A problem of major concern to the operators of nuclear reactors is the radiation-induced increase in the ductile-brittle transition temperature of the steel used for the primary pressure containment vessel of the reactors. Because of the self-shielding and attenuation properties of the vessel material, a nuclear reactor pressure vessel will have a neutron flux and spectrum variation across its thickness. As a result of this variation, a pressure vessel should show various degrees of neutron-induced embrittlement throughout its thickness, and it is postulated that the embrittlement will be greatest at the inner wall and least at the outer wall. This phenomenon has been investigated by the irradiation of a large block of A302-B steel at the core face of a pool reactor in a position simulating the location of an actual pressure vessel. The steel block, 6 in. thick, was made to accommodate five equally spaced assemblies of Charpy V-notch specimens which, in turn, represented the vessel material at comparable positions.
The notch ductility test results of the irradiated specimens demonstrate a significant degree of embrittlement as well as a significant decrease in the degree of embrittlement through the simulated pressure vessel wall. However, the observed decrease is small when related to the respective variation in neutron dosage.
A correlation of the notch ductility data developed in this study to that between test reactor experiments and in-service power reactor conditions is indicated. Neutron dosage values in terms of neutrons/cm2 ( > 1 Mev) determined for positions in the test block as well as similar positions in water alone form the basis for this correlation. Thus, the values obtained enhance the validity of the >1 Mev criterion for reporting neutron dose in radiation embrittlement studies.