Nuclear engineering and design/fusion : an international journal devoted to the thermal, mechanical, materials, structural, and design problems of fusion energy最新文献

The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory requires state-of-the-art structural-mechanics methods to deal with access constraints for plasma heating and diagnostics, alignment requirements, and load complexity and variety. Large interactive structures required an integrated analytical approach to achieve a reasonable level of overall system optimization. The Tandem Magnet Generator (TMG) creates a magnet configuration for the EFFI calculation of electromagnetic-field forces that, coupled with other loads, form the input loading to magnet and vessel finite-element models. The analytical results provide the data base for detailed design of magnet, vessel, foundation, and interaction effects.

A trade-off study has been carried out to compare the differential costs of using high-strength alloy copper versus oxygen-free, high-conductivity (OFHC) copper for the center legs of the toroidal field (TF) coils of an Ignition Spherical Torus (IST). The electrical heating, temperatures, stresses, cooling requirements, material costs, pump costs, and power to drive the TF coils and pumps are all assessed for both materials for a range of compact tokamak reactors. The alloy copper material is found to result in a more compact reactor and to allow use of current densities of up to 170 MA/m² versus 40 MA/m² for the OFHC copper. The OFHC conductor system with high current density is $24 million less expensive than more conventional copper systems with 30 MA/m². The alloy copper system costs $32 million less than conventional systems. Therefore, the alloy system offers a net savings of $8 million compared to the 50% cold-worked OFHC copper system.

Although the savings are a significant fraction of the center conductor post cost, they are relatively insignificant in terms of the total device cost. It is concluded that the use of alloy copper contributes very little to the economic or technical viability of the compact IST. It is recommended that a similar systematic approach be applied to evaluating coil material and current density trade-offs for other compact copper-TF-coil tokamak designs.

The first wall material problems and the structural mechanics studies being undertaken for the plasma facing components during the current pre-design phase of the Next European Torus (NET) project are described. The activities are aimed at generating the required materials data, setting up qualified analysis tools and criteria, and performing comparative design evaluations, thereby assisting in the final selection of materials and design concepts before embarking on the detailed design of NET at the end of the present decade.

The dynamic response of different water-cooled liquid Li₁₇Pb₈₃ breeder blanket modules has been calculated to investigate the integrity of these modules under conditions of coolant tube rupture. Numerical calculations with the code PISCES were carried out taking into account the fluid-structure interaction and the elasto-plastic behaviour of the structural material. The results show that for inert coolant characteristics the proposed conceptual designs for INTOR and NET have sufficient resistance against the pressure shock waves created by a coolant tube rupture, but when taking into account energy release due to chemical reaction of water with LiPb-alloy, up to doubling of the wall thickness has to be envisaged to guarantee structural reliability.

During the past year the Ignition Studies Project (ISP) at Princeton has been investigating design options and performing systems studies on science oriented DT ignition experiments. The goal has been to identify promising approaches leading to a device that could test the basic viability of the burning-plasma regime at a relatively early time and moderate cost. This is a substantial departure from the Tokamak Fusion Core Experiment (TFCX) which had both a physics and an engineering mission. This paper describes some of the approaches under consideration and highlights some of the tradeoffs that need to be understood for the development of an ignition experiment.

ASDEX Upgrade is a tokamak experiment with external poloidal field coils that is now under construction at IPP Garching. It can produce elongated single-null (SN), double-null (DN), and limiter (L) configurations. The SN is the reference configuration with asymmetric load distributions in the poloidal field (PF) system and the toroidal field (TF) magnet. Plasma control and stabilization require a rigid passive conductor close to the plasma. The design principles of the coils and support structure are described.

The paper reviews the configurational aspects and the structural mechanics of the overall engineering design of next-generation tokamak-type experimental power fusion reactors by describing and analysing two of the present major studies, INTOR and NET. Some safety aspects, related to containment and shielding during operation and maintenance, are also described. INTOR and NET are both considered because they constitute two different, significant approaches to defining the reactor mechanical configuration. The paper introduces the assumptions and the system integration procedures adopted to determine the mechanical configurations. The two designs are described and the structural and engineering problems related to the overall architectural aspects are identified. Finally, the main structural problems and the solution adopted are discussed.

The general design of the Joint European Torus (JET) is briefly described. The loads on its major structural components, at normal operation, and in cases of plasma instability and/or disruption, are discussed. The way these components have been assessed and optimised in relation to their loads is presented. A short account of mechanical design problems of auxiliary equipment is given. Finally, the state of operation of JET and its implications for the mechanical design is summarized.

The mechanically most important components of the JET device are the support structure of the toroidal magnet, the vacuum vessel, the coils of the magnets and the pedestals supporting the weight of the torus. These components all participate in resisting and transmitting the primary forces during operation.