High Neutral Pressure Research Articles

Access to an ELM-suppressed X-point radiator regime in TCV snowflake minus configurations

TCV’s operating regime with an X-point radiator (XPR) has been broadened by changing the magnetic geometry. XPRs have properties that could make them an attractive power exhaust solution for fusion reactors. These include the conversion of a high fraction of exhaust power into radiation. TCV had previously accessed the XPR regime only with difficulties, as predicted for plasmas where radiative losses are dominated by carbon impurities, that are ubiquitous in TCV. Guided by this theoretical model of the XPR, recent experiments employed TCV’s configurational versatility to demonstrate that XPR access can be facilitated by introducing a second X-point in the vicinity of the separatrix. This configuration, which has a snowflake-minus topology, features a particularly long magnetic connection length from the region just above the X-point to the outer midplane together with a wide geometrical interface with the private flux region that reaches high neutral pressures. Transitioning to this configuration in a high-power H-mode leads to a shift in the radiating region across the separatrix from the divertor to a volume above the X-point, i.e. within the last closed flux surface (LCFS). This displacement of the radiating region is co-incident with the disappearance of edge localised modes (ELMs), while retaining H-mode confinement, a behaviour only, to date, observed in devices with metallic walls. In contrast to observations in these other devices, on TCV, the primary strike points in these configurations remain attached. Detailed measurements of the plasma kinetic parameters inside and outside of the separatrix now challenge the models for access and stability of the XPR and ELMs alike.

Facilitated core-edge integration through divertor nitrogen seeding in the HL-2A tokamak

A divertor detachment with sustainable high-performance plasmas has been achieved through divertor nitrogen seeding in the HL-2A tokamak. The closed divertor structure facilitates achieving high divertor neutral pressure due to its efficient particle containment effect as demonstrated by both experimental and SOLPS simulation results, which aids in achieving the divertor detachment. The radiative divertor is conducive to high-frequency small edge localized modes (ELMs) operation, characterized by a reduced pedestal ion temperature gradient ∇Ti and the enhanced electron density gradient ∇ne . The presence of minority nitrogen at the edge plasma, which results from the divertor nitrogen seeding, contributes to the increase of ion temperature Ti in the confinement region by suppressing the turbulence through the impurity dilution effect and E×B shear. The increased ion temperature Ti and electron density ne compensate the energy loss resulting from the increased edge radiation during detachment, which contributes to the confinement sustainment combined with reduced pedestal energy loss caused by small ELMs. This work advances our understanding of the fundamental physics governing closed divertor detachment with sustainable high-performance plasmas through divertor nitrogen seeding.

Benign termination of runaway electron beams on ASDEX Upgrade and TCV

This paper discusses the development of a benign termination scenario for runaway electron (RE) beams on ASDEX Upgrade and TCV. A systematic study revealed that a low electron density (n e) companion plasma was required to achieve a large MHD instability, which expelled the confined REs over a large wetted area and allowed for the conversion of magnetic energy to radiation. Control of the companion plasma ne was achieved via neutral pressure regulation and was agnostic to material injection method. The neutral pressure required for recombination was found to be dependent on impurity species, quantity and RE current. On TCV, n e increased at neutral pressures above 1 Pa, indicating that higher collisionality between the REs and neutrals may lead to an upper pressure limit. The conversion of magnetic energy to radiated energy was measured on both machines and a decrease in efficiency was observed at high neutral pressure on TCV. The benign termination technique was able to prevent any significant increase in maximum heat flux on AUG from 200 to 600 kA of RE current, highlighting the ability of this approach to handle fully formed RE beams.

