74 results about "Fission products" patented technology

A nanofuel engine including receiving nanofuel (including moderator, nanoscale molecular dimensions & molecular mixture) internally in an internal combustion engine that releases nuclear energy, is set forth. A nanofuel chemical composition of fissile fuel, passive agent, and moderator. 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 method of operating a nanofuel engine loaded with nanofuel in spark or compression ignition mode. A method of cycling a nanofuel engine, including compressing nanofuel; igniting nanofuel; capturing energy released in nanofuel, which is also the working fluid; and using the working fluid to perform mechanical work or generate heat.

The invention discloses a rapid separation method of activated product gallium in a fission product. The rapid separation method comprises the following steps of: preparation of a gallium-containing radioactive solution, preparation of an eluting solution, filling of a chromatographic separation column, control of flow speed, separation of chromatographic, and the like. The rapid separation method comprises the following specific steps of: carrying out adsorption separation by enabling a radioactive solution which is enriched with the fission product and the activated product gallium to pass through a P2O4 chromatographic column, an Al2O3 chromatographic column and a TBP (Tri-Butyl-Phosphate) extracting chromatographic column at a certain flow speed; washing, desorbing the gallium in the TBP extracting chromatographic column by utilizing a low-acid solution; and further purifying an effluent through an Al2O3 and active carbon mixed column and collecting the purified effluent in a sample bottle to obtain a radioactive measuring solution of the activated product gallium. With the adoption of the rapid separation method disclosed by the invention, the radioactive solution enriched with the fission product and the radioactive solution enriched with the activated product gallium are separated; the recycling rate of the gallium is 80-90%; and the decontamination factor of the fission product is superior to 104 and the separation flow can be finished within 2 hours.

The invention belongs to the technical field of nuclear material extraction and recovery, and relates to a method for extracting and recovering plutonium from a plutonium-containing nitric acid solution. The method comprises the following steps: oxidizing plutonium in the spent fuel reprocessing plutonium-containing nitric acid solution to tetravalent with an oxidant, adding an organic solvent containing DMHMP, Di(1-methyl heptyl)methyl phosphonate for extraction under strong acidic conditions, collecting organic phase, adding a reducing agent, performing back extraction under dilute acidic conditions to selectively reduce the tetravalent plutonium in the organic phase to trivalent, and performing back extraction to form the aqueous phase. The method for extracting and recovering plutoniumfrom the plutonium-containing nitric acid solution can efficiently separate plutonium from other impurities, such as uranium, fission product elements of plutonium, strontium, and cesium, and concentrate plutonium to obtain the plutonium nitrate product solution with standard purity.

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF₆ and PuF₄. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF₆ (gas) that is formed to decompose back to PuF₄ (solid), and b) to maintain stability of the UF₆. Uranium (in the form of gaseous UF₆) is easily extracted and separated from the plutonium (in the form of solid PuF₄). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF₆ (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF₄ or NF₃ as the fluorine sources instead of F₂ or HF.

The invention discloses a fuel element as well as a preparation method and application thereof. The fuel element disclosed by the invention is a sphere which sequentially comprises a fuel layer, a non-fuel layer, and a housing layer which are concentric from inside to outside, wherein the non-fuel layer is made from conventional base materials in the field. The fuel element has the advantages of being small in density, having the non-fuel layer which can alleviate swell generated in the service process of the fuel layer, being not liable to rupture, high in power density, fast in heat dissipation and the like; and besides, the housing layer is compact in structure, high in compressive strength, and capable of preventing fused salt from infiltrating and fission products from releasing, and can be used for reactors. The preparation method is simple and low in cost.

The invention relates to a concentric spherical surface separation plate type spherical main container, and belongs to the technical field of molten salt reactors. With the molten salt depleted uranium reactor formed from the concentric spherical surface separation plate type spherical main container, the difficult problem of the purification of the fission product inside the spherical main container can be solved, the fission product purification system is not required, the burning mode that the direct input of the depleted uranium and the direct output of the spent fuel are performed on the reactor is achieved, the low-cost long-term safe and stable operating is achieved, and all the excellent performances of the molten salt depleted uranium reactor can be maintained. According to the present invention, the space is divided into the thin sheet by using the multi-layer concentric spherical surface thin separation plate having the material inlet and the material outlet, or the space is divided into the tubular channels by additionally using the radial thin separation plate, valves and the like are arranged on the material inlet, the material outlet and the channel to control the molten salt flowing mode, the internal separation plate uses the modes such as module design, non-welded stacking installation of various modules, and whole replacement of the module, the top portion is provided with the openable cover plate system, and the innermost layer, the outermost layer and the required middle layer are provided with the corresponding channels connected to the corresponding control device and other devices outside the reactor core; and the concentric spherical surface separation plate type spherical main container is mainly used as the neutron source and the small energy source.

