CA2741076C - Method for preparation of uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps - Google Patents
Method for preparation of uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps Download PDFInfo
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Abstract
A method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps is provided, which includes dissolving uranium by adding carbonate solution with hydrogen peroxide into the uranium oxide scraps (step 1); adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution (step 2); recovering the acid and the alkaline solutions used at step 2 (step 3); dehydrating agglomerated uranium oxide powder precipitated at step 2 under argon atmosphere, reductive heat treatment the agglomerated powder under hydrogen atmosphere, and passivation of forming a protective oxide film(step 4); milling the agglomerated uranium oxide powder prepared at step 4 (step 5); and mixing the uranium oxide power milled at step 5 and virgin uranium oxide powder (step 6).
Description
METHOD FOR PREPARATION OF URANIUM OXIDE POWDER FOR
NUCLEAR FUEL PELLET FABRICATION FROM URANIUM OXIDE
SCRAPS
BACKGROUND OF THE INVENTION
1. Field ofthe Invention The present invention relates to a method for preparation of uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps.
NUCLEAR FUEL PELLET FABRICATION FROM URANIUM OXIDE
SCRAPS
BACKGROUND OF THE INVENTION
1. Field ofthe Invention The present invention relates to a method for preparation of uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps.
2. Description of the related art Nuclear fuel pellet for heavy water reactor and light water reactor of nuclear power plant uses UO2 pellets with 0.7 wt% U-235 or UO2 pellets with 4-5 wt% of U-235.
The nuclear fuel pellets are formed in a series of fabrication steps including preliminary forming in which UO2 powder is mixed with various additives, granulation, lubricant adding, compacting, sintering at 1650 ¨ 1750 t under hydrogen atmosphere, and final grinding. In the fabrication process of nuclear fuel pellet, uranium oxide scraps are generated, which include defective UO2 pellets, or UO2 pellets containing impurities of grinding sludge, or UO2 powder with impurities.
The impurities mixed with the uranium scraps are metals including Cr, Fe, Ni, Mo, Al, Si or the oxides thereof.
To make UO2 powder as nuclear fuel pellet, it is important for the pellets to have 95% or more theoretical density. In many cases, nuclear fuel scraps are stored rather than recycled. However, considering the increasing use of atomic energy worldwide and subsequent demands for uranium consumption, there will be increasing demands for refining and recycling of uranium scraps generated during the fabrication process of uranium nuclear fuel pellet. Conventionally, recovering uranium from uranium oxide scraps involves dissolving uranium oxide scraps in high-temperature nitric acid (approximately of 100 C, 10 M HNO3), separating only the uranium by solvent extraction, recovering uranium as ammonia-uranium mixed precipitate ((NH4). 2U207 or U04 . 2NI-14NO3) using aqueous ammonia, decomposing into UO3 powder, and forming into UO2 powder by reductive heat treatment.
However, the conventional method to recover uranium from uranium oxide scrap has problems in the step of dissolving uranium oxide using high temperature nitric acid, which include corrosion of equipment, generation of NO gas, generation of a large amount of secondary waste such as organic liquid waste due to solvent extraction, and ammonia nitrogen solution that is under environmental regulation.
Therefore, the conventional method has shortcomings of deteriorated economics and environmental friendliness.
Accordingly, while studying the method to solve the problems associated with the conventional method for recovering uranium from uranium oxide scraps mixed with metallic impurities generated from nuclear fuel fabrication, which mainly include corrosion of equipment, generation of NO gas, deterioration of operation stability due to use of high temperature acid, generation of organic liquid waste due to solvent extraction, and deteriorated economics due to requirement for treatment facilities, the present inventors have developed a method of preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps with higher operation stability, and minimized generation of the secondary waste in the treatment of the uranium scraps to provide improved environmental friendliness compared to the conventional method, through the use of dissolution of the uranium scraps in carbonate solution with hydrogen peroxide at room temperature, U04 precipitate by regulating pH of carbonate containing uranium ion, and electrolytic recycle of used carbonates. To make UO2 powder for nuclear fuel pellet from the uranium precipitate(U04) obtained from uranium scraps, it is important for the U04 powder to convert to UO2 powder with suitable powder characteristics satisfying the requirement of UO2 fuel pellet such as theoretical density. In present invention, a method for preparing sinterable UO2 powder for fabrication of UO2 pellet from the U04 obtained from uranium oxide scraps is given.
SUMMARY OF THE INVENTION
In one embodiment, a method for preparing UO2 powder for nuclear fuel pellet fabrication from uranium oxide scraps is provided, which may include:
dissolving uranium by adding carbonate solution with hydrogen peroxide into the uranium oxide scraps (step 1); adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution (step 2); recovering the acid and the alkaline solutions used at step 2 (step 3); dehydrating agglomerated uranium oxide
The nuclear fuel pellets are formed in a series of fabrication steps including preliminary forming in which UO2 powder is mixed with various additives, granulation, lubricant adding, compacting, sintering at 1650 ¨ 1750 t under hydrogen atmosphere, and final grinding. In the fabrication process of nuclear fuel pellet, uranium oxide scraps are generated, which include defective UO2 pellets, or UO2 pellets containing impurities of grinding sludge, or UO2 powder with impurities.
The impurities mixed with the uranium scraps are metals including Cr, Fe, Ni, Mo, Al, Si or the oxides thereof.
To make UO2 powder as nuclear fuel pellet, it is important for the pellets to have 95% or more theoretical density. In many cases, nuclear fuel scraps are stored rather than recycled. However, considering the increasing use of atomic energy worldwide and subsequent demands for uranium consumption, there will be increasing demands for refining and recycling of uranium scraps generated during the fabrication process of uranium nuclear fuel pellet. Conventionally, recovering uranium from uranium oxide scraps involves dissolving uranium oxide scraps in high-temperature nitric acid (approximately of 100 C, 10 M HNO3), separating only the uranium by solvent extraction, recovering uranium as ammonia-uranium mixed precipitate ((NH4). 2U207 or U04 . 2NI-14NO3) using aqueous ammonia, decomposing into UO3 powder, and forming into UO2 powder by reductive heat treatment.
However, the conventional method to recover uranium from uranium oxide scrap has problems in the step of dissolving uranium oxide using high temperature nitric acid, which include corrosion of equipment, generation of NO gas, generation of a large amount of secondary waste such as organic liquid waste due to solvent extraction, and ammonia nitrogen solution that is under environmental regulation.
Therefore, the conventional method has shortcomings of deteriorated economics and environmental friendliness.
Accordingly, while studying the method to solve the problems associated with the conventional method for recovering uranium from uranium oxide scraps mixed with metallic impurities generated from nuclear fuel fabrication, which mainly include corrosion of equipment, generation of NO gas, deterioration of operation stability due to use of high temperature acid, generation of organic liquid waste due to solvent extraction, and deteriorated economics due to requirement for treatment facilities, the present inventors have developed a method of preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps with higher operation stability, and minimized generation of the secondary waste in the treatment of the uranium scraps to provide improved environmental friendliness compared to the conventional method, through the use of dissolution of the uranium scraps in carbonate solution with hydrogen peroxide at room temperature, U04 precipitate by regulating pH of carbonate containing uranium ion, and electrolytic recycle of used carbonates. To make UO2 powder for nuclear fuel pellet from the uranium precipitate(U04) obtained from uranium scraps, it is important for the U04 powder to convert to UO2 powder with suitable powder characteristics satisfying the requirement of UO2 fuel pellet such as theoretical density. In present invention, a method for preparing sinterable UO2 powder for fabrication of UO2 pellet from the U04 obtained from uranium oxide scraps is given.
