GB2257293A

GB2257293A - Method of volume-reducing vitrification of highlevel radioactive waste

Info

Publication number: GB2257293A
Application number: GB9212578A
Authority: GB
Inventors: Misato Horie
Original assignee: Doryokuro Kakunenryo Kaihatsu Jigyodan
Current assignee: Doryokuro Kakunenryo Kaihatsu Jigyodan
Priority date: 1991-06-13
Filing date: 1992-06-12
Publication date: 1993-01-06
Anticipated expiration: 2012-06-12
Also published as: JPH04366800A; GB9212578D0; FR2677798A1; FR2677798B1; JP2551879B2; GB2257293B

Abstract

A method of volume-reducing vitrification of a high-level radioactive waste comprises calcining the radioactive waste, heating in the presence of a reducing agent of boron or a boron compound such as boron nitride and a vitrifying agent of an oxide of boron or a borosilicate in a reducing atmosphere at a temperature of 1,000 DEG C or above to melt the calcined material and to bring platinum group elements and molybdenum contained in the calcined material into a metallic form. The resulting molten metal is separated from a molten layer of residual oxides, and the molten layer of the residual oxides is solidified to obtain a vitrified product with a highly reduced volume. <IMAGE>

Description

METHOD OF VOLUME-REDUCING VITRIFICATION OF HIGH-LEVEL RADIOACTIVE WASTE The present invention relates to a method of volume-reducing vitrification of a high-level radioactive waste generated in, for example, the step of reprocessing spent nuclear fuels. More particularly, the present invention is concerned with a method of volume-reducing vitrification of a high-level radioactive waste by treating a calcined material of a high-level radioactive waste in the presence of a reducing agent of boron or boron compound and a vitrifying agent of an oxide of boron at a high temperature to melt the calcined material and to bring molybdenum and platinum group elements contained in the calcined material into a metallic form, separating and recovering the resulting metal, and solidifying resulting residual oxides as a vitrified waste of a high degree of volume reduction.

In the reprocessing of a spent. fuel, fission products are stored as a high-level radioactive waste in the form of a nitric acid solution. This high-level radioactive waste is solidified in the future by vitrification or the like.

Besides glass, many materials such as synthetic rock (synrock) and the like have been studied as a solidifying medium. In any case, the fission products are incorporated in a large amount of the solidifying medium. The concentration of the fission products in the solidifying medium is limited to about 10% by weight from the viewpoint of problems such as the solubility of the fission products in the medium, chemical durability (leaching rate in water), and removal of decay heat.

An element which brings about a problem with respect to the solubility of the fission products in a glass is Mo (molybdenum). The molybdenum is soluble only by about 3% by weight in terms of MoO3, and Mo remaining undissolved forms a substance called "yellow phase" and inhibits the production of a homogeneous vitrified product. For this reason, the Mo content.in the vitrified product should be about 2% by weight or less in the form of MoO3 from the viewpoint of safety factor. Mo is one of major elements contained in the fission products and amounts to about 10% by weight of the fission products. Therefore, the content of the whole fission products in the vitrified product is limited to 15% by weight in view of the limited amount of Mo.In respect to the decay heat, the vitrified product should be maintained at a temperature not above 5500C which is the transition point of the vitrified product, in order to prevent the deterioration of the vitrified product during the storage. The allowable value of the calorific value per a unit of the vitrified product is about 2.5 kW. For this reason, in the production of the conventional vitrified product, the total content of the fission products. is limited to about 10% by weight of the vitrified product.

Amoung others, Cs (cesium) is a major element participating in the generation of decay heat, and its calorific value amounts to about 40% of those of the whole fission products.

Meanwhile, the high-level radioactive waste contains therein platinum group elements (Ru, Pd and Rh) which are useful but poor in natural resources. Various attempts have been made to recover these elements for years, and representative examples of the method known in the art include a solvent extraction method wherein the platinum group elements are separated from a nitric acid solution of a high-level- radioactive waste by using a phosphoric ester.

The volume of the solidified product should be as small as possible for the purpose of reducing the cost for storage and disposal thereof. For this purpose, it is necessary to increase the content of the fission products in the solidified product. At the present time, however, it is difficult to increase the content of the fission products due to the above-described restriction by the Mo content and the decay heat.

