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 object of this paper is to describe the thermal creep behavior and the lifetime prediction of materials subjected to non-stationary tensile loading conditions. The calculations are based on HART's tensile test equation and on a phenomenological cavitation damage model. From this model the life fraction rule (LFR) is derived. Analytical expressions for the lifetimes are derived, which contain only stationary stress rupture data. The creep behavior of non-cavitating and ideally plastic materials is derived from the solution of the tensile test equation for the particular loading conditions considered. Cavitation damage is known to influence the creep behavior by reducing the load bearing capability. The corresponding constitutive equation containing the loading conditions as well as the damage function is derived.

The following loading conditions were considered: (i) creep at constant load F and temperature T; (ii) creep at linearly increasing load and T = const.; (iii) creep at constant load amplitude cycling and T = const.; (iv) creep at constant load and linearly increasing T; (v) creep at constant load and temperature cycling and (vi) creep at superimposed load and temperature cycling.

The accumulation of radioactive corrosion and neutron sputtering products on the surfaces of components in fusion reactor coolant and tritium breeding systems can cause significant personnel access problems. Remote maintenance techniques or special treatment may be required to limit the amount of radiation exposure to plant operational and maintenance personnel. A computer code, RAPTOR, has been developed to estimate the transport of this activated material throughout a fusion heat transfer and/or tritium breeding material loop. A method is devised which treats the components of the loop individually and determines the source rates, deposition and erosion rates, decay rates, and purification rates of these radioactive materials. RAPTOR has been applied to the MARS and Starfire conceptual reactor designs to determine the degree of the possible radiation hazard due to these products. Due to the very high corrosion release rate by HT-9 when exposed to LiPb in the MARS reactor design, the radiation fields surrounding the primary system will preclude direct contact maintenance even after shutdown. Even the removal of the radioactive LiPb from the system will not decrease the radiation fields to reasonable levels. The Starfire primary system will exhibit radiation fields similar to those found in present pressurized water reactors.

TASKA-M is a conceptual design of a technology test tandem mirror. The design objective is to obtain a reasonable neutron flux and fluence for blanket and materials testing at the lowest possible cost. The configuration chosen utilizes an axisymmetric central cell with hot, mirror-confined ions, and yin-yang MHD anchors. The neutron wall loading in the central cell varies from 0.7 to 1.3 MW/m² and the estimated direct cost is about US $ 400 × 10⁶. This device can be built with only modest extrapolations in technology, e.g. magnets and neutral beams; it also represents a minimum extrapolation from the present experimental data base for tandem mirrors.

This paper describes a facility for generating engineering data on the nuclear technologies needed to build an engineering test reactor (ETR). The facility, based on a tandem mirror operating in the Kelley mode, could be used to produce a high neutron flux (1.4 MW/m²) on an 8 m² test area for testing fusion blankets. Runs of more than 100 hours, with an average availability of 30%, would produce a fluence of 5 MW·yr/m² and give the necessary experience for successful operation of an ETR.

A simplified finite element technique for analyzing friction joints in a laminated structure is presented. It is shown that by substituting the effects of adjacent layers by a set of proper nonlinear springs one can reduce a large three-dimensional analysis to an iterative solution of a two-dimensional problem which is much more efficient. The application of the technique to the analysis of high-field Bitter plate solenoids is presented.

The materials test module design for TASKA-M is presented. A representative 23000 specimen test matrix that utilizes two-thirds of the volume in a single module is described. The remaining test volume can be utilized for high heat flux or surface effects testing as well as dynamic in-situ or component testing. A peak dpa value of 78 can be achieved at the end of the 7.8 full power years of operation. The cumulative volume integrated damage level in the test modules is 4120 dpa·ℓ. TASKA-M, therefore, is most useful for interactive component testing and large specimen qualification testing.

The determination of the potential net energy gain from deuterium—tritium-fueled fusion power plants on the basis of tokamak reactors of present-day design is treated in four main sections. In the first the principle of net energy balancing is discussed in detail and the total net energy gain is deduced as a magnitude characterizing the exploitation of an energy source. Then the net energy balance for fusion power plants is determined by means of a method combining the advantages of the energetic input-output method with those of process chain analysis. Thirdly, these results are compared with LWR, HTR, FBR, and coal-fired power plants. Finally, a comparison of the results with those of other authors is made.

As a result, the potential total net energy gain from fusion power plants considerably exceeds that from LWR, HTR, and coal-fired power plants and is in the same range as that of FBR power plants.