108 results about "High-level waste" patented technology

A method is disclosed for the temporary or permanent storage of nuclear waste materials comprising the placing of waste materials into one or more repositories or boreholes constructed into an unconventional oil formation. The thermal flux of the waste materials fracture the formation, alters the chemical and / or physical properties of hydrocarbon material within the subterranean formation to allow removal of the altered material. A mixture of hydrocarbons, hydrogen, and / or other formation fluids are produced from the formation. The radioactivity of high-level radioactive waste affords proliferation resistance to plutonium placed in the periphery of the repository or the deepest portion of a borehole.

An apparatus and method for inter-unit transfer of spent nuclear fuel. In one aspect, the invention is a method of transferring high level radioactive waste comprising: a) loading high level radioactive waste into a water-filled cavity of a canister body having an open top end at a first location; b) coupling a lid to the canister body to enclose the open top end; c) removing a volume of water from the cavity so that a water level of the water within the cavity is above a top end of the high level radioactive waste and a space exists between the water level and a bottom surface of the lid; d) hermetically sealing the cavity; and e) transferring the canister to a second location, the water level remaining above the top end of the high level radioactive waste during the transfer.

The invention discloses a method for preparing a radioactive waste solidified body. The method is characterized by comprising the following steps of: finely grinding a raw material which comprises 9.53 to 29.41 weight percent of ZrSiO4, 23.36 to 32.12 weight percent of CaCO3, 31.07 to 38.46 weight percent of TiO2, 0 to 15.66 weight percent of Nd2O3, 0 to 29.09 weight percent of CeO2 and 0 to 3.97 weight percent of Al2O3; drying the raw material; and adding polyvinyl alcohol sol, granulating, screening, molding, discharging the sol, performing vacuum hot-pressing sintering, and thus obtaining the radioactive waste solidified body. By adoption of the method, the raw material is low in cost, a high-purity and high-density perovskite acorite and titanite combined mineral solidified body is prepared at low temperature by using a simple and practical hot-pressing sintering technology, and foundation is laid for engineering application to artificial rock solidification treatment of high-level waste.

The invention belongs to the technical field of nuclear waste treatment, and particularly relates to a system of simulating characteristics of a high-level waste geology treatment room, which aims to provide the system capable of simulating the heat-water-power-chemistry coupling characteristic of the high-level waste deep geology treatment room. The system comprises a water supplying and gas-supplying device (1), a temperature control device (8), a multi-field coupling experience device, a measuring system, a data collecting system (10) and a data processing system (19). The system uses the multi-field coupling experience device as a core, the outermost layer of the system is a barrel-like stainless cavity body (4) with a sealing cover, and the innermost layer of the system is a cylindrical shape for simulating a heater (3) of the high-level waste. The invention provides a bentonite heat-water-power-chemistry coupling experience device under the condition of the multi-field coupling, has the characteristics of high strength, small deformation, corrosion resistance, good sealing property, high measuring precision and the like, and is suitable for a mock-up long experience of the bentonite under the condition of a simulated treatment room.

The invention relates to a method for preparing high radioactivity waste solidification treating base material which comprises the following steps: preparing the material: using nature hyacinth, CaCO3 and TiO2 as raw material and extracting them by weight ration: nature hyacinth: CaCO3:TiO2 =26-32:29-35:35-42; correct grinding: loading each raw material into the correct grinding device, adding the grinding ball to do correct grinding 30-60 minutes; burning: burning at the temperature 1150 deg. to 1350 deg. 20-60 minutes to obtain the product.

A canister apparatus, basket apparatus and combinations thereof for transporting and / or storing high level radioactive waste, such as spent nuclear fuel. The canister apparatus comprises a cavity for receiving the spent nuclear fuel that is surrounded by two independent gas-tight containment boundaries. The structures that form the two independent gas-tight containment boundaries are in substantially continuous surface contact with one another, thereby facilitating sufficient heat removal from the cavity. In another aspect, the invention is a basket apparatus having a plurality of disk-like grates arranged in a stacked and spaced arrangement so that the cells of the disk-like grates are aligned. In still another aspect, the invention can be a basket apparatus having a disk-like grate having a ring-like structure encompassing a gridwork of beams specially arranged to achieve a unique cell configuration.

A hydrogen mixed gas explosion source item estimation method for a high-level waste liquid storage tank is provided. The method includes calculating the hydrogen generation rate of a high-level wasteliquid storage tank; calculating the explosion limit of mixed gas of hydrogen, nitrogen monoxide and nitrogen; calculating the energy for explosion of the high-level waste liquid storage tank; and calculating release environment source items of airborne radioactive materials after an explosion event. The method is suitable for accident consequence assessment of reprocessing plants and protection against accidents of reprocessing plants, provides technical supports for emergency action level preparation of the reprocessing plants and has important economic value and social value.

