV. The normalized H mode confinement time was 0.95. Second harmonic tritium heatingproduced energetic tritons above the critical energy. This scheme heats the electrons in JET,unlike in ITER where the lower power density will allow mainly ion heating. The invertedscenario of tritium minority ICRH in a deuterium plasma was demonstrated as a successfulheating method producing both suprathermal neutrons and bulk ion heating. Theoreticalcalculations of the DT reactivity mostly give excellent agreement with the measured reactionrates.

Tritium retention in tungsten exposed to intense fluxes of 100 eV tritons

Tungsten is a candidate material for the International Thermonuclear Experimental Reactor (ITER) as well as other future magnetic fusion energy devices. Tungsten is well suited for certain fusion applications in that it has a high threshold for sputtering as well as a very high melting point. As with all materials to be used on the inside of a tokamak or similar device, there is a need to know the behavior of hydrogen isotopes embedded in the material. With this need in mind, the Tritium Plasma Experiment (TPE) has been used to examine the retention of tritium in tungsten exposed to very high fluxes of 100 eV tritons. Both tungsten and tungsten containing 1% lanthanum oxide were used in these experiments. Measurements were performed over the temperature range of 423–973 K. After exposure to the tritium plasma, the samples were transferred to an outgassing system containing an ionization chamber for detection of the released tritium. The samples were outgassed using linear ramps from room temperature up to 1473 K. Unlike most other materials exposed to energetic tritium, the tritium retention in tungsten reaches a maximum at intermediate temperatures with low retention at both high and low temperatures. For the very high triton fluences used (>10 25 T/m 2), the fractional retention of the tritium was below 0.02% of the incident particles. This report presents not only the results of the tritium retention, but also includes the modeling of the results and the implication for ITER and other future fusion devices where tungsten is used.

Fast particles and the edge transport barrier

The role of fast particles in edge transport barrier formation is discussed. Analysis of recent experiments on JET that have been carried out in hydrogen, deuterium and tritium plasmas, and in a DT mixture, with NBI and ICRF heating is presented which supports the idea that the width of the edge transport barrier is controlled by the banana width of the fast beam ions in the case of NBI heating and by the banana width of the thermal ions in the case of ICRF minority heating. A simple theoretical model which can account for this effect is analysed.

Isotope scaling of the H mode power threshold on JET

Results are presented from a series of dedicated experiments carried out on JET in tritium, DT, deuterium and hydrogen plasmas to determine the dependence of the H mode power threshold on the plasma isotopic mass. The Pthr ∝ Aeff-1 scaling is established over the whole isotopic range. This result makes it possible for a fusion reactor with a 50:50 DT mixture to access the H mode regime with about 20% less power than that needed in a DD mixture. Results on the first systematic measurements of the power necessary for the transition of the plasma to the type I ELM regime, which occurs after the transition to H mode, are also in agreement with the Aeff-1 scaling. For a subset of discharges, measurements of Te and Ti at the top of the profile pedestal have been obtained, indicating a weak influence of the isotopic mass on the critical edge temperature thought to be necessary for the H mode transition.

Modelling of plasma tritium concentration and wall tritium inventory at JET

In this paper a method for the calculation of the wall particle inventory in a Tokamak is proposed and applied for a series of discharges carried out during the recent Joint European Torus (JET) D–T experiments. These results are subsequently used to validate a new model for the calculation and prediction of the wall particle inventory on the pulse by pulse time scale. This model, known as the `Wall Retention' model, includes the effect of codeposition and is able to reproduce both the total and the tritium wall inventory if it is assumed that between 0.8% to 2% of the ion flux to the divertor is codeposited. It is demonstrated that the Wall Retention model can be used predictively while maintaining good agreement with the experiments. Extrapolations to ITER give a tritium retention per pulse of 54 g.