The prior art method of recovering platinum group elements by the solvent extraction method have the following drawbacks. In the solvent extraction method for recovering platinum group elements, the phosphoric ester becomes a secondary waste which is different in kind from the solvent for extraction used in the reprocessing, i.e., TBP (tributyl phosphate). This makes it necessary to employ a novel method for processing waste solvent which is defferent from the processing method used in the treatment of waste TBP.

The cost required for the research and development on this novel processing method and the construction and operation of a processing plant is so high that the cost of the recovery of the platinum group elements is increased over that of the commerically available platinum group elements, and therefore the conventional solvent extraction method does not pay well. Further, a highly volume-reducing treatment of the high-level radioactive waste cannot be accomplished due to the generation of a large amount of a secondary waste.

A heat-melting treatment without the above-described problems has been proposed as a method of volume-reducing solidification of the residual fission products after the separation and recovery of platinum elements. This method comprises calcining a high-level radioactive waste, heating the calcined material up to about 30000C to remove volatile elements, such as Cs, and to reduce platinum elements, etc.

into a metallic form to settle the metal as a molten lower layer, separating the lower layer from other molten fission product elements left in the form of an oxide and bringing the molten oxide of the fission products into a crystalline ceramic solidified waste with a highly reduced volume (see British Patent application No. 9001722.9). Since the resultant crystalline ceramic has a crystalline structure, the properties thereof are determined by a crystalline grain structure formed in the step of cooling the melt. Therefore, in order to produce a solidified product having a uniform quality, it becomes necessary to strictly control the cooling step. This makes the operation troublesome, and an additional step of examining the quality or the like of the solidified product should be provided.

An object of the present invention is to eliminate the above-described problems of the prior art and to provide a method of volume-reducing vitrification of a high-level radioactive waste which can facilitate the recovery of platinum group elements without additionally generating a large amount of a secondary waste, and can accomplish a highly volume-reduced vitrification of a high-level radioactive waste and the production of a vitrified product having a stabilized quality without a strict control of the production conditions.

In the present invention, the above-described conventional heat-melting treatment is further improved and developed. According to the present invention, there is provided a method of volume-reducing vitrification of a high-level radioactive waste comprising: subjecting a highlevel radioactive waste to a calcination and vaporization treatment to produce a calcined material; heating the calcined material in the presence of a reducing agent of boron or a boron compound and a vitrifying agent of an oxide of boron in a reducing atmosphere at a temperature of 10000C or above to melt the calcined material and to bring platinum group elements and molybdenum contained in the calcined material into a metallic form; separating and recovering the resulting molten metal from a molten layer of residual oxides; and solidifying the molten layer of residual oxides to obtain a vitrified product with a'highly reduced volume.

Referring to the accompanying drawings: Fig. 1 is an explanatory diagram of the process according to the method of the present invention.

Fig. 2 is a schematic view of one embodiment of an apparatus used for practicing the present invention.

Fig. 3 is a schematic view of another embodiment of an apparatus used for practicing the present invention.

The present inventors have found that in the above-described conventional heat-melting treatment of a calcined material of a high-level radioactive waste, the presence of a suitable amount of a reducing agent of boron or a boron compound and a vitrifying agent of an oxide of boron enables the separation of Mo and platinum group elements and the production of a vitrified product with a highly reduced volume to be simultaneously accomplished.

These findings have led to the completion of the present invention.

A high-level radioactive waste is usually in the form of a nitric acid solution thereof obtained as an extraction residue in the step of reprocessing spent fuels and contains almost all of the fission products present in the spent fuels. In the present invention, as shown in Fig. 1, the high-level radioactive waste is heated, e.g. at 100 - 150 or to evaporate water and nitric acid and further heated at a higher temperature, e.g. at 500 - 1000 oC, thereby obtaining a calcined material. At that time, Cs which is a highly calorific element volatilizes. A reducing agent and a vitrifying agent are added to the calcined material, and the resultant mixture is heat-melted in a reducing atmosphere at 10000C or above.This causes Mo and platinum group elements contained in the calcined material to be reduced into a metallic form which settles and can be separated from a molten layer of residual oxides. A vitrified product with a highly reduced volume is produced by the solidification of the molten oxide layer. The reducing agent of boron or a boron compound and the vitrifying agent of an oxide of boron may be added in the melting step, as described above.

