JPH0810273B2 - Reprocessing of spent nuclear fuel - Google Patents

Description

Detailed Description of the Invention

TECHNICAL FIELD The present invention relates to a method and a facility for reprocessing spent nuclear fuel.

[Prior art]

The conventional technique will be described separately for the reprocessing method and the ruthenium removing technique. First, a conventional reprocessing method will be described with reference to FIG. In FIG. 3, except for the ruthenium removing step 12, a conventional reprocessing method (main steps only) is shown. As a conventional reprocessing method, for example, Manson Benedict et al., Nuclear Chemical Engineering 2nd Edition, 1981, page 467, Figure 10.2 or 492, Figure 10.11 (M
anson Benedict, et al., Nuclear Chemical Engineerin
g, Second Edition, 1981, p.467, Figure 10.2 or p.492, F
As described in igure 10.11), it was not considered to remove ruthenium from the reprocessing solution (providing a ruthenium removal step). The conventional reprocessing method (process) in which step 12 is deleted from FIG. 3 is mainly a dissolving step 11 in which spent nuclear fuel 10 is dissolved in nitric acid, and uranium and plutonium are removed from this dissolving solution.
Co-decontamination process 1 using TBP (tributyl phosphate) for extraction
3. It comprises a partitioning step 14 for separating uranium and plutonium extracted into TBP, and purification steps 15 and 16 for purifying the separated uranium and plutonium. In addition, a high-level waste liquid treatment process 17 for treating the waste liquid generated in the co-decontamination process 13 and a low-level waste liquid treatment process 18 for treating the waste liquid generated in the distribution and refining processes (14, 15, 16) are also attached. . Spent nuclear fuel contains uranium, plutonium, and dozens of fission products, but among the fission product elements, ruthenium poses a problem in the reprocessing step due to its unique chemical properties. For example, in the dissolution step 11, of ruthenium dissolved in a nitric acid solution, a ruthenium species that is oxidized to a valence of 6 to 8 may greatly increase the corrosion potential of the dissolution tank and accelerate the material corrosion of the dissolution tank. In the co-decontamination step 13, since the ruthenium complex coordinated with nitrate ions has the property of being easily extracted by TBP, the ruthenium extracted similarly in the first cycle of uranium and plutonium extraction is back-extracted. Will be difficult. The water phase side component in the co-decontamination step 13 contains tens of kinds of fission products separated from uranium and plutonium,
After being evaporated and concentrated in the high-level waste liquid treatment step 17, it is vitrified and stored. The ruthenium contained in this high-level liquid waste volatilizes at the time of evaporative concentration or vitrification and deposits in the off-gas system, so there is a risk of blocking the piping. Further, the evaporation can has the same problem of corrosion as the melting tank. Therefore, it is desirable to remove ruthenium, which causes troubles at each step of the reprocessing, as early as possible in the initial steps of the reprocessing process. Next, a conventional ruthenium removing technique will be described.
Currently known techniques for removing ruthenium ions from nitric acid solutions are as follows, not for the solution in the initial step of reprocessing but for radioactive waste solutions such as reprocessing waste solutions. All of these are technologies that require additives such as a reducing agent, an adsorbent, an oxidizing agent, and a solvent, and a ruthenium removal technology that does not add additives has not been known. According to JP-A-56-19500, a reducing agent such as hydrazine and copper ions are added to a treatment solution to form a cuprous oxide precipitate, and ruthenium is fixed thereon. In JP-A-57-50698, ruthenium is adsorbed and removed by passing the treatment solution through a column in which a metal powder such as iron and copper is mixed with activated carbon. In addition, a method of selectively extracting ruthenium ions using a special solvent such as thiophosphoric acid, conversion of ruthenium ions to volatile ruthenium tetroxide using ozone or permanganate or chromate oxidizers. A method of removing as a gaseous substance is known. Since the treatment solution also contains a radionuclide other than ruthenium, addition of an oxidizing agent, a reducing agent, or the use of an adsorbent as in the above-mentioned conventional technique is desirable from the viewpoint of increasing the amount of secondary radioactive waste. Absent.

[Problems to be Solved by the Invention]

The prior art described above does not consider the point of removing ruthenium contained in the solution such as the dissolution step and the co-decontamination step, and there is a risk that ruthenium causes trouble in each step of reprocessing. Also, regarding the ruthenium removal technology, no consideration was given to the fact that additives are not used as much as possible, and there was a problem that the amount of secondary radioactive waste increased. In particular, when ruthenium is removed in the initial step of the reprocessing process, it is necessary to avoid the influence of additives on coexisting substances such as uranium and plutonium and the generation of waste in the initial step as much as possible. Furthermore, since the point of recovering ruthenium as a valuable metal was not considered, there is a difficulty in recovering the removed ruthenium. An object of the present invention is to provide a method for reprocessing spent nuclear fuel that can reduce the amount of radioactive waste and prevent troubles due to ruthenium. Another object of the present invention is to provide a method for reprocessing spent nuclear fuel, which has a simple structure and can avoid the problem caused by iodine after the co-decontamination step, in addition to the above objects.

