JP3079979B2 - Treatment of metal waste from the nuclear industry - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention minimizes the amount of radioactive waste contained in "metal waste discharged in the nuclear industry", such as a core tube of a light water reactor for power generation whose service life has ended. The present invention relates to a method for treating metal-based waste from the nuclear industry, in which the restrictions on storage locations are reduced at once and valuable metals can be recovered.

[0002]

2. Description of the Related Art Currently, among the light water reactors for power generation that are currently operating in Japan, those that have been introduced at an early stage are approaching their service life, and are being rebuilt (replaced).
It's time to consider.

[0003] However, when the reactor equipment is replaced, a large amount of dead volume (dead sp
Existing equipment with ace) will be discharged as waste, but these will require long-term storage until the radioactivity levels are sufficiently attenuated. But,
In Japan, a serious problem is that it is not possible to secure sufficient storage locations for “used nuclear industrial waste” such as demolished existing nuclear reactor facilities.

[0004] Therefore, in order to further develop and mature the nuclear industry in Japan, a technology for reducing the volume of "equipment waste contaminated with radioactive materials" discharged at the time of replacement is indispensable. It is believed to be something.

[0005] However, conventionally, for the metal-based ones of "equipment waste contaminated with radioactive materials",
Attempts have been made to reduce the volume by melting it into an ingot. However, although this method can certainly reduce the volume significantly, it is still not sufficient considering that a large amount of waste will be discharged in the future.

[0006] In view of the above, an object of the present invention is to remarkably reduce the volume of metal-based waste discharged in the nuclear industry and to recover valuable metals from the waste at the same time. It was to establish a processing means that could do it.

[0007]

The present inventors have conducted intensive studies to achieve the above object.
A series of findings as shown in d) were obtained. a) Among nuclear industrial wastes, uranium and radioactive fission products that require special attention are all attached to the reactor core tube, steam and water supply pipes as oxides. Therefore, as long as only these oxides can be separated and recovered from the discarded furnace tube or the like, the volume of the waste requiring long-term storage should be significantly reduced. However, it is almost impossible industrially to completely separate it by mechanical means.

B) However, when the “halogen organic solvent method” known as a method for extraction separation and quantification of oxide inclusions in steel is applied, “oxidation of uranium and radioactive fission products by a“ halogen-organic solvent mixture ” Uranium and radioactive fission product oxides can be separated and recovered with a yield of 100% even in a trace area of 20 ppm or less through the melting step of the above-mentioned furnace tube or the like to which the substance has adhered. It is possible. In other words, spent metal-based nuclear industrial wastes such as reactor core tubes contain high levels of radioactive materials mainly composed of oxides, but when this is treated with a "halogen-organic solvent mixture", the metal components are dissolved. Since the oxide does not dissolve and remains as a residue, the insoluble oxide can be completely recovered by filtration, and thus can be separated into radioactive substances and non-radioactive substances.

C) In addition, in this case, a "halogen-organic solvent mixture" in which the furnace tube and the like are dissolved to separate insoluble oxides.
, The dissolved metal component can be effectively recovered as an electrolytic deposited metal containing no radioactive substance. As a result, it is possible to reduce the empty volume of a “metal-based structure contaminated with high-level radioactive materials”, such as a light water reactor component for a used power generation, which has a large dead space as it is. it can.

D) Furthermore, if the components of the electrolytic solution after the electrolytic recovery of the metal are adjusted with respect to the halogen and the organic solvent, this can be reused as a solution for the furnace tube and the like. The amount of time required for the process is extremely small, such as "dissolving metal members contaminated with high-level radioactive materials → recovery of oxides containing radioactive materials (all radioactive materials are included in oxides) → metal components Collect as metal →
A "system for treating metal-based wastes from the nuclear industry" that can be sufficiently industrially used, "reuse of electrolytes as solution" is established.

The present invention has been completed on the basis of the above findings and the like, and is based on "dissolving a metal-based waste containing a radioactive substance in the nuclear industry with a mixed solution of a halogen and an organic solvent and then dissolving the same. Then, the remaining solution is electrolyzed and the metal component is recovered as a metal. Alternatively, by further adjusting the component of the electrolyte and reusing it as a solution, the radioactive substance (oxide) and the non-radioactive substance ( Metal parts), and can be completely separated and recovered, and the volume of "nuclear industrial waste containing radioactive materials" that requires long-term storage can be significantly reduced. .

