KR101061481B1 - The method for preparation of uranium powder for nuclear power from uranium oxide scrap - Google Patents

Abstract

The present invention relates to a method for producing uranium powder for nuclear fuel from uranium oxide scrap, and more particularly, the present invention comprises the steps of dissolving uranium by adding a carbonate solution to the uranium oxide scrap (step 1); Adding an acid to the solution in which uranium is dissolved in step 1 to precipitate uranium, and then reacting the generated carbon dioxide with an alkaline solution to recover the carbonate solution (step 2); Recovering the acid and alkali solution used in step 2 (step 3); Dehydrating the uranium powder aggregate precipitated in step 2 in an argon atmosphere, and subjecting it to a reduction protective heat treatment in a hydrogen atmosphere and then to an oxide protective film treatment (step 4); It relates to a method for producing uranium powder for nuclear fuel from the uranium oxide scrap comprising the step (step 5) of milling the uranium powder aggregate prepared in step 4.

Description

The method for preparation of uranium powder for nuclear power from uranium oxide scrap}

The present invention relates to a method for producing uranium powder for nuclear fuel from uranium oxide scrap.

Nuclear and light reactor nuclear fuels use UO ₂ sintered with 0.7% U-235 for nuclear fission or UO ₂ shaped sintered with 4-5% U-235. These sintered bodies are manufactured as a nuclear fuel through a series of manufacturing processes in which various additives are mixed with UO ₂ powder, ie preforming, granulation, lubricant addition, compression molding, sintering at 1650 ° C. in a hydrogen atmosphere, and final grinding. do. The nuclear fuel manufacturing process the uranium oxide scraps (scrap) such as defective sintered UO ₂ or UO ₂ powder or UO ₂ sintered product impurities are impurities, grinding debris mixed with a mixture is produced. At this time, the impurities mixed with uranium scrap are metals such as Cr, Fe, Ni, Mo, Al, Si, or these oxides.

In order for UO ₂ powder to be made as a fuel, it is very important that the sintered body has a density of 95% or more of the theoretical density. Nuclear fuel scrap is stored without being recycled in many cases, but the demand for uranium is due to the increase in global nuclear power generation. Increasing demand will increase the need to purify and recycle uranium scrap from the uranium fuel manufacturing process. To recover uranium from conventional uranium oxide scrap, dissolve the uranium oxide scrap in high temperature nitric acid (about 100 ℃, 10 M HNO ₃ ), separate only uranium by solvent extraction, and then mix the uranium with ammonia-auranium using ammonia water. The precipitate is recovered as a precipitate ((NH ₄ ) 2U ₂ O ₇ or UO ₄ 2NH ₄ NO ₃ ) and decomposed to form a UO ₃ powder, followed by reduction heat treatment to prepare a UO ₂ powder. However, conventional methods for recovering uranium from scraps of uranium scrap include problems of device corrosion and NO _x gas in the process of dissolving uranium oxide using high temperature nitric acid, generation of organic waste liquid by solvent extraction, and ammonia nitrogen compounds as environmental regulatory substances. Due to the large amount of secondary wastes, economic and environmental friendliness is greatly reduced.

Therefore, the inventors of the present invention, in the uranium oxide scrap containing the grinding residue impurities generated in the nuclear fuel processing facilities for light reactors, corrosion of the device generated when recovering only uranium by the conventional method, NOx gas treatment, deterioration of operation stability by using high temperature acid, organic While studying how to solve the problems such as the generation of waste liquid and the economical deterioration due to treatment facilities, UO ₄ precipitation separation of dissolved uranium and the carbonate used by pH control of carbonate containing uranium ion in room temperature carbonate system Development of nuclear fuel uranium oxide powder from uranium oxide scrap, which has higher operational stability when processing uranium scrap than conventional methods and minimizes the occurrence of secondary waste, thereby increasing environmental friendliness. The invention was completed.

It is an object of the present invention to provide a method for producing uranium powder for nuclear fuel from uranium oxide scrap.

In order to achieve the above object, the present invention comprises the steps of dissolving uranium by adding a carbonate solution to the uranium oxide scrap (step 1); Adding an acid to the solution in which uranium is dissolved in step 1 to precipitate uranium, and then reacting the generated carbon dioxide with an alkaline solution to recover the carbonate solution (step 2); Recovering the acid and alkali solution used in step 2 (step 3); Dehydrating the uranium powder aggregate precipitated in step 2 in an argon atmosphere, and subjecting it to a reduction protective heat treatment in a hydrogen atmosphere and then to an oxide protective film treatment (step 4); It provides a method for producing a uranium powder for nuclear fuel from the uranium oxide scrap comprising the step (step 5) of milling the uranium powder aggregate prepared in step 4.

