EM Loads Research Articles

Specification of asymmetric VDE loads of the ITER tokamak

During asymmetric vertical displacement events (AVDEs) associated with the kink mode of the plasma two asymmetry phenomena were observed in existing tokamaks, in particular in JET [1]. The related halo currents flowing in the passive structure were identified as the cause of asymmetric EM loads on tokamak components. The first phenomenon is a toroidal peak of the poloidal halo current that flows in the passive structure. The second phenomenon is that the toroidal plasma current is not uniform toroidally, so a toroidally non-uniform current flows in the vessel [2]. The specification of the expected characteristics of both phenomena as well as of the consequent asymmetric loads in ITER are summarized here. The related loads are specified for likely, unlikely and extremely unlikely AVDEs.

Electromagnetic studies of the ITER generic upper port plug

Electromagnetic analysis of the ITER generic upper port plug was carried out to estimate the most severe load to which the in-vessel components would be subjected. Static and dynamic analyses based on expected EM loads were also performed to evaluate the structural behavior of the generic upper port plug. These analyses reveal that stress and displacement of the upper port plug structure exceed design guidelines when the structure is subjected to an upward fast VDE. Stiffening the upper port plug will decrease stress and the severity of potential displacement. The design of the middle part of the generic upper port plug was considered. Static and dynamic analyses for stress and displacement of the port plug indicate that the stiffened model satisfies structural design guidelines. Further improvements have been considered to achieve sufficient design margins for the generic upper port plug for ITER. A deep slit in the diagnostic first wall and the diagnostic shield module was designed to mitigate the EM load. Additional EM analysis confirmed the applicability of these proposals. Results confirm that the generic upper port plug structure is able to withstand plasma disruptions.

Nuclear technology aspects of ITER vessel-mounted diagnostics

ITER has diagnostics with machine protection, basic and advanced control, and physics roles. Several are distributed on the inner and outer periphery of the vacuum vessel. They have reduced maintainability compared to diagnostics in ports. They also endure some of the highest nuclear and EM loads of any diagnostic for the longest time. They include:Inductive sensors for time-integrated and raw inductive measurements;Steady-state magnetic sensors to correct drifts of the inductive sensors;Bolometer cameras to provide electromagnetic radiation tomography;Microfission chambers and neutron activation stations to provide fusion power and fluence;MM-wave reflectometry to measure the plasma density profile and the plasma-wall distance and;Wiring to service magnetics, bolometry, and in-vessel instrumentation.This paper summarises the key technological issues these diagnostics arising from the nuclear environment, recent progress and outstanding R&D for each system.

ITER vacuum vessel: Design review and start of procurement process

The ITER vacuum vessel (VV) is one of the most critical components in the ITER project. It is on the critical path in the construction schedule and it is also a safety important class component (SIC), providing the first confinement barrier. As a result of reviews and the latest physics analyses, design requirements have been updated (e.g. ELM/VS coils) and a few design changes have to be implemented. This paper covers the updates of the VV vertical and horizontal EM load conditions during asymmetric VDEs, the design analysis of the ELM/VS coils and their interfaces to the VV, the blanket manifold design and the preparation of the technical specification in preparation for the procurement arrangement to be signed.

Assessment of EM loads on the EU HCPB TBM during plasma disruption and normal operating scenario including the ferromagnetic effect

Static and transient analyses have been carried out to evaluate Electro Magnetic loads on the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) during a type II linear disruption (Lorentz forces) and (because of the presence of ferromagnetic structural material) during normal ITER operating scenario (Maxwell forces (MFs)). At last the influence of the toroidal field ripple on the MFs has been investigated.

Detailed electromagnetic numerical evaluation of eddy currents induced by toroidal and poloidal magnetic field variation and halo currents

A detailed evaluation of the EM loads in the ITER divertor during plasma disruptions is mandatory for the correct dimensioning of the divertor component. The EM loads during plasma disruptions are mainly produced by: (1) toroidal flux variation (TFV) during the thermal quench (TQ) and current quench (CQ); (2) halo currents (HC); and (3) poloidal flux variation (PFV) during TQ and CQ phase. The new ITER reference disruption and the last changes in the divertor design have been considered in the EM models created to calculate all the EM loads due to TFV, HC and PFV. All the analyses have been performed for the three different main design options of the divertor plasma facing units (PFU). The effects of PFV have been analyzed using an EM-zooming procedure that has allowed a good detail of the component model, while new numerical approaches have been developed for the evaluation of the effects due to TFV and HC maintaining the same detail for the divertor model. Separate models have been developed to evaluate the equivalent electrical resistivities of the various PFU options; this allows in the full 3D model a strong simplification of a geometry which would otherwise be very complex. The effect of an electrical surface bridging of the PFU castellation has also been taken into account.

Design, Analysis and R&D of the EAST In-Vessel Components

In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R&D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m3/s pumping rate at a pressure of 10−1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event (VDE) will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 °C. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.

Optimization of the IGNITOR operating scenario at 11MA

Several endeavours have been made, in the frame of the IGNITOR design, in order to minimize technological problems, reducing electromagnetic loads, power supply requirements while avoiding the use of not completely proven materials. The present design modifications comprise both the operating scenario for the poloidal field (PF) currents and the PF coils geometry. The new scenario and a relevant simulation of a related typical fast vertical disruption, in the most dangerous plasma conditions in IGNITOR, have been simulated with the MAXFEA code. In order to reduce the EM loads on plasma chamber and the related stresses due to plasma disruptions, a new approach has been investigated based on the possibility of mitigating these loads by using copper toroidal layers added to the plasma chamber in proper positions. We found that this expedient could be quite effective not only in increasing the time constant of the plasma displacement but even in reducing the vertical force and its combined effect with hoop force on the vessel during a vertical displacement event (VDE).

Selection of design solutions and fabrication methods and supporting R&D for procurement of ITER vessel and FW/blanket

The vacuum vessel (VV) design has progressed by selecting design options, such as an independent cooling configuration in the VV field joint regions, 9 lower ports instead of 18, a single-wall structure for the upper and equatorial ports, and the vacuum vessel gravity support located below the lower ports. Double curvature pressing is now selected instead of facet shape welding for inner and outer shells in the upper and lower inboard regions. Regarding the first wall (FW)/blanket system, the plasma-facing surface of the FW has been redefined to avoid protruding leading edges in the inboard area. FW panels are supported with a central leg, and selection of a race-track shape cross-section for the central leg provides a more robust structure against halo current loads and also leads to a new cooling configuration in the shield block, where the pressure drop is significantly reduced to ∼0.1 MPa. Detailed EM analysis has been performed by using a newly defined plasma current quench scenario (40 ms linear decay and 18 ms exponential decay), and EM loads due to eddy currents are reduced in the current design with deeper slits. R&D results confirm the basis for the selected design solutions and fabrication methods.

Design analysis of electromagnetic forces on the KSTAR vacuum vessel interfaces

The KSTAR vacuum vessel consists of double-wall configuration with major ports, reinforcing ribs and supporting structures. The vacuum vessel structure supports all in-vessel structures. The electromagnetic loads on the plasma facing components induced during plasma disruptions would be transferred to the vessel through the supporting structure and produce a large localized stress on the interface between the vessel and the supporting structure. Due to the element limitation of load calculation code, a simplified supporting structure model was defined to obtain the electromagnetic loads on the plasma facing components. A new transferring method of these loads to the interface area was developed. In this paper the electromagnetic loads on the in-vessel components were calculated and stress analyses were performed to help design the interface area based on the equivalent forces estimated using the developed EM load transferring method.