US3073698A - Dispersion hardening of uranium metal - Google Patents

Description

Jan. 15, 1963 W. ARBITER DISPERSION HARDENING OF URANIUM METAL Filed May 29, 1961 5 4 3 2 I 1 mma OOO mwmzomdi mm w IO OXIDE CONTENT, V/O

FIG. 2

INVENTOR WILLIAM ARBITER W MW FIG.3

3,73,fi% Patented Jan. 15, 1963 tine 3,073,698 DISPERSIGN HARDENHJG OF URANIUM METAL William Arbiter, Yonkers, N.Y., assignor to the United States of America as represented by the United States Atomic Energy Commission Filed May 29, 1961, Ser. No. 113,567 Claims. (Cl. '75-212) The present invention relates to a method for hardening uranium metal and more particularly to a method of hardening uranium metal by forming therein a fine and uniform dispersion of uranium dioxide.

Metallic uranium is a desirable nuclear fuel material but the use of this material is restricted to relatively low temperatures and burn-ups due to certain inherent metallurgical limitations. For example, uranium exists in three separate phases, designated alpha, beta and gamma, depending on the particular range of temperature applied. The low-temperature, alpha phase uranium is somewhat malleable and has been described as semiplastic due to its low elasticity. Uranium in this phase is dimensionally unstable and so is not completely suitable for reactor use in this state. The medium-temperature, beta phase uranium is brittle, while the high-temperature, gamma phase is plastic and somewhat more suitable for reactor use than the other phases, especially the low temperature alpha phase.

As an example of the problems presented when pure uranium is used as a reactor fuel, under normal reactor opearting conditions both alpha and beta phases can exist simultaneously in the metal. With fuel core temperatures in the beta temperature range and the fuel skin in the alpha temperature range, distortion and ultimate failure of the fuel can result. This is because of the increase in volume on transforming to beta in the core which puts the skin under severe tensile stresses. Furthermore, fission gas formation at high temperatures causes considerable pressure exceeding the creep strength of the metal.

It has been found possible to minimize the low temperature thermal and radiation induced dimensional instabilities in unalloyed uranium by providing fine, randomly oriented grains in the metal. This is done either by beta heat treatment of cold Worked uranium metal followed by quenching, or by the use of powder metallurgical techniques in the fabrication of the uranium. These measures, however, are only slightly helpful. Another approach is to make alloying additions to the uranium to stabilize the isotropic gamma phase, thereby avoiding the low temperature instability. However, since considerable quantities of molybdenum or niobium are required for stabilization of the gamma phase, these additions generally have an undesirable efiect upon neutron economy. Therefore, a uranium-rich metallic fuel material having good elevated temperature strength properties and capable of high burn-up is still not available.

It has been suggested that the mechanism of dispersion hardening be utilized to solve this problem. By this technique, a large number of fine particles of a refractory material are uniformly distributed throughout the matrix of the parent metal. In order to be eifective, however, the spacing between the particles of the dispersed phase must be small, about one micron or less. In addition, the dispersion must be stable, that is, not grow or agglomerate at the temperature of use. This requirement eliminates most precipitation-hardened alloys from consideration for high temperature applications, since in these alloys the precipitated phase is soluble in the matrix from which it was originally derived.

The best known dispersion-hardened product known is SAP, an aluminum base material made by means of powder metallurgy from surface oxidized fine aluminum. Attempts to apply this type of dispersion hardening to uranium metals have not proved successful, because no technique or method has been devised for successfully and accurately obtaining the proper degree of dispersion along with proper particle size as well as the desirable spacing of the dispersoid.

