Predictive Modeling for Performance Assessment of ITER-Like Divertor in China Fusion Engineering Testing Reactor

To facilitate the design of the China Fusion Engineering Testing Reactor (CFETR), predictive modeling for the assessment and optimization of the divertor performances is an indispensable approach. This paper presents the modeling of the edge plasma behaviors as well as the W erosion and transport properties in CFETR with ITER-like divertor by using the B2-Eirene/SOLPS 5.0 code package together with the Monte Carlo impurity transport code DIVIMP. As expected, SOLPS modeling of divertor-SOL plasmas finds that the peak heat flux onto the divertor targets greatly exceeds 10 MW/m², an engineering limit posed to the steady-state and/or long-pulse operation of the next-step fusion devices, for a wide range of plasma conditions, and thus modeling of Ar puffing by scanning the puffing rate for radiative divertor is performed. As the increase of the Ar puffing rate, the peak target heat fluxes and plasma temperature decreases exponentially,reflecting that Ar puffing is highly effective at power exhausting. Based on the ion fluxes from SOLPS, the W erosion is calculated by taking into consideration the bombardment of both D and Ar ions, and then the W plasma concentrations are calculated based on the W erosion fluxes using DIVIMP. The calculations show that if the Ar puffing only being used to reduce the divertor heat load, the W plasma contamination in the core plasma exceeds the tolerable value (<10^?5), which demonstrates that some further upgrading of the divertor geometry is still needed.     

Spectroscopic studies of fuel recycling and impurity behaviors in the divertor region of Wendelstein 7-X

The first divertor operation phase(OP1.2 a) was carried out on Wendelstein 7-X in the second half of 2017.Fuel recycling and impurity behaviors in the divertor region were investigated by employing a newly built ultraviolet–visible–near infrared overview spectroscopy system.The characteristic spectral lines of the working gases(hydrogen and helium),intrinsic impurities(carbon,oxygen and iron),and seeded impurities(neon and nitrogen) were identified and analyzed.The divertor electron temperature and density were measured using He I(667.8,706.5,and 728.1 nm) line intensity ratios.The Hα(656.3 nm),He I(587.6 nm),C II(514.5 nm),and O I(777.2 nm) emissions were investigated over a wide range of operating conditions.The results showed that fuel and impurity emissions in the divertor region exhibit a strong dependence on magnetic topology and plasma conditions.The levels of Hα,He I,C II,and O I emissions are all reduced moving from the standard configuration to the high mirror configuration,and even further reduced for the high iota configuration,which is associated with decreasing connection length in these island divertor configurations.The H/He influx ratio shows that the plasma is a mixture of helium and hydrogen.The neutral and impurity influxes from the divertor target tend to increase with increasing divertor electron temperature.     

Radiative Cooling Rates for Low-Z Impurities in Non-coronal Equilibrium State

The local or transient radiation losses in tokamak plasmas can greatly exceed those in the coronal equilibrium. This excess is especially pronounced at the plasma edge. The reason for the increase of radiation in a peripheral plasma is as follows. The impurities are lost fast from the plasma edge and the new impurity source is supplied to this region. The charged states of impurities, therefore, do not reach their coronal equilibrium ones. These impurity ions have more electrons than those in the coronal equilibrium, and as a result emit the higher radiation power. In the simplest case, the non-coronal radiative rate can be determined only by two parameters: the electron temperature \(T_{\text {e}}\) and the so-called “residence parameter” \(n_{\text {e}}\tau _{\text {i}}\), where \(\tau _{\text {i}}\) is the impurity residence time in the plasma. Despite the strong simplification, such an approach allows to do simple estimates of non-coronal radiation. In this paper, two dimensional polynomial fits describing radiative cooling rates and mean charge are obtained for eight impurity species: helium, lithium, beryllium, carbon, nitrogen, oxygen, neon, and argon. The results are presented in figures and tables. The figures show curves calculated from the original atomic database and least-squares polynomial fits to these curves. The tables contains coefficients for this fits. The obtained fits can be useful for qualitative estimates and simple numerical calculations.     

