A method for recovering uranium and plutonium from remix spent fuel without separation

Abstract

The present application relates to a kind of from REMIX spent fuel without separation recovery uranium, plutonium, belong to spent fuel reprocessing technical field, the method comprises the following steps: S1, the REMIX spent fuel assembly structure of to be handled is disassembled and sheared, and REMIX spent fuel pellet is obtained;S2, the spent fuel pellet is dissolved, and feed liquid is obtained;S3, the element of splitting in the feed liquid is decontaminated using co-extraction, co-reverse extraction circulation process, so that splitting purification meets product requirement;S4, after the feed liquid treated in step S3 is treated using tail end co-precipitation calcination process, and uranium-plutonium co-precipitation is obtained, and uranium-plutonium mixed oxide is obtained.The method provided by the present application can meet the requirement of 99.9% of uranium-plutonium yield under the condition of uranium-plutonium not separation.

Description

TECHNICAL FIELD

[0001] The application belongs to the technical field of spent fuel reprocessing, and particularly relates to a method for recovering uranium and plutonium without separation from REMIX spent fuel. BACKGROUND

[0002] Spent fuel reprocessing is a key link of a closed nuclear fuel cycle. A current commercial spent fuel reprocessing process is a PUREX process. The process of the PUREX process is a uranium-plutonium separation process, which is divided into element disintegration shearing, spent fuel dissolution, one-cycle co-extraction (1A), technetium washing (TcS), one-cycle oxidation-reduction stripping (1B), one-cycle uranium stripping (1C), uranium wire and plutonium wire. The process can scaleably process spent fuel, recover useful uranium, plutonium and other transuranium elements therein, and improve the utilization rate of nuclear materials. It is a reprocessing process commonly used in current reprocessing plants of various countries. However, a large amount of waste liquid, waste and waste gas is generated in the process, and the workload of waste treatment and geological disposal is large. Uranium-plutonium separation, uranium and plutonium product recovery and remaining radioactive waste solidification can easily cause nuclear material diffusion. In order to simplify the spent fuel reprocessing process and reduce the cost of spent fuel reprocessing and the amount of spent fuel storage, a new concept of REMIX technology (REgenerated MIXture of U and Pu oxides) is proposed by Russian researchers, and has been successfully implemented. The core idea of REMIX is to extract uranium and plutonium isotopes in spent fuel without separation, and then add a proper amount of high-concentration uranium or a certain amount of plutonium to make new nuclear fuel according to the composition ratio. In this way, the reprocessing process can be greatly simplified, most of the spent fuel reenters the REMIX cycle, and the amount of spent fuel storage is reduced. Moreover, plutonium is no longer extracted separately, and the risk of nuclear material diffusion is reduced.

[0003] The research object corresponding to the Russian REMIX technology is a VVER-1000 reactor. The REMIX technology will be verified on a M310 reactor type in China, and key nuclide data in spent fuel will be given. Based on the data and the currently scaled PUREX reprocessing process, a uranium-plutonium non-separation reprocessing process is designed, and verification is performed on a reprocessing simulation calculation software. SUMMARY

[0004] To solve the defects in the prior art, the purpose of the present application is to provide a method for recovering uranium and plutonium without separation from REMIX spent fuel. The method is applied to a M310 reactor type, and the method can meet the requirement of 99.9% uranium-plutonium yield under the condition of uranium-plutonium non-separation.

[0005] To achieve the above purpose, a technical solution adopted by the present application is as follows:

[0006] A method for recovering uranium and plutonium from REMIX spent fuel without separation, comprising the following steps:

[0007] S1, element disassembly shearing is performed on the structure of the REMIX spent fuel assembly to be treated to obtain REMIX spent fuel pellets;

[0008] S2, the spent fuel pellets are dissolved to obtain a feed liquid;

[0009] S3, the fissile element in the feed liquid is decontaminated and purified by using a co-extraction and co-stripping circulation process, so that the fissile element purification meets the product requirements;

[0010] S4, a tail-end co-precipitation calcination process is used to co-precipitate uranium and plutonium from the feed liquid after step S3 to obtain a uranium-plutonium mixed oxide.