In the present invention, it is possible to apply the metal-metal oxide dispersion technique to uranium metal. The inventive method permits the exercising of close, complete and accurate control over the proper degree of dispersion, particle size, and spacing of the dispersoid. The dispersoid, which is U0 in the preferred embodiment of this invention, instead of being inserted as a separate ingredient into the uranium particles, is formed chemically in the pure metal as a surface film on very pure, finely divided uranium powder prior to the consolidation of the powder by powder metallurgical techniques. In order to obtain the uranium at a very finely divided pure and active state, the uranium is initially, in accordance with this invention, hydrided to form uranium hydride which results in a large volume increase causing the uranium hydride to powder as it forms on the surface of the uranium. The particle size may be made smaller by decomposing the hydride and repeating the hydriding step. The hydride powder thus formed is very active due to its fine particle size and large surface area. As a result, it is seen that the uranium metal powder which is later formed from the hydride in accordance with this process is likewise extremely active because of its highly divided state and large surface area, a fact which results in its sintering at temperatures as low as 600 F. when the hydride is decomposed. This sintering or agglomeration of the decomposed hydride means that a comminution step is required to bring the material backto a fine particle size. it is diilicult, if not impossible, to perform such comminution and regain the fine particle size originally present in the hydride without contamination. However, to prevent such sintering from occurring, the hydride is exposed to a very dilute mixture of oxygen in an inert atmosphere to form initially on the fine uranium hydride particles a thin oxide film. The oxide skin thus formed prevents sintering or agglomeration of the uranium powder as it is formed during the decomposition of the hydride which is continued until completion. It is the presence of the thin oxide skin on each uranium particle which permits the retention of the fine particle size during the subsequent process steps and obtains many of the benefits of this invention as Will be later described in more detail.

it is, therefore, a first object of this invention to provide a method for dispersion hardening uranium metal.

Another object of this invention is to provide a method for preparing a dispersion hardened uranium metal based on a metal-metal oxide system.

Still another object of this invention is the provision of a method of preparing uranium metal powder for use in reactor applications.

Another object is the provision of a method for dispersing U0 in a uranium matrix.

Still another object is to provide a method for controlling the degree of dispersion, particle size, and spacing of a U0 dispersoid in a uranium dispersion medium.

Another object of this invention is to provide a method for preparing a finely divided pure uranium powder having a controllable amount of uranium oxide on the surface in the form of a thin film.

Other objects and advantages of this invention will hereinafter become more evident from the following description of a preferred embodiment of this invention with reference made to the accompanying drawing in which:

FIG. 1 is a schematic diagram of apparatus suitable for carrying out the preferred method of this invention;

FIG. 2 shows a chart of the Meyer hardness of uranium plotted as a function of oxide content; and

FIGQ3 is a photomicrograph of a sample of dispersion hardened uranium prepared in accordance with this invention.

In a preferred form of this method, uranium is first hydrided to form uranium hydride (UH This may be accomplished by loading uranium shot or lumps into a stainless steel container and then passing dry hydrogen into the container. Simultaneously, heat is applied to the container to carry out the desired reaction at 400 to 690 F. Hydriding of the uranium causes the final product to expand and powder with particles in the range of about 0.5 to 20 micron size. Repetition of the hydriding step after first decomposing the UH particles under heat brings the average particle size into the lower part of this range with somewhat greater uniformity in the final product.

The next step of the process is carried out in the apparatus illustrated in FIG. 1, where is shown a heat resistant glass column 10 provided at the bottom with a fritted glass filter 12 and below this a vacuum stopcock 14. At

' the top of column 10 is a ground glass joint 16 which permits column 10 to be sealed vacuum tight and in which a filter 18 is disposed. Column 10 exhausts through a bubble trap 19 as is understood in the art. The uranium hydride powder to be oxidized in accordance with this invention (as is to be more particularly described below) valve controlled manifold 32, while calibrated volume 26 l is provided with a pressure gauge 34.

In utilizing the apparatus of FIG. 1 to carry out the next step of the method of this invention, column 10 is charged with the uranium hydride powder (prepared as previously described) after the powder weight is first determined. With column 10 charged with the uranium hydride, the quantity of oxygen required to yield the desired oxide content is calculated and the calibrated volume 26 is filled with the required amount of oxygen at some desired pressure, i.e., 20 p.s.i.g. Then from manifold 24, argon is admitted under suitable pressure into the reaction column 10 and the fiow rate is establishedto provide agitation of the powder bed therewithin.

When the agitation is established, oxygen is permitted to bleed into the argon stream through calibrated leak 28 at a very low rate of flow to prevent burning. A thermocouple (not shown) attached to the outside surface of column 10 adjacent to the powder bed may be: used to record the temperature of the column. When the temperature of column 10 reaches a peak and begins to fall I off, the oxidation is interrupted by cutting off the supply of oxygen. If the total column weight of the powder is obtained again, the difference, of course, between the two weights represents the oxygen pickup. This initial oxidation step, which is more or less self regulating, may be referred to as the preoxidation step of this process, and

results in thin stoichiometric or non-stoichiometric uranium oxide coatings on-the UH particles in the powder.

