Method for recovering tritium and zirconium in nuclear fuel cladding

Abstract

The invention provides a method for recovering tritium and zirconium in a nuclear fuel cladding. The method comprises the following steps: oxidizing the nuclear fuel cladding with oxygen-containing gas at 800-1200 DEG C to obtain an oxidized nuclear fuel cladding, and collecting tritium nuclide; chlorinating the oxidized nuclear fuel cladding to obtain a zirconium chloride product; and reducing the zirconium chloride product to obtain zirconium metal. By means of the method, efficient recovery of tritium and zirconium can be achieved.

Description

Technical Field

[0001] The invention relates to the technical field of nuclear waste treatment, in particular to a method for recovering tritium and zirconium from nuclear fuel cladding. Background Art

[0002] Zirconium alloys have excellent performance in high temperature resistance and water vapor corrosion, so they are widely used as fuel cladding materials in second-generation (1960-1990) and third-generation (1990-2010) nuclear reactors. However, neutron irradiation can cause zirconium alloys to become brittle, and there will be irradiation elongation and distortion deformation and other performance changes, and finally the comprehensive performance of zirconium alloys will decrease. The cladding damaged by irradiation needs to be replaced along with the fuel replacement cycle, so the amount of zirconium alloy is very large. Zirconium not only plays a key role in the field of nuclear energy, but is also widely used in chemical equipment, spacecraft structural materials, and biomedicine. Each ton of sponge zirconium is worth about 20,000 to 30,000 US dollars. Tritium is not only used in nuclear fuel and nuclear devices, but also widely used in neutron generators in the civilian field, including neutron irradiation, neutron activation analysis, baggage inspection equipment, etc. At present, the only commercial source of tritium in the world is 19 Canadian uranium nuclear reactors, which are expensive and have limited production. Each gram of tritium is worth about 30,000 US dollars.

[0003] For every 1,000 tons of nuclear fuel processed, about 250 tons of zirconium alloy waste and 50-90 grams of tritium are generated, of which the tritium in the cladding accounts for 10-60% of the total tritium in the nuclear fuel. These zirconium alloy wastes contain more than 95% zirconium metal, of which the zirconium cladding accounts for 80-90% of the total zirconium alloy waste. Therefore, the recovery of tritium and zirconium in nuclear fuel cladding can not only reduce the space for waste disposal and reduce potential harm to the environment, but also help to realize the resource utilization of tritium and zirconium metals, and can also improve the economic efficiency of the reprocessing plant to a certain extent.

[0004] At present, the research on zirconium cladding recovery is mainly based on the chlorination method. Min Ku Jeon studied the chlorination behavior of fresh Zircaloy-4 to separate zirconium and recover ZrCl4 with higher purity. However, the nuclear fuel cladding contains a large amount of fission products and actinides, which affect the chlorination separation process. Among them, the tritium that penetrates into the cladding is captured in the form of tritiated zirconium and converted into tritium chloride (TCl) during the chlorination reaction. Part of TCl may form hydrogen bonds with ZrCl4 or other chloride products (such as SnCl4 or NbCl3) recovered by chlorination, increasing the complexity of zirconium separation. In addition, zirconium oxide (ZrO2) in the 20-30μm thick oxide layer formed in the reactor does not react directly with chlorine, resulting in about 8% of the total zirconium ZrO2 remaining in the ash, and the recovery of this part of the zirconium remaining in the ash requires additional process steps, and the separation is difficult.

[0005] Therefore, in order to reduce the complexity of subsequent separation and improve the recovery rate of tritium and zirconium, it is necessary to provide a new method for recovering tritium and zirconium in nuclear fuel cladding. Summary of the invention

[0006] In view of this, the main purpose of the present invention is to provide a method for recovering tritium and zirconium in nuclear fuel cladding with simple separation operation and high recovery efficiency.

