Processing method in PUREX process

Abstract

The invention relates to the technical field of uranium and plutonium separation, and provides a treatment method in a PUREX process, the treatment method comprises the following steps: adding a reducing agent, a characteristic complexing agent and a feed liquid containing uranium, plutonium and technetium into a plutonium back extractor, the plutonium in the feed liquid is reduced by the reducing agent, and technetium in the feed liquid and the characteristic complexing agent generate a liquid complex so as to jointly generate a plutonium-containing aqueous phase back extraction liquid; the feed liquid subjected to reduction back extraction generates uranium-containing organic phase back extraction liquid. Tc (V) and Tc (IV) are selectively complexed by the characteristic complexing agent, and the catalytic action of technetium can be inhibited or even blocked, so that technetium catalytic cycle failure is promoted, excessive use of a reducing agent can be avoided to a certain extent, the effective separation effect of uranium and plutonium is improved, the influence of technetium on the valence state of neptunium can be avoided to a certain extent, and the catalytic action of technetium can be inhibited or even blocked; the method is of great significance in controlling neptunium and technetium trend, improving uranium and plutonium product quality and reducing radionuclide harm.

Description

Technical Field

[0001] This application relates to the technical field of uranium-plutonium separation, and particularly to a processing method in the PUREX process. Background Art

[0002] This section aims to provide background or context for the implementation of this application. The descriptions herein are not admitted to be prior art merely because they are included in this section.

[0003] Technetium ( 99 Tc) is a long-lived radionuclide. It has a high fission yield in reactors and is prone to migration under geological disposal conditions. Therefore, technetium has attracted much attention in the reprocessing of reactor spent fuel.

[0004] The PUREX process is a process for separating and recovering uranium (U) and plutonium (Pu) from spent fuel through extraction, for example, using tributyl phosphate (TBP) as an extractant. In the PUREX process, technetium mainly dissolves in the spent fuel dissolution solution in the form of TcO4 - . In the co-decontamination separation cycle of the PUREX process, U(IV) (tetravalent uranium) and hydrazine are generally used as reducing agents to reduce Pu(IV) (tetravalent plutonium). However, in the presence of technetium, 5-8 times more U(IV) than the stoichiometric amount of chemical reaction must be added to ensure the effective reduction of Pu(IV). In addition, technetium will also undergo complex redox reactions with neptunium (Np), thereby changing the valence state of neptunium and affecting the fate of neptunium. Summary of the Invention

[0005] In view of this, the embodiments of this application are expected to provide a processing method in the PUREX process that can inhibit or even block the catalytic effect of technetium.

[0006] The embodiments of this application provide a processing method in the PUREX process. The processing method includes:

[0007] Adding a reducing agent, a characteristic complexing agent, and a feed solution to a plutonium stripping column. Among them, the feed solution contains uranium, plutonium, and technetium. The plutonium in the feed solution is reduced by the reducing agent, and the technetium in the feed solution forms a liquid complex with the characteristic complexing agent to jointly generate a plutonium-containing aqueous stripping solution; the feed solution after reduction stripping generates a uranium-containing organic stripping solution.

[0008] In some embodiments, the characteristic complexing agent includes at least one of ammonium thiocyanate, thiourea, and thiourea compounds.

[0009] In some embodiments, the concentration of the characteristic complexing agent is 0.01 mol / L - 0.30 mol / L.

[0010] In some embodiments, the plutonium stripping column has a cylindrical structure with its axis extending in the up-down direction.

[0011] In some embodiments, the plutonium stripping extractor forms a first inlet, a first outlet, a second inlet, a second outlet, and a complexing agent inlet. The height of the first inlet is higher than that of the first outlet. The heights of the second inlet and the second outlet are both lower than that of the first outlet. The height of the complexing agent inlet is lower than that of the first inlet and higher than that of the second inlet.

[0012] The first inlet is used to inject the reducing agent. The first outlet is used to discharge the uranium-containing organic phase stripping extract. The second inlet is used to inject the feed liquid. The second outlet is used to discharge the plutonium-containing aqueous phase stripping extract. The complexing agent inlet is used to inject the characteristic complexing agent.

[0013] In some embodiments, the reducing agent includes a supporting reducing agent and a reducing stripping agent, and the supporting reducing agent and the reducing stripping agent are separately added to the plutonium stripping extractor.

[0014] In some embodiments, the reducing stripping agent includes a first reducing stripping agent with a first concentration and a second reducing stripping agent with a second concentration. The first reducing stripping agent and the second reducing stripping agent are separately added to the plutonium stripping extractor, wherein the first concentration is greater than the second concentration.

[0015] In some embodiments, the treatment method includes:

[0016] The plutonium-containing aqueous phase stripping extract enters the uranium supplementary extraction column.

[0017] A supplementary extraction agent is added to the uranium supplementary extraction column, wherein the supplementary extraction agent extracts uranium from the plutonium-containing aqueous phase stripping extract and generates a supplementary extraction organic phase product, and the plutonium-containing aqueous phase stripping extract after supplementary extraction generates a plutonium primary product.

[0018] In some embodiments, the treatment method includes:

[0019] The supplementary extraction organic phase product is fed into the plutonium stripping extractor.

[0020] In some embodiments, the treatment method includes:

[0021] The uranium-containing organic phase stripping extract enters the uranium stripping extractor.

[0022] A dilute acid stripping agent is added to the uranium stripping extractor, wherein the dilute acid stripping agent strips uranium from the uranium-containing organic phase stripping extract and generates a uranium primary product, and the uranium-containing organic phase stripping extract after stripping generates a waste solvent.

