Passive safety system for a pressurized water nuclear reactor

Abstract

A passive safety system for a nuclear reactor is comprised of a first subsystem for circulating water solely by natural convection from a first branch to a second branch of a reactor coolant circuit with the circumvention of a steam generator for removing decay heat from the reactor coolant circuit at any pressure. The first branch guides heated water from the reactor vessel into the steam generator and the second branch guides cooled water from the steam generator into the reactor vessel. The first subsystem includes a heat exchanger for cooling water flowing from the first branch and prior to being introduced into the second branch and a first valve for allowing flow of water from the first branch into the second branch solely in response to a parameter value pertaining to operational safety. There is also provided a second subsystem for introducing stored cold water solely by gravity into the reactor vessel for making up for lost water in the reactor coolant circuit at any pressure. The second subsystem includes a valve for allowing the stored cold water to be introduced into the reactor vessel solely in response to a parameter value pertaining to operational safety. The passive safety system further has a third subsystem for introducing stored cold water solely by gravity into the reactor vessel when the pressure within the reactor coolant circuit is reduced at least approximately to the same pressure as in the containment, for flooding the containment to a height above the first and second branches of the reactor coolant circuit. The third subsystem includes a valve for depressurizing the reactor coolant circuit in response to a parameter value pertaining to operational safety.