[0011] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, element disassembly shearing is performed in step S1 by using a shearing device and process for Daya Bay pressurized water reactor spent fuel elements.

[0012] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, the spent fuel pellets are dissolved in step S2 by using a reinforced dissolution method to ensure the dissolution of insoluble substances, and then the feed liquid is filtered and clarified to obtain a 1AF feed liquid.

[0013] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, step S3 specifically includes the following process:

[0014] S31, the 1AF feed liquid enters a first circulation uranium-plutonium co-extraction section, and a high-acid extraction and high-acid washing process is used to decontaminate and purify the 1AF feed liquid to obtain a 1AP feed liquid;

[0015] S32, the 1AP feed liquid enters a first circulation uranium-plutonium co-stripping section, and a double-acid stripping process is used to strip the 1AP feed liquid to obtain a 1BP feed liquid;

[0016] S33, the 1BP feed liquid enters a second circulation uranium-plutonium co-extraction section, and a high-acid extraction and high-acid washing process is used to decontaminate and purify the 1BP feed liquid to obtain a 2AP feed liquid;

[0017] S34, the 2AP feed liquid enters a second circulation uranium-plutonium co-stripping section, and a double-acid stripping process is used to strip the 2AP feed liquid to obtain a 2BP feed liquid.

[0018] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, the process parameters involved in each cycle of co-extraction and co-stripping process are determined by simulation calculation in step S3, and the process parameters include the concentration of each reagent used, flow ratio, and feed stage number.

[0019] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, according to the simulation calculation results and future experimental data, if the fissile purification does not meet the product requirements after two cycles of co-extraction and co-stripping process, a third cycle of co-extraction and co-stripping process is added.

[0020] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, the nitric acid concentration of the 1AF feed solution is adjusted to be in the range of 3-5 mol / L, and the uranium concentration of the 1AF feed solution is controlled to be in the range of 200-250 g / L in step S31, the nitric acid concentration of the 1AS is 2-4 mol / L, and the organic extractant of the 1AX is 30% TBP-kerosene.

[0021] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, the double acid stripping process in step S32 is as follows: 1BX1 is high acid with small flow ratio, mainly used for adjusting the acidity of stripping nitric acid and stripping plutonium; and 1BX2 is low acid with large flow ratio, mainly used for stripping uranium.

[0022] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, the process flow of step S33 is basically the same as that of step S31, and the process flow of step S34 is basically the same as that of step S32, and the concentration of each reagent is adjusted adaptively according to the simulation calculation results.

[0023] Further, in the method for recovering uranium and plutonium from REMIX spent fuel without separation as described above, the uranium-plutonium mixed oxide is obtained by using the tail-end co-precipitation and calcination process of Daya Bay pressurized water reactor spent fuel reprocessing in step S4 to treat the feed solution after step S3.

[0024] The method for recovering uranium and plutonium from REMIX spent fuel without separation according to the present application has the following significant technical effects:

[0025] The method provided by this invention is applied to the M310 reactor type. Using the computer simulation software of the 1A process extraction unit as a tool, combined with the five cycles of REMIX fuel used in pressurized water reactors, and based on the spent fuel composition data calculated by simulation of a burnup of 42 GWd / tU, the method employs computer simulation technology and a co-extraction and co-back-extraction cycle process to decontaminate and purify the fragmented elements in the feed liquid. Under the condition that uranium and plutonium are not separated, the product requirement of 99.9% uranium and plutonium recovery rate is met. This avoids the problems of large amounts of waste liquid, waste, and waste gas generated by processes such as PUREX, as well as the large workload of waste treatment and geological disposal. At the same time, it greatly simplifies the reprocessing process, with most of the spent fuel re-entering the REMIX cycle, reducing the amount of spent fuel stored. Furthermore, plutonium is no longer extracted separately, reducing the risk of nuclear material proliferation. Attached Figure Description

[0026] Figure 1 This is a flowchart of a method for recovering uranium and plutonium from spent REMIX fuel without separation, provided in an embodiment of the present invention.

