Bulletin of the American Physical Society
60th Annual Meeting of the APS Division of Plasma Physics
Volume 63, Number 11
Monday–Friday, November 5–9, 2018; Portland, Oregon
Session GM9: Mini-Conference on Plasma–Material Interactions in Fusion Devices: ITER and Beyond. I. Integrated PMI Modeling and Analysis |
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Chair: IIon Joseph, Lawrence Livermore National Laboratory Room: OCC C123 |
Tuesday, November 6, 2018 9:30AM - 9:55AM |
GM9.00001: Operating range for the ITER tungsten divertor Richard A Pitts The first ITER tungsten (W) divertor is the largest and most complex of its kind ever to be constructed. It must survive an expected ~2500 hours of plasma exposure through the non-active to the first DT campaigns. A key parameter determining the operational range for stationary loading is the peak target power flux density, qpk, itself fixed by the allowable surface temperature to remain below W recrystallization. There is a strong relationship between qpk and the divertor neutral pressure, pn, the latter crucial for adequate He exhaust during burning plasmas and strongly influenced by the transparency for neutral recirculation between inner and outer targets. The main divertor design features were established using a database generated with the SOLPS code, which found an acceptable operational range within the constraints fixed by the input asumptions. However, some of the latter require modification in the light both of recent improvements in the physics model and the need for divertor component shaping. All tend to push the operating space to higher pn and higher concentration of extrinsic impurities with increased upstream densities. Nevertheless, there appears to be sufficient margin to avoid the deep detachment often associated with operational limits in today‘s devices. |
Tuesday, November 6, 2018 9:55AM - 10:20AM |
GM9.00002: Status of integrated simulation of PFC surface evolution within the PSI-SciDAC project John Canik, Ane Lasa, Sophie Blondel, Mark R Cianciosa, Davide Curreli, Jon T Drobny, Wael Elwasif, David L Green, Philip C Roth, Tim Younkin, Brian Wirth, Russ Doerner, Daisuke Nishijima, Matthew Baldwin A new simulation capability is being developed within the PSI-SciDAC project to simulate the long-term evolution of plasma-exposed surfaces. The integrated model includes a wide range of phenomena, including models for a) the scrape-off layer plasma including fuel ions and extrinsic impurities (using SOLPS[1]), b) sheath physics (using the hPIC code), c) transport and redeposition of eroded wall material (using the new Monte Carlo code GITR), d) implantation of ions into the wall and subsequent erosion (using F-TRIDYN, an extension of TRIDYN [2]), and e) dynamics of the subsurface (Xolotl, a new continuum cluster dynamics code). These components have been combined to predict the evolution of surface morphology, recycling and retention, and the impact of erosion and redeposition on these processes, assuming steady-state conditions and without feedback on the background plasma. After benchmarking against PISCES experiments, we have now applied this model to make predictions for the ITER divertor. Predictions for standard and helium operations, and for a partially and completely detached divertor, will be presented. [1] R. Schneider et al, Contrib. Plasma Phys. 46 (2006) 3. [1] W. Miller et al, Comp. Phys. Comm. 51 (1988) 355. |
Tuesday, November 6, 2018 10:20AM - 10:45AM |
GM9.00003: Coupled PIC-BCA simulations of the near-surface plasma at the ITER Outer Target Davide Curreli, Jon T Drobny, Ane Lasa, Sophie Blondel, John Canik, David L Green, Tim Younkin, Brian Wirth In this work we present the methodology adopted for modeling the near-surface plasma of ITER’s Outer Target including the sheath/presheath region and the surface gross erosion. The full-orbit Particle-in-Cell hPIC has been adapted to accept inputs from a SOLPS plasma background near the surface. The code resolves the structure of the magnetic presheath / Debye sheath, and produces energy-angle distributions of the ions impacting on the surface which are then passed as an input to the sputtering code Fractal-TRIDYN. Simulations have been run on a region spanning over two meters (poloidally) along the surface of ITER’s Outer Target. We report the calculated energy-angle distributions of the ions at the surface, the expected sputtering yields, and the sputtering fluxes produced in nominal ITER conditions for both helium and DT discharges. |
Tuesday, November 6, 2018 10:45AM - 11:10AM |
