Development of DTT single null divertor scenario

Abstract This paper focuses on scrape-off layer and divertor modelling of the medium-density single-null scenario of the Divertor Test Tokamak facility (DTT), under construction in Italy. The modelling was performed using the 2D coupled fluid-Monte Carlo code SOLEDGE2D-EIRENE. For DTT pump designing, neutral pressure at the pump aperture below the dome is calculated in deuterium-only cases as well as with impurity seeding with various puffing levels. This scenario analysis also allowed the characterization of detachment in DTT and the influence of pumping on detachment itself. Two different radiating impurities, neon and nitrogen, were tested in the high power scenario to evaluate the minimum impurity concentration required to achieve sustainable conditions at DTT divertor. The sensitivity of the model was studied by varying the impurity concentration; the model shows a hysteresis-like behaviour between the impurity influx and the total impurity content by which detachment is strongly influenced

Fluctuation characteristics of the TCV snowflake divertor measured with high speed visible imaging

Tangentially viewing fast camera footage of the low-field side snowflake minus divertor in TCV is analysed across a four point scan in which the proximity of the two X-points is varied systematically. The motion of structures observed in the post- processed movie shows two distinct regions of the camera frame exhibiting differing patterns. One type of motion in the outer scrape-off layer remains present throughout the scan whilst the other, apparent in the inner scrape-off layer between the two nulls, becomes increasingly significant as the X-points contract towards one another. The spatial structure of the fluctuations in both regions is shown to conform to the equilibrium magnetic field. When the X-point gap is wide the fluctuations measured in the region between the X-points show a similar structure to the fluctuations observed above the null region, remaining coherent for multiple toroidal turns of the magnetic field and indicating a physical connectivity of the fluctuations between the upstream and downstream regions. When the X-point gap is small the fluctuations in the inner scrape-off layer between the nulls are decorrelated from fluctuations upstream, indicating local production of filamentary structures. The motion of filaments in the inter-null region differs, with filaments showing a dominantly poloidal motion along magnetic flux surfaces when the X-point gap is large, compared to a dominantly radial motion across flux-surfaces when the gap is small. This demonstrates an enhancement to cross-field tranport between the nulls of the TCV low-field-side snowflake minus when the gap between the nulls is small.Comment: Accepted for publication in Plasma Physics and Controlled Fusio

3D numerical study of neutral gas dynamics in the DTT particle exhaust using the DSMC method

Recently the design of the divertor tokamak test (DTT) Facility divertor has been modified and consolidated. The new divertor design presents significant geometrical differences compared to the previous ITER-like one, including the presence of a more flattened dome shape. This paper presents a complete 3D numerical analysis of the neutral gas dynamics inside the DTT subdivertor area for the latest divertor design. The analysis has been performed based on the direct simulation Monte Carlo method by applying the DIVGAS simulator code. SOLEDGE2D-EIRENE plasma simulations have been performed for a deuterium plasma scenario at the maximum additional power in partially detached condition achieved by neon impurity seeding and the extracted information about the neutral particles has been imposed as incoming boundary conditions. The pumping efficiency of the DTT divertor is examined by considering various cases with respect to the pumping probability and the effect of the toroidal and poloidal leakages is quantified. The results show that a significant percentage of the incoming flux of neutrals returns back to the plasma site through the entry gaps (60% for deuterium and 40% for neon), and, consequentially, only a small percentage (∼2%–15%) of the incoming flux can be pumped out from the system. The toroidal leakages affect significantly the pumping performance of the divertor causing a significant decrease in the pumped flux (and also in the pressure at the pumping opening) about 37%–47% and 43%–56% for deuterium and neon respectively. It is discussed how many pumping ports are needed depending on the achievable pumping performance per port. The number can be reduced by closing the toroidal gaps. The analysis shows that enlarging the poloidal gaps by a factor of two causes a significant increase in the poloidal flux losses by a factor 1.7. It is also illustrated how the presence of the cooling pipes leads to conductance losses

Heat flux measurements and modelling in the RFX-mod experiment

The knowledge of edge plasma transport parameters and plasma edge phenomena is a key element in the design of the first wall for a magnetically confined fusion experiment. In RFX-mod heat flux measurement and edge transport modelling have been done to improve the understanding of this aspect. Heat flux deposition profiles have been evaluated from infrared temperature measurements of insertable graphite limiters. They were inserted up to 12 mm into the reversed field pinch plasma of ohmically heated discharges with Ip= 0.6÷1.0 MA, ne= 0.5÷3·1019 m−3 (n/nG< 0.7) and total power of about 10÷15 MW. Strong asymmetries in heat flux deposition have been measured in poloidal direction at low density between the electron and the ion drift side and smaller ones in toroidal direction when q(a)≠0. The poloidal asymmetry has been associated to the presence of superthermal electrons [1] while the toroidal one has been less clearly identified as due to the small toroidal extension of the limiters. To account for the 2D deposition nature of heat load on the surface of the employed limiters, a simple 3D code has been developed to evaluate heat flux from temperature data. In this way at the deeper limiter insertions a heat flux decay length of about 2 mm and 2.5 mm has been evaluated in electron and ion drift sides. Modelling of the evaluated heat fluxes has been done using the SOLEDGE2D-EIRENE edge code [2]. This fluid code is well suited for the RFX-mod wall limiter configuration because, thanks to the implemented penalization technique, the computational domain can be extended up to the entire first wall. Edge modelling has shown that measured decay lengths are compatible with energy diffusion coefficients in Scrape Off Layer (SOL) smaller than those commonly evaluated at plasma edge; the cause of the reduced diffusion in the SOL will be discussed in the paper