[0027] Figure 2 The interface of the computer simulation software for the process extraction unit;

[0028] Figure 3 This is a simplified flowchart of a method for recovering uranium and plutonium from spent REMIX fuel without separation, provided in an embodiment of the present invention.

[0029] Figure 4 for Figure 1 The process flow diagram of co-extraction and co-back-extraction cycle described in the method;

[0030] Figure 5 This is a simplified flow chart of the first cycle co-extraction and co-back-extraction process section. Detailed Implementation

[0031] The present invention will now be further described with reference to specific embodiments and the accompanying drawings.

[0032] To address the issues mentioned in the background art, such as the large amount of waste liquid, waste gas, and waste gas generated by the PUREX process, and the significant workload involved in the treatment of these three wastes and geological disposal, this invention utilizes 1A process extraction unit computer simulation software (… Figure 2 Using this tool and combining the five cycles of REMIX fuel used in pressurized water reactors, and based on spent fuel composition data calculated using a burnup of 42 GWd / tU, a REMIX reprocessing flow that achieves a 99.9% uranium-plutonium recovery rate under non-separation conditions is proposed using computer simulation technology. Unlike the traditional Purex process, the REMIX spent fuel reprocessing adopts a non-separation process route for uranium and plutonium, meaning it does not involve the redox computer simulation of the Purex process's 1B stage.

[0033] According to the technical concept, the method for recovering uranium and plutonium from REMIX spent fuel without separation provided in the embodiment of the application, Figure 1 and Figure 3 A flowchart of the method is shown, which comprises the following steps:

[0034] S1, element disassembly shearing is performed on the REMIX spent fuel assembly structure to be treated to obtain REMIX spent fuel pellets.

[0035] For the REMIX spent fuel assembly structure to be treated, element disassembly shearing is performed by referring to the shearing equipment and process of the Daya Bay pressurized water reactor spent fuel element to obtain REMIX spent fuel pellets.

[0036] S2, the spent fuel pellets are dissolved to obtain a feed liquid.

[0037] Since the REMIX spent fuel has a high plutonium content and the amount of fission fragment elements is slightly higher than that of the pressurized water reactor spent fuel element, the dissolution process is more difficult, the amount of insoluble residue increases, and secondary precipitation is prone to occur. Therefore, by comparing the dissolution process differences between the REMIX spent fuel and the pressurized water reactor spent fuel element, a reinforced dissolution method is used for the REMIX spent fuel in the application to ensure the dissolution of the difficult-to-dissolve substances, and then a feed liquid clarification device is used to filter the feed liquid to obtain 1AF feed liquid without precipitates.

[0038] S3, the fission fragment elements in the feed liquid are decontaminated and purified by using a co-extraction and co-stripping circulation process, so that the fission fragment purification meets the product requirements.

[0039] Figure 4 A flowchart of the co-extraction and co-stripping circulation process is shown, and the specific process flow of this step comprises the following steps:

[0040] S31, the 1AF feed liquid enters a first circulation uranium-plutonium co-extraction section, and a high-acid extraction and high-acid washing process is used to decontaminate and purify the 1AF feed liquid to obtain 1AP feed liquid;

[0041] S32, the 1AP feed liquid enters a first circulation uranium-plutonium co-stripping section, and a double-acid stripping process is used to strip the 1AP feed liquid to obtain 1BP feed liquid;

[0042] S33, the 1BP feed liquid enters a second circulation uranium-plutonium co-extraction section, and a high-acid extraction and high-acid washing process is used to decontaminate and purify the 1BP feed liquid to obtain 2AP feed liquid;

[0043] S34, the 2AP feed liquid enters a second circulation uranium-plutonium co-stripping section, and a double-acid stripping process is used to strip the 2AP feed liquid to obtain 2BP feed liquid.

