JPS6182200A - Method and device for solidifying and treating radioactive waste - 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 [Technical Field of the Invention] The present invention relates to a solidification treatment method and a solidification treatment apparatus for low-level radioactive concentrated waste liquid containing sodium nitrate generated in a nuclear fuel reprocessing facility.

[Technical Background of the Invention P1] Generally, spent nuclear fuel is generated as a result of the operation of a nuclear power plant.
It contains valuable substances such as Np, which can be used as fuel for fast breeder reactors, and these valuable substances are recovered at spent nuclear fuel reprocessing facilities.

``°U'' or ``pu'' from spent nuclear fuel is usually recovered by a method such as the purex method.

This method involves dissolving spent nuclear fuel in nitric acid and
U or "pu" is ionized, and this solution is brought into contact with a tributyl phosphate (hereinafter referred to as TBP)-n dodecane mixed organic solution to be recovered.

In other words, the ions of ``''() or 12@p u can easily form a complex with TBP depending on the temperature, acid concentration, and valence. transfer and further contact with water,
1°゛U or Ill p by adjusting the valence and acid concentration
U is recovered by back extraction.

The recovery rate can be increased by performing the above process in several stages in series.

This reprocessing process generates waste liquid containing TBP degraded by 1irl acid and chemical radiation irradiation, and the treatment of these waste liquids is divided into high-level radioactive waste liquid and medium to low-level radioactive waste liquid. They are treated differently as follows:

That is, high-level radioactive waste liquid is mainly waste liquid from the spent nuclear fuel melting process and the first stage extraction process, and as is well known, it is mixed with molten borosilicate glass and vitrified.

On the other hand, medium to low-level radioactive waste liquid contains low concentrations of radioactivity, nitric acid, and degraded TBP.

To prevent corrosion of pipes, etc., this radioactive waste liquid is made neutral or alkaline by adding sodium hydroxide to convert nitric acid into sodium nitrate, and then evaporated 111i!
The volume is reduced by a container and stored in a low-level radioactive waste liquid storage tank.

It is said that when 1 ton of spent nuclear fuel is processed, 500 yn' of low-level radioactive waste liquid is generated, and safe processing of these low-level radioactive waste liquids is desired.

As methods for treating this low-level radioactive waste liquid, currently being researched are the asphalt solidification method and the plastic solidification method.

In the desfunflu 1-solidification method, low-level radioactive waste liquid or dry powder of low-level radioactive waste liquid is fed into an askrut heated to 100 to 200°C to make it fluid,
This method fixes radioactivity in asphalt.

On the other hand, the plastic solidification method is a method in which low-level radioactive waste liquid is dried and solidified, often using thermosetting resin. All of these methods have the advantage that the water in the radioactive 13Fia evaporates, resulting in greater volume reduction.

[Problems with the Background Art] However, both the asphalt solidification method and the plastic solidification method were developed for the solidification of low-level radioactive waste fluid generated from nuclear power plants, and this method is not suitable for nuclear power plants. Manri reprocessing/+l! If this method were to be applied as is to the I liquid generated from the plant, the following problems would arise.

In other words, although the mechanical and chemical properties of the solidified bodies produced by these asphalt solidification methods or plastic solidification methods satisfy all the conditions required for radioactive waste solidification 141, Since the solidified material is an organic material, there is a problem that the solidified material deteriorates if kept at high temperature for a long time.

On the other hand, when looking at the durability of a solidified material from the type of radioactivity contained in the solidified material, the main nuclide in low-level radioactive waste fluid generated from nuclear power plants is CO, and its half-life is approximately 5 years. be.

Therefore, if the solidified material is managed for 50 years, the radioactivity will decrease by 1/2.
In 1000.100 years, it becomes 1/1000000.

On the other hand, most of the radioactivity contained in the radioactive waste fluid generated from nuclear fuel reprocessing FM facilities is fission products, including those with extremely long half-lives. for example,
'The half-life of C3 is 30 years, so the radioactivity can be reduced to 171
300 years to reach 000, 1/1000000
It took 600 years to complete this process. Therefore, over such a long period of time, it cannot be guaranteed that there will be no crustal deformation or heating of the solidified material due to fire, and therefore, long-term strict management is required, which requires a huge amount of expense.

In addition, a general problem with the asphalt solidification method and the plastic solidification method is that both require only drying treatment as pretreatment of waste, and there is no process for the waste itself to decompose. There is a problem in that it is inconvenient when volume reduction is required.

