Radiative Divertor Research Articles

Investigation of electron parallel pressure balance in the scrapeoff layer of deuterium-based radiative divertor discharges in DIII-D

Investigation of electron parallel pressure balance in the scrapeoff layer of deuterium-based radiative divertor discharges in DIII-D

Particle exhaust in radiative divertor experiments

Abstract Power exhaust is one of the critical questions to be solved for ITER, especially with the side condition of sufficient particle exhaust. Since the H-mode threshold might require a high power flux over the separatrix, only small power losses inside the bulk plasma might be permitted, and thus eventually a large fraction of the power must be removed in the scrape-off layer and divertor plasma. Neutral energy losses can only remove a small fraction of the power, and therefore most of the power flux has to be radiated, either by hydrogen isotopes, by the intrinsic impurities (mainly carbon) or by additionally seeded impurities. However, all these scenarios with reduced power flow to the divertor imply also a reduced particle flow into the divertor. This might be critical for particle exhaust, especially concerning the helium ash. Reduction of the power flow to the divertor by enhanced radiation losses has been investigated in the large divertox experiments like JET, DIII-D, ASDEX Upgrade, and JT-60U, and an experimental overview of the present status will be given in this paper, together with a discussion of impurity exhaust (mainly He and Ne) in these scenarios.

Routes to divertor detachment in ASDEX Upgrade

Abstract A simple analytic model of divertor radiation is used as a framework for identifying different routes to divertor detachment. Three means of achieving detachment are demonstrated experimentally including, increasing plasma density, purposely increasing the impurity level and increasing the scrape-off layer connection length. The resulting low electron temperatures in the divertor causes friction processes to dominate ionization in the recycling region, giving rise to plasma pressure loss along field lines. The pressure loss is in approximate agreement with simple analytic modelling.

First measurements of electron temperature and density with divertor Thomson scattering in radiative divertor discharges on DIII-D

We have obtained the first measurements of n e and T e in the DIII-D divertor region with a multi-pulse (20 Hz) divertor Thomson scattering (DTS) system. Eight measurement locations are distributed vertically up to 21 cm above the divertor plate. Two-dimensional distributions have been obtained by sweeping the divertor plasma across the DTS measurement location. Several operating modes have been studied, including Ohmic, L-mode, ELMing H-mode, and radiative divertor operation with puffing of D 2 and impurities. Mapping of the data to either the ( L pol , φ) or ( R, Z) planes with the EFIT equilibrium is used to analyze the 2D profiles. We find that in ELMing H-mode: n e , T e , and P e are relatively constant along field lines from the X-point to the divertor plate, especially near the separatrix field line. With D 2 puffing, the DTS profiles indicate that T e in a large part of divertor region below the X-point is dramatically reduced from ∼ 30–40 eV in ELMing H-mode to 1–2 eV. This results in a fairly uniform low- T e divertor, with an increased electron density in the range of 2 to 4 × 10 20 m −3. Detailed comparisons of the spatial profiles of n e , T e , and electron pressure P e , are presented for several operating modes. In addition, these data are compared with initial calculations from the UEDGE fluid code.

Stability of a radiative mantle in ITER

We report results of a study to evaluate the efficacy of various impurities for heat dispersal by a radiative mantle and radiative divertor (including SOL). We have derived a stability criterion for the mantle radiation which favors low Z impurities and low ratios of edge to core thermal conductivities. Since on the other hand the relative strength of boundary line radiation to core bremsstrahlung favors high Z impurities, we find that for the ITER physics phase argon is the best gaseous impurity for mantle radiation. For the engineering phase of ITER, more detailed analysis is needed to select between krypton and argon.

Radiative plasmas in TdeV

The main aspects of radiative plasma experiments carried out in the LH-heated TdeV tokamak are summarized. D 2, N 2 and Ne puffing, and divertor biasing, are assessed in terms of their ability to increase edge and divertor radiative losses and reduce neutralizer plate heat loads. Detailed patterns of radiation and plate power deposition are compared. Combined N 2 and D 2 puffing gives the best results. Biasing enhances divertor radiation but also causes an increase in power conducted and convected to one of the divertor legs. The improvements obtained are still insufficient due to disruptive instabilities that appear at high impurity puffing rates. The instability is triggered at the edge with N 2 and in the core with Ne.

