KR102436254B1 - Method for treating spent nuclear fuel comprising strontium, barium or cesium - Google Patents

Abstract

The present invention provides a method comprising the steps of: (a) converting a spent nuclear fuel oxide containing at least one of strontium, barium, and cesium into spent nuclear fuel chloride by performing a chlorination reaction in molten salt at 600° C. or less; and (b) recovering and distilling the converted spent nuclear fuel chloride by an electrochemical method.

Description

A method for treating spent nuclear fuel containing strontium, barium or cesium

The present invention relates to a method for the treatment of spent nuclear fuel comprising strontium, barium or cesium.

As of December 2019, a total of 482,592 bundles of spent nuclear fuel generated at the nuclear power plant are expected, and from 2021, the temporary storage facility in the Wolseong heavy water reactor is expected to be saturated. Although the government is making efforts to solve the national issue of spent nuclear fuel accumulation and secure a repository, it is difficult to secure good deep bedrock in a wide area due to the narrow land and high population density.

According to the prior art, in order to remove representative fission products in spent nuclear fuel, a process of volatilization and adsorption can be performed through high-temperature pretreatment. there was Therefore, as a representative fission product, there is a need for research to effectively treat spent nuclear fuel containing strontium, barium or cesium.

Domestic Patent Publication No. 10-0766516 (2007. 10. 05.)

The present invention provides a solid-liquid reaction to a spent nuclear fuel oxide comprising at least one of strontium, barium and cesium for treating at least one of strontium, barium and cesium with a high removal rate by subjecting a chlorination reaction to a high removal rate, (a) converting the spent nuclear fuel oxide containing at least one of strontium, barium and cesium into spent nuclear fuel chloride by performing a chlorination reaction in molten salt at a temperature of 600° C. or less; and (b) recovering and distilling the converted spent nuclear fuel chloride by an electrochemical method.

However, the technical problem to be achieved by the present invention is not limited to the above-mentioned problems, and other problems not mentioned will be clearly understood by those skilled in the art from the following description.

The present invention provides a method comprising the steps of: (a) converting a spent nuclear fuel oxide containing at least one of strontium, barium, and cesium into spent nuclear fuel chloride by performing a chlorination reaction in molten salt at 600° C. or less; and (b) recovering and distilling the converted spent nuclear fuel chloride by an electrochemical method.

According to the present invention, a chlorination reaction may be applied to a spent nuclear fuel oxide comprising at least one of strontium, barium and cesium as a solid-liquid reaction to treat at least one of strontium, barium and cesium with a high removal rate. Therefore, it is expected that the disposal area of spent nuclear fuel can be reduced and safety can be enhanced.

1 is a schematic diagram schematically illustrating a method for treating spent nuclear fuel according to an embodiment of the present invention.
2 (a) and (b) are graphs showing the Sr concentration or the Ba concentration by time using ICP as a result of performing the chlorination reaction according to Examples 1 and 2;

In particular, the spent nuclear fuel oxides generated in light water reactors mainly contain SrO, BaO, Cs ₂ O, and the like. Sr, Ba, or Cs, in particular, Sr and Ba, have a high boiling point, making it difficult to treat them through existing processes. Therefore, the present invention was completed by applying an optimized chlorination reaction to effectively treat Sr, Ba, and Cs.

As used herein, the term “spent nuclear fuel oxide” is characterized as an oxide containing at least one of cesium, strontium, and barium. In some cases, the spent fuel oxide includes at least two of strontium, barium, and cesium. It may be an oxide or an oxide containing all of strontium, barium and cesium. In this case, strontium or cesium corresponds to a highly exothermic fission product. In addition, the spent nuclear fuel oxide may include various metal oxides such as UO _{2 ,} U ₃ O ₈ , Ta ₂ O ₅ , and TiO ₂ . In this case, the spent nuclear fuel oxide may be generated in both light and heavy water reactors.

Hereinafter, the present invention will be described in detail.

The present invention provides a method comprising the steps of: (a) converting a spent nuclear fuel oxide containing at least one of strontium, barium, and cesium into spent nuclear fuel chloride by performing a chlorination reaction in molten salt at 600° C. or less; and (b) recovering and distilling the converted spent nuclear fuel chloride by an electrochemical method.

1 is a schematic diagram schematically illustrating a method for treating spent nuclear fuel according to an embodiment of the present invention.

As shown in FIG. 1 , the method for treating spent nuclear fuel according to an embodiment of the present invention includes the steps of pretreating the spent nuclear fuel oxide containing at least one of strontium, barium, and cesium (S10); converting it into spent nuclear fuel chloride by performing a chlorination reaction in molten salt at a temperature of 600° C. or less (S20); and distilling it after recovering it by an electrochemical method (S30).