The present invention discloses a method for performing MA / RE mutual separation to high-level waste, including: adjusting the nitrate concentration of the nitrate solution containing hypo-actinide element MA and rare-earth element RE; (2) the nitrates solution with adjusted nitrate concentration flowing through a chromatographic column filled with adsorbing substance; (3) using the eluent A to elute the chromatographic column to leach out the nitrates of rare-earth element, wherein, the eluate A is an aqueous solution containing NaNO3 and HNO3; (4) leaching the chromatographic column using the de-ionized water to leaching out the nitrate containing hypo-actinide element. The inventions has features of good separation effect, strong selection adsorbability of the adsorbing substance, capability of comparative thoroughly MA / RE mutual separation in the high-level waste, thoroughly security, economy and high efficiency.

The present invention relates to a treatment of high-level waste of radiochemical production containing radionuclides and macro-admixtures including sodium. The method of extraction of radionuclides by processing acidic aqueous waste solutions by extractants containing macrocyclic compounds selected from the group of crown ethers having aromatic fragments containing alkyl and / or hydroxyalkyl substituents of a linear and / or branched structure, and / or cyclohexane fragments containing alkyl and / or hydroxyalkyl substituents of a linear and / or branched structure, and / or fragments of —O—CHR—CH2O—, where R is the normal or branched alkyl or hydroxyalkyl in organic solvents containing polyfluorinated telomeric alcohol 1,1,7-trihydrododecafluoroheptanol-1 having the formula H(CF2CF2)nCH2OH, where n=3, and a mixture of polyoxyethylene glycol ethers of synthetic primary higher aliphatic alcohols of a fraction C12-C14 of a general formula CnH2n+1O(C2H4O)mH, where n=12-14, m=2 is proposed. A possibility of selection of quantitative and qualitative composition of the extractant depending on the waste composition and the object to be achieved of joint or individual extraction of radionuclides results in universal technology. A method of regeneration of extractants containing crown ethers allowing one to carry out the operation of extraction of radionuclides in a commercial scale under countercurrent dynamic conditions, with regeneration of the extractant for reuse is also proposed.

The invention belongs to the field of research of buffer material multi-field coupling performances in high-level waste geological disposal and in particular relates to a sensor arrangement structure and a method in a buffer material multi-field coupling experiment bench. The invention aims to provide necessary conditions for successfully constructing a buffer material thermal-water-force-chemical multi-field coupling large experiment bench and indicating behaviors of the buffer material under the multi-field coupling condition. According to the structure and the method, three different monitoring areas are divided in a compacted bentonite block layer from bottom to top according to the position of a heater in the experiment bench, different monitoring layers are finely divided in each monitoring area, and the monitoring layers are perpendicular to a cross section of a vertical axis of the experiment bench, wherein two monitoring layers are divided in a bottom monitoring area, four monitoring layers are divided in a middle monitoring area, two monitoring layers are divided in an upper monitoring area, a displacement sensor needs to penetrate through multiple monitoring layers, and a soil stress sensor, an interstitial hydraulic pressure sensor, a temperature sensor, a temperature and humidity sensor and an electrochemical corrosion sensor are positioned in the monitoring layers.

The invention belongs to the field of nuclear installation decommissioning and radioactive waste governance, and particularly relates to an evaluation method for high-level waste geological disposal storage yard address appropriateness on the basis of a Delphi method. The method comprises the following steps that: S1: determining the quantitative evaluation index of high-level waste geological disposal storage yard address appropriateness; S2: constructing an index quantification rule and standard; S3: using the Delphi method to determine the weight of the index; and S4: calculating the final value of the high-level waste geological disposal storage yard address appropriateness. By use of the method, the establishment of the evaluation index is taken as a basis, the Delphi method is taken as a means to determine the weight of the index, the construction of the index quantification rule and standard is taken as a core, the problem of the quantitative evaluation of high-level waste geological disposal storage yard address appropriateness is solved, and the method is suitable for the quantitative evaluation of storage yard address appropriateness of a large and complex high-level waste geological disposal storage yard with an ultralong evaluation period.

