JPH03179297A - Treatment of highly radioactive waste - Google Patents

Abstract

PURPOSE:To simplify a treatment process and to miniaturize a treatment apparatus by subjecting the calcined matter of highly radioactive waste to heating and melting treatment at high temp. of 1,000 deg.C or higher in a reductive state. CONSTITUTION:The calcined matter of highly radioactive waste is introduced into a melting container 10 to be subjected to heating reduction treatment at 1,000 deg.C or higher and molybdenum present in the calcined matter is reduced and alloyed with a platinum group element to be separated into a platinum group element layer 12 high in specific gravity and an oxide layer low in specific gravity. The platinum group element layer 12 and the oxide layer 14 successively flow down from the downflow nozzle 16 provided to the bottom part of the container 10 and are injected in separate containers to be solidified. By this method, the platinum group element can be separated and recovered without adding additives and a treatment process can be simplified and a treatment apparatus can be miniaturized. Since the oxide residue is solidified as it is, sharp volume reducing solidification as low as one over several tenths as compared with conventional glass solidifying treatment can be realized and storage disposal cost is reduced to a large extent.

Description

[Detailed Description of the Invention] [Industrial Application Field] The present invention provides a method for treating calcined bodies of high-level radioactive waste generated in spent fuel reprocessing processes at high temperatures.
The present invention relates to a treatment method for separating and recovering platinum group elements and converting residual oxides into highly sterile solidified waste.

[Prior Art] High-level radioactive waste generated in the reprocessing of spent fuel using the Burex method contains decay products (finsion products) in a nitric acid solution in the form of a solution or slurry. This highly radioactive waste will be solidified into glass and other materials in the future. This solidification method involves mixing decay products into the medium. Many different materials are being studied as media, including glass and synthetic rock (synrock). The concentration of the decay products in the medium is limited to about 10% due to the solubility of elements in the decay products in the medium and chemical durability (leaching rate into water). The volume of the solidified body should be reduced in order to reduce the cost of storage and disposal, and for this purpose it is necessary to increase the content of decay products in the solidified body, but this is currently difficult for the reasons mentioned above. It is.

On the other hand, highly radioactive waste contains platinum group elements such as Ru (ruthenium), Pd (palladium), RhC, and rhodium, which are useful elements.

Attempts to recover these platinum group elements from highly radioactive waste have been ongoing for many years, and the following three methods are known as prior art. These are: ■ Solvent extraction method that uses phosphate ester to separate target nuclides from a nitric acid solution of highly radioactive waste; ■ When melting highly radioactive waste into glass, lead is added and platinum group elements are transferred to the lead layer. Lead extraction method for separation ■This is an ion exchange method that performs ion exchange treatment on radioactive waste and separates the target nuclide.

[Problems to be Solved by the Invention] However, the conventional methods for recovering platinum group elements as described above each have the following drawbacks.

■In the solvent extraction method, phosphoric acid ester becomes a secondary waste, which is different from the TBP (butyl phosphate) used in reprocessing. Therefore, waste TBP and a separate waste solvent treatment method (research and development of treatment methods) and treatment plant construction, etc.). This expense is significant, raising the cost of the platinum group elements to be recovered above the commercial price, making recovery of the platinum group elements from highly radioactive waste economically unviable.

■The lead extraction method is advantageous in that it does not use any additives other than lead, but in order to increase the extraction efficiency, we use glass with a low viscosity, which is different from the glass used for the production of vitrified highly radioactive waste. There are many problems that must be solved, such as separating lead from platinum group elements and turning used lead into secondary waste.

■In the case of the ion exchange method, combustible substances are generated when the ion exchange resin comes into contact with nitric acid, which poses a safety problem.

Furthermore, no matter which of these methods is adopted, secondary waste is generated, and highly radioactive waste cannot be treated to reduce the volume of the waste.

The purpose of the present invention is to eliminate the drawbacks of the prior art as described above,
Platinum group elements can be easily recovered without adding additives,
Another object of the present invention is to provide a processing method that does not generate secondary waste and can achieve high volume reduction and solidification of highly radioactive waste.

[Means for Solving the Problems] The present invention, which can solve the above-mentioned technical problems, heats and melts a calcined body of highly radioactive waste at a high temperature of 1000°C or higher in a reduced state, and melts it in the calcined body. Mo (molybdenum) present in the oxide is reduced and alloyed with platinum group elements, the resulting platinum group alloy layer is separated from the oxide layer by sedimentation, and the residual oxide is solidified. This is a processing method.

Highly radioactive waste is usually a nitric acid solution obtained as an extraction residue in the spent fuel reprocessing process, and contains almost all the decay products in the spent fuel. In the present invention, this nitric acid solution is heated to exude moisture and nitric acid to obtain a calcined body. 1000 in the reduced state
When heated and melted at a high temperature of ℃ or higher, MO present in the calcined body is reduced and alloyed with platinum group elements, and the resulting platinum group alloy layer settles below the oxide layer and is separated from the oxide layer. Can be separated. The platinum group alloy is recovered and the residual oxide layer is solidified.