SUMMARY OF THE INVENTION
In one embodiment, a method for preparing UO2 powder for nuclear fuel pellet fabrication from uranium oxide scraps is provided, which may include:
dissolving uranium by adding carbonate solution with hydrogen peroxide into the uranium oxide scraps (step 1); adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution (step 2); recovering the acid and the alkaline solutions used at step 2 (step 3); dehydrating agglomerated uranium oxide
3 powder precipitated at step 2 under argon atmosphere, reductive heat treatment of the agglomerated powder under hydrogen atmosphere, and passivation of forming a protective oxide film (step 4); milling the agglomerated uranium oxide powder prepared at step 4 (step 5); and mixing the uranium oxide power milled at step 5 and virgin uranium oxide powder (step 6). The method for preparing UO2 powder for nuclear fuel pellet from uranium oxide scraps in one embodiment recovers uranium using simple precipitation method with high efficiency, greatly improves environmental-friendliness by not generating secondary waste due to circulation of inorganic salt solution including carbonate, acid, or alkali, etc., and satisfies the density specification of UO2 pellet for light water reactor and heavy water reactor of nuclear power plants using simplified process. Accordingly, it is possible to efficiently recycle the uranium oxide scraps mixed with metallic impurity sludge generated from nuclear fuel fabrication process as resources, which are otherwise disposed of as waste materials.
BRIEF DESCRIPTION OF THE DRAWINGS
FIG. I illustrates a flow diagram of a method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps according to an embodiment of the present invention;
FIG. 2 illustrates a flow diagram illustrating in detail a process of uranium precipitates and recovering carbonate in FIG. 1;
FIG. 3 illustrates a process of electrolysis of the acid and alkali recovering of FIG. 1;
BRIEF DESCRIPTION OF THE DRAWINGS
FIG. I illustrates a flow diagram of a method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps according to an embodiment of the present invention;
FIG. 2 illustrates a flow diagram illustrating in detail a process of uranium precipitates and recovering carbonate in FIG. 1;
FIG. 3 illustrates a process of electrolysis of the acid and alkali recovering of FIG. 1;
4 FIG. 4 illustrates a graphical representation of solubility of uranium oxide scraps containing impurities in hydrogen peroxide-carbonate solution;
FIG. 5 illustrates a graphical representation of uranium concentration varying according to the change in pH of uranyl peroxo carbonate complex;
FIG. 6 illustrates a graphical representation of the result of X-ray diffractometer (XRD) of uranium agglomerate precipitates;
FIG. 7 illustrates a graphical representation of change in carbonate concentration of uranium solution and change in carbonate concentration in NaOH solution used for absorbing carbon dioxide gas;
FIG. 8 illustrates a graphical representation of concentration change of acid and alkaline solution according to time using electrolytic dialysis system;
FIG. 9 is a graphic view showing powder after treatment at step 4 according to the present invention;
FIG. 10 is a flow diagram illustrating experimental process to confirm the density specification of nuclear fuel pellet for uranium oxide powder prepared at step
FIG. 5 illustrates a graphical representation of uranium concentration varying according to the change in pH of uranyl peroxo carbonate complex;
FIG. 6 illustrates a graphical representation of the result of X-ray diffractometer (XRD) of uranium agglomerate precipitates;
FIG. 7 illustrates a graphical representation of change in carbonate concentration of uranium solution and change in carbonate concentration in NaOH solution used for absorbing carbon dioxide gas;
FIG. 8 illustrates a graphical representation of concentration change of acid and alkaline solution according to time using electrolytic dialysis system;
FIG. 9 is a graphic view showing powder after treatment at step 4 according to the present invention;
FIG. 10 is a flow diagram illustrating experimental process to confirm the density specification of nuclear fuel pellet for uranium oxide powder prepared at step
5 according to an embodiment of the present invention;
FIG. 11 is a graphic view showing powder after treatment at step 5 according to the present invention; and FIG. 12 is a flow diagram illustrating an experimental process to confirm the density specification of nuclear fuel pellet fabricated from powder mixture of the Examples 1 to 6 in which UO2+x powder prepared from uranium oxide scraps and virgin UO2+, powder is mixed.
DETAILED DESCRIPTION OF EXEMPLARY EMBODIMENTS
In one embodiment, a method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps is provided, comprising the steps of:
(Si) dissolving uranium by adding carbonate solution with hydrogen peroxide into the uranium oxide scraps;
(S2) adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution;
(S3) recovering the acid and the alkaline solutions used at step 2;
(S4) dehydrating agglomerated uranium oxide powder precipitated at step 2 under argon atmosphere, reductive heat treatment of the agglomerated powder under hydrogen atmosphere, and passivation of forming an oxide protection film;
(S5) milling the agglomerated uranium oxide powder prepared at step 4; and (S6) mixing the uranium oxide power milled at step 5 and virgin uranium oxide powder.
The method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps according to the present invention will be explained in greater detail below with reference to the respective steps.
In one embodiment, at step 1, carbonate solution is added to uranium oxide scraps to dissolve uranium. At step 1, uranium is selectively separated from uranium scraps
FIG. 11 is a graphic view showing powder after treatment at step 5 according to the present invention; and FIG. 12 is a flow diagram illustrating an experimental process to confirm the density specification of nuclear fuel pellet fabricated from powder mixture of the Examples 1 to 6 in which UO2+x powder prepared from uranium oxide scraps and virgin UO2+, powder is mixed.
DETAILED DESCRIPTION OF EXEMPLARY EMBODIMENTS
In one embodiment, a method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps is provided, comprising the steps of:
(Si) dissolving uranium by adding carbonate solution with hydrogen peroxide into the uranium oxide scraps;
(S2) adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution;
(S3) recovering the acid and the alkaline solutions used at step 2;
(S4) dehydrating agglomerated uranium oxide powder precipitated at step 2 under argon atmosphere, reductive heat treatment of the agglomerated powder under hydrogen atmosphere, and passivation of forming an oxide protection film;
(S5) milling the agglomerated uranium oxide powder prepared at step 4; and (S6) mixing the uranium oxide power milled at step 5 and virgin uranium oxide powder.
The method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps according to the present invention will be explained in greater detail below with reference to the respective steps.
In one embodiment, at step 1, carbonate solution is added to uranium oxide scraps to dissolve uranium. At step 1, uranium is selectively separated from uranium scraps
6 mixed with impurities including grinding sludge generated in the fabrication process of nuclear fuel pellet, using alkali carbonic acid solution including hydrogen peroxide, instead of acid solution, to dissolve only uranium while leaving the most impurities including grinding sludge in the solution, because the impurities are hardly dissolved in the solution. In carbonate solution of pH 11-13, grinding sludge with impurities of metallic oxides (Fe, Ni, Mo, Al, Si) are hardly dissolved, and even if they do, these are precipitated as metal hydroxide by hydrolysis with very low solubility, so that only a slight amount of the impurities of metallic oxides are dissolved in the carbonate solution, while most of the impurities of metallic oxides as solid form in the solution. Meanwhile, uranium dioxide (UO2) is dissolved due to oxidation of hydrogen peroxide, and combined with carbonate ion to form uranium complex ion to be dissolved. As UO2 is dissolved, surface intermediate of mixed oxidation state is formed on the surface, and oxygen to convert UO2 into UO3 is supplied from water. Generally, UO2 is dissolved as reaction formula 1 below.
<Reaction formula 1>
UO2 UO2+x ¨> U0233 ¨> UO3 ¨* (U022)surface (UO2(CO3)xYl)bulk wherein x and y are positive integers.
To be specific, under oxidation condition, non-stoichiometric oxide remaining in grain boundary of UO2 is oxidized at first, and UO2 grain is oxidized into UO2 33(U307) in the sintering of nuclear fuel UO2. UO2.33 is mixed state of U6+ and U4+ or U5+ and U4+, and the generation of U0233 decreases U4+ in UO2. Finally, with oxidation state of +6 is formed, then it is converted into uranyl ion (i.e., U022+)
<Reaction formula 1>
UO2 UO2+x ¨> U0233 ¨> UO3 ¨* (U022)surface (UO2(CO3)xYl)bulk wherein x and y are positive integers.
