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This project involves the use of friction stir processing with a change in the number of passes on grain refinement of pure copper. FSP is a rising metal processing technique for manufacturing fine-grained structures. The project’s main goal was to determine the outcome of change in the number of passes and pass direction on the grain size and microstructure. The change in microstructure is accompanied by changes in properties like grain structure, tensile strength, and hardness. FSP has been done on 3-mm-thick pure copper plates, using a tungsten carbide tool of 18 mm diameter shoulder with a tapered pin of diameter 3 mm, a tool rotation speed of 1500 rpm, 2° tilt angle and tool traverse speed of 20 mm/min with the variable being the number of passes and pass direction. The optical microscopy, grain size measurements of the nugget region, and time–temperature plot is presented. The specimen with smaller grain sizes are further tested for tensile strength.

The COMPASS Upgrade tokamak (B = 5 T, IP = 2 MA, R = 0.894 m, a = 0.27 m) requires strengthening of the existing foundations due to larger electro-magnetic (EM) loads occurring during plasma disruption events. The focus is given to the VDE as they create the largest EM forces. Due to the fast rate of change of these forces, dynamic effects were considered. In order to find the time evolution of the electromagnetic forces during the VDE disruptions, a 2D electromagnetic model has been created. Current values in the central solenoid, poloidal field coils and the vacuum vessel were obtained from the FIESTA code and the plasma was modeled with finite number of filaments with prescribed currents to model the vertical motion. The time evolutions of the electromagnetic forces during 500 ms were imported to a 3D mechanical model built in Ansys Workbench. The transient analysis was solved with the mode superposition method based on 100, 400 and 1000 modes. The extremum reaction forces acting on the foundations were found to be -12.6 MN and +12.7 MN

The COMPASS-U is a new tokamak in the design phase at Institute of Plasma Physics, Prague, Czech Republic. It is a medium-sized tokamak with high magnetic field (Bt = 5 T) and high temperature vessel ( This paper outlines the latest status on the design activities of the vacuum vessel concept and its mechanical verification against the electromagnetic forces. The vessel design concept is developed considering the spatial constraints, fabrication simplicity, minimum cost and the design is verified for minimum deformation and stresses in the vessel. It also includes the design evolution of PSP, PSP support and VV support. The mechanical analyses are carried out for load combinations of plasma transient events.

Large electromagnetic forces acting on the vacuum vessel and passive stabilizing plates of the COMPASS-U tokamak (R = 0.894 m, a = 0.27 m, BT ≤ 5 T, IP ≤ 2 MA) due to disruption events are expected during its operation. An electromagnetic model based on the finite element method was developed using ANSYS software to determine the maximal possible forces on the vacuum vessel assembly that might occur. The effects of current quench, thermal quench, vertical displacement events and halo currents are all incorporated in the model. The toroidal eddy currents, the poloidal eddy currents (caused by thermal and current quench) and the poloidal halo current are taken into account. The model predicts that the induced toroidal current can reach the value up to 1.33 MA in the vacuum vessel shell and the value up to 0.52 MA in the passive stabilizing plates. Vertical force up to 4.1 MN acting on the entire vacuum vessel assembly can be expected and force up to 2 MN acting on the stabilizing plates, in particular. The results of the presented electromagnetic model are part of the load specification for the mechanical design of the vacuum vessel assembly of the COMPASS-U tokamak.

The COMPASS-U tokamak (Bt = 5 T, Ip = 2 MA, R = 0.894 m, a = 0.27 m) is currently under design at IPP CAS (Institute of Plasma Physics of the Czech Academy of Sciences). One of the design challenges is to design sufficiently strong foundations to support even the highest forces that are generated during plasma disruption events. The worst, found up to now, scenario has been considered in order to simulate the response of the main tokamak components and to calculate the force transmitted to the foundations. A 3D model has been built in the ANSYS Workbench software including the cryostat support, the vacuum vessel, the poloidal field coils and their supports. The forces have been calculated via global electromagnetic model. Integrated values of the forces and torques have been spread onto surfaces of the parts. The modal superposition method has been utilized for the integration, based on the first 100 vibration modes. Rayleigh damping model with the fraction of critical damping of 2% was used. The reaction forces transmitted to the foundations have been obtained providing input for the civil design work.

