Irradiation hardening simulation method in zirconium-based alloy based on rate theory and model system

Abstract

The invention discloses a zirconium-based alloy irradiation hardening simulation method and model system based on a rate theory, and relates to the technical field of nuclear material irradiation simulation. The key points of the technical scheme are as follows: establishing an evolution and irradiation growth model for simulating a dislocation structure; establishing a strain rate calculation model during irradiation growth; inputting the obtained material parameters into a strain rate calculation model during irradiation growth; simulating and calculating the influence relationship of the irradiation dose on the dislocation ring radius, the dislocation density and the growth strain according to the numerical simulation parameters through a strain rate calculation model during irradiation growth; calculating to obtain a hardening value caused by the point defect absorption trap through a pre-constructed hardening degree calculation model according to the influence relationship of analog calculation. According to the method, the influence of the irradiation dose on the dislocation ring radius, the dislocation density, the growth strain and irradiation hardening can be simulated and calculated, the influence of tin and niobium as alloy elements in the zirconium-based alloy on the mechanical property and irradiation growth can be simulated, the prediction range is wide, and the calculation result is accurate.

Description

technical field [0001] The invention relates to the technical field of nuclear material radiation simulation, more specifically, it relates to a method and a model system for simulating radiation hardening in zirconium-based alloys based on rate theory. Background technique [0002] The fuel element is the core component of the reactor, mainly composed of fuel pellets and cladding. Among them, the cladding material is one of the most important reactor structural materials and the first safety barrier of the nuclear reactor. The thermal neutron absorption cross section of metal zirconium is small, and the zirconium alloy made by adding alloy elements has good high temperature water corrosion resistance and good mechanical properties, so it is often used as cladding material for reactor fuel elements. Under severe reactor conditions (high temperature, high pressure), there are many failure modes of zirconium cladding, such as water side corrosion, hydrogen embrittlement, cree...

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Problems solved by technology

In terms of nuclear materials, French researchers established a cluster dynamics model of pure zirconium, and the results obtained are similar to the experimental trend, but it is still unable to accurately simulate the evolution data of defects in zirconium alloys under neutron irradiation

Domestically, the Institute of Solid State Physics of the Chinese Academy of Sciences and Fudan University have carried out related research using rate theory, but none of them can accurately simulate the concentration of irradiation defects in nuclear materials, let alone obtain data and parameters that can be in line with engineering or experiments.

[0005] After investigation, the radiation properties of zirconium alloys are extremely complex, and there are still many unsolved problems in radiation hardening and radiation growth mechanisms.

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