[0023] The processing method provided by the embodiments of the present application has the feature that the characteristic complexing agent selectively complexes Tc(V) and Tc(IV), which can inhibit or even block the catalytic action of technetium, thus causing the failure of the technetium catalytic cycle. This not only can avoid the excessive use of reducing agents to a certain extent and improve the effective separation effect of uranium and plutonium, but also can avoid the influence of technetium on the valence state of neptunium, inhibit or even block the catalytic action of technetium, which is of great significance for controlling the trend of neptunium and technetium, improving the quality of uranium and plutonium products, and reducing the harm of radionuclides. Technetium forms a liquid complex with the characteristic complexing agent and enters the aqueous phase. As a component of the back-extraction solution of the plutonium-containing aqueous phase, the characteristic complexing agent makes technetium present as a liquid complex. Since the liquid complex can enter the next process with the aqueous solution, it basically will not remain in the plutonium back-extractor and the pipeline, which can avoid the problems of cleaning and eliminating solid complexes, and there is no need to add devices such as a technetium washing tank upstream of the plutonium back-extractor, simplifying the process flow. BRIEF DESCRIPTION OF THE DRAWINGS

[0024] Figure 1 FIG. is a schematic structural diagram of a plutonium back-extractor, a uranium supplementary extractor, and a uranium back-extractor in some embodiments of the present application, where solid lines are used to indicate the flow direction of the aqueous phase, and dashed lines are used to indicate the flow direction of the organic phase;

[0025] Figure 2 is a curve showing the change of the concentration value of Np(V) with time;

[0026] Figure 3 is a curve showing the change of the concentration value of Tc(VII) with time. DETAILED DESCRIPTION OF THE EMBODIMENTS

[0027] The following further describes the embodiments of the present application in detail with reference to the drawings and examples. The following examples are used to illustrate the present application, but cannot be used to limit the scope of the present application.

[0028] In the various specific technical features and each embodiment described in the detailed description, without contradiction, they can be combined in any suitable manner. For example, different embodiments can be formed by combining different specific technical features / embodiments. To avoid unnecessary repetition, various possible combination methods of the various specific technical features / embodiments in the present application will not be described separately. In addition, terms such as "first" and "second" are only used for descriptive purposes and cannot be understood as indicating or implying relative importance. The present application will be further described in detail below with reference to the drawings and specific embodiments.

[0029] It should be noted that in the embodiments of the present application, "down" is the direction towards the ground, and "up" is the direction opposite to "down".

[0030] The applicant has found that in the co-decontamination separation cycle of the PUREX process, the key to the catalysis of technetium on hydrazine and U(IV) lies in that after Tc(VII) (septavalent technetium) is reduced to low-valent technetium (Tc(V) (pentavalent technetium) and Tc(IV) (tetravalent technetium)), the generated Tc(IV) can react quickly with Tc(VII) to generate Tc(V) and Tc(VI) (hexavalent technetium), thus forming a catalytic cycle, resulting in excessive consumption of the reducing agent hydrazine (N2H4) and U(IV). In addition, the presence of Tc(V) and Tc(IV) can also catalyze the reduction of Np(V) (pentavalent neptunium) to Np(IV) (tetravalent neptunium), changing the valence state of neptunium and thus affecting the separation of neptunium.

[0031] It has also been found through research that if a technetium washing tank is set before the uranium-plutonium separation unit to strengthen the removal of technetium. Since the distribution coefficient of technetium decreases with the increase of nitric acid concentration in the aqueous phase, high-concentration nitric acid is used in the technetium washing tank to wash the technetium extracted into the organic phase. The technetium washing process flows used in the French UP3 plant, UP2-800 plant and the Liusuo plant in Japan are basically the same, and the decontamination factor of technetium is 2.8 - 3.2, thus reducing the technetium entering the uranium-plutonium separation unit by about two-thirds. The above method requires an additional technetium washing process unit and cannot completely remove technetium, making the reprocessing process complex and increasing the amount of waste liquid. If DTPA (diethylenetriaminepentaacetic acid) and hydrazine are added to the uranium-plutonium separation unit, by using DTPA to complex Tc(IV) to inhibit the catalytic behavior of Tc, and at the same time complexing technetium into the aqueous phase, the separation coefficient SF(Tc / U) in the uranium-plutonium separation unit reaches 350 - 500. However, after adding DTPA, plutonium accumulation will occur periodically in the extractor, and at the same time, the solid compound formed by DTPA and Tc(IV) needs to be removed regularly, which leads to a more complex process flow. The above two methods cannot well inhibit the catalytic effect of technetium and cause problems such as complex process flow and increased radioactive waste.

[0032] The embodiment of the present application provides a processing method in the PUREX process. Please refer to Figure 1 , the processing method includes:

[0033] S10. Add a reducing agent, a characteristic complexing agent and a feed solution to the plutonium stripping extractor. Among them, the feed solution contains uranium, plutonium and technetium. The plutonium in the feed solution is reduced by the reducing agent, and the technetium in the feed solution forms a liquid complex with the characteristic complexing agent to jointly generate a plutonium-containing aqueous stripping extract; the feed solution after reduction stripping generates a uranium-containing organic stripping extract.

[0034] Spent fuel dissolution: Spent fuel refers to used nuclear fuel, usually produced by nuclear reactors in nuclear power plants, which contains a large amount of radioactive elements. Spent fuel dissolution is a step following the decanning of spent fuel, aiming to completely dissolve uranium, plutonium, etc. in the solid fuel core in a solvent and convert uranium, plutonium, and fission products into chemical forms conducive to separation, so as to separate the solid fuel core and the cladding as completely as possible.

[0035] The dissolution mechanism of the solid fuel core can vary depending on the different solid fuel cores such as metallic uranium, uranium oxide, uranium-plutonium mixed oxide, etc. The solid fuel core and the dissolving agent undergo a chemical reaction to obtain a spent fuel dissolution solution including uranium, plutonium, technetium, etc.