Overview over the neutral gas pressures in Wendelstein 7-X during divertor operation under boronized wall conditions

During the first test divertor campaign of the stellarator experiment Wendelstein 7-X (Pedersen et al 2022 Nucl. Fusion 62 042022), OP1.2b, 13 neutral gas pressure gauges collected data in different locations in the plasma vessel, enabling a detailed analysis of the neutral gas pressures, the compression ratios and the particle exhaust rates via the turbomolecular pumps in the different magnetic field configurations. In Wendelstein 7-X, the edge magnetic islands are intersected by the divertor target plates and used to create a plasma-wall interface. As the number and position of the magnetic islands varies in different magnetic field configurations, the position of the strike line on the target plates and thus the neutral gas pressure in the subdivertor differs between the configurations. Neutral gas pressures on the order of few 10−4 mbar were measured in the subdivertor region. The highest neutral gas pressure of mbar was obtained in the so-called high iota configuration featuring four edge magnetic islands per cross section. The neutral particle flux through the pumping gaps into the subdivertor volume was provided by EMC3-EIRENE simulations and allowed to analyze the relation between the particle flux entering the subdivertor and the pressure distribution in the subdivertor. Finite element simulations in ANSYS provide a detailed picture of the pressure distribution in the subdivertor volume and agree with the neutral gas pressure measurements in the subdivertor in the standard configuration featuring an island chain of 5 edge magnetic islands. Surprisingly high neutral gas pressures that were not predicted by the simulation were measured in the subdivertor region away from the main strike line for discharges in the most used magnetic configuration, the standard configuration. While the pressure ratio between the two sections of the subdivertor volume, the low and high iota section is 0.06 in high iota configuration, a ratio of 2–5 was obtained in the other configurations, indicating significant particle loads and exhaust rates on the high iota section of the subdivertor in magnetic configurations with the main strike line on the low iota divertor targets.

Numerical modelling of the impact of leakage under divertor baffle in WEST

In WEST experimental campaign C5, the divertor pumping capability has been improved by sealing the space between the divertor outer baffle and the vacuum vessel. It is expected that the degree of baffle leakage influences the transport of neutral particles inside the main chamber, which affects the detachment onset. Knowing the exact impacts of leakage and understanding the physical processes behind it are helpful for the study and control of plasma detachment. We investigate the impact of leakage by performing transport simulations through SOLEDGE-EIRENE code considering several cases with different leakage levels. Starting from the basic simulation case, non-constant radial transport coefficients obtained by the feedback control method are applied to achieve a better match with the experimental one (#54903) in L-mode. Based on the basic case, the evolution of plasma regimes from sheath limited regime to detached one in different wall geometries has been studied by ramping the upstream density. The numerical results show that the cases with closed or reduced leakage under the baffle have better performance in trapping the neutral particles and higher neutral pressure near the baffle. The neutral compression ratio is increased by a factor up to 4, leading to more significant momentum and power dissipation in the divertor, thus lowering the detachment threshold in n e,sep by up to 16%. At the same time, a much higher gas puff rate by a factor up to 22 is needed to maintain an equivalent n e,sep level in the case without or reduced leakage. For all the cases here, there exist characteristic parameters on which the baffle closure has no obvious influence on their value when plasma starts to detach. The evolution of radiator height as a function of target temperature shows no sensitivity to the leakage, which gives some insight into the stable detachment control strategy in the future. Finally, simulation results are compared with available neutral pressure measurements from WEST campaigns to verify the predictions from the simulation. • The impact of wall geometry (characterized by different sizes of leakage under the divertor baffle) on detachment properties in WEST is investigated using SOLEDGE-EIRENE transport code. • Radial non-constant transport coefficients obtained by the feedback control method are applied in the simulation, allowing a better match with experimental results in L-mode. • The case with less leakage has better performance in trapping the neutral particles and has higher neutral pressure near the target, leading to greater power dissipation in the divertor, thus decreasing the detachment threshold in upstream separatrix density. • The evolution of the radiation front and the sensitivity of the detachment process to multiple parameters under the influence of leakage have been investigated. • Simulation results are compared with available neutral pressure measurements from WEST campaigns to verify the predictions from the simulation with respect to the impact of leakage.