The plasma boundary in Single Helical Axis RFP plasmas

Single Helical Axis (SHAx) states obtained in high current reversed field pinch (RFP) plasmas display, aside from a dominant mode in the m=1 spectrum, also a dominant m=0 mode, with the same toroidal mode number as the m=1 one. The two modes have a fixed phase relationship. The island chain created by the m=0 mode across the reversal surface gives rise, at shallow reversal of the toroidal field, to an X-point structure which separates the last closed flux surface from the first wall, creating a divertor-like configuration. The plasma-wall interaction is found to be related to the connection length of the field lines intercepting the wall, which displays a pattern modulated by the dominant mode toroidal periodicity. This configuration, which occurs only for shallow toroidal field reversal, could be exploited to realize an island divertor in analogy to stellarators.Comment: 12 pages, 9 figures Submitted to Nuclear Fusio

Persistent storage of non-event data in the CMS databases

In the CMS experiment, the non event data needed to set up the detector, or being produced by it, and needed to calibrate the physical responses of the detector itself are stored in ORACLE databases. The large amount of data to be stored, the number of clients involved and the performance requirements make the database system an essential service for the experiment to run. This note describes the CMS condition database architecture, the data-flow and PopCon, the tool built in order to populate the offline databases. Finally, the first results obtained during the 2008 and 2009 cosmic data taking are presented.In the CMS experiment, the non event data needed to set up the detector, or being produced by it, and needed to calibrate the physical responses of the detector itself are stored in ORACLE databases. The large amount of data to be stored, the number of clients involved and the performance requirements make the database system an essential service for the experiment to run. This note describes the CMS condition database architecture, the data-flow and PopCon, the tool built in order to populate the offline databases. Finally, the first experience obtained during the 2008 and 2009 cosmic data taking are presented

Cross-code comparison of the edge codes SOLPS-ITER, SOLEDGE2D and UEDGE in modelling a low-power scenario in the DTT

As reactor-level nuclear fusion experiments are approaching, a solution to the power exhaust issue in future fusion reactors is still missing. The maximum steady-state heat load that can be exhausted by the present technology is around 10 MW m-2. Different promising strategies aiming at successfully managing the power exhaust in reactor-relevant conditions such that the limit is not exceeded are under investigation, and will be tested in the Divertor Tokamak Test (DTT) experiment. Meanwhile, the design of tokamaks beyond the DTT, e.g. EU-DEMO/ARC, is progressing at a high pace. A strategy to work around the present lack of reactor-relevant data consists of exploiting modelling to reduce the uncertainty in the extrapolation in the design phase. Different simulation tools, with their own capabilities and limitations, can be employed for this purpose. In this work, we compare SOLPS-ITER, SOLEDGE2D and UEDGE, three state-of-the-art edge codes heavily used in power exhaust studies, in modelling the same DTT low-power, pure-deuterium, narrow heat-flux-width scenario. This simplified, although still reactor-relevant, testbed eases the cross-comparison and the interpretation of the code predictions, to identify areas where results differ and develop understanding of the underlying causes. Under the conditions investigated, the codes show encouraging agreement in terms of key parameters at both targets, including peak parallel heat flux (1%-45%), ion temperature (2%-19%), and inner target plasma density (1%-23%) when run with similar input. However, strong disagreement is observed for the remaining quantities, from 30% at outer mid-plane up to a factor 4-5 at the targets. The results primarily reflect limitations of the codes: the SOLPS-ITER plasma mesh not reaching the first wall, SOLEDGE2D not including ion-neutral temperature equilibration, and UEDGE enforcing a common ion-neutral temperature. Potential improvements that could help enhance the accuracy of the code models for future applications are also discussed

Core integrated simulations for the Divertor Tokamak Test facility scenarios towards consistent core-pedestal-SOL modelling

Deuterium plasma discharges of the Divertor Tokamak Test facility (DTT) in different operational scenarios have been predicted by a comprehensive first-principle based integrated modelling activity using state-of-art quasi-linear transport models. The results of this work refer to the updated DTT configuration, which includes a device size optimisation (enlargement to R-0=2.19 a = 0.70 m) and upgrades in the heating systems. The focus of this paper is on the core modelling, but special attention was paid to the consistency with the scrape-off layer parameters required to achieve divertor plasma detachment. The compatibility of these physics-based predicted scenarios with the electromagnetic coil system capabilities was then verified. In addition, first estimates of DTT sawteeth and of DTT edge localised modes were achieved