Patent Number: 042474957
Section: description

DETAILED DESCRIPTION OF THE INVENTION The starting materials with their chemical formulas are indicated on the first line of the drawing. On the left-hand side of the drawing are shown the process steps (a), (b) and (c). These result in an intermediate product which is stored. The further processing of the intermediate product by pressing and sintering leads to Product I, shown in the lower left-hand corner of the drawing, which has the highest PuO.sub.2 content. A mixture of the intermediate product from process step (c) with UO.sub.2 granuleate, i.e. granules, is effected in process step (d). Two different mixing ratios of intermediate product to UO.sub.2 granules of 90% intermediate product to 10% UO.sub.2 granules and 10% intermediate to 90% UO.sub.2 granules are given in the drawing. This is to show that by means of process steps (c) and (d) practically any desired fuel specification can be realized as far as plutonium content is concerned, quickly and with a minimum of technical means. Products II and III, shown to the right of Product I in the drawing, are then generated, corresponding to the different mixing ratios. The process steps of milling, pressing and sintering under the designation Part 1 of the drawing are known and correspond to the state of the art as shown by German Published Prosecuted Application No. 1 571 343. However, one of the problems with which the invention is particularly concerned, namely substantially complete solubility in nitric acid suitable for reprocessing, is not mentioned in the German Publication. The following Comparative Example 1 shows that the solubility of the plutonium content attainable with these process steps of the prior art which is the problem to which this invention is addressed, is not satisfactory. COMPARATIVE EXAMPLE 1 Uranium oxide (UO.sub.2+x) with a percentage of about 75% and Pu dioxide with a percentage of about 25% were mixed in a mixer and subsequently milled together in a ball mill. The percentages are by weight based on the mixture. After a milling period of four hours, the powder was taken out of the mill and subsequently pressed without particular attention as to the form of the pressed powder or the density of the material, to obtain a readily transportable material with intimate contact of the different grains of UO.sub.2 and PuO.sub.2. Usual values of density were scattered between 5.0 and 6.0 g/cm.sup.3, but depending on the type of powder, other density values can also be obtained. These pellets and also fragments thereof were sintered in a sintering furnace in a reducing atmosphere (mixture of inert gas and hydrogen) at a temperature of 1700.degree. C. The holding time at sintering temperature was about four hours. The date, attainable with this procedure, regarding stoichiometry-oxygen content, O.sub.2+x, density in g/cm.sup.3 and solubility in nitric acid of the plutonium content are listed in the following Table I: TABLE I ______________________________________ Milling 1 Pressing 1 Sintering 1 ______________________________________ Stoichiometry 2.12 -- 1.97 Density -- 5.9 10.2 g/cm.sup.3 Solubility of the 2.6% -- 97.1% Pu Content ______________________________________ As is seen therefrom, the solubility of the plutonium content is only 97.1%. The process of the present invention starts with the sintered product of the prior art and subjects it to treatment as explained hereinafter to obtain the improved results. After the mentioned sintering process, the annealed material was comminuted down to a grain size of less than 1 mm. The comminuted material was placed in a ball mill and again milled for six hours down to a primary grain size of less than 2 .mu.m. After the milling, the powder was pre-compacted in a press. Contrary to the first-mentioned pressing, careful pressing for blank or pellet densities as uniform as possible is desirable here. The pressed blanks or pellets were subsequently comminuted to form a highly flowable granulate, i.e. free-flowing granules which readily pour, similar to fluid. This highly flowable granulate was subsequently formed into the nuclear fuel pellets to meet the requirements such as density of the pellets, height, special shapes (for instance, dishing at the two end faces). The so-called blanks were subsequently sintered in a second sintering processin an atmosphere of an inert gas/hydrogen mixture, in which again 1700.degree. C. was prescribed as the maximum temperature with a holding time of four hours. The results of this further treatment in the method in accordance with the invention are listed in the following Table II: TABLE II: ______________________________________ Milling 2 Pressing 2 Sintering 2 ______________________________________ Stoichiometry 2.06 -- 1.96 Density -- 7.9 10.6 g/cm.sup.3 Solubility of the 97.1 -- 99.8% Pu Content ______________________________________ It can be seen therefrom that now the solubility of the plutonium content is 99.8%. This procedure corresponds to that producing Product I as shown in the left-hand side of the drawing. In the process mentioned in German Published Prosecuted Application No. 1 571 343, in which, without reference to the tests performed here on the solubility of the plutonium dioxide component, renewed milling and sintering of already milled and sintered powder mixtures are carried out, so-called virgin powder U.sub.3 O.sub.x and PuO.sub.2 is added. No statement is made regarding the solubility of the nuclear fuel prepared by this method; however, the solubility must be less than