The thermal neutron absorption cross section of zirconium is very small, and it has good resistance to high temperature water corrosion and mechanical properties. Therefore, zirconium alloy is widely used as the cladding material of fuel rods and the structural component of nuclear reactor core in water-cooled nuclear reactors. With the development of nuclear power reactor technology towards the direction of increasing fuel consumption, reducing fuel cycle cost, improving reactor thermal efficiency and improving safety and reliability, higher requirements are put forward for the corrosion resistance, hydrogen absorption, mechanical properties and radiation dimensional stability of fuel element cladding material zirconium alloy, the key core components. Fuel elements are subject to creep and fatigue under service conditions (irradiation, high temperature, high pressure and complex stresses). The creep property of zirconium alloy is one of the important problems to be considered when working in water-cooled power reactor. At present, the most mature and widely used zirconium alloy called -L and Zr-4 alloy is the most widely used. But with the development of nuclear fuel assembly to long life and high burnup, zirconium alloy as the structural material of reactor must have better corrosion resistance, creep resistance, radiation growth resistance and other comprehensive properties. On this point, -L and Zr-4 alloy can not meet the requirements.
The preferred composition of the zirconium alloy material is: The composition of the zirconium alloy material is: Nb 0.9% ~ 1.1 %, Sn 0.35% ~ 0.5%, Fe 0.0.3% ~ 0.5%, Cr 0.21% ~ 0.25 %, Cu 0.08% ~ 0.11 %, 0 0.09% ~ 0.14% and allowance. The preferred zirconium alloy material composition is: Nb 1%, Sn 0.5%, Fe 0.35%, Cr 0.25%, Cu 0.1%, 0.99% ~ 0.14% and Zr allowance.
The general technical person in the field should be aware that the alloy formulation may also include some other unavoidable impurity components from the raw materials which, when present in unavoidable amounts, will not adversely affect the zirconium alloy of the invention.
The invention optimized the design of the traditional Zr-Nb-Sn alloy. When the total content of Sn and dog in the alloy was controlled within a certain range, the corrosion at the weld of the alloy could be greatly reduced by adding a specific amount of Cr, thus improving the safety of the nuclear fuel assembly. At the same time, a small amount of Cu can further improve the corrosion resistance of the alloy. In summary, the zirconium alloy of the invention has better corrosion resistance and reliability, and meets the requirements of nuclear reactor structural materials.
