The as-built stiffness in the ITER central solenoid (CS) modules (CSM1 thorough to CSM4 are currently manufactured) determines the range of vertical compression forces that can be tolerated by the CS modules during ITER operation. Since the as-built stiffness of the CS modules manufactured (∼32 GPa and ∼34 GPa for CSM1 and CSM2, respectively and similar for the other modules) has been reduced from the design value (53 GPa), the CS axial (vertical) force criteria have been updated assuming a conservative stiffness (25 GPa) with margins for all six CS modules. Initial analysis using the updated CS force criteria has revealed that this reduction affects only the plasma initiation with fully charged CS in the ITER 15 MA Baseline DT scenario, resulting in a slight reduction of poloidal magnetic flux, from 117.5 Wb to 116.2 Wb at initial CS magnetization. Therefore, the 15 MA Baseline scenario has been re-developed with an updated plasma start-up, and then the entire evolution of the CS and poloidal field coil parameters has been validated against all the coil currents, fields and forces criteria. To explore potential risks and opportunities for further optimization of scenarios, the equilibrium operational space (the plasma internal inductance versus the poloidal magnetic flux produced by the coils) at flat-top burn has been analyzed using the CORSICA and DINA codes. The three major ITER reference DT operation scenarios, 15 MA Q = 10 Baseline, 12.5 MA Q > 5 Hybrid and 10 MA Q ∼ 5 Steady-State, satisfy all the coil criteria including the CS force updated reflecting the as-built stiffness. The evolution of the plasma discharge parameters within the equilibrium operational spaces provided a guidance for potential optimization with margins.

The article is devoted to the calculation of equilibrium configurations of the MEPHIST-0 tokamak plasma with a poloidal magnetic system taking into account the part responsible for compensating the magnetic field of the central solenoid using the DINA plasma physics code. The parameters of the poloidal magnetic system necessary for the formation of initial equilibrium after the discharge breakdown are obtained. The results of calibration experiments without taking into account the vacuum chamber using a pulsed current source are presented, on the basis of which the dynamic problem of plasma equilibrium calculation is formulated. The simulation results are compared with the experimental data, and the influence of the chamber on the results is considered.

In ITER, tritium retention primarily occurs through co-deposition with beryllium. To avoid exceeding the strict tritium inventory limit, efficient tritium recovery techniques are essential. Baking is the ITER baseline for tritium recovery, but its effectiveness in removing tritium from thick beryllium layers is limited. A raised strike point scenario is considered an alternative method for removing tritium from the ITER inner vertical divertor target by heating components via plasma flux. This paper presents SOLPS-ITER code simulations conducted under various conditions, assessing the divertor performance and tritium outgassing of the raised strike point scenario. As the strike point is raised, recycled neutrals are not efficiently baffled by the dome and scrape-off layer, significantly changing the neutral trajectory and ionization source distribution. This improves detachment accessibility but worsens core-edge compatibility compared to the baseline scenario. However, in the partially detached condition, the impact of raising the strike point, perpendicular transport, and q (95) on target heat flux is not significant, as it primarily scales with the input power. Target heat flux is translated to target surface temperature using a simplified heat transfer model that considers the 3D target monoblock geometry and active cooling condition, excluding Be layer thermal properties. For partially detached divertor conditions, the bulk tungsten monoblock surface temperature remains below the baking temperature, which is insufficient for efficient tritium outgassing under the actively cooled ITER divertor condition. However, considering the potential thermal contact resistance between the beryllium and tungsten layers, which may significantly impact temperature distribution, the temperature of the beryllium layer can be raised to a level sufficient for efficient tritium outgassing. Therefore, the raised strike point scenario can be considered as an alternative in-vessel tritium removal technique.

