Hot-ion H-mode Research Articles

Alpha heating, isotopic mass, and fast ion effects in deuterium–tritium experiments

Alpha heating experiments in the Tokamak Fusion Test Reactor (TFTR) and in the Joint European Torus (JET) 1997 DTE1 campaign are reexamined. In TFTR supershots central electron heating of both deuterium only and deuterium–tritium supershots was dominated by thermal ion-electron heat transfer rate . The higher in deuterium–tritium supershots was mainly due to higher largely caused by isotopic mass effects of neutral beam—thermal ion heating. The thermal ion–electron heating dominated the electron heating in the center. The ratio of the predicted alpha to total electron heating rates is less than 0.30. Thus alpha heating (and possible favorable isotopic mass scaling of the thermal plasma) were too small to be measured reliably.The JET alpha heating Hot-Ion H-mode discharges had lower , and thus had lower and the deuterium–tritium DT discharges had higher than in TFTR. There were not enough comparable discharges to verify alpha heating. The high performance phases consisted of rampup to brief flattop durations. At equal times during the rampup phase central and were linearly correlated with the thermal hydrogenic isotopic mass which co-varied with beam ion pressure, the tritium fraction of neutral beam power, and the time delay to the first significant sawteeth which interrupted the increases.For both devices the expected alpha heating rate and the null hypothesis of no alpha heating are consistent with the measurements within the measurement and modeling uncertainties.

Fusion Nuclear Science Facility (FNSF) before Upgrade to Component Test Facility (CTF)

The compact (R0~1.2-1.3m) Fusion Nuclear Science Facility (FNSF) is aimed at providing a fully integrated, continuously driven fusion nuclear environment of copious fusion neutrons. This facility would be used to test, discover, and understand the complex challenges of fusion plasma material interactions, nuclear material interactions, tritium fuel management, and power extraction. Such a facility properly designed would provide, initially at the JET-level plasma pressure (~30%T2) and conditions (e.g., Hot-Ion H-Mode, Q<1)), an outboard fusion neutron flux of 0.25 MW/m2 while requiring a fusion power of ~19 MW. If and when this research is successful, its performance can be extended to 1 MW/m2 and ~76 MW by reaching for twice the JET plasma pressure and Q. High-safety factor q and moderate-plasmas are used to minimize or eliminate plasma-induced disruptions, to deliver reliably a neutron fluence of 1 MW-yr/m2 and a duty factor of 10% presently anticipated for the FNS research. Success of this research will depend on achieving time-efficient installation and replacement of all internal components using remote handling (RH). This in turn requires modular designs for the internal components, including the single-turn toroidal field coil center-post. These device goals would further dictate placement of support structures and vacuum weld seals behind the internal and shielding components. If these goals could be achieved, the FNSF would further provide a ready upgrade path to the Component Test Facility (CTF), which would aim to test, for ≤6 MW-yr/m2 and 30% duty cycle, the demanding fusion nuclear engineering and technologies for DEMO. This FNSF-CTF would thereby complement the ITER Program, and support and help mitigate the risks of an aggressive world fusion DEMO R&D Program. The key physics and technology research needed in the next decade to manage the potential risks of this FNSF are identified.

Neutral particle measurements of fusion tritons in JET

A neutral particle analyzer [A.D. Izvozchikov et al., JET Report No. JET-R(91)-12, 1991] operating in the MeV energy range was used to measure the flux of neutralized d-d fusion tritons emitted from the hot-ion H-mode deuterium plasma heated by deuterium neutral beams. It was found that tritons in the energy range of 0.3–1.1 MeV were largely neutralized by the beam atoms and the beam halo atoms. This enabled us to find the localized energy distribution function of the fusion tritons in the central plasma region. Simulation of the triton energy distribution function shows that MeV ions in the JET hot-ion H-mode plasma behave classically.