[0007] To this end, the present invention provides a method for recovering tritium and zirconium in nuclear fuel cladding, the method comprising: oxidizing the nuclear fuel cladding with an oxygen-containing gas at 800°C-1200°C to obtain an oxidized nuclear fuel cladding, and collecting tritium nuclides; chlorinating the oxidized nuclear fuel cladding to obtain a zirconium chloride product; and reducing the zirconium chloride product to obtain zirconium metal.

[0008] In some embodiments, the oxygen-containing gas is a mixed gas of nitrogen and oxygen.

[0009] In some embodiments, the volume fraction of oxygen in the oxygen-containing gas is 10%-100%.

[0010] In some embodiments, the oxidation temperature is 1000° C.-1200° C.; and / or the volume fraction of oxygen in the oxygen-containing gas is 30%-100%.

[0011] In some embodiments, collecting tritium nuclides comprises converting tritium into tritiated water over a catalyst and then capturing the tritiated water using a molecular sieve or a condenser.

[0012] In some embodiments, the catalyst is one or more selected from copper oxide, platinum, palladium, nickel, and iron-platinum alloy.

[0013] In some embodiments, the chlorination is performed in a chlorine atmosphere, wherein the volume fraction of chlorine is 10%-50%.

[0014] In some embodiments, the chlorination is performed in the presence of a first reducing agent, and the first reducing agent is a carbonaceous reducing agent.

[0015] In some embodiments, the carbonaceous reducing agent is selected from one or more of graphite, coke, and petroleum ether.

[0016] In some embodiments, the chlorination is performed at a temperature of 600° C.-800° C. for 2-6 hours.

[0017] In some embodiments, the chlorinated product is a fraction collected within a temperature range of 320°C to 340°C.

[0018] In some embodiments, the reducing comprises subjecting the chlorinated product to a reduction reaction in the presence of a second reducing agent to obtain zirconium metal.

[0019] In some embodiments, the second reducing agent is metallic magnesium.

[0020] In some embodiments, the reduction reaction is carried out at a temperature of 600° C.-1000° C. for 2-8 hours.

[0021] The method for recovering tritium and zirconium in nuclear fuel cladding provided by the present invention releases and collects tritium nuclides in the cladding by high-temperature oxidation, and at the same time, some semi-volatile fission products such as Rh, Ru, Tc, etc. will volatilize in the form of oxides such as RhO2, RuO4, Tc2O7, etc., which is conducive to reducing the impurity components in the nuclear fuel cladding. The oxidized nuclear fuel cladding is further subjected to chlorination treatment. During the chlorination process, the different sublimation temperatures of the chlorides of each element are utilized to achieve effective separation of zirconium chloride from most radioactive and non-radioactive impurity elements, and then the collected zirconium chloride is further purified to obtain high-purity metal zirconium. The method of the present invention achieves effective separation of radioactive and non-radioactive impurity elements and realizes the recovery of tritium and zirconium in the cladding through the combined action of high-temperature oxidation and chlorination. The method of the present invention can not only completely recover tritium in the nuclear fuel cladding, but also ensure the full recovery of high-purity zirconium, providing a new way to improve resource utilization. DETAILED DESCRIPTION

[0022] The following will be combined with the embodiments of the present invention to clearly and completely describe the technical solutions in the embodiments of the present invention. Obviously, the described embodiments are only part of the embodiments of the present invention, not all of the embodiments. Based on the embodiments of the present invention, all other embodiments obtained by ordinary technicians in this field without creative work are within the scope of protection of the present invention.

[0023] Throughout the specification, unless otherwise specifically stated, the terms used herein should be understood as meanings commonly used in the art. Therefore, unless otherwise defined, all technical and scientific terms used herein have the same meanings as those generally understood by those skilled in the art to which the present invention belongs. In the event of a conflict, the present specification takes precedence.