Alternatively, they may be previously added to the high-level radioactive waste or added in the step of calcination/ vaporization.

Examples of the boron compound as the reducing agent include boron nitride, boron carbide and the like. Boron nitride is most suitable because it is easy to handle, inexpensive and provides boric acid as a reaction product.

The use of boron or a boron compound in an amount of 10% by weight or less in terms of boron as a simple substance based on the calcined material will suffice. The addition of boron or a boron compound in a larger amount brings about an increase in the amount of waste and therefore is unfavorable. The'amount is preferably 5% or less. In the present invention, the formation of an eutectic crystal is most desirable for lowering the melting point of the platinum group alloys. However, a similar effect can be attained even when boron is added in an amount of 0.5%.

Therefore, the amount of addition of boron may be 0.5% or more, preferably 1% or more.

The reduction state in the heat-melting treatment of the high-level radioactive waste is controlled by the temperature, atmosphere and addition of a reducing agent.

The heating temperature is 10000C or above. When the heating temperature is below 10000C, Ru and Mo cannot be reduced into metallic state although Pd and Rh can be reduced. The temperature is thus preferably 15000C or above. Since Ru-, Pd-, Rh-, Mo- and B-base alloys melt at 20000C or below, there is no need to employ a temperature above 20000C. The control of the atmosphere is conducted for the purpose of accelerating the reduction reaction. In the present invention, the reaction is preferably carried out in an atmosphere of air having a reduced oxygen content, nitrogen or argon. In this case, a gaseous reducing agent such as hydrogen or carbon monoxide, and a reducing agent such as carbon or the like which gasify in a redox reaction may also be used.It is also possible to use substances such as metals, carbides and nitrides of aluminum and silicon, which do not have any adverse effect on the residual oxide phase as a waste even when they remain as an oxide. The conditions of the above-described temperature, atmosphere and reducing agent are properly combined with one another depending upon the reaction conditions.

Examples of the oxide of boron as the vitrifying agent include boron oxide, borosilicate glass, sodium borate and the like. Boron oxide and borosilicate glass are most suitable because they are easy to handle and inexpensive.

The amount of addition of these vitrifying agents is 15 to 85% by weight based on the resulting vitrified porduct. The use of the vitrifying agent in an amount exceeding the above-described range is unnecessary from the viewpoint of forming a vitrified product with a highly reduced volume.

Even when the amount of addition of the vitrifying agent is 85%, a remarkable reduction in the volume of the vitrified product can be attained as compared with the conventional vitrified product, since Mo, platinum group elements, etc., are removed from the fission products In the addition of the vitrifying agent, it is also possible to simultaneously add a silicon compound or an aluminum compound commonly used in the glass industry. The addition of these compounds contributes to an improvement in the properties of the resulting vitrified product.

The fission products present in spent fuels are broadly classified into (1) alkali metal elements, (2) alkaline earth metal elements, (3) rare earth elements, and (4) transition metal elements (including platinum group elements). A highly calorific element, Cs, which is an alkali metal element (1) is removed by heating the hihg-level radioactive waste.As a result, in the case of spent fuels of 45000 MWD/MTU in the burnup and five years in the cooling time, major components of the calcined material except for elements having a content 100 g/MTU or less are as follows: alkaline earth metals (Sr, Ba) ,,...3.3 kg/MTU 8.7 %- transition metals (Zr, Mo, Tc) ..... 10.5 kg/MTU 27.9 % platinum group elements (Ru, Rh, Pd) .....5.4 kg/MTU 14.3 26 rare earth elements (Y, La, Ce, etc.) .....18,5 kg/MTU 49.1 % Total 37.7 kg/MTU Platinum group elements are separated and recovered by further heat-melting the calcined material in the presence of a reducing agent of boron or a boron compound. It is known that the platinum group elements have a small free energy of the formation of an oxide thereof and are reduced into a metallic state when being heat-melted. The melting points of the platinum group elements are 15540C for Pd, 19630C for Rh, and 22540C for Ru. Ru and Rh are not completely dissolved in each other because they are different from each other in the crystal form. No alloy having a eutectic point is formed between Pd and Rh or Ru.

