Effect of reactor environment on embrittlement of Cr-Mo-V steel

The effect are studied of neutron radiation, applied strain and corrosion medium on the propagation of corrosion cracks in the materials of PWR pressure vessels. Tested were low alloy steels Cr-Mo-V and Cr-Ni-Mo-V. Tensile properties were tested on cylindrical samples with the diameter d=4 mm at the deformation rate 8.3x10-4 s-1. Cylindrical samples with a 60deg V notch, diameters d1=3.5 mm, d2=5 mm, l=17.5 mm and notch radius 0.5 mm were tested at constant load. Irradiation was carried out on the WWR-S reactor with neutron fluence 2.6 to 4.9x1023 n.m-2 (E>0.5 MeV) at temperatures 130, 180 and 290 degC. Irradiated and non-irradiated samples were charged with hydrogen in 1N H2SO4 for one hour before the test. The experiment determined the value of threshold stress under which samples of steel are not disturbed even at high current densities: for Cr-Mo-V steel, this is 210 MPa, for Cr-Ni-Mo-V it is 410 MPa. The susceptibility of unirradiated steel to hydrogen embrittlement following hydrogen charging occurs at a hydrogen content of over 2.5 ppm. The effect of radiation embrittlement on hydrogen embrittlement is indirectly proportional to the irradiation temperature. Hydrogen charging of samples irradiated at 130 and 180 degC with a fluence greater than 2.6x1023 n.m-2 leads to the superposition of hydrogen and radiation embrittlement thereby often resulting in total loss of plasticity. (Pu)