Experimental study of electron heat transport in ion heated H-modes in ASDEX Upgrade

The electron heat transport in low density H-mode plasmas heated by neutral beam injection (NBI) is investigated in ASDEX Upgrade using electron cyclotron heating (ECH) combining both steady-state and transient response analysis by modulating the ECH power. Under these conditions, more than 60% of the NBI power (>3 MW) is delivered to the ions, while approximately 20% (∼1 MW) is delivered to the electrons. In the confinement region, the electron-to-ion temperature ratio, Te/Ti, varies between 0.5 and 0.7 in the NBI-only phase and between 0.8 and 1.0 when the ECH is also applied. Due to the low collisional coupling, the power in the electron channel is locally more than doubled by applying up to the available 2 MW of ECH, while the power in the ion channel is locally increased by less than 30%. A dependence on the density of the reaction of the plasma parameters to the ECH is observed. For plasmas with average density (defined as ‘hot-ion’ H-modes), when the ECH is applied, Te increases, the central Ti drops and the density flattens. These effects disappear with increasing density and are not observed for (defined as ‘regular’ H-modes). Power balance analysis of both the hot-ion and regular H-modes points to a strong resilient behaviour of the Te profiles. In the hot-ion cases, the ECH heating induces a strong increase in transport in the ion channel. Power balance and transient response analysis of the regular H-modes are consistent with an inverse scale length transport model with a threshold in , above which the electron heat transport is increased. Comparison with recent studies in pure EC heated L-modes points to a stronger resilience of Te in the NBI heated H-modes.

Magnetohydrodynamic Stability of Improved Confinement Plasmas in JT-60U

Progress in the understanding of magnetohydrodynamic (MHD) stability is summarized on JT-60U tokamak discharges with improved confinement such as the (hot-ion) H-mode, high-βp mode, high-βp H-mode, and reversed shear discharges. Transport barriers, which are essential for the improved confinement, play key roles in the local and global MHD stability owing to the local large pressure gradient and the related bootstrap current. Disruptive β limits of these discharges are consistent with theoretical ideal kink-ballooning stability limits with low toroidal mode numbers n. Achievable β limit is improved by broadening of the pressure profile with high plasma internal inductance, plasma shaping, and wall stabilization. Edge localized modes (ELMs) and barrier localized modes (BLMs), which are associated with edge and internal transport barriers, respectively, are analyzed carefully. Resistive interchange modes with n ≤ 3 are excited in the negative shear region in reversed shear discharges with the internal transport barrier and lead to major collapse occasionally through nonlinear coupling with a tearing mode in the positive shear region. MHD characteristics of low m/n (m: poloidal mode number) tearing modes, which are attributed to the neoclassical tearing mode, are investigated. Stabilization of tearing modes and control of sawtooth activity are demonstrated using the fundamental O-mode electron cyclotron wave injection. Resistive wall modes associated with current-driven and pressure-driven low n external kink modes are identified.

Study of d–t neutron energy spectra at JET using natural diamond detectors

Four natural diamond detectors (NDDs) have been used for deuterium–tritium neutron spectrometry and flux monitoring during the 1997 tritium experiments (DTE1) carried out in the Joint European Torus (JET). Neutron energy spectra have been measured with three NDDs for discharge scenarios that included (a) hot ion H-mode studies using combined neutral-beam (NB) and ion cyclotron resonance frequency (ICRF) heating, (b) optimized shear experiments using combined NB and ICRF heating, (c) alpha-particles heating experiments with NB heating only and (d) ICRF heating studies without NB heating. Within the statistical accuracy of the data, the spectra can be adequately represented by Gaussian distributions, whose fwhm values provide effective ion temperatures that characterize the energy distributions of the ions taking part in fusion reactions.

Plasma current dependence of the edge pedestal height inJET ELM-free H-modes

Some models for the suppression of turbulence in the L to Htransition, suggest that the width of the H-mode edge barrieris either proportional or is of the order of the thermal or thefast-ion poloidal Larmor radius. This would require that thewidth of the edge barrier should depend on the plasma current.This dependence has been clearly verified at JET in experimentsdesigned to control the edge MHD stability of ELM-free hot-ionH-mode plasmas. The effects of isotopic mass and theapplicability of several edge barrier models to the hot-ionH-mode plasmas were analysed in (Guo H Y et al 2000 Edgetransport barrier in JET hot-ion H-modes Nucl. Fusion 40 69) using a large database containing both deuterium-onlyand deuterium-tritium plasmas. This database has nowbeen enlarged to include discharges from a plasma shape scan,allowing one to study the dependence of the pedestal height onthe edge shear. In addition, the range of plasma currents wasextended up to 6 MA. It is shown that the edge data are bestdescribed by a model where the edge barrier width is determinedby the fast ions weighted towards the components with largestpoloidal Larmor radii. However, it is not possible toconclusively eliminate the thermal ion model.