To be specific, under oxidation condition, non-stoichiometric oxide remaining in grain boundary of UO2 is oxidized at first, and UO2 grain is oxidized into UO2 33(U307) in the sintering of nuclear fuel UO2. UO2.33 is mixed state of U6+ and U4+ or U5+ and U4+, and the generation of U0233 decreases U4+ in UO2. Finally, with oxidation state of +6 is formed, then it is converted into uranyl ion (i.e., U022+)
7 if the surface is contacted with acidic solution, and converted into UO2(OH)3-ion if the surface is contacted with alkaline solution. If the solution is carbonate solution, UO2 dissolves with binding to C032- to form uranyl carbonate complex ions (UO2(CO3)22-, UO2(CO3)34-. Also, if H202 exists in the carbonate solution, uranium yx-is dissolved in the form of UO2(02)),(CO3)2-22y as described in reaction formula 2 below, and if H202 does not exist in carbonate solution, UO2(CO3)34- is formed, and the uranyl peroxo carbonate complex ion has a very high solubility compared to the uranyl carbonate complex. Therefore, the carbonate solution can use Na2CO3 solution of 0.1 ¨ 3 M including 0.1-5 M of H202.
<Reaction formula 2>
_32- + yH202 + 2y011- = [UO2(02)y(CO
UO2 + xCO oxi2.2x-zy + 2yH20 + 2e"
wherein y is 0, 1 or 2, and x/y= 1/2, 2/1 or 3/0.
According to the method for preparing uranium oxide powder for nuclear fuel pellet in one embodiment, step 2 includes processes of adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution.
At this time, if pH of carbonate solution in which uranium is dissolved is adjusted to about 0 ¨ 4, uranyl peroxo carbonate complex ion in the solution is decomposed. As described in reaction formula 3 below, the uranyl peroxo carbonate ion is precipitated in the form of uranyl peroxide (U04). At the same time, carbonate species (C032) of the uranyl peroxo carbonate complex ion are converted into
<Reaction formula 2>
_32- + yH202 + 2y011- = [UO2(02)y(CO
UO2 + xCO oxi2.2x-zy + 2yH20 + 2e"
wherein y is 0, 1 or 2, and x/y= 1/2, 2/1 or 3/0.
According to the method for preparing uranium oxide powder for nuclear fuel pellet in one embodiment, step 2 includes processes of adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution.
At this time, if pH of carbonate solution in which uranium is dissolved is adjusted to about 0 ¨ 4, uranyl peroxo carbonate complex ion in the solution is decomposed. As described in reaction formula 3 below, the uranyl peroxo carbonate ion is precipitated in the form of uranyl peroxide (U04). At the same time, carbonate species (C032) of the uranyl peroxo carbonate complex ion are converted into
8 carbon dioxide and released from the solution in a gaseous state. The precipitated uranyl peroxides are separated from the solution and dried to give U04 . xH20(U04 . 2H20 or U04 . 4H20) agglomerate form. The released carbon dioxide gas may be converted into carbonate solution in a gas absorption column in which alkaline solution (NaOH) is flowed down, and then recovered.
<Reaction formula 3>
UO2(02)x(CO3)3,2-2x-2y mu _ 2H20 -> UO2(02) . 4H20 + yH2CO3 wherein m is 4, 6 or 8 at y=0, 1 or 2, respectively. And x/y is 1/2, 2/1 or 3/0.
According to the method for preparing uranium oxide powder for nuclear fuel pellet in one embodiment, step 3 includes recovering the acid and the alkaline solutions used at step 2.
At step 3, the acid and the alkaline solutions used at step 2 are recovered by electrolytic dialysis which uses a pair of anion exchange membrane and cation exchange membrane. To be specific, as the residual solution after step 2 containing anion of acid used at step 2 and cation of carbonate solution used at step 1 is passed between the anion and cation exchange membranes of the electrolytic dialysis system, cation and anion pass the cation exchange membrane at cathode side and the anion exchange membranes at anode side, respectively, and are recovered as acid and alkaline. The acid and alkaline are circulated to step 2.
Also, after the acid and the alkaline solutions are recovered, a steps of washing the uranium precipitates, and separating solid and liquid may additionally be provided.
<Reaction formula 3>
UO2(02)x(CO3)3,2-2x-2y mu _ 2H20 -> UO2(02) . 4H20 + yH2CO3 wherein m is 4, 6 or 8 at y=0, 1 or 2, respectively. And x/y is 1/2, 2/1 or 3/0.
According to the method for preparing uranium oxide powder for nuclear fuel pellet in one embodiment, step 3 includes recovering the acid and the alkaline solutions used at step 2.
At step 3, the acid and the alkaline solutions used at step 2 are recovered by electrolytic dialysis which uses a pair of anion exchange membrane and cation exchange membrane. To be specific, as the residual solution after step 2 containing anion of acid used at step 2 and cation of carbonate solution used at step 1 is passed between the anion and cation exchange membranes of the electrolytic dialysis system, cation and anion pass the cation exchange membrane at cathode side and the anion exchange membranes at anode side, respectively, and are recovered as acid and alkaline. The acid and alkaline are circulated to step 2.
Also, after the acid and the alkaline solutions are recovered, a steps of washing the uranium precipitates, and separating solid and liquid may additionally be provided.
9 The uranium precipitates are washed by being contacted with distilled water two to three times, and solid and liquid are separated by centrifuge or filtration.
The uranium precipitates washed and separated from the solution are formed into agglomerated U04 xH20 powder in the form of U04- 2H20 or U04. 4H20 depending on the drying temperature.
According to the method for preparing uranium powder for nuclear fuel pellet in one embodiment, step 4 includes dehydrating the agglomerated uranium powder precipitated at step 2 under inert atmosphere, and reductive heat treatment under hydrogen atmosphere, and forming protective oxide film by passivation.
At step 4, the dehydrating the powder (U04 . 2H20 or U04 .41-120) precipitated at step 2 is performed under inert atmosphere, and preferably at 120-200 C under argon atmosphere. If the dehydrating temperature is less than 120 'C, water contained in uranium hydrate (i.e., U04 .4H20 powder) is not fully removed, and if the dehydrating temperature is above 200 C, U04 is oxidized into other oxide forms (i.e., U0).
Also, the reductive heat treatment at step 4 is desirably performed at 600-800 C
under hydrogen reducing atmosphere. If the treatment is performed at a temperature less than 600 C , the reduction reaction rate is low, and if the temperature exceeds 800 "C, powder with low compactability and sinterability is produced due to the aggregation of powder particles.
Further, the forming of the protective oxide film at step 4 includes forming the agglomerated uranium oxide powder with chemical composition of UO2+, (0<x<0.17) at 75-85 t under the oxygen partial pressure range of 1-3%.
According to the method for preparing uranium oxide powder for nuclear fuel pellet in one embodiment, step 5 includes milling of the agglomerated uranium powder prepared at step 4.
At step 5, uranium powder can be prepared by attrition mill or ball mill to improve compactability and sinterability.
According to the method for preparing uranium powder for nuclear fuel pellet in one embodiment, step 6 includes the mixing uranium powder milled at step 5 with virgin uranium oxide powder.
The milled uranium oxide powder satisfies the specification for use as nuclear fuel.
However, since there is not a plenty of uranium oxide scraps generated, it may be desirable to increase the sinterability of the uranium powder and reduce the sintering temperature and time by mixing with the virgin uranium powder, rather than directly recycling the final product of uranium oxide powder for the nuclear fuel from the uranium oxide scraps. The virgin uranium powder is desirably mixed by 10 - 50 wt%.
The present inventive concept provides a method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps, comprising the steps of:
(Step A) dissolving uranium by adding carbonate solution with hydrogen peroxide into the uranium oxide scraps;
(Step B) adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution, and recycling the carbonate for Step A;
(Step C) recovering the acid and the alkali solution used at step B to regenerate the acid and the alkaline solutions at an electrolytic dialysis system, and recycling the acid and alkali for step B;
(Step D) washing uranium agglomerate precipitated at step B and performing solid¨
liquid separation;
(Step E) dehydrating the uranium agglomerate prepared at step D under argon atmosphere, reductive heat treatment of the uranium agglomerate under hydrogen atmosphere, and forming a protective oxide film by passivation;
(Step F) milling the agglomerated uranium oxide powder prepared at step E and then milled powder mixing with virgin uranium oxide powder.