Measurement and analysis of radiated power components in the JET MkI divertor using VUV spectroscopy

This paper presents a spectroscopic study of deuterium and impurity radiation in the MkI divertor of JET in different experimental regimes. Radiated power measurements using absolutely calibrated VUV spectroscopy are consistent with bolometer measurements in a variety of impurity scenarios. These observations, combined with measurements of visible emission, are compared with code predictions of divertor radiation distributions, highlighting uncertainties in the description of the background plasma and of the carbon source, which affect detailed modelling of divertor impurity transport.

Modification and control of divertor detachment in Alcator C-Mod

The effect of adding different impurity gases and ICRF heating on divertor detachment has been compared for Ohmic, L-mode and high- q | (≥ 300 MW/ m −2), H-mode plasmas. The relative effect of all these external controls on divertor and core radiation is evaluated. Recycling gases (Ne and Ar) are found to primarily affect the main plasma as opposed to the divertor region. The lower Z non-recycling gases, N 2 and CD 4 are more efficacious in this regard. N 2 was found effective in inducing detachment for high- q |, H-mode plasmas. An additional advantage of N 2 over the higher Z gases is that its injection resulted in the smallest degradation in H-mode energy confinement (∼ 10%).

Impurity feedback control for enhanced divertor and edge radiation in DIII-D discharges

Radiated power fractions above 90% have been achieved in DIII-D using impurity feedback control. Heat flux reductions up to a factor of 6 and confinement enhancements of 1.6 X ITER-89P L-mode scaling have been obtained with a radiated power fraction above 90%. When the outer separatrix strike point was positioned for pumping with the DIII-D cryopump, ELMing H-mode (Type III ELMs) was triggered with neon feedback injection in conjunction with a constant deuterium gas puff rate. Changes in the ELM frequency, energy confinement and central neon concentrations as a function of radiated power are presented. Observations of detached discharges with H-mode confinement similar to the ASDEX-U CDH-mode are also discussed.

Comprehensive 2D measurements of radiative divertor plasmas in DIII-D

This paper presents a comparison of the total radiated power profile and impurity line emission distributions in the SOL and divertor of DIII-D. This is done for ELMing H-mode plasmas with heavy deuterium injection (partially detached divertor operation, PDD) and those without deuterium puffing. Results are described from a series of dedicated experiments performed on DIII-D to systematically measure the 2D ( R, Z) structure of the divertor plasma. The discharges were designed to optimize measurements with new divertor diagnostics including a divertor Thomson scattering system. Discharge sequences were designed to produce optimized data sets against which SOL and divertor theories and simulation codes could be benchmarked. During PDD operation the regions of significant radiated power shift from the inner divertor leg and SOL to the outer leg and X-point regions. D α emission shifts from the inner strikepoint to the outer strikepoint. Carbon emissions (visible CII and CIII) shift from the inner SOL near the X-point to a distributed region from the X-point to partially down the outer leg during moderate D 2 puffing. In heavy puffing discharges the carbon emission coalesces on the outer separatrix near the X-point and for very heavy puffing it appears inside the last closed flux surface above the X-point. Calibrated spectroscopic measurements indicate that hydrogenic and carbon radiation can account for all of the radiated power. L α and CIV radiation are comparable and when combined account for as much as 90% of the total radiated power along chords viewing the significant radiating regions of the outer leg.

Recent results from tokamak divertor plasma measurements

New diagnostics have been developed to address key divertor physics questions, including: target plate heat flux reduction by radiation, basic edge transport issues, and plasma wall interactions (PWIs) such as erosion. A system of diagnostics measures the target plate heat flux (imaging ir thermography) and particle flux (probes, pressure and Penning gauges, and visible emission arrays). Recently, Te, ne, and Pe (electron pressure) have been measured in two-dimensions (2D) with divertor Thomson Scattering. During radiative divertor operation Te is less than 2 eV, indicating that new atomic processes are important. Langmuir probes measure higher Te in some cases. In addition, the measured Pe near the separatrix at the target plate is lower than the midplane pressure, implying radial momentum transport. Bolometer arrays, inverted with reconstruction algorithms, provide the two-dimensional core and divertor radiation profiles. Spectroscopic measurements identify the radiating species and provide information on impurity transport; both absolute chordal measurements and tomographic reconstructions of images are used. Either intrinsic carbon or an inert species (e.g., injected Ne) are usually observed, and absolute particle inventories are obtained. Computer codes are both benchmarked with the experimental data and provide important consistency checks. Several techniques are used to measure fundamental plasma transport and fluctuations, including probes and reflectometry. PWI issues are studied with in situ coupons and insertable samples (DiMES). Representative divertor results from DIII-D with references to results on other tokamaks will be presented.

