The Features of Heat and Mass Transfer in the Operation of Nuclear-Powered Icebreaker Steam Generators

Abstract

The problem of hydrogen induced embrittlement of and crack formation in the piping systems of nuclear-powered icebreaker steam generators is considered. Information about the relevance of this problem for ensuring reliable operation of the nuclear power installations of transport facilities and similar systems such as floating nuclear power units is given. Hydrogen accumulation in the steam generator feedwater header pipes welding zone and their subsequent hydrogen-induces embrittlement lead to crack formation. For modeling the hydrogen transport processes from the reactor to the steam generator, two types of computer codes are jointly used: 1D approximation and 3D ones. The performed analysis has identified the main characteristic processes. For solving the problem, it was decomposed into subproblems. The transport of vapor-gas bubbles in the steam generator was analyzed with paying special attention to studying their behavior in the steam generator internal cylinder. It is shown that, despite the presence of structural obstacles in the inner cylinder, coolant convective motion and bubble floatation occur. An alternative hypothesis is put forward, according to which cracks may be caused by cavitation on the secondary system coolant side. To check the hypothesis, 3D modeling was carried out, the results of which did not confirm its validity. The obtained study results can be used to substantiate measures on preventing the hydrogen-induced embrittlement of and crack formation in the steam generator piping systems of nuclear-powered icebreakers, floating nuclear power units, and small capacity NPPs.