Three-dimensional modeling of sputtered materials transport in diagnostic ducts of fusion devices

Migration of plasma erosion products in plasma facilities is studied experimentally and numerically within the framework of modeling transport of plasma-facing materials in the diagnostic ducts of fusion devices. Material transport simulation is discussed for two cases of low and high background neutral gas pressures. Monte Carlo software KITe was used to simulate transport at a neutral gas background pressure 0.1–0.5 Pa—typical during steady-state tokamak operation and during pressure pulses caused by edge localized modes (ELMs). The simulation approach was implemented to describe experiments at the MAGNUM-PSI facility. Fluid dynamic code FLUENT is used to simulate transport during pressure surges as high as 1000 Pa, which can occur in the case of severe disruptions in tokamak plasma discharges, such as vertical displacement events (VDE) or accidental events. The hydrodynamic approach was verified in simulation of target sputtering in the QSPA plasma gun facility.

Reduction of pulsed particle load with dynamic pressure induced by transient recycled neutral flux

From pulsed plasma experiments focusing on neutral pressure dependence, the impacts of a transition from a low to a high recycling target on the particle load were investigated and discussed in the linear plasma device, Magnum-PSI. Time traces of the target ion flux were mitigated in high neutral pressure cases because of a plasma-neutral interaction. On the other hand, in low neutral-pressure cases, the target ion flux indicated partial suppression in the last part of the pulse. The Langmuir probe, located 200 mm upstream from the target plate, did not exhibit such a suppression. Pulse suppression can be expected from the localized interaction between recycled neutral flux and pulsed plasma in front of the target. The mean-free paths of recycled neutral particles regarding the charge exchange with pulse ions and elastic scattering with background neutral particles were compared. Modeling using a fluid code coupled with a neutral transport code was performed, and it was concluded that dynamic pressure induced by the transient recycled neutral flux caused sufficient momentum loss to stagnate the pulsed plasma toward the target plate.

Experimental study on the role of the target electron temperature as a key parameter linking recycling to plasma performance in JET-ILW* *This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive,

Changes in global and edge plasma parameters (H 98(y,2), dimensionless collisionality ν *, core density peaking, separatrix density n e,sep) with variations in the D2 fueling rate and divertor configuration are unified into a single trend when mapped to ⟨T e,ot⟩, the spatially averaged spectroscopically derived outer target electron temperature. Dedicated JET with the ITER-like wall (JET-ILW) experiments in combination with an extended JET-ILW database of unseeded low-triangularity H-mode plasmas spanning a wide range of D2 fueling rates, I p, B t and heating power have demonstrated the importance of ⟨T e,ot⟩ as a key physics parameter linking the recycling particle source and detachment with plasma performance. The remarkably robust H 98(y,2) trend with ⟨T e,ot⟩ is connected to a strong inverse correlation between ⟨T e,ot⟩, n e,sep and ν *, thus directly linking changes in the divertor recycling moderated by ⟨T e,ot⟩ with the previously established relationship between ν *, core density peaking and core pressure resulting in a degradation in core plasma performance with decreasing ⟨T e,ot⟩ (increasing ν *). A strong inverse correlation between the separatrix to pedestal density ratio, n e,sep/n e,ped, and ⟨T e,ot⟩ is also established, with the rise in n e,sep/n e,ped saturating at ⟨T e,ot⟩ > 10 eV. A strong reduction in H 98(y,2) is observed as ⟨T e,ot⟩ is driven from 30 to 10 eV via additional D2 gas fueling, while the divertor remains attached. Consequently, the pronounced performance degradation in attached divertor conditions has implications for impurity seeding radiative divertor scenarios, in which H 98(y,2) is already low (∼0.7) before impurities are injected into the plasma since moderate gas fueling rates are required to promote high divertor neutral pressure. A favorable pedestal pressure, p e,ped, dependence on I p has also been observed, with an overall increase in p e,ped at I p = 3.4 MA as ⟨T e,ot⟩ is driven down from attached to high-recycling divertor conditions. In contrast, p e,ped is reduced with decreasing ⟨T e,ot⟩ in the lower I p branches. Further work is needed to (i) clarify the potential role of edge opacity on the observed favorable pedestal pressure I p scaling; as well as to (ii) project the global and edge plasma performance trends with ⟨T e,ot⟩ to reactor-scale devices to improve predictive capability of the coupling between recycling and confined plasma fueling in what are foreseen to be more opaque edge plasma conditions.