[0044] Before steps S31-S34, first, the concentration of each reagent, flow ratio, feed stage and other parameters used in each cycle of co-extraction and co-stripping process need to be determined by simulation calculation. According to the simulation calculation results and future experimental data, if the purification of the fragments does not meet the product requirements (the recovery rate of uranium and plutonium reaches 99.9%) after two cycles of co-extraction and co-stripping process, a third cycle of co-extraction and co-stripping process can be added.

[0045] The product of REMIX spent fuel reprocessing is a uranium-plutonium mixed oxide fuel, which is different from the uranium-plutonium co-decontamination and uranium-plutonium separation process of the typical Purex process. The REMIX spent fuel reprocessing process generally includes two cycles. The first cycle is a uranium-plutonium co-extraction and co-stripping cycle, and uranium, neptunium and plutonium are in the same liquid stream. Part of the neptunium enters the 1AW waste liquid. The second cycle is similar to the first cycle, which is still a uranium-plutonium co-extraction and co-stripping cycle. The extraction process is further strengthened to decontaminate the fragments. Uranium, neptunium and plutonium are in the same liquid stream. Part of the neptunium enters the 2AW waste liquid.

[0046] Figure 5 A flow diagram of the first cycle of co-extraction and co-stripping process section is shown. The 1A co-extraction section adopts a high-acid extraction and high-acid washing process. First, the concentration of nitric acid in the 1AF feed liquid is adjusted to be in the range of 3-5 mol / L. The uranium concentration of the 1AF feed liquid is controlled in the range of 200-250 g / L. The 1AX organic extractant can be selected from TBP-kerosene. The concentration of nitric acid in the 1AS is 2-4 mol / L. The specific concentration range is determined according to simulation calculation. The 1AF feed liquid is treated by the 1A co-extraction process to obtain the 1AP feed liquid and the 1AW waste liquid. The 1AP feed liquid enters the 1B co-stripping section, which adopts a double-acid stripping process. The 1BX1 is high-acid with a small flow ratio, which is mainly used to adjust the acidity of the stripping nitric acid and strip plutonium. The 1BX2 is low-acid with a large flow ratio, which is mainly used to strip uranium. The concentration of nitric acid and the flow ratio, feed stage and other parameters of the 1BX1 and 1BX2 are determined by simulation calculation. The 1AP feed liquid is treated by the 1B co-stripping process to obtain the 1BP feed liquid and the 1BW waste liquid. The second cycle process is basically the same as the first cycle process. The concentration of each reagent is slightly adjusted according to the simulation calculation results to achieve the purpose of further purification of the fragment elements.

[0047] S4, the tail-end co-precipitation calcination process is used to co-precipitate uranium and plutonium from the feed liquid treated in step S3 to obtain a uranium-plutonium mixed oxide.

[0048] Specifically, the tail-end co-precipitation calcination process of spent fuel reprocessing in Daya Bay pressurized water reactor can be used to co-precipitate uranium and plutonium from the feed liquid treated in step S3, for example, 17% low-concentration uranium is added to the feed liquid to obtain a uranium-plutonium mixed oxide, which is used as nuclear fuel again to enter the REMIX cycle.

[0049] Example REMIX spent fuel reprocessing example

[0050] 1. Post-processing of initial 1AF feed solution source term data calculation

[0051] The input condition is the balanced cycle of M310 nuclear power unit, the fuel assembly is AFA3G type with improved grid, and the average discharge burnup is "44224" MWd / tU.

[0052] The spent fuel assembly is cooled for 5 years and then processed by post-processing. Only U and Pu isotopes are retained in the post-processing process, and the rest is removed. It is assumed that the post-processing removal rate is 100%, and there is no loss in the post-processing process.

