JP5544111B2 - Method for separating, recovering and treating radioactive solid waste mainly composed of silica component - Google Patents

Description

The present invention relates to a treatment method for separating and recovering a silica component and uranium from a radioactive solid waste containing a silica component as a main component and containing at least hundreds of ppm to several percent of uranium and purifying uranium. It is.

  Nuclear fuel processing facilities store various types of radioactive solid waste containing uranium. Radioactive solid waste is being examined by burying it, but the scale of underground burial facilities differs depending on the concentration of uranium contained in the waste. Expected to be needed. Therefore, in order to reduce the economic burden, it is considered preferable to reduce the concentration of uranium contained in the waste as much as possible.

  In nuclear fuel processing facilities, industrial water glass is added to radioactive waste liquid containing 50 to 200 ppm of uranium to produce agglomerated precipitates of silica components and uranium, and the agglomerated precipitates are slurried from the waste liquid by a solid-liquid separator. Separation, mineral acid represented by nitric acid is added to the aggregated precipitate in the slurry state, uranium in the solid content is eluted as uranyl ions to the solution side, and further contained in radioactive liquid waste by solid-liquid separation Most of the uranium that had been collected is recovered. The uranium recovered here is reused as a uranium resource through a purification process and the like. And in the silica residue which remained as solid content in the elution process, several mass% uranium remains, and this silica residue is stored as radioactive solid waste (for example, refer patent document 1). .

  In addition, caustic soda and aqueous alkaline solutions such as ammonia water are added to radioactive effluent with a dilute uranium concentration of 50 to 200 ppm to produce sodium heavy uranate and ammonium heavy uranate, which are used for rotary drum centrifugal filters, etc. The uranium contained in the radioactive liquid waste is recovered by solid-liquid separation using Here, since the concentration of the aggregated precipitate is dilute, it is easy to cause filtration leakage during the filtration operation, so filtration is performed using diatomaceous earth mainly composed of silica as a filter aid, and the recovered solid content is radioactive solid waste. It is stored as a thing.

JP-A-60-237398 (page 1, lower right column, line 4 to page 2, upper left column, line 1, FIG. 1)

  As a method of separating uranium from radioactive solid waste containing such a silica component as a main component and containing uranium at a rate of several hundred ppm to several percent, the concentration of the solution should be 2 mm or less in a nitric acid solution of 3N or more. A method is used in which finely pulverized radioactive solid waste is added, and the solution is heated to 60 ° C. or more to promote elution to separate uranium. When uranium elution is performed from radioactive solid waste using the above method, it can be eluted up to about 80 to 90% of uranium contained in the solid content, but it is still undissolved in the silica residue. Eluted uranium remains on the order of several hundred ppm to several thousand ppm and cannot be completely separated. Therefore, in order to achieve complete separation of uranium, the same elution operation has to be repeated several times.

  However, the elution ratio of uranium decreases as the elution procedure is repeated multiple times, and conversely, the silica concentration in the eluate increases as the number of elutions increases, reaching several thousand ppm to several percent. When separation and purification is carried out by the subsequent solvent extraction method as the silica concentration increases and the silica concentration increases, a third phase called an emulsion is likely to be formed at the interface between the solution and the solvent, which causes a hindrance to the separation and purification. There was a problem of generating secondary radioactive solid waste with high uranium concentration.

  As a measure to solve the above problems, an aqueous hydrofluoric acid solution is added to a radioactive solid waste containing a silica component as a main component and containing at least several hundred ppm to several percent of uranium, and both the silica component and uranium are combined. A method of completely dissolving and separating as an ionic state is conceivable.

  However, this measure requires an excess of hydrofluoric acid that is several times to 10 times the equivalent of the reaction equivalent of the silica component, and heat treatment is required to promote dissolution. The above equipment burden becomes a problem. In addition, since the silica component and uranium dissolved by this method are both anionic in the solution, a separation operation using an anion exchanger having high selectivity for uranium is required. . Anion exchangers are expensive, and there is a problem that used anion exchangers generate secondary radioactive solid waste having a high uranium concentration.

