JP2993486B2 - Radioactive waste filling container and solidified radioactive waste - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a radioactive waste filling container and a solidified radioactive waste, and more particularly to a radioactive waste filling container suitable for use in solidifying radioactive waste generated from a nuclear power plant or the like. And a solidified radioactive waste obtained using the radioactive waste filling container.

[0002]

2. Description of the Related Art In the treatment and disposal of so-called low-level radioactive waste such as concentrated waste liquid, spent ion exchange resin, radioactive miscellaneous solids, etc. generated from radioactive material handling facilities such as nuclear power plants and nuclear fuel reprocessing facilities, Cement, concrete, and hydraulic inorganic solidifying materials such as water glass (sodium silicate) have been used for solidifying containers, solidifying materials, backfilling materials, and disposal site structures.

The hydraulic inorganic solidifying material has advantages such as easy solidifying operation, low cost, and excellent radiation resistance, and is suitable for the treatment and disposal of low-level radioactive waste. Are suitable. Furthermore, for the treatment and disposal of low-level radioactive waste, the integrity can be maintained even under adverse conditions such as the solidified body or disposal facility being submerged.
It is necessary that the radionuclide inside has a property capable of greatly delaying the rate at which the radionuclide leaks out of the solidified product or the disposal facility.

Conventionally, as a method for securing long-term durability of a solidified body, there has been a method of adding glass fibers to a hydraulically solidified material as described in JP-A-60-202398. Since the fibrous substance has a tensile strength several times greater than that of the hydraulically solidified base material, the fibrous substance has an effect of reinforcing the hardened body, and the tensile strength and bending strength of the entire solidified body are significantly improved. Therefore, even if the volume change or external force of the packing is applied to the solidified body, the solidified body does not crack or break, and even if it is assumed that the solidified body is disposed of on land, the radioactivity is at a sufficiently low level. It is considered impossible that the solidified body deteriorates between several tens of years and several hundreds of years.

Further, as disclosed in JP-A-58-40000, a protective layer is provided on a container for solidifying radioactive waste to delay the release of radioactive nuclides, and the protective layer is ion-exchanged. There is a method of embedding a filler having a property and absorptivity.

[0006]

SUMMARY OF THE INVENTION Among the above prior arts,
Japanese Patent Application Laid-Open No. 60-202398 does not consider reducing the rate of leaching of radioactive waste from radioactive waste, and disposes of radioactive waste with a higher radioactivity level than the current level, and reduces the amount by half. When disposing of waste containing long-lived carbon-14 and technesium-99, there has been a problem that measures must be taken to reduce the radioactive leaching rate.

On the other hand, Japanese Patent Application Laid-Open No. 58-40000 does not have the property of improving the strength of the solidification container, so that the solidification container cracks due to the dry / humidity and warm / cold cycle of the disposal site, and the soundness cannot be maintained. There was a problem.

An object of the present invention is to provide a radioactive waste filled container capable of increasing bending strength, improving impact resistance and reducing radioactive leaching rate.

An object of the present invention is to provide a solidified body of radioactive leaching rate can be reduced, and radioactive waste-filled container and radioactive wastes can increase the filling amount of radioactive waste.

[0010]

According to a first aspect of the present invention, there is provided a radioactive waste filling container at least partially composed of concrete, wherein the concrete adsorbs a dissolved substance in a liquid. In that it contains a dissolved substance adsorption reinforcing material.

[0011] The action of the reinforcing material for adsorbing dissolved substances in concrete increases the bending strength of the radioactive waste filled container,
In addition, the impact resistance can be improved. For this reason, the occurrence of cracks in the concrete portion due to bending and impact can be significantly reduced. The probability that the radionuclide contained in the radioactive waste filled in the radioactive waste filling container leaks out is significantly reduced.

[0012] Since dissolved substances adsorbing reinforcement adsorbs radionuclides, radioactive waste filling container containing dissolved substances adsorbing reinforcement, it is possible to improve the distribution coefficient for radioactive nuclides,
The radioactivity leaching rate can be further reduced.

A second feature of the present invention that achieves the object of the present invention is that at least a part of the concrete is made of concrete, and the concrete contains a dissolved substance adsorption reinforcing material in which dissolved substances in a liquid are adsorbed. Using radioactive waste filled containers,
It is a solidified radioactive waste obtained by solidifying radioactive waste filled inside concrete with a solidifying material.