[0024] It should be noted that, in the embodiments of the present invention, the terms "include", "comprise" or any other variants thereof are intended to cover non-exclusive inclusion, so that a method or device including a series of elements includes not only the elements explicitly stated, but also other elements not explicitly listed, or also includes elements inherent to the implementation of the method or device. In the absence of further restrictions, an element defined by the sentence "includes a ..." does not exclude the existence of other related elements in the method or device including the element.

[0025] It should be noted that the terms "first\second\third" involved in the embodiments of the present invention are only used to distinguish similar objects, and do not represent a specific order for the objects. It is understandable that "first\second\third" can be interchanged with a specific order or sequence where permitted. It should be understood that the objects distinguished by "first\second\third" can be interchanged where appropriate, so that the embodiments of the present invention described herein can be implemented in an order other than those described herein.

[0026] In the prior art, the recovery of metallic zirconium in nuclear fuel cladding mainly adopts the chlorination method, but the tritium that penetrates into the cladding is captured in the form of tritiated zirconium and converted into tritium chloride (TCl) during the chlorination reaction. Some TCl may form hydrogen bonds with ZrCl4 or other chloride products (such as SnCl4 or NbCl3) recovered by chlorination, increasing the complexity of zirconium separation. Therefore, in order to reduce the complexity of subsequent separation and improve the recovery rate of zirconium, it is necessary to release the tritium in the nuclear fuel cladding before chlorination recovery.

[0027] In view of this, the present application provides a method for recovering tritium and zirconium in nuclear fuel cladding, the method comprising: oxidizing the nuclear fuel cladding with an oxygen-containing gas at 500°C-1200°C to obtain an oxidized nuclear fuel cladding, and collecting tritium nuclides; chlorinating the oxidized nuclear fuel cladding to obtain a zirconium chloride product; and reducing the zirconium chloride product to obtain zirconium metal.

[0028] In the present application, the nuclear fuel cladding is oxidized with nitrogen or oxygen-containing gas at a high temperature of 800°C-1200°C, so that the zirconium in the nuclear fuel cladding is converted into zirconium oxide, and the tritium nuclides in the cladding are released in the form of tritium gas and tritium hydride for collection and recovery. Other fission products and actinide elements in the cladding are also converted into oxides, among which some semi-volatile oxides of fission products such as RhO2, RuO4, Tc2O7, etc. are volatilized to achieve separation, while semi-volatile or difficult-to-volatile oxides such as U3O8, PuO2, CoO, Cs2O, etc., remain in the zirconium oxide. The zirconium oxide is further subjected to chlorination treatment, so that the oxides of zirconium oxide and impurity elements are converted into chlorides. In the chlorination process, the different sublimation temperatures of each chloride are used to achieve the separation of zirconium and other impurity elements. For example, the chlorides of nuclides such as U, Co, and Sn have a lower vapor pressure, so they can be evaporated and removed at a lower temperature; ZrO2 is converted into volatile ZrCl4 metal salts through chlorination treatment, and the volatile ZrCl4 metal salts can be collected and further reduced and recovered; for example, the chlorides of nuclides such as Cs and Pu, CsCl and PuCl4, sublime at temperatures above 460°C and 600°C, respectively, and therefore remain in the ash. Therefore, the method of the present application realizes the effective separation and treatment of various nuclides in the nuclear fuel cladding through the combined use of high-temperature oxidation and chlorination.

[0029] In the present application, the temperature of the high temperature oxidation is 800°C-1200°C. By making the temperature within the above range, it is beneficial for the nuclear fuel cladding to be fully oxidized to form oxides. Exemplarily, the temperature of the high temperature oxidation can be 800°C, 900°C, 1000°C, 1100°C, 1200°C or a value between the ranges consisting of any two of them. In some preferred embodiments, the temperature of the high temperature oxidation is 1000°C-1200°C.

[0030] In some embodiments, the high temperature oxidation time is 3-8 hours.