Overview of ITER physics deuterium-tritium experiments in JET

An overview of JET experimental results in DT plasmas directly relevant to ITER modes of operation is presented. Experiments in D:T mixtures varying from 100:0 to 10:90 and those carried out in hydrogen plasmas show that the H mode threshold power has an approximately inverse isotope mass dependence. Matching some of the key dimensionless parameters to the ITER values, the ITER similarity experiments with ITER shape and safety factor q show that the global energy confinement time is practically independent of isotopic mass (∼A0.03±0.08), where A is the atomic mass of the hydrogenic species. Subtracting the edge pedestal energy (which scales as ∼A0.57±0.2) from the total stored energy leads to a ∼A-0.17±0.1 dependence of confinement in the plasma core, very similar to that expected from the gyro-Bohm transport (∼ A-0.2) model. The observed scaling of the edge pedestal energy is consistent with a model in which the edge pressure gradient saturates at the ballooning limit over a region of width that scales as the ion poloidal Larmor radius governed by the average energy of the fast ions in the edge. The steady state total stored energy for a given input power in both ICRH and NBI discharges is the same despite the lower edge pedestal in the ICRH case, which is compensated for by more peaked power deposition profiles in ICRH. The ELM frequency is smaller with NBI; it decreases with isotopic mass in both NBI and ICRH discharges. A steady state, type I ELMy H mode discharge with ITER shape and q at 3.8 T/3.8 MA with an input power of 22 MW produced a Q ≈ 0.18 for 3.5 s and extrapolates well to ignition with ITER parameters. Here, Q is the ratio of fusion output power to input power. The thermal ELMy H mode confinement in both deuterium and tritium gas fuelled plasmas decreases significantly when the plasma density exceeds 0.75 of the Greenwald (nGW) limit, and the maximum density achieved is 0.85nGW. In L mode, the density limit decreases with increasing isotope mass roughly in accordance with code predictions. ITER reference ICRH scenarios have been evaluated. Second harmonic heating of tritium at the densities available in JET produces strong tails and heats electrons predominantly as expected. The 3He minority in 50:50 D:T and tritium dominated plasmas showed strong bulk ion heating leading to ion temperatures up to 13 keV with ICRH alone. Deuterium minority ion cyclotron heating in tritium plasmas at a power level of 6 MW produced steady state record values of Q ≈ 0.22 for more than 2.5 s.

Deuterium-tritium operation in magnetic confinement experiments: results and underlying physics

A review of experimental results obtained in JET D-T plasmas is presented. In discussing the underlying physics, results previously obtained on tokamak fusion test reactor (TFTR) are also taken into account. In JET, the maximum fusion power output of 16.1 MW has been obtained in an edge localized mode (ELM)-free hot-ion H-mode featuring an edge confinement barrier in a single-null divertor plasma with a where is the total input power to torus. A steady-state H-mode discharge, with plasma shape and safety factor q similar to that of ITER, produced 4 MW for 5 s (22 MJ). The steady-state results extrapolate well to ignition with ITER parameters using the normalized plasma pressure achieved on JET. Also, the advanced tokamak regime using optimized magnetic shear configuration featuring an internal transport barrier produced 8.2 MW of fusion power. With regards to reactor physics issues, a clear identification of electron heating by fusion born -particles has been made both in JET and TFTR. The JET experiments show that the H-mode threshold power has approximately an inverse isotopic mass dependence and that it does not depend on the method of auxiliary heating. The global energy confinement time in the TFTR D-T supershot regime scales as but in the JET H-modes, it is found to be practically independent of isotopic mass where A is the atomic mass of the hydrogenic species. In JET, the plasma core and the edge appear to have different underlying confinement physics, the former follows the gyro-Bohm transport model whereas the edge pedestal energy scales as . The maximum edge pressure in H-modes is analysed in relation to the ion poloidal Larmor radius at the edge. The fast ions driven by neutral beam injections (NBI) or ion cyclotron resonance heating (ICRH) could play an important role in setting the width of the edge pedestal. The thermal ELMy H-mode confinement both in D or T gas fuelled plasmas decreases significantly when the plasma density exceeds 0.75 of the Greenwald limit and the maximum density achieved is . The ICRH scenarios for a reactor have been evaluated. For example, minority in 50:50 D:T and tritium-dominated plasmas showed strong bulk ion heating leading to ion temperatures up to 13 keV with ICRH alone. Deuterium-minority ion cyclotron heating in tritium plasmas at a power level of 6 MW produced steady-state record values of for more than 2.5 s. Finally, the on-site closed-cycle tritium reprocessing plant and remote handling tools at JET have been used routinely and provided an integrated demonstration of safe and reliable operations of a tokamak device in reactor-relevant conditions.

H-mode power threshold and confinement in JET H, D, D-T and T plasmas

The isotope dependence of the H-mode power threshold and the energy confinement time have been examined in dedicated experiments on the Joint European Torus (JET) in hydrogen, deuterium and tritium plasmas. The results show that: (1) The H-mode power threshold is inversely proportional to the effective isotopic mass, M, of the plasma. (2) The edge electron pedestal temperature at the L to H-mode transition scales as . (3) The global thermal energy confinement time, , does not depend strongly on M, i.e. in the ohmic, L-mode, edge localized modes (ELMy) and ELM-free H-mode confinement regimes with , if the engineering parameters such as current, magnetic field, density, power and geometry are kept fixed. (4) In the ELMy H-mode regime the transport in the core and the edge of the plasma scale differently with M, i.e. the core transport increases weakly with M whereas the edge transport decreases strongly with M.