[0036] Uranium-plutonium co-extraction: The 1AF aqueous feed solution prepared from spent fuel through steps such as decanning, core dissolution, filtration, and adjustment is introduced into the 1A extraction column (co-decontamination column). Under appropriate process conditions, uranium, plutonium, neptunium, and technetium are extracted into the TBP organic phase. Among them, uranium, plutonium, and neptunium exist in the forms of UO2(TBP)2(NO3)2, Pu(NO3)4(TBP)2, and NpO2(TBP)2(NO3)2 respectively, while technetium forms a neutral molecule with metal ions in the form of TcO4 - and is extracted by TBP. The liquid stream 1AP loaded with uranium, plutonium, neptunium, and technetium enters the 1B column, such as the plutonium stripping column 1BX100, while most of the fission products with very small distribution coefficients remain in the aqueous extraction solution 1AW, thus realizing the separation of uranium-plutonium and most of the fission products.

[0037] Therefore, the feed solution HSP can be understood as the TBP organic phase loaded with uranium and plutonium flowing out from 1AP, which also contains neptunium and technetium. It can be understood that the feed solution HSP can also be other organic phase solutions containing uranium, plutonium, and technetium.

[0038] Plutonium Uranium Recovery by Extraction, abbreviated as PUREX, and the PUREX process can also be called the Purex process, mainly including a co-decontamination separation cycle, a uranium purification cycle, and a plutonium purification cycle. The co-decontamination separation cycle mainly produces the uranium primary product 1CUF and the plutonium primary product 1BSP.

[0039] The treatment method provided in the embodiments of the present application is used for the co-decontamination separation cycle. The feed solution HSP is in the organic phase, and the reducing agent is in the aqueous phase. The main chemical reactions occurring in the plutonium stripping column 1BX100 at least include:

[0040] The tetravalent plutonium in the feed solution HSP is reduced to trivalent plutonium by the reducing agent and enters the aqueous phase. The tetravalent and pentavalent technetium in the feed solution HSP form a liquid complex with the characteristic complexing agent and enter the aqueous phase. Thus, they jointly constitute an aqueous plutonium stripping solution.

[0041] That is to say, plutonium undergoes an oxidation-reduction reaction with a reducing agent, and tetravalent technetium and pentavalent technetium undergo complexation reactions with characteristic complexing agents.

[0042] The uranium in the feed solution HSP does not react with the reducing agent or the characteristic complexing agent. After reduction and back-extraction, at least part of the plutonium and technetium in the feed solution HSP are removed, and an organic phase back-extraction solution 1BXU containing uranium is generated.

[0043] The reaction temperature in the plutonium back-extractor 1BX100 can be room temperature. Preferably, the reaction temperature can be between 25°C and 35°C.

[0044] The basic principle of the reaction in the plutonium back-extractor 1BX100 is that the reducing agent reduces Pu(IV) in the feed solution to Pu(III), which is extremely difficult to be extracted, so that Pu(III) enters the aqueous phase, while the uranium in the feed solution remains in the organic phase, realizing the separation of uranium and plutonium.

[0045] In the treatment method provided by the embodiments of the present application, the characteristic complexing agent selectively complexes Tc(V) and Tc(IV), can inhibit or even block the catalytic effect of technetium, thus causing the technetium catalytic cycle to fail. This not only can avoid the excessive use of the reducing agent to a certain extent and improve the effective separation effect of uranium and plutonium, but also can avoid the influence of technetium on the valence state of neptunium, inhibit or even block the catalytic effect of technetium, which is of great significance for controlling the trend of neptunium and technetium, improving the quality of uranium and plutonium products, and reducing the harm of radionuclides. Technetium forms a liquid complex with the characteristic complexing agent and enters the aqueous phase. As a component of the aqueous-phase plutonium-containing back-extraction solution, the characteristic complexing agent makes technetium in the form of a liquid complex. Since the liquid complex can enter the next process with the aqueous-phase solution, it basically does not remain in the plutonium back-extractor 1BX100 and the pipeline, can avoid the problem of cleaning and removing solid complexes, and does not require adding a technetium washing tank and other devices upstream of the plutonium back-extractor 1BX100, simplifying the process flow.

[0046] In some embodiments, the characteristic complexing agent includes at least one of ammonium thiocyanate (NH4SCN), thiourea, and thiourea compounds.

[0047] In some embodiments, the characteristic complexing agent is ammonium thiocyanate, thiourea, or a thiourea compound.

[0048] In some embodiments, the characteristic complexing agent is two of ammonium thiocyanate, thiourea, and thiourea compounds.

[0049] In some embodiments, the characteristic complexing agent is three of ammonium thiocyanate, thiourea, and thiourea compounds.

[0050] In this embodiment, ammonium thiocyanate and technetium can undergo a complexation reaction to form a liquid complex. Thiourea and technetium can undergo a complexation reaction to form a liquid complex. Thiourea compounds and technetium can undergo a complexation reaction to form a liquid complex. Technetium can form stable liquid complexes with ammonium thiocyanate, thiourea, and thiourea compounds.

[0051] In some embodiments, the concentration of the characteristic complexing agent is 0.01 mol / L - 0.30 mol / L. Exemplarily, the concentration of the characteristic complexing agent is 0.01 mol / L, 0.05 mol / L, 0.10 mol / L, 0.15 mol / L, 0.20 mol / L, 0.22 mol / L, 0.25 mol / L, 0.26 mol / L, 0.28 mol / L, or 0.30 mol / L, etc.

[0052] It should be noted that the unit "mol / L" is moles per liter.

[0053] In some embodiments, adding a reducing agent, a characteristic complexing agent, and a feed solution to a plutonium stripping column includes:

[0054] S11. Adding the reducing agent and the feed solution to the plutonium stripping column;

[0055] S12. Adding the characteristic complexing agent to the plutonium stripping column.

