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Zirconium alloy for cladding material of nuclear fuel element in non-hydrogenated deoxidization pressurized water reactor

A technology for nuclear fuel elements and cladding materials, applied in the field of zirconium alloy materials, can solve problems such as corrosion acceleration and sensitivity to dissolved oxygen concentration, and achieve the effects of reducing tin content, excellent corrosion resistance, and good application prospects

Inactive Publication Date: 2017-08-11
SHANGHAI UNIV
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  • Summary
  • Abstract
  • Description
  • Claims
  • Application Information

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Problems solved by technology

In addition, the Zr-4 alloy also undergoes obvious corrosion acceleration in the 360 ​​°C / 18.6 MP a / 0.01 M LiOH aqueous solution outside the stack
The addition of alloying element niobium can significantly improve the boil-like corrosion resistance of Zr-4 alloys, but zirconium alloys containing niobium are sensitive to the concentration of dissolved oxygen in the corrosive medium

Method used

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  • Zirconium alloy for cladding material of nuclear fuel element in non-hydrogenated deoxidization pressurized water reactor
  • Zirconium alloy for cladding material of nuclear fuel element in non-hydrogenated deoxidization pressurized water reactor
  • Zirconium alloy for cladding material of nuclear fuel element in non-hydrogenated deoxidization pressurized water reactor

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Embodiment 1

[0016] Embodiment one: see table 1, wherein provides the zirconium tin series alloy of the present invention and the composition of Zr-4 alloy:

[0017] .

[0018] The alloy materials with the composition in Table 1 were prepared according to the following steps:

[0019] (1) According to the above formula and ingredients, entrust the factory to prepare alloy ingots with a weight of about 20 kg by conventional techniques;

[0020] (2) Forge the above alloy ingot at 950~1050 ℃ to form a billet, and at the same time break the coarse as-cast grain structure;

[0021] (3) After descaling and pickling, the billet is subjected to β-phase homogenization treatment at 1030-1050 °C in vacuum for 0.5-1 h and then air-cooled; then it is hot-rolled at 700-800 °C, and the oxidation is removed first after hot-rolling. Skin;

[0022] (4) After the blank is air-cooled, it is subjected to multiple cold rolling and intermediate annealing at 580 °C, and finally recrystallization annealing at...

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Abstract

The invention relates to zirconium alloy for a cladding material of a nuclear fuel element in a non-hydrogenated deoxidization pressurized water reactor, and belongs to the technical field of zirconium alloy materials. The zirconium alloy comprises, by weight, 0.73%-1.1% of Sn, 0.25%-0.6% of Fe, 0.1%-0.25% of Cr and the balance Zr and inevitable impurities. The zirconium alloy does not contain niobium and thus is insensitive to dissolved oxygen in corrosive media, and meanwhile, the alloy has extremely excellent corrosion resistance and is obviously superior to Zr-4 alloy under the corrosive condition of superheated steam at the temperature of 500 DEG C and under the pressure of 10.3 Mpa, the corrosive condition of superheated steam at the temperature of 400 DEG C and under the pressure of 10.3 Mpa, the corrosive condition of a LiOH water solution at the temperature of 360 DEG C and under the pressure of 18.6 Mpa / 0.01 M and the corrosive condition of deionized water at the temperature of 360 DEG C and under the pressure of 18.6 Mpa. The zirconium alloy is used for the cladding material of the nuclear fuel element in the non-hydrogenated deoxidization pressurized water reactor.

Description

technical field [0001] The invention relates to a zirconium alloy used as cladding material for nuclear fuel elements in a pressurized water reactor without hydrogenation and deoxygenation, and belongs to the technical field of zirconium alloy materials. Background technique [0002] Zirconium alloys have special nuclear properties (the thermal neutron absorption cross section is 0.18 barn), excellent corrosion resistance and moderate mechanical properties, and are widely used as cladding materials for water-cooled nuclear reactor fuel elements. Very important structural material. The zirconium alloy cladding is oxidized due to the corrosion of high temperature and high pressure water during work, which reduces the effective thickness of the zirconium alloy cladding and affects its service life. Zr-4 (Zr-1.5Sn-0.2Fe-0.1Cr) alloy has been widely used in water-cooled reactor nuclear power plants since the mid-1960s, and has shown excellent corrosion resistance. However, in o...

Claims

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Application Information

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Patent Type & Authority Applications(China)
IPC IPC(8): C22C16/00G21C3/07
CPCC22C16/00G21C3/07Y02E30/30
Inventor 姚美意黄娇周邦新张金龙李强
Owner SHANGHAI UNIV
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