Abstract: A passive safety system for a nuclear reactor coolant system including a water storage tank, a heat exchanger positioned within the water storage tank and connected to the hot leg, and a core make-up tank having a tank inlet connected to the heat exchanger and a tank outlet connected to a selected location of the nuclear reactor coolant system. Upon the occurrence of an abnormal condition, water flows by natural circulation from the hot leg through the heat exchanger so that heat may be transferred to the water in the water storage tank, and thereafter flows through the core make-up tank and into a selected location of the nuclear reactor coolant system to provide cooling for the reactor core.

Abstract: A pressurized water nuclear reactor pressure vessel has an upper core support plate with passageways for the passage of coolant, an upper support plate above the upper core support plate, and peripheral hollow support columns extending upwardly above the passageways in the upper core support plate for supporting the upper support plate above the upper core support plate and for guiding coolant from the passageways into the plenum. Each peripheral support column has a peripherally slotted upper portion for guiding the coolant in the peripheral support column into the plenum and an unslotted lower portion.

Abstract: A primary coolant system of a nuclear power plant for providing coolant to a reactor vessel, the primary coolant system comprises a primary loop for transferring heat away from the reactor vessel. An accumulator tank for containing water under a pressurized, gaseous atmosphere is attached to the primary loop via a coupling, which provides an outlet for the coolant from the accumulator tank to pass to the primary loop. A valve, which includes a pressure sensor that senses the pressure in the accumulator tank and opens or closes the coolant flow in the coupling in response to the sensed pressure, is attached to the coupling for sensing the pressure in the accumulator tank.

Abstract: Selectively and remotely disconnectable two-piece drive rod assemblies comprising elongated hub and drive rod portions interconnected by a selectively disconnectable coupling, connect rod clusters within the pressure vessel of a pressurized water reactor system to respectively corresponding drive means. The drive means are selectively operable, through the drive rod assemblies, to move the corresponding rod clusters between fully inserted positions within the lower barrel assembly of the pressure vessel, in telescoping relationship with fuel rod assemblies therein, and to a fully withdrawn position within an inner barrel assembly of the pressure vessel.