The method for producing uranium powder for nuclear fuel from uranium oxide scrap according to the present invention is not only recovering uranium with high efficiency using a simple precipitation method, but also secondary waste by circulation of inorganic salt solutions such as carbonate, acid, and alkali. It greatly improves environmental friendliness without any occurrence, and provides a manufacturing method of uranium powder for nuclear fuel that satisfies the nuclear fuel sintering density specification of light water reactor and heavy water reactor type nuclear power plant. Mixed uranium oxide scrap can be effectively recycled as a resource rather than as a waste.

1 is a process flow diagram illustrating a method for producing uranium powder for nuclear fuel from uranium oxide scrap according to the present invention;
FIG. 2 is a process flow diagram illustrating in detail the uranium precipitation and carbonate recovery steps in FIG. 1;
3 is a view showing the electrolytic reaction process of the acid, alkali recovery step of FIG.
4 is a graph showing the solubility of uranium oxide scrap containing impurities in hydrogen peroxide-carbonate solution;
5 is a graph showing the uranium concentration according to the pH change of the uranium oxo carbonate complex solution;
6 is a graph showing the results of X-ray diffraction analysis (XRD) of uranium aggregate precipitate;
7 is a graph showing the change in the carbonic acid concentration of the uranium solution and the change in the carbonic acid concentration of the NaOH solution;
8 is a graph showing the concentration change with time of the acid and alkali solution using the electrolysis dialysis tank;
9 is a photograph showing the powder after performing step 4 according to the present invention;
10 is a flow chart showing an experimental procedure for determining the nuclear fuel density specification of the uranium powder prepared in Example 1 according to the present invention;
11 is a photograph showing a powder after performing step 5 according to the present invention; And
12 shows an experimental procedure for determining the fuel density specification of the mixed powder of Examples 2 to 7 in which the UO _{2 + x} powder prepared in Example 1 and the pure UO _{2 + x} powder were mixed according to the present invention. It is a flow chart.

The present invention

Dissolving uranium by adding a carbonate solution to the uranium oxide scrap (step 1);

Adding an acid to the solution in which uranium is dissolved in step 1 to precipitate uranium, and then reacting the generated carbon dioxide with an alkaline solution to recover the carbonate solution (step 2);

Recovering the acid and alkali solution used in step 2 (step 3);

Dehydrating the uranium powder aggregate precipitated in step 2 in an argon atmosphere, reducing heat treatment under a hydrogen atmosphere, and then treating the oxide protective film (step 4); And

It provides a method for producing a uranium powder for nuclear fuel from the uranium scrap comprising the step (step 5) of milling the uranium powder aggregate prepared in step 4.

Hereinafter, a method for producing uranium powder for nuclear fuel from uranium scrap according to the present invention will be described in detail step by step.

In the method for producing uranium powder according to the present invention, step 1 is a step of dissolving uranium by adding a carbonate solution to the uranium oxide scrap.

Step 1 is a step for separating uranium from uranium scrap mixed with impurities such as grinding waste generated in the nuclear fuel manufacturing process, using most of the grinding residue using an alkaline carbonate solution containing hydrogen peroxide without using an acid solution. Can dissolve only uranium without dissolving impurities. In the carbonate solution of pH 11-13, the grinding residue impurity metals (Fe, Ni, Mo, Al, Si) oxides are hardly dissolved, but even when dissolved, they precipitate into metal hydroxides by hydrolysis reaction and have very low solubility so that only a small portion of carbonates are present. It will dissolve into solution and most will remain in solid form in the solution. On the other hand, uranium dioxide (UO ₂ ) is dissolved by the oxidation of hydrogen peroxide and is combined with carbonate ions to form uranium complex ions. Dissolution of the UO ₂ is dissolved as the intermediate product (intermediate surface) formed on the surface of the ⁺ U ^5, the oxygen for the UO ₂ to UO ₃ is supplied to the water. Commonly known dissolution of UO ₂ is shown in Scheme 1 below.