A method for making an improved nuclear fuel cladding tube includes reinforcing a Zr alloy tube by first winding or braiding ceramic yarn directly around the tube to form a ceramic covering, then physically bonding the ceramic covering to the tube by applying a first coating selected from the group consisting of Nb, Nb alloy, Nb oxide, Cr, Cr oxide, Cr alloy, or combinations thereof, by one of a thermal deposition process or a physical deposition process to provide structural support member for the Zr tube, and optionally applying a second coating and optionally applying a third coating by one of a thermal deposition process or a physical deposition process. If the tube softens at 800° C.-1000° C., the structural support tube will reinforce the Zr alloy tube against ballooning and bursting, thereby preventing the release of fission products to the reactor coolant.

A nanofuel engine including receiving nanofuel (including moderator, nanoscale molecular dimensions & molecular mixture) internally in an internal combustion engine that releases nuclear energy, is set forth. A nanofuel chemical composition of fissile fuel, passive agent, and moderator. 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 method of operating a nanofuel engine loaded with nanofuel in spark or compression ignition mode. A method of cycling a nanofuel engine, including compressing nanofuel; igniting nanofuel; capturing energy released in nanofuel, which is also the working fluid; and using the working fluid to perform mechanical work or generate heat.

A method for detecting a leak in a cladding tube in a nuclear reactor is described. The method is well-suited for use in a reactor having a plurality of cladding tubes housed in a plurality of linearly arranged channels for flowing coolant past the cladding tubes. The method includes monitoring the channels for the occurrence of an increase in radiation above a selected base line indicative of the presence of at least one fission product in the coolant in at least one of the plurality of channels, and monitoring the channels for the occurrence of time dependent changes in the strength of radiation in the coolant above the base line along the length of the at least one of the plurality of channels. The leak location is calculated by triangulating the radiation readings from a fixed linear array of detectors positioned adjacent to the channels to determine the location of the strongest radiation reading and the location along the length of the channel where the increase in radiation occurred.

The invention relates to a method for rapidly calculating the stock of a short-life inert gas fission product reactor core disk, and belongs to the technical field of reactor engineering.The method comprises the following steps that S1, an ignition consumption equation set is simplified by setting a simplification condition, and an analytical solution of the stock of the short-life inert gas fission product disk in different decay chain modes is obtained; s2, calculating reactor characteristic parameters by using an ignition consumption calculation program; s3, obtaining sub-parameters of the reactor through mathematical fitting according to the reactor characteristic parameters; s4, calculating reactor operation parameters according to the actual operation thermal power, the accumulated integral power and the fissible nuclide mass of the reactor; and S5, substituting the calculation results of the step S3 and the step S4 into the corresponding analytical solution in the step S1 to obtain the stock of the short-life inert gas fission product reactor core disk. The method provided by the invention can be used for rapidly calculating the short-life inert gas fission product disk stock of balanced cores or non-balanced cores of various types of nuclear fission reactors, and is accurate enough in engineering application.

The invention relates to a fission ionization chamber, in particular to a sampling fission ionization chamber and a fission total number measuring method based on the same. The problems that when an existing ionization chamber is used for measuring the fission yield, due to the fact that the ionization chamber has self-shielding and is poor in neutron energy spectrum consistency, large measurementuncertainty is introduced, gas fission products and short-life fission products cannot be measured, and the influence on the health of measuring personnel is large are solved. The ionization chamberis characterized by comprising a cathode support, a cathode and an anode support, wherein the cathode support and the cathode are coaxially arranged in sequence from bottom to top, and the anode support is arranged in an inner cavity of the cathode and is coaxial with the cathode. An annular sealing boss and an isolation boss are arranged on the lower bottom surface of the anode support. The sealing boss is positioned on the outer side of the isolation boss. The lower end surface of the sealing boss is hermetically and fixedly connected with the bottom inner surface of the cathode. A gap is formed between the lowermost end of the isolation boss and the inner surface of the cathode bottom. The sampling anode and the annular anode are attached to the lower bottom surface of the anode support, and the sampling anode is located in the isolation boss. The annular anode is located between the isolation boss and the sealing boss.