[0056] In the initial liquid addition stage of the plutonium stripping column, step S11 can be implemented first, and then step S12. Of course, due to the continuous recycling of the reprocessing process, during the continuous feeding process of the plutonium stripping column, the reducing agent, the feed solution, and the characteristic complexing agent can be added simultaneously.

[0057] In this embodiment, in the reprocessing process of spent fuel, the entire process is continuous. The spent fuel reprocessing co-decontamination and separation cycle includes 1A (co-decontamination tank), 1B (uranium-plutonium separation tank), and 1C (uranium stripping tank). Figure 1 Mainly shown are the processes of 1B and 1C, where the feed solution HSP is the organic phase liquid flow from the 1A tank. When the feed solution HSP flows into the plutonium stripping column 1BX100, the reducing agent also flows in, thus achieving countercurrent extraction. Technetium mainly exists in the form of pertechnetate ions, i.e., TcO4 - dissolved in the feed solution HSP. The characteristic complexing agent mainly complexes pentavalent technetium Tc(V) and tetravalent technetium Tc(IV), and most of the tetravalent and pentavalent technetium come from the reduction products of heptavalent technetium.

[0058] The reducing agent includes but is not limited to U(IV) (tetravalent uranium) and hydrazine.

[0059] In some embodiments, the reducing agent includes a supporting reducing agent and a reducing stripping agent, and the supporting reducing agent and the reducing stripping agent are added to the plutonium stripping column separately.

[0060] The supporting reducing agent includes hydrazine solution. The reducing stripping agent includes U(IV) and hydrazine solution.

[0061] In this embodiment, the supporting reducing agent 1BXH can be used to remove nitrous acid. The reducing stripping agent 1BXR is used to reduce plutonium, for example, reducing tetravalent plutonium to trivalent plutonium.

[0062] In some embodiments, the reducing stripping agent includes a first reducing stripping agent with a first concentration and a second reducing stripping agent with a second concentration. The first reducing stripping agent and the second reducing stripping agent are separately added to the plutonium stripper, wherein the first concentration is greater than the second concentration.

[0063] As an example, the first reducing stripping agent includes a U(IV) solution with a first concentration, and the first reducing stripping agent includes a U(IV) solution with a second concentration.

[0064] In this embodiment, during the mixing process of the reducing stripping agent 1BXR and the feed solution HSP, the reducing stripping agent 1BXR is divided into a first reducing stripping agent 1BSR2 with a different concentration and a second reducing stripping agent 1BSR1, so that the feed solution HSP at different heights in the up and down directions can contact and react with the reducing stripping agent 1BXR, which is beneficial to the reduction of plutonium.

[0065] In some embodiments, please refer to Figure 1 , the treatment method includes:

[0066] S20. The plutonium-containing aqueous stripping solution enters the uranium supplementary extractor;

[0067] S30. Add a supplementary extractant to the uranium supplementary extractor, wherein the supplementary extractant extracts uranium from the plutonium-containing aqueous stripping solution and generates a supplementary extraction organic phase product, and the plutonium-containing aqueous stripping solution after supplementary extraction generates a plutonium primary product.

[0068] The plutonium-containing aqueous stripping solution produced by the plutonium stripper 1BX100 is sent to the uranium supplementary extractor 1BS200 for further treatment.

[0069] Add the supplementary extractant 1BSS to the uranium supplementary extractor 1BS200. The supplementary extractant 1BSS contacts the plutonium-containing aqueous stripping solution to further separate uranium and plutonium.

[0070] The supplementary extractant 1BSS extracts uranium from the plutonium-containing aqueous stripping solution and generates a supplementary extraction organic phase product 1BSU, that is, uranium is further enriched in the organic phase.

[0071] The plutonium-containing aqueous stripping solution after supplementary extraction generates a plutonium primary product 1BSP, that is, plutonium is enriched in the aqueous phase.

[0072] It is understood that the technetium complex is dissolved in the plutonium-containing aqueous stripping solution. Therefore, the technetium complex can be discharged with the plutonium primary product 1BSP.

[0073] The supplementary extractant 1BSS is used for extracting uranium, and the types of the supplementary extractant 1BSS include but are not limited to tributyl phosphate (TBP).

[0074] In this embodiment, uranium in the plutonium-containing aqueous stripping solution is extracted by using the supplementary extractant 1BSS, so that uranium and plutonium are further separated, the content of plutonium in the aqueous phase is increased, and the obtained plutonium primary product 1BSP can be sent to the plutonium purification cycle for treatment.

[0075] In some embodiments, please refer to Figure 1 , the treatment method includes:

[0076] S40. Feed the supplementary extraction organic phase product into the plutonium stripping column.

[0077] The supplementary extraction organic phase product 1BSU contains uranium. Recycling the supplementary extraction organic phase product 1BSU to the plutonium stripping column 1BX100 can improve the separation and utilization rate of uranium.

[0078] In some embodiments, please refer to Figure 1 , the treatment method includes:

[0079] S50. Feed the uranium-containing organic phase stripping solution into the uranium stripping column.

[0080] S60. Add a dilute acid stripping agent to the uranium stripping column, wherein the dilute acid stripping agent strips uranium in the uranium-containing organic phase stripping solution to generate a uranium primary product, and the uranium-containing organic phase stripping solution after stripping generates a waste solvent.

[0081] The dilute acid stripping agent 1CX strips uranium, enabling uranium to be separated from the uranium-containing organic phase stripping solution 1BXU into the aqueous phase.

[0082] After stripping, the uranium content in the uranium-containing organic phase stripping solution 1BXU decreases, and the organic phase from which uranium is removed is the organic solvent, i.e., the waste solvent 1CW is generated.

[0083] Taking the feed solution HSP as the spent fuel dissolution solution as an example, if tributyl phosphate is used as the organic solvent to dissolve the spent fuel, the main component of the waste solvent 1CW is tributyl phosphate.

[0084] The types of the dilute acid stripping agent 1CX include but are not limited to nitric acid.