[0031] In some embodiments, the oxygen-containing gas is a mixed gas of nitrogen and oxygen. In the oxygen-containing gas, oxygen acts as an oxidant to react with metal elements in the nuclear fuel cladding to generate oxides, and nitrogen is an inert gas that does not participate in the oxidation reaction and can be used to adjust the oxygen content in the oxygen-containing gas, which is beneficial for controlling the high-temperature oxidation reaction rate as needed.

[0032] In some embodiments, in the oxygen-containing gas, the volume fraction of oxygen is 10%-100%. By making the volume fraction of oxygen in the oxygen-containing gas within the above range, it is beneficial for the high-temperature oxidation reaction to be fully carried out, and on the other hand, the rate of the high-temperature oxidation reaction can be controlled as needed. Exemplarily, in the oxygen-containing gas, the volume fraction of oxygen can be 10%, 20%, 30%, 40%, 50%, 60%, 70%, 80%, 90%, 100% or a value between the ranges composed of any two of them. In some preferred embodiments, in the oxygen-containing gas, the volume fraction of oxygen is 30%-100%.

[0033] In some embodiments, the collection of tritium nuclides includes converting tritium into tritiated water on a catalyst, and then capturing the tritiated water using a molecular sieve or a condenser. The mechanism for collecting tritium nuclides includes capturing the tritium nuclides in the zirconium cladding in the form of tritium gas or tritium hydride in a high temperature environment, and then reacting with oxygen in an oxygen environment under the action of a catalyst to generate tritiated water, and capturing and collecting the tritiated water by a molecular sieve or a condenser, thereby achieving the collection of tritium nuclides. In some embodiments, the catalyst is one or more selected from copper oxide, platinum, palladium, nickel, and iron-platinum alloys. In some embodiments, the molecular sieve is selected from 5A molecular sieves and 3A molecular sieves, and aluminosilicate molecular sieves composed of calcium carbonate, aluminum oxide and silicon oxide.

[0034] In some embodiments, the chlorination is carried out in a chlorine atmosphere, wherein the volume fraction of chlorine is 10-50%. By carrying out the chlorination treatment in chlorine, the oxides generated in the high temperature oxidation step are converted into chlorides, which is beneficial for utilizing the difference in the sublimation temperature of chlorides to achieve separation between elements.

[0035] In some embodiments, the chlorination is performed in the presence of a first reducing agent, and the first reducing agent is a carbon-containing reducing agent. Preferably, the carbon-containing reducing agent is selected from one or more of graphite, coke, and petroleum ether.

[0036] In some embodiments, the chlorination is performed at a temperature of 600-800° C. for 2-6 hours. By carrying out the chlorination treatment within the above temperature range, it is beneficial to fully convert the oxide into chloride, and it is beneficial for volatile and semi-volatile chlorides and zirconium chloride to volatilize in different temperature ranges, so as to achieve the separation of different chlorides and the collection of zirconium chloride; on the other hand, it is avoided that the temperature is too high, so that the chlorides of nuclides such as Cs and Pu are volatilized to affect the collection of zirconium chloride.

[0037] In some embodiments, the zirconium chloride product is a fraction collected within a temperature range of 320-340° C. The sublimation temperature of zirconium chloride is about 331° C. By collecting the fraction within the above temperature range, the zirconium chloride product and chlorides of other impurity elements in the nuclear fuel cladding can be effectively separated, and the zirconium element in the nuclear fuel cladding can be efficiently recovered through subsequent condensation, so as to subsequently reduce and recover the zirconium metal.

[0038] In some embodiments, the reduction includes subjecting the zirconium chloride product to a reduction reaction in the presence of a second reducing agent to obtain zirconium metal. Under the action of the second reducing agent, the zirconium chloride product is converted into metallic zirconium, thereby obtaining nuclear grade zirconium metal. In some embodiments, the second reducing agent is metallic Mg.

[0039] In some embodiments, the reduction reaction is carried out at a temperature of 600-1000° C. for 2-8 hours. By making the temperature and time of the reduction reaction within the above range, it is beneficial for the zirconium chloride product and the second reducing agent to fully react, thereby achieving efficient recovery of zirconium metal.

