On the role of finite grid extent in SOLPS-ITER edge plasma simulations for JET H-mode discharges with metallic wall

Peer reviewe

Towards radiation transport modelling in divertors with the EIRENE code

Opacity effects, in particular of Lyman lines, in divertors are believed to be relevant for plasma spectroscopy and for the overall divertor dynamics through possible redistribution of excited hydrogen atoms and radiation losses. Quite elaborate computational radiation transport tools have been developed, specialized for numerous applications. The task in fusion research has been adaptation to fusion edge plasma conditions. In this paper, we start from an existing kinetic neutral particle code already well adapted to divertor applications, and extend this from the 'particle' simulation to an analogue 'photon gas' simulation. It is shown how this can be achieved and that a quite flexible and detailed divertor radiation transport code can conveniently be obtained. We apply this to study Lyman opacity effects on population kinetics and hydrogen divertor radiation losses

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate.Peer reviewe

Perceptions of Business Skill Development by Graduates of the University of Michigan Dental School

Peer Reviewedhttps://deepblue.lib.umich.edu/bitstream/2027.42/153743/1/jddj002203372011754tb05074x.pd

On the role of finite grid extent in SOLPS-ITER edge plasma simulations for JET H-mode discharges with metallic wall

The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters lambda(div)(q). Otherwise and further upstream the mid-plane heat decay length lambda(q) parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that lambda(q) upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations lambda(ne)>lambda(Te)>lambda(q) hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of lambda(ne) into account, rather than lambda(q), as the former actually defines the required minimum upstream SOL grid extent.EURATOM 63305

Beryllium global erosion and deposition at JET-ILW simulated with ERO2.0

The recently developed Monte-Carlo code ERO2.0 is applied to the modelling of limited and diverted discharges at JET with the ITER-like wall (ILW). The global beryllium (Be) erosion and deposition is simulated and compared to experimental results from passive spectroscopy. For the limiter configuration, it is demonstrated that Be self-sputtering is an important contributor (at least 35%) to the Be erosion. Taking this contribution into account, the ERO2.0 modelling confirms previous evidence that high deuterium (D) surface concentrations of up to similar to 50% atomic fraction provide a reasonable estimate of Be erosion in plasma-wetted areas. For the divertor configuration, it is shown that drifts can have a high impact on the scrape-off layer plasma flows, which in turn affect global Be transport by entrainment and lead to increased migration into the inner divertor. The modelling of the effective erosion yield for different operational phases (ohmic, L- and H-mode) agrees with experimental values within a factor of two, and confirms that the effective erosion yield decreases with increasing heating power and confinement.EURATOM 63305

Characterisation of the deuterium recycling at the W divertor target plates in JET during steady-state plasma conditions and ELMs

