Abstract: A method for storing the energy of a nuclear power plant in which the nuclear core is cooled by gases or liquid heat transfer media. The hot heat transfer liquid is stored directly in storage tanks. When needed, it is used for heating a power plant. The heat of a compressed gas heat transfer medium such as helium is stored by passing the compressed gas through tanks filled with heat-resistant solids and recovered by passing the same type of gas in a second circuit in a reverse direction. Through the hot tanks to the power plant and back. This Abstract is not intended to define the invention disclosed in the specification, nor intended to limit the scope of the invention in any way.

Abstract: A steam generator system for a pressurized water reactor which employs an external to containment steam drum and recirculation loop piping. The steam generator system changes the arrangement of a typical pressurized water reactor recirculation steam generator by relocating the functions of steam separation and feedwater preheating outside of the reactor coolant system. The steam generator system and thermal hydraulic conditions are selected in order to minimize the size of the steam generator heat exchanger component volume inside of the containment. The external steam drum component can be isolated in accident conditions when desired and is used as a source of secondary fluid inventory for improved decay heat removal capability and tolerance for loss of feedwater events. Thus, the steam generator component volume inside of the containment is reduced and the amount of maintenance required for the reactor coolant system components are similarly reduced.

Abstract: A feed water and steam header equipped to a feed water nozzle or a steam nozzle of a reactor vessel in a steam generator for an integrated nuclear reactor is provided. The feed water and steam header may include a nozzle connection portion connected to a steam nozzle or a feed water nozzle of a reactor vessel, a header flange protruded outward from a lower part of the nozzle connection portion, and a tube connection portion disposed on two pipelines branched from the nozzle connection portion and connected to a tube of the steam generator.

Abstract: A pressurized water nuclear reactor (PWR) includes a once through steam generator (OTSG) disposed in a generally cylindrical pressure vessel and a divider plate spaced apart from the open end of a central riser. A sealing portion of the pressure vessel and the divider plate define an integral pressurizer volume that is separated by the divider plate from the remaining interior volume of the pressure vessel. An internal control rod drive mechanism (CRDM) has all mechanical and electromagnetomotive components including at least a motor and a lead screw disposed inside the pressure vessel. Optionally CRDM units are staggered at two or more different levels such that no two neighboring CRDM units are at the same level. Internal primary coolant pumps have all mechanical and electromagnetomotive components including at least a motor and at least one impeller disposed inside the pressure vessel. Optionally, the pumps and/or CRDM are arranged below the OTSG.

Abstract: Nuclear reactors are customarily cooled by water from a natural source in the area. Water brings impurities to and surrounds the fuel rods with a mix of materials, some of which react with the fuel rod contents. Changing the coolant to a pure, inert gas sourced from its cryogenic liquid form with ambient pressure gives greater control of the situation and enables running the reactor at the critical point of water so the cycle of coolant is released to the purifier carrying whatever material is expelled by the fuel rods and the steam cycle leaves the radiator from the reactor chamber as steam leaving little, if any, water release from the nuclear plant with no impurities but what is emitted by the fuel rods themselves contaminating the rod environment. Eliminating the hot water surrounding the plant, security of the Nuclear site is greater since infrared sighting is prevented with shielding just the reactor.

Abstract: Core barrel for a high-temperature nuclear reactor, including internal parts formed of an inner layer of graphite blocks stacked on each other and an outer layer of carbon blocks stacked on each other together forming columns, the blocks of the respective layers being staggered in height, offset relative to each other in azimuth and having indentations alternately formed therein causing each two layers of graphite blocks to be supported on a protruding layer of carbon blocks and each two layers of carbon blocks to be supported on a protruding layer of graphite blocks.