The role of recycling and impurity production in JET hot-ion H-modes

The effect of impurity production and recycling on JET hot-ion H-mode discharges is investigated for the MkI and MkII divertors. Analysis of the best of these discharges has been performed with the EDGE2D/NIMBUS codes and the results compared with a wide range of diagnostics. For the very low recycling hot-ion regime, the code predicts that Z eff at the edge (in the upstream SOL) increases with SOL density, and is higher for the MkII than for the MkI divertor. These predictions have been confirmed by the experimental data. The significance of the edge density and edge Z eff for the hot-ion regime is that the loss power has been predicted from neo-classical theory to scale as: P loss ∼ n edge 2 Z eff,edge. It has now been demonstrated that this is consistent with both the MkI and MkII experimental data. It has been observed that the carbon sources in the MkII divertor are increased relative to those in the MkI divertor. These data are compared with the modelled results from the EDGE2D code, taking into account the change in the chemical sputtering yield with the different base temperatures of the MkI (40°C) and MkII (270°C) divertors.

Analysis of bulk ion heating with ICRH in JET high-performance plasmas

Ion cyclotron resonance heating (ICRH) of JET high-performance (optimized shear and hot-ion H-mode) deuterium discharges, characterized by strong neutral beam injection (NBI), high ion and electron temperatures and high deuterium-deuterium fusion reactivity, is analysed using numerical simulations. The main aim is to investigate under which conditions good bulk ion heating can be obtained with ICRH tuned to the fundamental hydrogen resonance, coinciding with the second-harmonic deuterium resonance, near the plasma centre. The analysis has been facilitated by adding NBI source terms in the ion cyclotron range of frequencies (ICRF) code PION, which calculates the time evolution of the ICRF power deposition and the resonating ions distribution functions in a self-consistent way. The results show that damping ICRF power on deuterons results in a more central bulk ion heating profile than minority hydrogen ions, which is an advantage. The highest bulk ion heating fractions of up to 50% of the total ICRF power were obtained by using multiple frequencies, the effects of which have been simulated in detail for the first time. Furthermore, operating with high plasma densities has been found to be beneficial for maximizing the bulk ion heating.

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.

Nonperturbing meaurement of energy distribution function of confined D-T fusion alpha particles in Joint European Torus plasmas using impurity induced neutralization (abstract)

We present measurements of the energy distribution function of fusion α particles in the range 0.8⩽E(MeV)⩽3 in hot-ion H-mode plasmas, using a NPA. Two contributions to the measured atomic flux are distinguished: (i) alpha particles due to direct neutralization effected in double charge–exchange reactions with intrinsic plasma impurities C4+, Be2+, and He0, and (ii) MeV energy deuterons formed by elastic scattering on fusion alpha particles, their neutralization effected by reactions with C5+, Be3+, He1+, and D0. Application of the impurity induced neutralization model to the measurements is presented. Deduced fusion α particle and knock-on deuteron energy distribution functions are compared with Fokker–Planck simulations in which measured time-dependent α-particle source and collisionality are used.

ITG modes with finite-β effects and the upper ηi stability regime

It is shown that the toroidal ion temperature gradient (ITG) mode has an upper ηi (higher ηi) stability regime for experimentally relevant parameter values in addition to the lower regime with the stability threshold at ηi around one ( where Ln and LT1 are the characteristic lengths for the density and ion temperature gradients). The ITG mode is studied with a focus on the upper ηi stability regime and the β dependence (β = plasma pressure/magnetic pressure). The results of a gyrokinetic and a two-fluid model as well as a semilocal approximation are compared. It is shown that the upper stability threshold is very sensitive to and considerably reduced by finite-β effects. It is also sensitive to finite Larmor radius (FLR) effects and to (LB is the characteristic length for the toroidal magnetic field gradient). Predictions and comparisons are made with data from a joint European torus (JET) optimized shear discharge and a JET hot ion H-mode (high-performance mode) discharge.