Particle control in long-pulse discharge using divertor pumping in LHD

Density control is crucial for maintaining stable confined plasma. Divertor pumping, where neutral particles are compressed and exhausted in the divertor region, was developed for this task for the Large Helical Device. In this study, neutral particle pressure, which is related to recycling, was systematically scanned in the magnetic configuration by changing the magnetic axis position. High neutral particle pressure and compression were obtained in the divertor for a high plasma electron density and the inner magnetic axis configuration. Density control using divertor pumping with gas puffing was applied to electron cyclotron heated plasma in the inner magnetic axis configuration, which provides high neutral particle compression and exhaust in the divertor. Stable plasma density and electron temperature were maintained with divertor pumping. A heat analysis shows that divertor pumping did not affect edge electron heat conductivity, but it led to low electron heat conductivity in the core caused by electron-internal-transport-barrier-like formation.

Physics design of new lower tungsten divertor for long-pulse high-power operations in EAST

A new lower tungsten divertor has been developed and installed in the EAST superconducting tokamak to replace the previous graphite divertor with power handling capability increasing from <2 MW m−2 to ∼10 MW m−2, aiming at achieving long-pulse H-mode operations in a full metal wall environment with the steady-state divertor heat flux of ∼10 MW m−2. A new divertor concept, ‘corner slot’ (CS) divertor, has been employed. By using the ‘corner effect’, a strongly dissipative divertor with the local buildup of high neutral pressure near the corner can be achieved, so that stable detachment can be maintained across the entire outer target plate with a relatively lower impurity seeding rate, at a separatrix density compatible with advanced steady-state core scenarios. These are essential for achieving efficient current drive with low-hybrid waves, a low core impurity concentration and thus a low loop voltage for fully non-inductive long-pulse operations. Compared with the highly closed small-angle-slot divertor in DIII-D, the new divertor in EAST exhibits the following merits: (1) a much simpler geometry with integral cassette body structure, combining vertical and horizontal target plates, which are more suitable for actively water-cooled W/Cu plasma facing components, facilitating installation precision control for minimizing surface misalignment, achieving high engineering reliability and lowering the capital cost as well; (2) it has much greater flexibility in magnetic configurations, allowing for the position of the outer strike point on either vertical or horizontal target plates to accommodate a relatively wide triangularity range, δ l = 0.4–0.6, thus enabling to explore various advanced scenarios. A water-cooled copper in-vessel coil has been installed under the dome. Five supersonic molecular beam injection systems have been mounted in the divertor to achieve faster and more precise feedback control of the gas injection rate. Furthermore, this new divertor allows for double null divertor operation and slowly sweeping the outer strike point across the horizontal and vertical target plates to spread the heat flux for long-pulse operations. Preliminary experimental results demonstrate the ‘corner effect’ and are in good agreement with simulations using SOLPS-ITER code including drifts. The EAST new divertor provides a test-bed for the closed divertor concept to achieve steady-state detachment operation at high power. Next step, a more closed divertor, ‘sharp-cornered slot’ divertor, building upon the current CS divertor concept, has been proposed as a candidate for the EAST upper divertor upgrade.

The role of target closure in detachment in Magnum-PSI

A cylindrical target with a high degree of closure was exposed to ITER divertor-relevant plasmas with typical electron temperatures of 2 eV, electron densities of 5⋅1020 m−3, and heat fluxes up to 20 MWm−2 in the linear device Magnum-PSI. By terminating the plasma in an unpumped closed volume, neutral pressures were enhanced from about 0.5 to 20 Pa without any increase in the neutral flux returning to the plasma. Such pressures were sustained largely by the pressure exerted by the incoming plasma. By means of hydrogen gas injection, internal neutral pressures of up to 40 Pa were reached during plasma exposure. We find that at these high neutral pressures, a &amp;lt; 1 eV recombination front forms and expands from the back of the cylinder, so that downstream density drops dramatically. Furthermore, in these scenarios, heat deposition to the back plate vanishes and is redirected to the upstream part of the cylinder and to hot neutrals, which can carry 50% of the plasma input power. A power balance analysis reveals that even without additional gas puffing, only about 10% of the incoming heat load reaches the back plate for the 20 MWm−2 plasma. These results demonstrate the important role of closed target configurations and local gas puffing in mitigating plasma heat loads and indicate that the gained experience should be taken into account in next-generation divertor designs.

