Deuterium Gas Puffing Research Articles

The effect of gas injection location on a lithium vapor box divertor in NSTX-U

The lithium vapor box divertor is a proposed divertor design that will minimize contamination of the upstream plasma in a fusion device, while also ensuring protection of the target. In this design lithium is evaporated near the target by high temperature lithium surfaces, dissipating the plasma heat flux. The lithium vapor box has been predicted via the fluid-kinetic code scrape off layer plasma simulator (SOLPS-ITER) to achieve low (/0.05) upstream concentrations of lithium and low target heat fluxes. Here we compare two choices of deuterium gas puff location using SOLPS-ITER, the private flux region (PFR) and the common flux region (CFR), and find significant differences in the contamination level required to reach an acceptable target heat flux (defined here as q10 MW m−2). Deuterium gas puffing from the PFR is seen to better reduce upstream lithium contamination. The difference in puffing location is seen to cause changes in the upstream flow of lithium ions. The PFR puff, having better access to the separatrix, can better reduce the upstream-directed flow of lithium near the separatrix which is the primary source of contamination due to a large thermal force in this region. Puffing from the CFR, partially due to inefficacy at reducing separatrix lithium flow, has higher lithium concentration within the plasma. Solutions that reduce the heat flux to below 10 MW m−2 have a range of lithium concentrations between /0.01–0.12 depending on puff intensity, location, evaporator temperature and recycling at the various plasma facing components. The efficacy of the puffs is tested for sensitivity to deuterium recycling coefficient at the target, evaporators, and main chamber walls. A CFR located puff is found to be more dependent on the recycling coefficients used than a PFR located puff. regardless of the set of recycling coefficients chosen, PFR puffing achieves lower lithium contamination than CFR puffing for a given heat flux.

Predictive modeling of the deuterium gas puffing effect on impurity and heat flux compression for the HL-2M divertor by SOLPS

To study the effect of gas puffing on divertor operation regime and impurity level in the SOL region, we introduce deuterium gas puffing and pumping, based on the HL-2M lower single null (LSN) divertor configuration (Ip=2MA), using the SOLPS5.0 code. The simulation result indicates that the location of the gas puffing port impacts the divertor plasma condition, and puffing deuterium from the location near the divertor region can enhance the power dissipation and the divertor regime transformation from high recycling region to detachment. Besides, increasing deuterium gas puffing rate can strengthen the core confinement. Meanwhile, the impurity level in the SOL region is compressed rapidly, from Zeff=2.0 to 1.1, with a reasonable deuterium gas puffing rate of 2.5×1022s−1 puffed from the optimized port and a constant chemical sputtering yield Ychem of 1.0%, compared to the case with no gas puffing. The physics interpretation behind this phenomenon is also analyzed. It is demonstrated that the ionization rate distribution and the carbon ions’ poloidal velocity define the impurity flow in the SOL region. The higher ionized carbon ions (C4+ to C6+) retain in the ionization region and even flow downstream when they finish ionization before reaching the stagnation point with gas puffing. Then the impurity ions accumulate in the divertor region. The balance between the thermal and friction forces acting on the carbon ions results in the change of poloidal velocity. In general, deuterium gas puffing and pumping induces the modification of impurity flow and the decline of the Zeff. Furthermore, a chemical sputtering yield of 3.0% is also calculated to verify the efficiency of deuterium gas puffing on Zeff compression.

Systematic extraction of a control-oriented model from perturbative experiments and SOLPS-ITER for emission front control in TCV

Systematic extraction of locally valid dynamic models from experiments is necessary for controller design and the validation of high fidelity models. This paper describes the extraction of a dynamic model in the form of a transfer function, giving the dynamic response of the CIII (465.0 nm) emission front position to deuterium gas puffing in the TCV divertor during flattop, relevant for heat exhaust control. The model is extracted using frequency response data from both SOLPS-ITER simulations and perturbative experiments. We use the steady-state solutions of the model SOLPS-ITER to obtain an additional data point at the zero frequency, as the identifiable frequency range by perturbative experiments is lower bounded by discharge time. We specifically approach the problem from a control engineering point of view, aiming to develop control-oriented models for the systematic design of impurity emission front controllers. We find a transfer function structure based on a diffusive process to best describe the obtained frequency response data. The resulting transfer function model accurately reproduces the local dynamic response measured during experiments, so it can be used to assess new controllers offline for similar discharge scenarios.

