47 results about "Neptunium" patented technology

Methods of producing plutonium-238 are disclosed. One method includes dissolving neptunium-237 in a nitric acid solution to produce a neptunium target solution, subjecting the neptunium target solution to neutron radiation to produce plutonium-238, and removing the plutonium-238 from the neptunium target solution. A second method includes exposing a solution of neptunium-237 to neutron radiation to produce plutonium-238, complexing the plutonium-238 with an organophosphorus compound, and separating the plutonium-238 / organophosphorus compound complex from the solution of neptunium-237. A third method includes dissolving neptunium-237 to form a neptunium-237 target solution, exposing the neptunium-237 to thermal neutrons to produce plutonium-238, utilizing an organophosphorus compound to complex the plutonium-238 and the organophosphorus compound, extracting the plutonium-238 / organophosphorus compound complex from the irradiated neptunium target solution, and recovering the plutonium-238.

The invention concerns a vessel for processing, transfer, accumulation and / or transfer of material containing civilian or military origin plutonium, in the form of plutonium oxide, carbide and / or nitride. The plutonium is preferably present in a concentration not higher than a guaranteed maximum between 20 and 50% wt. in said material, possibly with other actinides such as americium, neptunium, or curium. Said material is preferably in form of powder, granulate and / or tablets. The vessel comprises a volume of 20 to 70 liters for containing said material, demarcated at least by two substantially parallel walls, these two walls being separated by a distance e between 8 and 15 cm. The invention concerns also an enclosure comprising analogous geometrically safe criticality-preventing means (bottom-catcher), a fuel production plant comprising a set of such vessels, preferably installed in a set of said enclosures, as well as a nuclear fuel production process carried out in a set of such vessels, preferably installed in a set of said enclosures.

A method for processing a spent nuclear fuel is disclosed which includes a step for decontaminating uranium(VI) from one or more actinides(IV) and more specially from neptunium and / or plutonium, by complexing this (these) actinide(s)(IV). This method includes a step for decontaminating uranium (VI) from at least one actinide(IV), which decontaminating step comprises at least one operation for stripping the actinide(IV) from an organic phase, not miscible with water, and wherein uranium(VI) and the actinide(IV) are present, by putting the organic phase into contact with an aqueous phase comprising nitric acid and at least one complexing agent which more strongly complexes actinides(IV) than uranium(VI), and then separating the organic phase from the nitric aqueous phase, wherein the at least one complexing agent is a diglycolamide.

The invention belongs to the technical field of analysis of uranium, neptunium and plutonium in radioactive liquid and particularly relates to a preparation method of an organic-phase plutonium and organic-phase neptunium standard solution. The preparation method includes the following steps of preparation of nitric acid-hydrofluoric acid, preparation of an organic-phase reagent, aampling of standard substances, dissolution of a standard plutonium substance sample, plutonium sample evaporation, preparation of an organic-plutonium standard, neptunium standard sampling, neptunium standard sample evaporation, organic-phase addition and preparation of organic-phase neptunium standard. The uncertainty of the organic-phase plutonium standard prepared by adopting the method is 0. 069%, and the uncertainty of the organic-phase neptunium standard is 1.08%. The preparation method can be used for preparation of relevant series standard solutions in a method for direct determination of the low-concentration uranium, neptunium and plutonium in posttreatment samples and thus meets the need for direct determination of organic phases uranium, neptunium and plutonium in posttreatment process feed liquors.

A method for preparing plutonium-238 (Pu-238) includes using a nuclear reactor in irradiating a liquid containing a neptunium-237 (Np-237) based compound with neutrons to convert a portion of the Np-237 based compound to a Np-238 based compound; retaining the liquid containing the resulting mixture for a sufficient amount of time for the Np-238 based compound to at least partially convert to a Pu-238 based compound; and separating the Pu-238 based compound from the neptunium based compounds using fractional distillation or a chemical separation method.

The invention relates to a method and a device for continuously adjusting the neptunium valence state and acidity in a 1CU feed liquid of a Purex process. The method comprises: adjusting the nitric acid concentration of the 1CU feed liquid to be 0.03-0.5mol / L; The neptunium in the liquid is oxidized to a high price; the 1CU feed liquid after the thermal insulation pretreatment is continuously cooled; the cooled 1CU feed liquid is continuously acid-adjusted to become the raw material liquid 2DF of the uranium purification cycle; in the above process, the heating and thermal insulation pretreatment The treatment process is carried out in the heating coil and the cooling process is carried out in the cooling coil to realize continuous production; the holding time of the heating and heat preservation pretreatment process is controlled by the flow rate of 1CU feed liquid, the diameter and length of the heating coil . The device includes a 1CU material and liquid temporary storage tank, a material and liquid transfer pump, a heating coil, a cooling coil and a static mixer. The beneficial effects of the present invention are as follows: the technical solution of the present invention can realize the continuous adjustment of neptunium valence state in 1CU, the process operation and control are very simple, and the tedious operation of the batch process is avoided.