Here, the kind of halogen constituting the "halogen-organic solvent mixture" is not particularly specified, and any halogen element may be applied, but bromine (Br) or iodine may be used. It can be said that (I) is preferable.

The organic solvent constituting the "halogen-organic solvent mixture" is not particularly limited. For example, methanol (MeOH), methyl acetate (AcMe), acetonitrile (AcN), dimethylformaldehyde ( DM
Various organic solvents such as F), tetrahydrofuran (THF) and the like can be applied.
It can be said that Me is more preferable.

Further, the mixing ratio between the halogen and the organic solvent does not need to be particularly limited. However, if handling is taken into consideration, the halogen concentration in the mixed solution is 10 to 20% (hereinafter, "%" representing the halogen concentration in the mixed solution). Is the volume%).

The following is a specific example of the process of the present invention relating to the treatment of nuclear industrial waste. 1) First, the above-mentioned waste is treated as “halogen-organic solvent (for example,
The metal component is dissolved by immersion in a “Br ₂ -methanol) mixed solution” (if necessary, it is effective to promote the dissolution by anodic electrolysis). 2) Then, “halogen-organic solvent” The “oxide containing radioactive substances” insoluble in the “mixture” is separated by filtration through a micropore filter (if there are many insolubles, remove most of the insolubles by centrifugation, etc., and then filter. The filter medium will have a longer life) and the metal component will be completely recovered from the mixed solution. 3) Next, the filtrate containing the dissolved metal component will be used as the electrolyte.
Electrodeposition and recovery of metal ions in the electrolytic solution as metal by cathodic electrolysis. 4) Further, if necessary, adjust the components of the electrolytic solution.
This is reused as a metal component solution of nuclear industrial waste.

[0016]

Next, the nuclear industrial waste disposal method according to the present invention is described as "a typical metallic material for a nuclear reactor, SUS316.
Austenitic stainless steel is used as a material to be treated (simulated nuclear industrial waste), and a halogen-organic solvent mixture is a halogen of Br and an organic solvent of MeOH, AcN, AcMe, D
A more detailed description will be given together with the operation based on "Test Examples Using a Treatment Solution that is MF or THF". Table 1 shows the chemical analysis values of the SUS316 steel used.

[0017]

[Table 1]

In the present invention, first, as a first step, metal-based waste from the nuclear industry is dissolved in a "mixed solution of halogen and an organic solvent". Is SUS316 steel) by the “halogen-organic solvent mixture” as is apparent from the following dissolution test results.

That is, in the melting test, a SUS316 steel test piece (a SUS316 pipe imitating a core tube of a light water reactor)
Five kinds of the following mixed solvents were prepared to dissolve. A) 10% Br ₂ -MeOH (methanol), B) 5% Br ₂ -THF (tetrahydrofuran), C) 10% Br ₂ -DMF (dimethylformamide), D) 10% Br ₂ -AcN (acetonitrile) _{, E) 10% Br 2 -AcMe} ( methyl acetate). Here, the Br ₂ concentration when the organic solvent is THF is 5%.
In this case, if the Br ₂ concentration is mixed at 10%, the Br _2
This is because it was dangerous that a phenomenon of rapid boiling was observed some time after the addition, and it was found that a considerable amount of Br ₂ was volatilized. Of course, there was no particular problem with other solvents, and it was possible to adjust a mixed solvent having a Br ₂ concentration of 5% or 10%.
A mixed solvent of 0% was prepared.

In the dissolution test, about 3.5 g of a sample was cut out from the SUS316 pipe (SUS316 steel test piece) simulating the core tube of a light water reactor, and this was mixed with the mixed solvent 1.
The method was carried out by placing the mixture in a flask together with 00 ml and leaving the mixture without heating. The amount of dissolution was calculated based on the measured value by taking out a sample for each elapsed time and weighing it. The dissolution test results are shown in Tables 2 and 3.