Development of an integrated core–edge scenario using the super H-mode

An optimized pedestal regime called the super-H (SH) mode is leveraged to couple a fusion relevant core plasma with a high density scrape-off layer appropriate for realistic reactor power exhaust solutions. Recent DIII-D experiments have expanded the operating space of the SH regime using advanced control algorithms and investigated optimization of impurity seeding, deuterium gas puffing, and 3D magnetic perturbations. Simultaneous real-time control of the pedestal density and radiated power with in-vessel coils and nitrogen seeding enable optimal coupled divertor and pedestal conditions. Four case studies are analysed with varied levels of radiated power in the divertor volume ranging from 0 (no seeding) to 8.5 MW radiated from carbon and nitrogen emission. Plasmas with a 4.5 MW radiated power target establish a radiative mantle, leading to divertor temperatures of ∼16 eV while maintaining SH-mode, and with only marginal impact on the pedestal and core performance. Increased levels of N2 seeding with a 7.5 MW radiated power target facilitate detachment onset and divertor temperatures <5 eV, with no degradation in stored energy and the operational point remaining inside the SH-mode channel for . Finally, a 8.5 MW radiated power target leads to partial detachment, which is so far associated with the loss of access to SH-mode pedestal conditions.

Results from core-edge experiments in high Power, high performance plasmas on DIII-D

Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND) hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P) ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5) due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot”) but also that heating near the slot opening is a significant source for impurity contamination of the core.

Impurity ion flow and temperature measured in a detached divertor with externally applied non-axisymmetric fields on DIII-D

Externally applied non-axisymmetric magnetic fields are shown to have little effect on the impurity ion flow velocity and temperature as measured by the multichord divertor spectrometer in the DIII-D divertor for both attached and detached conditions. These experiments were performed in H-mode plasmas with the grad-B drift toward the target plates, with and without n=3 resonant magnetic perturbations (RMPs). The flow velocity in the divertor is shown to change by as much as 30% when deuterium gas puffing is used to create detachment of the divertor plasma. No measurable changes in the C III flow were observed in response to the RMP fields for the conditions used in this work. Images of the C III emission are used along with divertor Thomson scattering to show that the local electron and C III temperatures are equilibrated for the conditions shown.

Bifurcation to a stable limiter MARFE in Tore Supra

This paper reports on the observation on the tokamak Tore Supra of a very high density, highly radiating edge regime featuring a high recycling divertor. This regime is stable over several energy confinement times, without any additional external control. It is achieved through strong deuterium gas puffing, in an ICRH heated plasma scenario. When the electron density reaches ∼1.1 times the Greenwald density, a bifurcation occurs, leading to strong edge asymmetries: electron density, enhanced edge radiated power localized close to the limiter surface, and a consistent reduction of the conducted power to the limiter. A Greenwald fraction up to ∼1.4 is then achieved, without any loss of confinement, and with low plasma core contamination.

Divertor Heat Flux Control Research on DIII-D

Divertor heat flux characterization and control results from DIII-D are summarized. The peak divertor heat flux is found to scale with a simple conduction model having perpendicular transport scaling with plasma current and heating power. In a double-null configuration, the heat flux sharing between divertors is very sensitive to the magnetic balance. Heat flux control in H-mode with edge-localized modes (ELMs) is obtained with deuterium gas puffing resulting in a partially detached divertor (PDD) regime. Important physical processes in the PDD regime include radiation from the intrinsic carbon impurity and deuterium, loss of electron pressure near the separatrix, parallel energy transport in the divertor dominated by convection, and particle flux reduction from deuterium recombination. Divertor neutral pressure is found to be an important control parameter to maintain the PDD regime. Divertor heat flux reduction is also obtained with impurity injection. In one approach divertor radiation is enhanced using induced scrape-off-layer flow to enrich divertor impurity concentration. Another approach uses seeded impurities to produce radiation inside the separatrix in a radiating mantle configuration. Observations of heat flux transients from ELMs and disruptions are summarized. Finally, the implications of these results for next-generation tokamaks are discussed.