Impact of the new TCV baffled divertor upgrade on pedestal structure and performance

Abstract A new set of carbon tiles, neutral beam heating optics and gas baffles were installed on TCV during the baffled divertor upgrade in early 2019. The installation of the baffles allows a deconvolution of the roles of main chamber and divertor neutral pressure on the H-mode pedestal structure. This physical barrier allows relatively high neutral pressures to be constrained to the divertor, thus preventing neutrals from entering the main chamber and potentially degrading core confinement. This study presents the experimentally measured and modelled pedestal heights and structure for a series of H-mode discharges prior to and after this upgrade. Increased pedestal performance at high divertor neutral pressure was observed after the baffled divertor upgrade. This was consistent across all triangularities and outer target locations investigated and is attributed to higher pedestal top temperatures being maintained at high gas injection rates. ASTRA simulations indicated beam heating power coupled to the plasma did not significantly vary after the baffled divertor upgrade or as a function of divertor neutral gas pressure. Analysis of the pedestal structure exposed a strong correlation between pedestal performance and the density pedestal position prior to and after the baffled divertor upgrade. The baffled divertor upgrade limited the outward shift of the density pedestal, thus maintaining higher pedestal performance at high divertor neutral pressures. Stability analysis indicated the majority of discharges studied were within 25% of the stability boundary. No correlation was found between the distance from the stability boundary and pedestal performance or structure. Comparison with the EPED1 model indicated that TCV discharges do not have a fixed dependence between pedestal β θ and pedestal width. A large variation in the EPED1 relating parameter was observed and found to vary with the density pedestal position.

Evidence for electron energization accompanying spontaneous formation of ion acceleration regions in expanding plasmas

We report experiments conducted in an expanding argon plasma generated in the inductive mode of a helicon source in the Hot hELIcon eXperiment–Large Experiment on Instabilities and Anisotropies facility. As the neutral gas pressure increases, the supersonic ion acceleration weakens. Increasing neutral pressure also alters the radial profile of electron temperature, density, and plasma potential upstream of the plasma expansion region. Langmuir probe measurements of the electron energy probability function (EEPF) show that heating of electrons at the plasma edge by RF fields diminishes with increasing gas pressure, yielding a plasma with a centrally peaked electron temperature, and flat potential profiles at higher neutral pressures. For neutral pressures at which ion acceleration regions develop in the expanding plasma plume, EEPFs reveal electrons with two temperature components.

Electron loss mechanisms in a miniature microwave discharge water neutralizer

This study analyzes the mechanism of electron loss at the discharge chamber wall of a microwave discharge neutralizer via three-dimensional particle-in-cell simulations with Monte Carlo collisions (PIC–MCCs). The neutralizer employs electron cyclotron resonance discharges with two ring-shaped permanent magnets and 4.2-GHz microwaves, where the plasma is confined by a magnetic mirror. The PIC–MCC simulation results show that the electron extraction efficiency of a water neutralizer can be increased by two times in an optimized magnetic field configuration, which is a higher increased rate than that of a xenon neutralizer. However, the efficiency of 20% is still low (e.g., less than half of the xenon one) because many electrons are lost to the magnet surface. The loss is determined to be due to approximately 5-times higher ratio of electrons inside the loss cone in the water neutralizer than that in the xenon neutralizer. The electron velocity distributions of each neutralizer clearly show that the water neutralizer has a larger fraction of electrons parallel to the magnetic field than the xenon neutralizer. This result is attributed to the large number of electron collisions in the water neutralizer owing to the high neutral gas pressure.

