The present invention is applied to a power system of a power plant including a three-winding transformer, and relates to a double incoming breaker system, including: a plurality of main circuit breakers respectively connected one by one to the plurality of first non-safety class high voltage buses and the plurality of second non-safety class high voltage buses; a plurality of auxiliary circuit breakers, one of which is connected in series to one of the plurality of main circuit breakers; a first power source measurer installed to correspond to the main circuit breaker and measuring a power source level of a non-safety class high voltage bus applied to the main circuit breaker; a second power source measurer installed to correspond to the auxiliary circuit breaker and measuring a power source level at an installed first point thereof; and a controller that outputs a first open signal to the main circuit breaker when an abnormal situation of the non-safety class high voltage bus is checked through the power source level measured by the first power source measurer, and outputs a second open signal to the auxiliary circuit breaker when it is determined that the main circuit breaker fails through the power source level at the first point measured by the second power source measurer after outputting the first open signal.

Steam generators in power plants exchange energy from a primary medium to a secondary medium for energy extraction. Steam generators include one or more primary conduits and one or more secondary conduits. The conduits do not intermix the mediums and may thus discriminate among different fluid sources and destinations. One conduit may boil feedwater while another reheats steam for use in lower and higher-pressure turbines, respectively. Valves and other selectors divert steam and/or water into the steam generator or to other turbines or the environment for load balancing and other operational characteristics. Conduits circulate around an interior perimeter of the steam generator immersed in the primary medium and may have different cross-sections, radii, and internal structures depending on contained. A water conduit may have less flow area and a tighter coil radius. A steam conduit may include a swirler and rivulet stopper to intermix water in any steam flow.

The fuel cartridge may include a plurality of fuel channels, a first header disposed on a first side of a fuel matrix, a second header disposed on a second side of the fuel matrix opposite to the first side, and a plurality of cooling tubes through which a working fluid flows. Each of the plurality of cooling tubes may pass through each corresponding cooling channel of the plurality of cooling channels, where each of the plurality of cooling tubes has a first end connected to the first header and a second end connected to the second header. The fuel cartridge may include an interior space for sealingly containing the fuel matrix may include a pressure boundary independent from an interior of the plurality of cooling tubes, such that the interior space is not in fluid communication with the plurality of cooling tubes.

A depressurisation valve for a cooling system comprising: a main chamber having a main valve, a pilot line having a secondary valve and a blowdown line; the main valve being located to seal a path of the coolant system of the nuclear reactor. The main chamber is connected to the cooling circuit via the pilot line allowing coolant to enter the main chamber, and the blowdown line allows coolant to escape from the main chamber, the pilot line having a lower fluid resistance than the blowdown line. The pressure of coolant in the main chamber maintains the main valve in a closed position, and under elevated temperature and/or pressure conditions fluid is prevented from entering the main chamber via a closure of the secondary valve on the pilot line and reduce the pressure from the valve, moving it to its open position.

An inadvertent actuation block valve includes inlet and outlet orifices being in selective fluid communication via a chamber. A disc is disposed within the chamber and a bellows is configured to contract at a predetermined pressure differential between reactor fluid entering a reference pressure orifice and control fluid entering the inlet orifice. When the bellows contracts, the disc engages the outlet orifice and isolates fluid communication between the inlet and outlet orifices. The inadvertent actuation block valve prevents inadvertent opening of an emergency core cooling valve when a reactor is at operating pressure that is above the predetermined set pressure range. The inadvertent actuation block valve permits the emergency cooling valves to open and to remain open when reactor pressure is below the predetermined set pressure range. The inadvertent actuation block valve does not impede long term emergency cooling that occurs when the reactor is at low pressure.

A mobile heat pipe cooled fast nuclear reactor may be configured for transportation to remote locations and may be able to provide 0.5 to 2 megawatts of power. The mobile heat pipe cooled fast reactor may contain a plurality of heat pipes that are proximate to a plurality of fuel pins inside the reactor. The plurality of heat pipes may extend out of the reactor. The reactor may be configured to be placed in a standard shipping container, and may further be configured to be contained within a cask and attached to a skid for easier transportation.

In view of above problems, an object of the invention is to provide a primary containment vessel venting system having a structure capable of continuously discharging vapor in a primary containment vessel out of the system and continuously reducing pressure of the primary containment vessel without discharging radioactive noble gases to the outside of the containment vessel and without using an enclosing vessel or a power source. In order to achieve the above object, an nuclear power plant of the invention includes a primary containment vessel which includes a reactor pressure vessel, a radioactive substance separation apparatus which is disposed inside the primary containment vessel and through which the radioactive noble gases do not permeate but vapor permeates, a vent pipe which is connected to the radioactive substance separation apparatus, and an exhaust tower which is connected to the vent pipe and discharges a gas, from which a radioactive substance is removed, to the outside.

A dynamic characteristic analysis method of DET and RELAP5 coupling based on a universal instrumental variable method includes steps of: constructing a DET simulation model of a discrete dynamic event tree and modifying TRIP cards of an input file by adding universal instrumental TRIP variables according to state transition types of DET simulation objects, the universal instrumental TRIP variable being variable type or logical type; setting a simulation time of the RELAP5, controlling a simulation step, and analyzing an output result file of each simulation step of the RELAP5; backtracking the RELAP5 according to state transition types of DET simulation objects. The dynamic characteristic analysis method has advantages of simplifying TRIP setting process and method of DET state transition objects in an input file of the RELAP5 required for the coupling of DET and RELAP5, reducing a modeling complexity and improving a modeling efficiency.

A full-digital control rod position measurement device and a method thereof. The full-digital rod position measurement device transforms the core process of rod position measurement that is normally processed by an analog circuit or analog-to-digital mixed circuit into a digital processing. The full-digital rod position measurement device comprises an excitation power supply, an integrated interface board, and a universal signal processor. The universal signal processor processes output signals of detectors according to a preset numerical processing algorithm and outputs Gray code rod position signals. The full-digital rod position measurement device and method disclosed by the present disclosure may effectively reduce the complexity of the primary excitation circuit and the secondary measurement signal processing circuit of the detectors, and improve the operation reliability and measurement accuracy of the rod position processing equipment.

A method comprising measuring a number of gamma-ray counts in a gamma-ray sensitive detector (60) that is placed outside a biological shield (10) near a primary cooling circuit (30) of a nuclear power plant, and determining the thermal power of the nuclear power plant based on the number of gamma-ray counts measured in the gamma-ray sensitive detector (60).