  In addition, as another measure for solving the above-mentioned problems, a radioactive solid waste containing a silica component as a main component and containing at least uranium in a ratio of several hundred ppm to several percent, a silica component contained in the waste and an appropriate amount of iron And an oxide of calcium, once heated to a high temperature of 1400 ° C. or higher, rapidly cooled to form a mineral crystal, which is completely dissolved with a mineral acid and separated.

However, in this measure, the solubility is improved by using a SiO ₂ —CaO—Fe ₂ O ₃ ternary mineral crystal, but the crystallization region is narrow due to the blend balance of the three elements. An amorphous mineral is generated, and an undissolved residue is generated, resulting in a secondary radioactive solid waste. Moreover, since it processes in a high temperature furnace, there exists a subject that the capital investment expense burden is large.

  The object of the present invention is to separate silica component and uranium with high efficiency from radioactive solid waste, which contains silica component as the main component of radioactive solid waste and contains uranium at a ratio of several hundred ppm to several percent. It is intended to provide a method for treating radioactive solid waste, which can suppress the uranium transfer rate to the silica component side to a very low ratio by collecting and greatly reduce the cost for waste disposal. .

Another purpose of this method is to separate and recover silica component and uranium from radioactive solid waste with high efficiency, thereby reducing the rate of silica component transfer to the uranium side to a very low rate. It proposes a method for separating, recovering, and treating radioactive solid waste that can be refined and can be recovered as reusable clean uranium. .

A first aspect of the present invention is a method for treating a radioactive solid waste containing a silica component as a main component and containing at least uranium in a proportion of several hundred ppm to several percent, wherein the radioactive solid waste powder is and mixing such that 1.5 or more ratio: the NH ₄ F · HF powder, molar reaction equivalent ratio of the silica component and the NH ₄ F · HF powder contained in the radioactive solid waste powder 1 Te The mixed powder is put into a sealed container, and an inert gas is introduced into the sealed container as a carrier gas to make the inside of the container an inert gas atmosphere, and the mixed powder is heated at a temperature of 270 to 370 ° C. for 30 minutes to 2 minutes. By heating for a period of time, the silica component contained in the mixed powder reacts with NH ₄ F · HF to volatilize the silica component as (NH ₄ ) ₂ · SiF ₆ , and is discharged from the sealed container ( NH _{4) 2} · SiF ₆ The carrier gas is passed through a cooling means to cool to the gas to 50 ° C. or less, including the majority of the contained in the gas (NH _{4) 2} · SiF ₆ to solidify, followed by passing through a cooling means gas The gas is brought into contact with the absorbent provided in the scrubber and brought into contact with the absorbent provided in the scrubber so that (NH ₄ ) ₂ .SiF ₆ remaining in the gas is absorbed by the absorbent, thereby being discharged from the sealed container (NH ₄ ) _2.・ Recover the entire amount of SiF ₆ and the non-volatile powder remaining in the sealed container after the heating process, recover the solid content after cooling the container, add the mineral acid solution to the recovered non-volatile powder and dissolve it And a step of purifying uranium contained in the solution using at least one method selected from the group consisting of an ion exchange method, a solvent extraction method, and a uranium peroxide aggregation precipitation method. And

The second aspect of the present invention is an invention based on the first aspect, wherein the mixing of the radioactive solid waste powder and the NH ₄ F · HF powder is performed using a mixing device having a stirring blade, It is a method of mixing with coarse pulverization by stirring with this stirring blade.

  A third aspect of the present invention is an invention based on the first aspect, wherein the scrubber absorbent is caustic soda or water.

  In the method for treating radioactive solid waste according to the present invention, the silica component and uranium are separated from the radioactive solid waste containing the silica component as a main component and containing at least a few hundred ppm to several percent of uranium. Can be separated and recovered with high efficiency, the uranium transfer rate to the silica component side can be suppressed to a very low rate, and the cost for waste disposal can be greatly reduced. In addition, the silica component transfer rate to the uranium side is suppressed to a very low rate, and uranium contained in radioactive solid waste, which has been unusable in the past, can be purified and recovered as reusable clean uranium. it can.

It is the schematic which shows the processing apparatus for enforcing the method of this invention.

  DESCRIPTION OF EMBODIMENTS Embodiments for carrying out the present invention will be described with reference to the drawings.