Since the concrete contains the dissolved substance adsorption reinforcing material, it is possible to increase the bending strength and the impact resistance of the radioactive waste filling container as described above, and as a result, the radioactive waste that can be filled inside the concrete is obtained. The quantity of things can be increased.

Further, the dissolved substance-adsorbing reinforcement in concrete, for adsorbing radionuclides in nuclear waste
Since the solidified radioactive waste produced using the radioactive waste filling container of the first invention can improve the distribution coefficient for radionuclides, the radioactive leaching rate can be further reduced.

[0016]

DESCRIPTION OF THE PREFERRED EMBODIMENTS One embodiment of the present invention will be described below with reference to the drawings.

FIG. 1 is a diagram showing a flow of a system for solidifying radioactive spent ion exchange resin (waste resin) generated from a nuclear power plant in a solidification container by a cement-based hydraulic hardening material.

A slurry of waste resin discarded from a nuclear power plant is supplied to the resin dehydrator 1 and dewatered by centrifugal dehydration so that the water content becomes about 50%. Then, the dehydrated waste resin is sent to the resin receiving tank 2, and a fixed amount is supplied to the kneading tank 4 by the feeder 3. This waste resin captures several types of radionuclides in the form of ions or solids. Next, kneading water is supplied to the kneading tank 4 by the addition water tank 7, cement powder is supplied by the cement silo 5 via the feeder 6, and fibrous activated carbon is supplied from the additive hopper 8 to the kneading tank 4. The mixture is forcibly kneaded with a stirring blade 9 to form a paste. Then, after the obtained paste is poured into the solidification container 10, it is cured and solidified to be a solidified body. When vibration is applied by the vibrator 11 at the time of injection, a good solidified body with few bubbles can be produced inside the solidified body. In addition to the above-described process, it is also possible to directly supply a dehydrated waste resin, cement powder, and fibrous activated carbon to a solidification container, and to solidify by stirring.

Next, this embodiment will be described in detail. In this example, two types of solidified bodies having the compositions shown in Table 1 were prepared.

[0020]

[Table 1]

Here, the cement powder used is a class C blast furnace cement, but cement glass, silica cement, alumina cement, fly ash cement, sulfate resistant cement which is a solidified material obtained by mixing sodium silicate powder with cement. And the like can also be used, and these are collectively referred to as a cement-based hydraulic hardening material. In addition, about 2% of a high-performance water reducing agent for cement (β-naphthalene sulfonate high-condensate) was added to the added water in advance. The fibrous activated carbon is made of tar and has a thickness of 15 mm and a length of 3 mm.
mm, the specific surface area is 1500 m ² / g, and the average pore diameter of the micropores on the surface is 20 °.

After a curing period of 4 weeks, the completed solid was subjected to core boring to prepare a large number of test pieces of about 0.1 liter, and a drying cycle test of water immersion → 70 ° C. air drying → water immersion was performed. . The compressive strength of the test specimen was measured for each cycle, and the soundness of the solidified specimen was examined. Whether or not cracks are generated in the solidified body due to the dry-wet cycle can be detected by a change in compressive strength.

FIG. 2 is a diagram showing a change in compressive strength with an increase in the number of dry and wet cycles. The plot of white coloring is CAS
The solidified body of E1 is shown, and the solid black plot shows the solidified body of CASE2. In the case of the solidified case without the addition of the fibrous activated carbon of CASE2, cracks began to appear gradually on the surface when the dry / wet cycle exceeded 5 cycles, and the solidified state collapsed in 10 cycles. On the other hand, in the case of solidification to which the fibrous activated carbon of CASE 1 was added, no decrease in compressive strength was observed even after more than 20 cycles, and the solidified body was sound.

Next, a part of the solidified product of CASE 1 and CASE 2 was sampled, and 1 g of a sample crushed in a mortar was subjected to 100 μCi of carbon-14 and cesium-134, respectively.
After dispersing in distilled water containing the mixture and shaking at 25 ° C. in a water bath incubator, the partition coefficient of carbon-14 (anion) and cesium-134 (cation) of each solidified CASE was measured. For the measurement of the partition coefficient, carbon-14
, A liquid scintillation counter was used, and for cesium-134, a pure Ge semiconductor detector was used. Table 2 shows the measured distribution coefficients (relative values). In general, it is known that the radioactive leaching rate from a solidified body when the solidified body is submerged is inversely proportional to the square root of the distribution coefficient of the solidified material.