The present invention has two concepts. The first is to solidify highly radioactive waste without adding additives that create secondary waste that is difficult to dispose of, and the second is to reduce the platinum group elements in highly radioactive waste at high temperatures. It is to separate it as a metal.

The redox state during heat treatment of highly radioactive waste is controlled by the heating temperature, heating atmosphere, and addition of a reducing agent.

The heating temperature is 1000°C or higher. At temperatures below 1000°C, Pd and Rh are reduced to metals, but Ru and Mo are not. Therefore, it is preferable to perform the heat treatment at 1500 to 2000°C.

At 2000°C or higher, Ru -P d -Rh -Mo
Since the alloys in the system melt, higher temperatures are not necessary.

The heating atmosphere is controlled to promote the reduction reaction. Although it is possible to carry out the reduction reaction under air by raising the temperature to a higher temperature, there are many difficulties in heating technology such as the structure of the melting furnace, the material of the furnace, and the material of the melting vessel at temperatures above 2000°C. It is necessary to process at lower temperatures. Therefore, in the present invention, it is preferable to carry out the process in an atmosphere of air, nitrogen, or argon with a reduced oxygen content.

The reducing agent is used to further promote the reduction reaction.

As reducing agents, we use gaseous reducing agents such as hydrogen and carbon monoxide to avoid producing secondary waste, reducing agents that gasify in redox reactions such as carbon, and oxidizing agents that become waste such as alkaline earth metals and rare earth elements. A reducing agent that is a constituent element of the material layer is used.

These heating temperatures, atmospheres, and reducing agents are set and used in appropriate combinations depending on the reaction conditions.

[Function] Nuclear fission products in spent fuel can be broadly classified into ■metallic elements, ■nonmetallic elements, and ■rare earth elements. Examples of the metal elements include alkaline earth metals, transition metals such as MO, and platinum group elements. By heating the calcined body of highly radioactive waste liquid at high temperature,
Most of the alkali metals in the nonmetallic elements (2) and the metallic elements (2) are removed. They are Sb and Te.

Cs, Rb, etc. As a result, the main components of the calcined body, in the case of spent fuel with a burnup of 45,000 MWD/MTU and a cooling period of 5 years, have a content of 100 g/MT [excluding elements below I] as follows.

・Alkaline earth metals (Sr, Ba)... 3. 3kg/MTU 8. 7%・Transition metal (Z r , Mo , T c )...10
．． 5kg/MTU 27. 9%・Platinum group elements (R
u, Rh, Pd) ... 5. 4kg/MTU 14. 3%・Rare earth elements (Y, La, Ce, Pr, Nd, Eu, Gd)・
...18.5kg/gate TL149.1% total ...
37, 7kg/MTυ Therefore, by further heating, sintering and melting this calcined body, a solidified body with a high volume reduction rate compared to the decay product content of about 10% of a normal solidified body of highly radioactive waste can be obtained. is obtained. The vitrified material is 10 times as heavy as the decay products and has a volume of several hundred liters per ton of spent fuel, but in the present invention, the solidified material can have a volume of several tens of liters. This is possible without producing secondary waste of a different type than that produced by Burex reprocessing.

Furthermore, in the present invention, the platinum group elements can be recovered and separated by placing the system in a reduced state by heating and melting the calcined body. It is known that platinum group elements have low free energy for oxide formation and are reduced to a metallic state when heated. Further, the calcined body contains MO, which has a relatively small free energy of oxide formation.

The melting point of MO contained in the calcined body is 2623°C, the melting point of the platinum group elements is 1554°C for Pd, and 1963°C for Rh.
, Ru is 2254'C.

Since Ru has a different crystal type from Rh, it does not completely form a solid solution, and Pd does not form an alloy having a eutectic point with Rh and Ru. Therefore, the melting point of platinum group elements and their alloys can exceed 2000°C, and it is difficult to separate the platinum group elements alone or as an alloy from the oxide residue by melting the calcined body. . Even if the phases are separated, the melting temperature must be extremely high to separate them into two layers as a melt.

However, when MO is present, it forms low melting point alloys with the platinum group elements. In alloys of MO and platinum group elements, the lowest melting point is 1948°C for MORu, Mo-Pd.
The temperature is 1737°C for the Mo-Rh system and 1940°C for the Mo-Rh system.

In this way, the melting point of a platinum group element such as Ru, which has a melting point of 2254° C., can be lowered by alloying it with MO.

The present invention takes advantage of this phenomenon to combine Ru, PdRh, and Mo with 2
The platinum group elements are recovered by reducing them to a form that melts at temperatures below 000°C, forming a molten alloy layer, and separating it from the oxide layer, and solidifying the remaining decay product oxides with a high degree of sterility. It is something you put into your body.

The composition of the oxide of this decay product Ili is as follows in the case of the components of the above-mentioned calcined body.