The main results of numerical simulation of transient electromagnetic processes in the modules of the first and second rows of TRT first wall panels are presented. The study was carried out using the TYPHOON software designed for numerical simulation of quasi-stationary eddy currents in conducting shells of complex shape randomly located in space, taking into account their multicoupling and branching. The basic design of the first wall panel and its five modifications are considered. It is determined that the electromagnetic forces and moments for the basic design reach extreme values, and the edge elements of the panel are the most loaded. Options with inserting electrical insulation at the places of contact of the plasma facing elements and the base of the panel, as well as incorporating a shunt between adjacent panels, make it possible to achieve the strongest reduction in loads. These options are accepted as basic for further modification of the panel design and conduction of additional EM analysis.

Long Pulse Scenarios (LPS) in ITER foreseen during the Pre-Fusion Power Operation (PFPO) phase of the ITER Research Plan (IRP) are assessed using 1.5D transport simulations within the ASTRA framework. Such assessment is required to predict the operational space for LPS operation in PFPO, as well as to evaluate which physics processes for LPS operation during Fusion Power Operation (FPO) could be studied during PFPO. An important aspect in the development of LPSs in PFPO is to minimize lifetime consumption of the Central Solenoid (CS) for these scenarios. The maximum pulse length achievable for LPSs in PFPO with no consumption of CS lifetime (currents in CS coils ⩽30 kA per turn) has been assessed for a range of heating schemes and heating mixes, confinement regimes (L-mode and H-mode) and for helium and hydrogen plasmas. The operational space of LPS and pulse length has been explored through density scans with the Heating and Current Drive mix required for the FPO Q ⩾ 5 steady-state plasma scenario (namely Neutral Beam Injection and Electron Cyclotron Heating) including acceptable shine through losses on the first wall for both helium and hydrogen plasmas. Fast particle physics aspects that are common between FPO plasmas and LPS PFPO H-mode plasmas at low densities are studied including MHD stability analysis with the KINX code and non-perturbative critical gradient model based on high-n Toroidal Alfven Eigenmodes (TAE) stability kinetic ballooning code HINST calculations.

An overview of the MEPhIST-0 educational and research small-scale spherical tokamak project is presented including the vacuum vessel, magnetic field systems, and diagnostics. In contrast to other small machines, it is an advanced tokamak with D-shaped plasma equipped with an electron-cyclotron resonance pre-ionization system for plasma startup and an ion-cyclotron resonance system for plasma heating and wall conditioning. The design choices taken are discussed from the perspective of a primarily educational installation. The machine design is simplified while remaining relevant to larger devices. First plasma results obtained in 2021 are presented.

This article is devoted to the development of a software and computing system for calculating the evolution of nominal currents and voltages in the coils of the poloidal magnetic system in order to ensure the equilibrium of the plasma of the T-15 tokamak in a given discharge scenario. The obtained evolutions of currents and voltages are necessary for the subsequent construction of the system of magnetic control of the current, shape, and position of the plasma during the discharge scenario. The physical parameters of the discharge and the required evolution of the plasma boundary during the discharge process are set using the graphical interface. The software package is developed on the basis of the DINA plasma physics code within the Matlab-Simulink system. Examples of using the graphical interface as well as the simulation results of the ohmic scenario with a plasma current of 2 MA are presented.

Systematic test particle tracing simulations for runaway electrons (REs) are performed for six post-thermal quench equilibria from DIII-D and ITER, where large scale, kink-like n = 1 (n is the toroidal mode number) magnetohydrodynamic (MHD) instabilities are found. The modeled particle guiding center orbits allow extraction of the effective diffusion–convection coefficients of REs in the presence of large three-dimensional (3D) perturbations up to 10% of the equilibrium toroidal field. With a fixed spatial distribution of the field perturbation, the RE transport coefficients along the plasma radial coordinate track reasonably well with the surface-averaged perturbation level. A substantial variation in the value of the transport coefficients—by three orders of magnitude in most cases, however, occurs with varying launching location of REs along the plasma radius. Large 3D perturbations almost always lead to comparable diffusion and convection processes, meaning that diffusion alone is insufficient to describe the particle motion. At lower (but still high) level of perturbation, the RE convection is found to be dominant over diffusion. A similar observation is made when the perturbation is too strong. In the presence of large perturbation, the dependence of the RE transport on the particle energy is sensitive to the spatial distribution of the perturbation. Based on numerically obtained RE transport coefficients, an analytic fitting model is proposed to quantify the particle diffusion and convection processes due to large MHD events in post-thermal quench plasmas. The model is shown to reasonably well reproduce the direct test particle tracing results for the RE loss fraction and can, thus, be useful for incorporating into other kinetic RE codes in order to simulate the RE beam evolution in the presence of large 3D perturbations.