[Objective of the Invention 1] The present inventors conducted 1:1 research in order to eliminate the drawbacks of such asphalt solidification method or plastic solidification method, and as a result, the present inventors have developed a method for handling release QJ materials such as nuclear fuel reprocessing facilities. 61!1M generated at the facility, containing sodium
It has been found that by drying the A4 shrinkage waste liquid and heating and melting it together with a vitrification material, a vitrified material with a high volume reduction rate and excellent mechanical and chemical properties can be obtained.

In other words, low-level radioactive waste fluid generated from nuclear fuel reprocessing facilities contains, in addition to sodium 5Flate, trace amounts of WM sodium and roughly trace amounts of sodium sulfate. The melting points of these sodium compounds are 308
'C is 1852°C and 884°C, respectively, and becomes a molten salt above these temperatures. Each of these sodium compounds is a material used in the glass industry, and is decomposed and vitrified by reaction with silicon oxide.

This vitrified material has excellent durability and is ideal for containing radiation for a long period of time.

The present invention has been made based on this knowledge, and an object of the present invention is to provide a solidification treatment method and a solidification treatment apparatus for radioactive waste generated from nuclear fuel reprocessing facilities and the like.

[Summary of the Invention] That is, the solidification treatment method for solid waste of the present invention involves drying radioactive concentrated waste liquid containing sodium nitrate generated at a nuclear fuel reprocessing facility to make it into a baby, and combining this dry powder with glass. It is characterized by heating and melting the vitrified material, processing the generated pyrolysis gas and discharging it outside the system, and filling the remaining vitrified material into a storage container.The solidification processing equipment is also used for nuclear fuel reprocessing. A radioactive waste liquid na treatment device that dry powders radioactive waste liquid containing sodium nitrate generated in the facility and mixes it with a vitrification material at a predetermined ratio, and the pretreatment 1! l! A glass melting furnace that heats and melts the mixture of dry powder and vitrification material produced in the apparatus, and a glass melting furnace that processes the pyrolysis gas generated in the glass melting furnace and disposes it as waste waste outside the system.
The 11th characteristic is that it is equipped with an off-gas treatment device for discharging 1, and a vitrified material filling device for filling a storage container with the radioactive waste vitrified in the glass melting furnace.

The method for solidifying radioactive waste of the present invention will be described in further detail below.

In the method of the present invention, first, a concentrated waste liquid containing helium (b1-1 acid) generated at a nuclear fuel reprocessing facility is first dried by a known vertical thin film drying method (92) and turned into powder.
rl! II! The dry powder containing disodium and the vitrification material are heated at 8 degrees.

The radioactive waste containing the above-mentioned sodium chloride also contains trace amounts of sodium carbonate and sodium sulfate in addition to sodium nitrate.

Further, as the vitrification material, any material can be used as long as it can produce a vitrified material with high R-region strength and water resistance, but silicon dioxide is particularly suitable.

In the presence of silicon dioxide, 61' 1M sodium, sodium carbonate and sodium sulfate are respectively (1
) to (3), it becomes soda-silicate glass and vitrifies.

2Na NO3-+2Na NO2+022Na NO
2-+Na 20-1-N2 +1/2 02Na 2
Q+3i 02-+Si 02 ・Na 2 Q...
・(1) N82 CO3-+Na 20+CO2, Na2O+5
102→5ioz・Na2O...(2) Na 2804 +Si 02-*Si 02 ・Na
20+SO2+Q...(3) Among the reactions of (1) to <3), (1) and (2)
) The reaction proceeds easily at about 1200°C.

In addition, in reaction (3), if the amount of sodium sulfate is large relative to the total 1-lium compound, it is necessary to add carbon to about 5% of sodium 11 fltiate (written by Naruse, Glass Industry, 1981, published by Kyoritsu Shuppansha) ).

However, if it is less than 5% of the total alkaline compounds (
The reaction 3) proceeds even without carbon; sodium sulfate decomposes into sodium oxide, which reacts with silicon dioxide and vitrifies.

Further, when boric acid, aluminum hydroxide, and calcium carbonate are used in combination with silicon dioxide as a vitrification agent, the vitrification reaction can proceed smoothly.

In other words, if only silicon dioxide is used as a vitrifying agent, the viscosity of the glass will increase, making it difficult to form a homogeneous glass, and the melt FIA temperature will gradually increase, but using boric acid will increase the viscosity of the glass. (it decreases and melting becomes even lower).