[0025]

[Table 2]

As can be seen from Table 2, the radioactive leaching rate from the solidified CASE1 is reduced to about 1/10 with carbon-14 and about 1/7 with cesium-134 as compared with the solidified CASE2. I have. That is, by adding a small amount (<5% by weight) of fibrous activated carbon to the conventional cement-based solidified material, the durability of the solidified body is improved and the radioactive leaching rate can be reduced.

Fibrous activated carbon has numerous micropores (pore diameter: 20 °) on its surface, and has the property of mainly physically adsorbing radioactive nuclides in an ionic or molecular state. For this reason, the adsorption characteristics have no polarity and basically have the function of adsorbing both nuclides in the form of cations and nuclides in the form of anions in a liquid, and delaying the elution thereof.
It has the property of having a high distribution coefficient even for carbon-14, which is in the form of an anion or organic carbon and for which there was no effective method of elution delay.

Further, as for the shape of the fibrous activated carbon, the larger the aspect ratio (fiber length / fiber thickness) is, the greater the reinforcing effect is, but the kneading property and injectability of the paste are deteriorated accordingly. Therefore, although the aspect ratio can be freely adjusted, 200 to 300 is desirable. Also, when used in combination with a cation adsorbent, the effect of further reducing the leaching rate of radionuclides in the form of cations is further exerted. In this embodiment, the fibrous activated carbon has been described. However, the same effect can be obtained by adding an ion exchange resin and an alkali metal titanate fiber.

The ion-exchange fiber is provided with an ion-exchange group (such as a sulfonic acid group, a carboxyl group, or a quaternary ammonium group) derived from a polymer as a main skeleton, and an ion of a radionuclide dissolved in water by an ion exchange reaction. Adsorb. Since the polarity of ions that can be adsorbed differs depending on the type of exchange group, almost all nuclides can be adsorbed by mixing cation exchange fibers and anion exchange fibers. The alkali metal titanate fibers are adsorbed by ion exchange between alkali metal ions present between the layers of the titania layered structure and nuclides in the form of cations in the liquid. Accordingly, only the cation nuclide has the effect of delaying elution, and has no effect on the anion nuclides such as carbon-14 and neutral molecules. Further, the alkali metal titanate fiber has a specific gravity of 3 or more,
It is desirable to adjust the apparent specific gravity to 1.5 to 2.5 because there is a possibility of settling in the hardening material paste.

The fibrous substances described above (fibrous activated carbon,
The ion-exchange fiber and the alkali metal titanate fiber) can be used as a blend of two or more kinds depending on the kind and amount contained in the waste contained therein.

Next, another embodiment of the present invention will be described. This embodiment is suitable for solidifying a used ion exchange resin generated from a nuclear power plant in a solidification container.

The system used in this embodiment is the same as that shown in FIG. From the additive hopper 8, a solidified body to which powdered activated carbon is added and a solidified body to which fibrous activated carbon is added are prepared. And the distribution coefficient of carbon-14 in the solidified material by appropriately changing the addition amount of powdered activated carbon and fibrous activated carbon,
The maximum fillable resin amount was determined by experiment. Note that the maximum amount of resin that can be filled is 1 unit for the solidified body prepared by curing for one month.
It means the maximum amount of resin that can be immersed in water for a period of 30 months to ensure a compressive strength of 30 kg / cm ² or more.

FIG. 3 shows the results of the above experiments, and shows the relationship between the amount of the additive, the resin filling rate, and the distribution coefficient of C-14. As for the solidified material to which the powdered activated carbon was added as shown by the black coloring plot, the carbon-14 was increased with the addition amount.
Increases the distribution coefficient of carbon-14 and is effective in reducing the leaching rate of carbon-14, but the maximum resin loading is reduced and the strength of the solidified material itself is reduced.

On the other hand, in the solidified product to which the fibrous activated carbon was added as shown by the whitening plot, the maximum resin filling rate and the distribution coefficient of carbon-14 increased with the increase of the added amount. . In other words, when the addition amount is 4% by weight, the maximum resin filling rate can be twice or more than that when no addition is made. Further, the distribution coefficient of carbon-14 is several times as effective as that of powdered activated carbon.