On ITER, plasma start-up will be performed in limiter configuration on the inboard equatorial beryllium first wall panels (FWP). In contrast to most present tokamaks, however, this ramp-up phase will be comparatively long (∼10 s) and the use of actively cooled components means that power flux management is key if FWP lifetime is not to be compromised. Shaping of the FWPs is mandatory to ensure that leading edges do not appear between neighbouring units. For the ITER inboard panels, this has been optimized to account for the discovery in recent years on current devices of narrow scrape-off layer power flux channels for inner wall limited plasmas. However, the shaping results in power densities which are particularly sensitive to the overall ‘longwave’ (LW) alignment of the central column FWP ring with the structure of the toroidal magnetic field (TF), placing tight constraints on the target alignment. This target is currently based on a pure n = 1 LW alignment, but simulations of TF coil (TFC) locking upon energization show that, depending on the initial configuration of the gaps between the TFC inner legs, the field structure can be more complex. Although the TFC manufacture and machine assembly strategy is to make every effort possible to approach the ideal TF structure, an NMR sensor-based TF mapping diagnostic will be implemented to measure the field structure during the first plasma and engineering operation phase. An analytic framework has been developed and verified against numerical simulations to assess the capability for measurements from a set of discrete sensors located on the vacuum vessel inner column to be used to reconstruct the field structure at the FWP locations, a further ∼60 cm radially inward. In parallel with the alignment optimization and TF mapping strategies, modified ramp-up scenarios are also being designed which may be used to reduce inner wall limiter power fluxes if this proves to be necessary during operation.

Mitigation of runaway electrons (REs) by three-dimensional (3D) magnetic field perturbations is numerically investigated for the ITER 15 MA baseline D–T scenario, utilizing the MARS-F code (Liu et al Phys. Plasmas 7 3681) with a drift orbit test particle tracing module. Considered are two types of 3D fields: the n = 3 ( n is the toroidal mode number) resonant magnetic perturbation (RMP) utilized for the purpose of controlling the edge localized modes in ITER, and perturbations generated by the n = 1 magneto-hydrodynamic (MHD) instabilities in a post-disruption plasma. The RMP field, applied to a pre-disruption plasma, is found to be moderately effective in mitigating the RE seeds in ITER when vacuum field model is assumed. Up to ∼40% loss fraction is possible at 90 kA-turn coil current. The mitigation efficiency is however substantially reduced, down to less than 5%, when the plasma response is taken into account. This is due to strong screening of the resonant magnetic field components by the plasma response resulting in much less field line stochasticity. On the other hand, the MARS-F modeling, based on the DINA-simulated post-disruption equilibria, shows that the n = 1 resistive kink instabilities develop in these plasmas, as the edge safety factor q a evolves and drops below integer numbers. RE mitigation by these MHD instabilities is sensitive to the eigenmode structure. The best mitigation is achieved as q a drops below 3, when a global kink instability occurs that encompasses both internal and external components. This global instability is found to be capable of mitigating over 80% MeV-level passing RE orbits at a field perturbation | δB |/ B 0 that is comparable to that observed in DIII-D experiments, and full mitigation if the perturbation amplitude is doubled. The ‘wetted’ area on the ITER limiting surface, due to MHD instability induced RE loss, generally increases with the perturbation amplitude (together with increasing loss fraction). At the highest perturbation level assumed in this study, the wetted area reaches ∼60% of the total limiting surface area. The lost RE orbits mainly strike the outer divertor region of the limiting surface, with some fraction also hitting a wide area along the inboard side of the surface.