First, the method for treating spent nuclear fuel according to the present invention may include a step [step (a)] of pretreating the spent nuclear fuel oxide containing at least one of strontium, barium, and cesium.

The pretreatment is to prepare the spent nuclear fuel oxide in powder form, and specifically, the pretreatment is performed in an air atmosphere having an oxygen concentration of 10% to 60%, and the temperature is increased at a rate of 1°C/min to 10°C/min. Next, it may be carried out through the step of heat treatment at a temperature of 300 ℃ to 700 ℃ for 10 minutes to 90 minutes. Through the optimized pretreatment conditions as described above, the spent nuclear fuel oxide containing cesium, strontium or barium can be controlled in a powder form having an average particle size of 10 nm to 100 μm, preferably 0.5 μm to 50 μm, There is an advantage that can maximize the chlorination reaction efficiency to be described later.

Next, the spent nuclear fuel treatment method according to the present invention performs a chlorination reaction in a molten salt of 600° C. or less on a spent nuclear fuel oxide containing at least one of strontium, barium, and cesium to give spent nuclear fuel chloride converting to .

The spent nuclear fuel oxide may be in the form of a powder having an average particle size of 10 nm to 100 μm, thereby maximizing the efficiency of the chlorination reaction.

The chlorination reaction corresponds to a solid-liquid reaction carried out in molten salt at 600° C. or less. It is characterized in that the spent nuclear fuel oxide is converted into the spent nuclear fuel chloride through the chlorination reaction. In particular, it is characterized in that at least one of strontium oxide, barium oxide and cesium oxide is converted into at least one of strontium chloride, barium chloride and cesium chloride. characterized in that In this case, the spent nuclear fuel chloride may be in a liquid state.

Specifically, through the chlorination reaction, strontium oxide can be converted into strontium chloride as shown in the following formula (1):

2 LiCl + SrO → Li ₂ O + SrCl ₂ (1)

As shown in Table 1 below, in the case of the strontium oxide, it is confirmed that the highest temperature at which G is a negative value (that is, a spontaneous reaction occurs) is 600° C., and at a temperature higher than that, it is confirmed that the spontaneous reaction does not occur.

Temperature (℃) ΔH (kJ) ΔS (J/K) ΔG (kJ) K 0 -22.897 -20.707 -17.241 1.983E+003 100 -23.906 -23.898 -14.989 1.254E+002 200 -24.568 -25.479 -12.513 2.407E+001 300 -25.121 -26.541 -9.909 8.001E+000 400 -25.621 -27.346 -7.213 3.628E+000 500 -25.942 -27.795 -4.452 1.999E+000 600 -25.879 -27.725 -1.671 1.259E+000 700 -65.194 -72.276 5.141 5.297E-001

In addition, through the chlorination reaction, barium oxide can be converted into barium chloride as shown in the following formula (2):

LiCl + BaO → Li ₂ O + BaCl ₂ (2)

As shown in Table 2 below, in the case of the barium oxide, when the temperature of the chlorination reaction is 600° C. or less, it is confirmed that the spontaneous reaction occurs sufficiently well. On the other hand, although a spontaneous reaction occurs even at a temperature higher than that, there is a problem in that the reactivity is lowered.

Temperature (℃) ΔH (kJ) ΔS (J/K) ΔG (kJ) K 0 -88.184 -28.048 -80.522 2.510E+015 100 -89.435 -31.994 -77.497 7.065E+010 200 -90.392 -34.271 -74.176 1.547E+008 300 -91.280 -35.976 -70.661 2.756E+006 400 -92.135 -37.351 -66.992 1.581E+005 500 -92.917 -38.436 -63.200 1.863E+004 600 -93.558 -39.218 -59.315 3.537E+003 700 -133.974 -84.958 -51.297 5.671E+002

In addition, through the chlorination reaction, cesium oxide can be converted into cesium chloride as shown in the following formula (3):

Cs ₂ O + 2LiCl → 2CsCl + Li ₂ O (3)

As shown in Table 3 below, in the case of the cesium oxide, it is confirmed that the spontaneous reaction occurs sufficiently well when the temperature of the chlorination reaction is 600° C. or less. On the other hand, although a spontaneous reaction occurs at a temperature higher than that, there is a problem in that the reactivity is relatively lowered.

