A hydrogen-free carbon film polymer lubricating material and a preparation method and use thereof are disclosed. In the method, a graphite target is used as the target material, and a magnetron sputtering deposition is performed on a surface of the polymer substrate, thereby physically depositing and forming a hydrogen-free carbon film on the surface of the polymer substrate, thereby obtaining a hydrogen-free carbon film polymer lubricating material.

Provided is a casing deformation amount measuring apparatus for a steam turbine including a casing, a plurality of nuts fixed to an outer surface of the casing, a plurality of bolts individually screwed, at each one side end portion thereof, in the nuts and projecting to an outer side in a diametrical direction of the casing from the outer surface of the casing, and a lagging material held by the plurality of bolts and covering the casing. The casing deformation amount measuring apparatus includes a target mounted on another side end portion of one of the bolts and exposed to the outer side in a diametrical direction of the casing with respect to the lagging material, and a distance meter that is disposed in an opposing relation to the target on the outer side in a diametrical direction of the casing and measures a distance to the target.

A dissymmetric RN,N-dialkylamides of formula (I) in which: R.sup.1 represents a linear C.sub.1 to C.sub.4 alkyl, R.sup.2 represents a linear C.sub.1 to C.sub.10 alkyl, and R.sup.3 represents a linear or branched C.sub.6 to C.sub.15 alkyl, where R.sup.3 is different from a n-octyl, n-decyl, n-dodecyl, 2-ethylhexyl and 2-ethyloctyl group when R.sup.1 represents a n-butyl group and R.sup.2 represents an ethyl group. A method for synthesising the N,N-dialkylamides, and uses of same for extracting uranium and/or plutonium from an aqueous acid solution or for fully or partially separating the uranium from the plutonium contained in an aqueous acid solution and a solution resulting from the dissolution of spent nuclear fuel in nitric acid. A method for treating an aqueous solution resulting from the dissolution of spent nuclear fuel in nitric acid, which allows the uranium and the plutonium contained in the solution to be extracted, separated and decontaminated in a single cycle.

The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

The invention is applicable to the corium localizing and cooling systems of a nuclear reactor designed for localization of severe beyond design-basis accidents, in particular, to the devices for directing corium of a nuclear reactor to the corium trap. The technical result of the claimed invention is to increase the efficiency of localization and cooling of the nuclear reactor core melt.

The goal of the invention is to eliminate the guide assembly failure due to the concentration of impact load in the conical part of the guide assembly and, therefore, the instantaneous penetration of the core, fragments of the reactor vessel internals and the reactor vessel head into the core catcher.

According to the claimed invention, the guide assembly of the nuclear reactor corium localizing and cooling system installed under the reactor vessel and supported on the cantilever truss contains a cylindrical part, a conical part with an opening made in it, with their walls covered with heat-resistant and fusible material and divided into sectors by bearing ribs arranged radially relative to the opening, and a bearing frame consisting of the outer upper bearing ring, the outer lower bearing ring, the inner bearing shell, the outer upper bearing shell, the middle bearing shell divided into sectors by bearing ribs, the outer lower bearing shell, support ribs, base, upper inclined plate connecting the conical head, bearing ribs and the middle bearing shell, lower inclined plate connecting the conical head, bearing ribs, the middle bearing shell and the outer upper bearing shell.

3 claims, 4 figures of drawings

The present invention relates to a method for welding a half coil of a reactor, which performs the gas metal arc welding (GMAW) along a half coil temporarily welded to a welding machine body using an automatic welding machine to increase a formation amount of back beads, thereby increasing coupling strength of the half coil, increasing airtightness of a fluid flowing along the half coil, securing uniform welding quality by automatic welding, increasing productivity, and reducing production costs.

The invention is applicable to the corium localizing and cooling systems of a nuclear reactor designed for localization of severe beyond design-basis accidents, in particular, to the devices for directing corium of a nuclear reactor to the corium trap.

The technical result of the claimed invention is to increase the efficiency of localization and cooling of the nuclear reactor core melt.

The goal of the invention is to eliminate the guide assembly failure due to the concentration of impact load in the conical part of the guide assembly and, therefore, the instantaneous penetration of the core, fragments of the reactor vessel internals and the reactor vessel head into the core catcher.

In accordance with the invention, the guide assembly of the corium localizing and cooling system installed under the reactor pressure vessel and resting on the cantilever truss apart from the load-bearing frame contains the thermal elements that in the aggregate allows providing guaranteed entry of core, debris of the internals and the head of the reactor pressure vessel into the corium trap by excluding melt-through of the walls of conical and cylindrical parts and by redistributing the corium jet streams.

A liquid metal cooled molten salt reactor having a liquid metal vessel connected to a gas chamber that is connected to a molten salt chamber that is connected with a hot liquid metal vessel. A fuel salt that is withdrawn from the fuel salt tank through a feeding tube into the molten salt chamber from which the fuel salt is withdrawn into a salt separator. A purging gas is inserted into the gas chamber and withdrawn. A liquid metal coolant is dispensed from the liquid metal vessel through a plurality of dispensing nozzles into the molten salt chamber. The liquid metal coolant flows through the molten salt into a hot liquid metal vessel and then through a liquid metal filter into a liquid metal pump. The liquid metal coolant flows through a thermal exchanger subsequently returning to the liquid metal vessel.

A method of obtaining transuranic elements for nanofuel including: receiving spent nuclear fuel (SNF); separating elements from SNF, including a stream of elements with Z>92, fissile fuel, passive agent, fertile fuel, or fission products; and providing elements. A method of using transuranic elements to create nanofuel, including: receiving, converting, and mixing the transuranic elements with a moderator to obtain nanofuel.

A multi-core thermocouple includes a plurality of wires, an insulation core surrounding the plurality of wires, a sheath surrounding the insulation core, and a plurality of electrical junctions. The plurality of electrical junctions may include a first electrical junction formed between a first wire of the plurality of wires and the sheath at a first axial mid-section of the multi-core thermocouple, the first electrical junction including a first swaged axial mid-section of the sheath and a second electrical junction formed between a second wire of the plurality of wires and the sheath at a second, different axial mid-section of the multi-core thermocouple, the second electrical junction including a second swaged axial mid-section of the sheath.