The invention discloses a method for separating element palladium from high-level waste, which includes the steps: adding concentrated nitric acid into nitrate solution of the high-level waste to adjust the nitric acid concentration to be 2mol / L; and allowing the high-level waste nitrate solution with the nitric acid concentration adjusted to pass through a chromatographic column filled with adsorbent to enable element palladium and sub-actinide elements to be adsorbed by the chromatographic column filled with the adsorbent; using nitric acid aqueous solution of thiourea to elute the chromatographic column with the adsorbed element palladium and sub-actinide elements, and eluting the element palladium out in a nitrate form, wherein the thiourea concentration of the nitric acid aqueous solution of thiourea is 0.2mol / L, and the nitric acid concentration is 0.1mol / L; and after the element palladium is eluted out in a nitrate form, using saltpeter solution with the pH (potential of hydrogen) being 6.5 to elute the chromatographic column with the adsorbed sub-actinide elements and eluting the sub-actinide elements out in a nitrate form. The method is simple, efficient, high in selectivity of the chromatographic column and effective in separation.

An isolated and purified bacterium is provided which was isolated from a high-level radioactive waste site of mixed waste. The isolate has the ability to degrade a wide variety of organic contaminants while demonstrating high tolerance to ionizing radiation. The organism is uniquely suited to bioremediation of a variety or organic contaminants while in the presence of ionizing radiation.

A method for preparing Gd2Zr2O7 nano-powder sol with a sol-gel method relates to a method for preparing nano-powder and comprises preparing, gelling and drying of original rare earth solution and a roasting process of dried gel. ZrOC12.8H2O and Gd2O3 are respectively prepared to be nitrate solution, after even mixing, chelant citric acid is added, and ethylene glycol is also added. On the condition of continuous mixing, the potential of hydrogen (pH) value of the system is adjusted by means of stronger ammonia water, absolute ethyl alcohol is added, and then the solution is heated in water bath to enable the solution to form vitreosol and gel. The dried gel is obtained by drying the gel, and after grinding, the dried gel is roasted for two hours to obtain the Gd2Zr2O7 nano-powder. The method is simple in process, the adopted raw materials are simple to obtain and less in loss, the method has short preparation period, the Gd2Zr2O7 nano-powder is small in grain size and even in distribution, is basically in the spherical shape, and has broad prospects when applied to the thermal barrier coating field and the high-radioactivity waste solidification field.

The invention discloses a glass solidified body and a preparation method thereof, which belong to the field of material preparation of solidified bodies. The glass solidified body comprises components: by mass percentage, 67-74% of Sio2, 13-15% of A12O3, 9-10% of B2O3, and 1-11% of CeO2. The preparation method includes: grinding the components according to proportion and mixing the same uniformly, melting the mixture for 2.5-3.5 hours at the temperature ranging from 1350 DEG C to 1400 DEG C, forming, annealing and cooling finally. The glass solidified body is used for solidification treatment of high-level waste, is fine in chemical stability, evident in solidifying effect and high in leaching rate resistance and has excellent application prospect. The preparation method of the glass solidified body is simple in process, low in melting temperature and easy to control during production.

A module for storing high level radioactive waste includes an outer shell, having a hermetically closed bottom end, and an inner shell forming a cavity and being positioned inside the outer shell to form a space therebetween. At least one divider extends from the top to the bottom of the inner shell to create a plurality of inlet passageways through the space, each inlet passageway connecting to a bottom portion of the cavity. A plurality of inlet ducts each connect at least one of the inlet passageways and ambient atmosphere, and each includes an inlet duct cover affixed atop a surrounding inlet wall, the inlet wall being peripherally perforated. A removable lid is positioned atop the inner shell and has at least one outlet passageway connecting the cavity and the ambient atmosphere, the lid and the top of the inner shell being configured to form a hermetic seal therebetween.

The invention discloses ceramic containing pyrochlore phase zirconic acid gadolinium powder. The ceramic is prepared from the following raw materials in percentage by weight: 10-100% of zirconic acid gadolinium, 0-90% of at least one of gadolinium oxide and zirconium oxide and 0-5% of a sintering additive. The invention further discloses a preparation method of the ceramic. The ceramic containing pyrochlore phase zirconic acid gadolinium powder has excellent properties such as high thermal stability, good sintering density, uniform microstructure and the like, and can be applied to catalysts, solid electrolyte, curing of high-level wastes, high-temperature thermal barrier coatings and the like.

The invention specifically relates to an apparatus and method for testing the swelling pressure and saturated infiltration performance of a buffer material, belonging to the field of experimental study of the swelling and infiltration of buffer materials used for geological disposal of high-level waste. The objective of the invention is to provide the testing method capable of acquiring the swelling and infiltration performance parameters of the buffer material bentonite and providing a parameter basis for barrier design of deep geological disposal repositories and projects of high-level wasteso as to overcome disadvantages in the prior art. The method has the following beneficial effects: (1) the swelling pressure of the bentonite under a constant-volume condition can be measured; (2) the swelling pressure and saturated infiltration performance of a sample under different temperature and exhausting conditions can be tested; and (3) the swelling pressure and infiltration coefficient of a same sample can be carried out at the same time, operation is convenient, measurement is fast, and test results are high in precision.