  The radioactive solid waste to be treated in the present invention is a waste powder containing a silica component as a main component and containing at least uranium in a proportion of several hundred ppm to several percent. The waste powder before mixing does not need to be pulverized to adjust the particle size.

In the treatment method of the present invention, first, NH ₄ F · HF powder, which is relatively inexpensive as a reagent product and easier to handle than other fluorine-based reagents, is added to and mixed with this radioactive solid waste powder. To prepare a mixed powder. The mixing ratio here is such that the molar reaction equivalent ratio between the silica component contained in the waste powder and the NH ₄ HF · HF powder is a ratio of 1: 1.5 or more, and the waste powder and NH ₄ F · HF. And mix.

  The molar reaction equivalent formula is as shown in the following formula (1).

SiO ₂ + 4NH ₄ F · HF →
(NH ₄ ) ₂ SiF ₆ (g) + 2NH ₄ F (g) + 2H ₂ O (g) (1)
When the molar reaction equivalent ratio is less than 1: 1.5, the reaction is difficult to proceed, and all the silica components contained in the waste powder cannot be reacted. Among these, mixing the waste powder and the NH ₄ F · HF powder so that the molar reaction equivalent ratio is a ratio of 1: 1.5 to 1: 2.0 is a reaction represented by the above formula (1). It is particularly preferable from the technical point of view, for example, that the amount of excess ammonium fluoride generated in the volatile recovery system is suppressed. The effect of the present invention does not change even if the molar reaction equivalent ratio exceeds 1: 2.0, but it is not preferable because NH ₄ HF · HF powder is used more than necessary and increases the processing cost. .

The waste powder and the NH ₄ F / HF powder are mixed by a mixing device having a rotating drum, or by using a mixing device having a stirring blade and mixing with coarse pulverization by stirring with the stirring blade. It is preferable.

Next, the mixed powder is put into a sealed container, and an inert gas is introduced into the sealed container as a carrier gas, and the container is heated to an inert gas atmosphere under a certain condition. The silica component contained is reacted with NH ₄ F · HF to volatilize the silica component as (NH ₄ ) ₂ · SiF ₆ .

The reaction shown in the above formula (1) is advanced, and the silica component in the mixed powder is reacted with NH ₄ F · HF to convert it into volatile (NH ₄ ) ₂ SiF ₆ gas. Separate from uranium. By this heating process, the uranium transfer rate to the silica component side can be kept to a very low rate, and the silica component transfer rate to the uranium side can be kept to a very low rate, so the silica component and uranium can be separated with high efficiency. can do.

  In this heating process, for example, as shown in FIG. 1, a sealed container 11 having a high sealing property is provided inside the heating furnace 10, and an inert gas such as dry nitrogen is supplied to the sealed container 11 as a carrier gas. It is preferable to use a heating means 14 having a structure in which a gas supply pipe 12 capable of discharging gas and a gas discharge pipe 13 for discharging gas from inside the container 11 are connected. The heating means 14 is provided with a thermocouple 15 that measures the temperature inside the container.

In the processing apparatus shown in FIG. 1, the mixed powder 16 is put into the sealed container 11 and the inert gas is continuously supplied from the gas supply pipe 12 into the container 11 as a carrier gas. and the mixed powder 16 with an active gas atmosphere by heating at a constant condition (NH _{4) 2} SiF ₆ cause gas discharged together with the carrier gas from the gas discharge pipe 13 (NH ₄₎ a ₂ SiF ₆ gas recovery unit 17 The cold trap 18 and the scrubber 19 are used for recovery.

The reason for using such a sealed container 11 and continuing to introduce an inert gas as a carrier gas into the container 11 to make the heating atmosphere an inert gas atmosphere is that air having an oxygen content or the like enters the container 11. And (NH ₄ ) ₂ SiOF ₂ and other intermediate compounds are easily produced in the compound group shown on the right side of the above formula (1) due to the influence of oxygen and moisture. This is because, in addition to being able to be discharged from the processing system as an unstable compound that cannot be recovered by the subsequent recovery means 17, there is a risk that the pipe will be clogged due to accumulation of white matter in the piping portion during recovery. Therefore, it is necessary to expel air from the container 11 before heating.