<Scheme 1>

_{UO 2 → UO 2 + x →} UO 2 .33 → UO 3 → (UO 2 2 +) surface → (UO 2 (CO 3) x y +) bulk

(Where x and y are positive integers)

More specifically, the non-stoichiometric oxide remaining at the grain boundary during the sintering of the nuclear fuel UO ₂ under oxidizing conditions is first oxidized, and the UO ₂ grain is UO _2.33 (U). ₃ O ₇ ). UO _2.33 is a mixture of U ⁶⁺ and U ⁴⁺ or a mixture of U ⁵⁺ and U ⁴⁺ , whereby U ⁴⁺ in UO ₂ is reduced by the production of UO _2.33 . When +5 in UO ₂ is finally oxidized to UO ₃ , the solution that contacts the surface becomes UO ₂ ^{2+, which} is an uranil ion if the acid is acidic, and changes to UO ₂ (OH) ^3- ion if alkaline. If the solution is a carbonate solution, it combines with CO ₃ ^2- to form a complex form of uranium carbonate complex ion (Uranium carbonato complex, (UO ₂ (CO ₃ ) ₂ ^2- , UO ₂ (CO ₃ ) ₃ ^4- ) In addition, when H ₂ O ₂ is present in the carbonate solution, uranium is in the form of an oxo carbonate complex (UO ₂ (O ₂ ) _x (CO ₃ ) _y ^2-2x-2y ) as shown in Scheme 2 below. When dissolved, and in the absence of H ₂ O ₂ , it is formed into a uranium carbonate complex (UO ₂ (CO ₃ ) ₃ ^4- ), and the oxo carbonate complex has a very high solubility compared to the uranium carbonate complex. 0.1 can be used, such as carbonate (Na ₂ CO ₃₎ solution of 0.1 - 3.0 M containing 5.0 M H ₂ O ₂ in.

<Scheme 2>

_{^{UO 2 + xCO 3 2 - +}} yH 2 O 2 + 2yOH - = [UO 2 (O 2) y (CO 3) x] 2-2x-2y + 2yH 2 O + 2e -

(Where y is 0, 1 or 2, and x is 1, 2 or 3, respectively).

Next, in the method for producing uranium powder according to the present invention, step 2 is to add an acid to the solution in which uranium is dissolved in step 1 to precipitate uranium, and then react the generated carbon dioxide with an alkaline solution to form a carbonate solution. Recovery step.

At this time, when the pH of the carbonate solution in which uranium is dissolved is adjusted to 0-4, the uranil oxo carbonate complex in the solution is decomposed. As shown in Scheme 3, the uranyl oxo carbonate complex precipitates in the form of uranyl peroxide (UO ₄ ), and at the same time the carbonate ion (CO ₃ ^2- ) separated from the uranyl oxo carbonate complex and free in solution ( Free carbonate ions are converted to carbon dioxide and discharged out of solution in gaseous form. The precipitated uranil peroxide is separated from the solution and dried to form a UO ₄ xH ₂ O (UO ₄ 2H ₂ O or UO ₄ 4H ₂ O) aggregate. In this case, the carbon dioxide gas discharged may be converted into a carbonate solution in a gas absorption tower in which an alkaline solution such as caustic soda (NaOH) is circulated and recovered.

<Scheme 3>

UO ₂ (O ₂ ) _x (CO ₃ ) _y ^{2 -2x-2y} + mH ⁺ + 2H ₂ O → UO ₂ (O ₂ ) 4 H ₂ O + yH ₂ CO ₃

(Where y is 0, 1 or 2, and m when x / y is 1/2, 2/1 or 3/0 is 4, 6 or 8, respectively.)

Next, in the method for producing uranium powder according to the present invention, step 3 is a step of recovering the acid and alkali solution used in step 2.

The recovery of step 3 may be performed by an electrolysis dialysis method using a cation exchange membrane and an anion exchange membrane. More specifically, while removing the trace amount of uranil ions remaining in the acidified solution uranium and carbonate is removed in the step 2 at the same time, the anion of the acid used in the step 2 and the cation of the alkaline solution is added to the electrolysis dialysis tank The acid may be recovered from the positive electrode through the silver anion exchange membrane, and the cation may be recovered as alkali from the negative electrode through the cation exchange membrane to circulate to step 2.

In addition, after recovering the acid and alkali solution may further comprise the step of washing the uranium precipitate and solid-liquid separation. The uranium precipitate may be washed by contacting the uranium precipitate with water two or three times, and the solid-liquid separation may be carried out by stirring the uranium precipitate, followed by centrifugation to separate the supernatant or filtering the uranium precipitate. Can be used.

In this case, the washed precipitate was separated from the uranium solution is provided by UO ₄ 2H ₂ O or UO ₄ 4H ₂ O in the form of UO xH ₂ O ₄ powder agglomerates according to the drying temperature.

Next, in the method for producing uranium powder according to the present invention, step 4 is a step of dehydrating the uranium powder aggregate precipitated in the above step 2 in an inert atmosphere, subjected to reduction heat treatment in a hydrogen atmosphere, and then to form an oxide protective film.

