Engineering topics

Wischmeier, M.

Publications and source records attributed to Wischmeier, M..

7 records

DOE OSTI2024

Overview of the EUROfusion Tokamak Exploitation programme in support of ITER and DEMO

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

ASDEX upgrade↗

DOE OSTI2024

Helium plasma operations on ASDEX Upgrade and JET in support of the non-nuclear phases of ITER

For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of the EUROfusion Consortium carried out dedicated experimental campaigns in He on the ASDEX Upgrade (AUG) and JET tokamaks in 2022, with particular emphasis put on the ELMy H-mode operation and plasma-wall interaction processes as well as comparison to H or deuterium (D) plasmas. Both in pure He and mixed He + H plasmas, H-mode operation could be reached but more effort was needed to obtain a stable plasma scenario than in H or D. Even if the power threshold for the LH transition was lower in He, entering the type-I ELMy regime appeared to require equally much or even more heating power than in H. Suppression of ELMs by resonant magnetic perturbations was studied on AUG but was only possible in plasmas with a He content below 19%; the reason for this unexpected behaviour remains still unclear and various theoretical approaches are being pursued to properly understand the physics behind ELM suppression. The erosion rates of tungsten (W) plasma-facing components were an order of magnitude larger than what has been reported in hydrogenic plasmas, which can be attributed to the prominent role of He 2+ ions in the plasma. For the first time, the formation of nanoscale structures (W fuzz) was unambiguously demonstrated in H-mode He plasmas on AUG. However, no direct evidence of fuzz creation on JET was obtained despite the main conditions for its occurrence being met. The reason could be a delicate balance between W erosion by ELMs, competition between the growth and annealing of the fuzz, and coverage of the surface with co-deposits.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

The role of isotope mass and transport for H-mode access in tritium containing plasmas at JET with ITER-like wall

The required heating power, $P_{\mathrm{LH}}$, to access the high confinement regime (H-mode) in tritium containing plasmas is investigated in JET with ITER-like wall at a toroidal magnetic field of $B_t = 1.8$ T and a plasma current of $I_p = 1.7$ MA. $P_{\mathrm{LH}}$, also referred to as the L-H power threshold, is determined in plasmas of pure tritium as well as mixtures of hydrogen with tritium (H-T) and mixtures of deuterium with tritium (D-T), and is compared to the L-H power threshold in plasmas of pure hydrogen and pure deuterium. It is found that, for otherwise constant parameters, $P_{\mathrm{LH}}$ is not the same in plasmas with the same effective isotope mass, $A_{\mathrm{eff}}$, when they differ in their isotope composition. Thus, $A_{\mathrm{eff}}$ is not sufficient to describe the isotope effect of $P_{\mathrm{LH}}$ in a consistent manner for all considered isotopes and isotope mixtures. The electron temperature profiles measured at the L-H transition in the outer half of the radius are very similar for all isotopes and isotope mixtures, despite the fact that the L-H power threshold varies by a factor of about six. This finding, together with the observation of an offset linear relation between the L-H power threshold, $P_{\mathrm{LH}}$, and an effective heat diffusivity, $\chi_{\mathrm{eff}}$, indicates that the composition-dependent heat transport in the low confinement mode (L-mode) determines, how much power is needed to reach the necessary electron temperatures at the edge, and hence $P_{\mathrm{LH}}$.

tritium plasmas↗

DOE OSTI2022

Characterisation of divertor detachment onset in JET-ILW hydrogen, deuterium, tritium and deuterium–tritium low-confinement mode plasmas