The invention relates to a method and device for continuously adjusting the valence state and acidity of neptunium in a Purex process 1CU feed liquid. The method comprises the steps that the nitric acid concentration of the 1CU feed liquid is adjusted to be 0.03-0.5 mol / L; the 1CU feed liquid subjected to acid adjustment is subjected to heating and heat preservation continuous pretreatment to oxidize neptunium to the high valence; the 1CU feed liquid subjected to heat preservation pretreatment is continuously cooled; the cooled 1CU feed liquid is subjected to continuous acid adjustment to form raw material liquid 2DF for uranium purification circulation; in the process, the heating and heat preservation pretreatment process is carried out in a heating coil, and the cooling process is carried out in a cooling coil, so that continuous production is realized; the heat preservation time of the heating and heat preservation pretreatment process is controlled through the flow of the 1CU feed liquid as well as the diameter and length of the heating coil. The device comprises a 1CU feed liquid temporary storage tank, a feed liquid conveying pump, the heating coil, the cooling coil and a static mixer. The method and the device have the beneficial effects that according to the technical scheme, continuous adjustment of the valence state of neptunium in the 1CU can be achieved, process operation and control are very easy, and tedious operation of a batch process is avoided.

The invention discloses a solution reactor used for the plutonium combustion and the neptunium-237 or the americium-241 transmutation. The solution reactor adopts plutonium or plutonium and americium in the spent fuel of the nuclear power plant as the fuel, and adopts neptunium-237 or americium-241 as the target. By adopting the solution reactor, the minor actinides which have long half life and high toxicity and can not be stored for a long time, such as the neptunium-237 or the americium-241, can be changed into nuclides with short half life and low toxicity through transmutation. Because the solution reactor adopts the spent fuel and can generate electricity at the low temperature, the cost-effectiveness and the security are high; the neutron provided by the plutonium combustion can extract plutonium-238 from the neptunium-237 solution target as the isotope battery, thereby the inexpensive energy is provided; the americium-242m extracted from the americium-241 solution target can be widely applied to the aerospace industry and the medical industry.

The invention discloses a device for adjusting the valence states of neptunium and plutonium in a hot chamber. The device comprises an inner cylinder, a clamping sleeve, a sample introduction system, a material discharge system, an online sample analysis system, a seasoning bubbling system, and a liquid level measurement system. systems, temperature measurement systems, pressure measurement systems. The device provided by the invention is suitable for use in radioactive and highly corrosive places, and can directly measure the light absorption value of the sample in the hot chamber, and use the spectral absorption peaks of different substances to judge the ion valence state and other information in the material liquid.

The invention belongs to the technical field of nuclear fuel reprocessing and discloses a process for recovering and purifying neptunium from waste liquor discharged from 2AW+2DW in Purex flow. The process comprises the primary neptunium purification process and the secondary neptunium purification process; the process comprises the following steps: by taking 30 volume percent of TBP-kerosene as an extraction agent, extracting Np, U and Pu to the organic phase, and allowing the fission fragments to enter the aqueous phase; adding a reducing agent and an extraction agent into the organic phase, reversely extracting the Np to the aqueous phase; further extracting uranium and plutonium, and performing oxalate precipitation, filtering and calcining on the obtained aqueous phase product to obtain the purified neptunium oxide solid. The process has the advantages that the neptunium recovery rate is over 98 percent, and the uranium and plutonium removal coefficient in neptunium is high.

The invention belongs to the technical field of spent fuel reprocessing, and in particular relates to a neptunium dioxide-plutonium dioxide dissolver for a hot cell, which is used for dissolving neptunium targets containing neptunium and plutonium elements in a hot cell, including strong acid resistance A high-temperature dissolving tank (1), the periphery of the dissolving tank (1) is provided with a heating device; the dissolving tank (1) is also pierced with a cleaning pipe (4) for adding nitric acid into the dissolving tank (1), for adding and dissolving The feed pipe (5) of liquid, the bubbling pipe (6) that is used to pass into the disturbing gas in the solution in the dissolution tank (1), is used to discharge the discharge pipe (7) of solution; The integral electrode (14) which is installed inside the dissolving tank (1) is replaced; it also includes a control system for remote operation of the replacement of the integral electrode (14). The invention can be applied in a hot chamber to dissolve a neptunium target; the dissolution reaction can be remotely operated by a manipulator in the hot chamber; the neptunium target can be fully dissolved.

The invention discloses a method for separating neptunium (Np) from a uranium product by a TEVA-UTEVA extraction chromatographic column. The method comprises the following steps: (1) adding an acidic solution and a reducing agent into the uranium product to stabilize the neptunium into Np (IV), separating the neptunium from the uranium by a TEVA levextrel resin, and desorbing the neptunium by the mixed solution of nitric acid and oxalic acid; and (2) allowing the eluent to pass through a UTEVA extraction chromatographic column to further purify the neptunium. According to the method provided by the invention, the two separation processes only need one-time price adjustment and are simple; the method is high in recovery rate, precision and separation efficiency and stable in recovery; and the neptunium is separated from the uranium product with the ratio of the uranium to the neptunium of 106:1 by the TEVA-UTEVA extraction chromatographic column.