In the dehydration of step 4, the powder aggregate precipitated in step 2 (UO ₄ 2H ₂ O or UO ₄ 4H ₂ O) is preferably carried out at 120 to 200 ° C. under an inert atmosphere, preferably argon atmosphere. If the dehydration temperature is lower than 120 ° C., the water contained in the uranium hydrate UO ₄ 4H ₂ O powder may not be sufficiently removed. If the dehydration temperature is higher than 200 ° C., UO ₄ may be an oxide form of UO _x . There is a problem of oxidation.

In addition, the heat treatment of step 4 is preferably carried out at 600 ~ 800 ℃ in a hydrogen reducing atmosphere. If the heat treatment is less than 600 ℃, there is a problem that the reaction rate is lowered, if it exceeds 800 ℃ there is a problem that the powder having a low sinterability due to the bonding between the powder particles.

Furthermore, the oxide protective film of step 4 may be formed at 75 to 85 ° C. at an oxygen partial pressure of 1 to 3% to prepare uranium powder aggregates having UO _{2 + x} (0 < _x ≦ 0.17).

Next, in the method for producing uranium powder according to the present invention, step 5 is a step of milling the uranium powder aggregate prepared in step 4.

Step 5 may be a uranium powder of UO _{2 + x} (0 < _x ≦ 0.14) using an attrition mill or ball mill to have sinterability to the powder agglomerates prepared.

Next, in the method for producing uranium powder according to the present invention, the method may further include mixing the powder prepared in step 5 with pure uranium powder.

The uranium powder produced by the production method according to the present invention satisfies the necessary specifications for use as a nuclear fuel material, but typically does not generate much uranium scrap and is purer than directly recycling the uranium powder obtained from the uranium scrap directly into the fuel material. By mixing with uranium powder, the sintering temperature and time can be shortened while increasing the stability of the uranium powder. At this time, the pure uranium powder is preferably mixed at 10 to 50% by weight.

In addition,

Dissolving uranium by adding a carbonate solution to the uranium oxide scrap (step A);

Adding acid to the solution in which the uranium is dissolved in step A to precipitate uranium, and then adding an alkali solution to the generated carbon dioxide to recover the carbonate solution, and reusing it as the carbonate of step A (step B);

Recovering the acid and alkali solutions used in step B and regenerating them in an electrolysis dialysis tank, reusing them with the acid and alkali of step B (step C);

Washing and solid-liquid separation of the uranium aggregates precipitated in step B (step D);

Dehydrating the uranium agglomerates prepared in step D under an argon atmosphere, subjecting to a reduction protective heat treatment under a hydrogen atmosphere, followed by an oxide protective film treatment (step E);

It provides a method for producing a uranium powder for nuclear fuel from the uranium scrap comprising the step of milling the uranium powder aggregate prepared in step E and mixing with pure uranium powder (step F).

Referring to Figure 1 showing a process flow diagram of a method for producing a uranium powder according to the present invention, by adding a hydrogen peroxide-carbonate solution to the uranium oxide scrap containing impurities, the impurities are precipitated (1) while dissolving only uranium, dissolved uranium The pH of the oxo carbonate complex solution is adjusted to precipitate uranium with UO ₄ (2), and carbon dioxide generated by precipitation of uranium in a gas absorption tower using NaOH is recovered (3). In addition, an acid and alkaline solution used in UO ₄ precipitation and carbonate recovery is recovered (4) using an electrolysis dialysis tank, and the UO ₄ precipitate is washed and solid-liquid separated (5). The separated powder agglomerate is formed from a UO ₄ 2H ₂ O or UO ₄ 4H ₂ O UO ₄ in the form of powder aggregates xH ₂ O (6) according to the drying temperature. The uranium precipitate (2) shown in FIG. 1 and the carbonate solution recovery (3) by the gas absorption tower are shown in more detail in FIG. As shown in FIG. 2, when the uranium oxo carbonate complex (UO ₂ (O ₂ ) _x (CO ₃ ) _y ^2-2x-2y ) contacts the acid (HNO ₃ ) solution in the UO ₄ precipitation tank, uranium precipitates as UO ₄ . carbonate ion (CO ₃ ^{2 -)} in the process and the carbon dioxide is converted to (CO ₂₎ is emitted out of the solution, the released carbon dioxide gas - enters the bottom of the gas absorber is a bead (bead) filling for the liquid in contact filled, Alkaline (NaOH) solution flows down from the gas absorption tower, and carbon dioxide is recovered as a carbonate solution (Na ₂ CO ₃ ) and discharged to the bottom of the gas absorption tower, and circulated to the uranium dissolved impurity precipitation (1). Figure 3 shows the principle of the electrolysis dialysis tank, the solution containing the Na 3 ⁺ cation generated from the acid and NO ^{3 −} anion generated from the acid used in the uranium precipitation is injected between the cation exchange membrane and the anion exchange membrane of the electrolysis dialysis tank Na ⁺ The cations pass through the cation exchange membrane to combine with OH ^- generated by electrolysis of water at the cathode to form NaOH, and the NO ^{3 -} anion passes through the anion exchange membrane to combine with H ⁺ produced by decomposition of water at the anode, leading to HNO ₃ The regenerated solution is recycled to the alkaline NaOH solution and the acid solution used for uranium precipitation, respectively, in the gas absorption tower. The powder aggregates (UO ₄ 2H ₂ O or UO ₄ 4H ₂ O) precipitated above were dehydrated at 120-200 ° C. under argon atmosphere, reduced heat treated at 600-800 ° C. in hydrogen atmosphere, and 75 at oxygen partial pressure of 1-3%. protective oxidation treatment to ~ 85 ℃ (7) to prepare a powder of uranium aggregate _{UO 2 + x (0 <x≤0.17} ). In addition, the powder agglomerates are milled (8) using an attrition mill or ball mill to have sinterability, and mixed (9) with pure uranium powder to prepare uranium powder for nuclear fuel.