A method of storing high level radioactive waste, and specifically a method of adjusting or controlling the temperature of ventilation air flowing through a storage cavity of a ventilated system. The method includes positioning a metal canister containing high level radioactive waste in a storage cavity of the ventilated system. The ventilated system includes a cask body, a cask lid, a plurality of inlet ducts, and at least one outlet duct so that ventilation air can flow from atmosphere into the storage cavity where it is heated and then back out to the atmosphere. The method includes progressively reducing a cross-sectional area of one or more of the inlet ducts and / or the outlet duct over time so that a rate at which the ventilation air is heated within he storage cavity is maintained above a predetermined threshold to mitigate the risk of stress corrosion cracking in the metal canister.

The invention relates to a multifunctional absorbing vessel used for absorbing an underwater highly radioactive foreign matter and a fishing technology. According to the fishing technology, an underwater dust collector is used cooperatively with a fishing tool to perform fishing of the highly radioactive foreign matter under a water shielding working condition, and the highly radioactive foreign matter can be preserved in the fishing tool so that analysis and evidence obtaining work can be conveniently carried out after the fishing process. According to the multifunctional absorbing vessel used for absorbing the underwater highly radioactive foreign matter and the fishing technology, a large-scale radioactive foreign matter shielding flask does not need to be prepared, the fished highly radioactive foreign matter can be conveyed to a spentfuel storage pool for temporary storage, and the highly radioactive foreign matter can be taken out from the spentfuel storage pool at any time when needed, so that the highly radioactive foreign matter is subjected to evidence obtaining and analysis. According to the fishing technology, long time limit for a project and a large amount of cost in processing equipment for highly radioactive waste in a later period are saved, and convenient conditions for technological analysis in the later period are provided.

The invention specifically relates to a device and method for measuring the heat conductivity coefficient of a buffer / backfill material, belonging to the field of testing of the thermal conductivity of buffer / backfill materials applied to geological disposal of high-level waste. The objective of the invention is to provide the measuring device and method capable of acquiring the thermal conductivity of the buffer / backfill material, controlling effective transmission of decay heat produced in a high-level waste tank and providing bases for engineering design of deep geological disposal repositories of high-level waste so as to overcome disadvantages in the prior art. The device comprises a gas loading device (1), a testing box (2), a sample holder (3), a screw (4), a sample (5), a test probe (6), a heater (7), a temperature sensor (8), a gas monitoring probe (9), a liquid (10), a valve A (11), a valve B (12) and a valve C (13).

The invention discloses a method for extracting and separating element palladium from high-level waste. The method comprises the following steps of: adding concentrated nitric acid into nitrate solution of the high-level waste, from which minor actinide elements have been separated, to adjust the concentration of nitric acid to be 0.42-5.11 mol / L to obtain solution A; mixing 2,6-bi (5,6-dinonyl-1,2,4-triazine-3-yl) pyridine, R-OH and alkanes of C6-C12 to obtain solution B, wherein R is alkyl groups of C6-C12, and the volume fraction of R-OH is 30-100 percent in terms of the total volume of the R-OH and alkanes of C6-C12; mixing the solution A and the solution B, oscillating and then demixing to obtain an oil phase and a water phase; separating the oil phase from the water phase to obtain palladium nitrate-containing oil phase and the nitrate solution of the high-level waste from which the element palladium is separated respectively. The method for extracting and separating element palladium from high-level waste provided by the invention has the advantages of high selectivity, high separation capacity and suitability for large-scale industrial application.

The invention discloses a preparation method for a zirconolite type manmade rock. The preparation method is characterized by comprising the following steps: preparing materials from CuO powder, CaO powder, Ti powder, TiO2 powder and ZrO2 powder; performing dry-process ball-milling on the raw materials, uniformly mixing the raw materials and enabling the raw materials to pass through a 100-mesh sieve; feeding 10-50g of the mixed materials into a stainless steel mould, and pressing the mixed materials into cylindrical blank; introducing the blank into a self-propagating quasi-isostatic device, igniting the blank through a tungsten filament to react, and applying pressure of 50-200MPa after the combustion reaction is ended, thereby preparing the high-compactness zirconolite type manmade rock. The preparation method disclosed by the invention can be used for synthesizing the zirconolite type manmade rock within several minutes by adopting a self-propagating high-temperature combined quick compacting process, has the advantages of being high in synthesis speed, high in efficiency, high in reaction temperature, high in product purity, low in energy consumption, simple, and the like; the prepared zirconolite type manmade rock is suitable for long-term safety solidification treatment and disposal of actinide in high-level waste.