Results of Recent Deuterium/Tritium Experiments in JET

During 1997, JET carried out a campaign of operation in deuterium/tritium. A total of 99 grams of tritium was admitted to the torus using gas puffing and neutral beam injection. With a site inventory of 20 grams of tritium, this required repeated re-processing of the gas recovered from the torus using the JET active gas handling plant. Around 220 tokamak pulses were carried out with tritium concentrations above 40%, during which a total of 2.5.1020 14 MeV neutrons were produced. Emphasis was placed on re-producing conditions close to those anticipated in the ITER experimental fusion reactor, in particular maintaining dimensionless parameters important in the physics of confinement. The experimental program included high fusion yield hot-ion and optimized shear scenarios in particular for the study of alpha particle physics. Achievements included a maximum fusion power of 16 MW in hot-ion H-mode at a Q of 0.6; first production of DT power (8 MW) in optimized shear; a Q of 0.2 for 5 seconds in an ITER relevant steady state ELMy H-mode at a fusion power of 4 MW; a Q of 0.22 in RF only discharges; and observation of alpha particle heating. Tritium was found to give a marked reduction in the H-mode threshold and an improvement in edge pedestal stability but no change in global confinement. The optimized shear scenario required re-optimization in tritium, only partially achieved. The results are generally consistent with ignition in ITER. Retention of tritium in the torus is much higher than anticipated and tritium recovery during the clean-up campaign was modest. The divertor tiles have since been replaced remotely with no personnel access to the torus. Tritium release and the dose to personnel have been well within the low approved levels.JET has successfully completed this tritium campaign, producing both physics and technical data invaluable to the design of next step devices. The results in particular demonstrate the importance of operations in tritium in reliably predicting the performance of future machines.

Transport properties and predictive modelling of JET H-mode and optimized shear discharges

An empirical model for anomalous transport in tokamak, supplemented by the proper boundary conditions, has been developed and applied to four different types of JET discharges: ELM-free hot-ion H-mode, L-mode, ELMy H-mode and discharges with optimized magnetic shear. The results of the numerical modelling show that the model gives quite reasonable agreement with experiment in a wide range of plasma parameters and operational regimes. Moreover, they reveal some important generic features of plasma transport which are model independent.

Prediction/modelling of the neutron emission from hot-ion H-mode discharges in JET

A simple zero-dimensional computer code has been prepared for predicting the neutron emission in the initial phase of beam-heated hot-ion H-mode discharges in the Joint European Torus (JET). It performs very well for discharges with low particle recycling up to the onset of MHD-related activity, which usually occurs within one second of commencing full power heating. It can also be used to model the entire discharge, including the termination, when fuelling of the plasma volume uniformly with cold neutrals is adopted as a first-order approximation for edge recycling effects. Specifically, given the time traces for the beam heating power, the global neutron emission, average electron density, and signals, the code can satisfactorily reconstruct the core volume rotation rate, ion temperature and electron temperature for the entire duration of the discharge, and it also provides an explanation for the anomalously low burnup of d - d fusion product tritons that is sometimes observed. The code provides insight into the relative intensities of the three contributing reaction mechanisms to the total neutron emission (thermal, beam - thermal and beam - beam reactions) as an aid to interpretation of data from the various neutron diagnostics on JET, including the neutron profile monitor and the neutron spectrometers. Tritium fuelling and beam injection are also implemented, so the code can be used to extrapolate the neutron emission from existing d - d discharges to that expected from similar d - t discharges.

