•First experimental study on HCFP using helium at prototypical pressure.•Thermal-fluids correlations derived from experimental data.•Thermal-fluids performance estimated at prototypical conditions.
...Various helium (He)-cooled solid-tungsten (W) divertor concepts have been proposed for long-pulse magnetic fusion energy reactors. Among these concepts, the He-cooled flat plate divertor (HCFP) modules have the largest plasma-facing surface area of ∼0.2 m2. Simulations have shown that the most recent version of the design can withstand heat fluxes as great as 8 MW/m2. Earlier experimental studies of a single shortened HCFP cooling unit with a slot length of 7.6 cm used air at ambient temperature and pressures below 0.6 MPa. Here, we present initial experimental studies of a copper alloy and steel test section modeling a shortened HCFP cooling unit using He at prototypical pressure of 10 MPa, inlet temperatures Ti ≤ 200 °C and steady-state incident heat fluxes q″ ≤ 1.2 MW/m2. Results for Nusselt number Nu as a function of Reynolds number Re were obtained for Re = 1.2×104−3.4×104, and used to develop a Nu(Re) correlation and validate numerical models of the test section using commercial computational fluid dynamics (CFD) software. Simulations with this model are performed to evaluate the effect of the shortened slot. These analyses are used to estimate the thermal-fluids performance of the HCFP under prototypical conditions.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UILJ, UL, UM, UPCLJ, UPUK, ZAGLJ, ZRSKP
Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field ...lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma E ⃗ × B ⃗ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.
Abstract
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.
Detailed analysis of convective fluxes caused by E × B drifts is carried out in a realistic JET configuration, based on a series of EDGE2D-EIRENE runs. The EDGE2D-EIRENE code includes all guiding ...centre drifts, E × B as well as ∇B and centrifugal drifts. Particle sources created by divergences of radial and poloidal components of the E × B drift are separately calculated for each flux tube in the divertor. It is demonstrated that in high recycling divertor conditions radial E × B drift creates particle sources in the common flux region (CFR) consistent with experimentally measured divertor and target asymmetries, with the poloidal E × B drift creating sources of an opposite sign but smaller in absolute value. That is, the experimentally observed asymmetries in the CFR are the opposite to what poloidal E × B drift by itself would cause. In the private flux region (PFR), the situation is reversed, with poloidal E × B drift being dominant. In this region poloidal E × B drift by itself contributes to experimentally observed asymmetries. Thus, in each region, the dominant component of the E × B drift acts so as to create the density (and hence, also temperature) asymmetries that are observed both in experiment and in 2D edge fluid codes. Since the total number of charged particles is much greater in the CFR than in PFR, divertor asymmetries caused by the E × B drift should be attributed primarily to particle sources in the CFR caused by radial E × B drift.
During the ITER design phase, the focus of ITER boundary plasma modeling activities has been on divertor performance under baseline H-mode, fusion power operation (FPO) conditions. However, early ...ITER operation will be primarily with hydrogen fuel in L-mode, in the pre-fusion power operation 1 (PFPO-1) phase. Here, the SOLPS-ITER code is used to evaluate divertor performance during this non-active phase. To verify the assumptions used in the existing high power simulation database, gas throughput scans were performed for two types of divertor surface material (beryllium and tungsten) and two gas puff locations (divertor and main chamber). The adoption of beryllium target surfaces simulates the effect of main chamber material erosion and migration and, along with main chamber gas injection, is the current default for the high power database. Depending on the divertor surface material, the atom to molecule ratio of the recycled neutral particles varies. This modifies the momentum and power loss mechanisms arising from plasma-neutral interactions. However, since the effect of atomic and molecular reactions are compensatory, the 'total' power and momentum losses are relatively insensitive to the target surface material. Similarly, the impact of gas puff location on divertor plasma parameters is not significant, though main chamber injection provides an additional ionization source in the upstream scrape-off layer (SOL) and leads to moderate changes in the upstream density and far SOL parameters. However, these effects can be neglected within the available range of the gas puff and pump rates in ITER. Since beryllium and tungsten are materials at both extremes in terms of surface reflection properties, the conclusions may be applicable to other divertor surface materials. An important additional finding of the study is that the insensitivity of upstream density to divertor neutral pressure found in the FPO database is also recovered in these PFPO-1 simulations.
The process of divertor detachment, whereby heat and particle fluxes to divertor surfaces are strongly diminished, is required to reduce heat loading and erosion in a magnetic fusion reactor to ...acceptable levels. In this paper, the physics leading to the decrease of the total divertor ion current (It), or 'roll-over', is experimentally explored on the TCV tokamak through characterization of the location, magnitude and role of the various divertor ion sinks and sources including a complete analysis of particle and power balance. These first measurements of the profiles of divertor ionisation and hydrogenic radiation along the divertor leg are enabled through novel spectroscopic techniques. Over a range in TCV plasma conditions (plasma current and electron density, with/without impurity-seeding) the It roll-over is ascribed to a drop in the divertor ion source; recombination remains small or negligible farther into the detachment process. The ion source reduction is driven by both a reduction in the power available for ionization, Precl, and concurrent increase in the energy required per ionisation, Eion: this effect of power available on the ionization source is often described as 'power starvation' (or 'power limitation'). The detachment threshold is found experimentally (in agreement with analytic model predictions) to be ~Precl/ItEion ~ 2, corresponding to a target electron temperature, Tt ~ Eion/γ where γ is the sheath transmission coefficient. The target pressure reduction, required to reduce the target ion current, is driven both by volumetric momentum loss as well as upstream pressure loss. The measured evolution through detachment of the divertor profile of various ion sources/sinks as well as power losses are quantitatively reproduced through full 2D SOLPS modelling through the detachment process as the upstream density is varied.
The effect of the upcoming TCV divertor upgrade on the distribution of neutrals and the onset of detachment is studied using 2D transport code simulations. The divertor upgrade is centered around the ...installation of a gas baffle to form a divertor chamber of variable closure. SOLPS-ITER simulations predict that the baffle geometry selected to be installed in TCV in 2019 increases the divertor neutral density by a factor ∼5 and the neutral compression by one order of magnitude in typical TCV single null, Ohmic heated scenarios (330 kW). The compression increases further with the addition of auxiliary heating systems (1.2 MW). Simulations show that volumetric power losses in the divertor increase giving access to deeper detachment for given upstream densities and heating power. Predictions for observations by various TCV diagnostics, including baratrons, divertor spectrometer and visible camera systems, are presented to guide the experimental verification of the efficiency of the divertor baffles.
The present vision for a plasma-material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma ...power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertor configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). This paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.
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