In addition, the glass formed according to the present invention has a high soda content in order to improve volume reduction properties, and therefore has low water resistance, but boric acid and aluminum hydroxide have the effect of improving the water resistance of the glass. do.

Calcium carbonate also has the effect of reducing the viscosity of the glass and improving its water resistance, and also has the effect of accelerating the decomposition of trace amounts of sodium sulfate contained in the dry powder.

In the present invention, suitable amounts of the dry powder and vitrification material are in the following ranges, for example.

Radioactive waste 1 liter (2) Dry powder...20-55% silicon oxide...2
5-65% boric acid ・・・・・・・・・・・・・・・
・・・・・・1~20% aluminum hydroxide "Kum...
・・・・・・1~5% calcium carbonate・・・・・・
...... Most of the chemical components contained in the low-level radioactive waste liquid generated from nuclear fuel reprocessing FM facilities, such as 1 to 15% or more, can be used as raw materials for glass as they are.
If radioactive waste liquid is dried to form radioactive powder, and silicon dioxide powder is added to these powders and melted under high pressure, the sodium compounds in the powder will easily decompose and become sodium oxide, reacting with silicon dioxide to form glass. become

In addition, in the present invention, by decomposing the sodium compound itself, gaseous components are released and the product is significantly reduced.
be done. Moreover, l! Since the released components in the IQ4 concentrated waste liquid are trapped in the vitrified material, a highly safe solidified product can be obtained.

The pyrolysis gas generated in the above reaction process is combusted by a conventional method, and after removing the Q = 1 property, it is released outside the system, and is also vitrified solid released Q = J waste material. is filled into a predetermined storage container.

Next, the solidification processing apparatus of the present invention will be explained.

As shown in FIG. 1, the volume reduction solidification processing system of the present invention is comprised of radioactive concentration 1 complete liquid moe treatment system I! ! ! The main parts thereof include an apparatus 1, a glass melting furnace 2, an off-gas treatment apparatus 3, and a solidified material removal apparatus 4.

The radioactive ia condensate liquid pretreatment device 1 is a device for pulverizing the radioactive concentrated waste liquid and mixing it with the vitrification material at a predetermined ratio.
The drying processing equipment is capable of storing the dried powder and discharging any amount of it, a powder storage tank with an angle of 0, and a storage tank capable of storing vitrification material and discharging the desired amount of it. Powder mixing (a
It consists of etc.

Volume reduction and vitrification of radioactive concentrated waste liquid can be carried out by directly feeding the radioactive concentrated waste liquid onto molten glass, but in order to increase the melting efficiency, the temperature of the furnace must be lowered by the evaporation of water in the waste liquid. It is desirable to dry and powder it in advance to prevent it. Drying and powdering in advance in this way makes it easier to measure and mix with powdered glass raw materials, and furthermore, the layer formed by the deposition of powder in the glass melting furnace contains pyrolysis gas. This has the advantage of acting as a filter.

The glass melting furnace 2 is a device that heats a mixture of a vitrification material and a dry powder containing sodium nitrate produced in the radioactive concentrated waste liquid pretreatment device 1 at a high temperature, decomposes it, and vitrifies it. A high-frequency heating device is suitable as a heat source for this glass melting furnace 2, and its power frequency is 10 to 60 cm when the inner diameter of the glass melting furnace 2 is in the range of 20 to 60 cm.
400KHz is suitable.

The high frequency coil is installed at the bottom of the glass)8 melting furnace, and the
A temperature 1σ gradient is generated such that the quality of the liquid decreases upward from pIf hl +.

High-frequency heating equipment is easy to raise the temperature and control temperature = 1-1-1, and has a long lifespan because the high-frequency coil, which is the heating source of the furnace, is completely separated from the radioactive atmosphere by the furnace material. It has the advantage of being easy to maintain.

Sodium nitrate is the main component of radioactive concentrated waste fluid from nuclear fuel reprocessing facilities, and sodium carbonate and ti! are minor components. Sodium chloride is melted in the glass melting furnace 2 to become a good electrical conductor, and the melt itself receives high frequency induction to generate heat and maintain a high temperature of 1100 to 1200°C.

The mixture of dry powder and vitrification material is supplied from the top of the glass melting furnace, and is vitrified according to the reactions shown in equations (1) to (3) described above, stably solidifying the radioactivity in the waste. do.

The pyrolysis gas generated here passes through the packed dry powder and glass mixture and rises, and at this time, this mixture acts as a filter and is contained in the gas generated by the glass reaction. Acts to capture fine dust.