analysis

1. Solubility Analysis of Uranium Oxide Scrap in Hydrogen Peroxide-carbonate Solution

UO ₂ type uranium scrap powder dissolved in 0.5 M Na ₂ CO ₃ and 1.0 MH ₂ O ₂ mixed solution to determine the solubility of uranium oxide scrap containing impurities in hydrogen peroxide-carbonate solution (average particle size: U concentration over time was analyzed while varying the amount of 10 μm), and the results are shown in FIG. 4.

As shown in FIG. 4, it can be seen that UO ₂ completely dissolves rapidly in water in a carbonic acid solution, where U ions are uranium oxo carbonate complexes (UO ₂ (O ₂ ) _x (CO ₃ ) _y in a carbonated solution. ^2-2x-2y ) and the solution color appeared red. The rate of dissolution and the final concentration of U dissolved in the carbonate solution of UO ₂ are greatly influenced by the concentration of hydrogen peroxide in the solution. At this time, the other elements Fe, Ni, Cr, and Al oxides mixed in the uranium oxide scrap powder are not dissolved in the carbonic acid solution and are not detected in the solution.Uranium is dissolved in the carbonic acid solution when dissolving the uranium scrap, but other impurities are not dissolved. Could be separated.

2. Uranium Oxo Carbonate Complex  Solution pH  Uranium Concentration as Changes

Uranium oxo carbonate complexes _{(UO 2 (O 2) x} (CO 3) y 2-2x-2y) solution (UO ₂ (U standards in 0.5 M Na ₂ CO ₃ the carbonate solution containing 1.0 MH ₂ O _2: 50 g / l) was dissolved completely) the concentration of uranium while reducing the pH with nitric acid (HNO ₃ ) was measured, the results are shown in FIG.

As shown in FIG. 5, when the pH of the carbonate solution containing U is lowered to 6 or less, the carbonate ions of the solution undergo a decarbonation process in which carbon dioxide ions are changed to carbon dioxide, and at the same time, uranium precipitation begins to be observed and the pH is 2-4. The most precipitates occurred at and the U concentration in the solution was less than 1 ppm. At this time, the concentration of uranium in the solution was about several ppm.

3. Phase analysis of uranium precipitate

After drying the uranium precipitate of Analysis 2, X-ray diffraction analysis (XRD, MAC Science, TX J-827) to analyze the phase, and the results are shown in FIG.

As shown in FIG. 6, the uranium precipitate was found to be UO ₄ 4H ₂ O and was confirmed to be uranium peroxide (UO ₄ ). The solubility product (K _sp ) of UO _{4 in} the solution was so low as 10 ⁻³ that the uranium concentration was less than 1 ppm at pH 2-4. Therefore, U dissolved in the form of uranium oxo carbonate complex (MO ₂ (O ₂ ) _x (CO ₃ ) _y ^2-2x-2y in carbonate solution can be recovered with high efficiency of 99.9% or more by adjusting the pH to 2-4 It can be seen.

4. Changes in Carbonic Acid Concentration in Uranium Solution NaOH  Change in Carbonic Acid Concentration of a Solution

In order to adjust the pH of the carbonate solution containing the uranium oxo carbonate complex (MO ₂ (O ₂ ) _x (CO ₃ ) _y ^2-2x-2y ), 1.0 M nitric acid is injected to change the carbonic acid concentration of the uranium solution and In order to absorb carbon dioxide, the change in carbonic acid concentration in a 1.0 M NaOH solution circulated to a gas absorption tower was analyzed, and the results are shown in FIG. 7.