Temperature (℃) ΔH (kJ) ΔS (J/K) ΔG (kJ) K 0 -319.235 -24.385 -312.574 6.009E+059 100 -320.353 -27.917 -309.936 2.451E+043 200 -321.198 -29.925 -307.039 7.929E+033 300 -322.157 -31.757 -303.955 5.054E+027 400 -323.368 -33.698 -300.684 2.159E+023 500 -338.574 -53.355 -297.323 1.228E+020 600 -339.484 -54.460 -291.932 2.923E+017 700 -338.770 -55.468 -284.791 1.939E+015 800 -337.827 -54.551 -279.286 3.937E+013 900 -336.209 -53.114 -273.898 1.572E+012 1000 -333.867 -51.203 -268.679 1.057E+011

In other words, the temperature of the chlorination reaction is 600 ° C or lower, preferably 500 ° C or lower, more preferably 400 ° C or lower while maintaining the molten salt in a molten state (that is, above the melting point of the molten salt). characterized in that

The molten salt for performing the chlorination reaction may be a eutectic salt including LiCl, and may react with at least one of strontium oxide, barium oxide, and cesium oxide present in the spent nuclear fuel oxide. Specifically, the eutectic salt may be a LiCl-KCl eutectic salt. Specifically, considering that the melting point of LiCl-KCl (eutectic salt) is 353 °C, the temperature of the chlorination reaction may be 353 °C to 600 °C or less. On the other hand, in the case of a eutectic salt that does not contain LiCl, for example, in the case of a KCl-NaCl eutectic salt, it does not thermodynamically react with at least one of strontium oxide, barium oxide, and cesium oxide, so it is a molten salt for performing the chlorination reaction. Can not use it. In addition, since the melting point of LiCl (solar molten salt) is 605°C, it cannot be used as a molten salt for performing the chlorination reaction.

Next, the method for treating spent nuclear fuel according to the present invention includes the step of recovering and distilling the converted spent nuclear fuel chloride by an electrochemical method [step (b)].

The converted spent nuclear fuel chloride can be recovered by reducing it by an electrochemical method, and the recovery by the electrochemical method is an electrochemical method, and the recovery is performed using a liquid electrode, but corresponding to the reduction potential of Li ⁺ cation About -2.5 V (vs. Ag/AgCl), which is a positive voltage, it may be carried out at -0.5 V to -2.5 V conditions. On the other hand, if it is performed at a voltage more negative than about -2.5 V (vs. Ag/AgCl) corresponding to the reduction potential of Li ⁺ cations, lithium is co-electrodeposited in addition to one or more of strontium, barium, and cesium, and then lithium is separated separately Only then can each element be recycled independently and disposal stability can be improved. Therefore, it is necessary to adjust the applied voltage so that lithium is not electrodeposited.

Specifically, after the chlorination reaction is completed, the converted spent nuclear fuel chloride dissolved molten salt (hereinafter, "extraction molten salt") is maintained at 500° C. to 600° C., and then, Bi, Zn, Sn Materials such as liquid anode, glassy carbon, Pt, etc. Anode and reference electrode are immersed, respectively. Then, a voltage more positive than the reduction potential of the extraction molten salt Li ⁺ cation is applied to the liquid negative electrode to form an intermetallic compound with only one or more of strontium, barium, and cesium with the material of the liquid electrode (Bi, etc.) can do it On the other hand, if a voltage negative than the reduction potential of the extracted molten salt Li ⁺ cation is applied to the liquid negative electrode, at least one of strontium, barium, and cesium is recovered together with the lithium-liquid electrode material (Bi, etc.). Meanwhile, chlorine gas may be generated at the anode.

Thereafter, the recovered material may be distilled. Considering that the recovered material is an intermetallic compound formed from a nuclear fission product and a material (such as Bi) of a liquid electrode, the distillation is It may be carried out at a temperature higher than the melting point of the material, and specifically, it may be carried out at a temperature of 250 °C to 1200 °C, preferably at a temperature of 300 °C to 1200 °C. Therefore, only one or more of strontium, barium and cesium can be finally recovered.

According to the present invention, a chlorination reaction can be applied to a spent nuclear fuel oxide containing strontium or barium as a solid-liquid reaction to treat strontium, barium, or cesium with a high removal rate. Therefore, it is expected that the disposal area of spent nuclear fuel can be reduced and safety can be enhanced.

Hereinafter, preferred examples are presented to help the understanding of the present invention. However, the following examples are only provided for easier understanding of the present invention, and the contents of the present invention are not limited by the following examples.

[Example]

Example 1

The pre-treatment was carried out on a powder-type sample containing SrO [average particle size = 0.5-50 μm], and a chlorination reaction was performed for 1 to 3 hours in LiCl-KCl eutectic salt at about 500 ° C (LiCl- Concentration of SrO in KCl eutectic salt = 0.064 to 0.557 wt%), analysis was performed using ICP, and the results are shown in FIG. 2(a).

As shown in FIG. 2(a), the initial (ie, before chlorination) concentration of Sr in the LiCl-KCl eutectic salt is 0.054 to 0.471 wt% (in this case, Sr is an element present in the starting material SrO). do. Under the premise that the starting material SrO is 100% converted to SrCl ₂ due to the presence of excess LiCl, the concentration of Sr after 1 hour in LiCl-KCl eutectic salt and the concentration of Sr after 3 hours in LiCl-KCl eutectic salt even if measurement error is taken into account Considering that the concentrations are similar to each other (in this case, Sr is an element present in the product SrCl ₂ ), it can be seen that the chlorination reaction was completed within 1 hour.