[0085] It is understood that the uranium primary product 1CUF may contain pertechnetate ions and the like.

[0086] In this embodiment, the uranium stripping column 1C300 can achieve the stripping of uranium.

[0087] The shape and structure of the plutonium stripping column 1BX100 are not limited. Exemplarily, in some embodiments, please refer to Figure 1 , the plutonium stripping column 1BX100 has a cylindrical structure with its axis extending in the vertical direction.

[0088] In some embodiments, please refer to Figure 1 , the plutonium stripping column 1BX100 forms a first inlet 101, a first outlet 102, a second inlet 103, a second outlet 104, and a complexing agent inlet 105. The height of the first inlet 101 is higher than that of the first outlet 102. The heights of the second inlet 103 and the second outlet 104 are both lower than that of the first outlet 102. The height of the complexing agent inlet 105 is lower than that of the first inlet 101 and higher than that of the second inlet 103.

[0089] The first inlet 101 is used to inject a reducing agent.

[0090] The first outlet 102 is used to discharge the uranium-containing organic phase stripping solution 1BXU.

[0091] The second inlet 103 is used to inject the feed solution HSP.

[0092] The second outlet 104 is used to discharge the plutonium-containing aqueous phase stripping solution.

[0093] The complexing agent inlet 105 is used to inject a characteristic complexing agent.

[0094] In this embodiment, the reducing agent is in an aqueous phase and enters the plutonium stripping column 1BX100 through the first inlet 101. The feed solution HSP is in an organic phase and enters the plutonium stripping column 1BX100 through the second inlet 103. The mass of the reducing agent is greater than that of the feed solution HSP. The reducing agent flows downward, and the feed solution HSP moves upward. The height of the first inlet 101 is higher than that of the second inlet 103, which is conducive to the two moving towards each other and promoting their contact.

[0095] The mass of the plutonium-containing aqueous phase stripping solution is greater than that of the uranium-containing organic phase stripping solution 1BXU. The uranium-containing organic phase stripping solution 1BXU is located above the plutonium-containing aqueous phase stripping solution. The height of the second outlet 104 is lower than that of the first outlet 102, which is conducive to the uranium-containing organic phase stripping solution 1BXU being discharged from the first outlet 102 and the plutonium-containing aqueous phase stripping solution being discharged from the second outlet 104. The complexing agent inlet 105 is located between the first inlet 101 and the second inlet 103, which is conducive to the characteristic complexing agent reacting with tetravalent technetium and pentavalent technetium in the feed solution HSP.

[0096] As an example, please refer to Figure 1, a first inlet 101 and a first outlet 102 are formed at the upper part of the plutonium stripping column 1BX100, and a second inlet 103 and a second outlet 104 are formed at the lower part of the plutonium stripping column 1BX100.

[0097] Furthermore, the first inlet 101 may be formed on the top surface of the plutonium stripping column 1BX100.

[0098] The upper part of the plutonium stripping column 1BX100 refers to the part above the midpoint of the plutonium stripping column 1BX100 in the up-down direction.

[0099] The lower part of the plutonium stripping column 1BX100 refers to the part below the midpoint of the plutonium stripping column 1BX100 in the up-down direction.

[0100] In some embodiments, the reducing agent includes a supporting reducing agent 1BXH and a reducing stripping agent 1BXR. The supporting reducing agent 1BXH and the reducing stripping agent 1BXR may separately enter the plutonium stripping column 1BX100 through the first inlet 101 successively. For example, the reducing stripping agent 1BXR first enters the plutonium stripping column 1BX100 through the first inlet 101, and then the supporting reducing agent 1BXH enters the plutonium stripping column 1BX100 through the first inlet 101. As another example, the supporting reducing agent 1BXH first enters the plutonium stripping column 1BX100 through the first inlet 101, and then the reducing stripping agent 1BXR enters the plutonium stripping column 1BX100 through the first inlet 101.

[0101] In some embodiments, the supporting reducing agent 1BXH and the reducing stripping agent 1BXR may also be first mixed in an intermediate container or pipeline, and then enter the plutonium stripping column 1BX100 through the first inlet 101 after mixing.

[0102] In some embodiments, please refer to Figure 1 , the plutonium stripping column 1BX100 forms a component inlet 106. The height of the component inlet 106 is lower than that of the first inlet 101 and higher than that of the complexing agent inlet 105. The reducing stripping agent 1BXR includes a first reducing stripping agent 1BSR2 and a second reducing stripping agent 1BSR1. The first reducing stripping agent 1BSR2 enters the plutonium stripping column 1BX100 through the component inlet 106, and the second reducing stripping agent 1BSR1 enters the plutonium stripping column 1BX100 through the first inlet 101.

[0103] As an example, the second reducing stripping agent 1BSR1 and the supporting reducing agent 1BXH may separately enter the plutonium stripping column 1BX100 through the first inlet 101 successively, or the second reducing stripping agent 1BSR1 and the supporting reducing agent 1BXH may also be first mixed in an intermediate container or pipeline, and then enter the plutonium stripping column 1BX100 through the first inlet 101 after mixing.

[0104] The shape and structure of the uranium supplementary extractor 1BS200 are not limited. Exemplarily, in some embodiments, refer to Figure 1 , the uranium supplementary extractor 1BS200 has a cylindrical structure with its axis extending in the up-down direction.

[0105] In some embodiments, refer to Figure 1 , the uranium supplementary extractor 1BS200 forms a third inlet 201, a third outlet 202, a fourth inlet 203, and a fourth outlet 204. The heights of both the fourth inlet 203 and the fourth outlet 204 are lower than the height of the third inlet 201, and the height of the third outlet 202 is higher than the height of the third inlet 201.

[0106] The third inlet 201 is used to inject the plutonium-containing aqueous counter-extraction liquid.

