JPS6034718B2 - Method for separating volatile fission products from radioactive nuclear fuel - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention provides a method for extracting N02 from radiated fuel pellets or
The present invention relates to a method for separating volatile fission products using N02 and its decomposition products, ie, a mixture of O2 and N02.

One of the problems that arises in the reprocessing of radioactive nuclear fuel is that it is extremely difficult to separate volatile fission products from the reaction solution of the reprocessing process.

This is particularly important with respect to tritium removal, in that tritium isotopes (tritium) are extremely difficult to separate from aqueous solutions. The hydrogen isotope tri-IJ tium has a half-life of 12.26K and is a ternary fission product produced in minute amounts of about 1/1000 of an ounce per ton of fuel. but,
Considering the amount of radioactive fuel expected to be produced to meet the world's electricity needs in the year 2000, the amount of tritium produced will be enormous. Recovering tritium in nuclear fuel reprocessing plants of conventional design is impractical because the amount of tritium per ton of fuel is very low. Tritium is tritiated water (
In the form of tritium and dwater), thousands of pounds were produced in the factory. completely mixed with treated water. Isotopic separation of the tritium from this large amount of water will be required before the water can be separated as a liquid or vapor. Since isotopic separation of tritium in such large quantities of water is impractical, it is necessary to develop techniques to remove volatile fission products from radioactive fuel prior to aqueous reprocessing. The problem of removing volatile fission products from radioactive fuels is particularly important when removing volatile fission products from radioactive uranium dioxide fuels and/or
Alternatively, it is encountered when reprocessing radioactive oxides, such as mixtures of U02 and Pu02.

The problem with removing this volatile fission product is due to the fact that the volatile fission products sink into the crystal lattice of the radiated fuel and therefore cannot be removed by grinding or crushing the fuel. It is increasing. A method for removing volatile fission products from radioactive fuel was developed by the Oak Ridge National Laboratory in the United States.
It is called voloxidation and is described in Oak Ridge National Laboratory Report ORNL-TN-3723.

This voloxidation
is a method of oxidizing radiated fuel at controlled temperatures in the presence of oxygen to form a fine powder of U308.
This voloxidation
It has been determined that the reaction temperature is temperature sensitive and that in large commercial plants the reaction temperature must be maintained within a narrow range of about 48,000 plus or minus 10 qC. This extremely narrow temperature range makes this process increasingly difficult to operate in large scale reprocessing plants. The present invention dissipates volatile fission products from radiated fuel in a commercially reasonable time and under commercially reasonable processing conditions, thereby greatly reducing the cost of reprocessing radiated nuclear fuel. It concerns a method that can be used to Generally speaking, the present invention is a method of oxidizing irradiated nuclear fuel to liberate volatile fission products, such as iodine, xenon, krypton, and tritium, comprising oxidizing irradiated uranium dioxide or a mixture of oxidized uranium dioxide. The fuel is oxidized with nitrogen oxide, i.e., nitrogen dioxide. Nitrogen dioxide is in equilibrium with its decomposition products, namely nitric oxide and oxygen, at the reaction temperature of this process. Therefore, as an oxidizing agent, nitrogen dioxide, a decomposition product of nitrogen dioxide (nitric oxide and oxygen), or a mixture of nitrogen dioxide and a decomposition product thereof may be added.
The oxidizing agent may be diluted with a gas, diluent such as nitrogen or nitric oxide, without substantial deleterious effects. The oxidation reaction temperature may be about 325C0 to 80000C, preferably between about 350QO and 780OOO. More preferably, the reaction is maintained at a temperature in the range of 350°C to 650qo. The above-mentioned oxidation reaction produces a radiated fuel pellet of 0.105 mm, preferably 0.1 mm.
0.045 mm, continued for a sufficient time to convert to a powder having an average particle size smaller than 0.045 mm.

Heating to reaction temperatures higher than 80,000 °C will result in agglomeration of the fine powder and therefore such temperatures should be avoided.

After removal from the reactor, the radiated fuel assembly is generally cooled, and then the fuel stubs are cut open or their contents exposed to prepare the radiated fuel for reprocessing.

This radiated nuclear fuel is then placed in a reaction vessel and nitrogen dioxide alone, or a mixture thereof with its decomposition products, is passed through the reactor. This system is 325q0 and 800o○
Heat to reaction temperature between . During this oxidation reaction, fission gases, especially tritium, are liberated in almost quantitative quantities into the gas removal system. During this reaction, U02 is oxidized to U03 and/or I8. According to the invention, radiated nuclear fuel, ie U02 or a mixture of U02 and PW02, is placed in a reaction vessel and nitrogen oxides, eg nitrogen dioxide, or nitrogen dioxide and its decomposition products, ie oxygen and nitric oxide, are introduced into the reaction vessel. , is oxidized by a mixture of

This oxidation is carried out by placing the radiated fuel in a reaction vessel and heating it to a reaction temperature of about 350 to 780 q0 in the presence of nitrogen oxides. The radiated fuel is heated for a sufficient amount of time to turn the radiated fuel into a fine powder (F08 and U03). The reaction time is highly dependent on the volumetric fuel piece size, UO subdivision size, temperature and gas composition of the materials in the reactor. The fine powder has an average particle size of 0.103 millimeters, preferably 4 degrees greater than 0.045 millimeters. The released tritium is collected as tritium gas or oxidized to the form of THO, cooled, and converted into a liquid (TNQ, H
It is collected as Tachibana (TO, etc.).