Further, as the amount of fibrous activated carbon increases, the distribution coefficient of carbon-14 can be increased, but the kneadability and injectability of the solidified material paste are reduced. Therefore, the amount of the fibrous activated carbon is preferably 10% by weight or less, and more preferably 5% by weight or less. The effect is as long as the fibrous activated carbon is added to some extent, and the lower limit is about 0.1% by weight. On the other hand, in the case of powdered activated carbon, it is possible to add up to about 10% by weight in relation to the resin filling rate.

As described above, the addition of powdered activated carbon and fibrous activated carbon has an effect of increasing the distribution coefficient of carbon-14 having a small distribution coefficient. Furthermore, in the case of fibrous activated carbon, the mechanical strength of the solidified body is increased, so that the filling rate of waste can be increased. And since these additives mainly capture nuclides by physical adsorption, cations,
Similar effects are obtained for radionuclides in the form of neutral molecules regardless of anions.

Another embodiment of the present invention will be described with reference to FIG. The present embodiment is suitable for solidifying and disposing of radioactive concentrated waste liquid generated from a nuclear power plant or a nuclear fuel reprocessing plant.

The concentrated waste liquid containing about 20% of sodium sulfate is temporarily stored in the waste liquid tank 12. First,
The concentrated waste liquid is supplied to the centrifugal thin film dryer 13, and the concentrated waste liquid is pulverized. Then, the dried powder is directly mixed in the kneading tank 18.
To a solidified material 22 by mixing with the solidified material. Alternatively, the paste 21 may be put into the solidification container 19 and a paste of a solidified material separately kneaded may be poured into the solidified body 21. For the solidified material tank 15, a class C blast furnace cement premixed with 4% by weight of fibrous activated carbon (diameter 10 μm, length 3 mm) as a nuclide-adsorbing reinforcing material was used. In place of this cement, it is needless to say that so-called cement glass, silica cement, alumina cement, fly ash cement, sulfate-resistant cement and the like, which are the above-mentioned cement-based hydraulic hardening materials, can also be used. This solidified material is kneaded in a kneading tank 17 with an appropriate mixing together with kneading water supplied from an additive water tank to form a paste. The added water contains β
It is desirable to add about 2% of a naphthalene sulfonate-based high-performance water reducing agent in advance.

This embodiment will be described specifically. In this embodiment, the dry powder of the concentrated waste liquid and the solidified material paste were directly mixed. The dry powder was changed in the range of 0 to 50% by weight, and a solidified body was prepared from a solidified material obtained by premixing 4% by weight of fibrous activated carbon and a solidified material without addition. The water / solids ratio was 0.1.
And 4. In addition, in the case of adding 10% of dry powder, ⁹⁹ TcO ₄ ⁻ as an anion nuclide was added as a tracer in an amount of 100 μCi per solidified body. Then, the compressive strength of each solid was examined, and a radioactivity leaching test was performed.

FIG. 5 shows the results of examining the compressive strength after immersion in water for one month, and FIG. 6 shows the results of the radioactivity leaching test. The case where fibrous activated carbon is added and the case where it is not added are shown in comparison.

As is apparent from FIG. 5, even if the addition amount of the dry powder is more than doubled by the addition of the fibrous activated carbon, sufficient strength is obtained. In addition, as is apparent from FIG. 6, the leaching rate of technesium-99 can be reduced by about one digit, which is particularly effective when solidifying waste at a higher level than in the past. In order to reduce leaching of cation nuclides, cation exchange fibers and alkali metal titanate fibers are effective and can be used as a substitute for fibrous activated carbon. Further, it is more effective to use a mixed fiber of a cation exchange fiber and an anion exchange fiber, or a mixed fiber of a fibrous activated carbon and an alkali metal titanate fiber.

When alkali metal titanate fibers are used, the specific gravity of the fibers is 3 or more, which is larger than that of general cement paste. Therefore, it is necessary to reduce the diameter of the fiber or shorten the length of the fiber. However, the strength of the fiber may be reduced or the advantage of preventing cracking may be lost. Therefore, as another method, sedimentation of the fiber can be prevented by injecting the viscosity of the cement paste and increasing the viscosity so as not to hinder kneading. In this case, the viscosity of the paste is desirably about 3000 to 5000 cp.

