Abstract: An electronic apparatus includes a display, a camera, a communication interface, a memory configured to store information, and a one or more processor configured to, based on the specific object being identified from an image obtained through the camera, identify a preset zone with respect to the identified specific object, based on a user being identified from the image obtained through the camera, identify whether the identified user is located within the preset zone, and based on a signal corresponding to a user interaction being received by the communication interface while the user is located in the preset zone, control the display to display a user interface screen corresponding to a type of the user interaction.

Abstract: Disclosed are a method and a device for recovering uranium (U) using a process for chemically treating washing wastewater of a uranium hexafluoride (UF6) cylinder. The method and the device are provided to separate uranium (U) from the wastewater released during a process of washing the uranium hexafluoride (UF6) cylinder and to release a filtrate that satisfies atomic energy licensing standards and environmental regulation standards using evaporation and condensation. Accordingly, an independent technology and process for treating the wastewater released during the process of washing the uranium hexafluoride (UF6) cylinder are ensured, which provides easier maintenance and greatly reduces costs compared to the purchase and operation of apparatuses manufactured by foreign makers.

Abstract: Disclosed are a method and a device for recovering uranium (U) using a process for chemically treating washing wastewater of a uranium hexafluoride (UF6) cylinder. The method and the device are provided to separate uranium (U) from the wastewater released during a process of washing the uranium hexafluoride (UF6) cylinder and to release a filtrate that satisfies atomic energy licensing standards and environmental regulation standards using evaporation and condensation. Accordingly, an independent technology and process for treating the wastewater released during the process of washing the uranium hexafluoride (UF6) cylinder are ensured, which provides easier maintenance and greatly reduces costs compared to the purchase and operation of apparatuses manufactured by foreign makers.

Abstract: A large-particle-size ammonium uranate hydrate crystal and a method of manufacturing the same, in which the reaction rate is controlled by injecting ammonia gas into a uranyl nitrate aqueous solution, thereby increasing the particle size of the ammonium uranate hydrate crystal, and the average particle size of a final ammonium uranate hydrate crystal is 9.32 to 14.68 ?m, which is 100 times or more that of a conventional crystal, and uranium content is less than 1 ppm. Since this ammonium uranate hydrate crystal has a very large particle size, handling thereof in subsequent filtration, drying and calcination/reduction processes is very easy, and moreover, the design of devices for subsequent processes is simple, and little uranium is contained in the filtrate of the filtration process, thus obviating an additional chemical treatment process for uranium recovery, which can greatly reduce facility investment costs.

Abstract: Disclosed is a method of predicting, calculating, or analyzing the sintered density of uranium oxide (UOx) before uranium oxide is added in the pelletizing process during a process of manufacturing nuclear fuel, the method including measuring the chromaticity of ammonium diuranate using a spectrophotometer. The present invention provides a simple and highly reliable method of predicting the sintered density of uranium oxide (UOx), which overcomes the problem with a conventional technology where the sintered density of uranium oxide (UOx) can be analyzed only in a pellet state and a subsequent treatment process needs to be performed according to the analysis result.

Abstract: Disclosed is a method of predicting, calculating, or analyzing the sintered density of uranium oxide (UOx) before uranium oxide is added in the pelletizing process during a process of manufacturing nuclear fuel, the method including measuring the chromaticity of ammonium diuranate using a spectrophotometer. The present invention provides a simple and highly reliable method of predicting the sintered density of uranium oxide (UOx), which overcomes the problem with a conventional technology where the sintered density of uranium oxide (UOx) can be analyzed only in a pellet state and a subsequent treatment process needs to be performed according to the analysis result.