[Means for solving problems]

A feature of the present invention that achieves the above-mentioned object is a ruthenium ion which is either a fuel solution on the upstream side of the co-decontamination process of a spent nuclear fuel re-treatment facility, or a re-treatment waste liquid generated in the co-decontamination process Is to electrolytically reduce a nitric acid aqueous solution containing a to deposit the ruthenium ions as metal ruthenium to remove the metal ruthenium from the nitric acid aqueous solution. Another feature of the present invention that achieves the above-mentioned other object is that an aqueous nitric acid solution containing iodine oxide ions and ruthenium ions is formed using an electrolyzer on the upstream side of the co-decontamination step of the spent fuel reprocessing facility. By reducing, the iodine oxide ions are discharged as iodine from the nitric acid aqueous solution, and the ruthenium ions are electrodeposited as metal ruthenium to remove the metal ruthenium from the nitric acid aqueous solution.

[Action]

Since ruthenium ions are removed from either the fuel solution on the upstream side of the co-decontamination step or the reprocessing waste liquid generated in the co-decontamination step, troubles caused by ruthenium after the removal can be prevented. Further, since ruthenium ions are electrodeposited as metal ruthenium by electrolytic reduction, ruthenium can be removed without using an additive and an adsorbent. Therefore, the amount of radioactive waste generated when ruthenium is removed can be reduced. Metal ruthenium, which is a valuable metal, can be recovered. Further, according to another feature of the present invention, since the electrolytic oxidation device is used to electrolytically reduce iodine oxide ions and ruthenium ions, the electrolytic device can be shared, and the device configuration for electrolytic reduction of iodine ions and ruthenium ions is complicated. No, the configuration is simple. Furthermore, since iodine is removed, problems caused by iodine after the co-decontamination step can be avoided. The logical proof of the present invention is based on the results of the inventors' systematic investigation of the reduction and deposition potentials of many metal ions contained in the reprocessed nitric acid aqueous solution. The results are shown in FIG. This shows that the reduction deposition potential of ruthenium, which is a noble metal element, is higher than the reduction deposition potential of other coexisting cations. This has been found to be very convenient for electrowinning of ruthenium ions at the cathode. Therefore, in the present invention, the reduction deposition potential of ruthenium is 0.
Ruthenium ions are selectively deposited on the cathode while adjusting the current so as to maintain the electrode potential corresponding to 45 V (relative to the standard hydrogen electrode). This cathode reaction is expressed by the following equation. Ru ² ++ 2e- → Ru (1) The electrode potential actually set is 0 in consideration of the activation overvoltage.
A range of ~ 0.45V (relative to a standard hydrogen electrode) is preferred.
At this potential, the nitrate complex of ruthenium ion, which is difficult to reduce, can be reduced. According to the present invention, since ruthenium is not contained in the solution transferred to the subsequent step of the reprocessing, various troubles caused by ruthenium in the reprocessing process can be prevented. For example, by removing ruthenium from the fuel solution, it is possible to avoid obstacles caused by ruthenium in the co-decontamination process and the high-level waste liquid concentration and solidification process. Further, by removing the ruthenium ions existing in the reprocessed nitric acid aqueous solution by the electrolytic reduction method, it is not necessary to add an additive or an adsorbent, so that the amount of secondary waste generated can be reduced.