[0021]

[Table 2]

[0022]

[Table 3]

From the results shown in Tables 2 and 3, SU
It can be confirmed that S316 steel dissolves in any of the mixed solvents. This result was the same even when the combination of the halogen and the organic solvent was variously changed.

However, as shown in Tables 2 and 3, SU
The dissolution rate of S316 steel varies greatly depending on the type of the mixed solvent, and is very fast for MeOH, AcN and AcMe (see Table 2). 3), the dissolution is slower than that of the above three mixed solvents. In the case of THF (see Table 3), the dissolution rate was very slow, probably because the Br ₂ concentration was 5%, which is half. “10% Br ₂ -AcMe”
Although the dissolution amount is smaller in the case of dissolution time of 10 minutes than in the case of MeOH or AcN, it is considered that this is because the surface passivation film of SUS316 steel is involved. Therefore, once the reaction occurs, the dissolution is rapidly progressing, and the solubility is “10
% Br ₂ -MeOH ”and“ 10% Br ₂ -AcN ”.

In the present invention, as a second step, insoluble residues (oxides containing radioactive substances) are separated from the solution of metal-based waste. As described above, uranium and radioactive fission products are converted into oxides in the reactor core tube. And the amount adhered to feed pipes of water, steam, and water, and the amount of the attached matter greatly varies depending on the position of the pipe. Therefore, the recovery rate cannot be quantitatively evaluated from these actual products. However, since a proper quantitative evaluation of the recovery rate can be performed by performing a recovery test of oxides contained in the same metal material as the furnace tube, a recovery confirmation test of oxides in steel of a SUS316 steel sample was performed. .

In the recovery confirmation test, a SUS316 steel sample having a known oxide form in steel was used, and the total oxygen (TO) in the steel was used.
The amount of - while seeking in "inert gas fusion thermal conductivity measurement method", was dissolved a certain amount of the sample in the "10% Br ₂ -MeOH" mixed solvent (60 ° C.), microporous filter insoluble residue And collected by filtration using ICP-AES (inductively coupled plasma).
Emission spectroscopy) to determine metal components, calculate the amount of oxygen from each oxide form, and compare the total amount with the previously determined total oxygen (TO) value in steel. OK.
Table 4 shows the results.

[0027]

[Table 4]

From the results shown in Table 4, the total oxygen in steel (T
It can be confirmed that the value of O) and the value converted from the oxide are in good agreement with each other, and therefore, the radioactive material (ie, oxide) adhering from the metal-based waste in the nuclear industry by this method can be confirmed. It can be seen that it can be almost completely recovered. In addition,
It was also confirmed that it was possible to separate oxides even in a trace area where the amount of oxides was 20 ppm or less, and that 100% of the oxides could be recovered.

In the present invention, as a third step, a solution (halogen-organic solvent mixture) from which insoluble residues (oxides) have been removed
The metal component is electrodeposited and recovered from SUS316. To confirm the effect, about 3 g of a SUS316 pipe was weighed with 10% Br ₂ -Me.
The solution dissolved in 100 ml of "OH" mixed solvent was used as an electrolytic solution, and electrolysis was performed using an electrolytic cell as shown in FIG. For the electrolysis, a platinum wire of 2 mmφ was connected to the anode 1 and 30 mm width × 80 mm.
A platinum plate having a length of 0.2 mm and a thickness of 2 mm was used as the cathode 2 and a direct current of 0.5 to 1 A was passed.

As a result, a black precipitate is obtained on the cathode,
It was confirmed that all of them adhered to the magnet. Also,
When the amount of deposits increased to some extent, it was also observed that the tip portion was dissolved again. Table 5 shows the chemical analysis values of the recovered black precipitate.

[0031]

[Table 5]

From the results shown in Table 5, it can be seen from the solution (halogen-organic solvent mixture) that the metallic components of the waste (Fe,
It can be seen that Ni and Cr) can be sufficiently recovered.