When the temperature peak mentioned above is reached the surface oxidation is completed and the reaction does not continue. This pre-oxidation step results in the preservation of the fine particle size of the uranium powder during the remainder of process because thefthin U coatings prevent sintering of the uranium metal which normally occurs at temperatures as low as 600 F. in this finely divided and active state. The preoxidation step l may be carried out at ambient temperatures and up to as high at 250 P. if desired.

In order to convert the UH particles contained within the U0 into pure uranium, the pre-oxidized uranium hydride powder is heated in vacuum to 600-700 F. During this treatment, the bulk of the hydrogen is removed but because of the high density oxide skin, the rate of the removal is slow. The elimination of the hydrogen from the lattice permits additional oxygen to be adsorbed during subsequent oxidation operations.

After the hydrogen is removed, the procedure outlined in the preoxidation step is repeated until all the premeasured amounts of oxygen are absorbed. With the hydrogen removed, the oxygen penetrates the U0 layers to oxidize the uranium directly underneath. It has been found that the final oxide content of the material can be controlled to within l-2 volume percent, while the final U0 content may be varied by this process from 1 to 30 volume percent. With the pre-oxidation step of the hydride being self-controlled, it is actually the subsequent, final oxidation step after the decomposition of the preoxidized hydride particles which permits the proper amount of oxygen to be added to obtain the final desired relationship between oxide and pure uranium.

After the UO -U powder is produced in accordance with this invention, the final dispersion hardened product is prepared by hot or cold pressing in accordance with known technques.

. EXAMPLES A large number of runs were made carrying out the process described above and they may be illustrated by the following examples.

Uranium shot was hydrided as described with reaction temperatures in the range of 400 to 690 .1 producing UH particles of 0.5 to 20 size, as already indicated. A typical example is one batch which was hydrided at a beginning temperature of 560 F., and final particle size of 4.5 1. Another example with a rehydriding step added produced a final product size of 1.9 It should also be pointed out that uranium in any form initially, such as 3" 'bars of uranium cylindrical stock, will produce the same eifect as far as'the UH powder is concerned'and may be used.

7 Table I lists data of several runs of uranium hydride prepared in column 10 (FIG. 1) in which the calibrated volume was filled with oxygen to 20 p.s.i.g. and the oxygen was permitted to bleed into the argon stream through the calibrated leak at from 10 to 12 0111. of oxygen per minute depending on the pressure. It will be seen that the total volume percent of U0 (v/o of U0 calculated and as desired are reasonably close. Whilethe oxidation steps shown in the table below were carried out at ambient temperatures, other tests indicate that these steps could, if desired, be carried out at elevated temperatures at least up to 250 F.

' Table I Preoxldation Final Oxidation Total v/o U02 Aw. calculated Aw. calculated g O/ g. U v/o U02 g O/g. U v/o U0: Desired Actual TESTS A series of tests'on the' dispersion hardened uranium metal prepared from powders made in accordance with 'i the method described above were performed to determine how the physical characteristics of pure uranium compared with the characteristics of the dispersion hardened materials. The UO -U dispersions described were prepared for these tests by hot pressing and extruding the powdered product.

For example, high temperature, or gamma phase, extrusions were prepared, and these were subject to analysis and testing. FIG. 2 is a graph showing the relationship between hardness of the final product after extrusion as a function of oxide content. Point A represents a sample which was selected for metallographic examination. In FIG. 3 is shown a photomicrograph of this specimen having 11.4 v/o U as polished, at 500x. This specimen shows a close approach to an ideal dispersion-hardened microstructure, although there are some areas relatively devoid of the oxide phase. Point B in FIG. 2 represents the hardness of unalloyed uranium without any oxide content for comparison with the samples with oxide content prepared in accordance with this invention.

Other tests on these samples were run, including alphabeta thermal cycling and transverse creep measurements under load. These may be briefly summarized in Table II in which a sample with 16.5 v/o of U0 prepared in accordance with this invention is compared to an unalloyed pure uranium control:

1 3.5 lb. load at 1500 F. applied for 15 minutes in vacuum. 2 Max. fibre stress 1330 p.s.i. at 1700 F. for 5 minutes. 3 Max. fibre stress 4040 p.s.i. at 1740 F. for 30 minutes.

