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

A major issue in the design of fusion reactor blankets is the trade-off between tritium breeding and other blanket design requirements. While net breeding is required, the blanket design should also ensure adequate heat removal, efficient power production and sufficient shielding. A novel aqueous self-cooled blanket concept (ASCB) based on lithium compounds dissolved in water has been proposed and analyzed using one-dimensional neutronics calculations. This concept utilizes zircaloy for the structural material and a vanadium alloy for the first wall. Light water as well as heavy water systems lead to an acceptable design with respect to tritium breeding. One-dimensional tritium breeding ratios in the range 1.1–1.2 seem feasible for the proposed concept. Contrary to conventional blanket designs, the 3-D tritium breeding ratio is expected to be comparable to the 1-D performance because of the additional breeding in water cooled duct shields and high heat-flux components. The resulting design is simple, utilizes materials with a large data base, does not require additional neutron multiplying materials, and satisfies the commonly proposed criteria for a fusion blanket.

The environmental issues in the R-project are mainly considered from radiological safety point of view. Supposing a new experimental site for the program to be constructed in Toki Area, land and water surveys have been made since 1982 on the background radiation levels and radioactive material concentrations. Results of in-situ observations using instruments of various types are given and compared, such as TLD's (exposed for every three months), a pressurized ionization chamber, a NaI(Tl) detector in DBM mode, and a portable Ge detector. Also shown are the data from laboratory analyses of sampled materials in α- and γ-spectrometries, and neutron activation method. Tritium concentrations in river water samples have been quarterly measured with a liquid scintillation counter. These kinds of data will further be compiled to define the regional characteristics of the site environments. In relating to the expected radiation yields in the R-project, the importance of monitoring methods effective to pulsed radiation components is particularly emphasized. Possible use of ionization chamber and TLD's for this purpose is discussed in some detail with the preliminary application to the existing tokamak experiments on JIPP T-IIU.

An approximate analytical solution of the problem of melting and evaporation during disruptions in magnetic fusion reactors is developed via a technique known as the heat balance integral method. The heat balance integral method has been used successfully in the past in analyzing a variety of heat transfer problems, and is particularly well suited to handle nonlinear boundary value problems such as the one considered in this paper. The results obtained with this approach are then compared with results generated via a numerical techniques based computer code developed elsewhere. It is found that the results obtained via this method are quite reasonable in view of the assumptions made in developing the solution. In addition, some tentative approximate scaling laws are derived.

For the study of the alpha particles behavior in DT plasma the ‘Reacting Plasma Project’ (R-project in short) was initiated in 1981. The design study of the ‘R-tokamak’ was performed from 1981 to 1985, and evolved through three versions. The first version is based upon the usual tokamak and is similar to TFTR. The remote maintenance and dismantling however is considered very difficult. For the reduction of the residual radiation level the design of DT tokamak with aluminum alloy was conducted. It was found that hand-on maintenance after 1000 DT shots was feasible. Another effect of the aluminum alloy is the increase of the shell effect and stabilization of the variously shaped tokamak. The 3rd version tokamak with the extremely shaped cross-section (the crescent tokamak) is proposed for tokamak improvement (higher beta, better confinement and low radioactivity).

Scoping studies for NET using the SUPERCOIL system code are described. Capital cost optimized devices satisfying constraints imposed on stresses/strains, fields, access, etc. are compared. The main objectives are to determine what impact the main design characteristics, performance objectives and underlying plasma physics assumptions have on the parameters and cost of NET. A complete picture for choosing the main parameters of NET is developed and illustrated by the main NET study points used during the conceptual design phase.

The FTU tokamak will have to operate steadily at high magnetic fields in order to reach the expected performance. Its magnet and vacuum chamber, due to thermal and electromagnetic loads, will undergo very high stresses for a large number of shots.

In order to assess the feasibility of the components, numerical codes to compute current, magnetic field and temperature distributions, and extensive three dimensional finite element stress analysis have been developed. The main results obtained are illustrated.