Referring FIG. 1 illustrating the flow diagram of a method for preparing uranium oxide powder (10) for nuclear fuel pellet according to an embodiment of the present invention, hydrogen peroxide-carbonate solution is added to uranium oxide scraps including impurities to dissolve only uranium, leaving the impurities to precipitate (1); pH of the dissolved uranyl peroxo carbonate complex solution is adjusted, leaving the uranium to precipitate as U04 (2); and the carbon dioxide gas, which is generated at the U04 precipitation step, is recovered as carbonate by a gas absorption column using NaOH (3). Further, by using electrolytic dialysis system, the acid and the alkali solution used in U04 precipitation and carbonate recovery are recovered (4); and the U04 precipitates are washed and solid-liquid separated (5).
Depending on the drying temperature, the separated powder agglomerate is formed U04 .
xH20 powder agglomerate in the form of U04. 2H20 or U04 . 4H20 (6). The precipitated powder agglomerate (U04.xH20) is dehydrated, reduced and passivated to form a protective oxide film (7). The uranium powder agglomerate is milled (8) and the milled powder is mixed with virgin uranium oxide powder (9) to provide uranium oxide powder (10). FIG. 2 illustrates in greater detail the uranium precipitation and the recovery of carbonate solution at the gas absorption column of FIG. 1. As shown in FIG. 2, in U04 settling tank, as uranyl peroxo carbonate complex ion solution(UO2(02)x(C00y2-2x-231.
) is contacted with acid (HNO3) solution, uranium is precipitated as U04, during which carbonate ion (C032") is converted into carbon dioxide (CO2) to be released from the solution. The released carbon dioxide is introduced into a bottom portion of the gas absorption column filled with beads, where alkali (NaOH) solution flows down from a top of the gas absorption column so that carbonate solution (Na2CO3) is recovered from carbon dioxide, released to the bottom portion of the gas absorption column, and circulated to the step of the uranium dissolution and impurity precipitation. FIG. 3 illustrates a principle of the electrolytic dialysis system, according to which solution including NO3- anion coming from acid used in uranium precipitation and Na + cation coming from the carbonate solution is introduced between the cation exchange membrane and the anion exchange membrane of the electrolytic dialysis system where Na + cation passing through the cation exchange membrane combines with OFF generated at cathode by electrolysis of water to give NaOH, and NO3- anion passing the anion exchange membrane combines with Hi- generated at anode by electrolysis of water to give HNO3. The regenerated solutions are respectively circulated as alkali NaOH
13a solution of the gas absorption column and as acid solution used for the U04 precipitation. The precipitated powder agglomerate (U04. 2H20 or U04 . 4H20) is dehydrated at 120-200 C under argon atmosphere, reduce at 600-800 C under hydrogen atmosphere, and passivated to form a protective oxide film at 75-85 C
under the oxygen partial pressure range of 1-3% to prepare uranium powder agglomerate with chemical composition of UO2 x(0<x<0.17). Also, milling is performed using attrition mill or ball mill to convert the agglomerated powder into fine powder with a good compactability and sinterability for final nuclear fuel pellet, and the milled powder is mixed with the virgin uranium oxide powder to provide uranium oxide powder for fabrication of sound nuclear fuel pellet and decreasing defective proportion.
Analysis 1. Dissolution of uranium oxide scraps in hydrogen peroxide-carbonate solution To investigate the solubility of uranium oxide scraps including impurities in hydrogen peroxide-carbonate solution, the amount of UO2 form of uranium scrap powder (average particle size: 10 gm) which is dissolved in the mixed solution of 0.5 M Na2CO3 and 1.0 M H202 was varied and U concentration are analyzed with time, and the result is illustrated in FIG. 4.
As illustrated in FIG. 4, UO2 was completely and fast dissolved in the carbonate solution within several minutes, in which U ion exists in carbonate solution as uranyl peroxo carbonate complex (UO2(02),(CO3)y2-2x-2y) and the solution color appears in red. The dissolution rate of UO2 in carbonate solution and the final concentration of U highly depend on the concentration of hydrogen peroxide in the solution.
Since the other elements including Fe, Ni, Cr, Al oxides mixed in the uranium oxide scrap powder were not dissolved in carbonate solution, these were not detected.
Since uranium was dissolved in the carbonate solution during the dissolution of the uranium scraps, while the other impurities remained undissolved, uranium and impurities can be separated from each other.
2. Uranium concentrations in uranyl peroxo carbonate complex solution with pH
When the pH of uranyl peroxo carbonate complex (UO2(02)(CO3)y2-2x-2y) solution in which UO2 (U basis: 50 g/t) is completely dissolved in 0.5 M Na2CO3 solution with 1.0 M H202 was controlled with nitric acid (HNO3), the concentration of uranium in the solution were measured and the result is illustrated in FIG. 5.
As illustrated in FIG. 5, when pH of the carbonate solution including U was lowered to 6, decarbonation occurs in which carbonate species of the uranyl peroxo carbonate complex was begun to be changed to carbon dioxide. At the same time, uranium precipitate was started. When the pH is 2-4, the maximum amount of the precipitate was observed, and the uranium concentrations of solution were several ppm.
3. Evaluation of phase of uranium precipitate After drying the uranium precipitate obtained at above analysis 2, the uranium precipitate was analyzed by XRD (MAC Science, TX J-827), and the result is shown in FIG. 6.
As illustrated in FIG. 6, the uranium precipitate is uranium peroxide of U04.4H20.
Since solubility product (Ksp) of U04 was as low as 10-3 at pH 2-4 in the solution, the concentration of uranium was a few ppm in the supernatant. Therefore, it was confirmed that U dissolved in the form of uranyl peroxo carbonate complex ion (UO2(02)x(CO3)y2-2x-2y) in the carbonate solution can be recovered with efficiency as high as 99.9%, by adjusting the pH 2-4.
4. Change in carbonate concentration in uranium solution and NaOH solution, respectively To adjust pH of the uranyl peroxo carbonate complex (UO2(02)x(CO3)5,2-2x-2y) at above analysis 2, when 1.0M HNO3 was added into the uranium solution, the changes in the carbonate concentrations in the uranium solution, and in 1.0 M
NaOH
solution, which was circulated within a gas absorption column to absorb carbon dioxide released from the uranium solution due to acidification of the uranyl peroxo carbonate complex solution ,were analyzed. The result is shown in FIG. 7.
As shown in FIG. 7, carbon dioxide released from the uranium solution during the U04 precipitation was recovered at a gas absorption column, in which NaOH is flowed, as 99% of the initial uranium carbonate solution. The recovered carbonate solution is recycled to the dissolution step of uranium scraps. Accordingly, in one embodiment, the uranium scraps can be dissolved without generating carbonate solution waste.
5. Recovery of acid and alkaline solution using electrolytic dialysis system Step 3 of using electrolytic dialysis system, in which cation exchange membrane was placed at cathode side and an anion exchange membrane was placed at anode side, to recover acid and alkaline solution was analyzed, and the result is illustrated in FIG. 8.
Since the residual solution in which uranium is precipitated as U04 includes a large amount of cation of the carbonate solution at step 1 (step of dissolving uranium) and anion of the acid used to adjust pH of uranium solution, the cation and anion are desirable to recover as acid and alkaline solutions, respectively.
FIG. 8 illustrates the concentration changes of HNO3, NaOH and NaNO3 measured at anodic and cathodic chambers of the electrolytic dialysis system, when 100 of 0.5 M of NaNO3 solution was circulated between the cation exchange membrane and anion exchange membrane of an electrolytic dialysis system. At this time, initial solutions (0.1 M NaOH, 0.1 M HNO3) were circulated to cathodic and anodic chambers, respectively, and cell voltage 15 volt was applied. The Na + and NO3" ions of the solution fed between the two ion exchange membranes migrated into the cathodic and anodic chambers, respectively, and regenerated most as 0.5 M
and 0.5 M NaOH solutions due to water splitting reaction in the each chamber.
The regenerated NaOH was recycled as alkali solution to be supplied to the gas absorption column to recover carbon dioxide generated in the step of uranium precipitation. The regenerated HNO3 solution was recycled to adjust pH of carbonate solution which contains uranyl peroxo carbonate complex ion in the uranium precipitation step.
The above analyses confirm that the method of dissolving uranium scraps and recovering U04 from uranium oxide scraps contaminated by impurities according to an embodiment can recover only uranium without generating waste.