Therefore, in the platinum group element and its alloy system, the melting point often exceeds 20000C. It is difficult to separate the platinum group element alone or in the form of an alloy from the residue in the form of an oxide by melting of the calcined material. In other words, even when they can be separated as a phase, a very high melting temperature is required for separating the two layers from each other in the molten state. Mo in the calcined material has a relatively small free energy of the formation of an oxide and forms a low-melting alloy with a platinum group element. Since, however, the contents of Mo and platinum group element in the fission products are dependent on the burnup or the like of the spent fuel, it is difficult to realize a.composition having the lowest melting point in the respective alloy systems comprising four components.

In the present invention, a reducing agent of boron or a boron compound is added. This leads to the formation of an alloy of Mo or a platinum group element with boron which melts at a low temperature. In general, numerous elements (M) combine with boron (B) to form an M/B or 2M/B compound.

This compound forms a eutectic crystal together with the element (M). The melting point of the eutectic crystal is much lower than those of the constituent elements. Since the atomic weight of boron is as small as about 11, the weight content of boron at a eutectic point with other element is 5% by weight at the highest. Therefore, the amount of boron to be added for the purpose of lowering the melting point of the platinum group elements and Mo may be very small. Thus, the platinum group elements and Mo are reduced at a temperature of 20000C or below into an easily meltable form. Since this molten alloy has a specific gravity larger than that of the residual oxides, it is not homogeneously dispersed but settles and separates at the bottom of a melting oven, so that the platinum group element can be recovered.

Further, in the present invention, a vitrifying agent of an oxide of boron is added. For this reason, the residual oxides are in the form of a glassy melt which can be solidified into a vitrified product. In this vitrified product, since Cs and Mo are separated and removed, the content of the fission products remaining in the glass is not limited to about 10% by weight which is a limitation of the content in the case of the conventional vitrified product of a usual high-level radioactive waste and may be up to about 80% by weight. In this case, it is thus possible to obtain a vitrified product having a remarkably reduced volume. In the case of the conventional vitrified product, the weight becomes 10 times as large as that of the fission product, and the vitrified product amounts to several hundreds of liters per ton of the spent fuel.On the other hand, in the method of the present invention, the volume of the vitrified product is several tens of liters.

Thus the present invention enables the reprocessing of spent fuels by purex process to be effectively conducted without generating any secondary waste other than the solid waste.

Fig. 2 is a schematic view of one embodiment of an apparatus for practicing the method of the present invention. This apparatus exemplifies a bottom flow type apparatus. A calcined high-level radioactive waste is thrown into a melting vessel 10. The calcined waste is melted and reduced under heating and separates into a molten layer 12 of platinum group element alloys having a higher specific gravity and a molten layer 14 of oxides having a smaller specific gravity. The platinum group element alloy layer 12 and the oxide layer 14 successively flow down through a flow-down nozzle 16 into another vessel 18 for solidification. The oxide layer 14 is solidified into a vitrified product.

Fig. 3 is a schematic view of another embodiment of an apparatus used for practicing the method of the present invention. This apparatus exemplifies an intermediate type apparatus comprising a combination of the overflow type apparatus with the bottom flow type apparatus. A calcined high-level radioactive waste is thrown into a melting vessel 20 from the top thereof and heat-melted. The melt separates into a platinum group element layer 12 located in the lower part and an oxide layer 14 located in the upper part. The platinum group element layer 12 flows down through a flow-down nozzle 22 located therebeneath and is received by a receiver 24 for metal and solifified. The oxide layer 14 overflows so as to pass through a flow path 26 shown by an arrow, flows down through a flow-down nozzle 28 into a receiver 30 for glass for vitrification.

For the calcination of the high-level radioactive waste, a rotary kiln system, a microwave heating system, etc., which are under research in relation to vitrification, can be used. For the heat treatment of the calcined waste, a heater system, a direct energization system, a high-frequency heating system, etc., may be employed.

Particular Experimental Examples will now be described.

Experimental Example 1 The composition of fission products contained in a spent fuel of 45000 MWD/MTU in the burnup and 5 years in the cooling time was calculated by using ORIGEN code to prepare a simulated waste solution of the corresponding high-level radioactive waste solution. The simulated waste solution was heated to 6000C to prepare a calcined material. 45 g of the calcined material was put in a crucible, and 5 g of boron nitride and 10 g of boron oxide were added thereto.