[0053] It is assumed that there is no element loss in the dissolution process, the uranium concentration of 1AF feed solution is controlled at 225g / L, and the nitric acid concentration is calculated at 2-4mol / L. The composition of 1AF feed solution is:

[0054] Table 1 Composition of 1AF feed solution

[0055]

[0056] 2. First cycle calculation results

[0057] Adjust the acidity of 1AF feed solution to 2mol·L-1HNO3 concentration, extract with 30% TBP-kerosene, 8-stage feeding, flow ratio 1AF:1AX:1AS=1:2.4:0.57, the concentrations of each element in the organic and aqueous phases after the first extraction are shown in the table:

[0058] Table 2 Concentrations of each element in organic and aqueous phases after 1A extraction

[0059]

[0060]

[0061] Stripping Pu with 4.5mol·L-1HNO3, stripping uranium with 0.01mol·L-1HNO3, flow ratio 1AP:1BX1:1BX2=1:0.1:1.3, the first time the co-stripping results are shown in the table:

[0062] Table 3 Concentrations of each element in organic and aqueous phases after first stripping

[0063] Element 1 BP aqueous phase concentration g / L 1 BW organic phase concentration g / L U 62.68 2.60E-10 Pu 0.703 2.39E-24 Np 3.62 5.39E-21 Tc 0.059 2.96E-9 Zr 9.68E-11 2.53E-48 Ru 6.31E-12 1.18E-17 HNO3 0.49 mol / L 7.06E-5 mol / L Flow rate 1.44 ml / min 0.96 ml / min

[0064] According to the above results, the uranium recovery rate of the first extraction is 99.9999%, and the plutonium recovery rate is 99.9999%. Most of the fragment elements such as Zr and Ru enter 1AW.

[0065] 3. Second cycle calculation results

[0066] The 1BP was used as the 2AF feed solution to perform two-cycle extraction, 8-stage feeding, and the flow ratio of 2AF:2AX:2AS was 1:0.66:0.3, and the results are shown in Table 4.

[0067] Table 4 Concentrations of various elements in the organic phase and the aqueous phase after 2A extraction

[0068]

[0069]

[0070] The Pu was stripped with 6 mol·L-1HNO3, and the U was stripped with 0.01 mol·L-1HNO3, and the flow ratio was 1AP:1BX1:1BX2=1:0.1:1.3, and the first stripping results are shown in the table:

[0071] Table 5 Concentrations of various elements in the organic phase and the aqueous phase after the first stripping

[0072] Element 2 BP aqueous phase concentration g / L 2 BW organic phase concentration g / L U 63.02 2.09E-9 Pu 0.707 7.14E-23 Np 1.98 2.88E-19 Tc 3.28e-5 2.47E-12 Zr 1.99E-20 3.03E-58 Ru 4.92E-38 7.76E-45 HNO3 0.65 mol / L 7.06E-5 mol / L Flow rate 1.44 ml / min 0.96 ml / min

[0073] According to the results, the uranium recovery rate in the second cycle extraction was 99.9999%, and the plutonium recovery rate was 99.999%. Most of the fission fragment elements such as Zr and Ru entered 1AW.

[0074] According to the results of the two cycles, the uranium and plutonium recovery rate reached the index of 99.9%, and most of the fission fragment elements were removed. After tail-end co-precipitation calcination, the uranium-plutonium mixed product can enter the next cycle to prepare fuel elements.

[0075] The method provided by the application does not separate and recover uranium and plutonium from REMIX spent fuel, is applied to M310 type, uses 1A process extraction unit computer simulation software as a tool, combines five cycles of REMIX fuel used in a pressurized water reactor, uses simulated calculation data of spent fuel composition according to a burnup of 42 GWd / tU, uses computer simulation technology, adopts a co-extraction and co-stripping cycle process to decontaminate and purify fission fragment elements in a feed solution, meets the product requirement of a uranium-plutonium recovery rate of 99.9% under the condition of no separation of uranium and plutonium, avoids problems such as a large amount of waste liquid, waste, and waste gas generated in a PUREX process, and a large workload of treatment and geological disposal of the three wastes; simultaneously, the reprocessing process is greatly simplified, most of the spent fuel reenters the REMIX cycle, the storage amount of spent fuel is reduced, and the plutonium is no longer extracted separately, so that the risk of nuclear material diffusion is reduced.

