Abstract: Nuclear reactors produce large amounts of spent fuel during operation. In addition to recyclable materials such as uranium and plutonium, spent fuel also contains significant amounts of fission products and highly radioactive transuranic (TRU) elements. Homogenization of nuclides in traditional post-processing technology hinders efficient fuel recycling. In order to improve the efficiency of spent fuel recycling, and to reduce the inevitable “highly radioactive waste” produced in existing spent fuel recycling processes, a novel method for pre-separating nuclides in spent fuel is disclosed. Specifically, we have developed a physical method, wherein an artificially created large temperature gradient drives the migration of fission gas bubbles in spent metallic nuclear reactor fuel. The fission gas bubbles preferentially carry fission products and transuranic elements to achieve effective pre-separation of these elements from spent fuel, lowering cost and improving efficiency of spent fuel recycling.

Abstract: A calibration method, a calibration device and a calibration system of a laser positioning system are provided. The calibration method includes: arranging calibration devices in a first chamber and a second chamber, respectively, each calibration device including a calibration die, and the relative position between the calibration die of the first chamber and a predetermined center point of the first chamber being the same as the relative position between the calibration die of the second chamber and a predetermined center point of the second chamber; and comparing the position of a first positioning mark on the calibration die of the first chamber and the position of a second positioning mark on the calibration die of the second chamber. The technical solution provided can avoid deviation of an irradiation position to a patient from a correct treatment position, and reduce or avoid unnecessary radiation dose to the patient.

Abstract: The present disclosure provides a method for pre-separating nuclides in spent fuel, comprising steps: S1. generating a temperature of 1200 to 1600° C. in a center of a fuel element, and at the same time introducing a cooling medium to an outer surface of the fuel element, to introduce a temperature difference to heat the fuel element; S2. soaking cooled spent fuel in an ionic liquid to selectively dissolve fission products and transuranic elements, so that the fission products and the transuranic elements are pre-separated from the spent fuel. The method achieves the effect of pre-separating part of the fission products and transuranic elements by physical methods, and solves the problem that homogenization in traditional post-processing technology causes obstacles to nuclide separation.

Abstract: A neutron capture therapy system, which can reduce secondary radiation and radiation damage caused by the action of recoil neutrons generated in the process of generating neutron beams by a beam transmission portion and a neutron beam generation portion. The neutron capture therapy system comprises an accelerator, the beam transmission portion, and the neutron beam generation portion; the accelerator accelerates charged particles to generate charged particle beams; the beam transmission portion transmits the charged particle beams generated by the accelerator to the neutron beam generation portion; the neutron beam generation portion generates neutron beams for treatment; and the neutron capture therapy system comprises a beam transmission shielding component for the beam transmission portion.

Abstract: The invention belongs to the technical field of nuclear reactor materials design, and discloses a method for improving the withstanding capability of the cladding material in the fast neutron irradiation environment, comprising the following steps: selecting the cladding material with the annular structure and placing it on the outer side of the metallic fuel slug, with leaving a 0.2-0.8 mm gap between the metallic fuel slug and the cladding material; processing the operation in a reactor subsequently, with an annealing process of the fast neutron reactor fuel during the operation of the reactor; improves the withstanding capability of the cladding material in the fast neutron irradiation environment.

Abstract: The invention relates to the technical field of nuclear reactor materials, in particular to a processing method for improving the corrosion resistance of iron and steel materials in lead or lead-bismuth, comprising the following steps: selecting iron and steel materials containing Mn and Cr elements, using high-energy fast neutrons generated by fission as the radiation source, and performing irradiation on the iron and steel material so that Mn and Cr elements diffuse to the surface of the iron and steel material to form a dense oxide film, so as to complete the improvement of the corrosion resistance of the iron and steel material. The invention enhances the formation of the dense-structured oxide layer by irradiation. The oxide layer has good protection and self-healing properties in irradiation environment, and a new solution is proposed for enhancing the corrosion resistance of steel in lead and lead-bismuth coolant fast reactors.

Abstract: The invention relates to a field of nuclear reactor fuel temperature measurement and discloses a method of measuring nuclear reactor fuel temperature, comprising: S1: collect fission gas produced by nuclear reactors through a gas collection device, S2: measure pressure value and temperature value of the fission gas through pressure and temperature sensors, S3: obtain the corresponding fuel temperature by calculating the pressure value and temperature value, and the invention provides a method of measuring nuclear reactor fuel temperature, and it collects the fission gas discharged by fuels through a fission gas collection device, utilizes the sensitive relevance in a specific temperature range between the release amount of metal fuel fission gas and fuel temperature changes, and makes the pressure of metal fuel fission gas correspond to fuel temperature, thus to convert fuel temperature measurement which is difficult to achieve into fission gas pressure measurement which is easy to achieve.

Abstract: The disclosure discloses a high-burnup fast reactor metal fuel, wherein the reactor core is loaded with metal fuel made of natural uranium alloy U-50Zr. The metal fuel manually controls the temperature to realize phase transition, increase burnup, and extend the service life of fuel; increases the fuel burnup to increase uranium utilization and reduce the pressure of disposing nuclear waste; extends the fuel life cycle to reduce nuclear power costs and improve the economy of nuclear energy; effectively carries out the timely release of fission gas and the periodic elimination of fuel defects, thus reducing the fuel-cladding mechanical interaction caused by swelling, and increasing the safety.

Abstract: The invention provides a method of forming a nuclear fuel pellet of a uranium containing fuel alternative to UO2, with the steps of obtaining a fuel form in a powdered state; coating the fuel form in a powdered state with at least one layer of a material; and sintering the powdered fuel form into a fuel pellet. Also provided is a sintered nuclear fuel pellet of a uranium containing fuel alternative to UO2, wherein the pellet is made from particles of fuel, wherein the particles of fuel are particles of a uranium containing moiety, and wherein the fuel particles are coated with at least one layer between about 1 nm to about 4 nm thick of a material using atomic layer deposition, and wherein the at least one layer of the material substantially surrounds each interfacial grain barrier after the powdered fuel form has been sintered.

Abstract: A method of forming a nuclear fuel pellet, with the steps of obtaining a fuel form in a powdered state; coating the fuel form in a powdered state with at least one layer of a material; and sintering the powdered fuel form into a fuel pellet. A sintered nuclear fuel pellet, wherein the pellet is made from a powdered fuel form, wherein the powdered fuel form is coated with at least one layer of a material, and wherein the at least one layer of the material substantially surrounds each interfacial grain barrier after the powdered fuel form has been sintered.