Thermal wall load control using fast gas puffing in the TPE-RX reversed-field pinch

In order to reduce the thermal wall load due to the phase- and wall-locked modes particularly in the high plasma current regime, neutral gas puffing experiments have been conducted in the TPE-RX reversed-field pinch. The temperature of the plasma-facing wall at the mode-locking position sometimes exceeds the melting temperature of stainless steel. It has been determined that the maximum temperature increment of the wall becomes almost half in the neon and deuterium gas puffing cases at a plasma current of 350kA. This indicates that neutral gas puffing can reduce the severe thermal wall load. The radiative cooling around the peripheral region as a result of neon gas puffing and the decrease in the heat flux along local magnetic field lines as a result of deuterium gas puffing are possible reasons behind the decrease in the rate of energy flow to the wall.

Particle transport and density profile analysis of different JET plasmas

Over the last two years, several experiments relevant for the study of particle transport and density profile evolution, have been performed at JET. They can be classified as stationary discharges with and without central particle source due to the beams, quasi-stationary discharges with deuterium gas puffing, deep pellet fuelled discharges and discharges perturbed by cold pulses obtained by shallow pellet injection. All these experimental scenarios have been simulated by means of the JETTO transport code, employing different transport models: purely empirical models and the semi-empirical mixed Bohm/gyro-Bohm transport model, both with the addition of different theory-based expressions for the anomalous particle pinch and the first principle Weiland transport model. The coefficients used to scale the pinch velocity in the purely empirical and in the mixed Bohm/gyro-Bohm model have been varied from shot to shot. In this paper, the results of the simulations are presented. The main conclusions are that, for the cases studied in this paper, the sawtooth activity is the main particle transport mechanism in the plasma centre (r/a ⩽ 0.5). Nevertheless, to reproduce the density profile in the gradient zone (0.5 ⩽ r/a ⩽ 0.9), an anomalous pinch seems to be necessary, at least for L-mode plasmas. This anomalous convective flux is well reproduced by the off-diagonal elements of the transport matrix given by the Weiland model.

Investigation of the hydrogen fluxes in the plasma edge of W7-AS during H-mode discharges

In the stellarator W7-AS the H-mode is characterized by an edge transport barrier which is localized within a few centimeters inside the separatrix. The corresponding L-H transition shows well-known features such as the steepening of the temperature and density profiles in the region of the separatrix. With a so-called sniffer probe the temporal development of the hydrogen and deuterium fluxes has been studied in the plasma edge during different H-mode discharges with deuterium gas puffing. Prior to the transition a significant reduction of the deuterium and also the hydrogen fluxes can be observed. This fact confirms the assumption that the steepening of the density profiles starts at the outermost edge of the plasma. Moreover, sniffer probe measurements in the plasma edge could therefore identify a precursor for the L-H transition. The analysis of the hydrogen neutral gases shows a distinct change of the hydrogen isotope ratio during the transition. This observation is in agreement with the change in the particle fluxes onto the targets and can also be seen in the reduced H x signals from the limiters. It is further demonstrated that significant improvement in the time resolution of the measured data can be obtained by deconvolution of the data with the apparatus function using Bayesian probability theory and the Maximum Entropy method with adaptive kernels.

H mode discharges with feedback controlled radiative boundary in the ASDEX Upgrade tokamak

Puffing of impurities (neon, argon) and deuterium gas in the main chamber is used to feedback control the total radiated power fraction and the divertor neutral particle density simultaneously in the ASDEX Upgrade tokamak. The variation of Psep=Pheat-Prad(core) by impurity radiation during H mode shows a similar effect on the ELM behaviour as that obtained by a change of the heating power. For radiated power fractions above 90%, the ELM amplitude becomes very small and detachment from the divertor plates occurs, whilst no degradation of the global energy confinement is observed (completely detached high confinement mode). Additional deuterium gas puffing is found to increase the radiated power per impurity ion in the plasma core owing to the combined effect of a higher particle recycling rate and a lower core penetration probability. The outer divertor chamber, which is closed for deuterium neutrals, builds up a high neutral pressure, the magnitude of which is determined by the balance of particle sources and pumping. For this particular situation, the effective pumping time of neon and argon is considerably reduced, to less than 0.3 s, mainly owing to an improved divertor retention capability. The radiation characteristics of discharges with a neon driven radiative mantle are modelled using a 1-D radial impurity transport code that has been coupled to a simple divertor model describing particle recycling and pumping. The results of simulations are in good agreement with experiment