A 14 MeV neutron spectrometer for the Joint European Torus deuterium-tritium experiments

A 14 MeV neutron spectrometer, based on the principle of the proton recoil telescope, has been installed as a deuterium–tritium (DT) plasma diagnostic at the Joint European Torus. The device comprises three identical telescope modules arranged one behind the other, to increase efficiency without worsening resolution. It views the plasma along a horizontal line of sight. The total efficiency is 7.05×10−5 counts per (neutron per cm2), and the calculated response function has a full width at half maximum of 2.2% at an incident neutron energy of 14.1 MeV. The spectrometer was operational throughout the recent DT experiment No. 1, and selected neutron spectra have been analyzed in terms of the contributions from the various neutron production mechanisms.

Neutron spectrometry of radio-frequency heated deuterium–tritium plasmas

Spectrometry of the neutron emission has been used to probe the nonthermal features of the fuel ion velocity distributions in tokamak plasmas, especially, those arising from auxiliary heating. The first time resolved measurements are reported from the use of the new magnetic proton recoil (MPR) spectrometer in observations of deuterium–tritium plasmas at JET with strong ion cyclotron resonance heating. Results from preliminary analysis data are presented and compared with simulated neutron emission spectra based on the assumptions of a thermal plasma (a single Maxwellian ion distribution) or anisotropic Maxwellians with parallel and perpendicular temperatures, beside their relative amplitudes. The diagnostic results attained are also used to assess the diagnostic capabilities and the potential for use of MPR based diagnostics on burning plasma tokamaks.

Pellet charge exchange measurements of confined alphas in deuterium–tritium plasmas (invited) (abstract)

Alpha particle confinement is essential to achieving ignition in deuterium–tritium plasmas, and measuring the energy and spatial distributions of fusion alphas is one of the most challenging tasks in plasma diagnostics research. Confined trapped-alpha energy spectra and radial density profiles in the Tokamak Fusion Test Reactor (TFTR) have been obtained using the pellet charge exchange (PCX) diagnostic, which measures high energy (Eα=0.5–3.5 MeV), trapped alphas (|v∥/v|=0.048) at a single time slice (Δt∼1 ms) with a spatial resolution of Δr∼5 cm. PCX measures the energy spectrum of energetic helium neutrals resulting from charge exchange interactions of alphas incident on the ablation cloud surrounding small boron and lithium pellets injected radially into TFTR. The success of PCX has led to a number of important results on the behavior of alphas in TFTR. The measured alpha energy spectrum in the plasma core of sawtooth-free discharges is consistent with the alphas being well confined and slowing down classically. Outside the plasma core, the trapped alphas show the effects of stochastic diffusion due to the toroidal magnetic field ripple, with the PCX measured profiles consistent with the functional dependence of the stochastic ripple diffusion on the alpha energy and the q profile. Large sawtooth instabilities result in radial redistribution of the trapped alphas to well outside the q=1 radius and beyond the stochastic ripple loss boundary. Broadening of the radial profiles of trapped alphas is also observed in reversed and reduced shear discharges on TFTR, with potential implications for reactor designs based on optimized shear configurations. Finally, radial redistribution of trapped alpha particles in the presence of core localized Toroidal Alfv́en Eigenmode activity is also observed. Application of PCX measurements to future experiments, including the International Thermonuclear Experimental Reactor will also be discussed.

Alpha particle-driven toroidal Alfvén eigenmodes in Tokamak Fusion Test Reactor deuterium–tritium plasmas: Theory and experiments

The toroidal Alfvén eigenmodes (TAE) in the Tokamak Fusion Test Reactor [K. Young et al., Plasma Phys. Controlled Fusion 26, 11 (1984)] deuterium–tritium plasmas are analyzed using the NOVA-K code [C. Z. Cheng, Phys. Rep. 211, 1 (1992)]. The theoretical results are compared with the experimental measurements in detail. In most cases, the theory agrees with the observations in terms of mode frequency, mode structure and mode stability. However, one mode with toroidal mode number n=2 is observed to be poloidally localized on the high field side of the magnetic axis with a mode frequency substantially below the TAE frequency.

Estimation of permeation probability in plasma driven permeation

Hydrogen permeation experiments were carried out with different thickness of Ni membranes; 20, 50 and 200 μm, pre-conditioned for the same surface conditions. Hydrogen plasma was created by ECR heating and plasma driven permeation (PDP) through the Ni membrane of 523 K was investigated. With gas driven permeation (GDP) experimental data, recombination coefficients of the upstream and the downstream surfaces (m 4 s −1) were calculated; 1.6×10 −29≤K l, 7.9×10 −30<K 2≤1.6×10 −29. The permeation rate and its probability were estimated by changing K 1 as a parameter. About 20% of tritium permeation probability; equivalent to≈2 Ci m 2 s −1, was expected when low energy (≈eV) tritium plasma was discharged at 1.33 Pa and 10 μm Ni membrane of 523 K, pre-conditioned for reducing K 1 by two orders of magnitude, was employed as a tritium membrane pump.

