Zirconium-tin-niobium corrosion-resistant zirconium-base alloy for nuclear fuel cans

A technology for zirconium-based alloys and nuclear fuels, which is applied in the field of corrosion-resistant zirconium-based alloys and zirconium-tin-niobium-based corrosion-resistant zirconium-based alloys for nuclear fuel cladding, can solve the problems of corrosion performance decline, grain boundary segregation, etc. The effect of corrosion resistance, excellent comprehensive performance and excellent creep resistance

Active Publication Date: 2010-08-18
西安西部新锆科技股份有限公司
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Problems solved by technology

[0019] In order to improve the creep resistance of zirconium alloys, most of the existing patents tend to add a small amount of sulfur (S) and silicon (Si) to zirconium alloys, but these

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  • Zirconium-tin-niobium corrosion-resistant zirconium-base alloy for nuclear fuel cans
  • Zirconium-tin-niobium corrosion-resistant zirconium-base alloy for nuclear fuel cans

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[0044] Example

[0045] Use nuclear grade sponge zirconium, Nb, MgO, CeO 2 , Y 2 O 3 , SiO 2 , V 2 O 3 , O and other elements are prepared in the form of master alloy according to the formula and pressed into electrodes, and are smelted three times in a vacuum consumable electric arc furnace to produce 6 alloy components of 3kg alloy ingots; sample the ingots for chemical composition analysis, see the alloy composition Table 1. The alloy ingots are forged at 980℃~1050℃ to make billets; they are heated in the β phase zone at 1015℃~1075℃ and then quenched by aqueous medium; the billets are hot-rolled below 620℃, and the deformation is greater than 60%. Intermediate annealing at 600℃, and then cold rolling for many times. The deformation of fire is more than 50%. Intermediate annealing is carried out at the same annealing temperature as that after hot rolling to form a plate. The final product is processed by recrystallization annealing at 580℃ to produce Obtain the zirconium alloy...

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Abstract

The invention discloses a zirconium-tin-niobium corrosion-resistant zirconium-base alloy for nuclear fuel cans, which comprises the following components in percentage by mass: 0.6-1.4% of Sn, 0.10-1.5% of Nb, 0.1-0.5% of Fe, 0.02-0.3% of Cr, 0.005-0.5% of MgO, 0-0.5% of CeO2, 0-0.5% of Y2O3, 0-0.015% of SiO2, 0-0.03% of V2O3, 0.07-0.15% of O, and balance of Zr and inevitable impurities. Compared with the prior Zr-4 alloy, the zirconium-base alloy of the invention has better high temperature creep resistance and excellent corrosion resistance, and is suitable for can materials, grids and other structural members for reactor fuel rods in a nuclear power station.

Description

technical field [0001] The invention belongs to the technical field of zirconium-based alloy materials, and relates to a corrosion-resistant zirconium-based alloy that can be used as fuel rod cladding materials, grids and structural parts in nuclear reactor cores of light water and heavy water nuclear power plants. Zirconium-tin-niobium corrosion-resistant zirconium-based alloys for material-strengthened nuclear fuel cladding. Background technique [0002] The thermal neutron absorption cross section of zirconium is very small, and has good high temperature water corrosion resistance and mechanical properties, so zirconium alloys are widely used as cladding materials for fuel rods and structural elements of nuclear reactor cores in water-cooled nuclear reactors. With the development of nuclear power reactor technology in the direction of increasing fuel burnup and reducing fuel cycle costs, increasing reactor thermal efficiency, and improving safety and reliability, higher r...

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Application Information

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IPC IPC(8): C22C16/00G21C3/07
CPCY02E30/40Y02E30/30
Inventor 周军李中奎张建军田锋石明华王文生
Owner 西安西部新锆科技股份有限公司
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