An analysis workflow has been developed to assess energy deposition and material damage for ITER vertical displacement events (VDEs) and major disruptions (MD). This paper describes the use of this workflow to assess the melt damage to be expected during unmitigated current quench (CQ) phases of VDEs and MDs at different points in the ITER research plan. The plasma scenarios are modeled using the DINA code with variations in plasma current I p, disruption direction (upwards or downwards), Be impurity density n Be, and diffusion coefficient χ. Magnetic field line tracing using SMITER calculates time-dependent, 3D maps of surface power density q ⊥ on the Be-armored first wall panels (FWPs) throughout the CQ. MEMOS-U determines the temperature response, macroscopic melt motion, and final surface topology of each FWP. Effects of Be vapor shielding are included. Scenarios at the baseline combination of I p and toroidal field (15 MA/5.3 T) show the most extreme melt damage, with the assumed n Be having a strong impact on the disruption duration, peak q ⊥ and total energy deposition to the first wall. The worst-cases are upward 15 MA VDEs and MDs at lower values of n Be, with q ⊥,max = 307 MW m−2 and maximum erosion losses of ∼2 mm after timespans of ∼400–500 ms. All scenarios at 5 MA avoided melt damage, and only one 7.5 MA scenario yields a notable erosion depth of 0.25 mm. These results imply that disruptions during 5 MA, and some 7.5 MA, operating scenarios will be acceptable during the pre-fusion power operation phases of ITER. Preliminary analysis shows that localized melt damage for the worst-case disruption should have a limited impact on subsequent stationary power handling capability.

Background Sustainable energy transition of a country is complex and long-term process, which requires decision-making in all stages and at all levels, including a large number of different factors, with different causality. The main objective of this paper is the development of a probabilistic model for decision-making in sustainable energy transition in developing countries of SE Europe. The model will be developed according to the specificities of the countries for which it is intended—SE Europe. These are countries where energy transition is slower and more difficult due to many factors: high degree of uncertainty, low transparency, corruption, investment problems, insufficiently reliable data, lower level of economic development, high level of corruption and untrained human resources. All these factors are making decision-making more challenging and demanding. Methods Research was done by using content analysis, artificial intelligence methods, software development method and testing. The model was developed by using MSBNx— Microsoft Research’s Bayesian Network Authoring and Evaluation Tool . Results Due to the large number of insufficiently clear, but interdependent factors, the model is developed on the principle of probabilistic (Bayesian) networks of factors of interest. The paper presents the first model for supporting decision-making in the field of energy sustainability for the region of Southeastern Europe, which is based on the application of Bayesian Networks. Conclusion Testing of the developed model showed certain characteristics, discussed in paper. The application of developed model will make it possible to predict the short-term and long-term consequences that may occur during energy transition by varying these factors. Recommendations are given for further development of the model, based on Bayesian networks.

Abstract Ion cyclotron resonance heating (ICRH) is one of the three additional heating schemes to be deployed on ITER. Its two antenna arrays, installed on the outboard midplane, will deliver 20 MW of RF power in the 40–55 MHz frequency range. The plasma-facing component of the antenna assembly is the Faraday screen, comprising beryllium (Be) tile armoured, actively cooled bars located only ~1 cm radially behind the innermost point of the shaped Be first wall panels (FWPs). As such they are in close proximity to the scrape-off layer (SOL) plasma and it is important to assess the maximum heat loads that the screen bars may experience during high power ITER operation. This paper provides a detailed assessment of these loads using the new 3D field line tracing and power deposition framework SMITER (Kos et al., 2019). The focus is on the H-mode, burning plasma scenario, taking into account both plasma heat loading (including average loading due to mitigated Type I ELMs) and the loads due to photonic impact (assessed with the optical ray-tracing package Raysect (Meakins and Carr, 2017)) from power radiated in the core obtained from integrated JINTRAC simulations. Calculations are also performed to assess the minimum allowed antenna to magnetic separatrix distances, for cases in which closer approach may be required to improve RF coupling.

Abstract The beryllium (Be) main chamber wall interaction during a 5 MA / 1.8 T upward, unmitigated VDE scenario, first analysed in [J. Coburn et al., Phys. Scr. T171 (2020) 014076] for ITER, has been re-evaluated using the latest energy deposition analysis software. Updates to the DINA disruption model are summarized, including an improved numerical convergence for the 0D power balance, limitations on the safety factor within the plasma core, and the choice to maintain a constant plasma + halo poloidal cross-section. Such updates result in a broad halo region and higher radiated power fractions compared to previous models. The new scenario lasts for ∼75 ms and deposits ∼29 MJ of energy, with the radial distribution of parallel heat flux q ‖ r resembling an exponential falloff with an effective λ q = 75 - 198 mm. A maximum halo width w h of 0.52 m at the outboard midplane is observed. SMITER field line tracing and energy deposition simulations calculate a q ⊥ , m a x of ∼83 MW/m2 on the upper first wall panels (FWP). Heat transfer calculations with the MEMOS-U code show that the FWP surface temperature reaches ∼1000 K, well below the Be melt threshold. Variations of this 5 MA scenario with Be impurity densities from 0 to 3∙1019 m-3 also remain below the melt threshold despite differences in energy deposition and time duration. These results are in contrast to the early study which predicted melt damage to the first wall [J. Coburn et al., Phys. Scr. T171 (2020) 014076], and emphasize the importance of accurate models for the halo width w h and the heat flux distribution q ‖ r within that halo width. The 2020 halo model in DINA has been compared with halo current experiments on COMPASS, JET, and Alcator C-Mod, and the preliminary results build confidence in the broad halo width predictions. Results for the 5 MA VDE are compared with those for a 15 MA equivalent, generated using the new DINA model. At the higher current, significant melting of the upper FWP is to be expected.

Abstract Built around the SMARDDA modules for magnetic field-line tracing [IEEE Tr. Plasma Sc. 42 (2014) 1932], the SMITER code package (SMARDDA for ITER) is a new graphical user interface (GUI) framework for power deposition mapping on tokamak plasma-facing components (PFC) in the full 3-D CAD geometry of the machine, taking as input a user-defined specification for parallel heat flux in the scrape-off layer (SOL) and a description of the equilibrium magnetic flux. The software package provides CAD model import and integration with the ITER Integrated Modelling and Analysis Suite (IMAS), parametric CAD components catalogue and modelling, CAD de-featuring for PFC surface extraction, meshing, visualization (using an integrated ParaView module), Python scripting and batch processing, storage in hierarchical data files, with several simulation cases in one study running in parallel and using message passing interface (MPI) for code speed-up. An integrated ParaView module can combine CAD geometry, magnetic field equilibrium, meshes and results for detailed setup analysis and a module is under development for full finite element computation of surface temperatures resulting from the power deposition patterns on 3-D PFCs. The code package has been developed for ITER, but can be deployed for similar modelling of any tokamak. This paper presents and discusses key features of this field-line tracing environment, demonstrates benchmarking against existing field-line tracing code and provides specific examples of power deposition mapping in ITER for different plasma configurations.