6 results about "Nuclear fuel cycle" patented technology

The present invention provides an operational plan generation device, etc., that generates operational plans that take into account the uncertainty of future conditions regarding nuclear fuel cycle facilities. [Solution] The operation plan generation device 10 includes a constraint acquisition unit 121 that acquires combinations of multiple patterns of constraints as operational constraints for the nuclear fuel cycle facility, an operation plan search unit 122 that searches for an operation plan for the nuclear fuel cycle facility toward the optimization of a predetermined objective function under the constraints of each of the aforementioned combinations, and an output control unit 126 that outputs the search results for the operation plan for each of the aforementioned combinations.

The application belongs to the technical field of nuclear fuel cycle and nuclear waste liquid treatment. The application reports a kind of carboxyl-substituted o-phenanthroline water-soluble ligand based on selective masking of part of trivalent actinide under high acidity conditions, and under optimal conditions, the separation coefficient of trivalent europium can reach 120 in 1.5M nitric acid; the separation coefficient of trivalent cerium can reach 4.4. The single crystal structure of the complex is a double metal dimer structure, which guarantees the high acidity tolerance and high selectivity of the ligand. The synthesis of the ligand involved in the application is simple, the extraction efficiency is high, and the extraction mode is novel; the related content will greatly promote the design and performance optimization of water-soluble lanthanum, lanthanum, actinium and actinium separation ligand.

This invention provides an asymmetric o-phenanthroline amide pyrazole compound, its preparation method, and its applications, belonging to the field of nuclear fuel cycle and high-level radioactive waste treatment technology. The asymmetric o-phenanthroline amide pyrazole compound provided by this invention has the structure shown in Formula I. The asymmetric o-phenanthroline amide pyrazole compound provided by this invention exhibits good performance against Am under high acidity conditions. 3+ The extraction partition ratio is high, and the separation effect is excellent. Under 3 mol / L nitric acid conditions, the asymmetric o-phenanthroline amide pyrazole compound provided by this invention has a high separation efficiency for Am. 3+ and Eu 3+ Extraction partition ratio ( D The separation factors were 3.32 and 0.09, respectively. SF Am / Eu It is 36.9, for Am 3+ It has a good separation effect.

This invention discloses a method for operating a high-burnup spent fuel transfer container with graded pressure monitoring and depressurization, belonging to the field of nuclear fuel cycle technology, and includes three stages: loading, transfer, and unloading. In the loading stage, the container is connected to the loading well through-hole and filled with water to balance the water level before loading fuel. After completion, leakage testing and radiation monitoring are performed. In the transfer stage, a qualified container is transported as a whole to the target facility using a tractor. In the unloading stage, before opening the lid, the pressure in the lid chamber and inner chamber is detected via a quick-connect coupling. If overpressure occurs, the pressure is controlled by connecting to the facility's exhaust gas system. After fuel assembly damage monitoring, the container's inner cavity is cooled and cleaned. Subsequent steps include precise positioning, inner lid disassembly and cleaning for protection, and closed-loop recording. This method solves the problems of low positioning efficiency, high high-altitude risks, high personnel radiation dose, and equipment vulnerability in existing operations. It achieves graded pressure control and systematic pollution prevention and control, improving the safety, accuracy, and efficiency of operations, and has strong compatibility with existing facilities.

This invention discloses a modular, controllable atmosphere high-temperature oxidation device for the recovery of substandard sponge zirconium, relating to the fields of nuclear fuel cycle and materials technology. The device includes: a furnace body system, comprising, from the outside to the inside, a steel outer shell, a refractory insulation layer, and a detachable furnace liner; multiple independently controlled heating zones arranged along the furnace body axis, each equipped with an independent heating element and thermocouple; an atmosphere control system including nitrogen and oxygen sources, an inert gas source, a mass flow controller, and a mixer; a sealing and cooling system employing a high-temperature sealing ring and a double-layer furnace shell design, with circulating cooling water flowing through the interlayer; and an automatic control system equipped with a temperature-atmosphere-time automated control program for fully automated operation and remote monitoring of the oxidation process. The device provided by this invention can achieve efficient and uniform oxidation treatment of irregular porous sponge zirconium within the range of 500-1000℃, transforming it into zirconium oxide with consistent physical properties.

A method for recovering plutonium from destructive analytical waste liquid of complex systems includes the following steps: S1: containing C2O4 2‑ Waste liquid containing SO4 was added to H2O2 solution. 2‑ and PO4 3‑ The waste liquid was treated with Fe(NO3)3 solution; the two solutions were combined and the acidity was adjusted to (7-8) mol / L; S2: Sodium nitrite was added to oxidize plutonium to plutonium(IV), and then plutonium was extracted from the aqueous phase at a certain ratio and acidity, followed by back-extraction at a certain ratio to obtain a plutonium solution; S3: The plutonium solution obtained in S2 was adjusted in valence and acidity, and plutonium was recovered by column separation. This invention can solve the problem of plutonium recovery from destructive analytical waste liquid containing complex anions during the nuclear fuel cycle, effectively reducing the loss of nuclear materials. The collected waste liquid was treated using a waste liquid pretreatment-extraction / back-extraction-column separation process, and the final plutonium recovery rate was greater than 90%.