<Example 1> Preparation of uranium oxide powder for nuclear fuel pellet 1 Step 1: Dissolving uranium of uranium oxide scraps To dissolve uranium from uranium oxide scraps mixed with impurities, an alkali carbonate solution at pH 12 with 0.5M Na2CO3 and 2 M H202 was used to dissolve only uranium from the uranium oxide scraps without dissolving metallic impurities.
Step 2: Precipitating uranium and recovering carbonate solution To the carbonate solution in which uranium was dissolved at step 1, HNO3 was added and pH was adjusted from 11 to 3. Uranyl peroxo carbonate complex ion in the solution was precipitated in the form of uranyl peroxides (U04), and at the same time, carbonate ion (C032") separated from uranyl peroxo carbonate complex ion was converted into carbon dioxide gas to be released. The precipitated uranyl peroxides was separated from the solution, and was formed into Ua4xH20 (U04=2H20 or U04-4H20) agglomerate. The released carbon dioxide was converted into sodium carbonate solution in a gas absorption column in which NaOH alkali solution circulates, and recovered.
Step 3: Recovering acid and alkali solution The solution which includes NO3- anion coming from the acid used in uranium precipitation and Na + cation coming from the carbonate solution was introduced between the cation exchange membrane and anion exchange membrane of the electrolytic dialysis system. Na + cation passes the cation exchange membrane, and combined with 01-F generated at cathode by water split reaction to give NaOH, and NO3- anion passes the anion exchange membrane, and combined with HI- generated at anode by water split reaction to be recovered as HNO3.
Step 4: Preparing reduced uranium oxide powder U04-4H20 powder recovered at step 3 was dried at 150 t for approximately 1 hour under inert atmosphere (i.e., argon atmosphere), to volatilize and remove water of crystallization, and form U04 powder agglomerate. After that, under 4% of H2-Ar atmosphere, the U04 powder was reduced at 700 t for 5 hours to form UO2 powder, and under 2% of 02¨Ar atmosphere, formation of a protective oxide film by passivation was conducted at 80 t for 2 hours to prepare UO2+x(x=0.08) powder (see FIG. 9).
Step 5: Milling The UO2+, powder agglomerate prepared at step 4 was ball-milled for 6 hours to obtain UO2 x(x=0.14) oxide powder with average particle size of 0.54 gm and specific surface area of 3.1 dig (see FIG. 11).
Step 6: Mixing virgin uranium oxide powder UO2+, powder prepared at step 5 and virgin uranium oxide powder were mixed in a weight ratio of 90:10 to prepare mixed UO2+ powder.
<Example 2> Preparation of uranium oxide powder for nuclear fuel pellet 2 UO2+ powder prepared at Example 1 of step 5 and virgin UO2+ powder were mixed in a weight ratio of 80:20 to prepare mixed UO2+, powder.
<Example 3> Preparation of uranium oxide powder for nuclear fuel pellet 3 UO2+, powder prepared at Example 1 of step 5 and virgin UO2+ powder were mixed in a weight ratio of 75:25 to prepare mixed UO2+, powder.
<Example 4> Preparation of uranium oxide powder for nuclear fuel pellet 4 UO2+ powder prepared at Example 1 of step 5 and virgin UO2+, powder were mixed in a weight ratio of 70:30 to prepare mixed UO2+, powder.
<Example 5> Preparation of uranium oxide powder for nuclear fuel pellet 5 UO2+ powder prepared at Example 1 of step 5 and virgin UO2+), powder were mixed in a weight ratio of 60:40 to prepare mixed UO2+õ powder.
<Example 6> Preparation of uranium oxide powder for nuclear fuel pellet 6 UO2+, powder prepared at Example 1 of step 5 and virgin UO2+, powder were mixed in a weight ratio of 50:50 to prepare mixed UO2+, powder.
The mixing ratio of UO2+,, powder and virgin UO2+ powder of Example 1 ¨ 6 are illustrated in Table 1.
Table 1 Mixing ratio (wt%) Specific Sintered Example prepared surface area density virgin UO2+x UO2+, (m2 /g) (g/cm3) Example 1 90 10 3.32 10.47 Example 2 80 20 3.54 10.53 Example 3 75 25 3.65 10.55 Example 4 70 30 3.75 10.57 Example 5 60 40 3.97 10.60 Example 6 50 50 4.19 10.62 <Experimental example 1> Density of nuclear fuel pellet fabricated from UO2+, powder prepared from uranium oxide scraps To confirm the density of nuclear fuel pellet fabricated from UO2+õ(x=0.14) powder prepared at step 5 of Example I, the experiment of FIG. 10 was conducted and the result was analyzed.
To confirm if UO2-Fx powder prepared at step 5 of Example 1 satisfies the nuclear fuel density specification, the powder was formed under 300 MPa and sintered at 1700 t for 4 hours under hydrogen atmosphere. Accordingly, sintered density of pellet was 10.42 g/cm3 (approximately 95.1% of theoretical density) which satisfied the density specification of nuclear fuel pellet. However, it is required to completely exclude the possibility of fabrication of nuclear fuel pellet below the required specification by preparing nuclear fuel pellets which have yet higher sintered density.
<Experimental example 2> Density of nuclear fuel pellet fabricated from powder mixture of recovered UO2+õ and virgin UO2+, To evaluate the density of nuclear fuel pellet fabricated from powder mixture of Examples 1 to 6 of the present invention, the experiment of FIG. 12 was conducted and the result was analyzed.
Powder mixture of Examples 1 to 6 was compacted at pressure of 300 MPa, and sintered at 1700 r for 4 hours under hydrogen atmosphere. The sintered density of pellet was increased from 10.47 g/cm3 to 10.62 g/cm3 when the ratio of added virgin UO2+, powder was changed from 10% to 50%, which was in the range of 95.5-96.9% of the theoretical density. Therefore, it was confirmed that the density specification of nuclear fuel pellet was fully satisfied.
The uranium precipitates washed and separated from the solution are formed into agglomerated U04 xH20 powder in the form of U04- 2H20 or U04. 4H20 depending on the drying temperature.
According to the method for preparing uranium powder for nuclear fuel pellet in one embodiment, step 4 includes dehydrating the agglomerated uranium powder precipitated at step 2 under inert atmosphere, and reductive heat treatment under hydrogen atmosphere, and forming protective oxide film by passivation.
At step 4, the dehydrating the powder (U04 . 2H20 or U04 .41-120) precipitated at step 2 is performed under inert atmosphere, and preferably at 120-200 C under argon atmosphere. If the dehydrating temperature is less than 120 'C, water contained in uranium hydrate (i.e., U04 .4H20 powder) is not fully removed, and if the dehydrating temperature is above 200 C, U04 is oxidized into other oxide forms (i.e., U0).
Also, the reductive heat treatment at step 4 is desirably performed at 600-800 C
under hydrogen reducing atmosphere. If the treatment is performed at a temperature less than 600 C , the reduction reaction rate is low, and if the temperature exceeds 800 "C, powder with low compactability and sinterability is produced due to the aggregation of powder particles.
Further, the forming of the protective oxide film at step 4 includes forming the agglomerated uranium oxide powder with chemical composition of UO2+, (0<x<0.17) at 75-85 t under the oxygen partial pressure range of 1-3%.
According to the method for preparing uranium oxide powder for nuclear fuel pellet in one embodiment, step 5 includes milling of the agglomerated uranium powder prepared at step 4.
At step 5, uranium powder can be prepared by attrition mill or ball mill to improve compactability and sinterability.
According to the method for preparing uranium powder for nuclear fuel pellet in one embodiment, step 6 includes the mixing uranium powder milled at step 5 with virgin uranium oxide powder.
The milled uranium oxide powder satisfies the specification for use as nuclear fuel.
However, since there is not a plenty of uranium oxide scraps generated, it may be desirable to increase the sinterability of the uranium powder and reduce the sintering temperature and time by mixing with the virgin uranium powder, rather than directly recycling the final product of uranium oxide powder for the nuclear fuel from the uranium oxide scraps. The virgin uranium powder is desirably mixed by 10 - 50 wt%.