Ignition in a tokamak reactor with INTOR-like parameters

The requirements for ignition in a tokamak reactor with INTOR-like parameters were studied using a one-dimensional transport code. With empirical electron energy diffusivity _e , ignition was obtained with 60–75 MW of neutral beam injection at a volume average pressure ratio =4–5% under a variety of conditions. Changing _e gave ignition at the same if the plasma minor radius varied asa _e ^1/2 . The maximum impurity concentration which allows ignition was found to be comparable to that for the much simpler case of a homogeneous plasma with radiative losses only. In long pulse simulations with efficient helium pumping, the maximum toroidal field ripple which allowed ignition was 2.0% (peak-to-peak) at the plasma edge. Ignition was maintained with over 99% recycling of helium ash using 5% less than maximum ripple.     

On the impurity problem in quasi steady-state toroidal plasma experiments and fusion reactors

A simple equation for estimating the impurity build-up in a plasma due to sputtering is discussed under various assumptions. It is shown that the D-T burning time in an experiment (or reactor) without a divertor or cold gas blanket is one particle confinement time at most. If the accumulation of impurities in the center of the plasma cannot be avoided, steady-state operation of a reactor even with a divertor will not be achievable. The effect of neutron sputtering is included in the discussion.     

Nuclear Characteristics of Gas-Suspended Boron Carbide Cooling Catalyzed D Fusion Reactor Blanket

Transient behaviors of plasma and in-vessel components have been investigated considering the divertor plasma state (detached/attached) transition. The SAFALY code consisting of a zero-dimensional plasma model and a one-dimensional heat transfer model of components has been modified to take account of the divertor plasma state transition on the basis of the updated divertor plasma physics. Several plasma events, i.e., over fueling, sudden auxiliary heating injection and Confinement improvement events which would be expected to result in overpower, were selected for the International Thermonuclear Experimental Reactor (ITER) and the transient behaviors were calculated on the assumption of a combined failure of plasma control and machine interlock in addition with a postulated plasma transient. The results show that plasma burning passively terminates due to sublimated impurity penetration from the carbon target surface, but there are possibilities of dry out of the coolant for the high heat flux in sudden attached state transition under the multifailure of plasma control. However, effects by the aggravating failure of the divertor are expected to be safely terminated by the confinement boundary, the vacuum vessel and its pressure suppression system.     

Beam plasma neutron sources based on beam-driven mirror

The design and performance of a relatively low-cost, plasma-based, 14-MeV D-T neutron source for accelerated end-of-life testing of fusion reactor materials are described in this article. An intense flux (up to 5×10¹⁸ n/m²·s) of 14-MeV neutrons is produced in a fully-ionized high-density tritium target (n _e 3×10²¹ m^–3) by injecting a current of 150-keV deuterium atoms. The tritium plasma target and the energetic D⁺ density produced by D⁰ injection are confined in a column of diameter 0.16 m by a linear magnet set, which provides magnetic fields up to 12 T. Energy deposited by transverse injection of neutral beams at the midpoint of the column is conducted along the plasma column to the end regions. Longitudinal plasma pressure in the column is balanced by neutral gas pressure in the end tanks. The target plasma temperature is about 200 eV at the beam-injection position and falls to 5 eV or less in the end region. Ions reach the walls with energies below the sputtering threshold, and the wall temperature is maintained below 740 K by conventional cooling technology.     

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n_e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n_e/n_GW  1), to enable n_e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n_e/n_GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m²) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.     

Impurity Measured by VUV Spectrometer and OMA on HL-2A Tokamak

Impurity is one of the key issues on a great impact to the quality of tokamak plasma.HL-2A is the first divertor tokamak in China. In this paper the experimental results are presented on impurity through the line emission measurement in the campaign in 2003 under the limiter and divertor configurations. The low-Z impurities such as carbon and oxygen are the most important components in the plasma, but their content are not so high to affect the discharge quality. The high-Z impurities such as copper and ferrum are not essential. The emission intensity of impurity is clearly decreased during the divertor configuration formed.     