[0076] The above examples are merely illustrative of the application and do not limit the scope of the application as described or defined by the appended claims. Thus, the examples should not be considered to limit the scope of the application in any way. The scope of the application is limited only by the claims and their equivalents.

Claims

1. A method for recovering uranium and plutonium from REMIX spent fuel of M310 reactor without separation, the method comprising the following steps: S1. Element disassembly shearing is performed on the REMIX spent fuel assembly structure to be treated to obtain REMIX spent fuel pellets; S2. The spent fuel pellets are dissolved to obtain a feed solution; S3. The fissile elements in the feed solution are decontaminated by using a co-extraction and co-stripping circulation process, so that the fissile element decontamination meets the product requirements; S4. The feed solution after step S3 is subjected to uranium-plutonium co-precipitation by using a tail-end co-precipitation calcination process to obtain a uranium-plutonium mixed oxide; Step S3 specifically comprises the following process: S31. The 1AF feed solution enters a first circulation uranium-plutonium co-extraction section, and the 1AF feed solution is decontaminated by using a high-acid extraction and high-acid washing process to obtain a 1AP feed solution; S32. The 1AP feed solution enters a first circulation uranium-plutonium co-stripping section, and the 1AP feed solution is subjected to reverse extraction by using a double-acid stripping process to obtain a 1BP feed solution; S33. The 1BP feed solution enters a second circulation uranium-plutonium co-extraction section, and the 1BP feed solution is decontaminated by using a high-acid extraction and high-acid washing process to obtain a 2AP feed solution; S34. The 2AP feed solution enters a second circulation uranium-plutonium co-stripping section, and the 2AP feed solution is subjected to reverse extraction by using a double-acid stripping process to obtain a 2BP feed solution; According to the simulation calculation results and future experimental data, if the fissile element decontamination does not meet the product requirements after two rounds of co-extraction and co-stripping process circulation, a third round of co-extraction and co-stripping process is added.

2. The method of claim 1, wherein the uranium and plutonium are recovered from the M310 REMIX spent fuel without separation. In step S1, the shearing equipment and process of the Daya Bay pressurized water reactor spent fuel element are used for element disassembly shearing.

3. The method of claim 1 or 2, wherein the method is characterized by, In step S2, the spent fuel pellets are dissolved by using a reinforced dissolution method to ensure the dissolution of the insoluble matter, and then the feed solution is filtered and clarified to obtain a 1AF feed solution.

4. The method of claim 3, wherein the uranium and plutonium are not separated from the M310 REMIX spent fuel. In step S3, the process parameters involved in each circulation co-extraction and co-stripping process are determined by simulation calculation, and the process parameters include the concentrations of each reagent, the flow ratio, and the feed stage number.

5. The method of claim 4, wherein the uranium and plutonium are not separated from the M310 REMIX spent fuel. In step S31, the nitric acid concentration of the 1AF feed solution is adjusted to be within the range of 3-5 mol / L, the uranium concentration of the 1AF feed solution is controlled to be within the range of 200-250 g / L, the 1AS nitric acid concentration is 2-4 mol / L, and the 1AX organic extractant is 30% TBP-kerosene.

6. The method of claim 5, wherein the uranium and plutonium are not separated from the M310 REMIX spent fuel. In step S32, the double-acid stripping process specifically is that 1BX1 is high-acid with a small flow ratio, which is mainly used for adjusting the acidity of the stripping nitric acid and stripping plutonium; and 1BX2 is low-acid with a large flow ratio, which is mainly used for stripping uranium.

7. The method of claim 6, wherein the uranium and plutonium are not separated from the M310 REMIX spent fuel. The process flow of step S33 is basically the same as that of step S31, and the process flow of step S34 is basically the same as that of step S32, and the concentrations of each reagent are adaptively adjusted according to the simulation calculation results.

8. The method of claim 1, wherein the method is characterized by, In step S4, the Daya Bay pressurized water reactor spent fuel reprocessing tail-end co-precipitation calcination process is used to co-precipitate the feed solution after step S3 to obtain the uranium-plutonium mixed oxide.

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