Bubble coalescence in solids is assumed to result when collisions occur between bubbles as they migrate by surface diffusion through the solid. The migration of an isolated pore is examined in detail and the results are applied in a finite-difference approach, programmed for a digital computer, to predict the variation of the number of bubbles per unit volume as a function both of bubble size and of post-irradiation annealing time. Two idealized models of bubble coalescence have been treated. The first model considers coalescence resulting from random migration of the bubbles, and the second considers coalescence resulting from llnidirectional, biased migration, such as would result in the presence of a thermal gradient. Both models are based on surface diffusion migration of randomly distributed bubbles in a perfect, infinite solid. The results give the predicted variation with time of the entire distribution of bubble sizes, so that various parameters, including the mean bubble radius and volume change, can be predicted. Comparison of the predicted values to parameters measured in appropriate experiments could provide a means of measuring fundamental properties of materials. The results can also be applied to the problem of swelling of reactor fuel materials because of trapped fission gases. Although the present analysis is limited by a number of simplifying assumptions, it is clear that the presence of a thermal gradient can greatly enhance the swelling rate.
Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of reactor pressure vessels because of their superior fracture toughness properties. However, exposure to high levels of neutron irradiation for extended periods leads to significant reduction in the fracture resistance of these steels. Experimental data are presented on fracture toughness and crack growth rates (CGRs) of austenitic SSs irradiated to fluence levels up to 2.0 x 10 21 n/cm 2 (E > 1 MeV) (ª3.0 dpa) at ª288°C. Crack growth tests were conducted under cycling loading and long hold time trapezoidal loading in simulated boiling water reactor (BWR) environments, and fracture toughness tests were conducted in air. Neutron irradiation at 288°C decreases the fracture toughness of the steels; the data from commercial heats fall within the scatter band for the data obtained at higher temperatures. In addition, the results indicate significant enhancement of CGRs of the irradiated steels in normal water chemistry BWR environment; the CGRs for irradiated steels are a factor of ª5 higher than the disposition curve proposed for sensitized austenitic SSs. The rates decreased by more than an order of magnitude in low-dissolved-oxygen BWR environment.
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