JPS61176888A - Pre-treatment method of light water reactor spent nuclear-fuel - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

[Detailed Description of the Invention] [Industrial Application Field] The present invention relates to a pretreatment method for separating spent nuclear fuel from a light water reactor into fuel cladding tubes and fuel pellets. The fuel cladding is heated to oxidize, and mechanical force is used to generate the buds. r light eggplant t /7'l a is there.

[Prior art] In the reprocessing of spent nuclear fuel, prior to the main process of separating and recovering unburned fissile materials and newly generated fissile materials, they are first decladded and stored in the fuel cladding tube and its interior. It is necessary to separate the fuel pellets that are Conventionally, mechanical methods and chemical methods have been used to de-clad the fuel cladding of spent nuclear fuel.

As a mechanical decoating method, there is a so-called "shear leach method" in which spent nuclear fuel is cut into several lengths while still being coated, and only the nuclear fuel is leached and dissolved in nitric acid, and this method is widely used.

On the other hand, the chemical decoating method is a method in which the entire spent nuclear fuel is immersed in a solution to dissolve it all, and then separated.

[Problems to be solved by the invention] In the chemical decladding method, in principle, all of the spent nuclear fuel is dissolved in the solution as described above, so a large amount of fuel cladding components are included. Therefore, only the components of the cladding tube must be separated after melting, which is very cumbersome.

On the other hand, the mechanical decoating method has the advantage that it reduces loss of nuclear fuel and generates less waste liquid than the chemical decoating method and is economically superior.

However, since the cut fuel is directly chemically dissolved, volatile nuclides are dissolved in the dissolution tank, resulting in various problems. Furthermore, since the gas generated from the dissolution tank is accompanied by acid, it becomes difficult to treat the exhaust gas such as recovering tritium, krypton, and xenon. Furthermore, the cladding tube, which is a dissolution residue, must be treated in a separate process.

As described above, conventional technology involves difficult problems such as spent nuclear fuel decladding, cladding treatment, exhaust gas recovery, etc., and new technology is needed to solve these problems and facilitate chemical dissolution in the main process. At present, development is strongly desired.

The purpose of the present invention is to eliminate the above-mentioned drawbacks of the prior art, to easily perform decladding of spent nuclear fuel, cladding treatment, exhaust gas recovery, etc. in a dry state, and to improve the subsequent main reprocessing process. The object of the present invention is to provide a method for preprocessing spent nuclear fuel in a light water reactor that is efficient and practicable, and that can prolong the life of the equipment used in the main process.

[Means for Solving the Problems] The invention that can achieve the above-mentioned purpose is to heat light water reactor spent nuclear fuel in the presence of oxygen, for example, in air to oxidize the fuel cladding, and then This is a pretreatment method in which force is applied to destroy the fuel cladding tube and separate it from the fuel pellets inside.

In light water reactor fuel, the fuel cladding is made of Zircaloy (zirconium alloy). Zirconium has a small thermal neutron absorption cross section and is suitable as a nuclear reactor material, but it tends to corrode in high-temperature pure water and is highly reactive with nitrogen in the air. Zircaloy was created by adding a small amount of tin to remove the harmful effects of nitrogen. Most of Zircaloy is Zr.
(zirconium), it becomes ZrO3 when heated in the presence of oxygen.In particular, it is easily oxidized by repeating a thermal cycle of increasing and decreasing the temperature several times in air or in an oxygen-enriched gas atmosphere.

As a result of experiments, it was found that the oxide thus obtained exhibited thermal expansion behavior similar to pure zirconia (zirconium oxide).

The coefficient of thermal expansion of pure zirconia is approximately 1, as shown in Figure 1.
Irregularities appear at 200°C, expansion stops and contraction begins. Thermal expansion becomes zero at about 1100°C, and when cooled to room temperature, the coefficient of thermal expansion becomes about 0.2%. Such a large difference in coefficient of thermal expansion means that cracks occur and fracture becomes easy. This can be seen from the fact that when it is heated again, it expands, but reproducibility cannot be obtained.

The oxidation causes many fine cracks in the fuel cladding, making it very brittle, and the fuel pellets and cladding material can be easily separated by applying a small amount of mechanical force.

[Operation] According to the present invention, as described above, the fuel cladding material and the fuel pellets can be easily separated without being chemically dissolved. Because the present invention includes a heating step, which liberates the gases in the fuel, exhaust gas recovery can occur simultaneously with stripping.

If the spent nuclear fuel is heated and roasted after decoating and before being dissolved in acid for the reprocessing process, volatile substances and gases mainly caused by nuclear fission contained in the spent nuclear fuel can be completely separated and recovered. , corrosion of equipment in the main process is reduced.

Moreover, this exhaust gas separation and recovery process is a dry process, so it can be easily performed. By further crushing the obtained fuel pellets, chemical dissolution in the main process can be carried out easily and quickly.

[Example] The present invention will be explained in more detail below.

As mentioned above, the present invention has the potential to accumulate spent nuclear fuel in a light water reactor.
The major feature of this method is that after the fuel injection, mechanical force is applied to destroy only the fuel cladding. In light water reactor fuels, the fuel cladding is typically made of Zircaloy, which can be heated to melting temperature, e.g.
If a thermal cycle treatment of heating and cooling from 00 to 1200°C is performed, it will easily oxidize and the material will change drastically. The oxidized Zircaloy becomes very brittle and can be easily crushed and separated from the fuel pellets with the application of slight mechanical force.