High performance Joint European Torus (JET) plasmas for deuterium–tritium operation with the MkII divertor

Planned experiments in the Joint European Torus [Plasma Physics and Controlled Fusion Research, Proceedings, 13th International Conference, Washington, D.C., 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, p. 27] (JET) with deuterium–tritium (D–T) plasmas require high fusion performance for α-particle heating studies and for investigation of isotope dependence in conditions relevant to the International Thermonuclear Experimental Reactor [Plasma Phys. Controlled Fusion 37, A19 (1995)]. In deuterium plasmas, the highest neutron rates have been obtained in the hot-ion high-confinement mode (H mode) which is ultimately limited by magnetohydrodynamic (MHD) phenomena when the pressure gradient approaches ideal ballooning and kink stability limits in the vicinity of the edge transport barrier. Results are reported confirming the MkII divertor’s increased closure and pumping in this regime, progress in understanding the MHD-related termination is discussed, and the use of ion cyclotron resonance heating (ICRH) in combination with high-power neutral beams to increase the neutron yield is described. In separate experiments internal transport barriers have been established through careful programming of the current ramp and heating waveforms, and neutron emission comparable with the best hot-ion H-modes achieved. Steady-state H-mode discharges exhibiting edge localized modes (ELMs) in reactor-like configurations and conditions have been demonstrated, including cases in which relevant dimensionless parameter values are preserved, ready also for testing in D–T.

Deduction of SOL transport coefficients using 2D modelling for hot-ion ELM-free H-modes in JET

Deduction of SOL transport coefficients using 2D modelling for hot-ion ELM-free H-modes in JET

Deduction of SOL transport coefficients using 2D modelling for hot-ion ELM-free H-modes in JET

Profiles of T e and J sat measured by Langmuir probes at the target plates of the JET MkI divertor and the target plate power loading are modelled via the EDGE2D/NIMBUS codes. The low-density ( n e ∼ 1 × 10 19 m −3 ) OH plasma preceding high-power NBI, as well as the ELM-free high performance H ∗ and high performance rollover H RO phases are considered. Experimentally, the power splitting between ions ( P i ≥ 7 MW) and electrons ( P e < 1 MW) in conjunction with the H-mode transport barrier effects that T e and J sat change little from OH to H ∗ conditions, although P in ∼ 2 → 20 MW and n e can increase up to a factor of three as the H ∗-phase progresses. In modelling χ i is taken as 1 m 2/s. To match the divertor T e - profiles, χ e must be varied from the strike point outwards as ∼ 0.5 → 3 m 2/ s in all phases. Perpendicular particle transport is assumed to be purely diffusive or with a pinch. Duplication of J sat ( R) requires: D ⊥ ∼ 0.03 → 0.015 m 2/s (OH → H ∗) , or ν pinch D ⊥ ∼ 15 → 45 → 25 m −1 (OH → H ∗ → H RO) with D ⊥ = 0.1 m 2/ s. Sensitivity studies involving different recycling scenarios, deep and shallow computational grids, wall material and the thermal transport barrier are carried out to judge the possible influences on the deduced transport coefficients.

JET results with the new pumped divertor and implications for ITER

This paper presents an overview of results of the 1994/95 experimental campaign on JET with the new pumped divertor and draws implications for ITER in the areas of detached and radiative divertor plasmas, the use of beryllium as a divertor target tile material, the confinement properties of discharges with the same dimensionless parameters (except for the dimensionless Larmor radius) as ITER and the effect of varying the toroidal magnetic field ripple in the ITER relevant range. Discharges with high fusion performance at high current, in steady-state with ELMs and in the ELM-free hot-ion H-mode are also reported. Limits to operations are discussed and projections to D-T performance are made.

Comparison between measured scrape-off layer plasma parameters and 2D model calculations for JET X-point discharges

Modelling with the EDGE2D/U-NIMBUS code of the measured scrape-off layer plasma parameters for JET divertor discharges is presented. Model results agree with the experiment if very small values (≲ 0.1 m 2/s in H-modes and low power L-modes) for the anomalous particle transport diffusion coefficient in the SOL are used. While the assumption of the power flow out of the main plasma being carried equally by the electrons and the ions describes satisfactorily the global power balance for Ohmic and L-mode discharges, more power flowing out through the ion channel is consistent with the power balance in hot ion H-modes. Some sensitivity studies of the code results on the modelling hypothesis are discussed.