[0040] The method of the present application is described in more detail below through specific examples.

[0041] Example

[0042] Example 1

[0043] The nuclear fuel cladding is oxidized in a 100% oxygen atmosphere at 800°C for 4 hours to obtain an oxidized nuclear fuel cladding. During the oxidation process, tritium volatilizes in the form of tritium gas (T2) and tritium hydride (HT). The volatilized tritium gas is guided to the catalytic device through the gas guide system, and tritium is oxidized to tritiated water (HTO) on the heated CuO catalyst, and then the tritiated water is captured using a 5A molecular sieve.

[0044] The above-mentioned oxidized nuclear fuel cladding is added with graphite as a reducing agent to further carry out chlorination reaction in a chlorine atmosphere with a chlorine gas volume fraction of 60%. The reaction temperature is controlled to be 335°C and the reaction time is 4 hours. The ZrCl4 product close to the sublimation temperature of 331°C is collected and a pure ZrCl4 crystalline product is obtained after condensation.

[0045] The ZrCl4 crystal product was subjected to a reduction reaction with a Mg reducing agent, in which the Mg reducing agent was added in an amount 2.5 times the mass of ZrCl4. The temperature of the reduction reaction was 900°C and the time was 6 hours, and nuclear-grade zirconium metal was recovered.

[0046] Embodiment 2-5

[0047] The nuclear fuel cladding is treated by a method similar to that of Example 1, the only difference being that the temperature of the high temperature oxidation is adjusted with reference to the value in Table 1.

[0048] Comparative Example 1

[0049] The nuclear fuel cladding was treated in a similar manner to that of Example 1, except that the temperature of the high temperature oxidation was set to 500°C.

[0050] The recovery rates of tritium and zirconium metal obtained in the above Examples 1-5 and Comparative Example 1 are recorded in Table 1. Wherein, tritium recovery rate = (mass of tritiated water captured by molecular sieve / 18) × 2 / total tritium release amount × 100% (1)

[0051] Recovery rate of zirconium metal = mass of metal zirconium recovered by reduction / mass of cladding × 100% (2).

[0052] It is understood that in the above formula (1), the total release of tritium is the mass of all tritium completely released in the cladding, and the tritium in the cladding can be completely released by methods known to those skilled in the art, such as long-term high-temperature oxidation treatment to release all tritium. In formula (2), since the mass of nuclides produced by impurities and fission in different zirconium claddings is difficult to quantify, the recovery rate of zirconium metal is calculated based on the mass of the cladding. Examples 1-5 and Comparative Example 1 are processed using claddings under the same working conditions, so the impurity content in the cladding is the same, which does not affect the judgment of the recovery rate of zirconium metal.

[0053] Table 1

[0054]

[0055] According to the above results, it can be seen that oxidation treatment with oxygen-containing gas at 800°C-1200°C and collection of tritium nuclides, followed by chlorination reaction and reduction to obtain zirconium chloride products have high zirconium recovery rates and tritium nuclides recovery rates. The method of the present invention can effectively recover tritium nuclides and zirconium metal.

[0056] Example 6

[0057] The nuclear fuel cladding is oxidized in a mixture of nitrogen and oxygen with an oxygen volume fraction of 90% at 1200°C for 4 hours to obtain an oxidized nuclear fuel cladding. During the oxidation process, tritium volatilizes in the form of tritium gas (T2) and tritium hydride (HT). The volatilized tritium gas is guided to the catalytic device through the gas guide system, and tritium is oxidized to tritiated water (HTO) on the heated CuO catalyst, and then the tritiated water is captured by a molecular sieve.

[0058] The above-mentioned oxidized nuclear fuel cladding is further subjected to chlorination reaction by adding graphite as a reducing agent in a chlorine atmosphere with a chlorine gas volume fraction of 60%. The reaction temperature is controlled to be 335°C and the reaction time is 4 hours. The ZrCl4 product close to the sublimation temperature of 331°C is collected and a pure ZrCl4 crystalline product is obtained after condensation.

