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 unique confinement properties of the poloidal-field-dominated Reversed-Field Pinch (RFP) are exploited to examine physics and technical issues related to a compact high-power-density fusion reactor. This resistive-coil, steady-state, toroidal device would use a dual-media (i.e., two separate coolants) power cycle that would be driven by a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, and coils) having a power density and mass approaching pressurized-water-fission reactor values. A 1000-MWe(net) base case is selected from a comprehensive trade-off study to examine technological issues related to operating a high-power-density FPC. A general rationale outlining the need for improved fusion concepts is given, followed by a description of the RFP principle, a detailed systems and trade-off analysis, and a conceptual FPC design for the ∼20-MW/m² (neutrons) compact RFP reactor, CRFPR(20). Key FPC components are quantified, and full power-balance, thermal, and mechanical FPC integrations are given.

The sensitivity of lifetime predictions for fusion reactor blanket structures is investigated by applying the Monte Carlo numerical technique. A structural computer code, Stress Analysis Including Radiation Effects (STAIRE), developed for the analysis of mirror fusion blankets, is used as a deterministic model for the prediction of the lifetime of semicircular coolant tubes. Uncertainties in material variables are treated as probabilistic inputs to the STAIRE code and output distributions are obtained.

Irradiation creep rates are shown to be sufficient for relaxation of swelling-induced stresses under most conditions. In absence of high stresses, the creep limit seems to be life-limiting, although this depends on the design-dependent swelling limit. In the case of the Mirror Advanced Reactor Study (MARS) blanket design, a lifetime of several hundred dpa is shown to be highly probable.

The Next European Torus (NET) is the major plasma device envisaged, in the European strategy towards fusion, between JET and DEMO. NET aims to produce a plasma with reactor like parameters and will adopt reactor relevant technologies. A reference parameter set is being defined which offers considerable safety margins against physics uncertainties and it is optimized for minimum cost. The degree of extrapolation between NET and DEMO appears to be acceptable.

JT-60 is one of the four large tokamaks aiming at the break-even plasma condition. Construction was started in April 1978 and finished in March 1985. The structure is not only large and complex but is also subjected to huge electromagnetic forces, heat load, etc. These conditions were compensated by the supporting method, the selection of material, the accuracy of fabrication, etc. In the fabrication, inspections of many kinds were carried out to guarantee the quality. After installation at the site performance tests at full power etc. confirmed the soundness of the JT-60 machine.

A quantitative comparison of the technology requirements, environmental and cost issues of DD. Compact Reversed-Field Pinch Reactor (CRFPR) relative to a DT/CRFPR has been performed. The first wall/blanket energy recovery cycle for the DD reactor is simpler and more efficient than the DT reactor. In other technology areas (such as magnets and vacuum systems) DD requirements are not significantly different than the DT reactor. Tritium technology for processing the plasma exhaust is required to DD reactors, but no tritium containment around the blanket or heat transport system is needed. Safety analysis shows similar consequences for the release of activated corrosion products or activated first wall/blanket structure. Consequences of all postulated DD accidents for tritium releases are significantly smaller than those from the DT reactor. Cost studies have been performed for a series of DD reactors and compared with the DT reactor.

An unsteady, two-dimensional heat conduction code has been used to study the performance of swirl-flow-based neutral particle beam targets. The model includes the effects of two-phase heat transfer and asymmetric heating of tubular elements. The calorimeter installed in the Medium Energy Test Facility, which has been subjected to 30-s neutral beam pulses with incident heat flux intensities of ⩾ 5 kW/cm², has been modeled. The numerical results indicate that local heat fluxes in excess of 7 kW/cm² occur at the water-cooled surface on the side exposed to the beam. This exceeds critical heat flux limits for uniformly heated tubes with straight flow by approximately a factor of 5. The design of a plasma limiter based on swirl flow heat transfer is presented.

The residual stresses in a bonded tile/substrate structure were analyzed using both analytical and finite element methods.

Beam theory and 2-D elasticity solutions were compared and the latter was found to be more accurate, due to inadequate boundary conditions used in beam theories. Agreement between variational elasticity and finite element solutions was favorable, but the increased flexibility of finite element codes makes them superior when non-linear problems are considered.

The response of the calculated stress states to changes in various model dimensions and material parameters was studied parametrically. In general, dimensional changes were found to be significant only for short, thin tiles.

SUPERCOIL is a code for the computational design of tokamaks, including in particular ignited next-generation devices. It takes into account all physical, technical and geometrical constraints relevant to the basic design of a tokamak. Among the solutions of the model equations that meet all constraints the one optimized with respect to a prescribed figure of merit (typically capital cost) is determined. The main components modelled are the plasma, blanket and shield, the TF system and the PF system. The main characteristics of the component models are summarized and a detailed description of the solution method is given. A typical NET study point is given as an illustrative example. The validity of the code is assessed by applying it to the ASDEX Upgrade tokamak.

The austenitic steel 1.4301 (X5 CrNi 189) and the two ferritic steels 1.4914 and 1.4923 (X18 CrMoVNb 121, X22 CrMoV 121) have been tensile tested in liquid lithium in order to get information about the mechanical properties of these alloys in liquid alkali metal environment. Specimens have also been tested in lithium and in air after a pre-exposure in lithium for 1000 h at a temperature of 550°C.

The results of the tensile tests indicate that the austenitic steel is insensitive to lithium environment, when not pre-exposed to this medium before. After a pre-exposure the specimens fail in a brittle manner when tested in lithium, and in a ductile manner when tested in air. The behaviour of the two ferritic steels can be regarded as quite similar when compared to each other. The tests indicate that the steels are affected in their tensile properties even without a pre-exposure (lower ductility, brittle/shear mixed fracture mode). The reactions of the lithium are due to grain boundary effects in all steels tested. These effects are determined by the metallurgic state of the grain boundaries. The observed LME effects might be of technical significance during the procedures of starting and interrupting the reactor operation, when the critical temperature ranges are passed. The material behaviour under operating conditions of a blanket at temperatures of 400–500°C and with creep deformation should not be influenced to that degree.