[0107] The third outlet 202 is used to discharge the supplementary extraction organic phase product 1BSU.

[0108] The fourth inlet 203 is used to inject the supplementary extractant 1BSS.

[0109] The fourth outlet 204 is used to discharge the plutonium primary product 1BSP.

[0110] In this embodiment, the mass of the plutonium-containing aqueous counter-extraction liquid is greater than the mass of the supplementary extractant 1BSS, and the height of the third inlet 201 is higher than the height of the fourth inlet 203, which is beneficial for the downward movement of the plutonium-containing aqueous counter-extraction liquid and the upward movement of the supplementary extractant 1BSS to increase the contact duration between the supplementary extractant 1BSS and the plutonium-containing aqueous counter-extraction liquid. The supplementary extraction organic phase product 1BSU is in the organic phase and has a relatively light mass, and its position in the uranium supplementary extractor 1BS200 is relatively high. The relatively high height of the third outlet 202 is beneficial for discharging the supplementary extraction organic phase product 1BSU. The plutonium primary product 1BSP is in the aqueous phase and has a relatively heavy mass, and its position in the uranium supplementary extractor 1BS200 is relatively low. The relatively low height of the fourth outlet 204 is beneficial for discharging the plutonium primary product 1BSP.

[0111] As an example, refer to Figure 1 , the third inlet 201 and the third outlet 202 are formed in the upper part of the uranium supplementary extractor 1BS200, and the fourth inlet 203 and the fourth outlet 204 are formed in the lower part of the uranium supplementary extractor 1BS200.

[0112] Furthermore, the third outlet 202 can be formed on the top surface of the uranium supplementary extractor 1BS200.

[0113] The upper part of the uranium supplementary extractor 1BS200 refers to the part of the uranium supplementary extractor 1BS200 above the midpoint in the up-down direction.

[0114] The lower part of the uranium supplementary extractor 1BS200 refers to the part of the uranium supplementary extractor 1BS200 below the midpoint in the up-down direction.

[0115] The shape and structure of the uranium stripping extractor 1C300 are not limited. Exemplarily, in some embodiments, please refer to Figure 1 , the uranium stripping extractor 1C300 has a cylindrical structure with its axis extending in the up-down direction.

[0116] In some embodiments, please refer to Figure 1 , the uranium stripping extractor 1C300 forms a fifth inlet 301, a fifth outlet 302, a sixth inlet 303 and a sixth outlet 304. The heights of both the sixth inlet 303 and the sixth outlet 304 are lower than the height of the fifth outlet 302, and the height of the fifth inlet 301 is higher than the height of the fifth outlet 302.

[0117] The fifth inlet 301 is used to inject the dilute acid stripping agent 1CX.

[0118] The fifth outlet 302 is used to discharge the waste solvent 1CW.

[0119] The sixth inlet 303 is used to inject the uranium-containing organic phase stripping liquid 1BXU.

[0120] The sixth outlet 304 is used to discharge the uranium primary product 1CUF.

[0121] In this embodiment, the mass of the dilute acid stripping agent 1CX is greater than the mass of the uranium-containing organic phase stripping liquid 1BXU, and the height of the fifth inlet 301 is higher than the height of the sixth inlet 303, which is beneficial to the downward movement of the dilute acid stripping agent 1CX and the upward movement of the uranium-containing organic phase stripping liquid 1BXU to increase the contact time between the dilute acid stripping agent 1CX and the uranium-containing organic phase stripping liquid 1BXU. The waste solvent 1CW is in the organic phase and has a relatively light mass, so its position in the uranium stripping extractor 1C300 is relatively high, and the height of the fifth outlet 302 is relatively high, which is beneficial to discharging the waste solvent 1CW. The uranium primary product 1CUF is in the aqueous phase and has a relatively heavy mass, so its position in the uranium stripping extractor 1C300 is relatively low, and the height of the sixth outlet 304 is relatively low, which is beneficial to discharging the uranium primary product 1CUF.

[0122] As an example, please refer to Figure 1 , the fifth inlet 301 and the fifth outlet 302 are formed in the upper part of the uranium stripping extractor 1C300, and the sixth inlet 303 and the sixth outlet 304 are formed in the lower part of the uranium stripping extractor 1C300.

[0123] Furthermore, the fifth inlet 301 can be formed on the top surface of the uranium stripping extractor 1C300. The sixth inlet 303 can be formed on the bottom surface of the uranium stripping extractor 1C300.

[0124] The upper part of the uranium stripping vessel 1C300 refers to the part of the uranium stripping vessel 1C300 above the midpoint in the vertical direction.

[0125] The lower part of the uranium stripping vessel 1C300 refers to the part of the uranium stripping vessel 1C300 below the midpoint in the vertical direction.

[0126] The influence of technetium on U(IV) is verified through Example 1 as follows:

[0127] Example 1

[0128] In this example, the Tc-U(IV)-N2H4-HNO3 system is adopted, and an experimental group and a control group are set up. The online ultraviolet-visible spectroscopy is used to measure the curve of the U(IV) concentration value changing with time in the experimental group and the control group.

[0129] The supporting reducing agent 1BXH includes an N2H4 + HNO3 solution.

[0130] The feed solution HSP includes a Tc(VII) solution.

[0131] The reduction stripping agent 1BXR includes a U(IV) solution.

[0132] The characteristic complexing agent is a 0.30 mol / L NH4SCN solution.

[0133] Experimental group: At room temperature, the N2H4 + HNO3 solution (supporting reducing agent 1BXH), the Tc(VII) solution (feed solution HSP), and the U(IV) solution (reduction stripping agent 1BXR) are preheated respectively, and the N2H4 + HNO3 solution, the Tc(VII) solution, and the U(IV) solution are all added into the reaction vessel. After mixing, the composition of the reaction solution in the reaction vessel is 1.2 mmol / L Tc(VII) + 0.02 mol / L U(IV) + 0.1 mol / L N2H4 + 1.0 mol / L HNO3. The reaction temperature is room temperature, such as 25 °C. Subsequently, an appropriate amount of 0.30 mol / L NH4SCN solution is added to the reaction solution.