Experimental and calculated basis of the lithium capillary system as divertor material

First results as experimental and calculated basis of a new concept are described in the paper. Experimental models of liquid lithium capillary structure have been tested at long-pulse high heat loads. The power loads on the capillary target up to 50 MW/m 2 were provided by an electron beam with electron energy ≤ 9 ke V in a longitudinal magnetic field of 0.25 T. Seven experiments were performed with the different capillary targets. The effects of disruption discharges in tokamaks have been simulated by means of magnetized plasma flows with pulse length of 0.2 ms, electron density of 10 22 m 3 and energy density up to 4 MJ/m 2. The plasma flow was generated by a quasistationary plasma accelerator and interacted with a lithium capillary structure. 2D modelling of the ITER divertor with a lithium target is presented as the first step in the validation of a new divertor concept. A lithium radiative divertor scenario has been examined for the ITER using DDIC95 code. First calculations have shown that thermal loads on the divertor plates are reduced down to 1.3 MW/m 2. The main power is radiated in the divertor.

Investigation of electron parallel pressure balance in the scrapeoff layer of deuterium-based radiative divertor discharges in DIII-D

Electron density, temperature and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: I p = 1.4 MA, q 95 = 4.2 and P input ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of 2 between the midplane separatrix and the X-point and then an additional ∼ 3–5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power (i.e., approximately twice the L-H-mode threshold power) maintain good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7–30 eV.

Particle exhaust in radiative divertor experiments

Power exhaust is one of the critical questions to be solved for ITER, especially with the side condition of sufficient particle exhaust. Since the H-mode threshold might require a high power flux over the separatrix, only small power losses inside the bulk plasma might be permitted, and thus eventually a large fraction of the power must be removed in the scrape-off layer and divertor plasma. Neutral energy losses can only remove a small fraction of the power, and therefore most of the power flux has to be radiated, either by hydrogen isotopes, by the intrinsic impurities (mainly carbon) or by additionally seeded impurities. However, all these scenarios with reduced power flow to the divertor imply also a reduced particle flow into the divertor. This might be critical for particle exhaust, especially concerning the helium ash. Reduction of the power flow to the divertor by enhanced radiation losses has been investigated in the large divertox experiments like JET, DIII-D, ASDEX Upgrade, and JT-60U, and an experimental overview of the present status will be given in this paper, together with a discussion of impurity exhaust (mainly He and Ne) in these scenarios.

A review on impurity transport in divertors

Power exhaust is one of the most crucial requirements for future fusion reactors, like ITER. It is widely recognized that impurity injection is needed to significantly reduce the heat load to the divertor plates. By controlled injection of light impurities, the compatibility of high confinement core plasma with strong radiative divertor has been successfully demonstrated in many tokamaks. However, the core impurity contamination was high ( Z eff ∼ 3) compared to the value required for ITER ( Z eff < 1.6). Therefore, a scheme for impurity retention in the divertor region should be established for fusion research. This paper reviews the recent progress in experiments and simulations which have been made for the purpose of understanding impurity transport in divertors. The issues contained in the paper are impurity generation, shielding and cross field diffusion. As for the impurity generation, chemical sputtering and wall source are discussed with emphasis on the characteristics of their transport and shielding. Impurity control with plasma flow induced by gas puffing and divertor pumping, and adequately designed divertor geometry is also presented.

Routes to divertor detachment in ASDEX Upgrade

A simple analytic model of divertor radiation is used as a framework for identifying different routes to divertor detachment. Three means of achieving detachment are demonstrated experimentally including, increasing plasma density, purposely increasing the impurity level and increasing the scrape-off layer connection length. The resulting low electron temperatures in the divertor causes friction processes to dominate ionization in the recycling region, giving rise to plasma pressure loss along field lines. The pressure loss is in approximate agreement with simple analytic modelling.

