A reactor water radioactivity concentration of a nuclear power plant can be predicted with high accuracy. First, a plant state quantity prediction value is calculated by using a physical model that describes plant state quantities of the power plant including a flow rate of feedwater and a metal corrosion product concentration in feedwater of the reactor water is calculated. Next, data for supervised learning is created, and the data for supervised learning includes the previously calculated plant state quantity prediction value and a plant state quantity such as the flow rate of feedwater, the metal corrosion product concentration in feedwater, a metal corrosion product concentration in reactor water, and a radioactive metal corrosion concentration of the reactor water in the reactor as input data and includes a radioactive metal corrosion concentration in the reactor water which is an actual measured value as output data, and a predictive model is trained.

A method for at least partially disrupting or removing radioactive deposits formed on a surface of a structure in a nuclear water reactor is disclosed. The method includes identifying the structure, taking the structure out of operational service, isolating the structure, contacting the surface of the structure with an aqueous solution, and adding an effective amount of an elemental metal in solid form to the aqueous solution. The effective amount includes an amount to at least partially disrupt or remove the radioactive deposits formed on the surface of the structure. The method is conducted at ambient temperature. The radioactive deposits include oxide-containing radionuclides deposited on the surface. The surface is a primary side structure in the nuclear water reactor.

The present invention relates to a method for identifying the unit causing a raw water leak in a condenser of a thermal power plant consisting of n units.

A nuclear power plant comprising a primary coolant circuit, a steam-water circuit separated from the primary coolant circuit and a steam generator connected to the primary coolant circuit and the steam-water circuit to transfer heat from the primary coolant circuit into the steam-water circuit has at least one dosing point in the steam-water circuit to inject a reducing agent into the steam-water circuit, wherein the reducing agent is an organic compound consisting of carbon, hydrogen and oxygen. Furthermore, a method for operating said nuclear power plant and the use of the method for downtime preservation of the secondary side of a steam generator of a nuclear power plant are provided.

A pressurized water reactor includes a primary reactor coolant circuit flown through by a primary reactor coolant during operation, and a chemical and volume control system for the primary reactor coolant. The chemical and volume control system includes, along the direction of flow of the primary reactor coolant, a letdown line, a high-pressure charging pump with a given discharge pressure, and a charging line leading to the primary reactor coolant circuit. The chemical and volume control system further includes a hydrogenation system with a hydrogen supply and a hydrogen feeding line. In order to achieve efficient and fast hydrogen injection into the primary reactor coolant, a high-pressure feeding pump is arranged in the feeding line to provide a gas pressure higher than the discharge pressure of the charging pump. The feeding line discharges into the charging line.

A film-forming apparatus is connected to a carbon steel cleanup system pipe of a BWR plant. Formic acid and hydrogen peroxide are injected into the circulation pipe of the film-forming apparatus. An iron elution accelerator aqueous solution containing 3000 ppm of formic acid and 1500 ppm of hydrogen peroxide is brought into contact with the inner surface of the cleanup system pipe, and Fe2+ is eluted from the cleanup system pipe by formic acid, and hydroxyl radicals generated from hydrogen peroxide. The film-forming aqueous solution produced from the iron elution accelerator aqueous solution by injecting the nickel formate aqueous solution is brought into contact with the inner surface of the cleanup system pipe, and the Ni ions incorporated into the inner surface by the substitution reaction are reduced by the electrons generated at the time of elution of Fe2+ to form a Ni metal film on the inner surface thereof.

A detection apparatus is a flow and temperature measuring device in the form of a tube having a channel positioned parallel to a coolant flow stream and contains a drogue that is firmly attached to a mount within the channel. The mount has a load detection apparatus firmly imbedded near the flow inlet to the tube. The drogue has at least neutral buoyancy in the fluid at the minimum fluid temperature of interest. A change in the buoyancy of the drogue as a function of temperature and a corresponding change in the strain detection apparatus output in the fluid is determined by a combination of simple physics and calibration measurements. The relationship between changes in strain detection apparatus signal output and flow-rate-induced forces applied to the drogue surface are also determined using a combination of simple physics and calibration measurements. A system and method are also described.

Device for producing energy by nuclear fission, and methods of using same. The device comprises a core container of a core container material, which core container encloses an inner tubing of an inner tubing material. The inner tubing and/or the core container has(have) an inlet and an outlet. The device also comprises a molten fuel salt with a fissionable material and a molten moderator salt comprising metal hydroxide(s), metal deuteroxide(s) or a combination thereof and a redox-element having a reduction potential, which is larger than that of the inner tubing material or of the inner tubing material and the core container material. The molten moderator salt is located in the core container, and the molten fuel salt is located in the inner tubing. Alternatively, the molten fuel salt is located in the core container, and the molten moderator salt is located in the inner tubing.

Intended is to improve the corrosion resistance of an overlay used in a nuclear power plant, and to reduce dissolution of cobalt from an overlay. The corrosion and wear resistant overlay 7 is formed along a surface of a base 2 by laser lamination modeling, and is configured from a plurality of metal layers 1a, 1b, 1c, and 1d of a Co-base alloy. The thickness of carbide eutectics that precipitate in the metal layers 1a, 1b, 1c, and 1d is the largest in the metal layer 1a closest to the base 2, and is gradually smaller in the metal layers 1b, 1c, and 1d farther away from the base 2. The intensity of the laser beam applied to form layers by laser lamination modeling is adjusted so that the carbide eutectics that precipitate in at least the outermost metal layer 1d have a controlled size of 10 μm or less.

The present invention relates to a method for identifying the unit causing a raw water leak in a condenser of a thermal power plant consisting of n units.