[0134] It should be noted that the unit "mmol / L" is millimole per liter.

[0135] Control group: In the control group, the composition of the reaction solution, the preheating process of the reaction solution, and the reaction temperature are all the same as those in the experimental group. The difference between the control group and the experimental group is that the control group does not add a 0.30 mol / L NH4SCN solution.

[0136] The U(IV) concentration in both the experimental group and the control group was measured online using ultraviolet-visible spectroscopy as it changed over time. A control group was also set up in the experiment: the reaction temperature and the composition of the reaction solution were the same as those in the experimental group, but the NH4SCN solution was not added. During the reaction process, the U(IV) concentration was measured online using ultraviolet-visible spectroscopy as it changed over the reaction time. The experimental results of the two groups are shown in Table 1, and Table 1 shows the U(IV) concentration values at different residence times.

[0137] Table 1

[0138]

[0139]

[0140] As can be seen from Table 1, when the residence time was 60 min, only 35.40% of U(IV) in the experimental group was consumed, while 62.68% of U(IV) in the control group was consumed. Therefore, adding NH4SCN can significantly reduce the dosage of the reducing stripping agent 1BXR U(IV).

[0141] It can be verified through Example 1 that by the complexation reaction of the characteristic complexing agent NH4SCN with technetium to form a liquid complex, the catalytic effect of technetium can be inhibited or even blocked, reducing the excessive consumption of the reducing stripping agent 1BXR U(IV).

[0142] It should be noted that the unit "min" is minutes.

[0143] The following verifies the effect of technetium on Np(V) through Example 2, which is specifically described as follows:

[0144] Example 2

[0145] In this example, a Tc-Np-N2H4-HNO3 system was adopted, and an experimental group and a control group were set up. The ultraviolet-visible spectroscopy was used to measure the curve of the Np(V) concentration value changing over time, and a liquid scintillation counter was used to measure the curve of the Tc(VII) concentration value changing over time.

[0146] The supporting reducing agent 1BXH includes an N2H4 + HNO3 solution;

[0147] The feed solution HSP includes Np(V) and Tc(VII) solutions.

[0148] The characteristic complexing agent uses a 0.05 mol / L NH4SCN solution.

[0149] Experimental group: The reaction solution consisted of 1.0 mmol / L Tc(VII) + 0.1 mol / L N2H4 + 1.0 mol / L HNO3 solution, and the reaction temperature was room temperature. At 40 min of the reaction, 0.05 mol / L NH4SCN solution was added, and then Np(V) solution was added to make its concentration 1.0 mmol / L.

[0150] Control group: In the control group, the composition of the reaction solution and the reaction temperature were the same as those in the experimental group. The difference between the control group and the experimental group was that the control group did not add 0.05 mol / L NH4SCN solution.

[0151] The ultraviolet-visible spectrum was used to measure the curve of the concentration value of Np(V) changing with time. The measurement results are shown in Figure 2 , Figure 2 which was the curve of the concentration value of Np(V) changing with time. Curve L1 was the curve of the experimental group, and curve L2 was the curve of the control group.

[0152] A liquid scintillation counter was used to measure the curve of the concentration value of Tc(VII) changing with time. The measurement results are shown in Figure 3 , Figure 3 which was the curve of the concentration value of Tc(VII) changing with time. Curve L3 was the curve of the experimental group, and curve L4 was the curve of the control group.

[0153] Through Figure 2 and Figure 3 it can be seen that after adding NH4SCN, both Tc(V) and Tc(IV) underwent complexation reactions with NH4SCN, resulting in Tc being unable to catalyze the reduction of Np(V). Therefore, the concentrations of Tc(VII) and Np(V) basically did not change with time.

[0154] The following further illustrates the processing method provided in the embodiments of the present application with a specific example:

[0155] A feed solution HSP containing plutonium, uranium, and technetium was prepared. The feed solution HSP included 0.5 g / L of plutonium (Pu), 88 g / L of uranium (U), and technetium (Tc), and the feed solution HSP was fed into the plutonium stripping column 1BX100 at a flow rate of 165 L / h.

[0156] It should be noted that the unit "g / L" is grams per liter, and the unit "L / h" is liters per hour.

[0157] The reducing stripping agent 1BXR included 1.5 mol / L of HNO3, 160 g / L of U(IV), and 0.2 mol / L of N2H4. The reducing stripping agent 1BXR was fed into the plutonium stripping column 1BX100 at a flow rate of 2.2 L / h. N2H4 in the reducing stripping agent 1BXR was beneficial for U(IV) to maintain a stable valence state.

[0158] The supporting reducing agent 1BXH includes H at 0.1 mol / L + and N2H4 at 0.2 mol / L. The supporting reducing agent 1BXH is fed into the plutonium stripping column 1BX100 at a flow rate of 15 L / h.

[0159] Ammonium thiocyanate is fed into the plutonium stripping column 1BX100 at a concentration of 0.01 mol / L to 0.30 mol / L.

[0160] In the plutonium stripping column 1BX100, tetravalent plutonium is reduced to trivalent plutonium and enters the aqueous phase, which contains a technetium complex; uranium remains in the organic phase, achieving the primary separation of uranium and plutonium.

[0161] The plutonium-containing aqueous stripping solution enters the uranium supplementary extraction column 1BS200.

[0162] The supplementary extractant 1BSS includes 30% TBP. The supplementary extractant 1BSS is fed into the uranium supplementary extraction column 1BS200 at a flow rate of 20 L / h.

[0163] The supplementary extraction organic phase product 1BSU is refluxed to the plutonium stripping column 1BX100.