  If the gas introduced into the container 11 directly hits the mixed powder 16, uranium in the fine powder state in the mixed powder 16 is likely to scatter and jump out into the exhaust system due to the air current. It is necessary to consider the direction of the tip of the gas supply pipe 12, the shape of the apparatus, and the like so that the generated gas does not directly hit the mixed powder 16 as much as possible.

The mixed powder 16 is heated by holding it at a temperature of 270 to 370 ° C. for 30 minutes to 2 hours. When the heating temperature is less than 270 ° C., the reaction shown in the above formula (1) is not accelerated, and the silica component remains on the non-volatile residue side (uranium side). Further, when the heating temperature exceeds 370 ° C., the reaction shown in the above formula (1) proceeds too quickly, and (NH ₄ ) ₂ SiF ₆ gas is generated excessively, and this excess (NH ₄ ) ₂ SiF ₆ gas is Since it is discharged from the gas discharge pipe 13 in accordance with the amount of carrier gas introduced into the container 11, in order to reliably recover the total amount of (NH ₄ ) ₂ SiF ₆ gas, the recovery means that follows the heating means 14 The equipment capacity of 17 needs to be increased more than necessary, which causes a problem that the equipment burden increases. Also, when the generation rate of volatile gas becomes excessively high and excessive volatile gas is generated, uranium in the fine powder state (particle size is submicron) in the mixed powder 16 is involved and jumps out into the recovery system. Cause malfunctions. In this case, the separation accuracy between the silica component and uranium is greatly reduced. Therefore, in this heating step, it is necessary not to apply excessive heat to the mixed powder more than necessary in order to increase the separation accuracy between the silica component and uranium.

  The heating and holding time varies somewhat because it depends on the equipment capacity and the heating temperature, but the separation operation is performed in a reaction time of about 30 minutes to 2 hours.

Next, the carrier gas containing (NH ₄ ) ₂ .SiF ₆ discharged from the sealed container 11 is passed through the cooling means (cold trap) 18 of the recovery means 17 to cool the gas to 50 ° C. or lower and discharge. Most of (NH ₄ ) ₂ .SiF ₆ contained in the gas thus formed is solidified, then the gas passed through the cooling means 18 is sent to the scrubber 19, and the gas is absorbed into the absorbent 19 a provided in the scrubber 19. The total amount of (NH ₄ ) ₂ .SiF ₆ discharged from the sealed container 11 is recovered by allowing the absorbent to absorb (NH ₄ ) ₂ .SiF ₆ remaining in the gas by contact.

(NH ₄ ) ₂ SiF ₆ generated in the sealed container 11 is volatilized at a temperature of 160 ° C. or higher, and is in a solid state at room temperature, and has a solubility in water of 50 g / 100 g H ₂ O (17.5 ° C.). And high properties.

After the (NH ₄ ) ₂ SiF ₆ gas having such properties is exhausted from the sealed container, the entire amount can be easily recovered by a method such as a cold trap or a scrubber.

However, if the exhausted (NH ₄ ) ₂ SiF ₆ gas is collected only by the scrubber, the (NH ₄ ) ₂ SiF ₆ solidifies and accumulates in the middle of the piping for introducing the gas component directly into the scrubber, and the piping May cause occlusion. Therefore, it is industrially advantageous to solidify most of the (NH ₄ ) ₂ SiF ₆ once with a cold trap or the like and collect it as a solid 20 and separate and collect it before reaching the scrubber.

  For the absorbent provided in the scrubber, if an acidic solution is used, free fluorine ions are generated and cause corrosion, etc., so caustic soda or water is often used. If possible, the scrubber is divided into two stages. It is preferable to use caustic soda for the former absorbent and water for the latter absorbent.

The uranium concentration in (NH ₄ ) ₂ SiF ₆ recovered by the above method is as low as several ppm, and a decontamination factor of 1000 or more is obtained as the decontamination efficiency.

  Furthermore, the non-volatile powder remaining in the sealed container after the heating process is recovered as a solid after cooling the container, and a mineral acid solution is added to the recovered non-volatile powder to dissolve it, and the ion exchange method, solvent extraction method And uranium contained in the solution is purified using at least one method selected from the group consisting of uranium peroxide flocculation and precipitation.