As shown in FIG. 7, it can be seen that carbon dioxide generated during precipitation of UO ₄ by adjusting pH in a carbonate solution containing uranium may be recovered as a 99% carbonate solution in a gas absorption tower in which NaOH flows. By recovering the recovered carbonate solution in the uranium scrap dissolving tank, it can be seen that the present invention can dissolve the uranium scrap without generating carbonate waste solution.

5. Electrolytic Dialysis tank  Used obstetrics Alkali  Recovery analysis of the solution

Step 3 of the present invention, in which an acid and alkaline solution was recovered using an electrolysis dialysis tank equipped with a cation exchange membrane on the negative electrode side and an anion exchange membrane on the positive electrode side, was analyzed, and the results are shown in FIG. 8.

At pH 2-4, uranium precipitated as UO ₄ and the remaining solution contained a large amount of cations of the carbonate solution used in step 1 of the present invention, which is a uranium dissolving step in the uranium solution, and an anion of the acid used to control the pH of the uranium solution. It is desirable to recover the acid and alkali solution.

100 ml of 0.5 M Na ⁺ , NO ₃ ^- solution was circulated between the cation exchange membrane and the anion exchange membrane of the electrolysis dialysis tank, and the initial 0.1 M NaOH solution and 0.1 M HNO ₃ solution were circulated to the cathode and the anode, respectively, at a voltage of 15 Volts. The HNO ₃ , NaOH and NaNO ₃ concentrations measured in the positive electrode, negative electrode and feed solution when is applied are shown in FIG. 8. It can be seen that Na ⁺ and NO ₃ ⁻ ions in the solution supplied between the two ion exchange membranes are moved to the cathode and the anode, respectively, and are regenerated into almost 0.5 M HNO ₃ solution and 0.5 M NaOH solution by water decomposition reaction, respectively. The regenerated NaOH is recycled to an alkali solution which is fed to the gas absorption tower to recover the carbon dioxide from the uranium precipitation step, and the regenerated HNO ₃ solution is recovered from the uranium oxo carbonate complex (UO ₂ (O ₂ ) _x ( CO ₃ ) _y ^2-2x-2y ) can be recycled to adjust the pH of the carbonate solution.

From the analysis, it can be seen that the method of dissolving uranium scrap and recovering UO ₄ can recover only uranium with little waste from uranium scrap contaminated with impurities.

Example 1 Preparation of Uranium Powder for Nuclear Fuel 1

Step 1: dissolving the uranium of the uranium oxide scrap

To dissolve uranium from a mixture of impurities, uranium has a pH of 12 and 0.2 M alkali carbonate (Na ₂ CO ₃ ) containing 0.2 M hydrogen peroxide, which does not dissolve most of the impurities such as grinding residues. Only the bay was dissolved.

Step 2: Precipitating Uranium and Recovering Carbonate Solution

HNO ₃ was added to the uranium solution in which the uranium was dissolved in step 1 to adjust the pH to 0-4. The uranyl oxo carbonate complex in the solution was decomposed, and the uranyl oxo carbonate complex precipitated in the form of uranyl peroxide (UO ₄ ) and at the same time separated from the carbonate ion (CO ₃ ^2- ) separated from the uranyl oxo carbonate complex. Free carbonate ions in the solution were converted to carbon dioxide and discharged out of solution in a gaseous state. The precipitated uranil peroxide was separated from the solution and dried to form UO ₄ xH ₂ O (UO ₄ 2H ₂ O or UO ₄ 4H ₂ O) aggregates, and the carbon dioxide released was discharged from the gas absorption tower where NaOH alkaline solution was circulated. It was converted to a carbonate solution and recovered.

Step 3: recovering acid and alkaline solution

A solution containing NO ^{3 −} anion from the acid used in the uranium precipitation and a Na ⁺ cation from the carbonate solution is injected between the cation exchange membrane and the anion exchange membrane in the electrolysis dialysis tank, while the Na ⁺ cation is passed through the cation exchange membrane to remove water from the cathode. It was combined with OH ⁻ generated by electrolysis to form NaOH, and the NO ^{3 −} anion was recovered through the anion exchange membrane and combined with H ⁺ produced by decomposition of water at the anode to recover HNO ₃ .

After recovery of the acid and alkali solution to further perform the step of washing the uranium precipitate to solid-liquid separation to form a UO ₄ 2H ₂ O or UO ₄ 4H ₂ O in the form of UO xH ₂ O ₄ powder agglomerates.