【Example】

An embodiment of the present invention will be described below with reference to FIGS. 1 and 3. Example 1 FIG. 1 shows an optimum apparatus configuration of means for removing ruthenium, which is a feature of the present invention. A reprocessed nitric acid aqueous solution containing ruthenium ions is supplied to the electrolytic cell 7 through the supply port 8. The electrolytic cell is provided with a working electrode 2, a counter electrode 3, a reference electrode 4 such as a saturated calomel electrode or a silver chloride electrode, and a nitrogen blowing pipe for purging. After the supply of the treatment liquid, the potential control device 1 performs electrolysis while adjusting the current value between the working electrode 2 and the counter electrode 3 so that the electrode potential with respect to the reference electrode 4 becomes a set value. The degree of progress of the electrolytic reduction reaction can be known by monitoring the current value with the potential control device 1. The treated liquid after the reaction is discharged from the drain port 9 and transferred to the next step. The gas generated during the electrolysis is exhausted to the existing off-gas processing system in the reprocessing facility through the exhaust port 6. In this embodiment, the spent nuclear fuel solution is supplied as the processing liquid. In this case, about 1% of iodine, which is one of the fission products, exists in the form of IO ₃ −. Since this iodine is troublesome to bring into the step after co-decontamination, it is desirable to remove it from the nitric acid solution before the co-decontamination step. A reprocessing method for removing ruthenium before the co-decontamination step will be described with reference to FIG. After the spent nuclear fuel 10 is dissolved in nitric acid in the dissolution step 11, ruthenium is removed from this solution in the ruthenium removal step 12 by the method shown in the present embodiment (some fission products may be mixed). To remove. By removing ruthenium, it is possible to prevent troubles caused by ruthenium in each step of reprocessing after the co-decontamination step 13. When ruthenium is removed before the co-decontamination step 13, it is necessary to consider iodine removal as described above. Therefore, first, the electrode potential that reduces IO ₃ − to I ₂ ,
That is, the potential of the working electrode 2 is set to 0.7 to 1.1 V (relative to the standard hydrogen electrode), and electrolysis is performed while purging with nitrogen gas purging to remove iodine. The electrode reaction at this time is as shown in equation (2). IO ₃ − ＋ 6H ＋＋ 5e− → (1/2) I ₂ ↑ ＋ 3H ₂ O …… (2) After that, change the set potential to 0.0 to 0.45V (relative to the standard hydrogen electrode) and continue electrolysis. Reduces ruthenium ions. In this case, the nitrogen gas purge may be stopped, but it is preferable to purge it for the stirring effect and removal of dissolved oxygen. Materials for the working electrode 2 and the counter electrode 3 are preferably those having corrosion resistance and good electrical conductivity, such as platinum and platinum-coated titanium are suitable, but graphite electrodes are suitable in view of hydrogen overvoltage. Even if the ruthenium metal is electrodeposited on the electrode, the electrode characteristics do not deteriorate so much.
Rather, the deposited ruthenium changes to RuO ₂ which is in the passive state, and the reduction efficiency can be expected to improve due to the self-catalytic action on the reduction of ruthenium ions. According to this example, ruthenium ions can be removed and recovered from the fuel solution without using an additive, and dissolved iodine can be expelled. Example 2 In this example, the apparatus of FIG. 1 is used to treat high-level waste liquid generated in the co-decontamination process. In this case, the set potential is 0.
0 to 0.45 V (relative to standard hydrogen electrode). According to this example, not only ruthenium, which is a valuable metal, can be recovered from the high-level waste liquid, but also it is possible to prevent contamination of the off-gas treatment system and corrosion of the evaporator due to volatilization of ruthenium in the evaporative concentration step.

【The invention's effect】

According to the present invention, since ruthenium is removed, troubles due to ruthenium after the removal can be prevented.
The amount of radioactive waste generated during the removal of ruthenium can be reduced. Furthermore, ruthenium metal, which is a valuable metal, can be recovered. According to another feature of the present invention, in addition to the above-mentioned effects, an effect that problems caused by iodine after the co-decontamination step can be avoided can be obtained. Furthermore, since the electrolysis device can be shared, the device configuration for electrolytically reducing iodine ions and ruthenium ions does not become complicated and has a simple configuration.

[Brief description of drawings]

FIG. 1 is a diagram showing an embodiment of a means for removing ruthenium, which is a feature of the present invention, FIG. 2 is a diagram showing a basic basis for devising the present invention, and FIG. 3 is a diagram showing the present invention. It is a figure which shows one Example.

[Explanation of symbols]

1: Potential control device, 2: Working electrode, 3: Counter electrode, 4: Reference electrode, 5: Nitrogen blowing pipe, 6: Exhaust port, 7: Electrolyzer, 8: Liquid supply port, 9: Discharge port, 1
0: spent nuclear fuel, 11: dissolution process, 12: ruthenium removal process, 13: co-decontamination process, 14 …… distribution process, 15: uranium purification process, 16: plutonium purification process, 17: high-level waste liquid treatment process , 18: Low level waste liquid treatment process, 19: Uranium product, 20: Plutonium product, 21: High level waste, 22: Low level waste, 23: Ruthenium

Claims (4)

[Claims] 1. A nitric acid aqueous solution containing ruthenium ions, which is either a fuel solution on the upstream side of a co-decontamination process of a spent nuclear fuel reprocessing facility or a reprocessing waste liquid generated in the co-decontamination process. A method for reprocessing spent nuclear fuel, which comprises electrolytically reducing the ruthenium ions to deposit as metal ruthenium and removing the metal ruthenium from the nitric acid aqueous solution. 2. The electrode potential of a working electrode immersed in the nitric acid aqueous solution to be treated in the electrolytic reduction step is set within a range of 0.0 to 0.45 V with respect to a similarly immersed standard hydrogen electrode. The method for reprocessing spent nuclear fuel according to item 1 of the above. 3. The iodine oxide ion is reduced by reducing an aqueous nitric acid solution containing iodine oxide ion and ruthenium ion using an electrolyzer at an upstream side of a co-decontamination step of a spent nuclear fuel reprocessing facility. A method for reprocessing spent nuclear fuel, which comprises discharging as ruthenium ions from the aqueous nitric acid solution and depositing the ruthenium ions as metallic ruthenium to remove the metallic ruthenium from the nitric acid aqueous solution. 4. When electrolytically reducing the iodine oxide ion, the electrode potential of the working electrode immersed in the aqueous nitric acid solution to be treated in the electrolytic reduction step is set to the standard hydrogen electrode similarly immersed. The method for reprocessing spent nuclear fuel according to claim 3, wherein the range is set to 0.7 to 1.1 V, and when the ruthenium ions are electrolytically reduced thereafter, the range is set to 0.0 to 0.45 V.