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

Using the INTOR tokamak study as an example, this paper aims at illustrating the various properties and requirements in an integrated form and this way to give reasons why certain performance data of certain components have to be achieved, and how the different requirements are inter-related.

The methods to estimate the effects of the eddy currents in the conductive components, such as the first wall, blanket, shield and coil vacuum chamber, on the vertical position control in tokamak reactors are presented. The classical PID feedback control is investigated and the control properties of the Fusion Experimental Reactor (FER) design of JAERI are discussed. The shell structures in the breeding blanket modules, shield structures and belljar type of coil vacuum chamber have important effects on the plasma vertical position controllability. In this FER design, the feedback control power with the coil inside the shield and inner belljar is smaller by a factor of − 10 than the control by the outside coil. The derivative action of the PID controller is essential to control the plasma vertical position stably.

Three miniature sheet-type tensile specimens and a miniature rod-tensile specimen are being used in irradiation-damage studies for fusion reactor materials. Miniature specimens are also being used to determine design properties of irradiated materials. Consequently, a comparison of the tensile properties determined with these different specimens was made by testing cold-worked and solution-annealed Type 316 stainless steel sheet and rod at room temperature, 300°C and 600°C. Good agreement was observed when the properties obtained with the different specimens were compared, and all of the specimens can be used for irradiation-damage studies. However, there were small differences in the properties obtained using the different specimens, and it was recommended that before any miniature specimens are used to determine design-type properties, further studies be conducted.

TASKA-M is a conceptual design of a low power, low cost test reactor based on a tandem mirror without thermal barriers utilizing physics which is presently in hand, or which should be available by 1985. The main purpose of the reactor is to study plasma engineering and provide a reasonable cost test bed for qualifying materials and investigating different blanket concepts. The central cell which is nominally 4.25 m long is divided into five zones: a liquid metal test module, a solid breeder test module, two material test modules and a central cell shield insert. This paper describes the structural, thermal hydraulics and neutronic aspects of the breeding test modules. Two liquid metal test blanket modules, one using Li₁₇Pb₈₃ and the other elemental Li are designed to fit into the same slot within the central cell and will, therefore, be tested consecutively. The solid breeder test module is designed to test all kinds of breeding materials rather than as a reactor prototypical module with a high breeding ratio. The main emphasis is on fast, easy insertion and withdrawal of the solid breeder material samples. A breeding ball blanket concept has been developed, where the Li-ceramic is contained in spherical stainless steel shells which are inserted into pressure tubes coiled around the plasma.

Several methods for developing design correlations for ferritic steels are discussed in this paper. Equations which describe swelling, embrittlement, and irradiation creep are reviewed. We develop design equations for use in inelastic structural mechanics applications, for the most important thermal creep parameters. Empirical correlations for creep rupture time and the complete description of elongation vs. time are presented. A phenomenological description of steady-state creep is also developed.

TASKA-M is a study for a mirror based D—T plasma device for fusion technology tests. Mature technology was applied whereever possible. The axial confinement time is relatively short, resulting in a large gas throughput compared to the fusion power level of 6.8 MW. The technological requirements of the exhaust system will not cause undue development problems in either of the two major areas: highly loaded dumps for the conversion of the escaping particle and plasma streams to thermal gas; vacuum pumping facilities.

It is shown that neglect of resonance self-shielding of fertile materials causes gross overestimation of fissile breeding in fission—fusion hybrids and threatens unexpected criticality of water cooled spallation target/blanket assemblies even at low enrichments. In the latter case the spallator's task, i.e. fissile breeding cannot be fulfilled. The importance of resonance self-shielding is fully confirmed in the performed calculations, primarily in soft spectrum blankets. Consequently, the necessity of consideration of the resonance self-shielding effects is proven and the resignation of use of moderators in fissile breeding systems is suggested.

The use of bellows in a magnetic fusion reactor gives, in many cases, more efficient penetration of the poloidal electric field and very efficient plasma heating in the first phase of operation. This is because bellows sections lead to higher electrical resistance of the vacuum vessel. There are many examples of its application to tokamak type experimental devices such as TFTR, JT-60 and so on. In addition to its higher electrical resistance, the bellows has an ability to absorb large deformation due to thermal expansion and radiation creep, and thus has been applied to the design of a first wall configuration in some conceptual designs of a commercial power reactor. Its application to a vacuum vessel of the Reacting Plasma machine, which was proposed by the Institute of Plasma Physics of Nagoya University, was examined for circular plasmas. When applying the numerical analysis to the bellows, a computational difficulty occurs. Too many unknows appear in a mesh division with a general shell element and these are not easily handled. Therefore in the present paper a finite element code is made to consider an axisymmetric structure loaded by nonsymmetric forces. In the code various types of electromagnetic forces are taken into account. These include the diamagnetic electromagnetic force, the toroidal-current-induced-electromagnetic forces, the electromagnetic forces due to toroidal coil current change and saddle shaped electromagnetic forces. Aluminum is selected as bellows material because of its low activation nature but there are large induced currents because of its higher electrical conductivity. Results of stress analysis show that a leak of the saddle shaped current into the thick aluminum vacuum vessel generates maximum bending stress near a joint between the bellows and the vessel. Stress due to other types of electromagnetic force are significantly smaller.

Ceramics are required to serve in a conventional role as electrical and thermal insulators and dielectrics in fusion power reactors. In addition, certain ceramic materials can play a unique role in fusion power reactors by virtue of their very low induced radioactivity from transmutation products produced by fusion neutron capture. The aspects of safety, long-term radioactive waste management, and personnel access for maintenance and repair can all be significantly improved by applying the low-activation ceramics to the first wall and blanket regions of a fusion reactor. This application imposes tensile, compressive, and shear structural loads and thermal stresses on the materials, and it is primarily in support of tensile stresses where problems in ceramic design lie.

Silicon carbide, carbon, and graphite materials are three primary candidate structural ceramics. Electrical insulators and radio frequency electromagnetic wave windows commonly employ ceramics, such as Al₂O₃, MgO, SiO₂, Si₃N₄ and glasses. Material properties characteristic of the radiation damaged state must be used. The structural failure modes of low-ductility ceramics are by immediate fracture when a critical stress is applied or by slow crack growth eventually propagating to fracture. Both failure modes follow a statistical distribution with a finite probability of failure with low applied loads. Proof testing, however, can reduce this probability of failure to zero below some threshold stress by eliminating the weaker components, thus easing design problems.

Design studies have been performed to develop conceptual designs of fusion power reactor components using low-activation ceramic and metallic materials. These components include limiters, first walls, blanket modules, shields, superconducting magnets, diagnostic instrumentation, electrical insulation, and radio frequency windows. Present day ceramics can fulfill all the functional requirements of these components without undue performance penalties. Improved ceramic materials, both monolithic and fiber composites, are being developed at a rapid pace, and these can readily be applied to improved fusion designs. It thus appears possible to design a fusion reactor using only low-activation ceramic materials, principally structural ceramics, in the high-neutron flux zones of the reactor. Presently operating fusion plasma devices employ graphite for limiters and armor, and ceramics for electrical insulators, providing a base for continued utilization when power-producing devices are built. The ultimate potential of fusion as an environmentally benign energy source with a high degree of safety and public acceptance is optimally achieved through the use of low-activation structural ceramics.