In the present embodiment, when the solidified material paste is injected and solidified after pelletizing the dried powder, it is necessary to consider the viscosity of the solidified material paste and the length of the fiber. In other words, in the case of spontaneous injection without giving vibration with a vibrator, the viscosity of the solidified material paste is 3000 cp or less,
Preferably, it must be 2000 cp or less. Even in the case of vibration filling, by setting the viscosity of the solidifying material paste to 5000 cp or less, it is possible to produce a dense solid without voids. Regarding the length of the fiber, when a fiber having a large aspect ratio is used, the fiber may be easily caught in the gap between the pellets at the time of injecting the solidifying material, and the fiber may be unevenly distributed. In this case, the reinforcing effect is slightly reduced, but it is desirable to adjust the aspect ratio to 100 or less before use.

Another embodiment of the present invention will be described with reference to FIG. This embodiment relates to a solidification container suitable for solidifying radioactive waste generated from a nuclear power plant. As shown in FIG. 7, a concrete container 24 is lined with an iron drum 23 and used as a solidification container. However, it is also possible to use only the concrete container 24 as a solidification container.

The concrete container 24 is composed of cement, fine aggregate, coarse aggregate, and a nuclide-adsorbing reinforcing material. Cement is a cement-based hydraulic hardening material described above,
In addition to so-called cement glass, silica cement, alumina cement, fly ash cement, sulfate cement, portland cement, blast furnace cement and the like can be used. As fine aggregate, river sand, silica sand, silica fume, fly ash, granulated blast furnace slag, chamotte, and the like are suitable. Gravel and crushed rock are suitable for coarse aggregate. As the nuclide-adsorbing reinforcing material, fibrous activated carbon, ion-exchange fiber, or alkali metal titanate fiber can be used.

Table 3 shows the standard composition of the concrete container 24 thus implemented.

[0047]

[Table 3]

The maximum size of the coarse aggregate is optimally about 1/2 of the fiber length in consideration of the reinforcing effect of the fiber, but this coarse aggregate can be omitted.

When the amount of fibrous activated carbon added and the aspect ratio of the fibrous activated carbon are changed in the composition shown in Table 3 above, the bending strength of the solidification container changes as shown in FIG. As the addition amount and the aspect ratio increase, the bending strength can be increased. However, in order to avoid a decrease in the kneading property of the paste and the injectability into the shape, the addition amount of the fibrous activated carbon is 5% by weight or less. The aspect ratio is preferably about 200 to 300.

According to this embodiment, the bending strength of the solidified container is increased, and the impact resistance and the crack resistance can be improved. In addition, since the solidification container can be provided with a radionuclide adsorption property, it is possible to reduce the leaching of nuclides from radioactive waste filled therein. Furthermore, since the conductivity of the solidification container can be increased, local corrosion of the drum can also be reduced. The PMM from the surface of the solidification container of this embodiment
By impregnating a polymer such as A (polymethyl methacrylate), the water stopping property and the low leaching property of the container are further improved.

Another embodiment of the present invention will be described with reference to FIG. This embodiment relates to a pit and a backfill material when radioactive waste is disposed of on land. As shown in FIG. 9, the solidified body 25 filled with radioactive waste and solidified
Is settled in the pit 27 and the gap between the solidified bodies is filled with the backfill material 26. For the pit 27, a reinforced concrete structure having the composition shown in Example 4 or a prestressed concrete structure is suitable.

As the backfill material, those having the compositions shown in Table 4 are preferable. Instead of the class C blast furnace cement, cement glass, silica cement, alumina cement, fly ash cement or sulfate-resistant cement, which is the above-mentioned cement-based hydraulic hardening material, may be used.

[0053]

[Table 4]

Further, a fine aggregate of the cement of the class C blast furnace may be partially replaced. The backfill material preferably has a high fluidity, and if the viscosity is 2000 cp or less, injection becomes easy. Also, as in this embodiment, the backfill material is
By adding natural minerals such as zeolite, bentonite, montmorillonite, vermiculite, kaolinite, clinoptilolite and clay minerals, leaching of cationic nuclides can be further reduced. Further, according to this embodiment, the durability and weather resistance of the pits and the backfill material at the disposal site are improved, and the leaching rate of radioactive nuclides such as carbon-14 can be further reduced.