The populations of excited levels of hydrogen-like and helium-like ions in plasmas traversed by neutral hydrogen beams

This paper examines the populations of excited levels of impurity ions in a spatially homogeneous plasma containing primarily thermal electrons and protons and monoenergetic neutral hydrogen atoms. Of special concern is the role of recombination which may include the radiative, three-body and dielectronic processes together with charge exchange capture from neutral hydrogen beams.The influence of these primary processes on the populations is modified by radiative transitions and redistributive transitions due to collisions with electrons and protons in the plasma. The behaviour of the populations of the ions C⁵⁺, C⁴⁺ and Ar¹⁶⁺ with variation of plasma parameters is explored in the present work. A bundled principal quantum level picture and a more elaborate LS resolved picture are used which allow investigation of the expected spectral emission and its sensitivity to uncertainty in the primary rates.The variation of the impurity ion spectrum in transiently recombining or ionising conditions is also considered.     

Latest Results from the Damavand Tokamak Experiments

Damavand is a small tokamak (a = 7 cm, R = 36 cm) with an elongated plasma cross section (k 2) and a poloidal divertor. Its passive coils within the vacuum chamber provide the plasma formation at the torus center and act as a passive stabilizer for the plasma current. The experimental measurements presented here are the latest results related to the Damavand discharge main behavior with graphite limiters (before the device modification) during the ohmic heating phase. In this respect, we have evaluated some of the characteristic parameters such as edge safety factor (q_a 3.1), energy confinement (_E = 0.4 ms), electron temperature (T_e = 161 eV), and so on, during normal operation of the plasma current. The assessment of disruption phase of the plasma current has been considered by estimating the characteristic parameters of thermal and current quenches to be about 6 eV and 6 MA/s, respectively. Here, also we have monitored the evolution of the line emission of impurity (C, O) ions in the central deuterium plasma. The relative density of carbon and oxygen impurity levels has been estimated to be 2.6 and 1.4%, respectively. It is concluded that the impurity radiation loss is much higher during disruption phase of plasma current.     

OEDGE modeling of plasma contamination efficiency of Ar puffing from different divertor locations in EAST

Modeling with OEDGE was carried out to assess the initial and long-term plasma contamination efficiency of Ar puffing from different divertor locations,i.e.the inner divertor,the outer divertor and the dome,in the EAST superconducting tokamak for typical ohmic plasma conditions.It was found that the initial Ar contamination efficiency is dependent on the local plasma conditions at the different gas puff locations.However,it quickly approaches a similar steady state value for Ar recycling efficiency ＞0.9.OEDGE modeling shows that the final equilibrium Ar contamination efficiency is significantly lower for the more closed lower divertor than that for the upper divertor.     

Thermal effects and component cooling in Magnum-PSI

Magnum-PSI is a linear plasma generator, built at the FOM-Institute for Plasma Physics Rijnhuizen. Subject of study will be the interaction of plasma with a diversity of surface materials. The machine is designed to provide an environment with a steady state high-flux plasma (up to 10²⁴ H⁺ ions/m² s) in a 3 T magnetic field with an exposed surface of 80 cm² up to 10 MW/m². Magnum-PSI will provide new insights in the complex physics and chemistry that will occur in the divertor region of the future experimental fusion reactor ITER and reactors beyond ITER. The conditions at the surface of the sample can be varied over a wide range, such as plasma temperature, beam diameter, particle flux, inclination angle of the target, background pressure and magnetic field. An important subject of attention in the design of the machine was thermal effects originating in the excess heat and gas flow from the plasma source and radiation from the target.     

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.     