  Nonvolatile metal elements such as uranium and iron remain in a fine powder state in the container 11 after the heating process. This non-volatile powder is recovered in the solid state after cooling the container 11. When a mineral acid solution, preferably about 3N nitric acid solution is added to the recovered non-volatile powder, it can be easily dissolved without heating and heating at about 60 ° C.

This is because, in radioactive solid waste, uranium and non-volatile metal elements generally exist as oxides, and it appears that there is a silica component around them. There reacted with NH ₄ F · HF, when the exit was volatilized in the form of _{(NH 4) 2 · SiF 6} , the oxide is left behind as a sparse state, easily react with mineral acid solution such as nitric acid It is thought that the main cause is that it exists in a state.

  Since there is a metal element such as iron in the nitric acid-dissolved uranyl nitrate solution, it can be reused as a clean uranium resource by performing purification and separation by ion exchange, solvent extraction or uranium peroxide precipitation. It is possible.

  Next, examples of the present invention will be described in detail together with comparative examples.

<Example 1>
A dried silica starch having a uranium concentration of about 1% by mass is prepared. NH ₄ F · HF powder is added to the dried silica starch, the silica component contained in the dried silica starch powder, NH ₄ F · HF powder, Were mixed so that the molar reaction equivalent ratios were 1: 1.0, 1: 1.5 and 1: 2.0, respectively, to prepare three kinds of raw material mixed powders. Further, the three kinds of prepared mixed powders were each divided into 5 equal parts.

  Then, about 7 g of the mixed powder is put into a sealed container provided in the heating furnace shown in FIG. 1, and dry nitrogen gas is used as a carrier gas while being supplied into the sealed container at a flow rate of about 0.5 L / min. The inside of the sealed container is heated to a temperature of 200 ° C., 230 ° C., 270 ° C., 320 ° C. and 370 ° C. at a rate of temperature increase of 5 ° C./min. After reaching these temperatures, the temperature is kept constant. Each test was held for 40 minutes.

  The sealed container after heating was cooled, the non-volatile residue remaining in the container was compared with the initial weight, and the silica component that did not volatilize was qualitatively evaluated to confirm the separation efficiency. Table 1 below shows the separation efficiency results.

表１から明らかなように、モル反応当量比率と加熱保持温度の組み合わせ条件としては、モル反応当量比率が１：１．５以上でかつ加熱保持温度が３２０℃以上の場合、モル反応当量比率が１：２．０でかつ加熱保持温度が２７０℃の場合に残存量１．０質量％未満と極めて良好な分離特性が得られた。なお、モル反応当量比率が１：１．５でかつ加熱保持温度が２７０℃の場合、シリカ成分が数％残留している結果となったが、放射性固体廃棄物の処理としては十分実用に耐え得るレベルであった。

As is apparent from Table 1, as a combination condition of the molar reaction equivalent ratio and the heating and holding temperature, when the molar reaction equivalent ratio is 1: 1.5 or more and the heating and holding temperature is 320 ° C. or more, the molar reaction equivalent ratio is When the temperature was 1: 2.0 and the heating and holding temperature was 270 ° C., the remaining amount was less than 1.0% by mass and very good separation characteristics were obtained. When the molar reaction equivalent ratio was 1: 1.5 and the heating and holding temperature was 270 ° C., the silica component remained several percent, but it was sufficiently practical for the treatment of radioactive solid waste. It was a level to get.

In addition, in the example where the separation efficiency was less than 1.0% and very good separation characteristics were obtained, there was no significant accumulation in the exhaust system, and the recovered compound was almost (NH ₄ ) ₂ SiF _6. It was.

The uranium concentration in the recovered (NH ₄ ) ₂ SiF ₆ is several ppm or less for all samples, and the uranium concentration in the waste scrubber solution is tens of ppb or less. As a result, high separation effects were obtained for uranium and silica on the volatile side.

  Furthermore, when the powdery substance remaining in the sealed container after the heating process is recovered after cooling the container and dissolved in a 3N nitric acid solution, the separation efficiency is less than 1.0% and very good separation characteristics are obtained. It was confirmed that all of the obtained examples were able to dissolve neatly, but in the case where the non-volatile content with a separation efficiency of several percent or more remained, the dissolution residue remained and uranium could not be completely dissolved. .