Another embodiment of the present invention will be described with reference to FIG. This embodiment relates to a reinforced concrete structure in contact with the sea. As shown in FIG. 10, seawater is provided by providing a layer of concrete to which the ion-adsorbing reinforcing material of the present invention is added on the side of a conventional reinforced concrete structure including a general concrete layer 29 and a reinforcing bar 30 in contact with the ocean. Corrosion of the rebar caused by diffusion and intrusion of chlorine ions in the inside can be mitigated, and there is an effect of extending the life of the structure.

[0056]

According to the first aspect of the present invention, it is possible to increase the bending strength of the radioactive waste filled container at least partially made of concrete, to improve the impact resistance and to reduce the radioactive leaching rate. it can.

According to the second aspect of the present invention, it is possible to increase the bending strength and the impact resistance of the radioactive waste filling container at least partly made of concrete, and as a result, radioactive waste that can be filled inside concrete. Can be increased. Further, since the dissolved substance adsorbed reinforcement in the concrete adsorbs radionuclides, it can reduce the radioactivity leaching rate.

[Brief description of the drawings]

FIG. 1 is a configuration diagram of a waste resin solidification system according to an embodiment of the present invention.

FIG. 2 is a characteristic diagram showing a change in compressive strength with an increase in the number of dry / wet cycles.

FIG. 3 is a characteristic diagram showing a relationship between an amount of an additive, a resin filling rate, and a distribution coefficient of C-14.

FIG. 4 is a configuration diagram of a concentrated waste liquid solidification system according to an embodiment of the present invention.

FIG. 5 is a characteristic diagram showing the relationship between the amount of dry powder added and the compressive strength of a solidified product.

FIG. 6 is a characteristic diagram showing the results of a leaching test for Tc-99.

FIG. 7 is a longitudinal sectional view of a solidification container according to another embodiment of the present invention.

FIG. 8 is a characteristic diagram showing the relationship between the bending strength of the solidification container, the amount of fibrous activated carbon added, and the aspect ratio.

FIG. 9 is a longitudinal sectional view of a disposal site structure according to another embodiment of the present invention.

FIG. 10 is a longitudinal sectional view of a marine offshore structure according to another embodiment of the present invention.

[Explanation of symbols]

DESCRIPTION OF SYMBOLS 1 ... Resin dehydrator, 2 ... Resin receiving tank, 4 ... Kneading tank, 8 ... Additive hopper, 10 ... Solidification container, 13 ... Centrifugal thin film dryer, 2
3: drum, 24: concrete container, 25: solidified body, 26: backfill material, 27: pit, 30: rebar.

──────────────────────────────────────────────────の Continuing on the front page (72) Inventor Tsukasa Baba 1168 Moriyama-cho, Hitachi City, Ibaraki Prefecture Inside Hitachi Energy Co., Ltd. (72) Inventor Koichi Chino 1168 Moriyama-cho Hitachi City, Hitachi Energy Research Laboratory (72) Inventor Takashi Ikeda 1168 Moriyama-cho, Hitachi City, Ibaraki Prefecture Inside Energy Research Laboratory, Hitachi, Ltd. (72) Inventor Jin Kikuchi 3-1-1 Sachimachi, Hitachi City, Ibaraki Prefecture, Hitachi, Ltd. 58) Field surveyed (Int.Cl. ⁶ , DB name) G21F 9/16 G21F 9/30 G21F 9/36

Claims (8)

(57) [Claims] 1. A radioactive waste filling container at least partially composed of concrete, characterized in that said concrete contains a dissolved substance adsorption reinforcing material for adsorbing a dissolved substance in a liquid . Filling container. Wherein said concrete is 5 wt% or less of the
The radioactive waste filling container according to claim 1, further comprising a dissolved substance adsorption reinforcing material. 3. The container according to claim 1, wherein said dissolved substance adsorption reinforcing material has micropores on its surface. 4. The container according to claim 3, wherein the micropores have an average pore diameter of 10 to 25 °. 5. The radioactive waste filling container according to claim 1, wherein the dissolved substance adsorption reinforcing material is a fibrous substance. 6. The fibrous substance has an aspect ratio of 200.
The radioactive waste container according to claim 5, wherein the number is from 300 to 300. 7. The container according to claim 5, wherein the fibrous substance is at least one selected from fibrous activated carbon, ion exchange fiber, and alkali metal titanate fiber. 8. A solidified radioactive waste obtained by solidifying a radioactive waste with a solidifying material inside the concrete of the radioactive waste filling container according to any one of claims 1 to 7.