that resulting from the process according to the present invention because of the admixture of virgin powders (raw powders) required by the German process. Our own tests have shown that in a nuclear fuel, of which about 40% consisted of already sintered and milled material and 42% of virgin UO.sub.2+x and approximately 18% of virgin PuO.sub.2, a maximum solubility of only about 72% could be achieved, referred to the total plutonium content present in the nuclear fuel. The method according to the invention, in contrast, offers not only the advantage of practically complete solubility of the plutonium content, but also that of uncomplicated production of practically any nuclear fuel specification by means of storing the intermediate product prepared in the process step (c). This will be illustrated by the following Example 2: EXAMPLE 2 Uranium dioxide and plutonium oxide were mixed, milled, pressed, sintered, milled again, pre-compacted and granulted in the same manner and in the same ratios as in Example 1. The granulate i.e. granules was homogenized in one lot and put into storage. Analyses were performed on representative samples for the purpose of determining all data required for the further processing, such as the plutonium isotope vector, the plutonium content, the uranium isotope vector, the uranium content, the bulk density, the sinterability, etc. At the same time, a UO.sub.2 granulate, i.e. granules, with specified powder date such as density of the granulate, grain shape of the granulate, sinterability of the granulate, uranium content and the isotope vector of the uranium, was produced in a UO.sub.2 processing plant and delivered. The specifications for the granulate depend in essence on the specifications required for the UO.sub.2 /PuO.sub.2 granulate. It was ensured thereby that the two granulates are compatible, i.e. pressed and sintered well. For manufacturing a mixed-oxide fuel, for instance, for thermal nuclear power plants, 17.5% of the UO.sub.2 PuO.sub.2 granulate and 82.5% of the UO.sub.2 granulate were weighted and mixed together. After mixing, the predetermined and adjusted Pu/U+Pu fission material content was checked once more. Thereupon the blanks of specified shape and dimensions were pressed and sintered. The measured product data are listed in the following Table III: TABLE III ______________________________________ UO.sub.2 /PuO.sub.2 Granulate as per process Nuclear Fuel UO.sub.2 Granulate step c Pellets ______________________________________ Density 6.5 7.0 10.4 g/cm.sup.3 Pu Content -- 22.0 3.9% by wt. U Content 87.8 66.0 84.3% by wt. Stoichiometry 2.05 2.10 1.98 Av.Granule Size - x.sub.50 = 180 .mu.m - x.sub.50 = 180 .mu.m -- Solubility 100 97.1 99.8% ______________________________________ The end product, which according to the flow sheet in the attached drawing falls in the product group II/III, therefore likewise exhibits a solubility of the plutonium content in nitric acid of 99.8%. The two examples mentioned are given only for better illustration; the data given therein such as sintering temperature, holding time at the sintering temperature, percent Pu content, etc. are not limiting. Thus, temperatures other than 1700.degree. C. are also possible, but a lower temperature than 1400.degree. C. is not desirable, as then the diffusion processes proceed too slowly. A temperature of 1800.degree. C. represents at present an upper limit due to the technical limitations in furnace design. Gas mxitures other than inert gas/hydrogen mixtures are also conceivable as the sintering atmosphere, but according to the present state of the art, the latter is preferable since furnaces of larger size for temperatures of 1700.degree. C. are usually laid out so that they can be operated only in a reducing atmosphere. For safety reasons (oxygen-hydrogen gas explosion), the hydrogen usually provided for reduction purposes must be diluted with the inert gas. In this connection, it should further be pointed out that the maximum plutonium content in the intermediate product of the process stage (c) should be smaller than 50%, as otherwise the risk of local plutonium enrichment due to the demixing processes i.e. separating effect of uranium and plutonium mentioned at the outset and thus, partial insolubility of the plutonium can occur in the reprocessing. Important advantages of the method according to the invention are: Simple procedure, because the specified fission material content is adjusted only shortly prior to the pressing of the nuclear fuel pellets by weighing the contents, calculated in advance, of the intermediate product after step (c) and UO.sub.2 granulate. From this it further follows that no special measures for cleaning the mills are required when the specified fission material is changed. The processing of the starting materials is uniform for the mixed-oxide nuclear fuel, for thermal nuclear power plants as well as for mixed-oxide nuclear fuels for fast reactors, particularly fast breeders. Substantial parts of the total process, namely, the process steps (a), (b) and (c), need not be carried out with the entire quantity of mixed-oxide nuclear fuel, so that only that part need to be processed under glove box conditions. If, for instance, a nuclear fuel for thermal nuclear power plants is to be manufactured, this corresponds only to about 10 to 15% of the total quantity of fuel. Reprocesing of the nuclear fuel is simplified, because in addition to uranium content, the plutonium content is also practically completely soluble in nitric acid without the use of additives.