A thermal control system for a reactor pressure vessel comprises a plate having a substantially circular shape that is attached to a wall of the reactor pressure vessel. The plate divides the reactor pressure vessel into an upper reactor pressure vessel region and a lower reactor pressure vessel region. Additionally, the plate is configured to provide a thermal barrier between a pressurized volume located within the upper reactor pressure vessel region and primary coolant located within the lower reactor pressure vessel region. One or more plenums provide a passageway for a plurality of heat transfer tubes to pass through the wall of the reactor pressure vessel. The plurality of heat transfer tubes are connected to the plate.

The present invention relates to an apparatus for efficiently and economically generating a cooling medium by using high-temperature and high-pressure steam generated in a nuclear power plant, and cooling method therefor. According to one embodiment of the present invention, the cooling medium generating apparatus provided in a containment vessel of a nuclear power generation facility so as to generate the cooling medium can comprise: a nuclear reactor for heating a coolant by using heat included in the heated coolant; a cooling module for generating the cooling medium by using the steam generated in the steam generator; and a cooling medium supplying pipe of which the end portion is connected to the outside of the containment vessel so as to supply the cooling medium, having been generated in the cooling module, to the outside of the containment vessel.

The invention relates to the nuclear energy field, including systems for passive heat removal from the pressurized water reactor through the steam generator. The invention increases heat removal efficiency, coolant flow stability and system reliability. The system includes at least one coolant circulation circuit comprising a steam generator and a section heat exchanger above the steam generator in the cooling water supply tank and connected to the steam generator through the inlet and outlet pipelines. The heat exchanger is divided into parallel sections wherein L/D20, L being the half-section length, D being the header bore, and includes an upper and lower header interconnected by heat-exchange tubes, startup valves with different nominal bores are installed on the outlet pipeline. The inlet and outlet pipeline sections of the circulation circuit comprise a set of branched parallel pipelines individually connected to each of the above heat exchanger sections.

A steam generator that reduces the thermal and hydraulic unevenness in the steam generator, improves the filling capacity of the steam generator with heat exchange tubes, organizes an economizer portion of the heat exchange surface in the steam generator, and reduces the concentration of corrosive impurities in the weld zone of the primary circuit to the horizontal shell. To solve the task in such steam generator containing the horizontal shell and other component, the heat exchange tubes are located in vertical planes, and the inlet and outlet manifolds of the primary circuit are arranged horizontally. The steam generator can also be equipped with at least two output manifolds of the primary circuit furthers, the feed water dispenser can be located below the heat exchange tubes of the steam generator.

A thermal control system for a reactor pressure vessel comprises a plate having a substantially circular shape that is attached to a wall of the reactor pressure vessel. The plate divides the reactor pressure vessel into an upper reactor pressure vessel region and a lower reactor pressure vessel region. Additionally, the plate is configured to provide a thermal barrier between a pressurized volume located within the upper reactor pressure vessel region and primary coolant located within the lower reactor pressure vessel region. One or more plenums provide a passageway for a plurality of heat transfer tubes to pass through the wall of the reactor pressure vessel. The plurality of heat transfer tubes are connected to the plate.

The invention relates to improved systems and methods for inspecting the tubes of a steam generator of a nuclear reactor that involves modeling the steam generator, comparing signals of a tube from an eddy current sensor with aspects of the model to determine whether further analysis is required, employing primary and secondary analysis processes, and producing a combined report of the primary and secondary analysis results.

A nuclear fusor apparatus includes magnetically conductive spherical shell, solid iron ball, annular ferromagnetic guide rails, orbiting members, ignition, and gas outlet for transmitting mixed gas of deuterium and tritium, wherein the annular ferromagnetic guide rails are symmetrically fixed on an outer wall of the spherical shell; the orbiting members are located above the annular ferromagnetic guide rails; the orbiting members cooperate with the annular ferromagnetic guide rails to generate magnetic levitation force and propelling force; a water pipe is positioned inside the spherical shell, an end of the water pipe is connected to a water pump, another end of the water pipe is connected to an air extraction unit; the ignition and the gas outlet are both fixed on the outer wall of the spherical shell, an ignition end of the ignition extends into the spherical shell, and an end of the gas outlet extends into the spherical shell.

A steam generator includes a shroud and an annular stepwise helical baffle extending along at least part of a length of the shroud. There is a riser located in a central region of the steam generator. The helical baffle is made up of at least one annuler sector of flat plates. The edges of the flat plates may be straight or corrugated.

The present invention discloses a heat exchange system and a nuclear reactor system. The heat exchange system includes: a heating device; a heat consuming device connected with the heating device through a pipe to form a loop; and a steam, which is in a wet steam state before being supplied to a heat source, and is supplied to the heat consuming device after becoming dry steam or superheated steam by exchanging heat with the heating device. Heat exchange efficiency and security of the nuclear reactor system are improved by adopting steam as a heat exchange medium.

A hybrid geothermal power system is discussed. The system includes a geothermal system including power plant (101) and pumping station (102) and a nuclear plant (103). Pumping station (102) is used to inject fluid from reservoir (104) through an injection well (105) into the bedrock (106) (also referred to as the hot dry rock HDR zone) and extracted via a secondary bore (extraction well) usually coupled to the power plant (101). In the present example however the injection well is linked to the extraction well (107). As fluid is injected into the bedrock a drop in temperature occurs due to heat transfer to the fluid. Nuclear plant (103) is utilized to combat this drop, the plant (103) has the fissionable components (1091, 1092, 1093) of the reactor positioned within bores (1081, 1082, 1083) within the HDR zone.