CS201977B1 - Process for preparing semiproducts and products of sinterable ceramic nuclear fuels - Google Patents

Description

The present invention provides a process for the preparation of intermediates and products of compressible ceramic nuclear fuels by direct thermal denitrification of colloidal aqueous suspensions of uranium ammonium polyanions or mixtures thereof with plutonium or thorium compounds formed concomitantly with nitric or ammoniacalization of the respective aqueous nitrate solutions.

Most common default! The raw material for the preparation of ceramic nuclear fuels is uranyl nitrate. Depending on the fuel type, the feed compound contains uranium in the usual isotope ratio, or uranium further enriched in isotope 235, enriched in plutonium 239, ep. The treatment of the starting salts into oxides is always preceded by jejiph transfer to the ammonium salt of the respective polyanions, their filtration from the aqueous phase, their subsequent thermal decomposition and reduction. The preparations are usually based on salt solutions having a concentration of about 10% by weight. and containing about 0.05 to 0.15 moles of nitric acid / kg. By adding about 11 moles of ammonia gas, a product is obtained at pH 3.5 which rapidly settles at 50 DEG C. and has good filtration properties. However, as the pH rises, the product filterability deteriorates extremely, and at pH values above 6.6, filtration is no longer operationally feasible. Nevertheless, it is precisely the preparations of these hardly separable suspensions that after chemical decomposition and reduction of the uranium oxide formed, generally the most valuable uranium dioxide, whose specific gravity is always greater than 10.5 g.cm &^lt; ³ & gt ^; (i.e., more than 97% of theory). The products obtained, on the other hand, reach the highest specific gravity at pH 3.5

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9.28 g.cm &^lt; ³ & gt ^; (i.e., about 85% of theory).

In this case, precipitation at a pH of 3.5 gives a yield of about 55% to 90% in each case, while the remainder must be recovered from the solutions at pH values greater than 7.

On a production scale, these difficulties necessarily lead to the application of the so-called two-stage process; however, the separation of the products in the second stage of precipitation is always all the more difficult as the lower rate of precipitation is reduced in the first stage.

Products dropped at pH 3.5 consist of particles having a size of 20-24 µm, while the polyurethanes obtained, for example, at pH 7.2, form clusters of amorphous particles having a size of not more than 3 µm. By repeated washing to remove co-precipitated ammonium nitrate, the precipitate additionally peptides and further deteriorates the filtration properties of the products, their further processing and yields.

The summary summarizes the current state of uranium oxide production, which can be represented by a number of patents, such as US patents.

394,997, 3,174,821, 3,846,520, 3,120,526, or Oap. U.S. Pat. 6,924,884, Fr. U.S. Pat. 1 152 140

221,395, Brit. U.S. Pat. 845 235, 893 320 and the like, which, with major or minor variations, have also been applied industrially.

The aforementioned disadvantages of the processes for treating uranyl nitrate into nuclear ceramic fuel eliminate the process according to the invention, characterized in that uranyl nitrate or uranyl nitrate is mixed with plutonium nitrate or thorium nitrate in aqueous solutions. ammonia at pH 7 to 9, the resulting colloidal suspension is subjected to a pH-ex. direct thermal degradation at 500 to 900 ° C and the reduced pellet product is then sintered in a known manner.

The process of the invention illustrates, for example, the following reaction scheme;

U0 ₂ (NO ₃ ) ₂ + 12 NH ₃ + 12 HgO -> 2 (NH ₄ ) ₂ U ₂ 0 ₇ + 8 NH ₄ NO ₃ + 6 H ₂ 0

NH ₄ NO ₃ . .............> 8 N ₂ 0 + 16 HgO (NH ₄ ) ₂ U ₂ 0 ₇ -4 U 0 ₃ + 4 NH ₃ + 2 H ₂ 0

U0 ₃ + 2 NH ₃ .. ...........-.....- of 4 U0 ₂ + NgO + 3 HgO

U0 ₂ (NO ₃ ) ₂ + 12 NH ₃ + 12 HgO -V 4 U0 ₂ + 9 NH ₃ + 2, NH ₃ + 27 H ₂ 0

A small excess of ammonia guarantees not only a reducing environment but also degradation of the present ammonium nitrate under completely safe conditions. Mixed solutions containing plutonium or thorium react with completely similar results.

The advantage of the process is in particular that the released gaseous emissions do not even contain toxic intravenous gases which would need to be disposed of for a longer period of time, but do not contain any particulate matter or aerosols. The process eliminates the aforementioned disadvantages of precipitation of ammonium polyuranate at various pH values; the product does not show inferior quality of the products precipitated at higher pH values of the blades 6. The process is suitable 1 for the treatment of the combined ammonium polyuranate amine and plutonium or thorium duaiinates.

The process according to the invention is further elucidated by means of exemplary embodiments which, however, have the possibility to omit the AA '

Example 1 g of crystalline uranyl nitrate was dissolved in water to give 200 g of solution. Ammonia gas was introduced into the solution with stirring until a pH of 7.0 was reached.

The resulting slurry was thermally degraded by gradual introduction into a crucible in an oven heated to 500 ° C.

A total of 28 g of uranium oxide was obtained with a probable composition of υ ^ Οθ, which corresponds to almost 100% yield.

Example. 2

The same solution as in Example 1 was adjusted to pH 8.7 with ammonia gas. The resulting suspension was thermally degraded at 850 to 900 ° C. A total of 27 g of uranium dioxide with the corresponding cubic crystalline structure was obtained, corresponding to a yield of 100%; product after sintering had a density of 10.64 g.cm ^- ^, which corresponds to approximately 98%. .

Exhalations after degradation did not contain detectable amounts of nitrous emissions or radioactive aerosols.

Example3

A solution containing a mixture of uranyl nitrate and plutonium nitrate containing 23 g uranium and 1.2 g plutonium in 200 g solution was treated in a similar manner to Example 2. The solution ammoniaized to pH 9 was thermally degraded at a temperature in the range of 780-910 ° C.

There were obtained 28.4 g of a mixed oxide (95: 5) of UOg-PuO (10: 5), having a specific gravity of 10.72 g.cm < -1 > corresponding to 97.5% of theory.

Example 4

In a similar manner to Example 3, a mixed model solution containing 48 g thorium nitrate and 5 g uranyl nitrate was treated. The pH of the solution was adjusted to 8.6 with ammonia gas to give a fine suspension.

Thermal degradation was performed at a temperature in the range of 780 to 850 ° C. The resulting exhalations did not contain toxic nitrous gases and the mixed product was a cubic system.

Claims (1)

Process for preparing intermediates and products of sinterable ceramic nuclear fuels from aqueous solutions of uranyl nitrate or mixtures of uranyl nitrate with plutonium nitrate or thorium nitrate, characterized in that the aqueous solutions are adjusted to a pH of 7-9 with ammonia, and the colloidal suspension is then subjected to a direct colloidal suspension. thermal degradation at temperatures of 500 to 900 ° C and the reduced powder product is finally tabletted and sintered.