The results produced by experimental campaigns of RFX-mod shows the need to improve passive MHD control and minimise braking torque on plasma. These improvements require major mechanical changes on the present components of the machine including first wall, vacuum vessel, and toroidal support structure (TSS). The vacuum vessel will be removed and the first wall tiles will be directly supported by the passive stabilising shell, so increasing the poloidal cross section by 28 mm radially and bringing the passive stabilising shell as close as possible to the plasma boundary. This paper presents the mechanical design modifications of the torus assembly. Composite rings are designed to support the passive stabilising shell on the TSS. The vacuum boundary is shifted at the TSS by developing different joint solutions compatible with the stringent requirements of the present components: ceramic-metal brazed rings at the two poloidal joints, fully welded solution at the external equatorial joint, and resistive weld plate at the internal equatorial joint. The vacuum vessel portholes for pumping, fuelling, diagnostics, and sensor cable routing are redesigned to be integrated onto the TSS. The design is supported by thermo-mechanical analyses and verifications carried out applying load combinations corresponding to the expected load cases.

After 10 years of operation since its major modification, an upgrade of the RFX-mod experiment is presently under design. The scientific objective is the improvement of 3D physics studies through a more robust transition to higher confinement regimes in both Reversed Field Pinch (RFP) and Tokamak configuration obtained thanks to an advanced system for the active control of MHD instabilities. The main design driver requirements for this machine upgrade are the removal of the present resistive vacuum vessel and the enhancement of the ‘shell-plasma proximity’, to reduce the deformation of the last close magnetic surface and to improve the self-organized helical plasma regimes. The fulfillment of these requirements implies a major change of the internal components of the machine such as the replacement of the whole first wall, the change of the support system of the stabilizing shell and the modification of the present toroidal support structure to provide the function of vacuum barrier. In combination, other components of the machine will be upgraded, such as magnets and power supply, diagnostic systems and a NBI will be integrated. The paper presents an overview of the engineering design of the new components and highlights the critical aspects of the new torus assembly.

This paper reports the main recent results of the RFX-mod fusion science activity. The RFX-mod device is characterized by a unique flexibility in terms of accessible magnetic configurations. Axisymmetric and helically shaped reversed-field pinch equilibria have been studied, along with tokamak plasmas in a wide range of q(a) regimes (spanning from 4 down to 1.2 values). The full range of magnetic configurations in between the two, the so-called ultra-low q ones, has been explored, with the aim of studying specific physical issues common to all equilibria, such as, for example, the density limit phenomenon. The powerful RFX-mod feedback control system has been exploited for MHD control, which allowed us to extend the range of experimental parameters, as well as to induce specific magnetic perturbations for the study of 3D effects. In particular, transport, edge and isotope effects in 3D equilibria have been investigated, along with runaway mitigations through induced magnetic perturbations. The first transitions to an improved confinement scenario in circular and D-shaped tokamak plasmas have been obtained thanks to an active modification of the edge electric field through a polarized electrode. The experiments are supported by intense modeling with 3D MHD, gyrokinetic, guiding center and transport codes. Proposed modifications to the RFX-mod device, which will enable further contributions to the solution of key issues in the roadmap to ITER and DEMO, are also briefly presented. © 2017 IAEA, Vienna.

25th Fusion Energy Conference (FEC), Govt Russian Federat, St ă Petersburg, RUSSIA, OCT 13-18, 2014; International audience; The RFX-mod device is operated both as a reversed field pinch (RFP), ă where advanced regimes featuring helical shape develop, and as a ă tokamak. Due to its flexibility, RFX-mod is contributing to the solution ă of key issues in the roadmap to ITER and DEMO, including MHD instability ă control, internal transport barriers, edge transport and turbulence, ă isotopic effect, high density limit and three-dimensional (3D) ă non-linear MHD modelling. This paper reports recent advancements in the ă understanding of the self-organized helical states, featuring a strong ă electron transport barrier, in the RFP configuration; the physical ă mechanism driving the residual transport at the barrier has been ă investigated. Following the first experiments with deuterium as the ă filling gas, new results concerning the isotope effect in the RFP are ă discussed. Studies on the high density limit show that in the RFP it is ă related to a toroidal particle accumulation due to the onset of a ă convective cell. In the tokamak configuration, q(a) regimes down to q(a) ă = 1.2 have been pioneered, with (2,1) tearing mode (TM) mitigated and ă (2,1) resistive wall mode (RWM) stabilized: the control of such modes ă can be obtained both by poloidal and radial sensors. Progress has been ă made in the avoidance of disruptions due to the (2,1) TM by applying ă q(a) control, and on the general issue of error field control. The ă effect of externally applied 3D fields on plasma flow and edge ă turbulence, sawtooth control and runaway electron decorrelation has been ă analysed. The experimental program is supported by substantial ă theoretical activity: 3D non-linear visco-resistive MHD and non-local ă transport modelling have been advanced; RWMs have been studied by a ă toroidal MHD kinetic hybrid stability code.