Zirconium alloy for nuclear power

Abstract

Disclosed is a zirconium alloy material for a nuclear power reactor core, falling within the technical field of special alloy materials, and comprising the following components in percentage by weight: 0.60-1.00% of Sn, 0.80-1.10% of Nb, 0.10-0.40% of Fe, 0-0.1% of Cu or Bi or Ge, 0-0.03% of Si or S, 0.06-0.15% of O, less than 0.008% of C, less than 0.006% of N and the balance being Zr. Based on a Zr-Sn-Nb system alloy, other components used for improving the performance of the alloy are added, and appropriate component contents thereof are selected, so that the performance of the alloy provided meets the requirements of high burnup of the nuclear power reactor on the structural materials of the core. A product prepared from this alloy improves uniform corrosion resistance in ex-core pure water, especially in a lithium hydroxide aqueous solution, and improves nodular corrosion resistance in high-temperature steam.

Description

technical field [0001] The invention belongs to the technical field of special alloy materials, and in particular relates to a zirconium alloy material for nuclear power. Background technique [0002] Zirconium alloys are widely used as nuclear power reactor fuel cladding and other reactor internals due to their low neutron absorption cross section, excellent corrosion resistance and mechanical properties. During the development of light water reactors, fuel design puts forward high requirements on reactor core structural components, such as fuel element cladding, grids, guide tubes, etc. In the early days, these components were usually made of Zr-4 alloy . The design of high fuel consumption requires prolonging the residence time of these components in the stack and increasing the temperature of the coolant, so that the zirconium alloy components face a more harsh corrosion environment. These high requirements promote the improvement of the corrosion resistance of the Zr-4...

AI Technical Summary

Problems solved by technology

Although people can accept the following point of view: the materials tested and tested in 360 ° C aqueous solution and 400 ° C steam can be used in pressurized water reactors, and those tested and tested in 360 ° C lithium-containing aqueous solution are more suitable for use in pressurized water reactors In high lithium concentration working conditions, those that pass the test in steam above 500°C can be used in boiling water reactors; moreover, in the above-mentioned published documents, it has been proved by test examples that the relevant zirconium alloys are more efficient than the previous ones. The zirconium-2 and zirconium-4 alloys have more excellent properties, but whether these alloys can be really applied in practice and show satisfactory technical effects is still unknown

Moreover, none of the above-mentioned literatures gives information about the behavior of the alloy against furuncle corrosion in 500°C steam.

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