[0059] The ZrCl4 crystal product was subjected to a reduction reaction with a Mg reducing agent, in which the Mg reducing agent was added in an amount 2.5 times the mass of ZrCl4. The temperature of the reduction reaction was 900°C and the time was 6 hours, and nuclear-grade zirconium metal was recovered.

[0060] Examples 7-10

[0061] The nuclear fuel cladding is treated by a method similar to that of Example 1, except that the volume fraction of oxygen in the oxygen-containing gas (mixed gas of nitrogen and oxygen) used in the oxidation step is adjusted with reference to the values ​​in Table 2.

[0062] Comparative Example 2

[0063] The nuclear fuel cladding is treated in a similar manner to that of Example 1, except that nitrogen gas (without oxygen) is used in the oxidation step.

[0064] The amounts of tritiated water obtained in Examples 6-10 and Comparative Example 2 and the recovery rates of zirconium metal are recorded in Table 2.

[0065] Table 2

[0066]

[0067] According to the results of the above examples, compared with the treatment in an oxygen-free nitrogen atmosphere in Comparative Example 2, the method of the present invention can effectively improve the recovery rate of tritium nuclides and zirconium metal by performing high-temperature oxidation treatment on the nuclear fuel cladding in an oxygen-containing gas environment.

[0068] The above description is only a preferred embodiment of the present invention, and does not limit the patent scope of the present invention. All equivalent structural changes made by using the contents of the present invention specification under the inventive concept of the present invention, or directly / indirectly applied in other related technical fields are included in the patent protection scope of the present invention.

Claims

1. A method for recovering tritium and zirconium from nuclear fuel cladding, characterized in that: The method comprises: oxidizing the nuclear fuel cladding with oxygen-containing gas at 800° C.-1200° C. to obtain an oxidized nuclear fuel cladding, and collecting tritium nuclides; chlorinating the oxidized nuclear fuel cladding to obtain a zirconium chloride product; and reducing the zirconium chloride product to obtain zirconium metal.

2. The method according to claim 1, wherein: The oxygen-containing gas is a mixed gas of nitrogen and oxygen; preferably, the volume fraction of oxygen in the oxygen-containing gas is 10%-100%.

3. The method according to claim 1 or 2, wherein: The oxidation temperature is 1000° C.-1200° C.; and / or the volume fraction of oxygen in the oxygen-containing gas is 30%-100%.

4. The method according to claim 1 or 2, wherein: The collecting of tritium nuclides comprises converting tritium into tritiated water on a catalyst, and then capturing the tritiated water using a molecular sieve or a condenser; preferably, the catalyst is one or more selected from copper oxide, platinum, palladium, nickel, and iron-platinum alloy.

5. The method according to claim 1 or 2, wherein: The chlorination is carried out in a chlorine atmosphere, in which the volume fraction of chlorine is 10%-50%.

6. The method according to claim 5, wherein: The chlorination is carried out in the presence of a first reducing agent, wherein the first reducing agent is a carbon-containing reducing agent. Preferably, the carbon-containing reducing agent is selected from one or more of graphite, coke, and petroleum ether.

7. The method according to claim 1 or 2, wherein: The chlorination is carried out at a temperature of 600° C.-800° C. for 2-6 hours.

8. The method according to claim 1 or 2, wherein: The chlorinated product is a fraction collected within a temperature range of 320°C to 340°C.

9. The method according to claim 1 or 2, wherein: The reduction comprises subjecting the chlorinated product to a reduction reaction in the presence of a second reducing agent to obtain zirconium metal; preferably, the second reducing agent is metallic magnesium.

10. The method according to claim 9, wherein: The reduction reaction is carried out at a temperature of 600° C.-1000° C. for 2-8 hours.

Images