ITER safety and operational scenario

The safety and environmental characteristics of ITER and its operational scenario are described. Fusion has built-in safety characteristics without depending on layers of safety protection systems. Safety considerations are integrated in the design by making use of the intrinsic safety characteristics of fusion adequate to the moderate hazard inventories. In addition to this, a systematic nuclear safety approach has been applied to the design of ITER. The safety assessment of the design shows how ITER will safely accommodate uncertainties, flexibility of plasma operations, and experimental components, which is fundamental in ITER, the first experimental fusion reactor. The operation of ITER will progress step by step from hydrogen plasma operation with low plasma current, low magnetic field, short pulse and low duty factor without fusion power to deuterium–tritium plasma operation with full plasma current, full magnetic field, long pulse and high duty factor with full fusion power. In each step, characteristics of plasma and optimization of plasma operation will be studied which will significantly reduce uncertainties and frequency/severity of plasma transient events in the next step. This approach enhances reliability of ITER operation.

D-T Fusion with Ion Cyclotron Resonance Heating in the JET Tokamak

Ion cyclotron resonance heating (ICRH) experiments have been carried out in JET D-T plasmas using scenarios applicable to reactors. Deuterium minority heating in tritium plasmas is used for the first time and produces 1.66 MW of D-T fusion power for an ICRH power of 6 MW. The Q value is 0.22, which is a record for steady state discharges. Fundamental He3 minority ICRH, in both 50:50 D-T and tritium dominated plasmas, generates strong bulk ion heating and ion temperatures up to 13 keV. Second harmonic tritium ICRH is seen to heat mainly the electrons as expected for JET conditions. All three schemes produce H−mode plasmas.Received 23 December 1997DOI:https://doi.org/10.1103/PhysRevLett.80.4681©1998 American Physical Society

Ion cyclotron range of frequencies heating and flow generation in deuterium–tritium plasmas

Recent radio-frequency heating experiments on the Tokamak Fusion Test Reactor (TFTR) [Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] have focused on developing tools for both pressure and current profile control in deuterium–tritium (DT) plasmas. A new antenna was added to investigate pressure profile control utilizing direct ion Bernstein wave (IBW) heating. This was the first time direct IBW heating was explored on TFTR. Plasma heating and driven poloidal flows are observed. Previously heating and current drive via mode-converted IBW waves had been demonstrated in non-DT plasmas but efforts in DT plasmas had been unsuccessful. This lack of success had been ascribed to the presence of a small Li7 minority ion population. In the most recent experiments Li6 was used exclusively for machine conditioning and mode-conversion heating consistent with theory is now observed in DT plasmas.

Toroidal Alfvén eigenmodes in TFTR deuterium–tritium plasmas

Purely alpha-particle-driven toroidal Alfvén eigenmodes (TAEs) with toroidal mode numbers n=1–6 have been observed in deuterium–tritium (D–T) plasmas on the tokamak fusion test reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)]. The appearance of mode activity following termination of neutral beam injection in plasmas with q(0)&amp;gt;1 is generally consistent with theoretical predictions of TAE stability [G. Y. Fu et al. Phys. Plasmas 3, 4036 (1996)]. Internal reflectometer measurements of TAE activity is compared with theoretical calculations of the radial mode structure. Core localization of the modes to the region of reduced central magnetic shear is confirmed, however the mode structure can deviate significantly from theoretical estimates. The peak measured TAE amplitude of δn/n∼10−4 at r/a∼0.3−0.4 corresponds to δB/B∼10−5, while δB/B∼10−8 is measured at the plasma edge. Enhanced alpha particle loss associated with TAE activity has not been observed.

Modeling of materials mixing behavior among beryllium, carbon and tungsten as plasma-facing materials under deuterium and tritium plasma bombardment

If in-vessel components in a magnetic fusion device employ two or more different plasma-facing materials, materials mixing can occur due to the impurity transport along with erosion and redeposition. Materials mixing alters surface characteristics, which in turn affects plasma interactions behavior with in-vessel components. Currently, the use of beryllium, carbon and tungsten as plasma-facing materials is under consideration for in-vessel components for the International Thermonuclear Experimental Reactor (ITER). However, there is no standard model applicable to analyze the materials mixing behavior. In this study a zero-dimensional materials balance model has been applied to analyze the behavior of deposition and removal of plasma impurities over in-vessel components under deuterium and tritium (DT) particles bombardment for the purpose of evaluating the possibility of materials mixing via impurity deposition. It is found that materials mixing between beryllium and carbon is likely whereas tungsten tends to remain without beryllium and carbon impurity deposition. In contrast, the model predicts that beryllium and carbon plasma-facing components will be highly susceptible to the deposition of tungsten impurities.