Nuclear Structural Engineering最新文献

The experimental studies were carried out on the influences of Mn and Si contents upon the impact strength of MnSi lime basis type weld metal in order to manufacture the arc welding electrode for the welding of the reactor pressure vessel of the Advanced Calder Hall type. The test results showed that it was possible to obtain the MnSi lime basis type weld metal with high impact strength by increasing the Mn content of weld metal in welding of heavy thick plate.

The evolution of concrete pressure vessel design is traced in this report, and a summary of the current applications of prestressed concrete for reactor pressure vessels is presented. Important design considerations and methods for design evaluation are discussed. It is concluded that a basis for design compatible with the ACI and ASME pressure vessel codes is evolving, but the maximum pressure limits for which prestressed concrete can be used are yet to be established.

Finite plastic expansion of a thick-walled spherical vessel is studied in presence of neutron radiation. Due to irradiation the yield stress and the hardening modulus vary across the shell wall. The pressure-expansion relation is obtained in case of combined linear hardening and linear attenuation.

The production of an operationally reliable pressure vessel for the reactor is one of the difficult tasks encountered in many projects of large nuclear power stations. These steel vessels have large dimensions (diameters from 3 m to 8 m) and a comparatively high internal pressure so that the basic wall thickness is sometimes amounting to several tens of centimetres. According to the type of nuclear reactor it is possible to divide the pressure vessels in the bodies working at high temperature (>250°C) and low temperature (<250°C). This paper concerns the second group of vessels, where it is not necessary to consider the creep of metals. In particular, the effect of size on quasi-brittle strength of pressure vessel steel and the effect of pulsating loads on the strength of vessels are discussed. It is emphasized, that the selection of materials, shape and size of large welded reactor pressure vessels only on the basic of standard tests of mechanical properties of materials performed with small test specimens is completely inadequate. Special attention is called to the great importance of the size factor.

The choice of reactor sites is determined by more numerous and diverse considerations than are involved in the site selection for a conventional industrial plant. The question of site selection for nuclear reactors and nuclear research centers is still a matter of much controversy. The present paper attempts to develop a systematic procedure for guidance in site selection considering hydrologic, geologic, topographic, and meteorologic conditions, as well as parameters of population density distribution, interests of neighbouring industries, agricultural use of the environments, and traffic situation. The task is to select in a systematic way, from a number of possible suitable sites the one representing an overall optimum. As illustrative examples some reactor sites in Germany are analysed using the suggested procedure. The intended use of the procedure is, of course, not aimed to the comparison of sites for different reactors, but to the comparison of different sites for a specific reactor facility.

A conceptual design of a horizontal cylindrical prestressed concrete reactor pressure vessel for housing a large gas-cooled high power-density reactor and power generating equipment is presented. The evolution of the geometry of the vessel design is described. Structural considerations for a full chamber-size opening are outlined. The choice of tendons and anchorages is discussed. Consideration is given to the practical aspects of construction.

A theoretical analysis for the determination of strains and stresses produced in long, annular reactor fuel elements is presented. In the analysis, primary effects of thermal-cycling growth, irradiation growth, swelling, creep and neutron flux levels developed in the fuel material are taken into account.

An exact solution based on a product of the modified Bessel functions and the Fourier cosine function, and a simplified, approximate solution of the parabolic function for thermal neutron flux distribution are obtained from the simple diffusion equations. After the rate of volumetric heat generation has been determined, the radial temperature distribution in the fuel is found by using Poisson's equation of heat conduction.

From some basic assumptions the fundamental equations of displacement-strain relations, compatibility, incompressibility, equilibrium, and yield criterion are established. The strain and stress equations for the fuel elements are derived. From the calculated results of an illustrative example, the neutron flux levels, degree of thermal and radiation dilatation, thickness and properties of the cladding material are found to have important effects on the strain and stress distributions produced in the fuel elements.

Based on the principle of complementary energy, a system of compatibility equations for axisymmetric structures is derived. Using the compatibility equations, an analysis method, suitable for use with the digital computer and for hand computations is developed. Equations for analysis of most of the usually encountered types of free bodies are given. Method for evaluation of matrix contributions is described, and a complete numerical sample is reproduced at the end of this article. Analysis as described herein has been programmed at The Franklin Institute and successfully used by the author for numerous reactor vessel, heat exchanger and other axisymmetric structure analyses.