The present inventive concept provides a method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps, comprising the steps of:
(Step A) dissolving uranium by adding carbonate solution with hydrogen peroxide into the uranium oxide scraps;
(Step B) adding acid to the solution in which uranium is dissolved at step 1, leaving uranium to precipitate, and absorbing carbon dioxide which is generated from solution during acidification of the uranium solution, with alkaline solution to recover carbonate solution, and recycling the carbonate for Step A;
(Step C) recovering the acid and the alkali solution used at step B to regenerate the acid and the alkaline solutions at an electrolytic dialysis system, and recycling the acid and alkali for step B;
(Step D) washing uranium agglomerate precipitated at step B and performing solid¨
liquid separation;
(Step E) dehydrating the uranium agglomerate prepared at step D under argon atmosphere, reductive heat treatment of the uranium agglomerate under hydrogen atmosphere, and forming a protective oxide film by passivation;
(Step F) milling the agglomerated uranium oxide powder prepared at step E and then milled powder mixing with virgin uranium oxide powder.
Referring FIG. 1 illustrating the flow diagram of a method for preparing uranium oxide powder (10) for nuclear fuel pellet according to an embodiment of the present invention, hydrogen peroxide-carbonate solution is added to uranium oxide scraps including impurities to dissolve only uranium, leaving the impurities to precipitate (1); pH of the dissolved uranyl peroxo carbonate complex solution is adjusted, leaving the uranium to precipitate as U04 (2); and the carbon dioxide gas, which is generated at the U04 precipitation step, is recovered as carbonate by a gas absorption column using NaOH (3). Further, by using electrolytic dialysis system, the acid and the alkali solution used in U04 precipitation and carbonate recovery are recovered (4); and the U04 precipitates are washed and solid-liquid separated (5).
Depending on the drying temperature, the separated powder agglomerate is formed U04 .
xH20 powder agglomerate in the form of U04. 2H20 or U04 . 4H20 (6). The precipitated powder agglomerate (U04.xH20) is dehydrated, reduced and passivated to form a protective oxide film (7). The uranium powder agglomerate is milled (8) and the milled powder is mixed with virgin uranium oxide powder (9) to provide uranium oxide powder (10). FIG. 2 illustrates in greater detail the uranium precipitation and the recovery of carbonate solution at the gas absorption column of FIG. 1. As shown in FIG. 2, in U04 settling tank, as uranyl peroxo carbonate complex ion solution(UO2(02)x(C00y2-2x-231.
) is contacted with acid (HNO3) solution, uranium is precipitated as U04, during which carbonate ion (C032") is converted into carbon dioxide (CO2) to be released from the solution. The released carbon dioxide is introduced into a bottom portion of the gas absorption column filled with beads, where alkali (NaOH) solution flows down from a top of the gas absorption column so that carbonate solution (Na2CO3) is recovered from carbon dioxide, released to the bottom portion of the gas absorption column, and circulated to the step of the uranium dissolution and impurity precipitation. FIG. 3 illustrates a principle of the electrolytic dialysis system, according to which solution including NO3- anion coming from acid used in uranium precipitation and Na + cation coming from the carbonate solution is introduced between the cation exchange membrane and the anion exchange membrane of the electrolytic dialysis system where Na + cation passing through the cation exchange membrane combines with OFF generated at cathode by electrolysis of water to give NaOH, and NO3- anion passing the anion exchange membrane combines with Hi- generated at anode by electrolysis of water to give HNO3. The regenerated solutions are respectively circulated as alkali NaOH
13a solution of the gas absorption column and as acid solution used for the U04 precipitation. The precipitated powder agglomerate (U04. 2H20 or U04 . 4H20) is dehydrated at 120-200 C under argon atmosphere, reduce at 600-800 C under hydrogen atmosphere, and passivated to form a protective oxide film at 75-85 C
under the oxygen partial pressure range of 1-3% to prepare uranium powder agglomerate with chemical composition of UO2 x(0<x<0.17). Also, milling is performed using attrition mill or ball mill to convert the agglomerated powder into fine powder with a good compactability and sinterability for final nuclear fuel pellet, and the milled powder is mixed with the virgin uranium oxide powder to provide uranium oxide powder for fabrication of sound nuclear fuel pellet and decreasing defective proportion.
Analysis 1. Dissolution of uranium oxide scraps in hydrogen peroxide-carbonate solution To investigate the solubility of uranium oxide scraps including impurities in hydrogen peroxide-carbonate solution, the amount of UO2 form of uranium scrap powder (average particle size: 10 gm) which is dissolved in the mixed solution of 0.5 M Na2CO3 and 1.0 M H202 was varied and U concentration are analyzed with time, and the result is illustrated in FIG. 4.
As illustrated in FIG. 4, UO2 was completely and fast dissolved in the carbonate solution within several minutes, in which U ion exists in carbonate solution as uranyl peroxo carbonate complex (UO2(02),(CO3)y2-2x-2y) and the solution color appears in red. The dissolution rate of UO2 in carbonate solution and the final concentration of U highly depend on the concentration of hydrogen peroxide in the solution.
Since the other elements including Fe, Ni, Cr, Al oxides mixed in the uranium oxide scrap powder were not dissolved in carbonate solution, these were not detected.
Since uranium was dissolved in the carbonate solution during the dissolution of the uranium scraps, while the other impurities remained undissolved, uranium and impurities can be separated from each other.
2. Uranium concentrations in uranyl peroxo carbonate complex solution with pH
When the pH of uranyl peroxo carbonate complex (UO2(02)(CO3)y2-2x-2y) solution in which UO2 (U basis: 50 g/t) is completely dissolved in 0.5 M Na2CO3 solution with 1.0 M H202 was controlled with nitric acid (HNO3), the concentration of uranium in the solution were measured and the result is illustrated in FIG. 5.
As illustrated in FIG. 5, when pH of the carbonate solution including U was lowered to 6, decarbonation occurs in which carbonate species of the uranyl peroxo carbonate complex was begun to be changed to carbon dioxide. At the same time, uranium precipitate was started. When the pH is 2-4, the maximum amount of the precipitate was observed, and the uranium concentrations of solution were several ppm.
3. Evaluation of phase of uranium precipitate After drying the uranium precipitate obtained at above analysis 2, the uranium precipitate was analyzed by XRD (MAC Science, TX J-827), and the result is shown in FIG. 6.
As illustrated in FIG. 6, the uranium precipitate is uranium peroxide of U04.4H20.
Since solubility product (Ksp) of U04 was as low as 10-3 at pH 2-4 in the solution, the concentration of uranium was a few ppm in the supernatant. Therefore, it was confirmed that U dissolved in the form of uranyl peroxo carbonate complex ion (UO2(02)x(CO3)y2-2x-2y) in the carbonate solution can be recovered with efficiency as high as 99.9%, by adjusting the pH 2-4.
4. Change in carbonate concentration in uranium solution and NaOH solution, respectively To adjust pH of the uranyl peroxo carbonate complex (UO2(02)x(CO3)5,2-2x-2y) at above analysis 2, when 1.0M HNO3 was added into the uranium solution, the changes in the carbonate concentrations in the uranium solution, and in 1.0 M
NaOH
solution, which was circulated within a gas absorption column to absorb carbon dioxide released from the uranium solution due to acidification of the uranyl peroxo carbonate complex solution ,were analyzed. The result is shown in FIG. 7.
As shown in FIG. 7, carbon dioxide released from the uranium solution during the U04 precipitation was recovered at a gas absorption column, in which NaOH is flowed, as 99% of the initial uranium carbonate solution. The recovered carbonate solution is recycled to the dissolution step of uranium scraps. Accordingly, in one embodiment, the uranium scraps can be dissolved without generating carbonate solution waste.
5. Recovery of acid and alkaline solution using electrolytic dialysis system Step 3 of using electrolytic dialysis system, in which cation exchange membrane was placed at cathode side and an anion exchange membrane was placed at anode side, to recover acid and alkaline solution was analyzed, and the result is illustrated in FIG. 8.
Since the residual solution in which uranium is precipitated as U04 includes a large amount of cation of the carbonate solution at step 1 (step of dissolving uranium) and anion of the acid used to adjust pH of uranium solution, the cation and anion are desirable to recover as acid and alkaline solutions, respectively.