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.

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 λq=75-198 mm. A maximum halo width wh of 0.52 m at the outboard midplane is observed. SMITER field line tracing and energy deposition simulations calculate a q⊥,max 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 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 wh 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.

The work is devoted to the analysis of the influence of poloidal current induced in a tokamak vacuum chamber due to change of paramagnetic plasma properties as a result of thermal quench on the magnitude and distribution of electromagnetic forces in a vacuum chamber. The work was performed numerically using the plasma-physics code DINA for conditions of major disruption in the plasma of ITER and T-15MD. A comparison is made of the electromagnetic effect obtained from the toroidal and poloidal currents induced in the vacuum chamber as a result of thermal quench with estimates of the electromagnetic forces due to only toroidal currents. It is shown that the integral radial component of the electromagnetic force in the elements of the vacuum chamber when the poloidal current is taken into account can be much lower than in the case of only the toroidal current induced in the chamber. The paper presents a model for calculating the poloidal current induced in a vacuum chamber first used in the DINA code.

Tokamak plasmas with vertically elongated cross-sections are unstable due to vertical displacements. Feedback stabilization of the ITER plasma current centre will be performed using the speed of its vertical displacements, dZ/dt, as input. Inevitable noise in the diagnostic signal for dZ/dt leads to "noisy" components in the feedback voltage and current in the stabilizing coils. This leads to noisy components in the vertical position of the plasma current centre and the divertor strike points. For burning plasma conditions on ITER, these strike point displacements may be a concern for thermal fatigue at the water cooling interface of the tungsten monoblocks constituting the divertor targets. A study is presented here in which this concern is examined for the first time. The baseline 15 MA scenario (fusion power of 500MW with a 500 s flattop and fusion power gain, Q(DT)= 10) is simulated with the DINA code, assuming low frequency noise in the dZ/dt diagnostic signal. The noise has uniform spectrum with a given root mean square (RMS) value (< dZ/dt>=0.6 ms(-1) or 0.2 ms(-1)) in the frequency band (0, 1 kHz). The results of these DINA simulations are combined with dissipative divertor plasma solutions obtained with the SOLPS-ITER plasma boundary code to provide a time dependent divertor target heat flux density profile. The latter is then imposed on a finite element model of the target monoblocks to assess the temporal evolution of the 3D temperature field in the block, including the Cu-W and Cu-CuCrZr joints at the cooling interface. These joints represent the points of the Cu and CuCrZr materials that see the largest temperature changes and are thus at greatest risk of failure. To evaluate an acceptance criterion on the non-cyclic and non-uniform thermal loads at the joints, an approach is developed which combines a Rainflow counting technique with Palmgren-Miner's rule for fatigue accumulation. Analysis of the temperature evolution at the CuW joint shows that the RMS value of noise < dZ/dt>similar to 0.6 ms(-1) is unacceptable from the point of view of thermal fatigue for the expected total exposure time under burning plasma conditions that the first ITER divertor must survive. The lower value (< dZ/dt>similar to 0.2 ms(-1)) is found to be acceptable. A subsequent parametric study concludes that < dZ/dt> less than or similar to 0.44 ms(-1) is just consistent with fatigue lifetime for the prescribed divertor power flux density profile, but should be kept lower than this to allow for some margin. Regarding the monoblock surface temperature, the natural power spreading caused by the separatrix movements is found to be beneficial from the point of view of recrystallization. For the level of noise in the dZ/dt diagnostic required to stay below joint fatigue limits, the average surface temperature on the most loaded monoblocks is reduced by similar to 100 degrees C compared to the case with stationary strike points and the amount of time spent at this temperature at any one block by over 80 %.