An in-vessel cooling and power generation system according to the present disclosure may include a small scale reactor vessel, a heat exchange section provided inside the reactor vessel, and formed to supply supercritical fluid to receive heat from a reactor coolant system in the reactor vessel, an electric power production section comprising a supercritical turbine formed to produce electric energy using the energy of the supercritical fluid whose temperature has increased while receiving heat from the reactor coolant system, a cooling section configured to exchange heat with the supercritical fluid discharged after driving the supercritical turbine to shrink a volume of the supercritical fluid, wherein the supercritical fluid that has received heat from the reactor coolant system in the heat exchange section is formed to circulate through the electric power production section, and the cooling section.

A method of operating a nuclear reactor is provided. The method includes defining a layer increment of a deposit layer modeling a deposit on a heat transfer surface of the nuclear reactor; periodically updating a thickness of the deposit layer by adding the layer increment to the deposit layer; recalculating properties of the deposit layer after each layer increment is added to the deposit layer; determining a temperature related variable of the heat transfer surface as a function of the recalculated properties of the deposit layer; and altering operation of the nuclear reactor when the temperature related variable of the heat transfer surface reaches a predetermined value. A method of modeling a deposit on a heat transfer surface of a nuclear reactor is also provided.

Methods for nuclear power plant maintenance which reduces outage time while a polar crane is maintained. In some examples, the method includes the steps of placing an auxiliary crane on a set of rails of a polar crane, using the auxiliary crane to perform maintenance work while the polar crane is maintained, and removing the auxiliary crane after maintenance work on the polar crane is completed. In some further examples, the method includes placing a second auxiliary crane on the set of rails. Still in other example, bridge motors are used to turn the rails, allowing the auxiliary crane to reach every area of the plant floor.

The present invention relates to filters used to remove debris from water being sucked into a piping system. It has particular application for use in nuclear power plants, which, after a loss of coolant accident, must pump cooling water back into the reactor core from a collection sump. This water may contain various types of debris that must be removed before the water is sent back into the reactor cooling system. Filtering of the debris is realized with the component known as “strainers”. There are restrictions on the space available for installing strainers. The vaned filtering element, for example a vaned fin, of the present invention is designed to reduce the space required for strainer installation by increasing strainer surface area per unit volume, while maximizing the quantity of debris that can be filtered from the water.

The present invention provides a corrosion-resistant structure for a high-temperature water system comprising: a structural material 1; and a corrosion-resistant film 3 formed from a substance containing at least one of La and Y deposited on a surface in a side that comes in contact with a cooling water 4, of the structural material 1 which constitutes the high-temperature water system that passes a cooling water 4 of high temperature therein. Due to above construction, there can be provided the corrosion-resistant structure and a corrosion-preventing method capable of operating a plant without conducting a water chemistry control of cooling water by injecting chemicals.

A remote integrated monitoring operation system includes: a unit integrated database for sequentially recording a name of each plant unit, a parameter indicating an event that has occurred in the plant unit, a state of the parameter, and warning classification indicated by the parameter and the state; an inter-unit influence degree evaluation database for recording influence of the event on the other plant unit; a restoration response guidance database for defining a response to the event; a per-unit urgency degree determination section for determining a degree of urgency of each plant unit; an inter-unit influence degree determination section for evaluating a degree of influence of the event on the other plant unit; and a priority determination section for determining priorities between the respective plant units from the degree of urgency and the degree of influence.

This invention relates to electric power industry, and can be used in horizontal steam generators for nuclear power plants (NPP) with a water-water energetic reactor (VVER). We claim a steam generator primary circuit coolant header with U-shaped tubes of a horizontal heat-exchange bundle designed as a thick-wall welded vessel with a perforated central cylindrical part designed so as to allow installation and fastening of a U-shaped heat-exchange tube bundle in the same, wherein the tubes are grouped into banks and separated by vertical inter-tubular tunnels, a lower cylindrical part designed so as to allow welded connection with the steam generator vessel connection pipe, and an upper cylindrical part with a conical adapter to the flange connection of the manhole with a lid, wherein primary circuit header outer diameter D.sub.head in the central part is selected based on formula.

The technical result of the invention involves assurance of strength of the header wall bridges between holes for fastening of heat-exchange tubes and leaktightness of heat-exchange tube connections with the header assuming that the outer surface of the perforated header part is more efficiently (fully) used for tubing.

This invention relates to electric power industry, and more particularly to horizontal steam generators for nuclear power plants with a water-cooled water-moderated power reactor (VVER) and to reactor plants with a VVER reactor and a horizontal steam generator. A reactor plant with a VVER reactor and a horizontal seam generator, including a nuclear reactor with four circulation loops, each comprising a steam generator with a horizontal bundle of heat-exchange tubes divided into banks by means of inter-tubular tunnels and connected to primary circuit coolant headers inside a cylindrical pressure vessel with elliptical bottoms, a reactor coolant pump, and a primary circuit coolant main circulation pipeline.

The invention relates to steam generators for VVER nuclear power plants. We claim a steam generator with a horizontal heat-exchange tube bundle comprising a welded cylinder vessel manufactured of steel shells and equipped with at least one feed water supply connection pipe and one steam removal connection pipe, and two elliptical bottoms, vessel internals, inlet and outlet headers connected to the heat-exchange tube bundle forming a heat-exchange surface of the steam generator, wherein inner diameter dvess of the steam generator vessel is selected based on a formula. The steam generator vessel is filled with heat-exchange bundle tubes from the bottom upwards to the height of three quarters of its inner diameter or less, the remaining space in the top part of steam generator vessel is left for steam drying. The technical result is creation of a steam generator with a decreased specific amount of metal per structure wherein the generated steam is dried in one with the heat-exchange surface.

A method of providing an emergency supply of services to a nuclear power plant having a cooling water cycle, the method including: situating a container at a remote location from the power plant, wherein the container has permanently integrated therein: a motor comprising a first shaft and a second shaft; a generator driven by the first shaft; a pump driven by the second shaft; a fuel tank connected to the motor, and supplying fuel to the motor; and a transformer connected to the generator; connecting the pump to the cooling water cycle of the nuclear power plant; using the pump to pump water from an external water feed into the cooling water cycle of the nuclear power plant.