Further, the molten vitrified material is taken out from a dedicated outlet nozzle attached to the lower part of the glass melting furnace in a manner that prevents the molten glass containing radioactivity from scattering.

In addition to the above-mentioned high-frequency heating device, the glass 78 melting furnace used in the present invention may be a melting furnace that is used in the ordinary glass industry, such as one that applies a voltage between electrodes and utilizes the electrical resistance of the molten glass. It is also possible to use an energizing furnace that heats the material by heating it, or an indirect heating melting furnace that energizes a resistance heating element such as a silicon carbide rod or nichrome wire and uses the heat generated to heat and melt the material.

The off-gas processing device 3 is a device that incinerates the combustible thermal decomposition gas generated in the glass melting furnace 2, removes radioactivity, and discharges the gas to the outside of the system. This off-gas treatment device 3 includes a filter chamber for removing radioactivity from pyrolysis gas,
It has a high-performance filter to completely remove radioactivity from the off-gas generated from this, and a measurement system to confirm the presence or absence of radioactivity in the off-gas.

The solid handling device 4 is a device that takes out vitrified radioactive waste and fills it into a storage container, and the glass melt taken out from the takeout nozzle at the bottom of the glass melting furnace 2 is injected into a special mold here. After cooling, it is transferred to a radioactive waste storage facility.

In the solidification treatment 1'l apparatus of the present invention, first, nuclear fuel reprocessing J! 11! Low-level radioactive concentrated waste liquid generated at the facility is dried and solidified in a radioactive waste surface treatment device 1. This dry powder mainly composed of sodium nitrate is charged into the glass melting furnace 2 together with the vitrification material, where it is heated and glassed, and generates pyrolysis gases such as N2, CO2, 02, etc. do.

In addition, since the electrical conductivity of glass is low in (I), it is desirable to add a substance with high electrical conductivity, such as silicon carbide, at the time of startup.

Silicon carbide does not inhibit the thermal decomposition reaction because its groove structure is 511-, which is involved in the vitrification reaction, and the silicon component is incorporated into the glass produced, and the carbon component is 6i W sodium. CO due to active oxygen generated from
It becomes two dregs and is discharged from the system.

Next, the ampholytic sodium, which has been melted and has a low electrical resistance, is further heated by high frequency heating to a temperature of 1000 to 1200'Q.
These are vitrified by keeping them at a high temperature.

FIG. 2 is a sectional view schematically showing the state of the charge in the glass melting furnace 2.

From the hopper installed at the top of the glass melting furnace 2, dry powder mainly composed of sodium nitrate and SHQ 2
A mixture with vitrification material of +7 is charged. □These charges are heated by the pyrolysis gas coming upward from the lower layer side in the upper layer A in the glass melting furnace 2, and are sequentially heated. It thermally decomposes while moving to the lower layer 81, producing N2.CO2,02
Reduce the volume without generating gases such as. Further, the layer 0 is heated by the high-frequency coil 5 to undergo a vitrification reaction, and the layer D, which is the lowest layer, becomes a stable molten glass layer.

Note that the mixture of dry powder and vitrification material in layer A acts as an IQ filter for thermal decomposition residue containing radioactivity and prevents radioactivity from leaving the system.

The pyrolysis gas generated in the glass melting furnace 2'c is sent to the off-gas treatment device 3, passed through a filter chamber to remove radioactivity, and after passing through a high-performance filter, the presence or absence of radioactivity is confirmed by a monitor. and released outside the system.

On the other hand, the vitrified material of layer D is discharged from the lower outlet nozzle 6, injected into a metal canister, solidified, and stored in a radioactive waste storage.

In the figure, reference numeral 7.8 denotes a heater and a cooling pipe for melting or solidifying the vitrified material 9 in the outlet nozzle 6 at any time to take out and seal the vitrified material.

[Embodiments of the invention] An embodiment embodying the present invention will be described below.

FIG. 3 is a diagram schematically showing the configuration of a radioactive waste solidification processing apparatus according to an embodiment of the present invention.

In the figure, reference numeral 11 indicates a waste liquid storage tank that stores concentrated waste liquid containing sodium nitrate generated at a nuclear fuel reprocessing facility.

The radioactive concentrated waste liquid in the waste liquid storage tank 11 is supplied to a vertical thin film dryer 13 by a waste liquid supply pump 12 .

