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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.

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The embrittling effect of neutron irradiation on reactor pressure vessel steel and welds has a significant impact on the regulation of reactors by the U.S. Nuclear Regulatory Commission (NRC). Because the NRC is concerned with the continuing integrity and safety of the reactor's primary system, the neutron-irradiation-induced loss of fracture toughness and ductility must be carefully documented, understood, and controlled to assure that despite such loss, there is always a sufficient reserve of toughness and ductility to preclude crack initiation and uncontrolled propagation in the unlikely event of an accident. Aspects of embrittlement important to the NRC, which are discussed herein, include prediction of the transition temperature and upper-shelf-energy levels at critical vessel locations for the setting of pressure-temperature limits, for analysis of accident loadings such as pressurized thermal shock, and for evaluation of flaws found in inspection. Also discussed are the use of embrittlement-related neutron dosimetry and evaluation of vessel annealing.