GM9.00004: Multiscale Modeling of Plasma Surface Interactions with Coupled Boundary Plasma Physics Brian Wirth This talk presents results of a coordinated, multiscale project to develop and deploy validated, high-performance computational codes to predict the boundary plasma and the evolving surfaces of plasma facing components (PFC). A focus is on the sub-surface gas dynamics in tungsten, which influence tritium retention and material surface evolution, as well as the computational framework for coupling the boundary plasma physics to the surface response. Our approach to modeling PFC surface response is based on simultaneously attacking this problem from both a “bottom-up” atomistic approach, as well as from a “top-down” continuum perspective that focuses on kinetic models of species reactions and diffusion. This presentation will also introduce a new, depth dependent surface characterization technique to validate the multiscale models of sub-surface D-He gas dynamics in tungsten. The presentation concludes with a perspective on divertor and PFC performance in fusion devices beyond ITER. |
Tuesday, November 6, 2018 11:10AM - 11:35AM |
GM9.00005: The influence of plasma-surface interactions on tungsten performance in ITER Gregory De Temmerman Tungsten (W) monoblocks in the ITER divertor targets will be exposed to particle fluences far above the highest values reached in today’s tokamak experiments. Such conditions can lead to significant surface modifications, with a key example being the formation of nano-bubbles and fuzz under He irradiation. To assess whether fuzz may form in ITER, a growth/annealing equilibrium model, including the effect of edge-localized modes (ELM) and the reduced thermal conductivity of fuzz, has recently been proposed which can explain the experimentally observed surface temperature window for fuzz formation. Applying this model to ITER reveals that for an ELM energy density of 0.1MJ.m-2 (lower than required to avoid monoblock edge melting and surface roughening), the maximum fuzz thickness is limited to 1-2 mm by the ELM-induced fast annealing. The mechanical and thermal properties of the W material can be modified during operations. In particular, W recrystallization, which leads to a decrease in material strength, is a key effect due the high surface temperatures in ITER. An approach is proposed to develop an operational budget for the W material, i.e. the time the divertor material can be operated at a given temperature before a significant fraction of the material is recrystallized. |
Tuesday, November 6, 2018 11:35AM - 12:00PM |
GM9.00006: What post mortem analysis tells us of plasma surface interactions in JET Anna Widdowson, JET Contributors Periodic removal of beryllium and tungsten plasma facing components (PFCs) from the JET tokamak allows for surface analysis to be carried out. Measurement of components from around the vessel enable both local erosion, deposition and fuel retention phenomena to be characterised and global material migration and fuel retention to be mapped. Whilst extraction of PFCs within operating periods and systematic parameter studies of materials are not possible on JET, the samples that are removed at the end of an operating period are unique in that they represent long term PFC exposure in a tokamak. The analysis and interpretation of JET PFCs for this reason is challenging as they result from a varied plasma programme. Despite this, valuable insight into the global parameters affecting material migration and fuel retention may be obtained and bench marking of models is possible. An overview of post mortem analysis in JET shows the influence of plasma configuration on deposition patterns - with fuel retention being dominated by co-deposition at the upper inner divertor, evidence of prompt deposition of eroded materials and longer-range migration to remote areas where mirrors are degraded, dust inventories and related sources and potential PFC lifetime issues from melt events and erosion. |
Tuesday, November 6, 2018 12:00PM - 12:20PM |
GM9.00007: On coupled plasma-wall instabilities. Sergei Krasheninnikov Although approximate, but rather simple approach to numerical study of coupled plasma-wall instabilities is suggested. One of important advantages of this approach is related to splitting of plasma and wall models, which allows using the results of different plasma and wall models without performing any additional simulations. For simplicity only synergistic effects caused by hydrogen fluxes are considered . However, this approach could be easily extended to account for impurity effects. Finally, we note that the response functions needed for dispersion equation could be deduced from experimental data in a specially design experimental set up. |
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