Research Progress on Neutron Radiation Damage Behavior of Several Types of Cladding Materials and Nuclear Fuels

Wei Zhou*, Jiting Tian, Qijie Feng, Yong Sun, Xiankun Liu, Jian Zheng, Bin Tang, Dazhi Qian, Shuming Peng

Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang, 621999

ABSTRACT: Some achievements on several types of fuel element cladding materials and radiation damage behavior of new nuclear fuels derived from China Mianyang Research Reactor are introduced. The neutron irradiation-induced hardening behavior of the zirconium alloy Zr-4 was studied through tensile experiments and transmission electron microscopy analysis. The results show that the yield strength and tensile strength of the irradiated material increase and the elongation decreases, which is attributed to dislocation loops in the zirconium matrix induced by irradiation. Two formation mechanisms of experimental-scale dislocation loops in the zirconium matrix are proposed by molecular dynamics simulations, i.e. directly generated by a single high-energy collision cascade and direct generated by a single intermediate-energy cascade induced by dislocations. In addition, it is found that the local strain generated by the accumulation of irradiation defects will promote the appearance of <c>-shaped vacancy loop nucleation on the basal plane of hcp-Zr. Preliminary neutron irradiation and tensile tests were carried out on the zirconium alloy coated by Cr, FeCrAl and SiC/SiC composite materials. The results indicate that the first two types of alloys shows obvious hardening, meanwhile the flexural strength of the SiC composite is slightly reduced, along with the debonding between the fiber and the matrix at fracture surface. The irradiation-induced swelling behavior of metallic fuel U-10Zr was studied. We find that when the burnup depth of ²³⁵U reached about 8%, the cylindrical U-10Zr pellets swell to about 12%. The reactor irradiation test of accident-tolerant UO₂-BeO and UO₂ fuel rods with large grain size have shown that when the burnup is about 5000MWd/tU, there are obvious cracks in the core of the former material with a swelling of about 5%. By contrast, there are no cracks and swelling with an unbroken core structure in the latter. Study of higher dose neutron radiation damage on accident-tolerant cladding materials and nuclear fuels will be carried out step by step, assisting the development of new nuclear energy materials.

Keywords: zirconium alloy; nuclear fuel; irradiation damage; neutron