In the results shown in Table 5, the total amount of Fe, Ni and Cr in the deposit is about 85%.
It is considered that C, O, etc. were contained in large amounts. In fact, the EPMA (electron probe microanalyzer) shown in FIGS.
Examination of the results of qualitative analysis by lysis (EDX) shows that these elements are detected. In addition, "Observation of Electrodeposition State and Electrodeposit by Video Microscope", "SEM of Electrodeposit by EPMA"
Investigations were also made taking into account the results of "observation" and "identification by X-ray diffraction method". As a result, it was found that the deposit was a metal containing α-Fe as a main component, and that Fe, Ni and Cr could be recovered. However, since this precipitate contains a large amount of Br, C, O, etc., it is necessary to purify the recovered metal for practical use.

When the halogen component and the organic solvent component of the electrolytic solution after the metal component of the waste are deposited and recovered are adjusted and reused as the waste solution, a particular disadvantage occurs. It was also confirmed that a sufficient effect was obtained. Next, the present invention will be described more specifically with reference to examples.

[0035]

EXAMPLE First, SUS3 having the chemical composition shown in Table 1 was used.
A 16-piece pipe (outer diameter 6 mmφ × wall thickness 0.5 mm) was prepared, and a halogen mixed solvent having two Br concentrations of “5% Br ₂ -MeOH” and “10% Br ₂ -MeOH” was prepared.

From the above pipe, length: 3.5 to 4.5 cm
(2.5 to 3.5 g) of the dissolved sample was cut into round slices, placed in a flask together with 20 ml of each mixed solvent, immersed and held at room temperature or 50 ° C. for a certain period of time, and then the dissolved sample was taken out and weighed to determine the amount of the dissolved sample. . Table 6 shows the measurement results of the dissolved amount at this time.

[0037]

[Table 6]

The following were confirmed from the results shown in Table 6. (a) The higher the Br ₂ concentration, the faster the dissolution of the sample and the amount of dissolution increases, and (b) Even at the same Br ₂ concentration, the higher the solution temperature, the higher the dissolution rate. ) The amount of dissolution is determined by the Br ₂ concentration regardless of the liquid temperature.
Steel) with Ni: 13%, Cr: 16%, Fe: 71%, and Br
Ni, Cr and Fe dissolved by NiBr ₂ , CrBr ₃ and Fe
If it is present as Br ₃ , the amount of the sample dissolved in 20 ml of a 10% Br ₂ mixed solvent is about 1.6 g, which indicates that SUS316 and Br react almost stoichiometrically. Therefore, the amount of dissolved sample is determined by the Br ₂ concentration.

Then, this time, as a solution, “10% Br _2-
Using the "MeOH mixed solvent", a treatment test of a sample imitating "metal-based waste containing radioactive materials in the nuclear industry" was performed.
Here, “a sample imitating“ metal-based waste containing radioactive materials in the nuclear industry ”” includes synthetic simulated fission products (CrO ₃ , Fe ₂ O ₃ , NiO, SeO ₂ , Rb ₂ O, StO, Y ₂ O
₃ , ZrO ₂ , MoO ₃ , TcO ₂ , RuO ₂ , Rh ₂ O ₃ , PdO,
Ag ₂ O, CdO, In ₂ O ₃ , SnO, Sb ₂ O ₃ , TeO ₂ , Cs _{2
O, BaO, La 2 O 3} , CeO 2, Pr 6 O 11, Nd 2 O 3, Sm 2
A mixture of 0.500 g of a fine powder of a mixture of O ₃ , Eu ₂ O ₃ , Gd ₂ O ₃ and ReO ₂ which had been roasted) and about 5 g of a cut piece of the SUS316 pipe were used. In addition,
"Samples simulating" metal-based waste containing radioactive materials in the nuclear industry "" were adjusted to three samples.
The total oxygen content of 0 g of each synthetic simulated fission product was determined by "inert gas melting-thermal conductivity measurement method", and the values were 0.236%, 0.238% and 0.234%, respectively.

This sample (fine powder of synthetic simulated fission product)
0.500 g and a mixture of about 5 g of a cut piece of a SUS316 pipe) in the treatment test,
And add 200 ml of “10% Br ₂ -MeOH mixed solvent”
After the addition, the mixture was heated at 60 ° C. for about 1 hour to completely dissolve SUS316.