Step 4: preparing a powder aggregate

The powder aggregate recovered in step 3 was dried at 150 ° C. for about 1 hour in an argon atmosphere, which is an inert atmosphere, to remove volatilized crystal water to form UO ₄ powder aggregates, and then to 700 ° C. for 5 hours under 4% hydrogen and argon atmosphere. To reduce UO _{2 + x} powder agglomerates, and to prepare UO _{2 + x} (x = 0.08) powder agglomerates having a composition of the fuel sintered raw material by treating the oxide film for 2 hours at 80 ° C. under 2% O ₂ and argon atmosphere. (See FIG. 9).

Step 5: milling step

And the milling step to a mean particle size of 0.54 ㎛ ball for a UO _{2 + x} powder agglomerates for 6 hours production at 4, having a specific surface area is to prepare a powder of UO _{2 + x} 3.1 ㎡ / g (see Fig. 11).

Example 2 Preparation of Uranium Powder for Nuclear Fuel 2

The UO _{2 + x} powder prepared in Example 1 and the pure UO _{2 + x} powder were mixed in a weight ratio of 90:10 to prepare a UO _{2 + x} mixed powder.

Example 3 Preparation of Uranium Powder for Nuclear Fuel 3

The UO _{2 + x} powder prepared in Example 1 and the pure UO _{2 + x} powder were mixed in a weight ratio of 80:20 to prepare a UO _{2 + x} mixed powder.

Example 4 Preparation of Uranium Powder for Nuclear Fuel 4

The UO _{2 + x} powder prepared in Example 1 and the pure UO _{2 + x} powder were mixed in a weight ratio of 75:25 to prepare a UO _{2 + x} mixed powder.

Example 5 Preparation of Uranium Powder for Nuclear Fuel 5

The UO _{2 + x} powder prepared in Example 1 and the pure UO _{2 + x} powder were mixed in a weight ratio of 70:30 to prepare a UO _{2 + x} mixed powder.

Example 6 Preparation of Uranium Powder for Nuclear Fuel 6

The UO _{2 + x} powder prepared in Example 1 and the pure UO _{2 + x} powder were mixed at a weight ratio of 60:40 to prepare a UO _{2 + x} mixed powder.

Example 7 Preparation of Uranium Powder for Nuclear Fuel 7

The UO _{2 + x} powder prepared in Example 1 and the pure UO _{2 + x} powder were mixed in a weight ratio of 50:50 to prepare a UO _{2 + x} mixed powder.

The mixing ratios of the UO _{2 + x} powders and the pure UO _{2 + x} powders of Examples 1 to 7 are shown in Table 1 below.

Yes Mixing ratio (% by weight) Specific surface area (m &lt; 2 &gt; / g) UO _{2 + x} according to the invention Pure UO _{2 + x} Example 1 100 0 3.10 Example 2 90 10 3.10 Example 3 80 20 3.32 Example 4 75 25 3.64 Example 5 70 30 3.75 Example 6 60 40 3.97 Example 7 50 50 4.19

Experimental Example 1 Nuclear Fuel Density Specification of UO _{2 + x} Powder

In order to determine the nuclear fuel density specification of the UO _{2 + x} (x = 0.14) powder prepared in Example 1 according to the present invention, the experiment shown in FIG. 10 was performed and analyzed.

In order to confirm that the prepared UO _{2 + x} powder satisfies the final fuel density specification, the final sintered density was 10.61 g / cm 3 (theoretical density 96.8) after molding at 300 MPa and sintering at 1750 ° C. for 10 hours in a hydrogen atmosphere. %), It can be seen that it satisfies the nuclear fuel density specification (95% or more of the theoretical density) of the light and heavy water reactor.

Experimental Example 2 Nuclear Fuel Density Specification of UO _{2 + x} Powder and Pure UO _{2 + x} Powder

In order to determine the nuclear fuel density specification of the mixed powder of Examples 2-7, which is a mixture of UO _{2 + x} powder and pure UO _{2 + x} powder prepared in Example 1 according to the present invention, the experiment shown in FIG. And analyzed.

Examples 2-7 mixed powder was molded at 300 MPa and sintered at 1700 ° C. for 4 hours in a hydrogen atmosphere. As the ratio of pure UO _{2 + x} powder was increased, the sintered density was increased from 10.47 g / cm 3 to 10.62 g / cm 3. As the theoretical density was in the range of 95.5 ～ 96.9%, it was confirmed that the fuel density specification of the light and heavy water reactors requiring more than 95% of the theoretical density was satisfied.