Vertical thin film dryer [13] is widely used in general chemical plants and q/l waste treatment (plastic solidification, pelletization), etc., and has an airtight interior, no leakage, and a simple internal structure. This has the advantage of easy decontamination and inspection.

The discharge port of this vertical thin film dryer 13 is connected to a powder storage tank 15 via a powder transporter 14.
5 is connected to a powder mixing tank 17 via a powder transporter 16.

Further, a raw glass material storage tank 18 such as Si 02 is also connected to the powder mixing tank 17 via a powder transport rock 19 .

In this powder mixing tank 17, by adjusting the supply amount of dry powder from the powder storage At1i tank 15 and the supply amount of vitrification material from the raw glass material storage tank 18, dry powder can be mixed at an arbitrary ratio. It is now possible to adjust the mixture with the vitrification material.

The outlet of the mixed powder mixer 117 is connected to a material input port opening into the upper part of the glass melting furnace 21 via the powder transporter 20.

A high frequency coil 2 for heating is installed at the lower side of the glass melting furnace 21.
2 is wound, and an outlet nozzle 23 for molten glass is opened at the bottom thereof, and this outlet nozzle 23 is connected to a mold 24 for forming a molten vitrified product. Note that 21L is a level meter consisting of a thermal lightning pair for detecting the liquid level of the molten vitrified material.

A pyrolysis gas outlet 25 is provided at the upper side of the glass melting furnace 21, and the pyrolysis gas outlet 25 is connected to a filter chamber 26 of an off-gas treatment device.

In addition, the section from the outlet nozzle 23 to the mold 24 is completely spaced, and the inside pressure is set to f; The generated gas is also filtered to the filter chamber 2.
6 to be processed.

The filter chamber 26 is composed of a ceramic filter, and an off-gas cooling device 27 is provided at the outlet of the filter chamber 26.
, a high-performance filter 28 are connected in this order, and a radioactivity monitor (not shown) is arranged immediately after the high-performance filter 28. 29 is an exhaust fan.

In the radioactive waste solidification processing apparatus of this embodiment, the radioactive concentrated waste liquid containing sodium nitrate in the waste liquid storage tank 11 is first dried by the vertical thin film dryer 13 to form a dry powder.

The concentrated waste liquid is supplied from the inlet of the vertical thin film dryer 13 and is heated and dried to form a powder, which falls onto the powder transporter 14 from the lower powder outlet and is temporarily stored in the powder storage tank 15.

Next, from the powder storage tank 15 and the raw glass material storage tank 18,
Dry raw material and vitrification material are supplied to a powder mixing tank 17 at predetermined ratios, and the mixed materials are sequentially charged into a material input port of a glass melting furnace 21.

Next, the mixture of these dry powders and the vitrifying material is heated by the high frequency coil 22 in the glass melting furnace 21, but when the glass melting furnace 21 is started, the vitrifying material mixed with silicon carbide is charged. Ru.

Furthermore, since it is desirable to form molten glass as quickly as possible upon startup, it is desirable to charge a mixture containing a large amount of sodium nitrate, which has a low melting point. Sodium 'fr1ate has a melting point of 308°C and can be easily melted by high frequency heating.

The temperature inside the glass melting furnace is 1000°C in the glass melting Fa layer.
As described above, the temperature is, for example, 1200° C., and these mixtures are sequentially inserted above the charged mixture of dry powder and vitrification material so that the temperature is slightly higher than room temperature.

As the heating time lj1 elapses, the mixture of sodium nitrate and vitrification material melts and vitrification begins. The glass-formed material has a high specific gravity, so it is molten FA! It separates from the bow and settles down.

On the other hand, IC pyrolysis gas generated by pyrolysis of the dry powder is led to an off-gas treatment device from a pyrolysis gas outlet 25 provided at the upper side of the glass melting furnace 21 . At this time,
The pyrolysis gas is heated while passing through a mixture of dry powder and vitrification material deposited in the glass melting furnace 21, and the radioactivity contained in the pyrolysis gas is removed by the filtering action of the mixture. Ru.

As described above, the entrainment rate of radioactivity in the solidification processing apparatus of the present invention can be kept extremely low due to the static thermal decomposition and the filtering effect of the radioactive waste in the upper part of the glass melting furnace.

The pyrolysis gas discharged from the glass melting furnace 21 is sent into the filter chamber 26.

The filter chamber 26 is made of a candle-shaped ceramic filter and removes dust and radioactivity contained therein from the hot gas. The temperature of the filter chamber 26 is 600 to 80
The temperature is maintained at 0°C, and any remaining flammable gas that has not been completely oxidized is completely oxidized.

