An external reactor vessel cooling and electric power generation system according to the present invention includes an external reactor vessel cooling section formed to enclose at least part of a reactor vessel with small-scale facilities so as to cool heat discharged from the reactor vessel, a power production section including a small turbine and a small generator to generate electric energy using a fluid that receives heat from the external reactor vessel cooling section, a condensation heat exchange section 140 to perform a heat exchange of the fluid discharged after operating the small turbine, and condense the fluid to generate condensed water, and a condensed water storage section to collect therein the condensed water generated in the condensation heat exchange section, wherein the fluid is phase-changed into gas by the heat received from the reactor vessel. The external reactor vessel cooling and electric power generation system according to the present invention can continuously operate even during an accident as well as during a normal operation to cool the reactor vessel and produce emergency power, thereby enhancing system reliability. The external reactor vessel cooling and electric power generation system according to the present invention can easily apply safety class or seismic design using small-scale facilities, and its reliability can be improved owing to applying the safety class or seismic design.

A steam generator comprising a vessel having an inlet and an outlet, and in use a primary fluid flow enters the vessel through the inlet and exits the vessel through the outlet. A plurality of modules are connected in series and at least partially housed within the vessel, and each module comprises at least one tube. The modules are arranged such that at least one tube of one module is coaxial with at least one tube of an adjacent module so as to define a conduit through which a secondary fluid can flow from one module to an adjacent module.

A nuclear power system includes a reactor vessel that includes a reactor core mounted within a volume of the reactor vessel, the reactor core including one or more nuclear fuel assemblies configured to generate a nuclear fission reaction; a containment vessel sized to enclose the reactor vessel such that an open volume is defined between the containment vessel and the reactor vessel; and a boron injection system positioned in the open volume of the containment vessel and including an amount of boron sufficient to stop the nuclear fission reaction or maintain the nuclear fission reaction at a sub-critical state.

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction; a riser positioned above the reactor core; a primary coolant flow path that extends from a bottom portion of the volume through the reactor core and through an annulus between the riser and the reactor vessel; a primary coolant that circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to generate electric power in a power generation system; and a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

A nuclear power system includes a reactor vessel that includes a reactor core mounted, the reactor core including nuclear fuel assemblies configured to generate a nuclear fission reaction; a riser positioned above the reactor core; a primary coolant flow path that extends from a bottom portion of the volume below the reactor core, through the reactor core, within the riser, and through an annulus between the riser and the reactor vessel back to the bottom portion of the volume; a primary coolant that circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the received heat to generate electric power in a power generation system fluidly or thermally coupled to the primary coolant flow path; and a control system communicably coupled to the power generation system and configured to control a power output of the nuclear fission reaction independent of any control rod assemblies during the normal operation.

A compact pressurized water nuclear reactor having connected to the reactor pressure vessel a plurality of horizontal pressure vessels, with all the horizontal pressure vessels connected to the reactor pressure vessel by a single connection between the respective nozzle of the reactor pressure vessel with the respective nozzle of each horizontal pressure vessel.

A nuclear steam supply system utilizing gravity-driven natural circulation for primary coolant flow through a fluidly interconnected reactor vessel and a steam generating vessel. In one embodiment, the steam generating vessel includes a plurality of vertically stacked heat exchangers operable to convert a secondary coolant from a saturated liquid to superheated steam by utilizing heat gained by the primary coolant from a nuclear fuel core in the reactor vessel. The secondary coolant may be working fluid associated with a Rankine power cycle turbine-generator set in some embodiments. The steam generating vessel and reactor vessel may each be comprised of vertically elongated shells, which in one embodiment are arranged in lateral adjacent relationship. In one embodiment, the reactor vessel and steam generating vessel are physically discrete self-supporting structures which may be physically located in the same containment vessel.

A control rod drive system (CRDS) for use in a nuclear reactor. In one embodiment, the system generally includes a drive rod mechanically coupled to a control rod drive mechanism (CRDM) operable to linearly raise and lower the drive rod along a vertical axis, a rod cluster control assembly (RCCA) comprising a plurality of control rods insertable into a nuclear fuel core, and a drive rod extension (DRE) releasably coupled at opposing ends to the drive rod and RCCA. The CRDM includes an electromagnet which operates to couple the CRDM to DRE. In the event of a power loss or SCRAM, the CRDM may be configured to remotely uncouple the RCCA from the DRE without releasing or dropping the drive rod which remains engaged with the CRDM and in position.

A method for heating primary coolant in a nuclear reactor system during system start-up. A primary coolant loop fluidly couples together a reactor vessel and a steam generating vessel. The primary coolant loop is filled with primary coolant. A portion of the primary coolant is taken from the primary coolant loop and placed into a start-up sub-system. The portion is heated while in the sub-system to form a heated portion of the primary coolant. The heated portion is returned into the primary coolant loop. The method allows for the primary coolant to be heated to a no-load operating temperature.

Disclosed is a method for managing stoppage of a pressurized-water nuclear reactor integrated into a submerged module for producing electrical power, in case of detection of a primary/secondary leak in a steam generator equipped with a safety valve, which generator is connected to the reactor and associated with a standby cooling unit. The method includes: detecting a primary/secondary leak in the steam generator; automatically stopping the reactor and isolating the broken steam generator; bringing the corresponding standby cooling unit online; monitoring the primary pressure and, once the primary pressure has passed below the set pressure of the safety valves of the steam generators, isolating the standby cooling unit of the broken steam generator; and continuing to passively cool the reactor with the remaining steam generators and cooling unit.