Combined cleanup and heat sink systems work with nuclear reactor coolant loops. Combined systems may join hotter and colder sections of the coolant loops in parallel with any steam generator or other extractor and provide optional heat removal between the same. Combined systems also remove impurities or debris from a fluid coolant without significant heat loss from the coolant. A cooler in the combined system may increase in capacity or be augmented in number to move between purifying cooling and major heat removal from the coolant, potentially as an emergency cooler. The cooler may be joined to the hotter and colder sections through valved flow paths depending on desired functionality. Sections of the coolant loops may be fully above the cooler, which may be above the reactor, to drive flow by gravity and enhance isolation of sections of the coolant loop.

Computing devices and methods for reading gauges are disclosed. A gauge reading method includes capturing image data corresponding to a captured image of one or more gauges, detecting one or more gauges in the captured image, cropping a detected gauge in the captured image to provide a use image including the detected gauge, and classifying the detected gauge to correlate the detected gauge with a template image. The gauge reading method also includes attempting to perform feature detection rectification on the use image to produce a rectified image of the detected gauge, performing template matching rectification on the use image to produce the rectified image responsive to a failure to perform the feature detection rectification, and estimating a gauge reading responsive to the rectified image. A computing device may implement at least a portion of a gauge reading method.

Disclosed herein is a method pertaining to a power distribution of a reactor core of a nuclear installation, the method being executed on a general purpose computer. The method comprises: measuring current values from a plurality of vanadium neutron detector assemblies which are disposed in the reactor core of the nuclear installation; determining a measured relative core power distribution based upon the measured current values; producing a measured core power distribution based upon the measured relative core power distribution; and verifying that the reactor is operating within the licensed core operating limits based at least in part upon the measured core power distribution. Also disclosed herein is a vanadium neutron detector assembly.

A combined power generation system and method of a small fluoride-salt-cooled high-temperature reactor and solar tower is provided, which belongs to the field of new energy and renewable energy application and includes: a nuclear reactor power generation system, a solar tower power generation system and a heat compensation system. Both the nuclear reactor power generation system and the solar tower power generation system adopt supercritical carbon dioxide Brayton cycle system to generate electricity efficiently; molten salt pool in the nuclear reactor power generation system stores high-temperature heat from the modular reactor, and multi-stage temperature heat is utilized for generating power and compensating heat required by the solar tower power generation system.

A heat exchanger includes a body having an inlet header through which a fluid is introduced, and an outlet header through which the fluid is discharged; and one or more plates accommodated in the body and provided with flow path modules providing flow paths for the fluid introduced through the inlet header to flow to the outlet header. The heat exchanger further includes at least one flow path adjuster each having at least a portion thereof accommodated in the body and being movable or rotatable to open or close a part or all of the flow paths or to change directions of the flow paths so that a flow of the fluid is adjusted.

Power conversion systems for converting thermal energy from a heat source to electricity are disclosed. In one exemplary embodiment, the power conversion system may include a substantially sealed chamber having an inner shroud having an inlet and an outlet and defining an internal passageway between the inlet and the outlet through which a working fluid passes. The sealed chamber may also include an outer shroud substantially surrounding the inner shroud, such that the working fluid exiting the outlet of the inner shroud returns to the inlet of the inner shroud in a closed-loop via a return passageway formed between an external surface of the inner shroud and an internal surface of the outer shroud. The power conversion system may further include a source heat exchanger disposed in the internal passageway of the inner shroud, the source heat exchanger being configured to at least partially receive a heat transmitting element.

The steam generating unit of dual circuit reactor with blowdown and drain system is implemented in the close loop, without any conventional blowdown expansion tanks and is designed for maximum pressure of the steam generator (SG) working medium. The SG blowdown water is combined into a single line, cooled down in the regenerative heat exchanger, then in the blowdown aftercooler and drain cooling line and taken out of the tight shell. Out of the tight shell, the SG blowdown water is supplied for treatment to the SG blowdown water treatment system designed for maximum pressure of the steam generator (SG) working medium. After treatment, the water returns to the tight shell and, via the regenerative heat exchanger, to the feed pipelines of each SG. The invention provides increased SG blowdown that leads to the accelerated chemical condition normalization even with considerable deviations.

A high-precision high-fidelity real-time simulation and behavior prediction method and device for a nuclear power station is provided. The method comprises the following steps: (1) constructing a nuclear power station simulator and a physical nuclear power station based on the same design parameters; (2) operating the nuclear power station simulator and the physical nuclear power station in parallel, and obtaining predicted parameters output by the nuclear power station simulator and operation parameters of the physical nuclear power station in real time; (3) comparing the predicted parameters and the operation parameters representing the same physical quantity one by one, and correcting prediction models in the nuclear power station simulator and input parameters of the prediction models by adopting a large-scale concurrent-parallel parameter search and correction algorithm and an artificial intelligence-based mode recognition and correction algorithm until the predicted parameters reach specified precision; and (4) operating the nuclear power station simulator according to a set operation condition to obtain the predicted parameters, thereby completing a behavior prediction of a physical nuclear power station system.

The invention relates in particular to a solid metallic component. This component (1) is particularly notable in that it comprises a core (5) and an external shell (3) which surrounds said core (5) in all directions, this core (5) and this shell (3) being made of different grades of steel, the steel of said core (5) having martensite and bainite critical cooling rates lower than those of the steel or steels of said shell (3).

A reactor pressure vessel includes a reactor pressure vessel body, a nozzle structure connected to the reactor pressure vessel body, a conduit structure connected to the nozzle structure, and a restraint device attached around a portion of the conduit structure. The restraint device includes collar parts that have cross sections corresponding to respective segments of a periphery of the portion of the conduit structure, brackets attached to the nozzle structure, and rods connecting the brackets to the collar parts. The collar parts are connected end-to-end to each other such that a cross section of the collar parts connected to each other corresponds to the periphery of the portion of the conduit structure. The collar parts are pinned to each other. The brackets spaced apart from each other around a periphery of the nozzle structure.