The combustion gas leaving the filter chamber 26 is cooled down to 200 degrees Celsius or less by being mixed with 3 to 4 times the atmospheric air in the 19th century, and is then passed through the high-performance filter 28, where the radioactivity is completely removed. will be removed.

The waste gas that has passed through the high-performance filter 28 is measured for the presence or absence of radioactivity by a radiation monitor, and after it is confirmed that it does not contain radioactive components, it is released from the stack into the environment.

As the process progresses, the amount of molten glass in the glass melting furnace 21 increases and when it reaches a certain level, the level meter 21L detects this and issues an alarm, and the molten glass is released from the outlet nozzle 23 at the bottom of the glass melting furnace 21. is extracted.

The removal of the molten glass is carried out by the same method as that used in a melting furnace for ordinary vitrification processing.

? However, during normal operation, the IC cooling II shown in Figure 2
The glass 1 in the outlet nozzle 23 is solidified by the I tube 8, and the outlet nozzle 23 is blocked.When taking out, the heater 7 is energized to melt the glass in the outlet nozzle 23 and the vitrified material is disposed of as waste. Discharge into a storage container.

Experimental Example A mixture of dry powder and vitrification material having the following composition was supplied to a glass melting furnace using the above-mentioned apparatus, and the temperature at the bottom was approximately 1.
The furnace was heated to 200°C to fill the vitrified material.

Radioactive concentrated waste liquid dry powder...30% (Silicon oxide containing 90% sodium liE acid...4
3% boric acid・・・・・・・・・・・・・・・・・・
・・・・・・13% aluminum hydroxide・・・・・・
...3% calcium carbonate...
...The composition and physical properties of the vitrified product produced at 11% were as follows.

Composition Si 02・・・・・・・・・60%Na2O・
・・・・・・・・・・・・19%B203・・・・・・・
...10%AnzO3 ...3% CaO ...896 Physical properties Glass transition point 530℃ Specific gravity 2.5 Sodium oxide elution amount 2x 1o-' (1/c+J ・tiay+00' in 25°C pure water
c boiling water sx 1o-5 (J/c sea day [Effect of the invention] As explained above, according to the radioactive waste solidification treatment method and solidification treatment apparatus of the present invention, radioactive waste can be achieved with a high volume reduction rate. The solidified material produced by processing materials has poor water resistance and heat resistance, but it does not deteriorate over time and has high chemical and mechanical stability. By using this material, the throughput and temperature of the glass melting furnace can be lowered, and the durability of the glass melting furnace can be increased.

[Brief explanation of the drawing]

Fig. 1 is a block diagram schematically showing the configuration of the radioactive waste solidification processing apparatus of the present invention, Fig. 2 is a sectional view of its glass melting furnace, and Fig. 3 is a solidification of Group 04 waste of the present invention. FIG. 3 is a diagram showing the configuration of a processing device. 1・・・・・・・・・fJIi radiation 11J1ii
Waste liquid pre-treatment device 2.21... Glass melting furnace 3... Off-gas treatment device 4...
......Solid handling equipment 5.22...High frequency coil 6.23...Outlet nozzle

Claims (4)

[Claims] (1) Radioactive concentrated waste liquid containing sodium nitrate generated at nuclear fuel reprocessing facilities is dried and powdered, this dried powder and vitrification material are heated and melted, and the generated pyrolysis gas is processed to create a system. A method for solidifying radioactive waste, which comprises discharging it outside and filling a storage container with the remaining vitrified material. (2) The method for solidifying radioactive waste according to claim 1, wherein the vitrification material is a mixture of silicon dioxide, boric acid, aluminum hydroxide, and calcium carbonate. (3) A radioactive concentrated waste liquid pre-treatment device that converts radioactive waste liquid containing sodium nitrate generated at a nuclear fuel reprocessing facility into a dry powder and mixes it with a vitrification material at a predetermined ratio, and a dry powder obtained by the pre-treatment device. a glass melting furnace that heats and melts a mixture of and a vitrification material, an off-gas treatment device that processes pyrolysis gas generated in the glass melting furnace and discharges it as waste gas outside the system, and vitrification in the glass melting furnace. A vitrified material filling device for filling a storage container with radioactive waste. (4) The radioactive waste solidification processing apparatus according to claim 3, wherein the glass melting furnace uses a high-frequency heating device as a heat source.