Fusion Engineering and Design最新文献

The minimum value of the safety factor profile is related to the magnetohydrodynamic (MHD) stability of the plasma confined in a tokamak. Therefore, active control of the minimum safety factor may mitigate MHD instabilities that can degrade or even terminate plasma confinement. Typically, in most tokamak scenarios, the minimum safety factor evolves spatially with time, i.e., the location at which the safety factor achieves the minimum value changes with time. In addition to the inherent nonlinearities in the minimum safety factor evolution, its spatial variation makes the control design challenging. In particular, complexity in control design may arise from the need for time-dependent nonlinear models that account for spatial variation of the minimum safety factor. Furthermore, the minimum safety factor may drift to locations where the actuator authority is low. The problem of minimum safety factor control with target location tracking and moving electron cyclotron current drive (ECCD) is addressed in this work. A nonlinear time-dependent model that incorporates the spatial variation of the minimum safety factor is presented. A nonlinear controller based on optimal feedback linearization is developed to track a target minimum safety factor. The proposed controller treats the ECCD position as a controllable variable. In other words, the controller prescribes the ECCD position (in addition to the non-inductive powers) in real time based on an optimal criterion that is defined a priori. This work also presents the steps necessary to integrate the minimum safety factor controller with a total energy controller to achieve multiple control objectives simultaneously. The proposed integrated control algorithm is tested using nonlinear simulations in the Control Oriented Transport SIMulator (COTSIM) for a DIII-D tokamak scenario.

A new self-cooled liquid metal blanket concept called TSLL (Toroidally Symmetric Lead-Lithium) blanket is proposed and assessed, including analysis for magnetohydrodynamic (MHD) flows, structural analysis, and heat transfer and neutronics assessments using the ARC reactor with demountable magnets designed by the Commonwealth Fusion Systems (CFS) as a testbed. The proposed blanket utilizes lead-lithium (PbLi) alloy as breeder/coolant and reduced activation ferritic/martensitic (RAFM) steel as structural material. A special feature of the new concept is the toroidally symmetric flow in the blanket integrated first wall and the breeding zone to reduce the MHD pressure drop, while using anchor links to strengthen the first wall construction. Provided analysis suggests acceptable MHD pressure drop, required mechanical integrity and high tritium breeding ratio of ∼ 1.64. As a result of these assessments, the new blanket concept can be recommended for more detailed studies as a promising blanket candidate for implementation in future fusion devices.

The feasibility of a radiatively cooled 3D-printable liquid metal heat pipe (HP) design is assessed. Using the design flexibility offered by 3D-printing, the design of the wick and geometry of the HP were optimised to meet the requirement of 20 $\mathrm{MW}/m^2$ heat load for a HP placed in a 1.5 T magnetic field. COMSOL was used to assess the operational limits of the HP, the thermal stresses in the wall, the thermally radiated power, and various materials for the HP. The main parameters are the diameter and spacing of the screen wires and the emissivity, 200 $\mu m$, 200 $\mu m$ and 0.86 respectively. Molybdenum was chosen as the wall material and lithium as the working fluid. The design was made in Siemens NX and then exported to COMSOL. From simulations it was concluded that a molybdenum HP with the final design was capable of handling a steady state heat load of 20 $\mathrm{MW}/m^2$.

The viability of the tokamak as a potential fusion reactor depends on the ability to keep the plasma in a stable regime while achieving temperatures, densities, and confinement times that are as high as possible. Tokamak scenario development attempts to find plasma regimes that achieve all of these conditions and are accessible with a given set of hardware constraints. This requires the ability to control plasma properties such as the normalized beta, the internal inductance, safety factor, rotation, etc. One property that has received less attention than some of the others, but is no less critical to achieving high performance, is the electron temperature ($T_e$) profile. In this work, Linear Quadratic Integral (LQI) control is used to develop a controller for the electron temperature profile in DIII-D. The controller is based on a linearized model derived from the transport equation that describes the evolution of the electron temperature, and includes contributions from the neural network surrogate models NubeamNet and MMMnet. The controller is tested in simulation using COTSIM, and is proven capable of tracking a target $T_e$ profile.