Measurements of the ion currents to and plasma conditions at the low-field side (LFS) divertor target plate in low-confinement mode plasmas in the JET ITER-like wall materials configuration show that the core plasma density required to detach the LFS divertor plasma is independent of the hydrogenic species protium, deuterium and tritium, and a 40 %/60 % deuterium–tritium mixture. This observation applies to a divertor plasma configuration with the LFS strike line connected to the horizontal part of the LFS divertor chosen because of its superior diagnostic coverage. The finding is independent of the operational status of the JET cryogenic pump. The electron temperature (T e ) at the LFS strike line was markedly reduced from 25 eV to 5 eV over a narrow range of increasing core plasma density, and observed to be between 2 eV and 3 eV at the onset of detachment. The electron density (n e ) peaks across the LFS plasma when T e at the target plate is 1 eV, and spatially moves to the X-point for higher core densities. The density limit was found approximately 20 % higher in protium than in tritium and deuterium–tritium plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

The radiated power limit in impurity seeded JET-ILW plasmas

The total radiated fraction is examined in high density H-mode plasmas (Greenwald fraction of about 85 %) in JET by the variation of the auxiliary heating power of Pheat = 14 MW-29 MW. An achieved radiation fraction of about 75 % at most has been observed in JET-ILW, which is less than the highest achievable (≈90 %) fraction in JET-C during the high radiative power scenarios with N2 seeding. It is shown that the maximal achievable total radiation fraction averaged over ELM cycles has a strong dependence on the radiation efficiency of the ELM energy, $θ_{rad}$: $\gamma^{max}_{rad,total}$ = 1 -$\frac {f_{ELM} x ΔW_{ELM}} {P_heat}$ (1- $θ_{rad}$). About 50 % and 16 % of the ELM induced diamagnetic energy drop (ΔW ELM ) radiates during the ELM in JET-C and JET-ILW, respectively, which corresponds to the maximum total radiated powers of $\gamma^{max}_{rad,JET-C}$ = 0.87 and $\gamma^{max}_{rad,JET-ILW}$ = 0.77. These values of the maximum of the radiative power fractions are in good agreement with $\gamma^{max}_{rad}$ experimentally observed in JET-C (90 %) and JET-ILW (75 %).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

SOLPS-ITER validation with TCV L-mode discharges

This work presents a quantitative test of SOLPS-ITER simulations against tokamak a configuration variable (TCV) L-mode experiments. These simulations account for drifts, currents, kinetic neutrals, and carbon impurities providing the most complete edge transport simulations for TCV to date. The comparison is performed on nominally identical discharges carried out to assess the effectiveness of TCV’s divertor baffles in the framework of the European Plasma Exhaust program and employs numerous edge diagnostics providing a detailed code-experiment benchmark for TCV. The simulations show a qualitative consistency, but the quantitative differences remain, which are assessed herein. It is found that, for a given separatrix density, the simulations most notably yield a colder, and denser, divertor state with a higher divertor neutral pressure than measured.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Parameter dependencies of the experimental nitrogen concentration required for detachment on ASDEX Upgrade and JET

While current tokamak experiments are beginning to use real-time feedback control systems to manage the plasma exhaust, future tokamaks still require validation of theoretical models used to predict the threshold impurity concentration required to sufficiently reduce the power and particle fluxes to the divertor. This work exploits new spectroscopic measurements of the divertor nitrogen concentration, c n , in partially detached N 2 -seeded H-mode plasmas on ASDEX Upgrade (AUG) and JET with the ITER-Like Wall (JET-ILW) to test the parameter dependencies of the power flowing to the outer divertor, P div,outer , and the separatrix electron density, n e.sep . A least-squares regression of the AUG measurements demonstrates that the threshold c n required for detachment scales as $c_n∝P^{1.19±0.32}_{div,outer}n^{-2.77±0.36}_{e,sep}$. This scaling of n e,sep is also consistent with the measurements from JET which, at constant P div,outer , show $c_n∝n^{-2.43±0.27}_{e,sep}$. The dependencies of P div,outer and n e.sep is demonstrated over at least a factor of two change in both parameters and indicates a stronger dependence on n e.sep in comparison to the Lengyel model, which could be due to the assumption in this model that the heat flux channel width is independent of density. This first assessment of detachment scaling with impurity seeding highlights the need for further analysis of the systematic uncertainties of the measurement and more consistent scenarios from more tokamaks to investigate the machine size scaling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