1: uranium dissolution-impurity precipitation
2: UO ₄ precipitation
3: carbonate recovery
4: acid and alkaline solution recovery
5: washing and solid-liquid separation
6: UO ₄ x H ₂ O
7: Dehydration reduction heat treatment oxidation protective film formation
8: milling
9: mixed
10: powder for nuclear fuel

Claims (17)

Dissolving uranium by adding a carbonate solution to the uranium oxide scrap (step 1);
Adding an acid to the solution in which uranium is dissolved in step 1 to precipitate uranium, and then reacting the generated carbon dioxide with an alkaline solution to recover the carbonate solution (step 2);
Recovering the acid and alkali solution used in step 2 (step 3);
Dehydrating the uranium powder aggregate precipitated in step 2 in an argon atmosphere, and subjecting it to a reduction protective heat treatment in a hydrogen atmosphere and then to an oxide protective film treatment (step 4); And
Method for producing a uranium powder for nuclear fuel from the uranium oxide scrap comprising the step (step 5) of milling the uranium powder aggregate prepared in step 4.
According to claim 1, wherein the carbonate solution of step 1 is characterized in that the pH range of 11 ~ 13, 0.1 ~ 3.0 M carbonate (Na ₂ CO ₃ ) solution containing 0.1 ~ 5.0 M H ₂ O ₂ Method for producing a uranium powder for nuclear fuel from the uranium oxide scrap.
The method of claim 1, wherein in step 2, the pH is adjusted to a range of 0 to 4 by adding an acid to the solution in which the uranium is dissolved.
The method of claim 1, wherein the recovery of step 3 is carried out by an electrolysis dialysis method using a cation exchange membrane and an anion exchange membrane, the method for producing uranium powder for nuclear fuel uranium oxide scrap.
The method of claim 1, further comprising the step of washing and solid-liquid separation of the remaining uranium precipitate aggregate after performing step 3. The method for producing uranium powder for nuclear fuel from the uranium oxide scrap.
The method of claim 1, wherein the dehydration of step 4 is carried out at 120 ~ 200 ℃ a method for producing a uranium powder for nuclear fuel from the uranium oxide scrap.
According to claim 1, wherein the heat treatment of step 4 is a method for producing uranium powder for nuclear fuel from the uranium oxide scrap, characterized in that carried out at 600 ~ 800 ℃.
The method of claim 1, wherein the oxide protective film of step 4 is formed by performing a uranium oxide scrap for nuclear fuel from uranium oxide scrap, characterized in that formed by performing at 75 ~ 85 ℃ at an oxygen partial pressure of 1-3%.
The method of claim 1, further comprising mixing 10 to 50% by weight of pure uranium powder after the step 5, the method for producing a uranium powder for nuclear fuel from the uranium oxide scrap.
Dissolving uranium by adding a carbonate solution to the uranium oxide scrap (step A);
Adding an acid to the solution in which the uranium is dissolved in step A to precipitate uranium, and then reacting the generated carbon dioxide with an alkaline solution to recover the carbonate solution, and reusing it as the carbonate of step A (step B);
Recovering the acid and alkali solutions used in step B and reusing them with the acid and alkali of step B (step C);
Washing and solid-liquid separation of the uranium aggregates precipitated in step B (step D);
Dehydrating the uranium agglomerates prepared in step D under an argon atmosphere, subjecting to a reduction protective heat treatment under a hydrogen atmosphere, followed by an oxide protective film treatment (step E);
Method of producing a uranium powder for nuclear fuel from the uranium scrap comprising the step of milling the uranium powder aggregate prepared in step E and mixing with pure uranium powder (step F).
The carbonate solution of step A is characterized in that the pH range of 11-13, 0.1 ~ 3.0 M carbonate (Na ₂ CO ₃ ) solution containing 0.1 to 5.0 M H ₂ O ₂ , characterized in that A method for recovering uranium powder from uranium oxide scrap.
The method for recovering uranium powder from uranium oxide scrap according to claim 10, wherein the pH is adjusted to a range of 0 to 4 by adding an acid to the solution in which uranium is dissolved in step B.
The method of claim 10, wherein the recovery of step C is carried out by an electrolysis dialysis method using a cation exchange membrane and an anion exchange membrane to prepare a uranium powder for nuclear fuel uranium oxide scrap.
The method of claim 10, wherein the dehydration of the step E is carried out at 120 ~ 200 ℃ method for producing a uranium powder for nuclear fuel from the uranium oxide scrap.
The method of claim 10, wherein the heat treatment of step E is carried out at 600 ~ 800 ℃ method for producing uranium powder for nuclear fuel from the uranium oxide scrap.
12. The method of claim 10, wherein the oxide protective film of step E is formed at 75 to 85 ° C. at an oxygen partial pressure of 1 to 3%.
The method of claim 10, wherein the pure uranium powder of step F is mixed with 10 to 50% by weight of the method for producing uranium powder for nuclear fuel from uranium scrap.

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