FIG. 8 illustrates the concentration changes of HNO3, NaOH and NaNO3 measured at anodic and cathodic chambers of the electrolytic dialysis system, when 100 of 0.5 M of NaNO3 solution was circulated between the cation exchange membrane and anion exchange membrane of an electrolytic dialysis system. At this time, initial solutions (0.1 M NaOH, 0.1 M HNO3) were circulated to cathodic and anodic chambers, respectively, and cell voltage 15 volt was applied. The Na + and NO3" ions of the solution fed between the two ion exchange membranes migrated into the cathodic and anodic chambers, respectively, and regenerated most as 0.5 M
and 0.5 M NaOH solutions due to water splitting reaction in the each chamber.
The regenerated NaOH was recycled as alkali solution to be supplied to the gas absorption column to recover carbon dioxide generated in the step of uranium precipitation. The regenerated HNO3 solution was recycled to adjust pH of carbonate solution which contains uranyl peroxo carbonate complex ion in the uranium precipitation step.
The above analyses confirm that the method of dissolving uranium scraps and recovering U04 from uranium oxide scraps contaminated by impurities according to an embodiment can recover only uranium without generating waste.
<Example 1> Preparation of uranium oxide powder for nuclear fuel pellet 1 Step 1: Dissolving uranium of uranium oxide scraps To dissolve uranium from uranium oxide scraps mixed with impurities, an alkali carbonate solution at pH 12 with 0.5M Na2CO3 and 2 M H202 was used to dissolve only uranium from the uranium oxide scraps without dissolving metallic impurities.
Step 2: Precipitating uranium and recovering carbonate solution To the carbonate solution in which uranium was dissolved at step 1, HNO3 was added and pH was adjusted from 11 to 3. Uranyl peroxo carbonate complex ion in the solution was precipitated in the form of uranyl peroxides (U04), and at the same time, carbonate ion (C032") separated from uranyl peroxo carbonate complex ion was converted into carbon dioxide gas to be released. The precipitated uranyl peroxides was separated from the solution, and was formed into Ua4xH20 (U04=2H20 or U04-4H20) agglomerate. The released carbon dioxide was converted into sodium carbonate solution in a gas absorption column in which NaOH alkali solution circulates, and recovered.
Step 3: Recovering acid and alkali solution The solution which includes NO3- anion coming from the acid used in uranium precipitation and Na + cation coming from the carbonate solution was introduced between the cation exchange membrane and anion exchange membrane of the electrolytic dialysis system. Na + cation passes the cation exchange membrane, and combined with 01-F generated at cathode by water split reaction to give NaOH, and NO3- anion passes the anion exchange membrane, and combined with HI- generated at anode by water split reaction to be recovered as HNO3.
Step 4: Preparing reduced uranium oxide powder U04-4H20 powder recovered at step 3 was dried at 150 t for approximately 1 hour under inert atmosphere (i.e., argon atmosphere), to volatilize and remove water of crystallization, and form U04 powder agglomerate. After that, under 4% of H2-Ar atmosphere, the U04 powder was reduced at 700 t for 5 hours to form UO2 powder, and under 2% of 02¨Ar atmosphere, formation of a protective oxide film by passivation was conducted at 80 t for 2 hours to prepare UO2+x(x=0.08) powder (see FIG. 9).
Step 5: Milling The UO2+, powder agglomerate prepared at step 4 was ball-milled for 6 hours to obtain UO2 x(x=0.14) oxide powder with average particle size of 0.54 gm and specific surface area of 3.1 dig (see FIG. 11).
Step 6: Mixing virgin uranium oxide powder UO2+, powder prepared at step 5 and virgin uranium oxide powder were mixed in a weight ratio of 90:10 to prepare mixed UO2+ powder.
<Example 2> Preparation of uranium oxide powder for nuclear fuel pellet 2 UO2+ powder prepared at Example 1 of step 5 and virgin UO2+ powder were mixed in a weight ratio of 80:20 to prepare mixed UO2+, powder.
<Example 3> Preparation of uranium oxide powder for nuclear fuel pellet 3 UO2+, powder prepared at Example 1 of step 5 and virgin UO2+ powder were mixed in a weight ratio of 75:25 to prepare mixed UO2+, powder.
<Example 4> Preparation of uranium oxide powder for nuclear fuel pellet 4 UO2+ powder prepared at Example 1 of step 5 and virgin UO2+, powder were mixed in a weight ratio of 70:30 to prepare mixed UO2+, powder.
<Example 5> Preparation of uranium oxide powder for nuclear fuel pellet 5 UO2+ powder prepared at Example 1 of step 5 and virgin UO2+), powder were mixed in a weight ratio of 60:40 to prepare mixed UO2+õ powder.
<Example 6> Preparation of uranium oxide powder for nuclear fuel pellet 6 UO2+, powder prepared at Example 1 of step 5 and virgin UO2+, powder were mixed in a weight ratio of 50:50 to prepare mixed UO2+, powder.
The mixing ratio of UO2+,, powder and virgin UO2+ powder of Example 1 ¨ 6 are illustrated in Table 1.
Table 1 Mixing ratio (wt%) Specific Sintered Example prepared surface area density virgin UO2+x UO2+, (m2 /g) (g/cm3) Example 1 90 10 3.32 10.47 Example 2 80 20 3.54 10.53 Example 3 75 25 3.65 10.55 Example 4 70 30 3.75 10.57 Example 5 60 40 3.97 10.60 Example 6 50 50 4.19 10.62 <Experimental example 1> Density of nuclear fuel pellet fabricated from UO2+, powder prepared from uranium oxide scraps To confirm the density of nuclear fuel pellet fabricated from UO2+õ(x=0.14) powder prepared at step 5 of Example I, the experiment of FIG. 10 was conducted and the result was analyzed.
To confirm if UO2-Fx powder prepared at step 5 of Example 1 satisfies the nuclear fuel density specification, the powder was formed under 300 MPa and sintered at 1700 t for 4 hours under hydrogen atmosphere. Accordingly, sintered density of pellet was 10.42 g/cm3 (approximately 95.1% of theoretical density) which satisfied the density specification of nuclear fuel pellet. However, it is required to completely exclude the possibility of fabrication of nuclear fuel pellet below the required specification by preparing nuclear fuel pellets which have yet higher sintered density.
<Experimental example 2> Density of nuclear fuel pellet fabricated from powder mixture of recovered UO2+õ and virgin UO2+, To evaluate the density of nuclear fuel pellet fabricated from powder mixture of Examples 1 to 6 of the present invention, the experiment of FIG. 12 was conducted and the result was analyzed.
Powder mixture of Examples 1 to 6 was compacted at pressure of 300 MPa, and sintered at 1700 r for 4 hours under hydrogen atmosphere. The sintered density of pellet was increased from 10.47 g/cm3 to 10.62 g/cm3 when the ratio of added virgin UO2+, powder was changed from 10% to 50%, which was in the range of 95.5-96.9% of the theoretical density. Therefore, it was confirmed that the density specification of nuclear fuel pellet was fully satisfied.
Claims (15)
1. A method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps, comprising:
dissolving uranium oxide by adding alkaline carbonate solution with pH of 11 to 13 with hydrogen peroxide into the uranium oxide scraps including the uranium oxide and impurities comprising at least one selected from the group of Cr, Fe, Ni, Mo, Al, Si, and the oxides thereof (step 1);
adding acid to the solution to adjust pH of the solution to a range of 0 to 4 in which the uranium oxide is dissolved at step 1, leaving the uranium oxide to precipitate as agglomerated uranium oxide powder form, and absorbing carbon dioxide which is generated from solution during acidification of the uranium oxide solution, with alkaline solution to recover the carbonate solution (step 2);
recovering the acid and the alkaline solutions used at step 2 (step 3);
dehydrating the agglomerated uranium oxide powder precipitated at step 2 under argon atmosphere, reductive heat treatment the agglomerated uranium oxide powder under hydrogen atmosphere, and passivation of forming a protective oxide film (step 4);
milling the agglomerated uranium oxide powder prepared at step 4 (step 5); and mixing the uranium oxide power milled at step 5 and virgin uranium oxide powder (step 6).
dissolving uranium oxide by adding alkaline carbonate solution with pH of 11 to 13 with hydrogen peroxide into the uranium oxide scraps including the uranium oxide and impurities comprising at least one selected from the group of Cr, Fe, Ni, Mo, Al, Si, and the oxides thereof (step 1);
adding acid to the solution to adjust pH of the solution to a range of 0 to 4 in which the uranium oxide is dissolved at step 1, leaving the uranium oxide to precipitate as agglomerated uranium oxide powder form, and absorbing carbon dioxide which is generated from solution during acidification of the uranium oxide solution, with alkaline solution to recover the carbonate solution (step 2);
recovering the acid and the alkaline solutions used at step 2 (step 3);
dehydrating the agglomerated uranium oxide powder precipitated at step 2 under argon atmosphere, reductive heat treatment the agglomerated uranium oxide powder under hydrogen atmosphere, and passivation of forming a protective oxide film (step 4);
milling the agglomerated uranium oxide powder prepared at step 4 (step 5); and mixing the uranium oxide power milled at step 5 and virgin uranium oxide powder (step 6).