Correction for Neutral Pressure-Driven Signal in Radiated Power Measurements on Proto-MPEX

A pinhole camera using four-channel resistive bolometers was designed and implemented on the Prototype Material Plasma Exposure eXperiment (Proto-MPEX) to measure the plasma radiated power. Quantification of the radiated power is required for comprehensive power balance assessments but faces unique challenges on linear plasma devices due to weak absolute signal, compared to typical use in toroidal fusion plasmas, and high neutral pressure changes. Around 1 mTorr, the pressure-driven signal contribution was shown to overwhelm the radiated power signal in Proto-MPEX, found to be ~3 μW. Since the pressure-driven signals could not be predicted from first principle arguments, the signals were corrected using empirical models that take two experimental pressure measurements and the raw bolometer signal to reveal the radiated power signals. Four “gas-only” Proto-MPEX shots were taken, without the radio frequency (rf) plasma discharge, to train, optimize, and validate empirical models. The data included two pressure measurements at different locations along with the bolometer signal. Multiple models were tested; partial least-squares regression with a cross correlation method to account for the time offset in the pressure measurements provided the best combination of accuracy and stability. The model was validated to have a modeling error of 0.0036 V, which was lower than 10% of the expected radiated power signal. The model was tested on a different “gas-only” shot with a complete time-series data. For plasma operation shots, the corrected bolometer measurements were compared with an absolute extreme ultraviolet photodiode measurement and similarities and differences are discussed.

Experimental investigation of sheath effects on I–V traces of strongly electron emitting probes

I–V traces of strongly emitting emissive probes are investigated in a multidiople filament discharge. It is found that at sufficiently high neutral pressure and emitting current, the variation of the I–V traces and their associated inflection points no longer follow the previous predictions of space charge limited (SCL) models. A new, steep slope region of the I–V trace appears near the plasma potential when the probe is strongly emitting, causing the inflection point and the floating potential to increase towards the plasma potential as emission current increases, rather than staying constant. This is, to our knowledge, the first experimental evidence that the effects predicted by Campanell et al’s inverse sheath theory (2017 Physics of Plasmas 24 057101) not only affect the floating potential but also a region in the I–V trace of an emissive probe. It is also found that the double inflection point structure when the probe is biased below the ionization energy of the working gas is highly likely to be an emission retardation effect from enhanced virtual cathode formation due to the increased local electron density. The implications of these findings on hot cathode sources are briefly discussed.

Scrape-off layer transport and filament characteristics in high-density tokamak regimes

A detailed cross-device investigation on the role of filamentary dynamics in high-density regimes has been performed within the EUROfusion framework, comparing the ASDEX Upgrade (AUG) and TCV tokamaks. Both devices run density ramp experiments at different levels of plasma current, keeping the toroidal field or q95 constant in order to disentangle the role of the parallel connection length and the current. During the scan at a constant toroidal field, in both devices the scrape-off layer (SOL) profiles tend to develop a clear SOL density shoulder at a lower edge density whenever the current is reduced. Different current behaviour is substantially reconciled in terms of the edge density normalized to the Greenwald fraction. During the scan at constant q95 AUG exhibits similar behaviour, whereas in TCV no upstream profile modification signature has been observed at lower current levels. The latter behaviour has been ascribed to the lack of target density rollover. The relation between the upstream density profile modification and detachment condition has been investigated. For both devices the relation between blob size and the SOL density e-folding length is found independent of the plasma current, with the observation of a clear increase in blob size and the edge density normalized to a Greenwald fraction. ASDEX Upgrade has also explored filamentary behaviour in the H-mode. The experiments in AUG have focused on the role of neutrals, performing discharges with and without cryogenic pumps, highlighting how high neutral pressure, not only in the divertor but also at the midplane, is needed in order to develop an H-mode SOL profile shoulder in AUG.