The mixture was heat-melted in an argon atmosphere at 18000C for one hour. After cooling, the contents in the crucible were observed and found to have a smooth surface and be in a glassy state. The crucible was broken and the contents were taken out. The contents were separated into two phases, and a metal mass was present in the bottom and could easily be separated from the residual vitrified portion. The metal mass was analyzed with an X-ray microanalyzer (EPMA). As a result, Ru, Rh, Pd and Mo were detected. The vitrified portion was subjected to the measurement of the leaching rate in water according to JIS R 3502. The leaching rate was 7 x 10-5 g/cm2.d and substantially the same as that of the conventional vitrification product.Thus it has been confirmed that the vitrified portion has a chemical durability sufficient as a high-level radioactive solid waste.

Experimental Example 2 The simulated high-level radioactive waste was treated in the same manner as that of Example 1 by putting 45 g of the calcined material in the crucible and adding thereto 5 g of boron nitride and 50 g of borosilicate glass. The heating temperature was 15000C. The results of observation after the treatment was the same as those of Experimental Example 1.

As has been described above, the method of the present invention comprises heat-melting a calcined high-level radioactive waste in a reducing atmosphere at a high temperature of 1000or or above in the presence of a reducing agent of boron or a boron compound and a vitrifying agent of an oxide of boron. This method makes it possible to separate and recover useful platinum group elements and Mo, simplify the treatment process and reduce the size of an apparatus for the treatment. Further, since Cs as a highly calorific element is removed by the calcination, the limitation on the content of the fission products in the vitrified product is eliminated. Further, since Mo is reduced and separated, the limitation on the solubility of the fission products with respect to the content of the fission products can be eliminated.Further, since the resulting residue of oxides is vitrified as it is, the vitrification is accompanied by such a remarkable volume reduction that the volume is below one-tenth of that of the conventional vitrification. This enables the cost of storage and disposal of the high-level radioactive waste to be remarkably reduced.

In the present invention, since an oxide of boron is added and the residual waste is vitrified, it becomes unnecessary to strictly control the production conditions as opposed to the formation of a crystalline ceramic in the conventional heat-melting treatment herein-before described, so that a homogeneous vitrified product having a stable quality can be easily produced at a high efficiency.

Further, in the present invention, the heat-melting can be conducted at a temperature of 20000C or below because boron or a boron compound is added to the calcined waste.

Therefore, it becomes possible to adopt a heat-treatment wherein heating is conducted with a heater without the necessity for using a special heating system (e.g., electron beam heating, plasma heating, etc.), and the material of the melting furnace may be zirconia, etc., without the necessity for using special high-melting materials (e.g., thorium oxide), which enables the facilities for treatment to be easily constructed at a low cost.

Claims

1. A method of volume-reducing vitrification of a high-level radioactive waste comprising: subjecting a high-level radioactive waste to a calcination and vaporization treatment to produce a calcined material; heating the calcined material in the presence of a reducing agent of boron or a boron compound and a vitrifying agent of an oxide of boron in a reducing atmosphere at a temperature of 10000C or above to melt the calcined material and to bring platinum group elements and molybdenum contained in the calcined material into a metallic form; separating and recovering the resultng molten metal from a molten layer of residual oxides; and solidifying the molten layer of residual oxides to obtain a vitrified product with a highly reduced volume.

2. The method according to claim 1, wherein the heating step is carried out at a temperature ranging from 10000C to 20000C.

3. The method according to claim 1, wherein the reducing agent is boron nitride.

4. The method according to claim 1, wherein the reducing agent is used in an amount of 0.5 to 10% by weight in terms of boron as a simple substance based on the calcined material.

5. The method according to claim 1, wherein the vitrifying agent is boron oxide or borosilicate glass.

6. The method according to claim 1, wherein the vitrifying agent is used in an amount of 15 to 85% by weight based on the vitrified product.

7. The method according to claim 1, wherein a silicon compound or an aluminum compound is added to the calcined material in addition to the reducing agent and the vitrifying agent.

8. A method of volume-reducing vitrification of a high-level radioactive waste, substantially as herein described and as illustrated in the accompanying drawings.