Summarizing the results of the samples and tests, the extent of the oxidation can be controlled to yield a product containing from 1 to 30% volume of U0 and the material thus formed is ideally suited for hot pressing and extrusion. This extrusion may be accomplished for example with a reduction ratio of ten to one at a temperature of 1500 to 1700 F. with pressures at 10,000 to 22,000 p.s.i. This achieves a dense dispersion hardened uranium having good hot strength, hardness and good thermal cycling and irradiation behavior.

It is thus seen that there has been provided a method for successfully producing a UUO dispersion in which the oxide content is controlled to a high degree. Further, the process retains fine particle size which is important for obtaining a matrix having good mechanical and nuclear properties. Further, it has been demonstrated in accordance with this invention that, after being hot pressed and extruded, under certain conditions, dense dispersion hardened compositions can be made containing up to 30% volume of oxide.

In effect, this invention permits the controlled oxidation of finely divided uranium particles accomplished in a uniform manner to an extent not heretofore found possible. This permits the use of unalloyed uranium as a metallic nuclear fission fuel material having good elevated temperature strength properties capable of high 5 burn-up. It also provides for the first time a way of applying the well-known dispersion hardening approach along with its advantages to the treatment of uranium. Also, the method permits to a certain extent the selection or the final physical characteristics because it permits the amount and thickness of the coatings of the U0 to be preselected in advance with the advantages already noted.

While only a preferred embodiment of this invention has been described, it is understood, of course, that the invention may be practiced in accordance with the scope of the appended claims.

What is claimed is:

1. The method of dispersing U0 in a uranium matrix comprising the steps of preparing a bed of fine particles of UH passing small amounts of oxygen mixed with an inert gas through said bed until all of said particles are coated with thin layers of U0 as a result of a chemical exchange, and heating said bed to decompose said UH particles causing the escaping hydrogen to pass out through said thin layers of U0 leaving a bed of unsintered fine uranium particles coated with U0 2. The method of claim 1 in which the oxygen is added in minute amounts at a temperature ranging from ambient up to 250 F.

3. The method of claim 1 in which the bed of coated UH particles is heated in a vacuum at 600-700 F. to remove the hydrogen.

4. The method of claim 1 in which additional oxygen is added to said bed after decomposition of said UH particles to further oxidize the coated uranium particles and thereby thicken the U0 coatings on said particles.

5. The method of claim 1 in which the sizes of the UH particles range from 0.5 to 20 1..

6. The method of preparing a dispersion of U0; in a uranium matrix comprising the steps of preparing an agitated bed of fine UH particles by flowing an inert gas therethrough, preoxidizing said particles by adding small amounts of oxygen to said inert gas until said UH particles are coated with U0 as a result of a surface chemical exchange, heating said coated UH particles at a temperature suflicient to decompose said UH until substantially all of the hydrogen has been freed thereby leaving a bed of unsintered uranium particles coated with U0 and further oxidizing said uranium particles by agitating and flowing further oxygen mixed with said inert gas through said bed until the desired content of U0 is reached.

7. The method of claim 6 in which the preoxidation of said particles is terminated when the temperature of said bed reaches a peak value and begins to decline.

8. The method of claim 6 in which the bed of UH particles is prepared from uranium which has been bydried to form UH which powders during its formation into a very fine particle size.

9. The method of claim 8 in which the UH is decomposed under heat and the hydriding step is repeated.

10. The method of dispersion hardening uranium metal comprising the steps of adding oxygen to a matrix of uranium hydride particles being 0.5 to 20p. in size sufiicient to form a uniform dispersion of U0; in the range of 1 to 30% by volume throughout said matrix and work hardening said dispersion by extruding the latter.

References Cited in the file of this patent UNITED STATES PATENTS 2,894,838 Gregory July 14, 1959

Claims (1)

1. THE METHOD OF DISPERSING UO2 IN A URANIUM MATRIX COMPRISING THE STEPS OF PREPARING A BED OF FINE PARTICLES OF UH3, PASSING SMALL AMOUNT OF OXYGEN MIXED WITH AN INERT GAS THROUGH SAID BED UNTIL ALL OF SAID PARTICLES ARE COATED WITH THIN LAYERS OF UO2 AS A RESULT OF A CHEMICAL EXCHANGE, AND HEATING SAID BED TO DECOMPOSE SAID UH3 PARTICLES CAUSING THE ESCAPING HYDROGEN TO PASS OUT THROUGH SAID THIN LAYERS OF UO2, LEAVING A BED OF UNSINTERED FINE URANIUM PARTICLES COATED WITH UO2.

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