The invention discloses a mobile integrated two-process gas cooled reactor system and a working method thereof. The system adopts a Brayton-organic Rankine combined cycle, and the whole system comprises a gas cooled reactor part, a main power generation part and an ORC waste heat power generation part; a core part of the gas cooled reactor part is composed of an inner fuel tank and an outer fuel tank which are different in internal coolant working medium pressure and opposite in flowing direction; the inner fuel tank and the outer fuel tank are arranged in a reflecting layer with coolant working medium flowing pore passages; the two axial ends of the reflecting layer are tightly connected with a gas turbine, a gas compressor and a power generation unit which are coaxially designed respectively, and the main power generation part is formed together with a heat regenerator; and the ORC waste heat power generation part consists of two stages of ORC waste heat power generation loops and is coupled with the main power generation part for discharging and recycling waste heat of the reactor core. According to the system, multiple sets of energy utilization equipment, an integrated double-process and fuel tank design are adopted, fission products can be contained, the equipment safety redundancy requirement is met, energy gradient utilization is achieved, and the system has the advantages of being compact, small in size, high in safety and high in energy utilization rate.

The steam generator is one of key equipment in a high-temperature gas cooled reactor nuclear power system. A steam generator heat transfer pipe is an important component of a loop pressure boundary and is an important barrier for preventing radioactive fission products from leaking. A heat transfer pipe is prone to causing serious damage accidents under the action of high temperature, high pressure and high radiation for a long time, and therefore blocking operation on the heat transfer pipe in the early stage of a fault is crucial. The invention provides a high-temperature gas cooled reactor steam generator water supply end laser welding pipe plugging device and method based on a multi-axis robot and a laser welding process aiming at a special structure of a high-temperature gas cooled reactor steam generator water supply end. By means of the device, the water supply end pipe plugging operation precision and the automation degree can be improved, and the manual operation time is shortened. The invention provides an innovative solution for laser welding pipe plugging at the water supply end of the steam generator of the high-temperature gas cooled reactor nuclear power station.

A post-accident fission product removal system may include an air mover, a filter assembly, and/or an ionization chamber. The air mover may be configured to move contaminated air through the filter assembly to produce filtered air. The ionization chamber may be connected to the filter assembly. The ionization chamber may include an anode and a cathode. The ionization chamber may be configured to receive the filtered air from the filter assembly and to ionize and capture radioisotopes from the filtered air to produce clean air.

The invention provides a method and a system for producing Mo-99 by using a liquid molten salt reactor. The method comprises the following steps: providing a liquid molten salt reactor, wherein a plurality of graphite moderation assemblies containing channels are arranged in the reactor core, the channels of the graphite moderation assemblies are filled with molten salt composed of low-enriched uranium and base salt, and Mo-99 is generated by fission in the liquid molten salt reactor. When the liquid molten salt reactor runs, an on-line solid-liquid separation method is adopted to separate indissolvable solid fission products on line, then a cooling method is adopted to reduce the radioactive activity of the indissolvable solid fission products, and finally a chemical separation method is adopted to separate and recover Mo-99 from the indissolvable solid fission products, so that the preparation of Mo-99 is realized. According to the invention, the production efficiency is improved, operation is convenient and fast, the economic cost is low, the fuel demand quantity is low, the radioactive shielding requirement is low, and the current Mo-99 supply demand problem can be effectively solved.

The invention discloses a fuel rod fission product release simulation device and a using method thereof. The simulation device comprises a heating furnace, a crucible and a fuel rod simulator, whereinthe fuel rod simulator is used for simulating a cladding and a pellet of a prototype fuel element; the crucible is used for supporting the fuel rod simulator and providing a gas circulation flow channel at the same time, the crucible can contain liquid melt when the fuel rod simulator is molten, and the melt is prevented from damaging the heating furnace; and the heating furnace is used for heating the fuel rod simulator. The invention provides a fission product release characteristic experimental device which does not contain radioactive isotopes and has a structure similar to that of a realfuel rod in physical and chemical processes for a fission product release characteristic experiment, and the fission product release characteristic experimental device can be used for researching release characteristics of various fission products with different reactor types and different serious burnup accidents.

The invention provides a metal-coated fuel and a preparation method thereof. The metal-coated fuel sequentially comprises a nuclear fuel core, a loose metal layer and a compact metal layer from insideto outside. The method comprises the following steps: S1, providing a nuclear fuel core, loading the nuclear fuel core into a high-temperature spouted bed, and introducing argon to enable the nuclearfuel core to be in a fluidized state; S2, introducing hydrogen or argon or a mixed gas of hydrogen and argon, and controlling the proportion of the precursor of the loose metal layer in the carrier gas to be 5-10% V/V, so as to coat the loose metal layer on the surface of the nuclear fuel core; S3, controlling the proportion of the precursor of the compact metal layer in the carrier gas to be 0.2-2% V/V so as to further coat the compact metal layer; and S4, stopping introducing the precursor, introducing argon, and cooling the fuel to obtain the lithium ion battery positive electrode material. The metal-coated fuel provided by the invention has the advantages of good thermal conductivity, strong fission product retention capacity, low breakage rate and the like, and can effectively improve the safety and economy of nuclear fuel.