A module for storing high level radioactive waste includes an outer shell, having a hermetically closed bottom end, and an inner shell forming a cavity and being positioned inside the outer shell to form a space therebetween. At least one divider extends from the top to the bottom of the inner shell to create a plurality of inlet passageways through the space, each inlet passageway connecting to a bottom portion of the cavity. A plurality of inlet ducts each connect at least one of the inlet passageways and ambient atmosphere, and each includes an inlet duct cover affixed atop a surrounding inlet wall, the inlet wall being peripherally perforated. A removable lid is positioned atop the inner shell and has at least one outlet passageway connecting the cavity and the ambient atmosphere, the lid and the top of the inner shell being configured to form a hermetic seal therebetween.

The invention discloses a sorbent for MA / RE mutual separation in high-level waste and a preparation method thereof. The sorbent is prepared from 2, 6-di-(5, 6-di-n-butyl-1, 2, 4-triazine-3-radical)-pyridine loaded on macropore SiO2 of a coated polymer, and the mass ratio of the macropore SiO2 to R-BTP is (1-20):1. The sorbent has good identification capability on MA, high adsorption capacity and high recovering rate; when the sorbent is used in an adsorptive separation process, the sorbent does not need to be diluted or dissolved by with diluent of special performance, and can avoid introducing large amount of toxic solvent substances or generating secondary radioactive waste; and the ion diffusion, adsorption and desorption rate of the sorbent are remarkably improved, and the sorbent can run safely under the condition of high flow rate and is a safe and economic sorbent for efficiently mutually separating MA / RE in the high-level waste.

The invention discloses a retrieving method for low-level waste in a waste pit. The method comprises the steps that waste in a pit opening of the waste pit is gripped to an orthographic projection area under the pit opening of the waste pit through a power manipulator, and then the waste in the orthographic projection area under the pit opening of the waste pit is gripped outside the waste pit and loaded in a waste package box or a transfer container through an electric gripper. According to the retrieving method for the low-level waste in the waste pit, the power manipulator and the electric gripper are combined together according to a certain operation sequence in the operating process, therefore, the technical problem of retrieving the low-level waste in the pit is solved, and imitation and reference value is provided for retirement of other pit type waste libraries in China. Compared with the prior art, the method has the advantages that by adopting a robot gripping mode, the retrieving efficiency of the low-level waste is improved; the exposure dose of operating personnel is decreased, and safety of the operating personnel is guaranteed.

The invention relates to a lead-free low-melting-point glass with an excellent comprehensive performance and an application method thereof. The oxide composition of the glass material is expressed as follows in molar percentage: 10-43% of ZnO, 9-20% of Fe2O3, 35-50% of P2O5, 0-6% of Al2O3, 0-10% of B2O3, 0-10% of SiO2, 0-7% of Bi2O3, 0-20% of Na2O, 0-20% of K2O, 0-15% of Li2O and 0-10% of MO, wherein MO is one or more of MgO, CaO, SrO and BaO, the sum of the molar percentages of ZnO, Fe2O3 and P2O5 is not less than 70%, the mixing sum of the molar percentages of alkali metal oxides, and the mixing sum of the molar percentages of alkali metals and alkaline-earth metals are both not greater than 20%. The lead-free low-melting-point glass prepared by the method disclosed by the invention is excellent in comprehensive performance, low in production cost, environment-friendly, and capable of sealing among glass, ceramics and metals, preparing electronic pastes, safely curing high-level wastes, and the like.

A ceramic waste immobilizing material for the encapsulation of high level radioactive waste (HLW), e.g. resulting from the reprocessing of irradiated nuclear fuel. The ceramic waste immobilising material enables waste ions from at least fission products in irradiated nuclear fuel to be dissolved in substantially solid solution form. The ceramic waste immobilising medium has a matrix comprising phases of hollandite, perovskite and zirconolite in which the waste ions are dissolved. The invention also includes a method of immobilizing HLW from reprocessed nuclear fuel assemblies comprising the steps of mixing a liquor containing the HLW with a precursor material comprising oxides or oxide precursors of at least titanium, calcium and barium to form a slurry, drying the slurry, and calcining the dried slurry under a reducing atmosphere to form a powder comprising 30–65 weight % waste.