Dependence of neutral pressure on detachment in the small angle slot divertor at DIII-D

Abstract Local neutral pressure measurements in the closed small angle slot (SAS) divertor [1,2] on DIII-D show a large increase when the divertor plasma shifts from high recycling into detachment. In-tile pressure gauges were installed to measure the pressure in the near- and far-SOL regions to examine the predicted high neutral pressures and compression. Cross-field drift effects lead to ∼ 10 × higher peak neutral pressure in detachment with the ion ∇B drift toward the divertor compared to out of the divertor, with similar pressures in attached conditions. Drifts also play a role in the neutral distribution in the slot while attached but become less pronounced in detachment once gradient drives reduce. Variation in the outer strike point location found higher neutral pressure and detachment at modestly lower main-plasma density with the strike point positioned away from the designed operation point. Reducing the fraction of the SOL width allowed into the slot increases neutral leakage into the main chamber and increases the main-plasma density required for detachment. Preliminary modeling with the SOLPS code without drifts over-predicts the neutral pressure in detachment by a factor of 2 with the strike point in the designed operation point while more significantly over-predicts the neutral slot compression; experiments show a broader distribution of neutrals through the slot. These measurements are used to help understand detachment and validate divertor design metrics.

Neutral pressure and separatrix density related models for seed impurity divertor radiation in ASDEX Upgrade

Analytical expressions for the divertor radiation around the onset of detachment based on the divertor neutral pressure and the upstream separatrix density are tested against measurements on strongly nitrogen seeded high confinement mode (H-mode) discharges in ASDEX Upgrade. The divertor pressure is used as a key experimental parameter as far as possible, since it can be regarded as engineering parameter based on the balance of puffing and pumping. For a tentative extrapolation to future tokamak power plants, limits set by the maximum core impurity concentration via the burn conditions are related to the divertor impurity content via the divertor enrichment factor. Combined with the H-mode density limit, which is most likely a limit for the separatrix density nesep/nGreenwald ≤ 0.5, estimates for the maximum separatrix power Psep compliant with the achievement of detachment are presented for a density related 2-point model and a flux related radiation model. The detachment, which is assumed necessary for the achievement of sufficiently long target lifetimes, requires in H-mode the dissipation of a major fraction of the power flowing to the divertor by impurity radiation. Favorable conditions for power exhaust feature a high Greenwald density / high divertor neutral pressure. The latter also facilitates a faster helium exhaust, which increases the margin for the seed impurity concentration in the core. Major uncertainties in the extrapolation of seeded scenarios regard the divertor impurity enrichment and the power width λ in the divertor for detached conditions.

Particle‐in‐cell/Monte Carlo simulation of electron and ion currents to cylindrical Langmuir probe

Electron and ion currents to a cylindrical Langmuir (electrostatic) probe were calculated using the particle‐in‐cell/Monte Carlo (PIC/MC) self‐consistent simulation for a neutral gas in the pressure range 2–3,000 Pa. The simulation enables us to calculate the probe currents even at high neutral gas pressures when the collisions of collected charged particles with neutral gas particles near the probe are important. The main aim of this paper is the calculation of probe currents at such high gas pressures and the comparison of the results with experimentally measured probe currents. Simulations were performed for two cases: (a) probes with varying radii in a non‐thermal plasma with high electron temperature at low neutral gas pressure of 2 Pa (in order to verify the correctness of our simulations), and (b) probe with the radius of 10 μm in the afterglow plasma with low electron temperature and a higher neutral gas pressure (up to 3,000 Pa). The electron probe currents obtained in case (a) show good agreement with those predicted by the orbital motion limited current (OMLC) theory for probes with radii up to 100 μm for the given plasma conditions. At larger probe radii and/or at higher probe voltages, the OMLC theory incorrectly predicts too high an electron probe current for the plasma parameters studied. Additionally, a formula describing the spatial dependence of the electron density in the presheath in the collisionless case is derived. The simulation at higher neutral gas pressures, i.e. case (b), shows a decrease of the electron probe current with increasing gas pressure and the creation of a large presheath around the probe. The simulated electron probe currents are compared with those of measurements by other authors, and the differences are discussed.