FIG. 2 is a process explanatory diagram showing an example of a pretreatment process for spent nuclear fuel in a light water reactor according to the present invention. The light water reactor spent fuel assembly 10 is sent to a cutting process, where stainless steel members 12 such as the upper and lower tie plates other than the fuel pins are removed as necessary, and the assembly is cut to a size that is easy to handle in the next thermal decladding process. Shredded. The obtained fragments 14 are subjected to a thermal cycle treatment in which the temperature is raised and lowered from 700 to 1200° C. in the air in a thermal decoating step. As a result, the zircaloy that makes up the fuel cladding tube oxidizes and turns into zirconium oxide (zirconia). It is desirable to apply the heating/cooling thermal cycle multiple times. The oxidized fuel cladding tube has many thin cracks and becomes very brittle, and easily peels off from the fuel pellet and becomes powder when a small amount of mechanical force is applied. The particle size of powdered zirconia (zirconia sand) varies depending on the thermal cycle conditions applied to the object to be processed, but is distributed within a range from powder to about 5 to 6 mm. The treated product 16 removed by the thermal decoating process as described above is therefore a mixture of pellets consisting primarily of uranium dioxide and zirconia sand.

The processed product 16 is sent to the next separation step. Note that before separation, roasting and pulverization treatments can be performed as necessary.

In order to separate the treated product 16 into the zirconia sand 18 and the fuel pellets 20, a specific gravity separation method or a sieving method can be used. In the case of uranium dioxide fuel pellets, the specific gravity is 10 to 11, whereas the specific gravity of zirconia is 5.68 to 6.27, which is a large difference. If zirconia is added, the lighter zirconia will float and the heavier fuel pellets will sink, allowing them to be separated.

Further, stainless steel members such as a rod inner spring and Inconel members such as an expansion spring used in the fuel pin can also be separated to the zirconia side.

Further, as mentioned above, the oxidized fuel cladding tube is crushed quite finely, whereas the fuel pellets are strong and do not get crushed under conditions where the fuel cladding tube is oxidized, but maintain their original shape. Therefore, when the treated product 16 is sieved, mainly the powdered zirconia sand 18 passes through, leaving mainly the fuel pellets 20 on the sieve.

The separated zirconia sand 18 is then sent to a waste treatment process, and the fuel pellets 20 are sent to a melting and refining process.

Figure 3 is a conceptual diagram showing an example of thermal decoating equipment.
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It includes the heating coil 24, a temperature sensor 26 that measures the temperature inside the furnace, and a control device 28 that controls the operation of the heating coil 24 according to the temperature information from the temperature sensor 26 and a preset temperature program. There is. Further, the heating furnace body 22 is provided with an air supply vent @30 and an exhaust gas passage 32.

Objects 34 to be processed, such as spent nuclear fuel pins and their fragments, are housed within the heating furnace body 22. The inside of the heating furnace body 22 is heated by a heating coil 24, and the temperature inside the furnace is measured by a temperature sensor 26.
The temperature information is sent to the control device 28. The control device 28 controls the amount of electricity supplied to the heating coil 24 according to the temperature information and a preset temperature program, and performs necessary temperature increase/decrease control. For example, 700℃
An operation is performed in which the temperature is repeatedly raised and lowered several times over a temperature range from 1200°C to 1200°C. During this time, air supply #
From the heating furnace body 22, oxygen necessary for oxidizing the workpiece 34 contained in the heating furnace body 22 is sent as air or oxygen-enriched gas into the air.

Gas containing volatile nuclides, tritium, etc. released from the object to be treated 34 by heating is sent to the exhaust gas treatment system through the exhaust gas passage 32.

As described above, in the present invention, it is possible to remove the coating very easily in a dry state. Therefore, if the obtained nuclear fuel pellets are heated and roasted before being sent to the main process such as extraction and separation, the volatile substances and gases contained therein, mainly caused by nuclear fission, can be completely separated and recovered in a dry process, and the exhaust gas Not only is the treatment easier, but the exhaust gases can be removed in advance, which reduces corrosion of equipment during the main process. Furthermore, if the decoated fuel pellets are finely pulverized in advance, chemical dissolution in the main process can be carried out easily and quickly.

Effects of the Invention J The present invention is a method for pre-processing spent nuclear fuel configured as described above, in which the fuel is removed by heating, oxidizing and mechanically destroying it, rather than by chemically dissolving the fuel. An excellent effect can be achieved in that the treatment of the 17 fuel cladding tube, which can be easily stripped, is also facilitated.

The present invention also has the effect that exhaust gas treatment can be performed relatively easily because exhaust gas containing volatile components can be recovered in a neutral and dry state.

In other words, since it is different from the so-called wet method that uses chemicals to dissolve, such as the conventional shear leach method or chemical uncoating method, it generates less waste (it also has the effect of significantly reducing pretreatment costs). be.

In addition, since the decoated fuel pellets are in a solid state, they can be roasted to separate volatile components and gases, or they can be crushed to an appropriate particle size and used in the main reprocessing process. The present invention has a number of advantages, including ease of chemical dissolution.

[Brief explanation of the drawing]

Figure 1 is an explanatory diagram showing the behavior of #i expansion of pure Zircaloy,
FIG. 2 is a process explanatory diagram showing an example of the pretreatment process according to the present invention, and FIG. 3 is a conceptual diagram showing an example of a #l decoating apparatus suitable for use therein. 22 heating furnace main body, 24... heating coil, 26 degree sensor, 28... control device, 34... processed object.

Claims (1)

[Claims] 1. After heating the light water reactor spent nuclear fuel in the presence of oxygen to oxidize the fuel cladding, mechanically destroying the fuel cladding,
A method for preprocessing spent nuclear fuel in a light water reactor, which is characterized by separating it from internal fuel pellets. 2. The pretreatment method according to claim 1, wherein the fuel cladding tube is made of Zircaloy and the fuel pellets are made of uranium oxide or uranium-plutonium mixed oxide.