Experiments in the JET tokamak equipped with the ITER-like wall (ILW) revealed that the inner and outer target plate at the location of the strike points represent after one year of operation intact tungsten (W) surfaces without any beryllium (Be) surface coverage. The dynamics of nearsurface retention, implantation, desorption and recycling of deuterium (D) in the divertor of plasma discharges are determined by W target plates. As the W plasma-facing components (PFCs) are not actively cooled, the surface temperature (Tsurface) is increasing with plasma exposure, varying the balance between these processes in addition to the impinging deuteron fluxes and energies. The dynamic behaviour on a slow time scale of seconds was quantified in a series of identical L-mode discharges (JET Pulse Number (JPN) # - 81938 73) by intra-shot gas analysis providing the reduction of deuterium retention in W PFCs by 1/3 at a base temperature (Tbase) range at the outer target plate between 65 °C and 150 °C equivalent to a Tsurface span of 150 °C and 420 °C. The associated recycling and molecular D desorption during the discharge varies only at lowest temperatures moderately, whereas desorption between discharges rises significantly with increasing Tbase. The retention measurements represent the sum of inner and outer divertor interaction at comparable Tsurface. The dynamic behaviour on a fast time scale of ms was studied in a series of identical H-mode discharges (JPN # - 83623 83974) and coherent edge-localized mode (ELM) averaging. High energetic ELMs of about 3 keV are impacting on the W PFCs with fluxes of 3 ´ 10 D s m 23 1 +- -2 which is about four times higher than inter-ELM ion fluxes with an impact energy of about Eim = 200 eV. This intra-ELM ion flux is associated with a high heat flux of about 60 MW m−2 to the outer target plate which causes Tsurface rise by Δ T = 100 K per ELM covering finally the range between 160 °C and 1400 °C during the flat-top phase. ELM-induced desorption from saturated nearsurface implantation regions as well as deep ELM-induced deuterium implantation areas under varying baseline temperature takes place. Subsequent refuelling by intra-ELM deuteron fluxes occurs and a complex interplay between deuterium fuelling and desorption can be observed in the temporal ELM footprint of the surface temperature (IR thermography), the impinging deuteron flux (Langmuir probes), and the Balmer radiation (emission spectroscopy) as representative for the deuterium recycling flux. In contrast to JET-C, a pronounced second peak, ; 8 ms delayed with respect to the initial ELM crash, in the Dα radiation and the ion flux has been observed. The peak can be related to desorption of implanted energetic intra-ELM D+ diffusing to the W surface, and performing local recycling.EURATOM 63305

Impact of the JET ITER-like wall on H-mode plasma fueling

JET ITER-like wall (ILW) experiments show that the edge density evolution is strongly linked with the poloidal distribution of the ionization source. The fueling profile in the JET-ILW is more delocalized as compared to JET-C (JET with carbon-based plasma-facing components PFCs). Compared to JET-C the H-mode pedestal fueling cycle is dynamically influenced by a combination of plasma–wall interaction features, in particular: (1) edgelocalized modes (ELMs) induced energetic particles are kinetically reflected on W divertor PFCs leading to distributed refueling away from the divertor depending on the divertor plasma configuration, (2) delayed molecular re-emission and outgassing of particles being trapped in W PFCs (bulk-W at the high field side and W-coated CFCs at the low field side) with different fuel content and (3) outgassing from Be co-deposits located on top of the highfield side baffle region shortly after the ELM. In view of the results of a set of well diagnosed series of JET-ILW type-I ELMy H-mode discharges with good statistics, the aforementioned effects are discussed in view of H-mode pedestal fueling capacity. The ongoing modelling activities with the focus on coupled core-edge plasma simulations and plasma–wall interaction are described and discussed also in view of possible code improvements required.EURATOM 63305

Recent progress of plasma exhaust concepts and divertor designs for tokamak DEMO reactors

The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (frad main = Prad main/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Prad div) for larger Psep/Rp ≥ 25MWm− 1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (frad div = Prad div/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm− 2 for the large frad div = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm− 2 . Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 ◦C and 200 ◦C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike point when the high qtarget of ~10 MWm− 2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 ◦C/ 5 MPa (EU-DEMO) and 290 ◦C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design

Effect of PFC Recycling Conditions on JET Pedestal Density

There is experimental evidence that the pedestal dynamics in type-I ELMy H-mode discharges is significantly affected by a change in the recycling conditions at the tungsten plasma-facing components (W-PFCs) after an ELM event. The integrated code JINTRAC has been employed to assess the impact of recycling conditions during type-I ELMs in JET ITER-like wall H-mode discharges. By employing a heuristic approach, a model to mimic the physical processes leading to formation and release (i.e. outgassing) of finite near-surface fuel reservoirs in W-PFCs has been implemented into the EDGE2D-EIRENE plasma-wall interaction code being part of JINTRAC. As main result it is shown, that a delay in the density pedestal build-up after an ELM event can be provoked by reduced recycling induced by depleted W-PFC particle near-surface reservoirs. However the pedestal temperature evolution is barely affected by the change in recycling parameters suggesting that the presented model is incomplete.EURATOM 63305