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 effects of radiation on the structural performance of fusion reactor structures is recognized as a major issue for the development of fusion reactor technology. Neutron irradiation changes the mechanical properties of structural components resulting in a general degradation of these properties. In addition to the mechanical loads (pressure and weight) and the thermal strains, non-uniform inelastic strain fields are induced by radiation swelling and creep in fusion structures. In this paper, we describe a new computer code, STAIRE, for $\text{ST}$ress $\text{A}$nalysis $\text{I}$ncluding $\text{R}$adiation $\text{E}$ffects. This code is based on standard beam theory for pipe-bends. The theory is modified in two areas: (1) consideration of the pipes' cross-section deformation as the radius of curvature changes; and (2) inclusion of inelastic radiation and thermal strains (swelling and creep). An efficient analytical/numerical approach is developed for the solution of indeterminate beam problems. As an application of the method, the stress distribution and deflections of toroidal blanket pipes in the Mirror Advanced Reactor Study (MARS) are evaluated. Swelling strains are identified as a major source of stress and deformation in the proposed blanket design, and possible solutions to the problem are outlined.

For the purposes of both developing appropriate alloys for fusion reactors and ultimately designing and building such devices, it is necessary to determine the mechanical response to high energy neutron irradiation of candidate structural materials. The current and near-term facilities for providing the requisite testing environment are either limited by dissimilarity to fusion reactor environments (i.e., fission reactors), or irradiation volume, or both. This necessitates the development of both a testing technology commensurate with the available facilities — i.e., primarily small specimen technology — and a correlation methodology which recognizes the limitations of the anticipated data base and maximizes the use of that data. In this paper we (1) describe a general philosophy for such correlation methodologies, (2) provide a brief overview of small specimen test techniques and data characteristics, and (3) demonstrate with several examples how correlations can be developed and applied.

Eddy current and magnetic force analysis during disruption of single-null divertor plasma has been carried out for a vacuum vessel of a tokamak device by using a code system, MATEX, which was developed for this purpose. The electromagnetic behavior during a disruption with plasma motion is emphasized in this study. The large upward or downward force induced on the vacuum vessel is due to midplane-asymmetric poloidal field configuration, and is intensified by vertical plasma motion in disruption. On the other hand, the saddle-like force at the inboard section of the flange is increased by the horizontal plasma motion during disruption. The results indicate that the analysis of the effect of plasma motion is very important to the design of a vacuum vessel and its support structure.

Pulsed operation of a tokamak reactor imposes cost penalties due to such problems as mechanical fatigue and the need to periodically transfer large amounts of energy to various reactor components. This study focuses on lifetime limitations and capital costs of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas include: fatigue in pulsed poloidal field coils; out-of-plane bending fatigue in toroidal field coils; electric power supply costs; and noninductive current driver costs. A capital cost comparison is made for tokamak reactors operating under the four distinct operating cycles which have been proposed. Since high availability and a low cost of energy will be mandatory for a commercial fusion reactor, we can characterize improvements in physics and technology which will help achieve these goals for different burn cycles. A key conclusion is that steady-state operation is likely to result in the least expensive tokamak reactor (perhaps 20% cheaper than the best pulsed reactor), provided noninductive current drive efficiency can be increased roughly four-fold over present-day experimental results.

The effects of helium on the high temperature mechanical properties of austenitic stainless steel, also known as high temperature helium embrittlement, are reviewed. Special emphasis is given to such experimental data which simulate a fusion reactor environment best, i.e., to creep and low frequency fatigue experiments at high helium levels (≥ 100 appm He). Correlations between mechanical properties data and microstructural data are presented and compared to recently developed models for helium embrittlement. The presented data are critically assessed in the light of fusion reactor materials needs and some suggestions for future work are made.

It is shown that the cyclic stress fatigue due to the over-turning force on the toroidal field coils in a tokamak fusion reactor can be greatly reduced by appropriately changing the plasma current in the recharging phase of the quasi-steady operation scenario with the use of lower hybrid wave current drive. This substantial reduction is made possible, if the characteristics of required variation of the equilbrium magnetic field for the dee-shaped divertor plasma is properly taken into account in adjusting the plasma current, in which the necessary equilibrium field becomes larger in low beta phase than in high beta phase for the same plasma current.

The paper compares the key features of four recent designs of tandem mirror test facilities (TASKA, TDF, TASKA-M and MFTF-B Upgrade) and points out similarities and differences in these designs. Some reference is also made to the relative cost of each device.

An analytic method is proposed to determine the startup scenario with the RF current drive. The scenario is obtained from the viewpoint of minimizing the RF injection energy by means of the calculus of variations. It is found that the scenario of RF current startup has distinct differences from that of inductive startup. For example, this analysis indicates that the poloidal beta is kept at the maximum value, the temperature is increased in an early startup phase to decrease the current drive power and the plasma current is ramped up slowly at high temperatures to suppress the return current induced. The numerical calculations reveal that the startup time and the RF injection energy are sharply decreased by the impurity contamination or the deterioration of the plasma confinement during the startup.