Design of an imaging bolometer system for the large helical device

We describe a radical design for a bolometer system employing infrared (IR) imaging of a segmented-matrix absorber in a cooled-pinhole camera geometry, which we will prototype and demonstrate on the large helical device (LHD).1 LHD will be operational in early 1998, with an l=2 superconducting winding, a major radius of 3.9 m, a minor radius of 0.5–0.65 m, and input powers ranging from 3 MW (steady state) to 30 MW (pulsed). The bolometer design parameters are determined by modeling the temperature of the foils making up the detection matrix using a two-dimensional time-dependent solution of the heat conduction equation. This design will give a steady-state bolometry capability, with modest (60 Hz) time resolution, while simultaneously providing hundreds of channels of spatial information. No wiring harnesses will be required, as the temperature-rise data is measured via a 12-bit, ±0.025 °C resolution, 3–5 μm band, 256×256 pixel IR camera. The spatial data will be used to tomographically invert the profile of the highly shaped stellarator main plasma and divertor radiation, in conjunction with more conventional fanned arrays of traditional bolometers.

Thomson scattering stray light reduction techniques using a CCD camera

The DIII–D Thomson scattering system has been expanded to measure divertor plasma temperatures from 1 to 500 eV and densities from 0.05 to 8×1020 m−3. To complete this system, a difficult stray light problem was overcome to allow for an accurate Rayleigh scattering density calibration. The initial stray light levels were over 500 times higher than the expected Rayleigh scattered signal. Using a charge-coupled device (CCD) camera, various portions of the vessel interior were examined while the laser was fired through the vessel in air at atmospheric pressure. Image relaying, exit window tilting, entrance and exit baffle modifications, and a beam polarizer were then used to reduce the stray light to acceptable levels. The CCD camera gave prompt feedback on the effectiveness of each modification, without the need to reestablish vacuum conditions required when using the normal avalanche photodiode detectors (APD). Once the stray light was sufficiently reduced, the APD detectors provided the signal time history to more accurately identify the source location. We have also found that certain types of high reflectance dielectric coatings produce 10–15 times more scatter than other types of more conventional coatings. By using low-scatter mirror coatings and these new stray light reduction techniques, we now have more flexibility in the design of complex Thomson scattering configurations required to probe the central core and the new radiative divertor regions of the DIII–D vessel.

Recent results from tokamak divertor plasma measurements (invited) (abstract)

New diagnostics have been developed to address key divertor physics questions, including: target plate heat flux reduction by radiation, basic edge transport issues, and plasma wall interactions (PWIs) such as erosion. A system of diagnostics measures the target plate heat flux (imaging ir thermography) and particle flux (probes, pressure and Penning gauges, and visible emission arrays). Recently, Te, ne, and Pe (electron pressure) have been measured in two-dimensions (2D) with divertor Thomson Scattering. During radiative divertor operation Te is less than 2 eV, indicating that new atomic processes are important. Langmuir probes measure higher Te in some cases. In addition, the measured Pe near the separatrix at the target plate is lower than the midplane pressure, implying radial momentum transport. Bolometer arrays, inverted with reconstruction algorithms, provide the two-dimensional core and divertor radiation profiles. Spectroscopic measurements identify the radiating species and provide information on impurity transport; both absolute chordal measurements and tomographic reconstructions of images are used. Either intrinsic carbon or an inert species (e.g., injected Ne) are usually observed, and absolute particle inventories are obtained. Computer codes are both benchmarked with the experimental data and provide important consistency checks. Several techniques are used to measure fundamental plasma transport and fluctuations, including probes and reflectometry. PWI issues are studied with in situ coupons and insertable samples (DiMES). Representative divertor results from DIII-D with references to results on other tokamaks will be presented.

A tangentially viewing visible TV system for the DIII-D divertor

A video camera system has been installed on the DIII-D tokamak for two-dimensional spatial studies of line emission in the lower divertor region. The system views the divertor tangentially at approximately the height of the X point through an outer port. At the tangency plane, the entire divertor from the inner wall to outside the DIII-D bias ring is viewed with spatial resolution of ∼1 cm. The image contains information from ∼90 deg of toroidal angle. In a recent upgrade, remotely controllable filter changers were added which have produced images from nominally identical discharges using different spectral lines. Software was developed to calculate the response function matrix of the optical system using distributed computing techniques and assuming toroidal symmetry. Standard sparse matrix algorithms are then used to invert the three-dimensional images onto a poloidal plane. Spatial resolution of the inverted images is 2 cm; higher resolution simply increases the size of the response function matrix. Initial results from a series of experiments with multiple identical discharges show that the emission from C II and C III, which appears along the inner scrape-off layer above and below the X point during ELMing H mode, moves outward and becomes localized near the X point in radiative divertor operation induced by deuterium injection.