When the same test was performed with the heating atmosphere changed from dry nitrogen gas to air, the separation efficiency showed almost the same tendency as the test performed in the dry nitrogen atmosphere. Because of this, many white deposits were observed in the piping part on the way from the heating means to the recovery means. The composition transferred to the recovery means and the composition of the accumulation were examined. As a result, (NH ₄ ) ₂ SiOF ₄ and (NH ₄ ) ₂ SiF ₆ were present, and the volatilized (NH ₄ ) ₂ SiF ₆ was air. It is presumed that the intermediate compound was produced by reacting with the oxygen contained therein.

Subsequently, a mixed powder was prepared using about 30 g of the dried silica starch, and about 75 g of the mixed powder was put into the processing apparatus shown in FIG. 1, and the heating and holding time was set to 2 hours. The separation efficiency showed the same tendency as in the above test, and the compound recovered by the recovery means was almost (NH ₄ ) ₂ SiF ₆ . From this result, it was confirmed that the same effect could be obtained even if the amount of treatment was increased.

From the above results, the waste powder and the NH ₄ F · HF powder are mixed so that the molar reaction equivalent ratio of the silica component and the NH ₄ F · HF powder is 1: 1.5 or more, and the heating step High efficiency with non-volatile uranium (uranium oxide) by volatilizing and separating silica components in the form of (NH ₄ ) ₂ · SiF ₆ by holding at 270 ° C or higher in an inert atmosphere In addition, by making the heating atmosphere an inert gas atmosphere, the volatilized (NH ₄ ) ₂ .SiF ₆ gas reacts with oxygen to form an unstable intermediate compound, which is used as a recovery means. It was confirmed that accumulation in piping along the way can be prevented. Furthermore, it was confirmed that the non-volatile uranium remaining on the sealed container side after the heating step was easily dissolved by nitric acid or the like, and could be purified and recovered.

DESCRIPTION OF SYMBOLS 10 Heating furnace 11 Sealed container 12 Gas supply pipe 13 Gas discharge pipe 14 Heating means 15 Thermocouple 16 Mixed powder 17 Recovery means 18 Cold trap 19 Scrubber

Claims (3)

A method for separating, recovering, and treating radioactive solid waste mainly composed of a silica component and containing at least uranium in a proportion of several hundred ppm to several percent,
The NH ₄ F · HF powder is used for the radioactive solid waste powder, and the molar reaction equivalent ratio of the silica component contained in the radioactive solid waste powder and the NH ₄ F · HF powder is 1: 1.5 or more. Mixing to a ratio;
The mixed powder is put into a sealed container, and an inert gas is introduced into the sealed container as a carrier gas so that the inside of the container is an inert gas atmosphere, and the mixed powder is heated at a temperature of 270 to 370 ° C. A step of reacting the silica component contained in the mixed powder with NH ₄ F · HF by heating for 30 minutes to 2 hours to volatilize the silica component as (NH ₄ ) ₂ · SiF ₆ ;
A carrier gas containing (NH ₄ ) ₂ .SiF ₆ discharged from the sealed container is passed through a cooling means to cool the gas to 50 ° C. or less, and (NH ₄ ) ₂ .SiF contained in the gas Most of ₆ is solidified, and then the gas passed through the cooling means is sent to the scrubber, and the gas is brought into contact with the absorbent provided in the scrubber to remain in the gas (NH ₄ ) ₂ Recovering the total amount of (NH ₄ ) ₂ .SiF ₆ discharged from the sealed container by absorbing the SiF ₆ into the absorbent;
The non-volatile powder remaining in the sealed container after the heating step is recovered as a solid content after the container is cooled, and a mineral acid solution is added to the recovered non-volatile powder to dissolve it, ion exchange method, solvent extraction And a step of purifying uranium contained in the solution using at least one method selected from the group consisting of a uranium peroxide coagulation precipitation method and a uranium peroxide coagulation precipitation method. Separating, recovering and treating radioactive solid waste. The method according to claim 1, wherein the radioactive solid waste powder and the NH ₄ F · HF powder are mixed using a mixing device having a stirring blade, and are mixed while being coarsely pulverized by stirring with the stirring blade.   The method of claim 1 wherein the scrubber absorbent is caustic soda or water.

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