After that, the extraction molten salt is maintained at about 500° C., and then, a Bi material liquid cathode, a glassy carbon material anode, and a reference electrode are immersed therein, respectively. Then, after applying a voltage of -0.5 to -2.5 V to the Bi-material liquid anode, it was distilled under reduced pressure at a temperature of about 300° C. to finally recover only Sr.

Example 2

The pre-treatment was carried out on a sample in the form of a powder containing BaO [average particle size = 0.5-50 μm], and a chlorination reaction was performed for 1 to 3 hours in LiCl-KCl eutectic salt at about 500 ° C (LiCl- Concentration of BaO in KCl eutectic salt = 0.153 to 1.363 wt%), analysis was performed using ICP, and the results are shown in FIG. 2(b).

As shown in FIG. 2(b), the initial concentration of Ba in the LiCl-KCl eutectic salt (that is, before the chlorination reaction) is 0.137 to 1.220 wt% (in this case, Ba is an element present in BaO, the starting material). do. Under the premise that the starting material BaO is 100% converted to BaCl ₂ due to the presence of excess LiCl, the concentration of Ba in the LiCl-KCl eutectic salt after 1 hour and the Ba concentration after 3 hours in the LiCl-KCl eutectic salt even if the measurement error is considered. Considering that the concentrations are similar to each other (in this case, Sr is an element present in the product BaCl ₂ ), it can be seen that the chlorination reaction was completed within 1 hour.

After that, the extraction molten salt is maintained at about 500° C., and then, a Bi material liquid cathode, a glassy carbon material anode, and a reference electrode are immersed therein, respectively. Then, after applying a voltage of -0.5 to -2.5 V to the Bi-material liquid anode, it was distilled under reduced pressure at a temperature of about 300° C. to finally recover only Ba.

Example 3

Through pretreatment, a chlorination reaction was performed for 1 to 3 hours in a LiCl _- KCl eutectic salt at about 500 ° C. Concentration of Cs ₂ O in LiCl-KCl eutectic salt = 0.124 to 1.113 wt%), analysis was performed using ICP. Although not shown in the drawings, it is confirmed that the chlorination reaction was all completed within 1 hour.

After that, the extraction molten salt is maintained at about 500° C., and then, a Bi material liquid cathode, a glassy carbon material anode, and a reference electrode are immersed therein, respectively. Then, after applying a voltage of -0.5 to -2.5 V to the Bi-material liquid anode, it was distilled under reduced pressure at a temperature of about 300° C. to finally recover only Cs.

The foregoing description of the present invention is for illustration, and those of ordinary skill in the art to which the present invention pertains can understand that it can be easily modified into other specific forms without changing the technical spirit or essential features of the present invention. will be. Therefore, it should be understood that the embodiments described above are illustrative in all respects and not restrictive.

Claims (9)

(a) converting a spent nuclear fuel oxide containing at least one of strontium, barium, and cesium into spent nuclear fuel chloride by performing a chlorination reaction in molten salt at a temperature of 600° C. or less; and
(b) recovering and distilling the converted spent nuclear fuel chloride by an electrochemical method;
In step (a), the spent nuclear fuel oxide is heated at a rate of 1°C/min to 10°C/min in an air atmosphere having an oxygen concentration of 10% to 60%, and then at a temperature of 300°C to 700°C for 10 minutes. A method of treating spent nuclear fuel that has been pre-treated through the step of heat-treating for 90 minutes.
According to claim 1,
The spent nuclear fuel oxide in step (a) is a method for treating spent nuclear fuel, characterized in that it is in the form of a powder having an average particle size of 10 nm to 100 μm.
delete According to claim 1,
In step (a), the molten salt is a eutectic salt containing LiCl, and it is characterized in that it reacts with at least one of strontium oxide, barium oxide and cesium oxide present in the spent nuclear fuel oxide. .
5. The method of claim 4,
The eutectic salt is a LiCl-KCl eutectic salt, characterized in that, spent nuclear fuel treatment method.
According to claim 1,
In the step (a), the spent nuclear fuel chloride is in a liquid state, characterized in that the spent nuclear fuel treatment method.
According to claim 1,
In the step (a), the spent nuclear fuel chloride comprises at least one of strontium chloride, barium chloride and cesium chloride.
According to claim 1,
The method of treating spent nuclear fuel, characterized in that the recovery by the electrochemical method in step (b) is performed using a liquid electrode, but under -0.5 V to -2.5 V conditions.
According to claim 1,
The method of treating spent nuclear fuel, characterized in that the distillation in step (b) is performed at a temperature of 250 °C to 1200 °C.

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