Next, the insoluble residue was separated from the solution by filtration with a microporous filter, washed, and dried to recover the oxide powder. Then, the constituent metal components of the oxide powder are quantified by “ICP (inductively coupled plasma) -emission spectroscopy” and “atomic absorption spectrometry”, and then converted into the respective oxide forms constituting the simulated fission product. And their oxygen content was determined. As a result, the calculated total oxygen amounts for the three samples were 0.238%, 0.236% and 0.235%, respectively.
And the total oxygen amount measured before the treatment showed a very good agreement, and it was confirmed that the oxide was quantitatively recovered without being dissolved. In addition, this point (the oxide was not dissolved) was confirmed from the fact that the recovered weights were 0.503 g, 0.497 g, and 0.508 g, respectively (note that oxides and the like in SUS316 were not dissolved). Corrected for insolubles).

Next, the solution obtained by filtering and separating the insoluble residue was subjected to electrolytic treatment in the electrolytic cell shown in FIG. The electrolysis conditions at this time were as follows. Electrolysis current: 0.5-1A, electrolyte temperature: 24 ° C, electrolysis time: about 150 minutes. As a result, a black precipitate was obtained on the cathode. The chemical analysis values of the deposit are as shown in Table 7, and the waste metal (metal component of SUS316 steel) dissolved from the solution was sufficient. It can be seen that it has been recovered.

[0043]

[Table 7]

Subsequently, Br ₂ was added to the electrolytic solution after the electrolytic treatment.
The source (Br ₂ 100%) and MeOH were replenished and the solution composition was changed to “10%
Br ₂ -MeOH "become as to component adjustment, was re-used as solution of the waste, it can be sufficiently utilized as a solution was confirmed.

SUS having an outer diameter of 6 mmφ and a thickness of 0.5 mm
The volume of 1 kg of 316 pipe is 375 cm ³ , and the net volume is 125 cm ³ even if the pipe is completely compressed. Therefore, from the above test results, the application of the method of the present invention is 3 times more than the case where the treatment is performed in the pipe shape.
It can be expected that the volume reduction can be expected to be 75 cm ³ / kg or more, and 125 cm ³ / kg or more as compared with the case of compressing and processing, and that the recovered metal can be sufficiently reused because the radioactive substance has been removed.

[0046]

[Summary of Effects] As described above, according to the present invention, it is possible to increase the volume reduction rate of waste containing radioactive materials in the nuclear industry to the utmost, and to solve the restrictions on storage locations all at once. Industrially extremely useful effects are obtained, for example, the construction of a new high-performance reactor can be facilitated.

[Brief description of the drawings]

FIG. 1 is a schematic explanatory view of an electrolytic cell used in a test.

FIG. 2 is a graph showing the results of qualitative analysis by EDX of electrolytic deposits deposited in the test.

FIG. 3 EDX of another part of the electrolytic deposit deposited in the test
6 is a graph showing the results of qualitative analysis by.

[Explanation of symbols]

 1 anode 2 cathode 3 electrolyte

────────────────────────────────────────────────── ─── Continued on the front page (51) Int.Cl. ⁷ Identification code FI G21F 9/30 561 G21F 9/30 561F // C22B 7/00 C22B 7/00 G (72) Inventor Yasuo Inokuma Amagasaki City, Hyogo Prefecture 1-8 Fuso-cho Sumitomo Metal Technology Co., Ltd. (58) Field surveyed (Int.Cl. ⁷ , DB name) G21F 9/28 G21F 9/30

Claims (2)

(57) [Claims] 1. Dissolving a metal-based waste containing a radioactive substance in the nuclear industry with a mixed solution of a halogen and an organic solvent, separating an insoluble residue, and then electrolyzing the remaining solution to recover a metal component as a metal. A method for treating metal-based waste from the nuclear industry. 2. After dissolving a metallic waste containing a radioactive substance in the nuclear industry with a mixed solution of a halogen and an organic solvent, an insoluble residue is separated, and then the remaining solution is electrolyzed to recover a metal component as a metal. And a method for treating metal-based wastes from the nuclear industry by adjusting the components of the electrolyte and reusing it as a solution.