2. The method of claim 1, wherein the carbonate solution at step 1 is 0.1 to 3M
of carbonate (Na2CO3) solution with pH of 11 to 13 containing 0.1 to 5M of H2O2.
of carbonate (Na2CO3) solution with pH of 11 to 13 containing 0.1 to 5M of H2O2.
3. The method of claim 1, wherein the recovery at step 3 is carried out by electrolytic dialysis using a cation exchange membrane and an anion exchange membrane.
4. The method of claim 1, further comprising washing and carrying out solid-liquid separation of uranium precipitate agglomerate remaining after step 3.
5. The method of claim 1, wherein the dehydrating at step 4 is carried out at 120 to 200°C.
6. The method of claim 1, wherein the reductive treating at step 4 is carried out at 600 to 800°C.
7. The method of claim 1, wherein the protective oxide film at step 4 is formed at 75 to 85°C under an oxygen partial pressure range of 1 to 3%.
8. The method of claim 1, wherein the virgin uranium powder at step 6 is added in an amount of 10 to 50 wt%.
9. A method for preparing uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps, comprising steps of:
dissolving uranium oxide by adding alkaline carbonate solution with pH of 11 to 13 with hydrogen peroxide into the uranium oxide scraps including the uranium oxide and impurities comprising at least one selected from the group of Cr, Fe, Ni, Mo, Al, Si, and the oxides thereof (Step A);
adding acid to the solution to adjust pH of the solution to a range of 0 to 4 in which the uranium oxide is dissolved at step A, leaving the uranium oxide to precipitate as agglomerated uranium oxide powder form, and absorbing carbon dioxide which is generated from solution during acidification of the uranium oxide solution, with alkaline solution to recover the carbonate solution, and recycling the carbonate for Step A (Step B);
recovering the acid and the alkali solution used at step B to regenerate the acid and the alkaline solutions at an electrolytic dialysis system, and recycling the acid and alkali for step B (Step C);
washing the agglomerated uranium oxide powder precipitated at step B and performing solid-liquid separation (Step D);
dehydrating the agglomerated uranium oxide powder prepared at step D under argon atmosphere, reductive heat treatment of the agglomerated uranium oxide powder under hydrogen atmosphere, and forming a protective oxide film by passivation (Step E);
milling the agglomerated uranium oxide powder prepared at step E and then milled powder mixing with virgin uranium oxide powder(Step F).
dissolving uranium oxide by adding alkaline carbonate solution with pH of 11 to 13 with hydrogen peroxide into the uranium oxide scraps including the uranium oxide and impurities comprising at least one selected from the group of Cr, Fe, Ni, Mo, Al, Si, and the oxides thereof (Step A);
adding acid to the solution to adjust pH of the solution to a range of 0 to 4 in which the uranium oxide is dissolved at step A, leaving the uranium oxide to precipitate as agglomerated uranium oxide powder form, and absorbing carbon dioxide which is generated from solution during acidification of the uranium oxide solution, with alkaline solution to recover the carbonate solution, and recycling the carbonate for Step A (Step B);
recovering the acid and the alkali solution used at step B to regenerate the acid and the alkaline solutions at an electrolytic dialysis system, and recycling the acid and alkali for step B (Step C);
washing the agglomerated uranium oxide powder precipitated at step B and performing solid-liquid separation (Step D);
dehydrating the agglomerated uranium oxide powder prepared at step D under argon atmosphere, reductive heat treatment of the agglomerated uranium oxide powder under hydrogen atmosphere, and forming a protective oxide film by passivation (Step E);
milling the agglomerated uranium oxide powder prepared at step E and then milled powder mixing with virgin uranium oxide powder(Step F).
10. The method of claim 9, wherein the carbonate solution at step 1 is 0.1 to 3M
of carbonate (Na2CO3) solution with pH of 11 to 13 containing 0.1 to 5M of H2O2.
of carbonate (Na2CO3) solution with pH of 11 to 13 containing 0.1 to 5M of H2O2.
11. The method of claim 9, wherein the recovery at step C is carried out by electrolytic dialysis using a cation exchange membrane and an anion exchange membrane.
12. The method of claim 9, wherein the dehydrating at step E is carried out at 120 to 200°C.
13. The method of claim 9, wherein the reductive heat treatment at step E
is carried out at 600 to 800°C.
is carried out at 600 to 800°C.
14. The method of claim 9, wherein the protective oxide film at step E is formed at 75 to 85°C under an oxygen partial pressure range of 1 to 3%.
15. The method of claim 9, wherein the virgin uranium powder at step F is added in an amount of 10 to 50 wt%.
Applications Claiming Priority (2)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| KR10-2010-0048576 | 2010-05-25 | ||
| KR1020100048576A KR101061481B1 (en) | 2010-05-25 | 2010-05-25 | The method for preparation of uranium powder for nuclear power from uranium oxide scrap |
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| Publication Number | Publication Date |
|---|---|
| CA2741076A1 CA2741076A1 (en) | 2011-11-25 |
| CA2741076C true CA2741076C (en) | 2015-07-14 |
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Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CA2741076A Expired - Fee Related CA2741076C (en) | 2010-05-25 | 2011-05-24 | Method for preparation of uranium oxide powder for nuclear fuel pellet fabrication from uranium oxide scraps |
Country Status (2)
| Country | Link |
|---|---|
| KR (1) | KR101061481B1 (en) |
| CA (1) | CA2741076C (en) |
Families Citing this family (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| KR101290151B1 (en) | 2013-02-20 | 2013-08-07 | 한국지질자원연구원 | Recovery method of uranium and vanadium from black shale ore |
| KR101291146B1 (en) | 2013-02-20 | 2013-08-07 | 한국지질자원연구원 | Extraction method of uranium and vanadium from black shale ore using sequential leaching process |
| CN109949960B (en) * | 2017-12-20 | 2023-01-03 | 中核四0四有限公司 | Recovery method for MOX fuel pellet returned material with unqualified density |
| CN116282176A (en) * | 2022-09-09 | 2023-06-23 | 哈尔滨工程大学 | Method for directly recovering high-purity uranium oxide from uranium-containing wastewater |
Family Cites Families (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| EP1041578A3 (en) | 1999-03-24 | 2002-01-16 | General Electric Company | Process for converting uranium metal alloys to UO2 powder and pellets |
| WO2001003143A1 (en) | 1999-07-02 | 2001-01-11 | Belgonucleaire S.A. | Method for making mixed oxide (u, pu)o2 nuclear fuel pellets from non-castable uo2 powder |
| JP4512029B2 (en) | 2005-12-01 | 2010-07-28 | 株式会社グローバル・ニュークリア・フュエル・ジャパン | Nuclear fuel uranium scrap recovery equipment |
-
2010
- 2010-05-25 KR KR1020100048576A patent/KR101061481B1/en active IP Right Grant
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2011
- 2011-05-24 CA CA2741076A patent/CA2741076C/en not_active Expired - Fee Related
Also Published As
| Publication number | Publication date |
|---|---|
| CA2741076A1 (en) | 2011-11-25 |
| KR101061481B1 (en) | 2011-09-02 |
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