・Alkaline earth metals (Sr, Ba)... 3. 3kg/MTU 11. 9%・Transition metals (Zr, Tc)...6. 0kg/MTU 21. 6%・Rare earth elements (Y, La, Ce, Pr, Nd, Eu, Gd)・
...18. 5kg/MTU 66. 5% total
...27, 8 kg/MTU Platinum group elements can be put to effective use by attenuating their radioactivity or separating radioactive isotopes, so they are subject to so-called "geological disposal" as radioactive waste. The remaining rare earth elements become oxides as main components. This is 27.8 kg per ton of spent fuel, and the actinide Np (0.7 k
g/MTU), Am (0.5kg/MTII)
Even if you add that, it's only about 30 kg. Their specific gravity is about 5
Therefore, the capacity is 6iMTU, which is 1/30 of the capacity of the vitrified material, which is 1801MTU.

As described above, the present invention makes it possible to significantly reduce costs in storing and disposing of highly radioactive waste.

In other words, by treating highly radioactive waste at high temperatures and controlling its redox treatment state, it is possible to extract volatile elements, which are βT nuclides, which are expected to be handled at lower costs in storage and disposal than alpha nuclides, and ■ useful metals. Radioactive waste can be roughly separated and disposed of into highly reduced volume solidified bodies of platinum group elements and oxides whose main components are rare earth elements.

[Example] FIG. 1 is a conceptual diagram showing an example of a processing apparatus for carrying out the method of the present invention. This is an example of a bottom flow type device. The calcined body of highly radioactive waste is placed in a melting container 10. The calcined body is subjected to a heat reduction treatment and is separated into a platinum group element layer 12 having a high specific gravity and an oxide layer 14 having a low specific gravity.

The layer 12 of the platinum group element and the layer 14 of the oxide flow sequentially down from the bottom flow nozzle 16 and are injected into a separate container and solidified.

FIG. 2 is a conceptual diagram showing another example of the processing apparatus used for carrying out the method of the present invention. This is an example of an overflow type apparatus. The calcined body of highly radioactive waste is placed in the center of the melting container 20 and heated and melted. The platinum group element layer 12 located below and the oxide layer 14 located above pass through channels 22 and 24 indicated by arrows, respectively, to a downstream nozzle 26.
．． 28 and poured into another container to solidify.

The device configuration is not limited to the above two examples, and an intermediate device configuration between a bottom flow type and an overflow type is also conceivable. That is, the platinum group element layer is caused to flow down by the bottom flow and is implanted and solidified, and the oxide layer is caused to flow down by the overflow and is implanted and solidified.

For the heat treatment of the calcined body, a single heater method, a direct energization method, a high frequency heating method, etc., which are used for vitrification of highly radioactive waste, can be applied.

Next, a specific experimental example will be described. Burnup degree 45000
? IWD/Montu calculates the composition of decay products in spent fuel with a cooling period of 5 years using the 0RIGEN code, combines a simulated highly radioactive waste fluid with the equivalent amount, heats this simulated waste fluid to 600°C, and temporarily calculates the composition. It was made into a fired body.

The calcined body was placed in a crucible containing 45 g of boron nitride, and heated and melted at 1800° C. for 1 hour in an argon atmosphere. After cooling, the crucible was destroyed and the contents were taken out. The contents were divided into two types, and there was a lump of metal at the bottom, which could be easily separated from the residue. As a result of analyzing the metal part with an X-ray microanalyzer (EPMA), Ru-P d
-Mo was detected. Rh is unconfirmed because its detection wavelength overlaps with that of Ru. The weight of this metal portion was about 90% of the weight of Ru-PdRh-Mo in the calcined body. Regarding the residue part, the leaching amount is determined according to JIS-R35.
Measurement was performed using a method similar to 02. Leaching rate is 8 x 10g/c
m''·d, which is approximately the same as that of the vitrified solidified material, and it was confirmed that it has the chemical durability as a highly radioactive solidified material.

[Effects of the Invention] As described above, the present invention is a method of heating and melting calcined bodies of highly radioactive waste at a high temperature of 1000°C or higher in a reduced state, so platinum group elements can be separated without adding additives. It can be recovered, simplifying the treatment process and downsizing the treatment equipment. In addition, since no additives are added, no secondary waste is generated, and since the residual oxide is solidified as it is, it is possible to achieve a significant volume reduction and solidification by several tens of minutes compared to conventional vitrification treatment. This will enable significant cost reductions in the storage and disposal of radioactive waste.

[Brief explanation of drawings]

FIG. 1 is a conceptual diagram showing an example of a processing device used to carry out the method of the present invention, and FIG. 2 is a conceptual diagram showing another example of the processing device. 10.20... Melting vessel, 12... Platinum group element layer, 14... Residual oxide layer, 16.26.28... Downflow nozzle.

Claims (1)

[Claims] 1. A calcined body of highly radioactive waste is heated to 100% in a reduced state.
The calcined body is heated and melted at a high temperature of 0°C or higher to reduce the molybdenum present in the calcined body and alloyed with platinum group elements, and the resulting platinum group alloy layer is collected by precipitation separation from the oxide layer, and the residue is recovered. A highly radioactive waste treatment method characterized by solidifying oxides. 2. The processing method according to claim 1, wherein the heating and melting treatment is carried out at 1500 to 2000°C.