The ITER cryogenic system consists of the Liquid Helium (LHe) plant, the Cryo-Distribution (CD) system, and the cryo-lines. The Auxiliary Cold Boxes (ACBs) dedicated to cooling the superconducting (SC) magnet system and the Cryoplant Termination Cold Box (CTCB) of the ITER CD system are in the factory acceptance phase. The internal components of ACBs, e.g., cryogenic valves, a cold compressor (CCp), heat exchangers, and a cold circulator (CCr), have been sized and assembled, ensuring their functionality. The interdependency of the functional parameters of one component over the others needs to be assessed, as their integrated performance under the dynamic heat load deposition from the SC magnets may impact the overall operation of the ITER cryogenic system. The ACBs are equipped with two helium baths having ∼ 1200 kg of He inventory and situated inside the Tokamak building. These baths act as a thermal buffer for the LHe plant, situated in the cryoplant building, allowing it to operate at a quasi-steady state despite heat load variation from the applications. Such a large helium inventory can challenge the secondary confinement system of ITER due to helium ingress accidental events and thus needs to be optimized. The integrated system-level simulation is therefore necessary for the safe and reliable operation of the cryogenic system under such demanding requirements. The present study summarizes the results obtained for ACBs dedicated to the magnet system, including CTCB for the enhanced ITER operation modes, and confirms the integrated performance of the system. The results show that the LHe baths inside the ACBs can be used as a thermal buffer with the proposed limit of initial filling and by keeping a constant opening of the respective J-T valves upstream of the LHe baths. The study outcome and the proposed recommendations would be beneficial to mitigate the pulsed heat load to the LHe plant while minimizing the helium inventory.

To explore innovative approaches for optimizing tokamak configurations and combining the advantages of both tokamaks and stellarators, the J-TEXT tokamak recently underwent an upgrade by installing the External Rotational Transform (ERT) coil system. This system consists of two rings for producing a helical magnetic field. Due to space limitations, the ERT coil system is installed inside the vacuum vessel. The ERT coils feature a modular rail structure designed to navigate the intricate vacuum vessel environment. The successful installation of the ERT coil system on J-TEXT has yielded preliminary experimental results that align with the design objectives.

Uneven heat load distribution on the divertor during the high power long-pluse discharge of the Experimental Advanced Superconducting Tokamak (EAST) leads to hot spot phenomena, potentially causing the Plasma Facing Component (PFC) material melting, flaking, and even penetration, which may trigger the in-vessel loss of coolant accident (LOCA). Continuous coolant intrusion could damage vacuum equipment, while flash evaporation may increase vacuum pressure, posing a potential threat to the safety of the device operation. In this research, the RELAP5/MOD3.4 program was employed to develope a model of the lower divertor primary heat transfer system (PHTS). Steady state analysis was conducted to obtain the key parameters of the system in comparison with the design parameters, and the results showed good consistency. Thermal-hydraulic analysis of the in-vessel LOCA is performed based on the design condition, quantitatively investigating the evolution of the breach discharge flow rate and vacuum pressure. An additional pneumatic isolation valve and check valve are proposed as an accident mitigation scheme, and the effectiveness is evaluated to provide a reference for the upgrade of EAST lower divertor PHTS.

To evaluate the liquid metal embrittlement (LME) susceptibility of F82H, a reduced activation ferritic-martensitic (RAFM) steel, a testing procedure using hollow cylindrical tensile specimens was used. Tensile tests are compared between specimens filled with argon and lithium at 200 °C. To validate the procedure, initial testing was performed on type 4340 steel, which is well-known to exhibit LME. Compared to 4340 steel, F82H only showed minor effects of Li exposure, including pre-testing exposures with Li at 400 °C for 1 h and 500 °C for 500 h. Furthermore, changing the strain rate or tensile test temperature also did not show significant embrittlement.

The European DEMOnstration Fusion Power Plant DEMO represents a significant milestone in the progression towards sustainable fusion energy and a critical phase between ITER and commercial fusion reactors, aiming to demonstrate sustained net positive electricity production. Thanks to its properties, tungsten is a promising material for divertor armor. Coupled with copper alloys as heatsinks, they offer robust thermal management properties to deal with intense thermomechanical loads and irradiation damage. Understanding the thermomechanical behaviour of tungsten-copper joints during their application is then necessary for divertor design.

This study presents experimental analysis on tungsten-copper brazed materials subjected to thermomechanical solicitations to simulate mono-block conditions with heat fluxes expected to reach 20 MW/m² and so to face potential creep-fatigue failure. The experimental tests were coupled with Digital Image Correlation up to 400 °C to analyse the thermomechanical behaviour of these joints, providing insights into their thermal behaviour, structural integrity, damage accumulation, joint failure and identification of strains required for creep-fatigue assessment using design codes.