To demonstrate the effectiveness of the present invention, fuel pellets were oxidized in conventional laboratory equipment.

The fuel pellet was placed into a sample board and inserted into the combustion tube. Thermocouples were used for temperature control.
A tube furnace was placed around the combustion tube to heat the reaction system.
The oxidizer was passed in gaseous form to the combustion tube. During the oxidation, tritium was liberated and converted to THO in the oxidizing steel furnace and condensed in the cooling tubes. Example 1 Radioactive uranium oxide (U02) fuel and mixed oxide* (
U02/Pu02) Samples of 1 gram each with fuel N
Oxidation was carried out for 4 hours at 400 pm using 02.

At the end of the four hour period, the radiated fuel was in the form of a very fine powder.

The resulting powder was dissolved in 15 ml of 8M HN03 at 10,000 ml for 4 hours, heated in an oven, and a sample of tritium remaining in the solution was taken. An additional 1 gram sample of each radiated fuel was added directly to 8 mol H without oxidation with N02 to prepare the base sample.
Dissolved in N03 at 100°C for approximately 4 hours. This was passed through a furnace and a tritium sample was collected. The remaining solid content in the furnace in both the standard sample and the test sample was further dissolved in 8 mol HN03 with an additive of 0.005 mol hydrofluoric acid, and the resulting solution was collected to determine the content of tritium in the solid content. The amount was measured.

The oxidation results for the two fuel samples are shown in Table 1 below. From the above results in Table 1, it can be seen that oxidizing the fuel with N02 before melting is effective in removing tritium from the fuel.

The mixed oxide solubilizer solution and the solids are
It contained 5% of J thium content. Example 2 A series of experiments were conducted using samples of U02 fuel pellets under isothermal conditions at temperatures between 300oo and 800qo to determine reaction rates.

When the oxidation was complete, the fuel pellet had turned into a finely ground powder and the fuel sample showed a weight gain of 4% by weight. The results shown in Figure 1 were observed.
In experiments conducted at temperatures between 32,500 and 60,000, the final product was a powder.

Experiments performed at 35,000, 50,000 and 60,000 appear to be nearly identical.

Example 3 A series of experiments with a constant heating rate of 30,000/hour were conducted to test the effect of diluting the oxidizer, ie N02, with nitrogen, with the results shown in FIG.

The fuel sample was U02 pellets.

It should be noted that the kinetics of oxidation of U02 by N02 is relatively insensitive to dilution with N2. Example
4 To determine the effect of dilution with NO on the oxidation of U02, experiments with a constant heating rate of 300 oo/hr were carried out using a mixture of N02 and N02 added to it as the oxidant mixture, and the results shown in Fig. 3 I got the result.

The oxidant used was a 1:1 mixture of N02 and NO.

It was observed that an increase in the NO nose caused a decrease in the oxidation reaction rate of U02 in N02.

[Brief explanation of the drawing]

FIG. 1 is a graph showing the relationship between the reaction time at various reaction temperatures and the weight change of the product in Example 2, and FIG. 2 is a graph showing the relationship between reaction time and weight change of the product at various reaction temperatures in Example 3. 3 is a graph showing the relationship between reaction temperature and product weight change in concentration, and FIG. 3 is a graph showing the relationship between reaction temperature and product weight change in Example 4.
2 is a graph showing the relationship between reaction temperature and product weight change at two concentrations. Figure 1 Figure 2 Figure 3

Claims (1)

[Claims] 1. Radiated nuclear fuel pellets and a nitrogen oxide oxidizer,
reacting at a reaction temperature between about 325°C and 800°C for a time sufficient to convert the pellet to a fine powder, thereby dissipating volatile fission products;
A method for separating volatile fission products from radioactive nuclear fuel. 2. The method of claim 1, wherein the radiated fuel is selected from the group consisting of uranium oxide, plutonium oxide, and mixtures thereof. 3. The method of claim 1, wherein the reaction temperature is maintained between about 350<0>C and 650<0>C. 4. The method of claim 1, wherein said oxidizing agent consists essentially of nitrogen dioxide. 5. The method of claim 1, wherein the oxidizing agent consists essentially of a mixture of nitrogen dioxide, oxygen, and nitric oxide. 6. The method of claim 1, wherein the volatile fission products are xenon, iodine, krypton and tritium. 7. the oxidizing agent is diluted with a gas diluent;
A method according to claim 1. 8. The method of claim 7, wherein the diluent is nitrogen. 9. The method of claim 7, wherein the diluent is nitric oxide.