[0164] The plutonium primary product 1BSP includes H at 1.7 mol / L + and U with a content of less than 0.4 g / L. The plutonium primary product 1BSP is discharged from the uranium supplementary extraction column 1BS200 at a flow rate of 20 L / h.

[0165] In the uranium supplementary extraction column 1BS200, TBP contacts the plutonium-containing aqueous stripping solution. Uranium in the plutonium-containing aqueous stripping solution enters TBP and is refluxed to the plutonium stripping column 1BX100. Trivalent plutonium remains in the aqueous phase and is discharged as the plutonium primary product 1BSP.

[0166] The uranium-containing organic stripping solution 1BXU enters the uranium stripping column 1C300;

[0167] The dilute acid stripping agent 1CX includes H at 0.01 mol / L + and the dilute acid stripping agent 1CX is fed into the uranium stripping column 1C300 at a flow rate of 180 L / h;

[0168] The uranium primary product 1CUF includes H at 0.05 mol / L + and U at 81 g / L. The uranium primary product 1CUF is discharged from the uranium stripping column 1C300 at a flow rate of 180 L / h.

[0169] The waste solvent 1CW includes 30% TBP and U with a content of less than 0.1 g / L. The waste solvent 1CW is discharged from the uranium stripping column 1C300 at a flow rate of 180 L / h.

[0170] In this embodiment, the plutonium primary product 1BSP contains a liquid technetium complex, and the uranium content in the uranium primary product 1CUF reaches 81 g / L. It can be seen that the separation of uranium and plutonium is basically achieved, and the dosages of the supporting reducing agent and the reduction stripping agent are not excessive. The characteristic complexing agent selectively complexes technetium, can inhibit or even block the catalytic action of technetium, thus causing the technetium catalytic cycle to fail, reducing the dosage of the reducing agent, and improving the effective separation effect of uranium and plutonium.

[0171] In the description of this specification, the descriptions referring to the terms "one embodiment", "some embodiments" and "exemplary" mean that the specific features, structures, materials or characteristics described in connection with the embodiment or example are included in at least one embodiment or example of the embodiments of the present application. In this specification, the schematic representations of the above terms do not necessarily refer to the same embodiment or example. Moreover, the specific features, structures, materials or characteristics described can be combined in any one or more embodiments or examples in a suitable manner. In addition, without contradiction, those skilled in the art can combine the different embodiments or examples described in this specification and the features of different embodiments or examples.

[0172] The various embodiments / embodiments provided by the present application can be combined with each other without contradiction. The above are only the preferred embodiments of the present application and are not used to limit the present application. For those skilled in the art, the present application can have various changes and modifications. Any modification, equivalent replacement, improvement, etc. made within the spirit and principle of the present application shall be included in the protection scope of the present application.

Claims

1. A processing method in a PUREX process, characterized in that: The processing method comprises: A reducing agent, a characteristic complexing agent and a feed liquid are added into a plutonium stripper, wherein the feed liquid contains uranium, plutonium and technetium, the plutonium in the feed liquid is reduced by the reducing agent, and the technetium in the feed liquid forms a liquid complex with the characteristic complexing agent to jointly form a plutonium-containing aqueous stripping liquid; the feed liquid after reduction stripping forms a uranium-containing organic stripping liquid.

2. The processing method according to claim 1, characterized in that: The characteristic complexing agent includes at least one of ammonium thiocyanate, thiourea and thiourea compounds.

3. The processing method according to claim 1, characterized in that: The concentration of the characteristic complexing agent is 0.01 mol / L-0.30 mol / L.

4. The processing method according to claim 1, characterized in that: The plutonium stripper is in a columnar structure with an axis extending in the up-and-down directions.

5. The processing method according to claim 1, characterized in that: The plutonium stripper forms a first inlet, a first outlet, a second inlet, a second outlet and a complexing agent inlet, wherein the height of the first inlet is higher than the height of the first outlet, the height of the second inlet and the height of the second outlet are both lower than the height of the first outlet, and the height of the complexing agent inlet is lower than the height of the first inlet and higher than the height of the second inlet; The first inlet is used to inject the reducing agent, the first outlet is used to discharge the uranium-containing organic reverse extraction liquid, the second inlet is used to inject the feed liquid, the second outlet is used to discharge the plutonium-containing water reverse extraction liquid, and the complexing agent inlet is used to inject the characteristic complexing agent.

6. The processing method according to claim 1, characterized in that: The reducing agent comprises a supporting reducing agent and a reducing stripping agent, and the supporting reducing agent and the reducing stripping agent are added into the plutonium stripping device separately.

7. The processing method according to claim 6, characterized in that: The reducing stripping agent includes a first reducing stripping agent of a first concentration and a second reducing stripping agent of a second concentration, and the first reducing stripping agent and the second reducing stripping agent are separately added to the plutonium stripping vessel, wherein the first concentration is greater than the second concentration.

8. The processing method according to claim 1, characterized in that: The processing method comprises: The plutonium-containing aqueous reverse extraction solution enters the uranium supplementary extractor; A supplementary extractant is added to the uranium supplementary extractor, wherein the supplementary extractant extracts uranium in the plutonium-containing aqueous reverse extraction liquid and generates a supplementary extraction organic phase product, and the plutonium-containing aqueous reverse extraction liquid after the supplementary extraction generates a plutonium primary product.

9. The processing method according to claim 8, characterized in that: The processing method comprises: The supplementary extraction organic phase product is fed into the plutonium stripper.

10. The processing method according to claim 1, characterized in that: The processing method comprises: The uranium-containing organic back-extraction liquid enters the uranium back-extraction device; A dilute acid stripping agent is added to the uranium stripping device, wherein the dilute acid stripping agent strips uranium in the uranium-containing organic stripping solution and generates a uranium primary product, and the uranium-containing organic stripping solution after stripping generates a contaminated solvent.

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