Erosion and re-deposition processes in JET tiles studied with ion beams

Plasma facing components in fusion reactor chambers will operate under extreme conditions. Among the processes with implications on the material lifetime are erosion and re-deposition due to plasma interactions.This work will address the behaviour of both JET divertor and outer poloidal limiters (OPL) under plasma irradiation. The limiters comprise about 50 pairs of tiles in a poloidal stack, each of which has a plasma facing surface about 25 mm (poloidal) by 350 mm (toroidal) and is about 50 mm thick. The divertor tiles are located at the bottom of the chamber and withstand high fluxes of radiation and heat. Standard carbon-fibre composite (CFC) tiles coated with a thin layer of W overlaid with a 10 μm layer of C were studied with RBS/PIXE to understand the erosion/re-deposition processes occurring in these regions of the reactor chamber. High resolution surface morphology was assessed through SEM with and without tilting of the sample. The retention of hydrogen isotopes in the tiles were studied combining NRA and ERDA techniques – this is mostly ²H from the fuelling gas, but ³H is also present as a result of ²H–²H fusion reactions, and ¹H coming from the atmospheric exposure.     

Study on the retention behavior of hydrogen isotopes and the change of chemical states of boron film exposed to hydrogen plasma in LHD

The behavior of hydrogen retention and the change of chemical states of boron film exposed to hydrogen plasma in LHD were investigated. The sample was prepared in LHD, and atomic concentrations for the boron film after hydrogen plasma exposure were changed from 75% for boron, 15% for carbon and 8% for oxygen to 53%, 18% and 22%, respectively. BC bond was a major chemical state of the boron film after hydrogen plasma exposure, although abundance of BB bond was the highest before the plasma exposure. Total hydrogen retention measured by TDS was evaluated to be 1.7 × 10²⁰ H m^?2, and the retentions of hydrogen as BHB, BH and BCH bonds were, respectively, 4.8 × 10¹⁹, 7.2 × 10¹⁹ and 5.2 × 10¹⁹ H m^?2. It was concluded that the hydrogen retention could be estimated by taking account not only of chemical states of impurities, but also of hydrogen depth profile.     

Saturation in deuterium retention of CFC graphite targets exposed to beryllium-seeded plasmas on PISCES-B

ITER strike-plates are foreseen to be of carbon-fiber-composite (CFC). In this study the CFC bulk deuterium retention in ITER-relevant conditions is investigated. DMS 701 (Dunlop) CFC targets were exposed to plasma in PISCES-B divertor plasma simulator. Samples were exposed to both pure deuterium plasma and beryllium-seeded plasma at high fluences (up to ) and high surface temperature (1070 K). The deuterium contents of the exposed samples have been measured using both thermal-desorption-spectrometry (TDS) during baking at 1400 K and ion beam nuclear reaction analysis (NRA). The total deuterium inventory has been obtained from TDS while NRA measured the deuterium depth distribution. In the analysed fluence range at target temperature of 1070 K, no fluence dependence was observed. The measured released deuterium is . In the case of target exposure with beryllium-seeded plasma no change in the released amount of deuterium was found. The deuterium concentration inside the samples is almost constant until the probed depth of ?m, except in the first 1 μm surface layer, where it is 5 times higher than in the bulk. No C erosion/redeposition was observed in the Be-seeded plasma cases. The measured retention, applied to 50 m² of ITER CFC surface, would imply a tritium saturated value of 0.3 gT, much lower than the ITER safety limit of 350 g.     

Experimental Study of the Atomic and Molecular Processes Related to Plasma Detachment in Steady-State Divertor Simulator MAP-II

Atomic and molecular processes relevant to the volumetric recombination phenomena were investigated in a linear divertor plasma simulator MAP-II. Volumetric recombination is induced in He plasma by puffing of He or H2. In the He puffing case, the reduction of the ion flux is dominated by the electron-ion recombination. In the H2 puffing case, however, it is dominated by the molecule-assisted recombination (MAR), which is characterized by the disappearance of the Helium Rydberg spectra and by the existence of the hydrogen negative ions. Current achievement and the future prospect are described.